Challenges and Future Needs for Independent Reactor Physics Simulations · PDF file · 2012-10-10Challenges and Future Needs for Independent Reactor Physics Simulations ... public

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.


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.


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

http://nuclearsafety.gc.ca/


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 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 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


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


Regulator Needs Technical Support

Licensing

Inspection

Enforcement

Independent Review

Research

Standards

Legal

Operations Branch Technical Support Branch


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.


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


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


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


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


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.


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


CNSC Approach and Codes

15

External Consultants: Independent Assessment and

Simulations

In-house Review and Simulations

HELIOS

NESTLE-C

DRAGON

MCNP

COMET

ATTILA


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


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


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


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


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


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


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.


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


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!


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


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


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


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


ASME V&V 10-2006

29


ASME Draft V&V 30 Standard for Verification and Validation of System Analysis and Computational Fluid Dynamics Software for Nuclear Applications

30


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


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


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


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


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



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


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


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


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


Cross Sections Representation in NESTLE-C


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



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.


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%


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


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

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Challenges and Future Needs for Independent Reactor Physics Simulations · PDF file · 2012-10-10Challenges and Future Needs for Independent Reactor Physics Simulations ... public