Design and safety approach of HTGRs
Jim C. Kuijper
IAEA Training Course on High-Temperature
Gas-Cooled Reactor Technology
Serpong, Indonesia, 19-23 October 2015
Introduction & outline (1)
Approach ::: Philosophy ::: Background
Context - Basic HT(G)R concepts
Some history
Origin of design criteria o Safety constraints, limits o Application & performance
Specific characteristics of HTGR systems (focus: reactor…) o Reactor physics (neutronics) o Material properties o Thermal hydraulics & heat transfer
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Introduction & outline (2)
HTGR core design characteristics o Core shape and dimensions o Fuel selection and (re-) load strategy
Concluding remarks…
Presentation is based on the chapter “(V)HTR in detail – Design & safety approach” of the JRC-IET book on Generation IV systems (to be published December 2015?)
State-of-the-art ~December 2014
With special thanks to the originators of the illustrations and other info, in particular to Prof.dr.ir. Jan-Leen Kloosterman (Delft University of Technology, NL) and Mr. Xavier Raepsaet (CEA Saclay, France)
For further (detailed/background) information, see the [References]
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Context (1)
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Context (2)
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Context (3)
(Very) High Temperature Gas-Cooled Reactor (Gen III+, IV)
(TRISO) coated particle fuel (He) gas-cooled Graphite moderator/reflector Epi-/thermal spectrum Pebble-bed and prismatic Influence of design choices on behaviour Some details of (V)HT(G)R (reactor) physics & thermal hydraulics HGTR fuel cycle – flexibility (other presentation) Not about: Calculation/analysis methods (other presentations)
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Basic HTGR concepts
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Pebble-bed fuel
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(J.L. Kloosterman, RAPHAEL Eurocourse, March 2007)
“Prismatic” fuel
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(W. Bernnat, RAPHAEL Eurocourse, March 2007)
Some history
Early gas-cooled reactors
HT(G)R plants constructed and operated
Former HGTR designs
“Current”HTGR designs
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Early gas-cooled reactors Name or acronym Oak Ridge Graphite
Reactor Windscale piles P2 Magnox AGR Tokai-1
Magnox
Location Oak Ridge, TN USA
Sellafield UK
Saclay France
UK UK Tokai Japan
Operation years 1943 - 1963 1950 1951 1956 - present 1962 - present
1966 - 1998
Fuel U metal cyl. Al cladding
Nat. U Nat. U metal Al cladding
Nat. U Nat. U Nat. U
Moderator Graphite Graphite Heavy water Graphite Graphite Graphite
Coolant Air air N2 (initially) CO2 (later)
CO2 CO2 CO2
Power [MWth/MWe]
1 – 4/- - 2 / - - / 166 - / 166
Reference [R.1.7] [R.1.8] [R.1.9]
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HTGR plants constructed & operated Name/ Acronym
DRAGON Peach Bottom
Fort St. Vrain
AVR THTR HTTR HTR-10
Location UK USA USA Germany Germany Japan China
Operation Years
1964 – 1975
1966 - 1974
1976 – 1989
1967 - 1988
1985 - 1991 1999 - present
2000 – present
Fuel element Cylinder Cylinder Cylinder in hex. block
Sphere Sphere Cylinder in hex. block
Sphere
Fuel coating TRISO BISO TRISO BISO BISO TRISO TRISO
Fuel kernel Carbide Carbide Carbide Oxide Oxide Oxide Oxide
Enrichment [%] 90 93 17
Power [MWth/MWe]
20 / - 115 / 40 842 / 330 46 / 15 750 / 300 30 / - 10 / -
Tin / Tout [oC] 350 / 750 377 / 750 400 / 775 270 / 950 270 / 750 395 / 950 300 / 700
He pressure [bar]
20 25 49 11 40 40 30
Power density [W/m3]
14 8.3 6.3 2.6 6.0 2.5 2.0
Reference [R.3.4] [R.3.4] [R.3.4] [R.3.4] [R.3.4] [R.3.4] [R.3.4]
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HTGR designs – Some basic data
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[H. Nickel, HTR/ECS, 2002]
Former HTGR designs …
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[W. von Lensa, HTR/ECS, 2002]
USA – From prototype to commercial design
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[W. von Lensa, HTR/ECS, 2002]
“Current” designs Name/ Acronym
MHTGR GT-MHR HTR-Modul PBMR HTR-PM
Location USA USA/Russia Germany South Africa China
Fuel element Compact in hex. block
Compact in hex. block
Sphere Sphere Sphere
Fuel coating TRISO TRISO TRISO TRISO TRISO
Fuel kernel UCO/Th-oxide UCO/MOX Oxide Oxide Oxide
Enrichment [%] 20 19.8 7.8 4.2 – 9.6 8.5
Power [MWth/MWe]
4x350/508 600 200 400/165 2 x 250/211
Tin / Tout [oC] 259 / 687 491 / 850 250 / 700 500 / 900 250 / 750
He pressure [bar]
63.7 70.7 60 90 70
Power density [MW/m3]
5.9 6.6 3.0 4.8 3.2
Remarks Preliminary design completed / Licensing process not finalised
Licensed Project terminated in 2009
Under construction
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Observations
Great variety in parameters (dimensions, power, He pressure, etc.)
Trend towards higher coolant outlet temperature (HTGR VHTR)
High degree of “passive safety” possible (if sufficiently low power density…)
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Origin of design criteria
INPRO (IAEA International Project on Innovative Nuclear Reactors and Fuel Cycles) [R.2.1]
Safety Constraints / limitations
Application & performance
Design...
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INPRO (1)
Comprehensive methodology for the assessment of safety and performance of an innovative reactor (so the HTGR/VHTR)
INPRO Volumes 1-9 [R.2.1]: 1. Overview of the methodology
2. Economics
3. Infrastructure
4. Waste management
5. Proliferation resistance
6. Physical protection
7. Environment
8. Safety of nuclear reactors
9. Safety of nuclear fuel cycle facilities
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INPRO (2)
Evaluation of innovative nuclear system designs: o Basic Principles
o User Requirements (safety, performance, ...)
o Criteria (= Indicator + Acceptance Limit)
Only a (very) small subset hereof in this presentation...
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Safety (1)
Many ways to define nuclear safety (IAEA, USNRC, ... [R.2.2])
INPRO volume 8 (reactors) and volume 9 (fuel cycle facilities)
Practical definition:
A nuclear reactor (system) is classified as “safe” if there is no health hazard to the public or personnel under all conceivable normal (operation, anticipated deviations) and off-normal (DBA, BDBA) sitiations.
Include “almost inconceivable” situations (“stress test”)???
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Safety (2)
This implies:
No need for off-site evacuation or taking shelter near the site boundary
No need for moving mechanical components to ensure this
Exposure to personnel significantly lower than current internationally accepted values
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Safety (3)
Safety design philosophy for a nuclear reactor: criteria at 3 distinct levels:
Top level criteria: o National regulatory body International/IAEA standards
o ALARA [R.2.3]
Basic safety functions: o Control of reactivity/criticality (ensure subcriticality)
o Removal of (decay) heat from (the fuel in) the core (temperature limit)
o Confinement of radioactive material
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Safety (4)
Defence-in-depth principle to prevent, mitigate and control any off-normal event. The IAEA formulation distinguishes 5 levels of defence: o Prevent deviations from normal operation
o Detect and control deviations
o Prevent core damage by incorporating safety features/systems
o Mitigate consequences (on-site and off-site)
o Mitigate radiological consequence (on-site and off-site)
Prevention and mitigation systems.
To be applied to all safety-related activities (organisational/behavioural/design) and all (reactor) system states (full power/low power/various shutdown states).
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Safety (5)
Alternative formulation of Defense-in-depth (USNRC [R.2.4][R.2.4a]):
“…, the philosophy ensures that safety will not be wholly dependent on any single element of the design, construction, maintenance, or operation of a nuclear facility”.
Both formulations share the notion of multiple, independent layers of defence or multiple barriers of protection against health hazard to public and personnel.
HTGR/VHTR can incorporate these barriers in a passive manner.
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Application & performance
Envisaged application & required performance (technical/economic), mostly connected to the possibility of high coolant outlet temperature:
Electricity and (process) heat
High fuel temperature basic design choices: o Refractory materials in the core: graphite o Use of (inert) gas as coolant: He, CO2, H2
o Coated particle ful (SiC, PyC)
Other requirements, e.g. high degree of sustainability:
In GIF usually attribited to fast spectrum systems (“closing the fuel cycle”) [R.2.5]
HTGR/VHTR may also provide a high degree of sustainability [R.2.6][R.2.6a][R.2.7]
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Design (1)
Several definitions of “design” possible... (see e.g. http://www.oxforddictionaries.com)
For a reactor system like a HTGR, “design” refers to the choice of materials, dimensions, and arrangement of structures and components and the choice of relevant (design) parameters (e.g. nominal power level, temperature, pressure, but also dimensions and dimensional changes) during normal operation and off-normal states.
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Design (2)
Application to structures and components at different levels:
Reactor system and structures’ dimensions, components and materials more or less fixed
Fuel elements (materials and dimensions) some details may change during the lifetime of the reactor
Core layout and loading scheme variation from cycle to cycle (or continuous...)
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Origin of design criteria – overview
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Application – Performance, e.g.:
Electricity production
CHP
District heating
H2-production (process heat in general) [R.3.3]
Pu/MA utilization/incineration [R.2.6][R.2.6.a][R.2.7]
Fuel utilisation – Waste production (sustainability), e.g.:
High burn-up
High conversion ratio
Reduction of Pu and/or MA [R.2.6][R.2.6a]
Minimise production of waste
Direct disposal
Operation, e.g.:
Cycle length
Maintenance
Direct/indirect cycle
Plant life
Multiple modules
Safety – inherently safe design (?)
Normal operation – Incidents/Accidents
Excess reactivity
Control rod worth – Shutdown margin
Xe-effect / Xe-oscillations
(Negative) temperature reactivity coefficient
Max. (fuel) temperature
Max. power per fuel element (particle, pebble, compact)
Fast fluence in fuel and core internals
Max. burn-up
Decay heat removal (passive)
???
Constraints (limits)
Reactor physics/neutronics
Thermal hydraulics
Material science
???
Design criterion ≈ Range of permissible parameter values
Specific characteristics of HTGR
Characteristics and associated constraints/limits:
Reactor physics/neutronics: mutual interaction of the materials in the reactor with the neutron field
Material properties
Thermal hydraulics and heat transfer
...
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Reactor physics/neutronics (1)
Neutronic properties of HTGR materials – which materials?
Interaction of materials with neutron field: o several reactions (fission, capture, scatter, (n,2n), (n,3n),…) possible per
nuclide o σ(E), η(E),... o material compositions/distribution changing with time
Radioactive decay
Moderator-to-fuel ratio (M/F ≈ C/U) => k
Material (nuclide) distribution over space → o Neutron distribution over energy (spectrum) o Neutron distribution over space => power distribution => temperature
distribution, decay heat distribution
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Reactor physics/neutronics (2)
Enrichment ε, burn-up, conversion ratio, control rod worth,…
Resonance self-shielding – Doppler effect – CP dimensions
Neutron leakage o H/D ratio (cylindrical core)
o Neutron streaming
o Minimum neutron leakage for cylindrical core: H/D ≈ 0.924
keff ≥ 1 (uncontrolled...)
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Materials in HTGR
Fissile: U235, U233, Pu239, Pu241 (usually oxide or (oxy-) carbide)
Fertile: U238, Th232 (usually oxide)
Moderator: graphite (C12)
Other materials in the fuel (e.g. inert matrix material)
Structural material: graphite, SiC
Control elements: steel, B4C
Fission products: Xe135, Sm149, Mo95, Cs133, Cs135, Tc99, Ag110, Nd145, Xe131, Rh103, Pm147, Eu153,…
Heavy isotopes: Pa233, U234, U236, Np239, Pu240, Pu242, Am…, Cm…, …
Coolant: He (“neutronically inert”)
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Neutron flux spectral distribution (1) Higher thermal and epithermal flux than in LWR
Dependent upon HM load, enrichment and burn-up
(W. Scherer, HTR/ECS, 2002)
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Neutron flux spectral distribution (2)
Dependent upon HM load, enrichment and burn-up [R.1.6]
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Neutron flux spectral distribution (3)
Dependent upon HM load, enrichment and burn-up [R.1.6]
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Controllability
Position of control structures (elements, absorber spheres,[R.1.5]) must be such that: o Sufficient subcriticality can be ensured at all times, even if the most reactive
element can not be inserted (“shutdown margin”).
o Sufficient margin for altering the reactivity (“control rod worth”) to counter changes in reactivity during operation
No real problem for prismatic block type HTGR
Limit on radial dimensions of core cavity for cylindrical pebble bed HTGR (on the other hand: THTR...)
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He coolant
Helium (mostly He4) is transparent to neutrons: ”No” change in reactivity from change in He temparature or density
DLOFC is almost purely thermal-hydraulic issue
No (positive) void coefficient No limits on the use of Pu fuel
Functions of coolant (He) and moderator (graphite, C12) not combined: No correlation between cooling geometry and M/F-ratio High degree of flexibility w.r.t. fuel management while retaining excellent safety characteristics
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M/F – Moderator-to-fuel ratio
M/F ≈ C/U (or C/HM or “light atoms”/HM)
Range: 500 to 3000
Under moderated?
[R.1.6]
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Coated particle size
Fuel in kernel of TRISO (or BISO) particle
Resonance self-shielding (hence resonance escape probability) depends on kernel and CP size
Double heterogeneity
[R.1.6]
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Coated particle size (2)
Harder spectrum and higher conversion for smaller particles
For smaller particles curve tend to the one for the homogeneous case
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Migration length – Prompt fission chain
“Neutronic” dimensions of the core: Compare actual dimensions with characteristic distances of neutronics (migration length, prompt fission chain length) o Migration length M o Prompt fission chain length:
(J. Keijzer, PhD thesis, Delft, 1996 [R.3.1a])
Prismatic type: M ~ 22 cm; lPFC = 668 cm
Pebble-bed type: M ~ 32 cm; lPFC = 972 cm
If characteristic dimension of the core > prompt fission chain length, axial Xe-oscillations may occur
Core height H limited to approx. 10 m
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2
PFC
M6l
Axial Xe-oscillations (1)
Xe-effect
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Axial Xe-oscillations (2)
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Temperature influence on reactivity
Mainly 3 temperature coefficients of reactivity:
W.r.t. temperature of the fuel (CP) (Doppler effect)
W.r.t. temperature of the moderator (in the core)
W.r.t. temperature of the reflector
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Doppler effect vs. burn up
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Graphite temperature effect
Moderator and reflector
Shift of Maxwellian peak to higher energy if temperature increases
Lower effective cross section (reaction rate) for “1/v” cross sections (thermal neutron energies)
Possibly higher effective cross section (reaction rate) for “non 1/v” or resonance cross sections (intermediate neutron energies)
Net effect may be positive or negative, depending on (local) circumstances (difference between reflector and moderator)
Example: equilibrium state (time-independent nuclide distribution)
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Temperature coefficients of reactivity (1)
PBMR-400 in equilibrium state (U-based fuel)
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Temperature coefficients of reactivity (2)
PBMR-400 in equilibrium state (Pu-based fuel)
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Material properties
Radiation damage o Fuel (coatings) o In-core structures (e.g. reflector) o Fast fluence (E > 0.1 MeV)
(Chemical) compatibility of coolant and other materials – not a big problem for He
Thermal stress & Chemical interactions with(in) CPs o Limit retention capability of FP and actinides o Max. acceptable release (R/B ratios) limit on burn-up
Limit on coated particle temperature (1600 oC/1250 oC)
Limits on material temperatures in general → limit on coolant output temperature
Wigner effect in graphite → minimum coolant entrance temperature of 200 oC
Many (other) material properties of materials used in HTGR components, with influence on performance and safety. Should be properly addressed in the design process
See e.g. [R.3.3] and [R.3.4] for applicable codes and standards
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Fuel at very high burn up
U-based fuel: 150 MWd/kg
Pu-based fuel: > 700 MWd/kg achieved in Dragon and Peach Bottom unit 1:
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Dimensional changes in graphite
30 years of irradiation at fast flux of 3x1013 cm-2s-1 (above 0.1 MeV) gives fast fluence of 2.5x1022 cm-2
Dimensional (and other) changes in graphite by radiation damage [R.1.5]
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Failure of coated particles – Fast fluence
Fast fluence < 5x1021 cm-2 (TRISO) (less restrictive nowadays?)
For CP failure mechanisms, see [R.3.1][R.3.5][[R.3.6][R.2.9]
[R.1.5]
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Failure of coated particles – Temperature (1)
> 1250 oC fission product attach on SiC layer
> 1600 oC decomposition effects and porosity in SiC layer
> 2000 oC thermal decomposition of SiC dominant mechanism
See [R.3.1][R.2.9]
Maximum design base event fuel temperature: 1600 oC
Maximum (peak) fuel temperature during normal operation: 1250 oC
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Failure of coated particles – Temperature (2)
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[R.3.7]
Thermal hydraulics
Transfer heat from fuel to a useful purpose (Tout?)
Heat conduction, convection and radiation [R.3.15]
Coolant properties (thermodynamic, fluid dynamic): He (other possibilities: CO2/H2)[R.1.4]
Cooling geometry o Pebble bed: fixed coolant volume fraction (CVF) o Prismatic block with coolant channels: CVF somewhat more flexible in the design phase
Coolant flow o Mass flow o Flow direction (upward/downward through the core) o Core pressure drop (← core dimensions, friction,…) o ∆p ~ H3
Passive (decay) heat removal capabilities (see [R.3.9][R.3.10][R.3.11][R.3.12][R.3.13] for background info on decay heat) o Core geometry and dimensions, H/D-ratio, friction, … o Free convection
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Pressure drop over the core
Pressure loss Δp over a pebble-bed core as function of core height (He-pressure is 40 bar) [R.1.5]
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Coolant volume fraction in core
CVF in pebble bed is more or less fixed (~0.39)
Larger range of possible CVFs in prismatic type (coolant channels in blocks), also < 0.39 → higher fuel density possible → higher power density possible
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Fluctuation of packing density (1)
Fluctuation of packing density has no significant influence on flux distribution or keff [R.3.17][R.3.18][R.3.19]
Stochastic nature of pebble bed has considerable influence on power- and temperature distribution [R.3.17][R.3.18][R.3.19]
Should be taken into account in the design extra margings for critical design parameters
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Fluctuation of packing density (2)
Axial packing fraction profile: measurement and simulation [R.3.17]
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HTGR core design characteristics
Core shape/dimensions and control structure positions – More or less fixed when design has been fixed
Fuel design and core (re-) loading, possibly including burnable poison – More or less flexible, even for a reactor already in operation.
Detailed design is a complex activity entailing many (other, even important) aspects well beyond this presentation, e.g. issues connected to the production, activation, transport, deposition and resuspension of (graphite) dust, especially in the case of pebble-bed HTGRs [R.4.3].
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Core shape and dimensions
Pebble-bed or prismatic – Coolant Volume Fraction
Cylindrical or annular core
Dimensions o Limits on core height H:
Xe-effect (oscillations) Coolant pressure drop over core ∆p ~ pumping power ~ H3
pumping power < ~5 % of electrical output ∆p < ~ 0.8 bar
o Limits on core diameter D: Control rod worth → annular core (control elements in central column) Passive heat removal → distance to core surface not too large
H/D-ratio o Minimum neutron leakage: H/D ≈ 0.9
Fixed or dynamic inner column
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Control rods – pebble bed
Control rods (usually) in reflector
Worth dependent upon thermal flux, hence reflector graphite temperature
Control rods in pebble bed not impossible (THTR), but not necessary
Reactivity requirements (feasible if cylindrical core radius not too large) o Control and hot shut down: ~4%
o Cold shut down: ~10%
(W. Scherer, HTR/ECS, 2002)
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Fuel selection and (re-) load strategy
HTGRs are very flexible with regard to fuel and fuel cycle o Uncoupling between and parameters characterising cooling geometry and
neutronics optimisation
o Solid moderator (no void effect)
Many types of fuel possible in principle – not all fuel designs have been qualified
High burn-up feasible – demonstrated (AVR, Peach Bottom, FSV, THTR)
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Physical reasons for flexibility w.r.t. fuel (cycle)
[R.1.6]
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Fissile and fertile materials
LEU cycle (enrichment 5 to 19%)
MOX cycle
Pu only
Th (HEU; MEU; Pu) [R.1.6]
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(Re-)load schemes
Pebble bed o MEDUL - continuous re-load o OTTO - Once Through Then Out N.B. Low excess reactivity for MEDUL/OTTO without BP o Peu-à-peu o cartridge o (spatial distribution of) burnable poison o ???
Prismatic o specific re-load scheme (axial and radial shift) o spatial distribution of fuel loading/enrichment o (spatial distribution of) burnable poison o ???
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Examples of reload schemes
OTTO vs. Medul (re-) loading scheme for pebble-bed HTGR
Comparison of radial loading patterns in “HTR-PM” (earlier 380 MW design with 2 m core radius)
Pebble-bed cartridge core with burnable poison (ACACIA)
Two-batch axially shifted re-fueling in GTHTR 300
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OTTO vs. MEDUL (re-) loading scheme
Comparison of axial power density distribution of pebble bed MEDUL (“Mehrfachdurchlauf”) and OTTO (“Einwegbeschickung”) [R.1.5]
Top Bottom
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Comparison of radial loading patterns in “HTR-PM”
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N.B. Earlier version with 2 m core radius and 380 MW thermal power
Radial power distribution “HTR-PM” (380 MW)
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Radial power distribution “HTR-PM” (1515oC)
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Pebble-bed cartridge core (ACACIA)
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ACACIA reactor with annular core (and BP)
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ACACIA – Initially homogeneous BP
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ACACIA – Initially inhomogeneous BP
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Two-batch axially shifted re-fueling in GTHTR 300
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GTHTR 300 core lay-out
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GTHTR 300 fuel layers and re-fueling scheme
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Axial power distribution in GTHTR 300
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Transient behaviour of GTHTR 300
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Concluding remarks
An overview was given on main considerations regarding design– and safety philosophy of present-day HTGR designs
Focus on main features and issues w.r.t. the reactor. Many (even important) issues have NOT been treated in this presentation
Once more it has been shown that HTGR core design is extremely flexible, although there ARE limits
Many different applications, fuels, fuel cycles, etc. possible within constraints of safe operation
Burnable poison can be used to limit excess reactivity while retaining core life
Developments are ongoing, e.g.: o Radial cooling [R.5.1] o “Wallpaper” fuel [R.5.2]
Development towards higher core outlet temperatures (> 1000 oC) is possible [R.2.9] Gen IV VHTR
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Coordinates
Dr.ir. J.C. Kuijper (Jim)
Nuclear Reactor Physics Expert
M +31 6 4022 9728
http://www.nuclic.eu (per 1 January 2016)
Associated with Nuclear-21.Net
http://www.nuclear-21.net
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Terima kasih
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Volume 2 - Economics
Volume 3 - Infrastructure
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[R.2.6a] J.C. Kuijper et al., “Pu and MA management in Thermal HTRs, Quo Vadis – Insights from the Euratom PUMA project”, Paper presented at the IAEA Technical Meeting “Deep Burn HTR”, IAEA, Vienna, Austria, 5-8 August 2013. Available at: http://inis.iaea.org/search. Corresponding paper to be published.
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[R.4.1] P. W. Humrickhouse, “HTGR Dust Safety Issues and Needs for Research and Development”, Report INL/EXT-11-21097, Idaho National Laboratory, Idaho Falls, Idaho 83415, USA, June 2011.
[R.4.2] M. Ragheb, “Nuclear, plasma and radiation science – Inventing the Future”, Web text, Lecture notes Nuclear Power Engineering NPRE 402, University of Illinois at Urbana-Champaign, USA, Spring 2015.
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[R.4.4] “PBMR COUPLED NEUTRONICS/THERMAL-HYDRAULICS TRANSIENT BENCHMARK - THE PBMR-400 CORE DESIGN - VOLUME 1: THE BENCHMARK DEFINITION”, Report NEA/NSC/DOC(2013)10, OECD Nuclear Energy Agency, Nuclear Science Committee, Paris, France, 17. July 2013.
[R.4.5] F. Li, X. Jing, “Comparison of loading pattern in HTR-PM”, Proc. 2nd International Topical Meeting of HTR Technology HTR 2004, Beijing, China, September 22 – 24, 2004.
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References (7)
[R.4.6] D.F. da Cruz, J.B.M. de Haas & A.I. van Heek, “ACACIA: A Small Scale Power Plant With Pebble Bed Cartridge Reactor”, Proc. International Congress on Advances in Nuclear Power Plants (ICAPP ’03), Cordoba, Spain, May 4 – 7, 2003, ISBN 0-89448-675-6.
[R.4.7] X. Yan, K. Kunitomi, T. Nakata, S. Shiozawa, “Design and development of GTHTR300”, Proc. 1st International Topical Meeting of HTR Technology HTR 2002, Petten, The Netherlands, April 22 – 24, 2002.
[R.5.1] B. Boer, J.L. Kloosterman, D. Lathouwers, T.H.J.J. van der Hagen, H. van Dam, “Optimization of a radially cooled pebble bed reactor”, Proc. 4th International Topical Meeting on High Temperature Reactor Technology HTR2008, Washington D.C., USA, September 28 - October 1, 2008.
[R.5.2] A. Marmier, M. Fütterer, K.Tucek, J.B.M. de Haas, J.C. Kuijper, and J.L. Kloosterman, “Revisiting the Concept of HTR Wallpaper Fuel”, Proc. 4th International Topical Meeting on High Temperature Reactor Technology HTR2008, Washington D.C., USA, September 28 - October 1, 2008.
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