ML080640587

Submittal of Relief Requests Associated with the Third and Fourth Inservice Inspection (Lsi) Intervals and the First and Second Containment Inservice Inspection (Cisi) Intervals

ML080640587
Person / Time
Site:	Peach Bottom
Issue date:	02/29/2008
From:	Cowan P Exelon Generation Co, Exelon Nuclear
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML080640587 (37)

Text

Exelon Nuclear www.exeloncorp.com 200 Exelon Way Nuclear Kennett Square, PA 19348 10 CFR 50.55a February 29, 2008 u.s. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555 Peach Bottom Atomic Power Station, Units 2 and 3 Renewed Facility Operating License Nos. DPR-44 and DPR-56 NRC Docket Nos. 50-277 and 50-278

Subject:

Submittal of Relief Requests Associated with the Third and Fourth Inservice Inspection (lSI) Intervals and the First and Second Containment Inservice Inspection (CISI) Intervals Attached for your review and approval are relief requests associated with the third (13R-45) and fourth (14R-08, 14R-25, 14R-44, 14R-46, 14R-47) Inservice Inspection (lSI) intervals for Peach Bottom Atomic Power Station (PBAPS), Units 2 and 3. The interval, ASME Boiler and Pressure Vessel (B&PV)Section XI Code Editions of compliance, start and end dates are as follows:

Interval Code Edition Start Date End Date Third lSI Interval 1989 Edition PBAPS, Unit 2 - November PBAPS, Unit 2 - November 4, 5, 1998 2008 PBAPS, Unit 3 - November 4, PBAPS, Unit 3 - August 15, 2008*

1998

• 13R-45 proposes to extend the PBAPS, Unit 3 third lSI interval by 13 weeks to Nov. 4, 2008 in order to create a common start and end date to the intervals.

Fourth Interval 2001 Edition PBAPS, Unit 2 - November PBAPS, Unit 2 - November 4, through 2003 5,2008 2018 Addenda PBAPS, Unit 3 - November PBAPS, Unit 3 - November 4, 5,2008 2018

RR Associated with Third and Fourth lSI Intervals and First and Second CISI Intervals February 29, 2008 Page 2 Also included for your review and approval are relief requests associated with the first (CRR-12, CRR-13) and second (CRR-13) Containment Inservice Inspection (CISI) interval for PBAPS, Units 2 and 3. The interval, ASME Boiler and Pressure Vessel (B&PV)Section XI Code Editions of compliance, start and end dates are as follows:

Interval Code Edition Start Date End Date First CISI Interval 1992 Edition Unit 2 - November 5, 1998 Unit 2 - November 4, 2008 through 1992 Addenda Unit 3 - November 5, 1998 Unit 3 - November 4, 2009*

• As requested in CRR-12, the second CISI interval for Unit 3 will overlap the first interval for approximately one year. This overlap is not perm itted in the 1992 Edition through 1992 Addenda of ASME Section XI.

Second CISI 2001 Edition Unit 2 - November 5, 2008 Unit 2 - November 4, 2018 Interval through 2003 Addenda Unit 3 - November 5,2008 Unit 3 - November 4,2018 We request your approval by February 28, 2009.

Should you have any questions concerning this letter, please contact Tom Loomis at (610) 765-5510.

Respectfully, ty rJ ,B

'if Y{/AtttJ/£(f{Ja<<

Pamela B. Cowan Director - Licensing & Regulatory Affairs Exelon Generation Company, LLC

Attachment:

1) Relief Requests Associated with the Inservice Inspection (lSI) and Containment Inservice Inspection (CISI) Programs for Peach Bottom Atomic Power Station, Units 2 and 3 cc: S. J. Collins, Regional Administrator, Region I, USNRC F. Bower, USNRC Senior Resident Inspector, PBAPS J. Hughey, Project Manager, USNRC R. I. McLean, State of Maryland R. R. Janati, Commonwealth of Pennsylvania

Relief Requests Associated with the Inservice Inspection (ISI) and Containment Inservice Inspection (CISI) Programs for Peach Bottom Atomic Power Station, Units 2 and 3 Inservice Inspection Relief Requests I3R-45 (Third ISI Interval)

I4R-44 (Fourth ISI Interval)

I4R-08 (Fourth ISI Interval)

I4R-25 (Fourth ISI Interval)

I4R-46 (Fourth ISI Interval)

I4R-47 (Fourth ISI Interval)

Containment Inservice Inspection (CISI) Relief Requests CRR-12 (First CISI Interval)

CRR-13 (First and Second CISI Intervals)

Relief Request I3R-45 for Synchronization of the Ten-Year ISI Intervals Between Units 2 and 3 for Class 1, 2, and 3 Components in Accordance with 10CFR50.55a(a)(3)(i)

(Page 1 of 3)

1. ASME CODE COMPONENTS AFFECTED:

Code Class: 1, 2, and 3

Reference:

IWA-2430 IWA-2432 Examination Category: All Item Number: All

Description:

Synchronization of Ten-Year Inservice Inspection (ISI)

Intervals between Units 1 and 2 for Class 1, 2, and 3.

Component Number: All Class 1, 2, and 3 Components

2. APPLICABLE CODE EDITION AND ADDENDA:

The current (third) Inservice Inspection (ISI) interval is based on the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code,Section XI, 1989 Edition.

3. APPLICABLE CODE REQUIREMENT:

Paragraph IWA-2430(b), Inspection Intervals, requires the inspection interval to be determined by calendar years following placement of the plant into commercial service.

Paragraph IWA-2432, Inspection Program B, requires that each inspection interval consist of a ten-year duration, except as modified by IWA-2430(d) which permits the inspection interval to be reduced or extended by as much as one year, provided that successive intervals are not altered by more than one year from the original pattern of intervals.

4. REASON FOR REQUEST:

Pursuant to 10CFR50.55a(a)(3)(i), relief is requested from the third ten-year interval requirements contained within IWA-2430(b) and (d) and IWA-2432 for the Peach Bottom Atomic Power Station, Units 2 and 3 ISI program on the basis that the proposed alternative would provide an acceptable level of quality and safety.

Specifically, relief is being sought to increase the duration of the Peach Bottom Atomic Power Station, Unit 3 third ISI interval end date from August 14, 2008 to November 4, 2008 in order to create a common ISI interval for both units. Currently, the Peach Bottom Atomic Power Station, Unit 2 third ISI interval is scheduled to end on November 4, 2008 and the Unit 3 third interval will end on August 15, 2008. This creates a gap between the two units ISI programs of approximately 13 weeks. This gap may result in different governing Code Editions in future intervals for a short period of time, which would in turn require different program requirements, and the need for different parallel implementing procedures.

Increasing the duration of the third ISI interval for Peach Bottom Atomic Power Station, Unit 3 by approximately 13 weeks will permit the commencement of its fourth ISI

Relief Request I3R-45 for Synchronization of the Ten-Year ISI Intervals Between Units 2 and 3 for Class 1, 2, and 3 Components in Accordance with 10CFR50.55a(a)(3)(i)

(Page 2 of 3) interval to coincide with the start of the fourth ISI interval for Peach Bottom Atomic Power Station, Unit 2 and hence will establish a joint inspection interval with common start and end dates. This will assure both ISI programs use the same Code Edition and Addenda for the next and successive intervals and will likewise establish a common program update and implementing procedures for future intervals for both units.

Increasing the interval by application of the extension guidance of IWA-2430(d) is not permitted for Peach Bottom because the 10-year ISI interval extension has been incorporated into the current ISI intervals for Units 2 and 3 (Reference 1).

Exelon Generation Company, LLC concludes that authorizing the proposed alternative as described herein provides an acceptable level of quality and safety, and does not adversely impact the health and safety of the public.

5. PROPOSED ALTERNATIVE AND BASIS FOR USE:

As an alternative to the third ten-year interval duration requirements of IWA-2430(b) and (d) and IWA-2432, Peach Bottom Atomic Power Station proposes to modify the interval end date of the Unit 3 third ISI interval to conclude on November 4, 2008. This will permit the subsequent ISI programs for both units to share a common inspection interval and to implement common Code Editions for Class 1, 2, and 3 components. The common code of record for the third interval ISI programs for both units is the 1989 Edition of ASME Section XI Code.

As a result of these requested interval modifications to extend the Unit 3 interval by approximately 13 weeks, the start date of the fourth interval ISI program for both Units 2 and 3 will be November 5, 2008. Using this date, the Peach Bottom Atomic Power Station, Unit 2 Fall refueling outage in September 2008 (P2R17) remains as currently scheduled in the third ISI interval, and the Peach Bottom Atomic Power Station, Unit 3 Fall refueling outage in September 2009 (P3R17) remains as currently scheduled, the first refueling outage of the fourth (next) ISI interval. Therefore, this change will not impact inservice inspections for the upcoming outages and is administrative in nature. As required by ASME Section XI, the intervals will be scheduled in 10-year increments from November 5, 2008 forward with the modifications allowed by IWA-2430 available to future intervals and periods.

6. DURATION OF PROPOSED ALTERNATIVE:

Relief is requested to extend the end date of the current Peach Bottom Atomic Power Station, Unit 3 third ISI interval by approximately 13 weeks. The current PBAPS, Unit 3 interval began on August 14, 1998. It is proposed that the fourth ISI interval begin November 5, 2008 and conclude November 4, 2018 for Peach Bottom Atomic Power Station, Units 2 and 3.

Relief Request I3R-45 for Synchronization of the Ten-Year ISI Intervals Between Units 2 and 3 for Class 1, 2, and 3 Components in Accordance with 10CFR50.55a(a)(3)(i)

(Page 3 of 3)

7. PRECEDENTS:

Relief requests that have been submitted or approved are:

1. Relief Request 3RR-10 was approved by the U. S. Nuclear Regulatory Commission for Susquehanna Steam Electric Station, Units 1 and 2 (U. S. Nuclear Regulatory Commission letter dated September 24, 2004, Susquehanna Steam Electric Station, Units 1 and 2 - Third 10-Year Inservice Inspection (ISI) Interval Program Plan (TAC NOS. MC1185, MC1186, MC1191, MC1192, MC1193, MC1194, MC1195, MC1196, MC1197, MC1198, MC1199, MC1200)).

2. Relief Request I3R-01 was approved by the U. S. Nuclear Regulatory Commission for Byron Station, Units 1 and 2 (U. S. Nuclear Regulatory Commission letter dated September 7, 2006, Byron Station, Unit Nos. 1 and 2 - Evaluation of Inservice Inspection Program Relief Request I3R-01 (TAC NOS. MD1209 and MD1210)).

3. Relief Request I3R-01 was approved by the U. S. Nuclear Regulatory Commission for Limerick Generating Station, Units 1 and 2 (U. S. Nuclear Regulatory Commission letter dated January 24, 2007, Limerick Generating Station, Units 1 and 2 - Relief Requests I3R-01 For Alignment of Inservice Inspection and Containment Inservice Inspection (TAC. NOS. MD2727 and MD2728)).

8.

REFERENCES:

1. G. A. Hunger (Exelon, formally PECO Energy Company) to U. S. Nuclear Regulatory Commission letter dated January 30, 1997 transmitting PBAPS, Units 2 and 3 Inservice Inspection (ISI) Program and Inservice Testing (IST) Program dates.

Relief Request I4R-44 for Alternate Risk-Informed Selection and Examination Criteria for Examination Category B-F, B-J, C-F-1, and C-F-2 Pressure Retaining Piping Welds In Accordance with 10CFR50.55a(a)(3)(i)

(Page 1 of 7) 1.0 ASME CODE COMPONENTS AFFECTED:

Code Class: 1 and 2

Reference:

Table IWB-2500-1, Table IWC-2500-1 Examination Category: B-F, B-J, C-F-1, and C-F-2 Item Number: B5.10, B5.20, B9.11, B9.21, B9.31, B9.32, B9.40, C5.11, C5.51, and C5.81

Description:

Alternate Risk-Informed Selection and Examination Criteria for Examination Category B-F, B-J, C-F-1, and C-F-2 Pressure Retaining Piping Welds Component Number: Unit 2 and Unit 3 Pressure Retaining Piping 2.0 APPLICABLE CODE EDITION AND ADDENDA:

The Inservice Inspection program is based on the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code,Section XI, 2001 Edition through the 2003 Addenda.

3.0 APPLICABLE CODE REQUIREMENT:

The following Code requirements are paraphrased from the 2001 Edition through the 2003 Addenda of ASME Section XI.

Table IWB-2500-1, Examination Category B-F, requires volumetric and surface examinations on all welds for Item B5.10 and surface examinations for all welds for Item B5.20.

Table IWB-2500-1, Examination Category B-J, requires volumetric and/or surface examinations on a sample of welds for Items B9.11, B9.21, B9.31, B9.32, and B9.40.

The weld population selected for inspection includes the following:

1. All terminal ends in each pipe or branch run connected to vessels.

2. All terminal ends and joints in each pipe or branch run connected to other components where the stress levels exceed either of the following limits under loads associated with specific seismic events and operational conditions:

a. primary plus secondary stress intensity range of 2.4Sm for ferritic steel and austenitic steel.

b. cumulative usage factor U of 0.4.

3. All dissimilar metal welds not covered under Examination Category B-F.

4. Additional piping welds so that the total number of circumferential butt welds, branch connections, or socket welds selected for examination equals 25% of the

Relief Request I4R-44 for Alternate Risk-Informed Selection and Examination Criteria for Examination Category B-F, B-J, C-F-1, and C-F-2 Pressure Retaining Piping Welds In Accordance with 10CFR50.55a(a)(3)(i)

(Page 2 of 7) circumferential butt welds, branch connection, or socket welds in the reactor coolant piping system. This total does not include welds exempted by IWB-1220.

Table IWC-2500-1, Examination Categories C-F-1 and C-F-2 require volumetric and/or surface examinations on a sample of welds for Items C5.11, C5.51, and C5.81. The weld population selected for inspection includes the following:

1. Welds selected for examination shall include 7.5%, but not less than 28 welds, of all dissimilar metal, austenitic stainless steel and high alloy welds (Examination Category C-F-1) or of all carbon and low alloy steel welds (Examination Category C-F-2) not exempted by IWC-1220. (Some welds not exempted by IWC-1220 are not required to be nondestructively examined per Examination Categories C-F-1 and C-F-2. These welds, however, shall be included in the total weld count to which the 7.5% sampling rate is applied.) The examinations shall be distributed as follows:

a. the examinations shall be distributed among the Class 2 systems prorated, to the degree practicable, on the number of nonexempt dissimilar metal, austenitic stainless steel and high alloy welds (Examination Category C-F-1) or carbon and low alloy welds (Examination Category C-F-2) in each system;

b. within a system, the examinations shall be distributed among terminal ends, dissimilar metal welds, and structural discontinuities prorated, to the degree practicable, on the number of nonexempt terminal ends, dissimilar metal welds, and structural discontinuities in the system; and

c. within each system, examinations shall be distributed between piping sizes prorated to the degree practicable.

4.0 REASON FOR REQUEST:

Pursuant to 10CFR50.55a(a)(3)(i), relief is requested on the basis that the proposed alternative utilizing Reference 1 along with two enhancements from Reference 4 will provide an acceptable level of quality and safety.

As stated in Safety Evaluation Report Related to EPRI Risk-Informed Inservice Inspection Evaluation Procedure (EPRI TR-112657, Revision B, July 1999) (Reference 2):

"The staff concludes that the proposed RISI program as described in EPRI TR-112657, Revision B, is a sound technical approach and will provide an acceptable level of quality and safety pursuant to 10CFR50.55a for the proposed alternative to the piping ISI requirements with regard to the number of locations, locations of inspections, and methods of inspection."

Relief Request I4R-44 for Alternate Risk-Informed Selection and Examination Criteria for Examination Category B-F, B-J, C-F-1, and C-F-2 Pressure Retaining Piping Welds In Accordance with 10CFR50.55a(a)(3)(i)

(Page 3 of 7)

The initial Peach Bottom Atomic Power Station Risk-Informed Inservice Inspection (RI-ISI) program was submitted during the first period of the third interval for Unit 2 and during the second period of the third interval for Unit 3. This initial RI-ISI program was developed in accordance with EPRI TR-112657, Revision B-A, as supplemented by Code Case N-578-1. The program was approved for use by the USNRC via a Safety Evaluation as transmitted to Exelon (Reference 5).

The transition from the 1989 Edition to the 2001 Edition through the 2003 Addenda of ASME Section XI for Peach Bottom Atomic Power Stations fourth interval does not impact the currently approved Risk-Informed ISI evaluation methods and process used in the third interval, and the requirements of the new Code Edition/Addenda will be implemented as detailed in the Peach Bottom Atomic Power Station ISI Program Plan.

Therefore, with the exception of specific weld locations that may have changed due to maintenance or modification activities, the proposed alternative RI-ISI program for the fourth interval is the same program as approved in Reference 5 for the third interval.

The Risk Impact Assessment completed as part of the original baseline RI-ISI Program was an implementation/transition check on the initial impact of converting from a traditional ASME Section XI program to the new RI-ISI methodology. For the fourth interval ISI update, there is no transition occurring between two different methodologies, but rather, the currently approved RI-ISI methodology and evaluation will be maintained for the new interval. As such, the original risk impact assessment is not a necessary element of the implementing process and is not required to be continually updated. This is because when updating the RI-ISI analysis for the fourth 10-year interval, the original risk impact assessment will also be updated to confirm the change in risk was maintained within the acceptance guidelines. The original methodology of the calculation has not changed, and the change in risk is simply re-assessed using the initial 1989 Section XI program prior to RI-ISI and the new element selection for the fourth 10-year interval RI-ISI program. This same process has been maintained in each revision to the Peach Bottom RI-ISI assessment that has been performed to date.

As an added measure of assurance, any new systems, portions of systems, or components being included in the RI-ISI program for the fourth interval will be added to the Risk Impact Assessment performed during the previous interval. These components will be addressed within the evaluation at the start of the new interval to assure that the new fourth interval RI-ISI element selection provides an acceptable overall change-in-risk when compared to the previous ASME Section XI population of exams which existed prior to the implementation of the first RI-ISI program.

The actual evaluation and ranking procedure including the Consequence Evaluation and Degradation Mechanism Assessment processes of the currently approved (Reference 5)

RI-ISI Program remain unchanged and are continually applied to maintain the Risk Categorization and Element Selection methods of EPRI TR-112657, Revision B-A.

These portions of the RI-ISI Program are reevaluated as major revisions of the site Probabilistic Risk Assessment (PRA) occur and modifications to plant configuration are

Relief Request I4R-44 for Alternate Risk-Informed Selection and Examination Criteria for Examination Category B-F, B-J, C-F-1, and C-F-2 Pressure Retaining Piping Welds In Accordance with 10CFR50.55a(a)(3)(i)

(Page 4 of 7) made. The Consequence Evaluation, Degradation Mechanism Assessment, Risk Ranking, and Element Selection steps encompass the complete program process applied under the Peach Bottom Atomic Power Station RI-ISI Program.

5.0 PROPOSED ALTERNATIVE AND BASIS FOR USE:

The proposed alternative originally implemented in the risk informed in-service inspection plan for Peach Bottom Atomic Power Station Units 2 and 3 (Reference 3),

along with the two enhancements noted below, provide an acceptable level of quality and safety as required by 10CFR50.55a(a)(3)(i). This original program along with these same two enhancements is currently approved for Peach Bottom Atomic Power Stations third inservice inspection interval as documented in Reference 5.

The fourth interval RI-ISI program will be a continuation of the current application and will continue to be a living program as described in the Reason For Request section of this relief request. No changes to the evaluation methodology as currently implemented under EPRI TR-112657, Revision B-A, are required as part of this interval update. The following two enhancements will continue to be implemented.

a. In lieu of the evaluation and sample expansion requirements in Section 3.6.6.2, "RISI Selected Examinations" of EPRI TR-112657, Peach Bottom Atomic Power Station will utilize the requirements of Subarticle -2430, "Additional Examinations" contained in Code Case N-578-1 (Reference 4). The alternative criteria for additional examinations contained in Code Case N-578-1 provide a more refined methodology for implementing necessary additional examinations. The reason for this selection is that the guidance discussed in EPRI TR-112657 includes requirements for additional examinations at a high level, based on service conditions, degradation mechanisms, and the performance of evaluations to determine the scope of additional examinations, whereas ASME Code Case N-578-1 provides more specific and clearer guidance regarding the requirements for additional examinations that is structured similar to the guidance provided in ASME Section XI, paragraphs IWB-2430 and IWC-2430. Additionally, similar to the current requirements of ASME Section XI, Peach Bottom intends to perform additional examinations that are required due to the identification of flaws or relevant conditions exceeding the acceptance standards, during the outage the flaws are identified.

b. To supplement the requirements listed in Table 4-1, "Summary of Degradation-Specific Inspection Requirements and Examination Methods" of EPRI TR-112657, Peach Bottom Atomic Power Station will utilize the provisions listed in Table 1, Examination Category R-A, "Risk-Informed Piping Examinations" contained in Code Case N-578-1 (Reference 4). To implement Note 10 of this table, paragraphs and figures from the 2001 Edition through the 2003 Addenda of ASME Section XI (Peach Bottom Atomic Power Stations Code of record for the Fourth Interval) will be utilized which parallel those referenced in the Code Case for the 1989 Edition. Table 1 of Code Case N-578-1 will be used as it provides a detailed breakdown for

Relief Request I4R-44 for Alternate Risk-Informed Selection and Examination Criteria for Examination Category B-F, B-J, C-F-1, and C-F-2 Pressure Retaining Piping Welds In Accordance with 10CFR50.55a(a)(3)(i)

(Page 5 of 7) examination method and categorization of parts to be examined. Additionally, Section 4 of EPRI TR-112657 states "Application of RI-ISI uses NDE techniques that are designed to be effective for specific degradation mechanisms and examination locations." Section 4 also identifies methods of examination for each degradation mechanism with the primary method being ultrasonic testing (UT) techniques.

However, EPRI TR-112657 does not identify the examination volumes for components without a degradation mechanism. In addition, EPRI TR-112657 does not specify examination volumes and methods for socket welds. Peach Bottom has requested to use the examination methods from Code Case N-578-1 instead of the methods from EPRI TR-112657, except that the volumetric method will be used to examine intergranular stress corrosion cracking (IGSCC). The examination figures specified in Section 4 of EPRI TR-112657 will be used to determine the examination volume based on the degradation mechanism and component configuration.

Peach Bottom uses UT techniques for RI-ISI volumetric examinations.

For the components addressed by the Risk Informed Inservice Inspection (RI-ISI) program, ASME Section XI focuses primarily on weld examinations. Risk Informed examination volumes also include portions of piping and fitting base materials that are susceptible to particular degradation mechanisms. The examination figures specified in Section 4 of EPRI TR-112657 differ from the examination figures in ASME Section XI for certain component configurations and evaluated degradations.

The ASME Section XI, Mandatory Appendix I, "Ultrasonic Examinations," specifies that UT examination procedures, equipment, and personnel used to detect and size flaws in piping welds shall be qualified by performance demonstration in accordance with ASME Section XI Appendix VIII, "Performance Demonstration for Ultrasonic Examination Systems." The RI-ISI program complies with Appendix VIII for weld examinations. In cases where the examination requirements cannot be met, Peach Bottom will submit a request for relief in accordance with 10CFR50.55a, "Codes and standards."

The examination methods are designed to be effective for specific degradation mechanisms and examination locations. The volumetric scanning will be in both axial and circumferential directions to detect the flaws in these orientations.

Additionally, all Peach Bottom dissimilar metals (DM) welds, as characterized in ASME Section XI IWA-9000, have been evaluated for failure potential and consequence of failure along with the other non-exempt piping. The piping segments containing the DM welds were classified into the appropriate RI-ISI categories, and appropriate elements were selected per the category requirements for examination during the third inspection interval.

Piping welds, including DM welds in vessel nozzles, that are susceptible to IGSCC (i.e.,

IGSCC Categories B through G, as applicable) and not subject to other degradation mechanism(s) are removed from the RI-ISI program population. They are contained in

Relief Request I4R-44 for Alternate Risk-Informed Selection and Examination Criteria for Examination Category B-F, B-J, C-F-1, and C-F-2 Pressure Retaining Piping Welds In Accordance with 10CFR50.55a(a)(3)(i)

(Page 6 of 7) the Peach Bottom ISI Augmented Program 01, "USNRC Generic Letter 88-01, Intergranular Stress Corrosion Cracking" and are subject to the inspection requirements of BWRVIP-75-A "BWR Vessel and Internals Project Technical Basis for Revisions to Generic Letter 88-01 Inspection Schedules". Furthermore, all piping welds and DM welds in vessel nozzles classified as Category A (resistant material) per BWRVIP-75-A are included in the RI-ISI program.

The Peach Bottom Atomic Power Station RI-ISI Program, as developed in accordance with EPRI TR-112657, Rev. B-A (Reference 1), requires that 25% of the elements that are categorized as High risk (i.e., Risk Category 1, 2, and 3) and 10% of the elements that are categorized as Medium risk (i.e., Risk Categories 4 and 5) be selected for inspection. For this application, the guidance for the examination volume for a given degradation mechanism is provided by the EPRI TR-112657 while the guidance for the examination method and categorization of parts to be examined are provided by the EPRI TR-112657 as supplemented by Code Case N-578-1.

In addition to this risk-informed evaluation, selection, and examination procedure, all ASME Section XI piping components, regardless of risk classification, will continue to receive Code required pressure testing as part of the current ASME Section XI program.

VT-2 visual examinations are within the ASME pressure boundary and are examined each refueling outage as part of the system leakage test required by ASME Section XI.

These examinations are scheduled in accordance with the Peach Bottom Atomic Power Station pressure-testing program, which remains unaffected by the RI-ISI program.

6.0 DURATION OF PROPOSED ALTERNATIVE:

Relief is requested for the Fourth Ten-Year Inspection Interval for Peach Bottom Atomic Power Station Units 2 and 3. The fourth ISI interval begins on November 5, 2008 for both Units as proposed in Relief Request I3R-45, and will conclude November 4, 2018 for Peach Bottom Atomic Power Station, Units 2 and 3.

7.0 PRECEDENTS

Similar relief requests have been approved for:

1. Peach Bottom Atomic Power Station Third Inspection Interval Relief Request RR-44 was approved per SER dated August 27, 2003. The Fourth Inspection Interval Relief Request will utilize the same RI-ISI methodology that was previously approved in the Third Inspection Interval.

2. Limerick Generating Station Second Inspection Interval Relief Request RR-32 was approved per SER dated March 3, 2003.

3. Susquehanna Steam Electric Station Third Inspection Interval Relief Request I3R-01 was approved per SER dated July 28, 2005.

Relief Request I4R-44 for Alternate Risk-Informed Selection and Examination Criteria for Examination Category B-F, B-J, C-F-1, and C-F-2 Pressure Retaining Piping Welds In Accordance with 10CFR50.55a(a)(3)(i)

(Page 7 of 7)

4. Dresden Station Fourth Inspection Interval Relief Request I4R-02 was approved per SER dated September 4, 2003.

5. Quad Cities Station Fourth Inspection Interval Relief Request I4R-02 was approved per SER dated January 28, 2004.

8.0 REFERENCES

1. Electric Power Research Institute (EPRI) Topical Report (TR) 112657 Rev. B-A, "Revised Risk-Informed Inservice Inspection Evaluation Procedure," December 1999.

2. W. H. Bateman (USNRC) to G. L. Vine (EPRI) letter dated October 28, 1999 transmitting "Safety Evaluation Report Related to EPRI Risk-Informed Inservice Inspection Evaluation Procedure (EPRI TR-112657, Revision B, July 1999)."

3. Initial Risk-Informed Inservice Inspection Evaluation - Peach Bottom Atomic Power Station Units 2 and 3, dated May 16, 2002.

4. American Society of Mechanical Engineers (ASME) Code Case N-578-1, "Risk-Informed Requirements for Class 1, 2, or 3 Piping, Method B."

5. James W. Clifford (USNRC) to John L. Skolds (Exelon) letter dated August 27, 2003 transmitting the Peach Bottom Atomic Power Station, Units 2 and 3 -

American Society of Mechanical Engineers Boiler and Pressure Vessel Code -

Relief for Risk-Informed Inservice Inspection of Piping (TAC Nos. MB5512 and MB5513).

Relief Request I4R-08 Due to the Limited Volumetric Examination of RHR Heat Exchanger Pressure Retaining Shell Circumferential Welds (Shell-to-Flange Welds)

In Accordance with 10 CFR 50.55a(g)(5)(iii)

(Page 1 of 3)

1. ASME CODE COMPONENTS AFFECTED:

Code Class: 2

Reference:

Table IWC-2500-1 Examination Category: C-A Item Number: C1.10

Description:

Limited Volumetric Examination of RHR Heat Exchanger Pressure Retaining Shell Circumferential Welds (Shell-to-Flange Welds)

Component Number: 10-2HXA(B, C, D)-1 (Unit 2) 10-2HXA(B, C, D)-1 (Unit 3)

Drawing Number: ISI-2-10-1 (Unit 2) and ISI-3-10-2 (Unit 3)

2. APPLICABLE CODE EDITION AND ADDENDA:

The Inservice Inspection program is based on the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code,Section XI, 2001 Edition through the 2003 Addenda.

3. APPLICABLE CODE REQUIREMENT:

The following Code requirements are paraphrased from the 2001 Edition through the 2003 Addenda of ASME Section XI.

Table IWC-2500-1 states that the heat exchanger shell welds require a volumetric examination in accordance with the examination requirements illustrated in Figure IWC-2500-1.

Per Table IWC-2500-1, examinations are limited to 100% of the pressure retaining shell circumferential welds at gross structural discontinuities of one (1) heat exchanger (or the equivalent of one heat exchanger) during the in-service inspection interval.

4. IMPRACTICALITY OF COMPLIANCE:

Pursuant to 10CFR50.55a(g)(5)(iii), relief is requested on the basis that conformance with these code requirements is impractical as conformance would require extensive structural modifications to the Residual Heat Removal heat exchangers.

Peach Bottom Atomic Power Station, Units 2 and 3, each have four (4) Residual Heat Removal heat exchangers. Code Examination Category C-A, Item Number C1.10, requires the volumetric examination of the equivalent of the welds of one heat exchanger per unit.

All eight RHR heat exchangers have the same configuration, which is shown in Figure No.

I4R-08-1. In accordance with ASME,Section XI Code requirements, examinations will be performed on shell-to-flange welds 10-2HXx-1 and 10-2HXx-2, which are the upper and lower circumferential shell welds in the A RHR heat exchanger in each unit. Upper shell-

Relief Request I4R-08 Due to the Limited Volumetric Examination of RHR Heat Exchanger Pressure Retaining Shell Circumferential Welds (Shell-to-Flange Welds)

In Accordance with 10 CFR 50.55a(g)(5)(iii)

(Page 2 of 3) to-flange weld 10-2HXA-1 can only be examined from one side of the weld due to the configuration of the flange. In addition, access for a one-sided examination is limited due to the weld crown configuration. Approximately 15% of the required examination volume for upper shell weld to flange welds 10-2HXA(B, C, D)-1 is inaccessible for examination due to the above conditions. Lower shell to flange welds 10-2HXA(B, C, D)-2 are accessible from both sides of the weld because the lower flange forging was supplied with a short integral extension.

5. BURDEN CAUSED BY COMPLIANCE:

Compliance with the applicable Code requirements can only be accomplished by redesigning and refabricating or replacing the subject heat exchangers. Based on this, the Code requirements are deemed impractical in accordance with 10CFR50.55a(g)(5)(iii).

6. PROPOSED ALTERNATIVE AND BASIS FOR USE:

As an alternative examination, Peach Bottom Atomic Power Station, Units 2 and 3, will examine the selected upper shell to flange weld (10-2HXA(B, C, or D)-1) to the extent practical. The expected code coverage is approximately 85% of the weld volume. The selected lower shell weld (10-2HXA(B, C, or D)-2) will achieve the required code coverage.

Additionally, a visual examination (VT-2) during system pressure testing per Examination Category C-H will be performed on the heat exchangers (once during each period) to verify leak tight integrity of these insulated welds.

7. DURATION OF PROPOSED ALTERNATIVE:

Relief is requested for the fourth ten-year inspection interval for Peach Bottom Atomic Power Station, Units 2 and 3. The fourth ISI interval begins on November 5, 2008 for both Units as proposed in Relief Request I3R-45, and will conclude November 4, 2018 for Peach Bottom Atomic Power Station, Units 2 and 3.

8. PRECEDENTS:

Similar relief requests have been approved for:

1. Peach Bottom Atomic Power Station Third Inspection Interval Relief Request RR-08 was authorized per SER dated July 31, 2000. The Fourth Inspection Interval Relief Request will utilize the same approach that was previously approved in the third inspection interval.

2. Limerick Generating Station Second Inspection Interval Relief Request RR-06 was authorized per SER dated May 2, 2002.

Relief Request I4R-08 Due to the Limited Volumetric Examination of RHR Heat Exchanger Pressure Retaining Shell Circumferential Welds (Shell-to-Flange Welds)

In Accordance with 10 CFR 50.55a(g)(5)(iii)

(Page 3 of 3)

FIGURE I4R-08-1 RESIDUAL HEAT REMOVAL HEAT EXCHANGER 1 0 -2 H X A (B , C , D )-1 1 0 -2 H X A (B , C , D )-2

Relief Request I4R-25 for Pressure Testing the RPV Head Flange Seal Weld Leak Detection System In Accordance with 10CFR50.55a(g)(5)(iii)

(Page 1 of 5)

1. ASME CODE COMPONENTS AFFECTED:

Code Class: 2

Reference:

Table IWC-2500-1 IWC-5200 Examination Category: C-H Item Number: C7.10

Description:

Pressure Testing the RPV Head Flange Seal Leak Detection System Component Number: Class 2 RPV Head Flange Seal Leak Detection System Unit 2: Piping 4DCN-1" Unit 3: Piping 4DCN-1" Drawing Number: ISI-351, Sht. 1 (Unit 2)

ISI-351, Sht. 3 (Unit 3)

2. APPLICABLE CODE EDITION AND ADDENDA:

The Inservice Inspection program is based on the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code,Section XI, 2001 Edition through the 2003 Addenda.

3. APPLICABLE CODE REQUIREMENT:

The following Code requirements are paraphrased from the 2001 Edition through the 2003 Addenda of ASME Section XI.

ASME,Section XI, Table IWC-2500-1, Examination Category C-H, Code Item No.

C7.10, requires the performance of a system leakage (visual (VT-2) examination) in accordance with IWC-5220 on Class 2 piping required to operate or support the safety function up to the first normally closed valve, or valve capable of automatic closure.

4. IMPRACTICALITY OF COMPLIANCE:

Pursuant to 10CFR50.55a(g)(5)(iii), relief is requested on the basis that compliance with Section XI requirements is impractical.

The Reactor Vessel Head Flange Leak Detection Piping is separated from the reactor pressure boundary by one passive seal, an O-ring located on the vessel flange. A second O-ring is located on the outer side of the vessel flange, on the opposite side of the tap in the vessel flange (See Figure I4R-25.1 and I4R-25.2). This piping is required during plant operation to indicate failure of the inner flange O-ring seal. Failure of the O-ring results in a High Pressure Alarm in the Main Control Room. Failure of the inner O-ring is the only condition under which this piping is pressurized.

The configuration of this system precludes manual pressure testing while the vessel head is removed because the orifice in the vessel flange and the associated 1 inch piping and

Relief Request I4R-25 for Pressure Testing the RPV Head Flange Seal Weld Leak Detection System In Accordance with 10CFR50.55a(g)(5)(iii)

(Page 2 of 5) components do not incorporate a means for pressure testing. As shown in Figure I4R-25.2, the configuration of the vessel tap, combined with the small size of the tap and the high-test pressure requirement (approximately 1045 psig), precludes the tap from being temporarily plugged or connected to other piping. The opening of the flange is only 3/16 of an inch in diameter and is smooth walled, making the effectiveness of a temporary seal at the required test pressure very limited. Failure of a temporary pressure seal could possibly cause ejection of the device.

The configuration also precludes pressure testing with the vessel head installed, because the seal prevents complete water filling of the piping due to the absence of a vent path.

Additionally, a pneumatic test performed with the head installed is not recommended due to the configuration of the top head. The top head of the vessel contains two grooves that hold the O-rings. The O-rings are in turn held in place within these grooves by a series of retainer clips that are housed in recessed cavities in the flange face. If a pressure test were performed with the head on, the inner O-ring would be pressurized in a direction opposite to what it would see in normal operation. This test pressure would result in a net inward force on the inner O-ring that would tend to push it into the recessed cavities that house the retainer clips. This inward force would very likely damage the thin O-ring material.

In addition to the problems associated with the O-ring design that preclude this testing, it is also questionable whether a pneumatic test is appropriate for this piping. The use of a pneumatic test performed at RPV nominal operating pressure would represent an unnecessary safety risk to personnel in the unlikely event of a test failure, due to the large amount of stored energy contained in pressurized air.

5. BURDEN CAUSED BY COMPLIANCE:

Pressure testing of this Class 2 piping during the Class 1 System Leakage Test is prevented because the piping is not connected to the source of pressure (the leak-off piping is only pressurized in the event of a failure of the inner O-ring). As discussed in section 4.0 above, the extensive modifications required to accommodate testing, without affecting the integrity of the O-ring seal, is impractical.

Based on the above, Peach Bottom Atomic Power Station requests the following alternative examination be performed on the Reactor Vessel Head Flange Seal Leak Detection System.

6. PROPOSED ALTERNATIVE AND BASIS FOR USE:

A visual examination (VT-2) on the Class 2 portion of the RPV Flange Leak Detection Piping will be performed during each refueling outage when the RPV head is off and the head cavity is flooded above the vessel flange. The static head developed with the leak detection piping filled with water will allow for the detection of any leaks in the piping.

This examination will be performed per the frequency specified by Table IWC-2500-1.

Relief Request I4R-25 for Pressure Testing the RPV Head Flange Seal Weld Leak Detection System In Accordance with 10CFR50.55a(g)(5)(iii)

(Page 3 of 5)

7. DURATION OF PROPOSED ALTERNATIVE:

Relief is requested for the fourth ten-year inspection interval for Peach Bottom Atomic Power Station, Units 2 and 3. The fourth ISI interval begins on November 5, 2008 for both Units as proposed in Relief Request I3R-45, and will conclude November 4, 2018 for Peach Bottom Atomic Power Station, Units 2 and 3.

8. PRECEDENTS:

Similar relief requests have been approved for:

1. Peach Bottom Atomic Power Station Third Inspection Interval Relief Request RR-25 was authorized per SER on July 31, 2000.

2. LaSalle County Station Second Inspection Interval Relief Request PR-04 was authorized per SER dated July 3, 1996.

3. Susquehanna Steam Electric Station Third Inspection Interval Relief Request 3RR-07 was authorized per SER dated September 24, 2004.

Relief Request I4R-25 for Pressure Testing the RPV Head Flange Seal Weld Leak Detection System In Accordance with 10CFR50.55a(g)(5)(iii)

(Page 4 of 5)

FIGURE I4R-25.1 REACTOR PRESSURE VESSEL HEAD FLANGE LEAK-OFF PIPING CONFIGURATION S E E F IG . I4R -25.2

Relief Request I4R-25 for Pressure Testing the RPV Head Flange Seal Weld Leak Detection System In Accordance with 10CFR50.55a(g)(5)(iii)

(Page 5 of 5)

FIGURE I4R-25.2 REACTOR PRESSURE VESSEL HEAD FLANGE LEAK-OFF PIPING DETAILS

Relief Request I4R-46 for Pressure Testing of Buried Piping In Accordance with 10 CFR 50.55a(a)(3)(ii)

(Page 1 of 4)

1. ASME CODE COMPONENTS AFFECTED:

Code Class: 3

Reference:

IWA-5244 Examination Category: D-B Item Number: D2.10

Description:

Alternative Examination Requirements of ASME Section XI, IWA-5244, "Buried Components" Component Number(s): Return Piping: 2-32HF-24, 3-32HF-24, 0-33HF-20 Drawing Number(s): ISI-330, Sheet 1; ISI-315, Sheet 1; ISI-315, Sheet 3

2. APPLICABLE CODE EDITION AND ADDENDA:

The Inservice Inspection program is based on the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code,Section XI, 2001 Edition through the 2003 Addenda.

3. APPLICABLE CODE REQUIREMENT:

The following Code requirements are paraphrased from the 2001 Edition through the 2003 Addenda of ASME Section XI.

IWA-5244(b)(1) requires buried components that are isolable by means of valves be tested to determine the rate of pressure loss. Alternatively, the test may determine the change in flow between the ends of the buried components and the Owner shall establish the acceptable rate of pressure loss or flow.

4. REASON FOR REQUEST:

Pursuant to 10 CFR 50.55a(a)(3)(ii), relief is requested on the basis that compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

The buried piping in question consists of one 20 diameter common (i.e., Unit 0) return header for Emergency Service Water (ESW) downstream of MO-0-33-0498 to the discharge pond; and two 24 diameter return headers for High Pressure Service Water (HPSW) downstream of MO-2-32-2486 and MO-3-32-3486 to the discharge pond (one for Unit 2 and one for Unit 3). These components are all buried between the Motor Operated Valves and the discharge pond with the exception of a valve pit that includes a manually operated gate valve and small amount of the associated piping on each of the HPSW and ESW returns to the discharge pond. There is no access to the buried sections without excavation. In addition, no annulus was provided during original construction that would allow for examination of these buried sections of piping.

Relief Request I4R-46 for Pressure Testing of Buried Piping In Accordance with 10 CFR 50.55a(a)(3)(ii)

(Page 2 of 4)

IWA-5244(b)(1) requires that the buried sections of piping be examined by a pressure decay test or a test that determines the change in flow between the buried ends. In order to perform a pressure decay test, it would be necessary to close the associated large gate valves1 to isolate the buried portion of each return header. This would also result in the isolation of portions of the Emergency Core Cooling Systems (ECCS), which would place Technical Specification limitations on the plant. These gate valves in the return headers are not expected to provide the leak tight capability which would be necessary to perform a pressure decay test due to the age of the valves. In order to perform a pressure decay test, it would be necessary to either replace these gate valves with valves that possess better leakage characteristics or to install blind flanges on the piping.

The other potential test would be a change in flow test. However, the buried ECCS return headers were not designed with the plant instrumentation and flow orifices that would be required to determine the flow rates. Installation of flow measurement devices would result in plant modifications.

Compliance with the specified requirements is a hardship without a compensating increase in the level of quality and safety as discussed above.

5. PROPOSED ALTERNATIVE AND BASIS FOR USE:

For the HPSW and ESW buried piping sections Peach Bottom Atomic Power Station proposes to utilize the requirements of IWA-5244(b)(2) along with additional data obtained during quarterly Inservice Testing (IST) trending to provide an adequate level of quality and safety. The IWA-5244(b)(2) requirements call for a test that confirms flow is unimpaired in nonisolable buried components. To confirm that flow is unimpaired in these buried pipes, Peach Bottom Atomic Power Station Inservice Testing procedures will be used to ensure adequate flow during operation.

Peach Bottom Atomic Power Station will use the Owner established minimum flow rate contained in the site IST surveillances as the acceptance criteria for IWA-5244 pressure testing of HPSW and ESW system buried piping.

If the minimum flow could not be achieved during the course of an IST surveillance, and the cause of the deviation was not attributed to the test instruments being used, the associated system would be declared inoperable as required under the IST surveillance and an Issue Report (IR) would be generated in accordance with the Exelon Corrective Action Program. Further corrective actions (i.e., maintenance on the pump, system walk downs, etc.) would be initiated as required to restore the pump and/or the system back to an operable condition.

1 Unit 2: HPSW: HV-2-032-22000 Unit 3: HPSW: HV-3-032-32200 Common: ESW: HV-0-033-11200

Relief Request I4R-46 for Pressure Testing of Buried Piping In Accordance with 10 CFR 50.55a(a)(3)(ii)

(Page 3 of 4)

Fig I4R-46-1

Relief Request I4R-46 for Pressure Testing of Buried Piping In Accordance with 10 CFR 50.55a(a)(3)(ii)

(Page 4 of 4)

6. DURATION OF PROPOSED ALTERNATIVE:

Relief is requested for the fourth ten-year inspection interval for Peach Bottom Atomic Power Station, Units 2 and 3. The fourth ISI interval is scheduled to begin on November 5, 2008 as proposed in relief request I3R-45 and will conclude November 4, 2018 for Peach Bottom Atomic Power Station, Units 2 and 3.

7. PRECEDENTS:

A similar relief request was previously approved for Byron and Braidwood Stations (M.

L. Marshall, Jr. (NRR) to C.M. Crane (Exelon), "Byron Station, Unit Nos 1 and 2, and Braidwood Station, Unit Nos. 1 and 2 Evaluation of Inservice Inspection Program Relief Requests I3R-07 and I2R-46 pertaining to Essential Service Water Buried Piping (TAC NOS, MD1757, MD1758, MD1760)," dated January 16, 2007).

Relief Request I4R-47 for Testing of Control Rod Drive Pressure Boundaries In Accordance with 10 CFR 50.55a(a)(3)(ii)

(Page 1 of 3)

1. ASME CODE COMPONENTS AFFECTED:

Code Class: 1

Reference:

IWB-5222(b)

Examination Category: B-P Item Number: B15.10

Description:

Alternative Examination Requirements of ASME Section XI, Table IWB-2500-1, "Pressure Retaining Boundary" Component Number(s): Class 1 piping between CV-2(3)-03A-13-127 (valves AA through HC inclusive, total of 185 valves) and HV-2(3)-

03A-13112 (valves AA through HC inclusive, total of 185 valves)

Drawing Number(s): ISI-357, Sheet 1 & 2

2. APPLICABLE CODE EDITION AND ADDENDA:

The Inservice Inspection program is based on the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code,Section XI, 2001 Edition through the 2003 Addenda.

3. APPLICABLE CODE REQUIREMENT:

The following Code requirements are paraphrased from the 2001 Edition through the 2003 Addenda of ASME Section XI.

IWB-5222(b) requires the pressure retaining boundary during the system leakage test conducted at or near the end of each inspection interval be extended to all Class 1 pressure retaining components within the system boundary.

4. REASON FOR REQUEST:

Pursuant to 10 CFR 50.55a(a)(3)(ii), relief is requested on the basis that compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

The piping in question is the Class 1 piping between the CV-2(3)-03A-13-127 valve (valves AA through HC inclusive, total of 185 valves) and the HV-2(3)-03A-13112 valve (valves AA through HC inclusive, total of 185 valves) for each of the 185 Control Rod Drive Mechanisms. During normal system lineup required for startup, the CV-2(3)-03A-13-127 valves are in the closed position. The HV-2(3)-03A-13112 valves are in the open position. The only time the CV-2(3)-03A-13-127 valves are open is during a plant scram or during CRD Scram Time testing.

During the performance of the system leakage test prior to plant startup following a refueling outage, the test boundary is the reactor coolant boundary with all valves in the position required for normal reactor operation startup (the CV-2(3)-03A-13-127 valves

Relief Request I4R-47 for Testing of Control Rod Drive Pressure Boundaries In Accordance with 10 CFR 50.55a(a)(3)(ii)

(Page 2 of 3) are closed). This test is conducted in accordance with Table IWB-2500-1 Item B15.10 and IWB-5222(a). As required by IWB-5222(a), the visual examination is to extend to the second closed Class 1 valve at the extremity of the boundary. This test is conducted in accordance with the code.

However, during the system leakage test conducted at or near the end of the ten-year ISI interval in accordance with IWB-5222(b), the pressure boundary extends to all Class 1 pressure retaining components which includes the piping between CV-2(3)-03A-13-127 and the Scram Discharge Volume (HV-2(3)-03A-13112). In order to pressurize the piping between the CV-2(3)-03A-13-127 valves and the HV-2(3)-03A-13112 valves and hold it for inspection, all 185 HV-2(3)-03A-13112 valves would have to be manually closed prior to inserting a scram which is a manpower intensive activity and creates hardship.

Alternately, the piping could be pressurized for testing by isolating each of the 185 segments of piping, and pressurizing with a manual hydro pump. This approach would involve filling and venting of the subject piping, and manipulating 4 valves and installing a fill hose at a threaded connection for each of the 185 piping segments. This activity is also manpower intensive, and creates hardship.

Compliance with the specified requirements is a hardship without a compensating increase in the level of quality and safety because this section of piping does not pressurize to nominal reactor operating pressure except for a very brief time during plant Scrams and Scram Time testing similar to normal operating conditions.

If this piping were to develop a leak, it would be identified during the Scram Time testing by the personnel performing the testing. The piping between the HV-2(3)-3A-13112 valves and the CHK-2(3)-3A-13114 valves is pressurized and tested during the completion of the Scram Discharge Volume System leakage tests.

Table IWB-2500-1 Item B 15.10, Note 2, states that the system leakage test shall be conducted prior to plant startup following each reactor refueling outage. Scram Time testing is required to be 100% complete for each outage prior to exceeding 40% Reactor Thermal Power per Peach Bottom Technical Specifications 3.1.4.

The piping in question is approximately 24 inches of 3/4 inch nominal OD schedule 80 stainless steel socket welded piping for each control rod drive. This piping is not susceptible to a corrosive environment nor is it susceptible to vibrations that would induce cracking. There have been no known leaks in this piping at PBAPS, Units 2 and 3.

Relief Request I4R-47 for Testing of Control Rod Drive Pressure Boundaries In Accordance with 10 CFR 50.55a(a)(3)(ii)

(Page 3 of 3)

Figure I4R-47-1

5. PROPOSED ALTERNATIVE AND BASIS FOR USE:

For the portion of the piping between the CV-2(3)-03A-13-127 valves and the HV-2(3)-

03A-13112 valves for each of the 185 Control Rod Drives, Peach Bottom Atomic Power Station proposes to use the Scram Time Testing that is performed for each Control Rod Drive prior to achieving 40% power at the conclusion of the outage at or near the end of the interval as a means to test this piping. A VT-2 qualified individual will be present during the Scram Time Testing.

6. DURATION OF PROPOSED ALTERNATIVE:

Relief is requested for the fourth ten-year inspection interval for Peach Bottom Atomic Power Station, Units 2 and 3. The fourth ISI interval is scheduled to begin on November 5, 2008 as proposed in relief request I3R-45 and will conclude November 4, 2018 for Peach Bottom Atomic Power Station, Units 2 and 3.

7. PRECEDENTS:

None

Relief Request CRR-12 for the Overlap of the First and Second Ten-Year CISI Intervals for Unit 3 for Class MC Components in Accordance with 10CFR50.55a(a)(3)(i)

(Page 1 of 5)

1. ASME CODE COMPONENTS AFFECTED:

Code Class: MC

Reference:

IWA-2430 IWA-2432 Examination Category: All Item Number: All

Description:

Overlap of the First and Second Ten-Year Containment Inservice Inspection (CISI) Intervals for Unit 3 Class MC Components.

Component Number: All Class MC Components

2. APPLICABLE CODE EDITION AND ADDENDA:

The Second Containment Inservice Inspection (CISI) interval will be based on the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV)

Code,Section XI, 2001 Edition through the 2003 Addenda.

The current Peach Bottom Atomic Power Station, Units 2 and 3 First CISI interval complies with the 1992 Edition through the 1992 Addenda of Subsections IWA and IWE.

3. APPLICABLE CODE REQUIREMENT:

The following Code requirements are paraphrased from the 2001 Edition through the 2003 Addenda of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components.

o Paragraph IWA-2430(b), Inspection Intervals, requires the inspection interval to be determined by calendar years following placement of the plant into commercial service.

o Paragraph IWA-2430(d)(2), Inspection Intervals, requires that examinations may be performed to satisfy the requirements of the extended interval in conjunction with examinations performed to satisfy the requirements of the successive interval.

However, an examination performed to satisfy requirements of either the extended interval or the successive interval shall not be credited to both intervals.

o Paragraph IWA-2432, Inspection Program B, requires that each inspection interval consist of a ten-year duration, except as modified by IWA-2430(d)(1) which permits the inspection interval to be reduced or extended by as much as one year, provided that successive intervals are not altered by more than one year from the original pattern of intervals. If an inspection interval is extended, neither the start and end dates nor the inservice inspection program for the successive interval need be revised.

The 1992 Edition through 1992 Addenda of the ASME B&PV Code Section XI, IWA-2430(d), Inspection Intervals, does not contain the statement that if an inspection

Relief Request CRR-12 for the Overlap of the First and Second Ten-Year CISI Intervals for Unit 3 for Class MC Components in Accordance with 10CFR50.55a(a)(3)(i)

(Page 2 of 5) interval is extended, neither the start and end dates nor the inservice inspection program for the successive interval need be revised. This provision would have permitted the overlap of the intervals, and avoided the need for this relief.

This relief request applies to the first CISI interval.

4. REASON FOR REQUEST:

Pursuant to 10CFR50.55a(a)(3)(i), relief is requested from the ten-year interval requirements contained within IWA-2430 (b) and (d) and IWA-2432 for the first 10-year Peach Bottom Atomic Power Station, Unit 3 CISI program on the basis that the proposed alternative would provide an acceptable level of quality and safety.

Specifically, relief is being sought to overlap the duration of the Peach Bottom Atomic Power Station, Unit 3 first and second CISI intervals as permitted in IWA-2430(d) of the 2001 Edition through the 2003 Addenda in order to commence the second interval on time and keep it aligned with the Unit 2 CISI interval while finishing the first CISI interval.

In addition, this will also allow the CISI and ISI intervals to remain aligned by start and end dates as well as on the same edition of the Code.

Initially, the Peach Bottom Atomic Power Station CISI first ten-year interval for both Units 2 and 3 was scheduled to end on November 4, 2008. Currently, the Peach Bottom Atomic Power Station, Unit 3 first CISI interval is scheduled to end on November 4, 2009 based on a one-year extension that was invoked in October 2007 in accordance with IWA-2430(d) of the 1992 Edition through 1992 Addenda. The one-year extension was required in order to complete a visual inspection (VT-3) of the Unit 3 Torus vent header downcomers, and pressure boundary structural attachment welds. These inspections were not completed in the fall 2007 outage (P3R16) due to poor water clarity. This creates a one-year gap between the two units CISI programs, which may result in different governing Code Editions, different program requirements, and the need for different parallel implementing procedures.

The following Tables 1 and 2 describe the current CISI interval dates and the proposed CISI dates as requested in this relief:

Table 1 - Current CISI Interval Dates First CISI Interval End Dates Second CISI Interval Start Dates Unit 2 - November 4, 2008 Unit 2 - November 5, 2008 Unit 3 - November 4, 2009 Unit 3 - November 5, 2009

Relief Request CRR-12 for the Overlap of the First and Second Ten-Year CISI Intervals for Unit 3 for Class MC Components in Accordance with 10CFR50.55a(a)(3)(i)

(Page 3 of 5)

Table 2 - Proposed CISI Interval Dates First CISI Interval End Dates Second CISI Interval Start Dates Unit 2 - November 4, 2008 Unit 2 - November 5, 2008 Unit 3 - November 4, 2009 Unit 3 - November 5, 2008*

• As requested in this relief, the second CISI interval for Unit 3 will overlap the first interval for approximately one year. This overlap is not permitted in the 1992 Edition through 1992 Addenda of ASME Section XI.

Any examinations required to complete the remainder of the Peach Bottom Atomic Power Station, Unit 3 first interval under the previous CISI interval Code (i.e., ASME Section XI, 1992 Edition through 1992 Addenda), will be completed in the next refueling outage scheduled to occur in Fall of 2009. The examinations will be conducted and credited under the rules of the code of record applicable to the first interval. The examinations, which are also required in the second interval under the second CISI interval Code (i.e., ASME Section XI, 2001 Edition through 2003 Addenda), will be scheduled to be completed at the end of the second interval. This method of scheduling will maintain the original sequence of examinations and thus will not affect the frequency of examination.

The supplementary information contained within Section 2.2 of former Final Rule (67 FR 60520) dated September 26, 2002, contains statements supporting the proposed alternative for modifying the CISI Interval. Specifically, the information pointed out that 10CFR50.55a(g)(4)(ii) does not prohibit licensees from updating to a later Edition and Addenda of the ASME Code midway through a ten-year IWE examination interval.

Additionally, the supplementary information advised that licensees wishing to synchronize their 120-month intervals may submit a request in accordance with Section 50.55a(a)(3) to obtain authorization to extend or reduce 120-month intervals.

Using the common interval date justified above based on the current Unit 2 CISI program end date of November 4, 2008, the code of record for the second interval CISI Program is to be set on November 4, 2007 (i.e., 12 months prior to the start of the successive interval in accordance with 10CFR50.55a(g)(4)(ii)). On November 4, 2007, the latest edition and addenda of the code incorporated by reference in 10CFR50.55a(b)(2) of the regulation was the 2001 Edition through the 2003 Addenda. Thus, Peach Bottom Atomic Power Station will utilize the 2001 Edition through the 2003 Addenda of ASME Section XI to develop the CISI program for the second CISI interval.

As noted in the Applicable Code Requirements, the 2001 Edition through 2003 Addenda of the Code does permit overlap of the intervals. However, the 1992 Edition

Relief Request CRR-12 for the Overlap of the First and Second Ten-Year CISI Intervals for Unit 3 for Class MC Components in Accordance with 10CFR50.55a(a)(3)(i)

(Page 4 of 5) through 1992 Addenda, which is applicable to the first CISI interval does not allow overlap of the intervals.

In conclusion, the proposed alternative as described herein provides an acceptable level of quality and safety, and does not adversely impact the health and safety of the public.

5. PROPOSED ALTERNATIVE AND BASIS FOR USE:

As an alternative to the full ten-year interval duration requirements of IWA-2430(b) and (d) and IWA-2432 for the Unit 3 first and second CISI intervals, Peach Bottom Atomic Power Station proposes to modify the interval dates of the Unit 3 first CISI Interval as discussed below. This will permit the Unit 2 and Unit 3 CISI programs to share a common inspection interval and to implement common Code Editions for Class MC components. The common code of record for the second interval CISI programs will be the 2001 Edition through the 2003 Addenda of ASME Section XI.

As a result of these interval modifications, the finish date of the first interval CISI Program will be November 4, 2009 for the Peach Bottom Atomic Power Station, Unit 3.

The start date for the second interval CISI program will be November 5, 2008 for Peach Bottom Atomic Power Station, Unit 3. Using this date, the Peach Bottom, Unit 3 Fall 2009 refueling outage (P3R17) will be the last refueling outage of the first interval and the first refueling outage of the second interval. The intervals will be scheduled in 10-year increments from this point forward with the modifications allowed by IWA-2430 fully available to future intervals and periods of the adjusted programs (Peach Bottom Atomic Power Station, Units 2 and 3 CISI) based on this new common interval date replacing the sequence previously established for the respective units.

Any examinations required to complete the remainder of the Peach Bottom Atomic Power Station, Unit 3 first interval under the previous CISI interval Code (i.e., ASME Section XI, 1992 Edition through 1992 Addenda), will be completed in the next refueling outage. The examinations will be conducted and credited under the rules of the code of record applicable to the first interval and not credited for the second interval. The examinations, which are also required in the second interval under the second CISI interval Code (i.e., ASME Section XI, 2001 Edition through 2003 Addenda), will be scheduled to be completed at the end of the second interval. This method of scheduling will maintain the original sequence of examinations and thus will not affect the frequency of examination.

6. DURATION OF PROPOSED ALTERNATIVE:

Relief is requested for the Peach Bottom Atomic Power Station, Unit 3 first CISI interval.

The second interval CISI program will begin November 5, 2008 and will conclude November 4, 2018 for Peach Bottom Atomic Power Station, Units 2 and 3 as requested in this relief request. On November 4, 2007, the latest edition and addenda of the code incorporated by reference in 10CFR50.55a(b)(2) of the regulation was the 2001 Edition through the 2003 Addenda.

Relief Request CRR-12 for the Overlap of the First and Second Ten-Year CISI Intervals for Unit 3 for Class MC Components in Accordance with 10CFR50.55a(a)(3)(i)

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7. PRECEDENTS:

Similar relief requests have been approved as follows:

1. Byron Station, Unit Nos. 1 and 2 Relief Request I3R-01 was approved in NRC Safety Evaluation Report (SER) dated September 7, 2006.

2. Limerick Generating Station, Units 1 and 2 Relief Request I3R-01 was approved in NRC SER dated January 24, 2007.

3. Susquehanna Steam Electric Station, Units 1 and 2 Relief Request 3RR-10 was approved in NRC SER dated September 24, 2004.

Relief Request CRR-13 for Examination of the N-3 Penetration (Containment Inservice Inspection Program) In Accordance with 10CFR50.55a(g)(5)(iii)

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1. ASME CODE COMPONENTS AFFECTED:

Code Class: MC

Reference:

IWE-1232 Examination Category: E-A Item Number: E1.11, E1.12

Description:

Alternative Examination Requirements of ASME Section XI, IWE-1232, "Inaccessible Surface Areas" Component Number(s): Penetration N-3 Drawing Number(s): 6280-S-53, 6280-C2-103-6, and 6280-C2-341-3

2. APPLICABLE CODE EDITION AND ADDENDA:

The first Containment Inservice Inspection (CISI) interval began on November 5, 1998 for PBAPS, Units 2 and 3, and complied with the ASME B&PV Code,Section XI, 1992 Edition through 1992 Edition Addenda.

The second CISI interval will begin on November 5, 2008, and will comply with the ASME B&PV Code,Section XI, 2001 Edition through 2003 Addenda.

The original construction code for the Drywell isSection III, Type B, 1965 Edition through Summer 1966 Addenda.

3. APPLICABLE CODE REQUIREMENT:

The following Code requirements are paraphrased from the ASME B&PV Code,Section XI, 2001 Edition through the 2003 Addenda.

IWE-1232(a) (Inaccessible Surface Areas) states that portions of Class MC containment vessels, parts, and appurtenances that are embedded in concrete or otherwise made inaccessible during construction of the vessel or as a result of vessel repair, modification, or replacement are exempted from examination, provided:

(1) no openings or penetrations are embedded in the concrete; (2) all welded joints that are inaccessible for examination are double butt welded and are fully radiographed and, prior to being covered, are tested for leak tightness using a gas medium test, such as Halide Leak Detector Test; and (3) the vessel is leak rate tested after completion of construction or repair/replacement activities to the leak rate requirements of the Design Specifications.

Relief Request CRR-13 for Examination of the N-3 Penetration (Containment Inservice Inspection Program) In Accordance with 10CFR50.55a(g)(5)(iii)

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4. IMPRACTICALITY OF COMPLIANCE:

Pursuant to 10CFR50.55a(g)(5)(iii), relief is requested on the basis that conformance with these code requirements is impractical as conformance would require extensive modifications to the primary containment.

When the drywell was being constructed, a 24-inch manhole was placed in the bottom head of the drywell. During construction, when the manhole was no longer needed, the penetration was seal welded, inspected, and embedded in concrete.

Based on the original construction drawings the manhole is a bolted, gasket connection that was seal welded, the handles were ground smooth and either a magnetic particle test or dye penetrant examination was performed. The N-3 manhole was seal welded and cannot meet the IWE-1232(a)(2) code requirement for a double butt weld.

5. BURDEN CAUSED BY COMPLIANCE:

Adding a double butt weld would involve a modification to the drywell that would require excavation of the concrete around the bottom head of the drywell or removal of the drywell floor thus making the code requirement impractical.

6. PROPOSED ALTERNATIVE AND BASIS FOR USE:

Integrated Leak Rate Testing will be performed in accordance with the station Appendix J Program, which is maintained independent of the ASME Section XI program.

7. DURATION OF PROPOSED ALTERNATIVE:

This relief is being requested for the first (previous) and second (upcoming) containment inservice inspection (CISI) interval.

The first CISI interval began on November 5, 1998 for PBAPS, Units 2 and 3, and complied with the ASME B&PV Code,Section XI, 1992 Edition through 1992 Edition Addenda. The second CISI interval will begin on November 5, 2008 for PBAPS, Units 2 and 3, and will comply with the ASME B&PV Code,Section XI, 2001 Edition through 2003 Addenda. Refer to CRR-12 for the associated start and end dates of the intervals.

8. PRECEDENTS:

None

Relief Request CRR-13 for Examination of the N-3 Penetration (Containment Inservice Inspection Program) In Accordance with 10CFR50.55a(g)(5)(iii)

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FIGURE CRR-13-1

Relief Request CRR-13 for Examination of the N-3 Penetration (Containment Inservice Inspection Program) In Accordance with 10CFR50.55a(g)(5)(iii)

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FIGURE CRR-13-2