PLA-5498, Proposed License Amendment Re Revision to the RPV Material Surveillance Program PLA-5498 for Susquehanna Steam Electric Station

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Proposed License Amendment Re Revision to the RPV Material Surveillance Program PLA-5498 for Susquehanna Steam Electric Station
ML022130555
Person / Time
Site: Susquehanna  Talen Energy icon.png
Issue date: 07/25/2002
From: Richard Anderson
Susquehanna
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
PLA-5498
Download: ML022130555 (19)
v • d • e


Text

S I j R. L. Anderson PPL Susquehanna, LLC Vice President - Nuclear Operations 769 Salem Boulevard Berwick, PA 18603 Tel. 570.542.3883 Fax 570.542-1504 rlandersonr@pplweb.com P '.t I I JUL 2 5 2002 STM U. S. Nuclear Regulatory Commission Attn.: Document Control Desk Mail Stop OP 1-17 Washington, DC 20555 SUSQUEHANNA STEAM ELECTRIC STATION PROPOSED AMENDMENT NO. 247 TO UNIT 1 LICENSE NO. NPF-14 AND PROPOSED AMENDMENT NO. 212 TO UNIT 2 LICENSE NO. NPF-22: REVISION TO THE RPV MATERIAL SURVEILLANCE PROGRAM Docket Nos. 50-387 PLA-5498 and 50-388

Reference:

1. Letter from W. H Bateman (USNRC) to C. Terry (BWR VIP Chairman) titled, "Safety Evaluation Regarding EPRI ProprietaryReport 'BWR Vessel and Internals Project, BWR IntegratedSurveillance ProgramPlan (BWRVIP-78)'and 'BWRVIP-86: BWR Vessel and Internals Project, BWR IntegratedSurveillance Program Implementation Plan,"' dated February1, 2002.
2. Regulatory Issue Summary No. 2002-05, "NRC Approval of Boiling Water Reactor Pressure Vessel IntegratedSurveillance Program," datedApril 8, 2002.

The purpose of this letter is to propose changes to the Susquehanna Steam Electric Station Final Safety Analysis Report (Susquehanna SES FSAR) for Unit 1 and Unit 2. This proposed change revises the Reactor Pressure Vessel Material Surveillance Program in accordance with References 1 and 2.

Attachment 1 to this letter is the "Safety Assessment" supporting this change.

Attachment 2 is the No Significant Hazards Considerations evaluation performed in accordance with the criteria of 10 CFR 50.92 and the Environmental Assessment.

Attachment 3 to this letter contains the applicable pages of the Susquehanna SES FSAR for Unit 1 and Unit 2, marked to show the proposed change.

Document Control Desk PLA-5498 The proposed change has been approved by the Susquehanna SES Plant Operations Review Committee and reviewed by the Susquehanna Review Committee.

Consistent with the process established between the NRC and the BWRVIP, this change is being processed as a license amendment to facilitate NRC review and approval.

PPL plans to implement the proposed changes in the Spring of 2003 to support deletion of work from the Unit 2 1 1 th Refueling and Inspection Outage. Therefore, we request NRC complete its review of this change by December 1, 2002 with the changes effective 30 days after approval.

Any questions regarding this request should be directed to Mr. Cornelius T. Coddington at (610) 774-4019.

Sincerely, R. L. Anderson Attachments: (1) Safety Assessment - Revision to the Reactor Pressure Vessel Material Surveillance Program (2) No Significant Hazards Considerations and Environmental Assessment (3) Final Safety Analysis Report Mark-Ups copy: NRC Region I Mr. D. J. Allard, PA DEP Mr. T. G. Colburn, NRC Sr. Project Manager Mr. S. L. Hansell, NRC Sr. Resident Inspector Mr. R. Janati, DEP/BRP Mr. E. M. Thomas, NRC Project Manager

BEFORE THE BEFORE THE UNITED STATES NUCLEAR REGULATORY COMMISSION In the Matter of PPL Susquehanna, LLC: Docket No. 50-388 PROPOSED AMENDMENT NO. 212 TO LICENSE NPF-22:

REVISION TO THE REACTOR PRESSURE VESSEL MATERIAL SURVEILLANCE PROGRAM UNIT NO. 2 Licensee, PPL Susquehanna, LLC, hereby files a revision to its Facility Operating License No. NPF-22 dated March 23, 1984.

This amendment involves a revision to the Susquehanna SES Final Safety Analysis Report.

PPL Susquehanna, LLC By:

R. L. Anderson Vice President - Nuclear Operations Sworn to and subscribed before me this dS'rday of ,2002.

Notarial Seal Nancy L.Garcia, Notary Public Salem Twp., Luzerne County My Commission Expires May 31, 2003 Member, Pennsylven' ssoc-atiin it Notaries Notary Public

BEFORE THE BEFORE THE UNITED STATES NUCLEAR REGULATORY COMMISSION In the Matter of PPL Susquehanna, LLC: Docket No. 50-3 87 PROPOSED AMENDMENT NO. 247 TO LICENSE NPF-14:

REVISION TO THE REACTOR PRESSURE VESSEL MATERIAL SURVEILLANCE PROGRAM UNIT NO. 1 Licensee, PPL Susquehanna, LLC, hereby files a revision to its Facility Operating License No. NPF-14 dated July 17, 1982.

This amendment involves a revision to the Susquehanna SES Final Safety Analysis Report Specifications.

PPL Susquehanna, LLC By:

R. L. Anderson Vice President - Nuclear Operations Sworn to and subscribed before me this 0 2.*day of 4 2002.

E NNotarial Seal SallemL.

Nancy Garcia, Twp., Luzrerne U.onty .. Ic NotaPrPubf.

2003 My commission Expires May 31, Membe" Pp*flsvh'anfl tssodiation ot Notaries Notary Public

Attachment 1 to PLA-5498 Safety Assessment Revision to the Reactor Pressure Vessel Material Surveillance Program

Attachment 1 to PLA-5498 Page 1 of 3 Safety Assessment Revision to the Reactor Pressure Vessel Material Surveillance Program The following provides the basis for the proposed revision to the reactor pressure vessel material surveillance program.

1.0 DESCRIPTION

OF THE PROPOSED CHANGE PPL Susquehanna, LLC (PPL) proposes to revise the licensing basis for Susquehanna Steam Electric Station Units 1 and 2 (SSES) by replacing the current plant-specific reactor pressure vessel (RPV) material surveillance program with the Boiling Water Reactor (BWR) Integrated Surveillance Program (ISP), which was approved by the NRC in its Safety Evaluation (SE) dated February 1, 2002 (Reference 1). The proposed revision to the SSES Final Safety Analysis Report reflecting this change is provided for information in Attachment 3.

2.0 REASON FOR THE PROPOSED CHANGE The BWR ISP was developed in response to an issue raised by the NRC staff regarding the potential lack of adequate unirradiated baseline Charpy V-notch (CVN) data for one or more materials in plant-specific RPV surveillance programs at several BWRs. The lack of baseline properties would inhibit a licensee's ability to effectively monitor changes in the fracture toughness properties of RPV materials in accordance with Appendix H to 10 CFR 50. The BWR ISP, as approved by the NRC, resolves this issue.

Implementation of the ISP will provide additional benefits. When the original surveillance materials were selected for plant-specific surveillance programs, the state of knowledge concerning RPV material response to irradiation and post irradiation fracture toughness was not the same as it is today. As a result, many facilities did not include what would be identified today as the plant's limiting RPV materials in their surveillance programs. Hence, this effort to identify and evaluate materials from other BWRs, which may better represent a facility's limiting materials, should improve the overall evaluation of BWR RPV embrittlement. Second, the inclusion of data from the testing of BWR Owners' Group (BWROG) Supplemental Surveillance Program (SSP) capsules will

Attachment 1 to PLA-5498 Page 2 of 3 improve the overall quality of the data being used to evaluate BWR RPV embrittlement. Finally, implementation of the ISP is also expected to reduce the cost of surveillance testing and analysis since surveillance materials that are of little or no value (either because they lack adequate unirradiated baseline CVN data or because they are not the best representative materials) will no longer be tested.

3.0 TECHNICAL ANALYSIS

Reference 1 concludes that the proposed ISP, if implemented in accordance with the conditions in the SE, has been determined to be an acceptable alternative to all existing BWR plant-specific RPV surveillance programs for the purpose of maintaining compliance with the requirements of Appendix H to 10 CFR Part 50 through the end of current facility 40 year operating licenses. Reference 1 requires that each licensee (1) provide information regarding what specific neutron fluence methodology will be implemented as part of participation in the ISP and (2) address the neutron fluence methodology compatibility issue as it applies to the comparison of neutron fluences calculated for its RPV versus the neutron fluences calculated for surveillance capsules in the ISP which are designated to represent its RPV. This information is provided in the following discussion.

The SSES Technical Specifications, as discussed in Amendment No. 200 to SSES Unit 1 Operating License (NPF-14) and Amendment No. 174 to SSES Unit 2 Operating License (NPF-22) require that new P-T curves be implemented based on updated fluence calculations by May 1, 2005 and May 1, 2006 (Unit 2 and Unit 1 respectively). See Reference 2 for additional information.

PPL intends to use the BWRVIP RAMA code or other NRC approved methodology to revise the calculations for both Units 1 and 2. The RAMA code will perform a full 3D-neutron transport solution to determine fluence within the vessel. The analysis will use the BUGLE-96 data library as recommended by Regulatory Guide 1.190. It will perform a full uncertainty analysis to determine the accuracy of the calculation.

The current schedule for completion of the BWRVIP RAMA code is December 2002. The BWRVIP intends to submit a topical report on the RAMA code to the NRC for review, with the objective of receiving a safety evaluation in 2003 approving use of the methodology.

Attachment 1 to PLA-5498 Page 3 of 3 The first surveillance capsule to be tested under the ISP is the River Bend 183' capsule. The test report is scheduled to be submitted to the NRC by February 2003. Coincidentally, these capsules, according to the ISP, are the substitute capsules for SSES Unit 2. Thus in accordance with the ISP, the SSES Unit 2 capsule will not be removed and tested.

The ISP requires the Unit 1 surveillance capsules be removed in 2012 and tested in 2013. The Unit 1 fluence calculations will be reevaluated both in 2006 and after this ISP testing.

REFERENCES:

1. Letter from W. H. Bateman (USNRC) to C. Terry (BWRVIP Chairman) titled, "Safety Evaluation Regarding EPRI Proprietary Report 'BWR Vessel and Internals Project, BWR Integrated Surveillance Program Plan (BWRVIP-78)'

and 'BWRVIP-86: BWR Vessel and Internals Project, BWR Integrated Surveillance Program Implementation Plan,"' dated February 1, 2002.

2. Letter from D. S. Collins (USNRC to R. G. Byram (PPL) titled, "Susquehanna Steam Electric Station Units 1 and 2 - Issuance of Amendment RE: Reactor Pressure Vessel Pressure-Temperature Limit Curves," dated February 7, 2002.

Attachment 2 to PLA-5498 No Significant Hazards Considerations and Environmental Assessment

Attachment 2 to PLA-5498 Page 1 of2 No Significant Hazards Considerations and Environmental Assessment The Commission has provided standards in 10 CFR 50.92(c) for determining whether a significant hazards consideration exists. A proposed amendment to an operating license for a facility involves no significant hazards consideration if operation of the facility in accordance with the proposed amendment would not (1) involve a significant increase in the probability or consequences of an accident previously evaluated; (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety.

PPL proposes to revise the licensing basis for SSES by replacing the plant-specific RPV material surveillance program with the BWR ISP. This change is acceptable because the BWR ISP has been approved by the NRC staff as meeting the requirements of paragraph III.C of Appendix H to 10 CFR 50 for an integrated surveillance program.

In accordance with the criteria set forth in 10 CFR 50.92, PPL has evaluated the proposed TS change and determined it does not represent a significant hazards consideration. The following is provided in support of this conclusion.

1. Does the proposedchange involve a significantincreasein the probability or consequences of an accidentpreviously evaluated?

No. The proposed change implements an integrated surveillance program that has been evaluated by the NRC staff as meeting the requirements of paragraph III.C of Appendix H to 10 CFR 50. Consequently, the proposed change does not significantly increase the probability of any accident previously evaluated. The proposed change provides the same assurance of RPV integrity. As a result, the consequences of any accident previously evaluated are not significantly increased.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposedchange create the possibility of a new or different kind of accidentfrom any accidentpreviously evaluated?

No. The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed) or changes in methods governing normal plant operation. The proposed change maintains an equivalent level of RPV material surveillance and does not introduce any new accident initiators. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

Attachment 2 to PLA-5498 Page 2 of 2

3. Does the proposedchange involve a significant reduction in a margin of safety?

No. The proposed change has been evaluated as providing an acceptable alternative to the plant-specific RPV material surveillance program that meets the requirements of the regulations for RPV material surveillance. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

ENVIRONMENTAL CONSIDERATION 10 CFR 51.22(c)(9) identifies certain licensing and regulatory actions, which are eligible for categorical exclusion from the requirement to perform an environmental assessment.

A proposed amendment to an operating license for a facility does not require an environmental assessment if operation of the facility in accordance with the proposed amendment would not (1) involve a significant hazards consideration; (2) result in a significant change in the types or significant increase in the amounts of any effluents that may be released offsite; or (3) result in a significant increase in individual or cumulative occupational radiation exposure. PPL has evaluated the proposed change and has determined that the proposed change meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Accordingly, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment needs to be prepared in connection with issuance of the amendment. The basis for this determination, using the above criteria, follows:

Basis

1. As demonstrated in the No Significant Hazards Consideration Evaluation, the proposed amendment does not involve a significant hazards consideration.
2. There is no significant change in the types or significant increase in the amounts of any effluents that may be released offsite. The proposed change does not involve any physical alteration of the plant (no new or different type of equipment will be installed) or change in methods governing normal plant operation.

There is no significant increase in individual or cumulative occupational radiation exposure. The proposed change does not involve any physical alteration of the plant (no new or different type of equipment will be installed) or change in methods governing normal plant operation.

Attachment 3 to PLA-5498 Final Safety Analysis Report Mark-Ups

SSES-FSAR NIMS Rev. 55 5.3.1.5.1.7 Reactor Vessel Annealing In-place annealing of the reactor vessel because of radiation embrittlement is unnecessary because the predicted end of life value of adjusted reference temperature will not exceed 200OF (see 10 CFR 50, Appendix G, Paragraph IV.C).

5.3.1.6 Material Surveillance 5.3.1.6.1 Compliance with "Reactor Vessel Material Surveillance Program Requirements" The materials surveillance program monitors changes in the fracture toughness properties of ferritic materials in the reactor vessel beltline region resulting from their exposure to neutron irradiation and thermal environment.

Materials for the program are selected to represent materials used in the reactor beltline region. The specimens are manufactured from a plate actually used in the beltline region and a weld typical of those in the beltline region and thus represent base metal, weld metal, and the transition zone between base metal and weld. The plate and weld are heat treated in a manner which simulates the actual heat treatment performed on the core region shell plates of the completed vessel.

The surveillance program includes three capsule holders per reactor vessel. Charov impact specimens for the reactor vessel surveillance programs are of the longitudinal orientation consistent with the ASME requirements prior to the issuance of the Summer 1972 Addenda and ASTM-E-185-82. Based on GE experience, the amount of shift measured by these irradiated longitudinal test specimens will be essentially the same as the shift in an equivalent transverse specimen.

The program for implementation of the scheduling and testing of the surveillance specimens is governed and controlled by BWRVIP-86, BWR Integrated Surveillance Program (ISP) Implementation Plan. The Unit 1 second holder (131C7717G2) will be pulled in accordance with the schedule in BWRVIP-86. For Unit 2. all the information will come from other plants in BWRVIP-86 ISP Program. No cagsules are scheduled to be withdrawn from Unit 2. Other plants will remove and test specimens in accordance with BWRVIP-86. The results from these tests will provide the necessary data to monitor embrittlement for Unit 2. Since the predicted adiusted reference temperature of the reactor vessel beltline steel is less than 100OF at end of life, the use of the capsules per BWRVIP-86 meets the requirements of 10 CFR 50, Appendix H. and ASTM-E-185-82. The withdrawal schedule and other requirements are provided in BWRVIP-86.

For the extent of compliance to 10 CFR 50, Appendix H, see Tables 5.3-1 b and 5.3-2b.

Rev. 55 5.3-8

,SSES-FSAR NIMS Rev. 55 Each holder is loaded with capsules which contain the following surveillance specimens and dosimeter wires:

First holder (131 C7717G3):

36 Charpy impact specimens including 12 base metal, 12 weld metal, and 12 heat affected zone metal specimens; 10 tensile specimens including 3 base metal, 4 weld metal, and 3 weld heat affected zone metal specimens; 9 metal wire dosimeters including 3 iron, 3 nickel, and 3 copper.

After the first capsule holders (for both Units 1 and 2) were withdrawn and the specimens tested (see references 5.3-4 and 5.3-5), the broken specimens were remachined as miniature specimens and reloaded in the vessels during the next refueling outages. The contents of the new "reconstituted" capsules (for both Units 1 and 2) are as follows:

2 Charpy specimen packets each containing 12 Charpy specimens - 1 packet for base metal specimens and 1 for weld metal specimens. (EXCEPTION: The Unit 1 weld metal capsule only has 11 specimens).

Copper, Iron and Niobium flux wires are included in the capsules with the Charpy specimens.

2 tensile specimen tubes - 1 containing one tensile capsule with four 0.113 inch diameter miniature tensile specimens, the other containing 1 capsule with one 0.113 inch diameter miniature tensile specimen and one 0.250 inch diameter original weld metal tensile specimen.

The new holders have the same geometry as the original capsule holders.

Second holder (131 C7717G2):

24 Charpy impact specimens including 8 base metal, 8 weld metal, and 8 weld heat affected zone metal specimens; 8 tensile specimens including 3 base metal, 3 weld metal, and 2 weld heat affected zone metal specimens; 6 metal wire dosimeters including 2 iron, 2 nickel, and 2 copper.

Third holder (131 C7717G1):

24 Charpy impact specimens including 8 base metal, 8 weld metal and 8 weld heat affected zone metal specimens; 6 tensile specimens including 2 base metal, 2 weld metal, and 2 weld heat affected zone metal specimens; 6 metal wire dosimeters including 2 iron, 2 nickel, and 2 copper.

A set of out-of-reactor baseline Charpy V-notch specimens is provided with the surveillance test specimens.

Rev. 55 5.3-9

,SSES-FSAR NIMS Rev. 55 Chary ipact spccimcns for the roactOr vessel su~'cillanec progarams arc of the lonituina oricntation consistent with the ASMVE rcguircrnnts prior to the issuanoc of amount of shaft mceasurFed by these irradiatcd longitudinal test .pcicn will be essentially the same as the shift in an cguilent traneycrs spccmcn The program includcs thrcc capsuics in the rcacter. Sinoc the prcdictcd adjusted rcfcrcncs tcmpraiturc of the rcaetor vcssel bcltline steel as less than 1002F at end of life, the use Of thrcc capsuics mocets the rcguirsmcnts of 10 CFR 50, Appendix H, and ASTMI E 185 73. The withdrawal schcdulc ispovidcd in Table 5. .

For the extent of complianoc to 10 CER 50, ,Appcndix H, see Tables 5.3 l b and 5.3 2b.

5.3.1.6.2 Neutron Flux and Fluence Calculations A description of the methods of analysis is contained in Subsections 4.1.4.5 and 4.3.2.8.

5.3.1.6.3 Positioning of Surveillance Capsules and Method of Attachment Surveillance specimen capsules are located at three azimuths at a common elevation in the core beltline region. The sealed capsules are not attached to the vessel but are in welded capsule holders. The capsule holders are mechanically retained by capsule holder brackets welded to the vessel cladding as shown in Figure 5.3-3. The capsule holder brackets allow the capsule holder to be removed at any desired time in the life of the plant for specimen testing. These brackets are designed, fabricated and analyzed to the requirements of Section III of the ASME Code. A positive spring-loaded locking device is provided to retain the capsules in position throughout any anticipated event during the lifetime of the vessel.

5.3.1.6.4 Time and Number of Dosimetry Measurements GE has provided neutron dosimetry wires in each of the specimen holders. In addition, one holder in each vessel is designed with a separately removable dosimeter, to be removed after one fuel cycle. The first cycle dosimeter was removed from Unit 1 in 1986 and analyzed. A first cycle dosimeter was not available for removal from Unit 2.

However, the first cycle dosimetry for Unit 1 provides a good estimate of flux for Unit 2, because vessel geometries and core power shapes are very similar.

The first cycle dosimetry provides a means of calibrating the flux distribution calculations to actual vessel conditions. Dosimetry will be updated as holders are removed and tested. The holder withdrawal schedule is listed in Table 5.3-3.

5.3.1.7 Reactor Vessel Fasteners Rev. 55 5.3-10

SSES-FSAR 5.3.4 References 5.3-1 Faynshtein, K., and D. R. Pankratz, "Power Uprate Engineering Report for Susquehanna Steam Electric Station, Units 1 and 2," General Electric Report NEDC-32161 P, as revised by PP&L Calculation EC-PUPC-1 001, Revision 0, March, 1994.

5.3-2 Carey, R. G., "Susquehanna Steam Electric Station Unit 1 Vessel Surveillance Materials Testing and Fracture Toughness Analysis,"

General Electric Report GE-NE-523-169-1292, Revision 1, October, 1993.

Attached to PP&L letter PLA-3953, R. G. Byram to C. L. Miller, NRC,

"*Susquehanna Steam Electric Station, Submittal of Reactor Vessel Material Surveillance Test Report per 1 OCFR50 Appendix H for Unit 1,"

April 8, 1993.

5.3-3 Contreras, G. W., "Susquehanna Steam Electric Station Unit 2 Vessel Surveillance Materials Testing and Fracture Toughness Analysis,"

General Electric Report GE-NE-523-107-0893, Revision 1, October, 1993.

Attached to PP&L Letter PLA-4126, R. G. Byram to C. L. Miller, NRC, "Susquehanna Steam Electric Station, Submittal of Revision to Reactor Vessel Material Surveillance Test Report per 10CFR50 Appendix H for Unit 2," May 19,1994.

5.3-4 DuBord, R.M., "Susquehanna Steam Electric Station Unit 1 Fabrication of New Surveillance Capsule with Reconstituted Charpy Specimens" General Electric Nuclear Energy report GE-NE-523-A054-0595, May 1995.

5.3-5 DuBord, R.M., "Susquehanna Steam Electric Station Unit 2 Fabrication of New Surveillance Capsule with Reconstituted Charpy Specimens" General Electric Nuclear Energy report GE-NE-523-A055-0595, May 1995.

5.3-6 Structural Integrity Associates Report No. SIR-00-167, Revision 0, "Revised Pressure-Temperature Curves for Susquehanna Units 1 and 2",

January 2001.

5.3-7 BWRVIP-86: BWR Vessel Internals Proiect. BWR Intearated Surveillance Proaram Implementation Plan. February 2002. includina the latest revisions.

Rev 55 5.3-19

TABLE 5.3-1b APPENDIX H MATRIX FOR SUSQUEHANNA SES UNIT 1 APPENDIX H TOPIC COMPLY ALTERNATE ACTIONS OR COMMENTS PARA. NO. YES/NO OR N.A.

I Introduction N/A IL.A Fluence <1017 n/cm2 - Surveillance Program Not Required N/A Noncompliance with ASTM El 85-82in that the surveillance specimens are not necessarily from the limiting beltline material. Specimens are from representative 11.8 Standards Requirements (ASTM) for Surveillance No beltline material, however, and can be used to predict behavior of the limiting material.

Heat and heat/lot numbers for surveillance specimens are to be supplied.

Noncompliance inthat specimens may not have necessarily been taken from along Surveillance Specimen Shall be Taken from Locations side specimens required by Section III of Appendix G and transverse CVNs may not be II.C.l the Fracture Test Specimens (Section Alongside Gl)B of No employed. However, representative materials have been used, and RTNDT shift Appendix G) appears to be independent of specimen orientation.

Code basis is used for attachment of brackets to vessel cladding. See Section II.C.2 Locations of Surveillance Capsules inRPV Yes 5.3.1.6.4.

Thr,. apc.l.i; planned. Starting RTNoT of limiting material is based on alternative II.C.3.a Withdrawal Schedule of Capsules, RTNDT<100'F Yes action (see Paragraph III.A of Appendix G). One capsule complete. Other ca=sules are scheduled and tested inaccordance with Reference 5.3-7.

II.C.3.b Withdrawal Schedule of Capsules, RTNDT<200°F N/A 0

II.C.3.c. Withdrawal Schedule of Capsules, RTNDT<200 F N/A III.A Fracture Toughness Testing Requirements of Specimens Yes See Section 5.3.1.6 111.1 MethodHAZof Determining Adjusted Reference Temp. for Base Yes See Section 5.3.1.5 Metal, and Weld Metal IV.A Reporting Requirements of Test Results Yes See Section 5.3.1.6 IV.B Requirement for Dosimetry Measurement Yes See Section 5.3.1.6.2, 5.3.1.6.4 IV.C Reporting Requirements of Press/Temp. Limits Yes See Section 5.3.2

SSES-FSAR TABLE 5.3-3 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM-WITHDRAWAL SCHEDULE I Specimen Holder Vessel Location Lead Factor

  • Withdrawal Time (EFPY9 UNIT 1 131C7717G1 3000 1.20 Spare 131 C7717G2 1200 1.20 4-522 6

1310C7717G3 300 1.20 (Actual Date - Spring 1992)

G3 Reconstituted 300 1.20 Spare Specimens UNIT 2 131C7717G1 3000 1.20 Spare 131 C7717G2 1200 1.20 -1r 6

131C771 7G3 300 1.20 (Actual Date - Fall 1992)

G3 Reconstituted 300 1.20 Spare Specimens SAt 1/4 T.

Note- The Unit 1 surefoillanco specimenRs at tho 300 location wore romoved froM the vesselfo testing during the Spring 1992 insp9ctonR outage and these 6pecmn wre rco:nstitutd and replaced back into the vessel 300 location dur*ing tho Fall 1923 ip eto utage (Ul -7R10). The Un~fit 2 surweillance Specmn were rmoved frm_ý1 the v~essel 300 location for testing duringth Fall 1-99-2 inspection outage and thoe se~oswr reconstituted and replaced back into the Vessel 9390 location duFrig the Spring 1994 npeto outage (UJ261410). Details of the reontiutonprocess and the capsule coententS c-an -be found- in Rieference 5.3 4 and 5.3-5-.

SWithdrawal Time is inaccordance wfth Reference-5.3-7.

Rev. 54,10/99 Page 8 of 8

TABLE 5.3-2b APPENDIX H MATRIX FOR SUSQUEHANNA SES UNIT 2 APPENDIX H TOPIC COMPLY PARA. NO. YES/NO OR N.A. ALTERNATE ACTIONS OR COMMENTS I Introduction N/A II.A Fluence <1017 n/cm 2 - Surveillance Program Not Required N/A II.B Standards Requirements (ASTM) For Surveillance No Noncompliance with ASTM E185-73 inthat the surveillance specimens are not necessarily from the limiting beltline material.

Specimens are from representative beltline material, however, and can be used to predict behavior of the limiting material. Heat and heatAot numbers for surveillance specimens are to be supplied.

I1.C.1 Surveillance Specimen Shall be Taken from Locations No Noncompliance inthat specimens may not have necessarily been Alongside the Fracture Test Specimens (Section 111.B of taken from alongside specimens required by Section IIIof Appendix Appendix G) G and transverse CVN's may not be employed. However, representative materials have been used, and RTNDT shift appears to be independent of specimen orientation.

II.C.2 Locations of Surveillance Capsules in RPV Yes Code basis is used for attachment of brackets to vessel cladding.

See Subsections 5.3.1.6.4.

II.C.3.a Withdrawal Schedule of Capsules, RTNDT<100*F Yes .Thr- plan.od. Starting RTNDT of limiting material is based on altemative action (see Paragraph III.A of Appendix G). One capsule comoleted, Other capsules are scheduled and tested in accordance with Reference 5.3-7.

II.C.3.b Withdrawal Schedule of Capsules, RTNDT<200 0 N/A II.C.3.c Withdrawal Schedule of Capsules, RTNDT<200 0 N/A III.A Fracture Toughness Testing Requirements of Specimens Yes See Section 5.3.1.6 1113. Method of Determining Adjusted Reference Temp. for Yes See Section 5.3.1.5 Base Metal, HAZ and Weld Metal IV.A Reporting Requirements of Test Results Yes See Section 5.3.1.6 IV.B Requirement for Dosimetry Measurement Yes See Section 5.3.1.6.2, 5.3.1.6.4 IV.C Reporting Requirements of Press/Temp. Limits Yes See Section 5.3.2

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