NUREG-1511. SuQV. 2

Reactor Pressure Vessel 

Status Report 

Supplement 2 

U.S. Nuclear Regulatory Commission 

Office of Nuclear Reactor Regulation 

Washington, DC 20555-0001

NUREG-1511 Sum 2
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212-642-4900
NUREG-1511, Supp. 2

Reactor Pressure Vessel 

Status Report

Manuscript Completed: September 2000 

Date Published: October 2000 

C. E. Carpenter, B. J. Elliot, A. L. Hiser, Jr., 

M. J. Khanna, A. D. Lee, J. Medoff, 

M.A. Mitchell, S. C. F. Sheng, K. R. Wichman, 

J. K Strosnider, W. H. Bateman 

Division of Engineering 

Office of Nuclear Reactor Regulation 

U.S. Nuclear Regulatory Commission 

Washington, DC 20555-0001
ABSTRACT

On May 18, 1995, the U.S. Nuclear 

Regulatory Commission (NRC) issued 

Generic Letter (GL) 92-01, Revision 1, 

Supplement 1, "Reactor Vessel Structural 

Integrity," requesting that addressees: (1) 

identify, collect, and report any new data 

pertinent to the analysis of structural integrity 

for the reactor pressure vessels (RPVs) at 

their nuclear power plants, and (2) assess the 

impact of those data on their RPV integrity 

analyses relative to the requirements of 

10 CFR 50.61, "Fracture Toughness 

Requirements for Protection Against 

Pressurized Thermal Shock Events," and to 

the requirements of Appendix G to 10 CFR 

Part 50, "Fracture Toughness Requirements," 

which is used to evaluate the upper shelf 

energy (USE) values for RPV materials and 

pressure-temperature (P-T) limits for RPVs. 

Since the issuance of GL 92-01, Revision 1, 

Supplement 1, the industry completed a 

major initiative to collect all available alloying 

chemistry and materials property data for the 

various forging, plate and weld materials 

used to fabricate the RPVs. This represents 

the first time that a comprehensive, 

integrated assessment of all alloying 

chemistry and surveillance data has been 

completed for the materials that have been 

used to fabricate the RPVs in U.S. light

water nuclear power plants. These efforts 

should mini'mize surprises regarding the 

alloying chemistries for domestic RPV beltline 

materials. In addition, as a result of the 

industry's efforts in response to GL 92-01, 

Revision 1, Supplement 1, licensees and staff 

will be able to perform reactor vessel integrity 

evaluations more efficiently and more 

effectively. However, the staff expects 

additional surveillance data will become 

available after fracture toughness testing is 

performed on surveillance capsules that are 

presently being irradiated in domestic RPVs. 

The staff will incorporate the additional

surveillance data into the existing database 

after it becomes available for review. 

In the summer and fall of 1998, the staff 

issued a series of requests for additional 

information (RAIs) regarding the industry's 

responses to GL 92-01, Revision 1, 

Supplement 1. In the RAIs, the staff 

requested that the addressees assess how 

the updated alloying chemistry and materials 

property data would affect the results of the 

RPV integrity analyses for their plants. 

This report summarizes both the industry's 

and the NRC's efforts to address how all of 

the new chemistry and surveillance data, 

when integrated, could affect the plant

specific RPV integrity analyses for the RPVs 

of U.S. light-water nuclear power plants. 

Specifically, this report discusses the 

following: (1) the basis for issuing and 

reviewing the responses to the RAIs on 

GL 92-01, Revision 1, Supplement 1, (2) the 

activities conducted by the industry owners 

groups to collect weld chemistry data for the 

materials used to fabricate the RPVs at U.S. 

nuclear plants, (3) the activities conducted by 

the industry owners groups to collect data 

from RPV surveillance capsules on behalf of 

the industry, (4) the staff's efforts to update 

plant-specific data into the Reactor Vessel 

Integrity Database, and to make Version 2 of 

the database accessible via the World-Wide

Web, (5) the staff's current regulatory and 

research activities regarding RPV integrity, 

(6) the staff's activities regarding RPV weld 

inspection reduction, as documented in the 

NRC final safety evaluation on Topical 

Report BWRVIP-05, (7) the staff's review of 

the thermal annealing project at the Marble 

Hill facility and the status of the Palisades 

thermal annealing application, and (8) the 

results of significant plant-specific RPV 

integrity reviews that could have an impact on 

the industry.

iii NUREG-1511, Supp. 2
CONTENTS 

Page ABSTRACT ............................................................... 

EXECUTIVE SUMMARY ..................................................... vii 

ACKNOW LEDGMENTS ..................................................... xi 

ABBREVIATIO NS ......................................................... xiii 

MATHEMATICAL AND SCIENTIFIC NOMENCLATURE ............................ xv 

1 Introduction ......................................................... 1-1 

1.1 Overview ........................ .......................... 1-1 

1.2 The Pressurized Thermal Shock Rule ................................ 1-1 

1.3 Requirements for Upper Shelf Energy and Pressure-Temperature Limits .... 1-2 

1.4 Generic Letter 92-01, Revision 1, NUREG-1511, and Supplements ........ 1-3 

2 Generic Letter 92-01, Revision 1, Supplement 1: Requests for 

Additional Inform ation .................................................. 2-1 

2.1 Background ................................................... 2-1 

2.2 Activities Regarding RPV Weld Chemistries ........................... 2-1 

2.3 Activities Regarding RPV Surveillance Data ........................... 2-4 

2.4 Requests for Additional Information ................................. 2-8 

2.4.1 Contents of the Requests for Additional Information ............... 2-8 

2.4.2 Status of the Staff's Review of the Responses to the RAIs .......... 2-8 

2.4.3 Closeout of Generic Letter 92-01, Revision 1, Supplement 1 ...... 2-10 

3 Significant Regulatory and Research Activities on RPV Integrity ................. 3-1 

3.1 Background ................................................... 3-1 

3.2 Research Activities ............................................... 3-1 

3.3 Codes and Standards Activities .................................... 3-2 

4 Recent Industry Efforts Regarding BWR RPV Weld Inspection Reduction 

(BW RVIP-05) ........................................................ 4-1 

4.1 O verview ...................................................... 4-1 

4.2 Background ................................................... 4-1 

4.3 NRC Assessment of BWRVIP-05 ................................... 4-3

V NUREG-1511, Supp. 2
Page 

5 Reactor Pressure Vessel Thermal Annealing ................................ 5-1 

5.1 Thermal Annealing Regulation ..................................... 5-1 

5.2 Regulatory Guides and Technical Codes and Standards for Thermal Annealing 5-1 

5.2.1 Regulatory Guide 1.162 .................................... 5-1 

5.2.2 ASTM Standard Procedure E-509 ............................ 5-1 

5.2.3 ASME Code Case N-557 on Thermal Annealing .................. 5-1 

5.3 Palisades Thermal Annealing Report ................................ 5-1 

5.4 Department of Energy/Industry Annealing Demonstration Project ........ 5-2 

6 Reactor Vessel Integrity Database ........................................ 6-1 

6.1 Development of the Reactor Vessel Integrity Database .................. 6-1 

6.2 Availability of the RVID on the World Wide Web ........................ 6-1 

6.3 Content of the RVID ............................................. 6-1 

6.4 Quality Assurance of RVID Data ................................... 6-2 

6.5 Future Revisions to the RVID ..................................... 6-2 

7 Significant Plant-Specific Reviews ........................................ 7-1 

7.1 Background ................................................... 7-1 

7.2 Palisades ..................................................... 7-1 

7.3 Beaver Valley Unit 1 ............................................. 7-2 

7.4 Fort Calhoun ................................................... 7-2 

7.5 License Renewal Plants: Calvert Cliffs Units 1 and 2, and 

Oconee Units 1, 2, and 3 ......................................... 7-2 

7.6 Reactor Pressure Vessels Fabricated by Multiple Vendors ................ 7-2 

7.7 Reactor Pressure Vessels Fabricated from Forging Materials ............. 7-3 

8 Summary and Conclusions .............................................. 8-1 

9 R eferences .......................................................... 9-1 

Appendix - PTS Tables for Pressurized Water Reactors and 

Limiting Material Tables for Boiling Water Reactors ..................... A-1

NUREG-1511, Supp. 2 vi
EXECUTIVE SUMMARY

This NUREG describes the actions taken by 

the U.S. Nuclear Regulatory Commission 

(NRC), as well as nuclear industry owners 

groups and individual licensees, regarding 

the ongoing assessment of reactor pressure 

vessel (RPV) integrity. Following the issuance 

of Generic Letter (GL) 92-01, Revision 1, in 

March 1992 ( Ref. 1), and NUREG-1511 in 

December 1994 (Ref. 2), the staff directed its 

efforts toward: (1) addressing an issue that 

some licensees were not aware of, or not 

using all of the available chemistry and 

surveillance data applicable to the 

evaluations of their RPV beltline materials, 

and (2) determining the generic implications 

of the larger-than-expected variability 

observed in the chemical compositions of 

RPV welds. To address these issues, the 

staff issued Supplement 1 to GL 92-01, 

Revision 1 (GL 92-01, Revision 1, 

Supplement 1), on May 18, 1995 (Ref. 3), 

and requested that the licensees to which the 

GL was addressed identify any new data that 

could be pertinent to, and impact the 

structural integrity analyses of, their RPVs 

relative to the requirements of Section 50.60 

of Part 50 of Title 10, Code of Federal 

Regulations (Ref. 4), Section 50.61 of Part 50 

of Title 10, Code of Federal Regulations 

(Ref. 5), and Appendices G and H to Part 50 

of Title 10, Code of Federal Regulations 

(Refs. 6 and 7).(1) 

Since the issuance of GL 92-01, Revision 1, 

Supplement 1, the industry owners groups 

(e.g., the Boiling Water Reactor Vessel and 

Internals Project (BWRVIP), the Combustion 

Engineering Owners Group (CEOG), and the 

Babcock and Wilcox Owners Group

Henceforth, Sections of Part 50 to Title 10 of the 

Code of Federal Regulations will be abbreviated 

10 CFR 50.XX or 10 CFR 50.XXX. Appendices 

to Part 50 will be abbreviated as 10 CFR Part 50, 

Appendix X designations.

(B&WOG)) have completed a major initiative 

to collect all available alloying chemistry and 

materials property data for the various 

forging, plate and weld materials used in the 

fabrication of U.S. RPVs. 

This represents the first time that a 

comprehensive, integrated assessment of all 

alloying chemistry and surveillance data has 

been completed for the materials that have 

been used to fabricate the RPVs in U.S. light

water nuclear power plants. These efforts 

should minimize surprises regarding the 

alloying chemistries for domestic RPV beltline 

materials. However, the staff expects that 

with the testing of future surveillance 

capsules, additional materials property and 

dosimetry data will become available. The 

Charpy materials property data and the 

reanalysis of capsule and RPV fluences 

(based on the dosimetry data) will be used to 

revise the existing database. 

All licensees responded to GL 92-01, 

Revision 1, Supplement 1. Although some 

licensees provided additional data that were 

not included in their responses to the initial 

version of the GL, the collective responses to 

GL 92-01, Revision 1, Supplement 1, 

demonstrated that there were no new RPV 

integrity issues that would be considered as 

an immediate safety concern. The majority of 

the licensees also indicated that they were 

participating in the activities of their 

respective industry owners groups to collect 

and analyze available RPV weld chemistry 

(specifically the copper and nickel contents) 

and surveillance data on behalf of their 

member utilities. The owners groups have 

now completed these initiatives. 

In the summer and fall of 1996, NRC staff 

issued closeout letters on GL 92-01, 

Revision 1, Supplement 1, which noted that 

no immediate safety issues were associated

vii NUREG-1511, Supp. 2
with the structural integrity assessments for 

U.S. light-water reactors (LWRs). However, 

the staff also acknowledged that ongoing 

RPV initiatives were being conducted by the 

industry owners groups, and informed the 

licensees that additional NRC work might be 

scheduled pending its review of the results of 

these initiatives. 

This review of the industry's RPV integrity 

initiatives led the NRC staff to conclude that 

the new data compiled by the owners groups 

could have an impact on the pressurized 

thermal shock (PTS) and pressure

temperature (P-T) limit assessments for 

some facilities; however, the staff did not 

consider this to be an immediate safety 

concern because these assessments are 

considered to be time dependent analyses.(2) 

As a result, the staff concluded that there was 

sufficient time for licensees to assess the 

impact of the new data on their PTS 

assessments (applicable to PWRs only) and 

P-T limit assessments prior to the expiration 

date of the operating licenses for their 

facilities.(2) Therefore, in 1998, the staff 

issued a series of requests for additional 

information (RAIs) to the majority of licensees 

that responded to GL 92-01, Revision 1, 

Supplement 1, and requested that the 

recipients assess the impact of the newly 

compiled chemistry and surveillance data on 

the PTS and P-T limit evaluations for their 

facilities. This report, in part, summarizes 

the staff's evaluation of the responses to 

the RAIs on GL 92-01, Revision 1, 

Supplement 1. The status of the staff's 

(2) The staff's generic assessment is documented in 

a memorandum from Jack R. Strosnider, Chief, 

Materials and Chemical Engineering Branch, 

Division of Engineering, Office of Nuclear 

Reactor Regulation, to Ashok C. Thadani, 

Associate Director for Technology, Office of 

Nuclear Reactor Regulation, dated May 5, 1995. 

This memorandum was included as part of 

Commission Paper SECY-95-119, "Status of 

Reactor Pressure Vessel Issues" (May 8, 1995, 

Ref. 8)

evaluations through its review of the 

responses to the RAIs on GL 92-01, 

Revision 1, Supplement 1, indicates that, with 

the exception of the RPVs for the Palisades 

and Fort Calhoun nuclear plants, all RPVs of 

light-water reactors in the U.S. will be 

in compliance with the requirements of 

10 CFR Part 50.61, and 10 CFR Part 50, 

Appendices G and H, throughout the terms of 

the operating licenses for the facilities. Both 

the licensees for the Palisades and Fort 

Calhoun nuclear plants have ongoing efforts 

to review the materials property and neutron 

fluence data for their RPVs, and to address 

compliance with the PTS requirements of 

10 CFR 50.61. 

Several developments have also occurred in 

the area of Codes and Standards activities 

and research developments that affect RPV 

integrity assessments. In particular, the 

American Society of Mechanical Engineers 

(ASME) Code Committees have passed 

Code Cases N-640 and N-588 (Refs. 9 and 

10), which have direct consequences on P-T 

limit assessments, and Code Case N-629 

(Ref. 11), which proposes a new 

methodology (the "Master Curve") for 

indexing material reference temperatures. 

Research developments have included work 

on new RPV embrittlement correlations, RPV 

flaw distribution studies, ongoing work to 

support the Master Curve methodology, and 

an effort by the NRC and industry to revise 

the PTS screening criteria. 

The staff has also considered proposals to 

reduce the scope of augmented inspections 

performed on the circumferential shell welds 

of boiling water reactor (BWR) RPVs. The 

basis for this proposed reduction in scope for 

the augmented inspections, which are 

required pursuant to paragraph (g)(6)(ii)(A) of 

10 CFR 50.55a (Ref. 12), was proposed by 

the BWRVIP and is documented in Topical 

Report BWRVIP-05 (Ref. 13). The staff 

approved this topical report on July 28, 1998 

(Ref. 14), and stated its position for reduced

NUREG- 1511, Supp. 2 viii
inspection scopes of BWR circumferential 

RPV shell welds in GL 98-05 (Ref. 15), which 

was issued on November 10, 1998. The staff 

has currently approved a number of 

proposals for reduced inspections of the 

circumferential RPV shell welds in BWR

designed facilities. 

With the adoption of 10 CFR 50.66 (Ref. 16), 

the staff has also established a regulatory 

framework for the thermal annealing of RPVs. 

Guidance for complying with 10 CFR 50.66 is 

documented in RG 1.162 (Ref. 17). The 

feasibility of thermal annealing was 

demonstrated by a joint Department of 

Energy (DOE)/Industry-sponsored annealing 

demonstration project (ADP); this ADP was 

performed at a decommissioned Marble Hill 

reactor unit, and applied indirect heating as 

the method for annealing the unit's RPV. 

A second ADP using an electrical resistance 

heating approach was terminated. 

The staff has also updated the reactor vessel 

integrity database (RVID). The RVID provides 

an efficient and effective means of storing 

and maintaining RPV vessel data relative

to the requirements of 10 CFR 50.60; 

10 CFR 50.61; and 10 CFR Part 50, 

Appendices G and H. The database can be 

used as an effective means of indicating 

compliance with these regulations and 

maintains safety through a comprehensive 

and integrated approach. RVID Version 2 

(RVID 2) was issued on the World-Wide-Web 

in June 1999. RVID 2 is a Windows 3.1 native 

application based on Microsoft Access 2 .

0 TM. 

The four RVID 2 diskettes can be 

downloaded from the homepage at 

(http://www.nrc.gov/NRR/RVID/index. html) 

which is linked to the NRC homepage. RVID 

was developed following the staff's review of 

licensee responses to GL 92-01, Revision 1. 

The database summarizes the properties of 

the reactor vessel beltline materials for each 

operating commercial nuclear power plant. 

RVID 2 reflects licensee responses to 

the RAIs on GL 92-01, Revision 1, 

Supplement 1. The database will be updated 

when sufficient amounts of new surveillance 

data, chemistry data, or fluence evaluations 

warrant a new revision.

ix NUREG-1511, Supp. 2
ACKNOWLEDGMENTS

We would like to thank Ms. Rene S. Cesaro for her valuable contributions to the report and time 

spent in reviewing, editing, and formatting the final product. We would also like to thank 

Mr. Michael E. Mayfield and Dr. Edwin M. Hackett of the NRC's Office of Research for their insights 

to the content of this report.

xi NUREG-1511, Supp. 2
ABBREVIATIONS

10 CFR Title 10 to the Code of Federal 

Regulations 

ADP annealing demonstration project 

ART adjusted reference temperature 

ASME American Society of Mechanical 

Engineers 

ASTM American Society for Testing and 

Materials 

B&W Babcock and Wilcox Company, now 

Framatome Technologies, Inc. 

B&WOG Babcock and Wilcox Owners Group 

BWR boiling water reactor 

BWR-4 boiling water reactor 4 design, one 

of General Electric's model designs 

for boiling water reactors 

BWROG Boiling Water Reactor Owners 

Group 

BWRVIP BWR Vessel and Internals Project 

CB&I Chicago Bridge and Iron Works 

CDF core damage frequency 

CE Combustion Engineering 

Corporation, which is currently 

named CE Nuclear Power LLC 

CEOG Combustion Engineering Owners 

Group 

CF chemistry factor 

CPCo Consumers Power Company 

CRD control rod drive

DOE U.S. Department of Energy 

EOL end of license 

EPRI Electric Power Research Institute 

GL Generic Letter 

HPCI high pressure coolant injection 

HPCS high pressure coolant spray 

IHI Ishikasajima-Hirama Heavy 

Industries 

IN Information Notice 

ISA independent safety assessment 

ISI inservice inspection 

LPCI low pressure coolant injection 

LPCS low pressure coolant spray 

LTOP low temperature overpressure 

protection 

LWR light water reactor 

MOU memorandum of understanding 

NEI Nuclear Energy Institute 

NRC U.S. Nuclear Regulatory 

Commission 

NSSS Nuclear Steam Supply System 

ORNL Oak Ridge National Laboratories 

P(FIE) conditional probability of failure 

P-T pressure-temperature 

PFM probabilistic fracture mechanics

NUREG-1511, Supp. 2 °,,i
PNNL Pacific Northwest National 

Laboratories 

PTS pressurized thermal shock 

PVRUF Pressure Vessel Research Users 

Facility 

PWR pressurized water reactor 

QA quality assurance 

RAI request for additional information 

RCIC reactor core isolation cooling 

RES Office of Nuclear Regulatory 

Research of the U.S. 

Nuclear Regulatory Commission 

RFO refueling outage

RG 

RPV 

RVID 

RVID 2 

SER 

SLC 

TAR 

USE 

UT 

VTT

Regulatory Guide 

reactor pressure vessel 

Reactor Vessel Integrity Database 

RVID Version 2 

safety evaluation report 

standby liquid control 

thermal annealing report 

upper shelf energy 

ultrasonic testing 

Technical Research Center of 

Finland (Valtion Teknillinen 

Tutkinuskeskus)

NUREG-1511, Supp. 2 xiV
MATHEMATICAL AND SCIENTIFIC NOMENCLATURE

A. as defined in equation 5 of 

10 CFR 50.61, the measured 

ARTNDT value for base metal 

materials, or for weld materials, the 

value of ARTNDT, as adjusted to 

account for the differences in the 

table CFs for the RPV and 

surveillance capsule weld materials 

ART adjusted reference temperature 

°C abbreviation for degrees C, a 

standard unit of temperature in the 

Centigrade temperature scale 

CF chemistry factor, which is a function 

of the copper and nickel alloying 

contents of reactor pressure vessel 

materials and is used in the 

calculations of ART and RTpTs 

Cu Periodic Table abbreviation for the 

element copper 

OF abbreviation for degrees F, a 

standard unit of temperature in the 

Fahrenheit temperature scale 

f projected neutron fluence value for 

a RPV material for neutrons having 

kinetic energies in excess of 1 MeV 

- reported in units of 1019 n/cm2 

(1019 neutrons per square 

centimeter) 

as defined in equation 5 of 

10 CFR 50.61, the neutron fluence 

value for the A, surveillance data 

point 

ft-lb an abbreviation for foot-pounds, a 

standard unit of work or energy in 

the English system of weights and 

measures

joules a standard unit of work or energy in 

the SI system of weights and 

measures

Kla lower bound crack arrest fracture 

toughness from Section XI of the 

ASME Code

Kic lower bound static initiation fracture 

toughness from Section Xl of the 

ASME Code 

M margin term to be added in the 

calculations of adjusted reference 

temperatures to account for 

uncertainties in the calculational 

procedures, the initial reference 

temperature, the copper and nickel 

contents of the vessel material, and 

the neutron fluence

Mn 

Mo

Periodic Table abbreviation for the 

element manganese 

Periodic Table abbreviation for the 

element molybdenum

MPa an abbreviation for megapascals, a 

unit of pressure or stress in the SI 

system of weights and measures

n as defined in equation 5 

10 CFR 50.61, the number 

surveillance capsule data points

of 

of

Ni Periodic Table abbreviation for the 

element nickel 

P(FIE) conditional probability of failure, a 

parameter used in probabilistic 

fracture mechanics calculations for 

reactor pressure vessel materials

XV NUREG-1511, Supp. 2
psig an abbreviation for pounds per 

square inch, a unit of pressure or 

stress in the English system of 

weights and measures 

RTNDT the reference temperature for a 

RPV material 

RTNDT(U) the initial reference temperature of 

the RPV material in the unirradiated 

condition 

RTPTS the reference temperature for 

pressurized thermal shock, which is 

equivalent to the adjusted reference 

temperature for a RPV material in 

PWRs at the end-of-license for the 

facility, as determined using the 

best estimate end-of-license 

neutron fluence for the material at 

the clad-base metal interface of the 

vessel 

RTTO as defined in ASME Code Case 

N-629, the fracture toughness curve 

indexing temperature based on use 

of the Master Curve methodology 

which replaces the use of RTNDT 

(mathematically defined as equal to 

To + 35°F) 

Table CFvessel chem. CF for a particular 

reactor vessel weld as determined 

from the tables in 10 CFR 50.61, 

and based on its copper and nickel 

alloying contents 

Table CFvessei suv. CF for a particular 

surveillance capsule weld material 

as determined from the tables in 

10 CFR 50.61, and based on its 

copper and nickel alloying contents

TO As defined in ASTM Standard 

Procedure E1921, and used in 

Master Curve methodology, a 

temperature defined to correspond 

to a fracture toughness of 

100 MPaVin 

ARTpTs the specific term for the mean value 

of the increase (shift) in the 

reference temperature for materials 

in PWR RPVs as a result of the 

effects of neutron irradiation on the 

materials, as used in the 

calculations for pressurized thermal 

shock (i.e., RTpTs calculations) 

ARTNDT the general term for the mean value 

of the increase (shift) in the 

reference temperature for RPV 

materials as a result of the effects 

of neutron irradiation on the 

materials 

,_ Symbol for a summation function in 

mathematics 

Ou standard deviation term for RTNDT(U), 

a parameter used in the calculation 

of margin terms (M) in plant specific 

evaluations for pressurized thermal 

shock (for the case of general ART 

calculations, this term is referred to 

as a, ) 

CFA standard deviation term for ARTNDT 

or ARTPTS, a parameter used in the 

calculation of margin terms (M)

NUREG-1511, Supp. 2 xvi
1 INTRODUCTION

1.1 Overview 

The Nuclear Regulatory Commission (NRC) 

has established regulations to address the 

implications of accumulated neutron 

irradiation on the structural integrity of the 

RPVs in the commercial nuclear industry. 

These regulations include 10 CFR 50.60 

(Ref. 4); 10 CFR 50.61, the Pressurized 

Thermal Shock (PTS) Rule (Ref. 5); and 

10 CFR Part 50, Appendices G and H 

(Refs. 6 and 7). 10 CFR 50.60 requires licensees to comply with the reactor coolant 

pressure boundary requirements and RPV 

material surveillance program requirements 

set forth in 10 CFR Part 50, Appendices G 

and H, respectively. 10 CFR 50.60, however, 

allows licensees to use an alternative to the 

requirements of 10 CFR Part 50, Appendices 

G and H, when the Commission grants an 

exemption under the requirements of 

10CFR50.12. Both 10 CFR 50.61, and 

10 CFR Part 50, Appendix G, establish limits 

on the degree to which the RPV may be 

embrittled as a result of neutron irradiation. 

Another regulation, 10 CFR Part 50, 

Appendix H, establishes the requirements for 

developing plant-specific RPV surveillance 

data that are used to monitor the 

structural integrity assessments required by 10 CFR 50.61, and 10 CFR Part 50, 

Appendix G. 

1.2 The Pressurized Thermal Shock 

Rule 

10 CFR 50.61, the PTS Rule, defines 

screening criteria for embrittlement of RPV 

materials in pressurized-water reactors 

(PWRs), as well' as the actions that are 

required if these screening criteria are

exceeded. The screening criteria limit the 

degree that a vessel material may increase in 

its reference temperature (RTPTs) following 

neutron irradiation of the RPV. The RTpTs 

values, which are based on the projected 

end-of-license (EOL) neutron fluence values 

for the RPV materials, are calculated in 

accordance with Equation 1-1: 

RTPTs = RTNDT(U) + ARTPTs + M (1-1) 

In this equation, RTNDT(U) represents the initial 

reference temperature of the vessel material 

in the unirradiated condition; ARTPTS 

represents the increase (shift) in reference 

temperature value for the material as a result 

of neutron irradiation of the RPV; and M 

represents the margin to be added in the 

calculations to account for uncertainties in the 

calculational procedures and in the methods 

for measuring the initial reference 

temperatures, copper and nickel contents, 

and neutron fluence values. Specifically, 

M = 2V/(au 2 + GA2), where ou is the standard 

deviation for RTNDT(u) and a. is the standard 

deviation for ARTPTS. The PTS rule requires 

licensees to calculate the shift in an RPV 

material's reference temperature value 

(ARTPTS) in accordance with Equation 1-2:

A RTPTs = CF x f(0.28-0.10x og(1)) (1-2)

In Equation 1-2, Y" represents the projected 

EOL neutron fluence (in units of 1019 n/cm2, for neutrons with kinetic energies greater 

than 1 MeV) for the material at the clad-base 

metal interface (i.e., at the inside surface of 

the base metal) for the vessel, and CF 

presents a proportionality factor, otherwise 

known as the chemistry factor. The PTS rule 

requires the CF for an RPV material to be 

determined by one of two methods:

1-1 NUREG-1511, Supp. 2
(1) by applying the methods of section 

(c)(1)(iv)(A) to 10 CFR 50.61, which 

gives the bases for using the tables in 

the rule to calculate the CF as a 

function of the copper and nickel 

alloying (chemistry) contents of the 

material; or 

(2) by applying the methods of sections 

(c)(2)(ii and iii) to 10 CFR 50.61 and the 

results of Charpy-V impact tests on 

surveillance capsule specimens 

removed in accordance with a utility's 

reactor vessel material surveillance 

program (i.e., the .10 CFR Part 50, 

Appendix H, program for the plant) if 

the testing data have been determined 

to be credible in accordance with the 

criteria of 10 CFR 50.61 (c)(2)(i)(A-E). 

The screening criteria in the PTS rule are 

132 0C (270 0 F) for plate, forging, and axial 

weld materials and 149 0C (300 0 F) for 

circumferential weld materials. When RTPTS 

values are projected to exceed these 

screening criteria, the rule requires that 

licensees perform neutron flux reductions, 

plant modifications, or additional plant

specific evaluations of their RPVs to justify 

continued operation of their reactors. 

1.3 Requirements for Upper Shelf 

Energy and Pressure-Temperature 

Limits 

10 CFR Part 50, Appendix G, contains 

screening criteria that limit the degree that an 

RPV material may drop in its upper shelf 

energy (USE) value following irradiation of 

the vessel. The regulation requires the USE 

for an RPV material (as measured from the 

results of Charpy-V impact tests) to be 

greater than 102 joules (75 ft-lb) when the 

material is in the unirradiated condition. The 

regulation also requires the USE of the 

material to remain above 68 joules (50 ft-lb) 

throughout the licensed life of the vessel.

If these conditions are not met, the regulation 

requires that additional fracture mechanics 

analyses be performed to demonstrate that 

sufficient margins of safety will exist for lower 

values of USE. These safety margins must be 

at least as conservative as those that would 

be obtained if the criteria in the edition and 

addenda of Appendix G to Section XI of the 

ASME Boiler and Pressure Vessel Code 

(Appendix G to the ASME Code, Ref. 18), as 

endorsed in 10 CFR 50.55a (Ref. 12), were 

used to satisfy the safety margin 

requirements. 

Through efforts coordinated by the owners 

groups, the industry has performed 

equivalent margins analyses to demonstrate 

that USE values below 68 J (50 ft-lb) will 

provide margins of safety against fracture 

equivalent to those required by Appendix G to 

the ASME Code. These analyses were 

performed for generic groupings of plants. In 

addition, some licensees performed plant

specific equivalent margins analyses. In 

NUREG/CR-6023 (Ref. 19), the NRC staff 

concluded that PWR and boiling water 

reactor (BWR) RPV materials could have 

EOL USE values less than 68 joules (50 ft-lb) 

and still provide the required margins of 

safety against fracture. On the basis of the 

industry's equivalent margins analyses and 

NRC's generic study, the staff concluded in 

NUREG-1511 (Ref. 2) that all RPVs will have 

adequate upper-shelf toughness throughout 

their current licensed operating life. 

10 CFR Part 50, Appendix G, also 

establishes requirements for the calculation 

of P-T limit curves that are used as a means 

of protecting the integrity of the RPV during 

normal operating conditions, anticipated 

operational occurrences, and pressure testing 

conditions. These P-T limits are used to 

establish low-temperature overpressure 

protection (LTOP) system setpoints for the 

plants. The regulation requires that the P-T 

limits must be at least as conservative as 

those that would be generated using the

NUREG- 1511, Supp. 2 1-2
methods of analysis and margins of safety of 

Appendix G to the ASME Code, as endorsed 

in 10 CFR 50.55a. 10 CFR 50.60 requires 

that licensees submit requests for exemptions 

to use less conservative P-T limits when this 

condition is not satisfied. Such requests are 

evaluated, and granted or denied by NRC, on 

a case-by-case basis. 

1.4 Generic Letter 92-01, Revision 1, 

NUREG-1511, and Supplements 

The "Reactor Pressure Vessel Status 

Report," NUREG-1511 (Ref. 2), described 

the RPV and discussed the effect of radiation 

embrittlement on RPV materials and the 

indicators for measuring embrittlement. 

NUREG-1 511 also summarized the results of 

the NRC staff's review of the industry's 

responses to Generic Letter (GL) 92-01, 

Revision 1 (Ref. 1), as well as plant-specific 

RPV evaluations for the 37 BWR plants and 

74 PWR plants in the United States. The data 

resulting from the staff's review are stored in 

NRC's RVID database. Following the 

issuance of GL 92-01, Revision 1, and 

NUREG-1511, the staff directed its efforts 

toward assuring that the licensees for U.S. 

light-water reactors were evaluating their 

RPVs by applying all available chemistry and 

surveillance data pertinent to the RPV 

assessments required by the confines of the 

current regulatory framework (e.g., within 

the regulatory framework of 10 CFR 50.60; 

10 CFR 50.61; and 10 CFR Part 50, 

Appendices G and H). To address this 

concern, the staff issued Supplement 1 

to GL 92-01, Revision 1 (Ref. 3), on 

May 18, 1995. In this supplement to the GL, 

the staff requested that the addressees 

identify, collect, and report any new data that 

could be pertinent to, and impact the 

structural integrity analyses of their RPVs 

relative to the requirements of 10 CFR 50.60; 

10 CFR 50.61; and 10 CFR Part 50, 

Appendices G and H.

All licensees responded to GL 92-01, 

Revision 1, Supplement 1. Although some 

licensees provided additional data that were 

not included in their responses to the initial 

version of the generic letter, all licensees 

indicated that there were no new RPV 

integrity issues that would have immediate 

safety concerns. The majority of the licensees 

also indicated that they were participating in 

the activities of their respective industry 

owners group (e.g., BWRVIP, CEOG, or 

B&WOG) to collect and analyze available 

RPV weld chemistry data (specifically, copper 

and nickel content data) and surveillance 

data on behalf of its member utilities. The 

owners groups have now completed these 

initiatives. This represents the first time that a 

comprehensive, integrated assessment of all 

alloying chemistry and surveillance data has 

been completed for the materials that have 

been used to fabricate the RPVs in U.S. light

water nuclear power plants. These efforts 

should minimize surprises regarding the 

alloying chemistries for domestic RPV beltline 

materials. However, the staff expects 

additional surveillance data will become 

available after fracture toughness testing is 

performed on surveillance capsules that are 

presently being irradiated in domestic RPVs. 

The staff will incorporate the additional 

surveillance data into the existing database 

after it becomes available for review. 

In October 1996 the NRC staff issued 

Supplement 1 to NUREG-1511 (Ref. 20). 

The updated "Reactor Pressure Vessel 

Status Report" discussed: (1) the basis for 

issuing GL 92-01, Revision 1, Supplement 1; 

(2) the status of licensee Responses 

to GL 92-01, Revision 1, Supplement 1; 

(3) the up-to-date status of licensee 

compliance with the PTS rule; (4) the NRC's 

establishment of a framework for the thermal 

annealing of RPVs; and (5) the staff's 

development of an updated version of the 

RVID. In the summer and fall of 1996, 

the staff issued its closeout letters on

1-3 NUREG-1511, Supp. 2
GL 92-01, Revision 1, Supplement 1. In 

these close-out letters, the staff noted that no 

immediate safety issues were associated with 

the structural integrity assessments for U.S. 

LWRs. However, the NRC staff also 

acknowledged that ongoing RPV initiatives 

were being conducted by the industry owners 

groups and informed the licensees that 

additional NRC work might be scheduled 

pending its review of the results of these 

initiatives. 

This review of the industry's vessel integrity 

initiatives led the NRC staff to conclude that

the new data compiled by the owners groups 

could have an impact on the PTS and P-T 

limit assessments for some facilities. 

Therefore, in 1998, the staff issued a series 

of RAIs to the majority of licensees that 

responded to GL 92-01, Revision 1, 

Supplement 1, and requested that the 

recipients assess the impact of the new 

chemistry and surveillance data on the PTS 

and P-T limit evaluations for the facilities. 

Chapter 2 of this report, in part, summarizes 

the staff's evaluation of the responses 

to the RAIs on GL 92-01, Revision 1, 

Supplement 1.

NUREG-1511, Supp. 2 1-4
2 GENERIC LETTER 92-01 

REVISION 1, SUPPLEMENT 1: 

REQUESTS FOR ADDITIONAL INFORMATION

2.1 Background 

The publication of Supplement 1 to 

NUREG-1511 (Ref. 20) in October 1996 and 

the issuance of administrative closeout letters 

to licensees in late 1996 and early 1997 

signified the administrative completion of the 

NRC staff's review of the industry's 

responses to GL 92-01, Revision 1, 

Supplement 1 (Ref. 3). At that time, some of 

the owners groups had ongoing activities to 

collect the best estimate chemistry values 

(specifically copper and nickel content data) 

and surveillance data for RPV weld materials. 

This data is important for the estimation of 

the fracture toughness properties of the 

materials, and particularly to the application 

of the chemistry factor ratio procedure 

methodology described in Position 2.1 of 

RG 1.99, Revision 2 (Ref. 21). Since that 

time, the owners groups have completed 

these activities and have submitted a number 

of topical reports to the staff which 

summarize these activities (Refs. 22-26). In 

addition, the NRC staff has conducted an 

inspection of Framatome Technologies, Inc. 

(FTI), to obtain all available RPV weld 

chemistry data for welds fabricated by Babcock and Wilcox (B&W). To ensure that 

licensees have considered the impact of 

these activities on their RPV integrity analyses, the NRC staff issued requests for 

additional information (RAIs) to specific 

licensees during 1998. 

This chapter describes industry and NRC 

activities relative to the nuclear industry's 

collection of best estimate RPV weld 

chemistry and RPV material surveillance 

data, and the issuance of the staff's RAIs. 

The discussion of the RAIs includes a 

description of the content of the RAIs and the

status of the NRC's review of the licensee 

responses to them. 

2.2 Activities Regarding RPV Weld 

Chemistries 

As a part of GL 92-01, Revision 1, 

Supplement 1, the NRC staff requested that 

licensees provide "a description of those 

actions taken or planned to locate all data 

relevant to the determination of RPV 

integrity..." The NRC staff's request was 

a direct result of observations that some 

licensees were not aware of or not using all 

the available chemistry and surveillance 

data applicable to the evaluations of their 

RPV beltline materials, as required by 

10 CFR 50.61 and 10 CFR Part 50, 

Appendices G and H. The staff made this 

observation after noticing that different 

licensees had reported significantly different 

copper and nickel content values (also 

referred to in this report as "chemistry 

values") for RPV welds made from the same 

heat of weld wire. The responses from the 

individual licensees to GL 92-01, Revision 1, Supplement 1, indicated that several owners 

group activities would be initiated to collect, analyze, and, in some cases, evaluate the 

impact of assessing all relevant RPV weld 

chemistry and surveillance data. The staff's 

RAIs were issued in part as a means of 

following up on a particular licensee's 

commitment to assess the impact of these 

owners group activities on the plant-specific 

RPV evaluations for its facility. 

Actions to address GL 92-01 issues were 

undertaken by three of the owners groups: 

B&WOG, CEOG, and BWRVIP. The latter 

is a technical group under the auspices 

of the Boiling Water Reactor Owners Group

NUREG-1511, Supp..2 2-1
(BWROG). These owners group activities 

culminated in the submission of several final 

reports to the staff. Individual licensees then 

referenced these reports in their docketed 

responses to the RAIs. 

The first report received from the CEOG was 

Topical Report CE NPSD-1039, Revision 2 

(Ref. 22). In this report, the CEOG described 

the methodologies used to assimilate all of 

the data on Combustion Engineering (CE) 

weld materials, to evaluate and screen data 

points, and to calculate the best estimate 

copper and nickel contents for each weld wire 

heat (or tandem wire combination) addressed 

in the report. The NRC staff examined the 

CEOG report, generally agreed with the 

approach taken, and developed comments on 

the methodologies proposed by the CEOG. 

These comments served as the basis for the 

staff's presentations in meetings with industry 

owners groups on GL 92-01-related topics 

(Refs. 27 and 28). The staff's comments 

were also formally forwarded to the 

CEOG (Ref. 29). 

The NRC's comments on CE NPSD-1039, 

Revision 2, are briefly discussed below and 

can be grouped into three broad categories: 

(1) the methods proposed for screening 

outliers from the weld wire chemistry 

database, (2) the use and development of 

generic best estimate chemistry values for a 

class of weld materials, and (3) the choice of 

a best estimate computational methodology 

based on the information available for a 

specific weld wire heat. The NRC staff 

emphasized that when proposing to exclude 

outlier data points from best estimate 

chemistry value evaluations, both statistical 

and physical bases should be provided to 

demonstrate that the data points should not 

be included in the evaluation data. On the 

use of generic best estimate chemistry 

values, the NRC staff affirmed its position 

that a single valid data point could serve as 

the basis for the determination of weld wire 

heat specific best estimate chemistry values,

but that when such limited data were cited, 

the NRC staff would consider the impact of 

using the generic chemistry values for the 

appropriate material class. This was done to 

provide confidence that sufficient margins 

existed in the RPV assessments in light of 

the variability reported in RPV weld 

chemistries. Finally, the NRC staff also 

noted that, in general, the use of coil

weighted or group-weighted averages for 

determining the best estimate chemistry 

values for a weld wire heat was preferable to 

the use of simple averages when chemistry 

data from several sources of data for that 

heat existed. A complete description of the 

meaning of simple, coil-weighted, and 

group-weighted averages is provided in 

CE NPSD-1 039, Revision 2. It is sufficient 

to say here that if chemistry data points 

existed from a number of surveillance welds 

for the same weld wire heat: 

"* The simple average best estimate 

composition is calculated by adding up 

each individual data point and dividing 

the sum by the total number data 

points. 

"* The group-weighted average best 

estimate composition is calculated by 

using the average chemistry values for 

each of the surveillance welds and 

averaging the sum of the individual 

average values. 

"* The coil-weighted average best 

estimate composition is calculated by 

using the average chemistry values for 

each of the surveillance welds, 

multiplying them by the number of weld 

wire coils used in each of the welds, 

totaling the resultant values, and 

dividing the sum by the total number of 

coils. 

As a result of NRC staff comments, the 

CEOG reevaluated some of the data 

reported in CE NPSD-1039, Revision 2.

NUREG-1511, Supp. 2 2-2
In July 1998, the CEOG provided a revised 

report to the NRC staff, Topical Report 

CE NPSD-1119, Revision 1 (Ref. 23). 

In this report, the CEOG supplied specific 

answers to comments presented in the NRC 

staff's letter of March 27, 1998 (Ref. 29), 

addressed questions that had been raised by 

the staff in the plant-specific RAIs, and 

provided a revised summary of best estimate 

values for CE RPV weld wire heats (and 

tandem wire combinations). As such, Topical 

Report CE NPSD-1 119, Revision 1, provides 

the final and most complete evaluation of 

weld chemistry values for CE fabricated 

vessels. 

Regarding the work by the B&WOG, initial 

evaluations of Babcock and Wilcox (B&W) 

RPV weld chemistry values were received 

from FTI in June and July of 1997 after the 

NRC had inspected the RPV data available at 

FTI's Lynchburg, Virginia, facilities (Ref. 30). 

These initial evaluations provided the raw 

B&W weld chemistry data and simple 

average best estimate chemistry values for 

B&W fabricated RPV welds. Subsequently, in 

May 1998, the B&WOG submitted a more in

depth report, BAW-2325, Revision 0. The 

B&WOG supplemented this report with the 

submittal of Topical Reports BAW-2325, 

Supplement 1 (Ref. 24), and BAW-2325, 

Revision 1 (Ref. 25), which were issued to 

incorporate comments and recommendations 

from the NRC staff. 

The approach taken by the B&WOG in 

the BAW-2325 reports differed somewhat 

from that taken by the CEOG in the 

CE NPSD-1 039 and CE NPSD-1 119 reports. 

At the November 1997 and February 1998 

meetings between the NRC staff and the 

industry, the staff not only addressed the 

evaluation of best estimate chemistry values, 

as mentioned previously, but also the 

evaluation of Charpy-based RPV surveillance 

data. Therefore, the B&WOG reports differed 

from the CE report in that it also included the

evaluation of surveillance data for B&WOG 

facilities and provided, for each licensee, an 

assessment of the data's impact on the RPV 

integrity assessments for its plant(s). The 

evaluations of the B&W best estimate weld 

chemistry values were based on the use of 

the group-weighted average methodology. 

The NRC staff did not raise any additional 

questions regarding the methodology used 

for evaluating the best estimate weld 

chemistry values cited in the B&WOG report. 

The final owners group assessment of RPV 

weld chemistry values was performed by the 

BWRVIP and submitted in Topical Report 

BWRVIP-46 (Ref. 26). In this report, the 

BWRVIP assessed the impact of new 

copper and nickel chemistry data on the 

RPV integrity assessments for the boiling 

water reactor (BWR) facilities. Since a large 

subset of the BWR vessels were 

manufactured by CE and B&W, much of the 

raw chemistry data had already been 

compiled and assessed in the CE and B&W 

reports. Additional data from the other 

principal BWR RPV manufacturer, Chicago 

Bridge and Iron (CB&I), were also reported 

and evaluated in Topical Report 

BWRVIP-46. 

The approach to assess the impact of the 

newly compiled chemistry data on BWR 

vessel integrity in the BWRVIP-46 report 

was also different from the approaches 

taken by the B&WOG and CEOG. The basis 

for the report was to examine the best 

estimate values of copper and nickel and the 

range of copper and nickel contents for the 

RPV weld material having the highest (most 

limiting) adjusted reference temperature 

(ART), as reported for each BWR RPV. If 

the copper variability for a limiting material 

was less than a characteristic value (i.e., 

0.05 percent, associated with the copper 

variability assumed in the PTS rule basis), 

no additional evaluation was performed. It 

was determined that the reported values for 

materials with less than 0.05 percent copper

2-3 NUREG- 1511, Supp. 2
variability would be sufficient to ensure that 

pressure-temperature (P-T) curves were 

adequate for RPV operation. If the variability 

was greater than 0.05 percent, then the upper 

bound data point from all of the chemistry 

data available for the limiting RPV weld wire 

heat was evaluated to determine whether the 

ART calculated from the use of that data 

point would indicate that the current basis for 

the licensee's P-T limit curves was non

conservative. This approach indicated that 

one insignificant change would occur for a 

single BWR licensee, a 20C (30F) change in 

the ART for Cooper Nuclear Station, from 

39 0C to 41°C (102 0 F to 105 0 F). The NRC 

staff reviewed the BWRVIP-46 report, 

confirmed the acceptability of the approach 

taken therein, and issued a letter (Ref. 31) to 

the BWRVIP accepting the report and closing 

out the staff's review. 

Since the information submitted in these 

reports was supplied by the owners groups 

and thus not directly associated with the 

licensing docket of any specific licensee, the 

NRC staff requested in its RAIs on GL 92-01, 

Revision 1, Supplement 1, that licensees 

review the information in these reports and 

determine its applicability to their facilities. 

Information submitted by the licensees in 

their responses to the RAIs consistently 

referenced the owners groups topical reports 

or the data presented in them. Therefore, the 

staff concluded that, as a result of the staff's 

GL 92-01 initiative and the owners groups 

activities in response to GL 92-01, a 

consistent industry-wide basis for the 

establishment of RPV weld best estimate 

chemistry values had, for the first time, been 

established. 

This represents the first time that a 

comprehensive, integrated assessment of all 

alloying chemistry and surveillance data has 

been completed for the materials that have 

been used to fabricate the RPVs in U.S. light

water nuclear power plants. These efforts 

should minimize surprises regarding the

alloying chemistries for domestic RPV 

beltline materials.

2.3 Activities Regarding 

Surveillance Data

RPV

As a part of GL 92-01, Revision 1, 

Supplement 1, the NRC staff requested that 

licensees provide "a determination of the 

need for use of the ratio procedure in 

accordance with Position 2.1 of RG 1.99, 

Revision 2, for those licensees that use 

surveillance data to provide a basis for the 

RPV integrity evaluation." The ratio 

procedure cited in the request refers to 

adjustments to the measured ARTNDT 

surveillance data to account for variances in 

the chemical compositions (specifically the 

copper and nickel contents) of the 

surveillance weld and the vessel weld. The 

staff included this request in the GL in order 

to assure that, if the best-estimate chemistry 

values for the RPV materials changed as a 

result of the newly reported chemistry and 

surveillance data, and were significantly 

different from the chemistry values for the 

surveillance specimens, the effects of the 

new data would be appropriately accounted 

for in the plant's PTS assessments and P-T 

limit curve calculations. The staff's request 

impacted all plants using surveillance data 

for their weld material evaluations, 

regardless of whether the data was from the 

plant's own surveillance program, an 

integrated surveillance program (in 

accordance with 10 CFR Part 50, Appendix 

H), or the surveillance program from another 

plant. The NRC staff provided an overview 

on use of surveillance data and the ratio 

procedure in public meetings held in 

November 1997 and February 1998. The 

overview also provided several examples of 

specific situations that could occur when 

applying the surveillance data to the PTS 

and P-T limit evaluations. 

In addition to making adjustments to the 

data in accordance with the ratio procedure,

NUREG-1511, Supp. 2 2-4
adjustments may be required to account for 

differences in the irradiation environment, and 

specifically for the difference in the irradiation 

temperatures for the surveillance capsule and 

the RPV inner wall. This temperature adjustment reflects observations that, within 

the temperature range for operation of U.S. 

nuclear power plants, a higher irradiation 

temperature results in reduced embrittlement, 

while a lower irradiation temperature results 

in higher embrittlement. This adjustment is 

made using the down-comer or cold-leg temperature as the reference temperature for 

the vessel wall and the surveillance capsules. The difference in the temperatures for the 

surveillance capsule and inner vessel wall is 

accommodated using a degree-per-degree 

approach, wherein the surveillance data for a 

surveillance capsule irradiated at a 

temperature of X0 above or below the vessel 

wall operating temperature would have X0 

added to or subtracted from each measured 

value of ARTNDT, respectively. 

As indicated in Regulatory Position 2.1 of 

RG 1.99, Revision 2, two or more credible 

surveillance data sets may be used to 

determine the ART values of the RPV beltline 

materials. Surveillance data are deemed 

credible in accordance with criteria described 

in the Discussion section of RG 1.99, Revision 2, and in 10 CFR 50.61 (c)(2)(i). 

10 CFR 50.61 and Regulatory Position 2.1 of 

RG 1.99, Revision 2, specify the process to 

be used to determine the ART with the 

availability of credible surveillance data. This 

position specifies that if there is clear 

evidence that the copper or nickel content of 

the surveillance weld differs from that of the 

vessel weld, the measured values of ARTNDT 

should be adjusted by multiplying them by the 

ratio of the chemistry factor (CF) for the 

vessel weld, as determined from Table 1 of 

the RG, to that for the surveillance weld, as 

determined from the same Table. This is 

indicated below in mathematical form by

Equation 2-1:

Ratio Adjusted ARTNDT = 

Table CVessei Chem x Measured ARTNDT Table CFs-,. Chem.

(2-1)

where Table CFvesseIChem. represents the CF 

for the vessel weld, as determined from 

Table 1 of the RG and based on its copper and nickel contents, and Table CFs5 ,v Chem. 

represents the CF for the surveillance weld, 

as determined from the same Table and 

based on its copper and nickel contents. The 

presumption made in this Regulatory Position is that a commonality exists in the 

weld wire heat number for the vessel and the 

surveillance welds. 

Regulatory Position 2.1 of the RG states the 

need for making adjustments to the 

measured ARTNDT surveillance data because 

there may be considerable variability in the 

copper chemistry contents when multiple welds are fabricated from a given material 

heat. The variability in the copper contents 

may be especially significant if the welds are 

made from copper-coated weld wire, where 

the thickness of copper coating may vary along the length of the wire spool. In 

contrast, except for welds fabricated by CE 

using an additional nickel wire feed, the 

nickel contents for welds fabricated from a 

given material heat do not normally vary as 

significantly as do the copper contents. 

Regulatory Position 2.1 does not specify any need for similar adjustments to the 

measured ARTNDT data for base metals (i.e., plate and forging materials), because base 

metals generally have more homogeneous alloying (chemical) contents than do welds. 

According to methods of Position 2.1 of RG 

1.99, Revision 2, an interim CF value based 

on the surveillance data is determined from 

the measured ARTNDT (base metals) or the 

ratio-adjusted ARTNDT (weld metals, from 

Equation 2-1) according to Equation 2-2:

NUREG-1511, Supp. 2 2-5
n [A x f,(o. 2 -. 01,togfi)] 

CFsu, = n 

[ V. f 0 56 - 0.20 X log i)] 

iY1

(2-2)

where n is the number of surveillance data 

points, A, is the measured ARTNDT (base 

metals) or the ratio-adjusted ARTNDT (weld 

metals), and f, is the fluence for the ith 

surveillance data point. The credibility of the 

CFsu, is determined by comparing the 

residuals of the measured ARTNDT data to the 

predicted ARTNDT data (as calculated using 

the CFsu,). If these residuals meet the 

credibility criteria of the RG, then the CFsu, 

becomes the CF for determining the ARTNDT 

value for the RPV beltline material, and the 

value of 0A used in the calculation of the 

margin term may be reduced by half. 

Five example cases concerning surveillance 

data were described in the public meetings in 

November 1997 and February 1998. The 

processes for evaluating the surveillance data 

in these cases are described below: 

Case 1: Credible surveillance data (weld 

material) only from the plant to be 

evaluated: 

a. Determine the interim CF, CFsur,, 

using as-measured ARTNDT data 

(Equation 2-2); 

b. Evaluate the credibility of CFsurv 

(compare measured ARTNDT to 

predicted ARTNDT using CF.,,, and 

check that the differences are less 

than aJ) - found to be credible in 

this case; 

c. Perform the ratio procedure 

adjustment to the measured 

ARTNDT data (to match the RPV 

weld chemistry); 

d. Determine that no irradiation 

environment (temperature) 

adjustments are required;

e. Reevaluate CFsur from Equation 2-2; 

f. Use CFsur, and reduce aA by half 

to evaluate status of RPV integrity 

per Position 2.1 of RG 1.99, 

Revision 2. 

Case 2: Noncredible surveillance data 

(base metal) from the plant to be 

evaluated and Table CF is 

conservative: 

a. Determine the interim CF, CFsu,, 

using as-measured ARTNDT data 

(Equation 2-2); 

b. Evaluate the credibility of CFs,,,

found to be noncredible in this 

case, as the differences between 

the measured ARTNDT and 

predicted ARTNDT using CFsur, 

exceed o; 

c. Determine that no chemical 

composition or irradiation 

environment (temperature) 

adjustments required; 

d. Compare the CFs•,from Equation 

2-2 to the Table CF - Table CF 

is higher and, therefore, 

conservative; 

e. Use the Table CF and full value of 

cr, to evaluate status of RPV 

integrity, since the surveillance 

data are noncredible and the 

Table CF is conservative. 

Case 3: Noncredible surveillance data 

(base metal) from the plant to be 

evaluated and Table CF is 

nonconservative: 

a. Determine the interim CF, CFs•,, 

using as-measured ARTNDT data 

(Equation 2-2); 

b. Evaluate the credibility of CFsu,

found to be noncredible in this 

case as differences between 

measured ARTNDT and ARTNDT 

predicted using CFsu, exceed oa;

NUREG 2-6 -1511, Supp. 2
c. Determine whether Table CF is 

conservative by comparing 

measured ARTNDT to ARTNDT 

calculated using Table CF and 

check if differences are less than 2 

aa - in this case, the Table CF is 

nonconservative, as some 

differences between measured and 

calculated ARTNDT exceed 2 on; 

d. Determine that no chemical 

composition or irradiation 

environment (temperature) 

adjustments are required; 

e. Use the CFsu,, and full value of oa 

to evaluate status of RPV integrity, 

since the surveillance data are 

noncredible and the Table CF is 

nonconservative. 

Case 4: Surveillance data (weld metal) 

from the plant to be evaluated and 

from other plants (different nuclear 

steam supply system or NSSS 

vendors): 

a. Determine the interim CF, CFsuv, 

using as-measured ARTNDT data 

only from the plant to be evaluated 

(Equation 2-2), since the in-vessel 

plant surveillance data do not 

require a temperature adjustment; 

b. Evaluate the credibility of CFsuv 

found to be credible. Surveillance 

data from other plants were not 

used because they require 

temperature adjustment, while the 

in-plant surveillance data are 

credible; 

c. Perform the ratio procedure 

adjustment to the measured 

ARTNDT data (to match the RPV 

weld chemistry); 

d. Determine that no irradiation 

environment (temperature) 

adjustments required; 

e. Reevaluate the CFsur, as 

determined from Equation 2-2;

f. Use CFsu, and reduce a. by half 

to evaluate status of RPV integrity 

per Position 2.1 of RG 1.99, 

Revision 2. 

Case 5: Surveillance data (weld metal) 

only from plants other than the 

plant to be evaluated (different 

NSSS vendors): 

a. Start with the surveillance data 

from plant(s) with the same 

NSSS vendor as the plant to be 

evaluated, since capsules from a 

plant with the same NSSS vendor 

operate in a similar nuclear 

environment; 

b. Evaluate the interim CF, CFsUN 

(Equation 2-2) using as-measured 

ARTNDT data identified in Step 

a-if the data are from multiple 

heats or plants, then chemical 

composition and irradiation 

environment (temperature) 

adjustments (to the average 

chemical composition and 

irradiation temperature of the 

surveillance data) may be 

required before evaluating CFsu,; 

c. Evaluate credibility of CFs., 

-found to be credible in this 

case. Surveillance data from other 

plants with different NSSS vendor 

were not used because they 

require temperature and other 

adjustments, while surveillance 

data from the same NSSS vendor 

are credible. 

d. Adjust the measured ARTNDT data 

in accordance with the ratio 

procedure to match the RPV weld 

chemistry and adjust the 

irradiation temperature of the 

plant to be evaluated; 

e. Reevaluate the CFsur as 

determined from Equation 2-2 

using the adjusted ARTNDT values 

from Step d, above;

2-7 NUREG-1511, Supp. 2
f. Use CFsurv and reduce a,, by half to 

evaluate status of RPV integrity 

per Position 2.1 of RG 1.99, 

Revision 2. 

These examples cases are an indication of 

the possible situations that may be 

encountered when applying plant-specific 

surveillance data to the structural integrity 

assessments for a plant's RPV. The staff 

recognizes that situations may arise where 

alternative methods of evaluating the 

surveillance data may be necessary. The 

staff will review any alternative methods of 

evaluating the surveillance data on a case-by

case basis. 

In BAW-2325, the B&WOG evaluated all of 

the available surveillance data for the weld 

metals in the B&WOG Reactor Vessel 

Working Group plants, which are plants with 

Westinghouse and B&W NSSS designs 

having B&W fabricated vessels. These welds 

were fabricated with the automatic 

submerged-arc process using copper-plated 

manganese-molybdenum-nickel (Mn-Mo-Ni) 

filler wire and Linde 80 flux.131 As a part of the 

B&WOG evaluation of the surveillance data 

for Linde 80 welds, all of the Charpy-V notch 

data were reevaluated using a hyperbolic 

tangent curve fitting program to achieve 

consistency in the interpretation of the data. 

In response to NRC questions, the B&WOG 

assembled Supplement 1 to BAW-2325, 

providing both the original interpretation of 

the surveillance data (generally from hand-fits 

to the data) and the hyperbolic tangent curve 

fitting interpretation. In the report, the 

B&WOG documented its systematic 

evaluation of the surveillance data for each 

(3) The identifying characteristic of these welds is the heat 

identification of the weld wire used for fabrication of the 

welds. The heat identification indicates the melt of 

material from which the wire was fabricated. All lengths 

(coils) of wire from the same melt material have the 

same heat identification number. Since many coils can 

be made from a single melt, the same heat 

identification number can be associated with the 

fabrication of multiple vessel and surveillance welds.

weld wire heat number, after considering 

many of the concepts described previously 

in the example cases. Revision 1 to 

BAW-2325 provides a final analysis of the 

data for each weld wire heat, after 

considering all corrections to the fitted 

surveillance data as well as appropriate 

adjustments to capsule fluences. 

2.4 Requests for Additional Information 

2.4.1 Contents of the Requests for 

Additional Information 

After reviewing the data collected by the 

owners groups, the staff issued a number of 

requests for information (RAIs) to certain 

pressurized water reactor (PWR) and BWR 

plants. These RAIs were issued from March 

to August 1998. The RAIs included blank 

tables for licensees to update the alloying 

chemistries of the beltline welds (or the 

limiting plate material if applicable), and to 

assess its surveillance data after reviewing 

the appropriate owners group topical report. 

The RAIs also asked the licensees to 

determine how the changes to the best 

estimate chemistry values would impact the 

structural integrity assessments for their 

facility relative to the requirements of 

10 CFR 50.60; 10 CFR Part 50, Appendices 

G and H; and 10 CFR 50.61. 

2.4.2 Status of the Staff's Review of the 

Responses to the RAIs 

All plants that received RAIs submitted their 

responses. The updated data and 

references submitted by licensees as a part 

of the GL 92-01, Revision 1, Supplement 1, 

review were entered into the newly

developed Microsoft Access® version of the 

Reactor Vessel Integrity Database (RVID). 

The RVID is discussed further in Chapter 6 

of this report.

NUREG- 1511, Supp. 2 2-8
As a result of its review of the responses to 

the RAIs on GL 92-01, Revision 1, 

Supplement 1, the staff identified a number of 

issues with the manner in which some of the 

licensees were applying the updated 

surveillance capsule and chemistry data to 

their PTS and/or P-T limit assessments. The 

issues are: 

(1) Several units have surveillance data 

where one or more data points do not 

meet the credibility criteria in RG 1.99, 

Revision 2, and 10 CFR 50.61 (the 

pressurized thermal shock rule). In 

some cases, licensees from the subject 

units used noncredible surveillance data 

to calculate the CFs, and a reduced 

margin term for calculating RTPTs. The 

staff's RAIs on GL 92-01, Revision 1, 

Supplement 1, contained the following 

statement: 

"....10 CFR 50.61(c)(2) specifies 

that licensees will consider plant

specific information (e.g., 

operating temperature and 

surveillance data) to verify that 

the RTNDT for each vessel beltine 

material is a bounding value. 

Regulatory Guide 1.99, Revision 

2, describes two methods for 

determining the amount of margin 

and the chemistry factor used in 

determining RTNDT. If the 

e valuation of the surveillance data 

indicate that the surveillance data 

set is not credible and the 

measured values of ARTNDT are 

less than the projected mean from 

the Tables plus the generic 2Ua, 

the chemistry factor may be 

calculated using either Position 

1.1 or Position 2.1; however, the 

full margin term must be applied.

The method chosen must bound all 

the surveillance data to be in 

compliance with 10 CFR 

50.61(c)(2)." 

Section 2.3 of this report presented 

five example cases as an indication of 

the typical situations that may arise 

when applying plant-specific 

surveillance data to the CF 

calculations. The staff used, where 

applicable, the full value of 0 A in this 

review when non-credible surveillance 

data were used to calculate the CF. 

(2) Two licensees with CE fabricated 

RPVs proposed to use their plant

specific chemistry data in their RPV 

integrity calculations; however, these 

data are nonconservative when 

compared with the CE topical reports. 

The staff used the data from the CE 

topical reports during the review. 

(3) Some licensees submitted fluence 

evaluations in topical reports that were 

reviewed as part of this effort. The 

submittal from one unit did not justify 

the proposed reduction in fluence. 

Although the current docketed fluence 

value was maintained in the RVID, the 

staff recommended that the licensee 

use calculational methods to verify the 

EOL fluence value when the next 

surveillance capsule is removed from 

the vessel and analyzed. 

(4) The response to the RAI on GL 92-01, 

Revision 1, Supplement 1, for a multi 

unit nuclear station did not reflect the 

fluence, RTNDT(U), and Ou values that 

were updated in its recent pressure

temperature (P-T) limits submittal. The 

staff used the information from the P-T 

limits submittal in the RPV integrity 

calculations.

NUREG-1511, Supp. 2 2-9
(5) One licensee was in dispute with the 

CEOG regarding the identity of the 

surveillance weld for its plant. Although 

this discrepancy did not impact the 

current P-T limits for the plant, it did 

affect the chemical composition of the 

subject weld reported by the CEOG. 

The licensee responded to the staff's 

closeout letter and indicated that the 

identity of the surveillance weld had 

been investigated and reconciled with 

the CEOG reports. Therefore this issue 

has been resolved. 

(6) A licensee submitted a revised RTNDT(u) 

value for a beltline weld in one of its 

RPVs; however, the licensee did not 

provide the basis for the change in the 

submittal. The staff informed the 

licensee that it would use the previous 

docketed plant-specific RTNDT(u) value. 

Therefore, the staff did not implement 

the change to the RTNDT(u) value when 

updating the RVID. 

(7) The shell courses of one licensee's 

RPV have differing thicknesses. The 

staff had previously contacted the 

licensee regarding this issue and 

reiterated the discrepancy during the 

GL review effort. This issue was 

documented in the updates to the RVID 

as an aid for reviewing any proposed 

changes to the P-T limit curves 

currently in the plant's Technical 

Specifications.

The staff identified these issues in the 

appropriate RVID reference sections and 

summary sheets during its update of the 

database. The staff also informed each 

licensee associated with these issues of its 

basis for making a change to the 

docketed data and inputting the amended 

data into the updated database. 

2.4.3 Closeout of Generic Letter 92-01, 

Revision 1, Supplement 1 

Each licensee received a letter to close out 

GL 92-01, Revision 1, Supplement 1. These 

letters were issued to licensees in 1999. The 

letters recommended that licensees review 

the RPV integrity data for their plants 

provided in the revised version of -RVID. 

Licensees may do so by downloading the 

RVID installation diskettes from a web site 

that is linked to the NRC homepage 

(http://www.nrc.gov/NRR/RVID/index.html). 

It should be noted that the website for the 

original version of the RVID was not linked to 

the NRC homepage. The closeout letters 

also informed licensees that the staff would 

assume that data entered into the RVID are 

acceptable if no comments were received by 

September 1, 1999. The closeout letters 

concluded by stating that future submittals 

on P-T limits, pressurized thermal shock 

(PTS), or upper shelf energy (USE) should 

reference the most current RPV integrity 

information.

NUREG-1511, Supp. 2 2-10
3 SIGNIFICANT REGULATORY AND RESEARCH 

ACTIVITIES ON RPV INTEGRITY

3.1 Background 

Since the issuance of NUREG-1511, 

Supplement 1 (Ref. 20), in October 1996, 

several developments have occurred that are 

expected to significantly affect the technical 

assessment of RPV integrity issues. This 

chapter will discuss these developments 

under two broad categories: (1) research 

activities to advance the understanding of 

RPV integrity parameters; and (2) consensus 

Codes and Standards activities to formalize 

alternatives to the established RPV integrity 

assessment methodologies. 

3.2 Research Activities 

Four topics will be discussed in the area of 

research activities that have advanced the 

understanding of RPV integrity parameters. 

The first is the continuing work by NRC staff, 

NRC contractors, and industry 

representatives (under the auspices of 

American Society for Testing and Materials 

(ASTM) Committee El 0) on the development 

of new embrittlement correlations for RPV 

materials. The NRC's Office of Nuclear 

Regulatory Research (RES) began this work 

in 1993 in conjunction with efforts by ASTM 

Committee E-10 in order to determine 

whether the existing database of commercial 

power reactor surveillance data (which 

currently consists of about 720 surveillance 

data points) could be used to improve upon 

the embrittlement correlations stated in US 

NRC Regulatory Guide (RG) 1.99, Revision 2 

(which were established based on a database 

of about 177 data points). This work may 

provide the basis for revising both RG 1.99 

and ASTM Standard E900 (Ref. 32). As part 

of the effort to revise these documents, the 

work on the new embrittlement correlations 

has also expanded to include an assessment

of the margin or uncertainty to be applied 

when using the new correlations and to 

develop a methodology for the assessment 

and use of plant-specific surveillance data. 

The initial analysis of the data conducted by 

an NRC contractor has been documented in 

Topical Report NUREG/CR-6551 (Ref. 33). 

Industry representatives on ASTM Committee 

El0 have supported this work by providing 

additional quality assurance reviews of the 

surveillance data in the NRC's Power Reactor 

Embrittlement Data Base and correcting or 

filling in missing data. The ASTM committee 

members have also reviewed the models 

developed in NUREG/CR-6551, and 

questions raised within the committee have 

led to the need for some reanalysis of the 

NUREG/CR-6551 models. Work on refining 

the models, completing the uncertainty or 

margins analyses, and assessing the 

surveillance data is expected to continue at 

least through fall 2000. 

The second topic has been the research 

performed to develop new RPV flaw 

distributions. Probabilistic fracture mechanics 

(PFM) evaluations of reactor pressure vessel 

integrity require characteristics of assumed 

flaws as one of the key input parameters. In 

lieu of deterministic applications of a fixed 

flaw size and geometry, flaw density and flaw 

size distributions are used to characterize a 

population of flaws for use in the PFM 

evaluations. For many years, the "Marshall 

distribution" (Ref. 34) has provided the basis 

for PFM calculations. Two recent initiatives 

by RES are focused on improving existing 

flaw density and flaw size distribution data. In 

the first initiative, researchers at the Pacific 

Northwest National Laboratories (PNNL) have 

been using state-of-the-art ultrasonic testing 

(UT) inspection equipment and evaluation 

tools to inspect the welds in vessels from

3-1 NUREG-1511, Supp. 2
decommissioned nuclear plants. These 

nondestructive evaluations have been 

accompanied by destructive confirmation of 

the indications identified by the UT 

inspections. Preliminary results of this work 

are provided in References 35 and 36. 

The second initiative refers to an expert 

judgement process that is used to classically 

resolve specific technical issues for which 

there is scientific uncertainty. In this case, 

the expert judgement process has been 

employed to review, interpret, and 

supplement available information on RPV 

fabrication processes and RPV flaw density 

and flaw size distributions. RES expects to 

issue a NUREG report on this work by the 

end of 2001. 

The third topic actually transcends both of the 

broad categories laid out at the beginning of 

this chapter, that is development of the 

Master Curve methodology. The Master 

Curve data methodology relies on the 

similarity of the fracture toughness

temperature transition behavior for ferritic 

steels, and provides a framework for enabling 

structural integrity assessments based on 

limited fracture toughness data sets from 

small test specimens (e.g., precracked 

Charpy-V specimens). This technology was 

originally developed at Valtion Teknillinen 

Tutimuskeskus (VTT, a Finnish Research 

Laboratory) (Ref. 37), and substantial funding 

for research in the United States has been 

provided by NRC's RES. A significant amount 

of additional research into the Master Curve 

methodology has also been supported by 

WOG, BWOG, CEOG, and EPRI. 

The final topic under research activities is a 

program initiated by RES to reevaluate the 

PTS screening criteria in 10 CFR 50.61. This 

program will combine the information from 

previous NRC work in the 1980s (i.e., the 

Integrated Pressurized Thermal Shock 

program) with improved flaw distributions, 

new embrittlement correlations, and a better

understanding of topics such as RPV weld 

chemistry variability. Issues regarding the 

identification of potential PTS transients, 

thermo-hydraulic calculations, and the 

development of acceptable risk metrics will 

also be addressed. This program has also 

been developed to include substantial 

participation by the industry and the public 

throughout the process. 

3.3 Codes and Standards Activities 

Several actions have also been taken by 

consensus Codes and Standards bodies to 

provide a framework for use of the Master 

Curve methodology. ASTM Standard E1921 

(Ref. 38) was passed in 1997 and provides 

methods for conducting Master Curve testing 

and data analysis. Subsequently, ASME 

Code Case N-629 (Ref. 11), which was 

approved in 1998, defines a relationship 

between the Master Curve test parameter (TO, 

a temperature defined to correspond to a 

fracture toughness of 100 MPafm) and the 

ASME Code methodology for indexing 

fracture toughness properties to a reference 

temperature. In this Code Case, a reference 

temperature, RTTO, as given by RTTO = To + 

35 0 F, is provided as an alternative for 

indexing temperature to RTNDT. During 

summer 1999, one licensee (Wisconsin 

Public Service Corporation, the licensee for 

the Kewaunee Plant) submitted a license 

amendment and requested an exemption for 

NRC approval to apply Code Case N-629 

and the Master Curve methodology to the 

analysis of their limiting RPV circumferential 

beltline weld (Ref. 39). The NRC's review of 

this submittal is expected to be completed by 

the end of Year 2000. 

In the area of consensus Codes and 

Standards activities, two additional ASME 

Code Cases have been passed that are of 

interest to RPV integrity evaluations. ASME 

Code Case N-588 (Ref. 10) was developed 

to permit licensees to postulate a 

circumferentially-oriented flaw when

NUREG- 1511, Supp. 2 3-2
evaluating a circumferential weld for RPV P-T 

limits. Previously, in the latest edition (1995, 

Edition through Summer 1996 Addenda) of 

Appendix G to the ASME Code, which has 

been invoked by reference in 10 CFR 50, 

Appendix G, licensees were required to 

postulate axially-oriented flaws in all RPV 

beltline materials when evaluating RPV P-T 

limits. Based on the metallurgical processes associated with welding, large axially-oriented 

flaws perpendicular to a circumferential weld 

seam are extremely unlikely. In fact, the 

ASME Code reference flaw (/ wall thickness 

in depth with a 6:1 aspect ratio) would extend 

not only across the width of a nominal RPV 

circumferential shell weld but into the base 

material beyond, making the axial flaw 

assumption even more non-physical for 

circumferential shell welds. 

The other ASME code case of interest is 

N-640 (Ref. 9, formerly listed as N-626), which permits the use of the lower bound 

static initiation fracture toughness curve (K1J 

instead of the lower bound crack arrest 

fracture toughness curve (Ka) for developing 

RPV P-T limits. The Kla fracture toughness 

curve, being based on dynamic and crack

arrest data, has provided a conservative 

basis in Appendix G to the ASME Code for 

establishing P-T limits. With the approval of 

this Code Case, the ASME Code concluded 

that the use of the Kic fracture toughness 

curve is technically justified, since the intent 

of the ASME Code procedures and the NRC 

regulations has been based on preventing 

crack initiation. 

10 CFR Part 50, Appendix G, requires that 

the P-T limits for nuclear power generation 

facilities must be at least as conservative as 

those that would generated if Appendix G to 

the ASME Code were used to establish the 

curves. Since the methods of Code Cases 

N-640 and N-588 will generate P-T limit 

curves that are less conservative than those 

that would generated using Appendix G to 

the ASME Code, licensees must request 

exemptions under the regulatory framework 

of 10 CFR 50.60(b) and 10 CFR 50.12 if they 

desire to generate their P-T limit curves using the Code Case methods. The staff will 

review applications for exemptions to use 

Code Case N-640 or Code Case N-588 on 

a case-by-case basis.

NUREG-1511, Supp. 2 3-3
4 RECENT INDUSTRY EFFORTS REGARDING BWR RPV 

WELD INSPECTION REDUCTION (BWRVIP-05)

4.1 Overview 

In 1995 the BWR Vessel and Internals 

Project (BWRVIP), a special industry 

technical review group formed by the 

BWROG to focus on resolution of reactor 

vessel and internals degradation issues, 

submitted a proprietary report, "BWR Vessel 

and Internals Project, BWR Reactor Pressure 

Vessel Shell Weld Inspection 

Recommendations (BWRVIlP-05)" (Ref. 13), 

for staff review and approval. The 

BWRVIP-05 report evaluated the current 

inspection requirements for the RPV shell 

welds in BWRs, formulated recommendations 

for alternatives to the inspection 

requirements, and provided a technical basis 

for the recommended alternatives. In the 

report, and as revised in subsequent 

submittals, the BWRVIP proposed to reduce 

the scope for augmented inspections of the 

circumferential RPV shell welds from "essentially 100 percent" of the welds (i.e., as 

defined in 10 CFR 50.55a(g)(6)(ii)(A)(2)) to 

essentially zero percent of the welds, except 

for the portions of the welds located at 

intersections of the axial and circumferential 

welds, where approximately 2-3 percent of 

the circumferential welds would be inspected. 

Revised criteria for the performance of 

successive and additional inspections were 

also recommended. 

4.2 Background 

On May 12,1997 (Ref. 40), the BWRVIP and 

NRC staff briefed the Commission on issues 

related to the requirements for a full 

inspection of reactor pressure vessel shell 

welds. The transcript of the Commission 

meeting of May 12, 1997, the Commission's 

Staff Requirements Memorandum (Ref. 41), 

and the meeting summaries related to this

issue are available in the Commission's 

Public Document Room, 2120 L Street NW., 

Washington, DC 20555. 

On August 7, 1997, the staff issued 

Information Notice (IN) 97-63, "Status 

of NRC Staff's Review of BWRVIP-05" 

(Ref. 42), regarding licensee requests for 

relief from the augmented inspection 

requirements of 10 CFR 50.55a(g)(6)(ii)(A). 

In the information notice, the staff stated that 

it would consider technically-justified requests 

for reliefs from the augmented examination 

requirements in accordance with the 

provisions for alternative programs stated in 

sections (a)(3)(i)and(ii) and (g)(6)(ii)(A)(5) to 

10 CFR 50.55a. Acceptably-justified requests 

were to be considered from BWR licensees 

who were scheduled to perform inspections 

of the RPV shell welds during the fall 1997 or 

spring 1998 outage seasons. In the 

information notice, the staff stated it would 

consider inspection delays of up to two 

operating cycles for the RPV circumferential 

shell welds only. The staff also stated that 

licensees will still need to perform the 

required augmented inspections of "lessentially 100 percent" of all axial RPV shell 

welds. 

The acceptability of such requests was based 

on plant-specific information submitted by the 

licensee. The staff granted schedular reliefs 

to defer the inspections of the RPV 

circumferential shell welds for four BWR units 

scheduled to enter refueling outages (RFOs) 

during the fall 1997 outage season, and for 

two BWR units scheduled to enter RFOs 

during the spring 1998 outage season. 

On August 14, 1997, the staff forwarded to 

BWRVIP its independent safety assessment 

(ISA) of the BWRVIP-05 document (Ref. 43). 

The staff's ISA was a multi-disciplinary, risk4-1 NUREG-1511, Supp. 2
informed review of the safety implications of 

reducing the inspection scope for the RPV 

circumferential shell welds as proposed in the 

BWRVIP-05 report. It provided a description 

of the two degradation mechanisms (fatigue 

and stress corrosion) that have the potential 

to initiate RPV cracking or to promote the 

growth of existing flaws, and of the limiting 

transients of concern. Also transmitted with 

the ISA was additional guidance on the type 

of information the staff would need to assess 

plant-specific requests for relief from the ISI 

requirements of 10 CFR 50.55a(g)(6)(ii)(A). 

Further work was performed by both the staff 

and the industry to assess more fully the risk 

associated with beyond-design-basis events 

for both the axial and circumferential welds at 

fluence levels projected to be reached later in 

life at some plants. This additional work 

included: (1) studies of potential precursor 

events to better quantify the potential for cold 

overpressure events in BWRs, (2) additional 

probabilistic fracture mechanics analyses 

both to understand the sensitivities to various 

parameters and to support an uncertainty 

analysis, and (3) an assessment of the 

proposed changes in inspection requirements 

relative to the probability of vessel failure. 

On May 7, 1998, the staff issued IN 97-63, 

Supplement 1 (Ref. 44), which informed BWR 

licensees that the staff was extending the 

applicable periods to the fall 1998 and spring 

1999 outage seasons. 

The staff concluded in its ISA that beyond

design-basis events occurring during plant 

shutdown (e.g., injection of cold water into the 

RPV at pressure or excessive pressurization 

of the cold vessel) could lead to cold 

overpressurization events that could 

challenge vessel integrity. Specifically, the 

staff identified a transient at a foreign BWR of 

U.S. design, in which the RPV was subjected 

to high pressure (7.9 MPa or 1150 psig) 

while at a low temperature (26 0C-310C 

or 79 0F-880 F). This cold overpressure

transient was not included as a design basis 

event for BWRs and was not considered in 

the BWRVIP-05 report, which focused only 

on design-basis events. However, the 

recognition of this transient led the staff to 

conclude that cold overpressure transients 

are safety significant and need to be 

considered. Accounting for these precursor 

and actual events, the staff estimated a 

frequency of cold overpressurization events 

that could challenge the RPV integrity at cold 

shutdown. 

The industry's response to the staff's ISA 

concluded that condensate and control rod 

drive (CRD) pumps could cause conditions 

that could lead to cold overpressure events 

that could challenge vessel integrity. 

Specifically, the industry gave the following 

justifications as the bases for concluding that 

most of the other BWR injection or spray 

systems would not contribute to the overall 

frequency for the occurrence of cold 

overpressurization events: 

1. The shutoff head to low pressure coolant 

spray or injection systems is low (e.g., 

shutoff heads to the LPCS/LPCI pumps), 

so that the RPV remains within the 

acceptable limits of the pressure

temperature (P-T) curves even at 

shutdown temperatures (i.e., these 

systems can be activated only under low 

pressure). 

2. Overfilling and pressurization to the 

shutoff head for high pressure coolant 

spray (HPCS) systems is very unlikely 

because the system automatically trips on 

the high-water level indications. 

3. The reactor core isolation cooling (RCIC) 

and high pressure coolant injection 

(HPCI) systems are steam driven and are 

not in use during cold shutdown 

conditions.

NUREG-1511, Supp. 2 4-2
4. Operation of the standby liquid control 

(SLC) system requires a series of 

deliberate operator actions (e.g., manual 

pump activation) and is unlikely to occur 

without adequate monitoring. 

Thus, RCIC, HPCI, HPCS, LPCI, LPCS, and 

the SLC systems were considered to have a 

negligible impact on the frequency for a cold 

overpressurization event. The BWRVIP's 

estimate of the frequency of 

overpressurization events that could 

challenge the RPV was 9.5 x 10-4/yr for 

BWR-4 designs and 9 x 10'/yr for BWR 

designs other than BWR-4. This frequency 

estimate is comprised of 1.5 x 104/yr from 

condensate injection, 7 x 10 4/yr from CRD 

injection (vessel pressure testing), and 

1 x 104/yr from loss of RWCU (3 x 10 5

/yr 

from loss of RWCU in BWR designs other 

than BWR-4). The staff considered the 

frequency contributions from these systems 

to be reasonable. 

The staff noted, however, that the BWRVIP 

concluded that the potential contribution from 

other injection sources (i.e., RCIC, HPCI, 

HPCI, HPCS, LPCI, LPCS, and SLC 

systems) would have a negligible impact on 

the frequency for the occurrence of a cold 

overpressurization event. However, because 

historical data indicated that actual 

inadvertent injections of these systems have 

occurred, the staff did not consider their 

contribution to the cold overpressurization 

frequency to be negligible. Therefore, the 

staff concluded that the NRC-estimated 

frequency for cold overpressurization 

resulting from inadvertent injections of these 

systems (6 x 1 0 5/yr) should be added to 

BWRVI P's estimate of about 9 x 1 0-/yr. The 

staff therefore estimated the total frequency 

for cold overpressurization to be 1 x 1 0 3

/yr. 

4.3 NRC Assessment of BWRVIP-05 

To estimate the conditional failure probability

(P(FIE)) for the BWR RPV vessel, the staff 

performed independent generic and plant

specific probabilistic fracture mechanics 

(PFM) analyses using the FAVOR Code. The 

FAVOR Code, which was developed by the 

Oak Ridge National Laboratory (ORNL), 

performs millions of random deterministic 

vessel simulations to determine the mean 

P(FIE) for a vessel subjected to a specific 

transient. To perform the independent PFM 

analyses, the staff used the following data as 

inputs to the simulations: (1) the operational 

data from the referenced foreign transient, 

(2) updated RPV flaw density and size 

distributions based on data developed at the 

NRC's Pressure Vessel Research Users 

Facility (PVRUF), and (3) the material 

chemistries and fluences developed by 

different vessel fabricators for the generic 

RPV evaluations and the material chemistries 

and fluences of bounding RPVs for the plant

specific evaluations. 

The BWRVIP failure frequency for the limiting 

circumferential welds was 9.0 x 10-1° /yr 

((9.0 x 10 4

/yr event frequency for a BWR-3) 

x (1.0 x 10-6 conditional probability of failure)). 

In contrast, the staff determined the limiting 

plant-specific failure frequency for 

circumferential welds at 32 effective full 

power years to be 8.2 x 10-8/yr ((1 x 10 3/yr 

event frequency) x (8.2 x 10' conditional 

probability of failure)). As depicted in 

NUREG-1560, Vol. I (Ref. 45), core damage 

frequencies (CDF) for BWR plants were 

reported to be approximately 10 7

/yr to 10 4

/yr. 

In addition, Regulatory Guide (RG) 1.154 

(Ref. 46) indicates that PWR plants are 

acceptable for operation if the plant-specific 

analyses predict that the mean frequency of 

through-wall crack penetration for pressurized 

thermal shock events is less than 5 x 1 0-/yr. 

Since the failure frequencies for 

circumferential welds in BWR RPVs were 

significantly lower than the failure probabilities 

specified in RG 1.154 and NUREG-1560, 

Vol. I, and since additional volumetric 

inspections would not significantly improve

NUREG-1511, Supp. 2 4-3
upon the already acceptably low failure 

probabilities for circumferential welds, the 

staff concluded that the proposal to eliminate 

the ISI requirements for RPV circumferential 

welds was justified. Therefore, in its final 

safety evaluation on BWRVIP-05 (Ref. 14), 

the staff informed the BWRVIP that the 

proposal to eliminate BWR vessel 

circumferential weld examinations is 

acceptable. 

On November 11, 1998, the staff issued 

GL 98-05 (Ref. 47) to inform licensees of 

BWRsthat they may request permanent 

relief from meeting the inservice inspection 

requirements of 10 CFR 50.55a(g)(6)(ii)(A) 

for circumferential RPV shell welds during the 

remainder of the current 40-year license 

terms for their facilities. In the GL, the staff 

stated that licensees could justify using 

the BWRVIP-05 report as the basis 

for reducing the inspection scope by 

(1) demonstrating that the circumferential 

RPV shell welds would continue to satisfy the 

limiting conditional failure probability for 

circumferential welds at EOL as cited in the 

staff's safety evaluation of July 28, 1998; and 

(2) demonstrating that they will have 

implemented operator training and 

established operating procedures to ensure 

that the probability for the occurrence of a 

cold overpressurization event would be 

limited to the frequency specified in the staff's 

safety evaluation of July 28, 1998. In the GL 

the staff also informed the licensees that they 

would still need to perform the required 

inspections of "essentially 100 percent" of all 

axial RPV shell welds. 

Unlike circumferential flaws, the failure 

frequencies estimated by the staff (Ref. 14) 

for BWR axial welds were relatively high, 

about 4.4 x 10 4/yr. Conservatisms in the 

analyses, such as using the limiting material

properties and chemistry for the inside 

surface flaws in the axial welds at the location 

of peak end-of-license (EOL) azimuthal 

fluence, excluded the RPV axial welds as a 

near-term safety concern. Nonetheless, to 

resolve the long-term potential safety 

concern, the staff pursued this subject with 

the BWRVIP to ensure that the estimated 

failure frequency for BWR axial welds is 

significantly lower than indicated by the staff's 

estimate for axial RPV shell welds. In May 

1999, the NRC staff met with the BWRVIP to 

discuss revised assumptions and criteria for 

the axial weld failure probability analyses. 

The BWRVIP submitted a revised 

assessment for BWR axial welds on 

November 12,1999 (Ref. 48). By letter dated 

March 7, 2000 (Ref. 49), the staff issued its 

evaluation of the revised assessment. In this 

letter, the staff concluded that the RPV failure 

frequencies for the limiting axial welds in the 

BWR fleet are below the applicable regulatory 

limits, given the assumptions used in the 

evaluation. However, the staff also noted that 

this conclusion applied only to initial 40-year 

license periods, and that consideration of 

BWR vessel welds for license renewal terms 

would require plant-specific assessments by 

the license renewal applicants. 

In the letter of March 7, 2000, the staff also 

noted that an expert judgement process has 

been employed to review, interpret, and 

supplement available information on RPV 

fabrication processes and RPV flaw density 

and flaw size distributions. The staff 

concluded that, should the results of the 

expert judgement process prove to be more 

conservative than the assumptions used in 

the BWRVIP's assessment (Ref. 48), the 

BWRVIP would be required to re-evaluate 

BWR axial welds using the results of the 

expert judgement process.

NUREG- 1511, Supp. 2 4-4
5 REACTOR PRESSURE VESSEL THERMAL ANNEALING

5.1 Thermal Annealing Regulation 

Thermal annealing is a process which is used 

to mitigate the effects of radiation 

embrittlement on RPV materials. The NRC 

has established a regulation, 10 CFR 50.66 

(Ref. 16), to provide the regulatory framework 

for the thermal annealing of RPVs. The 

regulation addresses the critical engineering 

and metallurgical aspects of thermal 

annealing and requires the following actions: 

(1) submittal of an engineering plan and 

analysis for thermal annealing in a thermal 

annealing report (TAR); (2) submittal of a 

report detailing the estimated degree of 

fracture toughness recovery that will be 

accomplished before implementing the 

annealing process; (3) submittal of a report 

confirming that the thermal annealing process 

was performed in accordance with the TAR; 

and (4) public meetings to be held both 

before and after the process is implemented 

to allow interested parties to make inquiries. 

5.2 Regulatory Guides and Technical 

Codes and Standards for Thermal 

Annealing 

5.2.1 Regulatory Guide 1.162 

RG 1.162 (Flef. 17) was issued by the NRC to 

provide guidance for the thermal annealing of 

RPVs. The RG contains a detailed listing of 

metallurgical and engineering issues that 

should be addressed in an application to 

implement a thermal annealing process. 

RG 1.162 presents three acceptable methods 

for estimating the degree of fracture 

toughness recovery resulting from a thermal 

annealing process: 

testing of RPV surveillance program 

materials;

"* removal of specimens from the RPV 

beltline; 

"* a generic computational method. 

5.2.2 ASTM Standard Procedure E-509 

General guidance for in-service annealing is 

given in ASTM Standard E509-86 (Ref. 50). 

Specifically, ASTM Standard E509-86 

prescribes general procedures for conducting 

an inservice thermal annealing of a RPV and 

for demonstrating its effectiveness and the 

degree of recovery in fracture toughness.

5.2.3 ASME Code Case 

Thermal Annealing

N-557 on

ASME Code Case N-557 (Ref. 51) was 

developed in 1995 by a special ASME task 

group to provide guidance specifically 

focused on the structural engineering aspects 

of thermal annealing. The Code Case 

addresses how annealing conditions 

(temperature and duration), temperature 

monitoring, evaluation of loadings, and 

nondestructive examination will be applied to 

the thermal annealing process. Code Case 

N-557 was formally approved by ASME in 

March 1996. 

5.3 Palisades Thermal Annealing Report 

In October 1995, CPCo (currently named 

Consumers Energy), the licensee for the 

Palisades plant, submitted a TAR for 

annealing of the RPV at the Palisades 

Nuclear Plant (Ref. 52). CPCo submitted the 

TAR after determining that the materials in 

the vessel might not satisfy the screening 

criteria of the PTS rule (10 CFR 50.61) 

through the end of license (EOL) for the unit. 

CPCo projected that an annealing treatment 

of the Palisades RPV would result in recovery

NUREG-1511, Supp. 2 5-1
of 80 to 90 percent of the fracture toughness 

loss due to neutron embrittlement. 

The staff completed a preliminary review of 

the Palisades TAR and requested additional 

information from CPCo. CPCo responded to 

the staff's request via a letter dated 

August 26, 1996 (Ref. 53). In April 1997, 

CPCo withdrew the TAR from the docket 

(Ref. 54), and instead elected to pursue other 

means of addressing the PTS issue for the 

Palisades RPV. 

On April 4, 1996, CPCo submitted revised 

neutron fluence estimates and PTS 

calculations for the Palisades RPV materials 

(Ref. 55). CPCo's revised PTS calculations 

showed that the Palisades RPV would 

remain below the PTS screening criteria 

of 10 CFR 50.61 through the end of 2007. 

The NRC staff reviewed the CPCo analysis 

and agreed with some aspects of their 

proposed fluence reduction while continuing 

its review of other aspects of CPCo's 

fluence methodology. On December 20,1996 

(Ref. 56), the NRC staff issued an interim 

safety evaluation that concluded that the 

Palisades RPV would remain below the PTS 

screening criteria through the end of 2003. 

Subsequently, the staff determined that no 

additional fluence reduction could be credited 

based on the original CPCo analysis 

methodology. However, CPCo has continued 

to refine their neutron fluence estimates and 

PTS calculations in an attempt to justify 

continued safe operation of the Palisades 

RPV through the end of its original operating 

license (2007). Chapter 7 of this report 

further summarizes both CPCo's and the 

staff's efforts to re-evaluate the neutron 

fluences for the Palisades RPV. 

5.4 Department of Energy/Industry 

Annealing Demonstration Project 

The Department of Energy (DOE) and the 

nuclear industry have sponsored engineering 

projects for demonstrating the feasibility of

thermal annealing in U.S. light water reactors. 

The contracts for two of these Annealing 

Demonstration Projects (ADPs) were 

announced on May 25, 1995. The Office of 

Nuclear Regulatory Research (RES) has 

represented the NRC's interests in these 

ADPs, which was conveyed to DOE in a 

memorandum of understanding (MOU) dated 

August 4, 1996, (Ref. 57). The feasibility of 

thermal annealing was demonstrated by a 

joint DOE/industry-sponsored ADP at the 

Marble Hill facility. This ADP employed an 

indirect gas-fired heating method. A second 

ADP using an electric resistance heating 

approach was planned for the Midland facility 

but subsequently canceled due to a lack of 

DOE funding. The Marble Hill and Midland 

facilities were decommissioned 

Westinghouse and B&W designed plants, 

respectively. 

An NRC Commission briefing was held on 

August 27, 1996, to provide the Commission 

with an update of the status of the ADPs and 

their relevance to the Palisades plant. DOE 

and the consortium associated with the 

Marble Hill ADP provided a summary of the 

activities that had occurred during the ADP 

(May 30, 1996 to July 19, 1996). The NRC 

staff, led by RES, also presented their 

observations on the Marble Hill ADP. 

Specifically, the NRC staff discussed the 

status of the plans being prepared by CPCo 

for the Palisades plant and how the 

information gathered from the ADPs would be 

used to support the Palisades TAR and the 

NRC's review of the annealing activities. 

However, as noted previously, the planned 

annealing for the Palisades RPV was 

ultimately terminated. 

Verification of the data and completion of a 

final report on the Marble Hill ADP were 

delayed because of the elimination of funds 

to support the projects beyond the end of the 

DOE fiscal year 1997. EPRI later provided 

sufficient funding to complete the Marble Hill 

report. EPRI published the final report on the 

Marble Hill ADP in March 1998 (Ref. 58).

NUREG-1511, Supp. 2 5-2
The NRC published its final report, NUREG/CR-6552, "Marble Hill Annealing Demonstration Evaluation," in February 1998 

(Ref. 59). Based on the review of the 

data provided to NRC following the annealing demonstration and the results of the thermal 

and structural analyses, it was concluded 

that the Marble Hill RPV was not adversely affected by the demonstration annealing

cycle. Although the ADP was successful, there was a concern with the unpredicted 

severity of the thermal gradient in the RPV 

between the nozzles and the RPV flange. Therefore, in NUREG/CR-6552, the staff 

recommended that reliable analytical models 

be developed and validated to address this 

concern.

NUREG-1511, Supp. 2 5-3
6 REACTOR VESSEL INTEGRITY DATABASE

6.1 Development of the Reactor Vessel 

Integrity Database 

The Reactor Vessel Integrity Database 

(RVID) was developed following NRC staff 

review of licensee responses to GL 92-01, 

Revision 1 (Ref. 1). The database was 

designed and developed to reflect the current 

status of reactor pressure vessel integrity, 

with the data consolidated in a convenient 

and accessible manner. Some of the data 

categories represent inputs of docketed 

information; other data categories are 

representative of computed values that are 

not necessarily docketed. The programming 

logic used for calculations in the database 

follows the methodology in NRC RG 1.99, 

Revision 2 (Ref. 21). 

RVID summarizes the properties of the 

reactor vessel beltline materials for each 

operating commercial nuclear power plant. 

For plants that ceased operation since 

issuance of the initial version of the RVID, the 

existing RPV data for those plants have been 

maintained within the RVID. 

6.2 Availability of the RVID on the World 

Wide Web 

The original RVID was a DOS application 

developed with FoxProTM software. The RVID 

was updated with new data and references 

resulting from the staff's review of the 

nuclear industry's responses to GL 92-01, 

Revision 1, Supplement 1 (Ref. 3). The staff 

released the database in June 1999 as RVID 

Version 2 (RVID 2), a Windows 3.1 native 

application based on Microsoft Access 2.OTM. 

The database runs in the higher Windows 

environments as well (e.g., Windows 95, 

Windows NT). The user is not required to 

have Microsoft AccessTM to run the database.

In addition, comments from licensees 

on the June 1999 data and information 

were used to update the RVID, and RVID 

Version 2.0.1 was released in July 2000. The 

four (4) RVID 2 diskettes can be downloaded 

from the RVID 2 homepage 

(http://www.nrc.gov/NRR/RVID/index.html). 

RVID 2 is also linked to the NRC World Wide 

Web homepage under "nuclear reactors" 

and the "U.S. Commercial Nuclear Plants" 

heading. 

6.3 Content of the RVID 

In addition to the licensee responses to 

GL 92-01, Revision 1, Supplement 1, and 

GL 92-01, Revision 1, the following 

documents were included in the review 

process and development of the RVID 2 

program: surveillance capsule reports; 

pressurized thermal shock and P-T limits 

submittals (as applicable); documents 

referenced in licensee-specific responses to 

GL 92-01, Revision 1 and GL 92-01, 

Revision 1, Supplement 1; and responses to 

the staff's requests for additional information 

(as applicable). RVID 2 has three tables for 

each plant: a PTS summary table for PWRs 

or a P-T limits summary table for BWRs, an 

USE summary table, and a surveillance data 

summary table. The surveillance data 

summary table is a new feature of the 

database. Additionally, RVID 2 includes sort 

and data search capabilities. The user can 

select a desired grouping of plants and then 

specify information categories to search and 

list. 

The RVID 2 program has references and 

notes that document the source(s) of data 

and provide supplemental information. As a 

result of comments received from the 

industry, the staff revised RVID 2 to include

NUREG-1511, Supp. 2 6-1
reference fields at the component level that 

highlight notes specific to the particular 

forging, plate, or weld. RVID 2 has the 

capability to apply the RG 1.99, Revision 2, 

ratio procedure to the calculation of the 

chemistry factor. The database can also 

automatically determine the credibility of 

surveillance data in accordance with 

the criteria in RG 1.99, Revision 2, and 

10 CFR 50.61 (Ref. 5). 

Significant effort has been made to ensure 

that the RVID 2 program is "user friendly" and 

that the programming logic is accurate. The 

staff used the information in BAW-2325, 

Revision 1, "Response to Request for 

Additional Information (RAI) Regarding 

Reactor Pressure Vessel Integrity" (Ref. 24), 

to highlight the relationship between the data 

in the B&W integrated surveillance program 

and the vessels to which they apply. Specific 

notes are included in the surveillance data 

and the component level note fields to relate 

the surveillance data to the applicable RPV 

materials. 

6.4 Quality Assurance of RVID Data 

The staff reviewed the docketed information

for each plant, and provided paper-markups 

of data to designated RVID editors. The two 

editors used the paper-markups to update the 

RVID, and consulted with each reviewer 

when necessary. Each editor performed 

quality assurance (QA) checks of the data in 

preparation for and after the update process. 

In addition, the editors required each original 

reviewer to QA the data before it was 

released on the RVID web site. 

Subsequently, in the GL 92-01 closeout 

letters, the staff requested that licensees 

review the information, and provide any 

comments back to the NRC. The staff will 

review any comments and recommended 

changes received from the industry and input 

the data into the RVID, as appropriate. 

6.5 Future Revisions to the RVID 

The RVID-2 program will be updated when 

sufficient amounts of new surveillance data, 

chemistry data, or fluence evaluations 

warrant a new revision. Revisions to the RVID 

database diskettes will continue to be 

released on the World Wide Web for 

downloading. The NRC technical monitor will 

mail the diskettes to requesters who do not 

have access to the World Wide Web.

NUREG-1511, Supp. 2 6-2
7 SIGNIFICANT PLANT-SPECIFIC REVIEWS

7.1 Background 

In the original NUREG-1511 (Ref. 2), the 

NRC staff noted that, as a result of 

information received in licensee responses to 

GL 92-01, Revision 1 (Ref. 1), several 

significant plant-specific issues were 

identified. These included the PTS 

evaluations of Palisades and Beaver Valley 

Unit 1, which showed that these facilities 

were projected to exceed the PTS screening 

criteria (refer to 10 CFR 50.61, Ref. 5) before 

the end of their current operating licenses. 

Additional information was also cited for 

Calvert Cliffs Units 1 and 2, Fort Calhoun, 

Indian Point Unit 3, and Zion Units 1 and 2. In 

NUREG-1511, Supplement 1 (Ref. 20), the 

NRC staff provided updated information on 

Palisades and Calvert Cliffs Units 1 and 2 

PTS assessments as well as information on 

the PTS assessment for Ginna. 

This chapter updates the information 

regarding the PTS assessments for the 

Palisades and Beaver Valley Unit 1 nuclear 

plants, and presents new information 

regarding the staff's structural integrity 

assessments of plants that have begun the 

license renewal process. Also included in this 

section is a brief discussion of an issue that 

developed with regard to the Fort Calhoun 

PTS assessment after the staff had 

completed its review of GL 92-01. 

7.2 Palisades 

At the time NUREG-1 511, Supplement 1, was 

published, the Consumers Power Company 

(CPCo, the licensee for the Palisades plant, 

currently named Consumers Energy) had 

submitted a revised evaluation of the 

projected end-of-license (EOL) neutron 

fluence for the Palisades RPV (Ref. 55). The 

licensee's analysis was performed by 

Westinghouse and based upon a combination 

of physical/geometrical refinements to the

neutron transport model, the use of plant

specific dosimetry data, and a specific 

spectral adjustment routine used in the 

Westinghouse computer code. The results of 

the analysis showed a decrease of 

approximately 25 percent in the EOL fluence 

for the limiting axial RPV weld. The licensee 

contended that, based on this reanalysis, the 

limiting axial welds in the Palisades RPV 

would remain below the PTS screening 

criteria through the end of its current 

operating license. 

The NRC issued an interim safety evaluation 

on this submittal on December 20, 1996 

(Ref. 56). With assistance from Brookhaven 

National Laboratory, the NRC staff concluded 

that the changes to the projected fluence due 

the physical/geometrical refinements to the 

neutron transport code model were 

acceptable. These changes led to a 

reduction in the fluence at the RPV of 

approximately 8 percent. The NRC staff also 

concluded that there was not sufficient 

justification to grant the remaining 17 percent 

reduction requested by the licensee on the 

basis of dosimetry data and spectral 

adjustments. The approved 8 percent 

reduction was sufficient to extend the date at 

which the limiting axial weld was projected to 

exceed the PTS screening criteria from 1999 

to the end of 2003. 

Since December 1996, CPCo and the NRC 

staff have held additional meetings to discuss 

whether there is a sufficient basis for granting 

the 17 percent reduction in the EOL neutron 

fluence for the Palisades plant. To date, the 

NRC staff continues to conclude that it is not 

possible to grant any additional reduction in 

the projected EOL fluences given the bases 

which have been presented by the licensee. 

In early 1999, CPCo did, however, present 

the NRC with a plan to take a different 

approach at demonstrating other ways in 

which a reduction in the projected fluence 

could be achieved. An initial submittal on this

7-1 NUREG-1511, Supp. 2
approach was received by the NRC in a letter 

from CPCo dated March 25, 1999 (Ref 60). 

After discussing the content of the submittal 

with the NRC staff, CPCo concluded that the 

adjustments to the fluence methodology 

described in the submittal were not likely to 

be approved without additional justification. 

CPCo therefore withdrew the submittal of 

March 25, 1999 until further additional 

information could be generated. CPCo 

submitted their most recent fluence analysis 

for the Palisades RPV on February 21, 2000 

(Ref. 61), with the objective of addressing the 

staff's concerns with the previous fluence 

methodology. The NRC staff's review of the 

new fluence methodology is in progress. 

7.3 Beaver Valley Unit 1 

The Duquesne Light Company (the licensee 

for the Beaver Valley Nuclear Station) 

submitted a revised PTS evaluation of the 

RPV for Beaver Valley Unit 1 to the NRC on 

August 2, 1996 (Ref. 62). The licensee 

contended that, as a result of their flux 

reduction effort and reanalysis of their 

existing Charpy data (using a hyperbolic 

tangent curve fitting program), the limiting 

plate material in the Beaver Valley Unit 1 

vessel was now below the PTS screening 

criteria through the end of the facility's current 

operating license. Although NRC staff 

disagreed with certain details of the 

licensee's evaluation, as cited in NRC's 

safety evaluation dated October 7, 1997 

(Ref. 63), the staff did concur with the 

licensee's conclusion that the limiting plate in 

the Beaver Valley Unit 1 RPV was no longer 

projected to exceed the PTS screening 

criteria before EOL. 

7.4 Fort Calhoun 

A neutron fluence evaluation submitted by the 

Omaha Public Power District, the licensee for 

Fort Calhoun Station Unit 1, has indicated 

that the licensee will need to take further

action to address PTS. The details of the 

Fort Calhoun issue are in the letter from 

L. Raynard Wharton to S. K. Gambhir dated 

November 30, 1999 (Ref. 64). 

7.5 License Renewal Plants: Calvert 

Cliffs Units 1 and 2, Oconee Units 1, 

2, and 3 

License renewal applications have been 

submitted by the licensees for the Calvert 

Cliffs and Oconee plants (Refs. 65 and 66). 

The NRC staff has reviewed these 

applications, in part, with respect to 

compliance with 10 CFR 50.61 requirements 

for meeting the pressurized thermal 

shock screening criteria and the Charpy USE 

analysis for compliance with Appendix G to 

10 CFR Part 50. The staff has issued safety 

evaluation reports on each application that 

indicate that the applicable regulations will be 

satisfied through the license renewal period 

for each unit (Refs. 67 and 68). 

7.6 Reactor Pressure Vessels Fabricated 

by Multiple Vendors 

In the course of recent reviews for several 

plants, the NRC staff has identified several 

plants with reactor pressure vessels 

fabricated by multiple vendors. A description 

of the circumstances in each case is provided 

below: 

"* Browns Ferry Unit 2: the axial welds 

within each shell course were fabricated 

by Babcock & Wilcox (B&W), and the 

circumferential welds were fabricated by 

Ishikasajima-Hirama Heavy Industries 

(IHI) of Japan. 

"* Browns Ferry Unit 3: the axial welds 

within each shell course were fabricated 

by B&W, and the circumferential welds 

were fabricated by IHI.

NUREG- 1511, Supp. 2 7-2
"* Peach Bottom Unit 2: the axial welds 

within each shell course were fabricated 

by B&W, and the circumferential welds 

were fabricated by Chicago Bridge & 

Iron (CB&I). 

"* Peach Bottom Unit 3: the axial welds 

within each shell course were fabricated 

by B&W, and the circumferential welds 

were fabricated by CB&I. 

"* Point Beach Unit 2: one of the beltline 

circumferential welds (nozzle belt to 

intermediate shell weld) was fabricated 

by B&W, and the other circumferential 

weld (intermediate to lower shell weld) 

was fabricated by Combustion 

Engineering (CE). 

"* Quad Cities Unit 2: the axial welds 

within each shell course were fabricated 

by B&W, and the circumferential welds 

were fabricated by CB&I. 

"* Surry Unit 1: the axial welds within 

each shell course and one of the 

beltline circumferential welds 

(SA- 1650) was fabricated by B&W, and 

the other beltline circumferential weld 

(J726) was fabricated by Rotterdam 

Dockyards. 

"* Surry Unit 2: the axial welds within 

each shell course were fabricated by 

B&W, and the beltline circumferential 

welds were fabricated by Rotterdam 

Dockyards.

7.7 Reactor Pressure Vessels Fabricated 

from Forging Materials 

Approximately 20 RPVs for US nuclear power 

plants (all PWRs) were fabricated from 

forgings. Although most of these forged 

RPVs were fabricated from SA-508 Class 2 

steel, the RPVs for Braidwood Units 1 and 2 

and Prairie Island Units 1 and 2 were 

fabricated from SA-508 Class 3 steel. A 

listing of the plants with forged RPVs and the 

forging fabricator is provided below: 

"* Braidwood 1 - Japan Steel Works 

"* Braidwood 2 - Japan Steel Works 

"* Byron 1- Ladish 

"* Byron 2 - Japan Steel Works 

"* Catawba 1 - Rotterdam Dockyard 

"* Davis-Besse - Ladish 

"* Ginna - Bethlehem Steel 

"* Kewaunee - Bethlehem Steel 

"* McGuire 2 - Rotterdam Dockyard 

* North Anna 1 - Rotterdam Dockyard 

* North Anna 2 - Rotterdam Dockyard 

"* Oconee 2 - Ladish 

"* Oconee 3 - Ladish 

"* Point Beach 2 - Bethlehem Steel 

"* Prairie Island 1 - Creusot-Loire 

"* Prairie Island 2 - Creusot-Loire 

"* Sequoyah 1- Rotterdam Dockyard 

"* Sequoyah 2- Rotterdam Dockyard 

"* Turkey Point 3 - Bethlehem Steel 

"* Turkey Point 4 - Bethlehem Steel 

"* Watts Bar - Rotterdam Dockyard

7-3 NUREG-1511, Supp. 2
8 SUMMARY AND CONCLUSIONS

Since the issuance of GL 92-01, Revision 1 

(Ref. 1), in March 1992, and NUREG-1511 

(Ref. 2), in December 1994, the staff has 

directed its efforts toward determining the 

generic implications of the larger-than expected variability observed in the chemical 

composition of RPV welds. To address this 

concern, the staff issued Supplement 1 

to GL 92-01, Revision 1 (Ref. 3), on 

May 18, 1995, and requested that the 

licensees to which the GL was addressed 

identify any new data which could be 

pertinent to and impact the structural 

integrity analyses of their RPVs relative 

to the requirements of 10 CFR 50.60, 

10 CFR 50.61, and 10 CFR Part 50, 

Appendices G and H (Refs. 4-7). 

The majority of the licensees responding to 

GL 92-01, Revision 1, Supplement 1, 

indicated that they were participating in the 

activities of their respective industry owners 

group (e.g., either the BWRVIP, CEOG, or 

B&WOG) to collect and analyze available 

RPV chemistry (specifically, copper and 

nickel content data) and surveillance data on 

behalf of its member utilities. The owners 

groups have now completed these initiatives. 

This represents the first time that a 

comprehensive, integrated assessment of all 

alloying chemistry and surveillance data has 

been compl.eted for the materials that have 

been used to fabricate the RPVs in U.S. light

water nuclear power plants. These efforts 

should minimize surprises regarding the 

alloying chemistries for domestic RPV beltline 

materials. However, the staff expects 

additional surveillance data will become 

available after fracture toughness testing is 

performed on surveillance capsules that are 

presently being irradiated in domestic RPVs. 

The staff will incorporate the additional 

surveillance data into the existing database 

after it becomes available for review.

The review of the industry's vessel integrity 

initiatives led the staff to conclude that the 

new data compiled by the owners groups could have an impact on PTS and P-T limit 

assessments for some facilities. Therefore, 

from summer to fall of 1998, the staff issued 

RAIs to the majority of licensees that 

responded to GL 92-01, Revision 1, Supplement 1, and requested that the 

recipients assess the impact of the newly compiled chemistry and surveillance data, if 

any, on the PTS and P-T limit evaluations for 

their facilities. The staff reviewed the licensee 

responses to the RAIs and all licensee 

responses to GL 92-01 Revision 1, and 

GL 92-01, Revision 1, Supplement 1, along with the applicable reports from the owners 

groups, to determine the appropriate chemical composition and neutron 

embrittlement trends for each RPV. In some 

cases, these reviews indicated the need for 

additional discussions with the licensee to 

clarify or justify assumptions and conclusions 

made in the submittals. The results of these 

reviews have been used to prepare revisions 

to the data contained in the RVID for each 

plant. 

Several developments have also occurred in 

the areas of Codes and Standards activities 

and research developments that affect RPV 

integrity assessments. In particular, the 

ASME Code Committees have passed Code 

Cases N-640 and N-588 (Refs. 9 and 10), which have direct consequences on P-T 

limit assessments, and Code Case N-629 

(Ref. 11), which proposes a new method (the "Master Curve") for indexing material 

reference temperatures. Research 

developments have included work on new 

RPV embrittlement correlations, RPV flaw 

distribution studies, ongoing work to support the Master Curve methodology, and an effort 

by the NRC and industry to revise the PTS 

screening criteria.

NUREG-1511, Supp. 2 8-1
The staff has also considered proposals to 

reduce the scope of augmented inspections 

performed on the shell welds of BWR RPVs. 

The basis for this proposed reduction in 

scope for the augmented inspections, which 

are required pursuant to Paragraph 

(g)(6)(ii)(A) of 10 CFR 50.55a, was proposed 

by the BWRVIP in Topical Report 

BWRVIP-05. The staff approved this topical 

report in its SER dated July 28, 1998 

(Ref. 14) , and informed the industry of its 

position to allow reduced inspection scopes 

for inspections of circumferential shell welds 

in BWR RPVs in GL 98-05, which was issued 

on November 11, 1998 (Ref. 47). The staff 

has currently approved a number of 

proposals for reduced inspections of the 

circumferential RPV shell welds in BWR

designed facilities. 

With the adoption of 10 CFR 50.66 (Ref. 16), 

the staff has also established a regulatory 

framework for thermal annealing of RPVs. 

The staff's regulatory position on thermal 

annealing is documented in RG 1.162 

(Ref. 17). 

The feasibility of thermal annealing was 

demonstrated by a DOE/Industry-sponsored

annealing demonstration project (ADP); this 

ADP was performed at a decommissioned 

Marble Hill reactor unit, and applied indirect 

heating as the method for annealing the unit's 

RPV. A second ADP using an electrical 

resistance heating approach was canceled. 

Currently, no licensee is planning to anneal 

the RPV of any U.S. facility. 

The staff has also updated the RVID. RVID 2 

was issued on the World Wide Web in June 

1999. RVID 2 is a Windows 3.1 native 

application based on Microsoft Access 2 .0TM. 

The four RVID 2 diskettes can be 

downloaded from the homepage at 

(http://www.nrc.gov/NRR/RVID/index.html) 

which is linked to the NRC homepage. The 

RVID was developed following the staff's 

review of licensee responses to GL 92-01, 

Revision 1. The database summarizes the 

properties of the reactor vessel beltline 

materials for each operating commercial 

nuclear power plant. RVID 2 reflects license 

responses to the RAIs on GL 92-01, Revision 

1, Supplement 1. The database will be 

updated when sufficient amounts of new 

surveillance data, chemistry data, or fluence 

evaluations warrant a new revision.

NUREG-1511, Supp. 2 8-2
9 REFERENCES

1. GL 92-01, Revision 1, "Reactor Vessel Structural Integrity," March 6, 1992. 

2. NUREG-151 1, "Reactor Pressure Vessel Status Report," December 1994. 

3. Generic Letter 92-01, Revision 1, Supplement 1, "Reactor Vessel Structural Integrity," 

May 18, 1995. 

4. Section 50.60 to Part 50 of Title 10 to the Code of Federal Regulations, "Acceptance Criteria for Fracture Prevention Measures for Lightwater Nuclear Power Reactors for Normal 

Operation." 

5. Section 50.61 to Part 50 of Title 10 to the Code of Federal Regulations, "Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events." 

6. Appendix G to Part 50 of Title 10 to the Code of Federal Regulations, "Fracture Toughness 

Requirements." 

7. Appendix H to Part 50 of Title 10 to the Code of Federal Regulations, "Reactor Vessel 

Material Surveillance Program Requirements." 

8. SECY-95-119, "Status of Reactor Pressure Vessel Issues," May 8, 1995. 

9. American Society for Mechanical Engineers Code Case N-640, "Alternative to Reference 

Fracture Toughness for Development of P-T Limit Curves for ASME Section XI, Division I," 

September 18, 1998. 

10. American Society for Mechanical Engineers Code Case N-588, "Attenuation to Reference Flaw Orientation of Appendix G for Circumferential Welds in Reactor Vessels, Section XI, Division 1," December 12, 1997. 

11. American Society for Mechanical Engineers Code Case N-629, "Use of Fracture Toughness Test Data to Establish Reference Temperature for Pressure Retaining Materials for Vessels 

of Section III, Division 1, Class 1," April 1998. 

12. Section 50.55a to Part 50 of Title 10 to the Code of Federal Regulations, "Codes and 

Standards." 

13. BWR Vessel and Internals Project Topical Report No. EPRI TR-105697, "BWR Reactor Pressure Vessel Shell Weld Inspection Recommendations (BWRVIP-05)," September 1995. 

14. Letter from G. C. Lainas, Acting Director - Division of Engineering, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, to C. T. Terry, Chairman - Boiling Water 

Reactor Vessel and Internals Project, "Final Safety Evaluation of the BWR Vessel and 

Internals Project BWRVIP-05 Report," July 28, 1998.

NUREG-1511, Supp. 2 9-1
15. Generic Letter 98-05, "Boiling Water Reactor Licensees Use of the BWRVIP-05 Report to 

Request Relief from Augmented Examination Requirements on Reactor Pressure Vessel 

Circumferential Shell Welds," November 10, 1998. 

16. Section 50.66 to Part 50 of Title 10 to the Code of Federal Regulations, "Requirements for 

Thermal Annealing of the Reactor Pressure Vessel." 

17. Regulatory Guide 1.162, "Format and Content of Report for Thermal Annealing of Reactor 

Pressure Vessels," February 1996. 

18. Appendix G to Section Xl of the American Society of Mechanical Engineers Boiler and 

Pressure Vessel Code, 1995 Edition through 1996 Addenda. 

19. NUREG/CR-6023, "Generic Analyses for Evaluation of Low Charpy USE Effects on Safety 

Margins Against Fracture of RPV Materials," Oakridge National Laboratory, July 1993. 

20. NUREG-151 1, Supplement 1, "Reactor Pressure Vessel Status Report," October 1996. 

21. Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," 

May 1988. 

22. CE NPSD-1039, Revision 2, "Best Estimate Copper and Nickel Values in CE Fabricated 

Reactor Vessel Welds," ABB Combustion Engineering Nuclear Operations, Windsor, CT, 

June 1997. 

23. CE NPSD-1 119, Revision 1, "Updated Analysis for Combustion Engineering Fabricated 

Reactor Vessel Welds Best Estimate Copper and Nickel Content," ABB Combustion 

Engineering Nuclear Operations, Windsor, CT, July 1998. 

24. BAW-2325, Supplement 1, "Supplement to the Response to Request for Additional 

Information (RAI) Regarding Reactor Pressure Vessel Integrity," Framatome Technologies, 

Lynchburg, VA, December 1998. 

25. BAW-2325, Revision 1, "Response to Request for Additional Information (RAI) Regarding 

Reactor Pressure Vessel Integrity," Framatome Technologies, Lynchburg, VA, January 1999. 

26. EPRI TR-109727, "BWR Vessel and Internals Project, Update of Bounding Assessment of 

BWR/2-6 Reactor Pressure Vessel Integrity Issues (BWRVIP-46)," Electric Power Research 

Institute, Palo, Alto, CA, December 1997. 

27. Memorandum from K. R. Wichman to E. J. Sullivan, "Meeting Summary for November 12, 

1997 Meeting with Owners Group Representatives and NEI Regarding Review of Responses 

to Generic Letter 92-01, Revision 1, Supplement 1 Responses," November 19, 1997. 

28. "NRC/Industry Workshop on Reactor Pressure Vessel Integrity, February 12 and 13, 1998," 

Nuclear Energy Institute, Washington, D. C.

NUREG-1511, Supp. 2 9-2
29. Letter from G. C. Lainas, Acting Director - Division of Engineering, U.S. Nuclear Regulatory Commission, to R. 0. Hardies, Baltimore Gas and Electric Company, "Combustion 

Engineering Owners Group Report CE NPSD-1 039 Revision 2, 'Best Estimate Copper and 

Nickel Values in CE Fabricated Reactor Vessel Welds' and CE NPSD-1039 Appendix A, Revision 2, 'CE Reactor Vessel Weld Properties Database Volumes 1 and 2'," 

March 27, 1998. 

30. Letter from S. C. Black, Deputy Director - Division of Licensing Program Management, U.S. 

Nuclear Regulatory Commission, to E. R. Kane, Framatome Services, "NRC Inspection of 

Framatome Technologies, Inc. (Inspection Report No.: 99901300/97-01)," January 28, 1998. 

31. Letter from G. C. Lainas, Acting Director - Division of Engineering, U.S. Nuclear Regulatory Commission, to C. D. Terry, Vice President- Nuclear Safety Assessment and Support, Niagra Mohawk Power Company, "Closeout for BWR Vessel and Internals Project, Update of 

Bounding Assessment of BWR/2-6 Reactor Pressure Vessel Integrity Issues (BWRVIP-46)," 

March 27, 1998. 

32. American Society for Testing and Materials E 900, "Standard Guide for Predicting Neutron 

Radiation Damage to Reactor Vessel Materials, E 706 (lII)." 

33. NUREG/CR-6551, "Improved Embrittlement Correlations for Reactor Pressure Vessel 

Steels," November 1998 

34. Marshall Committee, "An Assessment of the Integrity of PWR Pressure Vessels," United 

Kingdom Atomic Energy Authority, October 1, 1976. 

35. NUREG/CR-6471, Vol. 1, PNNL-1 1143, "Characterization of Flaws in U.S. Reactor Pressure 

Vessels, Density and Distribution of Flaw Indications in PVRUF," November 1998. 

36. NUREG/CR-6471, Vol. 3, PNNL-1 1143, "Characterization of Flaws in U.S. Reactor Pressure 

Vessels, Density and Distribution of Flaw Indications in the Shoreham Vessel," 

November 1999. 

37. Wallin, K., "Irradiation Damage Effects on the Fracture Toughness Transition Curve Shape for Reactor Vessel Steels," International Journal of Pressure Vessels & Piping, 55, pp. 61-79, 

1993. 

38. American Society for Testing and Materials Standard Procedure E 1921-97, "Standard Test 

Method for Determination of Reference Temperature, To, for Ferritic Steels in the Transition 

Range," February 1998. 

39. Letter NRC-99-039 from M. L. Marchi, Vice President - Nuclear, Wisconsin Public Service 

Company, to the U.S. Nuclear Regulatory Commission Document Control Desk, "Proposed Technical Specification Change Regarding Heatup and Cooldown Limit Curves and Request for Exemption to 10 CFR 50.60, 10 CFR 50.61, and Appendices G and H to Part 50," 

June 7, 1999. 

40. Transcript of the Commission Meeting dated May 12, 1999.

NUREG -1511, Supp. 2 9-3
41. Commission Staff Requirements Memorandum No. SRM M9705122B, May 30,1997. 

42. Information Notice 97-63, "Status of NRC Staff's Review of BWRVIP-05," August 7, 1997. 

43. Letterfrom the U.S. Nuclear Regulatory Commission to the Boiling Water Reactor Vessel and 

Internals Project, "Transmittal of NRC Staff's Independent Assessment of the Boiling Water 

Reactor Vessel and Internals Project BWRVIP-05 Report and Proprietary Request for 

Additional Information," August 14, 1997. 

44. Information Notice 97-63, Supplement 1, "Status of NRC Staff's Review of BWRVIP-05," 

May 7, 1998. 

45. NUREG-1560, Volume 1, "Individual Plant Examination Program: Perspectives on Reactor 

Safety and Plant Performance," December 1997. 

46. Regulatory Guide 1.154, "Format and Content of Plant-specific Pressurized Thermal Shock 

Safety Analysis Reports for Pressurized Water Reactors," January 1987. 

47. Generic Letter 98-05, "Boiling Water Reactor Licensees Use of the BWRVIP-05 Report to 

Request Relief from Augmented Inspection Requirements on Reactor Pressure Vessel 

Circumferential Shell Welds," November 11, 1998. 

48. Letter from Mr. V. Wagoner, Technical Chairman, BWRVIP Integration Committee to the U.S. 

Nuclear Regulatory Commission Document Control Desk, "Project No. 704 Response to NRC 

Comments on BWRVIP Analyses of BWR Reactor Pressure Vessel Axial Welds," 

November 12, 1999. 

49. Letter from J. R. Strosnider, Director of Engineering, U.S. Nuclear Regulatory Commission, 

to C. Terry, BWRVIP Chairman, "Supplement to Final Safety Evaluation of the BWR Vessel 

and Internals Project BWRVIP-05 Report," March 7, 2000. 

50. American Society for Testing and Materials, "Recommended Guide for In-Service Annealing 

of Water-Cooled Nuclear Reactor Vessels," ASTM E 509-86, Philadelphia, 1986. 

51. ASME Code Case N-557, "In Place Dry Annealing of a PWR Nuclear Reactor Vessel," 

March 1996. 

52. Letter from R. W. Smedley, Manager-Licensing, Consumers Power Company, to the U.S. 

Nuclear Regulatory Commission, "Docket 50-255-License DPR-20-Palisades Plant 

1OCFR50.61 -Preliminary Thermal Annealing Report Section 3," October 12, 1995. 

53.- Letter from T. C. Bordine, Manager - Licensing, Consumers Power Company, to the U.S. 

Nuclear Regulatory Commission Document Control Desk, "Response to Request for 

Additional Information Regarding the Preliminary Thermal Annealing Report.," 

August 26, 1996. 

54. Letter from T. C. Bordine, Manager - Licensing, Consumers Power Company, to the U.S. 

Nuclear Regulatory Commission Document Control Desk, "Preliminary Thermal Annealing 

Report - Withdrawal of Request to Review," April 24, 1997.

NUREG- 1511, Supp. 2 9-4
55. Letter from R. W. Smedley, Manager-Licensing, Palisades Plant, to the U.S. Nuclear 

Regulatory Commission Document Control Desk, "Docket 50-255, License DPR-20, 

Palisades Plant Updated Reactor Vessel Fluence Values, April 4, 1996. 

56. Letter from J. N. Hannon, Director - Project Directorate Il1-1, Division of Reactor Projects 

Ill/IV, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, to T. C. 

Bordine, Manager - Licensing, Consumers Power Company, "Palisades: Evaluation of 

Updated Reactor Pressure Vessel Fluence Values," December 20, 1996. 

57. Letter from J. M. Taylor, Executive Director for Operations, U.S. Nuclear Regulatory 

Commission, to Dr. T. R. Lash, Director - Office of Nuclear Energy, Science and Technology, 

U.S. Department of Energy, enclosing Two Copies of the Memorandum of Understanding for 

the DOE Annealing Demonstration Project, August 4, 1996. 

58. Electric Power Research Institute Report TR-108316, "Marble Hill Thermal Annealing 

Demonstration Project," March 1998. 

59. NUREG/CR-6552, "Marble Hill Annealing Demonstration Evaluation," February 1998. 

60. Letter from N. L. Haskell, Director of Licensing, Consumers Energy Company) to the U.S. 

Nuclear Regulatory Commission Document Control Desk, "Palisades Plant, Revised Best 

Estimate Fluence Evaluation Using Industry Data," March 25, 1999. 

61. Letter from D. G. Malone, Acting Director of Licensing, Consumers Energy, to the U.S. 

Nuclear Regulatory Commission Document Control Desk, "Palisades Reactor Vessel Neutron 

Fluence Reevaluation," February 21, 2000. 

62. Letter from S. C. Jain, Vice President - Nuclear Services, Duquesne Light Company, to the 

US Nuclear Regulatory Commission Document Control Desk, "Beaver Valley Power Station 

Unit No.1, 10 CFR 50.61 RT-PTS Submittal," August 2, 1996. 

63. Letter from D. S. Brinkman, Project Manager, U.S. Nuclear Regulatory Commission, to J. E. 

Cross, Chief Nuclear Officer, Duquesne Light Company, "Safety Evaluation Regarding 

Pressurized Thermal Shock (PTS) Assessment for Beaver Valley Power Station Unit No. 1 

.... " October 7, 1997. 

64. Letter from L. R. Wharton, Project Manager, U.S. Nuclear Regulatory Commission, to S. K. 

Gambir, Division Manager of Nuclear Operations, Omaha Public Power District, "Concerns 

with Fort Calhoun Station Basis for Prediction of RTPTs at Expiration of License . . . 

November 30, 1999. 

65. Letter from C. H. Cruse, Vice President - Nuclear Energy, Baltimore Gas and Electric 

Company, to the U.S. Nuclear Regulatory Commission Document Control Desk, "Application 

for License Renewal," April 8, 1998. 

66. Letter from M. S. Tuckman, Executive Vice President - Nuclear Generation, Duke Power 

Company, to the U.S. Nuclear Regulatory Commission Document Control Desk, "Oconee 

Nuclear Station, Unit 1, 2, and 3 ..... Application for Renewed Operating Licenses," 

July 6, 1999.

9-5 NUREG-1511, Supp. 2
67. NUREG-1705, "Safety Evaluation Report Related to the License Renewal of Calvert Cliffs 

Nuclear Power Plant, Units 1 and 2," December 1999. 

68. NUREG-1723, "Safety Evaluation Report Related to the License Renewal of Oconee Nuclear 

Station, Units 1, 2, and 3," March 2000.

NUREG- 1511, Supp. 2 9-6
APPENDIX

PTS TABLES FOR PRESSURIZED WATER REACTORS 

AND LIMITING MATERIAL TABLES FOR 

BOILING WATER REACTORS

A-1 NUREG-1511, Supp. 2
APPENDIX - PWR RTpts VALUES FOR LIMITING MATERIALS (RTpts ONLY APPLICABLE TO PWRs)

Oil T- U'W A?1111 1 ýI

z 

m 6) 

(I 

-c ri "1o 

n)

ARKANSAS NUCLEAR 1 

:ARKANSAS NUCLEAR 2 

* BEEAVR VALLEY 1 

.BEAYVER VALLEY 2 

BRAIDWOOD 1 

BRAIDWOOD 2 

BYRON 1 

BYRON 2 

CALLAWAY 

CALVERT CLIFFS 1 

CALVERT CLIFFS 2 

CATAWBA 1 

CATAWBA 2 

COMANCHE PEAK 1 

COMANCHE PEAK 2 

COOK 1 

COOK 2 

CRYSTAL RIVER 3 

DAVIS-BESSE 

DIABLO CANYON 1 

DIABLO CANYON 2 

FARLEY 1 

FARLEY 2 

FORT CALHOUN 

GINNA 

INDIAN POINT 2 

INDIAN POINT 3 

KEWAUNEE 

MCGUIRE 1 

MCGUIRE 2 

MILLSTONE 2 

MILLSTONE 3 

NORTH ANNA 1 

NORTH ANNA 2 

OCONEE 1 

OCONEE 2 

OCONEE 3 

PALISADES 

PALO VERDE 1.

UPPER/LOWER SHELL CIRC. WELD WF-112 

LOWER SHELL C-8010-1 

LOWER SHELL B6903-1 

INTERMEDIATE SHELL B9004-1 

LOWER NOZZLE BELT FORGING 

LOWER NOZZLE BELT FORGING 

INT. SHELL FORGING 

MIDDLE CIRCO WELD WF-447 

LOWER SHELL. R2708-3. 

LOWER SHELL AXIAL WELD 3-203A/C 

INTERMEDIATE SHELL D-8906-1 

LOWER SHELL 04 FORGING 

INTERMEDIATE SHELL B8605-2 

LOWER SHELL R108-1 

INT SHELL A3807-2 

INT. SHELL AXIAL WELDS 2-442 A,B,&C 

INTERMEDIATE SHELL PLATE 10-1 

LOWER SHELL AXIAL WELDS SA-1580 

UPPER/LOWER SHELL CIRC WELD WF-182-1 

LOWER SHELL AXIAL WELD 3-442C 

INTERMEDIATE SHELL B5454-2 

LOWER SHELL B6919-1 

INTERMEDIATE SHELL B7212-1 

LOWER SHELL AXIAL WELDS 3-410A/C 

NT/LOWER.SHELL CIRC. WELD SA-847 

INTERMEDIATE SHELL B2002-3 

LOWER SHELL B2803-3 

INT./LOWER CIRC. WELD 

LOWER SHELL AXIAL WELD M1.32 

LOWER SHELL 04 

LOWER SHELL C-506-1 

INTERMEDIATE SHELL B9805-1 

LOWER SHELL FORGING 03 

LOWER SHELL 03 

INTERMEDIATE SHELL AXIAL WELDS SA-1073 

MIDDLE CIRC WELD WF-25 

UPPER/LOWER SHL CIRC WELD (INSIDE 75%) 

WF-67 

LOWER SHELL AXIAL WELDS 3-112A/C 

INTERMEDIATE SHELL M-6701-3

406L44 237 300 

C-8161-2 123 270 

C6317-1 268 270 

C0544-1 153 270 

5P-7016 52 270 

5P-7056 70 270 

5P-5933 111 270 

442002 101 300 

C4499-1 115 270 

21935 240 270 

A-4463-1 189 270 

527708 60 270 

C0543-2 130 270 

C4464-1 100 270 

C5522-2 92 270 

13253/12008T 215 270 

C5556-2 216 270 

8T1762 211 270 

821T44 191 300 

27204 258 270 

C5168-2 211 270 

C6940-1 183 270 

C7466-1 205 270 

27204/12008(T) 268 270 

61782 254 300 

B-4922-1 230 270 

A-0512-2 265 270 

1P3571 277 300 

21935/12008 231 270 

411337-11 141 270 

C-5667-1 177 270 

C-4039-2 134 270 

990400/292332 184 270 

990533/297355 220 270 

1P0962 214 270 

299L44 273 300 

72442 236 300 

W5214 269 270 

C4188-1 123 270 2
APPENDIX - PWR RTpts VALUES FOR LIMITING MATERIALS (RTpts ONLY APPLICABLE TO PWRs)

PRAIRIE* ISLAND 1 

PRAIRIE ISLAND 2

ROBINSON 2 

SALEM 1 

SAILEM.21* ! ~ • ~ i... ........ ......... .. ISEA;BROOK ....... !SEQUOYAH 1 ...... 

:SHEARON HARRIS 

SONGS-2 

SON.GNGS-2..3........... .. SOUTH TEXAS1 

SOUTH TEXAS 2 

ST, LUCIE I. 

ST. LUCIE 2 

SUMMER 

.SURRY .. .. ...... ........ 1 ............... . .. 

.TM I-1 ................. STURKEY POINT 3 

TURKEY POINT, 4, 

VOTL-E' 2 

WATERFORD 3 

-WA-TTSBAR1..3 ........ 

NwOLF REEK. 

Z•ION~ 2 ..............

IN TERMIEDIATE SHELL F-765-6 

LOWER SHELL F-6411-2 

CIRCUMFERENTi!AL WE•D SA-1 101 

INTERMEDIATE TO LOWER SHELL C.RCWELD 

iSA1484 

....NOZZLE To NT. SHiELCL CIRC W ELD 

!NOZZLE SHELL TO INTERMEDIATE SHELL IRCt. 

WELD 

UPP RCIR WELD 10-273 

LOWER SHELL AXIAL WELDS 3-42 C 

LO W ER SHELL A XIAL WvELDS 3-442 A&C 

!LOWER SHELL R1808-1 

ICI ,RC WELD 

INTERMEDIATE SHELL FORGING 05 

INTERMEDIATE SHELL 

LOWER SHELL 7-40

INTERMEDIATE SHELL C-6802-1 

INTERMEDIATE SHELL R1606-3 

iNfERMEDIATES•HELL R2W5ý71 

LOWER SHELL AXIAL WELDS 3-203 

IN iTiERM.iEDIATE SHELL M-605-2..... 

LOWER SHELL 

LOWER SHELL AX!AL WELDS L2 1A5 226 

NTOLOWIE.R SH.E.LýL .CIROq..WEILD R3-0013 

LOWER-SH.ELL AXIAL WELD, (100% (SA-1526) INT. TO LOWER SHELL CIRC WELD SA-1101 ~iT/LOWER N........... SHE•ILL IRC WEL.D SA-I 1 01 ........ 

INTERMEDIATE SHELL B8805-2 

L LOWER SHELL R8-1 

.-LOWER SHELL -M-.1004-2

jINTERMEDIATE SHELL 05 

LOWER SHELL Ei • R 5 0 -8 .II................. .... 

MIDDLE CIR. WELD WF-70 

NTLOE SHL C.R WEL..... ,A1769

79745-1 

71249

......... 2 .. 2 6 9 ........... 

1752 

W5214 

3413009 21935/120080 

Di i08-3 

2.887157/981.105.7. 

B4197-2 

C-7585-1 

0-9195-2 

C-4326-2 ....... N iR 62 0 7-1 ... 

305424 

B-3416-2 

. 9923-1 

0227 RDAM 

-.L4 4 .......... 

! ... ... 2) ........... M .... .. 

71249 

C.0061.3-2 

0-4304-1 ... .. ..57286-1 ...... 

527536 

885-2 

-71249

68 

274 

288 

1-63 

150 

253 

227 

..120 

235 

152 

196 

146 

1253 

804 

679 

206.... 

163....... 

113....

z 

c 

m 

0i 

CA 

-c

A>
APPENDIX 

- BWR LIMITING MATERIALS 

. .... , A It

z 

C: 

m 

0O 

./) 

c--

"BROWNS FERRY 1 

BROWNS FERRY 2 

BROWNS FERRY 3 

BRUONSWICK 1 

BRUNSWICK 2 

CLINTON 

COOPER 

DRESDEN 2 

DRESDEN 3 

DUANE ARNOLD 

FERMI 2 

FITZPATRICK 

GRAND GULF 1 

HATCH 1 

HATCH 2 

HOPE CREEK 

LASALLE 1 

LASALLE 2 

LIMERICK 1 

LIMERICK 2 

MONTICELLO 

NINE MILE POINT 1 

NINE MILE POINT 2 

OYSTER CREEK 

PEACH BOTTOM 2 

PEACH BOTTOM 3 

PERRY 

PILGRIM 

QUAD CITIES 1 

QUAD CITIES 2 

RIVER BEND 

SUSQUEHANNA 1 

SUSQUEHANNA 2 

VERMONT YANKEE 

WNP-2

NOTES: 

1.) VALUES APPLY TO EXISTING LICENSE, AND WILL NEED TO BE RE-ASSESSED FOR LICENSE RENEWAL 

2.) FOR UPPER SHELF ENERGY INFORMATION, REFER TO NUREG 1511 (DECEMBER 1994) 

3.) HEAT ID's BEGINNING WITH "NA" ARE UNKNOWN (ELECTROSLAG WELDS)

4.) ART @ EOL IS THE VALUE MEASURED AT PEAK FLUENCE AT THE INSIDE SURFACE

I

CIRC WELD WF-154 

AXIAL WELDS 

AXIAL WELDS ES 

NOZZLE FORGING N16A 

NOZZLE FORGING N16B 

WELDS 

LOWER INTERMEDIATE SHELL G-2802-2 

LOWER SHELL AXIAL WELD 

LOWER INT. TO LOWER SHELL CIRC WELD 

LOWER INTERMEDIATE SHELL 1-20 

LOWER SHELL AXIAL WELDS 2-307AB,C 

LOWER SHELL AXIAL WELDS 2-233A/C 

#2 SHELL AXIAL WELDS 

LOWER INT. SHELL G-4804-2 

LOWER SHELL G6603-2 

INT-LOWER TO INT. SHELL CIRCUMFERENTIAL WELD 

MIDDLE SHELL AXIAL WELDS 3-308 A/C 

LOWER SHELL 21-2 

SHELL COURSE # 2 17-2 

SHELL COURSE # 1 14-2 

LOWER/INT. SHELL 1-14 

UPPER SHELL G-307-4 

NUMBER 2 SHELL 

LOWER-INT. SHELL G-8-6 

LOWER INTERMEDIATE SHELL 

LOWER INTERMEDIATE SHELL 6-139-10 

AXIAL WELD 

LOWER INT. SHELL AXIAL WELDS 1-338A,B,C 

LOWER AND LOWER INTERMEDIATE AXIAL WELDS 

LOWER INTERMEDIATE & LOWER SHELL AXIAL WELDS 

AXIAL WELDS 

LOWER INTERMEDIATE SHELL 22-3 

LOWER INTERMEDIATE SHELL 22-1 

LOCATION UNKNOWN 1-15 

#2 RING

- ARTOWL

147 

144 

144 __

406L44 

NA/W-A 

NA/W-A 

Q2Q1VW 

Q2Q1VW 

76492 

C2307-2 

PQ1092C-2 

299L44/8650 

B0436-2 

13253/12008 

27204/12008 

627260 

C4114-2 

C8553-1 

D53040 

1 P3571 

C9425-1 

C7677-1 

B3416-1 

C2220-1 

P2076 

C3147-1 

P2150-1 

C2873-1 

C2773-2 

627260/B322A27AE 

27204/12008 

PQ 1300 

PQ1300 

5P6756/0342(T) 

C2433-1 

C2421-3 

C3116-2 

B5301-1

59

58

68 

49

128 

72 

152 

74 

14 

93 

86 

95 

130 

165 

173 

59 

174

I UMMW MAltKtAL Stumnot ow

I
NRC FORM 335 U.S. NUCLEAR REGULATORY COMMISSION 1. REPORT NUMBER (2-89) (AssIgned by NRC, Add Vol., Supp., Rev., NRCM 1102, BIBLIOGRAPHIC SHEET and Addendum Numbers, If any.) 3201,.3202 BBIGAHCDATASHE 

(See instructions on the reverse) 

2. TITLE AND SUBTITLE Supplement 2 

Reactor Pressure Vessel Status Report 

3. DATE REPORT PUBLISHED 

MONTH YEAR 

October 2000 

4. FIN OR GRANT NUMBER 

5. AUTHOR(S) 6. TYPE OF REPORT 

C. E. Carpenter, B. J. Elliot, A. L. Hiser, Jr., M. K. Khanna, Technical 

A. D. Lee, J. Medoff, M. A. Mitchell, S. C. F. Sheng, K. R. Wichman, J. R. Strosnider, W. H. Bateman 7. PERIOD COVERED (Inclusive Dates) 

Inclusive of September 2000 

8. PERFORMING ORGANIZATION -NAME AND ADDRESS (If NRC, provide Division, Office or Region, U.S. Nuclear Regulatory Commission, and mailing address; if contractor, provide name and mailing address.) 

Division of Engineering 

Office of Nuclear Reactor Regulation 

U.S. Nuclear Regulatory Commission 

Washington, D.C. 20555-0001

9. S-'POISORING• OR'GANIZATION - NAME AN D ADDRESS (if NRC, type"Same as above;" ifcontractor, provide NRC Division, Office or Region, U.S. Nuclear Regulatory Commission, and mailing address.) 

Same as above

IV. OUrLCMj'dIMNT INtJIt

11. ABSTRACT (200 words or less) 

(Condensed Version of the Abstract from the Report) 

On May 18, 1995, the U.S. Nuclear Regulatory Commission (NRC) issued Generic Letter (GL) 92- 01, Revision 1, Supplement 1, "Reactor Vessel Structural Integrity," requesting that addressees: (1) identify, collect, and report any new data pertinent to the analysis of structural integrity for the reactor pressure vess els (RPVs) at their nuclear power plants, and (2) assess the impact of those data on their RPV integrity analyses relative to the requirements of 10 CFR 50.61, "Fracture 

Toughness Requirements for Protection Against Pressurized Thermal Shock Events, " and to the requirements of Appendix G 

to 10 CFR Part 50, "Fracture Toughness Requirements". Since the issuance of GL 92-01, Revision 1, Supplement 1, the industry completed a major initiative to collect all available alloying chemist ry and materials property data for the various 

forging, plate and weld materials used to fabricate the RPVs. This represents t he first time that a comprehensive, integrated assessment of all alloying chemistry and surveillance data has been completed f or the materials that have been used to fabricate the RPVs in U.S. light- water nuclear power plants. In the summer a nd fall of 1998, the staff issued a series of 

requests for additional information (RAIs) regarding the industry's responses t o GL 92-01, Revision 1, Supplement 1. In the 

RAIs, the staff requested that the addressees assess how the updated alloying c hemistry and materials property data would affect the results of the RPV integrity analyses for their plants. This report, in part, summarizes both the industry's and the NRC's efforts to address how all of the new chemistry and surveillance data, when integrated, could affect the plant-specific RPV integrity analyses for the RPVs of U.S. light-water nuclear power plants.

12. KEY WORDS/DESCRIPTORS (List words or phrases that will assist researchers in locating the report.) 13. AVAILABIUTY STATEMENT 

10 CFR 50.61; 10 CFR Part 50, Appendix G; 10 CFR Part 50, Appendix H; pressurized thermal shock unlimited 

(PTS); upper shelf energy (USE); reactor pressure vessel (RPV); material property surveillance data; 14. SECURITY CLASSIFICATION 

Reactor Vessel Integrity Database (RVID); best estimate chemistry; adjusted ref erence temperature (This Page) (ART or RTndt); adjusted reference temperature for pressurized thermal shock (R Tpts) ; initial RTndt unclassified 

(RTndt(u)); base metal materials (plates or forging materials); longitudinal (axial) weld materials; (This Report) circumferential annealing (girth) weld materials; screening criteria for pressurized ther mal shock; reactor vessel unclassified 

u c a sf e 15. NUMBER OF PAGES 

16. PRICE

I V. 0Vr"rLCPACNIII I AMI" T INU I CZ)•

NRC FORM 33512-89N
Federal Recycling Program
NUREG-1511, Supp. 2 REACTOR PRESSURE VESSEL STATUS REPORT OCTOBER 2000

UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON, D.C. 20555-0001