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EL EC TR IF Y T HE WO R L D = p el

BW RV I P BWR Vessel & Intemals Project 2005-126

March 11, 2005

Document Control DeskU. S. Nuclear Regulatory Commission11555 Rockville PikeRockville, MD 20852

Attention: Meena Khanna

Subject: Project No. 704 - BWRVIP-1 15NP: BWR Vessel and Internals Project, RAMAFluence Methodology Benchmark Manual - Evaluation of Regulatory Guide 1.190Benchmark Problems

Reference: Letter from Carl Terry (BWRVIP) to Document Control Desk (NRC), "Project 704- BWRVIP-1 15: BWR Vessel and Internals Project, RAMA Fluence MethodologyBenchmark Manual - Evaluation of Regulatory Guide 1.190 BenchmarkProblems," dated June 26, 2003.

Enclosed are two (2) copies of the report "BVWRVIP-1 15NP: BWR Vessel and Internals Project,RAMA Fluence Methodology Benchmark Manual - Evaluation of Regulatory Guide 1.190Benchmark Problems," EPRI Technical Report 1008063NP, March 2005. This is a non-proprietary version of the proprietary document submitted to the NRC by the letter referencedabove.

If you have any questions on this subject please call Robin Dyle (Southern Nuclear, BWRVIPIntegration Committee Technical Chairman) at 205.992.5882.

Sincerely,

William A. EatonEntergy Operations, Inc.Chairman, BWR Vessel and Internals Project

CORPORATE REAOOUARTERS

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BWRVIP-115NP: BWR Vessel andInternals ProjectRAMA Fluence Methodology Benchmark Manual-Evaluation of Regulatory Guide 1.190 BenchmarkProblems

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This report describes research sponsored by EPRI.

The report is a corporate document that should be cited in the literature in the following manner:

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iii

PRODUCT DESCRIPTION

This report describes the numerical and experimental benchmark analyses that were performed toqualify the RAMA Fluence Methodology for use in the evaluation of neutron fluence in BWRs.

Results & FindingsThe RAMA Fluence Methodology contains the following software components: the transportcode, parts model builder (PMB) code, state-point model builder (SMB) code, fluence calculator,and the nuclear data library. The methodology includes an advanced three-dimensional nuclearparticle transport theory code that performs neutron and gamma flux calculations. It couples athree-dimensional, multi-group deterministic nuclear transport theory method with acombinatorial geometry modeling capability to provide a flexible and accurate tool fordetermining fluxes for any light water reactor design. The code supports the method ofcharacteristics transport theory solution technique, a three-dimensional ray-tracing method,combinatorial geometry, a fixed-source iterative solution, anisotropic scattering, thermal-groupupscattering treatments, and a nuclear cross-section data library based upon the ENDF1B-VI datafile. The software is written in conformance to the Fortran 95 programming language standardfor ease of portability between computing platforms. The methodology adheres to therequirements set forth in NRC Regulatory Guide 1.190 for pressure-vessel neutron fluencedeterminations.

Challenges & ObjectivesThis project had the following objectives:

* Develop a state-of-the-art method for calculating fluence in a BWR

* Adhere to the requirements of NRC Regulatory Guide 1.190

* Validate the methodology against specific benchmark problems identified in the regulatoryguide and perform plant-specific analyses

* Develop a system of software codes for application by utilities

Applications, Values & UseThe RAMA Fluence Methodology software package is used to determine neutron fluence inBWR components in compliance with the requirements and guidelines provided in U.S. NuclearRegulatory Commission Regulatory Guide 1.190. RAMA, Version 1.0 is designed to calculatethe fluence for surveillance capsules, the reactor pressure vessel within the active fuel height, andthe core shroud within the active fuel height. Future versions of RAMA will be developed toextend the methodology to other internal components that are beyond the active fuel height.

v

EPRI PerspectiveAccurate neutron fluence determinations for BWRs are required for a number of reasons:

* To determine neutron fluence within the reactor pressure vessel (RPV) and at surveillancecapsule locations to address vessel embrittlement issues

* To determine neutron fluence on the core shroud in order to determine fracture toughness andcrack growth rate for use in flaw evaluation calculations

* To determine neutron fluence at other internal components for structural integrityassessments or to evaluate repair technologies

The RAMA Fluence Methodology is a state-of-the-art and versatile tool for calculating thefluence of the BWR pressure vessel and internals.

ApproachThe BWRVIP conducted an extensive review and evaluation of existing technologies employedto determine the fluence of light water reactors. The three-dimensional nuclear particle transporttheory, combinatorial geometry methods, and additional advanced features of the RAMAmethodology provide capabilities not available in other existing technologies to accuratelycalculate the fluence of complex BWR internal components. Thus, RAMA was selected as themethodology to address the needs of the BWRVIP. A key aspect of this work was to ensure thatthe RAMA methodology adheres to the requirements set forth in NRC Regulatory Guide 1.190for pressure vessel neutron fluence determinations. To accomplish this, the methodology wasverified and validated against specific benchmark problems identified in the regulatory guide andplant-specific analyses were performed.

KeywordsFluenceEmbrittlementBoiling water reactorVessel and internalsReactor pressure vessel

vi

ABSTRACT

This document discusses the efforts and the results that have contributed to the preliminaryverification and validation of the RAMA Fluence Methodology software. The RAMA FluenceMethodology contains the following software components: the transport code, parts modelbuilder (PMB) code, state-point model builder (SMB) code, fluence calculator and the nucleardata library.

The RAMA Fluence Methodology is used to determine neutron fluence in BWR Priority Icomponents in compliance with the requirements and guidelines provided in U.S. NuclearRegulatory Commission Regulatory Guide 1.190. The BWR Priority I components includesurveillance capsules, the reactor pressure vessel inside surfaces over the active fuel height, andthe core shroud over the active fuel height.

The RAMA Fluence Methodology includes an advanced three-dimensional nuclear particletransport theory code that performs neutron and gamma flux calculations. RAMA couples athree-dimensional, multi-group deterministic nuclear transport theory method with an arbitrarygeometry modeling capability to provide a flexible and accurate tool for determining fluxes forany light water reactor design. The code supports the method of characteristics integral transporttheory solution technique, a three-dimensional ray-tracing method, combinatorial geometry, afixed source iterative solution with anisotropic scattering, thermal-group upscattering treatments,and a nuclear cross-section data library based upon the ENDF/B-VI data file. The software iswritten in conformance to the Fortran 95 programming language standard for ease of portabilitybetween computing platforms.

vii

ACKNOWLEDGMENTS

The undersigned wish to acknowledge EPRI and the Boiling Water Reactor Vessel and InternalsProject (BWRVIP) for their support of the RAMA Fluence Methodology project. Special thanksare given to Robert Carter and Ken Wolfe of EPRI for their guidance, comments, and overallsupport in completing this document.

The undersigned also wish to acknowledge Dr. Mark Williams of Louisiana State University forproviding a thorough technical review of the information provided in this report. Steven Bakerand Charlotte Potze of TransWare Enterprises Inc. are acknowledged for their months of effortinvolved in the performance of the benchmarks documented in this report. Virginia Jones ofTransWare is acknowledged for the many hours that she contributed to making this a qualitydocument.

Dean B. Jones, TransWare Enterprises Inc.Kenneth E. Watkins, TransWare Enterprises Inc.

ix

CONTENTS

1 INTRODUCTION ........................................... 1-1

2 NUMERICAL AND EXPERIMENTAL BENCHMARKS ........................................... 2-1

2.1 Pool Critical Assembly Experiment Benchmark ..................................... 2-1

2.1.1 Summary of Results ............................................ 2-2

2.1.2 Purpose ........................................... . 2-2

2.1.3 Objectives ........................................... . 2-2

2.1.4 Problem Description ............................................ 2-3

2.1.4.1 Reactor System Geometry ........................................... 2-3

2.1.4.2 Reactor System Material Composition ........................................... 2-3

2.1.5 Calculations ........................................... . 2-6

2.1.5.1 Modeled Geometry Coordinate System ........................................... 2-6

2.1.5.2 RAMA Model ........................................... 2-6

2.1.5.3 RAMA Calculation Parameters ........................................... 2-8

2.1.6 Results ........................................... . 2-8

2.1.7 Sensitivity Analysis ............................................ 2-10

2.2 VENUS 3 Benchmark ........................................... 2-13

2.2.1 Summary of Results ............................................ 2-13

2.2.2 Purpose ........................................... . 2-13

2.2.3 Objectives ........................................... 2-14

2.2.4 Problem Description ............................................ 2-14

2.2.4.1 Reactor System Geometry ........................................... 2-14

2.2.4.2 Reactor System Material Composition ........................................... 2-17

2.2.5 Calculations ........................................... 2-18

2.2.5.1 Modeled Geometry Coordinate System ........................................... 2-18

2.2.5.2 RAMA Model ........................................... 2-19

2.2.5.3 RAMA Calculation Parameters ........................................... 2-19

2.2.6 Results ... 2-20

xi

2.2.7 Sensitivity Analyses ............................................ 2-22

2.3 H. B. Robinson 2 Pressure Vessel Benchmark . . .......................................... 2-24

2.3.1 Summary of Results ............................................. 2-25

2.3.2 Purpose ............................................ . 2-25

2.3.3 Objectives ............................................ 2-25

2.3.4 Problem Description ............................................. 2-25

2.3.4.1 Reactor System Geometry ............................................ 2-26

2.3.4.2 Reactor System Material Composition ............................................ 2-26

2.3.5 Calculations ............................................ 2-27

2.3.5.1 Modeled Geometry Coordinate System ............................................ 2-27

2.3.5.2 RAMA Model ............................................ 2-28

2.3.5.3 RAMA Calculation Parameters ............................................ 2-28

2.3.6 Results ............................................ . 2-28

2.3.7 Sensitivity Analyses ............................................. 2-31

2.4 BWR Numerical Benchmark ............................................ 2-35

2.4.1 Summary of Results ............................................. 2-35

2.4.2 Purpose ............................................ . 2-36

2.4.3 Objectives ............................................ 2-36

2.4.4 Problem Description ............................................. 2-36

2.4.4.1 Reactor System Geometry ............................................ 2-36

2.4.4.2 Reactor System Material Composition ............................................ 2-38

2.4.5 Calculations ............................................ 2-38

2.4.5.1 Modeled Geometry Coordinate System ............................................ 2-39

2.4.5.2 RAMA Model ............................................ 2-40

2.4.5.3 RAMA Calculation Parameters ............................................ 2-40

2.4.6 Results ............................................ . 2-41

2.4.7 Sensitivity Analyses ............................................. 2-45

3REFERENCES ............................................ 3-1

xii

LIST OF FIGURES

Figure 2-1 Planar View of the PCA Benchmark Facility ............................................................. 2-4Figure 2-2 Elevation View of the PCA Benchmark Facility ........................................................ 2-5Figure 2-3 Coordinate System for PCA Benchmark RAMA Model ............................................ 2-7Figure 2-4 Thermal Shield Mesh Sensitivity in Direction Perpendicular to Core Face ............. 2-11Figure 2-5 Pressure Vessel Mesh Sensitivity in Direction Perpendicular to Core Face ........... 2-11Figure 2-6 Water Gap Between Aluminum Window and Thermal Shield Mesh Sensitivity

in Direction Perpendicular to Core Face ................................................................. 2-12Figure 2-7 Water Gap Between Thermal Shield and Pressure Vessel Mesh Sensitivity in

Direction Perpendicular to Core Face ................................................................. 2-12Figure 2-8 Planar View of the VENUS-3 Benchmark Facility .................................................. 2-15Figure 2-9 Elevation View of the VENUS-3 Benchmark Facility .............................................. 2-16Figure 2-10 Coordinate System for VENUS-3 Benchmark RAMA Model ................................ 2-18Figure 2-11 Variation of Axial Distance Between Parallel Rays .............................................. 2-24Figure 2-12 Coordinate System for HBR-2 Benchmark RAMA Model ..................................... 2-27Figure 2-13 Variation of Planar Distance Between Parallel Rays ............................................ 2-33Figure 2-14 Variation of Axial Distance Between Parallel Rays .............................................. 2-33Figure 2-15 Variation of Convergence Criteria ................................................................. 2-34Figure 2-16 Variation of Angular Quadrature Order ................................................................. 2-34Figure 2-17 Elevation View of the BWR Numerical Benchmark Reactor ................................. 2-37Figure 2-18 Coordinate System for BWR Numerical Benchmark RAMA Model ...................... 2-39Figure 2-19 Variation of Planar Distance Between Parallel Rays ............................................ 2-47Figure 2-20 Variation of Convergence Criterion ................................................................. 2-47Figure 2-21 Variation of Angular Quadrature Order ................................................................. 2-48

xMUi

LIST OF TABLES

Table 2-1 Dosimeter Reaction, Energy Thresholds, and Fission-Spectrum WeightedCross Sections ................................................................... 2-2

Table 2-2 Material Compositions for Regions in PCA Benchmark Problem .............................. 2-5Table 2-3 Equivalent Fission Flux Results (C/M) ................................................................... 2-9

Table 2-4 Sensitivity Analyses ................................................................... 2-10

Table 2-5 Dosimeter Reaction, Energy Thresholds, and Fission-Spectrum WeightedCross Sections ................................................................... 2-13

Table 2-6 Material Compositions for Regions in VENUS-3 Benchmark Problem .................... 2-17

Table 2-7 Equivalent Fission Flux Results (C/M) by Region .................................................... 2-20

Table 2-8 Equivalent Fission Flux Results (C/M) by Detector Location ................................... 2-21Table 2-9 Equivalent Fission Flux Results (C/M) by Elevation ................................................ 2-22

Table 2-10 Sensitivity Analyses ................................................................... 2-23

Table 2-11 Material Compositions for Regions in HBR-2 Benchmark Problem ....................... 2-26Table 2-12 Specific Activities for Surveillance Capsule Dosimeters (in Bq/mg) Results

(C/M) Calculated Using the 8 State-point Operating Data for Cycle 9 ............................. 2-29

Table 2-13 Specific Activities for Cavity Dosimeters (in Bq/mg) Results (C/M) CalculatedUsing the 8 State-point Operating Data for Cycle 9 ......................................................... 2-30

Table 2-14 Specific Activities for Surveillance Capsule Dosimeters (in Bq/mg) Results(C/M) Calculated Using the Cycle 9 Average Operating Data Set ................................... 2-30

Table 2-15 Specific Activities for Cavity Dosimeters (in Bq/mg) Results (C/M) CalculatedUsing the Cycle 9 Average Operating Data Set ............................................................... 2-31

Table 2-16 Sensitivity Analyses ................................................................... 2-32Table 2-17 Material Compositions for Regions in BWR Numerical Benchmark Problem ........ 2-38

Table 2-18 Reaction Rate (in rps/atom) Comparison Results (RAMA/NUREG) forCapsule at 3 Degrees ................................................................... 2-41

Table 2-19 RPV Neutron Flux and DPA Comparison Ratio (RAMA/NUREG) at Elevationof Peak Flux ................................................................... 2-42

Table 2-20 RPV Neutron Flux and DPA Comparison Ratio (RAMA/NUREG) at Core Mid-plane ................................................................... 2-43

Table 2-21 Azimuthal Variation of RPV Neutron Flux Comparison Ratio (RAMA/NUREG)for Elevation of Peak Flux at OT for Energy >1.0 MeV ..................................................... 2-44

Table 2-22 Downcomer Neutron Flux and DPA Comparison Ratio (RAMA/NUREG) ............. 2-45

Table 2-23 Cavity Neutron Flux and DPA Comparison Ratio (RAMA/NUREG) atElevation of Peak Flux ................................................................... 2-45

xv

Table 2-24 Sensitivity Analyses ....................... 2-46

xvi

1INTRODUCTION

The BWR Vessel and Internals Project (BWRVIP) has developed the RAMA FluenceMethodology (hereinafter referred to as the Methodology) for use in calculating the neutronfluence in Boiling Water Reactors. The current version of the Methodology is applicable forcalculations at the surveillance capsule location as well as on the core shroud and within thereactor vessel over the active fuel height.

The Methodology includes a transport code, a parts builder code for use in generating geometricmodels, a state-point builder code for processing operating data, a code module for calculatingactivations, fluences and uncertainties, and a nuclear data library. The Methodology uses adeterministic, three-dimensional, multigroup nuclear particle transport theory code that performsneutron and gamma flux calculations. RAMA couples the nuclear transport method with ageneral geometry modeling capability to provide a flexible and accurate tool for determiningfluxes for any light water reactor design. The code supports the method of characteristicssolution technique, a three-dimensional ray-tracing method based on combinatorial geometry, afixed source iterative solution with anisotropic scattering, thermal-group upscattering treatments,and a nuclear cross-section data library based upon the ENDF/B-VI data file. The Methodologyand procedures for its use are fully described in the following separate reports: A TheoryManual [I], a User's Manual [2], and a Procedures Manual [3].

The Methodology was benchmarked against the requirements of U.S. NRC Regulatory Guide1.190 [4]. The results of the numerical and experimental benchmark cases prescribed by theGuide are described in Section 2 of this report.

1-1

2NUMERICAL AND EXPERIMENTAL BENCHMARKS

This section describes the numerical and experimental benchmark analyses that were performedto qualify the RAMA Fluence Methodology for use in the evaluation of neutron fluence inBWRs. The qualification benchmark requirements are prescribed in NRC Regulatory Guide1.190 [4]. Version 1.00 of the RAMA Fluence Methodology software was used in all analyses.

2.1 Pool Critical Assembly Experiment Benchmark

The Pool Criticality Assembly (PCA) Pressure Vessel Facility Benchmark is prescribed by theU. S. NRC for use in benchmarking pressure vessel neutron fluence prediction methodologies. Adescription of the dimensions, material compositions, and neutron source data required toperform the PCA benchmark are provided in NUREG CR-6454 [6]. Also provided are measureddosimeter reaction rates, expressed as equivalent fission neutron flux, at seven detector locations.The equivalent fission neutron flux utilizes an equivalent 2.. 5U fission spectrum dosimeter crosssection defined as:

#energxrvups (2-1)

g=l

where aR is the energy group dependent activation response cross section and X is the energy

group dependent 235U fission production spectrum. The equivalent 25U fission spectrum flux isthen defined as;

#eermeroups

E0 aRg ¢g

Oe= g=l (2-2)

v ¢eq

Measured reaction rates are reported for six dosimeter reactions: 23'Np(n,fission), 2"U(n,fission),"51n(nn')' Oin In, -8Ni(n p)58Co 2'Al(n,) 24Na, and '0 3Rh(n,n')'03wRh. The detectors are distributedthroughout the PCA geometry to provide spatial and spectral variations. Table 2-1 lists theenergy thresholds and fission-spectrum weighted cross sections for each dosimeter reactionexcept the 'IORh(n,n')'OmRh. See discussion on '0Rh(n,n')'0'nRh in Section 2.1.1 for anexplanation on its exclusion from the table.

2-1

- -

Numerical and Experimental Benchmarks

The PCA benchmark problem has been analyzed using the RAMA Fluence Methodology. Predicteddosimeter reaction rates at the various detector locations are compared to the measured reaction rates.Details of the model and comparison results are provided in the following subsections.

Table 2-1Dosimeter Reaction, Energy Thresholds, and Fission-Spectrum Weighted Cross Sections

Dosimeter Reaction Energy Threshold (in MeV) Fission-Spectrum WeightedCross Section (in barns)*

237Np(njfission) 0.7 1.3305E-00

238U(nfission) 1.5 3.0725E-01

115 1n(nn')" 5"in 0.6 1.8024E-01

'Ni(n,p)'Co 2.1 1.0701 E-01

2'A1(n, a)24Na 6.5 7.8206E-04

* The fission spectrum cross sections are flat weighted.

2.1.1 Summary of Results

The RAMA calculated results for the PCA benchmark are in excellent agreement with themeasurements.

Content Deleted -EPRI Proprietary Information

2.1.2 Purpose

The purpose of the PCA experiment is to simulate a PWR pressure vessel and downcomerconfiguration for benchmarking computational tools used in neutron fluence determination. ThePCA benchmark provides measured reaction rates inside a simulated pressure vessel, in the watergap in front of the pressure vessel, and in the thermal shield.

2.1.3 Objectives

The objectives of this study are to benchmark RAMA against the measurements performed at thePCA facility at the Oak Ridge National Laboratory and to assess the accuracy with which themethodology predicts the neutron flux distribution inside the pressure vessel.

2-2


2.1.4 Problem Description

The PCA Benchmark Facility consists of the PCA reactor and the ex-core components that areused to simulate pressure vessel surveillance configurations in light water reactors. The ex-corecomponents include the thermal shield (TS), a simulated pressure vessel (PV), and a void box(VB) that simulates the reactor cavity. The ex-core components are all rectangular prisms inshape.Various system configurations are supported by the PCA facility. The configuration that isapplicable to the PCA benchmark described in [6] is designated as the "12/13 configuration".

2.1.4.1 Reactor System Geometry

Figure 2-1 provides a plan view of the PCA benchmark facility in the 12/13 configuration.Figure 2-2 provides an elevation view of the PCA benchmark facility.

The PCA core consists of 25 material test reactor (MTR) curved-plate fuel elements arranged ina five by five array with a pitch of 7.71 cm by 8.10 cm. Three types of fuel elements are loadedin the core: standard (18 plate) fuel elements, Oak Ridge Research Reactor (ORR) (19 plate) fuelelements, and control rod fuel elements. The ORR fuel elements reside in locations C4 and C6(see Figure 2-1). The control rod fuel elements reside in locations B4, B6, D4 and D6. Thecontrol rod fuel element geometry consists of three distinct regions. The front and back regionscontain fuel, water, and aluminum. The middle region contains water and aluminum (with thecontrol rod withdrawn during operation).

2.1.4.2 Reactor System Material Composition

Table 2-2 provides the material composition for each region of the PCA benchmark facility. Thefuel plates in the fuel assemblies consist of highly-enriched uranium-aluminum alloy encased inan aluminum sheath. The void box consists of a void region (air) encased in an aluminum sheath.The thermal shield is stainless steel (SS304L) and the simulated pressure vessel is carbon steel(SA-36). All aluminum in the facility is type 6061-T6. The experiments are performed at anominal water temperature of 37.70 C. The facility operates at a nominal water temperature of37.70 C under normal atmospheric pressure.

2-3

- u


PAIM

COZE ^C

PYES5$5LWE v ey 1 *II

4.4.

40.1

DeI )9.

\'~ g zv

__ _ _ __I~.

Figure 2-1Planar View of the PCA Benchmark Facility

(dimensions are in cm)

2-4

Numerical and Experimental Benchnmarkls

~- RA-P

FoCLSeCT101"

(dimensions are in cm)

Figure 2-2Elevation View of the PCA Benchmark Facility

Table 2-2Material Compositions for Regions In PCA Benchmark Problem

Region Material Composition

Core 'U, 2'U. Al, Water (at 37.7C)

Aluminum Window Al

Thermal Shield Stainless Steel (SS304L)

Reactor Pressure Vessel Simulator SA-36 Carbon Steel

Void Box Al, Air

Water Regions Water at 37.7C

2-5


2.1.5 Calculations

The PCA benchmark is geometrically represented in RAMA by a three-dimensional model of thePCA facility. A two-dimensional model with planar meshing equivalent to the core mid-planeaxial node of the three-dimensional model is also used for selected sensitivity analyses.

2.1.5.1 Modeled Geometry Coordinate System

For reference purposes, a coordinate system is imposed upon the model. Figure 2-3 illustrates thecoordinate system in relation to the primary components of the PCA facility. The X axis isparallel to the edge of the core nearest the aluminum window. The Y axis is perpendicular to thecore edge and extends in a positive direction outward toward the ex-core components. Thepositive Z axis extends vertically upward. The origin of the coordinate system is at the center ofthe core edge facing the aluminum window.

2.1.5.2 RAMA Model

Content Deleted -

EPRI Proprietary Information

2-6


Plan View

Elevation View

PCA Core z

L-

E

C)

Vla C

3,

G F E D C

iE4jm

B A

C..In

c-

M

NOT TO SCALEFigure 2-3Coordinate System for PCA Benchmark RAMA Model

2-7


Content Deleted -


2.1.5.3 RAMA Calculation Parameters


2.1.6 Results

Content Deleted -


2-8


Table 2-3Equivalent Fission Flux Results (C/M)

Content Deleted -


2-9

-U


2.1.7 Sensitivity Analysis

Several sensitivity analyses were performed to evaluate the stability and accuracy of RAMA forthe PCA benchmark reference case with respect to mesh size and solution parameters. Asummary of these analyses is presented in Table 2-4.

Content Deleted -


Table 2-4Sensitivity Analyses

Content Deleted -


2-10

Numerical and Experimental Benchimarks


Figure 2-4Thermal Shield Mesh Sensitivity in Direction Perpendicular to Core Face

Content, Deleted -EPRI Proprietary Information

2-11

-



Figure 2-6Water Gap Between Aluminum Window and Thermal Shield Mesh Sensitivity in DirectionPerpendicular to Core Face

Content Deleted -


Figure 2-7Water Gap Between Thermal Shield and Pressure Vessel Mesh Sensitivity in DirectionPerpendicular to Core Face

2-12


2.2 VENUS 3 Benchmark

The VENUS-3 Benchmark is prescribed by the U. S. NRC for use in benchmarking pressurevessel neutron fluence prediction methodologies. A description of the dimensions and materialcompositions required to perform the VENUS-3 benchmark is provided in [7]. Also provided aremeasured dosimeter reaction rates, expressed as equivalent fission neutron flux, at thirtylocations and fourteen elevations. (See Section 2.1 of this manual for a definition of equivalentfission neutron flux.) Measured reaction rates are reported for three dosimeter reactions:58Ni(n,p)"Co, "31n(n,n')" 5'n1n, and "Al(n, c)24Na. The detectors are distributed throughout theVENUS-3 geometry to provide spatial and spectral variations. Table 2-5 lists the energythresholds and fission-spectrum weighted cross sections for each dosimeter reaction.

Table 2-5Dosimeter Reaction, Energy Thresholds, and Fission-Spectrum Weighted Cross Sections

Dosimeter Reaction Energy Threshold (in MeV) Fission-Spectrum WeightedCross Section (in barns)

All But Barrel Barrel

5Ni(np) 8Co 2.1 1.0701E-01 1.0617E-01

25An(n,n')"'In 0.6 1.8024E-01 1.7965E-01

2 Al(n, a)2Na 6.5 -7.8206E-04 7.3318E-04

Predicted dosimeter reaction rates using RAMA at the various detector locations are compared tothe measured reaction rates. Details of the model and comparison results are provided in thefollowing subsections.


The RAMA calculated results for the VENUS-3 benchmark are in good agreement with themeasurements.


2.2.2 Purpose

The purpose of the VENUS-3 experiment is to simulate the radial core shape and neutronspectrum of a typical pressurized water reactor (PWR). The experimental results provide a basisfor benchmarking computational tools used in neutron fluence determination. The VENUS-3experiment provides measured reaction rates for fourteen elevations distributed inside the core,in the inner and outer baffles, in the reflector region between the outer baffle and barrel, and inthe barrel.

2-13

-a


2.2.3 Objectives

The objectives of this study are to benchmark RAMA against the measurements performed at theVENUS Critical Facility at SCK/GEN, Mol (Belgium) and to assess the accuracy with which themethodology predicts the neutron flux distribution inside the core and in ex-core regions.


The VENUS-3 Benchmark Experiment consists of the VENUS reactor and the ex-corecomponents that are used to simulate pressure vessel surveillance configurations in light waterreactors. The ex-core components include a barrel, a neutron pad, a jacket that simulates thereactor cavity, and a simulated reactor pressure vessel (RPV).


Figure 2-8 provides a plan view of the VENUS-3 Benchmark Experiment showing the primarycomponents. Figure 2-9 provides an elevation view of the VENUS-3 Benchmark Experiment.

The VENUS-3 core is composed of 12 simulated PWR fuel assemblies. Each fuel assembly isrepresentative of a 15x15 PWR fuel assembly. There are a total of 2,548 fuel pins, 52 non-fuelpins, and a I Ox 10 pin area at the center of the core that is occupied by a core water hole and theinner baffle. The pin pitch is 1.26 cm. Four types of pins are loaded in the core: fuel 3/0 (3.3 wt.% 215U), fuel 4/0 (4.0 wt. % 235U), pyrex pins, and partial length shielded assembly (PLSA) pins.The pyrex pins simulate PWR control rod clusters. The PLSA pins consist of fuel 3/0 fuel pinsabove core mid-plane and stainless steel rods below core mid-plane that simulate a PWRpartially shielded peripheral assembly.

Between the core hole and the core is an inner stainless steel baffle. Outside the core is an outerstainless steel baffle that is typical of a PWR core baffle. A water reflector region occupies thespace between the outer baffle and the barrel. Beyond the barrel is a neutron pad. The remainingreactor components are the air-filled jacket and the pressure vessel.

The active core height is 50 cm. Below the active core height are a lower reflector region, abottom grid, the bottom support region, and a lower filling. Above the active core are anintermediate grid, an upper reflector, an upper grid, and an upper filling.

2-14


-4

ll

Planar Vie of th VEU- Benchmar !k Fclit

2-15


P?.1.sIi.s.. i

III

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I

Trhidbl. for octlv-tfon or IL.sion chmber JII Y.ter 1v-)

-- IjPClld (55I

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L_ 168 65

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-4' IMEtXUI

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I

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130 c.

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._11 --11

.VUH ' _ U-; L u 1t A.I -A L

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1 f-- II I

pwactor vess.j dlaeter 120 ct n l'r- I

Figure 2-9Elevation View of the VENUS-3 Benchmark Facility

2-16



Table 2-6 provides the material composition for each region of the VENUS-3 benchmarkfacility.

Table 2-6Material Compositions for Regions in VENUS-3 Benchmark Problem


Water Regions Water at 24.00C

Fuel 3/0 Pin 235U, 238U, Zircaloy

Fuel 4/0 Pin 235U, 238U, Stainless Steel

Pyrex Pin Pyrex, Stainless Steel

PLSA Pin 23U, 3U, Zircaloy, Stainless Steel

Inner and Outer Baffles Stainless Steel

Barrel Stainless Steel

Neutron Pad - Stainless Steel

Jacket Inner and Outer Walls Stainless Steel

Jacket Air

Vessel Stainless Steel

Lower Filling Water

Bottom Support Stainless Steel, Water

Bottom Grid Stainless Steel, Water

Lower Reflector Stainless Steel, Plexiglass, Water

Intermediate Grid Water, Plexiglass

Upper Reflector Stainless Steel, Plexiglass, Water

Upper Grid Stainless Steel, Water

Upper Filling Water

2-17


2.2.5 Calculations

The VENUS-3 benchmark is geometrically represented in RAMA by a three-dimensional model.An additional two-dimensional model with planar meshing equivalent to the core mid-plane axialnode of the three-dimensional model is used for selected sensitivity analyses.


For reference purposes, a coordinate system is imposed upon the model. Figure 2-10 illustratesthe coordinate system in relation to the primary components of the VENUS facility. The originof measurements, the neutron pad and the barrel are located in the northeast quadrant of themodel (i.e., the quadrant bounded by the positive X and Y axes). The positive Z axis extendsvertically upward.

yWater

Neutron Pad

Pressure Vessel

Inner BaffleXCore Hole Outer Baffle Barrel Jacket

Jacket Inner Wall Jacket Outer Wall

Figure 2-1 0Coordinate System for VENUS-3 Benchmark RAMA Model

2-18


2.2.5.2 RAMA Model

Content Deleted -


RAMA Calculation Parameters


2.2.5.3

2-19

haj


2.2.6 Results

Content Deleted -


Table 2-7Equivalent Fission Flux Results (C/M) by Region

Content Deleted -


2-20


Table 2-8Equivalent Fission Flux Results (C/M) by Detector Location

Content Deleted -


2-21

-


Table 2-9Equivalent Fission Flux Results (C/M) by Elevation


2.2.7 Sensitivity Analyses

Several sensitivity analyses were performed to evaluate the stability and accuracy of RAMA forthe VENUS-3 benchmark reference case with respect to mesh size and solution parameters. Asummary of these analyses is presented in Table 2-10.

Content Deleted -


2-22


Table 2-1 0Sensitivity Analyses


2-23


1.002

0 0.9980 4Wc uC

L U 0.996(hq,

LGUDCr

o 2 0.994U 'I

D 0.992a02

aa >t

.2 X 0.99I -

fI 0.988

0.986

0.9840 0.5 1 1.5 2 2.5 3 3.5

Axial Distance between Parallel Rays

Figure 2-11Variation of Axial Distance Between Parallel Rays

2.3 H. B. Robinson 2 Pressure Vessel Benchmark

The H. B. Robinson Unit 2 (HBR-2) Pressure Vessel Benchmark is prescribed by the U. S. NRCfor use in benchmarking pressure vessel neutron fluence prediction methodologies. HBR-2 is aWestinghouse pressurized water reactor that has been in operation since 1971. A description ofthe dimensions and material compositions required to perform the HBR-2 benchmark is providedin [9]. Also provided are measured reaction rates for surveillance capsule and cavity dosimeters.Measured reaction rates are reported for six dosimeter reactions: 2" 7Np(n,f)'l 7Cs, 238U(n,f)137Cs,

5 )8 O )4 46T n P46S 60ad13 U-58Ni(np)58Co, 5'Fe(np) Mn, Ti(n,p) Sc, and 63Cu(n, x)6"Co. The RAMA model for the HBR-2plant assumes octant symmetry and is described over azimuths 00 to 45°. The surveillancecapsule is at the azimuthal angle of 200 and the cavity dosimeters are located at the azimuthalangle of 0° in the model. The dosimeters were irradiated during operating cycle 9 only.

Predicted dosimeter reaction rates using RAMA at the various locations are compared to themeasured reaction rates. Details of the model and comparison results are provided in thefollowing subsections.

2-24



The RAMA calculated specific activities for the HBR-2 benchmark are in good agreement withthe measured values that were corrected for photofission effects and "Co impurities in thecopper. Calculations were performed using the eight state-point operating data provided for cycle9 as well as the cycle 9 average operating data set.


2.3.2 Purpose

The purpose of the HBR-2 benchmark experiment is to compare to measurements on the insideand outside of the reactor pressure vessel of a typical pressurized water reactor (PWR). Theexperimental results provide a basis for benchmarking computational tools used in neutronfluence determination. The HBR-2 experiment provides measured reaction rates for a radial andan axial location at the core mid-plane elevation.

2.3.3 Objectives

The objectives of this study are to benchmark RAMA against the measurements performed at theH. B. Robinson Unit 2 reactor facility and to assess the accuracy with which the methodologypredicts the neutron flux distribution inside and outside the pressure vessel.


The HBR-2 Benchmark Experiment uses the HBR-2 2300 MW pressurized light water reactor(PWR) having dosimeters in the surveillance capsule location (azimuthal angle of 200) inside thepressure vessel and in the cavity location (azimuthal angle of 00) outside the pressure vessel.

2-25

-A'



The HBR-2 core is composed of 157 fuel assemblies arranged within a 15 by 15 grid. Each fuelassembly is comprised of 225 fuel pins also arranged in a 15 by 15 array. Of the 225 pins in theassembly, there are 204 fuel pins and 21 water pins. The assembly pitch is 21.504 cm. The activecore height is 365.76 cm.

The core is surrounded by the core baffle, core barrel, thermal shield, pressure vessel, thermalinsulation, and biological shield. The surveillance capsule is located in the downcomer regionand is attached to the outside of the thermal shield. The cavity dosimetry is located in front of asteel wall cylinder and a detector well.


Table 2-1 1 provides the material composition for each region of the HBR-2 benchmark facility.

Table 2-11Material Compositions for Regions in HBR-2 Benchmark Problem


Core Water Regions Water (density 0.766 g/cm3)

Reactor Core UO~enriched to 2.9%, Stainless Steel,Zircaloy-4, Inconel-718

Core Baffle Stainless Steel SS-304

Bypass Water Regions Water (density 0.776 g/cm')(between Baffle and Barrel)

Core Barrel Stainless Steel SS-304

Downcomer Water Regions Water (density 0.787 g/cm3)(between Barrel and RPV)

Thermal Shield Stainless Steel SS-304

Surveillance Capsule Mounting Stainless Steel SS-304

Surveillance Capsule Content Steel A533B

Pressure Vessel Cladding Stainless Steel SS-304

Pressure Vessel Steel A533B

Insulation Stainless Steel SS-304, Air

Reactor Cavity Air

Biological Shield Concrete

2-26


2.3.5 Calculations

The HBR-2 benchmark is geometrically represented in RAMA by a three-dimensional model ofthe HBR-2 facility. An additional two-dimensional model based on a horizontal slice, withplanar meshing equivalent to the three-dimensional model, is used for selected sensitivityanalyses.


For reference purposes, a coordinate system is imposed upon the model. Figure 2-12 illustratesthe coordinate system in relation to the primary components of the HBR-2 facility. Assuminggeometrical symmetry, the measurement locations are represented in the north-northeast octantof the model. The positive Z axis extends vertically upward.

YLiner P ~ ~

Detector WVells

Steel Wall Cylinder W n DeteCavity Dosimetry ]

Insulation ------Pressure Vessel

Clad Biological ShieldThermal Shield

Core BarrelBaffle

Surveillance CapsuleDow ncomer

Reflector

Core

XFigure 2-12Coordinate System for HBR-2 Benchmark RAMA Model

2-27

-_'


2.3.5.2 RAMA Model


2.3.5.3 RAMA Calculation Parameters


2.3.6 Results

Predicted specific activities are determined using the computed flux distributions. The specificactivity values utilize the BUGLE-96 activity response cross sections. Predictive values aredetermined for all measured dosimeters at both detector locations. The RAMA calculatedspecific activities are in good agreement with the measured values.

2-28


Content Deleted -


Table 2-12Specific Activities for Surveillance Capsule Dosimeters (in Bq/mg) Results (C/M)Calculated Using the 8 State-point Operating Data for Cycle 9


2-29

-1I


Table 2-13Specific Activities for Cavity Dosimeters (in Bq/mg) Results (C/M) Calculated Using the 8State-point Operating Data for Cycle 9


Table 2-14Specific Activities for Surveillance Capsule Dosimeters (in Bq/mg) Results (C/M)Calculated Using the Cycle 9 Average Operating Data Set


2-30



Table 2-15Specific Activities for Cavity Dosimeters (in Bq/mg) Results (C/M) Calculated Using theCycle 9 Average Operating Data Set



Several sensitivity analyses were performed to evaluate the stability and accuracy of RAMA forthe HBR-2 benchmark reference case with respect to mesh size and solution parameters. Asummary of these analyses is presented in Table 2-16.

Content Deleted -


2-31



Content Deleted -


2-32


Content Deleted -


Figure 2-13Variation of Planar Distance Between Parallel Rays

Content Deleted -


Figure 2-14Variation of Axial Distance Between Parallel Rays

2-33


Content Deleted -


Figure 2-15Variation of Convergence Criteria


Figure 2-16Variation of Angular Quadrature Order

2-34


2.4 BWR Numerical Benchmark

The BWR Numerical Benchmark is prescribed by the U. S. NRC for use in benchmarkingpressure vessel neutron fluence prediction methodologies. A description of the dimensions andmaterial compositions required to perform the BWR Numerical Benchmark is provided in [10].

RAMA predicted values for the displacements per atom (DPA) rates and neutron flux for energy>1.0 MeV and energy >0.1 MeV are compared to the NUREG [10] calculated values. RAMApredicted reaction rates for capsule dosimetry are also compared to the values reported in [10].Details of the model and comparison results are provided in the following subsections.


The RAMA predicted reaction rates at the capsule location are in excellent agreement with thepredicted rates reported in [10].

Content Deleted -


2-35

- - u


Content Deleted -


2.4.2 Purpose

The purpose of the BWR Numerical Benchmark is to present a standard benchmark problem thatis representative of a typical boiling water reactor (BWR). The benchmark results provide a basisfor benchmarking computational tools used in neutron fluence determination. The BWRNumerical Benchmark provides calculated neutron flux values and DPA rates for the RPV at theelevation of peak flux and at the core mid-plane. These calculated values are provided at OT,1/4T, 1/2T, 3/4T and T locations for energy >1.0 MeV and energy >0.1 MeV. The BWRNumerical Benchmark also examines calculated neutron flux values and DPA rates for thedowncomer and cavity regions. Reaction rates are calculated for the simulated capsuledosimetry.

2.4.3 Objectives

The objectives of this study are to benchmark RAMA against the calculated values presented in[10] and to assess the accuracy with which the methodology predicts the neutron flux distributioninside the pressure vessel.


The BWR Numerical Benchmark uses a typical 3833 MW boiling water reactor (BWR) havingcalculated dosimeter reaction rates in the surveillance capsule location adjacent to the pressurevessel inner wall. There are 24 jet pump assemblies positioned every 15 degrees ofcircumference in the downcomer region.


Figure 2-17 provides an elevation view of the BWR Numerical Benchmark reactor.

The reactor core region is composed of 800 fuel assemblies. Regions outside the core consist ofthe shroud, downcomer containing jet pumps and risers, pressure vessel, mirror insulation, andan outer concrete biological shield. A stainless steel surveillance capsule is located on the insideRPV liner wall at 30 azimuth.

2-36

I


The active core height is 381 cm. Regions below the active core height include the inlet regionand core plate region. Regions above the active core include a top guide region and upperreflector region.

Upper Reflector Shroud Liner

485.99 - A______________467.47 - _______\ _430.43- Top Guide

Core - a C.,0 L. %

0 r U -

49.43

12:39 = Core Plate

o ~tr u r~ - v) o es cc o° Inlet oN N > n _ > -kn 0O 00 O 0\ N 1-° CO i-

" ! qt-in.NM rC1 oo , M en en M

RADIUS (cm)

- - NOT TO SCALEFigure 2-17Elevation View of the BWR Numerical Benchmark Reactor

2-37

-4'



Table 2-17 provides the material composition for each region of the BWR Numerical Benchmarkreactor represented in the RAMA model.

Table 2-17Material Compositions for Regions in BWR Numerical Benchmark Problem


Water Regions Water

Fuel Regions 23U, 23U, 0, Zr, Water

Jet Pump Water Water

Jet Pump Metal Chromium, Iron, Nickel

Jet Pump Riser Water Water

Jet Pump Riser Metal Chromium, Iron, Nickel

Reflector Water

Shroud Stainless Steel SS-304

Downcomer Water

Surveillance Capsule Stainless Steel SS-304

RPV Liner Stainless Steel SS-304

RPV Wall Steel

Cavity Air (Oxygen)

Insulation Liner Stainless Steel SS-304

Insulation Aluminum

Biological Shield Concrete

Inlet Water, Zr, SS-304

Core Plate Water, SS-304

Top Guide Water, Zr

Upper Reflector Water, Zr, SS-304

2.4.5 Calculations

The BWR Numerical Benchmark is geometrically represented in RAMA by a three-dimensionalmodel of the BWR Numerical Benchmark facility. An additional two-dimensional model withplanar meshing equivalent to the core mid-plane axial node of the three-dimensional model isused for selected sensitivity analyses.

2-38



For reference purposes, a coordinate system is imposed upon the model. Figure 2-18 illustratesthe coordinate system in relation to the primary components of the BWR Numerical Benchmarkreactor. The positive Z axis extends vertically upward. The axial geometry is shown inFigure 2-17.

0o

yBiological Shicld

Cavity

Liner

Clad

450

Jet Pump RiserShroud

Renector

N\ Pressure Vessel

Downcomer

Core

XFigure 2-18Coordinate System for BWR Numerical Benchmark RAMA Model

2-39


2.4.5.2 RAMA Model

Content Deleted -


RAMA Calculation Parameters


2.4.5.3

2-40


2.4.6 Results

Reaction rates predicted by RAMA for capsule dosimetry at four radial locations are comparedto the NUREG calculated values in this subsection. A comparison of the RAMA calculatedneutron fluxes and DPA rates to those calculated values presented in [1O] are provided in thissubsection for the RPV, downcomer, and cavity regions at various elevations and at energy >1MeV and energy >0.1 MeV. The RPV values are calculated at the OT, l/4T, 1/2T, 3/4T and Tradial thicknesses.


Table 2-18Reaction Rate (in rps/atom) Comparison Results (RAMA/NUREG) for Capsule at 3 Degrees

Content Deleted -


2-41


Content Deleted -


Table 2-19RPV Neutron Flux and DPA Comparison Ratio (RAMA/NUREG) at Elevation of Peak Flux


2-42


Content Deleted -


Table 2-20RPV Neutron Flux and DPA Comparison Ratio (RAMAINUREG) at Core Mid-plane

Content Deleted -


2-43

-


Content Deleted -


Table 2-21Azimuthal Variation of RPV Neutron Flux Comparison Ratio (RAMA/NUREG) for Elevationof Peak Flux at OT for Energy >1.0 MeV

Content Deleted -


2-44



Table 2-22Downcomer Neutron Flux and DPA Comparison Ratio (RAMA/NUREG)


Table 2-23Cavity Neutron Flux and DPA Comparison Ratio (RAMA/NUREG) at Elevation of Peak Flux

Content Deleted -



Several sensitivity analyses were performed to evaluate the stability and accuracy of RAMA forthe BWR Numerical Benchmark reference case with respect to mesh size and solutionparameters. A summary of these analyses is presented in Table 2-24.

2-45


Content Deleted -



Content Deleted -


2-46


Content Deleted -


Figure 2-19Variation of Planar Distance Between Parallel Rays

Content Deleted -


Figure 2-20Variation of Convergence Criterion

2-47


Content Deleted -


Figure 2-21Variation of Angular Quadrature Order

2-48

___ I

3REFERENCES

1. BWRVIP-1 14: BWR Vessel and Internals Project, RAMA Fluence Methodology TheoryManual, EPRI, Palo Alto, CA: 2003 1003660.

2. D. B. Jones et al., "RAMA Fluence Methodology User's Manual," EPRI, Palo Alto, CA,2003.

3. D. B. Jones et al., "RAMA Fluence Methodology Procedures Manual," EPRI, Palo Alto, CA,2003.

4. "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence,"Nuclear Regulatory Commission Regulatory Guide 1.190, March 2001.

5. D. B. Jones, et al., "RAMA Fluence Methodology Benchmark Manual - Evaluation ofSusquehanna Unit 2, Cycles 1-5" EPRI, Palo Alto, CA, 2003.

6. I. Remic and F. B. K. Kam, "Pool Critical Assembly Pressure Vessel Facility Benchmark,"NUREG/CR-6454, Oak Ridge National Laboratory, ORNLITM-13205, July 1997.

7. "Prediction of Neutron Embrittlement in the Reactor Pressure Vessel: VENUS-1 andVENUS-3 Benchmarks," NEA Nuclear Science Committee Task Force on ComputingRadiation Dose and Modelling of Radiation-induced Degradation of Reactor Components,OECD, 2000, p. 212.

8. "BUGLE-96: Coupled 47 Neutron, 20 Gamma-Ray Group Cross Section Library Derivedfrom ENDF/B-VI for LWR Shielding and Pressure Vessel Dosimetry Applications," RSICCData Library Collection, DLC-1 85, March 1996.

9. Remec and F. B. K. Kam, "H. B. Robinson-2 Pressure Vessel Benchmark,"NUREG/CR-6453, Oak Ridge National Laboratory, ORNL/TM-13204, February 1998.

10. J. F. Carew, et al., "PWR and BWR Pressure Vessel Fluence Calculation BenchmarkProblems and Solutions," NUREG/CR-6115, Brookhaven National Laboratory,BNL-NUREG-52395, September 2001.

11. J. E. White, D. T. Ingersoll, C. 0. Slater, and R. W. Roussin, "BUGLE-96: A RevisedMultigroup Cross Section Library for LWR Applications Based on ENDF/B-VI Release 3,"ANS Radiation Protection & Shielding Topical Meeting, Falmouth, MA, April 1996.

3-1

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Nuclear Power

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