Radiation Protection NPP

CONTENTS

CHAPTER 1 BASIC CONCEPT 1.1 Ra&oactiviry and Radiation ............................................. 1

1 . 1. 1 Natural Radioactivity .............a. ...-...................... 2 1.1 -2 Radiation ............................................................... 2 1.1 -3 Radiation Quantities and Units ................................. 4

1.2 Interaction between Rays and Matter -=- - - - * - - - * - -=*- - - * - - - * - - - * - - - - - - 12 1.2.1 X Rays and y Rays ................................................ 13 1.2.2 Neutrons ............................................................ 22 1.2.3 fl Rays ............................................................... 25

Fiwre .............................................................................. 32

CHAPTER 2 RADIATION DETECTION BASE 2.1 Radiation Detection ...................................................... 44

2.1.1 Introduction ......................................................... 44 2.1.2 Ionization Method ................................................ 45 2.1.3 Scintillation ......................................................... 68 2.1.4 Solid Detector ...................................................... 87 2.1 -5 Neutron Detector ..................... ........................... ... 97

2.2 Statistic of Radiation Measurement ................................. 102 2.2.1 Probability Distribution .........................................a 103 2.2.2 Standard Deviation ............................................a 123 2.2.3 Detection Limit and Sensitivity ................................. 135

Figure .............................................................................. 141

CHAPIXR 3 RADIATION PROIXCTION BASE 3.1 Radiation Act on Organism ............................................. 167

3.1.1 X Rays and Mankind undergone Radiation Damage in Earli ness ......................................................... 167

3.1 -2 Classification of Exposure to Radiation ..................... 169 3.1 -3 Classification of Radiation Damage........................... 169

3.2 Principles of Radiation Protection in Practice ..................... 172 3.2.1 Justification for Practic ......................................... 173

3.2.2 Radiation Protection Optimization ........................... 173 3.2.3 Dose Equivalent Limitation to Individual .................. 175

3.3 Shielding against Radiation and Safety Operation for Radiation ............................................................... 180 3.3.1 Influence of Factors on Exposure in Radiation Field.... .a 181 3.3.2 Protection of Various Rays .................................... 182

3.3.3 Protection of External Exposure for Gamma Rayk ...... 183

- 1 -

3.3.4 Protcction of Neutron External Exposure -*---------------• 198 FiSre .............................................................................. 215

CHAPTER 4 RADIATION PROTECTION OF NUCLEAR POWER PLANT AND REACTOR

4.1 General Desc-iption ...................................................... 223 4.1.1 Nuclear Power Plant ........................................me... 223 4.1.2 Radiation Protection Limits of Nuclear Power Plant ...... 224

4.2 Radiation Sources in Nuclear Power Plant .......=.....=.......-.. 227 4.2.1 Reactor Block ...................................................... 227 4.2.2 Coolant System ......................................en........ 231 4.2.3 Storage and Transportation of Spent Fuel .................. 23 3 ....................................... 4.2.4 Waste Treatment Systems 233

4.3 Radiation Detriment of Nuclear Power Plant ..................... 234 ....................................... 4.3.1 Occupational Exposures 234

.......................................... 4.3.2 Effects on Environment 240

4.3.3 Radiation Accident of Nuclear Power Plant - = - - - - - - - - - - - - - 244 4.4 Prevention Measures of Reducing Exposure Received

Occupational Radiation Workers ...............................an... 244 4.4.1 Control by Zoning ................................................ 244 4.4.2 Shielding .... ..................................................... 247 4.4.3 Ventilation ......................................................... 249 4.4.4 Measures for Reducing Activity of Radiation Sources ... 252 4.4.5 Planning. Organization and Trainning......*........*=..=.... 253

4.5 Protection Measures of Reducing Exposure of Public ............ 254 4.5.1 Siting a.............................................................. 254 4.5.2 Multibarrier of Preventing Release of Radioactive

Matters ............................................................ 258 4.6 Radiation Monitoring of Nuclear Power Plant .........*........... 258

4.6.1 Working Area Monitoring .................................... 259 4.6.2 Effluent Monitoring ............................................. 259

.. 4.6.3 Environment Monitoring .................................... 261 Fipre .............................................................................. 263

CHAPTER 5 RADIATION PROTECTION MONITORING 5.1 Radiation Monitoring Basis ............................................. 265

5.1.1 Ionization Methods Used to Measure X and y Ray Dose ......................................................... 265

5.1.2 Neutron Dose Measurement .................................... 269 ................................. 5.1 -3 Other Measurement Methods 271

5.2 Dosage Monitoring ...................................................... 280 5.2.1 Introduction ...................................................... 280

Light - - Pipes - -~ ~~

Tt is often unadvisable or even impossible to couple a PM tube directly to one face of a scintillator. For example, the size or shape of the scintillator may not conveniently match the circular photocathode area of commerciallv available PM tube. Better light collection efficiency usually can be achieved by using a transparent solid, know as a light pipe. to physically couple the scintillator to the PM tube and to act as a guide for the scintillation light.

For the simple case of a cylindrical scintillation crystal and tube of equal diameter, the light pipe can be a simple cylinder of the same diameter. More often, however, the light pipe cross-section shape must vary along its length in order t serve as a smooth transition between the scintillator exit surface and the PM tube end-window.

Photomultiplier Tube The widespread use of scintillation counting in radiation detection would be impossible without the availability of devices of convert the extremely weak light output of a scintillation pulse into a corresponding electrical signal. The photomultiplier (PM) tube accomplished this task remarkably wel1,converting light signals that typ- ically consist of no more than a few hundred photons into a usable current-pulse without adding a large amount of random noise to the signal.

The simplified structure of a typical photomultiplier tube is illustrated in Fig. 2.1-20. The shell of the PM tube is often made by the glass, and inside of it is in vacuum. The electrode structure is different for different type of the PM tube. The electrical lead of electrodes is connected with the tube pin through the base.

The two major elements consist of a photo sensitive layer called the photocathode, which is coupled to an electron multiplier structure. The photocathode serves to convert as many of the incident light photon as possible into low-enerm electrons. If the light consist of a pulse from a scintillation crystal, the photoelectrons produced will also be a pulse of similar time duration. Because only a few hundred photoelectrons may be involved in a typical pulse, their charge is too small at this point to serve as a detectable electrical signal. The electrum multiplier section in a PM tube provides an efficient collection geometry for the photoelectrons as well as serving as a near-ideal amplifier to greatly increase their number. After amplification through the multiplier structure, a typical scintillation pulse will give rise to 10~-10 '~ electrons, sufficient to easily C P ~ P a~ the r h g r g ~ . c_igzrl fs, Lb Ul;5inb: sciati1la:ion event. This charge

is conventionally collected at the anode or output stage of the multiplier

-81 -

structure. Most photomultiplier perform this charge amplification in a very

linear manner. producing an output pulse that remains proportional to the number of original photoelectrons over a wide range of amplitude. Much of the timing information of the original light pulse is also retained.

Photocathode The act of the photocathode is the conversion of -- -

incident light photons into electrons. First, the incident photons are absorbed and the energy is transferred to an electron within the photoemissive material, second, the electron migrates to the surface, and final the electron escapes from the surface of the photocathode.

Photocathodes can be constructed as either opaque or semitransparent layers. An opaque photocathode is normally fabricated with a thickness somewhat greater than the maximum escape depth, and is supported by a thick backing material. Photoelectrons are collected from the same surface on which the light is incident. Semitransparent photocathodes generally are no thicker than the escape depth and are deposit on a transparent backing (often the glass end window of the PM tube).

Presently available materials for photocathodes include a "multi-alkaliN material based on the compound Na,KSb. Prepared by activation with a small amount of cesium, this material was the frrst to show a relatively high quantum efficiency of up to 30 percent in the blue region of the spectrum. A later formulation based on K,CsSb activated with oxygen and cesium is given the name "bi-alkali" and can show an even high efficiency in the blue.

Multiple Stage I t is the electrode system for multiplicating electrons. The electrons are made by vaporizing Sb-Cs, K-Cs-Sb compound to the Ni base. These materials have higher secondary electron emission rate and lower thermionic electron emission rate. The number of multiple electrode is normally from a few to ten electrodes or more.

Electrons leaving the photocathode are attracted to the first dynode and produce 3-6 electrons for each incident photoelectron. The secondary electrons that are produced at the surface of the first dynode are quite easily guided by another electrostatic field established between the first dynode and a secondary similar dynode, and then produce 3-6 electrons at the surface of secondary dynode for one electron come from the first dynode. These process can be repeated many times until the elec-

trons reach the anode. I t the average multiplication coefficient of electron is m i for i-th dvnode, and the total ilurnkr of dynode is n, then the total gain of the PM tube is M = m, m,---m,.

Anode - - ~~

Anode is the electrode for collecting the electrons and giving output signal. The anode is made of the materials with large work function, far example, Ni, Mo and Nb.

PM Tube Specificauon Manufactures will conventionally quote performance of photomultiplier tubes in terms of certain characteristics which are defined here:

( 1 ) Overall luminous sensitivity Defined as the ratio of the measured anode current at operating voltage to the luminous flux from a tungsten light source of specified temperature incident on the photocathodes. This quantity is an overall measure of the expected current from the PM tube per unit incident light from a broad-band source. The units are amperes per lumen.

(2) Cathode luminous sensitivity Defined as above except that the current of photoelectrons leaving the photocathode is substituted in the numerator for the anode current. This quantity is again measured in amperes per lumen, is a characteristic only of the photocathode, and is independent of the electron multiplier structure.

(3) Overall radiant sensitivity This parameter is defined as the ratio of anode current to radiant power a t a given wavelength incident on the photocathode. Units are amperes per Walt.

(4) Cathode radiant sensitivity Defined as above, except that the photocathode current is substituted for the anode current.

(5) Dark current Normally specified in terms of anode current measured without photocathode illumination when the tube is operated to provide a given overall luminous sensitivity.

(6) Anode pulse rise time Quoted as the time taken for the output pulse to rise from 10% to 90% of the peak when the photocathode is illuminated by a flash of light of very short duration.

(7) Anode pulse width Normally quoted as the time width of the output pulse measured at half maximum amplitude, again for short-duration illumination of the photocathode.

Linearitv of PM Tube The electron multiplication factor in nearly all PM tubes remains constant for pulses that range in size from a single photoelectron to many thousands. Under these conditions the amplitude of the pulse collected at the anode is linearly related to the

- 8 3 -

number of photoelectrons, and consequently to the intensity of the scintillation light flash. Nonlinearities can arise for very large pulses due to space charge effects between the last dynode and anode where the number of electrons is greatest. The buildup of space charge affects the trajectories of electrons in this region and causes some to lost which would otherwise be collected. Another factor that can cause nonlinearities at high pulse amplitudes is any deviation of dynode voltages from their equilibrium value during the course of the pulse.

Under normal circumstances in scintillation pulse counting, these effects are seldom important with an adequately designed tube base and the photomultiplier tube remains in the linear range.

High-Voltage ~ ~ Supply and Voltage Divider An external voltage source must be connected to the photomultiplier tubes in such a way that the photocathode and each succeeding multiplier stage are correctly biased with respect to one another. Because electrons must be attracted, the first dynode must be held at a voltage that is positive with respect to the photocathode, and each succeeding dynode must be held at a positive voltage with respect to the preceding dynode. For efficient photoelectron collection, the voltage between photocathode and first dynode is often several times as great as the dynode-to-dynode voltage differences.

Tn the vast majority of cases, the voltage differences are provided by a resistive voltage divider and a single source of high voltage. Figure 2.1-21a shows a typical wiring diagram for the base of a photomultiplier tube in which a positive polarity high voltage is used. Tn this case the cathode of the photomultiplier tube is grounded, and the divider string supplies successively increasing positive voltages to each dynode down the multiplying string. The anode is a t a dc potential equal to the supply voltage, and signal pulses must therefore be capacitively coupled from the anode to allow the pulse component to be passed on at ground potential to succeeding electronic devices. The load resistor RL can be chosen by the experimenter so that the resulting anode circuit time constant is of proper magnitude. The anode capacitance C, usually is not a physical capacitor, but only the stray capacitance associated with the anode structure and connected cables. An analysis of the pulse shape expected from this anode circuit is given later.

Tn order to use the simplest and least expensive voltage supply. one would like to keep the current through she divider a t a minimum. Small

currents will also minimize problems due to heat dissipation in the divider resistors. However. the divider string current must always be kept large compared with the internal photomultiplier tube current represented by the pulse of electrons flowing from dynode to dynode. If the internal current at the peak of a pulse ever becomes comparable with the divider current, the voltage of the dynodes will begin to deviate from their equilibrium values, leading to drift of the PM tube gain. This problem is especially serious for the last few stages of the PM tube where peak currents are at a maximum. It is often advantageous to provide "stabilizing capacitors" (labeled as Cs in Fig. 2.1-2 1) to the stages of the divider string near the anode to help hold these latter dynode voltages at a constant value throughout the pulse. In order to prevent a more than 1 percent interdynode voltage change, the charge stored on the stabilizing capacitor (given by the product of capacitance value and the interdynode voltage) must be 100 times greater than the charge emitted by that dynode during the pulse.

Exactly the same interdynode voltages can be achieved by ground- ing the opposite (or anode) end of the divider string, and applying negative high voltage to the photocathode end. This latter arrangement is shown in Fig. 2.1-21b. It is therefore important that users be aware of which convention has been chosen by the manufacture of their own tube base before initial use of the equipment with a photomultiplier tube.

An advantage that stems from use of negative high voltage is the elimination of the coupling capacitor C, requires if positive polarity is used. The anode is now at ground potential, and signal pulse can be direct-coupled into subsequent measuring circuits. This advantage is particularly important for fast pulse applications in which it is often desirable to couple the anode directly into 504hm transmission line structure. However, operating the tube with negative high voltage means that the photocathode will be at the full high voltage supplied to the tube, and care must be taken to prevent spurious pulses due to high-voltage leakage through the glass tube envelope to nearby grounded structures.

Light and Magnetic Shielding The light to be accepted by the PM tube is very weak, therefore the PM tube can not be used under exposure to light. Intensive incident light will make the photocathode and multiplier fatigue due to the large current. therefore the power of emitting electrons for the PM tube may be lost and then the PM tube can not be used. The PM tube must be put in the shell which is in light-tight.

The electron optics within a photomultiplier tube are particularly sensitive to stray magnetic fie!ds because of the !ow average energy (on the order of 100 eV) of the electrons traveling from stage to stage. In situations in which the tube is likely to be physically moved or brought near equipment with stray magnetic fields, it is essential that a magnetic shield be provided to prevent gain shifts of the PM tube. The most common form conslsts of a thin cylinder of mu-metal which fits closely around the outside glass envelope of the PM tube. For most tube designs, this shield must be held at photocathode potential in order to avoid disturbing the electrostatic field between the photocathode and first dynode.

Output Pulse Shape of Scintillation Detector The anode circuit can be idealized as shown in Fig. 2.1-22. C presents the capacitance of the anode itself, plus capacitance of the connecting cable and input capacitance of the circuit to which the anode is connected. The load resistance R may be a physical resistor wired into the tube base or, if none is provided, the input impedance of the connected circuit. After one scintillation event occurs, the current reached at node is:

L --

i ( t )=io - e ro (2.1 - 17)

where 7, is the reciprocal of the scintillator decay constant i. The initial current 6 can be expressed in terms of the total charge Q collected over the entire pulse. Because

- - l o - T o

= N - M - e following formula can be obtained

Where N = the number of electron inverted a t the photocathode due to the

photons produced in the scintillator by one charged particle M =the multiplication factor of the PM tube 7, = the scintillation decay time

e - 1 . 6 ~ 1 0 - l ~ ~ . To derive the voltage pulse V(t) expected at the anode, we first note

that the current flowing into the parallel R-C circuit must be the sum of the current flowing into the capacitance i, and the current through the resistance i,:

The following formula can be obtained from above equation and initial condition V(0) = 0

If RC> >T,, that is to say, the anode time constant is larger than the scintillator decay time. then Eq. (2.1-22) can be become

If RC < < T,, then Eq. (2.1-22) can be become

0

The plot of these pulse forms is shown in Fig. 2.1-23. Application of Scintillation Detector There are wide application

for scintillation detector in nuclear power plant. Using it to monitor the total gamma radioactive in the water of the reactor primary loop, the failure fuel element in the reactor core can be predicted. Using it to monitor the gamma radioactive level in the water of the reactor secondary loop, the water leakage from primary into secondary can be predicted.

The scintillator whose effective atomic number is approximate to that of the air or body tissue can replace the ionization chamber. For example, the scintillation detector with 10cm3 scintillator will give the current from 1 0 - l ~ ~ to 5 x 1 0 + ~ for the exposure rate from lOmR / hr to 500 R / hr. The operation range for this devices is more wide than that for the air ionization chamber.

2.1.4 SOLID DETECTOR

In many radiation detection appiications, the use of a solid detection medium is of great advantage. For the measurement of high-energy electrons or gamma rays. detector dimensions can be kept much smaller than that of the equivalent gas-filled detector because solid densities are some 1000 times greater :han that for a gas.

Band --- Structure - -- - in Solids The periodic lattice of crystalline inaterials establishes allowed energy bands for electrons that exit within that solid. The energy of any electron within the pure material must be confined to one of these energy bands which may be separated by gaps or ranges of forbidden energies. A simplified representation of the bands of interest in insulators or semiconductors is shown in Fig. 2.1-24. The lower band, called the valence band, corresponds to those electrons that are bound to specific lattice sites within the crystal. In the case of silicon or germanium, they are parts of the covalent bonding which constitute the interatomic forces within the crystal. The next higher-lying band is called the conduction band and represents electrons that are free to migrate through the crystal. Electrons in this band contribute to the electrical conductivity of the material. The two bands are separated by the band gap, the size of which determines whether the material is classified as a semiconductor or an insulator. The number of electrons within the crystal is just adequate to completely fill all available sites within the valence band. In the absence of thermal excitation, both insulators and semiconductors would therefore have a configuration in which the valence band is completely full, and the conduction band completeiy empty. Under these circumstance, neither would theoretically show any electrical conductivity.

Charge Carriers At any nonzero temperature, some thermal energy is shared by the electrons in the crystal. I t is possible for a valence electron to gain sufficient thermal energy to be elevated across the band gap into the conduction band. Physically, this process simply represents the excitation of an electron which is normally part of a covalent band such that it can leave the specific bonding site and drift throughout the crystal. The excitation process not only creates an electron in the otherwise empty conduction band, but it also leaves a vacancy (called a hole) in the otherwise full valence band. The combination of the two is called an electron-hole pair and roughly the solid-state analog of the ion pair in gases. The probability per unit time that an electron-hole pairs is t'lermally generated is given by

where T = absolute temperature E, = band gap energy K = Baltzrnann constant C = proportionality constant characteristic of the material.

As reflected in the exponential term, the probability of thermal excitation is critically dependent on the ratio of the band gap energy to the absolute temperature. Materials with a large band gap will have a low probability of thermal excitation, and consequently will show the very low electrical conductivity characteristic of insulators. If the band gap is as low as several electron voltst sufficient thermal excitation will cause a conductivity high enough for the material to be classified as a semiconductor. In the absence of an applied electric field, the thermally created electron-hole pairs ultimately recombine, and an equilibrium is established in which the concenuation of electron-hole pairs observed at any given time is proportional to the rate of formation.

Migration of Charge Carriers in an Electric Field If an electric field is applied to the semiconductor material, both the electrons and holes will undergo a net migration. The motion will be the combination of a random thermal velocity and a net dnft velocity parallel to the direction of the applied field.

At low-to-moderate values of the electric field intensity, the drift velocity v is proportional to the applied field. Then a mobility p for both electrons and holes can be defined by

- +

v h = p h 0 E (2.1 - 26)

v e = p e 0 E (2.1 - 27)

where E is electric field magnitude. Numerical values for common semiconductor materials are given in Table 2.1-7.

Many semiconductor detectors are operated with electric field values sufficiently high to result in saturated drift velocity for the charge carriers. Because these saturated velocity are of the order 107cm / s, the time required to collect the carriers over typical dimensions of O.lcm or less will be under 10 nanoseconds.

Intrinsic ~ Semiconductors In a completely pure semiconductor, all the electrons in the conduction band and all the holes in the valence band would be caused by thermal excitation (in the absence of ionizing radiation). Under these conditions each electron must leave a hole behind, the number of electrons in the conduction band must exactly equal the number of the holes in the valence band. Such material is called an intrinsic semiconductor. That is to say,

The quantities ni and pi are known as the intrinsic camer densities. Its properties can be described theoretically, but in practice it is vir-

tually impossible to achieve. The electrical properties of real materials tend to be dominated by the very small levels of residual impurities.

Intrinsic hole or electron densities at room temperature are 1.5 x 10'~cm-~ in silicon. and 2.4 x 10 '~cm-~ in germanium.

n-Type Semiconductor To illustrate the effect of doping on semiconductor properties, we will use crystalline silicon as an example. Germanium and other semiconductor materials behave in a similar way. We first assume that the impurity is pentavalent, or is found in Group V of the periodic table.When present in small concentrations (of the order of a few pans million or less) the impurity atom will occupy a substitutional site within the lattice, taking the place of a normal silicon atom. Because there are five valence electrons surrounding the impurity atom, there is one left over after all covalent bonds have been formed. It takes very little energy to dislodge it to form a conduction electron without a corresponding hole. Impurity of this type are referred to as donor impurities because they readily contribute electrons to the conduction band. Because they are not pan of the regular lattice, the extra electrons associated with donor impurities can occupy a position within the normally forbidden gap. These very loosely bound electrons will have an energy near the top of the gap as shown in Fig. 2.1-25a.

Even though conduction electrons now greatly outnumber the holes, charge neutrality is maintained because of the presence of ionized donor impurities. These sites represent net positive charges which exactly balance the excess electron charges. They are not, however, to be con- fused with holes because the ionized donors are fixed in the lattice and can not migrate.

Table 2- 1-7 S ome Characteristic of S i and Ge

' atomic number

1 atomic weight I 28.09 72.60

atomic mass for stable 28, 29, 30 70, 72, 73, isotope I 74, 76

L

density (300K) (g/cmY) 2.33 5.33

number of atomic (l/cm3) 4.96X102' 4.41 X102z

dieletric constant 12 16 -

' energy band width(300K)eV I 1.115 0.665 - - - - -- - - - - P - -- - -

I

energy band width (0K)eV 1.165 I 0.745

concentration of 1.5 XIOIO 2.4 X1Ol3 intrinsic charge carrier (300K) (1,'cm3) I

: in t r i i s ic resistivity 2.3X106 47 , (O . cm)

1 electron mobility i (300K) (cm2/V sec)

I hole mobility 460 , (300K) (cm2(V . sec)

energy of producing one 3.62 pair of electron and hole (300K) eV

' Fano factor 0.08-0.14 0.05-0.12

The net effect in n-type material is therefore to create a situation in which the number of conduction electrons is much greater and the number of holes much smaller than in the pure material. The electrical con- duct.ivity is then deternlined almost exclusively by the flow of electrons. and holes play a very small role. In this case, the electrons are called the majority carriers and holes the minority carriers.

p-Type - - - Semiconductors -- . - The addition of a trivalent impurity such as an element from Group III of the periodic table to a silicon lattice results in a situation sketched in Fig. 2.1-25b. If the impurity occupies a substitutional site, it has one fewer valence electron than the surrounding silicon atoms and therefore one covalent bond is left unsaturated. This vacancy represents a hole similar to that left behind when a normal valence electron is excited to the conduction band, but its energy characteristics are slightly different. An electron filling this hole, although still bound to a specific location, is slightly less firmly attached than a typical valence electron. Therefore, these acceptor impurities also create electron sites within the normally forbidden energy gap. In this case, the acceptor lever lie near the bottom of the gap because their properties are quite close to site occupied by normal valence electrons, for example shown in Fig. 2.1-25b.

In p-type material, holes are called majority carriers and electrons minority carriers.

Compensated Material If donor and acceptor impurities are present in a semiconductor in equal concentration, the material is said to be compensated. Such material has some of the properties of an intrinsic semiconductor. At present, the only practical means for achieving com- pensation over large volumes in silicon or germanium is through the lithium ion dnfting process after the crystal has been fabricated.

"Ionization Energy" When a charged particle passes through a --

semiconductor with theband structure, the overall significant effect is the production of many electron-hole pair along the track of the particle. Regardless of the detailed mechanisms involved, the quantity of practical interest for detector application is the average energy expended by the primary charged particle to produce one electron-hole pair. This quantity, often loosely called the "ionization energyN and given the sym- bol W still.

The dominant advantage of semiconductor detectors lies in the smallness of the ionization energy. The values of W for either silicon or

germanium IS about 3eV. Fano Factor In addition to mean number the fluctuation or vari- - - - . - - - -

ance in the number of charge carriers IS also of primary interest because of the close connection of this parameter with energy resolution of the detector. The Fano factor F is introduced as an adjustment factor to re- late the observed variance to the Poisson predicted variance

observed statistical variance F = E

(2.1 - 29) -

W where the E is the incident panicle energy which is expended in the semiconductor. For good energy resolution, one would like the Fano factor to be as small as possible.

Semiconductor Detector of pn Junction The radiation detectors described here are based on the favorable properties which are created near the junction between n- and p-type semiconductor materials. Charge carriers are able to migrate across the junction if the region are brought together in good thermodynamic contact.

(1) Basic Junction Properties As an illustration, assume that the process begins with a p-type crystal that has been doped with a uniform concentration of acceptor impurity. In the concentration profile at the top of Fig. 2.1-26, this original acceptor concentration NA is shown as a horizontal line. We now assume that the surface of the crystal on the left is exposed to a vapor of an n-type impurity which diffuses some distance into the crystal. The resulting donor impurity profile is labeled

ND on the figure, and falls off as a function of distance from the surface. Near the surface, the donor impurity can be made to outnumber the acceptors, converting the left portion of the crystal to n-type material.

Because the concentration is different for the electron and hole at the two sides of the junction, any charge carriers that can migrate freely will has the net diffusion from the high concentration region to low concentration region.

The diffusion of conduction electrons out of the n-type material leaves behind immobile positive charges in the form of ionized donor impurities. A similar and symmetric argument leads to the conclusion that holes (the majority in the p-type material) must also diffuse across the junction because they also see an abrupt density gradient. Each hole

that is removed from the p side of'the junction leaves behind an acceptor site which has picked up an extra electron. and therefore represents a fixed and immobile negative charge. The combined effect is to build up a net negative space charge on the p side and a positive space charge on the n type of the junction.

The accumulated space charge creates an electric field that dimin- ishes t l e tendency for further diffusion. At equilibrium, the field is just adequate to prevent additional net diffusion across junction, and a steady-state charge distribution is therefore established.

The region over which the charge imbalance exists is called depletion region and extends into both the p and n sides of the junction. If the concentration of donors on the n-type material is higher than that of acceptor atoms in p-type, the electrons diffusing across the junction will tend to travel a greater distance into the p-type material before all have recombined with holes. In this case, the depletion region would extend farther into the p-type side.

(2) Properties of the Reverse Biased Junction If the voltage is applied at the pn junction and made the p side negative and the n side positive, that is to say, the junction has applied reverse bias. In this case the potential will tend to attract minority across the junction. Because the concentration of the minority is very low, the reverse current is very little.

When a reverse bias is applied to the junction, virtually all the applied voltage will appear across the depletion, because its resistivity is much higher than that of the normal n- and p-type material. In this case, the thickness of the depletion can be written

1

d z ( 2 - ~ - V - p - ~ ) ~ (2.1 - 30) where

p = resistivity of the doped semiconductor IL = mobility of the majority carrier V = applied reverse bias E = dielectric constant of the semiconductor. Because of the fixed charges that are built up on either side of the

junction, the depletion region exhibits some properties of a charge capacitor. If the reverse bias is increased, the depletion represented by the separated charges therefore decreases. The value of the capacitance

per un i t area is

N = dopant concentration (either donors or acceptors) on the side of the junction

e = charge of the electron. The meaning of other parameters is same as that in Eq. (2.1-30).

(3) Surface Barrier Detector There are many methods for fabn- cating the pn junction detectors. Surface barrier detectors fabricated by gold and n-type silicon are one typical detectors of the pn junction. A cross-sectional diagram of a typical mounting barrier detector is shown in Fig. 2.1-27. The interface between the silicon and the gold layer on silicon is the junction area. The large part of the junction thickness is in the silicon. The operating principle diagram for this detector is shown in Fig. 2.1-28, and the equivalent circuit for output is shown in Fig. 2.1-29.

The total charge of the electron-hole pairs produced by one charged particle is given by

Because the junction capacitance charges with the operating voltage, the pulse amplitude produced by the same enerm charged particle changes with the operating voltage, is ilot a fixed value. In order to make the output pulse amplitude is still proportional with the energy of the charged particle during the fluctuation of the operating voltage, the charge sensitive preamplifier shown in Fig. 2.1-30 must be used.

In the condition of (l+A)C,> >C,+C,, the pulse amplitude at the output terminal can be derived from Fig. 2.1-30

v -- - A - Q Output C* + C 3 +(1 + A ) - C ,

Selecting the capacitor with good stabiiity as the feedback capacitor C, the requirement that the pulse amplitude is proportional with the energy of rhe charged particle can be ~ealized, and the pulse amplitude is independent of the fluctuation of the external applied voltage.

(4) Collection Time for Electron-hole Pairs Normally, applied reverse bias is very high, and can reached to a few hundred volts to one thousand volts and over, it make the drift velocity of electron-hole pairs reach the maximum value. In this case, the collection time for eiectron-hole pairs is about lo-".

(5) Radiation Damage The radiation damage affects the properties for the semiconductor. Because exposure make the semiconductor material appear crystalline defects, and then make the resistivity and drift of charge carrier destructible, the application properties of semiconductor detector are affectzd. Therefore, semiconductor detectors are often used in measuring of the weaker radiation field.

High-Purity -- Germanium Detector The high-purity germanium detector is widely used for gamma spectroscopy and for nuclide analysis of the samples contained gamma radioactivity.

The high-purity germanium detector can be classified as two types: planar geometry and coaxial geometry.

The operating principle diagram of planar geometry high-purity germanium detector is shown in Fig. 2.1-3 1.

This is one kind of nfpp+ junction detector for the shown planar geometry high-purity germanium detector . The sensitive region of the detector extends from the nf boundary of the high-purity germanium region into high-purity germanium region. With increasing of the operating voltage, the sensitive region can extend to pf boundary of the high-purity germanium region. As good as pn junction detector, the relationship of thickness of sensitive region and the applied bias is the same as Eq. (2.1-30). The equation of the junction capacitor is the same as Eq. (2.1-31). When V is high enough and then making the value of d larger than the thickness of the high-purity germanium region, the sensitive region thickness of the detector reaches the maximum value. Therefore, during the high-purity germanium detector is operating, the operating voitage changes with the requirement of the sensitive region.

The junction capacitor decreases with the V increase. When V is over certain value. the capacitor is no longer decrease.

The operating principle dizgram of the coaxial geometry detector made by high-purity germanium is shown in Fig. 2.1-32.

The electric field configuration in the p region of the coaxial geometry high-purity germanium can be described by

where E = dielectric constant of the medium NA =concentration of acceptor impurity in the p-type high-purity

germanium r = perpendicular distance from axial line e = 1.6 x 10-19c. In order to ensure the sensitive regon thickness of the detector to

the maximum value r,-r, of the geometry thickness itself, the minimum value of operating voltage applied at the detector can estimated by

The junction capacitor of the coaxial geometry detector is decreasing with the V is increasing. When the operating voltage is over the V,,,, the value of the junction capacitance is no longer decreasing.

2.1.5 Neutron Detector Neutron are detected through nuclear reactions which result in en-

ergetic charged particles such as protons, alphas, and so forth. Virtually every type of neutron detector involves the combination of a target material designed to carry out this conversion together with one of the conventional radiation detectors discussed in earlier sections.

Nuclear Reactions of Interest in Neutron Detection In searching for nuclear reactions that might be useful in neutron detection, several factors must be considered. First, the cross sections for the reaction must be as large as possible so that efficient detectors can be built with small

chmensions. Second, the target nuclide should either be of high isotopic abundance in the natural element. Third. of principal important here is the Q-value of the reaction which determines the energy liberated in the reactions following neutron capture. The higher the Q-value, the greater will be the energy given to the reaction products, and the easier will be the task of discriminating against gamma ray events using simple amplitude discrimination.

10 (1) The B(n,a) reaction Probably the most popular reaction for the conversion of slow neutrons into directly detectable particles is the '%(n,a) reaction. The reaction may be written:

where the branching indicates that the reaction product 'Li may be left either in its ground state or in its first excited state. The excited lithium nucleus quickly returns (half-life of - 10-13s) to its ground state with the emission of a 0.48MeV gamma ray. When thermal neutrons (0.025eV) are used to induce the reaction, about 94 percent of all reactions lead to the excited state and only 6 percent directly to the ground state. For all excited state, ELi = 0.84MeV and E,= 1.47MeV.

Figure 2.1-33 is a plot of cross sections versus neutron energy for a number of nuclear reactions of interest in neutron detection. The thermal neutron cross section for the 'OB (n,a) reaction is 3840 barns. The cross section value drops rapidly with increasing neutron energy and is proportional to 1 / v (the reciprocal of the neutron velocity) over much of the range. The natural isotopic abundance of '9 is 19.8 percent.

( 2 ) The 6~i (n ,a ) Reaction The next most popular reaction for the detection of slow neutron is the 6~ i (n , a ) reaction:

Calculation of the neutron product energies for negligible incoming neutron energy yields the following:

The flying direction of alpha panicle and triton produced in the reaction must be opposite when the incoming neutron energy is low.

The thermal neutron cross section is 940 barns. 6 ~ i occurs with a

neutral isotopic abundance of 7.40 percent and is also widely available in separated form.

3 (3) The He(n,pj reaction The gas 3 ~ e is also widely used as a detection medium for neutrons through the reaction:

3 1 3

2 He + ,n -t H + p + 0.765MeV

1 (2.1 - 38) For reactions induced by slow neutrons, the Q-value of 0.765MeV !eads to opposite direction reaction products with energies:

The thermal neutron cross section for the reaction is 5330 barns. BF3 -- - - Proportional - - - Tube A widely used detector for slow

neutrons is the BF, proportional tube. In this devices, boron trifluoride serves both as the target for slow neutron conversion into secondary particles as well as a proportional gas. In nearly all commercial detectors, the gas is highly enriched in 'OB.

( I ) BF, tube pulse height spectra Figure 2.1-34a shows the ideal pulse height spectrum expected from a BF3 tube of very large dimensions. For a large tube, nearly all the reactions occur sufficiently far from the walls of the detector to deposit the full energy of the products within the proportional gas.

Once the size of the tube is no longer large compared with the range of the alpha particle and recoil lithium nucleus produced in the reaction, some events no longer deposit the full reaction energy in gas. If either particle strikes the chamber wall, a smaller pulse is produced. The cumu- lative effect of this type of process is known as the "wall effects" in gas counter. The range of the alpha particle produced in the reaction is on the order of a centimeter for typical BF, gas pressure.

Figure 2.1-34b shows the differential pulse height spectrum expected from a tube in which the wall effect is important. The primary change from the spectrum shown in Fig. 2.1-34a is the addition of a continuum to the left of the peaks corresponding to particle energy deposition in the gas of the tube.

The BF, tube is an example of a detector from which the differential pulse height spectrum tells us nothing about the enerm spectrum of the incident radiation, but is a function only of the size and geometry of the detector itself. We are likely to seek a stable operating

point or a counring plareau for which small drifts in operating parameters do not significantly affect the neutron sensitivity of the counter. That objective would be met by setting a fixed discriminator level a t the point labeled A in Fig. 2.1-34a. The flatten portion of that plateau should occur when the effective discrimination point is at point A.

(2) BF, tube construction The neutron detecting efficiency can be increased and the wall effect suppressed by making the rube large in dimension. Similar improvements can be achieved raising the pressure of BF, fillgas.

Tn common with most proportional counters, BF, tubes are universally constructed using cylindrical outer cathodes and small diameter central wire anodes. Aluminum is often the material of choice for the cathodes because of its low neutron interaction cross section. For low background application, other material such as stainless steel are preferred because aluminum normally shows a small amount of low level alpha activ~ty. With typical anode diameters of O.lmm or less, operating voltages tends to be about 2000-3000V.

The - ,He Proportional Counter With a cross section even higher than that of the boron reaction, the , ~ e ( n , ~ ) reaction is an attractive alternative for slow neutron detection. Unfortunately, because ,He is a noble gas, no solid compounds can be fabricated and the material must be used in gaseous form.

,He of sufficient purity will act as an acceptable proportional gas, and detectors based on this approach have come into common use. Be- cause the range of the reaction products is not always small compared with the dimensions of the proportional tube, however, the wall effect discussed earlier for a BF, tube can also be important for ,He proportional counters. The expected pulse height spectrum for tube of typical size is illustrated in Fig. 2.1-35. Only a signal full-energy peak should be expected for neutron energies which are small compared with 765keV. The step structure to the left of the peak is similar to that shown in Fig. 2.1-34b for a BF, tube. expect that the discontinuities will occur at energies corresponding to that of the proton (574keV) and triton (1 9 1 keV).

The continuum in the pulse height spectrum due to the wall effect is detrimental from several standpoints. The voltage range over which an acceptable counting plateau will be observed is reduced. and the small

-1 00-

pulse height for some neutron events will reduce the separation expected from low-amplitude, gamma-induced pulse.

Compared with the BF, tube, ' ~ e countex can be operated a t much hisher pressures with acceptable gas multiplica~ion behavior, and are therefore preferred for those applications in which maximum detection efficiency is important. The low Q-value of the ' ~ e reaction, however, makes gamma ray discrimination more difficult than that of an equivalent BF, tube.

Lithium-Containing -- Slow Neutron Detectors Because a stable lithium-containing proportional gas does not exist, a lithium equivalent of the BF, tube is not available. Because the lithium reaction goes exclusively to the ground state of the product nucleus, the same energy is always imparted to the reaction products for all slow neutron interactions. The resulting pulse height distribution in detectors which absorb all this energy is therefore a simple single peak.

Lithium-containing scintillators are quite common as also neutron detectors. A logical choice, because of its chemical similarity to sodium iodide, is crystalline lithium iodide. If a small amount (less than 0.1 percent of atoms) of europium is incorporated as an activator, light out- puts of about 35 percent of the equivalent NaI(T1) yield can be achieved. The scintillation decay time is approximately 0.3,s.

Similar to sodium iodide, lithium iodide is highly hygroscopic and can not be exposed to water vapor. Commercially available crystals are hermetically sealed in a thin canning material with an optical window provided on one face. Because of the high density of the material the crystal size need not be large for very efficient slow neutron detection. For example, a lOmm thick crystal prepared from highly enriched 6 ~ i ~ remains nearly 100 percent efficient for neutrons with energy from thermal through the cadmium cut off of 0.5eV.

Counters Based on Neutron Moderation The inherently low detection efficiency for fast neutrons of any above detector can be somewhat improved by surrounding the detector with a few centimeters of hydrogenous moderating material. The incident fast neutron can then lose a fraction of its initial kinetic energy in the moderator before reaching the detector as a lower energy neutron, for which the detector efficiency is generally higher.

The efficiency of a moderated slow neutron detector when used with a monoenergetic fast neutron source will show a maximum at a

specific moderator thickness. Assumins that the moderator is the usual choice of a hydrogenous material such as polyethylene or paraffin. the optimum thickness will range from a few centimeters for keV neutrons up to several tens of centimeters for neutrons in the MeV energy range.

When the properties of a small lithium iodide scintillator placed a t the center of polyethylene moderating spheres of different diameters were investigated, the experimental measurements were made of the sensitivity of the device to neutron of variable energy for sphere diameters ranging from 5.08cm to 30.48cm. Their results are shown in Fig. 2.1-36. The response curve of the 30.48cm sphere have a similar shape compared with the dose equivalent delivered per neutron as a function of energy. Because of the similarity of the two curves, the efficiency of the detector is high for those neutrons whose biological importance is high, and is low for neutrons that deliver less dose. Therefore, the overall count from the detector in a polyenergetic spectrum will automatically include proper weighting factors for all energies and give a meaningful measure of the combined dose due to all the neutrons. Typical intrinsic efficiency for a 30.48cm sphere, defined as the fraction of neutrons that strike the surface of the sphere which ultimately result in a count, is 2.5 x 10" a t the peak of the 30.84cm detector response. Translated into dose, the average response corresponding to about 3 x lo3 counts per mrem. It is virtually the only monitoring instrument that can provides realistic neutron dose estimation over the many decades of neutron energy ranging from thermal energies to the MeV range.

An alternate version of the spherical neutron dosimeter is shown in Fig. 2.1-37, A spherical 3 ~ e proportional counter has been substituted for the lithium iodide scintillator as the slow neutron detector. Used with the 20.8cm diameter polyethylene moderator, the energy response of the system to thermal and epithermal neutrons is higher than ideal so a spherical cadmium absorber, perforated with a number of holes, is placed around the 3 ~ e detector to shape the response curve. The detector underestimates a t thermal energy and overestimates in the keV range, so that the overall response to a mixed spectrum of neutrons should more realistically reflect the actual dose equivalent in most applications.

2.2 STATISTIC OF RADIATION MEASUREMENT Radioactive decay is a random process. Consequently, any

-102-

measurement which is based on observing the radiatiori emitted in nuclear decay is subject to some degree of statistical fluctuation. These inherent iluctuations represent an unavoidable source of uncertainty in all nuclear measurement, aiid oftea can be the predominant source of imprecision or error. Statistics of radiation measurement includes the framework of statistical analysis to process the results of nuclear experiments to make predictions about the expected precision of quantities derived from these measurements.

The value of statistic of radiation measurement falls into two general categories. The first is to serve as a check on the normal function of a piece of nuclear counting equipment. Here a set of measurements is recorded under conditions in which all aspects of the experiment are held as constant as possible.Because of the influence of statistical fluctuations, these measurements will not all be the same but will show some degree of internal variation. The amount of this fluctuation can be quantified and compared with predictions of statistical models. If the amount of observed fluctuation is not consistent with predictions, one can conclude that some abnormality exists in the counting system. The second application deals with the situation in which we have only one measurement. We can then use statistics of radiation measurement to predict its inherent statistical uncertainty and thus estimate a precision that should be associated with that single measurement.

2.2.1 PROBABILITY DISTRIBUTION Characterization of Data We will assume that we have a collec-

tion of N independent measurements of the same physical quantity: x,,x 2,..-.-. 7xi7.---.- PN

we will further assume that a single value xi from this set can only assume integer values so that the data might represent, for example, a number of successive reading from a radiation counter for equal time in- tervals. Two elementary properties of this data set are

X

"Sum": z = C x i I = 1

- r "Experimental mean1/: x = -

N

The experimental mean is written with the subscript e to distinguish it fi-om the mean of a particular statistical model which will be introduced later.

It is often convenient to represent the data set by a frequency distribution function F(x). The value of F(x) is the relative frequency. By definition

number of occurrences of the value "x" F(x) = (2.2 - 3) number of measurements ( = N)

The frequency distribution function F(x) is normalized, that is

3C

C F(x)= 1 (2.2 - 4) r - 0

As long as we d o not care about the specific sequence of the number, the complete data distribution function F(x) represents all the information contained in the original data set.

Table 2.2-1 gives a hypothetical set of data consisting of 20 entries. The corresponding values of F(x) are also shown in Table 2.2-1. Be- cause these entries range from 3 to 14, the data distribution function will have nonzero only between these extreme values of the argument x.

Table 2.2-1 Example of Data Distribution Function

tg Data i Frequency Distribution -- Function 14 p(3)= 1 / 20 = 0.05

A plot of the data distribution function for the example is~given in p~

Fig. 2.2-1. These data show an experimental mean, X e = 8.8. The dis--

tribution function is in some sense centered about experimental mean. Furthermore. the relative shape of the distribution function indicates qualitatively the amount of internal fluctuation in the data set.

Fig. 2.2-2 shows the shape of the distribution function corresponding to two extreme sets of data: one with large amounts of scatter about the mean and one with little. An obvious conclusion is that the width of the distribution function is a relative measure of the mount of fluctuation or scattering about the mean inherent in a given set of data.

It is possible to calculate the experimental mean by using the data distribution function, because the mean of any distribution is simply its first moment.

It is also possible to derive another parameter, known as sample variance, which will serve to quantify the amount of internal fluctuation in the data set. The first step is to define the deviation of any data point as the amount by which it differs from the mean value

-

e i = X i - X c (2.2 - 6)

The 20 numbers given in Table 2.2-1 is shown as the bar graph of Fig. 2.2-3a. The deviation of these value has been plotted on Fig. 2.2-3b. There must be an equal contribution of positive and negative deviation, so that

N

C E ~ = O (2.2 - 7) I - 1

If we take the square of each deviation, however, the result will always be a positive number. The graph are plotted for the example in Fig. 2.2-3c.

We can now introduce the sample variance s2 as

Which will now serve as single index of the degree of fluctuation inher-

ent in the original data. As long as the number of data entries N is reasonably large, the sample variance is essentially the average value of the squared deviation of each data point. To be precise, the sample variance is more fundamentally defined as the average value of the square of the deviation of each data point from the true mean value x which would be derived if an infinite number of data point were accumulated

Because we can not know x from a finite set of measurements, we use value X~ to calculate values for deviation. Use of the experimental

mean value will tend to reduce the average deviation and therefore result in smaller than normal variance. In statistical parlance, the number of degrees of freedom of the system has been reduced by one, and the -1 which appears in the denominator of Eq. (2.2-8) accounts for this self-minimizing effect.

The sample variance is an absolute measure of the amount of the scatter in the data, and does not, to first approximation, depend on the number of values in the data set. For example, of the data shown in Fig. 2.2-3 were extended by collecting an additional 20 values by the same process, we would not expect the sample variance calculated for the extended collection of 40 numbers to be substantially different from that shown in Fig. 2.2-3.

We can calculate the sample variance directly from the data distribution function F(x). Because Eq. (2.2-9) indicates that s2 is simply the average value of (X - we can write s2 as

c€

s2 = r (x-x12 - F(X) (2.2 - 10) r-0

An expansion of Eq. 2.2-10 will yield the well-known result

- 2 2 s = x - x 2 (2.2 - 1 I )

We can make two conclusions from above: a) Any set of data can be completely described by its frequency

distribution function F(x). b) Two properties of this frequency distribution function, the ex-

perimental mean and the sample variance. are particular important. Statistical Models Under certain conditions, we can predict the

dstribution function that will describe the results of many repeatitions of a given measurement. We will define a measurement as counting the number of successes resulting from a given number of trials. Each trail will be assumed to be a binary process in that onlv two results are possible: the trail either success or it is not a success. For everything that follows, we will assume that the probability of success is a constant for all :rials.

Table 2.2-2 gives three separate examples to show how these conditions apply to real situations. The third example indicates the basis for applying the theoretical framework that follows to the case of counting nuclear radiation events. In this case a trial is to observe a given radiation nucleus for a period of time t, the number of trials is equivalent to the number of nuclei in the sample under observation, and the measurement is to count those nuclei that undergo decay. We will identi- fy the probability of success of any one trial as p. In the experiment of

- ;.I radioactive decay, that probability is equal to ( I - e ), where i. is the decay constant of the radioactive sample.

Table 2.2-2 Example of Binary Processes

Tnal , Defin~tion of Success ' Probabil~ty of Success=p

I Tossing a coin "heads' 1 / 2 I Rolling a die 'a SLX" I

- - - . 1 / 6

Obscrv~ng a glven rad~oacuve I The nucleus decays d u n g

nucleus for a ume t Lhc obsemauon

(1) The Binomial Distribution. This is the most general model and is widely applicable to all constant-p processes. The binomial distribution is computational cumbersome in radioactive decay and is used only rarely in nuclear measurement because the number of nuclei is always very large in radioactive decay, In order to examine data acquired by counting a very short-lived radioisotope with a detector of high counting efficiency the binomial distribution must be used.

If n is the number of trials for which each trial has a success probability p, then the predicated probability of counting x success can be shown

-107-

n ! p(s) = --- p x (1 - p)n-x

(n - x)! x! p(x) is the predicated probability distribution function as given by the binomial distribution. and is defined only for integer value of n and x.

For the radioactive decay, following formula is given by

- Lt N(t) = No e

where No is the number of atoms of a given radioactive species existing at a starting time t = 0, N is the number of surviving atom at some time t, A is the disintegration constant. In the interval from t=O to t = t, the

- i~ number n of atomic disintegations would be given by N,(1 - e ).

Then the probability of any single atomic disintegration at some time t can be written as

- i t = I - e (2.2 - 13)

but the probability q of non-disintegrations, then the probability of the atomic disintegration of the atoms is as

- i t q = I - ~ = e

Substitute above p and q into the binomial disintegrations in the interval of time t can be written as

No ! - i t n - i.t No - n

p(n> = (No - n)! n! (1-e > ( e 1

We will show one example of an application of the binomial distribution. Imagine that we have an honest die so that the numbers 1 through 6 are all equally probable. Let us define a successful roll as one in which any of the numbers 3, 4, 5 or 6 appear. Because there are four of the six possible results, the individual probability of success p is equal to 4 / 6 or 0.667. We will now roll the die a total of ten times and record the number of rolls that result in success as defined above. The binomial distribution will now allow us to calculate the probability that exactly ?r

out of the ten trials will be successful, where x can vary between 0 and 10. Table 2.2-3 gives the value of the predicted probability distribution

from Eq. (2.2-1 2) For the parameters p = 2 / 3 and n - 10. The results are also plotted in Fig. 2.2-4.

Thus we can calculate the expected number of successes by multiplying the number of trial n by the probability p that any one trial will result in a success. In the example just discussed. an average number of successes is shown by

The mean value is obviously a very fundamental and important property of any predicated distribution.

(c) A predicted variance is a single parameter which can describe the amount of fluctuation predicted by a given distribution. We have already defined such a parameter, called the sample variance, for a set of experimental data as Eq. (2.2-10). By a n a l o a we will define a predicted variance 02 which will be a measure of the scatter about the mean predicted by a specific statistical model p ( ~ )

r - 0

conventionally, 02 is called the variance, we will emphasize the fact that it is associated with a predicted probability distribution function by call- ing it a predicted variance. It is also conventional to define the standard deviation as the square root of 02. The variance is in some sense a typical value of the square of the deviation from the mean. Therefore, a represents a typical value for deviation itself, hence is called as '{standard deviation''.

If we carry out the summation indicated in Eq. (2.2-18) for the specific case of p(x) given by the binomial distribution, the following result is obtained:

a - = n - p - ( 1 - p ) Because % = n p, Eq. 2.2-1 9 can be written as

o = u ' x * ( I - ~ ) (2.2 - 21) We have an expression which will give an immediate prediction of the amount of: fluctuation inherent in a given binomial distribution in terms of the basic parameters of the distribution, n and p, where % = n - p.

For the above example, we define success in such a way that

p = 2 / 3. and assumed 10 rolls of the die for the measurement of each group so thac n = 10. The predicted variance is given by

7

rs = n * p ( 1 - p) = 10 x 0.667 x 0.313 = 2.22 By taking the square root of the above value we get the predicted standard deviation:

a= \ / a2 =\12.22= 1.49 For this example, the mean value and the standard deviation are shown in Fig. 2.2-4.

(2) The Poisson Distribution. Many categories of binary process can be characterized by a low probability of success for each individual trial. For most nuclear counting experiments in which large numbers of nuclei are included in the sample, whereas a relatively small fraction of these give rise to recorded counts. Under these conditions the approximation that p < < 1 will holed and some mathematical simplification can be applied to the binomial distribution. I t can be shown that in this limicthe binomial distribution reduces to the form

Because p n = % holds for this distribution as well as for the binomial distribution, we can obtain:

which is the familiar form of the Poisson distribution. Important properties of the Poisson distribution are as follows: a. It is a normalized distribution:

n

I: p(x)= 1 (2.2 - 24) r - 0

b. The mean value of the distribution can be calculated by:

2

- s = 1 s - p ( x ) = p - n (2.2 - 25)

1-0

c. The predicted variance of the hstribution differs from that of the binomial and can be evaluated from this definition:

n 2 1

a I T: ( x - i ) - p (x )=pn (2.2 - 26) x - 0

This result can also be obtained at the condition p < < 1 from Eq.

(2.2-20). From Eq. (2.2-25), we can obtain: v&k;&a

L - G = x (2.2 - 27)

The predicted standard deviation is shown by a = dF (2.2 - 28)

The predicted standard deviation of any Poisson distribution is just the square root of the mean value which characterizes that same distribution. In the limit of p < < 1, Eq. (2.2-28) can also be obtained from Eq. (2.2-21).

We will now state the usage of the Poisson distribution with a example. We randomly select a group of 1000 people and define our measurement as counting the birthday among all member of that group. The measurement consists of 1000 trials, each of which is a success only if today is just his or her birthday. If we assume a random distribution of birthday in one year, then the probability of success p is equal to 1 / 365. Because p < < 1, this measurement can be described by the Poisson distribution. For this example, p = 1 / 365 = 0.00274, n = 1000, F = 2.74, a = 1.66. Thus, the distribution can be shown by

from wnich the value of Table 2.2-4 can be evaluated.

Table 2.2-4 Probability that x birthday will be observed from 1000 people in the same day

X P(X>

0 0.064

p(x) gives the predicted probability that x birthday will be observed from a random sampling of 1000 people. The value are plotted in Fig.. 2.2-5 and show that x = 2 is the most probable result. The mean value of 2.74 is also plotted in the figure, together with one value of the standard deviation on either side of the mean value. The distribution is centered about the mean value, but considerable asymmetry is evident for this low value of the mean.

(3) The Gaussian o r Normal Distribution The Poisson distribution is a mathematical simplification to the binomial distribution in the limit p < < 1. If the mean value of the distribution is large (for example, the mean value >20) Poisson distribution can further be simplified and then can lead to the Gaussian distribution:

1 (X - i)' - -

p(x) = e (2.2 - 29) t 2zx

This is again a pointwise distribution function defined only for integer value of x. I t shares the following properties with the Poisson distribution:

a. It is normalized:

b. The distribution is characterized by a single parameter x, which is given by the product np;

c. The predicted variance o' is given equal to the mean value F. The Ganssian distribution has two properties: a. The distribution is symmetric about the mean value X. Therefore

p(x) depends only on the absolute value of the deviance of any value x from the mean, defined as E = 1 x - ? 1 .

b. Because the mean value E is large, value of p(x) for adjacent value of x are not greatly different from each other, in other words, the distribution is slowly varying. These two properties suggest a recasting of the distribution as an explicit function of the deviation E (rather than of x) and as a continuous function (rather than a pointwise discrete function). These changes are accomplished by rewriting the Gaussian distribution as

where G(&)de is defined as the differential probability of observing a deviation in & about E . Because o = fi, Eq. (2.2-30) can be written as

2

2 c - -- ~~

G(E) = 7 e 2 b 2 V 2 n a

Comparing Eq. (2.2-30) with Eq. (2.2-29), a factor of 2 has entered in G(E) because there are two value of x for every value of the deviation E.

Now we return to the previous example of counting birthdays out of a group of randomly selected individuals, but now consider a group of 10000 people. For this example, p = 1 / 365 and n = 10000, so the predicted mean value of the distribution ?i= np = 27.4. Because the predicted mean is larger than 20, we can turn to the Gaussian distribution for the predicted distribution of the results of many measurements, each of which consist of counting the number of birthday found in a different group of 10000 people. The predicted probability of observing a specific count x is given by

and the predicted standard variance for the example is

a = 4 27.4 = 5.23 These results are shown graphically in Fig. 2.2-6a.

Fig. 2.2-6b shows the continuous form of the Gaussian distribution for the same example chosen to illustrate the discrete case. Comparing Fig. 2.2-6a and Fig. 2.2-6b, the scale factors for each abscissa are the same but the origin for Fig. 2.2-6b has been shifted to illustrate that a value of zero for deviation E corresponds to the position of the mean value ? on Fig. 2.2-6a. If a factor 2 difference in the relative ordinate scale is included, then the continuous distribution G(E) represents the smooth curve which connects the pointwise values plotted in Fig. 2.2-6a.

But the continuous form of the Gaussian distribution for often use is Eq. (2.2-29) (rather than Eq. (2.2-30)). Eq. (2.2-29) can be rewritten as

Because the Gaussian distribution is symmetric and continuous, we can use this property to conveniently evaluate the probability a(% _+ za) of the measurement value which appeared in the range of 2 + za is shown by

In order to conveniently calculate, we will change the distribution into the standardized form of the Gaussian distribution. Let

- X - X z --. --

G

then

In that case, the origin of scale z is in the position of the value ?r and the unit in a , that is, the size of a is equal to one. Eq. (2.2-33), the disuibu- tion of % = 0 and a = 1, is called as n(0,l) distribution and again as the standardized normal distribution.

A numerous calculations have been made using the standardized Gaussian distribution. For different value of z, we have evaluate the probability of the measurement values which appear in the range of 0 k z. For example, a t z = 1, the probability @ ( I ) is given by

= 0.683 (2.2 - 34) Eq. (2.2-34) shows that a group of the measurement value obeys the Gaussian distribution, and that 68.3 percent of all measurement values will deviate from the mean value by less than one value of the standard deviation. In the same way, we can calculate the probability of the measurement values which will deviate from the mean value by less than two or three values of the standard deviation. Table 2.2-5 gives the

-116-

probability for different value of z. Fig. 2.2-7 shows the graphical standard normal distribution.

Table 2.2-5 Value of the Probability @(z) Corresponding Values of z

APPLICATIONS OF STATISTICAL MODELS There are two major applications of measuring statistics in nuclear measurements. We will call the first application "Application AN. The first application involves the use of statistical analysis to determine whether a set of multiple measurements of the same physical quantity shows an amount of internal fluctuation which is consistent with statistical predictions. The usual purpose here is to determine whether a specific counting system is functioning normally. We will call the second application "Application B". In "Application B" we will examine the methods available to make a prediction about the uncertainty that should associate with single measurement to account for the unavoidable effects of statistical fluctuation.

Application A: Checkout of the counting system to see if observed

fluctuaticn are consistent with expected statistical fluctuation. Under keeping all experimental conditions as constant as possible we will periodically record a series of 20 to 50 successive counts from the detector system. By applying following analytical procedures, it can be determined whether the inherent fluctuation shown by these multiple measurements is consistent with a amount of fluctuation expected if statistical fluctuation is only origin. In this way abnormal amount of fluctuation can be detected which could indicate malfunctioning of some portion of the counting system.

Fig. 2.2-8 shows the chain of events that characterized this application. Properties of the experimental data are confined to the left half of the figure, whereas on the right-hand side are listed properties of an appropriate statistical model. Suppose that the data of N independent measurements is successive oneminute counts from a detector. The dis- eibution function F(x) as defined in Eq. (2.2-3) can be complied, then

mean value Y e and sample variance S' can be computed.

Next task is to match this experimental data with an appropriate statistical model. Almost universally we will want to match to either a Poisson or Gaussian distribution (depending on how large the mean value is). Either of the distribution is fully determined by it own mean value E. E e is only optimal estimate of the mean value for the distribu-

tion from which the data has been obtained. Setting E = Fe then pro-

vides the bridge from left to right in the figure, so that we have a fully specified statistical model. If we let p(x) represent the Poisson or Gaussian distribution with T7 = Kc, provided the statistical model

accurately describes the data distribution, then measured data distribution function F(x) should be an approximation to p(x). p(x) and F(x) are plotted in same figure to compare the shape and amplitude of the two dstribution.

A comparison of two function as shown above is only qualitative. I t is desirable to extract a single parameter from each distribution so that they can be compared quantitatively. The most fundamental parameter is the mean value, but these have already been compared and are the same as definition. A second parameter is the variance, and we can carry out the quantitative comparison by determining the predicted variance 2 of the statistical model and comparing with the measured sample va-

3.3.1 INFLUENCE OF FACTORS ON EXPOSURE IN RADIA- TION FIELD

For any regulation in radiation protection its basic principle is to decrease exposure. Either external exposure or internal exposure will be reduced to as low as reasonably achievable. If to remove the radiation is impossible or not practice it is necessary to consider the other measures to reach the aim of protecting the workers.

For a given radiation field the factors which determined the dose received by individual are as follows:

a. exposure time, b. distance from the radiation source, c. shielding. The time factor, more simply, is that the longer the staying time in

radiation field for individual, the higher the exposure received by individual. Some time, while dealing with an emergency event especially the job must be done in a strong radiation field. In this case it is necessary to plan the working procedural carefully in advance outside the radiation field so that the shortest time is used to perform the job. If the time is over long for one person performs this job and the exposure received will exceed the regulated limit from performing this job it must be to organize large numbers of person to perform this job, that is to say, using the small amount of exposure for many persons replaces the large amount of one person.

The radiation fluence rate decrease with the distance from the radiation source for this point. Suppose that the point considered is a penetration radiation source, then the radiation fluence rates of any cwo points are proportional inversely to the square of radial distance from these two points to radiation source. This law is only suitable for the dimension of source and detector is less than the distance between the source and detector, and this distance is measured in air or vacuum. For other non-point source radiation fluence rate will decrease with the in- crement of distance , but is not proportional inversely to square of distance.

While operating the radioactive source or doing the radiochemical experiment, using the long handle tools can greatly reduce the dose received by experimenters. But if the handle is long the operation is not flexible and the operation time may be enlarged, therefore the distance

-181-

protection and the rime protection musr be used overall planning and nimbly.

However, normally the worker faces the space problem or the worker must be close to the position of radiation source and then performs the job. Therefore, this case limits us to consider the problem from the distance. In this case one suitable shield material can be put in- to the place between the source and worker to eliminate the detriment. The selection of shield materials and its thickness depends on the type and energy of radiation.

In order to decreasing the detriment of external exposure as possible the factors of the time, distance and shielding may use separately and together.

While selecting the shielding materials the protection against the worker must be considered firstly. But the other factors influence the selection of materials also. For example, we must know if economy, if cumbersome, how large space the shield is allowed to occupy, we must know if the shield material has suitable construction strength. Besides, we must know if the material produces the poison and contamination and in order not to influence the precision measurement of various instruments we must know what level the radiation must be reduced to.

3.3.2 PROTECTTON OF VARIOUS RAYS There are external exposure and internal exposure for body expo-

sured to radiation. External exposure is exposure to radiation source outside the body for the body, and internal exposure is the exposure against body and induced by the radionuclides which come into body inside. The former is induced by the X, y rays, neutron beam, charged particles with high energy and P raysyxand the latter is induced by breathing, eating the radionuclides and absorbing the radionuclides through the good skin and the injurious skin. For these two kind of exposure mode there are completely different two protection means.

For protection of external exposure, normally one of these methods discussed in section 3.3-1 is adopted. The method of using shield materials between source and worker will be discussed in more detail in next section.

The protection method of internal exposure is completely different from the protection method of external exposure. For the protection method of internal exposure the most principal method is to reduce the

-1 82-

chance of radionuclides come into the body as possible. For example, laying down the reasonable health management system, ventilation, leaving with and operating radiation source in airtight case and individual protection. All of these measures are from this basic iciea.

For the radiation protection of reactors it is rare to concern the protection of external exposure of charged particles. Therefore. this problem will not be discussed in this book.

3.3.3 PROTECTION O F EXTERNAL EXPOSURE FOR GAMMA RAYS

The purposes of protection of external exposure are to control the radiation exposure against the body, and to make it keep all justifiable exposures as low as reasonably achievable. The dose calculation of external exposures is the basic of radiation protection and shielding calculation.

Dose Calculation of Point-Shape Gamma Source If the distance between the radiation source and one point in the radiation field is five times larger than the geometric dimension of radiation source itself, namely, the radiation source can be considered as the pointshape source, any shape radiation source can be considered as the summation of many point-shape sources, therefore, the case of point-shape source is only discussed here.

1. Kerma K The Kerma. K , is the quotient of dE,, by dm, where dE, is the

sum of initial kinetic energies of all the charged ionizing particles liberated by charged ionizing particles in a material of mass dm, namely

The concept of Kerma is only suitable for y rays, X rays and neutron. The initial kinetic e n e r g includes possible nuclear reactions, whether these kinetic energies deposit at any place.

The unit of Kerma is same with the unit of absorbed dose, i.e. J / kg. The special unit is Gy.

According to the definition of inass energy transfer coefficient. The Kerma of medium which is in position of radiation field of non-charged monoenergetic particles with fluence @ and energy E is

pen e where fx -- E ( - - - )a , , (T) is called exposure factor which repre- P a

sents the exposure related to unity photon fluence. Its unit is C kg-' m2. Table 3.3-1 lists the exposure factor, f,, for different energy gamma rays.

Table 3.3-1 Value of Exposure Factor

1 photon energy (MeV) I f, (C kg-' - m2) phocon energy (MeV) . f, (C . kg-' - m2) I

According to Eq. (3.3-8) if the fluence of gamma rays with energy E is known, as long as the mass energy transfer coefficient of gamma rays in air, (pen / P ) ~ ~ , can be looked up the exposure, X, can be calculated. The addend Table 1 . I lists the mass energy transfer coefficient of gamma rays with various energy in air. -

The exposure rate is

Example gamma ray energy emitted from ' 3 7 ~ s source is 0.662>1cV, the gamma ray fluence measured at point 1 meter from the source is 1 x 10'm--' s-', evaluating the exposure rate at this point.

Solving According ro Eq. (3.3-93, rne exposure rate is

From Table 3.3- 1 the f,eq uals 9.2 19 x 10-"C kg-' m2,then the exposure rate at the point I meter from source is

3. exposure rate constant The fluence rate , cp, for monoenergetic point-shape gamma radia-

tion source is

where A = activity of radiation source, Bq n = average emitted photon number for one disintegration r = distance from the radioactive source, m Normally, the gamma source does not only emit the gamma rays

with one kind of energy. Suppose that one gamma source emits gamma rays with K kinds of energy. The exposure rate at point whose distance from source is r should be the sum of exposure rates induced by the various energy at that point

It is convenient to calculate that defining the part of Eq. (3.3-1 1 ) as the exposure rate constant

where

-1 86-

ni E, =product of number and energy of i-th photon by considered nuclide in one disintegration. its unit is MeV Fjq-' s-'

' c n

(y =mass energy absorption coefficient of i-th photon in air.

mZ kg-' =exposure rate constant, C m2 kg-' B ~ - ' s-I.

Table 3.3-2 lists the values, T, of exposure rate constant for some radionuclides.

Table 3.3-2 I- Value of Radionuclides with y radioactivity

Nuclide T (C m2 kg-' Bq s-') 2 4 ~ a 3.532-18

Substituting Eq. (3.3-12) into Eq. (3.3-1 1) the exposure rate for the

-187-

source with activity 4 and at the point whose distance from source is s m is

1

- i The unit of exposure rate is C * kgui s . 4. air absorbed dose rate for gamma point-shape source Under the condition of charged particle equilibrium, from the defi-

nitions of absorbed dose and mass energy absorption coefficient the following equation can be gotten

where, ul is the energy fluence. pen/ p is the mass energy absorption coefficient of certain material for monoenergetic gamma rays.

It is known from above equation that while the energy fluence is constant the absorbed dose is proportional to mass energy absorption coefficient, pen / p , of certain material, therefore

p cn

Dl (-I1

- P

D 2 pen (- I 2

P The subscript, 1 and 2, represents the material one and material two respectively. Therefore, as long as the absorbed dose in one material is known the absorbed dose in other material can be gotten through the Eq. (3.3-1 5) under the condition of charged particle equilibrium.

According to Eq. (3.3-7) and Eq. (3.3-14) the relation between the exposure and absorbed dose in air under the condition of charged particle eqllibrium is

where D, is the absorbed dose in air at the same point. In the same way the relation between the air absorbed dose rate gamma rays and exposure rate can be gotten

Substituting Eq. (3.3-16) into Eq. (3.3-15) the following equation can bc zotten

where Dm = absorbed dose to be evaluated in certain material at the same

point in air, Gy X =exposure, C kg-' f m = transforming f ~ t o r from exposure in C kg-' to absorbed

dose in Gy, its unit is J C-'. Table 3.3-3 lists the fm values of water, soft-tissue and bone for

different energy. From Eq. (3.3-18), the relation between the absorbed dose rate of

material m at the same point in air and the exposure rate is given by

Example evaluating the exposure rate at the point from the source, meter for 6 0 ~ o source with activity 1 Ci, and the absorbed dose rate in air and the muscle beneath the skin.

Solving A = 1Ci = 3.7 x l o l b q , r = I m. From Table 3.3-2, l- =

2.50 x IO- '~C m2 / kg Bq s for 6 0 ~ o . Substituting these values into Eq. (3.3-13), then , , - A , \ -

- .-. 7

From Eq. (3.3-17) nL'

I

r , \ ;) a = 33.$5 x 9.25 x 10 -' , \,

- 6 = 3.13 x 10 G y / s

- 2 = 1.13 x 10 G y / h.

" 60 From Table 3.3-3 fmus,,,= 37.29 for Co. Substituting l,us,!, and X

into Eq. (3.3-19), then

=3.35 x I O - ~ G ~ / S

= 1.24 x l o 2 G Y / h . 5. absorbed dose rate and particle fluence for point-shape source From Eq. (3.3-14), the following equation can be gotten

where 0

D a =absorbed dose rate of gamma rays in air, Gy s-'

cp = photon fluence rate, m-' s-' E = photon energy, J. Narrow Beam Gamma Ray Attenuation Law The attenuation of

gamma rays in material obeys the simple exponential law, i.e.,

where No = photon number before passing through substance layer N =photon number after passing through substance layer d = thickness of substance layer p= linear attenuation coefficient of gamma rays in this substance. The characteristic of gamma rays is described by parameter p. The

physical meaning of parameter p has been described in first chapter, and pointed out that parameter, p, is the function of photon energy , E, and the atomic number, z, of substance.

Eq. (3.3-21) can be written as follows

where p = substance density, kg m-3

dm = d p. mass thickness of substance, kg m -2

if = mass attenua~ion coefficient 01. substance i b r given photon en- ?

ergy, rn' kg-' x

If decreasing the gamma rays for given energy to -- , the N o

thickness dm,, is needed for the substance one with mass attenuation

P coefficient (p ) , . and the thickness 4,. for substance two with (- ), . P .

.- P -

Then

d m,l P I ~- (3.3 - 23)

dm, P 2

The equivalent concrete thickness of brick and soil can be estimated from Eq. (3.3-23).

Broad Beam Gamma Rav Attenuation Law The narrow beam attenuation law is a simplification case. If the Compton scattering occurs the photon scattered may pass out of the substance. The energy and direction of photon changes, however the photon has not been absorbed truly by substance.

The radiation encountered in radiation protection is broad beam radiation for the most part, schemed in Fig. 3.3-1.

In this case the photon subjected to multiple scattering may still pass out the substance and reaches the point of interest. Therefore, there are not only the uncollided incident photons, but also photons undergone multiple scattering.

In order to consider the influence of multiple scattering, one correc- tion factor is introduced a t the right side of Eq. (3.3-21) for use with correcting the narrow beam attenuation law, i-e.,

where B is called buildup factor. There are different buildup factors for different radiation

quantities. The B in Eq. (3.3-24) is the buildup factor of photon number or fluence rate. In practical usage the exposure buildup factor, B,, is often used, i.e.,

. 0

where X and X , is the exposure rate with and without shield layer of

thickness d at the same point in radiation field respectively

Table 3.3-4 Parameter of Calculating Exposure Buildup Factor of Isotropic Point Source by Taylor's Formula

energy I ) energy I rnatenal A, -31, z2 material,

,MeV MeV I A, -2,

For isotropical point source the relation between buildup factor, B,, of medium and substance thickness, pd, can be expressed approximately by formulae. One of the more commonly used is Taylor's formula

where 11 =linear attenuation coefficient, m-' d = thickness of shield substance, m

For glven materials a , , a, and A , IS oniy <he function of gamma ray energy, the relative values can be seen In Table 3.3-4.

Parameters Used in Shield~ng Calculation The absorbed dose - - - - - -- - - - . . . - -. - - -- - - - -

rate or dose equivalent rate a t one point in radiation field is proportion- *

a1 to exposure rate, X , a t the same point. Therefore, after the broad

beam gamma rays pass out the shield layer the attenuation of absorbed dose rate or dose equivalent rate can be described in terms of exposure buildup factor, Bx(E,pd). From Eq. (3.3-25)

where

H , =dose equivalent rate a t one point without shield layer

H (d) = dose equivalent rate at same point with shield layer of thick-

ness d . 1. attenuation multiple Attenuation multiple is defined as

It represents the ratio of dose equivalent rate, H , , at one point in r aha -

tion field without shielding layer to dose equivalent rate, H(d), a t the

same point with shielding layer of thickness d, namely, represents the shielding power of this shielding material against the radiation. Attenuation multiple dose not have the hmension.

For given photon energy and material the linear attenuation coefficient, p, and buildup factor, B,(E,pd) is defined also. From the Eq. (3.3-28) the relation between K and d can be defined.

2. shielding transmission ratio Shielding transmission ratio, q, is defined as

. H (d) B ( E , P ~ )

- r l = , - 4

HO e

It represents that the power of radiation transmits the shielding material. The shielding transmission ratio does not have dimension, and is the reciprocal of attenuation multiple-There are many shielding transmission ratio curves for given shielding material and gamma ray energies.

3. half-value layer and tenth-value layer The definition of half-value layer, A! , is the shielding layer thick-

2

ness which can attenuate the incident photon number to half of original. The definition of tenth-value layer, A L , is the shielding layer

10

thickness which can attenuate the incident photon number to tenth of original.

The relation between A I and A 1 is 2 10

Because the attenuation of broad beam gamma rays in shielding medium is not a simple exponential law, the A1 and A L value is not a

2 10

constant in shielding medium for given radiation, they have slightly change with increasing of attenuation multiple K, and are used for assessing the shielding power of materials to radiation and calculation approximately shielding layer.

Shielding Calculation of Gamma Point Radiation Source The purpose of shielding calculation is as follows: what setting sufficient

shielding layer makes the sum of dose equivalent rate, H(d), induced for

various radiation at interest point in radiation field not exceeded

precondition control level, H , , i.e.,

If the primary gamma ray beam is the main factor of determining the shielchng layer thickness, then

The shielding transmission ratio for gamma rays pass through the shielding layer of thickness d is

or the attenuation multiple of shielding layer of thickness d for given gamma ray beam is

where A = activity of gamma source, Bq r =exposure rate constant, C kg-' m2 B ~ - ' s - 1

q =area occupancy factor, during the start on of radiation source the time fraction of worker occupancy a t the outside of shielding layer. In shielding design according to whole occupancy, partial occupancy and accidental occupancy a t the interest point the q can take 1, 1 / 4 and 1 / 16 respectively

r = distance from interest point to point source, m 0

H , =control level of dose equivalent rate a t interest point.

Either q is calculated o r K calculated, then the shielding layer thickness, d, can be gotten by looking up relative curve.

Exampic a 6 0 ~ o source with activity 3.7 x 10'"q is put into a lead container, requiring that the dose equivalent rate on the container surface is less than 2 x Sv h-'. The distance, r, from container surface to source is 25 cm. Evaluating the shielding layer thickness of container.

Solving knowing T = 2.503 x lo-'' C kg-' m2 B ~ - ' s-' from Table 3.3-2, taking q = 1. From Eq. (3.3-35)

6 =0.89 x 10

Looking up relation curve the lead shielding layer thickness is 23.7 cm, it is taken 24 cm in practice.

Material often Used in Shielding Gamma Rays While any material is chosen to be used of f i r shielding materialthe protection characteristic, construction property, stability and economic cost of material must be considered sytheticly from optimization principle. The shielding materials of shielding gamma rays are often used as follows.

Lead Its atomic number is 82 and its density is 11.34 g / cm3. Corrosion-resistant characteristic is well. It is damaged easily under exposure to radiation. Lead has high attenuation power for low energy and high energy gamma rays, therefore it is ideal material for shielding from gamma rays. But, its attenuation power is most weak for gamma rays with energy from 1 MeV to few MeV. Besides, the short coming of lead is its high cost, and is not temperature-resistant. I t is of used for making lead container, portable shielding and lead brick. While making the larger container and equipment the steel material must be used as the construction bone otherwise it can collapse because of weight itself.

Iron Its atomic number is 26 and density is 7.89 g / cm3 (for steel) or 7.2 g / cm3 (for iron). Its cost is low and available easily. Shielding characteristic is worse than that of lead. Normally, if the attenuation multiple is same the weight used of iron is about 30% weigh- tier than that of lead. The mechanism strength is very high. It is easy fabrication, and used for making shielding iron doors and covering plates of grooves.

Concrete The effective atomic number of normal concrete (cement 1 : sand 2 : break stone 2 : water 0.5) is 18 and density is 2.3 g / cm3. Its cost is very cheap, and has good construction characteristic. It is often used for fixed shielding layer in engineering. Sometimes, in order to decrease the thickness and volume of shielding layer, the high density concrete (called heavy concrete) is used. The density of heavy concrete can reach to 6.9 g / cm3, but its cost is very high.

Water The effective atomic number is 7.4 and density is 1 g / cm3. The characteristic for shielding from gamma ray is worse than that of three materials above mentioned. But it possesses with special

advantage. Its transparency is well and some materials can be put into it at will. Therefore, the water is often in form of water well or water pool to contain the gamma radiation source. While there is resolvable salt in water, the water can be made radiolysis under the action of strong exposure. Therefore, the deionization water is suitable for purpose of shielding.

The brick, sand and soil used in build may be used of for shielding from gamma rays. It is not special to use them for shielding, but they play a role in shielding from gamma rays objectively when they are used in build.

In some special cases, in order to decrease the weight and volume some expensiver, high density metals, for example, tungsten and uranium, are used as the partial shielding.

3.3.4 PROTECTION OF NEUTRON EXTERNAL EXPOSURE The neutron radiation source can be classified as four kinds by pro-

duction means of neutron, namely, radionuclide neutron source, accel- erator neutron source, reactor neutron source and plasma neutron source. The neutron sources can be classified as two kinds by its enera , namely, monoenergetic neutron source and polyenergetic neutron source. The neutron sources, for example, the radionuclide neutron source, have often concomitant gamma radiation.

Reactor Neutron Source The neutrons of reactor neutron source come from the self-sustaining nuclear chain reaction. The property of this neutron source is that: the neutrons are produced in a large number in unit time and the neutron spectrum distribution is very wide (from 0.025 eV to 17 MeV).

There are many kinds of radiation released from the reactor. The kinds and processes released from reactor are schemed in Fig. 3.3-2. Be- cause the transparent power of the neutron and gamma radiation is the most strong in these radiation as long as the shielding layer thickness is so sufficient that the neutron and gamma radiation can be reduced below the control level required, then the shielding problem of other radiations can be ignorable. For the shielding design of radiation the important radiation in nuclear reaction is as follows: fission prompt neutrons. fission prompt gamma rays, the gamma radiations induced by inelastic scattering of fast neutrons and the capture gamma radiations emitted from the reactor construction materials absorbed thermal neutrons.

-1 98-

Property of Interaction Between Neutrons and Body Tissue It is --- ..-.----ppp--p--pp----. - - - -----

important to consider the interaction between the neutrons and elements which consist of the body tissue.

weight of hydrogen, carbon, nitrogen and oxygen is 95% weight of body in body tissue. The number of hydrogen atom is over 60%of total number of the body tissue atom.

The fast neutrons gradually transfer the energy to tissue and then are moderated through the elastic and inelastic scattering with the nuclei of hydrogen, carbon, nitrogen and oxygen. The thermal neutrons moderated are absorbed by tissue through the reactions of ' ~ ( n , ~ ) 'H and

' 4 ~ ( n , p ) I4c. The energy of recoil protons (0.6MeV) and gamma rays released from reactions are finally absorbed by the tissue.

Calculation of Neutron Dose The following two methods are -- -

adopted in calculating the neutron dose. 1. Calculating of Kerma The Kerma of monoenergetic neutron can be expressed as follows

where

tr f k = (-) - E, Kerma factor of neutron, expressing the Kerma per

P unity neutron fluence

cD = monoenergetic neutron fluence in radiation field. If the Kerma of certain substance in neutron radiation field, K,,

has been known, then the Kerma of a small block tissue exposed to radiation at same point in radiation field, KT, can be gotten from following equation

If the charged particle equilibrium condition is satisfied, then absorbed dose of relative tissue is as follows

3 7)

the

After getting the monoenergetic neutron fuence from partial measurement or calculation and looking up the value of f, from appendant table 3.3 corresponding to the relative energy, then the Kerma, K, can be calculated. The value of Kerma can be seen as the approximate value of the absorbed dose. For the neutron whose energy is below 30 MeV the error introduced by this approximation can be ignored.

Table 3.3-5 Neutron Radiation Weighting Factor, w,, Neutron Dose Equivalent Factor, f,.,,

and Relative Neutron Fluence Rate (25,uSv h-') neutron radiatron dose neutron

En, McV weighting cquivalcnt factor , flucncc rate ,

factor w, (fAp-x 10-"SV . m') (CII-' . 5-') -. -- -

2.5 x lo-' 2 1.068 650.2

Example blood specimen of 1 ml is out into the nylon tube whose wall thickness is 3 mrn. Calculating the neutron absorbed dose for unity neutron fluence. The neutron energy is 14 MeV.

Solving while En = 14MeV, the range of recoil proton in nylon is 2 mm, the charged particle equilibrium condition can be satisfied for the wall thickness of nylon is 3 mm. Therefore, the absorbed dose of blood is equal the Kerma of blood, i.e.,

tr

- (7 ) b l d

- (fK)nylon pu (7 ) nylon

From the addend Table 3.3 (f,),,,,, = 0.658 x 1 0-lo Gy / crn2,

prn prn 0= lcm-2. The ratio of (--)blo, to (T)nylon equals 1 approximately. P

Therefore, the absorbed dose equals Kerma for blood, i.e.,

blood = blood = @ (f K 'nylon

= 1 x 0.658 x 10-lo

=. 6.58 x 10-l1 Gy 2. Calculation of Dose Equivalent The radiation weighting factors of neutron for different energy, w,,

and the dose equivalent factors for unity neutron fluence, f,,, are listed in Table 3.3-5. The neutron fluence rate relative with 25pSv h-', cp,,

are listed in the same table also. The neutron dose equivalent rate can be calculated by following equation

Example For the ' 'OPO-B~ neutron source with activity 3.7 x

I O " B ~ , calculatingthe neutron dose equivalent rate at the point from source 1 meter

Solving From the Table 3.3-12 getting emitting rate of neutron source

From Table 3.3-5 the f,,, of " V o - ~ e neutron source is 35.5 x

10-"SV - m2. According Eq. (3.3-39)

Attenuation ~ of Neutron - in Shielding ~ Laycr As viewed from neutron shielding, firstly, the fast neutrons become inro thermal neutrons through the elastic scattering and inelastic scattering between the substance and fast neutrons, secondly. the thermal neutrons are absorbed by substance.

The inelastic scattering is with threshold energy. The neutron inelastic scattering whose energy is higher than threshold energy may occur. For nuclei of mass number from 100 to 150, its level density near the ground state is very large and the level width is about 0.1 MeV. However, for light nuclei the corresponding level width is about 1 MeV. Once the inelastic scattering between neutrons and nuclei occurs, the neutron energy will be reduced in large amplitude. When neutron energy reduced to the energy which is less than that of threshold induced the inelastic scattering the neutrons rely on only elastic scattering to reduce its energy.

In elastic scattering the lighter the nuclei come into collision with neutrons, the more energy the neutron transfer to recoil nuclei. While neutrons come into collision with hydrogen nuclei the recoil nuclei get the energy in most. In one elastic scattering between neutron and hydrogen nucleus the neutron transfers its half energy to recoil proton on an average, even all energy at times. Therefore, the hydrogen is the best moderator for fast neutron with energy about 1 MeV. The average numbers of elastic scattering needed for reducing the neutron energy from 1 MeV to 0.025 eV in several substances are listed in Table 3.3-6.

Table 3.3-6 Average Numbers of Elastic Scattering Needed for Reducing the Neutr.on Energy from 1 MeV to 0.025 eV in Substance

H D He B , B e C 0 Fe U - -- - -- - - - -- - - - .- - - - - - - - - -

mass number 1 2 4 7 9 12 16 56 2381 I collision number 18 24 41 65 84 ' 1 1 1 146 485 20881

Table 3.3-7 Thermal Neutron Absorption Cross Section of Some Substance and Relative Maximum Energy of

Capture Gamma Radiation Thermal Maximum energy Thermal Max~mum energy

Element Neutron ( n , ~ ) of capture gamma Element ; Neutron (n.7) of capture gamma

cross section (b) nd~atlon (MeV) I cross sectior, (b) radiat~on (MeV)

H 0.332 2.23 Co 37.0 7.49 - - -- - P - -- P.. - -- log~) 3837 0.478 hTi ' 4.8 I 9.00

1) Cross section of (n,z) reaction a n d its concomitant gamma radiation energy

Although the thermal neutrons can be absorbed by substance, every one is not suitable for absorbin5 the thermal neutrons. After many materials absorbed the thermal neutron the energy capture gamma radiation are emitted. The thermal neutron absorption cross section for some materials and relative maximum energy of capture gamma radiation are given in Table 3.3-7.

Attenuation of Neutron Current in Shielding It is similar to -- - - - - -. - -. -- -- -- - -- - -- - - .. -

attenuation of gamma radiation in shielding, the attenuation of narrow neutron current in shielding layer obeys simple exponential law also, i.e.,

where cpd=neutron fluence rate at one point in radiation field without

shielding layer $,=neutron fluence rate at the same point in radiation field with

shielding layer C = total macroscopic cross section of shielding material for inci-

dent neutrons. In the same way the attenuation law of broad neutron current is

where B, is the buildup factor of broad neutron current. It is used to characterize the increasing fraction of neutron fluence rate at interest point behind shielding layer because of the polyscattering neutron produced in shielding layer.

Removal Cross Section Method for Calculating Attenuation of Broad Beam Neutron The basic content of removal cross section is as follows. To choose suitable shielding material makes the neutrons can be rapidly moderated in very short distance once the neutrons undergo scattering and ensures that the moderated neutrons are absorbed in shielding layer. That is to say, the neutrons undergone scattering are effective "removed" from the neutron beam of penetrating out the shielding layer, and the neutrons of penetrating out the shielding layer are those neutrons undergone interaction in shielding layer. In this case, the attenuation of neutrons? even if broad beam neutrons, satisfies the simple exponential law.

Using removal cross section the shielding material must satisfy fol-

lowing conditions. a. That the shielding layer is enough thick makes the dose equiva-

lent behind the shielding layer from the attributing to neutrons with very strong penetrating power in neutron beam.

b. Shielding layer contains the medium heavy o r heavy materials, for example. iron o r lead. Therefore. the neutron energy can be rapidly reduced to about 1 MeV.

c. The shielding layer must be enough hydrogenous so that it ensures that the neutrons can be made to reduce the energy from 1 MeV to thermal energy in very short distance and can be absorbed in shielding layer.

If the conditions mentioned above are satisfied the attenuation of broad beam neutron in shielding layer can be described by following equation

where 0

cp,, H , = neutron fluence rate and dose equivalent rate a t one

point in radiation field without shielding layer 0

cp,(d), H ,(d)=neutron fluence rate and dose equivalent rate a t

the same point in radiation field with shielding layer whose thickness is d

C,=macroscopic removal cross section of shielding material for neutron

d = thickness of shielding layer. For fission spectrum neutrons the relation between macroscopic

removal cross section and microscopic removal cross section is

where Z, = macroscopic removal cross section, cm M A = molar mass of nuclide, g mol-' p = material density, g -

If shielding material is mixture or compound, then total macroscopic ;emoval cross section is

Z , = Z N , a,, I

where Qi = weight percent for i-th nuclide in mixture M =molar mass of i-th nuclide, g mol-' *, a =microscopic removal cross section of i-th nuclide

A i

Qi - number of i-th nuclide in unit volume. 0.602 p - -

M Ai

Table 3.3-8 to Table 3.3-10 list macroscopic removal cross section of some materials and elements for fission spectrum neutrons. Although these values are for fission spectrum neutrons, they can be used to treat the shielding problem of radionuclide neutron source. Table 3.3-1 1 lists the microscopic removal cross section of some materials for different energy neutron.

Table 3.3-8 Macroscopic Removal Cross Section for Fission Spectrum Neutrons

Material Soil Contained Water 10%

Normal Concrete I 0.089

Polyethylene I 0.123 Iron 0.1576

Table 3.3-9 Macroscopic Removal Cross Section of Some Elements for Fission Spectrum Neutron

Macroscopic Dcnsity p

Z Elcmcnt A ! rcrnoval cross -3 , g - c m !

! section E,, cm-' - ----- .-

3 1 Li 6.940 , 0.534 ~ 0.0449

Tabie 3.3-10 Macroscopic Removal Cross Section of Some Compound for Fission Spectrum Neutron

Compound Density, g cmU2 i C, , an-'

H20 1 .oo 0.10 1 I D2O 1.10

-pp-pp

0.091 3 I 0.952

- - - 0.109

steel (1 % Carbon) 7.83 I , 0.1 63 --

sand 2.20 I 0.052

t- (CHJn petroleum 1

C8HH -~

FeB

I gravel 0 .O 92 I

LJ G-d

UO

G

.a L-d

U *

v3

2

r= 2 .g

C

d

the incident direction reduce to tenth of initial tluence is called rznrh-value layer. denoted by A I .

10

The curve of relative parameters for different neutron energy and dfferent materials can be found in relative books.

Neutron Shielding Calculation - - -- - - - - - -- - -- - -

The neutron shielding calculation of reactors is a complex problem. It is necessary to use complex code and computer for getting satisfactory results.

The shielding calculation of radionuclide is only discussed here from which the normal procedure of shielding calculation can be understood.

For hydrogenous materials whose thickness is not less than 20 cm, for example, water, paraffinum and polyethylene, the B, can take as 5. For lead B, = 3.5 and for iron B, = 2.6.

The shielding thickness, which can make the neutron tluence rate reduce to cp, (m-2 s-'), can be calculated from following equation

- Z , d v r = v r O B n e < V L (3.3 - 46)

Therefore

where d = thickness of shielding layer, cm C, =macroscopic removal cross section of shielding material, cm-' A =activity of radionuclide of neutron source, Bq y = yield of radionuclide neutron source, ~ q - ' s-' B, =buildup factor of neutron q =area occupancy factor r = distance from interest point to source, cm. Example Using special automobile to transport one Po-Be

neutron source whose activity is A = 3.7 x 10" Bq. The source is put in- to a shield tank made by paraffinum. Requiring dose equivalent rate at the position which driver sits on is below 2.5 pSv h-'. How many thickness is the paraffinum shielding needed.

Solving From Table 3.3-12 getting Ay = 2.5 x lo7 s-'. From Table 3.3-5 cp,= 1.97 cm-2 s-' for 2.5 pSv / h. From Table 3.3-8 get-

ting the macroscopic removal crcss section of paraffinum 1,=0.118 cm-'. Let B, = 5 and q - 1 .

From Eq. (3.3-47)

Example A 23!?u-~e neutron source is put at the central of a water bucket, shown in Fig. 3.3-3. Requiring the neutron fluence rate on the water surface of top is loss than 5.91 cm-2 s-'. How many depth is the water bucket needed.

Solving From Eq. (3.3-47)

Table 3.3-12 Neutron Source Characteristic cf Radionuclide

radio- rcacrlon Half- m a x ~ m u m - avcragc neutron neutron concomitant

name 1 nudlde ! type life, I energy of , energy of , ycld y, spmrum gamma

I T I neutron, neutron, ( x 10~s-' rad1a:ion I

I MeV ' MeV - Ba-' ! i Na-Be '%a n 15 Oh 0 83 3.51 3 . 7 6 ~ 10' (a) (b) 1 S b & 1 12%b , (.?,XI) 60.d , 0.029 1 5.14 1.33 x 10' (a) (b) 1

I Po-Be ' "90 (up) ' 138 4d 10 87 4.2 67.6 0.103 (c) (d) 1 I Ra-Be 2 " ~ a (a,n) 1620a 1 13.08 4.0 405 155 (c) (b) I 1 Pu-Be ' ( a . 87.753 11.3 4.5 54.1 11 .29 (c) (d) I ( P U - ~ e I (a,=) 243% 10 74 4 1 43 2 4 39 (c) (d) 1 1 Am-Be / "'.Am ( a s ) I 432a 11 5 4.5 541 < 2 J 8 (c) (4 I

(a) rnonoenerge~ic

(b) very strong

(c) continuous

(d) very low

From Table 3.3-8 getting macroscopic removal cross section Z, = 1.103-'. From Table 3.3-12 getting Ay = 1.6 x IO-~S-'. Let B, = 5 and q = 1. Then

Let Q , = e 0.103d 1 and Q2 = 7 x 1.0772 x lo6. Solving above two equa- d

tions in terms of graphic solution method, the d value can be gotten d = 56.5 cm

According to the attenuation of neutrons in materials many attenuation curves of some shielding materials have been drawn. From these curves the materials thickness needed can be looked up directly.

Materials often Used in Shielding ~ Neutron While any material and its thickness is chosen to be used for shielding material the protection characteristic, construction property, stability and economic cost of material must be considered sytheticly from optimization principle.

The optimum combination of shielding materials is that there is a certain amount of elements with atom number above medium and an amount of light element, especially hydrogen. The hydrogenous amount of some materials often used is given in Table 3.3-1 3. If an amount of '9 and 6 ~ i is added in shielding material the thermal neutron can be

effective absorbed and the capture gamma radiation can be decreased, therefore, the dimension of shielding layer can become thin.

Table 3.3-13 Hydrogenous Amount of Some often Used Shield- ing Materials

Material ~ Chemical Hydrogenous : composition ; amount

-~ ~. ~ 1 atom/cm3 .

*2O Water 6.7 x loZ2-

Paraffinum .- C , p 2 2 7.87 x Polyethylene (CH?)n 7.92 x 1 02'

Polyvinyl chloride ' (CH2CHC1), 4.1 x loZZ Organic glass

-- (C4H802)n 5.7 x loZ2 Gypsum -- C a S 0 4 * 2 H 2 0

--- 3.25 x 10"

Kaolin AI,O, - 2Si0, . ZH,O 2.42 x lo?' 9 2 % polyelhylene -8% B,C (CH,)-+B,C 7.68 x 10"

The shielding materials often used are as follows.

Water Water contains a lot of hydrogen. It is a very good kind of neutron moderator. The thermal neutron capture cross section of' hydrogen is 332b and the energy of capture gamma radiation is 2.2 MeV. The hydrogen contained in shielding. to moderate the neutrons. can capture the thermal neutrons in shielding. The water has no construction characteristic. It is often filled in a variety of container and made up of water doors and water boxes.

Concrete The normal concrete density is 2.3 g / cm3. I t is mixture of many elements shown in Table 3.3-14. It contains light element, heavy element and a certain amount of moisture content. Therefore, it has better shielding characteristic for neutrons and gamma rays. The concrete possesses good construction characteristic and is often used of for fixed shield. But when it is used for long term the moisture content can be lost, therefore, the protection characteristic for neutrons will be reduced.

Table 3.3-14 Element Composition of Concrete, x lo2' atom cm-) c1;

Carbonaceous concrete - - - - - -. - -- - . . - -- -- -

, Silanccous Element I 3.0%" 5.5%Q 8 . 0 % ~ concrete 5.0%~ ,

Q The density of concrete consisted of elements listed in table is 2.3 g cm-3

3 moisture content

Paraffinum It contains a lot of hydrogen, its cost is cheap and easily forming. It is good moderator of neutrons. But while air temperature is higher it is easily soften and out of shape. While air temperature is low it shrink and crack easily. Its construction characteristic is not good. I t has not fire resistance and is burnable. Its protection characteristic for gamma rays is poor. Therefore, it is often used with other mate-

rials. Polyethylene It contains plentiful hydrogen. It is better protec-

tion material for neutron. It is easy processing and shaping, but it is easy sof~en in high temperature. It is often used with mixing with other construction material.

Clay It contains larger water and is very cheap material. In order to use fully its protection characteristic, some times the neutron generating assemblies are built in basement.

Lithium and boron The thermal neutron absorption cross section of lithium and boron is 940b and 3837b respectively. The gamma radiation released after the neutrons are absorbed by lithium is very less and can be ignored. Although 95% capture events of boron release 0.47 MeV gamma radiations, they are shielded easily. Without special requirement the boric acid or borax with cheap cost is used. While requiring to decrease the shielding volume the B,C,which contains higher amount of boron, can be used. The lithium borate can be chosen for special case of requiring low yield of gamma radiation.

ul L-, *

i' V

; 2

O ru

V1 h

0 cu a

0%

0

FI w

.2 In

+-.a*

w Ccu

nk

w

a

fission I -

I (scattering)

-J rays 7- slow neutron

prompt fission compound nuclei inelastic scattering -frays in excitated state Y rays

radioactivite nuclei capture 1 "rays $particks decay

y rays brems strablung

I (absorption) (If Be, =H exist)

Compton scattering photoeleciric effect eleciron pair A I production neutron

7 -

A A A i recoil electrons photoelectrons positron electron

- brems strablung - annihilat.ion radiation

Fig 3.3-2 ionization radiation released in reactors

water.

neutron source

Fig 3.3-3 Schematic of Z39Pu-Besource in water shielding

Ap

pen

ded

tab

le 3

.1

Ker

ma

fact

oro

f n

eutr

on

in

som

esu

bst

ance

s fk

,Gy

.cm

2

. L ti

pp

roxi

mat

i bo

ne

I ln

usc

le

rele

ron

co

E./

M=

V

AE

.IM

IV

I ,, 1 dry

air

I wat

er

Kth

igbo

ne) -

.--

. nyl

on

org

anic

. I gl

ass

0.4

84

-11

0

.57

1-1

1

0.2

96

-11

0

.28

1-1

1

0.2

80

-I1

0

.11

2-1

1

04

-1

1

0.6

94

-11

0,1

15

-10

0

.20

2-1

0

0.3

57

-10

0

.61

2-1

0

0.1

04

-09

0

.17

9-0

9

0.2

97

-09

0

.46

2-0

9

0.E

58

--0

9

0.6

77

-04

0

.69

5--

09

0

.61

4-0

9

0.8

31

--0

9

0.60

1-:0

9 0

.70

1-0

9

0.7

40

-09

-0.1

29

-11

0

1

09

-11

0

.10

3-1

1

0.1

14

-11

0.1

50

.-1

1

0.2

30

-ll

0.3

87

-11

0

.65

0-1

1

0.1

12

--1

0

0.2

00

-10

0

.35

6-1

0

O.l

O6

-12

0

.17

8-1

2

0.3

08

-12

0

.52

5-1

2

0.8

11

-12

0

.16

6-1

1

0.2

86

-11

0

.62

0-l

l

0.9

06

-11

0

.16

4-1

0

0.2

42

-10

0

.50

3-1

O

0.8

53

-.1

0

0.1

48

-09

0

.24

6-0

9

0.3

82

-09

0.4

62

-09

6

.41

8-0

8

0.4

93

-01

) 0

.60

9--

09

0.5

23

-09

0

.61

8-0

9

0.6

82

-01

0

.61

5-0

9

0.1

17

-11

0

.12

2-1

1

0.1

12

-11

0

,12

2-1

1

0.1

56

--1

1

0.2

37

-11

0

.39

3-1

1

0.6

62

-ll

O.l

l4--

10

0

.20

4-1

0

0.3

62

--1

0

0.1

10

-0i

0.2

00

-04

0

.36

0-0

4

0.6

30

-04

0.1

10

--0

3

0.2

00

-03

0

.36

0-0

3

0.6

30

-03

0.1

10

-02

0

.20

0-0

2

0.3

60

-02

0.6

00

-05

0.

120-

0.1

0.2

00

-04

0

.34

07

04

0.6

00

-04

0

.12

0-0

3

0.2

00

-03

0

.34

0-0

3

0.6

00

-03

0

.12

0-0

2

0.2

00

-02

0.2

87

-10

0

.21

4-1

0

0.1

59

-10

0

.12

0-1

0

0.9

16

-11

0

.68

5-1

1

0.5

26

--1

1

0.4

21

-11

0.3

57

-11

0

.33

8--

11

0

.17

0--

11

0

.63

0--

02

0.1

10

-.0

1

0.~

00

,-0

1

0.3

60

--0

1

0,6

30

-01

0.8

20

-01

0

.86

0-0

1

0.9

00

-01

0

.91

0-0

1

0.9

80

-01

0

.10

5.1

-00

0

.11

5.(

-00

0

.12

5.1

.00

0.1

46

-12

0

.24

1-1

2

0.4

15

-12

0

.71

4-1

2

0.1

24

-11

. 0

.22

4-l

l 0

.40

2-1

1

0.7

01

-11

0.1

22

-10

0

.22

1-1

0

0.3

94

-10

O

.47

G-l

l

0.6

88

- 11

0

.10

8--

10

0

.17

0-1

0

0.2

56

-10

0.3

07

-10

0

.31

8-1

0

0.3

28

-10

0

.33

8-1

0

0.3

19

-10

0

.36

7-1

0

0.3

94

-.I0

0

.42

0--

10

0.6

77

-10

0.1

15

-09

0

.19

9-0

9

0.3

30

-09

0

.61

2-0

9

0.0

17

-09

0

.63

8-0

9

0.8

68

-09

0

.87

6-0

9

0.6

87

-00

0

.73

0-0

9

0.7

75

-09

0

.81

7-0

9

0.1

45

-ll

0.1

20

-11

0

.11

1-1

1

0.1

20

-11

0.1

54

-11

0

.23

3-1

1

0.3

86

--1

1

0.8

51

-11

0.1

12

-10

0

.20

0-1

0

0.3

56

-10

0

.62

2-1

0

0 6

11

-10

I

0.3

4d

-02

0.6

00

-02

0.

]2i)

-O1

0.2

00

-01

0

.34

0-0

1

0.(

00

-02

0

.10

0-0

2

0.4

00

-Q2

o.

(OO

-02

0.4

00

-02

0

.10

0-0

1

0.1

00

-01

0

.10

0-0

1

I 0

.12

7-1

1

0.1

06

-11

0

.96

9-1

2

0.1

01

-ll

0.1

21

-11

0

.17

1-1

1

0.2

69

-11

0

.44

0-l

l

0.7

42

-11

0

.13

2-1

0

10

.23

5-1

0

0.1

06

-09

0

.11

3-0

9

0.3

03

-09

0

.41

0-0

1

0.5

67

-03

0

.58

7-0

9

0.6

65

--0

9

0.6

24

-09

0.6

d1

-09

0

.67

2-0

9

0.7

13

-09

0

.75

2-0

0

0.1

04

-03

0

.17

9-0

9

0.2

97

-09

0

.46

2--

09

0.5

57

-09

0

.57

6--

08

0

.69

1-0

9

O.G

I3-0

9

0.6

30

-04

0

.66

0-0

9

0.7

00

-03

0

.73

8--

00

0.6

12

-10

0.1

01

-09

0

,18

0-0

9

0.2

98

-09

0

.46

3-0

4

0.5

58

-09

0

.57

7-0

9

0.5

96

--0

9

0.6

14

--0

9

0.6

31

-09

0

.66

1-0

9

0.7

01

--0

9

0.7

40

-09

0.3

99

-10

0,6

76

-10

0

.11

7-0

9

0.1

94

-09

0

.30

2-0

9

0.3

65

-09

0

.37

7-0

9

0.3

89

-09

0

.40

1-0

9

0.4

12

-09

0

.43

2-0

1

0.4

56

-09

0

.48

3-0

9

, I.. .

. . 8

0-Z

SI'O

8

0-8

r1.0

e

o-c

rr'o

80

--81

1'0

80

-11

1'0

8

0--2

Cf '0

8

0-8

21

.0

Po

-02

1'6

80

-91

1'0

8

0-2

11

'0

80

--60

1'0

P

O-S

OI'O

00

-1 OO

S'O

0

0 ( 0

9P

'O

00

-1-0

2P

.O

00

-I-OG

C'O

00

-1 0

Lr.o

0

0 1 o

sr'o

0

0 1

OE

E'O

00-1 O

IE'O

00

-1 O

GZ'O

0

0 -fO

LZ

'O

00

4-0

52

'0

00

IOE

Z'O

00

1-0

11

'0

00

CS

GI'O

o

oi s

u1

.o

00

-1 S

LI'O

con

t 'd

A E

./h

lcV

0.100--01

0.400-01

0.100-01

0.100-01

0,400-01

0,400-01

0..l00-0I

0.400 - 0 1

0. IUO 1

-00

0.100 1-00

0.100'1-00

0.100 1

00

0 100 boo

0. I00

k00

0.100-1-00

0.10

0 1-

00

0.100-bod

O.lU0 1.00

0.200 I-

00

0.200 t

oo

0 200 -1.00

0.200 -1-

00

0.200~l~00

0.200 k

00

refe

ren

ce

nin

n

dry

air

I w

ater

or

gan

ic

I n

ylon

g

loss

I

80

-95

5'0

6

0-lIC

'0

00

-lCS

'0

GO

-OZ

R'O

80

- 01

S'O

6

0- 8R

G

fl

RO

-ZO

S'O

6

0-

15K'O

I04

DciC

0

I0 I

OL

C'O

I0 l O

SC

'0

10

I oct n

A71V

/".7

con

t'd

--

CHAPTER 4 RADIATION PROTECTION OF NUCLEAR POWER PLANT AND REACTOR

4.1 GENERAL DESCRIPTION The nuclear power plant have had large scale development since the

first demonstration nuclear power plant came out in 1954. There are over 10 kinds of reactor used for electric power a t the moment, the more mature reactor of which is pressurized-water reactor (PWR), boiling water reactor (RWR), graphite-gas reactor, graphite-moderated water-cooled reactor and heavy-water-moderator reactor. The light water reactors (PWR and BWR) constitute majority in world nuclear power plant at the moment.

4.1.1 NUCLEAR POWER PLANT The nuclear power plants makes use of nuclear energy released

from the fission process of nuclide to generate electricity. For different kind of reactor the relative systems and equipment of nuclear power plant are large different.

The PWR nuclear power plant consists of the nuclear reactor, primary cooling system, secondary cooling system and other auxiliary systems. Fig. 4.1-1 gives the principle flow chart of the main systems in nuclear power plant.

The nuclear reactor is important equipment in nuclear power plant. Meanwhile, there are sustaining chain fission reaction in reactor, hence it is a radation source also. There are certain amount of nuclear fuels loaded in reactor. The heat energy released from fission process in nuclear fuels is removed out off the reactor by the coolant which flow through the reactor and then the coolant is fed into the steam generator.

The primary loop system consists of reactor, main circulating pump, pressurizer, steam generator, relative line pipes and valves and auxiliary equipment. The cooling water with high temperature and high pressure is pumped into reactor, after absorbing the heat energy released from the fission process in nuclear fuels the water flow into the steam generator. The heat quantity is transferred to the feed water of secondary cooling system in steam generator. Then, the coolant is pumped in- to reactor by main circulating pump. Such circulation forms one closed circulating loop.

-223-

The equipment of pnmary cooling system are put into the reactor containmellt by concentrated way. The containment is a large concrete building with about 30 m in inside diameter and about 60 n~ in height. Its function is to enclose the main equipment with raciioactive matters to prevent the radioactive matters diffusing outside. Even while the most serious accident occurs in nuclear power plant the radioactive matters are still enclosed safely in the containment and then the accident d o not affects the environment.

The secondary cooling system is the device which can transfer the heat energy to electricity. It consist of moisture separator, steam turbine, condenser, condensate pump, and feed pump. The feed water of secondary cooling system absorbs the heat quantity of primary cooling system and then become into steam, finally the steam come into the steam turbine and the motor is driven to generate the electricity. Because the reactor is a strong radiation source the coolant which flows through the reactor carries certain amount of activity. Especially while the fuel failure occurs the activity level of primary coolant is very high. Hence the coolant flowed off from reactor is not suitable for feeding the steam turbine directly, therefore PWR nuclear power plant has two loops. The building of the secondary cooling system for nuclear power plant is similar to the building of steam turbogenerator for normal power plant.

Besides the two cooling systems, there are chemical and volume control system, reactor safety system. fuel handling system, waste processing system and other systems in nuclear power plant.

4.1.2 RADIATION PROTECTION LIMITS O F NUCLEAR POWER PLANT

Dose Limits of Occupational Workers About basic limits with regard to occupational workers exposured, the rules of 60-th publication of ICRP have been described in Chapter 3. The limit of uniform exposure of whose body is of 0.02 Sv / a. In practice the exposure received by occupational workers is far more lower than this limit. The annual average dose equivalent of occupational workers of countries is 4.1 mSv in recent ten-odd years. The regulation of our country stipulated that under normal operation conditions the per caput effective dose equivalent of all occupational workers must be controlled below 5 mSv / a.

Dose -- Limits of Public Under Normal Operation Conditions The basic limits with regard to public, the rules of60-th publication of ICRP

is of 1 mSv, it is only half of natural radiation. But the authorized limit proposed by environmental protection department and the design rarget value proposed by management department of nuclear industry is far more lower than this value. The authorized limits and design target values for several countries are given in Table 4.1-1 and Table 4.1-2

respectively.

Table 4.!-1 Authorized Limits of Public Exposured Under Normal Operation Conditions

Country -- -

Dose equivalent limit, mSv / a (mrem / a) -- whole b o d ~ 0.25 (25)

- -- --- USA -- - thyroid gland 0.25 (25) -- 7'

-

other organ 0.25 (25) --

whole body 0.17 (1 7) UK - -- - - -- -- -- -- .- - -- -- -.

thyroid 1.0 (100) - - - - - -- - - -- - -

wasre gas: whole body 0.30 (30) Germany -- - waste liquid: whole body 0.30 (30) -- - - - - -- -- - -- - - - --

waste gas and liquid: thyroid gland 0.90 (90) 1 Sweden whole body 0.50 (50) I

Table 4.1-2 Design Target Values o f Public Exposured Under Normal Operation Conditions

Country Dose equivalent limit, mSv / a (mrem / a) - waste gas: whole body 0.05 (5), skin 0.15 (r --- -- -- ---

USA I and particulate: any organ 0.15 (1 5)

-- - waste liquid: whole body 0.03 (3), any organ 0.1 5 (1 5)

UK effective dose equivalent 0.25 (25) - Germany - whole body 0.10 (lo), thyroid gland 0.30 (30)

--

Sweden whole body 0.10 (10)

The relative standards of our country stipulates that the effective dose equivalent of any public induced from the radioactive substances released from one nuclear power plant to environment must be less than 0.25 mSv / a.

Release Limits In order to make the exposure undergone by pub- -- --

lic be less than the authorized limits or design target values the effluent amount of radioactive substances must be controlled. But the effluent amount is relative to the weather, the geographical features and other factors. Many countries stipulated general some control values for convenience of management, as known in Table 4.1-3.

Table 4.1-3 Radioactive Effluent Limits of PWR Nuclear Plant

i gaseous effluent, GBq i liquid effluent, GBq ( Country c- 7- - ( noblc gas , I plus particulate exception 3~ ! 3~

USA i ____2 1 3 . 7 ~ 1 0 ' ---- 1.85 x 10'

1 2.46x106 1 1 . 8 5 ~ 1 0 ~ I 1 . 4 8 ~ 1 0 ~ ' I . I I X I O ~ France ---- - ---

I 2.22x106 7 . 4 ~ 1 0 ' 1 . 4 8 ~ 1 0 ' 1 . 1 1 ~ 1 0 ~ j 1.11 x lo6 j 5.55 x 10' 1 . 8 5 ~ lo2 1 7.4

Germany ; 7 . 4 ~ los j 2.59 x 10' 7 . 4 ~ 10' 1 -- -

I I 3 . 7 ~ 10' ' 1 . 8 5 ~ lo2

The relative regulation stipulate that the effluent amount of gaseous and liquid effluent is controlled as following values for PWR nuclear power plant in our country.

gaseous effluent: noble gas 2.5 x 1 0 ' ~ ~ q

I 7.5 x 1 0 " ~ ~ particulate (T1 >8d) 2.0 x 1 0 " ~ ~

2

liquid effluent: 3~ 1.5 x 1014Bq

other nuclide 7.5 x 1 0 " ~ q . Dose Limits of Public Under Accident Conditions Strictly speak-

ing, any accidents which induce a large amount of exposure for public are allowed of no appearance, so ICRP has not regulated for this. But in order to give a reference to site assessment, to define design-base accident or emergency condition many countries stipulate the public dose equivalent limits under accident conditions, as shown in Table 4.1-4.

Table 4.1-4 Public Dose Limits Under Accident Conditions

1 Personal dose q l u v a l e n ~ limils. S v Collective dose equivalent i imib. man - Sv --

- ...a Country -- - - - - - --

who!e body lhyroid whole body thyroid ~-p .~

I Canada 0.25 2.50 1 x lo4 - - -- - -. - - -. - - -- - -. - - -- - - -- -- - -

Germany 0.05 I 0.15 1 - -- - - - - - - - I USA 0.25 3 .OO I

I lung 0.30 i 1

' bone marrow 0. I0 '

I

j 0.30 I

Finland 0.1 5 external radiation / 1 1 o f s k i n 0.90

The relative standards of our country stipulate that while one large accident occurs the effective dose equivalent received by any public must be controlled below 5 mSv and by thyroid below 50 mSv, and while one serious accident occurs the effective dose equivalent received by any public must be controlled below 0.1 Sv and by thyroid below 1 Sv.

4.2 RADIATION SOURCES IN NUCLEAR POWER PLANT The nuclear reactor is the device of generating nuclear energy in nu-

clear power plant. Hence it is one heating source, and at the same time, one radiation source with radioactive high level. The radiation emitted from reactor can be classified as primary radiation and secondary radiation. The radiations of fissionable nuclide products generated at fission and after fission are primary radiations. The radiations induced by the interactions between primary radiations and materials are called secondary radiations. Neutrons and gamma rays are the radiation with strongest penetrating power. The neutron and gamma ray sources are only discussed here.

4.2.1 REACTOR BLOCK Under Normal Operation Conditions -- Under normal operation

conditions the main neutron source is of fission neutron, and the main gamma source is of prompt gamma rays emitted at fissions and delayed

gamma rays released from fission products. 1 . neutron source One fission of ' 3 5 ~ released about 2.5 fis-

sion neutron in average, which carry about 5 iMeV energy. For one PWR with 900 MW electric power the prompt fission neuucn strength is 3 bout 4 x loZ0 MeV i s or 2.0 x lo2* n / s, and the neutron strength in unit volume is about 1.3 x 10'; MeV / (s cm3) or 6.5 x 1012 n / (s a crn3). The energy range of fission neutrons is from eV to 10 MeV, but the energy carried by neutrons above 14 MeV is less than 1 O/O of the total energy. Normal the 14 MeV is considered as the upper limits of fission neutron energy. The neutron spectrum distribution from 0.025 eV to 17 MeV can be expressed as follows

N(E) = 0.484 sh-expt - E) (4.2 - I ) where E in MeV.

The other neutron sources include delayed neutrons, neutrons of activation products and photoneutrons. The delayed neutrons are released during the fission products are decaying. The delayed neutrons released per fission are 0.0158 neutrons, and its energy is very low. When the water is used as the coolant the neutrons of activation products are the 1 MeV neutrons from 1 7 ~ are decaying, which are generated from

17 O(n, p) 1 7 ~ reaction. 2. gamma radiation sources The fission per 2 3 5 ~ releases 8.1

photons, which carry the energy 7.25 MeV, and the energy range is from 10 keV to 10 MeV. For one PWR nuclear power plant with 900MW electric power, its heat power is about 2600MW. The strength of prompt gamma radiation is

2.6 x lo9 x 3.1 x 10'' x 7.25 = 5.84 x lo2' MeV / s The core volume of such a reactor is about 31m3, therefore the strength of prompt fission gamma radiation for unit volume is about 1.89 x 1013 MeV / (s cm3).

The fission products is the mixture of gamma emitters whose half-life are from below 1s to few million years. The 6.65 MeV gamma enerm of " 5 ~ per fission is released after 1s of fission, and above three-fourth of this energy is released in 10-~s.

The other gamma radiation sources include thermal neutron capture gamma rays, fast neutron inelastic scattering gamma rays, gamma rays of nuclear reaction products, gamma rays of activauon products,

annihilation radiation and bremsstrahlung radiation. The amount and energy of these gamma radiation sources is very lictle. But capture gamma rays and inelastic scattering gamma rays can be produced in shielding. and the energy of capture gamma rays is from 6 MeV to 8 MeV, so when calculating the shielding these gamma rays must be considered.

Reactor Shutdown After reactor shutdown the main radiation - - - - - -- - - - - - . - - -

sources are gamma radiation released from fission products and activation products.

1. gamma radiation of fission products Normally, the energy of this gamma radiation is divided into seven energy groups:

1 0.1-0.4 MeV rz 0.4-0.9 MeV

r3 0.9-1.35 MeV r 4 1.35-1.80 MeV r5 1.80-2.20 MeV r6 2.20-2.60 MeV r7 > 2.60 MeV.

The Table 4.2-1 gives the dependence on time of gamma radiation strength of groups after shutdown for one reactor, whose power is 1 MW and operation time is one year.

2. activation products The materials (steel, water, zirconium and aluminum) in reactor will be with activity under neutron irradiation. Some parts, for example, fuel assembly, control rods, coolant and moderator, will be removed outside reactor, but some parts will left in reactor. The nuclear reactions often encountered are as follows: 160(n, p) 16N, 180(n, y) 190, " ~ a ( n , y) 2 4 ~ a , 2 7 ~ l ( n , a) 2 4 ~ a , 5 6 ~ e ( n , p) 5 6 ~ n , " ~ ( n , y) 5 9 ~ e , 5 8 ~ i ( n , p) 5 8 ~ o and 5 9 ~ o ( n , y) 6 0 ~ o .

Reactor -. Accident While reactor accident occurs, partial fission product will be released outside reactor.

1. noble gases The noble gases released mainly is K r and Xe. When the fuel elements are melted all of the Kr and Xe release from elements. Exception several nuclide, for example, lJ3xe, 1 3 5 ~ e , 3 5 ~ r, the half-life of other nuclides is very short in fission active gases, but as long as the fission activity gas can be detained for several hours the effects of inert gas can be greatly reduced.

hh

h

q~

"i1

X

XX

??

I-"

G

k X

XX

L

LL

-1 9

-n

nh

;IC X

IX

X

X

Lld

L

-

Table 4.2-3 Radioactive Level of Spent Fuel Storage and Transportation Container

lnmtory GRq (CI) Rdabrc to mrc (%) I Posltwn I UO, Fuel

UO, Fud pdkt Ga D To d Gap T o d

I I

I Corn'

(I) afier reactor shutdown by M m n

( 2 ) c a l ~ b y 2 / 3 o f w r c a s s m b t i c s . I/3ofthandeclybytbroc&ys, I/3ofthcm 150dap.

(3) calculaLcd by 1 / 2 of core arumblics, 1 / 3 of than decly by 6 day* 1 / 3 of them 160 days.

(4) 7 fud aoscmblics / c o n w a , decaying by I SO days.

(5) one fuc! ascmbly is conudcrd, dsapng by 3 days.

The amount of fission product in coolant is relative to the cladding materials and operating mode of reactor. Fot light water it is normally assumed that there are cladding failure of 1% of fuel elements at rate power, but in practice there is only 0.0 1 % fuel cladding failure.

Auxiliary Loops The radioactive concentration in liquid of auxiliary loops is relative to the purity power of the purity equipment (desalt- ing tower and filter) and the stay time of liquid in storages.

4.2.3 STORAGE AND TRANSPORTATION O F SPENT FUEL The radioactive substance exist mainly in the fuel elements. Speak-

ing of the radiation level, besides the reactor core the next are the spent fuel storage and container of the fuel transportation. The radioactive levels of spent fuel storage and container of the fuel transportation are given in Table 4.2-3.

4.2.4 WASTE TREATMENT SYSTEMS The activities of radioactive wastes are list in Table 4.2-4.

Table 4.2-4 Radioactive Wastes of PWR Nuclear Power Plan:

Systcrn Waste Swcilic act~vity GBq / rn' ~ ~ -- -- ~ ~~ -- ~ --

urilication system of primary wet ion exchange resin 3.7 x lo3-1.85 x 10' . -- - - - . - - --- - -- - - - - - - - A - -- - - - . -. - - - - - - -- - -

cooling sysl.crn filter unit 1.85 x 10'-1.85 x 10' .- . ~ --- I

concenlrated IIIJUILI waste of evaporator, Liquid wastc vw tmcnt 3 7 x 10-'-3 7 x 10'

residual sulphur, resln - - ~ - * ~~-~ ~ - - ~

high elXciency filter 3.7 x 10-'-3.7 x 10' waste gas and ventilation i---- - .. - -

I activated carbon filter 3 . 7 ~ 10- ' -3.7~ 10' - I waste paper, working clothes 7 4 x 10-'-7 4 x 10-I

I scrapped rnetailic pzrls and components 3.7 x lo0-3.7 x 10' Operation - I --- I

rod (after one year), start-up neutron source., 1 1 x 1 0 '

fuel assembly and space platc

4.3 RADIATION DETRIMENT OF NUCLEAR POWER PLANT The radiation detriment of nuclear power plant inciudes the expo-

sure on occupational workers and public and effects on environment

4.3.1 OCCUPATIONAL EXPOSURES The occupational exposures are relative to radiation level in plant,

type of work in production and operation in doing. Radiation Level in Nuclear Power Plant The zoning in

containment is given in Fig. 4.3-1, and the radiation levels in containment under normal operation are given in Table 4.3-1.

The radiation levels of main equipment are as follows. 1. reactor vessel After shutdown in few days the exposure rate in circular gap be-

tween reactor and primary shield near the highest of reactor core is 9-1 2 R / h which is induced by activation of carbon steel wall of reactor vessel and heat shield materials. The exposure rate at bottom of reactor vessel is 0.4-1.5 R / h. The exposure rate near the collecting water pit is 0.08-0.15 R / h. In addition, the neutron detectors arranged in this zone are become into strong radiation source by activation. The exposure rate on surface of detectors is up to 2-5 R / h.

L

'ON

"%

2. steam generator After shutdown the exposure r.ate of steam generator surhce is

10-100 m R / h. But because of the sediment of impurity in primary water the hot spot of strong activity is formed at local places. It is indicated by measuring the washer of manhole cover plate that main nuclides are 5 8 ~ o and 6 0 ~ o and their exposures are abou: 80% of total exposure.

3. main circulation pump After shutdown the exposure rate on surface of main circulation

pump is about decades mR / h. But there are some hot spot a t some places.

4. line pipes of primary loop After shutdown the exposure rate on surface of line pipe of primary

loop is about from decades mR / h to hundreds mR / h. Because of deposit of suspended matter the exposure rate on the top of pipe section is larger than that of the top of pipe section. There are some hot spot a t the bend and the interfaces between two pipes.

5. mixed bed column The mixed bed column is one equipment of chemical control

system, used for purifying the water of primary loop. The equipment arrangement is given in Fig. 4.3-2, and radiation levels of equipment and rooms are listed in Table 4.3-2. The second measurement of plant B is obtained under partial fuel cladding failure.

Spent Fuel Element Operation After shutdown in three days the exposure rate at 1 meter from the light water reactor fuel assembly is about lo5 R / h. While the spent fuel assembly is stored below water 4 meters the exposure rate is 1 R / h. The exposure rate is less than or equal 10 mR / h a t the upper of loading mechanism.

Occupational -- Exposure In Nuclear Power Plant The statistical data publicized by N R C of USA in 1981 give the exposure data about the PWR, shown as Table 4.3-3. From the data in this table it is can be seen that average rate power per reactor is increased year by year, average number of workers and average collective dose equivalent is increased year by year also but the average dose equivalent per worker reduce some, down to about 5 mSv / a.

Table 4.3-2

Radiation L

evel of Mixed B

ed Colum

n

On

surface of

ion

( exchan

ge colum

n

hlIeas11ring p

oint

N 0

.

En

try (uf m

ixed

bed coliimn

in

corridor for valve

operation

Measlirin

g position

Exp

osure rate,

mC

/(kg - h

) (R/h)

Corridor for valire

operation

a Up

per of

colum

n.

b Uown o

fcoln

n~

n.

Plan

t A

Plan

t I3

6.45 X10-'-1.29

X10-8

(0.026-0.006)

First

I Secon

d

I 'I'hi1-6

I

6.16 X lo-'

(0.02) 5.16

(23)

I -

1.29 X 10-'.-3.1 X

10da (0.00&

-0.012)

(n)

tlie

vnl

ue i

ll

the

bra

cket

s is

th

e r

enct

or

ni~

oll

~r

fo

r w

ork

ern

nn

ll~

r of m

sasi

~ra

ble

do

se an

d a

nnrr

al c

olle

ctiv

e d

oss

. (L

) th

o v

nlue

of

roil

lld

Lrn

cket

v is

re

lati

ve

t.o t

he

corr

espo

ndin

g v

alu

e of

po

inte

d re

ncto

rs

in

(a)

brn

cket

s.

-.

Dos

e sq

riiv

alen

t pe

r ti

nit

eloc

tric

en

emy

S

v/@

l W .

a)

[rer

n/(h

i W

a) ]

0.00

8 10

.61

0.02

2 12

.81

0.01

(1

.0~

--

0.01

5 [1

.5]

0.02

6 [2

.6]

0.01

11

.0)

0.00

7 10

.7)

0.01

11

.01

0.00

8 10

.81

0.00

8 [O

.Ej

--

0.01

2 (1

.21

0.01

3 (1

.31

J

r T

ota

l A

vera

ge

Tot

e1

4v

ern

ge

Ave

rnge

Ann

nct-

1969

1970

1w1

1072

-

1973

1974

1975

N~

un

ber

of

reac

tor"

'

4(3)

4(3)

4(3)

--

6(3)

12

20

26

Ave

rage

ra

te p

ower

pe

r re

acto

r (M

w)

349

349

399

4.16

633

619

699

1978

1919

.I 2

729

729

134.

69

1134

091

167.

13

(167

131

214.

37

(211

371

242.

66

(242

661

To

tal

elec

tric

en

erg

y

(hlW

. a

)

1079

979

1912

2544

3770

6824

1734

6

1984

0

1824

9

1828

7

3.96

[3

96]

4.29

14

291

6.10

1 [6

10]

6.78

(6

781

rnlm

ber

of

wo

rker

s fo

r m

e~

st~

r~

lle

do

se(k

"

(464

)

(134

0)

(406

)

(188

5)

9110

9697

0.00

66

[0.6

6]

0.00

66

(0.6

61

0.00

66

[O.Ii

6]

0.00

62

(0.6

21

613

419

2087

6

2672

0

3882

8

4t1'2

37

82.6

8 [8

268]

138.

07

[I38

071

P

676

614

696

924

1101

nrim

Ler

of

w

ork

ers

per

reac

tor

(161

)

(447

)

(228

)

(377

)

787

485

3.18

[3

18]

ppppp

4.80

[4

M]

1332

6

coll

ecti

ve

dose

eq

uiv

alen

t,

mnn

S

v (m

an . r

ern

J'"'

6.6l

(3

.63)

[6

a (

=a1

27.2

8 (1

0.99

) [B

28

(10

99)l

18.4

4 (9

.12)

I1

844

(012

))

--

37.0

8 (2

0.83

) 13

708

pm

3)l

93.9

9 [9

399]

66.2

7 16

6271

0.00

76

[0.7

6]

0.00

79

10.7

91

1768

8

coll

ecti

ve

dose

eq

uiv

alen

t pe

r re

acto

r S

vlr

em]

1.66

11

661

6.84

[6

84]

8.07

[

w]

4.64

(4

6i]

7.83

17

831

3.31

[3

31]

686

dobe

eq

uiv

alen

t pe

r m

an

Sv

[r

em]'b

'

(0.0

08)

[(0.

8)]

(0.0

082)

[(

0.82

)1

(0.0

101)

[(

i.oi

)]

--

(0.0

111)

l

l)

]

0.01

[1

.00]

0.

W

(0.6

81

Table 4.3-4 Collective Dose Equivalent Distribution by Type of Work in

Production in PWR Nuclear Power Plant(%)

I 1 Typeof workin production i PWR ; I

I

i Normal operation , 12.9

i Routine maintenance 1 24.7 ,

1 I Special maintenance i 34.1 , I

1 Inservice spection

Waste treatment 1 5 . 1 I

!

i Loading and unloading I

1 12.6

The distribution of collective dose in PWR by type of work in production is glvzn in Table 4.3-4. The 20%-25% of dose is received during reactor operation. The other 75-805G of dose is received duriilg reactor shutdown.

4.3.2 EFFECTS ON ENVTROSMENT The effeccs on environment discussed here point to the exposure on

public due to release little rzdioacuve substance. Effulent of Nuclear Power Plant The radioactive substances re-

- - - - - - - - - -- p- - - -

leased from nuclear power plant to atmosphere include the fission gases 4 1 (Kr and Xe), activation gases ( I4C and Ar), I, particulate and 3 ~ ,

and to water area include fission production, activation production and '11.

1. radioactive substances released to atmosphere For PWR nuclear power plant the activities of noble gases, I, 3~

and particulate released to atmosphere for generting unity electric ener- g y (GW(e) a) are listed in Table 4.3-5. There are not integral data about particulate nuclide component, but its components are almost same as the liquid effluent, including decades nuclides of fission products and activation productions. Unless the fuel cladding is greatly failure, in general, there are main activation products. The activity mainly come from the decay of 5 8 ~ o and 6 0 ~ ~ .

2. radionuclides released to water area For PWR nuclear power plant the activities of 'H and other sub-

stances released to water area for generating unity electric energy (GW(e) a) are listed in Table 4.3-6. The other nuclide components are listed in Table 4.3-7. The main corrosion products in effluent are 5 8 ~ o and '%o, next, are "Cr and fission products 1 3 4 ~ s , 1 3 7 ~ s and 13'1.

Collective Dose Equivalent of Public Under normal operation ~ - - - - - --

conditions the gaseous radioactive substances are very little, then the exposure received by public can not be measured. Assessing the effects on environment of liquid effluent is more difficult than that of gaseous effluent. Because the site conditions are different, one model can not be used for calculation of different sites.

Table 4.3-5 Radioactive Substances Released to Atomsphere for Unity Generation Output

J I I Averagereleasing quality for unity generation output., /

! i Radioactive 1 BGq/GW(e) . a (Ci/GW(e) a)

I substances 1975 1976 1 1977 1978 1979 1975-1979 I I I average

i Fission gas 401598 a 569430 I I

1 (10554) (15390)

-

Particulate 5.772 1.258 1 1.554 ; 1.85 1.295 2.146 I

1 (0.156) ' (0.034) 1 (0.042) i (0.05) (0.035) (0.058) I

* The values in table a r e the average value for PWR nuclear power plant in world-wide, but only the value of "C is the average value of Germany.

Table 4.3-6 Radioactive Substances Released to Wat,er Area for Average Unity

Generation Output. of PWR Nuclear Power Plant.

Averagereleasing quality f o r unity generat ion output I . . i Radioactive GBqlGW (e) a (Ci/GW (e) a ) I

1 substances 1975 1976 1977 1978 1979 1975-1979

1 average I

- - 1

I "H 52847.1 38661.3 39960 ' 32967 29970 37962 I 1 I (1428.3) ' (1044.9) (1080) (691) (810) (1026)

1 Nuclide 145.04 ' 309.69 305.62 125.80 74 183.59 1 e x c e p t 3 H (3.92) (8.37) (8.26) (3.40) (2.0) (4.97)

4.3.3 RADIATION ACCIDENT OF hiUCLEAR POWER PLANT .According to the probabiliry of accidenr, serious exrent or detri-

ment on public there are different classification methods in countries for possible occurring accidents of nuclear power plant. The accidents are classified as four kinds in China.

I . anticipated operational occurrences All operational processes deviating from normal operation which

are expected to occur once o r several times during the operating time life of the plant and which, in view of appropriate design provisions, d o not cause any significant damage to items important to safety nor lead to accident conditions, and do not cause radioactive substance to release greatly to environment.

2. unlikely accident The probability is less. If the relevant engineered safety features did

not function as per design intent this accident could lead to release of unacceptable quantities of radioaeiive materials to environment. While this accident occurs the dose received by individual of public wiil exceed 0.25 mSv, but and any individual of public will not exceed 5 mSv.

3. severe accident The probability of this accident is expected to be infrequent. Some

of the engineered safety features do not function, and lead to release of unacceptable quantities of radioactive materials. While this accident occurs the dose received by individual of public will exceed 5 mSv, but any individual of public will not exceed 0.1 Sv.

d. maximum credible accident This is a conceivable accident which leads to release of radioactive

materials in large scale and to produces serious sequence on environment. The probability of this accident is extreme little. This accident is used for radiation protection assessment in site selection. For PWR the maximum credible accident is core meltdown in large scale.

4.4 PREVENTION MEASURES OF REDUCING EXPOSURE RECEIVED BY OCCUPATIONAL RADIATION WORKERS

4.4.1 CONTROL BY ZONING To permit effective control over personal access to radiation areas

and to limit the spread of air or surface contamination. the building of

nuclear power plant must be divided into controlled areas, supervised areas and unsupervised areas. Site personal might receive doses in excess of three-tenths of the annual dose equivalent during the anticipated working - period in controlled areas. The radiatiori levels in supervised areas should be such that it is most unlikely that the annual exposure willexceed three-renths of the annual dose equivalent limits. The radiation levels in unsupervised areas shall be such that it is most unlikely that the annual exposure will exceed one-tenths of the annual dose equivalent.

To organize the operation of the controlled area in an effective way, this area is divided into zones based on radiation and radioactive contamination levels (dose rates and surface or airborne activity concentrations). The greater the radiation or contamination level of the zone, the greater is the operational control required to be exercised over the access of individuals in order to ensure compliance with individual annual exposure limits.

Zoning The methods of zoning for countries are different. The general practice is to divide the controlled area of nuclear power plants into three or more radiation and contamination zones. The zone classification can be seen in Table 4.4-1

Control by Zoning To minimize the radiation dose incurred by site personnel working in the controlled area, and the spread of contamination, the layout of the controlled area shall be such that personnel do not have to pass through areas of high radiation to gain access to an area of lower radiation, not through areas of high contamination to gain access to an area of lower contamination. The radiation areas, especially the areas whose dose equivalent is larger than 25 pSv / h, must be with significant sign. For high radiation areas prohibited to come into the door must be lock.

At the entrance to a controlled area, or the interface between contamination area and uncontamination area the charge rooms must be set. If necessary, at the interface between different contamination areas sub-change areas must be set.

Table 4.4-1 Zone Classification According to French Regulations

, Classification , Doseequivalentrate I DAC of zone (whole-body exposure, 2000hr/a) '

Monitored zone 2.5-7.5 li S v/h 0.1-0.3 (0.25-0.75 mremlh)

' Controlled zone GREEN 1

8 Controlled zone YELLOW

Controlled zone ORANGE

I

I Controlled zone RED

The layout and ventilation of the main change room shall be such as to prevent rhe spread of contamination from the controlled area to the uncontrolled area. Within this room, a barrier shall clearly separate the clean area from a potentially contamination area. The capacity of the main change room has to be large enough to meet requirements for work during shutdcwn periods and allow also for temporary personal from outside contractors.

Provision shall be made in the main change room for requirement by which to detect the external contamination of persons and equipment. The exit from the area shall be monitored or guarded to en- sxre that personal and equipment can only leave when a clearance signal is received from a body contamination monitor o r when permission is granted by the radiation protection personal.

In addition to radiation monitors, the main charge room shall be provided, a t the minimum, with following:

a. personnel decontamination facilities (showers and sinks). b. clean clothing and the necessary storage for it. c. contains for contamination clothing.

4.4.2 SHIELDING Properties -- of Reactor Shielding --A- In shielding design the reactor

shielding design is more complex. It has the properties as follows: a. complex radiation sources. For example, large activity, wide

energy range. There are neutrons, gamma rays and secondary gamma rays of neutrons, especially, after neutrons are absorbed in shielding materials the secondary gamma rays can be produced. The type, activity and spectrum has large different for reactor in operation and shutdown.

b. different shielding requirements. For example, it is requires that prevent equipment from damaging of irradiation, from activating of materials and from heating of shielding materials in technology. In radiation safety, according to the frequency and the occuqanncy time gained to access to the requirement for personnel, it is defined for different radiation levels and made the shielding design zone by zone.

c. complex shielding design. According to different objectives and requirements the different form shieldings are adopted, for example, integral shielding, partial shielding and shadow shielding. The pipe line penetration, especially local weak place induced by ventilation tubes passes through shielding wall, must be considered. The local high radia-

tion places induced by streaming effects of gaps, the secondary gamma rays produced in shielding layer and shielding heating must be considered also. For the entrance the special shielding forms must be adopted.

d. different shielding materials. Based on different radiation sources different materials are selected. For reactor block the steel and water are often used for shielding maierials. For cooling systems and auxiliary systems the concrete is often used.

Shielding -- --- of PWR Nuclear Power Plant In shielding design of - - -- - - -

PWR nuclear power plant the different forms and materials are adopted for different systems.

1. shielding of reactor block The reactor block shielding is called as primary shielding also.The

pipe lines and equipment of primary cooling system are arranged round the reactor. While the reactor in operation these equipment are with stronger radioactivity themselves and can not be accessed to. Therefore, the reactor block shielding function is to prevent these equipment and secondary coolant from activating and to ensure that the radiation levels outside shielding are lower than that of equipment themselves.

The reactor block shielding consists of multiplex steel-water layers in vessel and circular concrete of about 2 meters in thickness. Multiplex steel-water layers are core space plate, core cylinder, heat shielding, vessel and the water layers among them.

Besides protection aim for these shieldmg there are consideration in engineering, for example, the heat shielding can be used to prevent mechanical properties from change induced from extreme neutron irradiation, to reduce the heating in concrete, to prevent the equipment outside primary shielding from activating.

2. primary cooling system shielding The shielding of the primary cooling system is called as secondary

shielding also. The secondary shielding includes the wall, surrounded the primary cooling loop, which supports the crane, and the operation floor of concrete over the primary cooling loop. The secondary shielding includes the concrete construction also.

The important radiation source is 16N. The aim of the secondary shielding is to reduce the radiation of 16N to safety level, so that the personnel can come into the containment under operation in full power for necessary inspects and maintances. The secondary shielding has such function aiso that personnel may d o the routine works continuously

outside the containment under operation in full power. Because the containment thickness has certain thickness (one meter), therefore. when reactor melts and radioactive substances of large quantity come into containment it can make prevent surrounding public from excessive exposure.

3. shielding of fuel transportation The shielding of fuel transportation is to prevent personnel from

excessive exposure in processes of unloading, fuel transportation and fuel storage. The shieldings are water in unloading cavity and in spent fuel storage pool, the wall of unloading cavity, transportation passage and spent fuel storage pool and the metal transportation container. While unloading the upper space of reactor the vessel are filled with water. The thickness of water layer is to ensure that the exposure rate on surface of water layer is less than 2.5 mR / h.

4. shielding of auxiliary building The auxiliary building arranges with equipment, for example, chem-

ical and volume control system and waste treatment systems. Their radiation levels have large difference. These equipment must be shielded respectively to ensure that during operation period for adjacent equipment the personnel may come into this process room to make necessary maintenance.

5. movable shielding The movable shielding is used for the maintenance. The measuring

results indicate that the scattering rays are 15% - 44%, its average value is 38%. Hence, using simple movable shielding may significant reduce the dose received by personnel.

4.4.3 VENTILATION As far as the radiation protection is concerned, the primary objec-

tive of the ventilation system is to control the radioactive airborne contamination of the working environment in order to keep the radiation exposure and the intake of radionuclides as low as reasonably achievable for occupationally exposed persons and to ensure that appropriate limits are not exceeded. Other objectives of the ventilation system not necessa- rily related to radiation protection are:

a. To ensure employee comfort by providing air at suitable conditions.

b. To protect the plant structures and to enable the impaired func-

-249-

tioning of equipment by controlling the temperature and humidity of the ambient air as appropriate.

General Principle of Ventilation Design Tn order to reach the ob- - - - - - - - - - - . - - - - - - -

jectives of radiation protection the following measures are often adoptcd.

1. air change There must be enough air change rate in working areas and process

rooms to ensure environment conditions for workers can come into the working area and for making the equipment under normal operation conditions.

2. control of air flow direction In general, the air flow in the ventilation system should be directed

from regions of lower airborne contamination to regions of higher contamination. The pressure in areas with airborne contamination shall be lower than that in adjacent cleaner areas. A slightly negative gauge pressure is maintained in the contamination rooms and related buildings to permit control and monitoring of discharges of contaminated air.

3. control of air flow model in working areas The air feed holes and air exhaust holes must be arranged reasona-

bly. Considering the possible occurrence of hot and mechanical influence, if necessary, the local exhaust may be add to ensure however the contamination sources take place anywhere the contamination matter should be teken away. N o dead angles exist.

4. closed-loop recirculation In some areas it may be desirable to have closed-loop resirculation

including filters, dryers and coolers, as appropriate, with only a smail controlled-flow discharge.

5. purification According to need the air released to environment should be made

to decay, filter and remove iodine. When the radiation level in air reaches the regulated level it can be released.

6. minitoring and control The air released should be monitered. If necessary, the small con-

trolled-flow discharge is used to reduce the radioactive matter released to environment.

Ventilation for PWR Nuclear Power Plant The primary ventilation systems of PWR nuclear power plant are as follows:

1. ventilation of reactor block

It includes two parts of control rod drive mechanism ventilation and reactor sump ventilation. The reactor ventilation is aimed at removing the heat away to control rod mechanism and ionization chambers in reactor sump at normal operation condition.

2. containment ventilation It includes containment purification system, containment cooling

system, containment air control system and containment purge system. The containment purification system is a recirculation system in

continuous operation. I t exhausts the air from containment down, and makes the air pass through high effciency filter and remove iodine, then feed the air to the upper of containment operation plateform. In this way, the iodine and particulate in containment can be kept at certain level.

The containment cooling system is a recirculation system in continuous operation. It exhaust the air from upper of containment and feeds related rooms after cooling.

The containment air control system is to eliminate the hudrogen in upper of containment under loss-of-coolant accident to ensure that the concentration of hydrogen in containment is lower than the explosion limit. There is a hydrogen recombiner which can make the hydrogen become into water.

The containment purge system is used in reactor shutdown. It is started up before the persons come into containment to clear the containment and exchange the air in containment.

3. ventilation of auxiliary building The aim is to remove the heat away released from the equipment, to

make the working area at certain temperature and to keep that the contamination level in working area is lower than certain limits. This is a non-recirculation system. After filtering and cooling (or heating) the incoming wind is fed into low contamination, and then is exhausted from high contamination area, After filtering (if necessary, removing iodine) it is fed into stack to release.

4. ventilation of spent fuel buildmg The cleaning air is fed into the upper of fuel pit shielding slab where

the workers do some operations, then the air come into the upper of the pit through the slab gaps, the air is exhausted from the upper of pit and after filtering the air is fed into stack to release.

5. temporary ventilation

The air flow reguiating dampers or connection tubes can be set on ventilation tube to connect with the soft tube for ventilation.

One movable auxiliary ventilation system can be used. This system includes one air exhaust fan, high efficiency filter and activated carbon filter. The air released from this system can be directed to rooms, in this way, the equilibrium of ventilation sysem can not be distorted.

4.4.4 MEASURES FOR REDUCING ACTIVITY O F RADIATION SOURCES

The occupational exposure of persons in nuclear power plant large- ly comes from the maintanence. The main radioactive sources are 5 8 ~ o and 6 0 ~ o in primary cooling system. The 5 8 ~ o comes from reaction " ~ i ( n , p), however, the Ni exists in many alloies, especially in Inconel

of steam generator. The 6 0 ~ o is produced from the activation of 5 9 ~ o . The 5 9 ~ o as one kind of impurity (content less than 0.2%) exists in Inconel and stainless steel. In addition, the 5 9 ~ o is the main composition (content up to 50% or more) of surface harden material of wear-re- sisting parts in main pump, control rod drive mechanism and valves. In the initial term of reactor operaction the exposure contribution of 5 8 ~ o and 6 0 ~ o are almost same. Because of long half-life of 6 0 ~ o , therefore, after several years of operation the exposure of 6 0 ~ o is dominant. Hence how to reduce the content of corrosion products included 6 0 ~ o in primary cooling system has important significant for reducing occupational exposure.

To reduce the content of corrosion products in primary cooling system the following measures can be adopted.

a. Selecting the primary cooling system materials to reduce the radioactive corrosion products, espceially the formation of 6 0 ~ o , as low as possible.

b. Selecting resonably the operation condition (controlling the PH value of coolant) to reduce the deposit of corrosion products a t the equipment and pipe line.

c. Flitering the coolant to seperate the corrosion products from the primary cooling system.

d. Decatanminating to the systems to remove the deposited corrosion products away from the system.

Selection -- of Materials In the primary cooling system of PWR, the main sources of Co are the pipes made by Tnconel-600 in steam genera-

-252-

tor and surface harden matelals. The contacting area between the coolant and stainless steel is very little, therefore, the contribution of stainless steel for Co content in corrosion products is not large.

While looking for subtitute materials, it must ensure that the corrosion resistance and wear resistance is better than that of original materials, or at least not less than that of original materials.

Controlling the PH Value of Coolant To control the reactivity there is boric acid in coolant. To control the PH value of coolant the lithium hydroxides is added in coolant. In general, the PH value of coolant lies in range between 4.5 to 10.5. If regulating the concentration of boric acid and lithium hydroxides the PH value is controlled between from 6.8 to 7.2, then the generation and deposit on surface of equipment of corrosion products can be reduced, hence the radiation level in whole system can be reduced.

Filtering Filtering off corrosion products in coolant will reduce the radiation level.

Decontamination The decontamination can be classfied as decontamination of devices and whole system. The devices decontami- nated are steam generator, main pump and control rods and core monitering meters taken out from reactor. For whole system decontamination the following requirements must be satisfied:

a. effcacious removing off the radioactive deposited matter on surface of equipment of primary cooling system,

b. corrosion actions on system for deconmination reagent must be as small as possible,

c. waste produced in decontaminating processes must be as little as possible, and radioactive impurity in waste liquid can be seperated by ion exchange resin,

d. decontamination can be made under reactor loaded (boron content in coolant is 2000 ppm),

e. decontaminant is with good heat stability and radiolysis stability.

4.4.5 PLANNING, ORGANIZATION AND TRAI'IUXING To reduce the occupancy period of workers in radiation areas, it is

necessary to work out a well-conceived plan for maintance, to make strict organizing and trainning for maintance workers. For radioactive maintanence jobs, the following things must be done in advance: work-

-253-

ing out the job precedure, stipulating the time needed for jobs, estimating the abnormal cases appeared possibly and emergency measures adopted, defining the necessary measures of reducing the occupational exposure.

While working in strong radioactive areas the working time of workers must be limited, if necessary, the workers take turns, the wcrk- ers must be trained in advance, for example, emulating operations in models to make them skilfully master the operations and shorten the operation time as possible.

On the protection measures of reducing occupational exposure of workers, besides mention above, it is important for making careful design to systems and careful arrangement to requirement and raising the reliability of equipment in fabrication so that the frequencys of maintanence and exchange of equipment can be reduced and, hence, the maintanence time can be reduced.

4.5 PROTECTION MEASURES OF REDUCING EXPOSURE OF PUBLTC

There are many measures that can be used to reduce the exposure of public. Some of them are adopted in design stage, some of them are adopted in operation of nuclear power plant, some of them are adopted after the accident occurs. These measures concern waste treatment, safegear and emergency protection actions. The siting and multibarrier design which prevent releasing of radioactive substances are only dts- cussed herein.

4.5.1 SITING The site of nuclear power plant must be satisfied with requirements

of nuclear safety and radiation protection to ensure that under normal operation conditons and under accident condttions the individual dose equipment and collective dose equipment are lower than the regulated limits.

While knowing of one assessing site if suitable for building power plant the following effects on nuclear safety and radiation protection must be considered.

Effect of External Events on Plant In general, the radioactive -- -- - - -

risks induced by external events should not exceed that induced by internal accident. Hencc, it must make full investigation for effecting of'

-254-

external events on plant. That is to collect the data of occurrence frequency and severe extend, to analyze its reliability, precision and integi- ty and to define the design basis for external event by suitable methods..

For one external event (or combination of events), its reference value selected as design basis should ensure the integrity for structure, systems and parts which are relevant with chis event and important function to safety, and still do not lose its function. If recommended measures can not defend the plant from destroy of external events it is necessary to consider that this site is not suitable for building the plant.

The following aspects need to work out design basis for main external natural events.

a. Floodwater The definition of floodwater is any abnormal high level or overflow of river, flood-relief channel, lakes or seashore. These include river floodwater induced by precipitation and thaw, windstorm tide induced by hurticane, tsunami or seiche induced by earthquake and the floodwater induced by burst of a dyke, ice raft block and landslide.

b. Defect of geologrcal structure It includes breakage on earth surface,faults, landslide, mud-rock flow, snow slide, and caverns in soluble rocks, mine, oil well, gas well and water well which can induce sink, subside and swell of earth.

c. Earthquake It is necessary to understand seismic activity, earthquake formation and relation between both, and to define the ground motion about earthquake design basis vibration.

d. Tornado and typhoon. e. Events of effect on long term released heat. These events include

dry season water flow rate, lowest water level, and duration of lowest water level of cooling water source, and insufficient water supply of cooling water induced by river block, reservoir empty and ship collision.

f. Other natural events. For example, volcano activity, sandstorm, rainstorm, hailstone and thunder and lightning. If these events can effect the safety of plant it is necessary to define the relevant design basis.

The following aspects need to work out design basis external man-induced events.

a. Aircraft crash. Including impact, explosion and firing. b. Chemicals explosion. Including fabrication, treatment, storage,

transportation and usage of explosive, ammunition, chemicals and liquid or gaseous fuel. In addition, the missiles generated in explosion

-255-

should be considered. c. Fires. Including the fires generated by adjacent enterprises,

chemical plants, storage devices or oil and gas lines, and the fires of brush, forest and transportation accidents.

d. Other man-induced events. There are devices which storage, fabricate, transport or treat the matters with poison corrosion or activity.

Effect of Site Characteristics and Environment Characteristics on - - - -- .- - -- --

Radioactive Substance Migration For design and siting of plant, the effecting of effluents on environment, organisms' habits and public must be considered. The plant should ensure that under the conditions of maximum credible accidents and disadvantage diffusion it does not bring about unacceptable exposure to public.

Effecting of environment characteristics on radioactive substance are mainly dispersion of atmosphere, surface water and subterranean water.

a. Atmospheric dispersion. The region meteorological characteristics, meteorological data of site, and terrain and landform must be investigated. These data should observe on-site and the observation data must be with one year. Building an atmospheric dispersion mode of effluent is used to calculate diffusion factors of short term and long term to access effecting of effluent on public.

b. Dispersion of surface water. To access the dispersion characteristic of surface water, the water body position, water body size, water body form and its changes with time must be investigated. The following aspects must be investigated also: the flow rate of river, lake flow and sea flow, silt content in water body, water inlet of water supply system of user. The dispersion characteristic of water body must be studied on these bases to determine the migration mechanism of radioactive matters in water body, and to access the effecting of liquid effluent on public through the surface water.

c. Dispersion of subterranean water. To access dispersion characteristic of subterranean water in site, the following aspects must be investigated: non-saturation bed and water-bearing bed of stratum, water level isohypse and its changes with water level and meteorological phenomena, the direction and velocity of subterranean water motion, the utilize of subterranean water resources, the ways of persons contacting with subterranean, the hydraulic relation between the surface

water and subterranean water. In addition. the following aspects must be done: investigating phys-

ical and chemical characteristics of soil layers and rocks in site which are related with dispersion. building the model described mechanism of the radioactive nuclides pass through the water-bearing layers, defining the ways of the public receives the liquid effluent exposure, accessing the exposure received by public under accident conditions.

Population Distribution ~ and Resource Utilization To reduce the collective dose equivalent of public under accident conditions, to decrease the property loss and to evacuate public organized from contamination areas, the site of plant should be selected at the region where the population density is low in and is far from factories and mines and population centuries. The non-residence quarter should be set round the plant, and surrounding the non-residence quarter the limiting region should be set.

The non-residence quarter should be administered under plant. In general, the activities which are not relative to plant operation, should be limited. The activities, which are necessary in non-residence quarter, should not effect on plant operation absolutely, and should be governed under controlling of plant. The railways, highways and waterways are permitted to pass through the non-residence, but under accident conditions the plant should has power to control these traffic devices to protect the public.

The limiting re@on is limiting development area. Population growth and investment in construction should be given controlling, to make that the population and assets in this area is lower than that of surrounding areas, so that to reduce the effects of accidents of plant.

The data of present and programing population distribution should be collected. The population should include permanent population and transient population. The food and drink habits should be investigated for different resident group (for example, ages, nations, village and city resident), to access the ways for the radioactive matters come into body. The utility of earth and water resources should be investigated for site area. The investigation contents include the area of cultivated land and grazing land, kinds of products, industrial used water and public water supply. Especially, the directed and undirected ways of contamination of food chains should be paid attention to.

4.5.2 MULTIBARRIER O F PREVENTING RELEASE O F RADIOACTIVE MATTERS

To prevent radioactive matters release to environment, in general, the multibarrier is considered in design (see Fig. 4.5-1).

Pellet -. - - -- Fuel - . - The fuel itself can keep large part of fission products

generated by nuclear fission in inside of pellet fuel. Under normal opera- - tion conditions, only partial fission gases and volatility matters release from inside of pellet fuel. When the uranium nuclei near surface for few micrometers in depth fission the fission fragments can recoil from the surface.

Fuel Cladding To prevent the escape of radioactive fission products and chemical reactions between fuel and coolant the pellet fuels are usually put into the tubes which are as cladding. These tubes can bear certain pressure. The fission gases and volatility matters released from pellet fuels under normal operations may be held up in gaps between tubes and pellet fuels. Adopting the zirconium alloy as the cladding, the permeability quantity of 'H is only 0.013% - 1 % of content of 'H in fuel.

Primary Cooling System Whole cooling system, including vessel, --

pump, steam generator, pressurizer and their connecting pipes, consists of a closed system which can bear certain pressure. In this way, released fission products are held up in the cooling loop even in cladding failurz. Of course, because coolant loop is a high pressure loop it is not avoida- ble that there is small coolant leaking to containment and secondary loop, but this leakage is very small.

Containment - In general, while designing the containment the accident factors, for example, temperature and pressure generated in loss-of-coolant accident, earthquake, hurticane and fragments flew from inside and outside of containment, are considered. The volume of PWR containment can relax pressure rising in loss-of-coolant accident. In addition, there is spring system in containment. This system can make the steam released in accident condense to reduce the pressure in containment.

4.6 RADIATION MONITORING OF NUCLEAR POWER PLANT T o discover in time the abnormal radiation events, to ensure the ra-

diation safety of workers in plant and public, and to access the exposure received by occupational radiation workers and surrounding resident,

the radiation monitoring to workers, working areas and environment must be made.

4.6. i WORKING AREA MOhTITORING There are neutron radiation monitoring and gamma radiation mon-

itoring for working area monitoring. But large parts of them are continuous monitoring of gamma radiation. While gamma exposure rate is over present value the monitoring system can give alarm signal to prevent of workers from exceeding exposure.

The measurement points of area monitoring in PWR nuclear power plant are given in Table 4.6-1.

According practice arrangement there may be some areas needed to make area monitoring. On monitoring areas mentioned above, some areas of them need one and more detectors. Centre control room should set monitoring point, and the measured range is 0.1 - lo4 mR / h.

4.6.2 EFFLUENT MONITORNG There are two aims for making monitoring to effluent: providing

full data explains that release quantity is lower than release limit under normal operation conditions and when abnormal release occurs the monitoring assembly can give alarm signal in time to adopt necessary protection measures, providing some information tell if reactor system and waste treatment system are under normal states.

About radioactive effluent monitoring, the Safety Series NO. 46 published in 1978 proposed one recommendations. The monitoring programs of airborne effluent of plant in that recommendations is as follows.

a. The typical monitoring system for nuclear power plant includes continuous measurement of the noble gases and continuous sampling of 13'1 and particulates. The periodic measurement of the samples of 13'1

and particulates in the laboratory is, in general, considered sufficient and will be more accurate for low release rates.

b. Normally, only those mixtures of radionuclides and specific nuclides that are named in the authorization are routinely monitored. However, it may be necessary to carry out periodically a detailed analysis of the overall radionuclides composition. Additional monitoring may be required for special nuclides such as 3~ and I4c.

Table 4.6-1 Listing of Area Monitoring in PWE Nuclear Power Plant

XG. Position Range mC/(kg - h ) (mR/h)

--

1.0 Containment

I 1.1 Core instrument area 2.58X10-"-2.58 (0.1><104)

I 1.2 I Charge-discharge floor 2.58 X10-6-2.58 (0.1 X104)

I 1.3 , Worker access port (containment side)

I 2.0 Auxiliary

2.1 S olid waste storage area 2.58X10-"-2.58 (0.1-lo4)

2.2 Control panel area of 2.58 X10-6-2.58 (0.1-10") waste encapsulation station

I 2.3 Control panel area of radioactive waste

-

2.4 Residual heat removal pump 2.58 X 10-6-2.58 (0.1-10") and exchanger area

2.5 Maintanence building of 1 2.58 X 10-"-2.58 (0.1-10") radioactive compound

2.6 Radiochemical lab. I 2.58X10-"-2.58 (0.1-10")

2.7 Sampling station area 2.58 X10-5-2.58 (0.1-10") of primary water

2.6 Air exhaust fllter area 2.58X10-6-2.58 (0.1-10")

3.0 Turbine building

3.1 Evaporator of chemical 2.58X10-"-2.58 (0.1-los1) 8 waste liquid reproduction

4.0 Fuel building

4.1 Fuel storage area 2.58 X10-"-2.58 (0.1-lo4)

4.2 Operation area of fuel transfer flask

5.0 Other area

5.1 Equipment decontamination 2.58 X10-5-2.58 (0.1-10")

c. Any accident in a power reactor may involve the release of large quantities of noble gases. The continuous measurement of such gases required for normal operating conditions must also be adequate for the detection of accidental releases, and therefore, the measurement instruments must have a sufficiently wide measuring range.

d . Because of the different characteristics of the radionuclides that may be discharged to the atmosphere, several types of monitoring are required. Continuous monitoring is necessary for the noble gases, and continuous sampling for radioiodines and tritium and for particulates. For the direct monitoring of effluent release under normal operating conditions, continuous measurement of ' k r and of 13'1 at the stack is generally considered to be sufficient, For the nuclides sampled on a continuous basis, periodic measurements in the laboratory are necessary for 3 ~ , I4c, 12?, I3'I, the actinides and other beta/gamma-emitting

particulates. e. Such a monitoring system must also satisfy the requirements for

alarm monitoring in case of an accidental release. For liquid effluent, several requirements are in principle similar to

those of the monitoring programs for airborne effluent. In general the various types of liquid effluents produced are collec-

ted. according to their radioactive and chemical characteristics, in separate tanks and basins, and treated, as necessary, before discharge to the environment. The radioactive waste waters are then discharged on a discontinuous (batch) basis, to a pipe or canal which leads to a body of water.

The discharge of any such batch can be carried out only under appropriate control which ensure that the authorized discharge limit will not be exceeded. This requires taking of a representative sample of each batch and measuring its radioactivity content before discharge. Then the representative samples of each batch are made into composited sample in order of week, month or season and the composited sample is analyzed for its nuclides .

In addition, the continuous monitoring instruments must be set at end of discharge pipe or mixing well for discovering the abnormal discharge in time and giving the alarm signal to adopt necessary measures.

Suitable environmental monitoring equipment shall be provided. This equipment shall be able to detect any significant increase of radiation above background. Monitoring shall include the measurement of external radiation, airborne particulate and iodine concentrations and deposited activities, either by continuous measurement or by integration over specified periods of time.

Under normal conditions, main aim of environmental monitoring of nuclear power plant is providing full data prove that the environment round the plant is satisfying environmental quantity standard, and to estimate public exposure produced due to nuclear power plant operation, at least to be able to estimate upper limit of this exposure.

CHAPTER 5 RADTA'TION PROTECTION MONITORING

To estimate and control the radiation exposure level of working areas and environment, the measurements for radiation and radioactive matters are called as radiation monitoring. The radiation monitoring includes the accessment according to relevant stipulations of recommendations of ICRP and government authorities.

The radiation monitoring is main means for checking executive cases and knowing if protection devices are safety and reliable. Efficient radiation monitoring is helpful for discovering accidental omens as early as possible to adopt protection measures. Hence, radiation monitoring is very important in radiation protection.

5.1 RADIATION MONITORING BASIS Some changes induced by interactions between radiations and mat-

ters, including physical, chemical and biological changes, and concomitant appearing secondary phenomena become into the basis of radiation dose measurements. In addition, thermoluminescence detectors, glass phosphors and chemical method are often used in dose measurements.

5.1.1 IONIZATION METHODS USED TO MEASURE X AND y RAY DOSE

Standard Measurement Methods of Exposure -- Free air ionization chambers are used to measure the exposure of gamma rays and cavity ionization chambers are used to measure the absorbed dose.

1. Free air ionization chambers. The free air ionization chambers can be used to measure the expo-

sure and exposure rate accurately. Fig. 5.1-1 is a schematic diagram of a free air ionization chamber. The diaphragm F is used to limit the cross section a of incoming beam of X or y rays. The rays, which inject from F, pass through air plate ionization chamber, and then come out of hole 0. A is the high voltage electrode. B, and B, are protection electrodes. The electric potential of B, and B2 is close to that of electrode B. A group of protection wire is arranged in equal space between the electrode A and B to make an uniform distribution of electric potential dif-

-265-

ference between electrode A and B. The perpendicular distance from edge of X or y ray beam to electrodes. the distance L , from front edge to entrance of X and y rays and the distance L, from back edge of electrode B to exit of rays is at least larger than the average range of secondary electrons in air to keep that the secondary electrons can not impact with the electrodes and to satisfy the electron equilibrium condition. In this way, all ion pairs produced in volume between protection wires and electrode of A, B are generated by the secondary electrons released from X or y rays in measuring volume. The measuring volume is proportional to the electrode length L and the cross section of beam at the center of the volume. Because the area of beam changes with square of distance from radiation source, and the exposure is in inverse proportion to square of this distance. Hence, the total charges of one sign ions produced by the secondary electrons, which are generated in the measuring volume, is Q = X a, L p. The X is the exposure of X or y rays at the injecting diaphragm F, unit of exposure is C kg-'. a, is the cross section of injecting diaphragm, its unit is m2. L is the length of collection electrode, its unit is m. p is air density under normal condition, its unit is kg m-3. Hence, the exposure at injecting diaphragm can be expressed by following equation

Substituting p = 1.293 kg mP3 into Eq. (5.1-I), following equation can be obtained

In practice, measurement results obtained from above equation are needed to correct for following factors: ray decrement between diaphragm and collecting volume, loss of ion pairs induced in ion combination process and scattering of injecting radiation. Considering all of factors, accuracy of exposure measured by free air ionization chamber is in 1 O/o .

The free air ionization chamber under atmospheric pressure has been become as the standard device of measuring exposure. But the energy range of X or ;i rays is limited from 50 kev to 3 MeV. While the en- e r u is higher than 3 MeV, because the electron range is longer, to satisfy

the electron equilibrium condition it is necessary to build one very large free air ionization chamber or one ionization chamber filled with high pressure gas. But there is diff~culty in technology. While measuring the X or 7 rays with energy below 50 keV, because the air absorption is very severe the measurement error is larger. At present, the exposure measurements of X or :, rays with higher energy almost use cavi~y ionization chamber.

2. Cavity ionization chamber. Assuming one beam of uniform X o r y rays expose the solid me-

dium (see Fig. 5.1-2). There is a small cavity A fill with gas in the solid medium. The dimension of the small cavity is so small that the electron equilibrium condition can be satisfied. Then, if the small cavity is not present the relation between the absorbed dose Dm (J kg-') of the medium (m) a t the cavity position and the ionization produced by secondary electrons in gas in cavity (g) can be expressed by following equation

where q,=charge produced in cavity gas of unity mass by secondary elec-

trons, C kg-'

S =average mass collision stopping power ratio of matter m to m.g

-- gas g.

S equals ratio of average enerm imparted to unity mass me- m.a

dium m to average energy imparted to unity mass gas in cavity by secondary electrons respectively, i.e.

Eq. (5.1-3) is called Bras-Gray formula. This formula is a basic formula in dose measurement. I t is suitable for any medium and any gas

filled in small cavity. As long as knowing the value of sng and W,/ e,

the absorbed dose a t relevant position can be calculated according to Eq. (5.1-3) by measuring the ionization charges q, in cavity.

The cavity ionization chamber schemed in Fig. 5.1-3 is constructed according to Bragy-Gray principle. When assumption conditions men-

tioned above are satisfied the Eq. (5.1-3) may be rewritten as follows

where D, = absorbed dose of wall matter of ionization chamber a t the cav-

ity position

s =average mass stopping power ratio of wall matter to gas in ",R

cavity for secondary electrons. From Eq. (3.3-18) in Chapter 3 and Eq. (5.1-5), following equa-

tion can be obtained

Obviously, the cavity, which is used to measure the exposure, can be filled with any suitable gas, as long as the parameters in Eq. (5.1-6) may be known. The most ideal case is using air-equivalent ionization chamber, that is to say, the wall matters and the gas in cavity are equivalent with air. Then the Eq. (5.1-6) can be simplified as follows

In this case, air-equivalent ionization chamber is equivalent with free air ionization chamber, i.e. with Eq. (5.1-1).

In practice, it is quite difficult to require that the wall matter is completely equivalent with air. But under certain condition it is possible to make approximately equivalent.

- Under first degree approximation condition, the relation between

S in Eq. (5.1-6) and effective atomic number and atomic weight of "',g

relevant matters is as follows

Substituting above relation into Eq. (5.1-6). and assuming that the gas in cavity is air, then

If the wall of ionization chzmber is made from the low z materials

c c (z< 30), for example, carbon and aluminum, then (-)w / (-) z 1 in A A n

above equation. If the energy of gamma rays is in such range in which the Compton scatter between the gamma rays and wall of ionization

1' cn "cn chamber is dominant, then (-)a / (---)w z 1 in Eq. (5.1-7). There- P P

fore, the difference of result obtained from Eq. (5.1-7) and Eq. (5.1-1) is very small, i.e.

At present, the graphite ionization chamber act as the standard device to measure X and y rays exposure for all country in world. The measurement accuracy of the kind of these ionization chambers for measuring the gamma ray exposure of "CO is better than 0.7%.

5.1.2 NEUTRON DOSE MEASUREMENT The neutron dose measurement, in the same way as gamma ray

dose measurement, is based on the ionization effects of secondary charged particles produced from the interaction between neutrons (or gamma rays) and materials. But there are many kinds of secondary particles (for example, a particle, proton and photon) produced from interaction between neutrons and materials, and the changes of interaction cross section with neutron energy are quite complex. The radiation weight factor for different energy neutron have larger difference. Therefore, neutron dose measurement is more difficult than gamma ray dose measurement.

Measurement of Neutron Absorbed Dose Using cavity ionization chamber to measure the neutron dose is still the more accurate method in neutron absorbed dose measurement. But the range of secondary charged particles (for example, photon) due to interaction between

neutrons and materials is ver- short. In practice, i t is impossible to make that the dimension of ionization chamber is less more than the range of the secondary charged particles. Hence, it is necessary to use uniform medium ioni~ation chamber, i.e. the wall of chamber and the gas in cavity are with the same atoms. In this way, the wall of chamber and gas in cavity have the same mass energy transfer coefficient for injection neutrons, and have the same mass collision stopping power for secondary charged particles. If using the uniform medium ionization chamber, the neutron dose in tissue can be measured directly, i.e.

where q, = ionization charges produced in unity mass tissue equivalent gas

in cavity, C kg-'. The other parameters are same as above.

In practice measurement. to separate the neutron dose and gamma ray dose in mixed field of neutrons and gamma rays, the twin ionization chamber construction is used. One ionization chamber does not contain the hydrogenous materials, then it is not sensitivity to neutrons. Other ionization chamber contains hydrogenous materials then it is sensitivity to neutrons and gamma rays. Therefore using the reading difference of both of ionization chamber the absorbed dose of neutrons and gamma rays can be measured.

If using the proportional counter to measure the neutron dose, the pulse amplitude discrimination technique can be used to eliminate the effects of gamma rays.

Neutron Dose Equivalent Measurement In radiation protection, it is necessary to measure the dose equivalent in Sv. In principle, first, the neutron doses of different energy neutron are measured, by using the cavity ionization chamber, second, these doses are multiplied by relative radiation weight factors, and final the products are assumed. The result is the required dose equivalent. If the neutron spectrum distribution is not known, one special neutron detector, which can display the neutron dose equivalent value without relation with the neutron spectrum distribution, is used. The operation principle will be discussed briefly.

When neutron beam acts on the organism, according to neutron energy En and relevant radiation weight factor w, the dose equivalent

transfer factor f, in tissue of unity neutron flux for different energy can be calculated. The relation between f, and neutron energy En is shown in Fig. 5 . 1 4 . Therefore, as long as the neutron spectrum a,, is known, the dose equivalent can be calculated by following equation

But, in practice the neutron spectrum is not known. It is necessary to regulate the instrument respondency in certain energy range to make that the detecting efficiency q(E,) of instrument is proportional to f,. In this way, the neutron number, N,, detected by detector in radiation field is proportional to neutron dose equivalent H,.

According to this principle. there are two methods for measuring the neutron dose equivalent. One is using the recoil protons produced in hydrogenous materials by neutrons, that is to say, the neutrons are detected by recording the proton. Other is that after fast neutrons are slowed in hydrogenous materials, the LY particles which are produced by

6 the nuclear reactions (for example, '?B(n,cr) ? ~ i and Li(n,a) 3 ~ ) , are counted. The BF, counter and LiI, ZnS(Ag) scintillator belong in this kind of detectors.

The neutron dose equivalent meters fabricated by latter method are described herein. One slow neutron detector is positioned at the center, a layer of thermal neutron absorber (for example Cd or plastic contained B) with small holes enclose the detector. A thicker layer of hydrogenous material enclose the absorber. Incoming neutrons are subjected to slowdown, become into thermal neutron, and then recorded by the detector. Because the slowing-down layer is thicker and there is absorber in it, therefore, the detector has the higher efficiency for neutrons with higher ene re , and has the lower efficiency for neutrons with lower energy. Suitably regulating the thickness of slowing-down layer and the hole area opened on the absorber, the energy respondency can coincide with changing tendency of neutron dose equivalent transfer factor f, with neutron energy En. In this way, the detector can reach to tissue equivalent. and measure the dose equivalent value in Sv directly. The dose equivalent meter can be formed into sphere shape or cylinder shape.

5.1.3 OTHER MEASUREMENT METHODS

Application -- - of -. G-M Counter in Dose Measurement As one kind of sensitive radiation detector. the G-M counter have been used widely. But the respondency of G-M counter has no direct relation with absorbed dose D or exposure X. However, suitably selecting the wall materials of counter o r adding some shielding or filter outside the detector, then the respondency of counter can be proportional air absorbed dose, air Kerma or exposure.

Assumption that a beam of X or y rays with energy E, inject on to the G-M counter, photon flux rate is q and counter detecting efficiency for photon is q, then the counting rate is

n = q 0 q (5.1 - 10) Assumption the exposure rate of the position which the counter is

at is X , the mass energy absorption efficiency of X or y rays in air is

(pen / p),, and the bremsstrahlung radiation can be ignored, then from the Eq. (3.3-1 8) in Chapter 3 and Eq. (5.1-10) following equation can be obtained

From the Fig. 5.1-5, for the G-M counter fabricated by different cathode materials (Al, Cu, Pb) its detecting efficiency is almost proportional to photon e n e r a in certain energy range, that is to say, the ratio of q / E, in Eq. (5.1-1 1) is approximately a constant. In addition, the mass enerzy absorption coeficient of gamma rays in air is approximately a constant in cenain energy range. Hence, Eq. (5.1-1 1) can be rewritten into

where

Under the conditions mentioned above, K, and K, is approximately a constant, that is to say, exposure rate and air absorbed dose rate is roughly proportional to counting rate of counter. In this way, the exposure rate and air absorbed dose rate can be defined from the measured munting rate.

The respondencies of different cathode materials for X or y rays of unity exposure are shown in Fig. 5.1-6. From the Fig. 5.1-6, it can be seen that the deviation between the reading of Cu cathode G-M counter and its average value is in i 5% for energy range 0.5 MeV to 2.5 MeV. This result can satisfy the requirement of accuracy for radiation monitoring. In the mean time, because the photoelectric effect between the injection photons and cathode materials is dominant in low energy range, the counting rate for unity exposure rate and unity air absorbed dose rate sharp rise with energy E, decrement. Hence, the G-M counters are scarcely ever used to measure the radiation field of low energy X or y

rays. But if the G-M counter has been calibrated priori with a radiation source whose spectrum is the same for the spectrum of radiation field to be measured, and then the results obtained by using that G-M counter to measure the relevant parameters of radiation field to be measured are reliable. In addition, adding filtering foils on the outside of detector, the respondency of detector for low energy range can be improved, shown in Fig. 5.1-7.

The dead time of G-M counters are larger, they are not suitable for operation in high radiation field. If operating in high radiation field, the counting loss is quite severe, so the "Block" phenomenon occurs. But, because the sensitivity of G-hi counter is high and construction is simple, it can be fabricated into portable monitoring of low level dose. Therefore. i t is widely used in radiation monitoring at present.

When the atomic number of scintillator is closed to air effective atomic number, then the dependence of the ratio in Eq. (5.1-17) on energy of rays is very small. For most of the elements in inorganic scintilators, its atomic number are more larger than that of air (Z, =7.64). Hence, the change of respondency of inorganic scintilators with energy of photons is very large. But for organic scintillators, which consist of hydrogen and carbon can be considered as air-equivalent in large energy range. It can be seen from Fig. 5.1-8 that for NaI(T1) scintillator there is no flat in curve, and for anthracene scintillator the curve is flat between 0.2 MeV and 5 MeV. In practice works, the organic scintillator can be used to measure exposure, Kerma and absorbed dose.

Comparing with G-M counter, the advantages of using scintillation detector to measure dose are high sensitivity, and useful for field monitoring of low dose.

Film Dosimeters The film dosimeters are mainly used to measure - --

the personal dose of X or y rays. When charged particles pass through the nuclear emulsion, they interact with silver bromide grains. The change takeplaces and forms latent image on nuclear emulsion. After the nuclear emulsion is developed, the silver bromide grains formed into latent image on nuclear emulsion are reduced to silver atoms, and then the color of nuclear emulsion changes into black. According to blacking degree of nuclear emulsion the dose level can be judged. The blackness of nuclear emulsion can be measured by lumious densometer or nigrometer. Let I, express the lumious intensity injected on the nuclear emulsion, and the I express the lurnious intensity transmitted through the nuclear emulsion. The larger the dose, the blacker the emulsion, and the smaller the I, therefore, the ratio of I, to I can reflect the dose level. Normally, the logarithm of ratio of I, to I is defined as the blackness, i.e.

1 0 S = lg- I

The respondency relation between the blackness of emulsion and exposure is shown in Fig. 5.1-9. The measurement range of the film dosimeter is the dose range relevant with BC section in Fig. 5.1-9.

The films with different sensitivities have different measurement range. There are some differences in ene ru respondency for different film. In normal case, the film respondency to photon energy below 100

-275-

keV is many times higher than that to hizh energy photon. T o improve the film energy respondency, normally, a layer metallic filter (for exarn- ple. Cu, Cd, Pb, Sn and Al) covers outside film, to decrease the film sensitivity for low energy photons.

The inadvantage of film dosimeters is that the latent image formed on the film is easily affected by surrounding temperature and humidity. In high temperature and high humidity, the latent image can decline, so that the result are not correct. The maximum advantage of film dosimeter is that the treated films can be kept in files, if necessary it can be re- newed to access and to compare.

Glass Dosimeters The glass dosimeters are made by using the - - .- - --

principle of photoluminescence. The fluorescent glass is made of phos- phate of alkali metal glass or alkaline-earth metal glass in which is added slight silver metaphosphate. According to the atomic n u m k r of compositions contained in glass, the fluorescent glass can be classified as high z glass (for example, Ba glass) and low z glass (for example, Li glass).

Under action of radiation, the electrons produced by radiation in fluorescent glass come into the conduction band, and they are captured by some deeper traps. This kind of traps is the Ag ions doped into the glass. After capturing the electrons, the Ag ions become into sub-stable state and ~ g + ~ that forms the luminous centers. When the fluorescent glass formed luminous center is exposed under the exposure to ultraviolet rays with wavelength 465 nm, the captured electrons transit to excitation state, and then transit to ground state with emitting the fluorescence. In certain dose range, the concentration of luminous centers is proportional to dose subjected by glass. Hence the fluorescent intensity can be used to measure the dose subjected by glass.

The respondency of fluorescent glass is poor for low energy photons, for example, shown in Fig. 5.1-10, because the glass contains the composition with high atomic number, especially for Ba glass. Hence, when glass is used in low energy photon radiation field, it is necessary to add metallic filter (for example, Sn, A1 and Cd) with suitable thickness to improve energy respondency.

The fluorescent glass contains the 6 ~ i . '9, 'O'A~, therefore it can be used to measure the thermal neutron dose.

The measurement range of fluorescent glass is wider (104Gy - 102Gy). The fluorescent intensity can be measured repeatedly. After high

peaks can not very high and very low. For LiF, the better temperature is about 200 C . Tn practlce there are several peaks for many therrfloluminesence elements, the peak used in measurement is called as main peak. For LiF, the 5-th peak is used in measurement.

The special instrument is needed to measure the luminous intensity of thermoluminesence elements. For one given thermoluminesence element, at the same heating temperature the different areas are relation to different dose. Therefore, to measure the integral intensity of luminous peak under main peak, i.e. to measure the area of luminous peak under main peak, can determine the dose. This method is called as integral method. In addition, the dose can be defined by measuring the peak height. This method is called as the peak height method. But the consis- tence of heating speed for thermoluminesence element must be controlled, because the height of luminous peak depend on the heating speed.

The thermolurninesence dosimeter can be used to measure the dose of X, y rays, electron beam and fi rays with higher e n e r a . When using thermolurninesence element CaF,, CaSO, to measure the photons with low energy their thermoluminesence respondency has large energy dependence. In this case, it is necessary to use the filtering method to improve the energy respondency. For LiF, Li,B,O,(Mn) and B e 0 because they contain low z materials only, and their effective atomic number is closed to air and tissue, their energy respondency dependence to photon is smaller. If certain quantity 6 ~ i and '9 is contained in thermoluminesence materials, then they can be used to measure the neutron dose.

The thermolurninesence dosimeters have wider measurement range. For LiF the range is from 5 x I O - ~ G ~ . For Li,B,O,(Mn) the range is from Gy to 10, Gy. But the respondency of LiF is non-linearity over 5 Gy. The sensitivity of thermoluminesence element is high, its volume is small, and they can be used repeatedly. Its accuracy is high. The effects of environment changes on measuring result are small. But some thermoluminesence elements can decline under normal temperature, for example, the dose value can decline about 50% at the end of first month after radiation. In addition, the thermoluminesence elements are taken reading through heating, so the irradiation information stored in them disappear immediately, therefore, they can not be taken the reading re- pea tedl y.

'~l~

ern

~o

lr~

min

ex

en

ce

el

emen

t

Eff

ecti

ve

ntom

ic

nrlm

tmr

Wn

vo

len

~tt

i of

nits

xini

r~m

lr~

n~

ino

~~

s

pen

k(l

O-l

Orn

)

l'o

m~

~rn

ti~

re

of r

n'ni

n I

r~

n~

ino

~~

s

peak

('(.:

I

N~l

rnlr

dl. of

lr~

niin

oris

po

n k

I m

an

y

c ,, e

r~si

tiv

ity

for

1 .I.

rnye

of

''To

(rel

ativ

e to

LiF

)

Ln

erg

y r

espo

ndon

cy

NK

eV

/"T

o

Dec

line

in

fir

st.

horl

r,200

/,,

in f

irst

. th

enti

i;6

per

yea

r m

onth

,100

/,

in t

hre

e m

onth

e?, 1

VX

Th

ern

inl

neir

t.ro

n 4.

8X10

2 re

ap

nd

en

cy

"'

(1)

I~im

ino

risc

~~

inn

tity

ra

tio

of

lo

-" G

y th

erm

al n

elit

rons

(3.

7 ~,101° t

hern

ir\l

nor

~tr

onp/

cni'

) lo

lo-

" a

y Y

rn

y~

or '

"'(!n

.

5.2 DOSAGE MONJTORING

5.2.1 INTRODLJCTION Jn the system of dose limitation, the dose equivalent and other rele-

vant quantities H ~ , He and H,,,, provide the scale for accessing thc

damage made by radiation to body. Hence, the first task of monitorings, which are to search the safety conditions to be satisfactory, provide enough data to estimate the exposure workers by using those quantities expressed the basic limits. However, the average dose equivalent of one organ or tissue and the intake of radionuclides can not measured directly in practice. I t is necessary to estimate them according to that quantities which can be measured direct. Therefore, the measured results must be linked quantitatively the values to be estimated through mathematical model.

Due to the scale and properties of radiation protection have large difference under different cases, therefore, that the radioactive areas of rector building are divided into different zones has the practical meaning. In this way, on the one hand the controlled extent can be sim~lified, on the other hand it is easily to make plan which is economy, and in the mean time which does not reduce the safety standards.

The properties of radiation monitoring are as follows. The radioactive level of medium monitored is very wide. The different sensitivities of monitoring instruments are required. The low level radioactive measurement and microanalysis. The subjects monitored are complex, for example, air, water, living beings, soils,foods and surface of body. There are many interfluence factors in measurement. Hence, many kinds of relevant technique, for example, sample collection, sample treatment, measurement and analysis, are needed. In addition, the automatic monitoring and data processing systems are needed also.

The radiation protection monitoring can be classified as individual dose monitoring, working area monitoring and environmental radiation monitoring. According the different aims the individual monitoring and the the working area monitoring can be classified as routine monitoring, operational monitoring and special monitoring. The routine monitoring is relative to continuous jobs. The operational monitoring is for providing relevant data with one job. The special monitoring is used in practical existent abnormal case or doubtful case which may occur.

5.2.2 INDIVIDUAL DOSE MONITORING The individual dose monitoring is the monitoring of practical dose

received bv person. It includes individual external exposure monitoring. skin contamination monitoring and inside body contamination monitoring.

Individual - External Exposure Monitoring The individual exter- - - - -- -- - -. -

nal exposure monitorin is aim at accessing, recording and controlling the dose equivalent value received by personnel, or measuring and estimating the dose of exposured personnel when accident occurs. The individual monitoring is for such personnel which receive the dose that may to be higher than 3 / 10 of the dose equivalent limit.

Under external exposure. assuming that the simple protection measures have been taken, and the persons can not take the radioactive source by bare hands, then when the activity of radiation source is lower than following values the individual external monitoring may consider not to make. For y radiation source (all of its /I radiation has been shielded) the value is 50 MBq MeV, for /I radiation source (with or without y radiation) with energy Em,> 0.3 MeV the value is 5 MBq, and Em, <0.3MeV, the value is 50 MBq.

The basic means of individual dose monitoring is using personal dositmeters. The personal dosimeter is worn on the place of body surface where the measurement result is representative of dose received by whole body or local tissue as possible. For personal dosimeter, following requirement must be satisfied. Its energy respondency is well, its direction dependence is small, it has suitable measuring range, its measurement result is reliable, the dosimeter's volume is small, its weight is light, its structure is solid, and it is worn easily.

In individual dose monitoring the selection of personal dosimeter is based on the kind of radiation field, radiation energy, dose range and environmental conditions. For /I, X and y radiation field, normally, the ionization chamber type personal dosimeters, film dosimeters and thermoluminescence dosimeters are used. For neutron radiation field, the nuclear emulsion dosimeters and albedo neutron dosimeters are used. For the monitoring of persons worked in high dose areas, the dosimeter can give the dose value in time. and must have alarm indicator with sound and light. For neutrons, the induced radioactivity of hairs and metallic foil in clothing of exposed workers can be used to define the

neutron dose of local part of exposed workers. Using whole-body counter to measure the induced radioactivity inside man body can define the whole body dose received by personnel. Under large dose exposure, the chromosome aberration analysis has important meaning.

For the personal dosimeter worn on surface of personnel body, if there are tissue equivalent materials with thickness 10 mm or 0.07 mm covered on the detecting element, and this assembly has been calibrated, then the result measured by this dosimeter can be considered as deep personal dose equivalent H,(10) or as shallow dose equivalent HJ0.07). These two quantities can be considered as reasonable estimation of effective dose equivalent H E or deep dose equivalent index H,,, or shallow dose equivalent index H,,, under relevant exposure condition.

When large exposure accident occurs only, it is necessary to estimate in more detail absorbed dose received by organ or tissue. In this case to measure and to access the exposure received by parts of inside body, it is necessary to do mock-up experiments to define radiation fieid distribution.

Skin Contamination Monitorings Skin contamination is one p.p-p ~ pp.-p---p-p-

source of external exposure of man body, in the mean time, the radioactive materials contaminated skin can transfer into inside body to induce internal exposure. The surface contamination of skin is determined by using surface contamination monitor.

The measurement results are accessed by using the derived limits of surface contamination. If the contamination levels do not exceed these limits, normally, it is not necessary to estimate the radiation dose induced by contamination. If decontamination is difficult or initial level of contamination is very high, it is necessary to estimate the dose equivalent, although this estimation is not very accurate. If estimated dose equivalent value has been exceed 1 / 10 of relevant limits, then this value must be recorded into personal dosage file.

Inside Body Contamination Monitorings Normally, the following personnel need make inside body contamination monitoring.

1. The personnel of contacting large amount of radioactive gaseous volatility materials, the personnel of operating contained tritium and other luminous materials, and the personnel which may breathe in oxidate of tritium.

2. The personnel which are engaged in natural uranium treatment and riched uranium treatment and nuclear fuel fabrication.

3. The personnel which are engaged in plutonium and other transuranium clement treatment.

4. The personnel which are engaged in production of large amount of radionuclides.

When the monitoring results of working areas indicate that the personnel may intake large amount of radioactive materials or are in doubt whether the personnel may intake large amount of radioactive materials it is necessary to make special monitoring.

There are three kinds of method for inside body contamination monitoring. One is to estimate the accumulative quantities of radioactive for tissue through outside body measurement. Second is biological test, i.e. measuring the radioactivity of sample of man body discharge matters and saliva, perspiration, blood and hair, estimating the accumulative quantities inside body or in tissue from this measurement. The final is to measure radioactivity content in whole body, lung and thyroid gland directly, and it is the most suitable method to estimate the inside body contamination of nuclides which emit the gamma rays. Normally, the whole-body counter is used.

Frequency of inside body contamination monitoring is depend on retention time of intaken radioactive materials in body and detector sensitivity. The design of monitoring time interval and monitoring technique must be able to detect the all of main intaken radionuclides or large part of them.

For accessing inside body contamination, to establish a kind of model is necessary. The quantities measured in inside body contamination monitoring are in contact with the relevant secondary limit through this model. In order to establish this model, not only is it necessary to consider the radiological characteristics of radioactive materials and metabolize law in body of them, but also it is necessary to consider the intake mode of radioactive variation with time.

Once the relation mentioned above is established the reasonable accessment can be made for results of inside body contamination monitoring.

5.2.3 AREA MONITORING The area monitorings include the external exposure level moni-

toring of 0, y and neutron in working area, surface contamination monitoring in working area and the airborne rahoactive material concentra-

-283-

tion monitoring in air. External Exposure Monitoring of Working Areas Before any

--- ---p..ppppp-

new radiation device is delivered i t is necessary io estimate the level of external exposure radiation field induced by sourcc term in working area to provide basis for working out the design project of routine monitoring. The frequency of routine monitoring is depend on anticipated change of radiation environment. But if the change of radiation environment is slow, then it is only necessary to make periodic o r temporary test a t preset observation point, in this way, the change of radiation level can be fully reflected. If the radiation field is more stable it is seldom necessary to make routine monitoring of working area. If the radiation field changes easily, the change rate of radiation field is quicker and the severe extent of change can not be predicted, then one alarm system is needed. This alarm system is set in dose measurement room or working area, and is worn by personnel.

The working out of project of operational monitoring is depend in large extent on seeing if the radiation field in working area keeps constant throughout. Under constant condition, normally, it is enough to make a tour of inspection for radiation level of the area in which the personnel are working. If operation itself may induce the obvious change of radiation field it is necessary to do a series of continuous measurement for operational site during operational process.

The instruments used for monitoring of external exposure radiation field in working area are fixed o r portable radiation monitors. The detectors used for these types of instruments, normally, are ionization chambers, G-M counters and scintillation counter. The requirement of fixed monitors is able to give the alarm signal automatically when the radiation level exceeds the preset value.

However, because the characteristics and level of radiation in working area variate with space and time, and the active modes of personnel in working area either can not be predicted o r are difficult to understood exactly, hence the measurement results used to derive the exposed dose of personnel is very difficult. Therefore, one simplified assumption must be introduced.

In order to safety and convenience, it is assumed that the personnel is a t the point where the radiation level is highest in working area during whole operating time. In this way the upper limit value of exposure of personnel can be defined, the advantage is that as long as this upper lim-

it value is fBr lower than relevant limit, then it is not necessary to limit the activity of personnel iil working area, and to limit stay time of personnel in working area. In practice, in such working area the exposure of personnel will be far lower than tne relevant dose limit.

If i t is too difficult to keep the dose equivalent in working area at enough low level, then it is necessary to estimate dose disu-ibution in working area and to divide the working area into different zones, or even to limit the time closed to some radiation level area at a times.

During monitoring of operational process, the accessment work is often to define one special working time interval. In this time interval the exposure of' any individual can not exceed one level.

Surface ~ Contamination Monitoring in Working Area The main aims of surface contamination monitoring in working area are pre- ventin8 contamination dispersion, checking up contamination control is failure or not and the rules of operation are violated or not, the surface contaminations are limited in certain area and below certain level, and io provide the data for working out the project of individual monitoring and air monitoring, and rules of operation.

The contamination monitoring of working area where the radioactive materials may leak slowly can be realized through checking up the surface contamination of the filtering bag, working shoes, glove and clothing pockets. Although the method mentioned above neither can detect little amount of isolative occurred contamination events, nor can estimate quantitatively surface contamination level, it can give indi- cation of contamination level.

For the area where the radioactive contamination occurs easily, in order to prevent that personnel bring large amount of radioactive materials out of working areas it is necessary to set the contamination monitor at change room and exit of working areas, to be able to detect contamination accident. In some cases the monitoring results of operational process are of advantage to avoiding and limiting the contamination dispersion during operational process.

The main radiation kinds needed to monitor the surface contamination are a and p radioactivity. Their monitoring methods can be classified as direct measurement and indirect measurement.

The indirect measurement may use wipe method. It is mainly used for the surfaces which is inconveniently for direct measurement or for the monitored surface near which there is strong radiation background,

-285-

therefore the direct measurement can not be made. Sometimes in order to test the contamination extent of surface initially the indirect measurement can be used also.

1: must be noticed that in order to obtain the result it is necessary to select the measurement method and instrument correctly. In order to make the results reliable it is better to calibrate the instrument by using standard source which has the same nuclides with the monitored nuclides before measurement, and to make the geometry conditions same with calibration. The detectors used for contamination monitoring are G-M counters and scintillation counters.

The relation between surface contamination and exposure of personnel is quite complex. At present, the derived limits about surface contamination are arbitrary in some extent. In order to access the measurement result of surface contamination, it is necessary to make them in contact with the derived limits of surface contamination. If the level of surface contamination in working areas is lower, or more lower than relevant derived limit, then it is not necessary to make other form contamination monitoring. The derived limits of contamination for radioactive materials are listed in Table 3.2-3.

Air Contamination Monitoring It is necessary to make monitoring to air contamination for following cases: operating large amount of radioactive materials, the leakage of radioactive materials is possible and there are severe contamination in working areas. By this way, the extent of air contamination can be understood. the quantities of radioactive materials breathed in by personnel can be estimated in some cases and the leakage accident of a system or a part for reactor can be judged.

The mode of air contamination monitoring in most common is to use the air sampler. The air sampler is often set at the place which can represent the breathing zone of personnel. In order to detect unexpected air contamination it is necessary to set continuous monitoring devices which can sample and measurement continuously, and once the concentration exceed the preset value the monitoring devices can give the alarm signal.

The air contamination monitoring includes the radioactive aerosol measurement in air and radioactive gases measurement in air.

1 . radioactive aerosol measurement. Because the derived air contamination of aerosol is very low, the

concentrated method is used to coilect the aerosol. Because the concentration of i~atural-born radon. thoron and their daughters is rrlore higher than derived air concentration of art~ficial radionuclides, the air sample, which has just been collected. contains lot of radon's anci thoron's daughter which conceal the radioactivity of artificial nuclides to be measured very easily. In order to measure the contamination of the artificial radionuclides it is necessary to separate the artificial radionuclides from the natural-born radionuclides. At present the following methods are used: decaying method, energy discrimination method, false coinci- dence method and cc / ,!? ratio method. The decaying method is easily realized and used widely.

The decaying method makes use of the characteristic of short half-life for the daughters of radon and thoron. The sample is laid up for 4 days after sampling the short half-life daughter of radon and thoron can rough be considered to decay away. I n this way, measured concentration after 4 days is the concentration for long half-life artificial radionuclide aerosol.

2. radioactive gas measurement. 41 The air in reactor building can produce the radioactive gases Ar,

'%, 190 due to neutron activation. Once the accidents of reactor occur the fission product gases I , 3 5 ~ r and 13'xe can be discharged. Surrounding of heavy water reactors there are 3 ~ . Hence, the concentration of radioactive gases in air must be monitored. The measurement methods of radioactive gases should depend on physical and chemical properties of radioactive gases. For /I radioactive gases, gas flow ionization chambers, thin window proportional counters or G-M counters, and scintillation counters with thin plastic scintillators are often used.

3. accessment of results of air contamination monitoring. For accessing the results of air monitoring, it is necessary to under-

stand the physical and chemical properties of contamination materials, and the air sample can represent the breathing in level of personnel in how much extent. Normally, some simplified assumptions should be made. In practice, some simplification have been implied in ALI and DAC.

In some cases, routine operations can induce regular air contamination. In this time, it is necessary to make detail investigation for air in different stages of operation and a t the point whicn can represent the

-287-

breathing zones of personnel, and then the intake quantity of personnel is accessed for one whole operational period. Tf operational period mentioned above is the typical represent of normal operation for considcra- ble long time, then the total intake quantity for one person in corresponding time can be estimated.

Tf the air contamination sources are local, and they variate with time, even sampler close by the breathing zone of personnel the intake of personnel may not be given. Hence, the representative of sample collected by site sampler is depend on air sampling rate, characteristic of grain selecting and action of personnel in area in large extent, It is necessary to compare the long term monitoring results between personnel sampler and site sampler, or through other detail investigation, and then to estimate the representative of sampler. In this way, one derived limit of relevant site sampling can be obtained. If this comparison has not been done yet, then the suitabie method is assuming that the breathing in quantity of personnel is higher than about one order of magnitude for site monitoring result. It is notice thzt this simplification is not accurate. This method is suitable for accessing long term and average results, but it can not used to access the short time and single measurement results. Be- cause the difference of result between the personal sampler and the site sampler has two or three orders of magnitude. Therefore, the measurement results of site sampler in short time can not be used to derive the personal possible intake. As one replacement method, it is to set investigation level. Such investigation level must be enough low, so that some noticeable abnormal cases can be detected, but such investigation level should set at certain level to voide often to reach this level under normal cases. For the materials with low derived air concentration, choice of investigation level may be defined according to the reachable measuring sensitivity and possible existing (natural or artificial) radionuclide concentration in air.

5.2.4 ENVIRONMENTAL MONITORTNGS The environmental monitorings is one important link for protecting

the environment. Either it is basis of accessing effects of radioactive work on environment, or the accidents and hidden perils can be discov- ered. The environmental monitorings include background servey, routine monitoring in operation and accidental servey. The items of environment monitoring are air, water and soil contamination monitoring,

-288-

radionucl~de monitoring in animals and plants, a, D. y contamination test on ground surface and gamma radiation monitoring in environment.

Routine Monitoring The aims of routine monitoring are under- - - - . - - - -

standing contamination extent to environment. contamination laws and contamination tendency, estimating the dose equivalent commitment and collective dose equivalent of individual of public and accessing the detriment and far-reaching influence induced from contamination, testing the efficiency of waste management and providing the scientific basis for the measures of waste management.

The design of routine monitoring program includes the range of monitoring, arrangement and number of monitoring point, sampling and sampling frequency, sampling kinds, analyzing methods, nuclides measured and analyzed and quality assurance.

The general effluents for one plant are gas, aerosol and liquid. There are two methods for measuring effluents of plant. One is direct measurement, i.e. detectors are immersed in effluent or closed by the dischare pipe. Other is to measure and analyze the emuent after sampling. Ordinarily, the activity of a and is measured, sometimes the gamma spectrum is measured to define the concentration of gamma radionuclides. I n sampling measurement, it is necessary to make the sampling with representative. According to different cases the sampling can be classified as continuous sampling, periodic sampling and special sampling.

Emergency Monitoring The aims of the emergency monitoring under accident conditions are discovering the discharge quantity of harmful materials in time, getting quickly the data about the range and extent of environment contamination t o take the emergency measures, to reduce the detriment, to estimate the dose of public and to access the detriment of accident to environment and public, getting the scientific data of dispersion and transfer for harmful materials in environment.

The emergency monitorings mainly include building the monitoring and the alarm system of environmental contamination accidents, tracing discharged harmful materials, pre-accident measurement and post-measurement, monitoring of contamination tendency and accessment of accidental detriment. For emergency monitoring, there must be plan, preparedness and organization in advance, it must have idea of type and scale of possible occurring contamination accidents and

compositions, quantities, influence range and extent of discharged harmful materials for plant. The instruments must be simple, reliable and portable as possibie, and are often maintain to make them in normal operation state. The techniques and methods are the same as that of routine monitoring in many aspects. But quick respondency is required for emergency monitoring, there are different in some aspects for measurement techniques and methods.

Radioactive Background Servey of Environment The radioactive -- - - - -- - --

background servey is mainly for understanding the background level of harmful materials (included natural background, nuclear materials discharged by other enterprises near by the plant) and its change law before operation of plant to provide the basis for accessing the effects of after operation on environment.

Besides investigating relative data of design for plant itself, the contents of environment background investigation include natural environment (for example, terrain, landforms, water sources, hydrology, water quality, meteorological phenomena, living beings), its utilization (irrigation, aquiculture, herding and cultivation) and society conditions (resident distribution, dietetic custom and waste discharge conditions). Sometimes to obtain the data of dispersion and transfer for waste gases and waste water in environment, the dispersion experiments of atmosphere and water body in environment.

The objects of environmental background investigation include air, water, terrestrial and aquatic animals and plants, soil, water body properties, precipitate and foods. The harmful materials investigated include radionuclides, gamma radiation field and the non-radioactive materials which are relation to plant itself. The measuring methods and instruments used in environment background investigation should have enough sensitivity to ensure that the background levels of harmful materials can be measured accurately.

The time of environment background servey a t least lasts one year, and the samples collected must be retained to use later.

In environment background investigation the radioactive analytical instruments used in laboratory are cx counters (often used are ZnS(Ag)) scintillation counters, windowless or thin window gas flow proportional counters, silicon barrier detectors and nuclear track detectors, low background a and measuring devices and low background gamma spectrometers.

The data of environment background investigation is the important basis for accessing the result of routine monitorin_e.

Quality Assurance of Radiation Measurements To make the -- - -. ~ ~- - - ~ - - --

measurement results of radiation measurements reach tc certain accuracy, to reflect practical radiation level of' environment, the strict quality assurance is necessary. The quality assurance must be executed ehrough- out the whole process of environment monitoring, including sample collection, transportation, storage, treaunent, operation, purification, measurement, data processing and result explanation. Quality control measures, which are relative to measurement in laboratory, are simply introduced herein.

1. New installed or maintained measurement devices must be ad- justed, tested and calibrated.

2. For main characteristics of measurement devices used in routine measurements, they should be routine tested.

3. Standard sources and instruments used in test and calibration must be correctly use according the accuracy class.

4. The testing results of characteristics of measurement devices must recorded in the quality control fils, and drawn on the quality diagram. When the testing value fall out of outside of controlling limit which is corresponding to three times standard deviations, or fall out of outside of controlling limit which is corresponding to two times standard deviations for two times continuously, the reasons must be checked, an re-calibrated must be done.

5. To discover the system uncertainty produced in measurements and analyses of laboratory the contrast between native laboratories and this laboratory must be done.

6. The measurement conditions and quality control conditions (for example, standard sources, calibration of measurement instrument characteristics, testings, maintains, fabrication of reference sourcesj must be recorded and kept properly.

7. The personnel of engaging in measurement should have relevant knowledge and technical level for measurement devices and measurement methods. and is trained and examined periodically. -

Environment Quality Accessment The environment quality accessment is a important content of environment protection jobs, and sometimes is a important means of environment management jobs. The environment quality accessment is a quantitative description of good

and poor environment quality. Its aim is to master the law of environment quality change, to recognize the positions and roles of environment factors in man-induced environment, and then to provide the reliable basis for protecting the environment.

The exposure pathways to persons are variety for radioactive materials discharged to environment by plants, shown in Fig. 5.2-1 and Fig. 5.2-2. In these pathway, there are one pathway or two pathways that are more important than other pathwav. Such exposure pathways are called as "critical exposure pathwayN. The nuclides discharged to environment by plants are variety, but there must be one or two nuclides whose detriment to body through critical exposure pathway is more important than that for other nuclides. Such nuclides are called as "critical nuclideN. The dose equivalent produced by the nuclide through pathway for public is different. There must be one group of public whose exposure is higher than that of other group of public due to their occupation, living custom, residence position and age. This group of public is called as "critical groupN. T o get reliable data and to access easily, while working out the discrimination program it is necessary to consider the critical exposure pathway, the critical nuclides and the critical group.

For accessing effects of radionuclides discharged to environment on environmental quality. On the one hand is to estimate individual average dose equivalent and dose equivalent commitment, and on the other hand is also to estimate the collective dose equivalent and dose equivalent commitment for whole exposed groups and to compare to relevant dose limits. Therefore, it is necessary that the exposure pathway to body after radionuclides discharged to environment are described approximately by one model which is constructed by some reasonable assumptions. The accessment model should characterize the physical and chemical properties and states of radionuclides to be discharged, transportation ability and dispersion ability, the exposure pathways and characteristics of food chain, and the intake characteristic and supersession characteristics of body for radioactive materials.

Eff

icie

ncy

(%

)

I ' - / - anthracene,thickness Ocm E - - - anthracene, thickness 2cm j

0. J anthracene,thickness Ocm ! b) 1

i c - : ; , I - ' ! G o

photon energy, MeV

Fig 5.1-8 relative energy response for s c in t i l l a t o r s (Ail of the curves are normalized a s 1 a t 1 h4eV)

Fig 5.1-9 relationship between film blackness S and exposure

CHAPTER 6 RADIOACTIVE WASTE TREATMENT

6.1 GENERAL DESCRIPTION The discharging to environment of radioactive materials of nuclear

power plants, especially the environmental contamination induced by such discharge, is one of the important contents for accessing the safety of nuclear power plant.

The radioactive waste treatment system can collect, treat, storage and pack the waste by suitable means to avoid affecting the generation output or utilization ratio of nuclear power plant. This system can reduce the radioactive materials, discharged to environment or plant, to least. These radioactive wastes can induce over risk for health of public or personnel.

6.1.1 PRINCIPLES O F WASTE TREATMENT First, it is necessary to reduce the waste production quantities and

radioactive concentration as possible. This requires to ensure fuel clad- d n g for integrity and primary system for leak tightness as most as possible. In order to satisfy the requirements mentioned above, following measures are used: using zirconium alloy cladding, using corrosion-resistant materials to fabricate the equipment, pipes lines and valves, adopting leakless and testing leakage measures, Setting introduction leakage pipe lines in the parts which may produce the leakage, setting tight containment.

Second, the wastes which have been produced must be collected and controlled to avoid dispersion of radioactivity. For these aims, setting collection systems of wastes, classifying the wastes according their kind, property, and rahoactive level.

Final, the wastes which have been collected should be effectively treated. After treating the waste waters and waste gases should be utilized circularly as possible. For the wastes to be discharge and stored, their volume should be reduced as possible. Before discharge, the rahoactive levels must be measured, the radioactive levels of wastes must coincide with discharge standard.

The waste management is in close relationship with design, fabrication and operation of nuclear power plant, every stage must be paid

-399-

great attention to. Fig. 6.1-1 is the principle diagram of waste tieatment.

6.1.2 FUNCTTONS OF TREATMENT DEVICES The functions of waste treatment devices are as follows. First,

wastes are collected and then are stored, to make the nuclides with short half-life decay for reducing their radioactive levels. The wastes are treated bv different methods for different property wastes, to reduce their volumes. The radioactive levels of wastes are measured. For the wastes which can be used repeatedly they are fed to relevant systems and used again, for the wastes which can not be used repeatedly, they are discharged. but the activities of wastes to be discharged must be coincide with relevant standards.

6.2 CLASSTF'ICATION AND SOURCES OF WASTES

6.2.1 CLASSIFICATION O F WASTES According to the physical states of wastes, they can be classified as

liquid state waste, gaseous state waste and solid state waste. Hence, the waste treatment devices can be classified into liquid emuent treatment systems, waste gases treatment systems and solid waste treatment systems.

The liquid effluents are utilized again or discharged to river. The gaseous wastes are discharged to atmosphere and the solid wastes are stored by loading into drum.

6.2.2 WASTE SOURCES Radioactivity Sources During the reactor operation of nuclear

power plant, large amount of radioactive materials produced by nuclear fissions are the main sources of radioactivity for nuclear power plant. Besides these sources, capture reaction of neutrons form certain quantities of transuranium elements and activation productions of structural materials. These processes make the reactor become into one huge radioactive source. Table 6.2-1 lists the radioactivity accumulated quantities at the end of equilibrium circulated life for one PWR with 1000 MWe power.

Table 6.2-1 Equilibrium Radioactivity in Core of One PWR with 1000 MWe Power

Radioisotope Activlty (3.7 x 1 o ' ~ B ~ )

Flsslon products: -4 isotopes of Kr 1

I isotopes of Xe j 5030 I isotopes of I 595

r a r e e a r t h elements; 1110

other --

Total quantity of fission products Fp I 1 1970

I Actinide elements 361 1 I Neutron activation products i 12.9

Total I 15597

While the fuel element claddings are integrity, the fission products and the transuranium elements are closed in the fuel element claddings. While fuel failure, a part of fission products may come into the coolant. After the structural parts of reactor are corroded, the activation products of structural materials may come into coolant also. In addition, when the impure atoms in coolant pass through reactor core, they can be activated by neutron irradiation. When fuel failure is 1 % for PWR nuclear power plant with 1000 MWe the concentrations of fission products and activated corrosion products in coolant are listed in Table 6.2-2. The radioactive inventory and composition (without 3~ and noble gases) discharged to waste water of reactor outside from the coolant system of PWR nuclear power plant is listed in Table 6.2-3. The noble gases come into primary loop coolant system are about few 1016 Bq per year, most of them are short half-life Xe and Kr. They decay themselves, hence the Xe and Kr come into waste water are very small.

Table 6.2-2 Radioactive Concentration of Fission Products and .4crivated Corrosion Products in Primary Coolant of PWR Nuclear Power Plant with IOOOMWe

Radioisotope ~

half-lifc concentration (3.7 x 1 0 " ~ ~ / I ) --- - -~ ---.. ~ - - .

us 10.73 a. 1 . 1 0 ~ lo-' . 36% - - .. . ~~~~ ~ ~ ~ - - ~ --

noble ssm K r 1.18 hr. 3 6 - - ~- - . -

1.16x lo-'

gases 87 3bKr 76.31 min. 0.87 ~ x

produccd ;:'XC 5.29 d. 1.74 x lo-' .. - ~ .- --- - - -- ~ ~ -~ ..

in : i 3 " ~ e 2.19 d. 1 .97 x 1 0-3 ~ --

fission 135mXe 54 15.6 min.

~ ~

0.1 1 x lo-' - . -- - --

~ ;:"xc 17 min. . - --

0.36 x lo-'

2 : ~ n 312.5 d. 1.2 x ~- ~~ ~ . ~ ~-

activated :$ In 2.567 hr. 2.2 x lo-5 - -- - - -

corrosion 58 71.3 d. 27C0..- -

8.1 x

product ::FC 155.1 d. 1.8 x ~ -~ ~ - - ~ -- ---. ~... - - -

60 27C0 5.26 a. 1.1 x

Table 6.2-3 Radioactive Inventory Collected in Waste Water of PWR Nuclear Power Plant

Radio- Quantity I Radio- , Quantity

isotope (3.7 x 1 0 " ~ ~ / a) ! isotope (3.7 x 1 0 " ~ ~ / a)

91 3gsr 2.62 53 91 1

1 361

- -- - -. . - -. . -

3 1.12 :Z6cs 8 8

;;I- 22.2 ! :?CS 182 -- ~- ~

92 3gy 5.10 56 Ba 2.10

140

-.

9 2 r 1.75 : y ~ a 2.17 --

z r 1.1 1 G'CC 8.23 --

Total radioactivity 30969

Besides the fission products and activated corrosion products the primary coolant itself and some impure elements contained in coolant czn produce the radioactivity by neutron irradiation. The main nuclear reactions are ~ 6 ~ ( n , p ) ~ % . ,6~i(n,a):H and i0~(n,2a):H. The gamma ray energy of : 6 ~ is very high (Ey = 7.1 MeV), but its half-life is very short (T! = 7. Is), it disappears quickly after discharging to reactor out-

2

side. The radioactive energy of 3~ is more small, but its half-life is very long. The quantities and sources of 3~ in primary coolant of PWR nuclear power plant are listed in Table 6.2-4.

Table 6.2-4 3~ Yield Per Year in Coolant for PWR Nuclear Power Plant with 1000 MWe Power

Source I Yield (3.7 x 1 0 " ~ ~ / a)

tri-fission --

! 110

combustible poison rods 10

control rods ! 110 -

-. B in coolant I

-. 5 60

7 ,Li(n,na):H reaction i 11 ;Li(n,rx);~ reaction I 1 6

D(~,?)?H reaction I

I

Total I I 1108

Liquid Effluents The source of liquid effluent can be classified as three types.

1. primary loop drainage which can be utilized directly. The contained tritium water drawn out from primary loop do not

expose to air, but there are certain radioactive level in it. These types of water includes the water produced in flush or emptying of pressurizer relief tank, the water produced by leakage of containment and seal ring of main pumps. These waters are fed into boron recycling system.

2. possible reused drainage. These types of water includes the contained tritium water exposed

to air and water with certain radioactive level. They come from gaseous waste treatment system, overflow of primary loop drainage tank when its level is over high, drainage of all devices loaded the primary loop water. These waters are fed into liquid waste treatment system. They can

be reused or discharged. 3. no longer used waster liquid. Floor drainage, which is non-radioactive normal water (or washing

water withour detergence), is directly discharged after monitoring. But if the devices, which enclosed the radioactive water, leak, then the leaked water is fed into the liquid waste treatmertt system. The detoxicating water of auxiliary systems and their sampling water are chemical water with radioactivity. After measuring, if radioactive level of such water coincide with discharge standard, then the water can be directly discharged or discharged through filtering, if do not coincide with discharge standard, the water is fed into liquid waste treatment system. The bath water, washing and the detoxicating water with detergent belong to common waste water. Such waste liquids are with more weak radioactivity. If their radioactivities coincide with discharge standard after measuring then they are directly discharged. Otherwise, they are fed into radioactive liquid waste treatment system.

Waste Gases 'The waste gases can be classified as two types. 1 . hydrogeneous waste gases. These hydrogeneous radioactive gases (fission gases) come from

discharged gases and purge gas of containers of primary loop water and the storage tank of primary loop drainage. All of these gases are fed into gaseous waste treatment system.

2. oxygen-containing waste gases. They are produced from starting the storage tank and primary loop

vent system under atmospheric environment. The later may be contaminated unexpectedly by radioactive gases. These waste gases are directly fed into stack and discharged to atmosphere.

Solid Wastes Most of the solid wastes come from waste treatment system mentioned above. These wastes can be classified as three types:

a. waste resins. b. concentrated liquid after vaporing. c. failure filter element of filter. The parts and tools contaminated and protection devices on site are

belong to solid waste. All of these wastes are transported to solid waste treatment system

under biologic protection conditions, and then stored.

6.3 WASTE TREATMENT

6.3.1 LIQUID WASTE TREATMENT All of liquid waste will be treated by a certain treatment system.

Liquid waste treatment system consist of boron recovery system and waste liquid treatment system.

Boron Recovery System The operational schematic of this system is shown in Fig. 6.3-1.

Thc main functions for boron recovery system are storage, ueat- ment and monitoring of drainage for liquid wastes of primary loop.

The system can realize following functions in order: a. pre-storage of drainage. b. monitoring of physical properties. c. decontamination of drainage. d. degasification. e. temporary storage with decontamination monitoring function. f. separating two kinds of solution by vaporing method: the water

which satisfies the water quality requirement. Concentrated boron acid solution.

g. storage of above products before possible utilization. Waste Liquid Treatment System The operational schematic of

this system is shown in Fig. 6.3-1. This system has following functions: a. storage. b. monitoring. c. treatment. d. discharging to river. Methods of Waste Water Treatment There are many methods for

treating radioactive waste water, the effective in practice and widely used methods arc following kinds.

1. storage for decay. This is a simple and convenient and effective method which is suita-

ble method for radioisotopes with short half-life. For example, the storage time for several days can make the radioisotopes of I except 1311

and 12'1 decay away. But this method needs many large volume storage tanks, and the treatment effectiveness is not obvious for long half-life radioisotopes, using limitation is more larger.

2. vaporization. This is a kind of frequently used effective method. The

decontamination factor of radioactivity for this method can reach 1000 and over. The vaporization efficiency is mainly depend on the quantity of entraiqment of waste water carried by the secondary steam. Hence for easily bubbly waste water the decontamination efficiency is slightly low. Table 6.3-1 lists the decontamination factors for vaporing waste waters. Concentrated multiples of waste waters can reach few decades to few hundreds. The small concentrated liquid has concentrated large amount of radioactivity. Normal the concentrated liquid is treated by solidification to make it transport and storage easily. Most of nuclear power plant use the cement or chemical coagulant to solidify the concentrated liquid after vaporing, i.e. the cement or chemical coagulant is put into the drum loaded the concentrated liquid. and then it is enclosed until it harden naturally, and treated as solid waste.

Table 6.3- 1 Decontamination Factors of Elements for Vaporing Waste Waters

type of waste water clcrncnt I

I C s a n d N d Y Mo others I I ordinary waste water 1 o3 1 o4 lo4 lo4 lo4 1

Decontaminating, washing 1 02 1 02 10' 1 0 ' 10' 1

clothes and shower water I

1 Boronated wasre water 1 o2 lo3 lo3 lo3 lo3 1

3. ion exchange. The ion exchange is common method for purifying the waters. I t is

effective for removing radioactivity in low salt water. Strong acid and strong base type ion exchange resin have good effectiveness for removing radioactive I, Sr, Ni and Fe, but their removing power falls small short of for Sr, Na. M o and Y. The decontamination factors of ion exchange resins often used in different conditions are listed in Table 6.3-2.

Table 6.3-2 Decontamination Factors of' Mixed Bed (IT, OH) Ton Exchange Resins in Different Conditions

1 Waste water type positive ion negative ion CS. ~d 1 I - . - - - ---- - - -- - - - -

coolant 10 10 10 -- ---- - .-

condensate 103 ! 103 -- --

10 ----- ~

102 102 clean waste liquida : I I I --

I filthy waste liquidb 10 I 10 I 1

a. radioactive waste liquids which contain small chemical materials.

b. radioactive waste liquid which contain large chemical materials.

The reproduction effectiveness of the resin, which has been absorbed the radioactive materials, is more poor, long term irradiation can make its absorption capacity reduce, and processing the radioac" ~ive reproduction solution is a complex process. Therefore most of resins used to treat the radioactive waste liquid do not reproduce. After failure the resins are treated as the solid waste.

4. filtering. The power of removing radioactivity for filter is limited. On the one

hand, the filter can not remove the radioactive materials solved in water, on the other hand, the particulates of activated corrosion products, which do not solve in water. are very small and are hardly removed. Therefore, the filter of the waste water treatment system plays a auxiliary part in waste water treatment, for example, preventing breakdown resins from flowing away to ensure the normal operation of evaporator and resin beds.

The operational experiences indicate that the ion exchange resin bed has good filtering function. Its removing power is far superior to that of normal filters in filtering suspended solid micro particulates in water. In many conditions i t can have both ion exchange and filtering function. In normal condition it consists of following devices: one pre-bed filter which has fine mesh to ensure suspension for stay. One post-filter which is used to stay the break down resin carried in following liquid from demineralizer.

5. degasification. The function of degasification is removal the radioactive gases car-

ried in waste liquid. The degasification is made in steam. The separated

radioacuve gases are cooled and then fed into the gaseous treatment system. The waste liquid is cooled and then fed inro temporary storage rank.

Control ~ - ~ Discharge ... ~ of Waste ~ Water with Tritium Tritium can not separated by the methods mentioned above. Normally, the contained tritium water produced in light water nuclear power plants is not treated, and directly and in controlling discharged to environment by choosing advantageous discharging conditions. Most of the tritium produced in nuclear power plant is discharged to water system, therefore, it is necessary to choose the rivers and oceans which have larger diluting power and to discharge the contained tritium waters to them with strict controlling to make the tritium concentration of water systems and resi- dential areas is lower than the relevant standard.

6.3.2 WASTE GASES TREATMENT Generally speaking, the normal procedures for waste gases collected

are monitoring, treating with some choices and then discharging. Sources of Waste Gases The main sources of radioactive gases in

PWR nuclear power p l a y a r e the process gases discharged from the pressurizer, pressure reducing tank, volume control system, degasifier unit and temporary storage tank. When the fuel failure ratio is higher, the radioactive concentration in gas can reach few 3.7 x lo i0 Bq / 1. But the total volume of this gas is not large, about few cubic kilometers. The other sources of radioactive gases are gas of containment ventilation and auxiliary building ventilation, the gas discharged from radioactive liquid tank and working box, the gas discharged from main condenser spray system, the gas discharged from steam generator blow down tank and the gas of turbine building ventilation.

The radioactivity of former three kinds of gas is induced by the leakage of coolant. The later several kinds of secondary loop gas are with radioactivity only when evaporator tube failure occurs, and the primary loop coolant leaks to secondary loop. According to different cases the gas treatment devices are set respectively.

The containment gas treatment system must specially be paid great attention to. Besides it can normally purge the radioactive gases in containment, meantime it is a part of whole engineered safety features of reactor. When the loss-of-coolant accident occurs, it is together with the safetv spray system to prevent radioactivity dispersion outside. The

radioactivity discharged from containment appears as noble gases, radioactive aerosol, the steam of iodine and organic compounds of iodine.

Treatment . - -- of Process Waste Gases The gases introduced from the - - - -- - - -

devices mentioned above are fed into the buffer tank of gaseous waste treatment system, and then fed into decay tank under pressure of 1 MPa. The waste gases are stored for 60 days to 100 days to make the short half-life radionuclides decay away over 99.9%. After decay, the

133 radionuclides in waste gas is mainly : : ~ r and ,, Xe. Generally speaking, it is safe& to discharge waste gases decayed through the stack by diluting in controlling. After coming into atmosphere, the Z3xe decay away rapidly.

85 The ,,Kr, produced during nuclear power plant operation, need treat. The suitable method for treating : S r is absorption by using activated carbon under low temperature.

Treatment of Exhaust Air for Radioactive Building Normally two methods are used to treat rahoactive aerosol and iodine in exhaust air of building. The high efficiency filter is used to remove the aerosol, and the activated carbon is used to remove the iodine.

The structure of high efficiency filter used in radioactive building is similar to that of super-air-filter used in electronics industry. The glass fibre is used as the filtering medium. This medium can remove 99.97% and over of particulates whose diameter is larger than 0.3 pm. The activated carbon has good absorption function to iodine steam. In radioactive building, a part of iodine exists in form of organic compound. The absorption ability of activated carbon is more poor to organic iodine. Usually, the absorption characteristic of activated carbon is improved by using the method of macerating the activated carbon by using the iodine. The stable iodine macerated on the activated carbon is replaced with the radioactive organic iodine absorbed in air, therefore the absorption ability of activated carbon is enhanced.

In order to improve and maintain characteristic high efficiency filter and activated carbon filter, and to lengthen their using period, usually, it is necessary to set demister separator and prefilter to reduce the air humidity and to remove a part of impurity of large grain in air in front of high efficiency filter and activated carbon filter. Fig. 6.3-2 is typical schematic diagram of gaseous treatment system for nuclear power plant.

6.3.3 SOLlDWASTETREATMENT Kind. Quality and Radioactive Level of Solid Waste The main

-- - -- -- - - - -- - -

sources of solid-waste are failure resins and failure filter elements of coolant-purification system and waste water purification system, and solidification matter of evaporation residue. The radioactive level of these solid wastes is very high. During nuclear power plant operation, a small amount of solid waste, whose radioactive level is more high, is produced, for example, some activated component, probes of instrument. and parts in reactor. The other kind of solid wastes are contaminated matters, for example, contamination clothes, tools and arti- cles of protection use. The volumes of these contamination matters are more large, but their radioactive levels are not high. Table 6.3-3 lists the main kinds and quantities of solid wastes produced per year for one PWR nuclear power plant with 1000 MWe power.

Treatment -- of Solid Wastes For the loose solid wastes, normally, -- -

the method of' compression and the putting into the drum is used. There is a room in which the solid wastes are filled into the drum in nuclear power plant. The loose solid wastes are pressed into special drum by solid-waste press. In this way, the volume of solid wastes can be reduced, and it is for transportation and storage. The combustible solid wastes can be burned and become into carbon to reduce their volumes.

One nuclear power plant produce about 10 m3 failure resin. Its activity is about (1 1.1-14.8) x 10'' Bq / kg. Usually, the failure resin is stored for about one year in storage tank and water, its activity is reduced by decaying method, and then the failure resin is dewatered and put in metallic -or concrete drum. To ensure the safety of transportation, the failure resin is concreted by adding bonding agent and catalytic agent. After sealing of the drum, the drums are transported to solid waste vault by using isotope transport vehicle with shield. For the solid wastes with high radioactive level the treatment method is similar to that mentioned above.

Table 6.3-3 S olid Waste Quant,ity Produced fo'r One Typical PWR Nuclear

Power Plant with lOOOMWe Power

Kind

Quantity came Stored quantity come I

into solid waste from solid waste treatment system ; treatment system

(m3/a) , Volume Activity

(m3/a) (3.7 XlO'OBq/s)

j failure resin

i evaporation residue

! 'i filter failure ! / elements

I ; other I i I

i low : compressible ~ 70.6 I 11.1 i level 1 I

1 solid ) non I i

, / ( wasbe : compressible 1 28.1 1 28.1

6.4 RELEASE STANDARD 6.4.1 GENERAL DESCRIPTION

Before operation of nuclear power plant, the operating organiza- uon must submit the discharge limits of gaseous and liquid effluents to national nuclear safety department to gain the approbation. When sub- mitting these limits, the operating organization must follow the national relevant laws and replations, and stipulations.

The approved limits are stipulated by using the dose equivalent rate. But for the effluent limits, the approved limits are expressed through the activity of the radionuclides discharged in one year. The operating organization must work out the measures of controlling discharge to ensure coinciding these limits for effluent discharge and for "ALARA" principle.

Some exact measurement must be made before discharge to understand the activity of emuent to be discharged exactly as much as possible.

6.4.2 COMPOSITION OF EFFLUENT The radionuclides discharged to environment are fission products

and activation products in primary loop of nuclear power plant. 90 Fission Products ' 3 7 ~ s and Sr appear in form of solid, the for-

mer half-life is 30 years and the latter half-life is 28 years. The isotopes of Xe, Kr and I appear in form of gas, they mainly short half-life radionuclides, but the half-life of 8 5 ~ r is longer, its half-life is 10 years. The tritium gas is the product of tri-fission, it passes through the cladding materials by diffusion and then appears in water of primary loop. The half-life of tritium is 12 years, and it is difficult to treat it.

Activation Products After the corrosion products of stainless steel are activated, the following isotopes are generated: 6 0 ~ o (T! = 5a),

7

5 4 ~ n (T! = 280d) and 5 9 ~ e (TI =45d). The activation of Ar existed in 2 - 7

41 air produces the Ar (TI = 2hr). The activation of oxygen in water of Z

primary loop produces (T 1 = 7.1s). this materials is the main radia- Z

tion source under reactor operation. Because its half-life is very short. its effects disappears when reactor shutdowns. The activation of Li and B resolved in water produces the tritium also.

The radionuclides mentioned above will appear in effluents in nuclear power plant. The basis of working out release standards is the inherent characteristics of these radionuclides, for exampie, emitted ray lund, energy and half-life. These standards have regulated the maximum limits which can not be exceeded by the concentration of these radionuclides in air or water.

6.4.3 EFFLUENT MEASUREMENT Measurement on Storage Tank First, stirring the effluent to make

it homogenization, and then sampling the specimens and measuring. These measurements include: total P activity measurement, measuring activity of tritium by using the liquid scintillation detector, quantitative analyzing radionuclides in effluent by using the gamma spectrometer.

Measurements at Discharge Port The aim of these measurements is to monitor the effluent radioactive concentration. The monitoring instruments are with alarm device for abnormal condition.

During discharging the liquids, the P and y activity are measured. During discharging gases, the f l activity in gas filtered is measured using the difference ionization chamber installed in stack.

6.4.4 RELEASE STANDARD To reduce the exposure of public, the country all over the world

have regulated the strict release standard. According to different site conditions, climatic conditions and other reasons, the different design aim values of release when the nuclear power plants are designed. For liquid waste discharge and gaseous discharge, it is necessary to regulate that the concentrations of radioactive products in air or in water can not exceed the concentration limits.

Release Standard of Liquid Wastes France law regulate that the radioactivity concentration in water must be lower than lo-' Ci / m3 if the liquid wastes are discharged to river. This value is the mixed value for radionuclides. France health department requires that every nuclear power plant must consider the effects of deposition on earth in the range of standard mentioned above. For example, for CHOOE nuclear power plant, the limit of liquid waste discharge is lo-' Ci / m3.

Gaseous Waste Release Standard After waste discharge, the limit -

of radioactivity concentration in air is 3 x Ci / m3. For those inter- esting radionuclides, for example, 13'1, its limit of concentration in air is

6 x lo4 Ci / m3. From safety consideration, France health department requires that the radioactivity concentration of atmosphere surrounded the CHOOE nuclear power plant can not exceed I0 Ci / m3

Relevant - - Standard - - - about Solid Wastes In site. the storage of - - - - --

drums can not induce that the personnel is subjected to detnmental dose. For transportation, France relevant standard regulates that the maximum exposure rate at one meter from container is 10 mR / h.

6.4.5 RELEASE ACCESSMENT Taking generating units with 900 MWe power of PWR nuclear

plant as example estimates waste discharge quantity. This estimation is dependent on the operational conditions and transient conditions. There is 8000 m3 water which passes through boron recycle system. In addition, the radioactivity of primary. loop is dependent on fuel failure rate also.

Annual Discharge Quantity of Liquid The radioactivity dischareed to river is 10-20 Ci/ a, including the Cs, Sr in it. The discharge quantity of tritium is 1700 Ci / a.

The volumes of wastes are as follows: drain 300 m3 / a chemical waste liquid 600 m3 / a floor drains 600 m3 / a common waste water 1200 m3 / a Annual Discharge -- Quantity of Gases The main composition of

gas discharged is 1 3 3 ~ e . The total radioactivity in gas discharged to atmosphere is about 20000 Ci / a, 0.2 Ci is 13'1 in it. In addition, there is 500 Ci and over of tritium formed by steam.

Annual Discharge Quantity of Solid The volume of solid wastes loaded in metallic or concrete drum is about 50-100 m3 / a. After six months, the average activity is 10 Ci / drum.

I 71

deln

y ,d

ecay

fil

ter,

Fig

6.1

-1

Flo

w p

l-in

cipl

e d

iag

ram

of

was

tes

in n

ucl

ear

po

wer

pla

nt

1 te

mp

ora

ry Y

coll

ecti

on -

,R,P

c,

wat

er

spar

atln

g

&

sto

rag

e r,

bo

ron

fro

m

deg

asif

icat

ion

~

re-s

tora

ge

1 1 bo

ric

acid

bo

ron

I ~

euti

liza

tio

n

r- m

on

ito

rin

g

L

i st

ora

ge

I

wat

er

1 (e

vapo

ratio

n)- I

Fig

6.3

-1

Op

erat

ion

al p

rin

cip

le

dia

gra

m

of

bo

ron

rec

ycl

e sy

stem

qn~q merp rqanai ?nam?ear? p!nbn molJ

rna)sKs lor?oo;,

ua*bq ~ossardmoa 88% q6Bm amnloA pna TwKmaqJ morj rn-sds a1363ai no1oq morj

Date post:	11-Jul-2016
Category:	Documents
Upload:	seraj-ul-haq
View:	36 times
Download:	7 times

Radiation Protection NPP

Documents