ML081400722

Fourth Ten-Year Interval Inservice Inspection Program, Submittal of Relief Request Numbers 18, 19, 20, and 21

ML081400722
Person / Time
Site:	Ginna
Issue date:	05/10/2008
From:	Joseph Pacher Constellation Energy Nuclear Group
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML081400722 (39)

Text

Joe Pacher R.E. Ginna Nuclear Power Plant, LLC Manager, Nuclear Engineering Services 1503 Lake Road Ontario, New York 14519-9364 585.771.5208 585.771.3392 Fax joseph.pacher@constellation.com Co nstatel ti nergy Nuclear Generation Group May 10, 2008 U. S. Nuclear Regulatory Commission Washington, DC 20555 ATTENTION: Document Control Desk

SUBJECT:

R.E. Ginna Nuclear Power Plant Docket No. 50-244 Fourth Ten-Year Interval Inservice Inspection Program Submittal of Relief Request Numbers 18, 19, 20, and 21 Pursuant to 10 CFR 50.55a(a)(3)(i) and 10 CFR 50.55a (a)(3)(ii), R.E. Ginna Nuclear Power Plant, LLC (Ginna LLC) hereby requests NRC approval of the following requests for the Fourth Ten-Year Interval Inservice Inspection Program:

Ginna LLC is requesting relief to defer related reactor vessel Examination Category B-A, B-D, B-F, B-K, B-N-2 and B-N-3 examinations from the 2009 outage to the 2011 outage. The intent of the extension in the requests is to allow for deferment of the subject examinations by approximately six months beyond the Code allowed 12 month extension and to allow for the subject examinations to be performed at the same time, along with other reactor vessel related examinations as discussed with the staff on June 27, 2006. Inspecting all of these welds during the same outage will result in a reduction in man-rem exposure and examination costs. The details of the 10 CFR 50.55a requests are enclosed.

Ginna LLC requests NRC approval of the attached relief requests by April 2009 to support planning for the fall 2009 refueling outage.

Should you have questions regarding this matter, please contact Tom Harding (585) 771-3384, or Thomas.harding@constellation.com.

V y truly yours, oseph E. Pacher Enclosures cc: S. J. Collins, NRC D. V. Pickett, NRC Resident Inspector, NRC ,40(J7

RELIEF REQUEST NO. 18 R.E. Ginna Nuclear Power Plant - Fourth Interval ISI Program Defer RPV Category B-A and B-D Exams from 2009 to 2011 Outage

RELIEF REQUEST NO. 18 R.E. Ginna Nuclear Power Plant - Fourth Interval ISI Program Defer RPV Category B-A and B-D Exams from 2009 to 2011 Outage Proposed Alternative In Accordance with 10 CFR 50.55a(a)(3)(i)

-Alternative Provides Acceptable Level of Quality and Safety-

1. ASME Code Component(s) Affected The affected component is the R.E. Ginna reactor vessel, specifically the following American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (BPV) Code Section XI (Reference 1) examination categories and item numbers covering examinations of the reactor vessel (RV). These examination categories and item numbers are from IWB-2500 and Table JWB-2500-1 of the ASME BPV, Code Section XI.

Examination Category Item No. Description B-A B 1.1 I Circumferential Shell Welds B-A B 1.30 Shell-to-Flange Weld B-D B3.90 Nozzle-to-Vessel Welds B-D B3. 100 Nozzle Inner Radius Areas (Throughout this request the above examination categories are referred to as "the subject examinations" and the ASME BPV Code,Section XI, is referred to as "the Code.")

2. Applicable Code Edition and Addenda

ASME Code Section XI, "Rules and Inservice Inspection of Nuclear Power Plant Components," 1995 Edition with the 1996 Addenda.

3. Applicable Code Requirement

IWB-2412, Inspection Program B, requires volumetric examination of essentially 100% of reactor pressure vessel pressure retaining welds identified in Table IWB-2500-1 once each ten year interval.

IWA-2430(d) allows inspection intervals to be extended by as much as one year if this adjustment does not cause successive intervals to be altered by more than one year.

4. Reason for Request

Relief Request Number 18, which is being submitted along with Relief Request Numbers 19, 20 and 21, deals with associated Reactor Pressure Vessel examinations to be deferred from the 2009 Refueling Outage to the 2011 Refueling Outage. This request is for a less than six month interval extension beyond the Code allowed 12 month extension (IWA-2430(d)) to coincide with the next planned refueling outage in 2011. The intent of the extension in the requests is to allow for deferment of the subject examinations by one refueling cycle and allow for the subject examinations to be performed at the same time, along with other reactor vessel related examinations.

Page 1 of 5

RELIEF REQUEST NO. 18 R.E. Ginna Nuclear Power Plant - Fourth Interval ISI Program Defer RPV Category B-A and B-D Exams from 2009 to 2011 Outage

5. Proposed Alternative and Basis for Use R.E. Ginna Nuclear Power Plant proposes to perform the subject examinations for the fourth inspection interval one refueling cycle beyond the-end of the fourth interval. The fourth inspection interval for R.E.

Ginna started January 1, 2000 and is currently scheduled to end on December 31, 2009. The subject examinations are currently scheduled to be performed during the Fall 2009 refueling outage. The inspections are proposed to be performed in the subsequent refueling outage in Spring 2011. This inspection date is less than 6 months beyond the ten year Code inspection interval and the one year interval extension provided by IWA-2430(d).

In accordance with 10 CFR 50.55a(a)(3)(i), this interval extension is requested on the basis that the current inspection interval can be extended based on a negligible change in risk by satisfying the risk criteria specified in Regulatory Guide 1.174 (Reference 2).

The methodology used to demonstrate the acceptability of extending the fourth inspection interval for Category B-A and B-D welds based on a negligible change in risk is contained in the latest revision of WCAP- 16168-NP (Reference 3). This methodology was used to develop a pilot plant analysis for Westinghouse, Combustion Engineering, and Babcock and Wilcox reactor vessel designs and is an extension of the work that was performed as part of the NRC PTS Risk Re-Evaluation (NUREG- 1874, Reference 4). The critical parameters for demonstrating that this pilot plant analysis is applicable on a plant specific basis, as identified in the latest revision of WCAP- 161 68-NP, are identified in Table 1. By demonstrating that each plant specific parameter is bounded by the corresponding pilot plant parameter, the application of the methodology to the R.E. Ginna reactor vessel is acceptable as shown in Table I below.

Table 1 Critical Parameters for Application of Bounding Analysis Additional Evaluation Parameter Pilot Plant Basis Plant Specific Basis Required?

Dominant Pressurized Thermal NRC PTS Risk Study Ginna is bounded by No Shock (PTS) Transients in the NRC (Reference 4) PTS Generalization PTS Risk Study are applicable Study (Reference 5)

Through Wall Cracking Frequency 1.76E-08 Events per year 7.47E- 12 Events per No (Reference 3) year (Calculated per Reference 4)

Frequency and Severity of Design 7 heatup/cooldowns per year Ginna is bounded by No Basis Transients (Reference 3) 7 heatup/cooldowns per year Cladding Layers (Single/Multiple) Single Layer (Reference 3) Single Layer No Page 2 of 5

RELIEF REQUEST NO. 18 R.E. Ginna Nuclear Power Plant - Fourth Interval ISI Program Defer RPV Category B-A and B-D Exams from 2009 to 2011 Outage Additional information relative to R.E. Ginna reactor vessel inspections is provided in Table 2. This information confirms that satisfactory examinations have been performed on the R.E. Ginna reactor vessel.

Table 2 Additional Information Pertaining to Reactor Vessel Inspection Inspection methodology: Category B-A welds: ASME Section XI Appendix VIII (Reference 1)

Category B-D welds: Regulatory Guide 1.150 (Reference 6)

Information is for most recent inservice inspection performed in 1999.

Number of past inspections: A minimum of 3 inspections have been performed to date on each Category B-A and B-D weld.

Number of indications found: Zero reportable indications have been found to date. Any recordable indications have been acceptable per ASME Section XI IWB-3500. One recordable indication is in the inner 1/8th of the vessel inside diameter in the beltline region. The indication has a depth of 0.22", length of 1.54" and is 0.38" subsurface. This indication is acceptable per the flaw acceptance criteria in the voluntary PTS Rule, 10 CFR 50.61 a (Reference 7).

Proposed inspection schedule for Reactor Pressure Vessel Examination Category B-A and B-D are balance of plant life: currently scheduled for the 2009 outage within the Fourth Interval Inservice Inspection Program. These examinations for the Fourth Interval Inservice Inspection Program are proposed to be performed in the 2011 outage. Future interval examinations will revert back to the normal 10-year schedule.

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RELIEF REQUEST NO. 18 R.E. Ginna Nuclear Power Plant - Fourth Interval ISI Program Defer RPV Category B-A and B-D Exams from 2009 to 2011 Outage Table 3 provides additional information relative to the calculation of the Through-Wall Cracking Frequency (TWCF) for R.E. Ginna.

Table 3 Details of Through-Wall Cracking Frequency (TWCF) Calculation Inputs Reactor Coolant System Temperature, TRCs[°F]: 544.2 Twall (inches): 6.50 Fluence [1019 Region/Component Cu Ni P Mn Un-Irradiated Neutro[icm2, Description [wt%] [wt%] [wt%] [wt%] RTNDT(u) [°F] e> I MeV I E>1 MeV]

I Inter./Lower Circ. Weld Linde 80 .250 .560 .012 1.35 -4.8 5.42 2 Nozzle/Inter. Circ Weld Linde 80 .230 .590 .021 1.35 10.0 0.516 3 Nozzle Forging A508-2 .070 .680 .010 .700 30.0 0.516 4 Lower forging A508-2 .050 .690 .010 .700 40.0 5.42 5 Intermediate Forging A508-2. .070 .690 .010 .700 20.0 5.42 Outputs Methodology Used to Calculate AT30: NUREG- 1874 Region # Fluence [ 1019 * (flux)

(From RTMAX-XX [R] Neutron/cm2 , [Neutrons/

2 AT30 TWCF 95 -xx Above) E>l MeV] cm /sec] [OF]

Forging - FO 4 562.00 5.42 2.86E+10 62.31 7.63E-14 Circumferential Weld - CW 1 684.80 5.42 2.86E+10 229.91 3.40E-12 TWCF95_TOTAL (CXAwTWCF9 5 .AW + (XPLTWCF95_PL + c~cwTWCF95.cw + (xFoTWCF95_FO): 7.47E- 12 The methodology used to demonstrate the acceptability of extending the fourth inspection interval for the Reactor Pressure Vessel, Category B-A and B-D welds, is based on a negligible change in risk as contained in the latest revision of WCAP-16168-NP (Reference 3). By demonstrating that each plant specific parameter is bounded by the corresponding pilot plant parameter, the application of the methodology to the R.E. Ginna reactor vessel is acceptable as shown in the Tables above. WCAP-16168-NP was developed to decrease the frequency of examinations from once every ten years to once every twenty years. This request is for a less than a six month interval extension beyond the Code allowed 12 month extension (IWA-2430(d)) to coincide with the next planned refueling outage in 2011.

6. Duration of Proposed Alternative The alternative is requested for examinations under Category B-A and B-D to extend the fourth inservice inspection interval by less than six months beyond the ASME Code required 10-year inspection interval and Code allowed twelve month extension. This request is applicable to the R.E. Ginna fourth inspection interval only.

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RELIEF REQUEST NO. 18 R.E. Ginna Nuclear Power Plant - Fourth Interval ISI Program Defer RPV Category B-A and B-D Exams from 2009 to 2011 Outage

7. Precedents Similar relief with regards to extending the 10-year inservice inspection (ISI) interval for reactor vessel weld exams was approved for the Palisades Nuclear Plant in correspondence dated September 4, 2007 (ML071770387) and for the Indian Point Nuclear Generating Unit No. 2 in correspondence dated October 29, 2007 (ML072480249).

8. References

1. ASME Boiler and Pressure Vessel Code,Section XI, 1995 Edition with the 1996 Addenda, American Society of Mechanical Engineers, New York.

2. NRC Regulatory Guide 1.174, Revision 1, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis,"

November 2002.

3. WCAP-16168-NP, Revision 1, "Risk-Informed Extension of Reactor Vessel In-Service Inspection Interval," January 2006, and F. P. Schiffley to U. S. NRC, PWR Owners' Group Letter, OG-07-455 "Responses to the NRC Request for Additional Information (RAI) on PWR Owners Group (PWROG) WCAP-16168-NP, Revision 1, 'Risk-Informed Extension of Reactor vessel In-Service Inspection Interval', MUHP-5097/5098/5099, Tasks 2008/2059," October 16, 2007 and Enclosures 1 and 2.

4. NUREG- 1874, "Recommended Screening Limits for Pressurized Thermal Shock," March, 2007.

5. NRC Letter Report, "Generalization of Plant-Specific Pressurized Thermal Shock (PTS) Risk Results to Additional Plants," December 14, 2004.

6. NRC Regulatory Guide 1.150, Revision 1, "Ultrasonic Testing of Reactor Vessel Welds During Preservice and Inservice Examinations," February 1983.

7. SECY-07-0104, "Proposed Rulemaking - Alternate Fracture Toughness Requirements for Protection against Pressurized Thermal Shock," June 25, 2007, Enclosure 1.

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RELIEF REQUEST NO. 19 R.E. Ginna Nuclear Power Plant - Fourth Interval ISI Program Defer RPV Category B-F Exams from 2009 to 2011 Outage

RELIEF REQUEST NO. 19 R.E. Ginna Nuclear Power Plant - Fourth Interval ISI Program Defer RPV Category B-F Exams from 2009 to 2011 Outage Proposed Alternative In Accordance with 10 CFR 50.55a(a)(3)(i)

-Alternative Provides Acceptable Level of Quality and Safety-

1. ASME Code Component(s) Affected The affected components are the R.E. Ginna reactor vessel nozzle-to-safe-end/piping welds, specifically the following ASME Boiler and Pressure Vessel (BPV) Code Section XI (Reference 1) Examination Category and Item Number. This examination category and item number is from IWB-2500 and Table IWB-2500-1 of the ASME BPV Code Section XI.

Examination Category Item No. Description B-F B5.10 Pressure Retaining Dissimilar Metal Welds in Vessel Nozzles (Throughout this request the above examination category is referred to as "the subject examinations" and the ASME BPV Code Section Xl is referred to as "the Code". "Inspections" and "Examinations" may be used interchangeably.)

2. Applicable Code Edition and Addenda

The R.E. Ginna Fourth Interval Inservice Inspection (ISI) Program Plan is prepared to the ASME Section XI Code, 1995 Edition with the 1996 Addenda.

3. Applicable Code Requirement

IWB-2412, Inspection Program B, requires volumetric examination of essentially 100% of reactor pressure vessel pressure retaining welds identified in Table IWB-2500-1 once each ten year interval.

IWA-2430(d) allows inspection intervals to be extended by as much as one year if this adjustment does not cause successive intervals to be altered by more than one year.

4. Reason for Request

Relief Request Number 19, which is being submitted along with Relief Request Numbers 18, 20 and 21, deals with associated Reactor Pressure Vessel examinations to be deferred from the 2009 Refueling Outage to the 2011 Refueling Outage. This request is for a less than six month interval extension beyond the Code allowed 12 month extension (IWA-2430(d)) in order to allow the subject examinations to be performed at the same time as the reactor vessel weld examinations (Relief Request Number 18), along with other reactor vessel related examinations.

Page 1 of 22

RELIEF REQUEST NO. 19 R.E. Ginna Nuclear Power Plant - Fourth Interval ISI Program Defer RPV Category B-F Exams from 2009 to 2011 Outage

5. Proposed Alternative and Basis for Use R.E. Ginna Nuclear Power Plant proposes to perform the subject examinations for the fourth inspection interval one refueling cycle beyond the end of the fourth interval. The fourth inspection interval for R.E.

Ginna started on January 1, 2000 and will end on December 31, 2009. The subject examinations are currently scheduled to be performed during the Fall 2009 refueling outage. The inspections are proposed to be performed in the subsequent refueling outage in Spring 2011. This inspection date is less than 6 months beyond the ten year Code inspection interval and the one year interval extension provided by IWA-2430(d).

In accordance with 10 CFR 50.55a(a)(3)(i), this interval extension is requested on the basis that the current inspection interval can be extended while providing an acceptable level of quality and safety.

The technical justification for the deferral of the subject examinations consists of four areas. These are:

A.) PWR Service Experience B.) R.E. Ginna Inservice Inspection History C.) Deterministic Flaw Growth Analysis D.) Change-in-Risk Estimate Page 2 of 22

RELIEF REQUEST NO. 19 R.E. Ginna Nuclear Power Plant - Fourth Interval ISI Program Defer RPV Category B-F Exams from 2009 to 2011 Outage A.) PWR Service Experience Ginna Station reactor vessel has six (6) dissimilar metal Examination Category B-F welds from the reactor vessel nozzles to the safe-ends/piping/elbows. These welds exist on the reactor vessel outlet nozzles (2), reactor vessel inlet nozzles (2), and the reactor vessel safety injection nozzles (2). These welds are stainless steel welds that do not contain any Alloy 82 or 182 weld material. Therefore, these welds are not susceptible to primary water stress corrosion cracking.

To date, all known incidents of cracking in the PWR fleet in reactor vessel Category B-F welds have been attributed to PWSCC in susceptible Alloy 82 and 182 weld materials. There have been no known incidents of cracking in non-Alloy 82/182 reactor vessel Category B-F welds.

Given this history, it is not expected that cracking will occur in these welds at Ginna Station.

B.) R.E. Ginna Inservice Inspection History R.E. Ginna Nuclear Power Plant is currently in the Fourth Interval ISI Program. The subject examinations have been performed three times for Inservice Inspection. Most recently, these examinations were performed in April of 1999 in accordance with the 1986 Edition of the ASME Boiler and Pressure Vessel Code, Section Xl. Table 1 provides a summary of the inservice inspection results from the last inspection. Due to the improvements in inspection technology with time, the most recent inspection is considered to be of the greatest quality of the three inservice inspections performed. No indications were identified as reflected by Table 1 below.

Table 1: R.E. Ginna Examination Category B-F Inser vice Inspection Results I Weld ID Description PL-FW-II 300 Outlet Nozzle-to-safe-end from nozzle and safe-end side (ISI Summary # I 002100)

AC-1003-1 1050 SI Nozzle-to-safe-end from nozzle and safe-end side (ISI Summary # I 003300)

PL-FW-VII 1500 Inlet Nozzle-to-safe-end from nozzle and safe-end side (ISI Summary # I 003000)

PL-FW-IV 2100 Outlet Nozzle-to-safe-end from nozzle and safe-end side (ISI Summary # I 002700)

AC-1002-1 2850 SI Nozzle-to-safe-end from nozzle and safe-end side (ISI Summary # I 003600)

PL-FW-V 3300 Inlet Nozzle-to-safe-end from nozzle and safe-end side (ISI Summary # I 002400)

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RELIEF REQUEST NO. 19 R.E. Ginna Nuclear Power Plant - Fourth Interval ISI Program Defer RPV Category B-F Exams from 2009 to 2011 Outage C.) Deterministic Flaw Growth Analysis As shown in the inservice inspection results in paragraph "B", there are no known indications in the nozzle-to-safe-end/piping/elbows welds for which the extension is requested. The welds for which the subject examinations are conducted are similar metal low alloy steel welds which are not susceptible to PWSCC. Absent PWSCC, the most credible mechanism for flaws to initiate and grow in these welds is fatigue due to thermal and mechanical cycling from operational transients. ASME cumulative fatigue usage factors were calculated for these locations using a very conservative design duty cycle where the design duty cycle is the combination of the transient characteristics (pressure and temperature with time) and the number of design cycles.

The calculated fatigue usage factors are much less than the ASME Code design limit of 1.0 after 40 years of operation, and typically less than 0.1 in the region. Further, as R.E. Ginna enters the extended license period, these calculated fatigue usage factors will not exceed 1.0 after 60 years of operation since the originally specified number of cycles for 40 years of operation will now be used for the 60 year life. For this reason it is very unlikely that a flaw would have initiated during the 10 years since the last inservice inspection. Given the very small number of transients from the design duty cycle that may occur over the period of the requested extension, it is even more unlikely that any flaws will initiate during the requested extension.

In the unlikely event that a flaw was either missed in the previous inservice inspections (discussed in paragraph "B") or a flaw was initiated since the last inspection, the growth of any existing flaw is expected to be very small over the life of the reactor vessel. For example, flaw evaluation handbooks have been developed and submitted to the NRC for Westinghouse 2 loop plants (References 2 and 3) that have comparable geometries and loading conditions to that of R.E. Ginna. These evaluations, which take into consideration a very conservative design duty cycle, show that even if a surface flaw with an aspect ratio of 6 (I/a) and initial depth of 20%

through wall (a/t) is assumed to exist in any of the subject welds, the flaw will remain acceptable for at least 20 years per the ASME Code,Section XI. Due to the age of R.E. Ginna (nearing 40 years) compared to the earlier flaw growth studies, a change in risk estimate approach was also applied.

D.) Change-in-Risk Estimate To estimate the change in risk associated with extending the inspection interval, computer runs were performed using the Westinghouse Win-SRRA Code. This code was developed for the purpose of estimating piping weld failure probabilities as documented in WCAP-14572, Revision 1-NP-A, Supplement 1, and has been reviewed and approved for use by the NRC (Reference 4). Runs were created for the safety injection, reactor vessel inlet, and the reactor vessel outlet nozzle-to-safe-end/piping/elbows welds using R.E. Ginna specific inputs. The Win-SRRA Code determined the cumulative failure probability over a 40 year plant life. For the runs performed, "failure" was chosen to be a "small leak" which is equivalent to a through wall flaw. The inputs and outputs are identified in Attachment 1 of this request. These inputs were developed based on R.E. Ginna specific geometry and operating conditions in accordance with WCAP-14572, Revision 1-NP-A, Supplement 1 (Reference 4). The inputs used in this evaluation take into Page 4 of 22

RELIEF REQUEST NO. 19 R.E. Ginna Nuclear Power Plant - Fourth Interval ISI Program Defer RPV Category B-F Exams from 2009 to 2011 Outage consideration a recently implemented extended power uprate at R.E. Ginna and are appropriate for the license renewal period.

As mentioned, the Win-SRRA Code only determines failure probabilities out to 40 years. In order to estimate the failure probability into the beginning years of the extended operating license, two different approaches can be taken. The first approach is to divide the cumulative failure probability by 40 years and linear extrapolate the failure probability per year (failure frequency) into the extended license period. The more conservative approach is to base the projection on the maximum yearly increase determined between 0 and 40 years.

As shown by the cumulative failure probability results for the safety injection nozzle-to-safe-end weld, the probability of failure (with ISI) at year 11 is 2.38E-10 while the probability of failure at year 12 is 1.60E-09. This yearly change is the largest experienced in the 40-year life. By taking the difference in the probabilities for years 11 and 12, the change in failure probability for the safety Injection nozzle-to-safe-end weld for the requested interval extension can be conservatively estimated as 1.36E-09 with ISI. Similarly, the change in failure probability for the RV inlet nozzle-to-safe-end weld for the requested interval extension can be estimated as 2.13E-08 based on the difference in failure probability between years 29 and 30. The change in failure probability for the RV outlet nozzle-to-safe-end weld for the requested interval extension is estimated as 5.27E-09 based on the difference in failure probability between years 20 and 21.

If the conditional large early release probability associated with a weld failure is conservatively assumed to be 1.0, the one year change in large early release frequency (LERF) for one safety injection nozzle-to-safe-end weld can be estimated to be equal to 1.36E-09 events per year.

Likewise, the one year change in LERF can for the RV inlet and outlet nozzle-to-safe-end welds can be estimated to be equal to 2.13E-08 events per year and 5.27E-09 events per year, respectively. Since there are two of each type of nozzle, the total change in LERF can be taken as the sum of twice the change in LERF for each nozzle-to-safe-end weld. The change in LERF is then 5.59E-08 events per years. This is less than the Regulatory Guide 1.174 (Reference 5) criteria of 1.OE-07 events per year for a very small increase in LERF. Since the Regulatory Guide 1.174 (Reference 5) criteria for a very small increase in core damage frequency (1.OE-06) is higher than that for LERF, the LERF analysis is bounding.

Based on PWR Service Experience, Ginna Station Inservice Inspection History, Deterministic Flaw Growth Analysis and Change-in-Risk Estimate technical justifications above, the proposed extension for Ginna Station Fourth Interval Inservice Inspection Program on the subject examinations provides an acceptable level of quality and safety.

6. Duration of Proposed Alternative The alternative is requested to extend the Fourth Inservice Inspection Interval by less than 6 months beyond the ASME Code required 10-year inspection interval and Code allowed twelve month extension. This request is applicable to R.E. Ginna fourth inspection interval only.

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RELIEF REQUEST NO. 19 R.E. Ginna Nuclear Power Plant - Fourth Interval ISI Program Defer RPV Category B-F Exams from 2009 to 2011 Outage

7. References

1. ASME Boiler and Pressure Vessel Code, Section Xl, 1995 Edition with the 1996 Addenda, American Society of Mechanical Engineers, New York.

2. WCAP-10363, "Handbook on Flaw Evaluation for Prairie Island Units 1 & 2 Reactor Vessels,"

December 1984.

3. WCAP-1 1477, Revision 1, "Handbook on Flaw Evaluation for Point Beach Units 1 & 2 Reactor Vessels," July 1990.

4. WCAP-14572, Revision 1-NP-A, Supplement 1, "Westinghouse Structural Reliability and Risk Assessment (SRRA) Model for Piping Risk Informed Inservice Inspection," February 1999.

5. NRC Regulatory Guide 1.174, Revision 1, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis,"

November 2002.

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RELIEF REQUEST NO. 19 R.E. Ginna Nuclear Power Plant - Fourth Interval ISI Program Defer RPV Category B-F Exams from 2009 to 2011 Outage Attachment 1 Win-SRRA Inputs and Outputs Page 7 of 22

RELIEF REQUEST NO. 19 R.E. Ginna Nuclear Power Plant - Fourth Interval ISI Program Defer RPV Category B-F Exams from 2009 to 2011 Outage Westinghouse Structural Reliability And Risk Assessment Model (Win-SRRA)

Inservice Inspection Worksheet Input Report Worksheet Name: RV Inlet Nozzle SE System: Reactor Coolant Segment: RV Inlet Nozzle Safe End Weld Failure Modes: Thermal Fatigue Location: None Comments:

Failure Size(s) Small Leak: False Large Leak: True Full Break: False System Disabling Leak(s)

Value (GPM) Basis 1 Conservative leak for large early release 0

0 0

Snubber Locking Up Under Thermal Conditions: False Fatigue Stress Range 0 Snubber Failure Probability 0 Unexpected Design Limiting Event: False Design Limiting Stress 0 Unexpected Dynamic Event Probability 0 Leak Detection: False Minimum Detectable Leak 0 SRRA Input Case Description Event Probability RV Inlet Nozzle SE(OO-LL-Base-0001) 1.OOE+00 2008/02/07 Win-SRRA 2.0.2 Page 8 of 22

RELIEF REQUEST NO. 19 R.E. Ginna Nuclear Power Plant - Fourth Interval ISI Program Defer RPV Category B-F Exams from 2009 to 2011 Outage Westinghouse Structural Reliability And Risk Assessment Model (Win-SRRA)

Inservice Inspection Worksheet Input Report Worksheet Name: RV Inlet Nozzle SE System: Reactor Coolant Segment: RV Inlet Nozzle Safe End Weld Failure Modes: Thermal Fatigue Location: None Comments:

No. Input Parameter Value Basis 1 Type of Piping Steel Material 316 St. UFSAR Table 5.2-2 2 Crack Inspection Interval 10 ASME Section Xl 3 Crack Inspection Accuracy 0.24 UT Examination 4 Temperature at Pipe Weld 544.2 FSAR T inlet 5 Nominal Pipe Size 34 CN-RCDA-05-10 6 Thickness to O.D. Ratio .096 CN-RCDA-05-1 0 7 Normal Operating Pressure 2.235 FSAR 8 Residual Stress Level 20 SRRA Guidance Document 9 Initial Flaw Conditions 1 Assumed RT preservice 10 DW and Thermal Stress Level 0.17 SRRA Guidance Document, High Temp 11 Stress Corrosion Potential 0.001 SRRA Guidance Document 12 Material Wastage Potential 0.001 SRRA Guidance Document 13 Vibratory Stress Range 0.001 SRRA Guidance Document 14 Fatigue Stress Range 0.3 SRRA Guide. Doc. Normal Operation 15 Low Cycle Fatigue Frequency 5 SRRA Guide. Doc. Normal Operation 16 Design Limiting Stress 0.1 SRRA Guide. Doc. Normal Operation Material Flow Stress 53.6674262 at Input Temperature 2008/02/07 Win-SRRA 2.0.2 Page 9 of 22

RELIEF REQUEST NO. 19 R.E. Ginna Nuclear Power Plant - Fourth Interval ISI Program Defer RPV Category B-F Exams from 2009 to 2011 Outage Westinghouse Structural Reliability And Risk Assessment Model (Win-SRRA)

Reference Case Inputs Report Case Name: RV Inlet Nozzle SE(00-LL-Base-0001) 316 ST Steel Pipe Weld Large Leak Exceed Disabling Leak Rate or Break Event Probability: 1.00E+00 NOCASE=11 NTRIAL = 50000 NUMISI = 5 NUMFMD= 5 NCYCLE = 40 NOVARS = 28 NUMSSC = 6 NFAILS = 400 NUMSET = 6 NUMTRC = 6 Shift MV/SD INT%DEPTH = 2 L/D-RATIO = 2 SCC-COEFF = 0 FCG-COEFF = 1 No. Name Distribution Median Value Deviation or Factor 1 PIPE/ODIA Normal 3.4000E+01 3.9000E-02 2 WALL/ODIA Normal 9.6000E-02 2.9760E-03 3 SRESIDUAL Log Normal 2.OOOOE+01 1.4142E+00 4 INT%DEPTH Log Normal 2.1991 E+00 1.6476E+00 5 LID-RATIO Log Normal 6.OOOOE+00 1.7126E+00 6 FLAWS/iN Constant 6.8711E-02 0.0000E+00 7 FIRST-ISI Constant 5.0000E+00 0.OOOOE+00 8 FREQ-ISI Constant 1.OOOOE+01 0.OOOOE+00 9 EPST-PND Constant 1.OOOOE-03 0.0000E+00 10 ASTAR-PND Constant -2.4000E-01 0.OOOOE+00 11 ANNU-PND Constant 1.6000E+00 0.OOOOE+00 12 HOURS/YR Log Normal 7.4473E+03 1.0500E+00 13 PRESSURE Log Normal 2.2350E+00 1.0323E+00 14 SIG-DW&TH Log Normal 9.1235E+00 1.2599E+00 15 SCC-COEFF Log Normal 3.2310E-12 2.3714E+00 16 SCC-EXPNT Constant 2.161 OE+00 0.OOOOE+00 17 WASTAGE Log Normal 1.2740E-12 2.3714E+00 18 DSIG-VIBR Log Normal 1.6667E-04 1.3465E+00 19 CYCLES/YR Constant 5.OOOOE+00 0.OOOOE+00 20 DSIG-FATG Log Normal 1.61 OOE+01 1.4142E+00 21 FCG-COEFF Log Normal 9.1401E-12 2.8508E+00 22 FCG-EXPNT Constant 4.OOOOE+00 0.OOOOE+00 23 FCG-THOLD Constant 1.5000E+00 0.OOOOE+00 24 LDEPTH-SL Constant 0.OOOOE+00 0.OOOOE+00 25 SIG-FLOW Normal 5.3667E+01 3.2000E+00 26 STRESS-DL Log Normal 5.3667E+00 1.4142E+00 27 B-SDLEAK Constant 1.9613E+00 0.OOOOE+00 28 B-MDLEAK Constant 1.0681 E+02 0.0000E+00 2008/02/07 Win-SRRA 2.0.2 Page 10 of 22

RELIEF REQUEST NO. 19 R.E. Ginna Nuclear Power Plant - Fourth Interval ISI Program Defer RPV Category B-F Exams from 2009 to 2011 Outage Westinghouse Structural Reliability And Risk Assessment Model (Win-SRRA) Detailed Results Report Case Name: RV Inlet Nozzle SE(00-LL-Base-0001) 316 ST Steel Pipe Weld Large Leak Exceed Disabling Leak Rate or Break Number of Failures= 20 Number of Trials= 50000 Probabilities of Failure Run Date 2/7/08 10:47:13 AM Event Probability 11.OOE+00 Total Failure Probability Without ISI 3.54E-07 Total Failure Probability With ISI 5.59E-08 End of Cycle. Failure Probability Without Failure Probability With IS1 ISI Cumulative Total Cumulative Total 0 0.OOE+00 0.00E+00 1 0.OOE+00 0.00E+00 2 O.00E+00 0.00E+00 3 7.01E-11 7.01 E-11 4 7.01 E-11 7.01 E-11 5 7.01E-11 7.01 E-1 1 6 2.32E-10 7.05E-11 7 2.32E-10 7.05E-1 1 8 2.32E-10 7.05E-1 1 9 2.32E-10 7.05E-1 1 10 2.32E-10 7.05E-1 1 11 2.32E-10 7.05E-1I 12 2.32E-10 7.05E-11 13 2.32E-10 7.05E-11 14 2.32E-10 7.05E-1 1 15 2.32E-10 7.05E-11 16 2.32E-10 7.05E-11 17 2.32E-10 7.05E-11 18 6.92E-09 1.33E-09 19 6.93E-09 1.33E-09 20 6.93E-09 1.33E-09 21 6.93E-09 1.33E-09 22 1.46E-08 3.05E-09 2008/02/07 Win-SRRA 2.0.2 Page 11 of 22

RELIEF REQUEST NO. 19 R.E. Ginna Nuclear Power Plant - Fourth Interval ISI Program Defer RPV Category B-F Exams from 2009 to 2011 Outage Westinghouse Structural Reliability And Risk Assessment Model (Win-SRRA) Detailed Results Report Case Name: RV Inlet Nozzle SE(00-LL-Base-0001) 316 ST Steel Pipe Weld Large Leak Exceed Disabling Leak Rate or Break Number of Failures= 20 Number of Trials= 50000 Probabilities of Failure Run Date 2/7/08 10:47:13 AM Event Probability 1.OOE+00 Total Failure Probability Without ISI 3.54E-07 Total Failure Probability With ISI 5.59E-08 End of Cycle. Failure Probability Without Failure Probability With ISI ISI Cumulative Total Cumulative Total 23 1.46E-08 3.05E-09 24 1.46E-08 3.05E-09 25 1.46E-08 3.05E-09 26 2.33E-08 3.06E-09 27 2.35E-08 3.06E-09 28 2.35E-08 3.06E-09 29 6.73E-08 5.47E-09 30 2.48E-07 2.68E-08 31 2.48E-07 2.68E-08 32 2.48E-07 2.68E-08 33 2.77E-07 4.40E-08 34 3.09E-07 5.55E-08 35 3.09E-07 5.55E-08 36 3.09E-07 5.55E-08 37 3.09E-07 5.55E-08 38 3.09E-07 5.55E-08 39 3.53E-07 5.59E-08 40 3.54E-07 5.59E-08 2008/02/07 Win-SRRA 2.0.2 Page 12 of 22

RELIEF REQUEST NO. 19 R.E. Ginna Nuclear Power Plant - Fourth Interval ISI Program Defer RPV Category B-F Exams from 2009 to 2011 Outage Westinghouse Structural Reliability And Risk Assessment Model (Win-SRRA)

Inservice Inspection Worksheet Input Report Worksheet Name: RV Outlet Nozzle SE System: Reactor Coolant Segment: RV Outlet Nozzle Safe End Weld Failure Modes: Thermal Fatigue Location: None Comments:

Failure Size(s) Small Leak: False Large Leak: True Full Break: False System Disabling Leak(s)

Value (GPM) Basis 1 Conservative leak for large early release 0

0 0

Snubber Locking Up Under Thermal Conditions: False Fatigue Stress Range 0 Snubber Failure Probability 0 Unexpected Design Limiting Event: False Design Limiting Stress 0 Unexpected Dynamic Event Probability 0 Leak Detection: False Minimum Detectable Leak 0 SRRA Input Case Description Event Probability RV Outlet Nozzle SE(00-LL-Base-0001) 1.OOE+00 2008/02/07 Win-SRRA 2.0.2 Page 13 of 22

RELIEF REQUEST NO. 19 R.E. Ginna Nuclear Power Plant - Fourth Interval ISI Program Defer RPV Category B-F Exams from 2009 to 2011 Outage Westinghouse Structural Reliability And Risk Assessment Model (Win-SRRA)

Inservice Inspection Worksheet Input Report Worksheet Name: RV Outlet Nozzle SE System: Reactor Coolant Segment: RV Outlet Nozzle Safe End Weld Failure Modes: Thermal Fatigue Location: None Comments:

No. Input Parameter Value Basis 1 Type of Piping Steel Material 316 St. UFSAR Table 5.2-2 2 Crack Inspection Interval 10 ASME Section Xl 3 Crack Inspection Accuracy 0.24 UT Examination 4 Temperature at Pipe Weld 615.8 FSAR T outlet 5 Nominal Pipe Size 35.5 CN-RCDA-05-1 0 6 Thickness to O.D. Ratio .092 CN-RCDA-05-1 0 7 Normal Operating Pressure 2.235 FSAR 8 Residual Stress Level 20 SRRA Guidance Document 9 Initial Flaw Conditions 1 Assumed RT preservice 10 DW and Thermal Stress Level 0.17 SRRA Guidance Document, High Temp 11 Stress Corrosion Potential 0.001 SRRA Guidance Document 12 Material Wastage Potential 0.001 SRRA Guidance Document 13 Vibratory Stress Range 0.001 SRRA Guidance Document 14 Fatigue Stress Range 0.3 SRRA Guide. Doc. Normal Operation 15 Low Cycle Fatigue Frequency 5 SRRA Guide. Doc. Normal Operation 16 Design Limiting Stress 0.1 SRRA Guide. Doc. Normal Operation Material Flow Stress 50.6753338 at Input Temperature 2008/02/07 Win-SRRA 2.0.2 Page 14 of 22

RELIEF REQUEST NO. 19 R.E. Ginna Nuclear Power Plant - Fourth Interval ISI Program Defer RPV Category B-F Exams from 2009 to 2011 Outage Westinghouse Structural Reliability And Risk Assessment Model (Win-SRRA)

Reference Case Inputs Report Case Name: RV Outlet Nozzle SE(00-LL-Base-0001) 316 ST Steel Pipe Weld Large Leak Exceed Disabling Leak Rate or Break Event Probability: 1.OOE+00 NOCASE=1 NTRIAL = 50000 NUMISI = 5 NUMFMD= 5 NCYCLE = 40 NOVARS = 28 NUMSSC = 6 NFAILS = 400 NUMSET = 6 NUMTRC = 6 Shift MV/SD INT%DEPTH = 2 L/D-RATIO = 2 SCC-COEFF = 0 FCG-COEFF = 1 No. Name Distribution Median Value Deviation or Factor 1 PIPE/ODIA Normal 3.5500E+01 3.9000E-02 2 WALUODIA Normal 9.2000E-02 2.8520E-03 3 SRESIDUAL Log Normal 2.OOOOE+01 1.4142E+00 4 INT%DEPTH Log Normal 2.1991 E+00 1.6476E+00 5 UD-RATIO Log Normal 6.OOOOE+00 1.7126E+00 6 FLAWS/IN Constant 6.8892E-02 0.OOOOE+00 7 FIRST-ISl Constant 5.OOOOE+00 0.OOOOE+00 8 FREQ-ISl Constant 1.OOOOE+01 0.OOOOE+00 9 EPST-PND Constant 1.0000E-03 0.OOOOE+00 10 ASTAR-PND Constant -2.4000E-01 0.OOOOE+00 11 ANNU-PND Constant 1.6000E+00 0.OOOOE+00 12 HOURS/YR Log Normal 7.4473E+03 1.0500E+00 13 PRESSURE Log Normal 2.2350E+00 1.0323E+00 14 SIG-DW&TH Log Normal 8.6148E+00 1.2599E+00 15 SCC-COEFF Log Normal 3.2310E-12 2.3714E+00 16 SCC-EXPNT Constant 2.161 OE+00 0.OOOOE+00 17 WASTAGE Log Normal 1.2740E-12 2.3714E+00 18 DSIG-VIBR Log Normal 1.6667E-04 1.3465E+00 19 CYCLES/YR Constant 5.OOOOE+00 0.OOOOE+00 20 DSIG-FATG Log Normal 1.5203E+01 1.4142E+00 21 FCG-COEFF Log Normal 9.1401 E-1 2 2.8508E+00 22 FCG-EXPNT Constant 4.OOOOE+00 O.OOOOE+00 23 FCG-THOLD Constant 1.5000E+00 0.OOOOE+00 24 LDEPTH-SL Constant 0.OOOOE+00 0.OOOOE+00 25 SIG-FLOW Normal 5.0675E+01 3.2000E+00 26 STRESS-DL Log Normal 5.0675E+00 1.4142E+00 27 B-SDLEAK Constant 1.9213E+00 0.OOOOE+00 28 B-MDLEAK Constant 1.1 153E+02 0.OOOOE+00 2008/02/07 Win-SRRA 2.0.2 Page 15 of 22

RELIEF REQUEST NO. 19 R.E. Ginna Nuclear Power Plant - Fourth Interval ISI Program Defer RPV Category B-F Exams from 2009 to 2011 Outage Westinghouse Structural Reliability And Risk Assessment Model (Win-SRRA) Detailed Results Report Case Name: RV Outlet Nozzle SE(00-LL-Base-0001) 316 ST Steel Pipe Weld Large Leak Exceed Disabling Leak Rate or Break Number of Failures= 15 Number of Trials= 50000 Probabilities of Failure Run Date 2/7/08 10:48:29 AM Event Probability 1.OOE+00 Total Failure Probability Without ISI 2.48E-07 Total Failure Probability With ISI 7.16E-09 End of Cycle. Failure Probability Without Failure Probability With ISI ISI Cumulative Total Cumulative Total 0 0.OOE+00 0.OOE+00 1 0.OOE+00 0.OOE+00 2 0.OOE+00 0.OOE+00 3 7.01E-11 7.01 E-11 4 7.01E-11 7.01 E-11 5 7.01 E-11 7.01 E-11 6 7.01E-11 7.01 E-1 1 7 2.32E-10 7.17E-1 1 8 2.32E-10 7.17E-11 9 2.32E-10 7.17E-1 1 10 2.32E-10 7.17E-1 1 11 2.32E-10 7.17E-11 12 2.32E-10 7.17E-1 1 13 2.32E-10 7.17E-1 1 14 2.32E-10 7.17E-11 15 2.32E-10 7.17E-11 16 2.32E-10 7.17E-1 1 17 2.32E-10 7.17E-11 18 2.32E-10 7.17E-1 1 19 2.32E-10 7.17E-11 20 2.32E-10 7.17E-1 1 21 6.92E-09 5.34E-09 22 6.92E-09 5.34E-09 2008/02/07 Win-SRRA 2.0.2 Page 16 of 22

RELIEF REQUEST NO. 19 R.E. Ginna Nuclear Power Plant - Fourth Interval ISI Program Defer RPV Category B-F Exams from 2009 to 2011 Outage Westinghouse Structural Reliability And Risk Assessment Model (Win-SRRA) Detailed Results Report Case Name: RV Outlet Nozzle SE(00-LL-Base-0001) 316 ST Steel Pipe Weld Large Leak Exceed Disabling Leak Rate or Break Number of Failures= 15 Number of Trials= 50000 Probabilities of Failure Run Date 2/7/08 10:48:29 AM Event Probability 1.OOE+00 Total Failure Probability Without ISI 2.48E-07 Total Failure Probability With ISI 7.16E-09 End of Cycle. Failure Probability Without Failure Probability With Is' IS' Cumulative Total Cumulative Total 23 6.93E-09 5.34E-09 24 6.93E-09 5.34E-09 25 6.93E-09 5.34E-09 26 6.93E-09 5.34E-09 27 1.46E-08 5.37E-09 28 1.46E-08 5.37E-09 29 1.46E-08 5.37E-09 30 1.46E-08 5.37E-09 31 1.46E-08 5.37E-09 32 2.33E-08 5.91E-09 33 2.33E-08 5.91 E-09 34 2.35E-08 5.91 E-09 35 2.37E-08 5.96E-09 36 6.73E-08 6.01 E-09 37 2.48E-07 7.16E-09 38 2.48E-07 7.16E-09 39 2.48E-07 7.16E-09 40 2.48E-07 7.16E-09 2008/02/07 Win-SRRA 2.0.2 Page 17 of 22

RELIEF REQUEST NO. 19 R.E. Ginna Nuclear Power Plant - Fourth Interval ISI Program Defer RPV Category B-F Exams from 2009 to 2011 Outage Westinghouse Structural Reliability And Risk Assessment Model (Win-SRRA)

Inservice Inspection Worksheet Input Report Worksheet Name: Si Nozzle SE System: Safety Injection Segment: Safety Injection Nozzle Safe End Weld Failure Modes: Thermal Fatigue Location: None Comments:

Failure Size(s) Small Leak: False Large Leak: True Full Break: False System Disabling Leak(s)

Value (GPM) Basis 1 Conservative leak for large early release 0

0 0

Snubber Locking Up Under Thermal Conditions: False Fatigue Stress Range 0 Snubber Failure Probability 0 Unexpected Design Limiting Event: False Design Limiting Stress 0 Unexpected Dynamic Event Probability 0 Leak Detection: False Minimum Detectable Leak 0 SRRA Input Case Description Event Probability Si Nozzle SE(OO-LL-Base-0001) 1.OOE+00 2008/02/07 Win-SRRA 2.0.2 Page 18 of 22

RELIEF REQUEST NO. 19 R.E. Ginna Nuclear Power Plant - Fourth Interval ISI Program Defer RPV Category B-F Exams from 2009 to 2011 Outage Westinghouse Structural Reliability And Risk Assessment Model (Win-SRRA)

Inservice Inspection Worksheet Input Report Worksheet Name: Si Nozzle SE System: Safety Injection Segment: Safety Injection Nozzle Safe End Weld Failure Modes: Thermal Fatigue Location: None Comments:

No. Input Parameter Value Basis 1 Type of Piping Steel Material 316 St. UFSAR Table 5.2-2 2 Crack Inspection Interval 10 ASME Section Xl 3 Crack Inspection Accuracy 0.24 UT Examination 4 Temperature at Pipe Weld 544.2 FSAR T inlet 5 Nominal Pipe Size 4.125 CN-RCDA-05-10 6 Thickness to O.D. Ratio .128 CN-RCDA-05-1 0 7 Normal Operating Pressure 2.235 FSAR 8 Residual Stress Level 10 SRRA Guidance Document 9 Initial Flaw Conditions 1 Assumed RT preservice 10 DW and Thermal Stress Level 0.17 SRRA Guidance Document, High Temp 11 Stress Corrosion Potential 0.001 SRRA Guidance Document 12 Material Wastage Potential 0.001 SRRA Guidance Document 13 Vibratory Stress Range 0.001 SRRA Guidance Document 14 Fatigue Stress Range 0.3 SRRA Guide. Doc. Normal Operation 15 Low Cycle Fatigue Frequency 5 Normal Operation 16 Design Limiting Stress 0.1 SRRA Guide. Doc. Normal Operation Material Flow Stress 53.6674262 at Input Temperature 2008/02/07 Win-SRRA 2.0.2 Page 19 of 22

RELIEF REQUEST NO. 19 R.E. Ginna Nuclear Power Plant - Fourth Interval ISI Program Defer RPV Category B-F Exams from 2009 to 2011 Outage Westinghouse Structural Reliability And Risk Assessment Model (Win-SRRA)

Reference Case Inputs Report Case Name: Si Nozzle SE(00-LL-Base-0001) 316 ST Steel Pipe Weld Large Leak Exceed Disabling Leak Rate or Break Event Probability: 1.OOE+00 NOCASE=1 NTRIAL = 50000 NUMISI = 5 NUMFMD= 5 NCYCLE = 40 NOVARS = 28 NUMSSC = 6 NFAILS = 400 NUMSET = 6 NUMTRC = 6 Shift MV/SD INT%DEPTH = 2 L/D-RATIO = 2 SCC-COEFF = 0 FCG-COEFF = 1 No. Name Distribution Median Value Deviation or Factor 1 PIPE/ODIA Normal 4.6250E+00 1.5000E-02 2 WALL/ODIA Normal 1.2800E-01 3.9680E-03 3 SRESIDUAL Log Normal 1.OOOOE+01 1.4142E+00 4 INT%DEPTH Log Normal 1.6137E+01 1.3102E+00 5 L/D-RATIO Log Normal 6.OOOOE+00 1.7126E+00 6 FLAWS/IN Constant 3.0670E-03 0.OOOOE+00 7 FIRST-ISI Constant 5.OOOOE+00 0.OOOOE+00 8 FREQ-ISI Constant 1.OOOOE+01 0.OOOOE+00 9 EPST-PND Constant 1..OOOOE-03 0.OOOOE+00 10 ASTAR-PND Constant -2.4000E-01 0.OOOOE+00 11 ANNU-PND Constant 1.6000E+00 0.OOOOE+00 12 HOURS/YR Log Normal 7.4473E+03 1.0500E+00 13 PRESSURE Log Normal 2.2350E+00 1.0323E+00 14 SIG-DW&TH Log Normal 9.1235E+00 1.2599E+00 15 SCC-COEFF Log Normal 3.2310E-12 2.3714E+00 16 SCC-EXPNT Constant 2.1610E+00 0.OOOOE+00 17 WASTAGE Log Normal 1.2740E-1 2 2.3714E+00 18 DSIG-VIBR Log Normal 5.3680E-04 1.3465E+00 19 CYCLES/YR Constant 5.OOOOE+00 0.OOOOE+00 20 DSIG-FATG Log Normal 1.6100E+01 1.4142E+00 21 FCG-COEFF Log Normal 9.1401E-12 2.8508E+00 22 FCG-EXPNT Constant 4.OOOOE+00 0.OOOOE+00 23 FCG-THOLD Constant 1.5000E+00 0.OOOOE+00 24 LDEPTH-SL Constant 0.OOOOE+00 0.OOOOE+00 25 SIG-FLOW Normal 5.3667E+01 3.2000E+00 26 STRESS-DL Log Normal 5.3667E+00 1.4142E+00 27 B-SDLEAK Constant 2.0378E+00 0.OOOOE+00 28 B-MDLEAK Constant 1.4530E+01 0.OOOOE+00 2008/02/07 Win-SRRA 2.0.2 Page 20 of 22

RELIEF REQUEST NO. 19 R.E. Ginna Nuclear Power Plant - Fourth Interval ISI Program Defer RPV Category B-F Exams from 2009 to 2011 Outage Westinghouse Structural Reliability And Risk Assessment Model (Win-SRRA) Detailed Results Report Case Name: Si Nozzle SE(00-LL-Base-0001) 316 ST Steel Pipe Weld Large Leak Exceed Disabling Leak Rate or Break Number of Failures= 90 Number of Trials= 50000 Probabilities of Failure Run Date 2/7/08 10:42:22 AM Event Probability 1.OOE+00 Total Failure Probability Without ISI 3.38E-08 Total Failure Probability With ISI 1.91 E-09 End of Cycle. Failure Probability Without Failure Probability With Is' Isl Cumulative Total Cumulative Total 0 0.OOE+00 O.OOE+00 1 1.55E-13 1.55E-13 2 7.82E-12 7.82E-12 3 7.87E-12 7.87E-12 4 7.87E-12 7.87E-12 5 8.86E-12 8.86E-12 6 1.47E- 11 8.87E-12 7 4.90E-10 2.94E-1 1 8 7.79E-10 3.59E-1 1 9 8.47E-10 3.75E-1 1 10 8.59E-10 3.76E-1 1 11 2.32E-09 2.38E-10 12 1.02E-08 1.60E-09 13 1.02E-08 1.60E-09 14 1.17E-08 1.78E-09 15 1.17E-08 1.78E-09 16 1.18E-08 1.78E-09 17 1.18E-08 1.78E-09 18 1.18E-08 1.78E-09 19 1.26E-08 1.79E-09 20 1.32E-08 1.79E-09 21 1.36E-08 1.79E-09 22 1.46E-08 1.80E-09 2008/02/07 Win-SRRA 2.0.2 Page 21 of 22

RELIEF REQUEST NO. 19 R.E. Ginna Nuclear Power Plant - Fourth Interval ISI Program Defer RPV Category B-F Exams from 2009 to 2011 Outage Westinghouse Structural Reliability And Risk Assessment Model (Win-SRRA) Detailed Results Report Case Name: Si Nozzle SE(00-LL-Base-0001) 316 ST Steel Pipe Weld Large Leak Exceed Disabling Leak Rate or Break Number of Failures= 90 Number of Trials= 50000 Probabilities of Failure Run Date 2/7/08 10:42:22 AM Event Probability 1.00E+00 Total Failure Probability Without ISI 3.38E-08 Total Failure Probability With ISI 1.91 E-09 End of Cycle. Failure Probability Without Failure Probability With isI ISI Cumulative Total Cumulative Total 23 1.52E-08 1.80E-09 24 1.52E-08 1.80E-09 25 2.06E-08 1.86E-09 26 2.06E-08 1.86E-09 27 2.20E-08 1.86E-09 28 2.20E-08 1.86E-09 29 2.23E-08 1.86E-09 30 2.23E-08 1.86E-09 31 2.23E-08 1.86E-09 32 2.87E-08 1.91 E-09 33 2.87E-08 1.91 E-09 34 2.88E-08 1.91 E-09 35 2.88E-08 1.91 E-09 36 3.01 E-08 1.91 E-09 37 3.02E-08 1.91 E-09 38 3.03E-08 1.91 E-09 39 3.03E-08 1.91 E-09 40 3.38E-08 1.91 E-09 2008/02/07 Win-SRRA 2.0.2 Page 22 of 22

RELIEF REQUEST NO. 20 R.E. Ginna Nuclear Power Plant - Fourth Interval ISI Program Defer RPV Category B-K Exams from 2009 to 2011 Outage and Perform Volumetric Instead of Surface Examinations

RELIEF REQUEST NO. 20 R.E. Ginna Nuclear Power Plant - Fourth Interval ISI Program Defer RPV Category B-K Exams from 2009 to 2011 Outage and Perform Volumetric instead of Surface Examinations Proposed Alternative In Accordance with 10 CFR 50.55a (a)(3)(ii)

-Compliance with the Specified Requirements Would Result in Hardship or Unusual Difficulty without a Compensating Increase in the Level of Quality and Safety -

1. ASME Code Component(s) Affected The R.E. Ginna Reactor Vessel has two (2) Examination Category B-K welds. These welds attach the reactor vessel supports to the reactor vessel. The welds are located on the reactor vessel at approximately 90 degree (Weld RPV-VSL-1) and 270 degree (Weld RPV-VSL-2) azimuths of the vessel.

The plate and gusset material are fabricated from SA-21, Gr. B carbon steel and attached to the reactor vessel via a double-bevel full penetration groove weld with carbon steel weld filler metal.

The ASME Boiler and Pressure Vessel (BPV) Code Section Xl (Reference 1) Examination Category and Item Number of Table IWB-2500-1 are:

Examination Category Item No. Description B-K B10.10 Pressure Vessel Welded Attachments (for Reactor Pressure Vessel only)

(Throughout this request the above Examination Category and Item Number description are referred to as "the subject examinations" and the ASME BPV Code Section XI is referred to as "the Code".)

2. Applicable Code Edition and Addenda

The R.E. Ginna Fourth Interval Inservice Inspection (ISI) Program Plan is prepared to the ASME Section Xl Code, 1995 Edition with the 1996 Addenda.

3. Applicable Code Requirement

IWA-2430 (d), each inspection interval may be reduced or extended by as much as one year.

Adjustments shall not cause successive intervals to be altered more than one year from the original pattern of intervals. Additionally, Table IWB-2500-1, Examination Category B-K, Item Number B10.10 requires a surface examination of essentially 100% of each of the reactor pressure vessel integral attachment welds once each ten year interval.

Page 1 of 4

RELIEF REQUEST NO. 20 R.E. Ginna Nuclear Power Plant - Fourth Interval ISI Program Defer RPV Category B-K Exams from 2009 to 2011 Outage and Perform Volumetric instead of Surface Examinations

4. Reason for Request

Relief Request Number 20 is being submitted along with Relief Request Numbers 18, 19 and 21 which deal with associated Reactor Pressure Vessel examinations to be deferred from the 2009 Refueling Outage to the 2011 Refueling Outage. The intent of this request is for a six month interval extension beyond the Code allowed 12 month extension (IWA-2430(d)) and to perform volumetric examination of these two (2) welds from the vessel inside diameter (ID) instead of an outside surface examination (Table IWB-2500-1, Examination Category B-K, Item Number B10.10) to coincide with the above reactor vessel weld examinations (Relief Request Number 18).

The reactor pressure vessel body was designed and constructed to ASME Section Ill, 1965 Edition.

Early Codes that were used in the construction of Ginna Station did not contain requirements to ensure that items be made accessible for future examinations. Due to the limitations of early construction codes, the two reactor pressure vessel welded attachments (welds RPV-VSL-1 and RPV-VSL-2) are not accessible from the outside. The only examination that can be performed is by the ultrasonic technique from the inside of the vessel. This examination would be a best effort examination like the examinations that were performed in the past intervals.

Performing a volumetric examination from the ID of the Vessel on the two Reactor Pressure Vessel Welded Attachments during the same outage as the reactor vessel shell welds will result in a reduction of man-rem exposure. The performance of this examination would require the removal of all fuel and the core barrel from the reactor vessel. An unnecessary risk is created by the removal of the core barrel more than once within an inspection interval to perform associated vessel examinations without a compensating increase to quality or safety. The previous average dose rate in the general area of the vessel supports was 145 mRem per hour. The highest measured dose rate was 232 mRem per hour with a majority of the workers receiving 130-165 mRem per hour dose rates.

Significant radiation exposure reduction (multiple Rem savings) can be realized since the same equipment and personnel used for the volumetric examination of the vessel shell welds and nozzle welds from the vessel interior can be used to examine the two vessel support welded attachments. The reactor pressure vessel welded attachments volumetric examinations (Table IWB-2500-1, Examination Category B-K, Item Number B10.10) have historically been performed during the same outage at the end of the inservice inspection interval.

5. Proposed Alternative and Basis for Use R.E. Ginna Nuclear Power Plant proposes to perform volumetric examinations from the vessel ID for the fourth inspection interval one refueling cycle beyond the end of the fourth interval. The fourth inspection interval for R.E. Ginna started January 1, 2000 and will end December 31, 2009. The subject examinations are currently scheduled to be performed during the Fall 2009 refueling outage.

Volumetric examination of the two reactor pressure vessel welded attachments (welds RPV-VSL-1 and RPV-VSL-2) are proposed to be performed in the subsequent refueling outage in Spring 2011. This Page 2 of 4

RELIEF REQUEST NO. 20 R.E. Ginna Nuclear Power Plant - Fourth Interval ISI Program Defer RPV Category B-K Exams from 2009 to 2011 Outage and Perform Volumetric instead of Surface Examinations inspection date is less than 6 months beyond the ten year Code inspection interval and the one year interval extension provided by IWA-2430(d).

In accordance with 10 CFR 50.55a (a)(3)(ii), performance of a volumetric ID examination of the attachment welds instead of surface examinations due to access limitation is consistent with the examinations performed in the past and that the interval extension of performing the examinations of the reactor vessel welded attachments separate in time from the reactor vessel shell welds and nozzle welds would result in hardship without a compensating increase in quality or safety.

The technical justification for the deferral of the subject examination is as follows:

A.) PWR Service Experience There are no known incidents of cracking or degradation of these supports in either the domestic PWR or BWR Fleet. Therefore, there is a low probability that cracking will occur in the R.E. Ginna reactor vessel support welds during the time period that this relief is requested.

B.) R.E. Ginna Inservice Inspection History The attachment welds were volumetrically examined three times at R.E. Ginna in addition to a pre-service examination. Most recently, these examinations were performed in April of 1999 in accordance with the 1986 Edition of the ASME Boiler and Pressure Vessel Code,Section XI.

Table 1 provides a summary of the inservice examination results from the last examination. Due to the improvements in inspection technology with time, the most recent examination is considered to be of the greatest quality of the three inservice inspections performed.

Table 1: R.E. Ginna Examination Category B-K Inservice Inspection Results Weld ID Description # of # of recordable reportable indications indications 90° 90 degree vessel support weld 0 0 270" 270 degree vessel support weld 0 0 Page 3 of 4

RELIEF REQUEST NO. 20 R.E. Ginna Nuclear Power Plant - Fourth Interval ISI Program Defer RPV Category B-K Exams from 2009 to 2011 Outage and Perform Volumetric instead of Surface Examinations C.) R.E. Ginna Access Limitation The reactor pressure vessel body was designed and constructed to ASME Section III, 1965 Edition.

Early Codes that were used in the construction of Ginna Station did not contain requirements to ensure that items be made accessible for future examinations. Due to the limitations of early construction codes, the two reactor pressure vessel welded attachments (welds RPV-VSL-1 and RPV-VSL-2) are not accessible from the outside of the vessel. The only examination that can be performed is by the ultrasonic technique from the inside of the vessel.

Volumetric examination of the two reactor pressure vessel welded attachments has been performed several times at R.E. Ginna Nuclear Power Plant with no relevant indications noted. The examinations were last performed during the 1999 refueling outage with acceptable results. Due to the vintage of Ginna Station, Construction Codes that were used did not have requirements to ensure that items be made accessible for future examinations. Volumetric examination from the vessel ID is the only alternative. Also, there are no known incidents of cracking or degradation of these support welds in either the domestic PWR or BWR Fleet. Therefore, there is a low probability that cracking will occur in the R.E, Ginna reactor vessel support welds during the time period that this relief is requested

6. Duration of Proposed Alternative The alternative is requested to extend the fourth inservice inspection interval by approximately 6 months beyond the ASME Code required 10-year inspection interval and Code allowed twelve month extension and to perform a volumetric examination on the two reactor pressure vessel welded attachments from the vessel ID instead of an outside surface examination due to access limitation. This request is applicable to R.E. Ginna fourth inspection interval only.

7. References

1. ASME Boiler and Pressure Vessel Code, Section Xl, 1995 Edition with the 1996 Addenda

2. ASME Boiler and Pressure Vessel Code,Section III, 1965 Edition Page 4 of 4

RELIEF REQUEST NO. 21 R.E. Ginna Nuclear Power Plant - Fourth Interval ISI Program Defer Category B-N-2 and B-N-3 Exams from 2009 to 2011 Outage

RELIEF REQUEST NO. 21 R.E. Ginna Nuclear Power Plant - Fourth Interval ISI Program Defer Category B-N-2 and B-N-3 Exams from 2009 to 2011 Outage Proposed Alternative In Accordance with 10 CFR 50.55a (a)(3)(ii)

-Compliance with the Specified Requirements Would Result in Hardship or Unusual Difficulty Without a Compensating Increase in the Level of Quality and Safety -

1. ASME Code Component(s) Affected The affected component is the R.E. Ginna Nuclear Plant Reactor Vessel (RV).

The ASME Boiler and Pressure Vessel (BPV) Code Section XI, Examination Category and Item Number of Table IWB-2500-1 addressed in this request are:

Examination Cateqory Item No. Description B-N-2 B13.60 Interior Attachments Beyond Beltline Region B-N-3 B13.70 Core Support Structure (Throughout this request the above Examination Category and Item Number descriptions are referred to as "the subject examinations" and the ASME BPV Code Section Xl is referred to as "the Code".)

2. Applicable Code Edition and Addenda

The R.E. Ginna Fourth Interval Inservice Inspection (ISI) Program Plan is prepared to ASME Section XI Code, 1995 Edition with the 1996 Addenda.

3. Applicable Code Requirement

IWA-2430 (d), each inspection interval may be reduced or extended by as much as one year.

Adjustments shall not cause successive intervals to be altered more than one year from the original pattern of intervals. Additionally, Table IWB-2500-1, Examination Category B-N-2 and B-N-3, Item Number B13.60 and B13.70 requires a visual examination of the interior attachments beyond the beltline region and a visual examination of the core support structure of the reactor pressure vessel once each ten year interval.

4. Reason for Request

Relief Request Number 21 is being submitted along with Relief Request Numbers 18, 19 and 20 which deal with associated Reactor Pressure Vessel examinations to be deferred from the 2009 Refueling Outage to the 2011 Refueling Outage. The intent of this request is for a less than a six month interval extension beyond the Code allowed 12 month extension (IWA-2430(d)) in order to allow the subject examinations to be performed at the same time as the reactor vessel examinations (Relief Request Number 18), along with other reactor vessel related examinations.

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RELIEF REQUEST NO. 21 R.E. Ginna Nuclear Power Plant - Fourth Interval ISI Program Defer Category B-N-2 and B-N-3 Exams from 2009 to 2011 Outage During the Third Interval ten year inservice inspection of the reactor vessel welds in 1999, R.E. Ginna performed visual examinations of the reactor vessel interior attachments and the core support structure.

Since the core support structure (called Core Barrel on Westinghouse designed plants) requires removal to facilitate examination of the reactor vessel shell, nozzle and safe end welds, the visual examinations of ASME Category B-N-2 and B-N-3 have historically been performed during the same outage at the end of the inservice inspection interval.

5. Proposed Alternative and Basis for Use R.E. Ginna Nuclear Power Plant proposes to perform the subject examinations for the fourth inspection interval one refueling cycle beyond the end of the fourth interval. The fourth inspection interval for R.E.

Ginna started on January 1, 2000 and will end on December 31, 2009. The subject examinations are currently scheduled to be performed during the Fall of 2009 refueling outage. Visual examinations under ASME Category B-N-2 and B-N-3 are proposed to be performed in the subsequent refueling outage in Spring 2011. The inspection date is less than 6 months beyond the ten year Code inspection interval and the one year interval extension provided by IWA-2430 (d).

In accordance with 10 CFR 50.55a (a) (3) (ii), this interval extension is requested on the basis that performing the examination of the reactor vessel interior attachments and core support structure separate in time from the reactor vessel examinations would result in hardship or unusual difficulty without a compensating increase in quality or safety.

The examination required by ASME Category B-N-2 and B-N-3 requires the removal of all the fuel and the core barrel from the reactor vessel. An unnecessary risk is created by the removal of the core barrel more than once within an inspection interval to perform associated vessel examinations without a compensating increase in quality or safety. If this request is not accepted, then the core barrel would be required to be moved during the 2009 Refueling Outage and again in the 2011 Refueling Outage.

Visual examinations of the reactor pressure vessel interior attachments and the core support structure have been performed several times at R.E. Ginna Nuclear Power Plant with no relevant indications noted during the examinations. The examinations were last performed during the 1999 refueling outage with acceptable results. Additionally, review of industry surveys indicate that these examinations have been performed many times by the industry without any significant findings.

During the 2009 Refueling Outage, R.E. Ginna Nuclear Power Plant will perform the ASME Category B-N-1 visual examination. This examination will include the space above and below the reactor core that is made accessible for examination by the removal of components during normal refueling outages.

This examination, along with previous history for Category B-N-2 and B-N-3 examinations of no relevant indications will provide reasonable assurance of structural integrity.

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RELIEF REQUEST NO. 21 R.E. Ginna Nuclear Power Plant - Fourth Interval ISI Program Defer Category B-N-2 and B-N-3 Exams from 2009 to 2011 Outage

6. Duration of Proposed Alternative The alternative is requested to extend the fourth inservice inspection interval by less than 6 months beyond the ASME Code required 10-year inspection interval and Code allowed twelve month extension. This request is applicable to R.E. Ginna fourth inspection interval only.

7. References

1. ASME Boiler and Pressure Vessel Code, Section Xl, 1995 Edition with the 1996 Addenda Page 3 of 3