PWROG Cl-'11 IELECTRIC RESEARCH POWER INSTITUTE · 2012-12-04 · Stephen Byrne (Westinghouse) and...

PWROG

Owners G)

Cl-'11 IELECTRIC POWERRESEARCH INSTITUTE

MRP

PWR IndustryReactor Vessel Roadmap

Maurice DingierEPRI TSC Chairman

PWROG Vice ChairmanEPRI/PWROG/NRC Meeting

January 6, 2011 1

Reactor Vessel Roadmap Overview

* Objectives

* Goals and Timing

* Drivers

* Project Activities

* Issue Management Table Gaps

2

Reactor Vessel Roadmap Objective

Roadmap is a vision statement to develop strategic plans andobjectives from and to:

- Coordinate the Activities between Stakeholders* EPRI MRP* PWROG* NRC* International Community

- Conservatively address Reactor Vessel Integrity* Optimize Operational Flexibility* Improve the analytical tools

- Position the Industry for life beyond 60 years of Operation

3

Reactor Vessel Roadmap Goals

Goals:Near Term (1 to 3 years)

* Apply risk insights to ease operating constraints

* Develop a framework for knowledge retention and technical basesmaintenance addressing Vessel integrity

* Initiate research and testing to objectively assess irradiation damage through80 years of operation

Medium Term (3 to 6 years)" Provide power reactor data to improve accuracy and understanding of high

fluence irradiation damage" Address the applicability of irradiation damage approaches to Gen 3 plants0 Implement a RPV knowledge retention plan

Long Term (6 to 10 years)" Maintain regulatory stability for current fleet and Gen 3 plants* Resolve RV integrity issue without adverse impact on operations

4

Reactor Vessel Roadmap Drivers

Drivers:- Asset Management through. 80 years of Operations

* Pro-actively address irradiation damage effects* Optimize operating flexibility

- Maintain regulatory stability

5

Reactor Vessel Roadmap Project Plan

Project Plan:- Operational Support through 60+ Years

* Coordinated RV Surveillance Capsule Program* 10CFR50.61(a) Implementation Guidance0 'Extended Beltline' Assessment- Effect on Vessel Integrity0 Optimize Operational Flexibility* Regulatory Interaction:

- GALL review- RG 1.99- 10 CFR 50 Appendix G and H; ASME Appendix E and K- Master Curve (Code Case N-629)

- Modeling of Data* Vessel Nozzle Stress Intensity and Flaw Assumption for use in the

'Extended Beltline'

° Embrittlement Correlation from Surveillance Capsule Database° Fracture Toughness Irradiation Damage Model

* Fluence Attenuation Model 6

Reactor Vessel Roadmap Project Plan

Project Plan:- Testing

* Utilize ATR-2 Test Samples and Coordinated Surveillance CapsuleProgram results to assess 'late blooming phase' embrittlement trends

- Archival Basis Retention° Maintenance of archive reactor vessel materials* Reactor Vessel integrity knowledge retention

7

Reactor Vessel Roadmap

0 Program InterfacesReactor V4essel Integrity Program

and E R ule.ent

Develop Coordinated SurveileancetIATCapsule ProgramI o o

Develop"KnoweLge Managtement system B f looity E mt•ineers tPae

Develop Archie Mt~enals Storage rogr . ...mP ement Ar.chve Cr ate a IncorStorage Programn-Industry T

Coria andIncr1o99eRev 3 and GARLesuLt

Exedde!Lin'e SImIpactI: •,'lw nEt~elBiln

SDevelop• Material Property aaae ].. .. . Refi ne. 13t a. Cotrelations .. = • , • .. .,

[Attequationt'Model C rrelations•

IDevelop Fracture Toughnes.•s Embrittlement Model •;

.. .. . ' :' • ~Irr~adiatedl Material Testing (IAEA, CRIEP, etc.) . . . ;

F7 !;•:": I"NL-A;TR~trradiationTestrogram" ' ::; :1

L ~ Late Blooming Embrittlement Phase Testing

Coordinate and Incorporate Non-Industry Test Results •

I•[ ~Coordinate and Incorporate INL Test Results J

LTOPS Margin Improvement

L• Test/Power Reactor Damage Comparison Testing

C/)

CU

a-

Z--

2011

2010

2012 2013 2014 2015 2016 2017 2018 2019

2020

8


IMT Gap ID Applicable IMT Gap Responsibility

EPRI MRP: LeadP-AS-04 Neutron Embrittlement of RPV Steels PRG MSC

PWROG MSC

P-AS-05 Neutron Dose Rate Effects on Low Alloy EPRI MRP: LeadSteels PWROG MSC

P-AS-06 Pressurized Thermal Shock Evaluation EPRI MRP & PWROG MSC

P-AS-27 10 CFR 50 Appendix G EPRI MRP: LeadPWROG MSC

P-AS-28 Upper Shell Course / Nozzle Forging Neutron PWROG MSC: LeadEmbrittlement EPRI MRP

P-AS-37 and EPRI MRP: Lead

P-RG-11 80-Year Material Surveillance Program MSC support

P-RG-10 MRP License Renewal Open Items EPRI MRP & PWROG MSC

9

Reactor Vessel Integrity Topics

Implementation of the Roadmap - Current Programs

o 10 CFR 50 Appendix H Revision

* Regulatory Guide 1.99 Revision 3 and NRC RPV Surveillance DatabaseUpdate,

e Extended Beltline and Primary Nozzle Assessment

e PTS Rule Implementation Guidance

9 Surveillance Capsule Pull Recommendations and Regulatory Guide 1.99Revision 4

o Pressure-Temperature Limit Topics

e 10 CFR 50 Appendix G and Risk-informed Appendix G

* Closing Comments

10


Questions?

11

PWROG C I ' 1 •,ELECTRIC POWERm urRESEARCH INSTITUTE

MRP-Ow

10 CFR 50, Appendix H RevisionRecommendations

Stephen Byrne (Westinghouse) and Sarah Davidsaver (AREVA)

January 6, 2011

I

10 CFR 50, Appendix H Revision Recommendations

Background -1. 10 CFR 50 Appendix H issued in 1973 was tied to surveillance program

design and post-irradiation testing requirements ofASTM E 185-73.2. Subsequent revisions to Appendix H and E 185 were in parallel through

E 185-823. No E 185 updates to Appendix H after E 185-82.

4. Creation of separate standards, E 185 and E 2215, in 2002 and updatedin 2010.

5. Purpose is to present recommendations for changes to Appendix Hkeyed to the 2010 versions of E 185 and E 2215.

2


Summary of Issues:° Issue #1 - Confusion between testing and design requirements" Issue #2 - Post-irradiation testing refinements after 1982 were never

adopted in Appendix H.* Issue #3 - Enhancements to the design for new plants was constrained

by the tie to existing programs." Issue #4- Surveillance program guidance for 60-year operation in

Standard Review Plans is not consistent with ASTM standards orAppendix H.

* Issue #5 - Application of design enhancements for new plants washampered by the lack of an Appendix H reference to E 185-02 or -10.

3


Comparison of key changes ASTM E 185-82 to E185-10

Surveillance capsule withdrawal schedulemodified- Simplifies application to a variety of plant designs while retaining

the underlying (original) basis for withdrawal frequency.

- Makes schedule applicable to 60-year (or 80-year) operating periodand to newer plants.

* Removed requirement to provide or to test HAZspecimens.

4


Comparison of key changes ASTM E 185-82 to E 185-10

(cont'd)

* Fracture toughness test specimens are now required fornew plants.

* Limits on lead factor have been modified.

• In 2002, ASTM E 185 was split into two separatestandards to eliminate confusion in Appendix H- One standard to address requirements for the design of new plant

programs - ASTM E 185.

- Other standard to address requirements for testing of capsules fromexisting programs - ASTM E 2215.

5


Comparison of key changes ASTM E 2215-10 andE 185-82 (post-irradiation testing)

" Removed requirement to test HAZ specimens.° Added fracture toughness testing to optional tests in

ASTM E 185-98- More emphasis placed on measuring fracture toughness to supplement the

information from Charpy impact tests.

6

Significant Improvements Adopted in ASTM E 2215-10 and E 185-10 Compared to 2002 Versions

The staff evaluated adoption of ASTM E 2215-02 and E 185-02 into1OCFR50 Appendix H

- Significant improvements have been adopted in ASTM E 2215-10 and E 185-10

- Improvements to ASTM E185-10 compared to E185-02

1. In 3, added definitions for limiting material' and 'standby capsule';'beltline' definition clarified; modified EOL to 'end-of-license'

2. In 5.2.1-5.2.3, clarified required materials that are to be included inthe program; if limiting material is outside beltline, that must beincluded with limiting beltline weld and base metal.

3. In 5.5.2.1, lead factor recommendation was changed from 1 to 3 to1.5 to 5

7

Significant Improvements Adopted in ASTM E 2215-10 andE 185-10 Compared to 2002 Versions (cont'd)

1. In 5.8.2, changed required number of capsules from 3-5 to 4-5 based onpredicted shift; simplified and improved the withdrawal schedule:

E185-02

TABLE I Suggested Withdrawal Schedule

Sequence Target Ftuence Priority

First 5X 1018 n cm 2(5X10 n/rm2) 2 (Required if ARTNDT> W6'Cfor PWVRs; E>1 MeV [1O00'F])

Second EOL Y4-T 1 (Required for all materials)Third EOL ID 1 (Required for all materials)Fourth (EOL 1/-T - 1st Capsule)12 3 (Required if ARTD-r> 111*C

[200°FSubsequent Supplemental Evaluations Not Required

E185-10

TABLE 1 Suggesied. Wi-thdraw&l Schedule

SNL~arne Taiget FRuereNO~

First ',"EQL ID Ttefirog RequfredlSEDNO Il IEQL I D T2MfRg Renq&Ed~i' lPrnjeCte~

1,R7ND-a-Ill 111C(20GrFIThIM dY ~ EMID TESUr~gReqffl-edl

FOW11MECOL0 Te-ril R~eqttfred.SammydL e, 2 EOL ID TeIrig Not RecqUlred

a. Increased first capsule target fluence because 5.5E 18 was too low of a fluence

b. Simplified and set target fluence as a fraction of EOL ID fluence for allcapsules

c. Accommodated 60 or 80 year vessel surveillance program.

8


Recommended Enhancements and Changes to 10 CFR 50Appendix HSeparate the requirements for reactor vessel surveillance programdesign from post-irradiation testing and evaluation:

- Make direct reference to the post-irradiation testing and evaluationrequirements contained in ASTM E 2215-10.

- Provide for new plant design by direct reference to the surveillanceprogram design requirements contained in ASTM E 185-10.

- Clarify that Appendix H revision does not require change to designof existing surveillance program design.

* Eliminate the requirement to provide and test HAZ specimens to beconsistent with practice over the past 25 years.

9


Recommended Enhancements and Changes to 10 CFR 50Appendix H (cont'd)

• Specify the use of the new surveillance capsule withdrawalschedule contained in both E 185-10 and E 2215-10 to:- Modify the required frequency of surveillance capsule evaluations

to better accommodate operation beyond 40 years and to align withthe guidance in the GALL report.

- Simplify the requirements for surveillance capsule evaluationfrequency.

- Differentiate the required number of capsules based on theprojected sensitivity of the limiting vessel material to neutronirradiation.

10


Recommended Enhancements and Changes to 10 CFR 50Appendix H (cont'd)

* Introduce wording in the revised Appendix H to address the eventualtransition from Charpy impact to fracture toughness testing inGeneration III plant surveillance programs:

- E 185-10 specifies the inclusion of fracture toughness specimens(in addition to Charpy impact specimens).

- E 2215-10 cites fracture toughness testing as "optional" toaccommodate older plants without fracture toughness specimens inthe capsules.

- New U.S. designs specify that fracture toughness specimens beincluded in the surveillance program. Existing plants rely on theuse of Charpy impact testing in their original programs.

11


Summary and Conclusions

NRC has reviewed ASTM E 185-02 and E 2215-02; recommendations

for changes to 10 CFR 50, Appendix H based on the 2010 standards:

- Change 10 CFR 50, Appendix H to adapt withdrawal schedules to 60-year

license and to reduce confusion over requirements in existing plants. by

direct reference to E 185-10 and E 2215-10 for surveillance program

design, post-irradiation testing and surveillance capsule withdrawal

schedule.

Clarify Appendix H text to differentiate between the three (design, testing

and withdrawal schedule).

Incorporate the use of fracture toughness specimens in Generation III

plant designs to be consistent with current practice and to enhance design

requirements for new plants.

12


Summary and Conclusions (cont'd)" Need for specific wording in Appendix H to clarify that

revision does not require surveillance program designchanges for existing plants.

" Need for subsequent meeting to address details?

13

PWROG rý2Ia-P

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ELECTRIC POWERRESEARCH INSTITUTE

MRP

Regulatory Guide 1.99 Revision 3 andNRC Reactor Pressure Vessel

Surveillance Database

Stephen Byrne (Westinghouse) and Sarah Davidsaver (AREVA)

/ January 6, 2011

I

Regulatory Guide 1.99 Revision 3 and NRC ReactorPressure Vessel Surveillance Database

Objective

- Present discussion points on Regulatory Guide 1.99Revision 3.

- Present discussion points on surveillance capsule andvessel database efforts

Goal

- Initiate discussion and obtain feedback on discussionpoints

2


Introduction* RG 1.99, Rev 4 will be discussed under the Surveillance

Capsule Pull topic, so will defer discussion of highfluence/long time effects.

" For RG 1.99, Rev 3, talking points are provided with theintent of identifying actions needed by industry to preparefor subsequent issue of the Revision 3 document.

" For the NRC RPV database, talking points are alsoprovided with the intent of identifying actions needed byindustry

3


Regulatory Guide 1.99 Revision 3 Discussion Points:

1. Application of surveillance data

- Carry-over of Position 2.1 CF derived per Reg.1.99 Rev. 2

Trending analysis application of Alternate PTSsurveillance data credibility bases

Protocol for linking sister plant data

Guide

Rule;

4


Regulatory Guide 1.99 Revision 3 Discussion Points (cont'd):

2. New embrittlement trend curve application features

- neutron attenuation and margin

- Tcold for surveillance data; time-averaged value;securing data for establishment of Tcold

- Importance of establishing manganese and phosphorusvalues for RPV beltline materials

- Charpy upper shelf energy decrease projection method(new?) and Appendix G screening criteria (current?)

5


Regulatory Guide 1.99 Revision 3 Discussion Points(cont'd):

3. Future number of operative embrittlement trendcurves

- Impact of using Alt. PTS Rule for Reg. Guide 1.99Rev. 3 on existing PTS Rule

- Direction of Reg. Guide 1.99 Rev. 4 development

6


NRC Reactor Vessel Materials Database Discussion Points:

1. NRC plans on RPV surveillance materials database;- timing for formation of industry panel to discuss protocols for

creating, accessing and maintaining database

- use of January 30-to-February 1 ASTM E 10.02 meeting inBaltimore to continue industry workshop discussions

2. Common basis for surveillance capsule fluence;- Example:

Original Fluence Updated Fluence

Capsule 1 1.78 E 18 n/cm 2 1.47 E18 n/cm 2

Capsule 2 1.24 E19 n/cm2 1.05 E19 n/cm2

Position 2.1 CF increases due to fluence decreasing (shift versusfluence relationship changes)

7


NRC RPV Database Discussion Points (cont'd):3. Expansion of surveillance database (discussed later)

Surveillance database groupings- Forgings:

* SA-508 Class 2andClass 3, Cu>0.10 wt%; Cu< 0.lOwt%- Plates:

" A-302GrB,Cu>0.10wt%" SA-533 GrB CI.1 andA-302-B Mod., Cu<0.10 and> 0.10 wt%

- Welds:

" Linde 124, Cu < 0.10 wt%* Linde 0091, Cu < 0.10 and > 0.10 wt%" Linde 1092, Cu > 0.10 wt%" Linde 80, Cu < 0.10 and > 0.10 wt%" SMIT 89 and UM89; Grau Lo LW320 and E8018 Electrodes

8


NRC RPV Database Discussion Points (cont'd):

3. Process to populate RPV "licensing database- Roll-over from RVID or issue of Generic Letter?

- Raw versus processed data (e.g., based on industry fundedassessments)

4. Database Needs, Quality and Consistency- Value of documenting chemistry data for expanded list of

elements (manganese, phosphorus, etc.)

- Use of Charpy curve -fitting for consistency of transitiontemperature shift data

- Documentation of Tcold values for each surveillance capsule

- Action - Define next steps for Database Review Panel

9

PWROG 8812 ELECTRIC POWERRESEARCH INSTITUTE

Ownrs GMRP

Extended Beltline and PrimaryNozzle Assessment

PWROG MSC

Carol Heinecke , Westinghouse

January 6,2011

1

Extended Beltline and Primary Nozzle Assessment

Background

License Renewal is Producing the need to Extend the Life of the ReactorVessel

o The region of embrittlement concern may be extended to regions outside ofthe traditional beltline.

o 10 CFR 50 Definition of Beltline -" Region of the reactor vessel.., thatdirectly surrounds the effective height of the active core and adjacent regions... that are predicted to experience sufficient neutron damage to be consideredin the selection of the most limiting material with regard to irradiationdamage."

• Pressure-Temperature Limits - 10 CFR 50 Appendix G, ASME XI

Appendix G

o Typically only considers the traditional beltline.

2

Extended Beltline and Primary NozzleAssessment

_ v

1I --

3


Background UAI

A fluence of 1.0 E+ 17 l/cm2 from the GALL Report has been used asthe threshold for consideration of embrittlement.

* By 80 years, embrittlement of the nozzles and the lower head may reachthe threshold fluence and their consideration may be necessary.

* The extended beltline regions have different geometry and thereforedifferent stress intensity factor correlations.

* Material property data for extended beltline regions is often notavailable.

* The combination of RTNDT and higher stress intensity factor correlationsin the extended beltline regions could make the extended beltline regionsmore limiting for pressure-temperature limits than consideration of onlythe traditional beltline materials.

4


Scoping Study (Phase I) for The Extended BeltlineTechnical Approach

* Purpose- Identify the scenarios in which the Extended Beltline couldpotentially be more limiting than the traditional Beltline for P-T limits,

* Approach-

o Obtain Reactor Vessel Data including the Extended Beltlineo Choose Stress Intensity Factor Correlations for Extended Beltline

Regions as needed.o Perform Parametric Evaluations

5


Detailed Approach

* Obtain Reactor Vessel Data

o Lack of materials data will lead to a need to assign genericproperties for extended beltline materials

o Postulated Flaw size 1/4T and location* Choose Stress Intensity Factor Correlations for Extended Beltline

Regions as needed.

o Nozzle correlation - WRC-B 175 expression, NRC/ORNLexpression

o Other correlations possibly needed - Nozzle to Shell Junction,Thickness transition.

6


Detailed Approach

* Parametric Evaluations- Matrix of Cases

o Least vs. Highest Embrittled Reactor Vesselso Well defined properties vs. Generic Propertieso Nozzle and other extended beltline geometries, as applicable

* Parametric Evaluations - Goalo Show that the traditional beltline is limiting for PT limitso If not possible, show the scenarios in which it could be limiting

7


Current Status

" Initial Material Property data gathering complete

o Some plants have limited extended beltline material properties* Stress Intensity Factor Correlations

o nozzle comer expressions

Scheduleo Phase I Scoping Study to be completed by October 2011

o Phase II schedule to be determined

8


* Phase Io Definition of Extended Beltline - Recommendationo Nozzle expression

o Nozzle-Shell Junctions, and Thickness Transitionexpression

o Postulated Flaw Size

* Future Discussion Areas

o Neutron attenuation in the Extended Beltline region

9

PWROG 8~2I ELECTRIC POWERRESEARCH INSTITUTE°0

,ne-s

Coordinated U

MRP

OS. PWR ReactorVessel Surveillance Program

(CRVSP)

Tim Hardin - EPRI

January 6, 2011

1

Purpose

" Capsule management plan that yields high fluenceCharpy data to fill gaps in the PWR surveillance database(SDB)

" Data obtained from this program can then be used toinform an embrittlement trend curve (ETC) appropriate forU.S. PWR vessels at high fluence

" The ultimate goal is to support the development of the RG1.99 Rev. 4 ETC

2

Background

* Irradiated surveillance data are used to predict changes inRPV transition temperature shift due to irradiationembrittlement

* Current ETCs based on SDB data < -3x1019 n/cm 2

* 69 US PWRs

Peak RPV

Fluence Range Average

60-year 1.5-7x10' 9 n/cm 2 4.1x1 019 n/cm 2

80-year 2-9xl019 n/cm 2 5.6x10 19 n/cm2

i Limited amount of irradiated U.S. LWR surveillance dataat fluences above -3x10 19 n/cm 2

3

Strategy

" Obtain as much high fluence data (3-10 xl 019 n/cm 2) aspractical by 2025

" Maximize the quantity and quality of high fluence data forall types of materials in service while-minimizing the burden on the utilities and- maintaining compliance with the requirements of

Appendix H and consistency with the guidance of theGALL report

" Program implements this strategy by- Deferring withdrawal of selected capsules and-Withdrawing additional capsules not required under

current Appendix H programs

4

General Approach

" Review each plant RVSP-Document contents of remaining surveillance capsules

-Group capsule materials based on product form(forging, plate, weld) and chemical composition (Cu, Ni,P, and Mn) (characterizes susceptibility to irradiationinduced embrittlement)

-Assess current site specific RVSP withdrawal schedule

" Identify high fluence gaps

" Identify recommended changes to existing withdrawalschedules

5

Material Groupings

" Plates- SA-302 Grade B (all > 0.10 wt% Cu)- SA-533 Grade B Class 1 and SA-302 Grade B Modified (> 0.10

wt% Cu)- SA-533 Grade B Class 1 (< 0.10 wt% Cu)

" Forgings- SA-508 Class 2 and Class 3 (< 0.10 wt% Cu)- SA-508 Class 2 and Class 3 (> 0.10 wt% Cu)

" Welds- Linde 80 (> 0.10 wt% Cu)- Linde 80 (< 0.10 wt% Cu)- Linde 1092 (all > 0.10 wt% Cu)- Linde 0091 (> 0.10 wt% Cu)- Linde 0091 (< 0.10 wt% Cu)- Linde 124, (all < 0.10 wt% Cu)- SMIT 89 and UM 89Grau Lo LW320 6

Plate (example)

SA-533 Cu_< 0.10%

Current Plan

12

10

I0

C.)

2

OE'

3-4 4-5 5-6 6-7 7-8 8-9 9-10 10+

Capsule Fluence at Withdrawal (0l019 n/cmt)

7

Forgings (example)

8

Weld (example)

Linde 1092 Cu > 0.10%

Current Plan

0

0.

U

0

6

5

4

3

2

1

0.

o Tested , Planned

3-4 345-8 6-7 7-8

Capsule Fluence at Withdrwal (X1019 fl/cfn2)

4-5 8-9 9-10

"C', -7 1-,-- 1'

Linde 1092 Cu > 0.10%

Recommended Plan

5-

4

• 33FLC)

3-4 4-5 5-6 6-7 7-8 8-9 9-10

Capsule Fluence at Withdrawal (xlO'9 n/cm2) 9

Approach to Withdrawal Schedule Changes

* Adjust withdrawal schedules to- obtain high fluence PWR surveillance data for the full

range of materials across the entire PWR fleet-Obtain high fluence Charpy data in a timely manner

while remaining compliant with Revision 2 of the GALLreport for a 60-year license and 1 OCFR50 Appendix H

-Target fluence range is from 3x1 019 n/cm 2 to 1 OxI 019

n/cm 2

e Encompasses the highest projected 80 year peakRPV fluence for the U.S. PWR fleet

10


" All remaining RVSP capsules were screened for potentialto obtain high fluence Charpy data by the year 2025

- Capsules that can not reach the target fluence by theyear 2025 were not considered directly for this effort

" If EOL capsule not yet tested at 60-year peak RPVfluence, test at 2x60 (consistent with GALL Rev 2)

- If this cannot be achieved by 2025, then test at lowerfluence, e.g., 80-year peak RPV fluence

11


* In some cases, plants that already tested the 60 yearpeak RPV fluence capsule are being asked to test anadditional capsule at a higher fluence (e.g., 80-year or2x60-year peak RPV fluence)

* 19 units have already tested their 60-year capsule* Surveillance materials at these plants were assessed for

the value they would bring to the program1 .How well the base metal and weld flux material

categories will be populated with high fluence data bythe coordinated RVSP

2.The divergence of the embrittlement trend curves at theprojected withdrawal fluences

* 10 of the 19 plants were selected to test an additionalcapsule based on these criteria

12

Example: Additional Capsule Test NotRecommended for Plant A

500

450

400

350

300

o 25020-200

150

100

50 60-yearPeak I2xO-year Peak

RPV Fluence RPV Fluence

O.OE+O0 2.0E÷19 4.OE+19 6,0E+19 8.OEu19 1.OE+20 1.2E+20 1.4E+20

Fluence [n/cm2l

High fluence data well populatedLinde 1092 weld metal

for SA-533 base metal (high Cu), but not for

For Plant A weld metal, projected embrittlement shift using expected RG 1.99Rev. 3 or Kirk model is bound by RG 1.99 Rev.2 13

Example: Additional Capsule TestRecommended for Plant B

2500

225 CNlire: 056-er~

RPVFI~~ P: 0.0

0.O E.0 1.O .99 Re 3 (.0E.-19 6 . E1a IME-- 1.3 E2 0 . E 2 . E 2

122

100ce[/c

250

225

200

175

'150

a125

100

75

so25

0.

---- RG 1.99 Rev 2 Ni- O.07P: 0.008

_RG 1.99 Rev 3 (10CFR50.61a) Mn: 1.17T.- 643 F

Rux f Rec Cap: 1.24E11

Based on Test Reactor Data Mt: CE

60-yar Peak "ar .PeakRPV Flu.-no RPV Ru~ne-

400 2.OE-194.OE+19 6.OE+19

Fluence

8.OE-19 1.0E&20 1.2E+20 1.4E+20

[n/cm2]

SA-533 Base Metal Linde 91 Weld Metal

High fluence data well populated for SA-533 base metal (low Cu), but notfor Linde 0091 weld metal (low Cu)For Plant B base and weld metal, projected embrittlement shift usingexpected RG 1.99 Rev. 3 or Kirk model significantly exceeds the ETC ofRG 1.99 Rev.2

14


* CRVSP recommends that two plants move a capsule fromstorage for further irradiation to fill specific material/fluence data gaps

- In both cases, plant has withdrawn a capsule that hasexceeded the 60-year peak RPV fluence and placed itin storage

- Neither plant has tested a 60-year capsule-Re-irradiation not required to meet the requirements of

a 60-year license, but would presumably be required tomeet an 80-year license

-Highest fluence capsule in storage should be insertedback in the reactor and then tested @ 2x60-year peakRPV fluence

15

Summary of Plants Affected

* No change to the RVSP of 30 plants

* 10 plants that have already tested their 60-year capsuleare being asked to test an additional capsule

* 2 plants are being asked to take a capsule from storageand put it back in for further irradiation

* Remaining 27 plants will be asked to test their finalplanned capsule at a later date (80-year or 2x60)

* Based on current draft of CRVSP, final.numbers subjectto change 16

Data Summary - Current vs. Recommended

Current Plana Tested m= 2015o []2020 o] 2025

30

25-

S20-

015

10F

3-4 4-5 5-6 6-7 7-8 8-9 9-10 10+

Capsule Fluence at Withdrawal (xlO' 9 n/cm2

)

" To date-34 capsules have been

tested at >3x1 019 n/cm 2

and only 8 capsules at>5x1 09vn/cm2

" Current Plan-30 more capsules above

>3x1 019 n/cm 2 and only5 capsules at >6x10'9n/cm 2

* Recommended Plan-43 more capsules above

>3x1 019 n/cm 2 and 22capsules at >6x1 019n/cm 2 17

Recommended Plan

Ei Tested m2015 o2020 o2025

30-

25-

'o 20-

a 10-

5-

3-4 4-5 5-6 6-7 7-8 8-9

Capsule Fluence at Withdrawal (xl0' n/cm')

9-10 10+

Results

* Number of additional capsules to be tested by 2025 at orabove the stated fluence

Capsule Fluence Current Plan Recommended Plan

3xl019 n/cm2 30 43

6xl01 9 n/cm 2 5 22

8xl0 19 n/cm2 0 7

9xl019 n/cm 2 0 3

18

Status & Implementation Plans

" Draft report is under internal review

" Intent is to present to Utility management in 2011 forapproval and implementation action

-Current recommended implementation - revision toplant Appendix H program

" After approval, affected plants submit requests to amendAppendix H programs per recommended schedule

- 50.59 review for compliance with 1OCFR50 App Hand GALL Rev 2

- Plants submit letter notification to NRC

- NRC respond/accept

19

Regulatory Interaction

NRC respond/accept licensee capsule withdrawalschedule change

- Support timely review and approval of Appendix Hprogram changes

- License renewal review consistent with GALL Rev. 2

- Support Coordinated Surveillance Program

20

Questions?

21

PWROG I

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0Own e so

ELECTRIC POWER

RESEARCH INSTITUTE

MRP

Alternate PTS Rule ImplementationGuidancePWROG MSC

Nathan Palm, Westinghouse

January 20411

1

Alternate PTS Rule Implementation Guidance

Background

e Alternate PTS Rule 10 CFR 50.6 1 a was issued on January 4, 2010.° Corrections issued February 3, 2010 and November 26, 2010" Alternate PTS Rule has several areas without precise regulatory requirements

" (e)(1) - "Test results from the volumetric examination may be adjusted toaccount for the effects of NDE-related uncertainties."

* (e)(4) - "The licensee shall perform analyses to demonstrate that the reactorvessel will have a TWCF of less than 1 x 10-6 per reactor year if the ASMECode, Section XI volumetric results..." do not satisfy the flaw limits.

" (f)(6)(vi) - "If any of the criteria described in paragraph (f)(6)(v) (surveillancedata checks) ..are not satisfied... The licensee shall propose AT30 and RTMAx-xvalues considering their plant specific surveillance data..."

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Alternate PTS Rule Implementation Guidance

I implementation Guidance* What are NRC plans for developing regulatory guidance?

e Timeframe?* PWROG and MRP would welcome opportunity to contribute to Reg. Guide

development.

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ELECTRIC POWERRESEARCH INSTITUTE

MRP

Pressure-Temperature Limit Topics

Chris Kiefer, Ameren

PWROG Core Team Chair

January 6, 2011I


* Background

Pressure-Temperature Limits are producing more restrictive operatingwindows for utilities as reactor vessels are subject to increasing levels offluence.

o Very conservative methodology of 10 CFR 50 Appendix G and ASME SectionXI Appendix G result in very conservative P-T Limits.

o Conservative P-T limits are protected by the Low Temperature OverpressureProtection System, further restricting the operating window.

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Pressure-Temperature Limit TopicsFlange Notch

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"Flange Notch"

* Results from 10 CFR 50 Appendix G.* Based on KIA instead of the more recently accepted Kic methodology.• WCAP- 15315 provides the technical justification for the eliminationof the "Flange Notch" with the KIc approach.

* The LTOPS is often limited by the "Flange Notch."* A 10% relaxation on the LTOPS pressure is permitted with the use ofthe KIA methodology, but not the Kic per Code Case N-64 1. The10% cannot be applied to the "Flange Notch" with Kic and oftenresults in overly-conservative allowable pressures.

* The NRC has allowed a flange notch exemption on a plant-specificbasis, but not generic.

" Risk-informed Appendix G does not address the "Flange Notch."

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o Discussion

o Status of Rulemaking on 10 CFR 50 Appendix G?

o Risk-informed

a Flange Notch

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PWROG Cl-'11 IELECTRIC RESEARCH POWER INSTITUTE · 2012-12-04 · Stephen Byrne (Westinghouse) and...

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Transcript of PWROG Cl-'11 IELECTRIC RESEARCH POWER INSTITUTE · 2012-12-04 · Stephen Byrne (Westinghouse) and...