Robert Montgomery1, C.R. Stanek2

W. Liu3, B. Kendrick2

Presented at the

IAEA Technical Meeting “Modelling of Water-Cooled Fuel Including

Design-Basis and Severe Accidents”

28 October – 1 November 2013, Chengdu, China

US DOE CASL Program Fuel Performance Modeling for Steady State

and Transient Analysis of LWR Fuel

This work was supported by the U.S. Department of Energy, Office of Nuclear Energy.

3

1

2

Presentation Overview

Overview of CASL Program Fuel Modeling Efforts

Accident fuel behavior and importance of modeling/modeling needs

Peregrine and MAMBA development activities

Lower length modeling/methods development

Data Needs for Fuel Modeling Support

Lower length modeling/methods development

Transient modeling activities in CASL

Summary

“Energy Innovation Hub” Consortium for Advanced Simulation of LWRs

CASL will apply existing modeling and simulation (M&S) capabilities and

develop advanced capabilities to create a usable environment for

predictive simulation of light water reactors.

CASL partners are from national labs, universities and industry.

CASL consists of 6 technical

“focus areas”:

• Materials Performance

• Thermal Hydraulics

• Neutronics

• Validation/Uncertainty

• Model Applications

• Virtual Reactor development

Longer-term priorities (years 6–10) Near-term priorities (years 1–5)

CASL vision: Develop & apply a “Virtual Reactor” to assess fuel design, operation and safety

Deliver improved predictive simulation of PWR core, internals, and vessel

Couple VR to evolving out-of-vessel simulation capability

Maintain applicability to other NPP types

Execute work in 5 technical focus areas to:

Equip the VR with necessary physical models and multiphysics integrators

Build the VR with a comprehensive, usable, and extensible software system

Validate and assess the VR models with self-consistent quantified uncertainties

Expand activities to include structures, systems, and components beyond the reactor vessel

Established a focused effort on BWRs and SMRs

Continue focus on delivering a useful VR to:

Reactor designers

NPP operators

Nuclear regulators

New generation of nuclear energy professionals

Fuel Behavior is a Combination of Complex Interactions: Modeling is Required*

! "#$%&' () "%$(8@%'@3%K (Z%*(<<<( Z%*(<[ (

89: ;<! =(4 "$560123#2(4 %-0. +(C"*%(Q;EA(FGHG( \ (

! "#$%&' () "%$(8@%'@3%K (Z%*(<<<( Z%*(<[ (

8 9: ;<! =(4 "$560123#2(4 %-0. +(C"*%(Q;EA(FGHG( \ (

! "#$%&' () "%$(8@%'@3%K (Z%*(<<<( Z%*(<[ (

8 9: ;<! =(4 "$560123#2(4 %-0. +(C"*%(Q;EA(FGHG( \ (

! "#$%&' () "%$(8@%'@3%K (Z%*(<<<( Z%*(<[ (

89: ;<! =(4 "$560123#2(4 %-0. +(C"*%(Q;EA(FGHG( \ (

* Rashid, Yagnik, and Montgomery, JOM 63, no. 8 (2011)

Outcomes and Impact Approach

Multi-Physics/Scale Material Modeling Enabling Improved Fuel Performance through Predictive Simulation

• Provide physics-based materials models of fuel/clad/internals property evolution to enable predictive modeling of CRUD, GTRF and PCI within 3D, multi-physics, virtual reactor simulator

• Improved physics and chemistry insight delivered via constitutive relations

• MPO is comprised of a diverse group of computational materials scientists with a wide range of capabilities

• Predictive models of fuel failure, that quantitatively define operating margins & lifetime limits

• Validated predictions of fuel failure conditions

• Power uprates & increased fuel utilization

Challenging, multiscale processes

impact nuclear fuel performance

Fuel Performance Modeling in CASL for PCI and CRUD

A series of microscale activities provide mechanistic/physical

insight into complex degradation phenomena

CRUD

MAMBA (MPO Advanced Model

for Boron Analysis)

PCI

Peregrine (Fuel Performance)

Copyright by ASTM Int'l (all rights reserved); Fri Aug 14 20:28:09 EDT 2009Downloaded/printed byDion Sunderland (ANATECH Corp) pursuant to License Agreement. No further reproductions authorized.

PCI in Fuel Behavior Modeling: Why it is important?

PCI failure potential limits reactor performance associated with power uprates, higher burnup, fuel rod manufacturing quality and operating flexibility during power changes

Requires new 3D multi-physics simulation capability to reduce uncertainties in assessing PCI failure conditions during normal operation and in the presence of anomalies`

Material

Properties &

Characteristics

Reactor

Neutronics

Thermal

Hydraulics

PCI

Fuel Behavior

Analysis and Modeling

Methodology

PCI is controlled

by local effects

PCI is possible in many

rods and assemblies

PCI has system wide

influence

CRUD Buildup and Boron Deposition Is Important for Plant Operation

Large Mass of Material

In Reactor System for

Corrosion

CRUD deposition is

wide spread with non-

uniform effects

Reaction kinetics occur

at lower microscale

Transient Fuel Behavior: Reactivity Insertion Accidents (RIA)*

Pellet Thermal Expansion

- Pellet-Cladding Contact

- PCMI Loading

Cladding Failure by Hydrogen-Induced

Embrittlement

Phase 1

Heat Conduction to the Cladding

- Increase Cladding Temperature

- Initiate DNB

- Decrease Cladding Strength

Grain Boundary Cracking and Fission Gas Release

- Increase Rod Internal Pressure

- Additional Radial Deformation

Phase 2

Time During Power Pulse

Pow

er

Modeling identified the need for specific tests to address experimental

observations and determine applicability to in-reactor behavior

* Sunderland, Montgomery, Ozer, ANS LWR Fuel 2004

RIA Fuel Behavior Is Function of Prior Irradiation and Event Conditions

Cladding Temperature

Effect on Permanent Strain

Oxide/Hydrogen Content

Effects on Clad Cracking

Improved multi-physics and multi-scale modeling methods are necessary to

understand:

• Non-uniform power deposition on failure modes

• Cracking and spalling of the corrosion layer under rapid loading conditions

• Tighter coupling between void volume pressure and structural deformations

• Cladding to coolant heat transfer and high temperature material behavior

Data needs and model development for transient fuel performance modeling Key phenomena to model (Zr clad &/or

advanced cladding)

Fuel & Clad phase changes

O/H reactions & diffusion, hydrogen uptake/hydride precipitation

Pellet – clad chemical/mechanical interactions (RIA)

Thermal profiles, including axially dependent decay and reaction heat

Large strain plasticity & failure (including ballooning, secondary hydride cracking and fragmentation of fuel & clad)

Quench loads and post-quench embrittlement

Data Needs (Thermo-Chemical-Mechanical, separate & integral)

Temperature & state (BU, dpa, oxide)-dependent thermal (conductivity) & mechanical (creep, fracture) properties

Temperature, pressure & steam quality dependent thermochemical reactions (oxidation kinetics, hydrogen fate, breakaway oxidation mechanisms?) & fuel-clad chemical interaction

Thermal-mechanical processes (ductility/creep, fracture) at T & post-rewetting quench & as a function of state (dpa, oxide, H, …)

Embrittlement mechanisms & integral performance

Peregrine: Advanced Fuel Rod Modeling Capability for LWRs

Purpose

Enhance the modeling of thermal, mechanical, and chemical behavior of LWR fuel using multi-physics and multi-scale methods to reduce uncertainties in performance and safety margins

Approach

Based modern finite element computational framework for 2-D/3-D representation of a single fuel rod

Uses versatile material properties and behavior model constitutive structure for UO2 pellets and zirconium-alloy tubes.

Designed to leverage results from lower length scale models/methods

Benchmark and validation efforts working in parallel with development activities

Peregrine: Code Structure and Bases

- Elements provided by MOOSE/ELK/FOX systems (INL)

Problem Setup

•Power Density

•Burnup Distribution

•Fast/Thermal Flux Dist.

•Thermal B/C

•Mechanical B/C

Material Properties (e.g.)

•Specific Heat

•Density

•Conductivity

•Elastic/Plastic Behavior

•Creep Rate

Behavior Models (e.g.)

•Swelling/Densification/Cracking

• Irradiation Creep/Growth

•Grain Growth

•Fission Gas Release

•Corrosion

Physics Solution

•Thermal Transport

•Mechanical Forces

•Chemical Transport

Global Parameters Integration

•Rod Pressure

•Fission Gas Composition

•Linear Power

Result Output

•Write global variables

•Shift Internal State Variables

Peregrine Interface with other CASL Material Modeling Efforts and Core Simulator

Peregrine (Fuel Rod

Performance)

Neutronics/

Isotopics Subchannel T-H

Cladding Creep & Growth

Stress Corrosion Cracking

Validation Application

Hydrogen Diffusion &

Precipitation

Cladding Corrosion/H-

Pickup

¢¢¢f , f, Ci ¢¢q ,Tcool,Hcool,Pcool,Sitox, DD, ¢¢H

H,ZrH2

ecr, eplPf , t f , s cr

Virtual Reactor

Core Simulator

MAMBA

HYRAX

VPSC

MAMBA: Advanced CRUD Buildup and Boron Deposition Modeling

Purpose

Develop a multi-scale multi-physics reaction rate modeling approach for CRUD formation and growth beyond current industry practices

Approach

Construct 2-D & 3-D representation of the cladding surface for a single fuel rod with refined detail to capture local fluid/temperature effects

Species concentrations for each phase (particulates, solubles, vapor, and CRUD solids) within the CRUD layer are considered

Calculate mass and heat transport as function of local rod surface heat flux, sub-cooled boiling, CRUD microstructure, and chemical kinetics

Include conversion of species between the phases due to surface deposition/release, precipitation/ dissolution, sub‐cooled nucleate boiling, diffusion, and convection.

MAMBA flow chart: phases, interfaces, and submodels

Primary Species:

C1 = Ni

C2 = Fe

C3 = Zr

C4 = Zn

C5 = B

C6 = Li

C7 = Cr

C8 = Co

O and H implied

Chemical kinetics mechanisms

will compute species concentrations

versus time for “local conditions” at

each “node” within computational cell

Vapor (SNB)

Solubles

CSi

Particulates

CPi

CRUD

CMi

CILC

Rates

Coolant Flow

and Chemistry

Transport

CIPS

MAMBA CRUD deposition model: 2D View

Colored contours:

normalized boron

concentration

Compute node and

volume element

Heat Transport between nodes: 3D non-linear,

iterative, numerical solution at each time step,

with local “sinks” due to boiling

CRUD/coolant interface is time

dependent (adaptive):

deposition & “erosion”

Thermodynamics and chemical kinetics

computed at each node and time step

Boiling induced convective &

diffusive transport between

nodes and coolant

Cladding Surface

Coolant Flow

MAMBA Material Framework for CRUD Modeling

MAMBA MPO CRUD model

ChemPaC LANL/NNSA

BOA Westinghouse/NNL/EPRI

Boron/H2 Model MIT

NiO Model

NC State

Thermodynamics

ORNL/LANL

Thermal cond./

Sub-cooled Boiling

LANL

MAMBA = MPO Advanced Model for Boron Analysis

Lower Length Scale Modeling Activities in CASL – Important for Transients

High temperature thermal creep and creep rupture

Thermochemistry of coolant particulates/chemistry and cladding reactions for CRUD layer formation

CRUD layer deposition model that captures constituents, morphology, and thickness

Water side corrosion kinetics for oxide layer crystallography, morphology, and thickness

Hydrogen uptake, hydride formation, and redistribution

Xenon diffusion and bubble growth behavior in irradiated UO2 material

We can now predict defect-

dislocation interaction with high

accuracy, including effects that go

beyond anisotropic linear elasticity

BO transforms into

BS at ~ 0.008

Comparison of atomistics and continuum defect

energies for interstitials. Excellent agreement is

obtained for strains up to ~ 5% (full range not shown)

Types of self-interstitials

Defect energy in the field of a dislocation.

Away from the core, atomistics matches

continuum solution

Atomistics

captures

details

missed by

theory

Atomistic characterization of point defects in Zr

First Principles Study of CRUD solvation ■ Thermodynamic stabilities can

be used to determine which

species are driven to deposit as

CRUD.

■ First principles quantum

chemical methods are used to

compute free energies of

formation of clusters/species

(ΔGFP)

MDFPtot GGG

HOHNiOH2NiO

OHNiOHNiO

OHNiHNiO

OHNiH2NiO

32

0

22

2

2

cr

cr

cr

cr

Prior model reactions for

dissolution of NiO in water

- (A)

- (B)

- (C)

- (D)

Electron density

iso-surface for

Ni+2(H2O)6

(NWCHEM)

B3LYP/ LANL2DZ

ΔGMD

MD simulations for solvation free

energies of Ni+2(H2O)6 clusters

First Principles Analysis of Intermediate Temperature Mechanisms

Traditionally explained by enhanced

vacancy concentration according to:

However, this explanation relies

on a particular set of data that,

according to calculations, may not

be accurate.

D2 s2JVV

V Virr K /JV Z

TkFD

B

2.1expconstant2

Applying the same assumption as above

yields yields 3.17 eV. Interestingly this is

close to the high temperature regime.

Obtained by assuming:

JV exp 2.4

kBT

JXe exp Em

XeU2O EB

XeU2O

kBT

From DFT:

JV exp Em

VU2

kBT

Further analysis required, alternative mechanisms

may be important for intermediate T and high T.

25

3rd Annual DOE Review of CASL, Oak Ridge National Laboratory, Aug 13-14, 2013

• Evaluate Peregrine to calculate the thermal, mechanical, and irradiation behavior of UO2/Zr-alloy fuel rods

• Fuel temperature, cladding displacements, and fission gas release measurements from 14 fuel rods used in assessment

• Compare to Falcon results to understand performance of material and behavior models

• Peregrine is on par with industry standard codes (e.g. FALCON) up to 75 GWd/tU

• Reliably calculates fuel temperatures, cladding deformations, and fission gas release as function of environmental conditions

• Versatile framework to evaluate critical parameters and improved modeling approaches

Objective and Approach

Peregrine: Validation and Benchmark Evaluation of Integrated Fuel Performance Modeling Using Test Reactor Data and Falcon

Assessment

• Enhanced mechanistic models for UO2 thermal conductivity, Zircaloy creep and growth, pellet cracking and fragment relocation.

• Use lower length scale modeling to inform fission gas release kinetics

• Expand to 3-D and local effects modeling

Path Forward

CASL Milestone: L1.CASL.P7.02 Protected under CASL Multi-Party NDA No. 793IP

5-8

Figure 5-11 shows the comparison between the Peregrine calculations and the experimental

measurements following the ramp test for AN2. Overall, the model predictions track the axial variation of the diameter change reasonably well, although an apparent under-prediction of

cladding diameter change at ~200 mm corresponding to peak power location.

Figure 5-12 shows the comparison between the Peregrine predictions and the experimental

measurements following the RISØ ramp test for AN8. Similar to the result of AN2, the Peregrine

calculation agrees with the axial variation of cladding diameter changes but clearly under-predicts the deformation of the lower portion of the rod.

Figure 5-11 Peregrine Calculation of Post-ramp Cladding Diameter Change for AN2 Test Rod

Compared to the Measurement

Figure 5-12 Peregrine Calculation of Post-ramp Cladding Diameter Change for AN8 Test Rod

Compared to the Measurement

, 0.12!

, 0.1!

, 0.08!

, 0.06!

, 0.04!

, 0.02!

0!

0.02!

0! 100! 200! 300! 400! 500! 600! 700!

Chan

ge-in

-Cla

ddin

g-D

iam

eter

-(m

m)-

Axial-Position-(mm)-

Experiment!

Peregrine!

, 0.12!

, 0.1!

, 0.08!

, 0.06!

, 0.04!

, 0.02!

0!0! 100! 200! 300! 400! 500! 600! 700!

Chan

ge-in

-Cla

ddin

g-D

iam

eter

-(m

)-

Axial-Position-(mm)-

Experiment!

Peregrine!

CASL Milestone: L1.CASL.P7.02 Protected under CASL Multi-Party NDA No. 793IP

6-18

Figure 6-18 Comparison of Fission Gas Release Fraction between the Measurement and

Peregrine Calculations

Overall, the preliminary benchmark results for the tests case covering a number of operating

conditions have shown that Peregrine code has the general capability of modeling fission gas release. However, the modeling of fission gas release has been a challenging part in nuclear

fuel performance codes; existing models in fuel performance codes generally do not provide a

versatile capability of predicting fission gas release which could result from different fuel

materials at various operating and accident conditions. It can be seen that the combination of several models is necessary to account for different mechanisms involved in the fission gas

release process. Expansion of the validation database to include different operating regimes is

necessary to further test the capabilities of the mechanistic models in the Peregrine code and to identify the missing mechanisms in current models to make improvements. For example, the

current benchmark cases do not have cases with appreciable amount of fission gas release in

steady state operations except for IFA 505.5 rod1. Such cases, however, could be of more

relevant to the operating conditions of commercial reactor fuels. Further development of more physics-based fission gas release models is also of interest to reduce the reliance on the

empirical knowledge in current models and to develop/improve the predictability of fission gas

release model.

Current model validation is largely based on the total amount of measured FGR at the end of

life, which could have missed some important aspects of the fission gas release process such as the incubation period of thermal release, contribution of fission gas release from different fuel

microstructures, fuel restructuring at high burnup, et. al. Detailed PIE data of relative fractions of

isotopes and distribution of gaseous products in the fuel matrix are available in some

experiments. They have not been used in the benchmark due to the limitation of current models and the processing capability of the Peregrine code. They provide valuable information

regarding the fission gas release behavior; and it would be of interest to use such information to

assist the development of advanced modeling capability.

600 800 1000 1200 1400 1600 1800Measured Temperature (K)

600

800

1000

1200

1400

1600

1800

Calc

ula

ted

Tem

per

ature

(K

)

Peregrine

Falcon+/- 50 Degrees

Calculated vs. Measured Temperature

550 temperature

measurements

0 2 4 6 8 10 12Burnup (MWd/kgUO

2)

-30

-20

-10

0

10

20

30

40

Fuel

-Cla

ddin

g G

ap W

idth

(m

icro

ns)

Peregrine

Falcon

Annular FuelColumn

Clad Tube

Upper Plenum

High temperature Region

r

z

Not to scale

Thermo-mechanical modeling shows excellent agreement with temperature measurements

• Results comparable to industry

codes

• Accommodates burnup effects (e.g.

material property and geometric

changes) on thermal performance

• JNFK solver able to support

complex power vs. time functions

CASL Milestone: L1.CASL.P7.02 Protected under CASL Multi-Party NDA No. 793IP

4-35

Figure 4-5 Comparison of Peregrine and Falcon Calculated Temperatures with Measured

Temperatures as a Function of Burnup for IFA-562.1 Rod 5 at the Lower Thermocouple

Figure 4-6 Comparison of Peregrine and Falcon Calculated Temperatures with Measured

Temperatures as a Function of Burnup for IFA-562.1 Rod 5 at the Upper Thermocouple

600 800 1000 1200 1400 1600 1800Measured Temperature (K)

600

800

1000

1200

1400

1600

1800

Cal

cula

ted

Tem

per

ature

(K

)

Peregrine

Falcon+/- 50 Degrees

Calculated vs. Measured Temperature

550 temperature

measurements

CASL Milestone: L1.CASL.P7.02 Protected under CASL Multi-Party NDA No. 793IP

4-34

Figure 4-3 Comparison of Peregrine and Falcon Calculated Temperatures with Measured

Temperatures as a Function of Burnup for IFA-505, Rod 1

Figure 4-4 Comparison of Peregrine and Falcon Calculated Temperatures with Measured Temperatures as a Function of Burnup for IFA-515, Rod A1

IFA-515 Rod A1 IFA-562.1 Rod 5

Lower T/C

FGR Modeling in Peregrine: Current state

CASL Milestone: L1.CASL.P7.02 Protected under CASL Multi-Party NDA No. 793IP

6-10

Figure 6-9 Fission Gas Release Fraction Calculated using Different Models in Peregrine for

RISØ AN8 Test Rod in Base Irradiation

Results of fission gas release using Falcon and Peregrine code calculations for AN2 and AN8

test rods at the end of base irradiation are shown in Table 6-3. Measured fission gas release fraction based on sibling rods is also shown in Table 6-3 for comparison.

Falcon code predicts a much lower fission gas release fraction of 0.01% compared to the

measured 0.2%, which could be primarily caused by the athermal fission gas release. Falcon calculation is also lower than the athermal fission gas release of 0.07% calculated using the

Sifgrs model. Both the Sifgrs model and the Forsberg-Massih tend to predict higher fission gas

release; the amount of the thermal release alone is higher than the measurement.

Table 6-3 Comparison of FGR between Falcon, Peregrine, and Measurements for RISØ AN2

and AN8 Test Rods at the End of Base Irradiation

Calculated FGR (%) Measured

FGR (%) Falcon Peregrine

Sifgrs Sifgrs

athermal

Forsberg-Massih

AN2 0.01 1.0 0.07 0.6 0.2

AN8 NA 0.6 0.07 0.3 0.2

FGR RISØ AN8 Test

Rod in Base Irradiation

CASL Milestone: L1.CASL.P7.02 Protected under CASL Multi-Party NDA No. 793IP

6-18

Figure 6-18 Comparison of Fission Gas Release Fraction between the Measurement and

Peregrine Calculations

Overall, the preliminary benchmark results for the tests case covering a number of operating

conditions have shown that Peregrine code has the general capability of modeling fission gas release. However, the modeling of fission gas release has been a challenging part in nuclear

fuel performance codes; existing models in fuel performance codes generally do not provide a

versatile capability of predicting fission gas release which could result from different fuel

materials at various operating and accident conditions. It can be seen that the combination of several models is necessary to account for different mechanisms involved in the fission gas

release process. Expansion of the validation database to include different operating regimes is

necessary to further test the capabilities of the mechanistic models in the Peregrine code and to identify the missing mechanisms in current models to make improvements. For example, the

current benchmark cases do not have cases with appreciable amount of fission gas release in

steady state operations except for IFA 505.5 rod1. Such cases, however, could be of more

relevant to the operating conditions of commercial reactor fuels. Further development of more physics-based fission gas release models is also of interest to reduce the reliance on the

empirical knowledge in current models and to develop/improve the predictability of fission gas

release model.

Current model validation is largely based on the total amount of measured FGR at the end of

life, which could have missed some important aspects of the fission gas release process such as the incubation period of thermal release, contribution of fission gas release from different fuel

microstructures, fuel restructuring at high burnup, et. al. Detailed PIE data of relative fractions of

isotopes and distribution of gaseous products in the fuel matrix are available in some

experiments. They have not been used in the benchmark due to the limitation of current models and the processing capability of the Peregrine code. They provide valuable information

regarding the fission gas release behavior; and it would be of interest to use such information to

assist the development of advanced modeling capability.

PK2-3 Test Rod during

Power Ramp

Peregrine has general capabilities to model fission gas release

Pastore model improvement for high temperature fission gas release during rapid power maneuvers

Existing FGR models have limited versatility to consider local physical processes active during operating and accident conditions -> high uncertainty

Requires improved understanding of the spatial distribution of fission gas (i.e. intragranular vs. intergranular)

MAMBA Benchmark/Validation

MAMBA: Multi-rod CRUD Modeling

“MAMBA” calculates cladding

surface heat flux, crud surface

temperature, and is coupled

DeCART (neutronics) / STAR-

CCM+ (thermal-hydraulics)

calculation

Typical crud loading in a PWR fuel

assembly ( NEI, 2012)

The simulation produced findings useful to PWRs :

• Significant azimuthal temperature variations on the

cladding surface (see right)

• Varying crud deposition and erosion rates resulting in

streak deposits (observed in operating PWRs, see right)

• Cladding “hot spots” were observed for thicker crud

This new coupled simulation capability is still under

development and hence has only been qualitatively

validated. Comparisons with out-of-pile data currently

underway

Flow

Transient Modeling in CASL – RIA Demo

Pellet Burnup - 75 GWd/tU Deposited Enthalpy ~120 cal/gm

Fast Fluence ~ 1.2x1023 n/cm2-s

Transient Modeling Activities in CASL

Coupling with pin-resolved neutronics to evaluate impact of non-uniform power deposition within the pellet

Requires localized 3-D modeling of temperature, mechanical loads, and

Coupling MAMBA and Peregrine to improve corrosion and CRUD layer distributions

Needed to evaluate the consequences of cracking and spalling of the corrosion layer thickness under rapid loading conditions

Coupling with subchannel thermal-hydrauliccs and computational fluid dynamics

Improve local heat transfer conditions between cladding to coolant

Developing more mechanistic material models for high temperature mechanical and chemical reaction behavior

High strain rate and large deformation response

Fracture/Failure mode response

Mechanistic fission gas diffusion, bubble growth, interlinkage, and release

Establish distribution of fission gas prior to a transient

Improve fission gas release kinetics for rapidly changing conditions

Summary/Conclusions - 1

CASL is developing state-of-the-art fuel behavior modeling capabilities using a multi-scale, multi-physics approach

Identify the conditions and important fundamental mechanisms that influence material performance or behavior response

Use microscale modeling to improve representations for key material properties

Applying this approach to PCI failure and CRUD formation

CASL is working on capabilities for advanced fuel behavior modeling for off-normal conditions, particularly RIA and LOCA

Mechanistic cladding deformation model

coupling Peregrine and MAMBA to provide a more accurate distribution of cladding corrosion and CRUD layer distribution

improved fission gas diffusion kinetics needed to define intragranular and intergranular fission gas bubble distributions

Summary/Conclusions - 2

Peregrine and MAMBA benchmarking and validation activities provide confidence in the development of the fuel performance modeling capabilities

Quantitative and qualitative evaluations with measured data and industry codes show excellent results

Comparisons with data highlight the challenges modeling complex thermal and mechanical behavior of irradiated fuel performance

Areas of improvement include, pellet cracking and relocation, fission gas retention and release, cladding creep and growth, cladding oxidation and hydride formation and growth

Application of advanced modeling methods to LWR transient conditions could address

the impact of non-uniform power deposition within the pellet

the consequences of cracking and spalling of the corrosion layer thickness under rapid loading conditions,

the important role of cladding to coolant heat transfer

high temperature mechanical and chemical reaction behavior

Modeling Transient Behavior: Loss-of-Coolant Accidents

Proceedings of the 2005 Water Reactor Fuel Performance Meeting

Kyoto, Japan, October 2-6, 2005

Paper 1076/Track 5

6. JAERI LOCA Tests

An experimental program related to the high

burnup fuel behavior during a LOCA has been

initiated at the Japan Atomic Energy Research

Institute (JAERI). The objectives of this program are

to evaluate the influence of high burnup effects on

fuel behavior under LOCA conditions and to provide

basic data to assess applicability of the current

criteria for cladding embitterment to a higher burnup

range. The program consists of Zircaloy cladding

oxidation tests, integral tests of rod-burst, oxidation

and rod thermal shock by quenching, and related

material properties tests. As a part of the program,

integral thermal shock tests simulating the whole

LOCA sequence were conducted with Zircaloy-4 fuel

rod cladding, irradiated to 39 and 44GWd/tU in a

PWR. The test rod was quenched with axial restraint

to represent a possible condition of fuel rod locked

between two spacer grid positions. In the present

study, the maximum load was limited to about 540 N

and was maintained by the automatic load adjusting

operation of the tensile testing machine. The 540 N is

based on the measurement of resistant load between

deformed or chemically interacted cladding and

spacer grid14. For these tests, the UO2 pellets were

removed and alumina dummy pellets were loaded in

the 190mm-long cladding, and the rod was

pressurized to about 5 MPa with Ar gas at room

temperature. To date, three tests have been conducted

on irradiated claddings with the initial oxide

thickness between 18 to 25 µm and hydrogen

concentrations is estimated to be 200 to 300 ppm.

7. FALCON analysis of JAERI LOCA Tests

A PWR fuel rod, irradiated to 44 GWd/tU at

Takahama Unit-3 reactor was selected for analysis

with FALCON to demonstrate the ability of the code

to calculate the cladding ballooning and burst

behavior as well as to calculate the thermal shock

load resulted from the axial restraints during the

quenching. An axisymmetric r-z finite element grid

was used for this analysis. Axial constraint boundary

conditions are applied at the lower endcap to

represent the lower mechanical connection to the

universal testing machine. The top endcap is free to

move in the axial direction during the heat up and

isothermal oxidation period. At the beginning of the

cooling phase, the FALCON calculation was

restarted with a fully restrained boundary condition at

the top end cap, which restricts the movement of the

top and bottom endcaps in the axial direction.

FALCON’s versatile and highly user-oriented restart

capability allows the user to retrieve the necessary

data from a previous analysis and to continue the

analysis with a new set of boundary conditions. As a

result, it is possible to model the fully restrained axial

boundary conditions at the beginning of the cooling

phase and to calculate the tensile loading resulting

from the restriction on the cladding shrinkage during

the quench process. However, at this time FALCON

cannot model the partially restrained boundary

condition of the JAERI experiment. Further

modification of the upper end plug displacement

constraint boundary conditions is necessary to model

the partial restrained conditions in order to limit the

maximum axial loading, as done in several of the

JAERI tests.

Figure 11 compares the FALCON calculation of

cladding temperatures and cladding axial loading

with the measurements. FALCON calculates the

cladding temperatures at the thermocouple locations.

Four thermocouples are spot-welded on the outer

surface of the cladding; thermocouple 2 is at the mid-

height position, TC1 and TC3 are 40 mm above and

below from the rod middle respectively, and TC4 is

above 20 mm from the rod mid section.

Time (s)

0 100 200 300 400 500 600

Cla

dd

ing

Te

mp

era

ture

(oC

)0

200

400

600

800

1000

1200

1400

Ax

ial

Lo

ad

(N

)

0

500

1000

1500

2000

2500

3000

Measured Temp

Calculated TC 3

Calculated TC 2

Calculated TC 4

Calculated TC 1

Measured Load

Calculated load

Figure 11. FALCON Comparison to JAERI Test-2

Temperature and Axial Load Measurements

The cladding temperature is nearly uniform in the 40-

mm region at rod center, and the axial temperature

difference is about 5oC in this region. The cladding

tube balloons and ruptures at measured temperatures

ranging between 780 to 830oC during the heat up.

The rod is isothermally oxidized after the rupture.

Isothermal oxidation temperature and time ranges

from 1160 to 1200oC and from 130 to 300 s. The rod

is cooled in a steam flow to about 700oC and is

finally quenched with water flooding from the

bottom. The results presented in Figure 11 are for a

test rod that was quenched under partially restrained

conditions, where the maximum loading was limited

to 540 N. The FALCON calculated axial load reaches

about 2500 N at the end of the quenching using full

axial restraint. Fully restrained experiments

performed in the JAERI LOCA test facilities

9 808

0.00E+00

1.00E-05

2.00E-05

3.00E-05

4.00E-05

5.00E-05

6.00E-05

7.00E-05

8.00E-05

0 40 80 120 160 200 240 280 320

Inn

er

alp

ha

la

ye

r th

ickn

es

s (

m)

Time (sec)

HBR Zry-4 Test ID 7

Bond layer = 15 micron

Bond layer = 7 micron

HBR Zry-4 Test ID 8

Alp

ha L

ayer R

an

ge

for

Initia

l Oxid

e T

hic

kness

Fig. 6 Prediction of inner s-phase layer formation of high

burnup fuel cladding in one-sided steam oxidation

221

(a)

(b)

Figure 157. Comparison of Zry-4 samples oxidized (one-sided) in steam at 1200ºC to 5% CP-ECR: (a)

as-fabricated HBR-type Zry-4 at 4 mm from sample midplane and (b) high-burnup HBR Zry-4 at 4 mm

from sample midplane.

Nagase, JNST Vol 42, no. 2, Feb 2005

NUREG/CR-6967

Jahingir, WRFPM 2005

Liu, WRFPM 2011

Fuel Behavior Mechanisms Important for RIA:

Fuel rod initial conditions RIA Phase 1

Near-Adiabatic Energy Deposition

RIA Phase 2

Energy Deposition with Heat Transfer

Residual pellet-cladding gap thickness/fuel-

clad bonding

Thermal expansion of cracked pellet structure

(crack volume closure rates)

Fission gas behavior in both grain boundary

and intragranular bubbles

Cracked pellet state/fragment distribution Gap mechanical contact and thermal

conductance

Transient fission gas release kinetics

Fission gas distribution in pellet: grain

boundaries and grain concentrations

Fuel fragmentation and dispersal kinetics after

cladding fracture

Axial gas transport kinetics within/adjacent to

fuel column (localized pressurization)

Fission gas release into the gap and plenum Fuel fragment coolant interaction and pressure

generation

Fuel fragmentation and dispersal kinetics

Fissile isotope/burnup distribution Effective fuel thermal conductivity

Cladding corrosion and CRUD layer build up Cladding fast strain rate behavior/plasticity Cladding thermal annealing/plastic behavior

recovery

Cladding hydrogen uptake/content, distribution

of hydrides, etc.

Fracture and spallation of oxide and crud layer Thermal creep and creep rupture

Cladding creep deformations Cladding crack formation and growth

(displacement controlled deformation)

High temperature oxidation (Oxygen uptake

and diffusion in cladding)

Cladding Irradiation damage (dislocation

densities, second phase particle

amorphization, etc.)

Thermal shock stress distributions in cladding

during quench/rewetting

Boron deposition/Coolant chemistry Material mechanical behavior as function of

phase (ZrO2, high O alpha, and prior beta)

Post-Departure from Nucleate (DNB) clad-to-

coolant heat transfer

Interfaces to couple Fuel Performance with Neutronics and Thermal-Hydraulics

Neutronics Thermal-

Hydraulics

Coolant Gamma Heating

Moderator Temperature

and Density

Hea

t Tra

nsfe

r C

oeffi

cien

ts

Coo

lant

Tem

pera

ture

s

Total Pow

er Density (W

/m3)

/Fast N

eutron flux (>1 M

ev)

Fuel a

nd clad

tem

perature

distribution

(radia

l ave

rage)

Peregrine

Cla

ddin

g Su

rface

Tem

pera

ture

Cla

ddin

g Su

rface

Hea

t Flu

x