Licensing Issues and the PIRT
Frederik Reitsma
Oct 22-26, 2012
IAEA Course on High temperature Gas Cooled Reactor Technology
Oct 22-26, 2012 IAEA Course on High temperature Gas Cooled Reactor Technology 2
Content / Overview
– A few ideas to stimulate discussions:
– Safety assessment criteria
– Safety analysis
– Treatment of uncertainties
– NRC Advanced Reactor Policy
– Evaluation models
– PIRT
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Safety assessment • Combination of both deterministic and probabilistic methods
• South African National Nuclear Regulator has set the following limits:
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Safety analysis • Combination of both best estimate and conservative deterministic models and
analysis – best estimate analysis, the best estimate material properties and plant parameters are used
as inputs to the analysis. Other sources of uncertainty may be addressed more conservatively
– Conservative analysis results are achieved using conservative inputs in conservative models. Sensitivity analyses are often used to ensure parameters are set to ensure pessimistic results with respect to the acceptance criterion.
• Best estimate analyses are used: – to demonstrate As Low As Reasonably Achievable (ALARA) for Anticipated Operational
Occurrences (AOOs) and Design Basis Accident (DBAs).
– to determine the expected consequences of more hypothetical accidents, i.e. Beyond Design Basis Accidents (BDBAs)
– to provide the Probabilistic Risk Assessment (PRA) with expected or realistic consequences instead of conservatively biased ones.
• Conservative analyses are used: – To demonstrate compliance with regulatory limits for AOOs and DBAs
Oct 22-26, 2012 IAEA Course on High temperature Gas Cooled Reactor Technology 5
Treatment of uncertainties • Combination of both best estimate and conservative deterministic models and analysis
– best estimate analysis, the best estimate material properties and plant parameters are used as inputs to the analysis. Other sources of uncertainty may be addressed more conservatively
– Conservative analysis results are achieved using conservative inputs in conservative models. Sensitivity analyses are often used to ensure parameters are set to ensure pessimistic results with respect to the acceptance criterion.
• Best estimate analyses are used: – to demonstrate As Low As Reasonably Achievable (ALARA) for Anticipated Operational Occurrences
(AOOs) and Design Basis Accident (DBAs).
– to determine the expected consequences of more hypothetical accidents, i.e. Beyond Design Basis Accidents (BDBAs)
– to provide the Probabilistic Risk Assessment (PRA) with expected or realistic consequences instead of conservatively biased ones.
• Conservative analyses are used: – To demonstrate compliance with regulatory limits for AOOs and DBAs
• Methodology based on the Code Scaling, Applicability, and Uncertainty (CSAU) process has been adopted to quantify uncertainties in best-estimate calculations – Can by used to show the margin of conservative models and analysis
– The GRS-SUSA code to be used as point of departure.
• A new IAEA Coordinated Research Project on HTGR Reactor Physics, Thermal-hydraulics and Depletion Uncertainty Analysis is being performed.
Oct 22-26, 2012 IAEA Course on High temperature Gas Cooled Reactor Technology 6
NRC Advanced Reactor Policy Statement (1/2)
“Among the attributes which could assist in establishing the acceptability or licensability of a proposed advanced reactor design, and which therefore should be considered in advanced designs, are:
• Highly reliable and less complex shutdown and decay heat removal systems. – The use of inherent or passive means to accomplish this objective is
encouraged (negative temperature coefficient, natural circulation).
• Longer time constants and sufficient instrumentation to allow for more diagnosis and management prior to reaching safety system challenge and/or exposure of vital equipment to adverse conditions.
• Simplified safety systems which, where possible, reduce required – operator actions, – equipment subjected to severe environmental conditions, – and components needed for maintaining safe shutdown conditions. – Such simplified systems should facilitate operator comprehension,
reliable system function, and more straight-forward engineering analysis for analysis.
Oct 22-26, 2012 IAEA Course on High temperature Gas Cooled Reactor Technology 7
NRC Advanced Reactor Policy Statement (2/2)
• Designs which minimize the potential for severe accidents and their consequences by providing sufficient inherent safety, reliability, redundancy, diversity, and independence in safety systems.
• Designs that provide reliable equipment in the balance of plant, (or safety-system independence from balance of plant) to reduce the number of challenges to safety systems.
• Designs that provide easily maintainable equipment and components. • Designs that reduce radiation exposure to plant personnel. • Designs that incorporate defense-in-depth philosophy by maintaining
multiple barriers against radiation release, and by reducing the potential for consequences of severe accidents.
• Design features that can be proven by citation of existing technology or which can be satisfactorily established by commitment to a suitable technology development program.”
FR Vol 73 No. 199, pg. 60612-60616, Oct. 14, 2008
Oct 22-26, 2012 IAEA Course on High temperature Gas Cooled Reactor Technology 8
Evaluation Models
Oct 22-26, 2012 IAEA Course on High temperature Gas Cooled Reactor Technology 9
PBMR Evaluation Models
• Pre-Break – Focuses on the expected pre-break conditions, just
before a break in the helium pressure boundary occurs.
– The time phase for this Evaluation Model ends when such a break occurs.
– It does not include any of the phenomena that might occur during a pressure boundary break or later.
– Software used: VSOP, MCNP, FLOWNEX, Fluent, NobleG, FIPREX/GETTER, RADAX.
Oct 22-26, 2012 IAEA Course on High temperature Gas Cooled Reactor Technology 10
PBMR Evaluation Models
• Reactivity transients
– Calculates the transient reactor response for reactivity transient scenarios.
– Number of calculation models: 2
– Software used: VSOP, TINTE.
• Thermal transients
– Calculates the transient reactor temperatures for forced cooling and loss of forced cooling scenarios.
– Number of calculation models: 2.
– Software used: TINTE, FLOWNEX.
Oct 22-26, 2012 IAEA Course on High temperature Gas Cooled Reactor Technology 11
PBMR Evaluation Models • Normal operation release
– Calculates the release of activity during normal operation as a result of Helium Pressure Boundary leakage.
– Number of calculation models: 22 – Software used: VSOP, MCNP, FLOWNEX, ASTEC, Fluent,
NobleG, FIPREX/GETTER, RADAX. • Maintenance Dose
– Calculates the dose received by maintenance workers during maintenance periods due to the dust in the MPS and the activation of components that took place during normal operation.
– Number of calculation models: 15 – Software used: MCNP, RADAX, SCALE, FISPACT,
MicroShield.
Oct 22-26, 2012 IAEA Course on High temperature Gas Cooled Reactor Technology 12
PBMR Evaluation Models
• Overall Worker and Public Dose – Calculates the doses received by the public and worker for
Category A, B and Beyond B accidents that involve a breach in the pressure boundary.
– Provides the source term for each of these accidents as well as the resulting doses.
– Provides source terms that are used as an input into the Probabilistic Risk Assessment (PRA) for Category C events.
– Number of calculation models: 28
– Software used: VSOP, TINTE, MCNP, ASTEC, FLOWNEX, Fluent, PC COSYMA, NobleG, FIPREX/GETTER, RADAX.
– Main components: Pre-break, Initial release, Delayed release, Air ingress, Confinement, Atmospheric dispersion
Oct 22-26, 2012 IAEA Course on High temperature Gas Cooled Reactor Technology 13
PIRT
Phenomena Identification Ranking Tables
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PIRT background • Phenomena Identification and Ranking Tables are increasingly used in the nuclear
industry. • PIRT were initially used to identify the thermal-hydraulic processes that were most
important for a safety analysis computer code to simulate with acceptable accuracy, so that a limited set of sensitivity analysis could be performed to help quantify the uncertainty in the safety analysis results.
• PIRT are now recognized as a valuable tool to help prioritise efforts associated with safety analysis, development and assessment of codes and models, and specification of scaling or other requirements for tests and experiments.
• There is limited industry knowledge and experience with HTGR accident analysis, relative to that for LWRs.
• Since operating history is not available to provide the same valuable data, an accepted and auditable method of making early decisions related to analysis is needed.
• The PIRT process can be used as: – a tool and a guide to help prioritise software and model V&V efforts, – to agree on the necessary degree of conservatism to include in analysis assumptions and initial
conditions, – and to focus the available resources on the phenomena believed to be most important to the
safety analysis process.
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PIRT - Process • Process:
– Step 1. Define the PIRT objectives and plant design – Step 2. Define the accident or transient scenario – Step 3. Define figures of merit – Step 4. PIRT team review available data – Step 5. Partition scenario into convenient time
phases – Step 6. Identify involved and affected SSC – Step 7. Identify phenomena by time phase and SSC – Step 8. Rank importance of components and
phenomena with confidence levels – Step 9. Finalize and document PIRT for subject
scenarios and plant designs
– Repeat periodically
Oct 22-26, 2012 IAEA Course on High temperature Gas Cooled Reactor Technology 16
PIRT Status Decision Chart Rank
Confidence
in Rank
Confidence
in Value
(High / Low)(Sure /
Unsure)
(Sure /
Unsure)
8 High Unsure Unsure
Phenomenon is perceived as
significant but is not well
known.
High priority requirement for
analysis and validation.
7 High Sure UnsurePhenomenon is significant
and confidence in value is low.
High priority requirement for
validation.
6 High Unsure Sure
Phenomenon is significant
and the confidence in rank is
low.
High priority requirement for
analysis.
5 High Sure SurePhenomenon is significant
and well known.
Should be well represented in
the model. Should be readily
validated.
4 Low Unsure UnsurePhenomenon is not significant
but not well known.
Requires analysis and
validation to determine rank
and value.
3 Low Sure Unsure
Phenomenon is not significant
and the confidence in value is
low.
Low priority requirement for
validation.
2 Low Unsure Sure
Phenomenon is not significant
and the confidence in rank is
low.
Low priority requirement for
analysis.
1 Low Sure SurePhenomenon is well known
and is not significant.
May be modelled without
validation.
Status Symptom Action required
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PIRT - examples
• Source Term (Pre-Break)
• Reactivity transients
• Thermal transients
• Overall Worker and Public Dose PIRTs – Initial release
– Delayed release
– Air Ingress
– Confinement
– Atmospheric Dispersion
• Non-MPS leaks and spills
Oct 22-26, 2012 IAEA Course on High temperature Gas Cooled Reactor Technology 18
PIRT example Transport phenomena (from pre-break PIRT)
• Radionuclide plateout: Adsorption/ desorption processes Penetration/evaporation Diffusion into material Chemical characteristics of radionuclide Laminar or turbulent flow
• Dust deposition and lift-off: Agglomeration of dust particles Brownian diffusion Electrostatic forces Inertial separation Laminar or turbulent flow Saffman lift force Sedimentation Thermal gradient (thermophoresis) Plant operational transients Plant vibration Monolayer or multilayer resuspension
Oct 22-26, 2012 IAEA Course on High temperature Gas Cooled Reactor Technology
Reactivity Transient PIRT Example for the Reactor Unit
Ref
Number
System subsystem component Process Phenomena Rank
(3 - high,
2 -
medium,
1 - low)
Confidenc
e (Sure/
Unsure)
state of
knowledg
e
Rationale Notes
RT-RU-001 RU Pebble Bed Fuel Kernel Fission Fission Power
production in the fuel
High Sure High The fission process provides the
primary source of power production
in a reactor.
This is due to fission,
RT-RU-002 RU Pebble Bed Fuel Kernel Fission Temperature
influence on fission
(doppler effect)
High Sure High Kernel does get hot. Flux
dependence on group structure is
important. Temperature
dependence of resonance cross
sections important.
Consider Multi-group vs 2-group
neutron flux representation. The current
2 group misses out on resonance
treatment. The physics is understood,
but it comes down to modelling issues
and complexity.
RT-RU-003 RU Pebble Bed Fuel Pebble Fission Moderation -
temperature
dependence
High Sure High Moderator feedback effect is well
known and temperature interaction
with fission is well modelled
The moderator feedback effect is
known and temp interaction with fission
is currently well modelled.
RT-RU-004 RU Pebble Bed Coolant Fission Nitrogen inventory
reduction
High Sure Med Nitrogen is an absorber. The
reduction in the nitrogen inventory
could contribute to reactivity
addition.
This is particularly applicable to the
start-up sequence, the the reduction in
nitrogen could contribute to reduced
absortion. The effect of SAS removal
with nitrogen in the core needs to be
checked to see if the core approaches
criticality. The modelling of the scenario
has never been attempted but physics
capability is available in TINTE.
RT-RU-005 RU Pebble Bed Fuel Kernel Fission Non-local Power
Production
Med Sure High The negative coefficient of
reactivity dependence on
temperature is assisted by non-
local heating which makes this
phenomenon important.
The non-local power is as a result of
absorption of gamma radiation, fast
neutrons. Currently when modelling,
the fission process is adequately
captured, but gamma modelling is
approximated.
RT-RU-006 RU Pebble Bed Fuel Kernel Fission Fuel Burn-up Low Sure High Low for perturbations in burn up
from the anticipated core state.
Refers to the average value of Fuel
Burn-up (there will be differences when
considering the start-up or equilibrium
core)
RT-RU-007 RU Pebble Bed Fuel Kernel Fission Spatial distribution of
burn-up
Low Sure High Low for perturbations in burn up
from the anticipated core state.
Currently the axial distribution within
the core is modelled.
RT-RU-008 RU Pebble Bed Reactor core
incl graphite
SSC
Fission Nuclei Breeding Low Sure High The total effect on the FOM is
minimal
Nuclei Breeding is captured in the
characterisation of power production.
The effect is predominantly considered
over the whole core and not to the
kernel level. 19
Oct 22-26, 2012 IAEA Course on High temperature Gas Cooled Reactor Technology 20
PIRT – Iterative process • Positive outcomes:
– Good basis for identifying EM development requirements
– Easier to justify modelling assumptions • Challenges:
– In a new technology experts are not readily available • Spent a lot of time on irrelevant phenomena • Ranking may be incorrect • If nobody knows then typically have to spend a lot of time to
find out … but in the end this is positive – Include external experts
• Improves credibility (but also complexity…) – Perform hierarchical breakdown – Revised ranking bins
• Difficult to decide what to do with “medium” bins • Too many combinations of uncertainties available to
adequately action resolution
Oct 22-26, 2012 IAEA Course on High temperature Gas Cooled Reactor Technology
• Source material used: – HTGR Technology Course for the Nuclear Regulatory Commission,
May 24 – 27, 2010
– HTR/ECS 2002 High temperature Reactor School, 2002
– MUA 784: Reactor Physics, F Reitsma, Mechanical Engineering Post-Graduate: Nuclear Theme, University of Pretoria, 2012
– Workshop at PHYSOR 2010 – Advances in Reactor Physics to Power the Nuclear Renaissance: The Pebble Bed Modular Reactor: From V.S.O.P. (Very Superior Old Product) to Generation–IV candidate.
– “Safety Analysis Software Development and V&V”, Peter Robinson, Workshop on Safety Aspects of Modular HTGRs, October 2007, Beijing China
– Radionuclide transport during normal operation conditions”, Lize Stassen, Pieter Goede, Gen-IV CMVB Chemistry and Transport Workshop, Centurion, South Africa, 12 – 13 January 2009
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