NL-11-0242, Proposed Alternative in Accordance with 10 CFR 50.55a(a)(3)(i), HNP-ISI-ALT-11

Proposed Alternative in Accordance with 10 CFR 50.55a(a)(3)(i), HNP-ISI-ALT-11

ML111380211
Person / Time
Site:	Hatch
Issue date:	05/17/2011
From:	Ajluni M Southern Nuclear Operating Co
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
HNP-ISI-ALT-11, NL-11-0242
Download: ML111380211 (49)

Text

Mark J. Ajluni, P.E. Southern Nuclear Nuclear Licensing Director Operating Company. Inc.

40 Inverness Center Parkway Post Office Box 1295 Birmingham, Alabama 35201 Tel 205.992.7673 Fax 205.992.7885 May 17, 2011 SOUTHERN'\

COMPANY Docket Nos.: 50-366 NL-11-0242 U. S. Nuclear Regulatory Commission ATIN: Document Control Desk Washington, D. C. 20555-0001 Edwin I. Hatch Nuclear Plant - Unit 2 Proposed Alternative in Accordance With 10 CFR 50.55a(a)(3)(i)

HNP-ISI-ALT-11 Ladies and Gentlemen:

Pursuant to 10 CFR 50.55a(a)(3)(i), Southern Nuclear Operating Company (SNC) hereby requests approval of an alternative to the requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code)

Section XI, 2001 Edition through 2003 Addenda for Edwin I. Hatch Nuclear Plant (HNP) Unit 2. Currently, examination of reactor pressure vessel (RPV) circumferential shell welds is required in Table IWB-2500-1, category B-A, Item B1.11.

SNC requests NRC approval of Alternative HNP-ISI-ALT -11, Version 1 to continue the elimination of the RPV circumferential shell weld examinations as previously allowed by relief request RR-38 and the NRC's safety evaluation (SE) dated January 28, 2005, from July 31, 2007 through the period of extended operation (PEO), which expires on June 13, 2038.

This alternative is requested because compliance with the specified requirements would result in hardship without a commensurate increase in the level of quality and safety. According to 10 CFR 50.55a(a)(3), alternatives to the requirements of paragraph 50.55a(a)(3) may be used, when authorized by the NRC, if an applicant demonstrates that the proposed alternatives would provide an acceptable level of quality and safety.

By letter dated March 29, 2004, as supplemented by letter dated September 13, 2004, SNC submitted a proposed alternative (RR-38) to the requirements of the 1989 Edition of Section XI of the ASIVIE Code for HNP Units 1 and 2. This alternative proposed the elimination of the circumferential shell weld examinations based on BWRVIP-05. During the review of RR-38, the NRC staff and SNC agreed that RR-38 should be amended to include the PEO (i.e.,

through its current 60-year renewed license).

U.S. Nuclear Regulatory Commission NL-11-0242 Page 2 The NRC's response was communicated in a January 28,2005 safety evaluation (ADAMS Accession Number ML050130317). The NRC concluded that SNC's proposed alternative provided an acceptable level of quality and safety and it was authorized pursuant to 10 CFR 50.55a (a)(3)(i). However, the NRC staff determined that the neutron fluence values used in the alternative were based on a neutron fluence code (RAMA) that had not been reviewed and approved by the NRC staff. The NRC concluded that the fluence values were conservative through July 31,2007. Therefore, the !\IRC staff limited the duration of RR-38 until July 31, 2007 pending SNC's use of an NRC approved fluence calculation methodology.

The NRC has addressed the use of RAMA fluence modeling methodology as defined in a May 13, 2005 letter (TAC NO. MB9765) from W.H. Bateman (NRC) to Bill Eaton (BWRVIP Chairman). The letter granted conditional approval for the use of RAMA as indicated in Enclosure 4.

After approval of RAMA, Alternative HNP-ISI-ALT-8 (which is similar to this proposed Unit 2 alternative) was submitted to the NRC for HNP-Unit 1. This alternative was approved by !\IRC letter dated December 6, 2007. (ADAMS Accession No. ML073130188). Approval for HNP-Unit 2 was not requested at that time because the HNP Unit 2 fluence evaluation had not been completed. A fluence evaluation for HNP Unit 2 using the currently approved RAMA code has been completed. The report indicated that there is a relatively small increase in projected fluence at the end of 60 years (50.1 Effective Full Power Years (EFPY)). Based upon the completion of the HNP Unit 2 fluence evaluation, SNC is requesting approval to continue the elimination of the HNP Unit 2 RPV circumferential shell weld examinations until the end of the period of extended operation; June 13,2038.

Additional supporting information is provided in the Enclosure 1.

SNC requests approval of this proposed alternative by May 1, 2012 to support the Spring 2013 examinations required during the 3rd period of the 4th interval.

This letter contains no NRC commitments. If you have any questions, please contact N. J. Stringfellow at 205-992-7037.

Respectfully submitted, M. J. Ajluni Nuclear Licensing Director MJAlPAH/lac

U.S. Nuclear Regulatory Commission NL-11-0242 Page 3

Enclosures:

1. Proposed Alternative in Accordance With 10 CFR 50.55a(a)(3)(i),

HNP-ISI-ALT-11

2. Southern Nuclear Relief Request RR-38

3. NRC Evaluation of RR-38 (TAC Nos. MC2381 and MC2382) dated January 28, 2005

4. NRC Safety Evaluation of Proprietary EPRI Reports (TAC No.

MB9765) dated May 13, 2005

References:

1. NL-04-1764 dated 09/13/1974, Third Ten-Year Inservice Inspection Program, Submittal of Revised relief request RR-38."

2. NRC Letter of 01/28/2005, "Evaluation of Relief Request (RR) Number 38 (TAC Nos. MC2381 and MC2382).

3. NRC Letter of May 13, 2005, [Safety Evaluation of proprietary EPRI Reports, "BWR Vessel and Internals Project, RAMA Fluence Methodology Benchmark manual (BWRVIP-114," "RAMA Fluence Methodology Benchmark Manual-Evaluation of Regulatory Guide 1.1990 Benchmark Problems (BWRVIP-115)," "RAMA Fluence Methodology-Susquehanna Unit 2Surveillance Capsule Fluence Evaluation For cycles 1-5 (BWRVIP-117), " AND "RAMA Fluence Methodology Procedures Manual (BWRVIP-121)," and "Hope Creek Flux Wire Dosimeter Activation Evaluation for Cycle 1 (TWE-PSE-001 R-001)" (TAC No. MB9765).

4. BWR-VI P-05, "Structural Integrity Calculation Package HTCH-@ !Q 302R1, "Hatch Revised material Data Review and ART Calculations."

5. Transware Enterprises Document SNC-FLU-002-R-001, Revision 0, "Edwin I. Hatch Unit 2 reactor Pressure Vessel Fluence Evaluation at the End of Cycle 18 and 50.1 EFPY" dated April 23, 2007.

6. Final SER of the BWR Vessel Internals Project BWRVIP-05 Report, July 28, 1998. (TAC No. MA93925).

7. Supplement to Final Safety Evaluation of the BWR Vessel and Internals Project BWRVIP-05, dated March 7, 2000. (Adams Accession No. ML031430372).

cc: Southern Nuclear Operating Company Mr. J. T. Gasser, Executive Vice President Mr. D. R. Madison, Vice President - Hatch Ms. P. M. Marino, Vice President - Engineering RTYPE: CHA02.004 U. S. Nuclear Regulatory Commission Mr. V. M. McRee, Regional Administrator Mr. P. G. Boyle, NRR Project Manager- Hatch Mr. E. D. Morris, Senior Resident Inspector - Hatch

Edwin I. Hatch Nuclear Plant-Unit 2 Proposed Alternative in Accordance With 10 CFR SO.SSa(a)(3)(i)

HNP-ISI-ALT-11 Enclosure 1 Proposed Alternative in Accordance With 10 CFR SO.SSa(a)(3)(i)

HNP-ISI-ALT-11

Enclosure 1 Edwin I. Hatch Nuclear Plant - Unit 2 Proposed Alternative in Accordance with 10 CFR 50.55a{a){3){i)

HNP-ISI-ALT-11

1. Requested Date For Approval and Basis Approval is requested by May 1, 2012 to support the Spring 2013 examinations required during the 3rd Period of the 4th Interval.

2. ASME Code Component(s) Affected Class 1 Category B-A Pressure retaining circumferential welds in the Reactor Pressure Vessel (RPV).

3. Applicable Code Edition and Addenda

ASME Section XI, 2001 Edition through the 2003 Addenda.

4. Applicable Code Requirements Examination of RPV circumferential shell welds as required in Table IWB-2500-1, Category B-A, Item B1.11.

5. Reason for Request

Southern Nuclear Operating Company (SNC) proposes to continue the elimination of the RPV circumferential shell weld examinations as previously allowed by Relief Request RR-38 (Enclosure 2) and the NRC's safety evaluation (SE) dated January 28, 2005 (Enclosure 3) from July 31, 2007 through the period of extended operation (PEO). The PEO is the period of the HNP Unit 2 renewed license, which expires June 13, 2038.

6. Backqround Per the NRC SE dated July 28, 1998 and Generic Letter 98-05, it was indicated that BWR licensees could request relief from the inservice inspection requirements of 10 CFR 50.55a(g) for volumetric examination of circumferential reactor pressure welds (ASME Section XI Code, Table IWB 2500-1, Examination Category B-A, Item 1.11, Circumferential Shell Welds) by demonstrating:

1. At the expiration of their license, the circumferential welds satisfy the limiting conditional failure probability for circumferential welds in this evaluation.

2. Licensees have implemented operator training and established procedures that limit the frequency of cold over pressure events to the amount specified in this report.

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Based on these two requirements, SNC:

Was previously granted approval for permanent deferral (during the initial 40-years of operation) of the HNP Unit 2 augmented examination requirements for the circumferential welds pursuant to 10 CFR 50.55a(g)(6)(ii)(A)(5).

Justified relief from the volumetric examination of the circumferential RPV welds during the PE~.

Subsequently, by letter dated March 29, 2004, as supplemented by letter dated September 13, 2004, SNC submitted a proposed alternative (RR-38) to the requirements of 1989 Edition of Section XI of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) for Edwin I. Hatch Nuclear Plant (HNP), Units 1 and 2. This alternative proposed the elimination of the circumferential shell weld examinations based on BWRVIP-05. During the review of RR-38, the NRC staff and SNC agreed that RR-38 should be amended to include the PEO (Le., through its current 60-year renewed license).

The NRC's response was communicated in a January 28, 2005 safety evaluation (ADAMS Accession Number ML050130317). The NRC concluded that SNC's proposed alternative provided an acceptable level of quality and safety and it was authorized pursuant to 10 CFR 50.55a (a)(3)(i).

However, the NRC staff determined that the neutron fluence values used in the alternative were based on a neutron fluence code that had not been reviewed and approved by the NRC staff. The NRC concluded that the fluence values were conservative through July 31, 2007. Therefore, the NRC staff limited the duration of RR-38 to July 31, 2007 pending SNC's use of an NRC approved fluence calculation methodology.

7. Proposed Alternative and Basis for Use Proposed Alternative SNC proposes an alternative to continue the elimination of the RPV circumferential shell weld examinations for HNP Unit 2 as previously allowed by Relief Request RR-38 and the NRC's safety evaluation for a period of time from July 31, 2007 through the PE~.

Basis for Use The NRC has now addressed the approval and use of RAMA fluence modeling methodology as defined in a May 13, 2005 letter (TAC NO.

MB9765) from W.H. Bateman (NRC) to Bill Eaton (BWRVIP Chairman). The letter granted conditional approval for the use of RAMA as indicated in Enclosure 4 of the May 13, 2005 letter. The NRC staff's conditions and SNC's response to the conditions are given below.

1. Condition - For plants that are similar in core, shroud and downcomer vessel geometry to that of the Susquehanna and Hope Creek plants, i.e.,

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BWR-IV plants, the RAMA methodology can be applied without a bias for the calculation of vessel neutron fluence.

Response - HNP - Unit 2 is a BWR-IV plant that is similar in core, shroud and downcomer-vessel geometry to that of the Susquehanna and Hope Creek plants (see Enclosure 4, Section 4.1). Therefore, the RAMA methodology can be applied without a bias for the calculation of vessel neutron 'fluence.

2. Condition - For plants (or plant groups) with a different geometry than that of the Susquehanna or Hope Creek plants, a plant-specific application for RPV neutron fluence is required to establish the value of a bias.

Response - Not applicable to HNP - Unit 2.

3. Condition - Relevant benchmarking will be required for shroud and reactor internals applications.

Response: Not applicable for this alternative.

A fluence model using the currently approved RAMA code was constructed and results reported in TransWare Enterprises Inc. report SNC-FLU-002-R 001 revision 0 dated 4/23/07. The report indicated that there is a relatively small increase in projected fluence at the end of 60 years (50.1 Effective Full Power Years (EFPY>>.

Based on this data and updated Adjusted Reference Temperature (ART) calculations performed by Structural Integrity Associates (SIA), the HNP Unit 2 circumferential welds and axial continue to satisfy the limiting conditional failure probability at the end of the PEO. (Note: HNP Unit 2 is currently in its 31 st year of commercial operation).

Circumferential Welds The projected fluence value calculated by RAMA for the limiting Circumferential weld at the end of the period of extended operation (PEO) at HNP Unit 2 is higher than that previously reported for HNP Unit 2 in of Relief Request 38 (RR-38). RR-38 used a GE developed fluence methodology, which was approved on a limited time basis (through July 2007). As shown in the table below, the fluence at the end of the PEO was previously stated in RR-38 as 0.244 x 10 19 n/cm 2 while the RAMA calculated fluence used for this alternative at the end of the PEO is 0.324 x 10 19 n/cm 2

• The end of the PEO for HNP Unit 2 in RR-38 was based on an assumed 90% capacity factor, or 54 EFPY (effective full power years of operation). For the RAMA calculations, a more detailed evaluation of the capacity factor was performed, which used actual power history thru cycle 18 and resulted in the end of the PEO being defined at 50.1 EFPY.

For comparison purposes, updated values for the limiting HNP Unit 2 circumferential shell weld at the end of the PEO and values specified for E1-3

Combustion Engineering (CE) fabricated RVs in Table 2.6-5 of the July 28, 1998, NRC SER are shown below in Table 1. Also included for comparison are the values used as the basis for RR-38.

As demonstrated in Table 1, the projected fluence using the RAMA methodology increased. However, the HNP Unit 2 RTNOT is still bounded by the mean RTNOT calculated using either CE(VIP) or CE(CEOG) chemistry data as provided in the 1998 NRC SER for BWRVIP-05.

Table 1 HNP Unit 2 HNP Unit 2 Limiting 64 EFPY Limiting 64 EFPY Limiting Circ. Limiting Circ.

CE-VIP Case CEOG Case Weld at License Weld at License Study Table 2.6-5 Study Table 2.6-5 Expiration (54 Expiration (50.1 of NRC SE for of NRC SE for EFPY) from RR EFPY) from SIA BWRVIP-05 BWRVIP-05 38 ART Calculation Cu% 0.13 0.183 0.047 0.047 Ni% 0.71 0.704 0.049 0.049 Chemistry 151.7 172.2 31.0 31.0 Factor (CF)

Fluence 0.40 0.40 0.244 0.324 (10 19 n/cm 2 )

Delta RT NOT 113.2 128.5 19.2 18.9 Initial RT NOT 0 0 -50.0 -50.0 Mean RTNDT 113.2 128.5 -30.8 -31.1 P(F/E) *4 ......... _.........

A

-4 4.38 X 10-4 ------

NRC P(F/E)

BWRVIP Axial Welds The projected fluence value calculated by RAMA for the limiting axial weld at the end of PEO at HNP Unit 2 higher than that previously reported for HNP Unit 2 in Enclosure 2 of RR-38. RR-38 used a GE developed fluence methodology, which was approved on a limited time basis (through July 2007). As shown in Table 2, the fluence at the end of the PEO was previously stated in RR-38 as 0.244 x 1019 n/cm 2 while the RAMA calculated fluence used for this alternative at the end of the PEO is 0.299 x 1019 n/cm 2

• The end of the PEO for HNP Unit 2 in RR-38 was based on an assumed 90% capacity factor, or 54 EFPY (effective full power years of operation). For the RAMA calculations, a more detailed evaluation of the capacity factor was performed, which used actual power history thru cycle 18 and resulted in the end of the PEO being defined at 50.1 EFPY.

Table 2 has been provided to compare the axial weld information provided in RR-38 versus the RAMA information. The NRC issued a revised SER on BWRVIP-05 on March 7, 2000 and it indicated that the limiting axial welds should be compared with data in Table 3 of that document (Mod 2 in Table E1-4

below). As demonstrated in Table 2, the RTNDT(U) remained unchanged and the Mean RTNOT decreased; therefore, the values continue to be bounded.

In addition, axial welds and intersecting portions of circumferential welds will be examined to the extent practical, dependent upon interference by another component or restrictions due to the geometrical configuration. For those cases where the reduction in coverage is greater than 10%, relief will be requested pursuant to 10 CFR 50.55a requirements.

Table 2 HNP Unit 2 Limiting HNP Unit 2 Limiting Axial Weld at Axial Weld at License Mod 2 License Expiration Expiration (54 EFPY)

(50.1 EFPY) from from RR-38 SIA ART Calculation Cu% ------ 0.216 0.216 Ni% ------ 0.043 0.043 Chemistry

---

98 98 Factor (CF)

Fluence ------ 0.244 0.299 (10 19 n/cm 2)

Delta RTNDT 116 60.6 56.3 Initial RTNOT -2.0 -50.0 -50.0 Mean RTNDT 114 10.6 6.3 P(F/E)

NRC 5.02 x 10-6 ------ ------

P(F/E)

___ w ___

BWRVIP Conclusion The HNP Unit 2 circumferential and axial welds continue to satisfy the limiting conditional failure probability at the end of the PEO and SNC has previously demonstrated that operator training and established procedures limit the frequency of cold over pressure events. Therefore, this alternative will continue to provide an acceptable level of quality and safety and approval is requested per 10 CFR 50.55a(a)(3)(i).

8. Precedents Southern Nuclear Relief Request RR-38 (Enclosure 2)

A similar request was approved for HNP Unit 1 on December 6, 2007.

(ADAMS Accession Number ML073130188) (Note: HNP Unit 2 relief was not requested at the same time as HNP Unit 1 due to the Fluence Evaluation not being completed for HNP Unit 2.)

9. References E1-5

BWR-VIP-05, Structural Integrity Calculation Package HTCH-21 Q-302R1, "Hatch Revised Material Data Review and ART Calculations" Transware Enterprises Document SNC-FLU-002-R-001 Revision 0, "Edwin I.

Hatch Unit 2 Reactor Pressure Vessel Fluence Evaluation at the End of Cycle 18 and 50.1 EFPY", April 23, 2007 Final Safety Evaluation of the BWR Vessel Internals Project BWRVIP-05 Report, July 28, 1998. (TAC No. MA93925)

Supplement to Final Safety Evaluation of the BWR Vessel and Internals Project BWRVIP-05, March 7,2000 (ADAMS Accession No: ML031430372)

10. Status Awaiting NRC Approval E1-6

Edwin I. Hatch Nuclear PlantMUnit 2 Proposed Alternative in Accordance With 10 CFR SO.SSa{a){3){i)

HNpMISI-ALT-11 Enclosure 2 Southern Nuclear Relief Request RR-38

H.L~Jr.

Vice President Hatch Project 0,.....

SoIItiIrINuc...

c..p..,.rIC.

Po5t 0I'rIa! b 1295 Birmingham. Alabama 35201 Tel 205.992.1219 September 13, 2004 E--cI" Snw Kr.r'firlJ Docket Nos.: 50-321 NL-04-1164 50-366 U. S. Nuclear Regulatory Commission A1TN: Doc:ument Control Desk Washinston. D. C. 20555-0001 HaIda Nuclear Plant Third IQ..Ywlptmal Insmice Jnpctioa Proanm SubmjUal of Revised ReJiefltcQ.voG Rll*31 Ladies aad Ocntlemea:

By letter dated Mardl29. 2004 Southern Nuclear Operating Company (SNC) submitted RR-38 to allow the deletion of the Section XI requiRd IlPV circumferential sheIL weld cxaminations(during the remainder oltho 40 yell' initial license) bued OD NRC approved BWRVIP..oS (BWR Iteactar Pressure Vcucl Shell Weld Inspection R.ecommcndatiou).

During the review of relief request RR-38. the NRC and SNC agreed that the relief request should be amended to iDcludc the period of ex.tcndcd operation (PEO).

Accordingly. the attached revised relief request extends the requested duration to include thePEO.

This letter contains no NIlC commitments. If you have any questions, please advise.

Sinecroly.

~$'~

H. L. Sumner, Jr.

HLS/illdaj

Attachment:

Revised RcliefRequest RR-18 cc: Southern Nuclear QlJ1SDljna Company Mr. J. T. Gasser Executive Vice President Mr. G. R. FftlClerick, Genera] Manager- Plant Hawb RTYPE: CHA02.004 U. S. Nuclear RuuLttnry Commission Dr. W. D. Travers, Regional Administrator Mr. C. Gflitton, NRR Projca Manager - Hatch Mr. D. S. Simpkins. Senior R.esident Inspector - Hatch E2-1

SOUTHBRN NUCLEAR. OPERATING COMPANY EDWlNLHATCHNUCLEARPLANT.UNrrS I AND 1 THIRD 10.YEAR INTERVAL REQUEST FOR RELIEF NO. RR.-38 I. System/Component for Which Reilen, Requested: This Relief Request applies to the Reactor Pressure Vessel (RPV) circumfenmtial shell weld examinadORl for ~h Unill I 8Ild 2.

II. Cqde ReqpirmJents: The foJlowing 1989 Edition ofASME Scc:tion XI Code requirements apply to this request.

• IWB-2S00 requires componems to be examiBed as specified in Table IWB-2SOQ..I.

• Table JWB.2S00-1, Category B-A. Item No. BI.l1 requires that all circumferential welds be essentially 100% examined.

III. Code BeAyiremeqt ftom Which Relief is Requested: Southem Nuclear Operating Company (SNC) proposes to penunently exclude the examination ofR.PV circumferential shell 'WCIIda as n:quircd in Table IWB*2S()()"I. Category B-A, Item No. B1.11 (This request is applicable for the current 4Q..year license and the Period ofExtend cd Operation (PEa)].

IV. Background Information: By letta'dated September 28. 1995 the Boiling Water Reactor Vessel and Internals Project (BWRVIP) submitted BWRVIP-OS (BWR Ractor Pressure Vessel Shell Weld Inspection Rccommendad.ons) to dw NRC. BWRVIP-OS initially proposed to reduOD the inspection covCl8ge ofthe Bft UV shell welds from essentially 100% of aU RPV shell welds to SO% of the longiwdlnal welds and 0% ofdw circumferaltial welds. By Ieaer dated October 29,1996 the BWRVJP modified the recommendation in BWRVIP~S to examine essentially 100% ofthe longitudinal welds and 0% of1he circumferential welds (except for that portion ofa circumferential weld iDlersecting with the longitudinal weld beiDg examined).

The NRC issued their final safety evaluation (SE) for BWRVIP-OS by letter dated July 28, 1998.

The SE stated that, "BWRIk:cnaca may request rclief fi'om the insorvice iDSpOCtion Rl'quirements of 10 CFR SO.SSI(S) for volumetric examination ofcinmtnferential reactor pressure welc:b (ASME Code Section XI, Table JWB..2SOQ..1. Examination Category B-A, Item 1.11.

Circumferential Shell Welds) by demonstrating: (1) at the expiration of their license, the circumferential welds satisfy 1he limiting (;Onditional failure pobability for eirewnferentia} welds in this evaluation, and (2) tIley have implemented operator training and established procedures that limit the tRquency ofcold over pressure events to the amount specified ill this report." The SE indicated that the NRC staff' concluded that a near-term safety concern did not exist; however.

the NRC staff'identified a need to evaluate the high conditional failure probabilities for axial welds. In a request for additional information, the NRC requested the BWRVIP to provide a more realistic potential for axial weld failwu due to (;Old over-pressure events mel 10 provide the failure fi'equency ofaxial welds based on NRC recommendations.

On November 10, 1998 the NRC issued Generic lAtter 98-0S (Boilinl Water Reactor Licensees Usc of tho BWRVIP..oS Report 10 Request Relioftiom Ausmentcd ExaminMion RoquilelllCDts on Reactor Pressure Vessel Circumferential She]) Welds) to povide guidaacc for licensees to request relieffrom the augmented oxamination requiremenu for circumferential RPV shell welds.

[By letter dated December 2, 1998 SNC requested approval to pennaRCntly exclude the ex.aminatlon ofthe ffatch Unit 1 RPV cirewnferential shell welds, based on this guidance, and by letter dated March 11. 1999 the NRC isslllhtl SE for Hatch Unit l gnntina this request pursuant to 10 CPR SO.5Sa(aX3Xi).]

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SOUTHERN NUCLEAR OPERATING OOMPANY EDWIN I. HATCH NUCLEAR PLANT, UNITS I AND 2 THIRD 100YEAllINTERVAL REQUEST FOR RELIEF NO. 0-38 (cont.)

By letters dated December 15.1991 and November 12, 1999 the BWRVIP supplied additional information reprdinl axial weld failwe probabi1ities. By letter dated March 7. 2000 the NRC issued a supplement to the July 28. 1998 SE concludina that, "'the RPV failure fi:equcncy due to the failure of tile limiting axial welds in the BWJl neet are below 5 x 10-4 per reactor-)'aI'.

consistent with RG 1.154, Biven the assumptions described in the atbIched S8." Therefore. the issue with axial wolds wu resolved.

By Ictta' dated January 31, 2001, in response to a request for additioaal inforrnedon (RAJ) for the Hatch Units I and 2 Licen. R.enewaJ Application (LM). SNC supplied Hatch Units I and 2 RPV weld conditioaal fai1ure probebilities and informatioa reprdinl cold over-pressure wen.. to the NllC. Tho NRC concluded in Section 4.6.2 of the October 5. 2001 Safety BvalUidon Report that SNC hujustified reUcffrom the ioservico inspection rcquiremontl of 10 CFR. 50.SSa(a) for volumetric examination ofcu.:umferential RPV welda during the PBO. The information supplied to the NRC in responllC to the RAts is provided in Enclosure 1. Because of iNuos associated with the conditional failure probabiU~ of uIaI welds durina the PEO. conditioMl failure probabilities for axial welds were also provided to tho NRC lind are included in thil relief request (for information purposes) u Enclosure 2.

V. To!;bnjcal Bui!: Per the NRC SE dated July 28. 1998 and Generic Letter 91-05. BWRIiCIODIOCI may request relief from the inserYioo inspection roquirancats of 10 CFIt. 50.5SI(a) for volumetric examination of' circumfereo1ial1aCtor presaure welds (ASME Soctioa XI Code, Table IWB-2500-1. Examination Category B-A, Item l.ll. CircumftNDtial Shell Welds) by demonstrating:

I. At the expindon of their licen1c, the cin:umferential welds satisty thCllimiting conditional failure probability far circwnfercntial welds in this evaluation.

2. Licensees have implemented operator traininl and csa.blisbed prcK:edures that limit the frequeacy ofcold over pRllure events to the amount specified in dais report.

Based on thCIIC two rcquirwncnts, the NRC bu previously:

• Granted approval for permanent deferral (durina tho iDitia140-yean of operation) of the Hatch Unit I auamented examination requirements for tho cu.:umferential welda pursuant to 10 CFR SO.S5e(g)(6XiiXAXS).

Indicated that SNC bas justified relief fiom the volumetric examination of the circumferential RPV welds during tile PEO.

Hatch Unit. I and 2 arc bounded by the NRC ualysit for circumferential weld limitina conditional failure probabilities durin. and at the ad of1IIe PEO... shown in Enc:Jo.ure 1.

Therefore. at the expiratioa of tho inilial40-year license period. tho Hllch Units I and 2 circumferential weld. also will satisfy the IlmitiDl conditioaal failure probability for circumferential welds. (Note: Hatch Unit 1 is currently in its '2!J't year ofcommercial operation and Hatcb Unit 2 is currently in its 25* year ofcommercial operation).

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SOUTIlERN NUCLEAR OPERATING COMPANY EDWIN J. HATCH NUCLEAR PLANT, UNITS 1AND 2 nnan IO-YEAR INTERVAL REQUEST FOR REUEF NO. RIt-38 (conL)

SNC has previously demonstrated that operator training and established procedures limit the frequency ofcold over pressure events. This information was supplied to the NRC in the December 2, 1998 Hatch Unit 1 submittal (for the permanent deferral ofthe auptented.

examination requirements). which was subsequently approved by the NRC in the March 11. 1999 SE. This information was later referenced by SNC in the January 31. 2001 response to License Renewal RAIs, where. it was also noted that tho operator training and proc:edures for Hatch Units t and 2 are the same. Extracts of this information are shown in Enclosure 3.

VJ. Alternatiye Examinations: Axial welds and intersecting portions ofcircumfercntial welds will be examined to the extent practical, dependent upon interference by another component or restrictions due to the geometrical coDfiguration. For those cases where the reduction in coverage is greater than 10%, relief will be requested pursuant to 10 CFR 50.55a requirements.

VII. Justjfication for ARProval: At the expiration of the PEO (60 years) and therefore the initial 40 year license period as well (which corresponds to the start ofthe PEO), tho Hatch Units I and 2 circumferential welds will satisfY the limiting conditional failure probability for ci~umferential welds. Procedures and training used to limit cold over-pressure events are the same for both Hatch units (approved for Hatch Unit 1 by NRC letter dated March II, 1999). The NRC has previously concluded that elimination ofthe Hatch Units I and 2 circumferential weld examinations during the PEO is justified and the NRC has previously granted approval for the permanent defenal ofthe auamented ci~umfcreDtial weld examination requirements for Hatch Unit I. Therefore, approval should be grwlted to eliminate the examination of the Hatch Units I and 2 RPV circumferential shell welds pursuant to 10 CFR 50.55a(a)(3)(i).

VIU. lmplementatign Schedule: Required for theHatch 2 RPV weld examinations during the 18111 Refueling Outage (currently scheduled to begin in February 2(05).

IX. Relief Request Status: Awaiting NRC approval.

E2-4

SOlJl'HERN NUCLEAR OPERATING COMPANY EDWIN I. HATCH NUCLEAR PLANT, UNITS 1AND 2 1lIIRD lO-YEARINTERVAL REQUEST FOR RELIEF NO. RR-38 (coot)

ENCLOSURE 1 EVALUAnON OF LIMITING CONDmONAL FAILURE PROBABILITrES FOR HATCH CIRCUMFERENTIAL WELDS DURING TIlE PERIOD OF EXTENDED OPERATION By letter dated January 31, 200 I, in response to a request for additional infonnatioo (RAJ) for the Hatch License Renewal Application (LRA). SNC supplied Hatch RPV weld conditional failure probabilities to the NRC. RAI 4.6-1 addressed the circumferential welds. and as shown below, the Hatch RPV conditional failure probability for circumferential welds is bounded by tho NRC analysis.

"The Hatch limiting circumferential weld properties from Tables 3-1 and 3-2 ofthc LRA Appendix E are compared to the information in Table 2.6-4 and Table 2.6-5 from the stBffSERon BWRVIP-05."

"The NRC staffused materials and flucncc data in Tables 2.6-4 and 2.6-5 to evaluate failure probability of BWR circumferential welds at 32 and 64 EFPY. Tho NRC used Mean RTNDT for the comparison. Mean RTNOT is defined as: RTHD1' +

6RTMDT. The Mean RTtIDT used by the NRC have been compared to the Hatch values derived using Appendix E of the LRA. The Hatch 1 and Hatch 2 values at 54 EFPY are bounded by the 32 EFPY analysis by the NRC by at least 40 or,

. and almost 75 Op at 64 EFPY. Although. conditional failure probability has not been calculated, the fact that the Hatch 54 EFPY valuo is boWJded by the 32 and 64 EFPY value the staff used leads to the conclusion that Hatch RPV conditional failure probability is bounded by the NRC analysis."

See the table below for the comparison ofvalues.

Group CE(V1P) CE(CEOG) CE(VIP) CE(CI:OG) Bate.. l U.klll llEFPY llEFPY MEFPY MEFPY 54EFPY 54EFPY Cu". 0.13 0.183 0.13 0.183 0.197 0.047 Ni% 0.71 0.704 0.71 0.704 0.060 0.049 CF 151.7 172.1 IS1.7 172.2 91.0 31.0 Flueac:e 0.20 0.20 0.40 0.40 0.236 0.244 (10" weml )

86.4 98.1

~~ 113.2 128.S 55.5 19.2 0 -10 R~

0 0 0 -SO Mea~TND1' 86.4 98.1 113.2 128.5 45.5 -30.1 P(FIE)

NRC 2.8IE-5 6.34E-S 1.99&4 4.38E-4 - --

P(FIE)

BWRVIP No Failure E2-5

SOUTHERN NUCLEAR OPERATING COMPANY EDWIN r. HATCH NUCLEAR PLANT, UNITS 1AND 2 THmDl~YEAkThITERVAL REQUEST FOR RELIEF NO. RR-38 (cont.)

ENCLOSURE 1 (Continued)

EVALUAnON OF LIMITING CONDITIONAL FAILURE PROBABILITIES FOR HATCH CIRCUMFERENTIAL WELDS DURING THE PERIOD OF EXTENDED OPERAnON

References:

1. Hatch License Renewal Application, Appendix E, Tables 3-1 and 3-2.

2. Final SEa ofthe BWR Vessel and Internals Project BWRVIP-05 Report (TAC No. M93925),

dated July 28, 1998.

3. GE-NE-AOO-05389-08, July 1995 Power UpnIe Evaluation TaSk Report for Edwin I. Hatch Plant Units 1 and 2, 110% Power Uprate Revised Impact on Vessel Fl'ICture Toughness.

4. GE-NE-A 13~0402-9, March 1998 Extended Power Up,..te Evaluldion Task Report for Edwin I.

Hatch Plant Units 1 and 2 Revised ImpactOD Vessel FrKturoToughncss.

5. BWRVIP BWI. Vessel and Internals Project BWR Reactor Frasure Vcsscllnspcction and Flaw Evaluation Guidclinos, TIt-llH96.

6. Struc:turallntegrity Associates Letter, SIR-OO-I60, Rev. O. December 18, 2000.

E2-6

SOUTHERN NUCLEAR. OPEltATlNO COMPANY EDWIN I. HATCH NUCLEAR. PLANT, UNITS I AND 2 TII1RD IO-YEAR INTERVAL REQUEST FOR RELIEF NO. RR-38 (cont.)

ENCLOSURE 2 EVALUATION OF LlMITINO CONDmONAL FAILURE PROBABIUTJES FOR HATCH AXlAL WELDS DURING THE PERIOD OF EXTENDED OPERATION In a response to RAI4.6-1. SNC supplied Hatch RPV axial weld conditional failure probabilities to the NRC. AI shown below. the Hateh RPV conditional failure probability for axial welds is bounded by the NRC analysis. RAI-4.6-2 states (in part): .

"The SEll in the May 7. 2000 letter supcrccdes the anaIysil in the July 28, 1998Ieuer. Therefore, the appliQnt should revise its analysil to compare the mean RTNDl' for the Plant Hatch axial welds to the mean RTI'IM for Pilgrim Mod 2,"

In response. SNC stated:

"The Hatch limitina axial weld }X'Opertiee from Table 3*1 and 3-2 of Appendix E ue compared to the information In Table 2.6-4 and Table 2.6-5 from the staff'SER on BWRVIP..o5. The NRC noted that it issued a revised SEa on BWR.VIP-oS on March 7. 2000 and that the limido, axial welds should be eompared with datil in Table:) ofthat document (Mod 2 in Table below). Mean RTHOT is defined 811: Mean RTmrr= RTJGT+ b.R.TND'I'. The MeanRTHD'rused bytbeNRC have been compand to the Hatch valuel deri'f'ed usina Appendix E of the LRA. A comparison of the Mean RTMDT vafuee from the NRC report with the Hatch data shows that the NRC analysis bounds the Hatch welds. Although. conditional failure probability has not been calculated. the fact that the Hatch 54 EFPY value is less than the 64 EFPY value the staft'uscd leads to the conclusion that HIII:h is bounded by the NRC lUIalysis."

Group Hod 2 Hate'" aatela2 54EFPY 54EFPY C ..% 0.316 0.216 Ni% 0.724 0.043 CJ' 219 98.0 Flueace 0.347 0.244 I (10"Dle..~

I ARkNDT ISS.l 60.6 R~ -2 -50 -so I M"~TND1' 114 10s.I 10.6 P(FIE)

NRC S.02E-6 - -

P(I'IE)

BWRVIP

References:

See cireumferential weld references.

E2-7

SOUTIIERN NUCLEAR OPERATING COMPANY

• EDWIN l. HATCH NUCLEAR PLANT. UNITS J AND 2 rnIRD I()'YEAR INTERVAL REQUEST FOR RELl£F NO. RR-38 (COOL)

ENCLOSURE)

EVALUAnON Of OPERATOR TRAINING AND ESTABLISHED PROCEDURES Plant Hatch hal procedures in placc which monitor and control reactor press~ temperature. and water inventory durin, all aspects of cold shutdown and retV.elin, operations which minimizes the likclibood ofa Low Temperature Over-Pressurization (LTOP) event fiom happenina. In addition to procedW'll collb'Ols. periodic Licensed OperatorTrainina further reduces the possibility ofoccurrence of LTOP events. Initial Licensed Operator Training and Simulator Trainina of plant hcatup and cooldown events includes perfonnance of SW'Yeillance tests and monitoring wltfch enlW'e pressure-temperature cW'\'o complian<:e. In addition, periodic operator trainina reinforces manapment*s expectations for strict procedural compl~.

finally. Southern 'Nuclear operaling personnel eontinuously review industry operatina experiences to ensure that Plant Hatch procedures consider the implCt of actual eveats. including LTOP events. Appropriate changes to proceduIa and training are then implemented to preclude similar situations &om occurring at Plant Hatch.

Based on the above. tho probability ofan LTOP event at Plant Hatch is considered to be less than or equal to that used in the 'NRC evaluation.

E2-8

Edwin I. Hatch Nuclear Plant-Unit 2 Proposed Alternative in Accordance With 10 CFR 50.55a(a)(3){i)

HNP-ISI-ALT-11 Enclosure 3 NRC Evaluation of RR-38 (TAC Nos. MC2381 and MC2382) dated January 28, 2005

January 28, 2005 Mr. H. l. Sumner, Jr.

Vice President - Nuclear Hatch Project Southern Nuclear Operating Company, Inc.

Post Office Box 1295 Birmingham, Alabama 35201-1295

SUBJECT:

EDWIN I. HATCH NUCLEAR PLANT. UNITS 1 AND 2 - EVALUATION OF RELIEF REQUEST (RR) NUMBER 38 (TAC NOS. MC2381 AND MC2382)

Dear Mr. Sumner:

By letter dated March 29, 2004, as supplemented by letter dated September 13. 2004.

Southern Nuclear Operating Company, Inc. (SNC or the licensee), submitted proposed altematives to the requirements of Section XI of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) under the provisions of Title 10 of the Code of Federal Regulations (10 CFR) Section 50.55a(a)(3}(ii) for the Edwin I. Hatch Nuclear Plant (Hatch), Units 1 and 2. SNC proposed two additional relief requests, RR-39 and RR-40, in the March 29, 2004, letter. The Nuclear Regulatory Commission (NRC) staff reviewed RR-39 and RR-40, and provided the results of that review in a letter to you dated January 7,2005.

The following paragraphs summarize the NRC staffs findings regarding RR-38:

In RR-38, the licensee proposed an alternative inspection program for the circumferential welds in the reactor vessels (RVs) of Hatch, Units 1 and 2. The alternative program applies a probabilistic fracture toughness analysis to justify eliminating volumetric examinations of the RV circumferential welds, as required in Table IWB-2500-1 of Section XI of the ASME Code, and the augmented volumetric inspections for these welds, as required in 10 CFR 50.55a(g)(6)(ii)(A)(2). The NRC staff completed their review and determined that the licensee's proposed alternative provides an acceptable level of quality and safety. Therefore. the licensee's alternative is authorized pursuant to 10 CFR 50.55a(a)(3)(i).

The March 29, 2004. letter requested that RR-38 be approved for the remainder of the 40-year initial license. During review of the RR. the NRC staff and SNC agreed that the RR should be amended to include the period of extended operation (I.e.* through its current 60-year renewed license). Upon further review. the NRC staff determined that the neutron f1uence values used in the RR were based on a calculational code (i.e., RAMA) that had not been reviewed and approved by the NRC staff. The licensee had initially antiCipated that the NRC staffs review and acceptance of the RAMA code for calculating reactor pressure vessel neutron fluence would be complete by the time this RR was needed. The NRC staff is currently reviewing. but has not yet approved the use of the RAMA code for neutron fluence calculations. As a result, the NRC staff reviewed the licensee's current. NRC approved neutron f1uence analyses and concluded that the fiuence values calculated using this methodology are conservative through July 31.2007. Therefore, the NRC staff limits the duration of RR-38 until July 31,2007.

E3-3

H.L. Sumner. Jr -2 Please note that if you plan to seek approval for RR-38 beyond July 31,2007, you must submit a revised RR based on an analysis that uses an NRC staff-approved neutron fluence calculation methodology (e.g., RAMA, If approved for use by the NRC staff at that time). This Issue was discussed with SNC staff in a teleconference on September 20,2004.

The NRC staffs Safety Evaluation is enclosed. If you have any questions, please contact Christopher Gratton at 301-415-1055.

Sincerely, IRAJ John A. Nakoski, Chief. Section 1 Project Directorate II Division of licensing Project Management Office of Nudear Reactor Regulation Docket Nos. 50-321 and 50-366

Enclosure:

As stated cc w/end: See next page E3-2

' Ml050130317 NRR-D28 OFFICE PDII-1/PM PDII-11lA SRXB{B) EMCBlSC(A) OGC PDII-1/SC NAME CGratton CHawes JUhle BEliiot MZobler JNakoski DATE 1/27/05 1/27/05 01106105 1111/05 1118105 1/28105 Edwin I. Hatct! Nuclear Plant, Units 1 & 2 cc:

Lau rence Bergen Chairman Oglethorpe Power Corporation Appling County Commissioners 2100 E. Exchange Place Cou nty Courthouse P.O. Box 1349 Baxley, GA 31513 Tucker, GA 30085-1349 Mr. Jeffrey T, Gasser Mr. R.D. Baker Executive Vice President Manager* Licensing Southern Nuclear Operating Company, Inc, Southern Nuclear Operating Company. Inc. P.O. Box 1295 P.O. Box 1295 Birmingham, AL 35201-1295 Birmingham. Al 35201*1295 Mr. G. R. Frederick, General Manager Resident Inspector Edwin I. Hatch Nuclear Plant Plant Hatch Southern Nuclear Operating Company, Inc.

11030 Hatch Parkway N. U.S. Highway 1 North Baxley, GA 31531 P.O. Box 2010 Baxley. GA 31515 Harold Reheis, Director Department of Natural Resources Mr. K. Rosanski 205 Butler Street, SE., Suite 1252 Resident Manager Atlanta. GA 30334 Oglethorpe Power Corporation Edwin I. Hatch Nuclear Plant Steven M. Jackson P.O. Box 2010 Senior Engineer - Power Supply Baxley. GA 31515 Municipal Electric Authority of Georgia 1470 Riveredge Parkway. NW Atlanta, GA 30328-4884 Mr. Reece McAlister Executive Secretary Georgia Public Service Commission 244 Washington St., SW Atlanta. GA 30334 Arthur H. Domby. Esq.

Troutman Sanders Nations Bank Plaza 600 Peachtree St. NE. Suite 5200 Atlanta, GA 30308-2216 E3-4

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO RELIEF REQUEST (RFn NO. RR-38 SOUTHERN NUCLEAR OPERATING COMPANY, INC.

EDWIN I. HATCH NUCLEAR PLANT, UNITS 1 AND 2 DOCKET NOS. 50-321 AND 50-366

1.0 INTRODUCTION

In Letter No. NL-04-0478 dated March 29. 2004, as supplemented by Serial Letter No.

NL-04-1764 dated September 13. 2004, Southern Nuclear Operating Company, Inc. (SNC, the licensee), proposed an alternative to the requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI and applicable Addenda (henceforth Section XI), regarding the volumetric examination requirements for reactor vessel (RV) circumferential shell welds at the Edwin I. Hatch Nuclear Plant (Hatch), Units 1 and 2. The licensee also requested relief from the requirements of Title 10 of the Code of Federal Regulations (10 CFR) Section 50.55a(g)(6)(H)(A)(2), as they pertain to performing augmented volumetric inspections of the RV circumferential welds at Hatch. Units 1 and 2.

Nuclear Regulatory Commission (NRC, the Commission) staff approval of the RR would authorize the use of a proposed alternative to these volumetric examinations requirements in accordance with the alternative probabilistic fracture mechanics methods discussed in Electric Power Research Institute Topical Report. "BWR Vessel and Internals Project. BWR Reactor Pressure Vessel Shell Weld Inspection Recommendations (BWRVIP-05)." and with the NRC's guidelines for proposing these alternative programs, as established in NRC Generic Letter (GL) 98-05, "Boiling Water Reactor licensees Use of the BWRVIP-05 Report to Request Relief from Augmented Examination Requirements on Reactor Pressure Vessel Circumferential Shell Welds."

2.0 REGULATORY EVALUATION

2.1 Inservice Inspection Requirements Inservice inspection (lSI) of the ASME for Code Class 1,2, and 3 oomponents is performed in accordance with Section XI of the ASUE Code and applicable addenda as required by 10 CFR 50.55a(9). except where specifIC relief has been granted by the Commission pursuant to 10 CFR 50.55a(g)(6)(I). 10 CFR 50.55a(a)(3) states that altematives to the requirements of Enclosure E3 5

-2 paragraph (g) may be used, when authorized by the NRC, if: (i) the proposed alternatives would provide an acceptable level of quality and safety or (ii) compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

Pursuant to 10 CFR 50.55a(g)(4), ASME Code Class 1. 2. and 3 components (including supports) shall meet the requirements, except the design and access provisions and the pre-seNice examination requirements. set forth in the ASME Code,Section XI, "Rules for InseNice Inspection [lSI] of Nuclear Power Plant Components," to the extent practical within the limitations of design, geometry, and materials of construction of the components. The regulations require that inseNice examination of components and system pressure tests conducted during the first 10-year inteNal and subsequent inteNals comply with the requirements in the latest edition and addenda of Section XI of the ASME Code incorporated by reference in 10 CFR 50.55a(b) 12 months prior to the start of the 120-month interval. subject to the limitations and modifications listed therein. The applicable Code of record for the second 10-year lSI for Hatch, Units 1 and 2 is the 1989 Edition of the ASME Code,Section XI.

2.2 Augmented Inservice Inspections Requirements for RV Shell Welds 10 CFR 50.55a(g)(6)(ii)(A)(2) requires licensees to augment their reactor vessel examinations by implementing, as part of the 151 interval In effect on September 8, 1992. the examination requirements for RV shell welds specified in Item B1.10,Section XI, Table IWB-2500-1, Examination Category B-A, *Pressure Retaining Welds in Reactor Vessel."Section XI Item B1.1 0 includes the volumetric examination requirements for both circumferential RV shell welds, as specified in Section XI Item B1.11 , and longitudinal RV shell welds, as specified in Section XI Item B1.12. 10 CFR SO.S5a(g)(6)(ii)(A)(21 defines "essentially 100% examination" as covering 90 percent or more of the examination volume of each weld.

2.3 Additional Regulatory Guidance 2.3.1 BWRVIP-05 Report By letter dated September 28. 1995, as supplemented by letters dated June 24 and October 29, 1996, May 16, June 4, June 13, and December 18, 1997, and January 13, 1998, the Boiling Water Reactor (BWR) Vessel and Internals Project (BWRVIP). a technical committee of the BWR Owners Group (BWROG), submitted proprietary report BWRVIP-05. The BWRVIP-05 report evaluates the current inspection requirements for RV shell welds in BWRs, formulates recommendations for altematlve inspection requirements, and provides a technical basis for these recommended requirements. As mOdified, the 8WRVIP-05 proposed to reduce the scope of inspection of BWR RV welds from essentially 100 percent of all RV shell welds to examination of 100 percent of the axial (i.e., longitudinal) welds and essentially zero percent of the circumferential RV shell welds, except for the intersections of the axial and circumferential welds. In addition, the report includes proposals to provide alternatives to ASME Code requirements for successive and additional examinations of circumferential welds, provided in paragraph IW8-2420 and IWB-2430 respectively, of Section XI of the ASME Code.

E3-6

-3 On July 2S, 1995, the NRC staff issued a Safety Evaluation Report (SER) on BWRVIP-05.

This evaluation concluded that the failure frequency of RV circumferential welds in BWRs was sufficiently low to justify elimination of lSI of these welds. In addition, the evaluation concluded that the BWRVIP proposals on successive and additional examinations of circumferential welds were acceptable. The evaluation Indicated that examination of the circumferential welds will be performed if axial weld examinations reveal an active degradation mechanism. The NRC staff supplemented this SER in an SER to the BWRVIP dated March 7, 2000. In this SER, the NRC staff updated the Interim probabilistic failure frequencies for RV axial shell welds and revised the Table 2.6-4 to correct a typographical error in the 32 effective full power years (EFPY) Mean RTNOT value cited for the limiting Chicago Bridge and Iron (CB&I) case study for circumferential welds. The correction changed the 32 EFPY Mean RTNOT value for the CB&I case study from 109.5*F to 134.9 *F.

In the BWRVIP-05 report, the BWRVIP committee concluded that the conditional probabilities of failure for BWR RV circumferential welds are orders of magnitude lower than that of the axial welds. As a part of its review of the report, the NRC conducted an independent probabilistic fracture mechanics assessment of the results presented In the BWRVIP-05 report. The NRC staff's assessment conservatively calculated the conditional probability of failure values for RV axial and circumferential welds during the initial (current) 40-year license period and at conditions approximating an SO-year vessel lifetime for a BWR nuclear plant. The failure frequency is calculated as the product of the frequency for the critical (limiting) transient event and the conditional probability of failure for the weld.

The NRC staff determined the conditional probability of failure for axial and circumferential welds in BWR vessels fabricated by CB&I, Combustion Engineering (CE), and Babcock and Wilcox (B&W). The analysis identified a cold overpressure event that occurred in a foreign reactor as the limiting event for BWR RVs, with the pressure and temperature from this event used in the probabilistic fracture mechanics calculations. The NRC staff estimated that the probability for the occurrence of the limiting overpressurization transient was 1 x 1crJ per reactor year. For each of the vessel fabricators, Table 2.6-4 of the NRC staffs SER of March 7, 2000, identifies the conditional failure probabilities for the plant-specific conditions with the highest projected reference temperature (for that fabricator) through the expiration of the initial 40-year license period. Table 2.6*5 of NRC staffs SER of July 2S, 1998, identifies the conditional failure probabilities for the plant-specifIC conditions with the highest projected reference temperature (for that fabricator) through the expiration of an SO-year license period, which constitutes the licensing basis if two 20-year extended periods of operation have been granted for a BWR-<lesigned nuclear power plant.

2.3.2 Generic Letter 9S-05 On November 10, 1998, the NRC issued GL 98-05 that states that BWR licensees may request permanent (i.e., for the remaining term of operation under the existing, initial license) relief from the lSI requirements of 10 CFR 50.55a(g) for the volumetric examination of circumferential reactor pressure vessel welds (ASME Code Section XI, Table IWB-2500-1, Examination Category B-A, Item No. B1.11, 'Circumferential Shell Welds") by demonstrating conformance with the following safety criteria:

E3-7

• 4 (1) At the expiration of the operating license, the licensees will have demonstrated that limiting probability of failure for their limiting RV circumferential welds will continue to satisfy (I.e., be less than) the limiting conditional failure probability for circumferential weld assessed in the applicable BWRVIP-05Iimiting case study.

(2) licensees have implemented operator training and established procedures that limit the frequency of cold overpressure events to the amount specified in the NRC staffs July 28. 1998. SER.

In GL 98-05, the NRC staff stated that licensees applying the BWRVIP-05 criteria would need to continue performing the volumetric inspections of all axial RV shell welds that are required by the ASME Code.Section XI. Table IW8-2500-1,lnspectlon Category B-A, Item B1.12, and the augmented volumetric inspectIons of the RV axial shell welds that are required under 10 CFR 50.55a(g)(6)(U)(A)(2). For plants that are currently licensed to operate in accordance with their initial40-year operating licenses. the limiting case studies are provided in Table 2.6-4 of the revised SER on BWRVIP-05 dated March 7. 2000. For plants that have been granted operating licenses to operate for an extended period of operation. the limiting case studies are provided in Table 2.6-5 of the NRC staffs SER of July 28. 1998.

3.0 TECHNICAL EVALUATION

3.1 Code Requirement for Which Relief is Requested The licensee requested relief from the following requirements in the 1989 Edition of Section Xl of the ASME Code (Section XI):

Subarticle IWB-2500, Table 1W8 2500-1, Examination Category B-A, "Pressure Retaining Welds In Reactor Vessel," No. 81.11, *Circumferential Shefl b

Welds.

3.2 Licensee's Proposed Alternative to the ASME Code Using the guidelines of GL98-05 and Topical Report 8WRVIP-05 and the NRC staff's determination in its July 28.1998. SER on 8WRVIP-05, the licensee proposed to use a probabilistic fracture mechanics evaluation for the circumferential shell welds in the Hatch.

Units 1 and 2 RVs as the basis for eliminating the required volumetric examinations and augmented volumetric examinations for the welds through the expiration of the extended periods of operation for Hatch, Units 1 and 2.

The licensee proposed the following alternative in lieu of performing the required volumetric examinations of the RV circumferential shell welds:

Axial welds and intersecting portions of circumferential welds will be examined to the extent practical, dependent upon interference by another component or restrictions due to the geometrical conftguration. For those cases where the reduction in coverage is greater than 10%. relief will be requested pursuant to 10 CFR 50.55a requirements.

E3-8

-5 In SNC letter No. Nl-04-1764. dated September 13, 2004, the licensee clarified that the alternative inspection program in RR*38 is requested through the expiration of the periods of extended operation for Hatch, Units 1 and 2.

3.3 licensee's Bases for Alternative The licensee based RR-38 on the NRC's RR provisions of GL 98-05 and the guidelines of BWRVIP-05. The licensee cited the following acceptance criteria as the bases for evaluating the acceptability of RR-38.

Per the NRC SE dated July 28, 1998 and Generic letter 96*05. BWR licensees may request relieffrom the inservice Inspection requirements of 10 CFR 50.55a(g) for volumetric examination of circumferential reactor pressure welds (ASME Section XI Code, Table IWB-2500-1. Examination Category B-A. Item 1.11. Circumferential Shell Welds) by demonstrating:

1. At the expiration of their license, the circumferential welds satisfy the limiting conditional failure probability for circumferential welds in this evaluation ....[Gl 98-05 Safety Criterion 1).

2. licensees have implemented operator training and established procedures that limit the frequency of cold over pressure events to the amount specified in this report ....[Gl 96-05 Safety Criterion 2).

3.3.1 License Basis for Conforming with Gl 98-05 Safety Criterion 1 - Criterion for Conditional Probabilities of Failure In letter dated March 29. 2004. the licensee provided its 54 E FPY Mean RTNOT calculations for the limiting circumferential welds in the Hatch. Units 1 and 2 RVs (Refer to Enclosure 1 of RR-38) in order to support its basis for meeting GL 96-05 Safety Criterion 1 and to demonstrate that the 54 EFPY Mean RTNOT values for Hatch. Units 1 and 2 are bounded by the Mean 64 EFPY RTNOT value for the limiting CE-V1P case study.

3.3.2 license Basis for Conforming with Gl 98-05 Safety Criterion 2 - Criterion on Mitigating the Probability of Cold Overpressurization Events The licensee provided the following technical basis for meeting Gl 98-05 Safety Criterion 2:

SNC has previously demonstrated that operator training and established procedures limit the frequency of cold over pressure events. This information was supplied to the NRC in the December 2.1998 Hatch Unit 1 submittal (for the permanent deferral of the augmented examination requirements), which was subsequently approved by the NRC in the March 11. 1999 SE. This information was later referenced by SNC In the January 31,2001 response to license Renewal RAls. where, it was also noted that the operator training and procedures for Hatch Units 1 and 2 are the same. Extracts of this information are shown in Enclosure 3 [of SNC Letter NL-04-478, dated March 29. 2004).

E3-9

-6 4.0 NRC STAFF EVALUATION As discussed in Section 2.3.2 of this SE, GL 98-05 provides two criteria that BWR licensees requesting relief from lSI requirements of 10 CFR 50.55a(g) for the volumetlic examination of circumferential RV welds (ASME Code,Section XI, Table IWB-2500-1, Examination category B-A, Item No. B1.11, Circumferential Shell Welds) must satisfy. These criteria are intended to demonstrate that the conditions at the applicanfs plant are bounded by those in the safety evaluation. The licensee will still need to perform the required inspections of "essentially 100 percenf of all axial welds.

4.1 Neutron Fluence Calculation for RR-38 For any given RV circumferential or axial weld material, the conditional probability of failure increases with the material's neutron fluence value and mean RTNDT value, as projected to the expiration of the operating license for the facility. Gl 98-05 stipulates that, at the expiration of the operating license, the mean RTNOT estimates for circumferential welds should satisfy the limiting conditional failure probability for the weld materials, as stated in the NRC staffs SER of July 28, 1998. The neutron fluence values for the RV circumferential welds at the inside surface of the RV are critical inputs to the mean RT NOT estimate calculations.

SNC's current method for calculating neutron fJuence does not conform with the NRC staffs recommended methodology in Regulatory Guide (RG) 1.190, "Calculational and Dosimetry Methods for Determining Pressure VEssel Neutron Fluence: SNC had committed to have a conforming neturon fluence calculational methodology (i.e., the RAMA code) for Hatch, Units 1 and 2 approved by December 15, 2004. However, the approval of the RAMA code has been delayed. In a letter dated July 13, 2004 (SNC letter No. Nl-04-1123), SNC requested to revise its commitment date for having a neutron fluence calculational methodology that is compliant with RG 1.190, from December 15, 2004, to July 31, 2007. SNC stated that the previous commitment date was selected arbitrarily (based on the expectation that RAMA would be approved by this date), and that there was sufficient conservatism in the pressure-temperature (P-T) curves calculated using the current fluence methodology to allow the plant to operate safely until July 31,2007. The NRC approved the commitment change request in a letter to SNC dated September 9, 2004. The NRC staff is currently reviewing the RAMA code and anticipates that the review will be complete before July 31, 2007.

NRC staff experience has shown that neutron fluence values that are calculated using methods that do not conform with RG 1.190 are typically within +1- 40 percent of the values that would be obtained using the recommended methodology of RG 1.190. As stated previously, SNC's current fluence methodology does not conform with RG 1.190. SNC stated in a Letter No.

Nl-04-11S2. also dated July 13, 2004. that based on the current fluence methodology, the estl mated fluence values on August 1, 2007, for Hatch Units 1 and 2, will be 24.2 and 22.1 EPFY, respectively. These values are 44.8 percent and 40.9 percent of the 54 EFPY neutron fluence values estimated by the current methodology for the Hatch, Units 1 and 2. respectively.

As a result, neutron fluence estimates for August 1. 2007, are more conservative than the values estimated for the fJuence at 54 EFPY. even considering the 40 percent adjustment for the nonconforming methodology.

E3-I0

-7 Therefore, the NRC staff finds the licensee's neutron fluence estimates are acceptable to warrant approval of fluence values used in the 54 EFPY Mean RTNOT analyses. However. the NRC staff is limiting the acceptance of RR-38 to the period through July 31, 2007. SNC also stated in NL-04-1152 that the P-T limit curves will be evaluated and revised, if necessary at that time.

4.2 Circumferential Weld Conditional Failure Probability The NRC staffs SER for the BWRVIP-05 report evaluated the conditional failure probabilities for axial and circumferential shell welds in the limiting BWR RV designs manufactured by different vendors, including RVs manufactured using by CEo CB&I, and B&W. The SER also reported the Mean RTNOT calculations and values that were derived from the conditional failure probabilities for the limiting case studies. For a plant granted a renewed operating license, the evaluation criteria for the limiting conditional failure probabilities and Mean RTNDT values are those listed for the limiting case studies specified in Table 2.6-5 of the staffs SER of July 28, 1998. The associated limiting case studies, conditional failure probabilities, and Mean RTNOT values listed in Table 2.6-5 of the NRC staffs SER are limited to plants that have accumulated no more than 64 EFPY of power operation.

The renewed operating licenses for Hatch, Units 1 and 2 were approved and issued by the NRC on January 15, 2002. In the renewed operating licenses, the staff granted power operation through August 6, 2034, for Hatch, Unit 1 and June 13,2038. for Hatch, Unit 2. which represent operations through 54 EFPY of power operation. The period of applicability in Table 2.6-5 of the SER on BWRVIP-05 Is bounding for operations of the Hatch, Units 1 and 2 reactors to the expiration of the extended operating licenses and is representative of the evaluation for RR-38.

Since the Hatch, Units 1 and 2 RVs were fabricated by CE, the CE-VIP limiting case study in Table 2.6-5 provides the applicable case-study conditional probability of failure value and Mean RTNOT value criterion for the evaluation of RR-38.

In the license renewal application for the Hatch, Units 1 and 2 reactors, SNC identified the calculation of the Mean RTnell values for the Hatch, Units 1 and 2 RV circumferential welds as a time-limiting aging analysis (TLAA) for the application. In the staffs evaluation in Section 4.6 of NUREG-1803, *Safety Evaluation Report Related to the License Renewal of the Edwin I. Hatch Nuclear Plant. Units 1 and 2 (December 2001)". the NRC staff concluded that SNC had performed a valid TLAA analysis to justify re-submittal of the alternative inspection proposal for the Hatch, Units 1 and 2 RV circumferential welds to the expiration of the extend periods of operation for the reactor units. SNC's submittal of RR-38 on March 29, 2004, as amended in SNC Letter No. NL-04-1764, dated September 13, 2004, was performed to justify elimination of the volumetric examinations and augmented volumetric examinations for the RV circumferential welds through the expiration of the extended periods of operation for Hatch, Units 1 and 2.

The NRC staff performed an independent calculation of the Mean RT NOT values for the limiting Hatch, Units 1 and 2 RV circumferential welds through 54 EFPY. Table 4.1-1 on page 8 of this SE provides a summary of the Mean RT NOT values calculated by the staff for the Hatch, Units 1 and 2 RV through 54 EFPY and a comparison of the staffs Mean RTNOT values to both the corresponding Mean RTNDT values calculated by SNC and the Mean RTNDT value criterion for the limiting CE-VIP case study at 64 EFPY.

E3-11

-8 The results in Table 4.1-1 demonstrate that the Mean RTHOI' values calculated by the licensee for the Hatch. Units 1 and 2 RV circumferential welds are less than that for the limiting CE-VIP case study and are in agreement with those calculated by the NRC staff. Based on this analysis, the NRC staff concludes that SNC has provided a valid basis for conciudlng that the conditional probability of failure values for the Hatch. Units 1 and 2 RV circumferential welds are sufficiently low to justify elimination of the volumetric examinations that are required for these welds through 54 EFPY. However. for the reasons stated in Sections 4.1 and 5.0 of this S E.

the NRC staff is limiting its authorization for elimination of the required examinations until July 31. 2007.

Table 4.1-1 Comparison of NRC and SNC 54 EFPY Mean RTNOT Calculations to the 64 EFPY Mean Calculations for the Limiting CE-VIP Case Stud on BWRVIP-05 NRC 54EFPY SNC 54EFPY NRC 54EFPY SNC 54 EFPV Limiting Mean RTNDf Mean RTNOr Mean RTNOT Mean RTHOf 84 EFPY Calculations Calculations Calculations Calculations CE*VIP for Hatch, for Hatch. for Hatch. for Hatch.

Case Study Unit 1 Unit 1 Unit 2 Unit 2 (Note 1) (Note 1) (Note 1) (Note 1)

Alloy %Cu ." 0.197 0.19T 0.047 0.047 Alloy °Ao Ni 0.71 0.060 0.060 0.049 0.049 RTIIIJt(UI ('f) 0 -10.0 *10.0 -50.0 -50.0 Fluence I 0.400 0.236 0.236 0.244 0.244 (10'* n/cma)

Chemistry Factor 151.7 91.4 910 30.7 31.0 ART..m ('F) 113.2 55.7 55.5 19.0 '9.2 Mean RT"DT (-F) 1'3.2 45.7 45.5 -31.0 *30.8 NRC Establiahed 1.99E.... Mean RTNor is Mean RTNOT is Mean RTNOT is Mean RTNOT is Probability of Faijure (no fa~ure: Lower than Lower than Lower than Lower than

[ P(FIE) ICriterion for No1e2) Case Study Case Study Case Study Case Study Case I Result for Plant Mean RTNOT : Mean RT"IlT: Mean RTICIT : Mean RTNOT:

Specifi: Calculation Criterion Is Criterion is Criterion Is Criterion is met. (Nole2) met. (Note 2) met. (Note 2) mel. (Note 2)

Noles: 1. For the Hatch. Units 1 and 2 RV** Ihe Imiting circumferential weld materials delermined by the staff were equivalent to those determined by SNC. For Hatch.1, the limiting RV circumferential weld is

, -313A. which was fabricated from weld heat No. 90099. For Hatch-2. the limiting RV circumferential weld is 301-871, which was fabriCated from weld heat No. 4P6052.

2. If the plant-specific Mean RT""T is leu than the Mean RTNOT aSllOCia\ed with Limiting Case Study, the staff concludes that probability of failure lor ttle plant*specifc circumferential weld under review will be less that for the limiting c:ircumferential weld In the Limiting Case Study. BWR plants thai meet this cmeticn may conclude that the probability of faiu re for the limiting circumfenlntial RV welds Is sufficient to justify elimination of the volumetric examinations required by SectiOn XI of the ASME Code (Examination Category B-A. Item 8.1.11) and augmented volumetric examinations for the circumferential welds requiled by 10 CFR 50.55afg)(6)(il)(A}(2).

E3 12

-9 4.3 Minimizing the Possibility of Low Temperature Overpressurization The licensee stated that its bases for meeting Acceptance Criterion 2 of GL 98-05 and for demonstrating that the licensee has implemented acceptable procedures and controls for mitigating a low-temperature-overpressurization event are given in the "Consideration of Low Temperature - Over Pressurization Events* section of the licensee's enclosure to SNC Serial LeHer HL 5710, dated December 2, 1998, and are applicable to the evaluation of RR-38. In this letter, the licensee stated that the following operational controls, procedural controls, and staff training practices are in place to minimize the possibility of a low temperature overpressurization event.

4.3.1 Operational and System Design Considerations From an operational basis, the reactor feedwater system (RFS), high pressure coolant injection system (HPCI), reactor core isolation cooling system (RCIC), and standby liquid control system (SLC) are the high pressure systems that provide coolant or have the potential to provide coolant at high pressure into the RV. The HPCI and RCIC pumps are steam driven and cannot function during cold shutdown. The RFS pumps automatically trip during transient and postulated loss-of-coolant accidents (LOCA) conditions, and are manually tripped during routine reactor shutdowns. Since RFS pumps are steam driven, they cannot be operated during cold shutdown condition. Although not addressed in the licensee's submittal, the NRC staff noted that SLC is solely a manual injection system; there are no automatic starts associated with SLC at Hatch, Units 1 and 2. Operator initiation of the SLC occurs only in accordance with applicable Hatch, Units 1 and 2 emergency operating procedures and would not occur during normal utility planned shutdowns of the reactors or during transient operating conditions. SLC might be manually initiated during a postulated design basis LOCA; however, if manually initiated, the SLC injection rate of approximately 40 gallons per minute (gpm) would allow operators sufficient time to control RV water level and pressure during the postulated event.

The core spray system (CS), low-pressure coolant Injection/residual heat removal system (LPCIIRHR). control rod drive system (CRD), and reactor water cleanup system (RWCU) can also inject coolant into the reactor. CS and LPCIIRHR are low pressure emergency core cooling systems (ECCS), whose pumps' create a shutoff head of 375 pounds per square inch differential (psid) for CS and 223 psid for LPCI/RHR. Should either of these systems be started (Le., inject as designed) during cold shutdown, the resulting reactor pressure and temperature would be below the P-T limits. CRD and RWCU use a feed-and-bleed process to control RV level and pressure during normal cold shutdown conditions.

4.3.2 Procedural Considerations Plant-speCific normal operating and transient operating procedures have been established to provide gUidance to the operators regarding compliance with the Technical Specification P-T limits. These procedures direct operators to respond to any unexpected or unexplained rise in reactor water level. which could result from spurious actuation of an Injection system. The procedural actions include preventing condensate pump injection, securing ECCS system injection, tripping CRD pumps, terminating all other injection sources, and lowering RV level via the RWCU system. In addition, plant-specific emergency operating procedures have been E3-13

- 10 established to ensure that proper operating actions are followed during postulated design-basis LOCAs. The emergency procedures include control of reactor water level, reactor pressu re, and reactor temperature during these postulated events and instructions on how and when SLC should be manually initiated.

4.3.3 Operator Training The licensee emphasized that training and testing of control room operators is an integral part of ensuring the abilities of the operators to Implement these procedures. On the basis of the P-T limits of the operating systems, operator training, and es1ablished plant-specific procedures, the licensee determined that a nondeslgn-basis cold overpressure transient is unlikely to occur.

4.3.4 Staff Determination on SNC's Basis for Meeting Acceptance Criterion 2 of GL 98-05 The staff concluded that, based on the licensee's information provided about the systems that inject at high pressures, operator training, and plant-specific procedures at Hatch, Units 1 and 2, the possibility of a low temperature overpressurization event will be minimized, and thus, the licensee has provided a sufficient basis to support the NRC staffs approval of the alternative examination request for circumferential shell welds in the Hatch, Units 1 and 2 RVs.

5.0 CONCLUSION

FOR RR-38 The NRC staff has completed its review of the licensee's submittal and determined that the licensee conforms to the applicable safety evaluation criteria In NRC GL 98-05 and in the BWRVIP-05 report. The NRC staff has also determined that the licensee has acceptably demonstrated that the conditional probability of failure values for the Hatch, Units 1 and 2 RV circumferential welds are sufficiently low to justify the elimination of the augmented volumetric examinations required by 10 CFR 50.55a(g)(6)(ii)(A)(2), and the volumetric examinations required by the ASME Code,Section XI, Table IWB-2500-1, Examination Category B-A, Item No. B1.11.

Based on this analysis, the NRC staff concludes that the licensee's proposed altemative will provide an acceptable level of quality and safety in lieu of performing the required volumetric examinations. Therefore, the licensee's altemative is authorized pursuant to 10 CFR 50.55a(a)(3)(i).

SNC's letter dated March 29, 2004. requested that RR-38 be approved for the remainder of the 40-year initial license. During review of the RR, the NRC staff and SNC agreed that the RR should be amended to include the period of extended operation. Upon further review. the NRC staff determined that the neutron fluence values used in the RR were based on a calculational code (RAMA) that had not been review and approved by the NRC staff. The licensee had initially anticipated that the NRC staffs review and acceptance of the RAMA code for calculating reactor pressure vessel neutron fluence would be complete by the time this RR was needed. The NRC staff is currently reviewing, but has not yet approved the use of the RAMA code for neutron f1uence calculations. As a result. the NRC staff reviewed the licensee's current, NRC-approved neutron fluence analyses and concluded that the f1uence values calculated using this methodology are conservative through July 31. 2007. Therefore, the NRC staff limits the duration of RR-38 until July 31,2007.

E3-14

- 11

• Additional requirements of the ASME Code,Section XI for which relief has not been specifically requested and approved by the NRC staff remain applicable. including third party reviews by the Authorized Nuclear Inservice Inspector, Principal Contributor. J. Medoff. DE L. LOis. OSSA Date: January 28. 2005 E3 15

Edwin I. Hatch Nuclear Plant-Unit 2 Proposed Alternative in Accordance With 10 CFR 50.55a(a)(3)(i)

HNP-ISI-ALT-11 Enclosure 4 NRC Safety Eval uation of Proprietary EPRI Reports (TAC No. MB9765) dated May 13, 2005

BWRVI P 2005-20.98 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 2OII5HOO1 May 13.2005 Bill Eaton, BWRVIP Chairman Entergy Operations, Inc.

Echelon One 1340 Echelon Parkway Jackson, MS 39213-8202

SUBJECT:

SAFETY EVALUATION OF PROPRIETARY EPRI REPORTS, "BWR VESSEL AND INTERNALS PROJECT, RAMA FLUENCE METHODOLOGY MANUAL (BWRVIP~114)" -RAMA FLUENCE METHODOLOGY BENCHMARK MANUAL EVALUATION OF REGULATORY GUIDE 1.190 BENCHMARK PROBLEMS (BWRVIP-115),- -RAIIAA FLUENCE METHODOLOGY-SUSQUEHANNA UNIT 2 SURVEILLANCE CAPSULE FLUENCE EVALUATION FOR CYCLES 1*5 (BWRVIP-117): AND -RAMA FLUENCE METHODOLOGY PROCEDURES MANUAL (BWRVIP-121 ),. AND "HOPE CREEK FLUX WIRE DOSIMETER ACTIVATION EVALUATION FOR CYCLE 1 (TWE-PSE-001-R-001)*

(lAC NO. MB9765)

Dear Mr. Eaton:

By letters dated June 11. 2003. June 26. 2003, August 5, 2003. October 29, 2003, and March 24.2004. respectively, the Boiling Water Reactor Vassel and Intemals Project (BWRVIP) submitted the following Electric Power Research Institute (EPRI) proprietary reports for staff review and approval, -BWR Vessel and Internals Project, RAMA Fluence Methodology Manual (BWRVIP-114).- ~MA Auenoe Methodology Benchmark Manual--Evaluation of Regulatory Guide 1.190 Benchmark Problems (BWRVIP-115},-"RAMA Fluence Methodology Susquehanna Unit 2 Surve.lance Capsule Fluence Evaluation for Cycles 1*5 (BWRVIP-117).~

"RAMA Fluence Methodology Procedures Manual (BWRVIP-121); and "Hope Creek Flux Wire Dosimeter Activation Evaluation for Cycle 1 (TWE-PSE-001-R-D01):

The reports listad above provide and support a methodology which Is a new approach to neutron transport that has been developed by the BWRVIP for detennlnlng neutron fluance to the reactor pressure vessel (RPV) and internal components of BWR plants. The Radiation Analysis Modeling Application (RAMA) code will be applied In 1he reactor beltllne region denned by the top and bottom planes of the active fuel and the Inner wall of the biological shield. The methodology employs the RAMA computer code for evaluating the neutron nux from the core through the downc:omer. vessel internals, and through the RPV wall.

E4-1

B. Eaton -2 The staff has completed its review of the proposed methodology and ftnds that the methodology perfonns as described; however, the BWRVIP did not quanUfy the bias and uncertainty required for the quallftcatlon of the methodology, as stated In RG 1.190, -Radiation EmbrftUement of Reactor Vessel Materials.* Therefore, the staffs approval is conditional based on the following criteria: (1) for plants that are similar in core, shroud and downcomer-vessel geometry to that of the Susquehanna and Hope Creek plants, the RAMA methodology can be applied without a bias for the calaJlation of vessel neutron ftuence, (2) for plants (or plant groups) with a diffarent geometry than that of the Susquehanna or Hope Creek plants, a pfant-speclftc application for RPV neutron f1uenca is required to estabUsh the value of a bias, and (3) relevant benchmarking will be required for shroud and reactor Intemals appUcations.

The staff evaluation of the proposed RAMA methodology Is attached. Please contad Meana Khanna of my staff at 301*415-2150 if you have any further questions regarding this subject.

Sincerely, p-eL---/?/~/

William H. Bateman, Chief Materials and Chemical Engineering Branch Division of Engineering Offlce of Nudear Reactor Reguletlon

Enclosure:

As stated cc: BWRVIP Service Ust E4-2

U.S. NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULAIION SAFETY EVALUATION OF BWR VESSEL ANP INTERNALS PROJECT.

SAFETY EVALUATION OF PROpRIETARY EPRI REPORTS. -BWR VESSEL AND INTERNALS PROJECT, BAMA FLUENCE METHODOLOGY MANUAL (BWRVIP-114) "BAMA FLUENCE METHODOLOGY BENCHMARK MANUAL-EVALUATION OF REGULAIORY GUIDE 1.190 BENCHMARK PROBLEMS (BWRVIP-1151." *RAMA FLUENCE MEJHQOOLQGY-SUSQUEHANNA UNIT 2 SURVEILLANCE CAPSULE ELUENCE EVALUATION FOR CYCLES 1-5 (BWRVIP-l17l."*RAMA fLUENCE METHODOLOGY PROCEDURES MANUAL (BWRVIP-1211." AND *HOPE CREEK FLUX WIRE DOSIMETER ACTIVATION EVALUATION FOR cycLE 1 awe-PSE-001-R-OO11*

1.0 INTRODUCTION

1.1 Background By letters dated June 11, 2003, June 28, 2003, August 5, 2003, October 29, 2003, and March 23, 2004, respectively, the Bolling Water Reactor Vessel and Internals Project (BWRVIP) submitted the following Elecbic Power Research Institute (EPRI) proprietary reports for staff review and approval, "BWR Vessel and Internals Project, BAMA Fluence Methodology Manual (BWRVIP-114),* *RAMARuence Methodology Benchmark Manual-Evaluatlon of Regulatory Guide 1.190 Benchmark Problems (BWRVIP-115),* *RAMA Fluence Methodology Susquehanna Unit 2 Surveillance Capsule Fluence Evaluation for Cycles 1-5 (BWRVIP-l17),"

"RAMA Fluence Methodology Procedures Manual (BWRVI P-121),* and "Hope Creek Rux Wire Dosimeter Activation Evaluation for Cycle 1 (TWE-PSE-001-R-OOl )," These reports were supplemented by letter dated September 20, 2004, in response to the staffs request for addlttonal information (RAI) dated April 20, 2004.

The BWRVIP-114 report desaibes the theory of the neutron transport calculation methodology and the uncertainty analysis. The BWRVIP-115 report documents benchmarking of the neutron fluence calculation methodology against two reactor pressure vessel (RPV) simulator measurements, a PWR surveillance capsule measurement and a calculational benchmark. The BWRVIP-117 and TWE*PSE-001-R-OOl reports present plant-specific surveillance capsule neutron fluence benchmark comparisons for the Susquehanna and Hope Creek plants, respectively. The BWRVIP-121 report provides the standard procedures for carrying out neutron fluence calculations using this methodology.

The proposed methodology is essentially a new approach that has been dev~loped by the BWRVlP for determining the fast (E :I: 1.0 MeV) neutron fluence accumulated by the RPV and intemal components of BWR plants. The methodology employs the RAMA computer code for evaluating the neutron flux from the core through the downcomer, vessel internals and through the RPV wall. Iv! important feature of the methodology is that the neutron transport calculation is 3-dlmensional, rather than a synthesis of two 2-dlmensional calculations that is used In the finite differences method on which presently approved methodologies are based. An additional feature of this approach is that the computer modeling of the physical geometry is represented without approximation. The RAMA cade will be applied In the reactor beltllne region deflned by the top and bottom planes of the active fuel and the inside surface of the biological shield. The methodology employs the most recent BUGLE-98 nudear transport and reaction-specific ENCLOSURE E4-3

-2 measurad activity cross section data. The BWRVIP calculation and uncertainty methodology is summarized In Section 2. The technical evaluation is presented in Section 3, and the limitations and conclusions are provided in Section 4.

1.2 Purgose The staff reviewed the reports discussed above to detennine whether the BWRVIP's proposed methodology will provide an acceptable method for detennlnlng the fast (E ;;e 1.0 MeV) neutron nuence accumulated by the RPV and Internal components of BWR plants.

1.3 Regulatory Evaluation The basis for this review is Regulatory Guide (RG) 1.190, "RadiatIOn Embrittlement of Reactor Vessel Materfals.- RG 1.190 Is based on General Design Criterion (GDC) 14.30 and 31. and describes the attrtbutes of neutron transport methodologies which are acceptable to the staff.

The basic feature of an acceptable methodology is that the code is benchmarked by acquiring and evaluating a statistically significant database of measurement-to-calculation ratios and the resulting bias and uncertainty are within certain limits.

2.0

SUMMARY

OF THE EPRI BWRVIP VESSEL NEUTRON FlUENCE METHODOLOGY 2.1 RPV Neutron Fluence Ca!cyIaUpn Methodology The BWRVIP neutron nuence calculational methodology employs the RAMA code to evaluate the neutron flux through the core, vessel Internals, and vessel geometry. The code uses the BUGlE-98 cross-section library to calculate Ihe neutron transport and to detennlne the reactlon-speclftc measured activities. The RAMA code employs a combinatorial geometry method which allows an exact representation of geometrically complex components. This is accomplished by building the desired intemal component using vanous primitive geometry elements (Ref. B).

The neutron transport calculation Is basad on the foRowing: (1) the three-dImensional transport equation Is integrated by attenuating the neutron fluence along discrete rays according to the macroscopic cross-section and optical path in the intersected region, (2) a set of parallel rays are chosen in both a radial and axial plane and the neutron fluance Is detennlned on this grid, (3) to account for the various possible directions of partICle transport, rays are defined on a discrete set of angular quadratures. and (4) anisotropiC scattering Is treated using a legendre expansion of the neutron scattering cross-section.

The neutron soun::e is determined based on the core power density and the region-wise power distribution. The RAMA source accounts for the exposure dependence of the core neutron source and allC1NS tor a detailed pin power description of the sourca distribution. Typically, reftective boundary conditions are applied on the planes that define the angular sector of the geometry being calculated (typically, a core octant or quadrant), and vacuum boundary conditions are applied at the outer radal boundary (e.g., the outside wall of the RPV) and on upper and lower axial boundaries.

In order to fadlltate comparisons of meesurements to calculated values (as instructed by RG 1.190). RAMA calculates the corresponding quantities for the measured reaction rates. RAMA E4-4

-3 determines the time-dependent neutron flux and tracks the target and reaction product nuclides.

The RAMA methodology includes a detailed neutron fluence uncertainty analysis. The parameters making a significant contribution to the neutron fluence calculation uncertainty are identified and RAMA is used to determine numerical sensitivity coefficients for these p.ameters. The uncertainty contribution from these parameters is detennlned by combining the numerical sensitivities with the estimates of the input parameter uncertaintieS. When making comparisons to benchmark measurements, the calculatlon-to-measurement (CIM) differences are combined using a covariance matrix to determine the uncertainty contribution from the measurements. The overall calculation uncertainty and bias are determined based on the CJM differences and the calculation Input parameter uncertainties.

2.2 Calculation of the BpY Benchmarks In validating the RAMA methodology, comparisons of BAMA predictions were performed for the following four benchmarks: (1) the Oak Bidge National Laboratory (OBNL) Pool Crtllcal Assembly (PCA) benchmark experiment (Bef. 9), (2) the VENUS-3 engineering benchmark experiment (Bef. 10). (3) the H. B. Bobinson-2 (HBB-2) BPV benchmark measurement (Bef.

11). and (4) the Brookhaven National Laboratory (BNL) BPV calculation benchmark of NUBEG-6115 (Bef. 12). The PCA and VENUs-a experiments are well-documented RPV mock ups, including high accuracy dosimetry measurements. The PCA core includes twenty-five material test reactor (MTR) curved-plate type fuel assemblies and the simulator geomauy includes a thermal shield. BPV. and void box outside the BPV. The PCA dosimetry measurements were made at positions In front and behind the thermal shield, at locations in front and behind the RPV, and at BpY Intemals locations. The PCA dosimetry measurements include the Np-237 (n, f). U-238 (n, f),ln-115 (n, n'), Ni-58 (n, p) Co-58 and AJ..27 (n. a) Na-24 reactions. The BAMA model is 3-dlmenslonal and Indudes a radial quadrant of the PeA geometry, the full height of the core and the regions above and below the core. Detatled comparisons presented for both the thermal shield (or core shroud) and RPV locations Indicate good agreement with the doSimetry measurements.

The VENUS-3 core consists of twelve 15x15 pressurized water reactor (PWR) fuel assemblies and the simulator geometry Includes the baffle. core barrel. neutron pad and RPV simulator.

The VENUS-3 dOSimetry measurements Include the NI*58 (n, p) Co-58,10-115 (n, n'), and AJ..Z7 (n. a) Na-24 reactions. The BAMA model Is 3-dimenslonal and Includes a radial quadrant of the simulator geometry. the full height of the core, and the regions above and below the core.

Detailed comparisons are presented for the core. bafffe, and core barrel and Indicate good agreement with the measurements.

The HBR-2 benchmark experiment provides a wel!-documented set of dosimetry measurements for a full-height operating PWR. Including core barrel, thermal shield and RPV. The HBR-2 dosimetry measurements include Np-237 (n, f), U-238 (n. f), Ni-58 (n. p) Co-58, Fe-54 (n. p) Mn

54. Ti-46 (n, p) Sc-46 and Cu-63 (n. a) C0-60. The measurements were made at an in-vessel capsule and at a cavity location. The HBR*2 BAMA. model is 3-dirnenslonal and provides a detailed I'8presentation of an octant of the problem geometry for a centrally-located axial region of the core. The model extends from the center of the core out to the outer surface of the biological shield. Detaiied comp.isons are presented for both the In-vessel surveillance capsule and the cavity measurements, and Indicate good agreement with the measured data.

E4 5

-4 BNL NUREG-6115 provides the detailed specification and corresponding numerical solutions for a BWR RPV neutron fluence benchmark problem. The benchmark problem provides a reference calculation for a configuration that is typical of an operating BWR which includes the downcomer and RPV neutron nuences and the dosimeter response at an in-vessel surveHlance capsule. The surveillance capsule dosimetry includes the Np-237 (n, f). U-238 (n. f). NI-58 (n, p) Ccr58, Fe-54 (n, p) Mn-54. Ti-46 (n, p) Sc-46, and Cu-63 (n, a) Co-6O reaction rates. The RAMA model Is 3-dlmenslonal and provides a detailed representation of an octant of the problem geometry over an axial region that includes the core as well as the regions above and below the core. The model extends from the center of the core out to the outer surface of the biological shield. Detailed comparisons are presented for both the RPV neutron fluences and the dosimetry reaction rates. The surveillance capsule comparisons Indicate good agreement for all reaction rates. The downcomer and RPV neutron f1uence comparisons Indicate that RAMA is conservative relative to the reference solution.

2.3 Calculation oUbe Sysquehanna Neutron Buence Measyrements As part of the RAMA plant-spedflc qualification, RAMA transport calculations have been performed for the Susquehanna Unit 2 surveillance capsule that was removed at the end of Cycle 5. In order to validate the fast (E ~ 1.0 MeV) neutron fluence evaluations of the Susquehanna RPV, comparisons of the calculated and measured neutron fluence have been made to determine the neutron nuence calculational uncertainty and to identify any systematic bias in the neutron f1uence predictions. The Cyde 5 surveillance capsule was located in the downcomer, radially at a pOSition close to the innerwall of the RPV, and azimuthally 30* from the core flats. The surveillance capsule Included three each of the following dosimeter wires:

copper. nickel, and Iron. The measured activities included the Cu-63 (n, a) C0-60. Ni-58 (n, p)

Co-58, and Fe-54 (n, p) Mn-54 dosimetry reactions. The measurements were of high quality and were reported to have uncertainties on the order of a few percent The RAMA calculational model was based on detailed plant data provided by the Pennsylvania Power and Light (PPL) Company. The geometry data were taken from plant drawings and used to model the surveillance capsule and various core, core shroud, jet pump/riser and RPV components. RAMA provided a geometry model of high accuracy in which both the Cartesian geometry of the core boundary and the cylindrical geometry of the Jet pump/riser components were represented without approximation. The RAMA modellnduded a one-eighth (45°)

azimuthal sector and the radial geometry from the center of the core out to the inner wall of the biological shield.

The cora neutron source was based on the Susquehanna Cycles 1-5 operating history.

Three-dimensional power, void and exposure distributions were constructed from the plant operating history files. The pin-wise gradient and exposure dependence of the neutron source for the fuel assemblies on the core periphery were Included. Each cycle was desaibed by a representative set of operating state-points. Tho neutron f1uence accumulated by the capsule dosimeters was E4-6

-5 determined by an appropriate weighting of the RAMA state-point calculations. An extensive set of sensitivity calculations was also performed to ensure the stability and convergence of the numerical solution.

RAMA calculations of the dosimeter activities were petformed and compared with the measurements (dpslg). The average CIM overall measurement was found to be very close to unity Indicating that there Is no Significant bias in the RAMA neutron lIuence predictions. The standard deviation of all CIM values was less than 20% as recommended In RG 1.190 (Section 1.4.3). In order to provide an independent assessment of the accuracy of the RAMA neutron fluence prediction. a detailed analytic uncertainty analysis was also performed. The Important Input parameter uncertainties were Identified and an estimate of the uncertainty in each parameter was detennined. The uncertainty in each parameter was propagated through the RAMA calculation using numerical sensitivity calculations. The resultant analytic estimate of the RAMA neutron fluence calculation uncertainty, corresponding to the observed CIM standard deviation. was also shown to be less than 20"0, 2.4 Calculation of the Hope Creek Neutron Fluence Measurements RAMA transport calculations were perfonned for the surveillance capsule removed from the Hope Creek RPV at the end of the fil'8t cycle. In order to validate the fast (E ~ 1.0 MeV) neutron fluence evaluations of the RPV. comparisons of the calculated and measured neutron fluence have been made to determine the neutron fluence"<:alculational uncertainty and to identify any systematic bias in the neutron fiuence predictions. The first cycle surveillance capsule was located in the downcorner, radially at a position close to the innerwall of the RPV.

and azimuthally at 33- from the core flats. It is noted that two additional capsules are located at 121- and 299-. The surveillance capsule included three copper and three Iron flux wires. The measured activities included the Cu-63 (n. a) Co-60 and Fe-54 (n. p) Mn-54 dosimetry reactions. 'rhe measurements were reported to have uncertainties on the order of a few percent. The copper activity was corrected for the presence of Co*59 Impurity of about 0.25 parts per million (ppm).

The RAMA calculational model was based on detailed plant data. The geometry data were taken from plant drawings and used to model the surveillance capsule. the core, core shroud.

jet pump/rtser. and RPV components. RAMA provided a geometry model of high accuracy in which both the Cartesian geometry of the core boundary and the cylindrical geometJy of the jet pump/riser components were represented without approximation. The RAMA model included a one-eighth (45-) azimuthal sector and the radial geometry from the center of the core to the biological shield.

The core neutron source was based on the first cycle's operating history. Three-dimensional power. void. and exposure distributions were constructed from the plant operating history fiies.

The pin-wise gradient and exposure dependence of the neutron source for the fuel assemblies on the core periphery were included. The neutron fluence accumulated by the capsule dosimeters was determined by an appropriate weighting of the RAMA state-point calculations.

An extensive set of sensitivity calculations was also perbrmed to ensure the stability and convergence of the numerical solution.

RAMA calculations of the dosimeter activities were perfonned and compared with the measurements (dpslgm). The average ClM overall measurement was found to be very dose to E4 7

unity indicating that there is no significant bias in the RAMA neutron fluence predictions. The standard deviation of all CIM values was less than 20% as recommended In RG 1.190 (Section 1.4.3). In order to provide an Independent assessment of the acCuracy of the RAMA neutron fluence prediction. a detailed analytical uncertainty analysis was also perfonned. The Impor\ant input parameter uncertainties were Identified and an estimate of the uncertainty in each parameter was determined. The uncertainty In each parameter was propagated through the RAMA calculatlon using numertcal sensitivity calculations. The resultant analytical esUmate of the RAMA neutron fluence calculation uncertainty. corresponding to the Observed CIM standard deviation, was also shown to be less than 20%.

3.0 TECHNICAL EVALUATION

The staff's review of the BWRVIP neutron fluence methodology focused on the details of the application of the neutron auence calculation methodology and the qualification of the methodology provided by the benctvnark comparisons and the plant~specific CIM database.

3.1 Rev Nautron Fluence Calculation Methodology In tha RAMA transport calculation, the neutron flux Is determined by summing the contributions from a set of particle ray tracings through the problam geometry. Tha accuracy of this technique dapends on the specific problam geometry, as well as the number and distribution of the rays used to track the neutrons through tha geometry. In addition, the components that are associated with the problem geometry are represented with a discrete set of spatial regions (I.a., a spatial mash). Because the neutron nux Is averaged over these regions, a mesh--related uncertainty is Introduced Into the calculation. Sinca both of these numerical uncertainties are sensitive to the problem geometry. they require an evaluation that accounts for the geometry.

By letter dated April 20, 2004, the staff requested that the BWRVIP address the specific tests and critaria used to assure tha adequacy of the number of rays and volumes used in the RAMA neutron nuence calculations for plant-specific applications. By letter dated September 29, 2004, the BWRvtP indicated that In plant-specffic modal applications of the RAMA nuance methodology, numerical sensitivity calculations win be perfonned to assure the adequacy of the number of partlcla tracking rays and the number of volumes used to represent component geometry In the RAMA neutron fluence evaluations. The staff found this approach acceptable.

The RAMA geometry model represents the individual components and regions of the problem geometry USing a library of pre-calculated geometry elements. The modeling of the reflector region surrounding the core Is partlculaliy complicated In that it Involves geometry elements that have both planar and cylindrical side boundali.,. However. RAMA pr'CMdes an exact representation of the true geomatry (i.e., preserves the exact location, orIantation and shape of all surfaces deftnlng the physical geometry). For example, In the case of these reflector regions, the BWRvtP Indicated in its letter dated September 29, 2004, that the geometry modal allows for complex geometries, including the transition between the rectangular core and the cylindrical core shroud, to be pracisely represented.

The RAMA code has the necessary mechanisms for geometrical representation. neutron scattering and neutron transport approximations. Therefore. the staff finds the RAMA code acceptable. based on its slructural features.

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-7 3.2 Calculation of the RPV Benchmarks The RPV benchmark calculations are performed to evaluate the accuracy of RAMA and to identify any systematic bias in the proposed licensing methodology. In order for the benchmark comparisons to reflect the difference between the benchmark and the proposed methodology, the methods used in the benchmark calculations must be the same as the proposed licensing methods. By letter dated April 20, 2004. the staff requested that the BWRVIP identify the differences between the methods used in performing the RAMA benchmark analyses in the BWRVIP-115 report and the methods that will be used in performing the calculatfons of the RPV and core shroud neutron ftuence. By letter dated September 29, 2004, the BWRVIP Indicated that the methods used in performing the RAMA benchmark analyses are the same as the methods that will be used in performing BWR RPV and core shroud neutron fluence calculations. The staff found this acceptable in that there would be no inconsistencies in the methods used.

The BWRVIP-115, BWRVIP-117, and TWE-PSE-001-R-001 reports present the RAMAanalysls of a set of simulator calculations and operating reactor benchmarks which provide the basis of the Susquehanna and Hope Creek applications of the RAMA neutron f1uence methodology.

However, it Is expected that as additional surveillance capsules are removed, new benchmark CIM data will become available. RG 1.190 requires that as new measurements become available, they shall be incorporated into the CIM database and the neutron fluence calculational bias and uncertainty estimates shall be updated as necessary.

By letter dated April 20, 2004, the staff requested that the BWRVIP address how it will ensure that new measurements are incorporated In the ClM database and that the neutron ftuence bias and uncertainty will be updated in a timely manner. In its response by letter dated September 29, 2004. the BWRVIP stated that comparisons to measured surveillance capsule and benchmark dosimetry are maintained in a database that is updated as additional plant capsule evaluations are performed using the RAMA methodology. In addition. the BWRVIP stated that currently. TransWare Enterprises, Inc. (a primary contractor to the BWRVIP) maintains a surveillance capsule and benchmark dosimetry measurement database. The BWRVIP further stated that it would consider options of establishing a mechanism to collect and evaluate new CIM data. Based on the above, the staff found the BWRVIP's response acceptable.

111e staff's review of this section established that the RAMA methodology is applied to the benchmarks in the same manner (apprOximations, cross-sections, etc.) as applied in plant-specific applications, therefore. the staff Is In agreement that if a bias exists In the proposed code. It should appear in the benchmarks.

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-8 3.3 Results of the Susquehanna Dosimetry Measurements The Susquehanna, Unit 2 surveillance capsule contained three of each of the following dosimeter wires; copper, iron and nickel. The RAMA calculated ratios and the corresponding measured specific activity (dpslg) CIM ratios are close to unity and display very good agreement The individual ratios are wfHl within the 20% limit specified In RG 1.190. In addition, the standard deviation is just a few percent In accordance with the guidance in RG 1.190, the BWRVIP-117 report includes an analytical neutron f1uence uncertainty analysis. This analysis is important since it provides an independent estimate of the plant-specific Susquehanna RAMA neutron f1uence calculational uncertainty. The uncertainty analySis requires that estimates of the major components of the uncertainty be determined and the uncertainties be propagated through the RAMA neutron nuence calculation. The uncertainty propagation Is performed using numerical component sensitivity as calculated by RAMA. The important uncertainty components have been identified and include the following: (1) capsule and flux wire locations, (2) RPV inner radius, (3) core void fraction, (4) peripheral bundle power, and (5) iron cross-sections. In order to make an accurate determination of the RAMA uncertainty, reliable estimates of the component uncertainties are required.

By letter dated April 20, 2004, the staff requested that the BWRVIP discuss the basis for the parameter uncertainty for the componentsllocations listed above. In its letter dated September 29, 2004. the BWRVIP Indicated that the uncertainty estimates for these components/locations Is based on the following: (1) as-bullt measurements, (2) design drawing tolerances, (3) experience estimates of :i:5% variation in computed void fraction. (4) reported aCQJracy of core simulation analysis, and (5) experience estimates of :i:5% In the cross section, respectively.

In addition, the staff noted that Table 5-3 of the BWRVIP-117 report provided the values of the calculated bias and total uncertainty. The BWRVIP also displayed the calculation of the total uncertainty and bias from the elM and the analytic uncertainty with weighting factors Inversely proportional to the analytic and CIM variances In the BWRVIP-117 report. The staff finds the BWRVIP's response to the staffs request for addltlonallnfonnatlon and the values of the bias and uncertainty, as provided in the BWRVIP-117 report, acceptable because the values are well within the limits set forth in RG 1.190.

3.4 Results of the HODe Creek Dosimetry Measurements The Hope Creek surveillance capsule contained three copper dOSimeter wires and three iron dosimeter wires. The surveillance capsule was irradiated during the first cycle for 377.9 effective full power days. The RAMA code calculated the specific dosimeter activity to the corresponding meas....ed specific activity (dps/g). The CIM ratios are close to unity and displayed very good agreement. The individual dosimeter ratios are weg within the 20% limit. as specified in RG 1.190, and the standard deviation is Just a few percent However, it was noted that unlike the Susquehanna case, the Hope Creek calculaUon does not indude an analytical uncertainty and bias calculation.

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4.0 CONCLUSION

4.1 BWR RPV Neutron Fluence Based on the staffs review of the BWRVIP*114, *115. -117. and -121 reports. the TWE-PSE-001-R-OO1 report. and the supporting documentation, the staff concludes that the BWRVIP methodology. as described in these reports, provides an acceptable best-estimate plant-specific prediction of the fast (E :t 1.0 MeV) neutron fluence for BWR RPVs. This acceptance is limited to the axial region defined by the core active fuel height. The best estimate RPV neutron fluence prediction is determined using the RAMA transport code, detailed plant-specifIC geometry. core operating history. and the BUGLE-96 nuclear data library with a minimum of a P3 Legendre polynomial approxlmaUon in the iron Inelastic scattering.

With respect to the calculation of BWR RPV neutron f1uence, the staff concludes that based on the plant-specific benchmark data presently available, no calculational bias Is required for the application of the methodology to plants of similar geometrical design to Susquehanna and Hope Creek. I.e., BWR-IV plants. However. in order to provide continued confidence In the proposed neutron fluenca methodology for the BWR RPVs. the acceptance of this methodology is subject to the following conditions for plants which do not have geometries similar to the cHed BWR-IV's:

• To apply the RAMA methodology to plant groups which have geomemes that are different than the cited BWR-IV's. at least one plant-speciflc capsule dosimetry analysis must be provided to quantify the potential presence of a bias and assure that the uncertainty is within the RG 1.190 limits and

• Justification Is necessary for a specific application based on geometrical similarity to an analyzed core, core shroud. and RPV geometry. That is, a licensee who wishes to apply the RAMA methodology for the calculation of RPV neutron fluence must reference. or provide, an analySis of at least one surveillance capsule from a RPV with a similar geomeby.

4.2 Reactor Internals EPRl's stated objective for this submittal included neutron fluence calculations for reactor internals. Neutron ftuence values for reactor intemal components are used to either quantify irradiation assisted stress corrosion cracking (IASCC) susceptibility, or to quantify helkJm formation which could affect the weldability of reactor intemals components. IASCC depends on fast (E ~ 1.0 MeV) neutron ftuence, while helium formation is a function of thermal.

epithermal. and fast neutron fluence. The calc ulational accuracy requirements for reactor Internals are not the same as those for the RPV. and are not covered by the guidance in RG 1.190. In addition. the submittal does not include any bend1ma~ing for reactor internals' neutroo fluence calculations. Therefore. the staff will review quaNfication of RAMA for reactor Internals applications on a case-by-case basis, based on consideration of CIM vaJues and the associated accuracy requirements.

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-10 Licensees who wish to use the RAMA methodology for the calculation of neutron fluence at reactor internals locations must reference, or provide, an analysiS which adequately benchmarks the use of the RAMA methodology for uncertainty and caloolational bias based on the consideration ot. (1) the location at which the neutron nuence is being calculated, (2) the geometry of the reactor. and. (3) the accuracy required for the application. In addition, If a licensee qualifies RAMA for calculating, for example. helium generation at one location (e.g.,

the core shroud), this qualifies RAMA for the same reactor and purpose at other reactor Internals locations (e.g., at the location of the Jet pumps).

4.3 Assembling a Statistically Significant PatabUI EPRI stated that efforts are underway to assemble a database which will enable the staff to remove any limitations placed on the use of the RAMA methodology. For such an effort to be successful, the staff expects that the neutron fluence uncertainty analysis and detennlnation of the caloolatlonal bias for the relevant fleet of plants wMI be updated. as additional measurements are taken and as additional data become available. The results of the updated analysis.

including the CIM ratios, should be submitted to the stall for review and approval.

5.0 REFERENCES

1. Letter from C. Terry. Electric Power Research Institute to US Nudear Regulatory Commission "Projed No. 704* BWRVIP*114: BWR Vessel and Internals Project, RAMA Fluence Methodology Manual" June 11, 2003.

2. BWRVIP-115, -BWR Vessel and Internals Projed. RAMA Fluence Methodology Benchmark Manual. Evaluation of Regulatory Guide 1.190 Benchmark Problems," EPRI, Palo Alto, CA: June 26, 2003.

3. BWRVIP-117. "BWR Vessel and Internals Project. RAMA Fluence Methodology Susquehanna Unit 2 Surveillance Capsule Fluence Evaluation for Cycles 1 - 5.* July 2003.

4. BWRVIP-121, -BWRVIP Vessel and Internals Project RAMA FIU8nce Methodology Procedures Manual," August 2Q03.

5. Letter from W. A. Eaton: Hope Creek Flux Wire Dosimeter Activation Evaluation for Cycle 1 Using the RAMA Fluence Methodology, March 23.2004.

6. J. E. White. D. T. Ingersoll, C. O. Slater, R. W. Roussin. -BUGLE-96: A Revised multigroup Cross-section Library for LWR AppllcatJons Based on ENDF/B-VI Release 3, presented at the American Nuclear Society Radiation & Shielding Topical Meeting, April 21-25. 1996, Falmouth, MA, April 1996.

7. Office of Nuclear Regulatory Research, *Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence. Regulatory Guide 1.19Q. U.S. Nuclear It Regulatory Commission. April 2001.

8. J. T. West. M. B. Emmett. "MARS: A Multiple Array System Using Combinatorial Geometty,* Oak Ridge National Laboratory, Radiation Shielding Information Center Report. December 1980.

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9. W. N. Mcelroy, EdHor, "LWR Pressure Vessel Surveillance Dosimetry Improvement Program: PCA Experiments and Blind Test," NLiREGlCR-1861 (Hanford Engineering Development Laboratory, HEOL-TME 80-87). July 1981.

10. P. O'hondt, et al., "Contributions of the Venus-Engineering Mock-Up Experiments to the LWR-PV Surveillance, 'proceedings of the 7th ASTM-Euratom Sympgsjum on Reactor DOSimetry,'* Strasbourg. France. August 1990.

11. I. Remee and F. B. K. Kam. "H. B. Robinson-2 Pressure Vessel Benchmark," NUREGJCR 6453 (Oak Ridge National Laboratory, ORNUTM-13204). February 1998.

12. "Pressure Vessel F1uence Calculation Benchmark Problems and Solutions," NUREGJCR*

6115 (BNL NUREG-52395), Brookhaven National laboratory* September 2001.

13. Letter from S. Coffin, US Nuclear Regulatory Commission to B. Eaton, Chairman BWRVIP, "Request for Additional Information - Review of BWR Vessel and Intemals Project Reports, BWRVIP*114. BWRVIP-115, BWRVIP-117, and BWRVIP-121, and TransWare Enterprises Inc. Report TWE-PSE-001-R-001, Revision 0," April 20, 2004.

14. letter from WA Eaton to US Nuclear Regulatory Commission Document Control Desk

• ProJect No. 704

• BWRVIP Response to NRC Request for Addltlonallnfonnation on BWRVlP-114. -115.117, and -121,D September 29, 2004.

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