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nuclearsafety.gc.ca 2nd October, 2012
Challenges and Future Needs for Independent Reactor Physics Simulations – CNSC Approach and Practice
Dr. Dumitru Serghiuta Canadian Nuclear Safety Commission
IAEA WORKSHOP ON ADVANCED CODE SUITE FOR DESIGN, SAFETY ANALYSIS AND OPERATION OF HEAVY WATER REACTORS
Ottawa, Canada
Canadian Nuclear Safety Commission 2
Outline – Key Messages
o The experience and practice to date, the current and future industry needs and nowadays more demanding safety expectations increase the complexity of regulatory technical assessments and the role of independent confirmatory simulations.
o Reactor configurations are becoming more heterogeneous both in composition and in the distribution of power throughout the core. This makes predictions more difficult when assessing core behaviour and determining safety parameters, such as reactivity coefficients, that dictate transient behaviour.
o Thermal-hydraulic and neutronic codes are now being coupled in order to address safety issues. The modern use of advanced computational methods to refine safety analyses and safety margins has emphasized the need for more detailed data and advanced analytical methods, as well as the need for uncertainty assessments and new approaches in the area of validation and qualification.
Canadian Nuclear Safety Commission 3
Outline – Key Messages o CNSC has taken a proactive approach in the area of independent
reactor physics simulations and plays a catalyzing role in the development and introduction of advanced codes and use of modern V&V and UQ methods.
o Consolidation and enhancement of independent capability for reactor
physics analysis of HWR needs development of an integrated, state-of-the-art, user-friendly system of neutronic methods and computer codes which incorporates uncertainty assessment features and can be used for a large variety of applications, including reactor simulations and safety applications.
o HWR specific international benchmarking activities under the auspices of IAEA would significantly contribute to sharing of experience and harmonization of expectations and practices related to advanced codes and code suites for HWR independent confirmatory simulations.
Canadian Nuclear Safety Commission 4
Established May 2000, under the Nuclear Safety and Control Act
Replaced the AECB, established in 1946, Atomic Energy Control Act
Regulates the use of nuclear energy and materials to protect the health, safety and security of Canadians and the environment; and to implement Canada’s international commitments on the peaceful use of nuclear energy.
Canadian Nuclear Safety Commission
Canada’s independent nuclear regulator
65 years of experience
Established May 2000, under
the Nuclear Safety and
Control Act
Replaced the AECB of the 1946
Atomic Energy Control Act
Canadian Nuclear Safety Commission University of Saskatchewan 10.03.25 - 3
Canadian Nuclear Safety Commission 5
Typical share of nuclear energy in total electricity generation
Ontario - 52% New Brunswick - 30%
Quebec - 3% Canada - 14.7%
Gentilly QC
Point Lepreau
NB
Operable status (Average age – 25 Years)
In service / Returned to service
Safe storage state
In refurbishment
In service within design life
Bruce ON Darlington ON Pickering ON
A2 A4 A3 A1
B5 B8 B6 B7
In service 1971 Safe
storage state
In service 1971/2003 Mwe 515
In service 1972 Safe
storage state
In service 1971/2005 Mwe 515
In service 1983
Mwe 516
In service 1986
Mwe 516
In service 1984
Mwe 516
In service 1985
Mwe 516
1 2
3 4
In service 1993
Mwe 881
In service 1993
Mwe 881
In service 1992
Mwe 881
In service 1990
Mwe 881
A2 A4 A3 A1
B5 B8 B6 B7
In service 1977
Mwe 750
In service 1979/2003 Mwe 750
In service 1978/2003 Mwe 750
In service 1977
Mwe 750
In service 1985
Mwe 882
In service 1987
Mwe 882
In service 1984
Mwe 882
In service 1986
Mwe 882
In service 1983
Mwe 635
In service 1983
Mwe 635
Canada’s Nuclear Energy Profile
Canadian Nuclear Safety Commission 6
Regulatory Approach
o Approach stems from the Nuclear Safety and Control Act (NSCA) and CNSC regulations
o Three principles:
• Plan for the complete life of the facility
• Multi-barriers for the protection of people and the environment
• Defence in depth – never rely on a single system or process for protection
Canadian Nuclear Safety Commission 7
Regulator Needs Technical Support
Licensing
Inspection
Enforcement
Independent Review
Research
Standards
Legal
Operations Branch Technical Support Branch
Canadian Nuclear Safety Commission 8
Regulatory Independent Technical Assessment
o A core competency essential to the Canadian Nuclear Safety Commission (CNSC) mission is independent evaluation of licensees’ submittals in support of their safety cases.
o Reactor physics analysis, criticality safety, fuel and material science are among the core competencies essential to the pursuit of the CNSC mission.
o Although the licensees have the responsibility for safety and supporting technical analyses, the licensees and, indeed, the public expect the regulator to have capabilities in these areas that are at or very near the state-of-the-art and to use objective evidence, including advanced computational methods and accurate experimental evidence.
o Therefore, CNSC has enhanced over the last two decades its technical capability in these areas to make technically sound regulatory recommendations.
Canadian Nuclear Safety Commission 9
Technical Safety Objective
To prevent with high confidence accidents in nuclear plants; to ensure that, for all accidents taken into account in the design of the plant, even those of very low probability, radiological consequences, if any, would be minor; and to ensure that the likelihood of severe accidents with serious radiological consequences is extremely small.
Verify and confirm the technical claims in the licensees’ safety case
Defence in Depth Several failures must happen before radioactivity is released! Proven Engineering Practices Technically sound, adequately qualified and verifiable methods and practices Margins and Standards
Canadian Nuclear Safety Commission 10
Safety (Risk) Assessment
Kaplan & Garrick’s Questions
What can happen? Design assessment + PSA + Expert
judgment
How likely is it to happen? PSA
What are the consequences if it
happens?
Analytical simulations + Expert
judgment +Simplified
phenomenological or empirical models
How much confidence exists in the
answers to the above questions?
Design assessment + Model V&V +
Experimental database + Uncertainty
Quantification + Expert judgment
Canadian Nuclear Safety Commission 11
Reactor Physics and Regulatory Technical Assessment
o Key component in review and assessment of:
– core design and core nuclear performance
– safe operating envelope
– compliance with safe operating envelope limits
o Industry reactor physics codes are reviewed, but not approved
Canadian Nuclear Safety Commission 12
Outcome of Technical Assessments
o Recommendations for licensing actions, regulatory positions, regulatory expectations, standards
o Identification and recommended resolution of specific and generic safety issues
o Potential modifications, including changes to: – Processes
– Safety cases
• Limiting conditions for operation
– Plant design
– Operations & procedures
• Operating, monitoring and surveillance procedures
• Maintenance procedures
• Overriding safety devices
Canadian Nuclear Safety Commission
CANDU Industry Reactor Physics Codes
13
Infinite Lattice Void Reactivity
Delta 1/k in units of mk
4
9
14
19
-2000 0 2000 4000 6000 8000 10000 12000 14000
Burnup MWd/te(initial heavy elements)
Vo
id r
eacti
vit
y m
k
WIMS-AECL E5L
WIMS-AECL E6L
HELIOS
WIMS-UK
PPV-JG
PPV-AECB
MCNP
1997 PPV Benchmarking (AECB Research Project)
Simulation of 1996 G-2 Loss of Class IV Event
Design and safety analysis codes
POWEDERPUFS-V, RFSP,
MULTICELL, and SMOKIN
replaced with Industry Standard
Toolset (IST) suite of codes:
WIMS (with ENDF/B-VI Release 8),
DRAGON, RFSP (2g diffusion), and MCNP.
CNSC GAI 99G02 was raised to track the
progress of the transition and the validation
and verification of the IST reactor physics
codes. Complementary licensing action on
core surveillance and monitoring software
(fuel management) open in GAI 01G01.
Continuous effort on development and
validation of IST reactor physics: new
versions, like WIMS 3.1, expected to be
introduced soon.
Canadian Nuclear Safety Commission 14
Why Continuing
Research? R&D in CANDU Reactor
Physics area remains a high
priority issue
Main Drivers: Aging
Periodic Safety Review:
Effectiveness of Special Safety
Systems and Safety Margins
Improvements in operation
performance and flexibility
Refurbishment: Nuclear Design
Reconstitution
New Designs
Main areas for continuing R&D: Generic Safety Issues
Continuous development of the
capability of methods and computer
codes
Assessment and validation for
power reactor conditions
Standards
Support continued safe operation under new
conditions due to aging
Support life extension
Support new safety analysis methodologies
proposed by the Industry
Improvement of the understanding of the
phenomena of interest for safety
Better assessment and confirmation of the
design margins
Canadian Nuclear Safety Commission
CNSC Approach and Codes
15
External Consultants: Independent Assessment and
Simulations
In-house Review and Simulations
HELIOS
NESTLE-C
DRAGON
MCNP
COMET
ATTILA
Canadian Nuclear Safety Commission
Challenges
16
Challenging Physics Issues: Impact of ageing Reactivity Feedback – VR and PCR Power pulse Compliance with physics and fuel limits
Challenges for Physics Simulations: Modeling of real system Approximations Characterization of errors and uncertainties Qualification of confidence in predictions
Canadian Nuclear Safety Commission
Traditional Computational Procedure
17
Coupled diffusion / transport method
Transport Theory:
Homogenize the fuel lattice using a lattice depletion transport method
Solve the multigroup transport equation in the lattice assuming no neutron leakage
Use the neutron flux to generate homogenized multigroup cross sections
Diffusion Theory:
Model the core as a collection of homogenized boxes and solve the diffusion equation to obtain the flux distribution
HELIOS
DRAGON
Homogenized 2 and 4 groups cross-sections
and incremental
cross-sections
NESTLE-CANDU TH Bundle-Wise Data
Canadian Nuclear Safety Commission
Refined Computational Procedures
18
HELIOS MCNP
Multigroups
Multi-groups Cross-
Sections
MCNP 2 &4 groups
Cross-Sections
NESTLE
NESTLE Node Albedos MCNP
MCNP Node
Albedos HELIOS
Customized to help investigate: - Effect of certain
approximations in traditional methods
- Effect of core environment
Canadian Nuclear Safety Commission
Challenges in Modeling - Ageing
19
HELIOS Lattice Cell Models for Creep (5%) and Sag (7.7 cm)
MCNP Models used in Sub-region and Full Core simulations
Canadian Nuclear Safety Commission
Challenges in Modeling – Pin Power In Power Pulse
20
NESTLE 2-groups albedo for the node correspond to bundle with maximum enthalpy MCNP pin tally calculations
Canadian Nuclear Safety Commission
Challenges in Modelling of Full System
21
1995 Gentilly-2 loss of class 4 power event Simultaneous trip of PHT and LZC pumps Reactivity insertion initiated by drainage of LZC and accelerated by boiling Transient terminated by SDS1 within 2 sec of event initiation Peak reactor power ~ 110% FP
Process of selection of models Expert judgment and level of knowledge State of the art models necessary
Uncertainties related to process system performance models
Uncertainties specific to time frame of the transient Potential significant residual uncertainty even in areas where large amount of data is available
Canadian Nuclear Safety Commission 22
Stylized CANDU Core Benchmark Problems
A 3D (half-core) benchmark problem has been
developed based on a generic model of a CANDU
reactor. The benchmark problem was designed to
satisfy several conditions:
It retains the dominant characteristics of CANDU
reactors.
It maintains a realistic degree of heterogeneity
without being unnecessarily burdened by details
that do not affect the underlying physics of the
reactor.
The resulting problem is therefore referred to as a
stylized model since it captures the fundamental
properties of CANDU reactors (from a neutronics
point of view) without being constrained to a particular
realization of an operating CANDU plant. The development of transport theory benchmark problems in
reactor physics serves two primary purposes:
(1) Model test cases in the development and testing of new
and existing transport theory methods. Benchmark problems
are necessary to properly vet these methods and compare them
against existing solution techniques.
(2) Reference points for assessing existing reactor physics
methodologies or to determine the capabilities and limits of
existing core analysis techniques. Accurate benchmark
solutions, in this case, become a tool to isolate and quantify
approximations present in the existing methodologies.
Canadian Nuclear Safety Commission 23
Why Uncertainty Quantification?
o In spite of the wide spread use of Modeling and
Simulation (M&S) tools it remains difficult to
provide objective confidence levels in the
quantitative information obtained from numerical
predictions
o Use of M&S predictions in high-impact decisions
require a rigorous evaluation of the confidence
Canadian Nuclear Safety Commission 24
V&V and UQ
o The accepted process of evaluating M&S tools and solutions is based on the general concept of Verification and Validation (V&V)
o The last step of the process is invariably based on comparisons between numerical predictions and physical observations
o Precise quantification of the errors and uncertainties is required to establish predictive capabilities: UQ is a key ingredient of validation!
Canadian Nuclear Safety Commission 25
Plant and in-core diagnostics
PIE of used fuels
In- and out-of-pile testing of prototypic fuels
Separate-effect tests
Integral effect tests
Operational conditions
Fuel performance (crud, corrosion, fretting, failures)
Fuel behavior and capacity limits
“Elementary” physics (T/H, S/M, M/S)
Multi-physics and system dynamics
Multi-Physics Sub-
system and System
V&V and UQ
There is a growing realisation of the importance of uncertainty in simulator predictions Can we trust them? Without any quantification of output uncertainty,
it’s easy to dismiss them
Evolve to the next levels of V&V and UQ
Canadian Nuclear Safety Commission 26
Generic Challenge
System simulation requires the use of multi-physics code
systems.
The current qualification procedures of coupled multi-
physics code systems are still based on the verification and
validation of separate physics models/codes.
Although some V&V of the coupling methodologies of the
different physics models is possible, it may be too limited,
because of availability of experimental data (integral-effect
test data).
New processes and guidelines for V&V and UQ for multi-
physics code systems are necessary
Canadian Nuclear Safety Commission 27
Challenge: Process and Guidelines
More than 20 conceptually conflicting V&V standards have been produced Concepts and definitions Software reliability vs M&S credibility Error vs Uncertainty Process and methods
Canadian Nuclear Safety Commission 28
Concepts: Verification & Validation
Verification: The process of determining that a model implementation accurately represents the developer’s conceptual description of the model.
Validation: The process of determining
the degree to which a model is an
accurate representation of the real world
for the intended uses of the model
Verification aims at answering the
question “are we solving the equations
correctly?” – it is an exercise in
mathematics
Validation aims at answering the
question “are we solving the correct
equations?” – it is an exercise in physics
Canadian Nuclear Safety Commission
ASME Draft V&V 30 Standard for Verification and Validation of System Analysis and Computational Fluid Dynamics Software for Nuclear Applications
30
Canadian Nuclear Safety Commission 31
Qualify the level of
confidence in
predictions: Put the
“error bars” on
simulations’ results!
Sensitivity analysis (SA) investigates the
connection between inputs and outputs of
a (computational) model
The objective of SA is to identify how the
variability in an output quantity of interest
is connected to an input in the model; the
result is a sensitivity derivative
SA allows to build a ranking of the input
sources which might dominate the
response of the system
Strong large sensitivities derivatives do
not necessarily translate in critical
uncertainties because the input variability
might be very small in a specific device
of interest.
Characterization of Uncertainty
in Outputs:
1. Data assimilation:
characterize uncertainties in the
inputs
2. Uncertainty propagation:
perform simulations accounting
for the identified uncertainties
3. Certification: establish
acceptance criteria for
predictions
Canadian Nuclear Safety Commission 32
Qualify the level of
confidence in
predictions: Put the
“error bars” on
simulations’ results!
GPT (TSUNAMI) Requires adjoint code
development
Must linearize non-linear
equations
Number of necessary code runs
depends on the number of
responses of interest
Provides cross-section sensitivity
information
Careful use of “cross-sections”
adjustment
Careful use of data from
“experimental tests”
Stochastic Method Forward Propagation of
uncertainties Can be implemented as a
“black box”
Can handle non-linear
systems and non-Gaussian
probability distribtutions.
Number of code runs required
depends only on the level of
statistical accuracy desired
Difficult to generate sensitivity
information
Canadian Nuclear Safety Commission 33
Coupled in-core neutronics,
TH, and fuel performance
Containment System
Increased predictive capability: integrated,
more science-based multi-physics models
Avoid diffusion theory and homogenization:
Whole-core transport methods
Improve computational efficiency of fine-
mesh transport methods using new and
sophisticated acceleration techniques
Variant (sub-element method – explicit
modeling of fuel pin) Coarse-mesh transport
method without homogenization
Integrated, efficient, easy to use
predictive capability
Integrate SA and UQM
Multidisciplinary design team:
include IT experts!
Sub-system and system
validation
Learn to live with uncertainties!
Needed Evolution of Predictive Capability
Canadian Nuclear Safety Commission 34
Power Pulse
0
50
100
150
200
250
300
350
400
0 0.2 0.4 0.6 0.8 1 1.2 1.4 1.6 1.8 2
Time (s)
Relative co
re p
ow
er (%
)
Reference
Most favourable combination
Most conservative combination
LOCA Margins Power pulse simulation methodology In a CANDU reactor with its large positive CVR response, the simultaneous occurrence of severe positive reactivity insertion and severe degraded core cooling can result from a single potential initiating event i.e., LLOCA. Hence, the control, cool and contain safety functions of a CANDU reactor must be designed to cope with the consequences of both limiting events at the same time and effectiveness of shutdown systems is essential.
Integrated Multi-Physics Simulations: -LOCA -LORCA -LOF
State of the art
predictive capability
SA and UQ capability
Sub-system and
system V&V
Power Pulse Issue
Core Neutronic
Performance and
Effectiveness of
Shutdown Systems
Canadian Nuclear Safety Commission 35
Power Pulse
0
50
100
150
200
250
300
350
400
0 0.2 0.4 0.6 0.8 1 1.2 1.4 1.6 1.8 2
Time (s)
Relative co
re p
ow
er (%
)
Reference
Most favourable combination
Most conservative combination
Physics Input to Power Pulse Simulations
Reference bounds for
void reactivity, beta, FTC,
PCR
Conservative
bounding values How to do it
How to implement it
Uncertainty
What UQ Method is: sufficient
feasible
Canadian Nuclear Safety Commission 36
Stochastic-deterministic
approach based on
representation of
uncertainty in output by
subjective probabilities
No statement about
values of physics
parameters implied
Develop a framework for sensitivity and uncertainty
analysis for a general reactor design and analysis code
system based on well-established techniques for
uncertainty and sensitivity analysis for deterministic
methods (i.e., deterministic core simulators) and
probabilistic models (i.e., MCNP) to achieve the
following goals in a computationally efficient manner:
(a) Identify the most influential cross-sections
perturbations in the point-wise energy format on the
responses of interest.
(b) Propagate available cross-sections uncertainty
information through a probabilistic model (i.e. MCNP) to
the responses of interest
(c) Transfer sensitivity and uncertainty information at the
interface between probabilistic and deterministic models
(i.e., between MCNP and deterministic core simulator
codes)
(d) Propagate uncertainties of cross-sections in the group-
wise representation through the deterministic code
(e) Develop a variance reduction technique for
probabilistic models that accelerates the perturbation runs
required for sensitivity and uncertainty analysis.
(f) Determine the uncertainty contribution due to missing
and inaccurate cross section uncertainty information.
(g) Determine the numerical methods and modeling
(N&M) error uncertainty distribution by using stochastic
results as a surrogate for experimental data.
Potential Approaches
Expensive?
Work in progress
Canadian Nuclear Safety Commission 37
Conservative approach based
on representation of
uncertainty by subjective
probabilities
What do we know? -Range of most of initial
conditions
-Mean realistic value of void
reactivity less than value for
fresh fuel composition
-Mean realistic value of beta
higher than value for discharge
burnup composition
-Mean realistic value of FTC
negative for fresh fuel and
positive for discharge burnup
composition
What we do not know? -Impact of uncertainty in
nuclear cross-sections
-Structural (model) uncertainty
-Covariance and correlations
Canadian Nuclear Safety Commission 38
Operating-center
CANDU-6 LOCA case CATHENA time-dependent node-
wise thermal hydraulic
parameters
Fixed trip time actuation
Cross-sections changes from
lattice-cell branch-off
calculations used to simulate
“structural uncertainty”
Estimated maximum bundle energy deposition during power
pulse for several combinations (512) “equivalent” to: -Any value of void reactivity
-Any value of beta
-Any value of FTC
-(Any value of generation time)
HELIOS – NESTLE-C
How we do it
Canadian Nuclear Safety Commission 40
What we get
Power Pulse
0
50
100
150
200
250
300
350
400
0 0.2 0.4 0.6 0.8 1 1.2 1.4 1.6 1.8 2
Time (s)
Rela
tiv
e c
ore p
ow
er (
%)
Reference
Most favourable combination
Most conservative combination
A subjective probability is
based on some knowledge
often referred to as the
background knowledge.
The background
knowledge comprises
assumptions and
suppositions, models, etc.
Among the earlier
probability theorists, a
subjective probability is
linked to betting.
For many the focus is on
decision-making under
uncertainty, and then the
betting situation applies:
the distinction between
uncertainty assessment
and value judgments is
not important.
Canadian Nuclear Safety Commission 41
How we interpret it?
0
0.5
1
1.5
2
2.5
3
3.5
0.4 0.6 0.8 1 1.2 1.4 1.6
Normalised Bundle Enthalpy
No
rm
al
Dis
trib
uti
on
0
0.5
1
1.5
2
2.5
3
3.5
Oc
cu
re
nc
e d
en
sit
y f
ro
m S
imu
lati
on
s
Normal Distribution (mean=1,
sigma=0.14)
Occurence density from
Simulations
Assume a normal
probability distribution
function with the mean
= realistic value (~ 50%
confidence level)
The test results appear
to support this
assumptions
Statisticalprocessing
Potential Energy
Deposition = Realistic
Value + 60%
Canadian Nuclear Safety Commission 42
How would
one use
this
approach? Statistical Estimate with Imperfect
Information Tolerance limit and order statistics
Use of “expensive” approach to quantify
uncertainties in outputs
Best-Estimate with an Estimate
of Uncertainty
∆Esys = [ERealistic+conservative assumption – ERealistic]
∆Ei = Ei-th perturbed parameter - Enominal
BE-E = ERealistic + k [(∆Esys)2 + ∑ (∆Ei)
2]1/2
Where k defines the fraction of all LOCAs
with Es bounded by BE-E
Canadian Nuclear Safety Commission
Final Remarks
o Consolidation and enhancement of independent capability for reactor physics analysis of HWR needs development of an integrated, state-of-the-art, user-friendly system of neutronic methods and computer codes which incorporates uncertainty assessment features and can be used for a large variety of applications, including reactor simulations and safety applications.
o HWR specific international benchmarking activities under the
auspices of IAEA would significantly contribute to sharing of experience and harmonization of expectations and practices related to advanced codes and code suites for HWR independent confirmatory simulations: Benchmarking
Physics benchmark problems Postulated transients Plant transients
Specific expectations and recommendations for V&V of M&S for HWR
43