NL-08-177, Response to Request for Additional Information on Request for Relief to Extend Inservice Inspection Interval for the Reactor Vessel Weld Examination and Request for License Amendment

Response to Request for Additional Information on Request for Relief to Extend Inservice Inspection Interval for the Reactor Vessel Weld Examination and Request for License Amendment

ML090050020
Person / Time
Site:	Indian Point
Issue date:	12/23/2008
From:	Robert Walpole Entergy Nuclear Northeast
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NL-08-177, TAC MD9196, TAC MD9197
Download: ML090050020 (26)

Text

Entergy Nuclear Northeast

-- Entergy Indian Point Energy Center 450 Broadway, GSB Buchanan, NY 10511-0249 Robert Walpole Licensing Manager Tel 914 734 6710 December 23, 2008 Re: Indian Point Units 2 and 3 Docket No. 50-247 and 50-286 NL-08-177 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Mail Stop O-P1-17 Washington, DC 20555-0001

Subject:

Response to Request For Additional Information on Request For Relief To Extend The Unit 2 and 3 Inservice Inspection Interval For The Reactor Vessel Weld Examination And Request For License Amendment For Submittal of ISI Information and Analyses (TAC Nos. MD9196-MD9197)

References:

1. NRC Letter dated November 20, 2008, Request for Additional Information Regarding Relief Request for Vessel Weld Inspection Extension (TAC Nos.

MD9196 and MD9197)

2. Entergy Letter dated July 8, 2008 regarding Request For Relief To Extend The Unit 2 and 3 Inservice Inspection Interval For The Reactor Vessel Weld Examination And Request For License Amendment For Submittal of ISI Information and Analyses (NL-08-096)

Dear Sir or Madam:

Entergy Nuclear Operations, Inc. (Entergy) hereby submits in attachment 1 a response to the NRC request for additional information (Reference 1) regarding relief requests (RR) RR-76 and RR-3-43(I) made in Reference 2.

There are no new commitments being made in this submittal.

If you have any questions or require additional information, please contact me at (914) 734-6710.

Robert Wal Licensing Manager Indian Point Energy Center cc: next page AzCi4T

NL-08-177 Docket 50-247 and 50-286 Page 2 of 2 Response to Request for Additional Information Regarding The Request For Relief To Extend The Inservice Inspection Interval For The Reactor Vessel Weld Examination cc: Mr. John P. Boska, Senior Project Manager, NRC NRR DORL Mr. Samuel J. Collins, Regional Administrator, NRC Region 1 NRC Resident Inspector - IP2 Mr. Robert Callender, Vice President, NYSERDA Mr. Paul Eddy, New York State Dept. of Public Service

ATTACHMENT 1 TO NL-08-177 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION REGARDING THE REQUEST FOR RELIEF TO EXTEND THE INSERVICE INSPECTION INTERVAL FOR THE REACTOR VESSEL WELD EXAMINATION ENTERGY NUCLEAR OPERATIONS, INC.

INDIAN POINT NUCLEAR GENERATING UNIT NOS. 2 AND 3 DOCKET NOS. 50-247 and 50-286

NL-08-177 Docket 50-247 and 50-286 Attachment 1 Page 1 of 23 Response to Request For Additional Information Regarding The Request For Relief To Extend The Inservice Inspection Interval For The Reactor Vessel Weld Examination Entergy Nuclear Operations, Inc. (Entergy) submitted Relief Request 76 (RR-76) for Indian Point Unit No. 2 (IP2) and Relief Request 3-43(I) (RR-3-43(I)) for Indian Point Unit No. 3 (IP3) by letter dated July 8, 2008 (ADAMS Accession No. ML081980058).

These relief requests are from the requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components," and would apply for the Third 10-year Inservice Inspection (ISI) Interval for the Reactor Vessel (RV) Weld examinations. The NRC has requested additional information on these RR. These requests and the response follow.

The NRC questions note that the estimated bounding IP2 and IP3 through-wall cracking frequency (TWCF) is about an order of magnitude above the screening criteria change in TWCF for Westinghouse plants that was reported in WCAP 16168-NP-A, Revision 2, Table 4-1. This indicates that IP's design and operating characteristics are not bounded by the bounding evaluation relied on by the NRC staff to endorse increasing the inspection intervals for reactor vessel welds from 10 to 20 years.

Question 1 Please provide a plant-specific estimate of the change in risk associated with the requested interval extension for IP2 and IP3. This change in risk estimate may use the methods endorsed by the NRC staff in approving the WCAP (which were developed as part of the technical basis for the pressurized thermal shock (PTS) rulemaking).

Response

The results of the Indian Point specific change-in-risk estimate evaluation are shown in Table 1. As will be discussed in the response to RAI 2, this change-in-risk estimate was calculated consistent with the approach approved in WCAP-1 6168-NP-A, Revision 2, but with Indian Point specific inputs. This evaluation was performed for Indian Point Unit 3 since, as shown in Table 3 of the requests for relief, it was determined to have the more limiting through-wall cracking frequency. The evaluation was performed for 60,000 vessel simulations with version 06.1 of the FAVOR Code. As shown by Figure 1, the solution was converged for this number of simulations. As shown in Table 1, the bounding difference in risk estimated for Indian Point Unit 3 is 2.15E-08 events per year, which is about a factor of 5 below the criteria in Regulatory Guide 1.174 for an acceptably small change in large early release frequency.

NL-08-177 Docket 50-247 and 50-286 Attachment 1 Paqe 2 of 23 Table 1 - Indian Point Unit 3 Change-in Risk Estimate 10-Year ISI Only (Mean Value / Standard Error) 8.42E-08 / 4.78E-09 Upper Bound Value 9.38E-08 ISI Every 10 Years (Mean Value / Standard Error) 7.91 E-08 / 3.40E-09 Lower Bound Value 7.23E-08 Bounding Difference in Risk 2.15E-08 Results are in "Events Per Year' Figure 1 - Change-in-Risk for 1P3 @ 48 EFPY 1.40E-07 O 1.20E-07

• .1.00E m 8.OOE-08 6.OOE-08 ---------

uJ U. 4.OOE-08 2.00E-08 O.OOE+O0 ...

0 10000 20000 30000 40000 50000 60000 Iterations

- 10 Year ISI Only - Upper Bound ISI Every 10 Years - Lower Bound Question 2 Please summarize the principal steps in the methodology used in the WCAP and confirm that the Entergy methodology is consistent with the WCAP or justify any differences.

The summary should briefly describe the major steps in the methodology and should demonstrate that Entergy appropriately applied the WCAP methodology in its calculations.

Response

The approach used to estimate the Indian Point Unit 3 specific change-in-risk is consistent with that used for the pilot plants in WCAP-1 6168-NP-A, Revision 2. Two cases were evaluated using the FAVOR probabilistic fracture mechanics code. One of these cases considered inservice inspection performed on a 10 year interval while the second case considered elimination of inservice inspection following the first 10-year

NL-08-177 Docket 50-247 and 50-286 Attachment 1 Page 3 of 23 inservice inspection. These cases are referred to as "ISI Every 10 Years'" and "10-Year ISI Only", respectively. The approach for performing these FAVOR evaluations is shown in Figure 3-5 of the WCAP, and is discussed on pages 3-14 to 3-16 of the WCAP. As discussed on page3-16, the effects of inservice inspection and fatigue crack growth were modeled by using the Westinghouse developed PROBSBFD computer code to modify the surface breaking flaw distribution file, S.dat, input to the FAVOR Code. A surface breaking flaw distribution file was created for each of the cases evaluated. Two executions of the FAVOR Code were then performed, one utilizing the "ISI Every 10 Years" surface breaking flaw distribution file and the other utilizing the "10-Year ISI Only" surface breaking flaw distribution file. For each FAVOR execution, a mean value of through-wall cracking frequency was obtained along with values of standard error.

Consistent with the approach in the WCAP, to calculate a change-in-risk, a change in failure frequency was conservatively calculated based on the difference between the "Upper Bound" and "Lower Bound" values. The Lower Bound value was determined by subtracting 2 times the standard error as output by the FAVOR Code from the mean value of the "ISI Every 10 Years" case. The Upper Bound value was determined by adding 2 times the standard error as output from the FAVOR Code to the mean value of the "10-Year ISI Only" case. The difference between the upper and lower bound values was then compared against the risk criteria of Regulatory Guide 1.174 to determine the acceptability of the extension in inspection interval.

Question 3 Please describe which results from the PTS technical basis results were used as input to IP2 and IP3's change in risk calculations, and which inputs were developed as IP2 and IP3 specific inputs.

Response

The inputs to the change-in-risk estimate evaluation for Indian Point Unit 3, which also bounds Unit 2, are discussed below:

PTS Transients- The PTS transient frequencies and pressure, temperature, and heat transfer versus time definitions used in the plant specific evaluation were the same as those used for the Westinghouse pilot plant in WCAP-1 6168-NP-A, Revision 2. The applicability of these transients to Indian Point Unit 3 is justified in the response to RAI 4.

Design Basis Transients- 7 heatup/cooldown cycles per year was used in determining the fatigue crack growth of the surface breaking flaws. This number was considered in the WCAP to be bounding of all design basis transients for the domestic Westinghouse fleet and was therefore determined to be applicable for Indian Point Unit 3.

Probabilityof Detection - The probability of detection curve used was with the same as that used for the WCAP pilot plant applications.

NL-08-177 Docket 50-247 and 50-286 Attachment 1 Page 4 of 23 Fatigue Crack Growth Rate - The correlation of fatigue crack growth rate with change in stress intensity factor used was the same as that that used for the WCAP pilot plant applications.

Reactor Vessel Dimensions - The reactor vessel dimensions used in the calculation of the stresses and stress intensity factors by the FAVLOAD module of the FAVOR Code were Indian Point Unit 3 specific reactor vessel dimensions. Furthermore, the Indian Point Unit 3 reactor vessel dimensions were used to calculate the weld fusion areas for each subregion in the input to the FAVPFM module of the FAVOR Code.

Surface Breaking Flaw DistributionFiles - New files were created to reflect the geometric conditions of the Indian Point Unit 3 reactor vessel which differs slightly from those of the Westinghouse pilot plant. As shown in Figure 3-5 of the WCAP, the cyclic loads used by the PROBSBFD Code are determined using the FAVLOAD module of the FAVOR Code. The Indian Point specific dimensions of reactor vessel base metal thickness, cladding thickness, and inside radius were used as input to the FAVLOAD module to generate Indian Point Unit 3 specific transient loads. Furthermore, the initial flaw size input to the PROBSBFD Code was the Indian Point Unit 3 cladding thickness rounded up to the nearest 1% of the wall thickness. This approach to initial flaw depth is consistent with the approach used for the WCAP pilot plant analyses. The initial flaw density was consistent with that used in the technical basis for the proposed alternate PTS Rule.,

Plate and Weld Embedded Flaw DistributionFiles - The 1000 distributions of embedded flaw sizes and densities for plates and welds used in the Indian Point evaluation were

,consistent with those used in the technical basis for the PTS Rule and for the WCAP pilot plant analyses. However, new flaw input files were created to account for the differences in reactor vessel wall thickness between the Westinghouse pilot plant and that of Indian Point Unit 3. These flaw input files were created using the program VFLAW and the generic vessel input that was used to create the flaw distribution files used in the development of the technical basis for the proposed PTS Rule.

Fluence and Materials Inputs - An Indian Point Unit 3 specific input file was created for the FAVPFM module of the FAVOR Code. This input file contained 13 major regions for each of the 13 beltline materials. Major regions were divided into 66 subregions with location specific fluence values identified for these subregions. While this model does not have the same level of refinement with regard to fluence mapping as those used in the technical basis for the proposed PTS Rule and the WCAP pilot plants which used over 15,000 subregions, this approach yields conservative results because the limiting fluence is assumed for the entire subregion. Fluence values for each subregion were determined from WCAP-16251-NP (Reference 4). The fluence values used corresponded to 48 EFPY of operation as opposed to the 60 EFPY used for the WCAP pilot plant analyses. However, 48 EFPY still bounds Indian Point 3 until the end of a potential license renewal period through 60 years of operation. A map of the Indian Point Unit 3 major regions and subregions is shown in Figure 2. In this figure, the major region numbers are indicated along the perimeter while the subregion numbers are indicated directly on the map. Shaded areas represent weld regions while areas that are

NL-08-177 Docket 50-247 and 50-286 Attachment 1 Page 5 of 23 not shaded represent plate regions. Axial weld regions are 1.50 in width which corresponds to a weld width of 1.31". The plate subregions adjacent to the axial welds are 10 in width. These plate subregions are assigned the fluence values of the adjacent axial welds. The material properties for the major regions are identified in Table 2.

While these properties are consistent with the properties identified in Table 3 of the Indian Point relief requests (RR-76 and RR-3-43(l)) to implement the extended inservice inspection interval, the regions are numbered differently.

Coolant Temperature - The reactor vessel coolant temperature included in the input file for the FAVPFM module of the FAVOR Code was based on Indian Point Unit 3 specific operating data.

ConditionalProbabilityof Large Early Release - For the change-in-risk calculations, it was conservatively assumed that a through-wall crack results in a large early release.

Therefore, the conditional large early release probability (CLERP) is 1.0.

NL-08-177 Docket 50-247 and 50-286 Attachment 1 Page 6 of 23 Table 2 - Major Region Material Composition Peak Fluence@ 48 FAVOR Relief EFPY [1019 Major Request Cu Ni P Mn RTNDT(u) Neutron/cm 2, Region # Table 3# Description [wt%] [wt%] [wt%] [wt%]' [IF] E>1 MeV]

1 12 Lower Shell Axial Weld 3-042B 0.192 1.007 0.012 1.450 -56 1.11 2 13 Lower Shell Axial Weld 3-042C 0.192 1.007 0.012 1.450 -56 1.56 3 11 Lower Shell Axial Weld 3-042A 0.192 1.007 0.012 1.450 -56 1.11 4 9 Inter. Shell Axial Weld 2-042B 0.192 1.007 0.012 1.450 -56 1.24 5 10 Inter. Shell Axial Weld 2-042C 0.192 1.007 0.012 1.450 -56 0.727 6 8 Inter. Shell, Axial Weld 2-042A 0.192 1.007 0.012 1.450 -56 1.24 7 7 Circumferential Weld 9-042 0.221 0.732 0.023 1.450 -54 1.56 8 2 Intermediate Shell B2802-2 0.220 0.530 0.015 1.630 -4 1.56 9 3 Intermediate Shell B2802-3 0.200 0.490 0.011 1.630 17 1.56 10 1 Intermediate Shell B2802-1 0.200 0.500 0.010 1.635 5 1.56 11 6 Lower Shell B-2803-3 0.240 0.520 0.012 1.630 74 1.56 12 5 Lower Shell B2803-2 0.220 0.520 0.011 1.630 -5 1.56 13 4 Lower Shell B2803-1 0.190 0.470 0.012 1.630 49 1.56 Note 1: Manganese content was intended to be based on the conservative percent weight estimates in Table 4 of the proposed alternate PTS Rule, 10 CFR 50.61 a. However, the value of 1.45 for plates was inadvertently used for the welds and the value of 1.63 for welds was inadvertently used for the plates.

While this causes the FAVOR code to calculate values of RTNDT(u)+AT 3 0 that are 2 to 3 percent lower than intended for the welds, the values calculated for the plates are more conservative. As shown in Table 2.2-3, the plate region is controlling of all RTMAX-XX values. The FAVPOST output files also indicate that flaws in the plates contribute more than 90% of all through wall cracking in the reactor vessel. Therefore, the use of the higher manganese content value for the plates results in more conservative values from FAVOR and for the change-in-risk calculations.

NL-08-177 Docket 50-247 and 50-286 Attachment 1 Page 7 of 23 600 1800 3000 600 1050 2250 3450 Figure 2 - Indian Point Unit 3 FAVOR Beltline Map (Note: Subregions not drawn to scale)

NL-08-177 Docket 50-247 and 50-286 Attachment 1 Page 8 of 23 Question 4 If IP2 and IP3 used the PTS sequence frequency and binning from one of the reference plants in the WCAP, please justify that these results are bounding compared to the IP specific values, or that any differences between IP2 or IP3 and the reference plant are not expected to appreciably increase IP's estimated change in risk if IP2 or IP3 specific frequency and binning were used. A methodology to compare plant-specific PTS characteristics to the characteristics of a reference plant is described in Letter Report, "Generalization of Plant-Specific Pressurized Thermal Shock (PTS) Risk Results to Additional Plants" (ML042880482).

Response

The objective of the evaluation in the letter report cited in RAI 4, "Generalization of Plant-Specific Pressurized Thermal Shock (PTS) Risk Results to Additional Plants," (referred to as the "Generalization Study" for the remainder of this section) was to determine whether the results from-the detailed pilot plant analyses from the technical basis for the proposed PTS Rule were applicable to a wider number of PWRs. To make this determination, a set of five additional PWRs for which PTS may be important was determined. Potentially important design and operational features of the five additional PWRs were then compared to the same features of the three reference plants to determine the extent that these features were similar or different. Based on these comparisons, judgments were made to the appropriateness of treating the three pilot plants as being representative of the PWR fleet. The overall conclusion of the generalization study was that the results indicated that the through wall cracking frequency (TWCF) estimates produced by the detailed reference plant analyses were sufficient to characterize the TWCF estimates for the five additional plants and by inference, the remainder of the PWRs.

The five additional plants considered in the generalization study included three plants which were Westinghouse 4-loop designs, as are Indian Point Units 2 and 3. While the generalization study determined that there were differences in plant and operational features between the three 4-loop plants and the Westinghouse reference plant (Beaver Valley Unit 1), no differences were found that would cause significant differences in the progression or frequencies of the PTS transients used in these analyses. Furthermore, the downcomer temperature as a result of these transients was considered to be about the same for the additional plants as that analyzed for the reference plants. The conclusions of the generalization study should be applicable to Indian Points 2 and 3 if it can be shown that Indian Point Units 2 and 3 are similar to the three Westinghouse 4-loop plants in terms of plant features and operation.

Several PTS significant plant parameters identified in Appendix A of the generalization study are provided in Table 3 for Indian Point Units 2 and 3 along with a comparison to the three Westinghouse 4-loop units. The information regarding Indian Point is taken directly from the plant Updated Final Safety Analysis Reports (UFSAR) for Units 2 and 3.

NL-08-177 Docket 50-247 and 50-286 Attachment 1 Page 9 of 23 As shown in Table 3, the parameters of Indian Point Units 2 and 3 are consistent with, comparable to, or bounded by those of the Westinghouse 4-loop units as identified in Appendix A of the Generalization Study.

The emergency operating procedures (EOPs) used at Indian Point Units 2 and 3 were systematically developed through a program, which included phases of validation, verification, training and operator feedback. This program met the requirements of NUREG-0737 and utilized the guidance of NUREG-0899, NRC Standard Review Plan 13.5.2, and the Westinghouse Owners Group (WOG) Emergency Response Guidelines.

These generic WOG Emergency Response Guidelines, which were evaluated by the NRC in a December 26, 1985 Supplemental Safety Evaluation Report, were identified in Appendix A as the basis for the operator actions for at least one of the three Westinghouse 4-loop plants in the Generalization Study. It is reasonable to conclude that the EOPs for Indian Point Units 2 and 3 are comparable to those of the generalization plant since they are developed based on the same guidance.

Based on this and the similarities in plant characteristics identified in Table 3 and similarities in the bases for operator actions between Indian Point Units 2 and 3 and the three Westinghouse 4-loop plants in the generalization study, the conclusions of the generalization study regarding the applicability of the PTS transients to other PWRs are appropriate for Indian Point Units 2 and 3. Therefore, the use of the Westinghouse pilot plant PTS transients in developing a plant specific change-in-risk estimates for Indian Point Units 2 and 3 is appropriate.

Table 3 - Comparison of PTS Significant Plant Parameters Comparison to Westinghouse Generalization Study Parameter Indian Point Unit 2 Indian Point Unit 3 Plants

1) Number of MSIVs 1 isolation-type MSIV and 1 1 isolation-type MSIV and 1 Consistent with reverse flow check valve reverse flow check valve generalization study plants.

per each of 4.steamlines per each of 4 steamlines

2) Isolation capability with If all MIlVs close, the If all MSlVs close, the Consistent with regard to other possible steam dump/turbine steam dump/turbine generalization study plants.

secondary valve open bypass valves will be bypass valves will be paths. isolated. isolated.

The ADVs (one per The ADVs (one per steamline) are upstream of steamline) are upstream of the MSlVs. Each has a the MSIVs. Each has a block valve for isolation block valve for isolation purposes. purposes.

The SRVs (5 per The SRVs (5 per steamline) do not have steamline).do not have isolation capability, isolation capability,

3) Location/Size of Each SG is fitted with flow Each SG is fitted with flow Consistent with steamline flow orifices restrictors at the outlet restrictors at the outlet generalization study plants.

nozzle limiting the flow to nozzle limiting the flow to 1.4ft 2 . 1.4ff.

NL-08-177 Docket 50-247 and 50-286 Attachment 1 Page 10 of 23 Table 3 - Comparison of PTS Significant Plant Parameters Comparison to Westinghouse Generalization Study Parameter Indian Point Unit 2 Indian Point Unit 3 Plants

4) How are MFW/AFW SI signal will close all MFW SI signal will close all MFW; Consistent with controlled during steamline control valves and close control valves and close generalization study plants.

break? the MFW pump discharge the MFW pump discharge valves, which in turn would valves, which in turn would trip the MFW pumps. trip the MFW pumps.

MSIVs will close on high MSIVs will close on high steam flow with either low steam flow with either low RCS av. Temp. or low RCS av. Temp. or low steam line pressure or high steam line pressure or high containment pressure containment pressure signals. signals.

SI signal does not SI signal does not automatically isolate AFW, automatically isolate AFW, therefore AFW would have therefore AFW would have to be manually isolated. to be manually isolated.

5) Information on the feed Condensor hotwell Condensor hotwell Consistent with capabilities to the SGs inventory is 114,000 inventory is 114,000 generalization study plants.

including source inventory gallons. 2 steam driven gallons. 2 steam driven of water available to MFW pumps rates at MFW pumps rates at continue MFW or AFW if 15,300 GPM each. 15,300 GPM each.

were not isolated. There are 2 motor driven There are 2 motor driven AFW pumps @400 GPM AFW pumps @400 GPM each and one turbine each and one turbine driven AFW pump @ 800 driven AFW pump @ 800 GPM. AFWs draw from the GPM. AFWs draw from the condensate storage tank condensate storage tank which has a minimum which has a minimum volume of 360,000 gallons. volume of 360,000 gallons.

6) Information on normal For 100% Power, water For 100% Power, water Informationis comparable to steam generator inventory volume is 1599 ft3 and volume is 1627 ft3 and that of generalization plants steam volume is 3128 ft3 . steam volume is 3100 ft3 . when multiplied by For 0% power, water For 0% power, water appropriate steam and water volume is 2779 ft3 and volume is 2666 ft3 and densities.

steam volume is 1949ft3 . steam volume is 2061ft3 .

7)Information on possible MFW normal temp is MFW normal temp is Consistent with feed temperatures for all 436.2F. Minimum Hotwell 433.6F. Minimum Hotwell generalization plants.

feed sources temp is approximately 70F. temp is approximately 70F.

Minimum Condensate Minimum Condensate storage tank temp is storage tank temp is approximately 40F. approximately 40F.

8) Allowable range of safety RWST temperature range RWST must remain above Temperatures are within injection water is 40F to 11 F. freezing. range of those for temperatures generalization plants.

9) Safety injection water Normal RWST volume is Normal RWST volume is RWST volumes are slightly source size 345,000 gallons. There are 342,200 gallons. There are less than those of 4 SI accumulators with a 4 SI accumulators with a generalization plants.

volume of 1100 gallons volume of 1100 gallons Therefore, the generalization each. each. plants should be bounding.

10) How many and which SI signal will actuate 3 of 3 SI signal will actuate 3 of 3 Consistent with pumps auto start on SI pumps and 2 of 2 RHR SI pumps and 2 of 2 RHR generalization plants.

ESFAS/SI signal in pumps. pumps.

response to a LOCA

11) Charging, HPI, LPI SI Pump (HPI): 3550ft SI Pump (HPI): 3500ft Bounded by shutoff heads of Recirc. Pump: 476ft Recirc. Pump: 460ft

NL-08-177 Docket 50-247 and 50-286 Attachment 1 Page 11 of 23 Table 3 - Comparison of PTS Significant Plant Parameters Comparison to Westinghouse Generalization Study Parameter Indian Point Unit 2 Indian Point Unit 3 Plants shutoff heads RHR Pump (LPI): 390ft RHR Pump (LPI): 390ft generalization plant pumps

12) Actuation requirements Containment spray Containment spray Containment spray actuation for containment spray and actuated on a high-high actuated on a high-high occurs at slightly higher flow rate once running containment pressure containment pressure containment pressure than signal (-24 psig). There signal (-24 psig). There generalization plants. Flow are 2 containment spray are 2 containment spray rates are comparable.

pumps, each with a design pumps, each with a design flow rate of 2600 GPM. flow rate of 2600 GPM.

13) Accumulator discharge Nitrogen cover pressure is Normal operating pressure Pressures are comparable to pressure maintained between 598 is 650 psig those of generalization psig and 685 psig. plants.

14) Number and sizes of 2 PORVs with a capacity of 2 PORVs with a capacity of Consistent with PORVs on the pressurizer 179 klb/hr each and 3 179 klb/hr each and 3 generalization plants.

SRVs with a capacity of SRVs with a capacity of 420 klb/hr each 420 klb/hr each.

15) Instrumentation All engineered safety All remote operated valves Consistent with available to identify open system valves have have visual position generalization plants, PORVs or SRVs and to position indication on the indication in the Control multiple indications exist.

notice if they have reclosed control board to show Room. All of the PORVs proper positioning of the and their associated motor-valves. Acoustic sensors operated block valves have installed on the code safety been provided with an valves discharge lines acoustic monitoring system provide indication in the for position indication.

central control room of the Should there be any "flow" or 'nonflow" significant leakage from condition of line safety any of these valves, this valves. The PORVs have a system initiates an alarm in direct valve position the control room.

indication in the central control room.

16) Number of AFW 2 motor driven AFW pumps 2 motor driven AFW pumps Consistent with pumps/flow paths versus each feeding 2 SGs. One each feeding 2 SGs. One generalization plants.

minimum success criteria turbine driven AFW pump turbine driven AFW. pump for adequate feed to the feeding all 4 SGs. feeding all 4 SGs.

SGs Successful AFW flow is 1 Successful AFW flow is 1 motor driven pump to 2 motor driven pump to 2 SGs. SGs.

ADV = Atmospheric Dump Valve PORV = Power Operated Relief Valve AFW = Auxiliary Feedwater RCS = Reactor Coolant System GPM = Gallons per minute RHR = Residual Heat Removal HPI = High Pressure Injection RWST = Reactor Water Storage Tank LPI = Low Pressure Injection SG = Steam Generator MFW = Main Feedwater SI = Safety Injection MSIV = Main Steam Isolation Valve SRV = Safety Relief Valve Question 5 Section 5 of Relief Request (RR)-76 for IP2 proposed, "To bound IP2 the Westinghouse pilot plant was reevaluated at a value of EFPY (Condition B) that is well beyond 60 EFPY." Here, EFPY stands for effective full-power years, and Condition B stands for the

NL-08-177 Docket 50-247 and 50-286 Attachment 1 Page 12 of 23 extended embrittlement level with a mean TWCF around 1x10 6 as defined in NUREG-1806, "Technical Basis for Revision of the Pressurized Thermal Shock (PTS) Screening Limit in the PTS Rule (10 CFR 50.61)." The approach described is a significant extension of the analyses previously accepted by the NRC staff when WCAP-1 6168 was approved. This extension could bring the results much closer to the risk-informed decision-making criteria in Regulatory (RG) 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis." If Entergy continues to pursue the alternative, please reevaluate the potential sources of uncertainty in the calculation of TWCF and ATWCF (as defined in WCAP-16168, "Risk-Informed Extension of the Reactor Vessel In-Service Inspection [ISI]

Interval,") to provide confidence that the results which are proposed to bound IP2 and IP3 are consistent with RG 1.174.

Response

Entergy is no longer pursuing this alternative and has elected to respond to RAIs 1 through 4.

Question 6 Table 3 of RR-76 and RR-3-43(l) contains plant-specific information for the TWCFs that you calculated for the IP2 and IP3 reactor pressure vessels (RPVs). To complete the review, the NRC staff requests you:

1. Explain how the neutron fluence and flux values for the IP2 and IP3 RPVs were obtained, including references to these calculated results where necessary.

2. Identify source documents for determining the manganese content for the IP2 and IP3 RPV materials and describe how you averaged them when more than one data source is available.

3. Provide your calculations for obtaining the embrittlement shift, AT30, for the circumferential weld of the IP2 RPV. The NRC staff has verified your AT30 values within 1 degree F for the axial weld and the plate of the IP2 RPV, but is not able to conclude that the AT30 for the circumferential weld is the same as the AT30 for the plate.

4. Provide your calculations for obtaining the reference temperature for the axial weld (RTMAX-AW) and the reference temperature for the circumferential weld (RTMAX-CW) of the IP2 RPV. Please note that determination of RTMAX-AW and RTMAx-CW requires consideration of adjacent plate or forging.

5. Section 3.4 in the staff's safety evaluation (SE) of WCAP-16168 indicated the RTMAX-x and the AT30 value must be calculated using the latest approved methodology or other NRC approved methodology. The latest approved methodology is documented in Regulatory Guide 1.99, Revision 2. This RG also specifies a methodology for evaluating reactor vessel surveillance material data. An alternate methodology for calculating AT30 value is documented in an alternative PTS rule, 10 CFR 50.61 (a),

which was published for public comment in the Federal Registeron August 11, 2008.

The alternative PTS rule also documents a methodology for evaluating reactor

NL-08-177 Docket 50-247 and 50-286 Attachment 1 Page 13 of 23 vessel surveillance material data. If the licensee determines RTMAx-X and AT30 values using either the methodologies in RG 1.99, Revision 2 or the alternative PTS rule, the licensee must provide, the NRC the analysis of its reactor vessel materials surveillance data.

Response

1. The fluence values were calculated based on the latest available nuclear cross-section data derived from ENDF/B-VI and made use of the latest available calculational tools for neutron source generation and neutron transport. Furthermore, this neutron transport methodology follows the guidance and meets the requirements of Regulatory Guide 1:190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence". The results are documented in WCAP-16752-NP for Unit 2 and WCAP-16251-NP for Unit 3. These references provide the calculated maximum neutron fluence at the clad to base metal interface for azimuthal positions of 00, 150, 30°, and 450. For the TWCF calculation, the fluences used for the axial welds were based on the welds' azimuthal positions. For the plate materials and circumferential weld, the limiting fluence occurring anywhere in the plate or circumferential weld was used. Since the references cited above only provide data to 48 EFPY, the 48 EFPY values were linearly extrapolated to 60 EFPY.

2. The manganese content for the IP2 and IP3 materials were assumed based on the "Conservative Estimates for Chemical Element Weight Percentages" in Table 4 of the proposed alternate PTS Rule.

3. The AT30 identified in Table 3 of the Indian Point Unit 2 request for relief for RTMAX-CW is the AT30 for the limiting adjacent plate since the determination of RTMAx-cw requires consideration of all adjacent plates or forgings. The AT30 for the circumferential weld is 269.640 F. The circumferential weld has a mean initial RTNDT(u) of -56 0 F.

Therefore, the (RTNDT(u) + AT 3o) for the circumferential weld is 213.640 F. The AT30 for the limiting adjacent plate based on the limiting fluence for the circumferential weld is 228.73°F and the mean initial RTNDT of the plate is 21 °F. Therefore, the (RTNDT(u) +

AT30) for the adjacent plate is 249.73°F and is the limiting value for RTMAX-CW. The calculations for the circumferential weld and adjacent plates are shown in Table 4.

The equations used for calculating AT30 were taken directly from page 35 of NUREG-1874.

NL-08-177 Docket 50-247 and 50-286 Attachment 1 Page 14 of 23 Table 4 - Calculation of RTMAX-CW for Indian Point Unit 2 Circumferential Adjacent Adjacent Adjacent Adjacent Adjacent Weld Plate 1 Plate 2 Plate 3 Plate 4 Plate 5 Copper (weight %) 0.190 0.190 0.170 0.250 0.200 0.190 Phosphorous (weight %) 0.010 0.010 0.014 0.011 0.010 0.011 Nickel (weight %) 1.010 0.650 0.460 0.600 0.660 0.600 Manganese (weight %) 1.630 1.450 1.450 1.450 1.450 1.450 Max Fluence (n/cm 2) 2.38E+19 2.38E+19 2.38E+19 2.38E+19 2.38E+19 2.38E+19 Mean Initial RTNDT(u) (OF) -56 34 21 21 20 -20 t (sec) 1.89E+09 1.89E+09 1.89E+09 1.89E+09 1.89E+09 1.89E+09 Coolant Temperature (OF) 523 523 523 523 523 523 (n/cm 2/sec) 1.26E+1 0 1.26E+10 1.26E+10 1.26E+10 1.26E+10 1.26E+10 4Ote(n/cm2) 3.30E+19 3.30E+19 3.30E+19 3.30E+19 3.30E+19 3.30E+19 CUe 0.19 0.19 0.17 0.2435 0.2 0.19 Max(Cue) 0.30 0.2435 0.37 0.2435 0.2435 0.2435 f(CUe,P) 0.24 0.244 0.224 0.313 0.257 0.245 g(Cue,Ni,Ote) 0.98 0.99 0.99 0.99 0.99 0.99 A 1.417E-7 1.561 E-7 1.561E-7 1.561 E-7 1.561 E-7 1.561 E-7 B 155 135.2 135.2 135.2 135.2 135.2 MD 99.48 104.91 110.49 106.30 104.91 106.30 CRP 170.16 101.39 71.47 122.43 108.32 95.81 AT 30 (OF) 269.64 206.30 181.96 228.73 213.22 202.11 RTNDT(u) + AT 30 (OF) 213.64 240.30 202.96 249.73 233.22 182.11 RTMAX-CW (°F/0 R) 249.73 / 709.42

NL-08-177 Docket 50-247 and 50-286 Attachment 1 Page 15 of 23

4. Please see Table 4 of the response to bullet 3 for the calculation of RTMAX-CW for IP2.

Tables 5 and 6 provide the calculations for RTMAX-AW for IP2. Table 5 provides the calculations for the plates adjacent to the axial welds. Table 6 provides the calculations for the axial welds with consideration of the adjacent plates from Table 5.

NL-08-177 Docket 50-247 and 50-286 Attachment 1 Page 16 of 23 Table 5 - Calculation of RTNDT(u) + AT30 for Adjacent Plates for RTMAX-AW for Indian Point Unit 2 Axial Weld # 7 8 9 10 11 Adjacent Plate # 1 2 1 3 2 3 4 5 4 5 Copper (weight %) 0.190 0.170 0.190 0.250 0.170 0.250 0.200 0.190 0.200 0.190 Phosphorous (weight %) 0.010 0.014 0.010 0.011 0.014 0.011 0.010 0.011 0.010 0.011 Nickel (weight %) 0.650 0.460 0.650 0.600 0.460 0.600 0.660 0.600 -0.660 0.600 Manganese (weight %) 1.450 1.450 1.450 1.450 1.450 1.450 1.450 1.450 1.450 1.450 Max Fluence (n/cm 2) of 1.62E+19 1.62E+19 1.62E+19 1.62E+19 8.27E+18 8.27E+18 1.32E+19 1.32E+19 1.32E+19 1.32E+19 Adjacent Axial Weld Mean Initial RTNDT(u) (OF) 34 21 34 21 21 21 20 -20 20 -20 t (sec) 1.89E+09 1.89E+09 1.89E+09 1.89E+09 1.89E+09 1.89E+09 1.89E+09 1.89E+09 1.89E+09 1.89E+09 Coolant Temperature (OF) 523 523 523 523 523 523 523 523 523 523 (n/cm 2/sec) 8.55E+09 8.55E+09 8.55E+09 8.55E+09 4.37E+09 4.37E+09 6.98E+09 6.98E+09 6.98E+09 6.98E+09 Ote(n/cm 2) 2.48E+19 2.48E+19 2.48E+19 2.48E+19 1.51E+19 1.51E+19 2.13E+19 2.13E+19 2.13E+19 2.13E+19 Cue 0.19 0.17 0.19 0.2435 0.17 0.2435 0.2 0.19 0.2 0.19 Max(CUe) 0.2435 0.37 0.2435 0.2435 0.37 0.2435 0.2435 0.2435 .0.2435 0.2435 f(Cue,P) 0.244 0.224 0.244 0.313 0.224 0.313 0.257 0.245 0.257 0.245 g(Cue,NiOte) 0.98 0.98 0.98 0.98 0.97 0.97 0.97 0.98 0.97 0.98 A 1.561 E-7 1.561E-7 1.561 E-7 1.561 E-7 1.561 E-7 1.561 E-7 1.561 E-7 1.561 E-7 1.561 E-7 1.561 E-7 B 135.2 135.2 135.2 135.2 135.2 135.2 135.2 135.2 .135.2 135.2 MD 90.92 95.76 90.92 92.13 74.69 71.86 84.33 85.46 84.33 85.46 CRP 100.67 71.06 100.67 121.76 69.83 119.79 107.05 94.74 107.05 94.74 AT3o (F) 191.59 166.82 191.59 213.89 144.52 191.65 191.38 180.19 191.38 180.19 RTNDT(u) + AT 3o (F) 225.59 187.82 225.59 234.89 165.52 212.65 211.38 160.19 211.38 160.19 Max. RTNDT(u) + AT 30 (OF) 225.59 234.89 212.65 211.38 211.38 -

NL-08-177 Docket 50-247 and 50-286 Attachment 1 Page 17 of 23 Table 6 - Calculation of RTMAX-AW for Indian Point Unit 2 Axial Weld # 7 8 9 10 11 Copper (weight %) 0.210 0.210 0.210 0.210 0.210 Phosphorous (weight %) 0.021 0.021 0.021 0.021 0.021 Nickel (weight %) 1.010 1.010 1.010 1.010 1.010 Manganese (weight %) 1.630 1.630 1.630 1.630 1.630 Max Fluence (n/cm 2) 1.62E+19 1.62E+19 8.27E+18 1.32E+19 1.32E+19 Mean Initial RTNDT(u) (OF) -56 -56 -56 -56 -56 t (sec) 1.89E+09 1.89E+09 1.89E+09 1.89E+09 1.89E+09 Coolant Temperature (OF) 523.00 523.00 523.00 523.00 523.00 (n/cm2/sec) 8.55E+09 8.55E+09 4.37E+09 6.98E+09 6.98E+09

£Pte(n/cm 2) 2.48E+19 2.48E+19 1.51 E+19 2.13E+19 2.13E+19 Cue 0.21 0.21 0.21 0.21 0.21 Max(Cue) 0.30 0.30 0.30 0.30 0.30 f(Cue,P) 0.29 0.29 0.29 0.29 0.29 g(Cue,Ni,yte) 0.97 0.97 0.94 0.96 0.96 A 1.417E-7 1.417E-7 1.417E-7 1.417E-7 1.417E-7 B 155 155 155 155 155 MID 102.35 102.35 79.83 94.94 94.94 CRP 199.81 199.81 193.47 198.29 198.29 AT3o(OF) 302.16 302.16 273.30 293.23 293.23 RTNDT(u) + AT 30 (IF) 246.16 246.16 217.30 237.23 237.23 Max. RTNDT(u) + AT 30 of Adjacent 225.59 234.89 212.65 211.38 211.38 Plate (From Table 5) (OF)

Max. RTNDT(u) + AT 30 for Axial Weld 246.16 246.16 217.30 237.23 237.23 or Adjacent Plate .1 RTMAx-AW (OF/°R) 246.16 / 705.85

NL-08-177 Docket 50-247 and 50-286 Attachment 1 Page 18 of 23

5. As a condition of the alternate PTS Rule 10 CFR 50.61 a, the licensee must consider plant-specific information that could affect the use of the proposed equation for the determination of a material's AT30 value. In order to make this determination, the proposed alternate PTS Rule provides requirements for evaluation of reactor vessel surveillance capsule data. These requirements are specified in paragraphs (f)(6)(ii),

(f)(6)(iii), and (f)(6)(iv) of the proposed alternate rule. In summary the requirements consist of a Mean Deviation Test, a Slope Deviation Test, and an Outlier Deviation Test. The equations for performing these tests are included as Equations 8 through 12 of the proposed alternate rule. These tests must be performed when three (3) or more measurements of surveillance data are available for any of the reactor vessel beltline materials. The results of the Mean Deviation, Slope Deviation, and Outlier Tests must satisfy the limits of Tables 5, 6, and 7 of the proposed alternate rule, respectively.

The Indian Point Unit 2 reactor vessel surveillance program includes Intermediate shell plates B-2002-1, B-2002-2, and B-2002-3. The surveillance program also includes a weld material that is a heat match to the intermediate and lower shell axial welds (Heat #

W5214). However, three or more measurements of surveillance data are only available for plates B-2002-2 and B-2002-3. The program only includes longitudinal samples for the plate materials. The results of the most recent surveillance capsule evaluation are documented in Southwest Research Institute Final Report 17-2108 (Reference 5).

The Indian Point Unit 3 reactor vessel surveillance program includes lower shell plates B-2803-3 and intermediate shell plates B-2802-1 and B-2802-2. The surveillance program also includes a weld material but this weld material is not a match to any of the materials in the Indian Point Unit 3 reactor vessel beltline welds. Three or more measurements of surveillance data are available only for the lower shell plate B-2803-3.

The surveillance program includes both longitudinal and transverse samples for plate material B-2803-3. The results of the most recent surveillance capsule evaluation are documented in WCAP-16251 -NP (Reference 4).

For the materials where three (3) or more surveillance measurements were available, the inputs and results of the three surveillance data tests are contained in Tables 7 through 9. As can be seen from these tables, the surveillance data satisfies the criteria in the proposed alternate PTS Rule for all three tests. Therefore, the use of the equations contained in NUREG-1 874 for calculation of AT30 is acceptable for Indian Point Units 2 and 3.

NL-08-177 Docket 50-247 and 50-286 Attachment 1 Page 19 of 23 Table 7 - 10 CFR 50.61a Surveillance Data Evaluation for Indian Point Unit 2 Plate B-2002-2 Capsule T Z V Specimen Orientation Longitudinal Longitudinal Longitudinal Input Data Copper (Weight %) 0.1701 0.1701 0.1701 Phosphorous (Weight %) 0.014 0.014 0.014 Nickel (Weight %) 0.4601 0.4601 0.4601 Manganese (Weight %) 1.300 1.300 1.300 Fluence (xl0 1 9 n/cm 2, E > 1.0MeV) 0.253 1.020 0.492 EFPY 1.40 5.20 8.60 Time Averaged Coolant Temperature (OF) 543.00 531.00 528.00 Measured AT30 Transition Temperature (OF) 95.00 120.00 77.00 Calculated Values Predicted AT30 Transition Temperature (OF) 72.38 120.21 107.10 Residual (r) - 22.62 -0.21 -30.10 Maximum Heat-Average Residual Test Slope Deviation Test Outlier Deviation Test Mean Deviation -2.56 T-Statistic -0.45 Allowable Actual Maximum Mean Residual +/-28.52 Critical T-Statistic +/-31.82 Largest r* +/-2.71 -1.42 Pass/Fail? Pass Pass/Fail? Pass Second largest r* +1.55 1.07 Pass/Fail? Pass Note 1: Copper and nickel content are based on-average values for surveillance capsule charpy specimens as identified in Reference 5.

NL-08-177 Docket 50-247 and 50-286 Attachment 1 Page 20 of 23 Table 8 - 10 CFR 50.61a Surveillance Data Evaluation for Indian Point Unit 2 Plate B-2002-3 Capsule T I Y Z Specimen Orientation Longitudinal Longitudinal Longitudinal Input Data Copper (Weight %) 0.2501 0.2501 0.2501 Phosphorous (Weight %) 0.011 0.011 0.011 Nickel (Weight %) 0.6001 0.6001 0.6001 Manganese (Weight %) 1.290 1.290 1.290 Fluence (xl0 19 n/cm 2, E > 1.0MeV) 0.2530 0.455 1.020 EFPY 1.40 2:30 5.20 Time Averaged Coolant Temperature (OF) 543.00 543.00 531.00 Measured AT30 Transition Temperature (OF) 115.00 145.00 180.00 Calculated Values Predicted AT3o Transition Temperature (OF) 11 1.2ý2 134.91 168.32 Residual (r) 3.78 10.09 11.68 Maximum Heat-Average Residual Test Slope Deviation Test Outlier Deviation Test Mean Deviation 8.52 T-Statistic 2.22 Allowable Actual Maximum Mean Residual +/-28.52 Critical T-Statistic +/-31.82 Largest r* +/-2.71 0.55 Pass/Fail? Pass Pass/Fail? Pass Second largest r* +1.55 0.48

_______________________________________________________ -Pass/Fail? Pass Note 1: Copper and nickel content are based on average values for surveillance capsule charpy specimens.as identified in Reference 5.

NL-08-177 Docket 50-247 and 50-286 Attachment 1 Page 21 of 23 Table 9 - 10 CFR 50.61a Surveillance Data Evaluation for Indian Point Unit 3 Plate B-2803-3 Capsule T Y Z X T Z X Specimen Orientation Transverse Transverse Transverse Transverse Longitudinal Longitudinal Longitudinal Input Data Copper (Weight %) 0.240 0.240 0.240 0.240 0.240 0.240 0.240 Phosphorous (Weight %) 0.012 0.012 0.012 0.012 0.012 0.012 0.012 Nickel (Weight %) 0.520 0.520 0.520 0.520 0.520 0.520 0.520 Manganese (Weight %) 1.300 1.300 1.300 1.300 1.300 1.300 1.300 Fluence (x1019 n/cm2, E > 1.0MeV) 0.263 0.692 1.040 0.874 0.263 1.040 0.874 EFPY 1.40 3.20 5.50 15.50 1.40 5.50 15.50 Time Averaged Coolant Temperature (OF) 543.00 543.00 542.00 540.50 543.00 542.00 540.50 Measured AT30 Transition Temperature (OF) 105.90 148.90 157.90 158.20 139.40 167.80 159.60 Calculated Values Predicted AT30 Transition Temperature (OF) 104.89 136.58 I 148.29 151.09 104.89 148.29 151.09 Residual (r) 1.01 12.32 9.61 7.11 34.51 19.51 8.51 Maximum Heat-Average Residual Test Slope Deviation Test Outlier Deviation Test Mean Deviation 13.22 T-Statistic -0.59r Allowable Actual Maximum Mean Residual +/-18.67 Critical T-Statistic +/-3.36 Largest r* +/-2.98 1.63 Pass/Fail? Pass Pass/Fail? Pass Second largest r* +/-2.00 0.92

__Pass/Fail? Pass

NL-08-177 Docket 50-247 and 50-286 Attachment 1 Page 22 of 23 Question 7 During a teleconference between the licensee and the NRC staff on November 5, 2008, the licensee indicated that the current applicable American Society of Mechanical Engineers (ASME) Code of record for inservice examination is the 2001 Edition with 2002 Addenda for Indian Point Unit No. 2 and the 1989 Edition, no Addenda for Indian Point Unit No. 3.

Considering the unique nature of the RRs, which were intended to apply to the end of the current license, the NRC staff requests you revise the RRs according to the following:

" The Edition and Addenda of the Code should be the recent year ASME Code that you would use for the fourth interval for other ASME Code applications.

• Table 2 of the RR provides the results from prior reactor vessel examinations. This table indicates that the inspection method for the prior examinations were performed in accordance with RG 1.150. This is an acceptable methodology for prior examinations; however, the staff requires that all future examinations be qualified in accordance with ASME Code, Section Xl, Appendix VIII.

• Enclosures 1 and 2 to NL-08-096 indicate that the fourth reactor vessel inspection is proposed to be completed by 2032 for Indian Point Unit No. 2 and by 2035 for Indian Point Unit No. 3. These dates are beyond the current licensing term for each of these units. The NRC can only approve a RR until the end of the licensee's current license term. Therefore, the RR should be modified to identify all inspections that are planned within the current licensing term. If Indian Point Unit Nos. 2 and 3 have their licenses extended in the future, a new RR will be needed for the extended license terms.

Response

The Edition and Addenda of the Code that both Indian Point Unit 2 and Indian Point Unit 3 are requesting relief from is ASME Code Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components," Code 1989 Edition, No Addenda.

It is understood that Table 2 provides the results from prior reactor vessel examinations which were performed in accordance with RG 1.150. All future examinations will be qualified in accordance with ASME Code,Section XI, Appendix VIII.

The Relief Requests in enclosures 1 and 2 to NL-08-096 will be modified to identify that they only apply to inspections which will be performed within the current licensing term.

NL-08-177 Docket 50-247 and 50-286 Attachment 1 Page 23 of 23

References:

1. WCAP-16168-NP-A, Revision 2, "Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval," June 2008.

2. NUREG-1874, 'Technical Basis for Revision of the Pressurized Thermal Shock (PTS) Screening Limit in the PTS Rule (10 CFR 50.61)," 3/1/07 (ADAMS Accession Number ML070860156).

3. WCAP-16752-NP, Revision 0, "Indian Point Unit 2 Heatup and Cooldown Limit Curves for Normal Operation," January 2008.

4. WCAP-16251-NP, Revision 0, "Analysis of Capsule X for Entergy's Indian Point Unit 3 Reactor Vessel Radiation Surveillance Program," July 2004.

5. Southwest Research Institute Final Report, SwRl Project No. 17-2108, "Reactor Vessel Material Surveillance Program for Indian Point Unit No. 2, Analysis of Capsule V," October 1988.