NRC 2003-0018, Request for Exemptions to 10 CFR 50.61, Appendices G and H to 10 CFR 50, and Approval of PTS Application for Pbnp Unit 2 and Proposed Heatup and Cooldown Limit Curves for Pbnp Units 1 & 2

Request for Exemptions to 10 CFR 50.61, Appendices G and H to 10 CFR 50, and Approval of PTS Application for Pbnp Unit 2 and Proposed Heatup and Cooldown Limit Curves for Pbnp Units 1 & 2

ML080440221
Person / Time
Site:	Point Beach
Issue date:	03/03/2003
From:	Cayia A Nuclear Management Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NRC 2003-0018, TAC MA8585
Download: ML080440221 (128)

Text

NKC>

Commrned to Nuclear Excellence Point Beach Nuclear Plant Operated by Nuclear Management Company, LLC NRC 2003-0018 10 CFR 50.12 March 3,2003 U S . Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 DOCKETS 50-266 AND 50-301 POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 REQUEST FOR EXEMPTIONS TO 10 CFR 50.61, APPENDICES G AND H TO 10 CFR 50, AND APPROVAL OF PTS APPLICATION FOR PBNP UNIT 2 AND PROPOSED HEATUP AND COOLDOWN LIMIT CURVES FOR PBNP UNITS 1 AND 2 In accordance with the provisions of 10 CFR 50.12, "Specific Exemptions", Nuclear Management Company, LLC (NMC) is submitting a request for permanent exemption from certain requirements of I O CFR 50.61, "Fracture Toughness Requirements for Protection Against Thermal Shock Events", 10 CFR 50, Appendix G, "Fracture Toughness Requirements", and 10 CFR 50, Appendix H, "Reactor Vessel Material Surveillance Program Requirements", for the Unit 2 reactor vessel. Attachment 1 to this letter provides the justification for this exemption request.

Attachment 2 provides a summary of the regulatory commitments made in this submittal.

The requested exemptions would allow use of a different method, the Master Curve Methodology, for determining the adjusted RTNDT(reference nil-ductility temperature) of the Point Beach Nuclear Plant Unit 2 (PBNP Unit 2) reactor vessel limiting circumferential weld metal. This method is used for the Pressurized Thermal Shock (PTS) screening evaluation.

The NRC granted similar exemptions to the Kewaunee Nuclear Power Plant on February 21,2001 (TAC NO.MA 8585).

In association with this request for exemptions, NMC has reassessed PBNP Unit 2 compliance with 10 CFR 50.61 for end of license (EOL) conditions. The new PTS evaluation is provided in AT1 Consulting Report 021-030-2003-1, "Master Curve Fracture Toughness Application for Point Beach Nuclear Plant Unit 2", dated January 2003. A copy of the AT1 Consulting report is provided as Enclosure 1.

NMC hereby requests the NRC to review and approve the revised PTS evaluation at EOL for Unit 2.

To demonstrate consistency with the Master Curve Methodology used in the PTS evaluation for Unit 2, NMC has also reassessed the PBNP Units 1 and 2 compliance with 10 CFR 50.60 Appendix G for EOL. Enclosure 2 to this letter is a copy of WCAP-15976. Revision 0, "Point Beach Units 1 and 2 Heatup and Cooldown Limit Curves for Normal Operation", dated February 2003, which provides new 34 EFPY (EOL) pressure and temperature limit cuwes (PR curves).

6590 Nuclear Road Two Rivers, Wisconsin 54241 Telephone. 920.755 2321 AOQl

NRC 2003-0018 Page 2 NMC hereby requests the NRC to review and approve the new P/T curves at EOL for Units 1 and 2.

It is noted that although the Master Curve toughness application for the PBNP Unit 2 limiting circumferential weld has been considered in the P/T cume evaluation, there is no effect on the resulting P/T curves since the limiting materials for these curves are the intermediate and lower shell axial welds from PBNP Unit 1.

NMC requests approval of these exemptions by October 2003. The currently calculated fluence values for operation of Unit 1 are valid until October 30, 2003 and are valid for Unit 2 until October I , 2008. The currently approved methodology can be used to extend the effective date; however, the exemptions are desired prior to expiration of these values to allow issuance of the new P / l curves should consideration of the new Master Curve methodology be required.

Prior to crediting use of the new PTT curves, a revised reference in PBNP Technical Specifications (TS) 5.6.5,Reactor Coolant System (RCS) Pressure and Temperature Limits Report will be needed. A license amendment request to affect this administrative change in TS 5.6.5 will be submitted separately.

1 declare under penalty of perjury that the foregoing is true and correct.

Executed on March 3.2003.

Attachment:

1 Justification for Exemption Request 2 List of Regulatory Commitments

Enclosures:

1 AT1 Consulting Report 021-030-2003-1, January 2003 2 WCAP-15976, February 2003 cc: (w/ enclosures)

Project Manager, Point Beach Nuclear Plant, NRR, USNRC (w/o enclosures)

Regional Administrator, Region I l l , USNRC NRC Resident Inspector - Point Beach Nuclear Plant PSCW

JUSTIFICATION FOR EXEMPTION REQUEST POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2

NRC 2003-00l8 Attachment 1 Page 1 of 7

1.0 INTRODUCTION

In accordance with the provisions of 10 CFR 50.12, Specific Exemptions, Nuclear Management Company, LLC (NMC) is submitting a request for exemption from certain requirements of 10 CFR 50.61, Fracture Toughness Requirements for Protection Against Thermal Shock Events, 10 CFR 50, Appendix G, Fracture Toughness Requirements, and 10 CFR 50, Appendix H, Reactor Vessel Material Surveillance Program Requirements. The requested exemptions would al!ow a different method, the Master Curve Methodology, for determining the adjusted RTNoT(reference nil-ductility temperature) of the Point Beach Nuclear Plant Unit 2 (PBNP Unit 2) reactor vessel limiting circumferential weld metal. This method is used for the Pressurized Thermal Shock (PTS) screening evaluation.

2.0 BACKGROUND

10 CFR 50.60 and 10 CFR 50.61 establish criteria that ensure that each reactor vessel has adequate fracture toughness. When these rules were first promulgated fracture toughness specimens were too large to be used in reactor vessel radiation surveillance capsule programs.

Therefore, smaller Charpy V-notch specimens were used to estimate and monitor fracture toughness.

The latest Charpy-based toughness evaluation following current regulations for the PBNP Unit 2 limiting circumferential weld metal indicates that the projected value of RTprs at end-of-license (EOL) is close to (but below) the pressurized thermal shock (PTS) screening criterion of 300°F

( I O CFR 50.61). For bounding and evaluation purposes, conditions at end of license extended (EOLE) were also projected. The PTS screening criteria will be exceeded for the projected EOLE fluence, which assumes removal of hafnium fluence suppression assemblies and planned power up-rates. Therefore, NMC performed a bounding evaluation of Master Curve fracture toughness data for assuring reactor pressure vessel (RPV) integrity for PBNP Unit 2 at EOL and out to EOLE. This application represents the secund use of the Master Curve Methodology in the nuclear industry for a reactor pressure vessel with a beltline weld as the limiting material. The Kewaunee Nuclear Power Plant, also operated by NMC, was the first application.

This document summarizes the technical basis and justifications for the exemption requests to use ASME Code Cases N-629 and N-631, ASTM E185-98. ASTM E-1921-02, and the methodology described in AT1 Consulting Report 021-030-2003-1 for establishing EOL and EOLE indexing reference temperature values for assessment of the integrity of the PBNP Unit 2 reactor vessel.

3.0 PROPOSED EXEMPTIONS The three exemptions requested by NMC address portions of the following regulations:

(IAppendix

) G to IO CFR Part 50, which sets forth fracture toughness requirements for ferritic materials of pressure-retaining components of the reactor coolant pressure boundary of light water nuclear power reactors to provide adequate margins of safety during any condition of normal operation, including anticipated operational occurrences and system hydrostatic tests, to which the system may be subjected over its service lifetime;

NRC 2003-0018 Page 2 of 7 (2) 10 CFR 50.61, which sets forth fracture toughness requirements for protection against pressurized thermal shock (PTS); and (3) Appendix H to 10 CFR Part 50, which requires the establishment of a RPV material surveillance program.

The exemption from Appendix G to 10 CFR 50 is to replace the required use of the existing Charpy V-notch and drop-weight-basedmethodology and allow the use of an alternate methodology to incorporate the use of fracture toughness test data for evaluating the integrity of the PBNP Unit 2 circumferential beltline weld based on use of the 2002 Edition of American Society for Testing and Materials (ASTM) Standard Method E 1921 (E 1921-02) and American Society for Mechanical Engineering (ASME) Code Case N-629. The exemption is required since Appendix G to Section XI of the ASME Code pursuant to 10 CFR 50.55(a) requires the use of a methodology based on Charpy V-notch and drop weight data.

The exemption from 10 CFR 50.61 is to use an alternate methodology to allow the use of fracture toughness test data for evaluating the integrity of the PBNP Unit 2 limiting circumferential beltline weld based on the use of ASTM E 1921-02 and ASME Code Case N-629. The exemption is required because the methodology for evaluating RPV material fracture toughness in 10 CFR 50.61 requires the use of Charpy V-notch and drop weight data for establishing the PTS reference temperature (RTprs).

Finally, the exemption from Appendix H to 10 CFR 50 is to modify the basis for the PBNP Unit 2 surveillance program to allow the acquisition and use of fracture toughness data instead of the Charpy V-notch impact testing required by Appendix H to 10 CFR 50. The exemption is required because Appendix H to 10 CFR 50 does not address the testing of surveillance specimens for direct measurement of fracture toughness. A second reason for the exemption relates to a supplemental surveillance capsule. Due to the need for additional fracture toughness data for the PBNP Unit 2 weld metal at fluence levels extending out to EOLE, a supplemental capsule has been added to the surveillance program for PBNP Unit 2. This capsule has been installed in the highest lead factor location and includes other RPV beltline materials. The capsule is designed for Master Curve fracture toughness testing and evaluation at t h e projected EOLE fluence, so that the integrity of the RPV will be directly validated with the testing of this capsu!e should extended operation be considered. The composition of materials, specimen types, and estimated schedule for removal of this new capsule are addressed in AT1 Consulting Report 021-030-2003-1.

A tabular summary of the requested exemptions and the proposed alternatives are shown below.

NRC 2003-0018 Page 3 of 7

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Description Existing Requirement Proposed Alternative Determination of 10 CFR 50.61 and Appendix ASME Code Case N-629, adjustedlindexing reference G to 10 CFR Part 50 ASME Code Case N-631, temperatures AT1 Consulting Report 021-030-2003-1, and WCAP-15976 I

Use of the latest edition of App H to Part 50 specifies (1) ASTM E185-98 allows ASTM E185-98 use of ASTM E185-73,-79, - use of ASTM E 1921-02 for 82 for testing of surveillance testing of surveillance materials capsule material; (2) Use fracture toughness surveillance data from PBNP Unit 2 supplemental surveillance capsule for verification of EOLE toughness properties Alternative testing methods Appendices G and H to Part ASTM E1921-02, AT1 for determination of fracture 50 specifies Charpy V-Notch Consulting Report 021-030-toughness impact and drop weight 2003-1 and WCAP-15976 testing

4.0 TECHNICAL ANALYSIS

The attached AT1 Consulting Report 021-030-2003-1 provides the detailed technical analysis and basis for the Master Curve application at PBNP Unit 2. A general summary is presented here.

The PBNP Unit 2 RPV limiting weld metal heat 72442 was not included in the current surveillance program for PBNP Unit 2, but it was irradiated as part of the 6&W Owners Group integrated surveillance program. The latest projections based on Charpy impact testing, when analyzed following NRC guidelines and rules, indicate that this weld will reach the PTS screening criterion limit before EOLE. Therefore, fracture toughness testing of other irradiated surveillance specimens (from two different welds fabricated using weld wire 72442) has been performed and analyzed using the Master Curve methodology following ASME Code Cases N-629 and N-631. The evaluation performed involves extrapolation to EOL and EOLE fluences and shows that the RPV limiting weld metal has more than adequate toughness for operation out to EOLE and beyond. These projections will be confirmed by additional testing of weld heat 72442 from the B&W Owners Group Master Integrated Reactor Vessel Materials Surveillance Program (MIRVP) prior to reaching the EOL fluence at PBNP Unit 2. A supplemental surveillance program will be designed and implemented at PBNP Unit 2 that includes the limiting weld metal for future evaluation using the Master Curve methodology. Should extended operation be considered, the testing of this supplemental capsule at a fluence corresponding to EOLE will confirm the toughness condition for the PBNP Unit 2 RPV weld at about 38 EFPY, which is well before EOLE is reached.

NRC 2003-0018 Page 4 of 7 The following observations and conclusions are documented in the attached AT1 Consulting Report 021-030-2003-1 for the PBNP Unit 2 limiting beltline weld metal:

The latest Charpy-based toughness evaluation following current regulation for the PBNP Unit 2 limiting circumferential weld metal indicates that the PTS screening criterion of 3OO0Fwill be reached before EOLE when future plant operation is considered (removal of hafnium fluence suppression and planned power up-rates).

Application of the Master Curve methodology for the PBNP Unit 2 weld metal requires extrapolation (from the three available surveillance irradiations) to the RPV EOLE fluence. The extrapolation can be performed following several different approaches.

Three approaches were evaluated: ( I ) use of measured initial RTT, and adding Charpy shift; (2) use of measured initial RTToand adding the shift in RTro due to irradiation; and (3) use of the measured irradiated R T T values

~ directly without projection from zero fluence. All methods show that the EOLE R T ~value T ~ is less than the PTS screening limit of 30OOF. Method 1 somewhat follows the current regulatory practice and is the most conservative. Method 2 was evaluated following the Kewaunee SEI and the resulting projections in ART were substantially less than Method 1. Method 3 is the most accurate method, and the results obtained applying this direct measurement approach reveal that Method 2 is quite conservative.

The Margin term was chosen depending upon the analysis approach discussed above.

For Method 1, Margin was based on three uncertainties: material variability based on a Monte Carlo study from BAW-2308 of weld heat 72442 non-irradiated data (oMc= 9.3"F),

the uncertainty in determining To from ASTM E 1921-02 (oT0= 7.4"F), and the current regulatory value for weld metal Charpy shift (oA= 28°F);oMcand C T T ~are combined to give a measure of the uncertainty in initial properties (01 = 11.9"F). Method 2 used the Margin specified by the NRC in the Kewaunee SE, which used a larger oI(14°F) and the same aAof 28°F. Method 3 used a more complete uncertainty analysis: material variability ( Q M ~= 9.3"F as above), determination of irradiated To ( C T T ~= 10.7"F).

Cu content (ocu= 1.6-1.7"F), Ni content (aN,= 4.1-4.2"F), irradiation temperature (oTln= 6.9-8.9"F), fluence (aot= 13.2-125°F). and fluence projection (oPq = 1.0-1.6"F).

Remaining consistent with industry practice, an approximate 95% statistical level (or two sigma) Margin was chosen, where the individual uncertainties were combined as the square root sum of the squares.

Since there was a need to extrapolate to higher fluence levels (higher than where current fracture toughness measurements exist) to assess PTS and pressure-temperature operating curves, the current Regulatory fluence function for CVN-based predictions was used for the Master Curve approach.

The supplemental surveillance program utilizes irradiation of the limiting weld metal heat in a new capsule that will be available for testing near the time corresponding io 38 EFPY for the RPV. The direct measurement of fracture toughness for key weld metal will be evaluated at a fluence near to the projected EOLE. Fracture toughness data from the B&W Owners Group on this same weld metal will be available around 2008. This B8W Owners Group data should correspond closely to the PBNP Unit 2 EOL fluence for the limiting RPV weld.

NRC 2003-0018 Page 5 of 7 Conclusion Use of the Master Curve methodology, extrapolated to EOL and EOLE fluences, shows that the RPV limiting weld metal meets PTS screening criteria out to EOLE and beyond. These projections will be confirmed by additional testing of weld heat 72442 from the B&W Owners Group MIRVP prior to reaching the EOL fluence at PBNP Unit 2. A supplemental surveillance program will be designed and implemented at PBNP Unit 2 that includes the limiting weld metal for future evaluation using the Master Curve methodology. Should extended operation be considered, the testing of this supplemental capsule at a fluence corresponding to EOLE will confirm the toughness condition for the PBNP Unit 2 RPV weld at about 38 EFPY, which is well before EOLE is reached.

5.0 REGULATORY ANALYSIS

5.1 No Significant Impact Determination and Environmental Evaluation In accordance with the provisions of 10 CFR 50.12, 'Specific Exemptions", Nuclear Management Company, LLC (NMC) is submitting a request for exemption from certain requirements of 10 CFR 50.61, "Fracture Toughness Requirements for Protection Against Thermal Shock Events", 10 CFR 50, Appendix GI'Fracture Toughness Requirements", and 10 CFR 50, Appendix H, 'Reactor Vessel Material Surveillance Program Requirements". The requested exemptions would allow a different method, the Master Curve Methodology, for determining the adjusted RTNoT(reference nil-ductility temperature) of the PBNP Unit 2 reactor vessel limiting circumferential weld metal. This method is used for the Pressurized Thermal Shock (PTS) screening evaluation.

NMC has evaluated the proposed exemption against the criteria in 10 CFR 51.32 and has determined that the operation of PBNP in accordance with the proposed exemption presents no significant impact. Use of the Master Curve methodology, which has previously been approved by the Commission for use at the nearby Kewaunee Nuclear Power Plant, shows that the RPV limiting weld metal meets PTS screening criteria. The underlying purpose of 10 CFR 50.61 and Appendices G and H to 10 CFR 50, which is to establish criteria that ensure that each reactor vessel has adequate fracture toughness, continues to be achieved.

Operation of PBNP in accordance with the proposed exemption will not significantly increase the probability or consequences of accidents, no changes are being made in the types of effluents that may be released off site, and there is no significant increase in occupational or public radiation exposure. Therefore, operation of PBNP in accordance with the proposed exemption does not result in any significant radiological environmental impacts.

With regard to potential nonradiological impacts, the proposed action does not have a potential to affect any historic sites. It does not affect nonradiological plant effluents and has no other environmental impact. Therefore, there are no significant nonradiological environmental impacts associated with the proposed action.

NRC 2003-0018 Page 6 of 7 Conclusion Since there are no significant radiological or nonradiological environmental impacts associated with the proposed action, we conclude that the proposed exemption will not have a significant effect on the quality of the human environment. Therefore, as provided in 10 CFR 51.32, an environmental impact statement need not be prepared.

5.2 Commitments The actions committed to by NMC in this document are listed in Attachment 2. Any other statements in this submittal are provided for information purposes and are not considered to be regulatory commitments.

5.3 Applicable Regulatory Requirements 10 CFR 50.12(a) states that the Commission may grant exemptions from the requirements of the regulations contained in 10 CFR 50 that are:

(1) authorized by law; (2) will not present an undue risk to the public health and safety; (3) consistent with the common defense and security; and (4) special circumstances, as listed in 10 CFR 50.12(a)(2), are present.

This exemption request meets the criteria set forth in 10 CFR 50.12, as discussed herein.

Additional technical bases for the proposed exemptions are provided as Enclosure 1 to this letter.

1. The requested exemption is authorized by law.

No law exists which precludes the activities covered by this exemption request. 10 CFR 50.60(b) allows the use of alternatives to 10 CFR 50.Appendices G and H when an exemption is granted by the Commission under 10 CFR 50.12.

2. The reauested exemDtion does not present an undue risk to the public health and safety.

10 CFR 50 Appendices G and H specify that surveillance capsules shall be tested in accordance with ASTM E 185-73, -79, and -82. The latest version of ASTM E-185-98 encourages that supplemental fracture toughness testing be conducted in accordance with procedures and requirements of Practice E636, Method E-1820, or Method E-1921 when the surveillance materials exhibit marginal properties. Fracture toughness testing of weld metal heat 72442 has been performed to satisfy the requirements established in accordance with ASTM E-1921.

The use of this proposed approach ensures that the intent of the requirements specified in 10 CFR 50.61 are satisfied. Therefore, this exemption does not present an undue risk to the public health and safety.

3. The requested exemption is consistent with the common defense and security.

The common defense and security are not endangered by this exemption request.

NRC 2003-0018 Page 7 of 7

4. Special circumstances are present which necessitate the reauest for an exemption to the requlations of 10 CFR 50.61.

Pursuant to 10 CFR 50.12(a)(2), the Commission will consider granting an exemption to the regulations of 10 CFR 50 if special circumstances are present. This exemption request meets the special circumstances described in 10 CFR 50.12(a)(2)(ii):

Application of the regulation in the particular circumstances would not serve the underlying purpose of the rule or is not necessary to achieve the underlying purpose of the rule.

The underlying purpose of 10 CFR 50.61 and Appendix G to 10 CFR 50 is to establish criteria that ensure that each reactor vessel has adequate fracture toughness.

Westinghouse Electric Corporation and AT1 Consulting have prepared the enclosed reports to assess and document the integrity of the PBNP Unit 2 reactor vessel relative to the requirements and underlying purpose of 10 CFR 50.61, and Appendices G and H to 10 CFR 50. These reports provide the technical justification for the exemption requests. AT1 Consulting Report 021-030-2003-1 provides an updated PTS evaluation using Master Curve methodology for PBNP Unit 2, showing compliance with PTS screening criteria through EOLE. Should extended operation be considered, a supplemental surveillance program to validate EOLE fracture toughness properties is also described. For consistency, the Master Curve methodology has been considered in the development of updated PTT curves for PBNP Units 1 and 2, presented in WCAP15976. Together, these reports demonstrate that the alternate methodology (Master Curve methodology) achieves the underlying purposes of the regulatory tules from which exemptions are requested; therefore, the exemption requests are justified.

NRC 2003-0018 Page 1 of 1 LIST OF REGULATORY COMMITMENTS The following table identifies those actions committed to by NMC in this document. Any other statements in this submittal are provided for information purposes and are not considered to be regulatory commitments.

COMMITMENTS Due DatelEvent NMC will supplement this submittal with a license amendment April 2003 request to revise the reference in PBNP Technical Specifications (TS) 5.6.5, Reactor Coolant System (RCS) Pressure and Temperature Limits Report.

NMC will confirm the projections for the RPV limiting weld metal by Prior to reaching the additional testing of weld heat 72442 from the 68W Owners Group EOL fluence at Master Integrated Reactor Vessel Materials Surveillance Program. PBNP Unit 2 06/2OlO NMC will design and implement a supplemental surveillance program 02/2006 at PBNP Unit 2 that includes the limiting weld metal for future evaluation using the Master Curve methodology.

Should extended operation be considered, NMC will test the Should extended supplemental capsule at a fluence corresponding to EOLE to confirm operation be the toughness condition for the PBNP Unit 2 RPV weld at about 38 considered EFPY, which is well before EOLE is reached.

NNAL REPORT A Ti-021-030-2003-1 MASTER CURVE FRACTURE TOUGHNESS APPLICATION FOR POINT BEACH NUCLEAR PLANT UNIT 2 ATI Consulting W. L. Server J. R. Pfefferle January 2003 AT1 Consulting 6773 Sierra Court, Suite C Dublin, CA 94568

AT1 Consulting Report ATI-021-030-2003-1 Fracture Toughness Material Property Report for PBNP-2 EXECUTIVE

SUMMARY

The latest Charpy-based toughness evaluation following current regulations for the Point Beach Nuclear Plant, Unit 2 (PBNP-2) limiting circumferential weld metal indicates that the pressurized thermal shock (PTS) screening criterion of 300'F (10 CFR 50.61) is close to the projected value of RTms at end-of-life (EOL), but will be exceeded before end-of-life extension (EOLE) when future plant operation is considered (removal of hafnium flux reduction and planned power up-rates). This report summarizes application of Master Curve fracture toughness data for assuring reactor pressure vessel (RPV)integrity for PBNP-2 at EOL and out to EOLE. This application represents the second use of the Master Curve Methodology in the nuclear industry for a reactor pressure vessel with a beltline weld as the limiting material. The Kewaunee Nuclear Power Plant, also operated by Nuclear Management Company, LLC, was the first application. The fracture toughness data for the limiting weld metal presented in this report were generated in part under the B&W Owners Group Master Integrated Reactor Vessel Surveillance Program and additionally funded by Nuclear Management Company, LLC.

The Kewaunee Nuclear Power Plant received approval from the Nuclear Regulatory Commission to apply the Master Curve Methodology using the ASME Code alternative RTT, indexing parameter for the limiting weld metal in that RPV. EPRl has published two support documents (EPRI TR-I000707 and EPRI TR-108390, Rev. 1) endorsing and validating the use of the Master Curve approach. These documents provide the technical bases for ASME Code Cases N-629 and N-63 1 in which the alternative reference temperature, RTT,, is defined and used.

This report provides a summary of the RTT, methodology used to determine the adjusted reference temperature (ART) for the irradiated RPV weld metal in PBNP-2. Three approaches were evaluated: (1) use of measured initial RT-r, and adding Charpy shift; (2) use of measured initial RTT, and adding the shift in RTT, due to irradiation; and, (3) use of the measured irradiated RTT, values directly projecting from the measured data to higher fluence levels. All three methods show that the EOLE R T m value is less than the PTS screening limit of 300°F. Method 1 follows the current regulatory practice except use of the initial non-irradiated RTT,, and is the most conservative. Method 2 follows the NRC evaluation in the Kewaunee SE, and the resulting projections in ART are substantially less than Method 1. Method 3 is the most accurate method, and the results obtained applying this direct measurement approach reveal that Method 2 is quite conservative. Several key technical areas are addressed in this report; a bias correction for three-point bend fracture toughness specimens, accuracy and variability of To values measured using ASTM E l 921-02 and Monte Carlo simulations, and development of a margin approach for RTT, for all three methods that is consistent and meets the intent of current regulatory methods.

Due to the need for additional fracture toughness data for the PBNP-2 weld metal at fluence levels extending out to EOLE, a supplemental capsule has been added to surveillance program for PBNP-2. This capsule has been installed in the highest lead i

AT1 Consulting Report ATI-021-030-2003-1 Fracture Toughness Material Property Report for PBNP-2 factor location and includes other RPV beltline materials. The capsule is designed for Master Curve fracture toughness testing and evaluation at the projected EOLE fluence, so that the integrity of the RPV will be directly validated with the testing of this capsule.

The composition of materials, specimen types, and estimated schedule for removal of this new capsule are addressed in this report. Additional fracture toughness data on weld heat 72442 from the B&W Owners Group also will be available around 2008, which will provide results at an intermediate fluence results near EOL.

The projections for PTS are provided in this report based on applying the Master Curve methodology to the limiting weld. Using this methodology, the PBNP-2 RPV is projected to stay below the PTS screening criteria through and beyond EOLE. These PTS projections are based on the most limiting assumptions for future plant operation, including a substantial upgrade of thermal power and removal of hafnium flux reduction assemblies from future core designs.

ii

AT1 Consulting Report ATI-021-030-2003-1 Fracture Toughness Matenal Property Report for PBNP-2 ACKNOWLEDGMENTS The authors wish to acknowledge the help of Brian Burgos of AT1 Consulting and Randy Lott of Westinghouse Electric Company, LLC in helping to consolidate the fracture toughness test data and in reviewing technical details in this document.

Ill

AT1 ConsuRing Report ATl-021-030-2003-1 Fracture Toughness Material Property Report for PBNP-2 TABLE OF CONTENTS EXECUTIVE

SUMMARY

I ACKNOWLEDGMENTS iii LIST OF TABLES V LIST OF FIGURES vii 1 INTRODUCTION 1-1 2 REVIEW OF CVN-BASED PTS STATUS FOR PBNP-2 2- 1 2.1 REVIEW OF BELTLTNE MATERIALS AND FLUENCES 2- 1 2.2 ART PROJECTIONS FOR EOL AND EOLE 2-2 3 FRACTURE TOUGHNESS TESTING AND RESULTS 3-1 4 APPLICATION OF FRACTURE TOUGHNESS RESULTS TO PBNP-2 4-1 4.1 MASTER CURVE APPLICATION METHODOLOGIES 4-1 4.2 MEASURED INITIAL RTT, AND CHARPY-SHIFT APPROACH 4-3 4.3 MEASURED INITIAL RTT, AND RTT,-SHIFT APPROACH 4-4 4.4 DIRECT MEASUREMENT OF IRRADIATED RTT, 4-7 5

SUMMARY

OF SUPPLEMENTAL SURVEILLANCE PROGRAM 5- 1 5.1 HISTORY OF EXISTING PROGRAM 5-1 5.2 REVIEW OF REMAINING CAPSULES 5-1

5.3 DESCRIPTION

OF NEW SUPPLEMENTAL SURVEILLANCE CAPSULE 5 -2 5.4 SUPPLEMENTAL CAPSULE IRRADIATION AND WITHDRAWAL SCHEDULE 5 -2 6

SUMMARY

AND CONCLUSIONS 6- 1 7 REFERENCES 7- 1 iV

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AT1 Consulting Report ATI-021-030-2003-1 Fracture Toughness Material Property Report lor PBNP-2 LIST OF TABLES Table 2-1 Summary of the PBNP-2 RPV Beltline Materials 2-4 Table 2-2 Projections for Fluence for the PBNP-2 RPV at EOL and EOLE 2-4 Table 2-3 Summary of Weld Wire 72442 CdNi Measurements 2-5 Table 2-4 ART/RTms Projections at EOL (34 EFPY) Based on CVN Methodology for the PBNP-2 W V 2-7 Table 2-5 ART/RTns Projections at EOLE (53 EFPY) Based on CVN Methodology for the PBNP-2 RPV 2-8 Table 3-1 Fracture Toughness Testing for Weld Heat 72442 3-3 Table 3-2 Test Results for Unirradiated Weld Metal SA-1484 (0.394T-3PB Specimens, No Sidegrooves) 3-3 Table 3-3 Test Results for Irradiated Weld Metal SA-1484, Capsule DB-A3 3-4 Table 3-4 Test Results for Unirradiated Weld Metal WF-67 (50 Hour Stress Relief) 3-5 Table 3-5 Test Results for Unirradiated Weld Metal WF-67 (1 1 Hour Stress Relief)

(0.394T-3PB Specimens, No Sidegrooves) 3-5 Table 3-6 Test Results for Irradiated Weld Metal WF-67, Capsule DB-Ll 3-6 Table 3-7 Test Results for Irradiated Weld Metal WF-67, Capsule CR3-LG2 3-6 Table 3-8 To Determinations from Measured Fracture Toughness Data for Weld 72442 3-7 Table 3-9 Comparison of To Values from Different Specimen Types with No Bias Correction for Weld 72442 3-7 Table 4-1 Comparisons of Reference Temperature Methods and Results 4-1 1 Table 4-2 NRC Method for Correcting ARTT, Values for Chemistry and Irradiation Temperature Differences 4-12 Table 4-3 ASTM E 900-02 Corrections of ARTT, Values for Chemistry and Irradiation Temperature Differences 4-12 V

AT1 Consulting Report ATI-021-030-2003-1 Fracture Toughness Matenal Properly Report for PBNP-2 Table 4-4 Summary of ARTTOValues Corrected for Differences in Chemistry and Irradiation Temperature and Calculation of CFT, 4-12 Table 4-5 NRC Method for Determining ART at EOL Fluence 4-13 Table 4-6 NRC Method for Determining ART at EOLE Fluence 4-13 Table 4-7 Calculations of Prediction Uncertainties for ASTM E 900 Model Parameters at EOL and EOLE 4-13 Table 5-1 PBNP-2 Surveillance Program 5-4 Table 5-2 Types and Number of Specimens in the PBNP-2 Surveillance Test Capsules 5 -4 Table 5-3 Remaining MIRVP Capsules Containing Heat 72442 5-5 Table 5-4 PBNP-2 and Other Materials and Specimen Types in Supplemental Capsule 5-5 Table 5-5 PBNP-2 Fuel Cycles and Estimated Surveillance Capsule Fluence 5-6 vi

7- - -- -- -

AT1 Consulting Report ATI-021-030-2003-1 Fracture Toughness Matenal Property Report for PBNP-2 LIST OF FIGURES Figure 4-1 Comparisons of Projected ART from Master Curve Methods and Charpy Methodologies for Weld Wire 72442 (Measured Master Curve Data Do Not Have Margin Added) 4-14 Figure 5-1 Arrangement of Surveillance Capsules in the Reactor Vessel 5-7 vii

AT1 Consulting Drafl Report. January 2003 Fracture Toughness Materml Properties for PBNP-2 1 INTRODUCTION Nuclear Management Company, LLC (NMC) currently operates Point Beach Nuclear Plant Unit 2 (PBNP-2). The operators of this nuclear power plant have proactively assessed its condition relative to reactor pressure vessel (RFV) neutron embrittlement over the years of operation. Surveillance Charpy V-notch ( C W ) testing of the limiting vessel circumferential weld metal, contained in the integrated surveillance capsule program and irradiated in the B&W Owners Group program, has shown a high degree of neutron embrittlement. These results have led to projections of end-of-life (EOL) reference toughness, termed RTns, that approach the pressurized thermal shock (PTS) screening criterion of 300F in the Code of Federal Regulations, 10 CFR 50.61 [l]. For EOL extension (EOLE), the latest projections of RTpls exceed the screening criterion.

Section 2 of this report summarizes the projection of CVN results for the PBNP-2 limiting weld metal and other beltline materials.

In an attempt to control the embrittlement of the RPV weld metal at the peak flux locations, flux reduction methods were employed at PBNP-2 using hafnium rods in the peripheral fuel bundles. These localized flux reduction measures have forced a non-optimum core flux profile and dramatically increased fuel cycle costs. Recently, NMC began consideration of the elimination of the hafnium flux reduction, since embrittlement projections with the hafnium removed demonstrate that the beltline weld metal will meet the PTS screening criterion for the current EOL corresponding to 34 effective full power years (EFPY). For an end-of-life extension (EOLE) operating period of 53 EFPY, even with hafnium flux reduction left in the core, the PBNP-2 RPV will not meet the PTS screening criterion using the current CVN-based technology. Use of the Master Curve approach will provide assurance that the PBNP-2 RPV will be in compliance with the PTS rule through EOLE. For current operating life, the use of the Master Curve methodology for the evaluation of material toughness properties provides additional margin for PTS, and this same methodology has been considered in new heat-up and cool-down curves [2].

To better define the condition of the RPV, and to provide better stability in defining the best estimate of R T n s for EOL and EOLE, NMC has obtained and generated fracture toughness data for two different welds fabricated from the limiting Linde 80 flux weld wire heat 72442. These two welds (SA-1484 and WF-67) have been evaluated in the non-irradiated condition and were irradiated to essentially the same fluence of 1.25 x IO d c m 2 (E > 1 MeV) in two different irradiations (six fuel cycles) at the Davis Besse (DB) reactor. The test samples available from each capsuIe were five 0.936T round compact tension (0.936T-RCT) specimens and eight to twelve precracked Charpy three-point bend (0.394T-3PB) specimens; the non-irradiated weld test specimens included 0.394T-3PB and various compact tension size s ecimens. The WF-67 weld was also irradiated to a P

higher fluence of 1.66 x 10 d c m (E > 1 MeV) at the Crystal River 3 (CR3) reactor, but only four fracture toughness specimens (OST-CT) were initially available for testing.

Additional testing of two 0.394T-CT and two 0.936T-RCT specimens from the CR3 capsule has been completed to produce a valid measurement of transition temperature, To, following the Master Curve methodology prescribed in ASTM Test Method E 1921-02 1-1

AT1 Consulting Draft Report, January 2003 Fracture Toughness Matenal Properties for PBNP-2

[3]. Section 3 of this report describes all measured fracture toughness results, corresponding to non-irradiated and irradiated (less than the projected RPV EOL fluence) conditions.

The Master Curve fracture toughness approach has precedent with the Nuclear Regulatory Commission (NRC) through the non-irradiated application for the Zion RPVs by the B&W Owners Group [4] and in the Safety Evaluation (SE) issuance for the Kewaunee RPV [5]. Additionally, the B&W Owners Group has submitted a report covering non-irradiated fracture toughness for all of the welds in B&W-fabricated RPVs

[6J, and First Energy Nuclear Operating Company has produced a WCAP report for application for the Beaver Valley Unit 1 RPV [7].

To meet the intent of current regulations that use the CVN-based approach, NMC recognizes that the currently completed fracture toughness testing allows only the projection of an R T m value for EOL and EOLE fluences using the ASTM E 1921-02 transition temperature, To [3], and the ASME Code defined transition temperature, RTT,,

based on To [8,9]. To achieve full compliance for EOL and EOLE fluences, this same approach will be applied by performing Master Curve testing to measure To at higher fluence values. Four (4) more surveillance capsules \vi11 become available in the future from Davis Besse (2), Crystal River 3, and PBNP-2. Two of the capsules from Davis Besse and Crystal River 3 contain specimens that will validate material properties approaching the EOL fluence, and the PBNP-2 specimens will verify material properties through the EOLE fluence value. The supplemental surveillance capsule that has been fabricated and installed in PBNP-2 contains the limiting PBNP-2 weld metal heat 72442 (WF-67), as well as a weld and plate for the PBNP-1 RPV and a weld for the Davis Besse RPV. The testing of the supplemental capsule as part of a revised surveillance program will allow direct measurement of fracture toughness at the fluence corresponding to EOLE, thus eliminating the need to extrapolate using lower fluence data.

As indicated earlier, Section 3 includes presentation of the individual baseline non-irradiated and irradiated Master Curve fracture toughness results for weld wire heat 72442. Three methodologies are used to estimate the adjusted reference temperature (ART) at EOL and EOLE utilizing RTT, and a suitable margin. These methodologies are: (1) the approach endorsed in B&W-2308 where the initial RTT, is established using a large amount of fracture toughness data and a Charpy shift is used to account for the effects of irradiation with a suitable margin added for uncertainties in the Charpy shift, determination of RT-ro,and material variability; (2) the approach used by the NRC in approving the Master Curve application for the Kewaunee RPV; and (3) the direct measurement approach, which focuses on the irradiated fracture toughness measurements and explicitly considers the individual uncertainties to determine an appropriate margin.

This latter determination is the most realistic value and the other two have added conservatisms built into the methods. All of the methods result in an EOLE ART value (equivalent to R T ~ s of ) less than 300°F. The calculations and projections for ART (and RTms) at EOL and EOLE are presented in Section 4.

1-2

AT1 Consulting Draft Report. January 2003 Fracture Toughness Material Properties for PBNP-2 Details of the supplemental surveillance program for PBNP-2 are summarized in Section

5. The future withdrawal schedule for the PBNP-2supplemental capsule is presented and discussed with regard to future fracture toughness testing and ARTRTns validation. In addition, applicable surveillance capsules being irradiated as part of the B&W Owners Group integrated surveillance program are described.

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AT1 Consulting Draft Report January 2003 Fracture Toughness Material Properties for PBNP-2 2 REVIEW OF CVN-BASED PTS STATUS FOR PBNP-2 2.1 REVIEW OF BELTLINE MATERIALS AND FLUENCES The beltline region of an RPV, per ASTM E185-82 [lo], is defined as the irradiated region of the reactor vessel (shell material including weld regions and plates or forgings) that directly surrounds the effective height of the active core and adjacent regions that are predicted to experience sufficient neutron damage to warrant consideration in the selection of the surveillance material. Babcock and Wilcox Company (B&W) fabricated the PBNP-2 RPV using three shell forgings welded circumferentially at two locations.

The final girth weld (nozzle to intermediate shell) was actually fabricated by Combustion Engineering (CE) using a Linde 1092 flux weld. Table 2-1 provides a summary of the PBNP-2 RPV beltline materials, heat numbers, initial RTNDTvalues, and the nominal chemistry for the beltline materials.

The Nuclear Steam Supply System ( N S S S ) vendor, Westinghouse Electric Company, developed the original surveillance program for the PBNP-2 RPV. The original surveillance program was designed under ASTM E 185-66, but subsequent testing has followed the latest version of ASTh4 E 185 that has been approved by the NRC (ASTM E 185-82). A description of the surveillance program and the pre-irradiation mechanical properties of the reactor vessel materials are presented in WCAP-7712 [ 1 I]. Based on the measured chemistry, initial mechanical properties, and projected fluence, lower and intermediate shell forging heats 123V500 and 122W395 and a submerged arc weld metal (similar to the vessel intermediate to lower shell girth weld seam) were selected to be in the reactor vessel surveillance program. The actual weld metal corresponding to the limiting intermediate to lower shell girth weld seam, heat 72442 (flux lot designation SA-1484), was not included in the original surveillance program, however Wisconsin Electric Power Company (the company operating PBNP-2 at that time) was able to gain access to key data on this weld metal by joining the B&W Owners Group, which had been irradiating weld wire heat 72442 in a host reactor program. Four (4) capsules have been withdrawn and tested to date from the PBNP-2 surveillance program, which provide data on the two forging materials and the fludfluence exposure. Additionally, data have been generated on Heat 72442 weld metals (SA-1484 and WF-67) through the B&W Owners Group, which provide the basis for RPV projections for the most limiting material at PBNP-2.

The best estimate fluence projections for the operating life for PBNP-2 were determined as shown in Table 2-2. The conditions for the projections are based on a power uprate of the reactor core from 1518.5 MWt to 1678.0 MWt that began October 1,2001 and a future cumulative capacity factor of 95%. In addition, the hafnium flux reduction has been removed, and the physics calculation process for fluence does not take credit for use of the FERRET Code. The FERRET Code is used to combine the results of neutron measurements with neutron transport calculations to establish best estimates of the neutron exposure. However, the FERRET Code has not been reviewed and approved by the NRC, and currently cannot be used in a licensing submittal. EOL corresponds to 2-1

AT1 Consulting Draft Report. January 2003 Fracture Toughness Material Properties for PBNP-2 33.86 EFPY (rounded to 34 EFPY) and EOLE for 20 additional years of operation occurs at an additional 19.0 EFPY; therefore, EOLE is projected to be 53 EFPY.

2.2 ART PROJECTIONS FOR EOL AND EOLE In 1985, the NRC issued a formal rule on PTS, 10 CFR 50.61. It established the screening criteria for pressurized water reactor (PWR) vessel embrittlement as measured by the reference temperature termed RTprs. Screening criteria were set corresponding to EOL plant operation for beltline axial welds, forgings, and plates at 270"F, and at 300°F for beltline circumferentia1weld seams. All PWR vessels in the United States have been required to evaluate vessel embrittlement in accordance with these criteria through EOL or beyond.

The NRC amended its regulations for PWR plants to change the procedure for calculating radiation embrittlement RTns values. The revised PTS Rule was published in the Federal Register, May 15, 1991 with an effective date of June 14, 1991, and later updated on December 19, 1995 with an effective date of July 29, 1996. These amendments made the procedure for calculating R T n s values consistent with the method given in Regulatory Guide 1.99, Revision 2 [ 121. The PTS Rule states:

0 The screening criteria for the reactor vessel beltline region are:

270°F for plates, forgings, and axial welds, and 300°F for circumferential welds.

0 The following equations must be used to calculate R T m values (the value of ART at the fluence corresponding to current EOL) for each weld, plate or forging in the reactor vessel beltline:

where IRT is the initial RTNDTand M is a required margin term equal to an assumed two standard deviation (20) for the combined uncertainty in IRT and ARTNDT; ARTNDT= CF * [4t] (0 28 - 0 1 log I#])

where CF is the chemistry factor and Qt is the fluence at the inside surface dcm'; E > 1 MeV).

0 All values of RTns must be verified to be the most limiting for the specific reactor vessel. In doing this, each plant should consider plant-specific information that could affect the level of embrittlement. This includes determination of best estimate mean values of Cu and Ni for the vessel that considers all sources and welds made using the subject weld wire heat.

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AT1 Consulting Draft Report. January 2003 Fracture Toughness Material Properties for PENP-2 Plant-specific PTS safety analyses are required before a plant is within three years of reaching the screening criteria, including analyses of alternatives to minimize the PTS concern.

Prior NRC approval is required for operation beyond the screening criteria.

In complying with the provisions of the PTS Rule, RTms calculations have been performed for PBNP-2. Table 2-3 identifies all of the Cu and Ni measurements for weld Wire heat 72442. The following designations are used in the table: WQ is weld qualification; CR-3 ND is Crystal River Unit 3 nozzle dropout; MD-1 ND is Midland Unit 1 nozzle dropout, and MV is Mount Vernon, a B&W facility. Three different heatlflux combinations were used to fabricate welds using weld wire heat 72442. The overall source mean values are shown at the end of the table and the values round to 0.26 wt% Cu and 0.60 wt% Ni. The measured chemistries for the individual welds will be used later to make adjustments to the shift data from measured fracture toughness data.

The results of the calculation of R T m are shown in Tables 2-4 and 2-5 for EOL (34 EFPY) and EOLE (53 EFPY), respectively. The details of the calculations are explained in the notes and are based on the summary given in BAW-2325 [ 131. The best estimate chemistry for the CE weld heat 21935 was obtained from CE NPSD-1039, Rev. 2 [14].

As shown in Table 2-5, the EOLE fluence yields an R T m value of 3 16°F when using Charpy based methods for the limiting weld for a power uprate to 1678.0 M Wt and removal of the hafnium flux reduction assemblies. Therefore, to obtain extra margin relative to EOL and to reach EOLE, the Master Curve approach is being implemented to properly define a fracture toughness-based transition temperature. The additional fracture toughness testing and evaluation using the Master Curve approach provides a technically superior method for assessing radiation damage to the limiting PBNP-2 weld.

Application of the Master Curve approach also should be considered during current operating life for heat-up/cool-down curves and low temperature over-pressure protection (LTOP) to assure continuity of the Master Curve approach. Having a better knowledge of the fracture-toughness based transition temperature, economic and timing decisions regarding the need to replace or remove the hafnium inserts, reactor power uprating, and license renewal can be made in a more prudent manner.

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AT1 Consulting Draft Report, January 2003 Fracture Toughness Matertal Properties for PBNP-2 Table 2-1 Summary of the PBNP-2 RPV Beltline Materials Heat Initial W% M%

RPV Component Description (Heat/Lot) RTNDT cu Ni (OF)

Nozzle Shell Forging 123V352 +40 0.11 0.73 Intermediate Shell Forging 123V500 +40 0.09 0.70 Lower Shell Forging 122W195 +40 0.05 0.72 Nozzle to Intermediate Shell 21935 -56 0.18 0.70 CircumferentialWeld IntermediateTo Lower Shell 72442 (SA- -5 0.26 0.60 CircumferentialWeld 1484)

Table 2-2 Projectionsfor Fluence for the PBNP-2 RPV at EOL and EOLE EOLE (53 EOLE (53 EFPY)

EFPY)

Component Fluence Fluence (X 1019 Description Facto Factor(a) n/cm2)

Nozzle Shell 123v352 0.335 0.699 0.550 0.833 Forging Intermediate Shell 123V500 1.319 5.385 1.417 For in Lower Shell 122,195 1.313 5.315 1.414 For in ~~

Nozzle to Intermediate 21935 Shell Circ.

I I 0.335 I 0.699 0.550 0.833 Intermediate 1.301 5.085 1.406 Shell Circ. 1484)

Weld a) Fluence factor = [+t] ( 0 2 & 0 1 where $t is fluence 2-4

AT1 Consulting Draft Report, January 2003 Fracture Toughness Material Properties for PBNP-2 Table 2-3 Summary of Weld Wire 72442 Cu/Ni Measurements WF-67 8669 2-5

AT1 Consulting Draft Report. January 2003 Fracture Toughness Material Properties for PBNP-2 2-6

AT1 Consutting DraR Report, January 2003 Fracture Toughness Material Propertles for PBNP-2 Table 2-4 ART/RTPTSProjections at EOL (34 EFPY) Based on CVN Methodology for the PBNP-2 RPV I I Material and Mal lin Inside Surface ART/ RTPTS RPV ProDerties Determination 1I I1 Heat ~~

Om ne (Heat/Lot) Initial cF Margin Description See Fluence ARTNDT RTNDT Notes (OF) (OF) (OF) Factor ("F)

Nozzle Shell Forging 23v352 +40 76 34 A I 0.699 I 53.12 I 127 Intermediate Shell 123V500 +40 58 34 B 1.319 76.48 150 Forging Lower Shell Forging 122w195 +40 42.8 17 C I 1.313 I 56.20 I 113 Nozzle to Intermediate 21935 -56 170 65.51 D 0.699 118.82 128 Shell Circ.

Weld Intermediate To Lower 72442 (SA-

-5 180 68.47 E 1.301 234.14 298 Shell Circ. 1484)

Weld Notes-A Measured value of IRT (Initial RTNDT)with u1= 0 and table value for CF with oA= 17OF; Margin = 2 (a: + o,')'~ = 2 [(O)' + (17)'l'" = 34'F.

B Measured value of IRT (Initial RTNDT)with crI = 0 and table value for CF since PBNP-2 surveillance data are non-credible and table value of CF is conservative with oA= I P F ;

Margin = 2 (012 + = 2 [(O)' + (17)2]'n= 34'F.

C Measured value of IRT (Initial RTNDT) with a, = 0 and calculated value for CF from credible surveillance data with oA= 8.5'F; Margin = 2 (012 + cr,')ln = 2 [(O)' + (8,5)2]'n=

17'F.

D Generic value of IRT (Initial R T ~ Tfor ) a Linde 1092 weld with crI = 17OF and table value for CF with oA= 28'F; Margin = 2 (0: + 02) = 2 [(17)2+ (28)2]'n = 65.51'F.

E Generic value of IRT (Initial RTNDT)for a Linde 80 weld with cr, = 19.7OF and table value for CF since BBWOG surveillance data are noncredible and table value of CF is conservative with oA= 28'F; Margin = 2 (a: + = 2 [(19.7)* + (28)2]'R= 68.47'F.

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AT1 Consulting Draft Report. January 2003 Fracture Toughness Matenal Properties for PBNP-2 Table 2-5 ART/RTPT~ Projections at EOLE (53 EFPY) Based on CVN Methodology for the PBNP-2 RPV Material and Margin Inside Surface ART/ RTPTS RPV Properties Determination Heat Component Initial cF Margin (HeatlLot) See Fluence ARTNDT RTPTS Description RTNDT ( O F )

f°FI

("F) Notes Factor (OF) (OF)

Nozzle Shell 76 34 A 0.833 63.30 137 123V352 +40 Forging Intermediate Shell Forging 123V500 +40 I 158 34 1 B I 1.417 I 82.17 I 156 1 Lower Shell 42.8 17 C 1.414 60.53 118 122W195 +40 Forging Nozzle to Intermediate 21 935 -56 I I 170 65.51 I D I 0.833 1 141.58 I 151 I

I  :

Shell Circ.

Intermediate To Lower Shell Circ.

72442 (SA-1484)

-5 I 180 I 68.47 I E I 1.406 I 253.02 I 316 I

Notes:

A Measured value of IRT (Initial RTNDT) with ui = 0 and table value for CF with uA= 17OF; Margin = 2 (0: + u ~ ) l =R2 [(0)2+ (17)2]1R= 34OF.

B Measured value of IRT (Initial RTmT) with u, = 0 and table value for CF since PBNP-2 surveillance data are noncredible and table value of CF is conservative with ob= 17OF; Margin = 2 (0: + ~ 2 )= 2' [(Of ~ + (17)2]1n = 34OF.

C Measured value of IRT (Initial RTNDT)with u, = 0 and calculated value for CF from credible surveillance data with = 8.5OF; Margin = 2 (012 + 02)'~ = 2 [(0)2+ (8.5)2J'R=

17OF.

D Generic value of IRT (Initial RTNDT)for a Linde 1092 weld with 0 ,= 17'F and table value for CF with u,, = 28OF; Margin = 2 (012 + at)1R = 2 [(17)2+ (28)2]'n = 65.51OF.

E Generic value of IRT (Initial RTNDT)for a Linde 80 weld with ul = 19.7OF and table value for CF since B&WOG surveillance data are noncredible and table value of CF is conservative with uA= 28OF; Margin = 2 (012 + 0 2 ) '=~2 [(19.7)2+ (28)2]'n = 68.47'F.

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AT1 Consulting Draft Report. January 2003 Fracture Toughness Materml Properties for PBNP-2 3 FRACTURE TOUGHNESS TESTING AND RESULTS This section presents the measured Master Curve fracture toughness data for the PBNP-2 weld metal 72442. The testing that was performed followed the testing requirements of ASTM E 1921-02 [3]. All ofthe fracture toughness testing was performed by BWXT Services, Inc. (formerly McDermott Technologies and B&W) as part of the B&W Owners Group activities and under contract from NMC. BWXT Services, Inc., has been involved in Master Curve testing since the middle 1990s and has a proven record for producing consistent and reliable data [6,7].

Table 3-1 shows the fracture toughness testing performed to date on weld wire heat 72442. The testing program has included the following test specimen types: standard %T compact tension (OST-CT); standard 0.394T-CT, precracked Charpy three point bend (0.394T-3PB); and 0.936T round compact tension (0.936T-RCT). The individual results from the specimen types and irradiation conditions are listed in Tables 3-2 through 3-7.

Note that the determination of KJvalues were calculated using the plane stress method in the earlier 1997 version of ASTM E 1921 since that method was being used when the fracture toughness testing was performed. The new 2002 version of ASTM E 1921 changes the calculation method to a plane strain approach, which reduces the values of To by a few degrees Fahrenheit. Since the values determined using the plane stress method are higher and thus more conservative, these are the values reported and used here. The values reported in Tables 3-2 through 3-7 were obtained from B&W Owners Group reports BAW-2254 [ 151, BAW-2308 [6], BAW-23 13, Rev. 2 [161, BAW-2400 [ 171, and BAW-2412 [18]. Further testing and calculation details are contained in those reports.

Except as noted above, the determination of To from the fracture toughness data follows the multi-temperature method presented in ASTM E 1921-02. Table 3-8 lists the To values determined for the six available data sets (three unirradiated and three irradiated).

The average of the fluences from the individual specimens within a capsule has been used, and the different type and specimen size data have been combined as indicated. All of the non-irradiated 0.394T-3PB specimen data have been corrected to include a bias adjustment of 18'F to adjust all of the data to a similar constraint level (consistent with CT specimens, which were the primary basis for the ASME Code K I curve). ~ The irradiated 0.394T-3PB specimen data were adjusted by 8'F, consistent with the NRC approach of using 8.5"F approved for the Kewaunee vessel [ 5 ] .

The need for a bias adjustment appears to be dependent upon the degree of loss of constraint from testing small three-point bend specimens versus CT (pure bend) specimens. The CT specimen geometry maintains a higher level of constraint than the 3PB specimen geometry. This difference seems to be due to the difference in pure bending for the CT-type specimen versus the combined bending plus a small amount of shear loading for 3PB testing. The effect of specimen size appears to be reconciled through the Master Curve normalization to 1T size, but the loss of constraint from the specimen loading geometry is not. Recent finite element studies have compared the three-point bend versus the CT loading for non-irradiated ferritic material flow properties and found that a difference of 18'F can be expected in measured values of To[ 191. Note 3-1

AT1 Consulting Draft Report, January 2003 Fracture Toughness Materlal Properties for PBNP-2 that the bias is expected to decrease with the increased yield strength and different strain hardening exponent indicative of irradiated materials, but the detailed calculations have not been performed for the irradiated case. The strain hardening exponent n = 0.1 (N =

10) for unirradiated RPV materials typically changes to n = 0.07 (N=14.3) for irradiated materials. There are very limited experimental data for making a comparison in the irradiated case, since there is rarely enough irradiated material to assess this constraint difference using different specimen types. Evaluations of To using two of the irradiated data sets developed here are presented in Table 3-9; these comparisons suggest that the 8°F bias adjustment is very conservative. Note that the 0.936T-RCT data sets do not meet the ASTM E 192 1-02 validity requirements for number of specimens. These two invalid data sets suggest a reversed bias effect for the irradiated data sets.

A large amount of non-irradiated experimental data supports the need for a bias adjustment of about lS°F for unirradiated Toresults [7]. The two unirradiated determinations of To using 0.394T-3PB specimens support the need for a bias correction to match the determination from the CT specimens (see Table 3-8). The last row in Table 3-9 reflects the combination of all unirradiated specimen tests with the 1S°F bias included. The combined result is a Toof -73'F.

The B&W Owners Group has recently submitted a topical report on redefining the initial reference temperature using the Master Curve approach [6]. The approach taken by the B&W Owners Group relative to this specific heat of weld material is similar to what is presented here (including the 1 8 O F bias for 3PB specimens), except they propose an additional adjustment based on a loading rate effect. ASTM E 1921-02 allows a range of quasi-static loading rates with no adjustment identified. With the loading rate adjustment included, the B&W Owners Group determined a non-irradiated Tovalue of -65°F [6],

which is conservatively 8'F higher than the -73'F value indicated in Table 3-8 (where no loading rate adjustment is used).

3 -2

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AT1 Consulting Draft Report, January 2003 Fracture Toughness Materlal Properties for PBNP-2 Table 3-1 Fracture Toughness Testing for Weld Heat 72442 Weld Code Capsule and Fluence Test Specimen Number of Test (Flux Lot) (IO" nlcm', E > 1 MeV) Type Specimens Unirradiated; 51 h stress relief 0.394T-3PB 9 (see Table 3-2)

Capsule A3: I.246(a) 0.394T-3PB 12 (see Table 3-3)

Capsule A3: I.261(a) 0.936T-RCT 5 (see Table 3-3) 0.394T-CT 4 (see Table 3-4)

I Unirradiated; 50 h stress relief O.5T-CT 1 (see Table 3-4) 0.936T-RCT 2 (see Table 3-4)

Unirradiated; 11 h stress relief 0.394T-3PB 8 (see Table 3-5)

WF-67 CaDsule L1: 1.169(a) 0.394T-3 PB 8 (see Table 3-61 (8669) L Capsule L1: 1.392(a) 0.936T-RCT 5 (see Table 3-6)

Capsule LG2: 1.59@) O.5T-CT 4 (see Table 3-7)

Capsule LG2: 1.28 & 1.93(b) 0.394T-CT 2 (see Table 3-7)

Capsule LG2: 1.86 0.936T-RCT 2 (see Table 3-7)

Table 3-2 Test Results for Unirradiated Weld Metal SA-1484 (0.394T-3PB Specimens, No Sidegrooves) [6, 181 Test Specimen Temperature Jc KJ Identification (OF) (in-lblin2) (ksidin)

RSIOI I -100 I 177.0 I 72.9 RS102 I -100 I 55.8 I 40.9 RS103 I -100 I 511.0 I 123.8 RS104 1 -100 I 382.0 I 107.0 RS105 1 -100 I 456.0 I 116.9 RS106 I -100 I 129.0 I 62.2 RS107 I -100 I 581.0 I 132.0' RS108

~

I -100 I 393.0 I 108.6 RSIOO -75 Censored in determiningTo 3-3

AT1 Consulting Draft Report, January 2003 Fracture Toughness Malenal Properties for PBNP-2 Table 3-3 Test Results for Irradiated Weld Metal SA-1484, Capsule DB-A3 [6,18]

t included in determining To

• • Censored in determining To 3-4

AT1 Consulting Draft Report. January 2003 Fracture Toughness Material Properties for PBNP-2 Table 3-4 Test Results for Unirradiated Weld Metal WF-67 [6,16]

(50 Hour Stress Relie0 Test Specimen Tem perat ure Jc KJ Identification (OF)

. - (i n-lblin2) (ksidin)

I I I r

0.394 T-CT (No sidegrooves)

PT064

. . - - I -75 I 93.0 I 52.7 PT060 -75 220 81.O PT065 -75 408 110.4 PT061 -75 711 145.7 PT029 -65 I 415 I 111.2 PT140 -50 145 65.7 RS117 -50 133 62.9 Table 3-5 Test Results for UnirradiatedWeld Metal WF-67 [6]

(7 7 Hour Sfress Relie0 (0.394T-3PB Specimens, No Sidegrooves)

Test Specimen Temperature Jc KJ Identification (OF) (in-1b/in2) ( k s i h )

I NA I -120 I NA I 45.2 I -120 I NA I 57.8 I NA I -120 I NA I 66.5 I NA 1- -120 I NA I 84.1 I NA 1 -120 I NA I 98.1 I NA I -120 I NA I 106.4 I NA I -120 I NA I 109.7 NA I -120 NA 130.8*

3-5

AT1 Consulting Draft Report. January 2003 Fracture Toughness Material Properties for PENP-2 Table 3-6 Test Results for Irradiated Weld Metal WF-67, Capsule DB-L1 [6,17]

I-

~~~ ~~

Test Specimen Temperature Jc KJ Identification I OF) (in-Iblin2) (ksidin) 0.394T-3PB (No sidegrooves); 1.769 x IOi9 n/iimz DE132 I -30 I 70.2 I 45.6 DE143 1 5 I 503 DE213 30 315 96.1 DE214 30 328 98.1 DE212 30 388 106.7 DE210 30 456 115.6 Table 3-7 Test Results for Irradiated Weld Metal WF-67, Capsule CR3-LG2 [6,15]

Test Specimen Temperature Jc KJ Identification (OF) (in-Iblin2) (ksidin)

NA I 35 NA I 70.2

~ ~

PT020 0 81.I 48.9 PT021 0 86.8 50.6 PT023

. I 0 178.3 72.5 J PT022 1 0 I 254.6 I 86.6 0.936T-RTCT (10% sidegrooves); 1.86 x IO dcm -

NA 90 NA 139.8 NA 90 NA 102.1 3 -6

AT1 Consulting Drafl Report. January 2003 Fracture Toughness Material Propertles for PBNP-2 Table 3-8 To Determinations from Measured Fracture Toughness Data for Weld 72442 Weld Code Average Fluence Test Specimen (Flux Lot) (10" n/cm2,E > 1 MeV) TY Pes Unirradiated 0.394T-3PB SA-1484 (51 hour5.902778e-4 days <br />0.0142 hours <br />8.43254e-5 weeks <br />1.94055e-5 months <br /> stress relief)

(8579) 0.394T-3PB and 1.25 log@'

0.936T-R CT 0.394T-CT, 0.5T-Unirradiated CT, and -60 (50 hour5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br /> stress relief) 0.936T-RCT Unirradiated 0.394T-3PB (11 hour1.273148e-4 days <br />0.00306 hours <br />1.818783e-5 weeks <br />4.1855e-6 months <br /> stress relief)

WF-67 0.394T-3PB and 42@'

1.255 0.936T-RCT 0.394T-CT, 0.5T-1.66 CT, and 61 0.936T-RCT (8579)

All Unirradiated Combined See above -73 (a) Includes three-point bend specimen bias of 18°F.

(b) Includes three-point bend specimen bias of 8°F.

Table 3-9 Comparison of To Values from Different Specimen Types with No Bias Correction for Weld 72442 Weld Code Capsule and Fluence Test Specimen To, O F (IO" n/cm2, E > 1 MeV) Type Capsule A3: 1.246 0.394T-3PB 105 SA-I484 -

Capsule A3: 1.261 0.936T-RCT 102*

Capsule L1: 1.I 69 0.394T-3 PB 48 WF-67 Capsule L1: 1.392 0.936T-RCT 28*

3-7

AT1 Consulting Draft Report. January 2003 Fracture Toughness Material Properties for PBNP-2 4 APPLICATION OF FRACTURE TOUGHNESS RESULTS TO PBNP-2 4.1 MASTER CURVE APPLICATION METHODOLOGIES The methodology used to examine the integrity of the PBNP-2 RPV is based on application of either ASME Code Case N-629 [8] or N-63 1 [9]. Current regulations are based on the ASME reference fracture toughness curves (Klc and ,)&I which are indexed by the reference temperature RTNDT.Because direct determinations of RTNDTat all projected fluences are not feasible for irradiated materials at this time, current regulations employ estimates based on a combination of unirradiated R T ~ values T and irradiation-induced shifts in Charpy 30 fi-lb transition temperatures. ASME Code Case N-63 1 allows an alternative definition of the initial reference temperature (RTT,) to be used as R T ~ T The. regulatory parameters, RTprs and ART are determined by adding an additional margin to these estimated RTNDTvalues to assure that the resulting indexed fracture toughness curve provides a lower bound to the fracture toughness data. ASME Code Case N-629 provides an alternative method for indexing the ASME reference fracture toughness curves, based directly on fracture toughness measurements of irradiated surveillance capsule materials. As in the traditional regulatory approach, appropriate margin is needed to establish the regulatory parameters, R T m and ART at the EOL fluence using measured values of RTT,. Two EPRI reports have reviewed the Master Curve technical issues associated with the application of the Master Curve methodology and provided technical justification for the two ASME Code Cases, including an assessment of uncertainties and margin consideration [20,21].

Three approaches are presented here for assessing the fracture toughness-based values of ART and R T m at EOL and EOLE for the limiting weld metal heat 72442. The three methods are presented in the order that produces the highest degree of extra inherent margin. The first two methods rely upon the shift-based approach of adding either a Charpy shift or Master Curve To(and RTT,) shift to an initial measure of reference toughness (RTT,). The first approach uses the measured value of initial RTT, and adds a Charpy shift. The second approach also uses the measured value of initial RTT,, but adds a fracture toughness-based shift in RTT,. This method is the one that was used by the NRC in approving the Kewaunee Master Curve application. The third approach, which produces the most accurate determination, uses only the irradiated RTT, values directly as measurement of the irradiated reference temperature.

Application of direct measurements of fracture toughness to RPV analysis requires a method for transforming the measured values to equivalent values at the fluence of interest, if measurements are not directly made at that fluence. Current regulations employ trend equations for the Charpy transition temperature shifts, which are provided in 10 CFR 50.61 [I] and NRC Regulatory Guide 1.99, Revision 2 [I23 to accomplish this transformation. Recently, ASTM E 900-02 [22]was passed using a new trend equation that is mechanistically-guided, as described in an EPFU report [23]. However, there is no equivalent trend curve for Master Curve measurements. In the submittal for the Kewaunee FWV [24], the available data spanned the fluence of interest at EOL and 4-1

AT1 ConsuHing Draft Report, January 2003 Fracture Toughness Matenal Properties for PBNP-2 EOLE, and the transformation required only a small interpolation. However, application for the Beaver Valley Unit 1 RPV required an extrapolation to the EOLE fluence 171.

Application for the PBNP-2 circumferential weld is similar to the Beaver Valley Unit 1 case in that the current available data do not reach the current EOL or EOLE fluence levels. Thus, the EOL and EOLE fracture toughness trend must be inferred by extrapolating RTT, data from the three available irradiation conditions. This extrapolation will be confirmed by future fracture toughness surveillance testing of the new capsule inserted in PBNP-2 and irradiated to the EOLE RPV fluence. Future testing of additional B&W Owners Group integrated program capsules at intermediate fluences is also scheduled.

As mentioned previously, application of the Master Curve technology requires the development of a margin strategy. Although the ASME Code Cases provide a reference temperature (RTT~),which can be used as an alternative to RTNDT,they do not provide guidance on the margin required to determine corresponding values of RT~Tsor ART.

The ASME Code reference temperature, RTNDT,is designed to describe the ductile-to-brittle transition of ferritic steels. By itself, RTNDTdefines a degree of unspecified inherent conservatism in the ductile-to-brittle transition temperature since it is based on bounding values of nil-ductility transition temperature and 50 A-lb / 35 mils (lateral expansion) CVN temperature. In terms of actual fracture toughness, RTNDTis merely an indicator of underlying material properties. However, the relationship defined in the ASME Code between RTNDTand the reference toughness curves does imply real inherent conservatism. By choosing to index the reference toughness curves in a manner such that they provide lower bounds to the existing fracture toughness data, the ASME Code has sought to provide a conservative method for estimating fracture toughness values to be used in RPV integrity analysis. When used for unirradiated properties, where RTNDTis measured following the ASME Code procedure, no additional margin is generally required in the analytical process.

Although the definition of RTNDTis not limited in application to unirradiated materials, the amount of material required makes direct determinations of RTNDTin irradiated materials impractical. Therefore, current regulations employ ART to index the reference toughness curves.. The ART value is defined for a specific neutron fluence, which is generally taken as the EOL or EOLE fluence. However, the general form of the definition may be described as:

ART = RT(4t) + Margin (3) where RT(4t) is the estimated RTNMvalue as a function of fluence, and Margin is to account for the uncertainties in the estimation process.

The methodology being applied for the PBNP-2 RPV is dependent on the Master Curve application method being used. Each of the methods and the determination of Margin are

• Although the definitions for ART and RTnS appear in different places (one in an NRC Regulatory Guide

[ 121 and the other in an NRC Regulation [l I), they are identical in computation. For simplicity, ART generally is employed in the following equations and discussion to describe both values.

4 -2

AT1 Consulting Draft Report. January 2003 Fracture Toughness Material Properties for PBNP-2 described in the following sections. Table 4-1 summarizes the results from all of the methods.

4.2 MEASURED INITIAL RTT, AND CHARPY-SHIFT APPROACH In current regulation, the reference temperature is estimated as:

RT(+t) = RTND~O+ ART(4t) (4) where RTNDT(U) is the initial (unirradiated) RTNDTvalue, and ART(4t) is the Charpy 30 ft-lb transition temperature shift from Regulatory Guide 1.99, Revision 2 [12] or ASTM E 900-02 [22].

Under current regulations, generic values of RTNDT(U) may be used, but the associated uncertainties must be included in the Margin term. An alternative formulation of Equation 4 is required for Master Curve applications.

ASME Code Case N-63 1 provides a means of measuring an alternative reference temperature, RTT,, for non-irradiated materials, while ASME Code Case N-629 provides a means of directly measuring an alternative reference temperature, RTT,, for irradiated materials (and non-irradiated materials also). This reference temperature is defined as:

RTT, = To+ 35OF (5) where To is defined using the ASTM E 1921 [3] test procedure.

The ASME Code Cases are constructed to allow RTT, to be used in place Of RTNDTas an indexing temperature for the ASME reference toughness curves. These Code Cases clearly anticipate that this alternative reference temperature acts in a manner similar to that defined in Equation 3 for determining ART.

In the application using the alternative definition for RTNDT(U) using Master Curve data and ASME Code Case N-63 I, Equation 4 becomes:

RT(4t) = RT.rom + ART(4t) (6) where RTT,(v, is the initial (unirradiated) RTT, value (Equation 5 ) , and ART(@) is the Charpy 30 ft-lb transition temperature shift (Equation 2).

Substituting Equation 6 into Equation 3 gives:

ART = RTT,(u)+ ART(4t) + Margin (7)

Table 4-1 shows the calculation for Equation 7 as compared to the current Regulatory approach shown previously in Tables 2-4 and 2-5 using RTNDTCLI) and Charpy shift 4-3

AT1 Consulting Draft Report. January 2003 Fracture Toughness Material Properties for PBNP-2 (ACVN). The RTT'(u) selected is based on the combined Tovalue of -73'F (To+ 35'F =

-38'F) listed at the end of Table 3-8. The Margin term was determined based on the BAW-2308 analysis of uncertainties in the determination of RTT,(u, for weld heat 72442

[6] combined with the current Regulatory uncertainty in Charpy shift (06). The uncertainty for initial (unirradiated) R T T (01) ~ was determined for weld wire 72442 as the combination of uncertainties associated with material variability based on a Monte Carlo analysis of the actual data (CTMC) and the uncertainty in the measurement of To (oro)from ASTM E 1921-02 [3]:

For this application, the BAW-2308 Monte Carlo evaluation produced a OMC of 9.3'F for the 72442 weld metal and there were a total of 24 valid tests resulting in a C J T ~of 7.4'F; therefore, (31 = [(9.3)* + (7.4) 23 I R -- 11.9'F. Margin is then determined following the Regulatory Guide 1.99, Rev. 2 [ 121 method at an approximate 95% (or 20) statistical level with the uncertainty in shift included :

Margin = 2 [or2 + aA2]lR (9)

Using the Regulatory Guide 1.99, Rev. 2 value for CTA [32], Margin = 2 [(l 1.9)2 + (2S)2]IR

= 603°F. This value is 7.7'F lower than that used in the current licensing approach using a generic vaIue of RTNDT(u), which is logical since heat-specific RTT, data are being used instead of a generic value of RTNM.

As indicated in Table 4-1, the values of ART (and RTprs) determined using the B&W Owners Group approach are 40'F lower than from the current licensing approach, with the EOLE value now well below the PTS screening limit of 300'F.

4.3 MEASURED INITIAL RTT, AND RTT,-SHIFT APPROACH Measurements of irradiated specimens in accordance with Code Case N-629 represent direct determinations of the function RT($t) at specific fluences. Because they are direct measurements, they provide far more accurate values than the indirect estimation procedure used in current regulation or using the approach just discussed. However, most reactor vessel integrity analyses require the evaluation of RT(4t) at EOL and EOLE fluences, which can best be accomplished by fitting a curve to the measured data with some form of a fluence function.

The Regulatory Guide 1.99, Revision 2 [12] prediction curve fits CVN 30 ft-lb transition temperature (ACVN) data to a function of the form in Equation 2, which is restated here as:

ACVN = CF

• FF(+t) (10)

AT1 Consulting Draft Report, January 2003 Fracture Toughness Material Properties for PBNP-2

where, ACVN is the Charpy 30 ft-lb temperature shift assumed equal to ARTNDT, CF is the chemistry factor, and FF(4t) = fluence function.

While the magnitude of the shift is determined by CF, the shape of the curve is determined by FF($t). FF(4t) is the same for all materials; hence any material specific information is contained in the chemistry factor. Although Regulatory Guide 1.99, Revision 2 provides tables that allow determination of CF on the basis of the material form and composition, when credible surveillance data are available, CF can be determined by a fit to the data.

The use of CVN shifts to adjust the reference temperature is implicitly based on the assumption that there is equivalence between the fracture toughness shift and the CVN shift. This equivalence has been shown to be true when comparing ACVN and ATo (or ARTT,) [6,25]. It is, therefore, reasonable to apply the same form of Equation 10 to assess the shift in fracture toughness transition temperature, ARTT,:

where C F T is ~ the effective chemistry factor for fracture toughness shift.

When the Regulatory Guide 1.99, Revision 2 fluence function is used, CFT, may be determined by fitting ARTT, measurements. In this case, the reference temperature may be estimated as:

Substitution of this relationship into Equation 7, defines the Master Curve ART value:

ART = RTT,(v, + C F T *~FF(4t) + Margin . (1 3)

This equation may then be evaluated at the EOL and EOLE fluences to provide the final ART values. Evaluation of this equation requires determination of three basic parameters, RTT~(v,, CFT,, and Margin.

There are three other important aspects to be considered before Equation 13 is used.

First, any specimen bias correction to the measured To data should be made. These constraint corrections already have been included in the results in Table 3-8, and these corrections are technically more realistic than those assumed by the NRC for the Kewaunee SE [ 5 ] . Additionally, the Kewaunee SE only considers use of 0.394T-3PB tests with an appropriate bias. In the data used here, more emphasis has been placed on CT specimens where no bias is needed. Use of all measured data is the preferred approach. Second, adjustments to reflect the differences in irradiation conditions between the surveillance specimen and the actual RPV must be made. For application to the PBNP-2 W V ,there are differences in the temperature of irradiation that need to be 4-5

AT1 Consuktng Draft Report, January 2003 Fracture Toughness Material Properties for PBNP-2 considered. The average cold leg temperature for the PBNP-2 RPV is conservatively estimated to be 530°F at EOL and EOLE, while the cold leg temperature for the B&W Owners Group host reactors (DB and CR3) is higher at 556'F. This 26OF difference needs to be considered when evaluating the fracture toughness data irradiated in DB and CR3 as applied to the PBNP-2 vessel. Third, any chemistry differences in the best estimate for the PBNP-2 RPV weld and the two surveillance welds must be included.

The NRC developed a mathematical form of Equation 13 for the Kewaunee SE that included the three adjustments described above [SI. The only real difference between the NRC method and the CFT,-method in Equation 13 is use of an average value of non-irradiated RTT, versus a specific value for each weldment; the NRC approach allows use of individual weld initial RTT, values, whereas the approach in Equation 13 uses an average value. In the end, the NRC method averages the individual projections of ART, which is essentially the equivalent of averaging the initial values as used in Equation 13.

The method to adjust for irradiation temperature differences endorsed by the NRC uses a degree of shift adjustment per degree of temperature difference. When applied, the ratio of the adjusted shift value (measured plus 26°F) divided by the measured value is what is termed the temperature adjustment. This adjustment is shown in Table 4-2 for the measured shifts using the individual weld measurement of initial RTT, (the two measures for weld WF-67 were averaged). A similar adjustment based on the new ASTM E 900-02 correlation [22] is shown in Table 4-3. The NRC adjustment is much more conservative than that from ASTM E 900-02. The NRC adjustment is a function of the measured shifts, which is not consistent with the adjustment developed using the E 900-02 correlation.

Also shown in Tables 4-2 and 4-3 is the adjustment to account for differences in the best estimate chemistries for the PBNP-2 RPV weld and the two surveillance welds. The NRC approach uses the ratio of CF values from Regulatory Guide 1.99, Rev. 2 (Le., CF for the PBNP-2 RPV divided by the CF for the surveillance weld). The ASTM E 900-02 approach uses the ratio of the predicted shifts for the differing chemistries between the RPV and the surveillance welds. The two methods give similar results with the ASTM E 900-02 method giving slightly higher adjustments. The ASTM E 900-02 approach is superior since it was developed using combined mechanistic and statistical criteria and utilized a larger and more current database as compared to the Regulatory Guide 1.99, Rev. 2 CF values.

Overall, however, the adjusted ART,, values are lower using the new ASTM E 900-02 approach. Thus, the NRC adjustment method has added more conservatism into their methodology.

The measured ART,, values developed using the fixed value of initial R T T of ~ -38°F are summarized in Table 4-4. These values differ by only a few degrees from those listed in Table 4-2 and 4-3. Adjustments to reflect the NRC predicted differences are also shown, and the CFT, value is calculated in an analogous manner to the CVN-based approach in I O CFR 50.61 and Regulatory Guide 1.99, Rev. 2. The calculated CFT, is 163.3'F as 4-6

AT1 Consulting DraR Report. January 2003 Fracture Toughness Material Properties for PBNP-2 indicated in Table 4-1, which is less than the Charpy-based CF value of 180'F from the weld table in Regulatory Guide 1.99, Rev.2 and 10 CFR 50.61. This lower value of CFT0 illustrates even more conservatism for the 72442 weld wire heat since the fracture toughness shift is less than the Charpy shift.

The Margin specified in the Kewaunee SE is essentially the same as presented earlier in Equations 8 and 9. For the Kewaunee RPV, the NRC estimated a value of 14'F for 01, which is slightly more conservative than the actual value of 1 1.9'F determined for weld 72442 [6]. The effect of uncertainty in shift (a*)should be less significant for weld heat 72442 since the best estimate Cu (0.26 wt%) is above the embrittlement saturation level for Linde 80 flux welds of 0.25 wt% Cu [22,23]. Thus, a conservative level of uncertainty is being applied when the value of 06 is set at 28'F, based on the methodology used by the NRC. The NRC allowed a 2'F reduction in the inherent margin in determining RTT, (Le., NRC allowed RTT, = To + 33°F). This reduction in the definition of RTT, is too small to have any practical importance, but it is used here. The overall margin considering these parameters is a value of 60.5'F (Margin = 2 [(14)2 +

(28)*]" - 2).

Using Equation 13, with the values just discussed to be consistent with the NRC approach, the fifth and sixth rows of Table 4-1 illustrate the calculation of ART at EOL and EOLE. When the NRC approach is applied to the three irradiated weIds tested, the ART values at EOL and EOLE are 235°F and 252'F, respectively, as determined in Tables 4-5 and 4-6. Note that the results are the same. Conservatisms can be adjusted as follows: 01reduced to 11.9'F from 1 4 T based on the measured fracture toughness data; chemistry and temperature adjustments that follow the ASTM E 900-02 correction approach; and, inherent margin of 2 O F given back (small conservatism added rather than reduced). With these changes, the EOL and EOLE projections for ART are reduced 13-14'F to 222'F and 238'F, respectively.

The results presented for the use of initial RTT, and ARTT, shift are presented here to illustrate the conservatism in the initial RTT, and ACVN approach described in the previous section. There is extra margin of at least 20°F even when large uncertainties in the parameters are assumed. All of the projections for ART and R T ~ Tat s EOLE fluence are well below the PTS screening limit of 30OoF.

4.4 DIRECT MEASUREMENT OF IRRADIATED RTT, The direct measurement approach was originally submitted to the NRC for the Kewaunee RPV application [24]. The NRC did not accept this approach since it is a change to the current regulatory practice of using the shift-based approach in Regulatory Guide 1.99, Rev. 2 and the PTS Rule (10 CFR 50.61). The application for PBNP-2 requires extrapolation to higher fluences to assess PTS at EOL and EOLE fluences. The direct use of the irradiated RTT~ values w a s the intent of ASME Code Case N-629, but the selection of margin to address uncertainties intentionally was left to be defined in regulatory space. The first application for Kewaunee saw the NRC add higher margin for Master Curve use than that needed using the current regulatory practice. The issues of 4 -7

AT1 Consulting Drafl Report. January 2003 Fracture Toughness Matenal Properties for PBNP-2 using direct measurement and development of a more reasonable margin will be addressed in several international forums over the next few years.

The simplicity of the direct measurement approach is discussed next. The three measures of irradiated RTT, (properly adjusted for 3PB bias) are used without extrapolating from zero fluence. The three measured values cover a small range of fluence, so the three values of irradiated RTT, are adjusted for chemistry and temperature using the ASTM E 900-02 approach described in section 4.1 -2 and then averaged, as well as the fluence. The result is RTI, = 129'F at a fluence of 1.39 x l O I 9 n/cm2. This value is then extrapolated to the higher EOL and EOLE fluence using the same fluence function as used for CVN data. The FF(4t) relationship from Regulatory Guide 1.99, Rev. 2 is used since it has been shown to be essentially the same shape as the more complex relationship in the ASTM E 900-02 correlation model [22]. This functional form has also been shown to match the database of irradiated Linde 80 weld fracture toughness data in the BAW-2308 study [6].

The ART using the direct measurement approach involves a form similar to that in Equation 13, except the extrapolation is from the measured RTi, to the EOL and EOLE fluences:

ART = RTj, + CF

• AFF(4t) + Margin (14) where CF is taken as the Charpy-based CF of 180°F,which is conservative relative to the fracture toughness-based CF values shown earlier. This extrapolation from RTi, is small compared to starting at the unirradiated condition ( R T T , ~ ) AFF(+t)

is the difference in FF(+t) between the irradiated point and the EOL and EOLE fluences. The resulting RTi, (EOL) and RTi, (EOLE) are listed in the last rows of Table 4-1.

The development of the Margin is based on the uncertainties in the important parameters affecting irradiation changes in fracture toughness. These parameters and their corresponding uncertainties are: copper content, nickel content, irradiation temperature (T,m),and fluence ($t). Other uncertainties that also need to be included are material variability (OMC), accuracy of the measured irradiated To (GT~),and projection to higher fluences (Opmj). The irradiation sensitive parameters can be determined from the ASTM E 900-02 model for embrittlement. The way that an uncertainty in an independent variable propagates through the model to produce an uncertainty in the predicted value can be estimated through a simple error analysis. The effect of the uncertainty in the independent variable on the prediction uncertainty can be determined by taking the partial differential of the rediction equation. The model provides a functionf(xl,xz,xJ, ...), where xl represents the 'i independent variable. The contribution of uncertainty in x, to the uncertainty in the prediction, dP, is:

4-8

AT1 Consulting Draft Report, January 2003 Fracture Toughness Matenal Properties for PBNP-2 If the uncertainties in the independent variables are truly independent, then the composite uncertainty can be determined by taking the square root of the sum of the squares of the individual contributions. The uncertainties in the predictions are not constant and can vary as a function of fluence. Values of dxi must be established from measurements or other methods of assessing the uncertainty of the parameters. Table 4-7 summarizes the estimates of uncertainties in the E 900-02 model parameters and the calculated uncertainties in the prediction of RTi,. These prediction uncertainties, and the uncertainties associated with material variability, the determination of RTT,, and projections from the measured fluence where RTT, has been determined, can be used to assess Margin:

CMC is the value determined in BAW-2308 [6] from Monte Carlo runs using the non-irradiated 72442 weld data (9.3'F). CT, is the uncertainty calculated as p / N l R , where N is the number of valid test results used to determine the irradiated To,and p is the statistical quantity from ASTM E 1921-02 [3]; for an individual data set of 10 valid tests, G T is ~

10.7"F. oproj is determined by the uncertainty in CF that can be generated by the uncertainties in Cu, Ni, irradiation temperature, and fluence in Table 4-7; the result is an uncertainty in CF of 5'F, which gives Opmj of 1.O°F at EOL and 1.6'F at EOLE. ocU,(TNi, (TTin, and a g t are the prediction values from Table 4-7and are based on the statistics of the measured Cu and Ni values from Table 2-3 and estimates of how well known the irradiation temperature and fluence values are known.

The Margin values are listed in Table 4-1 for EOL and EOLE. These Margins are added to the projections for RTi, at EOL and EOLE to give the results for ART. These low ART values illustrate that there is even more conservatism in the B&W Owners Group and NRC methods of determining ART and R T m .

Note that some of the assumptions used to develop the Margin term from Equation 16 also may have additional conservatisms. For example, the fluence projections for the PBNP-2 circumferential weld are based upon calculations [26] that involve conservative assumptions. These assumptions include using calculations rather than measurement adjustments, not adjusting calculation results by use of the FERRET code, removal of hafnium flux reduction from the PBNP-2 core, implementation of the maximum power up-rate under consideration, and a future capacity factor of 95%. All of these factors together will produce conservative estimates of projected fluences at EOL and EOLE.

Therefore, using the qtterm derived from the ASTM E 900-02 model overstates the true uncertainty.

Figure 4-1 compares the Master Curve projection of ART using all three Master Curve methods with the projection based on the current regulatory Charpy methodology. The Master Curve approaches do not approach the PTS screening criterion even well beyond EOLE. The Master Curve data points are plotted in Figure 4-1 without added Margin and are all below the Master Curve-based ART projections. The Master curve data have been corrected for temperature and chemistry to match the PBNP-2 vessel weld using the 4-9

AT1 Consulting Draft Report. January 2003 Fracture Toughness Material Properties for PBNPZ ASTM E 900-02 adjustments. Further validation of these results will come from future testing of weld metal fracture toughness specimens irradiated to near EOL and EOLE fluences.

4-10

AT1 Consulting Draft Report, January 2003 Fracture Toughness Matenal Properties for PBNP-2 Table 4-1 Comparisons of Reference Temperature Methods and Results I

Method of Analysis Term Initial Material and Margin Properties RTNDT CF Margin See I Inside Surface ART= RTprs Fluence Dete rminat In ARTNOT or (OF) (OF) Notes Factor (OF)

RTTO 0

Current EO1 -5 180 68.5 A 1.301 234.1 298 Regulation EOLE -5 180 I 68.5 I A 1 1.406 I 253.0 316

-38 1.301 234.1 257

-38 1.406 253.0 276

-38 1.301 212.4 235 ARTT, +

Margin -38 1.406 229.5 252 Direct Use 209 of Irr. RTT, 229 Notes:

A Generic value of IRT (Initial R T N ~for ~ )a Linde 80 weld with oI= 19.7OF and table value for CF since BSWOG surveillance data are non-credible and table value of CF is conservative with oA= 28OF; Margin = 2 (a; + = 2 [(19.7)' + (28)']IR = 68.5"F.

B B&W Owners Group (BSWOG) approach [Errorl Bookmark not defined.] using measured combined Initial.RTTo(RTT~(u)) from Table 3-8 (To+ 35OF) with o,= 11.9OF (based on combined material uncertainty, oMC,of 9.3OF and uncertainty in the determination of To, oT0,of 7.4OF) and table value for CF since B&WOG surveillance data are non-credible and table value of CF is conservative with C J =~ 28OF; Margin = 2 (0: +

= 2 [(I1.9)' + (28)']'" = 60.8OF.

C Procedure matching the Kewaunee SE calculation approach [5] with measured Initial 2 tR-RTTo,CJ, = 14OF, a9 = 28OF,and credit for inherent margin of 2OF;Margin = 2 (a: + oA )

2 = 2 [(14)' + (28) ]In - 2 = 60.5OF.

D The most comprehensive approach using the measured irradiated RTTovalues and projecting them forward without reliance on the non-irradiated condition using realistic uncertainties to obtain a Margin term; Margin = 2 (oMc2+ CST: + oc: + ON? + '

' ~ EOL, o ~ , ~ , ' ) at  ;

o + CJ +

r Margin = 2 I(9.3)' + (10.7)* + (l.S)'+ (4.1)2 + (6.9)' + (13.2)2+'(l.O) 1]In 12

= 42.1OF; at EOLE, Margin = 2 [(9.3)'+ (10.7)' + (1.7)' + (4.2)'+ (8.9)2 + (12.5)+

(1 .6)2]1n= 42.9OF 4-1 1

AT1 Consulting Draft Report. January 2003 Fracture Toughness Material Propertiesfor PBNP-2 Table 4-2 NRC Method [5] for Correcting ARTT, Values for Chemistry and Irradiation Temperature Differences Material & Measured Tempera tu re Chemistry Adjusted Capsule ARTTO (OF) Adjust ment* Adjustment** ARTT, ( O F )

SA-1484lA3 183 1.I42 1.008 210.7 WF-67IL1 113 1.230 1.078 149.8 WF-67LG2 132 1.197 1.078 170.3 Table 4-3 ASTM E900-02 [22] Corrections of ARTT, Values for Chemistry and Irradiation Temperature Differences 7 Measured 132 I Temperature Adjustment*

I..---

ncx 1.095 I

I Chemistry I.096 Ratio 01 shift prediction with dtfl'erent irradiation temperatures considered using ASTM E 900-Adjusted 02

• • Ratio of shift predictions with difference copper and nickel chemistries using ASTM E 900-02 Table 4 4 Summary of A R T T ~Values Corrected for Differences in Chemistry and Irradiation Temperature and Calculation of CFTO Material & FIuence, FIuence Adjusted Capsule Measured Adjusted** I O q 9 nlcm2 Function Shift FF~

ARTT, (OF)" ARTT~ (OF) (E>1 MeV) (FF) X FF 182 209.7 1.25 1.062 222.7264 1.1282135 115 152.0 I .255 1.063 161.5929 1.1305620 134 172.5 1.66 1.I40 196.5448 1.2988859 C= 580.6441 3.5576614 1 .-

These shift values are based on the measured irradiated RTToand the combined non-

- 1 irradiated RTTo

• • Adjustments are made using the NRC methodology for chemistry and temperature.

4-12

AT1 Consulting Draft Report. January 2003 Fracture Toughness Material Properties for PBNP-2 Table 4-5 NRC Method [5]for Determining ART at EOL Fluence Chemistry Materia' ' and Fluence Temperatu re Function Adjusted Total Adjusted Margin** ART Adjusted Ratio' ARTTo(OF) RTi, ( O F ) (OF)

ARTTO(OF)

SA-I 484lA3 210.7 1.225 258.0 219.0 60.5 279.5 WF-6 7lL1 149.8 1.223 183.3 147.3 60.5 207.8 WF-67/LG2 170.3 1.I41 194.4 158.4 60.5 218.9 Average 235 Table 4-6 NRC Method [5] for Determining ART at EOLE Fluence I Chemistry I I and Fluence Total Material & Adjusted Margin** ART Temperature Function Adjusted Capsule Adjusted Ratio* ARTT, (OF) RTirr (OF) (OF)

ARTT~ (OF)

SA-14841A3 210.7 1.323 278.8 239.8 60.5 300.3 WF-67/Ll I 149.8 1.322 I 198.1 I 162.1 60.5 222.6 WF-67/LG2 I 170.3 1.233 I 210.0 1 174.0 60.5 234.5 Average 1 252 Ratio of the fluence functions between the RPV at EOLE and the surveillance capsule

• • The margin established by NRC in the Kewaunee SE (Margin = 2 [(14)' + (28)'I- 2)

Table 4-7 Calculations of Prediction Uncertainties for ASTM E 900-02 Model Parameters at EOL and EOLE Uncertainty in Uncertainty in Uncertainty in Parameter Parameter Prediction at EOL (OF) Prediction at EOLE (OF) cu 15.3% I.6 I.7 Ni 0.023 wt% 4.1 4.2 I Tim I 5OF I 13.2 I 12.5 4-13

AT1 Consulting Draft Report, January 2003 Fracture Toughness Material Properties for PBNP-2 350 300 I

250 I

200 G

t c

a 4

150 100 - -Master Curve Method 1

-Master Curve Method 2

-Master Curve Method 3

-Current Regulation 50

- A Adjusted SA-1484 Data 0 Adjusted WF-67 Data EOLE 0 1 2 3 4 5 6 7 Fluence (x10" nlcm')

Figure 4-1 Comparisons of Projected ART from Master Curve and Current Regulation for Weld Wire 72442 (Adjusted Master Curve Data Do Not Have Margin Added) 4-14

AT1 Consulting Draft Report, January 2003 Fracture Toughness Material Properties for PBNP-2 5 SUPPLEMENTAL SURVEILLANCE PROGRAM 5.1 HISTORY OF EXISTING PROGRAM The original PBNP-2 surveillance program was prepared in accordance with ASTM E 185-66 and consisted of six surveillance capsules attached to the outside of the reactor internals thermal shield. Each capsule contained mechanical specimens, dosimetry, and thermal monitors. The mechanical specimens were fabricated from material representative of the PBNP-2 W V . The materials included the lower and intermediate shell forging materials and a weld similar in chemistry to the circumferential weld connecting these two shell courses. A pre-irradiation (baseline) evaluation of the strength and Charpy toughness of the surveillance materials was performed.

Westinghouse Electric Company developed the original surveillance program for the PBNP-2 reactor vessel. Although the original program was in accordance with ASTM E 185-66, subsequent testing has followed the latest version of ASTM E 185 that was been approved by the NRC, through ASTM E 185-82. A description of the surveillance program and the pre-irradiation mechanical properties of the reactor vessel materials are presented in WCAP-7712 [l 11. Four surveillance capsules have been withdrawn and tested to date. One of the standby capsules has been removed and is being stored at Point Beach. Table 5-1 presents a summary of the PBNP-2 surveillance program and materials. A summary of the surveillance test results for the forging materials is provided in Section 2.

The surveillance materials are contained in capsules positioned in the reactor between the thermal shield and vessel as shown in Figure 5-1. This figure also includes the numbering system for the capsule specimens and their locations. The irradiation conditions (temperature, neutron spectrum, and flux) for the capsule are very similar to those of the reactor vessel. Each capsule contains Charpy V-Notch, tensile, and 1X-Wedge-Opening-Loading fracture toughness specimens in the quantities identified in Table 5-2.

Since the actual heat of the limiting weld metal is not in the PBNP-2 surveillance program, participation in the B&W Owners Group allowed access to irradiated surveillance data of 72442 welds. The results from these irradiations have been documented elsewhere [13]. The weld metal results are summarized in Section 2.

5.2 REVIEW OF REMAINING CAPSULES The remaining standby PBNP-2 surveillance capsules (Capsules N and P) do not include weld heat 72442. Therefore, they are of no use in validating the Master Curve projections in Section 4. However, surveillance capsules are being irradiated in the B&W Owners Group Master Integrated Reactor Vessel Surveillance Program (MIRVP) and include Heat 72442 [27]. Details on these capsules are provided in Table 5-3.

5- 1

AT1 Consulting Draft Report, January 2003 Fracture Toughness Material Properties for PBNP-2 These capsules contain several Charpy V-notch and compact fracture toughness specimens of the WF-67 weld material. Two of these capsules have a target fluence of 3.0 x 1Oi9 n/cm2, which is approximately the projected EOL fluence for PBNP-2.

Capsule A I was scheduled for removal from Davis Besse in 2008. Capsule A4 in Crystal River 3 should be available at about this same time depending upon the actual operating schedule. The exact status for capsule A I will depend upon a revised operation schedule at Davis Besse once the issues associated with the reactor vessel head are resolved. Since these two capsules are essentially redundant to each other, the loss of one would not affect the ability to produce Master Curve fracture toughness results applicable to the PBNP-2 vessel. Capsule L2 in Davis Besse has a lower target fluence and has little relevance for the PBNP-2 vessel. When any or all of these specimens are tested, the new results will be integrated with the existing data described in Section 4 to further assess RPV integrity.

5.3 DESCRIPTION

OF NEW SUPPLEMENTAL SURVEILLANCE CAPSULE In addition, a new PBNP-2 surveillance capsule has been installed for the purpose of obtaining relevant fracture toughness data at the EOLE fluence. The new PBNP-2 surveillance capsule contains surveillance specimens that will be used to directly measure the fracture toughness of the PBNP-2 weld metal heat 72442 as well as materials from PBNP-I and Davis Besse. The supplemental capsule contains the materials and specimens identified in Table 5-4.

The target fluence for the PBNP-2 supplemental surveillance materials will correspond to the PBNP-2 peak reactor vessel fluence at EOLE. Surveillance data obtained from this capsule will provide direct fracture toughness measurements for the 72442 weld metal near the maximum fluence at EOLE. These data will provide direct evidence to validate previous reactor vessel life assessments and a measure of the actual margins available for the PBNP-2 RPV.

5.4 SUPPLEMENTAL CAPSULE IRRADIATION AND WITHDRAWAL SCHEDULE The supplemental surveillance capsule for PBNP-2 will be irradiated to a target fluence equivalent to the extended end of operating license life (EOLE) fluence for the limiting weld metal. Irradiation to this fluence will allow fracture toughness measurements to be directly obtained to demonstrate adequate reactor vessel toughness throughout the license renewal term.

As discussed in Section 2, the peak PBNP-2 reactor vessel fluence will change significantly from previous estimates because of the possible decision to eliminate the hafnium flux reduction program and to implement power up-rates. The revised EOLE peak fluence estimate for the PBNP-2 circumferential weld is 5.085 x d c m 2 and considers the affects of hafnium removal and power up-rate.

5 -2

AT1 Consuting Draft Report. January 2003 Fracture Toughness Materlal Properties for PBNP-2 The supplemental surveillance capsule for PBNP-2 was installed following Cycle 25.

Table 5-5 provides details of the PBNP-2 supplemental surveillance capsule fluence projections.

Based on the fuel management strategies and power up-rates planned for PBNP-2, the supplemental surveillance capsule should be removed and tested at just over 38 EFPY, which is expected to be obtained by the outage following Cycle 33 (based on implementation of 2-year refbeling outage intervals for future plant operation). This reheling outage is estimated to occur during the year 2017.

It is recommended that the PBNP-2 peak reactor vessel fluence and the fluence for the supplemental surveillance capsule be re-evaluated in the future to reflect actual reactor operation. As further reactor operation occurs, better vessel and capsule fluence estimates can be made and a more definitive capsule withdrawal schedule may be established.

5-3

AT1 Consulting Draft Report, January 2003 Fracture Toughness Material Properties for PBNP-2 Table 5-1 PBNP-2 Surveillance Program Lead I - --_-.

Factor Removal Time Comments 1.52 EFPY v 3.37 (End of first core cycle) 4.74 x 10l8n/cm2, E > 1 MeV T 1.94 3.45 EFPY 9.45 x 10 n/cmL, E > 1 MeV R 3.37 5.1 EFPY 2.01 x 10 n/cm2. E > 1 MeV S 1.79 14.8 EFPY 3.47 x 10 n/cmz, E > 1 MeV Capsule was removed and stored in spent fuel pool.

P 1.94 19.5 EFPY Target fluence was 3.40 x IO n/cm2, E > 1 MeV Tar et Fluence is N 1.79 Standby 5.00 x I O?9 n/cm2. E > IMeV Table 5-2 Type and Number of Specimens in the PBNP-2 Surveillance Test Capsules Specimen Types and Number Material CVN Tensile 1X-WOL Weld WF-193 8 3 (O)* 3 (0)

Forging 123V500 12 3 (4)* 3 (4)*

Forging 122W195 12 3 (5)* 3 (5)

HAZ,Heat 8 0 0 122W195

• The first number represents the number of specimens in Capsules R, S , and V. The number in parentheses represents the number of specimens in Capsules N, P, and T.

AT1 Consulting Draft Report. January 2003 Fracture Toughness Material Properties for PBNP-2 Table 5-3 Remaining MIRVP Capsules Containing Heat 72442 Specimen Target Material Quantities in Capsule 1D Host Plant Fluence Each Capsule Davis Besse 1 3.00x 1019 W F-67 4 -Tensile Davis Besse 1 1.70 x 1019 WF-67 12 - CVN 5 - 0.936 TRCT A4 Crystal River 3 WF-67 Table 5 4 PBNP-2 and Other Materials and Specimen Types in Supplemental Capsule Number and Type of S2ecimens Material CVN 'AT-CT Tensile Weld Metal Heat 72442 (WF-67) 10 12 2 PBNP-2 Intermediate to Lower Shell Circumferential Weld Plate Material Heat A9811-1 10 0 2 PBNP-I Intermediate Shell Weld Metal Heat 71249 (SA-I 101) 10 12 2 PBNP-I Intermediate to Lower Shell Circumferential Weld FENOC Weld Metal Heat 821T44 (WF-182-1)

Davis Besse Intermediate to 14 8 2

~ Lower Shell Circumferential I

Weld L

Total Specimens 44 32 8 5-5

AT1 Consulting Draft Report January 2003 Fracture Toughness Matenal Properties for PBNP-2 Table 5-5 PBNP-2 Fuel Cycles and Estimated Surveillance Capsule Fluence I Projected RPV Fluence at 53 EFPY: 5.085 x IO nlcm I (a) RPV accumulated fluence was determined from Reference 1261;RPV cycle fluence is the arithmetic difference from the preceding accumulated fluence value.

@) The lead factor for the supplemental surveillance capsule is 3.37. The capsule cycle fluence is the product of the lead factor and the RPV cycle fluence. Capsule accumulated fluence is the sum of capsule cycle fluences from time of capsule insertion to the EFPY of interest.

5-6

AT1 Consulting Draft Report, January 2003 Fracture Toughness Material Properties for PBNP-2 Figure 5-1 Arrangement of Surveillance Capsules in the Reactor Vessel REACTOR VESSEL THERMAL SHIELD 0"

90"

\ T (1.94)

\-v (3.37) 5-7

AT1 Consulting Draft Report, January 2003 Fracture Toughness Material Properties for PBNP-2 6

SUMMARY

AND CONCLUSIONS This report summarizes the fracture toughness testing that has been conducted to date on the PBNP-2 RPV limiting weld metal heat 72442. This weld metal was not included in the current surveillance program for PBNP-2, but it was irradiated as part of the B&W Owners Group integrated surveillance program. The latest projections based on Charpy impact testing, when analyzed following NRC guidelines and rules, indicate that the RPV material will reach the PTS screening criterion limit before EOLE. Therefore, fracture toughness testing of other irradiated surveillance specimens (from two different welds fabricated using weld wire 72442) has been performed and analyzed using the Master Curve methodology following ASME Code Cases N-629 and N-631. Use of the Master Curve methodology involves engineering assessment in applying the ASME Code Cases to an actual RPV. The evaluation performed here involves extrapolation to EOL and EOLE fluences and shows that the RPV limiting weld metal has more than adequate toughness out to EOLE and beyond. These projections will be confirmed by additional testing of weld heat 72442 from the B&W Owners Group MIRVP prior to reaching the EOL fluence at PBNP-2. A supplemental surveillance program also has been designed and implemented at PBNP-2 that includes the limiting weld metal for future evaluation using the Master Curve methodology. The testing of this supplemental capsule at a fluence corresponding to EOLE will confirm the toughness condition for the PBNP-2 RPV weld at about 38 EFPY, which is well before EOLE is reached.

The following observations and conclusions were reached from this current analysis of the limiting beltline weld metal:

The latest Charpy-based toughness evaluation following current regulation for the PBNP-2 limiting circumferential weld metal indicates that the PTS screening criterion of 3OO0F will be reached before EOLE when future plant operation is considered (removal of hafnium flux reduction and planned power up-rates).

Current application of the Master Curve methodology for the PBNP-2 weld metal requires extrapolation (from the three available surveillance irradiations) to the RPV EOLE fluence. This extrapolation can be performed folIowing several different approaches. Three approaches were evaluated here: (1) use of measured initial RTT, and adding Charpy shiA; (2) use of measured initial RTT, and adding the shift in RTT, due to irradiation; and, (3) use of the measured irradiated RTT, values directly without projection from zero fluence. All methods show that the EOLE RTpz value is less than the PTS screening limit of 30OoF. Method 1 somewhat follows the current regulatory practice and is the most conservative.

Method 2 was evaluated following the Kewaunee SE, and the resulting projections in ART were substantially less than Method 1. Method 3 is the most accurate method, and the results obtained applying this direct measurement approach reveal that Method 2 is quite conservative.

The unirradiated fracture toughness was evaluated using measurements from three weldments of the 72442 weld wire. When precracked Charpy (0.394T-3PB) specimens were used, a bias correction of 18'F was applied; when CT specimens of various sizes were used, no bias correction was used since the assumed level of 6-1

AT1 Consulting Draft Report. January 2003 Fracture Toughness Materlal Properties for PBNP-2 constraint is typical of the CT specimen. The three Toresults were within typical data scatter, and all of the data were combined and a single To value determined.

Irradiated fracture toughness data were evaluated for the three different irradiations, all within a small fluence range (less than the RPV weld EOL fluence). A bias correction of 80F was applied to 0.394T-3PB test results for the two data sets containing these types of specimens. This bias correction for irradiated 3PB specimens appears to be overly conservative based on a comparison of the limited results from the irradiated 0.394T-3PB and 0.936T-RCT tests.

0 The Margin term was chosen depending upon the analysis method. For Method 1, Margin was based on three uncertainties: material variability based on a Monte Carlo study from BAW-2308 of weld heat 72442 non-irradiated data (OMC =

9.3'F), the uncertainty in determining Tofrom ASTM E 1921 -02 ( O T ~= 7.4OF),

and the current regulatory value for weld metal Charpy shift (06 = 28'F); o h f c and C Q - ~are combined to give a measure of the uncertainty in initial properties ((TI =

1 1.9'F). Method 2 used the Margin specified by the NRC in the Kewaunee SE, which used a larger q (14OF) and the same oAof 28'F. Method 3 used a more complete uncertainty analysis: material variability (OMC = 9.3"F as above),

determination of irradiated To((TT~= 10.7'F), Cu content (ocu = 1.6-1.7'F), Ni content (ON,= 4.1-4.2'F), irradiation temperature (onn= 6.9-8.goF), fluence (cot =

13.2-1 2S°F), and fluence projection (crproj = 1 .0-1.6'F). Remaining consistent with industry practice, an approximate 95% statistical level (or two sigma)

Margin was chosen, where the individual uncertainties were combined as the square root sum of the squares.

Since there was a need to extrapolate to higher fluence levels (higher than where current fracture toughness measurements exist) to assess PTS and pressure-temperature operating curves, the current Regulatory fluence function for CVN-based predictions was used for the Master Curve approach.

The supplemental surveillance program utilizes irradiation of the limiting weld metal heat in a new capsule that will be available for testing near the time corresponding to 38 EFPY for the RPV. The direct measurement of fracture toughness for key weld metal will be evaluated at a fluence near the projected EOLE. Fracture toughness data from the B&W Owners Group on this same weld metal will be available around 2008. This B&W Owners Group data should correspond closely to the PBNP-2 EOL fluence for the limiting RPV weld.

6-2

__ ~~ ~~~

AT1 Consulting Draft Report, January 2003 Fracture Toughness Matenal Properties for PBNP-2 7 REFERENCES

1. Title 10 Code of Federal Register Part 50, Section 50.61, Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events, June 14, 1991.

2. T. J. Laubham and B. Server, Point Beach Units I and 2 Heatup and Cooldown Limit Czcrvesfor Normal Operation, WCAP-15976, January 2003.

3. ASTM Standard Test Method E 192 1-02, Test Method for the Determination of Reference Temperature, To, for Ferritic Steels in the Transition Range, Annual Book ofASTMStandards, Vol. 03.01, American Society for Testing and Materials, West Conshohocken, PA.

4. K. K. Yoon, Fracture Toughness Characterization of IVF-70 Weld Metal, )

B&W Owners Group Report, BAW-2202, September, 1993.

5. Safety Evaluation by the Ofice of Nuclear Reactor lncltiding the Use of a Master Curve-based MethodologVfor Reactor Pressure Vessel Integrity Assessment, Docket No. 50-305, May 2001.

6. B&W-2308, Initial RTNDTof Linde 80 Weld Materials, July 2002.

7. W. L. Server et al., Master Curve Fracture Toughness Application for BVPS-I, WCAP-15624, November 2001.

8. Use of Fracture Toughness Test Data to Establish Reference Temperattirefor Pressure Retaining Materials,Section XI, Division I , Code Case N-629, ASME Boiler and Pressure Vessel Code, American Society of Mechanical Engineers, New York, USA, 1998.

9. Use of Fracture Toughness Test Data to Establish Reference Temperaturefor Pressure Retaining Materials OIher Than Boltingfor Class I VesselsSection III, Division 1,Code Case N-63 1, ASME Boiler and Pressure Vessel Code, American Society of Mechanical Engineers, New York, USA, 1998.

10. ASTM Standard E 185-82, Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels, 1998 Annual Book of ASTMStandards, Vol. 12.02, American Society for Testing and Materials, West Conshohocken, PA.

11. S. E. Yanichko and G. C. Yula, Wisconsin Michigan Power Co.and the IVisconsin Electric Power Co.,Point Beach Unit No. 2 Reactor Vessel Radiation Surveillance Program, WCAP-77 12, June 1971.

12. Regulatory Guide 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials, US Nuclear Regulatory Commission, July 1988.

13. Response to Request for Additional Information (RAI)Regarding Reactor Pressure Vessel Integrity, BAW-2325, May 1998.

14. Best Estimate Copper and Nickel Values in CE Fabricated Reactor Vessel Welds, CE NPSD-1039, Revision 2, June 1997.

15. M. J. DeVan, K. K. Yoon, and S . Q. King, Test Results of Capsirle CR3-LG2 B& IV Owners Group Master Integrated Reactor Vessel Materials Surveillance Program, BAW-2254, October 1995.

16. M. J. DeVan, B&I.VOG Reactor Vessel Working Group Reactor Vessel Materials and Stirveillance Data Information, BAW-23 13, Revision 2, December 1999.

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AT1 Consunrng Draft Report, January 2003 Fracture Toughness Matertal Properties for PBNP-2

17. J. B. Hall and J. W. Newman, Jr., Analysis of the B&W Owners Group Capsule L l , BAW-2400, March 2002.

18. J . B. Hall and J. W. Newman, Jr., Analysis of the B& W Owners Group Capsule A3, BAW-2412, April 2002.

19. M . Scibetta, 3 - 0 Finite Element Simulation of the PCCv Specimen Statically Loaded in Three-Point Bending, Report BLG-860, SCWCEN, Mol, Belgium, March 2000.

20. W. L. Server et al., Application of Master Curve Fracture Toughness Methodologyfor Ferritic Steels, Revision 1 (PWRMRP-Ol), EPRI Report TRI 08390, Rev, 1, Palo Alto, CA, 1999.

21. W. L. Server et a]., Validation of Master Curve Fracture Toughness Methodology for RPVIntegrity Assessment (PWXkfRP-26),EPRI Report 1000707, Palo Alto, CA, 2000.

22. ASTM Standard Guide E 900-02. Predicting Radiation-Induced Transition Temperature Shift in Reactor Vessel Materials, E706 (IIF), Annual Book of ASTM Standards, Vol. 12.02, American Society for Testing and Materials, West Conshohocken, PA

23. W. L. Server et al., Charpy Embrittlernent Correlations - Status of Combined Mechanistic and Statistical Bases for US,Pressure Vessel Steels (MRP-45), PWR Materials Reliability Program (PIVRMRP), EPRI Report 1000705, Palo Alto, CA, 2001.

24. R. G. Lott et al., Master Curve Strategiesfor RPVAssessment, WCAP-15075, November 1998.

25. M. A. Sokolov and R. K. Nanstad, Comparisonof Irradiation-Induced ShiJs of K J and

~ Charpy Impact Toughnessfor Reactor Pressure Vessel Steels, NUREG/CR-6609, November 2000.

26. LTR-REA-02-23, Pressure Vessel Neutron Exposure Evaluations, Point Beach Unit 1 and 2, Rev. 0, Westinghouse Electric Company, February 2002.

27. BAW-l543A, Revision 4, Supplement 4, May 2002.

7-2

Westinghouse Non-Proprietary Class 3 WCAP-15976 February 2003 Revision 0 Point Beach Units I and 2 Heatup and Cooldown Limit Curves for Normal Operation

WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-15976,Revision 0 Point Beach Units 1 and 2 Heatup and Cooldown Limit Curves for Normal Operation T. J. Laubham Bill Server (ATI)

FEBRUARY 2003 Prepared by the Westinghouse Electric Company LLC for the Nuclear Management Company, LLC Approved:

J. A. Gresham, Manager Engineering and Materials Technology Westinghouse Electric Company LLC Energy Systems P.O. Box355 Pittsburgh, PA 15230-0355 02003 Westinghouse Electric Company LLC

WESTINGHOUSE NON-PROPRIETARY CLASS 3 All Rights Reserved

ii PREFACE This report has been technically reviewed and verified by:

J.H. Ledger RECORD OF REVISION Revision 0: Original Issue

iii TABLE OF CONTENTS LIST OF TABLES ........................................................................................................................................ iv LIST OF FIGURES ....................................................................................................................................... v EXECUTIVE

SUMMARY

.......................................................................................................................... vi 1 INTRODUCTION ............................................................................................................................ 1 2 FRACTURE TOUGHNESS PROPERTIES .................................................................................... 2 3 CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURERELATIONSHIPS.................... 4 4 CALCULATION OF ADJUSTED REFERENCETEMPERATURE.............................................. 8 5 HEATUP AND COOLDOWN PRESSURE-TEMPERATURELIMIT CURVES ........................ 23 6 REFERENCES ............................................................................................................................... 37 APPENDIX A MASTER CURVE EVALUATION ON PT. BEACH UNIT 2 GIRTH WELD ..............A-1 APPENDIX B DATA POINTS FOR CIRC-FLAWPT CURVES ....................................................... B-l

iv LIST OF TABLES Table 1 Summary of the Best Estimate Cu and Ni Weight Percent. Initial RTmTValues and Chemistry Factor values for the Point Beach Unit I and 2 Reactor Vessel Materials ........3 Table 2 Summary of the Calculated Peak Pressure Vessel Neutron Fluence Values at 34 and 53 EFPY used for the Calculation of ART Values (dcm.E > 1 .0 MeV) ...........9 Table 3 Summary of the Calculated Fluence Factors used for the Generation of the 34 and 53 EFPY Heatup and Cooldown Curves ........................................................................... 11 Table 4 Calculation of the Pt.Beach Unit 1 ART Values for the 1/4T Location @ 34 EFPY .......14 Table 5 Calculation of the Pt .Beach Unit 1 ART Values for the 3/4T Location @ 34 EFPY ....... 15 Table 6 Calculation of the Pt . Beach Unit 1 ART Values for the 1/4T Location @ 53 EFPY ....... 16 Table 7 Calculation of the Pt.Beach Unit 1 ART Values for the 3/4T Location @ 53 EFPY .......17 Table 8 Calculation of the Pt . Beach Unit 2 ART Values for the 1/4TLocation @ 34 EFPY .......18 Table 9 Calculation of the Pt . Beach Unit 2 ART Values for the 3/4T Location @ 34 EFPY .......19 Table IO Calculation of the Pt . Beach Unit 2 ART Values for the 1/4T Location @ 53 EFPY .......20 Table 11 Calculation of the Pt . Beach Unit 2 ARTValues for the 3/4T Location @ 53 EFPY .......21 Table 12 Summary of the Limiting ART Values Used in the Generation of the Pt.Beach Unit 1 and 2 Heatup/Cooldown Curves ....................................................................................... 22 Table 13 34 EFPY Heatup Curve Data Points Using 1996 App.G gC ASME Code Case N-64 1 (without Uncertainties for Instrumentation Errors) ....................................................... 27 Table 14 34 EFPY Cooldown Curve Data Points Using 1996App.G & ASME Code Case N-64 1 (without Uncertainties for Instrumentation Errors) ....................................................... 29 Table 15 53 EFPY Heatup Curve Data Points Using 1996App. G & ASME Code Case N-641 (without Uncertainties for InstrumentationErrors) ....................................................... 33 Table 16 53 EFPY Cooldown Curve Data Points Using 1996App. G & ASME Code Case N-641 (without Uncertainties for InstrumentationErrors) ....................................................... 35

V LIST OF FIGURES Figure 1 Point Beach Units 1 and 2 Reactor Coolant System Heatup Limitations (Heatup Rates of 60 & I OOOFhr) Applicable for the First 34 EFPY (Without Margins for Instrumentation Errors) Using I996 App. G Methodo fogy & ASME Code Case N-64 I 2 5 Figure 2 Point Beach Units 1 and 2 Reactor Coolant System Coo1down Limitations (Cooldown Rates up to 10O0Fhr) Applicable for the First 34 EFPY (Without Margins for Instrumentation Errors) Using 1996 App. G Methodology & ASME Code Case N-641 .26 Figure 3 Point Beach Units I and 2 Reactor Coolant System Heatup Limitations (Heatup Rates of 60 & 1OOOFhr) Applicable for the First 53 EFPY (Without Margins for Instrumentation Errors) Using 1996 App. G Methodology & ASME Code Case N-641 .3 1 Figure 4 Point Beach Units 1 and 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 100"Fhr) Applicable for the First 53 EFPY (Without Margins for Instrumentation Errors) Using 1996 App. G MethodoIogy & ASME Code Case N-64 1 .32

vi EXECUTIVE

SUMMARY

This report provides the methodology and results of the generation of heatup and cooldown pressure temperature (PT) limit curves for normal operation of the Point Beach Units 1 and 2 reactor vessels. The PT curves were generated based on the latest available reactor vessel information and updated calculated fluence projections that included the impact of the power uprating and removal of the Hafnium absorber rods.

The new Point Beach Unit I and 2 heatup and cooldown pressure-temperature limit curves were generated using adjusted reference temperature (ART) values that bound both units. The highest ART values from the hvo units were from the Unit 1 and Unit 2 intermediate to lower shell girth welds, however the limiting materials are actually the intermediate and lower shell axial welds from Unit 1, depending on the vessel thickness (% thickness or % thickness location). The axial welds become limiting over the girth weld through use of circ-flaw methodology from ASME Code Case N-588.This methodology is less restrictive than the standard axial-flaw methodology from the 1995 ASME Code,Section XI through the 1996 Addenda. In addition to the use of Code Case N-588,the PT curves also made use of ASME Code Case N-640,which allows the use of the KI,methodology. Both ASME Code Case N-588and N-640were joined together under ASME Code Case N-641.

The PT limit curves were generated for 34 and 53 EFPY using heatup rates of 60 and 1OO°F/hr and cooldown rates of 0,20,40,60and 100°F/hr. These curves can be found in Figures 1 through 4.

1 1 INTRODUCTION Heatup and cooldown limit curves are calculated using the adjusted R T ~ (reference T nil-ductility temperature) corresponding to the limiting beltline region material of the reactor vessel. The adjusted R T ~ ofT the limiting material in the core region of the reactor vessel is determined by using the unirradiated reactor vessel material fracture toughness properties, estimating the radiation-induced ART-, and adding a margin. The unirradiated RT- is designated as the higher of either the drop weight nil-ductility transition temperature (NDTT) or the temperature at which the material exhibits at least 50 A-lb of impact energy and 35-mil lateral expansion (normal to the major working direction) minus 60°F.

RT- increases as the material is exposed to fast-neutron radiation. Therefore, to find the most limiting R T m at any time period in the reactors life, A R T ~ due T to the radiation exposure associated with that time period must be added to the unirradiated R T m ( I R T ~ T ) .The extent of the shift in R T ~ isT enhanced by certain chemical elements (such as copper and nickel) present in reactor vessel steels. The Nuclear Regulatory Commission (NRC) has published a method for predicting radiation embrittlement in Regulatory Guide 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials.l Regulatory Guide 1.99, Revision 2, is used for the calculation ofAdjusted Reference Temperature (ART) values (IRTmT + ARTmr + margins for uncertainties) at the 1/4T and 3/4T locations, where T is the thickness of the vessel at the beltline region measured from the cladhase metal interface.

The heatup and cooldown curves documented in this report were generated using the most limiting ART values and the NRC approved methodology documented in WCAP-I 4040-NP-A, Revision 2I2],

Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldowvn Limit Curves with exception of the following: 1) T h e fluence values used in this report a r e calculated fluence values (i.e. comply with Reg. Guide 1.190), not the best estimate fluence values.

2) The KI,critical stress intensities are used in place of the K b critical stress intensities. This methodology is taken from approved ASME Code Case N-641I3] (which covers Code Cases N-640 and N-588). 3) The 1996 Version ofAppendix G to Section XI[41will be used rather than the 1989 version. 4)

PT Curves were generated with the most limiting circumferential weld ART value in conjunction with Code Case N-58833.These curves, which are included in Appendix A, are bounded by the curves using the standard axial flaw methodology from ASME Code 1996 App. G with the ART from the intermediate or lower shell axial welds depending of the flaw location ...1/4T versus 3/4T.

The purpose of this report is to present the calculations and the development o f the Point Beach Unit 1 and 2 heatup and cooldown curves for 34 and 53 EFPY. This report documents the calculated ART values and the development of the PT limit curves for normal operation. The PT curves herein were generated without instrumentation errors. The PT curves include a hydrostatic leak test limit curve from 2485 psig to 2000 psig, along with the pressure-temperature limits for the vessel flange region per the requirements of 10 CFR Part 50, Appendix G].

2 FRACTURE TOUGHNESS PROPERTIES The fracture-toughness properties of the ferritic materials in the reactor coolant pressure boundary are determined in accordance with the NRC Standard Review Plan[']. The beltline material properties of the Point Beach Unit 1 and 2 reactor vessels are presented in Table 1.

The chemistry factors were calculated using Regulatory Guide 1.99 Revision 2, Positions 1.1 and 2.1.

Position 1.1 uses the tables from the Reg. Guide along with the best estimate copper and nickel weight percents. Position 2.1 uses the surveillance capsule data from all capsules withdrawn to date.

3 TABLE 1 Summary of the Best Estimate Cu and Ni Weight Percent, Initial RTmTValues and Chemistry Factor values for the Point Beach Unit 1 and 2 Reactor Vessel Materials Material Description wt. % Cu wt. % Ni Initial R T W CF Nozzle Belt Forging (122P237)

Inter. Shell Plate (A981 1-1)

Lower Shell Plate (C1423-1)

Nozzle Belt to Intermediate Shell 1 0.19 y 0.07 0.57 1 -S°F 77°F 152.4"F I

Circ. Weld (8T1762)

Intermediate Shell Axial Weld - 0.17 0.52 I -5°F 138.2"F ID 27% (1 PO8 15)

Intermediate Shell Axial Weld - OD 157.6OF 73% (1P0661)

Intermediate to Lower Shell 167.6"F Girth Weld (71249) 1

~~

Lower Shell Axial Weld (61782) 0.23 0.52 -5°F Point Beach Unit 2 Nozzle Belt Forging (123V352) 0.11 0.73 40°F 76°F Inter. Shell Forging (123V500) 0.09 0.70 40°F 58°F Lower Shell Forging (122W195) 0. os 0.72 40°F 31 "F/43"F Nozzle Belt to Intermediate Shell 0.18 0.70 -56°F") 1 70°F Girth Weld (21935)

Intermediate to Lower Shell Circ. 0.26 0.60 -5°F 1 80°F Weld (72442)

NOTES:

(a) Data within this table can be found in WCAP-15121, Rev. as directed by NMC in Reference 8.

(b) Per Regulatory Guide 1.99, Rev. 2 Position 2.1.

(c) Generic Value of RTNDT.

4 3 CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS 3.1 OVERALL APPROACH The ASME approach for calculating the allowable limit curves for various heatup and cooldown rates specifies that the total stress intensity factor, KI,for the combined thermal and pressure stresses at any time during heatup or cooldown cannot be greater than the reference stress intensity factor, Kl,,for the metal temperature at that time. Kk is obtained from the reference fracture toughness curve, defined in Code Case N-64 1, Alternative Pressure-Temperature Relationship and Low Temperature Overpressure Protection System RequirementsSection XI, Division 17ps:41 of the ASME Appendix G to Section XI.

The KI,curve is given by the following equation:

where, Klc = reference stress intensity factor as a function of the metal temperature T and the metal reference nilductility temperature R T m This K1, curve is based on the lower bound of static critical KIvalues measured as a function of temperature on specimens of SA-533 Grade B Classl, SA-508-1, SA-508-2, SA-508-3 steel.

3.2 METHODOLOGY FOR PREXWRE-TEMPERATWE LIMIT CURVE DEVELOPMENT The governing equation for the heatup-cooldown analysis is defined in Appendix G of the ASME Code as follows:

where, Kh = stress intensity factor caused by membrane (pressure) stress KI, = stress intensity factor caused by the thermal gradients KI, = function of temperature relative to the RTmT of the material C = 2.0 for Level A and Level B service limits C = 1.5 for hydrostatic and leak test conditions during which the reactor core is not critical

For membrane tension, the corresponding KIfor the postulated defect is:

K I m = Mm x (pR,f t ) (3) where, M, for an inside surface flaw is given by:

M, = 1.85 for 4 <2, M, = 0 . 9 2 6 J for 2 1 fi I3.464, M, = 3.21 for fi >3.464 Similarly, M, for an outside surface flaw is given by:

M, = 1.77 for & <2, M, = 0.893& for 2 5 & 13.464, M, = 3.09 for 6> 3.464 and p = internal pressure, Ri = vessel inner radius, and t = vessel wall thickness.

For bending stress, the corresponding KI for the postulated defect is:

Kb = Mb

• Maximum Stress, where M bis two-thirds of M, The maximum KI produced by radial thermal gradient for the postulated inside surface defect of G-2120 is K1,

=0.953~10 ',

x ~CR x t2 where CR is the cooldown rate in OFhr., or for a postulated outside surface defect, K1, = 0.753~10"x HU x t2',where HU is the heatup rate in "Fhr.

The through-wall temperature difference associated with the maximum thermal KI can be determined from Fig. G-2214-1. The temperature at any radial distance from the vessel surface can be determined from Fig. G-2214-2 for the maximum thermal KI.

(a) The maximum thermal KIre'lationship and the temperature relationship in Fig. G-2214-1 are applicable only for the conditions given in G-2214.3(a)(l) and (2).

(b) Alternatively, the KIfor radial thermal gradient can be calculated for any thermal stress distribution and at any specified time during cooIdown for a %-thickness inside surface defect using the relationship:

f i i = (I.0359Co+ 0.6322C1+0.4753Cz + 0.3855C3)*& (4)

6 or similarly, K m during heatup for a %-thickness outside surface defect using the relationship:

fit = (1.043Co+O.630Ci+ 0.481C2+0.401C3) *& (5) where the coeficients Co, C1, C2 and Cj are determined from the thermal stress distribution at any specified time during the heatup or cooldown using the form:

a(x) = c o + Cl(X / a) + C2(x /a) + C3(X / c7)3 (6) and x is a variable that represents the radial distance from the appropriate (i.e., inside or outside) surface to any point on the crack front and a is the maximum crack depth.

Note, that equations 3,4 and 5 were implemented in the OPERLIM computer code, which is the program used to generate the pressure-temperature (P-T)limit curves. No other changes were made to the OPERLIM computer code with regard to P-T calculation methodology. Therefore, the P-T curve methodology is unchanged from that described in WCAP-I 4040, Methodology used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves12]Section 2.6 (equations 2.6.2-4 and 2.6.3-1) with the exceptions just described above.

At any time during thc heatup or cooldown transient, K1,is determined by the metal temperature at the tip of a postulated flaw at the 1/4T and 3/4T location, the appropriate value for RThDT, and the reference fracture toughness curve. The thermal stresses resulting from the temperature gradients through the vessel wall are calcuIated and then the corresponding (thermal) stress intensity factors, KI,, for the reference flaw are computed From Equation 2, the pressure stress intensity factors are obtained and, from these, the allowable pressures are calculated.

For the calculation of the allowable pressure versus coolant temperature during cooldown, the reference flaw ofAppendix G to the ASME Code is assumed to exist at the inside of the vessel wall. During cooldown, the controlling location of the flaw is always at the inside of the wall because the thermal gradients produce tensile stresses at the inside, which increase with increasing cooldown rates.

Allowable pressure-temperature relations are generated for both steady-state and finite cooldown rate situations. From these relations, composite limit curves are constructed for each cooldown rate of interest.

The use of the composite curve in the cooldown analysis is necessary because control of the cooldown procedure is based on the measurement of reactor coolant temperature, whereas the limiting pressure is actually dependent on the material temperature at the tip of the assumed flaw. During cooldown, the 1/4Tvessel location is at a higher temperature than the fluid adjacent to the vessel inner diameter. This condition, of course, is not true for the steady-state situation. It follows that, at any given reactor coolant temperature, the AT (temperature) developed during cooldown results in a higher value of &,at the 1/4T location for finite cooldown rates than for steady-state operation. Furthermore, if conditions exist so that the increase in KI,exceeds Kl,,the calculated allowable pressure during cooldown will be greater than the steady-state value.

7 The above procedures are needed because there is no direct control on temperature at the 1/4T location and, therefore, allowable pressures may unknowingly be violated if the rate of cooling is decreased at various intervals along a cooldown ramp. The use of the composite curve eliminates this problem and ensures conservative operation of the system for the entire cooldown period.

Three separate calculations are required to determine the limit curves for finite heatup rates. As is done in the cooldown analysis, allowable pressure-temperature relationships are developed for steady-state conditions as well as finite heatup rate conditions assuming the presence of a 1/4T defect at the inside of the wall. The heatup results in compressive stresses at the inside surface that alleviate the tensile stresses produced by internal pressure. The metal temperature at the crack tip lags the coolant temperature; therefore, the KI, for the 1/4Tcrack during heatup is lowcr than the KI, for the 1/4T crack during steady-state conditions at the same coolant temperature. During heatup, especially at the end of the transient, conditions may exist so that the effects of compressive thermal stresses and lower K,,values do not offset each other, and the pressure-temperature curve based on steady-state conditions no longer represents a lower bound of all similar curves for finite heatup rates when the 1/4T flaw is considered. Therefore, both cases have to be analyzed in order to ensure that at any coolant temperature the lower value of the allowable pressure calculated for steady-state and finite heatup rates is obtained.

The second portion of the heatup analysis concerns the calculation of the pressure-temperature limitations for the case in which a 1/4T flaw located at the 1/4T location from the outside surface is assumed. Unlike the situation at the vessel inside surface, the thermal gradients established at the outside surface during heatup produce stresses, which are tensile in nature and therefore tend to reinforce any pressure stresses present. These thermal stresses are dependent on both the rate of heatup and the time (or coolant temperature) along the heatup ramp. Since the thermal stresses at the outside are tensile and increase with increasing heatup rates, each heatup rate must be analyzed on an individual basis.

Following the generation of pressure-temperature curves for both the steady-state and finite heatup rate situations, the final limit curves are produced by constructing a composite curve based on a point-by-point comparison of the steady-state and finite heatup rate data. At any given temperature, the allowable pressure is taken to be the lesser of the three values taken from the curves under consideration. The use of the composite curve is necessary to set conservative heatup limitations because it is possible for conditions to exist wherein, over the course of the heatup ramp, the controlling condition switches from the inside to the outside, and the pressure limit must at all times be based on analysis of the most critical criterion.

3 3 CLOSURE HEADNESSEL FLANGE REQUIREMENTS IO CFR Part 50, Appendix GIs1addresses the metal temperature of the closure head flange and vessel flange regions. This rule states that the metal temperature of the closure flange regions must exceed the material unirradiated RTmT by at least 120°F for normal operation when the pressure exceeds 20 percent of the preservice hydrostatic test pressure (3107 psi), which is 621 psig for Pt. Beach Unit 1. The limiting unirradiated RTmTof 60°F occurs in the vessel flange of the Pt. Beach Unit 2 reactor vessel, so the minimum allowable temperature of this region is 180°F at pressures greater than 621 psig. This limit is shown in Figures 1 and 3 wherever applicable.

8 4 CALCULATION OF ADJUSTED REFERENCE TEMPERATURE From Regulatory Guide 1.99, Revision 2, the adjusted reference temperature (ART) for each material in the beltline region is given by the following expression:

ART = Initial RTmT + ART- + Margin (7)

Initial RTmT is the reference temperature for the unirradiated material as defined in paragraph NB-233 1 of Section 111 of the ASME Boiler and Pressure Vessel Code[]. If measured values of initial RTmT for the material in question are not available, generic mean values for that class of material may be used if there are sufficient test results to establish a mean and standard deviation for the class.

A R T ~ isT the mean value of the adjustment in reference temperature caused by irradiation and should be calculated as follows:

To calculate A R T ~ at T any depth (e.g., at 1/4T or 3/4T), the following formula must first be used to attenuate the fluence at the specific depth.

where x inches (vessel beltline thickness is 6.5 inches) is the depth into the vessel wall measured from the vessel cladbase metal interface. The resultant fluence is then placed in Equation 8 to calculate the ARTNDT at the specific depth.

The Westinghouse Radiation Engineering and Analysis Group evaluated the vessel fluence projections and the results of the calculated peak fluence values at various azimuthal locations on the vessel cladhase metal interface are presented in Table 2. The evaluation used the ENDFB-VI scattering cross-section data set. The calculated fluence projections were determine using methods consistent with Regulatory Guide 1 .I 90, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence.rol Table 3 contains the 1/4T and 3/4Tfluence factors, per the Regulatory Guide 1.99, Revision 2, used to calculate the ART values for all beltline materials in the Pt. Beach Unit 1 and 2 reactor vessels.

9 TABLE 2 Summary of the Calculated Peak Pressure Vessel Neutron Fluence Values at 34 and 53 EFPY used for the Calculation ofARTValues (n/cm2, E > 1.O MeV)

Component Description I Surface) I 1/4T(b) i 3/4T(b)

PBNP Uni! I : 34 EFPY (End oJ Current License)

Nozzle Belt Forging (122P237) 0.25 0.17 0.08 Inter. Shell Plate (A981 1-1) 3.38 2.29 1 .os Lower Shell Plate (C1423-1) 3.04 2.06 0.94 Nozzle Belt to Intermediate Shell 0.25 0.17 0.08 Circ. Weld (8T1762)

Intermediate Shell Axial Weld - ID 2.1 9 1.48 0.68 27% (1 PO8 15)

Intermediate Shell Axial Weld - 2.19 1.48 0.68 OD 73% ( 1 PO661)

Intermediate to Lower Shell Circ. 3.05 2.07 0.95 WeId (71249)

Lower Shell Axial Weld (61782) 2.08 1.41 0.65 PBNP Uni! I : 53 EFPY (End of License Elrrension)

Nozzle Belt Forging (122P237) 0.42 0.28 0.13 Inter. Shell Plate (A981 1-1) 536 3.56 1.63 Lower Shell Plate (C1423-1) I 4.79 I 3.24 1 I .49 Nozzle Belt to Intermediate Shell 0.42 0.28 0.13 Circ. Weld (8T1762)

Intermediate Shell Axial Weld ID - 3.44 2.32 1.07 27% (1P0815)

Intermediate Shell Axial Weld - 3.44 2.32 1.07 OD 73% (I PO66 I )

Intermediate to Lower Shell Circ. 4.91 3.32 1.52 Weld (71249)

Lower Shell Axial Weld (61782) 3.37 2.28 1.os

10 TABLE 2- Continued Summary of the Calculated Peak Pressure Vessel Neutron Fluence Values at 34 and 53 EFPY used for the Calculation of ART Values (n/cm2, E > 1 .O MeV) 1

~ ~ ~~~

Component Description Surface() 1/4T@ 3/4T@

I PBi3 Nozzle Belt Forging (123V352) 0.34 0.23 0.1 1 Inter. Shell Forging (123V500) 3.38 2.29 1.05 I 1

~~ ~~~

Lower Shell Forging (122W195) 3.30 2.23 1.02 Nozzle Belt to Intermediate 0.34 0.23 0.1 1 Shell Circ. Weld (21935)

Intermediate to Lower Shell 3.13 2.12 0.97 Circ. Weld (72442)

I PBNP Utrif2: 53 EFPY (End of License Extension)

I Nozzle Belt Forging (123V352) 0.55 I 0.37 0.1 7 I Inter. Shell Forging (123V500) 5.39 3.65 1.67 Lower Shell Forging (122W195) 5.32 3.60 1.65 Nozzle Belt to Intermediate 0.55 0.37 0.1 7 Shell Circ. Weld (21935)

Intermediate to Lower Shell 5.09 3.46 1.58 Circ. Weld (72442)

NOTES:

(a) These fluence values are the calculated fluence values considering the power uprate without hafnium suppression rods.

@) Neutron attenuation per Reg. Guide 1.99,Rev. 2 Contained in Table 3 is a summary of the fluence factor (FF) values used in the calculation of adjusted reference temperatures for the Pt. Beach Unit land 2 reactor vessel beltline materials for 34 and 53 EFPY.

TABLE 3

11 Summary of the Calculated Fluence Factors used for the Generation of the 34 and 53 EFPY Heatup and Cooldown Curves Lower Shell Axial Weld 1.41 x lot9 1.10 0.65 x 10" 0.88 (61782)

Nozzle Belt Forging (122P237) 0.28 iot9 0.65 0.13 x lot9 0.47 Inter. Shell Plate (A981 1-1) I 3.56 1019 I 133 I 1.63~IOl9 I

Lower Shell Plate (C1423-1) 3.24 10'~ 1.31 1.49 x ioq9 1.11 Nozzle Belt to Intermediate 0.28 1019 0.65 0.13 iot9 0.47 Shell Circ. Weld (8T1762)

Intermediate Shell Axial Weld - 2.32 x IOt9 1.23 1.07 x loL9 NIA ID 27% (IP0815)

Intermediate Shell Axial Weld - 2.32 iot9 NIA 1.07 1oL9 1.02 OD 73% (1P0661)

Intermediate to Lower She11 3.32 iot9 1.31 1.52 x iot9 1.12 Circ. Weld (71249)

Lower Shell Axial Weld 2.28 1019 1.22 1.0s iot9 1.01 (61782)

12 TABLE 3 - Continued Summary of the Calculated FJuence Factors used for the Generation of the 34 and 53 EFPY Heatup and Cooldown Curves N o d e Belt Forging 0.23 1019 0.60 0.11 1019 0.44 (1 23v352)

Inter. Shell Forging (123V500) 2.29 1019 1.22 1.os 1019 1.01 Lower Shell Forging 2.23 1019 1.22 1.02 1019 1.01 (122W 195)

Nozzle Belt to Intermediate Shell Circ. Weld (21935) 0.23 1019 I OAO I 0.11 x 1019 1 Oi4 Intermediate to Lower Shell Circ. Weld (72442) 2.12 1019 I 1.20 I

~_____

PBNP Unit 2: f License Extension) 0.97 1019 I 0.99 Nozzle Belt Forging 037x 1019 0.17 x 1019

( 123V3 52)

~~ ~

Inter. Shell Forging (123V500) 3.6s 1019 1.67 x l O I 9 Lower Shell Forging 3.60 1019 1.33 1.65 x 1019 1.14 (122W 195)

Nozzle Belt to Intermediate 0.37 1019 0.73 0.17 x 1019 0.53 Shell Circ. Weld (21935)

Intermediate to Lower Shell 3.46 1019 1.32 1.58 x 10 1.13 Circ. Weld (72442)

13 Margin is calculated as, M = 2 Jcrt + cri . The standard deviation for the initial R T ~ margin T term, is cr, 0°F when the initial R T ~ isTa measured value, and 17°F when a generic value is available. The standard deviation for the A R T ~ margin T term, nA,is 17°F for plates or forgings, and 8S°F for plates or forgings when surveillance data is used. For welds, oAis equal to 28'F when surveillance capsule data is not used, and is 14°F (halfthe value) when credible surveillance capsule data is used. uAneed not exceed 0.5 times the mean value of A R T ~ T .

Contained in Tables 4 through 1 1 are the calculations of the 34 and 53 EFPY ART values used for generation of the heatup and cooldown curves.

14 TABLE 4 Calculation of the Pt. Beach Unit 1 ART Values for the 1/4T Location @ 34 EFPY Materiol Nozzle Belt Forging (122P237)

Inter. Shell Plate (A981 1-1)

~ ~~

Lower Shell Plate (C1423-1)

RG 1.99 R2 Method Position 1.1 Position 1.1 Position 2.1 CF (F) 77 88 79.3 FF 0.53 1.22 JRT&)

(F) 50 1

ART~T(C)

(OF) 40.8 107.4 1 ::::

u1 (OF) 0 26.9 1.22 1 96.7 26.9 OA (OF) 17 17 8.5 17

~

Margin (F) 34 63.6 56.4 63.6 I-....................

1 I I 1 1 6 643.0. 4 8.5 56.4 100 Nozzle Belt to Intermediate Shell Circ. Weld (8T1762)

Position 1.1 152.4 0.53 -5 80.8 19.7 28 68.5 1 144 I Intermediate Shell Axial Weld Position 1.1 138.2 1.1 I -5 153.4 19.7 28 68.5

- ID 27% (1P0815)

Intermediate Shell Axial Weld Position 1.1 157.6 NIA -5 NIA 19.7 28 68.5

- OD 73% (1P0661)

Intermediate to Lower Shell Circ. Weld (71249)

Position 1.1 167.6 1.20 10 201.1 0 28 56 I 267 I Lower Shell Axial Weld (61782) Position 1.1 157.4 1.10 Position 2. I 163.3 1.10

-5

-5 173.1 19.7 28 179.6 19.7 (a) Initial RTmT values are generic except the Nozzle Belt Forging and the Intermediate to Lower Shell Circ. Weld. See WCAP-15121, Rev. ]Iq.

14 T I

(b) ART = lnitial R T N DC~A R T ~ +T Margin (F)

(C) ARTNM= CF FF

TABLE 5 Calculation of the Pt. Beach Unit 1 ART Values for the 3/4TLocation @ 34 EFPY 1 R Method G (OF) I T OA (F)

Margin (F)

ART'^)

Nozzle Belt Forging(122P237) I Position 1.1 I 77 I 0.37 50 28.5 0 17 34 113 Inter. Shell Plate (A981 1-1) Position 1.1 1 88.9 26.9 154 Position 2.1 1 80.1 26.9 138 Lower Shell Plate (C1423-1) Position 1.1 1 54.2 26.9 17 119 Position 2.1 1 35.1 26.9 8.5 93 Nozzle Belt to Intermediate Shell Position 1.1 152.4 0.37 28 68.5 120 Circ. Weld (8T1762)

Intermediate Shell Axial Weld Position 1.1 138.2 N/A 28 68.5 NIA ID 27% (1 PO815)

Intermediate Shell Axial Weld OD 73% (1P0661)

I Position 1.1 1 157.6 I ~ ~ 0.89 -5 1 140.3 -1 19.7 28 68.5 204 Intermediate to Lower Shell Circ. Weld (7 1249) 1 Position 1 .I 1 167.6 I 0*99 10 I 165.9 1 0 28 56 232 Lower Shell Axial Weld (61782) Position 1.1 138.5 68.5

............. .............202 Position 2.1 143.7 48.3 I87 Notes:

(a) Initial R T ~ values T are generic except the Nozzle Belt Forging and the Intermediate to Lower Shell Circ. Weld. See WCAP-15121, Rev. l?

(b) ART = Initial RTNDT + A R T ~ +T Margin (OF)

(c) A R T ~ =TCF FF

TABLE 6 Calculation of the Pt. Beach Unit 1 ART Values for the 1/4TLocation @ 53 EFPY I

Material RG 1.99 R2 CF FF I R T ~ART&) ~ JI OA Margin ART@)

Method (OF) (F) (OF) (F) (OF) (OF) r Nozzle Belt Forging (122P237) Position 1.1 77 0.65 50 50.1 0 17 34 I34 Jnter. Shell Plate (A9811-1) Position 1.1 88 I .33 1 117.0 26.9 17 63.6 182 Position 2.1 79.3 1.33 1 105.5 26.9 8.5 56.4 163.

Lower Shell Plate (C1423-1) Position 1.1 55.3 1.31 1 72.4 26.9 17 63.6 137 Position 2.1 35.8 1.31 1 46.9 26.9 8.5 56.4 104 Nozzle Belt to Intermediate Shell Position 1.1 152.4 0.65 -5 99.1 19.7 28 68.5 163 Circ. Weld (8T1762)

Intermediate Shell Axial Weld Position 1.1 138.2 1.23 -5 170.0 19.7 28 68.5 234

- lD27%(1P0815)

Intermediate Shell Axial Weld Position 1.1 157.6 NIA -5 NIA 19.7 28 68.5 NIA

- OD 73% (lPO661)

Intermediate to Lower Shell Position 1.1 167.6 1.31 IO 219.5 0 28 56 286.

Circ. Weld (71249)

Lower Shell Axial Weld (61782) Position 1.1 157.4 1.22 -5 192.0 19.7 28 68.5 256 Position 2.1 163.3 I .22 -5 199.2 19.7 14 48.3 243 Notes:

17 TABLE 7 Calculation of the Pt. Beach Unit 1 ART Values for the 3/4TLocation @ 53 EFPY Material RG 1.99 R2 CF FF IRT&) ART,.,,?) 61 OA Margin AR~~)

Method (OF) (OF) (OF) (9 (OF) (OF)

I Nozzle Belt Forging (1 22P237) Position 1.1 77 0.47 50 36.2 0 17 34 120 Inter. Shell Plate (A981 1-1) Position 1.1 88 1.13 1 99.4 26.9 17 63.6 164 Position 2.1 79.3 1.13 I 89.6 26.9 8.5 56.4 147 Lower Shell Plate (C1423-1) Position 1.1 55.3 1.1 1 1 61.4 26.9 17 63.6 126 Position 2.1 35.8 1.1 1 1 39.7 26.9 8.5 56.4 97 Nozzle Belt to Intermediate Shell Position 1.1 152.4 0.47 -5 71.6 19.7 28 68.5 135 Circ. Weld (8T1762)

Intermediate Shell Axial Weld Position 1.1 138.2 N/A -5 NIA 19.7 28 68.5 N/A

- lD27%(1P0815)

Intermediate Shell Axial Weld Position 1.1 157.6 I .02 -5 1 60.8 19.7 28 68.5 224

- OD 73% (lP0661)

Intermediate to Lower Shell Position 1.1 167.6 1.12 IO 187.7 0 28 56 254 Circ. Weld (7 1249)

Lower Shell Axial Weld (61782) Position 1.1 157.4 1.01 -5 159.0 19.7 28 68.5 223 Position 2.1 163.3 1.01 -5 164.9 19.7 14 48.3 208 Notes:

TABLE 8 Calculation of the Pt. Beach Unit 2 ART Values for the 1/4TLocation @ 34 EFPY Material RG 1.99 R2 CF FF IRThmT(I) ARTp,n7(C) 01 UA Margin ART@)

Method (OF) (OF) (F) (OF) (OF) - (bF)

Noale Belt Forging (l23V352) Position 1.1 76 0.60 40 45.6 0 17 34 120 Position 1.1 58 1.22 40 70.8 0 17 34 145 Lower Shell Forging (122W195) Position 1.1 31 1.22 40 37.8 0 17 34 112 Position 2.1 43 1.22 40 52.5 0 8.5 17 1 IO Nozzle Belt to Intermediate Shell Position 1.1 170 0.60 -56 102 17 28 65.5 112 Circ. Weld (21935)

Intermediate to Lower Shell Position 1.1 180 1.20 -5 216 19.7 28 68.5 280 Circ. Weld (72442) (220) (a

~~ ~ ~~ ~~ ~

Notes:

(a) Tnitial RTNDT values are measured values except for the welds. See WCAP-15121,Rev.

(b) ART = Initial R T ~ +TA R T ~ +TMargin (F)

(C) A R T ~ = T CF

• FF (d) Value in parenthesis was calculated using Master-Curve Technology. See Appendix A.

19 TABLE 9 Calculation of the Pt. Beach Unit 2 ART Values for the 3/4T Location @ 34 EFPY I

Material Nozzle Belt Forging (123V352)

RG 1.99 R2 Method Position 1.1 CF (dF) 76 FF 0.44 IRT~.#

(OF) 40 ART^$

(OF) 33.4 oI (F) 0 UA (F) 17 Margin (OF) 34 ART@

107 I

Inter. Shell Forging (123V500) Position 1.1 58 1.01 40 58.6 0 17 34 133 Lower Shell Forging (1 22W 195) Position 1.1 31 1.01 40 31.3 0 17 34 Position 2.1 43 1.01 40 43.4 0 8.5 17 Nozzle Belt to Intermediate Shell Position 1.1 170 0.44 -56 74.8 17 28 65.5 Circ. Weld (2 1935)

Intermediate to Lower Shell Position 1.1 180 0.99 -5 178.2 19.7 28 68.5 Circ. Weld (72442) (185)(*

20 TABLE 10 Calculation of the Pt. Beach Unit 2 ARTValues for the 1/4TLocation @ 53 EFPY

21 TABLE 11 Calculation of the Pt. Beach Unit 2 ART Values for the 3/4T Location @ 53 EFPY Material RG 1.99 R2 CF FF IRT~?) ART,,^) <T[

Method (OF) (0 (OF) (OF)

Nozzle Belt Forging (123V352) Position 1.1 76 0.53 40 40.3 0 Inter. Shell Forging (123V500) Position 1.1 58 1.14 40 66.1 0 Lower Shell Forging (122W195) Position 1.1 31 1.14 40 35.3 0 17 34 109 Position 2.1 43 1.14 40 49.0 0 8.5 17 I06 Nozzle Belt to Intermediate Shell Position 1.1 170 0.53 -56 90.1 17 I I 100 Circ. Weld (21935)

Intermediate to Lower Shell Circ. Weld (72442)

Position 1.1 180 1.13 -5 203.4 19.7 zrlzr 28 65s (207)

22 Contained in Table 12 is a summary of the limiting ARTs to be used in the generation of the Pt. Beach Units 1 and 2 PT limit curves. The limiting curves, which bound both units, will be presented in Section 5 . When considering the master-curve ART for the Unit 2 intermediate to lower shell girth weld, the highest ART would then come from the Pt. Beach Unit 1 intermediate to lower shell Circumferential weld. Thus, the Circ-Flaw methodology from Code Case N-641 (Also known as CCN-588) will be used in conjunction with the Pt. Beach Unit 1 Circ-Flaw ART for generating PT curves. However, the highest Axial-Flaw material must also be run to check for limitations in the P T curves. The most limiting Axial-Flaw ART comes from the Pt. Beach Unit 1 lower shell axial welds (for 1/4T)and the intermediate shell axial welds (for 3/4T).

TABLE 12 Summary of the Limiting ART Values Used in the Generation of the Pt. Beach Unit 1 and 2 Heatup/Cooldown Curves EFPY Limiting Circ-Flaw ART (I) Limiting Axial-Flaw ART (OF)

%T(F) %T (F) %T(F) %T (F) 34 267 232 223 204 53 286 254 243 224 34 280 (220) ( 242(185)(d 145 133 53 3 0 1 (239)(dl 267 (207)(d 152 140 Notes:

(a) Pt. Beach Units 1 and 2 Limiting Circ. Flaw ART comes from the Intermediate to lower shell circumferential welds (Heat #s 71249 and 72442, respectively).

(b) The Axial-Flaw ARTS for Pt. Beach Unit 1 are from the lower shell axial welds (1/4T) and the intermediate shell axial welds (3/4T).

(c) The Axial-Flaw ARTs for Pt. Beach Unit 2 are fiom the intermediate shell forging 123V500.

(d) Value in parenthesis was calculated using Master-Curve Technology. See Appendix A.

23 5 HEATUP AND COOLDOWN PRESSURE-TEMPERATURE LIMIT CURVES Pressure-temperature limit curves for normal heatup and cooldown of the primary reactor coolant system have been calculated for the pressure and temperature in the reactor vessel beltline region using the methods discussed in Sections 3.0 and 4.0 of this report. This approved methodology is also presented in WCAP-14040-NP-A7Revision 2 with exception to those items discussed in Section 1 of this report.

Master-Curve technology is also presented within this report, however its use does not alter the results, as discussed hereafter.

Figures 1 and 3 present the limiting heatup curves without margins for possible instrumentation errors using heatup rates of 60 and 100°F/hr applicable for the first 34 and 53 EFPY,respectively. These curves were generated using a combination of the1996 ASME Code Section XI, Appendix G with the limiting ART values from the Unit 1 intermediate and lower shell longitudinal welds and the ASME Code Case N-641. These heatup curves bound those generated using the Circ-flaw methodology portion ofASME Code Case N-641 with the limiting circ-weld ART value from the Unit 1 intermediate to lower shell girth weld (See data points in Appendix B). Additionally, the limiting heatup curves presented in this section would also be more limiting than curves generated with limiting circ-weld ART value from the Unit 2 intermediate to lower shell girth weld; assuming the Master-Curve technology was not implemented (Again, see Appendix B for comparison).

Figures 2 and 4 present the limiting cooldown curves without margins for possible instrumentation errors using cooldown rates of 0,20,40, 60 and 100Fhr applicable for 34 and 53 EFPY, respectively. Again, these curves were generated using a combination of the1996 ASME Code Section XI, Appendix G with the limiting ART values from the Unit I intermediate and lower shell longitudinal welds and the ASME Code Case N-641. These cooldown curves bound those generated using the Circ-flaw methodology portion ofASME Code Case N-641 with the limiting circ-weld ART value from the Unit 1 intermediate to lower shell girth weld (See Appendix 3). Additionally, the limiting cooldown curves presented in this section would also be more limiting than curves generated with limiting circ-weld ART value from the Unit 2 intermediate to lower shell girth weld; assuming the Master-Curve technology was not implemented (See Appendix 3).

Allowable combinations of temperature and pressure for specific temperature change rates are below and to the right of the limit line shown in Figures 1 through 4. This is in addition to other criteria, which must be met before the reactor is made critical, as discussed below in the following paragraphs.

The reactor must not be made critical until pressure-temperature combinations are to the right of the criticality limit line shown in Figures 1 and 3. The straight-line portion of the criticality limit is at the minimum permissible temperature for the 2485 psig inservice hydrostatic test as required by Appendix G to 10 CFR Part 50. The governing equation for the hydrostatic test is defined in Code Case N-641 as follows:

24

where, KI, is the stress intensity factor covered by membrane (pressure) stress, KIE= 33.2 + 20.734 e 1 0 0 2 c T - R T ~ ) 1 ,

T is the minimum permissible metal temperature, and RTmTis the metal reference nil-ductility temperature.

The criticality limit curve specifies pressure-temperature limits for core 0perat.m to prov de ai jitiona margin during actual power production as specified in Reference 5. The pressure-temperature limits for core operation (except for low power physics tests) are that the reactor vessel must be at a temperature equal to or higher than the minimum temperature required for the inservice hydrostatic test, and at least 40°F higher than the minimum permissible temperature in the corresponding pressure-temperature curve for heatup and cooldown calculated as described in Section 4.0 of this report. For the heatup and cooldown curves without margins for instrumentation errors, the minimum temperatures for the in service hydrostatic leak tests for the Point Beach Unit 1 and 2 reactor vessels at 34 and 53 EFPY are 273'F and 293OF, respectively. The vertical line drawn from these points on the pressure-temperature curve, intersecting a curve 40°F higher than the pressure-temperature limit curve constitutes the limit for core operation for the reactor vessel.

Figures 1 through 4 define all of the above limits for ensuring prevention of nonductile failure for the Point Beach Unit 1 and 2 reactor vessels at 34 and 53 EFPY. The data points used for the heatup and cooldown pressure-temperature limit curves shown in Figures 1 through 4 are presented in Tables 13 through 16.

MATERIAL PROPERTY BASlS LlMlTING ART VALUES AT 34 EFPY 1/4T, 223°F (LOWER SHELL LONG. WELDS) 3/4T, 204°F (INTER. SHELL LONG. WELDS) 2500

-loperlln 7

ersion 5 1 Run 5655 2250 -- -

- I Leak Test Limit I

2000 -- --

1750 -----

A s!

1500 E3 VI VI 2 1250 -160 D e s . F/Hr&

n mQ)

-=

c.

0

-0 1000 -

0" 750 -.

500 -__ - -,

Boitup - Criticality Limit based on 250 Temp. inservice hydrostatic test temperature (273 F) for the service period up to 34 E F P Y 0 -rr, r , , ,

I , t r , ,

I, , , r .

I , , , , [ , , , ,, , ,

0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)

Figure 1 Point Beach Units 1 and 2 Reactor Coolant System Heatup Limitations (Heatup Rates of 60 & 100°F/hr) Applicable for the First 34 EFPY (Without Margins for Instrumentation Errors) Using 1996 A p p G Methodology & ASME Code Case N-641

26 MATERIAL PROPERTY BASIS LIMITING ART VALUES AT 34 EFPY: 1/4T, 223°F (LOWER SHELL LONG. WELDS) 3/4T, 204'F (INTER.SHELL LONG. WELDS) 2500 2250 2000 1750 J-1500 1250 1000 250 0

0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)

Figure 2 Point Beach Units 1 and 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 100°F/hr) Applicable for the First 34 EFF'Y (Without Margins for lnstrumentation Errors)Using 1996 App.G Methodology & ASME Code Case N-641

TABLE 13 34 EFPY Heatup Curve Data Points Using 1996 App. G & ASME Code Case N-641 (without Uncertainties for Instrumentation Errors) 60" h r. Critica y Limit 100' Ihr. Critica' y Limit Leak Test Limit Temp. Press. Temp. Press. Temp. Press. Temp. Press. Temp. Press.

(OF) (PSW (OF) (PSid (OF) (Psig) (OF) (PSif9 (OF) 60 0 273 0 60 0 273 0 254 60 62 1 273 62 1 60 62 1 273 62 1 273 65 62 1 273 62 1 65 62 1 273 62 1 70 62 1 2 73 62 1 70 62 1 273 62 1 75 62 I 273 62 1 75 62 1 273 62 1 80 62 1 273 62 1 80 62 1 273 62 1 85 62 1 273 62 1 85 62 1 273 62 1 90 62 1 273 62 1 90 62 I 273 62 1 95 62 1 273 62 1 95 62 1 273 62 1 100 62 1 273 62 1 100 62 1 273 62 1 105 62 1 273 62 1 105 62 1 273 62 1 I10 62 1 2 73 62 1 I10 62 1 273 62 1 115 62 1 273 62 1 1 I5 62 1 273 62 I 120 62 1 273 62 1 120 62 1 273 62 1 125 62 1 273 62 1 125 62 1 273 62 I 130 62 1 273 62 1 130 62 I 273 62 1 135 62 1 273 62 1 135 62 I 273 62 1 140 62 1 273 62 1 140 62 1 273 62 1 145 62 1 273 62 I 145 62 1 273 62 1 150 62 1 273 62 1 150 62 1 273 62 1 155 62 1 273 62 1 155 62 1 273 62 1 160 62 I 273 62 1 160 62 1 273 62 1 165 62 1 273 62 1 165 62 1 273 62 1 170 62 1 273 62 1 170 62 I 273 62 1 175 62 1 273 836 175 62 1 273 750 180 62 1 273 856 180 62 1 273 765 180 836 273 878 180 750 273 782 185 856 273 902 185 765 273 80 1 I90 878 273 929 190 782 273 82 1 195 902 273 959 195 80 1 273 845 200 929 273 992 200 82 1 273 870 205 959 273 1028 205 845 273 898 210 992 273 1068 210 870 273 929 215 1028 273 1112 215 898 273 964 220 1068 273 1161 220 929 273 1002 225 1112 275 1215 225 964 275 1044 230 1161 280 1274 230 1002 280 1090 235 1215 285 1340 235 1044 285 1141 240 1274 290 1411 240 1090 290 1198 245 1340 295 1477 245 1141 295 1260 250 1411 300 1549 250 1198 300 I329 255 1477 305 1628 255 1260 305 1405

28 TABLE 13 (Continued) 34 EFPY Heatup Curve Data Points Using 1996 App. G & ASME Code Case N-64 1 (without Uncertainties for Instrumentation Errors)

Critical y Limit Critica ty Limit Leak Test Limit Temp. Press. Temp. Press.

(F) (psig) (F) (PSifZ) 260 1549 310 1716 260 1329 310 1489 265 I628 315 1813 265 1405 315 1582 270 1716 320 1920 270 1489 320 1684 275 1813 325 2038 275 1582 325 1797 280 I920 330 2168 280 1684 330 1921 285 2038 335 2312 285 1797 335 2059 290 2168 340 2471 290 1921 340 2210 295 23 12 295 2059 345 2377 300 2471 300 2210 I

305 2377

29 TABLE 14 34 EFPY Cooldown Curve Data Points Using 1996 App. G & ASME Code Case N-64 1 (without Uncertainties for InstrumentationErrors)

Steac State 200 br. do0 9 r. 60' hr. 100' 'hr.

T (OF) 60 p( P 0

W 60

-LEL -

p(

0 PW 60 p(

0 PW T ("0p ( P W 60 0 T (OF) 60 p (Pa) 0 60 621 60 621 60 621 60 621 60 563 65 621 65 621 65 621 65 621 65 564 70 621 70 621 70 621 70 621 70 566 75 621 75 621 75 621 75 62I 75 568 80 621 80 621 80 621 80 621 80 57I 85 621 85 621 85 621 85 621 85 574 90 621 90 621 90 621 90 621 90 577 95 621 95 621 95 621 95 621 95 580 100 621 100 621 100 62I 100 621 100 584 105 621 105 621 105 621 105 621 105 589 110 621 110 621 110 621 1 IO 621 I IO 594 1 I5 621 115 621 115 621 1 I5 621 115 599 120 62I 120 621 120 62I 120 621 120 606 125 62I 125 621 125 621 125 621 125 613 I30 621 130 621 130 62I 130 621 130 621 135 621 135 621 135 621 135 62I 135 621 140 621 140 621 140 621 140 62I 140 621 145 621 I45 621 145 621 145 62I 145 621 150 621 150 621 150 621 I50 621 150 621 155 621 155 621 155 621 155 621 155 621 160 62I I60 621 160 621 160 621 160 621 165 62I 165 621 165 621 165 621 165 62I 170 62I I70 621 170 621 170 621 170 62I 175 621 175 621 175 621 175 621 175 621 180 62I 180 621 180 621 180 621 180 621 180 873 I80 852 180 83 1 180 810 180 770 185 892 185 872 185 852 185 833 185 795 190 914 190 895 190 876 190 858 190 824 195 937 195 919 195 902 195 886 195 855 200 963 200 947 200 931 200 916 200 889 205 992 205 977 205 963 205 950 205 928 210 1023 210 101 1 210 999 210 988 210 970 215 1058 215 1048 215 1038 215 1030 215 1017 220 1097 220 1089 220 1082 220 1076 220 1070 225 1140 225 1 I34 225 1 I30 225 1 I27 230 1187 230 1184 230 1183 235 1239 240 1297 245 1361 250 I43 I 255 1509 260 1595

30 TABLE 14 - (Continued) 34 EFPY Cooldown Curve Data Points Using 1996 App. G & ASME Code Case N-641 (without Uncertainties for Instrumentation Errors)

. Steady State 2O"Flhr. 4O"Fhr. 60"Fhr. 100"Fhr.

T(OQ p(psig) TPF) P(psig) T(OF) P(psig) T(OQ P(psig) T(OQ P(psig) 265 1690 270 1795 275 1911 280 2040 285 2182 290 2338

31 MATERIAL PROPERTY BASIS LIMITING ART VALUES AT 53 EFPY: 1/4T,243"F (LOWER SHELL LONG.WELDS) 3/4T,224"F (INTER. SHELL LONG. WELDS) 2500 2250 2000 1750 CI 2

1500 2

3 v) v)

2 1250 n

D 0

-mJ c

-m 1000 0

750 500 .-

250 lnservice hydrostatlc test t e m p e r a t u r e (293 F) f o r t h e s e r v i c e p e r i o d up t o 5 3 EFPY 1-I I I I 0 I I I ~ ~ L I I ~ I 0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)

Figure 3 Point Beach Units 1 and 2 Reactor Coolant System Heatup Limitations (Heatup Rates of 60 & 100°F/hr) Applicable for the First 53 EFPY (Without Margins for Instrumentation Errors) Using 1996App.G Methodology & ASME Code Case N-641

32 MATERIAL PROPERTY BASIS LIMITING ART VALUES AT 53 EFPY: 1/4T,243"F (LOWER SHELL LONG. WELDS) 3/4T,224"F (INTER. SHELL LONG. WELDS) 2500 2250 Unacceptable 1750 c--

0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)

Figure 4 Point Beach Units 1 and 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 100°F/hr) Applicable for the First 53 EFPY (Without Margins for Instrumentation Errors) Using 1996 A p p G Methodology & ASME Code Case N-641

33 TABLE 15 53 EFPY Heatup Curve Data Points Using 1996 App. G & ASME Code Case N-64 1 (without Uncertainties for Instrumentation Errors) 60°F/hr. Critica y Limit 100°F/hr. Criticality Limit Leak Test Limit Temp. Press. Temp. Press. Temp. Prcss. Temp. Press. Ternp. Press.

(OF) (psig) (OF) (PW (OF) Wg) (OF) 7 (PSk) (OF) 60 0 293 0 60 0 293 0 274 60 62 1 293 62 I 60 62 1 293 62 1 293 65 62 1 293 62 1 65 62 1 293 62 1 70 62 1 293 62 1 70 62 1 293 62 1 75 62 1 293 62 1 75 62 1 293 62 1 80 62 1 293 62 I 80 62 1 293 62 1 85 62 I 293 62 I 85 62 I 293 62 1 90 62 1 293 62 1 90 62 1 293 62 1 95 62 1 293 62 1 95 62 1 293 62 1 100 62 1 293 62 1 100 62 1 293 62 I 105 62 1 293 62 1 105 62 1 293 62 I 1 IO 62 I 293 62 I 110 62 1 293 62 1 1 I5 62 I 293 62 1 1I5 62 1 293 62 1 120 62 1 293 62 1 120 62 1 293 62 1 125 62 1 293 62 1 125 62 I 293 62 1 130 62 1 293 62 1 130 62 I 293 62 1 135 62 1 293 62 1 135 62 1 293 62 1 140 62 1 293 62 1 140 62 1 293 62 I 145 62 1 293 62 1 145 62 I 293 62 1 150 62 1 293 62 1 150 62 I 293 62 1 155 62 1 293 62 1 155 62 1 293 62 1 160 62 1 293 62 1 160 62 1 293 62 1 165 62 1 2 93 62 I 165 62 1 293 62 I 170 62 1 293 62 1 170 62 1 293 62 1 175 62 1 293 772 175 62 1 293 698 180 62 I 293 785 180 62 1 293 708 180 772 293 800 180 698 293 720 185 785 293 816 185 708 293 732 190 800 293 834 190 720 293 746 195 816 293 854 195 732 293 761 200 834 293 876 200 746 293 778 205 854 293 900 205 761 293 797 210 876 293 926 210 778 293 817 215 900 293 956 215 797 293 840 220 926 293 989 220 817 293 866 225 956 293 1025 225 840 293 893 230 989 293 1065 230 866 293 924 235 1025 293 1109 235 893 293 959 240 1065 293 I I57 240 924 293 996 245 1109 295 121 1 245 959 295 1038 250 1157 300 1270 250 996 300 1084 255 1211 305 1336 255 1038 305 1135

34 TABLE 15 - (Continued) 53 EFPY Heatup Curve Data Points Using 1996 App. G & ASME Code Case N-641 (without Uncertainties for Instrumentation Errors) 6OOFhr.

Temp.

265 1 Press.

1336 Criticality Limit Temp.

315 I Press.

1477 Temp.

265 100 Vh r.

Press.

(PW 1084 1135 Criticalitv Limit Temp.

315 I Press.

1253

+

Leak Test Limit Temp. Press.

270 1408 320 1548 270 1191 320 1321 275 1477 325 1628 275 1253 325 1397 280 1548 330 1715 280 1321 330 1480 285 1628 335 1812 ~ 285 1397 33s 1572 290 1715 340 1919 290 1480 340 1673 295 1812 345 2037 295 1572 345 1785 300 1919 350 2167 300 1673 350 1909 305 2037 355 2310 305 1785 355 2045 310 2167 360 2469 ~ 310 1909 360 2195 315 2310 1 315 2045 365 2361 320 2469 320 2195

~ 325 2361

35 TABLE 16 53 EFPY Cooldown Curve Data Points Using 1996 App. G & ASME Code Case N-64 I (without Uncertainties for Instrumentation Errors)

--Steac State 2OC h r. 40" h r. 60' 'hr. 100 lh r.

33 - p (PW T (OF) -

p (Pig) ZXlL p (Psi!) -m -

T (OF) p ( P W p (psigl 60 0 60 0 60 0 60 0 60 0 60 62 1 60 62 1 60 62 1 60 615 60 555 65 62 1 65 62 1 65 62 1 65 616 65 556 70 62 1 70 62 1 70 62 1 70 617 70 557 75 62 1 75 62 1 75 62 1 75 619 75 559 80 62 I 80 62 1 80 62 1 80 620 80 560 85 62 I 85 62 1 85 62 1 85 62 I 85 562 90 62 1 90 62 1 90 62 1 90 62 1 90 564 95 62 I 95 62 1 95 62 I 95 62 1 95 566 100 62 1 100 62 1 IO0 62 I 100 62 1 IO0 568 105 62 1 105 62 1 I05 62 1 105 62 1 105 571 1 IO 62 1 110 62 1 1 IO 62 1 110 62 1 110 5 74 1 I5 62 1 1 I5 62 1 115 62 I 115 62 1 I15 578 120 62 I 120 62 1 120 62 1 120 62 I 120 5 82 125 62 1 125 62 I 125 62 1 125 62 1 125 586 130 62 I 130 62 I 130 62 1 130 62 1 130 591 135 62 I 135 62 1 135 62 1 135 62 1 135 597 140 62 1 140 62 1 140 62 1 140 62 1 140 603 145 62 1 145 62 1 145 62 1 145 62 1 145 610 150 62 1 150 62 1 I50 62 1 150 62 1 150 618 155 62 1 155 62 1 155 62 1 155 62 1 155 62 1 160 62 1 160 62 1 160 62 1 160 62 1 160 62 I 165 62 1 I65 62 1 165 62 1 165 62 1 165 62 I 170 62 1 170 62 1 170 62 1 170 62 1 170 62 I 175 62 1 175 62 1 175 62 1 175 62 1 175 62 1 180 62 1 180 62 1 180 62 1 180 62 1 180 62 1 180 813 180 788 180 763 180 738 180 689 185 826 185 802 185 778 I85 753 185 706 190 840 190 817 190 793 I90 770 190 724 195 856 195 833 195 81 1 I95 789 195 745 200 873 200 852 200 830 200 809 200 768 205 892 205 872 205 852 205 832 205 794 210 914 210 894 210 875 210 857 210 822 215 937 215 919 215 902 215 885 215 853 220 963 220 946 220 931 220 915 220 888 225 992 225 977 225 963 225 950 22 5 927 230 1023 230 I010 230 998 230 987 230 969 235 I058 235 1047 235 1038 235 1029 235 1017 24 0 1097 240 1088 240 1081 240 1075 240 1069 245 1140 245 1134 245 1129 245 1 I27 250 I I87 250 1184 250 1 I83

TABLE 16 -(Continued) 53 EFPY Cooldown Curve Data Points Using 1996 App. G & ASME Code Case N-641 (without Uncertainties for Instrumentation Errors)

Steac State 20° 100' T (OF) 255

-To p (PSW 1239 T("F) P(psig) T (OF) 260 1297 265 1361 270 1431 275 1509 280 1595 285 1690 290 1795 295 191I 300 2040 305 2182 310 2338

37 6 REFERENCES

1. Regulatory Guide 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials, U.S.

Nuclear Regulatory Commission, May 1988.

2. WCAP-1404O-NP-A, Revision 2, Methodology used to Develop Cold Overpressure Mitigating system Setpoints and RCS Heatup and Cooldown Limit Curves, J.D. Andrachek, et. al., January 1996.

3. ASME Code Case N-64 1, Alternative Pressure-Temperature Relationship and Low Temperature Overpressure Protection System RequirementsSection XI, Division l, January 17,2000.

[Sub Refirence I: ASME Code Case N-640, AlternativeRejrence Fracture Toughnessfor Development of P-T Limit Curvesfor Section XI, Division I , Februav 26, 1999.1

[Sub Reference 2: ASME Boiler and Pressure Vessel Code, Case N-588.Attenuationto Reference Flaw Orientation of Appendrx Gfor Circumferential WeIds in Reactor Vessels,Section XI, Division I ,

Approved December 12, 1997.1

4.Section XI of the ASME Boiler and Pressure Vessel Code, Appendix ci Fracture Toughness Criteria for Protection Against Failure. Dated December 1995, through 1996 Addendum.

5. Code of Federal Regulations, 10 CFR Part 50,Appendix G, Fracture Toughness Requirements, U.S. Nuclear Regulatory Commission, Washington, D.C., Federal Register, Volume 60, No. 243, dated December 19,1995.

6. Fracture Toughness Requirements, Branch Technical Position MTEB 5-2, Chapter 5.3.2 in Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, LWR Edition, NUREG-0800, 1981.

7. WCAP-15 121, Revision 1, Point Beach Units 1 and 2 WOG Reactor Vessel 60-Year Evaluation Minigroup Heatup and Cooldown Limit Curves for Normal Operation, J.H. Ledger, April 2001.

8. Letter NPL 2002-0300, Design Input letter from Brian Kemp of MNC, Dated December 16,2002.

9. 1989 Section 111, Division 1 of the ASME Boiler and Pressure Vessel Code, Paragraph NB-233 1, Material for Vessels.

10. Regulatory Guide 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence, March 2001.

A-1 APPENDIX A MASTER CURVE EVALUATION ON PT. BEACH UNIT 2 GIRTH WELD

A-2 EVALUATION Master Curve fracture toughness testing has been performed on the Point Beach Nuclear Plant Unit 2 (PBNP-2) girth weld metal [A-I]. The results from this testing on weld wire heat 72442 are being incorporated into the determination of heatup and cooldown pressure-temperature (P-T)curves for the PBNP-2 reactor pressure vessel (RPV). The approach taken using Master Curve data utilizes the projected values of RTT, at end-of-life (EOL)and EOL extension (EOLE) maximum fluence levels for the girth weld to determine the value of adjusted reference temperature (ART). ART is the regulatory index value for use of the ASME Code reference fracture toughness curves. The methodology used for determining ART follows the approach that was the basis for the Kewaunee safety evaluation (SE) for the limiting girth weld in the Kewaunee RPV [A-21.

The calculation of heat-up and cool-down curves requires ART values at the %-thickness (%T) and %-

thickness (%T) through wall locations corresponding to the peak fluence for the girth (circumferential) weld. The attenuation of fluence (Ut) through the wall of the RPV was determined using the method in Regulatory Guide 1.99, Rev. 2 and the optional method in A S T M E 900-02 [A-31, which relies on an exponential decrease through the RPV wall:

where x is the distance into the vessel wall from the inside surface at the clad-base metal interface in units of inches. This attenuation formula for fluence is based upon dpa rather than neutron fluence with E > 1 MeV. Equation A 1 is applied for determining the fluences at %-thickness and %-thickness in the 6.5-in PBNP-2 RPV [A-4]. The ART values are then calculated using the Method 2 approach presented in Reference A-1, which follows the NRC evaluation method for the Kewuanee RPV [A-21. Tables A-I and A-2 list the results of these calculations for the %-thickness location at EOL and EOLE,and Tables A-3 and A 4 list the %-thickness location results for EOL and EOLE. The three lines in each of the tables are the projected calculations corresponding to the measured values of RTT, from the specific capsules.

Chemistry, temperature, and fluence differences between the capsule results are adjusted to reflect the RPV weld. The average of the three calculations at the end of each table is the ART value used for the P-T curve determination representing the girth weld metal in the PBNP-2 RPV.

A-3 Table A-I NRC Method [A-21 for DeterminingART at I/-T for EOL Fluence IChemistry and1 Temperature Fluence Total Material & Adjusted Adjusted Function Adjusted Capsule RTirr ( O F )

Ratio* ARTTOeF)

I I

\VF-67/L 1 I 2 10.7 149.8 I 1.132 238.9 169.7 199.9 133.7 60.5 260.4 194.2 WF-67nG2 I 170-3 I 179.9 143.9 204.4 Average

• Ratio of the fluence functions between the RPV at EOL and the surveillance capsule.

I 220

++ The margin established by NRC in the Kewaunee SE (Margin = 2 [(14)2 + (28)2]'R- 2)

Table A-2 NRC Method [A-2] for Determining ART at '/-T for EOLE Fluence Cbemistry and Temperature Fluence Total Material & Adjusted Adjusted Function Adjusted Capsule RTirr (OF)

Ratio* ~ T TCF) o ARTTO eF)

SA-1484lA3 210.7 I 1.245 WF-67L1 7 149*8 I 1.244 186.4 I I5Oa4 60s I 210*9 WF-67LG2 1.161 60-5 Average I 222.2

• Ratio of the fluence functions between the RPV at EOLE and the surveillance capsule.

++ The margin established by NRC in the Kewaunee SE (Margin = 2 [(14)2 + (28)2]'R- 2)

Table A-3 NRC Method [A-21 for Determining ART at %T for EOL Fluence Ihcmistry and Tempe rat ure Fluence Total Material & Adjusted Margin**

Adjusted Function Adjusted Capsule RTirr CF)

ARTTO@)

(OF)

Ratio*

SA-14841A3 210.7 0.934 196.8 218.3

~

WF-67LI 149.8 0.933 139.8 I 60s 164.3 WF-67LG2 170.3 0.870 148.2 I 6os Average 172.7 185 Ratio of the fluence functions between the RPV at EOL and the surveillance capsule.

• + The margin established by NRC in the Kewaunee SE (Margin = 2 [(14)2 + (28)2]tR- 2)

Table A-4 NRC Method [A-21 for DeterminingART at %-Tfor EOLE Fluence I 1Chemistr-y anc Fluence Total Function Adjusted Ratio* ~ T TeF) D I SA-1484lA3 1 210.7 1.060 223.4 184.4 60.5 I 244*9 149.8 1.059 158.7 122.7 183.2 170.3 0.988 168.3 132.3 192.8 Average I 207

• Ratio of the fluence functions behveen the RPV at EOLE and the surveillance capsule.

• • The margin established by NRC in the Kewaunee SE (Margin = 2 [(14)2 + (28)]R - 2)

A-5 APPENDIX A: REFERENCES A- 1 W. L. Server and J. R. Pfefferle, Master Curve Fracture ToughnessApplication For Point Beach Nirclear PIanf Unit 2, AT1 Consulting Report ATI-02 1-030-2003-1,January 2003.

A-2 Safev Evaluation by the Ofice of Nuclear Reactor Including the Use of a Master Curve-based Methodology f o r Reactor Pressure GsseI Integrity Assessment, Docket No. 50-305, May 2001.

A-3 ASTM Standard Test Method E 1921-02, Test Method for the Determination of Reference Temperature, To,for Ferritic Steels in the Transition Range, Annual Book ofASTMStandards, Vol. 03.01, American Society for Testing and Materials, West Conshohocken, PA.

A4 Instruction Manual, Reactor Vessel, Point Beach Nuclear Plant No. 2, Combustion Engineering, CE Book #4869, October 1970.

B-1 APPENDIX B DATA POINTS FOR CIRC-FLAW PT CURVES

[This Appendix contains PT curves that were developed using the Circ-Flaw Methodology from ASME Code Case N-588,which has been incorporated into Code Case N-641.There are two sets of Circ-Flaw curves within this Appendix. Curves in set one are those developed using the Circ-Flaw ART from PBNP Unit I. The second set of curves of those developed using the Circ-Flaw ART from the PBNP Unit 2 intermediate to lower shell circ. weld assuming the master-curve technology was not implemented. All ART values can be found in Table 12 of Page 22. Note that since neither set becomes limiting compared to the axial-flaw curves presented in the main body of this report, then only the data points will be presented in this Appendix for comparison purposes.]

TABLE B-1 34 EFPY Heatup Curve Data Points Using Circ. Flaw Methodology, ASME Code Case N-641 & Unit 1 Circ. Weld ART Values (without Uncertainties for Instrumentation Errors) 60 01r. Criticality Limit 1 OOOFhr. Criticality Limit Leak T t Limit Temp. Press. Temp. Press. Temp. Press. Temp. Press. Temp. Press.

(OF) 60 (PSk) 0 (OF) 229 (psig) 0 (0 - - -- -

60 (PW 0

(F) 229 (Psi!)

0 (F) 128 (PSk) 2000 60 62 1 229 62 1 60 62 1 229 62 1 229 2485 65 62 1 229 62 1 65 62 1 229 62 1 70 62 1 229 62 1 70 62 1 229 62 1 75 62 1 229 62 I 75 62 1 229 62 1 80 62 1 229 62 1 80 62 1 229 62 1 85 62 1 229 62 I 85 62 1 229 62 1 90 62 1 229 62 1 90 62 1 229 62 1 95 62 1 229 62 1 95 62 1 229 62 1 100 62 1 229 62 1 100 62 1 229 62 1 105 62 1 229 62 1 IO5 62 1 229 62 1 110 62 1 229 62 1 110 62 1 229 62 1 115 62 1 229 62 1 115 62 I 229 62 1 120 62 1 229 62 1 120 62 I 229 62 1 125 62 1 229 62 1 125 62 I 229 62 1 130 62 1 229 62 1 130 62 1 229 62 1 135 62 1 229 62 I 135 62 I 229 62 I 140 62 1 229 62 1 140 62 1 229 62 1 145 62 1 229 62 1 145 62 1 229 62 1 150 62 1 229 62 1 150 62 1 229 62 1 155 62 I 229 62 1 155 62 1 229 62 1 160 62 I 229 62 1 160 62 1 229 62 1 165 62 1 229 62 1 165 62 1 229 62 1 170 62 1 229 62 1 170 62 1 229 62 1 175 62 1 229 1517 175 62 1 229 1377 180 62 I 229 1540 180 62 1 229 1394 180 1517 230 1565 180 1377 230 1413 185 1540 235 1593 I85 I394 235 1434 190 1565 240 1623 190 1413 240 1458 195 1593 245 1657 195 1434 245 1484 200 1623 250 1695 200 1458 250 1512 205 1657 255 1736 205 1484 255 1544 210 1695 260 1782 210 1512 260 1580 215 1736 265 1833 215 1544 265 1619 220 1782 270 1874 220 1580 270 1662 225 I833 275 1919 225 1619 275 1710 230 1874 280 1969 230 1662 280 1763 235 1919 285 2025 235 1710 285 1822 240 I969 290 2086 240 1763 290 1886 245 2025 295 2153 24 5 1822 295 1958 250 2086 300 2228 250 1886 300 2036

8-3 TABLE B-I - (Continued) 34 EFPY Heatup Curve Data Points Using Circ. Flaw Methodology, ASME Code Case N-64 1 & Unit 1 Circ. Weld ART Values (without Uncertainties for Instrumentation Errors) 60 % r. Criticality Limit 1OO°F/hr. Criticality Limit Leak Test Limit Temp. Press. Temp. Press. Temp. Press. Temp. Press. Temp. Press.

(OF) (PSiP) (OF) (PW (0 (psi!) (F) (PSW (OF) (PW 255 2153 305 2310 255 1958 305 2124 260 2228 310 240 1 260 2036 310 2220 265 2310 265 2 124 315 2326 270 2401 270 2220 320 2443 215 2326 280 2443

~~~

B-4 TABLE B-2 34 EFPY Cooldown Curve Data Points Using Circ. Flaw Methodology, ASME Code Case N-64 1 & Unit 1 Circ. Weld ART Values (without Uncertainties for Instrumentation Errors)

~ _ _ ~ _ ~

Steac State 20" h r. 40' 'hr. 60' 'hr. 100 Ihr.

T (OF) 60 P( P W 0

T ('0 p (

60 0 PW 60 p (Pig:

0 T ("0 p ( P W 60 0 60 3xL p ( P a 9 0

60 621 60 621 60 621 60 621 60 621 65 621 65 621 65 621 65 621 65 621 70 621 70 621 70 621 70 621 70 621 75 621 75 621 75 621 75 621 75 621 80 621 80 621 80 621 80 621 80 621 85 621 85 621 85 621 85 621 85 621 90 621 90 621 90 621 90 621 90 621 95 621 95 621 95 621 95 621 95 621 100 621 100 621 1 00 621 100 621 100 621 105 621 105 621 105 621 105 621 105 621 110 621 110 621 110 62I 110 621 110 621 115 621 115 621 115 621 115 621 1 15 621 120 621 120 621 120 62I 120 621 120 62I 125 621 125 621 125 621 125 621 125 621 130 621 130 621 130 621 130 621 130 62I 135 621 135 621 135 621 135 621 135 621 140 621 140 621 140 621 140 621 140 621 145 621 145 621 145 621 145 621 145 621 150 621 150 62I 150 621 150 621 150 62I 155 621 155 621 155 621 155 621 155 62I 160 621 160 621 160 621 160 621 160 621 165 621 165 62I 165 621 165 621 165 621 170 621 170 621 170 621 170 621 170 621 175 621 175 621 175 621 I75 621 175 62I 180 621 180 621 180 621 180 621 180 621 180 1602 180 1544 180 1486 180 1427 180 1309 185 1619 185 1562 185 1504 185 1446 185 1330 190 1637 190 1581 190 1524 190 1468 190 1354 195 1657 195 1602 195 1547 195 1491 195 1380 200 1680 200 1626 200 1571 200 1517 200 1409 205 1705 205 1652 205 1599 205 1546 205 1442 210 1732 210 1681 210 1629 210 1578 210 1478 215 1763 215 1713 215 1663 215 1614 215 1518 220 1796 220 1748 220 1701 220 1654 220 1562 225 1833 225 1787 225 1742 225 1697 225 161 1 230 I874 230 1830 230 1788 230 1746 230 1666 235 1919 235 1878 235 1838 235 1800 235 1727 240 1969 240 193 1 240 1894 240 1859 240 1794 245 2025 245 1990 245 1956 245 1925 245 1868 250 2086 250 2054 250 2025 250 1998 250 1951 255 2153 255 2126 255 2101 255 2079 255 2043 260 2228 260 2205 260 2185 260 2168 260 2144 265 2310 265 2293 265 2278 265 2267 265 2256 270 2401 270 2390 270 2381 270 2377 270 2371

B-5 TABLE B-3 53 EFPY Heatup Curve Data Points Using Circ. Flaw Methodology, ASME Code Case N-641 & Unit 1 Circ. Weld ART Values (without Uncertainties for Instrumentation Errors) 60° hr. Critical y Limit 100 /hr. Critical v Limit Leak Test Limit Temp. Press. Temp. Press. Temp. Press. Temp. Press. Temp. Press.

(OF) 60 (PSk) 0 (OF) 248 (PW 0

(OF) 60 7

(PSiI3) 0 (F) 248 (psig) 0 (OF) 147 (PSk!)

2000 60 62 1 248 62 1 60 62 1 248 62 1 248 2485 65 62 1 248 62 1 65 62 1 248 62 1 70 62 1 248 62 1 70 62 I 24 8 62 1 75 62 1 248 62 1 75 62 1 248 62 1 80 62 1 248 62 1 80 62 1 248 62 1 85 62 1 248 62 1 85 62 1 248 62 1 90 62 1 248 62 1 90 62 1 248 62 1 95 62 1 248 62 1 95 62 1 248 62 1 I00 62 1 248 62 1 100 62 1 248 62 1 105 62 1 248 62 1 105 62 1 248 62 1 110 62 1 248 62 1 110 62 1 248 62 1 I15 62 1 248 62 1 115 62 1 248 62 1 120 62 1 248 62 1 120 62 1 248 62 1 125 62 1 248 62 1 125 62 1 248 62 1 130 62 1 248 62 1 130 62 1 248 62 1 135 62 1 248 62 I 135 62 1 248 62 1 140 62 1 248 62 1 140 62 1 248 62 1 145 62 1 248 62 1 145 62 1 24 8 62 1 150 62 1 248 62 1 150 62 1 24 8 62 1 155 62 1 248 62 1 155 62 1 24 8 62 1 160 62 1 248 62 1 160 62 1 24 8 62 1 165 62 1 24 8 62 1 165 62 1 24 8 62 1 170 62 1 248 62 1 170 62 1 24 8 62 1 175 62 1 24 8 1437 175 62 1 24 8 1313 180 62 1 248 1452 180 62 1 24 8 I324 180 1437 248 1468 180 1313 24 8 1336 185 1452 248 1485 185 1324 24 8 1349 190 1468 248 1505 190 1336 24 8 1363 195 1485 248 1527 195 1349 24 8 1380 200 1505 250 1550 200 1363 250 1398 205 1527 255 1577 205 1380 255 1418 210 1550 260 1606 210 1398 260 1441 215 1577 265 1639 215 1418 265 1465 220 1606 270 1674 220 1441 270 1493 225 1639 275 1714 225 1465 275 1523 230 1674 280 1758 230 1493 280 1557 235 1714 285 1806 235 1523 285 1594 240 1758 290 1860 240 1557 290 1636 245 1806 295 1919 245 1594 295 1681 250 1860 3 00 1980 250 1636 300 1732

E6 TABLE B (Continued) 53 EFPY Heatup Curve Data Points Using Circ. Flaw Methodology, ASME Code Case N-641 & Unit 1 Circ. Weld ART Values (without Uncertainties for InstrumentationErrors)

I 60"F/hr. Criticality Limit 100"Fhr. Criticalitv Limit Leak TI Temp. Press. Temp. Press. Temp.

("F) 255 (PSk) 1919 (OF) 305 (PSk) 1787

- (OF) 260 1980 310 2099 260 1732 310 1849 265 2036 315 2168 265 1787 315 1917 270 2099 320 2244 270 1849 320 1992 275 2168 325 2328 275 1917 325 2075 280 2244 330 2421 280 I992 330 2167 285 2328 285 2075 335 2268 290 2421 290 2167 340 2380 295 2268 300 2380

E7 TABLE B-4 53 EFPY Cooldown Curve Data Points Using Circ. Flaw Methodology, ASME Code Case N-641 & Unit 1 Circ. Weld ART Values (without Uncertainties for Instrumentation Errors)

---Stead State 20" hr.

- 40" h r.

- 60" h r. 100' /hr.

LEL 60 p (psig) 0 LEL 60 fJ (PW 0

T (OF) 60 p(PW 0

r60n -

P(PW 0

T (OF) 60 p( P W 0

60 621 60 621 60 621 60 621 60 621 65 621 65 621 65 62I 65 621 65 621 70 621 70 621 70 621 70 621 70 621 75 62I 75 621 75 621 75 621 75 621 80 621 80 621 80 621 80 621 80 621 85 621 85 621 85 621 85 621 85 621 90 621 90 62I 90 621 90 621 90 621 95 621 95 621 95 621 95 621 95 62I 100 621 100 621 100 621 I00 621 100 621 105 621 I05 621 105 621 105 62I 105 621 110 621 110 621 110 621 1 IO 621 110 62I 115 621 115 621 115 621 115 621 I15 621 120 621 120 621 120 621 I20 621 120 62I 125 621 125 621 125 621 125 621 125 621 130 621 130 621 130 62I 130 621 130 62I 135 621 135 621 135 621 135 621 135 621 140 621 140 621 140 62I 140 621 140 621 145 621 145 621 145 621 145 621 145 621 150 621 150 621 150 621 150 621 150 621 I55 62I 155 62I 155 621 155 621 155 621 I60 621 160 621 160 621 160 621 160 621 165 621 165 621 I65 621 165 621 I65 621 170 621 I70 62I 170 621 I70 621 170 621 175 621 175 621 175 621 I75 621 175 621 180 621 180 621 180 621 180 621 I80 621 I80 1552 180 1491 180 1430 180 1368 180 1242 185 1563 185 1503 185 1442 185 1380 I85 1256 I90 I576 190 1516 190 1456 190 1395 190 1271 I95 1590 I95 1531 195 1471 195 1410 195 1289 200 1605 200 1547 200 1487 200 1428 200 1308 205 I622 205 1564 205 1506 205 1448 205 1330 210 1641 210 1584 210 1527 210 1469 210 1354 215 1662 215 1606 215 1550 215 1493 215 1381 220 1685 220 1630 220 1575 220 1520 220 I41 1 225 1710 225 1657 225 1603 225 1550 225 1444 230 1738 230 1686 230 1634 230 1583 230 1481 235 1769 235 1719 235 1669 235 1619 235 1522 240 1803 240 1755 240 1707 240 1660 240 1568 245 1841 245 1795 245 1749 245 1704 245 1618 250 1883 250 1839 250 1796 250 1754 250 1674 255 1929 255 1888 255 1848 255 1809 255 1736 260 1980 260 1942 260 1905 260 1870 260 1805 265 2036 265 2002 265 1968 265 1937 265 1882

TABLE B (Continued) 53 EFPY Cooldown Curve Data Points Using Circ. Flaw Methodology, ASME Code Case N-64 I & Unit 1 Circ. Weld ART Values (without Uncertainties for Instrumentation Errors) 7 Steady Slate 2O"Fhr. 40"Fhr. ' 60"Fhr. 100"Fhr.

T(T) P(psig) T ( T ) P(psig) T("F) P(psig) T("F) P(psig) T ( O F ) P(psig) 270 2099 270 2068 270 2039 270 2012 270 i966 275 2168 275 2141 275 2116 275 2095 275 2060 280 2244 280 2222 280 2202 280 2186 280 2164 285 2328 285 2311 285 2297 285 2287 285 2279 290 2421 290 2410 290 2402 290 2399 290 2399

B-9 TABLE B-5 34 EFPY Heatup Curve Data Points Using Circ. Flaw Methodology, ASME Code Case N-641& Unit 2 Circ Weld ART Values (Le. No Master Curve Technology)

(without Uncertainties for Instrumentation Errors) 60' hr. Criticality Limit 100 Vh r. Criticality Limit Leak Test Limit

~~

Temp. Press. Temp. Press. Temp. Press. Temp. Press. Temp. Press.

(OF) (PSk) (OF) (psig) (OF) (psig) (OF) (PSk) (OF) (Psi&

60 0 245 0 60 0 245 0 141 2000 60 62 1 242 62 1 60 62 I 242 62 1 242 2485 65 62 I 242 62 I 65 62 1 242 62 1 70 62 1 242 621 70 621 242 62 1 75 62 1 242 62 1 75 62 1 242 62 I 80 62 1 242 62 1 80 62 1 242 62 1 85 62 1 242 62 1 85 62 1 242 62 1 90 62 1 242 621 90 62 1 242 62 1 95 62 1 242 62 1 95 62 1 242 62 I 100 62 1 242 62 1 100 62 1 242 62 I 105 62 1 242 62 1 105 62 1 242 62 1 1 IO 62 1 242 62 1 1IO 62 1 242 62 I 115 62 1 242 62 1 115 62 1 242 62 1 I20 62 1 242 621 120 62 1 242 62 I 125 62 1 242 62 I 125 62 1 242 62 I 130 62 1 242 62 1 I30 62 1 242 62 I 135 62 1 242 62 1 135 62 1 242 62 I 140 62 1 242 62 1 I40 62 1 242 62 1 145 62 1 242 62 1 145 62 1 242 62 I 150 62 1 242 62 1 150 62 1 242 62 1 155 62 1 242 62 1 155 62 1 242 62 1 160 62 1 242 62 1 160 62 1 242 62 1 165 62 1 242 62 1 165 62 1 242 62 1 170 62 1 242 62 1 170 62 1 242 621 175 62 1 242 1477 175 62 1 242 1345 180 621 242 1495 180 62 1 242 1358 180 1477 242 1515 180 1345 242 1374 185 1495 242 1538 185 1358 242 1391 190 1515 242 1563 190 1374 242 1410 195 1538 245 1591 195 1391 24 5 1431 200 1563 250 1621 200 1410 250 1454 205 1591 255 1655 205 1431 255 1480 210 1621 260 1692 210 1454 260 1509 215 I655 265 1734 215 1480 265 1541 220 1692 270 1776 220 1509 270 1576 225 1734 275 1810 225 1541 275 1615 23 0 1776 280 1849 230 I576 280 1658 235 1810 285 1892 235 1615 285 1706 240 1849 290 1939 240 1658 290 1759 24 5 1892 295 1991 245 1706 295 1817 250 1939 300 2048 250 1759 300 1881

B-10 TABLE B-5 (Continued) 34 EFPY Heatup Curve Data Points Using Circ. Flaw Methodology, ASME Code Case N-641& Unit 2 Circ Weld ART Values (Le. No Master Curve Technology)

(without Uncertainties for Instrumentation Errors)

I 6O"Fhr. I CriticalitvLimit I 100"F/hr. I Criticality Limit I Leak Test Limit I 1 Temp.

(OF) 255 Press.

tPW 1991 Temp.

(OF) 305 Press.

(PW 2112 Temp.

(OF) 255 Press.

(psig) 1817 Temp.

("0 305 Press.

(PSk) 1952 Temp.

(OF)

Press.

(Pig) 260 2048 310 2182 260 1881 310 203 1 265 21 12 315 2260 265 1952 315 21 18 270 2182 320 2346 270 203 1 320 2214 275 2260 325 244 1 275 2118 325 2319 280 2346 280 2214 285 244 1 285 2319 I

290 2436

B-11 TABLE B-6 34 EFPY Cooldown Curve Data Points Using Circ. Flaw Methodology, ASME Code Case N-641 & Unit 2 Circ Weld ART Values (Le. No Master Curve Technology)

(without uncertainties for Instrumentation Errors)

- Stead Slate 20° h r. 40' IIr. 60' h r. 100' h r.

-T (OF) 60 633 0

-T (OF) 60

-XEL --

P (psig) 0 60 p (psig) 0 T (OF) 60

-T p[ P W 0

(OF) 60 p (PSilz) 0 60 62 1 60 62 1 60 62 1 60 62 1 60 62 1 65 62 1 65 62 1 65 62 1 65 62 1 65 62 1 70 62 1 70 62 1 70 62 1 70 62 1 70 62 1 75 62 1 75 62 1 75 62 1 75 62 1 75 62 1 80 62 1 80 62 1 80 62 1 80 62 1 80 62 1 85 62 1 85 62 1 85 62 1 85 62 1 85 62 1 90 62 I 90 62 1 90 62 1 90 62 1 90 62 1 95 62 1 95 62 1 95 62 1 95 62 1 95 62 1 100 62 1 100 62 1 100 62 1 IO0 62 1 100 62 1 105 62 I 105 62 1 105 62 I I05 62 1 105 62 1 1 IO 62 1 110 62 1 1IO 62 1 I10 62 1 110 62 1 1 I5 62 1 115 62 1 115 62 1 115 62 1 115 62 1 120 62 1 120 62 1 120 62 1 I20 62 1 120 62 1 125 62 1 125 62 1 125 62 1 125 62 1 125 62 1 130 62 1 130 62 1 130 62 1 130 62 1 130 62 1 135 62 1 135 62 I 135 62 I 135 62 1 135 62 1 140 62 1 140 62 1 140 62 1 140 62 1 140 62 1 145 62 1 145 62 1 145 62 1 145 62 1 145 62 1 150 62 1 150 62 1 150 62 1 150 62 1 150 62 1 155 62 1 155 62 1 I55 62 1 155 62 I 155 62 1 160 62 1 160 62 1 160 62 1 160 62 1 160 62 1 165 62 1 165 62 1 165 62 I 165 62 1 165 62 1 170 62 1 170 62 1 170 62 1 170 62 1 170 62 1 175 62 1 175 62 1 175 62 I I75 62 1 175 62 1 180 62 1 180 62 1 180 62 1 180 62 1 180 62 1 180 1566 180 1506 I80 I445 180 1384 180 1260 185 1579 185 1519 185 1459 185 1399 185 1276 I90 1593 190 1534 190 1475 190 1415 190 1294 195 1609 195 1550 195 1492 195 1433 195 1314 200 1626 200 1568 200 1511 200 1453 200 1336 205 1645 205 1588 2 05 1532 205 1475 205 1361 210 1666 210 1611 210 1555 210 1499 210 1388 215 1690 215 1635 215 1581 215 1527 215 1419 220 1715 220 I662 220 1610 220 1557 220 1453 225 1744 225 1693 225 1641 225 1591 22 5 1490 230 1776 230 1726 230 1677 230 1628 230 1532 235 1810 235 1763 235 1715 235 1669 235 1578 240 1849 240 1803 240 1759 24 0 1715 24 0 1630 245 1892 245 1849 245 1806 24 5 1765 245 1687 250 1939 250 1898 250 1859 250 1821 250 1750 255 1991 255 1953 255 1918 255 I883 255 1821 260 2048 260 2014 260 I982 260 1952 260 1899

8-12 TABLE B (Continued) 34 EFPY Cooldown Curve Data Points Using Circ. Flaw Methodology, ASME Code Case N-641& Unit 2 Circ Weld ART Values (Le. No Master Curve Technology)

(without Uncertainties for Instrumentation Errors)

Steady State 2OOFhr. 40"FIhr. 60°F/hr. 1OO°F/hr.

T(OF) P(psig) T(OF) P(psig) T(OF) P(psig) T(OF) P(psig) T ( T ) P(psig) 265 2112 265 2082 265 2054 265 2028 265 1985 270 2182 270 2156 270 2133 270 2112 270 2081 275 2260 275 2239 275 222 1 275 2206 275 2187 280 2346 280 2330 280 2318 280 2309 280 2304 285 244 1 285 2431 285 2425 285 2423 285 2423

E13 TABLE B-7 53 EFPY Heatup Curve Data Points Using Circ. Flaw Methodology, ASME Code Case N-64I & Unit 2 Circ Weld ART Values (ie. No Master Curve Technology)

(without Uncertainties for Instrumentation Errors) 60" h r. Critica y Limit- 100 ?hT.

-- Critics y Limit Leak Test Limit Temp. Press. Temp. Press. Temp. Press. Temp. Press. Temp. Press.

("0 (PSiP) (OF) (PSk) (OF) (psig) (OF) (Psig) (OF) (PSk) 60 0 263 0 60 0 263 0 162 2000 60 62 1 263 62 1 60 62 1 263 62 I 263 2485 65 62 1 263 62 1 65 62 1 263 62 1 70 62 1 263 62 1 70 62 1 263 62 1 75 62 1 263 62 1 75 62 1 263 62 1 80 62 I 263 62 1 80 62 1 263 62 1 85 62 1 263 62 1 85 62 1 263 62 1 90 62 I 263 62 1 90 62 1 263 62 1 95 62 1 263 62 1 95 62 1 263 62 1 100 62 1 263 62 1 100 62 1 263 62 1 105 62 1 263 62 1 105 62 1 263 62 1 110 62 1 263 62 I 1 IO 62 1 263 62 1 1 I5 62 1 263 62 I 115 62 1 263 62 I 120 62 1 263 62 1 120 62 I 263 62 1 125 62 1 263 62 1 125 62 1 263 62 I 130 62 1 263 62 1 130 62 1 263 62 1 135 62 1 263 62 1 I35 62 1 263 62 1 140 62 1 263 62 1 140 62 1 263 62 1 145 62 1 263 62 1 I45 62 1 263 62 1 I50 62 1 263 62 I 150 62 1 263 62 1 I55 62 1 263 62 I 155 62 1 263 62 1 I60 62 1 263 62 1 160 621 263 62 1 165 62 I 263 62 1 165 62 1 263 62 1 170 62 I 263 62 1 170 62 1 263 62 1 175 62 1 263 1404 175 62 1 263 1287 180 62 1 2 63 1415 180 62 1 263 1295 180 1404 263 1428 180 1287 263 1304 185 1415 263 1441 185 1295 263 1313 190 1428 263 1456 190 1304 263 1324 195 1441 263 I472 195 1313 263 1337 200 1456 263 1491 200 1324 263 1351 205 1472 263 151 1 205 I337 263 1366 210 1491 263 1533 210 1351 263 1383 215 151 1 265 1558 215 1366 265 1402 220 1533 270 1586 220 1383 270 1423 225 1558 275 1616 225 1402 275 1446 230 1586 280 1650 230 1423 280 1472 235 1616 285 1687 235 1446 285 1500 240 1650 290 1728 240 I472 290 1532 245 1687 295 1773 245 1500 295 1567 250 1728 300 1824 250 1532 300 1605

E14 TABLE B (Continued) 53 EFPY Heatup Curve Data Points Using Circ. Flaw Methodology, ASME Code Case N-64 1 & Unit 2 Circ Weld ART Values (Le. No Master Curve Technology)

(without Uncertainties for Instrumentation Errors)

I

~~

6O"Fhr. Criticality Limit 1 0O"Fhr. Criticality Limit Leak Test Limit I

Tem D. Press. TemD. Press. Temp. Press. Temp. Press. Temp. Press.

-(an 255 (PSiP) 1773 (OF) 305 (psig) 1879 (OF) 255 (Psi@

1567 (OF) 305 (PSM 1648 T 6 - k 260 1824 310 1929 260 1605 310 1695 265 I879 315 I980 265 1648 315 1748 270 1929 320 2036 270 1695 320 1805 275 I980 325 2099 275 1748 325 1869 280 2036 330 2168 280 1805 330 1939 285 2099 335 2244 285 1869 335 2013 290 2168 340 2328 290 1939 340 2103 295 2244 345 242I 295 2017 345 2198 300 2328 300 2103 350 2303 3 05 3 05 2198 355 2418 3 10 2303 3 15 2418

B-15 TABLE B-8 53 EFPY Cooldown Curve Data Points Using Circ. Flaw Methodology, ASME Code Case N-641 & Unit 2 Circ Weld ART Values (Le. No Master Curve Technology)

(without Uncertainties for Instrumentation Errors)

Steac State 2oa h r. 40' h r. 60' br. 100' Yh r.

-- 60 0

- - 32 T ("0 p (psi&?) T (OF) 60 p( P W 0 60 p( P W 0

T ("0p (PSif9 60 0 60 T ("0p (Psi@

0 60 62 1 60 62 I 60 62 1 60 62 1 60 62 1 65 62 1 65 62 1 65 62 1 65 62 1 65 62 1 70 62 1 70 62 1 70 62 I 70 62 I 70 62 1 75 62 1 75 62 1 75 62 I 75 62 1 75 62 1 80 62 1 80 62 1 80 62 1 80 62 1 80 62 I 85 62 1 85 62 1 85 62 1 85 62 I 85 62 1 90 62 1 90 62 1 90 62 1 90 62 1 90 62 1 95 62 1 95 62 1 95 62 1 95 62 I 95 62 1 100 62 I 100 62 1 100 62 1 100 62 I 100 62 1 105 62 I 105 62 1 105 62 1 I05 62 I 105 62 1 110 62 1 110 62 1 110 62 1 110 62 1 I10 62 1 115 62 1 115 62 I 115 62 1 115 62 1 1 I5 62 1 120 62 1 120 62 1 120 62 1 120 62 1 120 62 1 125 62 1 125 62 1 125 62 1 I25 62 1 125 62 I 130 62 1 130 62 1 130 62 1 130 62 1 130 62 I 135 62 1 135 62 1 135 62 1 135 62 1 135 62 I I40 62 1 140 62 1 140 62 1 140 62 1 140 62 I I45 62 1 145 62 1 145 62 1 145 62 1 145 62 1 150 62 I I50 62 1 150 62 1 150 62 1 150 62 1 155 62 1 155 62 1 155 62 1 155 62 1 155 62 1 160 62 1 160 62 1 160 62 1 160 62 I 160 62 1 165 62 1 165 62 1 165 62 1 165 62 1 165 62 1 170 62 I 170 62 1 170 62 1 170 62 1 170 62 1 175 62 1 175 62 I 175 62 1 175 62 1 175 62 1 180 62 1 180 62 I 180 62 1 180 62 1 180 62 1 180 1524 180 1461 180 1398 180 1334 180 1204 185 1532 185 1470 185 1407 185 1343 185 1214 I90 1542 190 1480 190 1417 190 1354 190 1225 195 1552 195 1490 195 I428 195 1365 195 1238 200 1563 200 1502 200 1440 200 1378 200 1252 205 1576 205 1515 205 1454 205 1392 205 1267 210 1590 210 1530 210 1469 210 1408 210 1285 215 1605 215 1546 215 1486 215 1426 215 1304 220 1622 220 I564 220 1505 220 1445 220 1326 225 1641 225 1583 225 1525 225 1467 225 1350 230 1662 230 1605 230 1548 230 149 1 230 1377 235 1685 235 1629 235 1574 235 1518 235 1407 240 1710 240 1656 240 1602 240 1548 240 1441 245 1738 245 1685 245 1633 245 1581 24 5 1478 250 1769 250 1718 250 I667 250 1617 250 1519 255 1803 255 1754 255 I706 255 1658 255 1564 2 60 1841 260 1794 260 I748 260 1703 260 1615 265 1883 265 1838 265 1795 265 1752 265 1671

TABLE B (Continued) 53 EFPY Cooldown Curve Data Points Using Circ. Flaw Methodology, ASME Code Case N-641 & Unit 2 Circ Weld ART Values (Le. No Master Curve Technology)

(without Uncertainties for Instrumentation Errors)

Stead State 2OoF/hr. 4OoF/hr. 6OoF/hr. 1OO°F/h r.

T (OF) p (PSid T r F ) P(psig) T r F ) P(psig) T ( T ) P(psig) T(OF) 1 P(psig) 270 1929 270 1887 270 1847 270 1807 270 I 1734 275 1980 275 1941 275 I904 275 1868 275 1803 280 2036 280 2001 280 1967 280 1936 280 1879 285 2099 285 2067 285 2038 285 201 1 285 1964 290 2168 290 2140 290 21 15 290 2093 290 2059 295 2244 295 2221 295 2202 295 2185 295 2163 300 2328 300 2311 300 2297 300 2286 300 2278 305 2421 305 2410 305 2402 305 2399 305 2399