International Conference Nuclear Energy in Central Europe 2001 Hoteli Bernardin, Portorož, Slovenia, September 10-13, 2001 www: http://www.drustvo-js.si/port2001/ e-mail: [email protected]tel.:+ 386 1 588 5247, + 386 1 588 5311 fax:+ 386 1 561 2335 Nuclear Society of Slovenia, PORT2001, Jamova 39, SI-1000 Ljubljana, Slovenia 806.1 ANALYSIS OF RADIATION CHARACTERISTICS FOR CASKS LOADED WITH SPENT RBMK-1500 NUCLEAR FUEL Arturas Smaizys, Povilas Poskas Lithuanian Energy Institute Nuclear Engineering Laboratory 3 Breslaujos str., 3035 Kaunas, Lithuania [email protected], [email protected]ABSTRACT The objective of this paper is to present the analysis of radiation characteristics for the ductile cast iron CASTOR RBMK-1500 and heavy concrete CONSTOR RBMK-1500 casks loaded with spent nuclear fuel from Ignalina NPP RBMK-1500 reactors. These casks are designed for an interim storage (up to 50 years) of spent nuclear fuel at Ignalina NPP. Computer calculations have been performed using SCALE4.3 computer codes system. The dose rate calculations have been performed on the sidelong, upper and lower surface of the casks and for certain distance at the beginning of spent nuclear fuel storage in the casks and after 50 years of interim dry storage. The results obtained results show that dose rate values on the surface of the cask are much less than the design criteria value 1000 µSv/h when the average burn-up of fuel assembly is 20 GWd/tU. It was revealed that CONSTOR RBMK- 1500 cask has better shielding characteristics than CASTOR RBMK-1500 cask. 1 INTRODUCTION Initially when Ignalina nuclear power plant (INPP) was built, it was not planned to have interim storage of spent nuclear fuel (SNF). It was supposed that after a cooling period of 3-5 years the SNF would be transported to Russia for reprocessing or disposal. However after independence reestablishment in 1991 the question of SNF storage became Lithuanian problem. Various options have been analyzed and it was decided to use dry storage technology for interim (up to 50 years) storage of SNF at INPP. GNB (Germany) storage casks have been chosen. Part of them are ductile cast iron CASTOR RBMK-1500 casks and the remaining ones are – metal-concrete CONSTOR RBMK-1500 casks. Safety analysis of SNF storage casks involves a wide spectrum of problems. In order to evaluate radiation characteristics of the casks computer modeling (calculations of the equivalent dose rates, activities of nuclides, etc.) was performed. The results obtained are presented in this paper and they show that equivalent dose rate values on the surface of the casks are less than the design criteria value of 1000 µSv/h. 2 CALCULATION METHOD AND ASSUMPTIONS CONSTOR cask (Fig. 1,a) has cylindrical shape with diameter more than 2 meters, height more than 4 meters, total thickness of the side-wall 0,43 meter. The side-wall of the cask consists of two carbon steel cylinders and the heavy concrete layer between them. The
Transcript
International ConferenceNuclear Energy in Central Europe 2001Hoteli Bernardin, Portorož, Slovenia, September 10-13, 2001www: http://www.drustvo-js.si/port2001/ e-mail:[email protected].:+ 386 1 588 5247, + 386 1 588 5311 fax:+ 386 1 561 2335Nuclear Society of Slovenia, PORT2001, Jamova 39, SI-1000 Ljubljana, Slovenia
806.1
ANALYSIS OF RADIATION CHARACTERISTICS FORCASKS LOADED WITH SPENT RBMK-1500 NUCLEAR FUEL
Arturas Smaizys, Povilas PoskasLithuanian Energy Institute
The objective of this paper is to present the analysis of radiation characteristics for theductile cast iron CASTOR RBMK-1500 and heavy concrete CONSTOR RBMK-1500 casksloaded with spent nuclear fuel from Ignalina NPP RBMK-1500 reactors. These casks aredesigned for an interim storage (up to 50 years) of spent nuclear fuel at Ignalina NPP.Computer calculations have been performed using SCALE4.3 computer codes system. Thedose rate calculations have been performed on the sidelong, upper and lower surface of thecasks and for certain distance at the beginning of spent nuclear fuel storage in the casks andafter 50 years of interim dry storage. The results obtained results show that dose rate valueson the surface of the cask are much less than the design criteria value 1000 µSv/h when theaverage burn-up of fuel assembly is 20 GWd/tU. It was revealed that CONSTOR RBMK-1500 cask has better shielding characteristics than CASTOR RBMK-1500 cask.
1 INTRODUCTION
Initially when Ignalina nuclear power plant (INPP) was built, it was not planned to haveinterim storage of spent nuclear fuel (SNF). It was supposed that after a cooling period of 3-5years the SNF would be transported to Russia for reprocessing or disposal. However afterindependence reestablishment in 1991 the question of SNF storage became Lithuanianproblem. Various options have been analyzed and it was decided to use dry storagetechnology for interim (up to 50 years) storage of SNF at INPP. GNB (Germany) storagecasks have been chosen. Part of them are ductile cast iron CASTOR RBMK-1500 casks andthe remaining ones are – metal-concrete CONSTOR RBMK-1500 casks. Safety analysis ofSNF storage casks involves a wide spectrum of problems. In order to evaluate radiationcharacteristics of the casks computer modeling (calculations of the equivalent dose rates,activities of nuclides, etc.) was performed. The results obtained are presented in this paperand they show that equivalent dose rate values on the surface of the casks are less than thedesign criteria value of 1000 µSv/h.
2 CALCULATION METHOD AND ASSUMPTIONS
CONSTOR cask (Fig. 1,a) has cylindrical shape with diameter more than 2 meters,height more than 4 meters, total thickness of the side-wall 0,43 meter. The side-wall of thecask consists of two carbon steel cylinders and the heavy concrete layer between them. The
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bottom consists of the same materials and arrangement as the side-wall. The lid of the cask ismade from carbon steel. The geometrical shape and dimensions of the internal cavity, thegeometrical parameters of the basket and design materials are the same for both casks. Themain differences of dimensions and materials are in the side-walls and bottoms of the casks.In the case of CASTOR cask (Fig. 1,b), they are made of ductile cast iron and in the case ofCONSTOR – of two carbon steel layers and concrete layer between them. The lids are madefrom the same material.
a) CONSTOR b) CASTOR
1
2
3
4
5
1
2
3
4
Figure 1: Radial and axial cross-sections of the CONSTOR (a) and CASTOR (b) casks a) 1– lid; 2 – carbon steel cylinders; 3 – heavy concrete; 4 – basket with fuel half-assemblies; 5 –
cavity filled with helium; b) 1 – lid; 2 – cask body from ductile cast iron; 3 – basket withfuel half-assemblies; 4 – cavity filled with helium.
RBMK-1500 fuel assembly is made of two parts that are joined axially. Before storagein the casks, fuel assemblies are cut through the middle. The bottom part is turned upsidedown before storage while position of upper part is unchanged, i.e. the cut ends are alwaysfacing down in the cask. Active length of each half fuel assembly is 3,41 m. Internal basketcontains 102 stainless steel tubes each with half fuel assembly inside.
Computer modeling of radiation characteristics deals with two problems. Firstly thecharacteristics (concentrations of fission products, actinides, neutron and gamma sourceemissions, etc.) of irradiated fuel assembly must be calculated. Secondly, when properties ofSNF are obtained, equivalent dose rate calculations on the surface and at some distance fromthe casks must be done. Sequences SAS2H [1] and SAS4 [2] from SCALE4.3 computer codewere used for solution of these problems.
SAS2H computes neutron and gamma source spectrum and evaluates equivalent doserates from SNF storage/transport casks using a 1-D transport shielding analysis. SAS2Hexecutes codes BONAMI, NITAWL-II, XSDRNPM, COUPLE and ORIGEN-S for cross-section processing and fuel burn-up; radiation source calculations; the radial storage/transportcask shielding analysis applying the calculated SNF composition of nuclides and sources;determination of dose rates by XSDOSE from the angular flux leakage.
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SAS4 calculates radiation doses exterior to a storage/transport cask using a three-dimensional Monte Carlo method. The sequence executes BONAMI, NITAWL-II,XSDRNPM and MORSE-SGC for cross-section processing; for the radiation transport andradiation dose calculation.
The main input data for SAS2H are material composition and geometrical parameters offuel assembly, fuel channel in the reactor, storage casks; material concentrations andtemperatures; reactor power; irradiation and cooling periods of fuel assembly. SAS4 inputdata are nuclide composition of irradiated and cooled fuel assembly; material composition andgeometry of storage cask; temperatures; radiation source characteristics; locations of radiationdose detectors.
As it was mentioned above, calculations of radiation characteristics deal with twoproblems: 1) modeling of fuel assembly irradiation; 2) equivalent dose rate calculations.
The main assumptions for the modeling of fuel assembly irradiation were following:a) One fuel rod of a total of 18, that is in the RBMK-1500 fuel assembly, in reactor
channel was described as an element of 5 zones (Fig. 2) with infinite height;b) Radius of three exterior zones were calculated in such a way to have 1/18 part of the
appropriate components: water, technological channel, graphite;c) Fuel enrichment 2.0% 235U; average burn-up 20 GWd/tU, irradiation time 730 days
(2 years); cooling time 1826 days (5 years);d) For dose rate calculations axial burn-up distribution of fuel assembly was not taken
into account.
Figure 2: Five zones model used in calculations.
Assumptions for equivalent dose rate calculations were as follows:a) 3-D description of cask geometry in SAS4 code and 2-D description in SAS2H;b) Source emission was modeled as homogenous cylindrical body which contains 102
spent nuclear fuel half assemblies and basket internals;c) Locations of point detectors were at a middle of sidelong surface, at a center of the
lid and the bottom. Location distances - 0, 1 and 2 meters;d) Calculations were performed for two cases:
o at the beginning of SNF storage in the casks;o after 50 years of interim storage.
Using the appropriate computer codes and assumptions mentioned above, nuclidecomposition, concentrations, activities, neutron and gamma sources of irradiated nuclear fuelassembly and dose rate values on the surface and at some distance of the cask were calculated.Additionally radiation source dependence on axial burn-up distribution of fuel assembly hasbeen investigated and possible impact on dose rate has been evaluated.
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3 RESULTS
Results of fuel assembly irradiation modeling are presented in Tables 1 and 2. Theseresults were obtained when axial burn-up distribution of fuel assembly was not taken intoaccount. There are a lot of fission products in SNF. Nuclides presented in Table 1 have morethan 1% of total activity of fission products after 5 years cooling time. Set of the mostimportant actinides and their activities is presented in Table 2.
Table 1: Activities of fission products.
NuclideActivity of fuelassembly, Bq
Activity of source inthe cask, Bq
Kr-85 1,73 ⋅ 1013 8,83 ⋅ 1014
Sr-90 1,79 ⋅ 1014 9,13 ⋅ 1015
Y-90 1,79 ⋅ 1014 9,13 ⋅ 1015
Ru-106 4,91 ⋅ 1013 2,51 ⋅ 1015
Rh-106 4,91 ⋅1013 2,51 ⋅ 1015
Cs-134 4,63 ⋅ 1013 2,36 ⋅ 1015
Cs-137 2,43 ⋅ 1014 1,24 ⋅ 1016
Ba-137m 2,30 ⋅ 1014 1,17 ⋅ 1016
Ce-144 4,28 ⋅ 1013 2,18 ⋅ 1015
Pm-147 1,56 ⋅ 1014 7,96 ⋅ 1015
Table 2: Activities of actinides.
NuclideActivity of fuelassembly, Bq
Activity of source inthe cask, Bq
U-234 3,83 ⋅ 1009 1,95 ⋅ 1011
U-235 4,32 ⋅ 1007 2,20 ⋅ 1009
U-236 6,81 ⋅ 1008 3,47 ⋅ 1010
U-238 1,37 ⋅ 1009 6,98 ⋅ 1010
Np-237 3,98 ⋅ 1008 2,03 ⋅ 1010
Pu-238 2,60 ⋅ 1012 1,33 ⋅ 1014
Pu-239 6,56 ⋅ 1011 3,35 ⋅ 1013
Pu-240 1,49 ⋅ 1012 7,62 ⋅ 1013
Pu-241 2,21 ⋅ 1014 1,13 ⋅ 1016
Pu-242 5,02 ⋅ 1009 2,56 ⋅ 1011
Am-241 2,18 ⋅ 1012 1,11 ⋅ 1014
Am-242 9,47 ⋅ 1009 4,83 ⋅ 1011
Cm-242 3,52 ⋅ 1010 1,80 ⋅ 1012
Cm-244 1,31 ⋅ 1012 6,68 ⋅ 1013
Equivalent dose rate calculation results are presented in Figures 3-5. These figures showtotal equivalent dose rate values in various directions and distances of the casks at thebeginning and after 50 years of storage. As can be seen, dose rate values for CONSTOR caskin all cases are less than for CASTOR. The smallest difference of dose rate is at the lid points,because lid systems are made from the same material for both casks. Only total thickness ofthe lid system for CONSTOR cask is a little bigger. Material composition and thickness of theside walls and bottoms are different, so total equivalent dose rate differs from 2 to 7 times.
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0
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CONSTOR CASTOR
Equi
vale
nt d
ose
rate
, µ µµµSv
/h
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CONSTOR CASTOR
Equi
vale
nt d
ose
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, µ µµµSv
/h
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a) beginning b) after 50 yearsFigure 3: Total equivalent dose rate values on the cask lid.
0
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Equi
vale
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, µ µµµSv
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0102030405060708090
CONSTOR CASTOR
Equi
vale
nt d
ose
rate
, µ µµµSv
/h
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a) beginning b) after 50 yearsFigure 4: Total equivalent dose rate values on the cask side wall.
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0102030405060708090
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CONSTOR CASTOR
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vale
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ose
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, µ µµµSv
/h
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a) beginning b) after 50 yearsFigure 5: Total equivalent dose rate values on the cask bottom.
Total equivalent dose rate is formatted by neutrons and gamma radiation. Percentageof each component at the beginning and after 50 years of storage is presented in Figure 6 andFigure 7 respectively. At the beginning of storage gamma radiation is the dominatingcomponent at the side walls and bottoms for both casks. Influence of neutrons on total
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equivalent dose rate for CONSTOR and CASTOR casks is the same at the lid points, butcompletely different at side walls and bottoms (Figure 6).
25,3
98,6 96,5
74,7
1,4 3,50
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Lid Side Wall Bottom
%
GammaNeutron
31,1
69,162,2
68,9
30,937,8
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%
GammaNeutron
CONSTOR CASTOR
Figure 6: Percentage of gamma and neutron dose rates at the beginning of storage.
According to dose rate calculation results (Figures 3-5), total equivalent dose ratedecreases about 10 times after 50 years of interim storage. But after such period influence ofneutrons and gamma radiation on total dose rate for CONSTOR and CASTOR casks isdifferent (Figure 7). Equivalent dose rate caused by gamma radiation is dominating at the sidewall and bottom of CONSTOR cask, but dose rate caused by neutrons for CASTOR cask isdominating in all directions.
5,8
93,2
76,9
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32,524,5
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67,575,5
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%
GammaNeutron
CONSTOR CASTOR
Figure 7: Percentage of gamma and neutron dose rates after 50 years of interim storage.
As it was mentioned above, axial burn-up distribution of fuel assembly in equivalentdose rate calculations was not taken into account. Additionally calculations of sourceemission, when burn-up distribution is taken into account, were done. Axial distribution ofburn-up and calculated distributions of neutron and gamma sources are presented in Figure 8.Numerical data of neutron and gamma emissions from the source are presented in Table 3.Gamma and neutron emissions increase by 2,6% and 103% respectively, when burn-updistribution is taken into account. Such increase of source emission will raise equivalent doserates. Dose rates caused by gamma radiation will change a little, but dose rate caused by
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neutrons will increase significantly. On the other hand, dose rates caused by neutronsdominate only on the lid of the casks. Even if we assume that neutron emission is 2 timesbigger, permissible value of 1000 µSv/h will not be exceeded.
0,00
0,50
1,00
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2,00
2,50
3,00
3,50
0 100 200 300 400 500 600
Length of fuel assembly, cm
Nor
mal
ized
fact
or
Gamma source
Neutron source
Burn-up
Figure 8: Normalized axial distributions of burn-up, neutron and gamma sources.
Table 3: Neutron and gamma emission from homogeneous source in the cask (I – no axialdistribution of burn-up; II – burn-up distribution taken into account).
Analysis of radiation characteristics shows following:o CONSTOR RBMK-1500 cask has better shielding characteristics than CASTOR
RBMK-1500 cask;o During 50 years of interim storage period the total dose rate is decreasing from 5
to 20 times, depending on the location of the point detector and type of the cask;o After 50 years of interim storage percentage of neutron dose rate is increasing
and percentage of gamma dose rate is decreasing.
REFERENCES
[1] Hermann O. W., Parks C. V. SAS2H: A Coupled One-Dimensional Depletion AndShielding Analysis Module, Rev.5, Oak Ridge National Laboratory, March 1997.
[2] Tang J. S. SAS4: A Monte Carlo Cask Shielding Analysis Module Using An AutomatedBiasing Procedure, Rev.5, Oak Ridge National Laboratory, March 1997.
International ConferenceNuclear Energy in Central Europe 2001Hoteli Bernardin, Portorož, Slovenia, September 10-13, 2001www: http://www.drustvo-js.si/port2001/ e-mail:[email protected].:+ 386 1 588 5247, + 386 1 588 5311 fax:+ 386 1 561 2335Nuclear Society of Slovenia, PORT2001, Jamova 39, SI-1000 Ljubljana, Slovenia
808.1
RISK ASSESSMENT OF HAZARDOUSWASTE TRANSPORT – PERSPECTIVES OF GIS
APPLICATIONRoxana Elena Lazar, Maria Dumitrescu, Ioan Stefanescu
National R&D Institute of Cryogenics and Isotopic Technologies – ICSI Rm. Valcea,PO Box 10, 1000 Valcea, Romania
Due to the increasing public awareness of the potential risks associated with wastetransport, the environmental impact assessment of this activity has become an issue of majorimportance. This paper presents a project proposal, which can establish a national action planfor waste transport evaluation. Such a programme is sustained by the necessity to obtain anadequate method for the rapid and efficient estimation of individual and social risks due to thetransport of hazardous substances in Romania. The main objective is to develop regionalstrategies for risk assessment in comprising: establishing the areas that must be investigatedand their particular characteristics; identifying the transport activities in the areas; determininghazards; establishing the analysis criteria and prioritizing the study areas; evaluatingcontinuous emissions; studying major accidents; studying population health; classifying therisks; establishing regional strategies; implementing political and regulatory measures. Theproject expectation is to provide a decision tool for risk managers and authorities in order tocontrol or limit transportation and the storage of hazardous substances.
1 INTRODUCTION
In risk assessment the analysis of hazardous material transport and storage is an elementoften neglected. The potential accidental pollution of the environment is high for hazardousmaterial transport and its analysis is much more difficult than for a fixed plant taking intoaccount the various initiation conditions of an accident and their consequences.
One possibility to minimize the risk in hazardous material transport is to develop aregional strategy for monitoring industrial activities, hazardous materials and their transportroutes, and to collect all the necessary data for a complete transport database.
In a national research programme, ICSI Rm. Valcea will initiate a project with the aimof defining a possible strategy to reduce risks in transport and storage of hazardous materials,to establish the most adequate analysis methods and to illustrate the advantages of usingGeographic Information System (GIS) in such an analysis.
This paper attempt to highlight a framework for a regional strategy and to explore thebenefits of using GIS in risk analysis for hazardous materials transport and storage withregard to the Romanian territory.
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2 RISK ANALYSIS AND MANAGEMENT
Any human activities involve risk and risk estimation, assessment and management hasbecome a problem involving a multidisciplinary scientific approach. The goal of riskassessment is a humanitarian one because it is very important to prevent human and economicinjuries and damage.
The general steps of risk assessment, highlighted in Figure 1, refer to system/activitydefinition, hazard identification, quantifying the frequency and consequences of accidentsequences, risk analysis, estimation of risk levels and improvement proposals (processimprovement, intervention after accidents, etc.). [1,2]
Figure 1: General steps in risk analysis
Risk management represents a way in which the evaluation results are used to controlthe risk, to identify the places where the safety of the activity was satisfactory or not, and toelaborate solutions for risk reduction. The risk management approach is illustrated in Figure2.
Figure 2: General approach in risk management
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3 METHODOLOGICAL FRAMEWORK
This paper has the aim of presenting a general method and its associated procedures tochoose the priorities in the frame of different risk sources for hazardous materials transportand storage to facilitate a detailed risk assessment on a priority basis. This will representassessment of major accident risks caused by transportation of hazardous substances by road,railway and pipelines.
The risks considered are those for public health due to explosions, fires and emissionsof dangerous substances under established parameters.
The transport risk assessment method includes the main tasks necessary to classify therisks and to achieve a graduated priorities scheme as follows: a list of hazardous substances, aclassification of transport and storage activities, a selection of activities that must be studied,establishing the effects category, estimating consequences to the population and the frequencyof accidents, estimating the social risk, risk hierarchisation depending on priorities.
4 PROCEDURAL STEPS FOR RISK ASSESSMENT
For a complete identification of the characteristics of transport and storage activitiesover the whole of a given territory the study must be performed in separate zones ofapproximately 100 km2, with important industrial activity and a substantial number of people.
The main steps of the regional risk assessment process in a integrated manner arehighlighted in Figure 3 and Figure 4, representing a possible strategy for hazardous materialstransport and regional storage risk assessment. [3]
For a better understanding, a short presentation of the procedural steps for hazardousmaterials transport and storage are illustrated:
P.1. Establishing the boundary and the main characteristics of the studied areaP.1.1. Boundary area definitionP.1.2. Area descriptionP.1.3. The area map
P.2. Defining and centralizing the transport and storage of hazardous materials in the areaOnce the studied area and its main characteristics are established a database regarding
hazardous materials transport activities must be made. It is important to collect informationregarding names, localization, type, production, storage and/or transport conditions, physicalcharacteristics and quantity of substances and an activities criteria selection depending on thedistance to populated areas, traffic density, etc.
Those activities that do not represent a direct danger for the population, because ofdistance or low transport rate, were excluded from the hazardous substance classificationscheme. The studied areas (roads, railways, and pipes) were divided into 1 km parcels. Thoseportions that don’t match with the distance criteria face on populated area are ignored.
The nearest places to populated area were selected for every parcel and for railtransport special attention was paid to marshalling yards.
P.2.1. Classification of activities by type;P.2.2. Data collection regarding hazardous activities in the area (place, environmental
impact, etc.);P.2.3. Exclusion of those activities with low impact;P.2.4. Hazardous substances inventory compilation and activities enumeration;
estimation of the maximum quantity of hazardous substances released in an accident.
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Figure 3: Steps in the integrated risk assessment process for transport activities
P.3. Estimation of accident frequencyThe frequency estimation for different accident events can be made using:
- statistical data for hazardous substances storage and transport;- expert experience;- failure modes and effects analysis method;- failure tree method; others.
P.4. Estimation of consequencesThe estimation of consequences for every selected activity is performed after a
sufficient set of data has been gathered for the hazardous activities in the area. Theconsequences of an accident means the number of fatalities at population level who live andwork in the neighbourhood of a storage plant or near a road, railway or pipeline for hazardoussubstances transport.
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P.4.1. Activity selectionP.4.2. Separate analysis of substances
A group of substances that act together are considered as one substance. If aninflammable substance is also toxic, both effects are considered.
P.4.3. Activity classificationUsing the activity classification tables the negative activity type will be established, as
well as the physical and chemical characteristics of the substances (boiling point, molecularweight, heat of evaporation, liquid density, compressibility coefficient, dynamic viscosity,heat of combustion, upper and lower explosion limits, MAC value, probit coefficients). Theeffects categories take into account the maximum effect distance and affected perimeter.
P.4.4. Determining the maximum effect distance and the affected areaP.4.5. Estimating the population distribution in the affected areaP.4.6. Establishing the meteorological conditions (wind direction, wind speed class,
atmospheric stability class)P.4.7. Establishing a dispersion modelP.4.8. Evaluating the consequences
Figure 4: Possible strategy for regional risk assessment
P.5. Estimation of social riskFor every analyzed activity (storage plant or part of a road, railway or pipeline) one or
more estimates of the fatality number and the frequency of major accidents will be achieved.
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The risk for the population can be estimated using these values, representing themeasure of activity’ frequency and negative consequences.
P.5.1. Establishing the risk estimation mode: qualitative/quantitativeP.5.2. Establishing the risk level for every activity using a scale of consequence
classes and one of the probability classesP.5.3. If an activity represents an increased risk for the population due to utilization of
many different hazardous substances that can cause accidents independently one another, therisk will be the amount of risk value determined by substances from the same consequenceclass.
P.5.4. Representation of the results of activity classification depending on risk in agraphical mode
P.6. Risk hierarchisation depending on prioritiesP.6.1. A criteria for social risk acceptance is definedThe acceptability criteria can be established by many methods:
- establishing a threshold only for the probability;- establishing a threshold only for the consequences;- taking into account both the probability and consequences.
An event with lower frequency and greatest consequences is more important for studythan one with a lower frequency and lower consequences.
P.6.2. Identifying the activities having an unacceptable risk level.
5 USING GIS IN RISK ANALYSIS FOR HAZARDOUS MATERIALSTRANSPORT
The necessity for GIS in analysis of hazardous materials transport is sustained by therapid and efficient individual and social risk achievement. One important role of GIS is theability to bring together data from a variety of sources, for example, health, socioeconomicdata and environmental data, within a common framework. Once a map has been created inthe GIS it can be integrated with other information, such as census data. The programmecould provide a preliminary hazards evaluation allowing the decision factors to asses wherethe transport activity is acceptable or not, or a detailed analysis is required. Providingalternative solutions such as changing the transport routes or modes and identifying the leastdangerous one is also permitted.
The main ideas to be developed in this programme are:- data collection reduction and the activity organization by using average values for some
parameters;- separate estimation of consequences for every accident states depending on
meteorological conditions and their introduction in a database;- comparing the estimated individual and group risks with a standard one, established at
national or international level, to a rapidly information achieving regarding the risk level.The hardest task in the program’s development is data collection that for every route
must contain information regarding vehicle types, transported substances, emission scenarios,meteorological conditions, accident rates and population density.
To define the accident scenarios an emissions classification can be drawn updepending on the released quantity of substance: without emission, low emission (<10 L),medium emission (10-200 L) and major emission (>200 L). Taking into account that in mostcases a low emission (such as a substance released by a small leak) leads to a negligiblehazard which can be treated as a no emission case. In the other cases we can presume thatonce the accident occurs the product is released by a circular break on the bottom of a tank.
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The meteorological data are provided for a 10 year period, by month, and includeaverage values for minimum and maximum temperatures, humidity, wind direction and thepersistence of certain wind speed classes.
Figure 5: Data distribution in a geographical information system
The accident rate used in hazardous substances transport risk analysis must beexpressed as accidents per kilometer and per vehicle (generally, the existing data areexpressed as number of accidents per kilometer and per year).
Figure 6: South-East Romania population density
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The data regarding the population density are to obtained from census data. Inprinciple, every county must be divided in two zones: one that includes the towns with anintensely populated area and the other outside the towns with lowest population density. Anexample of population density classes is presented in Table 1. Figure 6 illustrates thepopulation density in South-East Romania.
Table 1. Population density classThe population density
classThe population density
(/km)Isolated area 0 – 49Rural area 50 – 149
Suburban area 150 – 499Urban area 500 - 3000
The greatest attention must be paid to the selection of substances included in thedatabase. All results regarding the individual and social risks should be compared withstandards. Unfortunately Romania doesn’t have its own set of limit values used to define riskacceptability.
6 CONCLUSIONS
From the perspective of Romanian sustainable development there exists an increasedinterest regarding the impact of human activities on the environment taking into account thatany human action or activity represents a degree of risk and risk assessment and managementis a multidisciplinary problem. The application at national level of such a programme can leadto increased awareness, risk reduction of hazardous substance storage and transport, andimprove the legal framework. Such an analysis provides many advantages. The routeidentification for hazardous substance transport is just one of them. Developing the accidentscenarios, collecting data and analyzing them permits us to explore at maximum potential ageographical information system to assess the risk for transport. The GIS benefits are betterand easier work for the analyst, permitting rapid route modifications and evaluation of anaccident’s possible impacts. If the geographical information systems would be applied inRomania on a large scale by staff in management, a better control of transport activity couldbe provided, in order to minimize risks.
REFERENCES
[1] R. E. Lazar, M. Dumitrescu, I. Stefanescu, Importanta evaluarii riscului in stocarea sitransportul substantelor periculoase. Identificarea surselor de risc, Report ICSI Rm.Valcea, 2000.
[2] Manual for classification and prioritisation of risks due to major accidents in process andrelated industries, IAEA, Vienna, 1993
[3] R. E. Lazar, M. Dumitrescu, Tehnica evaluarii potentialelor hazarduri. Scenarii deaccident, Report ICSI Rm. Valcea,2000
