C1_HTGR_design_safety_JCK

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
19-10-2015 IAEA Training Course on HTGR Technology, Serpong, Indonesia, 19-23 October 2015 2

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]

Context (1)

Context (2)

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)

Basic HTGR concepts

Pebble-bed fuel
(J.L. Kloosterman, RAPHAEL Eurocourse, March 2007)

“Prismatic” fuel
(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

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]

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]

HTGR designs – Some basic data
[H. Nickel, HTR/ECS, 2002]

Former HTGR designs …
[W. von Lensa, HTR/ECS, 2002]

USA – From prototype to commercial design
[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

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…)

Origin of design criteria
 INPRO (IAEA International Project on Innovative Nuclear Reactors
and Fuel Cycles) [R.2.1]
 Safety  Constraints / limitations
 Application & performance
 Design...

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

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...

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”)???

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

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

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).

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.

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]

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.

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...)

Origin of design criteria – overview
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
 ...

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

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...)

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”)

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)

 Dependent upon HM load, enrichment and burn-up [R.1.6]

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...)

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

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]

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]

Coated particle size (2)
 Harder spectrum and higher conversion for smaller particles
 For smaller particles curve tend to the one for the homogeneous case

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


2
PFC
M6
l

Axial Xe-oscillations (1)
Xe-effect

Axial Xe-oscillations (2)

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

Doppler effect vs. burn up

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)

Temperature coefficients of reactivity (1)
PBMR-400 in equilibrium state (U-based fuel)

Temperature coefficients of reactivity (2)
PBMR-400 in equilibrium state (Pu-based fuel)

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

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:

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]

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]

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

Failure of coated particles – Temperature (2)
[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

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]

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

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

Fluctuation of packing density (2)
Axial packing fraction profile: measurement and simulation [R.3.17]

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].

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

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)

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)

Physical reasons for flexibility w.r.t. fuel (cycle)
[R.1.6]

Fissile and fertile materials
 LEU cycle (enrichment 5 to 19%)
 MOX cycle
 Pu only
 Th (HEU; MEU; Pu) [R.1.6]

(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 ???

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

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

Comparison of radial loading patterns in “HTR-PM”
N.B. Earlier version with 2 m
core radius and 380 MW
thermal power

Radial power distribution “HTR-PM” (380 MW)

Radial power distribution “HTR-PM” (1515oC)

Pebble-bed cartridge core (ACACIA)

ACACIA reactor with annular core (and BP)

ACACIA – Initially homogeneous BP

ACACIA – Initially inhomogeneous BP

Two-batch axially shifted re-fueling in GTHTR 300

GTHTR 300 core lay-out

GTHTR 300 fuel layers and re-fueling scheme

Axial power distribution in GTHTR 300

Transient behaviour of GTHTR 300

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

Coordinates
Dr.ir. J.C. Kuijper (Jim)
Nuclear Reactor Physics Expert
E jimkuijper@gmail.com
M +31 6 4022 9728
http://www.nuclic.eu (per 1 January 2016)
Associated with Nuclear-21.Net
http://www.nuclear-21.net

Terima kasih

References (1)
[R.1.1] Generation IV International Forum, “Technology Roadmap Update for Generation IV Nuclear Energy Systems”, January 2014.
[R.1.2] Generation IV International Forum, “Annual Report 2013”.
[R.1.3] “The Very High Temperature Reactor: A Technical Summary”, Rev. 0, June 2004, MPR Associates Inc.
[R.1.4] C.W. Forsberg, P.F. Peterson, P.S. Pickard, “Molten-salt-Cooled Advanced High-Temperature Reactor for Production of Hydrogen and
Electricity”, Nuclear Technology, Vol. 144, No. 3, Pages 289 302, December 2003.
[R.1.5] K. Kugeler & R. Schulten, “Hochtemperatur-reaktortechnik”, Springer-Verlag 1989, ISBN 3-540-51535-6.
[R.1.6] RAPHAEL Eurocourse on V/HTR Technology, IKE, Stuttgart, Germany, 27-29 March 2007, see
ftp://ftp.cordis.europa.eu/pub/fp6-euratom/docs/raphael-eurocourse-seminar-folder-mar-2007_en.pdf, containing:
J.C. Kuijper, “HTGR core design criteria “;
X. Raepsaet, “Physics of the High Temperature Gas-Cooled Reactor Core “.
[R.1.6a] J.L. Kloosterman, “Design criteria for the HTR core”, Lecture, Delft University of Technology, Delft, The Netherlands, 20 November
2008.
[R.1.7] M.W. Rosenthal, “An Account of Oak Ridge national Laboratory’s Thirteen Nuclear Reactors”, ORNL/TM-2009/181, Oak Ridge
National Laboratory, Oak Ridge, TN, USA, March 2010.

References (2)
[R.1.8] S.E. Jensen, E. Nonbøl, “Description of the Magnox Type of Gas Cooled Reactor (MAGNOX)”, Risø National Laboratory, Roskilde,
Denmark, 1999, ISBN 87-7893-050-2.
[R.1.9] E. Nonbøl, ”Description of the Advanced Gas Cooled Type of Reactor (AGR)”, Risø National Laboratory, Roskilde, Denmark,
November 1996, ISBN 87-550-2264-2.
[R.1.10] Fütterer, M.A., et al., “Status of the very high temperature reactor system”, Progress in Nuclear Energy (2014),
http://dx.doi.org/10.1016/j.pnucene.2014.01.013
.
[R.2.1] “Guidance for the Application of an Assessment Methodology for Innovative Nuclear Energy Systems, INPRO Manual”, IAEA-
TECDOC-1575 Rev. 1, November 2008.
Volume 1 - Overview of the Methodology
Volume 2 - Economics
Volume 3 - Infrastructure
Volume 4 Waste Management
Volume 5 Proliferation Resistance
Volume 6 Physical protection
Volume 7 Environment
Volume 8 Safety of Nuclear Reactors
Volume 9 Safety of Nuclear Fuel Cycle Facilities

References (3)
[R.2.2] N. Prasad Kadambi, “Defence in depth in nuclear safety”, Nuclear Engineering International, 30 January 2013.
[R.2.3] IAEA, "Safety of Nuclear Power Plants: Design", Revision DS-414 to IAEA NS-R-1.
[R.2.4] USNRC, "White Paper on Risk-Informed and Performance-Based Regulation," Staff Requirements Memorandum Regarding SECY-98-
144, March 1, 1999.
[R.2.4a] USNRC, "Strategic Plan: Fiscal Years 2008-2013 (Updated)", NUREG-1614, Vol. 5, February 2012.
[R.2.5] Generation IV International Forum, “Proceedings GIF symposium, Paris, France, 9-10 September 2009”, ISBN 978-92-64-99115-6.
Available at: https://inis.iaea.org/search.
[R.2.6] J.C. Kuijper et al., “Plutonium and Minor Actinide Management in Thermal High-Temperature Gas-Cooled Reactors – Publishable
Final Activity Report”, Report PUMA-1006-D411g, Euratom 6th Framework Program contract no. FP6-036457, November 2010.
Available at: https://inis.iaea.org/search.
[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.
[R.2.7] F. Venneri et al., “The Deep Burn project: 2010 overview and progress“, in: “Actinide and Fission Product Partitioning and
Transmutation - Eleventh Information Exchange Meeting”, San Francisco, CA, USA, 1 – 4 November 2010, OECD/NEA Nuclear
Science and Nuclear Development 2012, ISBN 978-92-64-99174-3.
[R.2.8] “Uranium 2011, 2012. Resources, Production and Demand. A Joint Report by the OECD Nuclear Energy Agency and the International
Atomic Energy Agency”, OECD, Paris, France.
[R.2.9] K. Verfondern, H. Nabielek, J.M. Kendall, “Coated particle fuel for high temperature gas cooled reactors”, Nuclear Engineering and
Technology, Vol. 39, No. 5, October 2007.

References (4)
[R.2.10] S. Knol et al., “The ARCHER Project”, Proc. 7th International Topical Meeting on High Temperature Reactor Technology HTR 2014,
INET, Tsinghua University, Beijing, China.
[R.3.1] R.N. Morris, D.A. Petti, D.A. Powers, B.E. Boyack, “TRISO-Coated Particle Fuel Phenomenon Identification and Ranking Tables
(PIRTs) for Fission Product Transport Due to manufacturing, Operations and Accidents – Main Report”, Report NUREG-6844 Volume
1, US NRC, July 2004.
[R.3.1a] J. Keijzer, “Investigations of spatial effects in nuclear reactor kinetics “, PhD thesis, Delft University of Technology, Delft, The
Netherlands, 5 April 1996.
[R.3.2] A. Meier, W. Bernnat, K. Hossain, G. Lohnert, “Analyses of high temperature pebble bed reactors with plutonium fuel”, Proc.
International Conference on Mathematics, Computational Methods & Reactor Physics (M&C 2009), Saratoga Springs, New York,
May 3-7, 2009.
[R.3.3] D. Buckthorpe, J.-S. Genot, “RAPHAEL: Synthesis of achievements on materials and components and future direction”, Proc. 5th
International Topical Meeting on High Temperature Reactor Technology (HTR 2010), Nuclear Engineering and Design, Vol. 251, Pages
330 – 343, October 2012.
[R.3.4] B.K. McDowell, M.R. Mitchell, R. Pugh, J.R. Nickolaus, G.L. Schweringen, “High Temperature Gas Reactors: Assessment of
Applicable Codes and Standards”, Report PNNL-20869, Prepared for the US NRC, Pacific Northwest National Laboratory, October
2011.
[R.3.5] B. Boer, A.M. Ougouag, J.L. Kloosterman. G.K. Mills, “Stress analysis of coated particle fuel in graphite of high-temperature
reactors”, Nuclear Technology, Vol. 162, June 2008.

References (5)
[R.3.6] J. Jonnet, J.L. Kloosterman, B. Boer, “Performance of TRISO particles fuelled with Plutonium and Minor Actinides in a PBMR-400
core design”, Nuclear Engineering and Design, Vol. 240, Pages 1320 – 1331, 2010.
[R.3.7] M.P. LaBar, A.S. Shenoy, W.A. Simon, E.M. Campbell, Y. Hassan, “Nuclear Energy Materials and Reactors Volume II The Gas-
Turbine Modular Helium Reactor”, Encyclopaedia of Life Support Systems (EOLSS).
[R.3.8] J.J. Duderstadt & L.J. Hamilton, “Nuclear reactor analysis”, John Wiley & Sons, New York, ISBN 0 471 22363 8, 1976.
[R.3.9] “Decay heat power in light water reactors”, Standard ANSI/ANS-5.1-1994, American Nuclear Society, USA, 1993.
[R.3.10] “Calculation of the decay power in nuclear fuels of light water reactors - Part 1: Uranium oxide nuclear fuel for pressurized water
reactors (Berechnung der Zerfallsleistung der Kernbrennstoffe von Leichtwasserreaktoren - Teil 1: Uranoxid-Kernbrennstoff für
Druckwasserreaktoren)”, Standard DIN 25463-1:2014-02.
[R.3.11] “Calculation of the decay power in nuclear fuels of light water reactors - Part 2: Mixed-uranium-plutonium oxide (MOX) nuclear
fuel for pressurized water reactors (Berechnung der Zerfallsleistung der Kernbrennstoffe von Leichtwasserreaktoren - Teil 2: Uran-
Plutonium-Mischoxid (MOX)-Kernbrennstoff für Druckwasserreaktoren)”, Standard DIN 25463-2:2014-02.
[R.3.12] ”Decay heat power in nuclear fuels of high-temperature reactors with spherical fuel elements (Berechnung der Nachzerfallsleistung
der Kernbrennstoffe von Hochtemperaturreaktoren mit kugelförmigen Brennelementen)”, Standard DIN 25485-1:1990-05.
[R.3.13] K. Tasaka et al., “Recommendations on Decay Heat Power in Nuclear Reactors”, Journal of Nuclear Science and Technology, Vol. 28,
Issue 12, 1991.
[R.3.14] R.B. Bird, W.E. Stewart, E.N. Lightfoot, “Transport Phenomena”, John Wiley & Sons, 3rd printing, February 1963.
[R.3.15] “Heat Transport and Afterheat Removal for Gas Cooled Reactors Under Accident Conditions”, Technical Report IAEA-TECDOC-1163,
IAEA, Vienna, Austria, 2000.

References (6)
[R.3.16] G. Melese, R. Katz, “Thermal and Flow Design of Helium-cooled Reactors”, ISBN 0894480278, 1984.
[R.3.17] G.J. Auwerda, ”Core Physics of Pebble Bed High Temperature Nuclear Reactors”, PhD thesis, Delft University of Technology, Delft,
The Netherlands, ISBN 978-94-6295-047-4, 22 December 2014.
[R.3.18] G.J. Auwerda et al., “Effects of random pebble distribution on the multiplication factor in HTR pebble bed reactors”. Annals of
Nuclear Energy, Vol. 37, 1056, 2010.
[R.3.19] G.J. Auwerda et al., “Effect of non-uniform porosity distribution on thermal hydraulics in a pebble bed reactor”, Proc. NURETH-14,
Toronto, Ontario, Canada, 2011.
[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.
[R.4.3] B.M. Tyobeka, “ADVANCED MULTI-DIMENSIONAL DETERMINISTIC TRANSPORT COMPUTATIONAL CAPABILITY FOR SAFETY ANALYSIS
OF PEBBLE-BED REACTORS”, PhD thesis, The Pennsylvania State University, The Graduate School, Department of Mechanical and
Nuclear Engineering, August 2007.
[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.

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.

C1_HTGR_design_safety_JCK

Recommended

Recommended

More Related Content

Similar to C1_HTGR_design_safety_JCK

Similar to C1_HTGR_design_safety_JCK (20)

C1_HTGR_design_safety_JCK