ML061090832

Proposed License Amendment to Change Technical Specification 3.6.A.2, Pressure-Temperature Limit Curves

ML061090832
Person / Time
Site:	Pilgrim
Issue date:	04/12/2006
From:	Balduzzi M Entergy Nuclear Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
2.06.018
Download: ML061090832 (61)

Text

Entergy Entergy Nuclear Operations, Inc.

Pilgrim Nuclear Power Station 600 Rocky Hill Road Plymouth, MA 02360 Michael A. Balduzzi Site Vice President April 12, 2006 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555-0001

SUBJECT:

Entergy Nuclear Operations, Inc.

Pilgrim Nuclear Power Station Docket No. 50-293 License No. DPR-35 Proposed License Amendment to Change Technical Specification 3.6.A.2, Pressure-Temperature Limit Curves LETTER NUMBER: 2.06.018

Dear Sir or Madam:

Pursuant to 10 CFR 50.90, Entergy Nuclear Operations Inc. (Entergy) proposes to amend the Pilgrim Station Facility Operating License, DPR-35.

The proposed license amendment revises the Pressure - Temperature (P-T) curves, Technical Specification 3.6.A.2, Figures 3.6-1, 3.6-2, and 3.6-3, applicable for Hydrostatic and Leak Tests, Subcritical Heatup and Cooldown, and Critical Core Operation, respectively. The revised P-T curves are based on reactor pressure vessel neutron fluence calculated using NRC approved Radiation Analysis Modeling Application (RAMA) methodology, as specified in Regulatory Guide 1.190.

The current P-T curves expire at the end of the current Operating Cycle 16. Accordingly, NRC approved revised P-T curves are required for startup from refueling outage (RFO)-1 6. RFO-1 6 is currently scheduled to commence on or about April 15, 2007.

Entergy requests approval of the proposed amendment by March 15, 2007. Once approved, the amendment will be implemented within 60 days.

Entergy has reviewed the proposed amendment in accordance with 10 CFR 50.92 and concludes it does not involve a significant hazards consideration.

ADC

Entergy Nuclear Operations, Inc. Letter Number: 2.06.018 Pilgrim Nuclear Power Station Page 2 Attachment 1 provides an evaluation of the proposed change. Attachment 2 provides marked-up Technical Specification and associated Bases pages. The marked-up Bases pages are provided for information only. Attachment 3 provides description of methodology and results for the revised P-T curves.

There are no commitments made in this submittal If you have any questions or require additional information, please contact Bryan Ford at (508) 830-8403.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on the /. f* of ogre/= - 2006.

Sincerely, Michael Balduzzi WGUdm Attachments: 1. Evaluation of the Proposed Change - 6 pages

2. Proposed Technical Specifications (mark-up) - 14 pages

3. Structural Integrity Report, SIR-00-108, Rev. 2, Revised Pressure-Temperature Curves for Pilgrim, dated March 10, 2006 (37 pages) cc: Mr. James Shea, Project Manager Ms. Cristine McCombs, Director Office of Nuclear Reactor Regulation Mass. Emergency Management Agency Mail Stop: 0-8C-2 400 Worcester Road U.S. Nuclear Regulatory Commission Framingham, MA 01702 1 White Flint North 11555 Rockville Pike Rockville, MD 20852 U.S. Nuclear Regulatory Commission Mr. Robert Walker, Director Region 1 Radiation Control Program 475 Allendale Road Commonwealth of Massachusetts King of Prussia, PA 19406 90 Washington Street Dorchester, MA 02121 Senior Resident Inspector Pilgrim Nuclear Power Station

ATTACHMENT 1 Evaluation of the Proposed Chanae

Subject:

PROPOSED LICENSE AMENDMENT TO CHANGE TECHNCIAL SPECIFICATIONS 3.6.A.2, PRESSURE-TEMPERATURE LIMIT CURVES

1. DESCRIPTION

2. PROPOSED CHANGE

3. BACKGROUND

4. TECHNICAL ANALYSIS

5. REGULATORY SAFETY ANALYSIS 5.1 No Significant Hazards Consideration 5.2 Applicable Regulatory requirements and Criteria 5.3 Environmental Consideration

6. REFERENCES

Attachment 1 Page 1 of 6

1. DESCRIPTION Pursuant to 10 CFR 50.90, Entergy proposes to amend the Technical Specifications (TS) for Pilgrim Nuclear Power Station. The proposed license amendment revises the Reactor Pressure Vessel (RPV) Pressure - Temperatures (P-T) curves, Figures 3.6-1, 3.6-2, and 3.6-3, included in TS 3.6.A.2, applicable to Hydrostatic and Leak Tests, Subcritical Heatup and Cooldown, and Critical Core Operation, respectively. The revised P-T curves are based on reactor pressure vessel neutron fluence calculated using NRC approved Radiation Analysis Modeling Application (RAMA) methodology, as specified in Regulatory Guide (R.G.) 1.190.

2. PROPOSED CHANGE The current P-T curves identified in Figures 3.6-1, 3.6-2, and 3.6-3 are replaced by the revised figures.

The title blocks inthe revised P-T curves are revised to include:

"Curve Applies Through 34 EFPYs" The "intentionally left blank" pages 3/4.6-9 through 13 are deleted as they are no longer needed and subsequent pages are renumbered.

TS Bases pages B3/4.6-2 and 3 are revised to reflect the revised P-T curves and are included for information only.

3. BACKGROUND 10 CFR 50 requires that light-water nuclear power reactors meet the fracture toughness requirements for the reactor coolant pressure boundary set forth in 10 CFR 50 Appendix G. Appendix G contains the regulatory basis for the P-T curves for light-water reactors.

It specifies fracture toughness requirements for ferritic materials and pressure-retaining components of the reactor coolant pressure boundary to provide an adequate margin of safety during any condition of normal operation, including anticipated operational occurrences and system hydrostatic and leak tests, to which the pressure boundary may be subjected over its service life. Appendix G requires that the reference temperature and Charpy Upper-Shelf energy for reactor vessel beltline material account for the embrittlement caused by neutron fluence over the life of the vessel. Appendix G also requires that the P-T limits for an operating plant be at least as conservative as those that would be generated if the methods of ASME Code, Section Xl, Appendix G were applied. Furthermore, 10 CFR 50 Appendix G requires reactor vessel beltline material be tested in accordance with the surveillance program requirements of 10 CFR 50 Appendix H. In accordance with Pilgrim License Amendment 209, Pilgrim is participating in the Boiling Water Reactor Vessel and Internal Project (BWRVIP)

Integrated Surveillance program to comply with the requirements of 10 CFR 50 Appendix H.

Both ASME Section Xl Appendix G and 10 CFR 50 Appendix G provide the methodology for generating the P-T curves. The 1998 Edition of ASME Section Xl including 2000 Addenda is used for generating the P-T curves. This Edition has been endorsed in 10 CFR 50.55a. The methodology requires a safety factor of 2 on stress

Attachment 1 Page 2 of 6 intensities resulting from reactor pressure during normal and transient operating conditions and a safety factor of 1.5 for hydrostatic testing curves.

R.G. 1.190, "Calculation and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence", March 2001, describes methods and assumptions acceptable to the NRC staff for determining the pressure vessel neutron fluence. R.G. 1.99, "Radiation Embrittlement of Reactor Pressure Vessel Materials", Rev. 2, describes general procedures acceptable to the NRC staff for calculating the effects of neutron radiation embrittlement of the low-alloy steels currently used for light water-cooled reactor vessels.

10 CFR 50 Appendix G requires that the reference temperature and Charpy Upper Shelf Energy for RPV beltline materials account for the embrittlement caused by neutron fluence over the life of the vessel. The embrittlement is estimated by the shift (ARTNDT) in the initial Reference Temperature (RTNDT) due to neutron fluence 2 1Mev, which is based on a chemistry factor determined for the weld or plate material. The chemistry factor is dependent upon the copper and nickel contents of the material, and is determined from Tables 1 and 2 in R.G. 1.99, Rev. 2, or from actual surveillance data.

The fluence factor is dependent upon the neutron fluence at the maximum postulated (1/4T) flaw depth, and is determined using R.G. 1.99, Rev. 2 relationships or actual surveillance data.

Pilgrim's current P-T curves (Figures 3.6-1, 3.6-2, and 3.6-3) were approved by License Amendment 197 extending the applicability of the curves through the Operating Cycle (OC) 16 because the plant-specific calculations for the original fluence value were outdated, even though sufficient conservatism was contained in the calculations.

Amendment 197 specifically stated that in order for Pilgrim to operate beyond OC 16, a vessel fluence calculation which complies with the guidance of R.G. 1.190 should be performed. Also, Amendment 209 approved Pilgrim's participation in the BWRVIP Integrated Surveillance Program to collect new surveillance data to verify adjustments to the P-T curves for vessel heat-up and cool-down applications. In support of Amendment 209, Pilgrim committed to perform new neutron fluence calculations using R.G. 1.190 methodology and develop revised P-T curves for operation beyond OC 16 and submit the revised P-T curves in 2006 for NRC approval. This submittal fulfills that commitment.

Pilgrim has generated revised P-T curves for 24, 34, 44, and 54 EFPYs of operation.

The P-T curves were developed in accordance with 10 CFR 50 Appendix G and ASME Section Xl Appendix G,with material reference temperature shifts computed in accordance with RG 1.99, Rev. 2. At this time, Entergy proposes to operate Pilgrim on the 34 EFPY curves, which conservatively bounds operation through the end of the current operating license that expires on June 8, 2012. The actual EFPY at the end of the current operating license is estimated to be 28.2.

4. TECHNCIAL ANALYSIS Pilgrim has performed revised fluence calculations using the R.G. 1.190 guidance and generated P-T curves in accordance with 10 CFR 50 Appendix G and ASME Section Xl Appendix G with material reference temperature shifts computed in accordance with R.

G. 1.99, Rev. 2. Pilgrim neutron fluence calculations are updated using the NRC-approved RAMA Fluence Methodology (Ref. 1) for application in accordance with R.G. 1.190. The RAMA Methodology is a 3-D discrete ordinates code developed by the Electric Power Research Institute (EPRI) and BWRVIP for the purpose of calculating

Attachment 1 Page 3 of 6 neutron fluence in Boiling Water Reactors (BWRs). Benchmark testing has been performed using the methodology for several surveillance capsule and RPV fluence calculations. Results of these benchmark efforts show that the methodology accurately predicts fluence in the RPV and surveillance capsule components of BWRs. The Pilgrim fluence calculation results show the vessel would experience peak fluence for 54 EFPY 2 location and 1.28E+18 n/cm 2 at the lower of 1.14E+18 n/cm at weld 1-338AIC intermediate shell location, respectively. The N2 nozzle peak fluence at 54 EFPY was calculated to be 2.81 E+1 7 n/cm 2.

Attachment 3 describes the methodology used based on the requirements of Appendix G to 10 CFR 50 and Appendix G to ASME Section Xl to develop the revised P-T curves.

As presented in Attachment 3, RPV beltline, bottom head, and feedwater nozzle/upper vessel regions are evaluated. These regions bound all other regions with respect to brittle fracture, including the N2 nozzles. In addition, limiting stresses for the bottom head (CRD penetration) region were selected from all applicable design basis transients in order to accommodate any potential inadvertent bottom head cooldown events.

The proposed P-T curves for hydrostatic and leak tests (Fig. 3.6-1), subcritical heatup and cooldown (Fig. 3.6-2) and critical core operation (Fig. 3.6-3) were generated with fuel in the vessel for 34 effective full power years. The other changes proposed (e.g.

deletion of unused pages) are purely editorial in nature and have no impact on safety.

5. REGULATORY ANALYSIS 5.1 No Significant Hazards Consideration The proposed license amendment requests the use of revised pressure-temperature (P-T) curves, which are generated using Nuclear Regulatory Commission (NRC) approved methodology. Entergy has evaluated whether or not a significant hazards consideration is involved with the proposed amendment extending the applicability of the Pressure-Temperature curves in Figures 3.6-1, 3.6-2, and 3.6-3 by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of Amendment," as discussed below:

1. Does the proposed change involve a significant increase inthe probability or consequences of an accident previously evaluated?

Response: No.

The proposed License Amendment (LA) does not involve a significant increase inthe probability or consequences of an accident previously evaluated. There are no physical changes to the plant being introduced by the proposed changes to the pressure-temperature curves. The proposed change does not modify the reactor coolant pressure boundary, (i.e., there are no changes in operating pressure, materials, or seismic loading). The proposed change does not adversely affect the integrity of the reactor coolant pressure boundary such that its function in the control of radiological consequences is affected.

The proposed pressure-temperature curves are generated in accordance with the fracture toughness requirements of 10 CFR 50 Appendix G, and American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code, Section Xl, Appendix G and Regulatory Guide (R.G.)

1.99, Revision 2. "Radiation Embrittlement of Reactor Vessel Materials."

Attachment 1 Page 4 of 6 A best-estimate calculation of reactor vessel 34 effective full power years (EFPYs) neutron fluence and associated uncertainty has been completed for Pilgrim using the Radiation Analysis Modeling Application (RAMA) methodology. This methodology was previously approved by the NRC. The resulting reactor vessel neutron fluence value was then used in conjunction with R.G. 1.99, Rev. 2 to determine the adjusted reference temperature (ART) and with ASME Section Xl Appendix G to develop revised P-T curves. This provides sufficient assurance that the Pilgrim reactor vessel will be operated in a manner that will protect it from brittle fracture under all operating conditions. This proposed license amendment provides compliance with the intent of 10 CFR 50 Appendix G and provides margins of safety that assure reactor vessel integrity.

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

2. Do the proposed changes create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed license amendment does not create the possibility of new or different kind of accident from any accident previously evaluated. The revised pressure-temperature curves are generated in accordance with the fracture toughness requirements of 10 CFR Part 50 Appendix G and ASME Section Xl Appendix G. Compliance with the proposed pressure-temperature curves will ensure the avoidance of conditions in which brittle fracture of primary coolant pressure boundary materials is possible because such compliance with the pressure-temperature curves provides sufficient protection against a non-ductile-type fracture of the reactor pressure vessel.

No new modes of operation are introduced by the proposed change. The proposed change will not create any failure mode not bounded by previously evaluated accidents. Further, the proposed change does not affect any activities or equipment and is not assumed in any safety analysis to initiate any accident sequence. This provides sufficient assurance that Pilgrim reactor vessel will be operated in a manner that will protect it from brittle fracture under all operating conditions.

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

3. Do the proposed changes involve a significant reduction in a margin of safety?

Response: No.

The proposed license amendment requests the use of revised P-T curves that are based on established NRC and ASME methodologies. A best-estimate calculation of reactor vessel neutron fluence and associated uncertainty has been completed for Pilgrim through 34 EFPY using the NRC approved RAMA methodology. The 34 EFPY reactor vessel neutron fluence value was used in conjunction with R. G. 1.99, Rev. 2 to compute reference temperature shift, and with ASME Section Xl Appendix G to develop revised P-T curves. This provides sufficient margin such that the Pilgrim reactor

Attachment 1 Page 5 of 6 vessel will be operated in a manner that will protect it from brittle fracture under all operating conditions. Operation within the proposed limits ensures that the reactor vessel materials will continue to behave in a non-brittle manner, thereby preserving the original safety design bases. No plant safety limits, set points, or design parameters are adversely affected by the proposed changes.

Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

Based on the above, Entergy concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and accordingly, a finding of "no significant hazards consideration" is justified.

5.2 Applicable Regulatory Requirements/Criteria The Pilgrim revised P-T curves were developed in accordance with NRC approved methodology and provide adequate protection against brittle fracture of the reactor pressure vessel.

10 CFR 50.60, 'Acceptance Criteria for Fracture Prevention Measures for Lightwater Nuclear Power Reactors for Normal Operation," provides the requirement that the pressure and temperature limits as well as the associated vessel surveillance program are consistent with 10 CFR 50, Appendix G,"Fracture Toughness Requirements," and 10 CFR 50, Appendix H, 'Reactor Vessel Material Surveillance Program Requirements."

10 CFR 50 Appendix G and Appendix H describe specific requirements for fracture toughness and reactor vessel material surveillance that must be considered in establishing P-T limits. As an alternative to the Appendix H requirements, NRC approved the Pilgrim participation in the BWRVIP Integrated Surveillance Program.

10 CFR 50 Appendix G specifies that the fracture toughness and testing requirements for reactor vessel material are determined in accordance with ASME Section Xl Appendix G. 10 CFR 50 Appendix G also requires the prediction of the effects of neutron irradiation on the vessel embrittlement. The effects of vessel embrittlement are determined in accordance with the methods provided in Regulatory Guide 1.99, Revision 2, which require the calculation of the reference temperature for nil-ductility transition (RTNDT), and other vessel embrittlement related values.

In conclusion, based on the considerations discussed above, (1)there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3)the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

5.3 Environmental Consideration Entergy has determined that the proposed amendment would change a requirement with respect to the use of a facility component located within the restricted area, as defined in 10 CFR 20. However, the proposed amendment does not involve: (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in

Attachment 1 Page 6 of 6 10 CFR 51 .22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

6. REFERENCES

1. Letter from W. H. Bateman (NRC) to W. Eaton (BWRVIP Chairman), Radiation Analysis Modeling Application (RAMA) Safety Evaluation Report, dated May 13, 2005.

ATTACHMENT 2 Proposed Technical Specifications (Marked-up pages)

(14 pages)

Marked-up TS Paaes TS page 3/4.6-9 TS page 314.6-10 TS page 314.6-11 TS page 3/4.6-12 TS page 314.6-13 TS page 314.6-14 TS page 314.6-15 TS page 314.6-16 INSERT New TS Paaes TS page 3/4.6-9 TS page 314.6-10 TS page 34.6-11 Marked-uD TS Bases Paaes TS Bases Page B3/4.6-2 TS Bases Page B3/4.6-3 Insert WAN to B3/4.6-2 att2OOl.PDF

C This page is intentionally left blank.

Amendment No. 20, 60, 93, 161, 195/ 3X4.6-9

c This page is intentionally left blank.

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Amendment No. 2O-460-93, 46, 195 3/4.6-1 0

C This page is intentionally left blank.

(f' Amendment No. 20. 10, 690,11, 153, 195 3/4.6-11t

This page is intentionally left blank.

Q.1 Q 3, 195 Amendment No. 20, 40, 60, 3/4.6-12

(

ThIs page Is intenfonadly left blank I& X

( /

Amendment 82. lI4, 188,. 209 3/4.6-13

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PILGRIM REACTOR VESSEL PRESSURE-TEMPERATURE HYDROSTATIC AND LEAK TESTS LIMITS

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.:- _-' - 1 11 1 1 1- Pilgrim Reactor Pressure Vessel Pressure-Temperature 200 Umits for Hydrostatic Leak Tests 7-

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. . . ........ I _.I-50 60 70 80 90 100 110 120 130 140 150 160 170 180 190 200 210 220 230 240 250 260 270 TEMPERATURE ('F)

Amendment No. 28 , O2.04r440,197 3/4.6-14

PILGB.IM REACTOR VESSEL PRESSURE-TEMPERATURE SUBCR~mCAL HEATUP AND COOLDOWN UMITS 12100_.-...- . = . - .:=

o0 -. -. I_-L---.---.--_

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50 80 90 100 110 120 130 140 10. 160 170 180 190 200 210 220 230 770 240 250 260 270

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~TEMPERATURE ( F)

Amendment No. 2 8 r 82 4

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PILGRIM REACTOR VESSEL PRESSURE-TEMPERAMRE UMITS CRMCAL CORE OPERAMON I

1,0 1,100 1,000 oa 2: 700 900

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600 500 I

400 300 200 I

FIGURE 3.63 _

Plfgrim Reactor Pressure Vessel Pressure-Temperature -

100 1-

- Umits for Critical Core Operation The curve applies through Operating Cycle 16

. . I -. .I I .I . I I I I 50 60 80 90 100 110 120 130 140 150 160 170 180 190 200 210 220 230 240 250 280 270 TEMPERATURE ('F)

Amendment No. 440,197 314.6-16

PILGRIM REACTOR VESSEL PRESSURE-TEMPERATURE UMMS HYDROSTATIC AND LEAK TESTS I'm00 E -BHM 1,100 --

900 800-I I

I Lu 6o0 S00 300 :4.4--

200 FIGURE 3.6-1 .

P9IndAn RAe or Pressure Vesel Pressure-Temperaturenfts 100 for HydrstIc Loek Tes

- l -. Curve Appes Through 34 EFPYs 0 pm. , . .

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50 60 70 80 90 100 110 120 130 140 150 160 170 180 190 200 210 220 230 240 250 260 270 TEMPERATURE (on 3/4.6-9 Amendment No. 28, 824,41 140, 1 04,

EILGRIM REACTOR VESSEL PRESSURE-TEMPERATURE UNITS SUBCRMCAL HEATUP AND COOLDOWN m *- *- p- U - . m I - I - I -

1,100 owl lml HM 1,000 900 600 700 w

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400 200 100 0 anew_- - -

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FIGURE 3.02 Vsel PressureTempertu for Subcrical Hea ad Cooldown Cudve AppIles Throuh 34 EFPYS I Oi t i + - - -a i.-.+-

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50 60 70 80 90 100 110 120 130 140 150 160 170 180 190 200 210 220 230 240 250 260 270 TEMPERATURE ( F)

Amendment No. 2 I2 8 I44,149h 3/4.6-10

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PILGRIM REACTOR VESSEL PRESSURE-TMPEAMRE UMIM CRITICAL CORE OPERATION i20O *1 1,100 1,000 80 700 U'

e 600.

400 300 200 t'- F-+- -I 4 I - t -- I -4 -- FIGURE 36-3 Plftfrlm B.No-I Raetor Pressur Vona P. se-Temerature U nfts 100 for C~ricul Core Opeaton L--IYJI--47 -LI I- VT- Ca" Applles Through 34 EFPYs 0 -Y q- I- m9 4II +i It I  ! I iI--i I4 I I 1-4--+-4-*--F-4---I----- -s -- -- ,4 t 'a I '

50 60 70 60 90 100 110 120 130 140 150 160 170 180 190 200 210 220 230 240 250 210 270 TEMPERATURE (F)

Amendment No. 28,J82,01,140,1074.

.3/4.-11

BASES:

314.6 PRIMARY SYSTEM BOUNDARY (Cont)

C A. Thermal and Pressurization Limitations (Cont)

For critical core operation when the water level Iswithin the normal range for power operation and the pressure is less than 20% of the preservice system hydrostatic test pressure (313 psq, the minimum permissible temperature of the highly stressed regions of the closure flange IsRT,, + 60 degrees F = 70 degrees F; thus, a cutoff limit of 70 degrees Fwas chosen as shown on Figure 3.6-3 and as required by 10CFR5O Appendix G,paragraph IV.A3. This same cutoff is conservatively included inthe limits for hydrostatic and leak tests and for non-critical operation, as shown on Figures 3.6-1 and 3.6-2, respectively, Inorder to be consistent with the limits for critical operation.

The closure region Ismore limiting than the feedwater nozzle with respect to both stress Intensity and RT,. Therefore, the pressure-temperature limits of the closure are controlling.

The minimum bolt-up temperature Isthe minimum allowable nil ductility transition temperature (RT,) at pressures below the 20% of pre-operatlonal system hydrostatic test pressure that bolt pre-load stress can be applied to the reactor vessel closure regio. I Isdefined as the Initial RT,, of the higher stressed component of the reactor vessel that Includes the vessel head, head flange and shell adjacent to the vessel flange. The maxdmum RT,, Is+ 10 degrees F. For conservatism a minimum boft-up temperature of 55 degrees F Ischosen because this temperature provides sufficient margin between the lowest service temperature at 20% of the pre-operational hydrostatic test pressure prior to pressurization. CAt 41A e

-rtor M-a0 X bC 11ca6}

ul The adjusted reference temperature shft Isbased on Regulatory Guide 1.99, Resdef 2;

-0 U, the analytical resufts of __t _-M61E MC+ Ae .on. 1,datsdJan&Rt'2.M5. regarding projected neutron fluence; as A Bc I1fun-W1 t Smperaure and p-esro ofGI l 2 1~S B. Coolant Chemistry ( Ad4 r "StAA

.4,.1012.00 C v b i or)

The reactor vessel coolant chemistry requirements are discussed in Subsection 4.2 the FSAR.

A radioactivity concentration of 20 , C~mI total Iodine can be reached if there Is significant fuel failure or If there Is a failure or a prolonged shutdown of the cleanup demineralizer. Calculations performed by the AEC staff for this activity level results in a radiological dose at the site boundary of 8 rem to the thyroid from a postulated rupture of a main steam line assuming a 5 second valve closing time and a coolant Inventory release of 3 x 104 nbs.

A reactor sample will be used to assure that the limit of Specification 3.6.6.1 is not exceeded.

e t/4.6-3

INSERT "A" to B3/4.6-2 The bottom head defined as the spherical portion of the reactor vessel located below the lower circumferential weld, was also evaluated. Reference transition temperatures (RTS) were developed for the bottom head and the resulting pressure vs. tempeAature curves plotted on Figures 3.6-1, 3.6-2, and 3.63. Stress results specific to the bottom head from Attran Report 93177-TR-03, Pilgrim Reactor Cyclic Load Analysis, August 1994 were combined with the pressurization temperatures necessary to maintain fracture toughness requirements in accordance with ASME Boiler and Pressure Vessel Code, Section Xl, Appendix G, the criteria of 10CFR50, Appendix G, and the requirements of Reg. Guide 1.99, Rev 2. Limiting stresses for the bottom head (CRD penetration) region were selected from all applicable design basis transients in order to accommodate any potential inadvertent bottom head cool down events.

ATTACHMENT 3 Structural Integrity Report, SIR-00-108. Rev. 2.

Revised Pressure-Temperature Curves for Pilgrim. dated March 10, 2006 (37 pages) att300l.PDF

Structural Integrity Associates, Inc.

6855 S. Havana Street Sulte 350 Centennial, CO 80112-3868 Phone: 303-792-0077 Fax: 303-792-2158 WWw.Suctbtcom gstevensmstructitacon March 10, 2006 GLS-06-013 SIR-00-108, Rev. 2 Mr. George Mileris Entergy Operations, Inc.

Pilgrim Nuclear Power Station 600 Rocky Hill Road Plymouth, Massachusetts 02360-5599

Subject:

Revised Pressure-Temperature Curves for Pilgrim

Reference:

Entergy Purchase Order No. 4500544139 dated 10/26/2005.

Dear George:

The attachment to this letter documents the revised set of pressure-temperature (P-T) curves developed for the Pilgrim Nuclear Power Station (PNPS), in accordance with SI's Quality Assurance Program. This work was performed in accordance with the referenced contract, and includes a full set of updated P-T curves (i.e., pressure test, core not critical, and core critical conditions) for PNPS for 24, 34, 44, and 54 effective full power years (EFPYs). The curves were developed in accordance with ASME Code,Section XI, Appendix G (1998 Edition including the 2000 Addenda) and U.S. 10CFR 50 Appendix G.

The inputs, methodology, and results for this effort are summarized in the attachment.

Please don't hesitate to call me if you have any questions.

Prepared By: _ W IW Reviewed By: lzee H. L. Gustin, P. E.

Gary L. Stevens, P. E.

Senior Associate Associate Approved By: _ $, h (O GaryIY Stevens, P. E.

Senior Associate gls Attachments cc: PNPS-22Q-106 (cover only), PNPS-03Q-106 (cover only), PNPS-03Q-403 Austin,TX Clarlofe NIC ILStonington, Cr San Jose, CA Slier Spdng, ED Sundse, FL UnwontnOH Witter, CA 512-59191 70497-554 950-5994050 40697880 3014450 954-572-2902 330-99753 562-944-20

ATTACHMENT REVISED P-T CURVES FOR PILGRIM 1.0 Introduction This attachment documents the revised set of pressure-temperature (P-T) curves developed for the Pilgrim Nuclear Power Station (PNPS), in accordance with Si's Quality Assurance Program.

This work includes a full set of updated P-T curves (ie., pressure test, core not critical, and core critical conditions) for PNPS for 24, 34, 44, and 54 effective full power years (EFPYs).

The curves were developed using the methodology specified in ASME Code,Section XI, Appendix G (1998 Edition including the 2000 Addenda) [2], 10CFR50 Appendix G [3], and WRC-175 [4].

2.0 ARTNDT Values Adjusted reference temperature (ARTNDT) values were developed for all limiting PNPS reactor pressure vessel (RPV) materials' in the Reference [1] calculation.

3.0 P-T Curve Methodology The P-T curve methodology isbased on the requirements of References [2] through [4]. The supporting calculations for the curves are contained in References [5] and [6]. There are three regions of the reactor pressure vessel (RPV) that are evaluated: (1) the beltline region, (2) the bottom head region, and (3) the feedwater nozzle/upper vessel region. These regions bound all other regions with respect to brittle fracture. In addition, limiting stresses for the bottom head region were selected from all applicable design basis transients in order to accommodate any potential inadvertent bottom head cooldown events.

The approach used includes the following steps:

a. Assume a fluid temperature, T. Determine the temperature at the assumed flaw tip, T,/4, (ie., 1/4t into the vessel wall). For the purposes of this assessment, T,,4 ,

is conservatively assumed to be equal to the fluid temperature.

b. Calculate the allowable stress intensity factor, As, based on T/4, using the relationship from ASME Code,Section XI, Appendix G [2], as follows:

2

'Section 7 of Reference [71 also reports fluence values for the recirculation inlet (N2) nozzles as I .30x 1017 n/cm 2 2 for 20.7 EFPY and 2.8lxlO7 n/cm for 54 EFPY. Both of these values are greater than the l.Ox lol n/cm threshold fluence value. Evaluation of these nozzles was addressed in Reference [8], and it was determined that there was no impact on the P-T curves.

Attachment to SIR-00-108, Rev. 2 1 Structural Integrity Associates, Inc.

K = 20.734 e[J02(TI4,ARTNA)I + 33.2 where: TI/4t = metal temperature at assumed flaw tip (OF)

ARTNDT = adjusted reference temperature for location under consideration and desired EFPY (IF)

= allowable stress intensity factor (ksiIinch)

c. Calculate the thermal stress intensity factor, Kvr from ASME Code,Section XI, Appendix G [2] for the beltline region, or from finite element results for the bottom head and feedwater nozzle/upper vessel regions.

d. Calculate the allowable pressure stress intensity factor, KIp, using the following relationship:

KIp = (Kic-Krr)/SF where: Krp = allowable pressure stress intensity factor (ksWinch)

SF = safety factor

= 1.5 for pressure test conditions (Curve A)

- 2.0 for heatup/cooldown conditions (Curves B and C)

e. Compute the allowable pressure, P, from the allowable pressure stress intensity factor, VC4p.

f. Subtract any applicable adjustments for temperature and/or pressure from T and P, respectively.

g. Repeat steps (a) through (f) for other temperatures to generate a series of P-T points.

The following additional requirements were used to define the P-T curves. These limits are established in Reference [3]:

ForPressure Test Conditions(Curve A):

• If the pressure is greater than 20% of the pre-service hydro test pressure, the temperature must be greater than ARTNDT of the limiting flange material + 901F.

• If the pressure is less than or equal to 20% of the pre-service hydro test pressure, the minimum temperature is must be greater than or equal to the ARTNDT of the limiting flange material + 600F. This limit has been a standard recommendation for the BWR industry for non-ductile failure protection.

Attachment to SIR-00-108, Rev. 2 2 C Structural IntegrityAssociates, Inc.

For Core Not CriticalConditions (Curve B):

• If the pressure is greater than 20% of the pre-service hydro test pressure, the temperature must be greater than ARTNDT of the limiting flange material +

120 0F.

• If the pressure is less than or equal to 20% of the pre-service hydro test pressure, the minimum temperature must be greater than or equal to the ARTNDT of the limiting flange material + 60 0F. This limit has been a standard recommendation for the BWR industry for non-ductile failure protection.

ForCore CriticalConditions (Curve C!:

• Per the requirements of Paragraph IV.A.2 of Reference [3], the core critical P-T lnits must be 40'F above any Pressure Test or Core Not Critical curve limits.

Core Not Critical conditions are more limiting than Pressure Test conditions, so Core Critical conditions are equal to Core Not Critical conditions plus 40'F.

• Another requirement of Paragraph IV.A.2 of Reference [3] (or actually an allowance for the BWR), concerns minimum temperature for initial criticality in a startup. Given that water level is normal, BWRs are allowed initial criticality at the closure flange region temperature (ARTNDT + 60'F) if the pressure is below 20% of the pre-service hydro test pressure.

• Also per Paragraph IV.A.2 of Reference [3], at pressures above 20% of the pre-service hydro test pressure, the Core Critical curve temperature must be at least that required for the pressure test (Pressure Test Curve at 1,100 psig). As a result of this requirement, the Core Critical curve must have a step at a pressure equal to 20% of the pre-service hydro pressure to the temperature required by the Pressure Test curve at 1,100 psig, or Curve B + 400 F, whichever is greater.

The resulting pressure and temperature series constitutes the P-T curve. The P-T curve relates the minimum required fluid temperature to the reactor pressure.

4.0 P-T Curves Tabulated values for the P-T curves are shown in Tables 1 through 16. The resulting P-T curves are shown in Figures I through 12. The P-T curves are shown using Technical Specification formatting requirements in Appendix A.

Attachment to SIR-00-108, Rev. 2 3 Structural Integrity Associates, Inc.

5.0 References

1. Structural Integrity Associates Calculation No. PNPS-22Q-301, Revision 0, "ARTNDT and ART Evaluation."

2. ASME Boiler and Pressure Vessel Code,Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, Nonmandatory Appendix G, "Fracture Toughness Criteria for Protection Against Failure," 1998 Edition, 2000 Addenda.

3. U. S. Code of Federal Regulations, Title 10, Part 50, Appendix G, "Fracture Toughness Requirements," January 2005.

4. WRC Bulletin 175, "PVRC Recommendations on Toughness Requirements for Ferritic Materials," PVRC Ad Hoc Group on Toughness Requirements, Welding Research Council, August 1972.

5. Structural Integrity Associates Calculation No. PNPS-03Q-301, Revision 1, "Development of Pressure Test (Curve A) P-T Curves."

6. Structural Integrity Associates Calculation No. PNPS-03Q-302, Revision 1, "Development of Heatup/Cooldown (Curves B & C) P-T Curves."

7. TransWare Enterprises Report No. ENT-FLU-001 -R-001, Revision 0, "Pilgrim Nuclear Power Station Reactor Pressure Vessel Fluence Evaluation at End of Cycle 15 and 54 EFPY," 12/20/05, SI File No. PNPS-22Q-201.

8. Structural Integrity Associates Calculation No. PNPS-22Q-302, Revision 0, "N2 Nozzle Evaluation."

Attachment to SIR-00-108, Rev. 2 4 Structurallntegrity Associates, Inc.

Table 1 Tabulated Values for Beltline Pressure Test Curve (Curve A) for 24 EFPY Pressure-Temnerature Curve Calculation (Pressure Test = Curve A)

Inputs Plant = Pilgrim Component = Beltilne Vessel thickness. t = 5.5312 inches, so ft = 2.352 flinch Vessel Radius, R = 113.91 Inches ARTNDT = 68.1 -F = => 24 EFPY Cooldown Rate, CR = 0 F/hr KIT = 0.00 ksl*inch'12 (From Appendix G, for cooldown rate abowe)

AT14=- 0. 0 *F (no thermal for pressure test)

Safety Factor = 1.50 (for pressure test)

Mm = :2.178 (From Appendix G. for inside surface axial taw)

Temperature Adjustment = 0.0 *F Height of Water for a Full Vessel = 507.5 Inches Pressure Adjustment = :18.3 psig (hydrostatic pressure for a full %esselat 70OF)

Hydro Test Pressure = 1,565 psig Flange RTNDT = 10.0 OF Fluid Calculated Adjusted Adjusted Temperature 114t Pressure Temperature Pressure for T Temperature Kic Kip P for P-T Curve P-T Curve (F) 70F0 (ksrinch1 2) (ksrinch1 2) psls) ('F) (PSg) 70.0 70.0 54.74 36.49 0 70.0 0 70.0 70.0 54.74 36.49 814 70.0 795 75.0 75.0 57.00 38.00 847 75.0 829 80.0 80.0 59.51 39.67 885 80.0 866 85.0 85.0 62.27 41.51 926 85.0 907 90.0 90.0 65.33 43.55 971 90.0 953 95.0 95.0 68.71 45.81 1021 95.0 1,003 100.0 100.0 72.44 48.30 1077 100.0 1,058 105.0 105.0 76.57 51.05 1138 105.0 1,120 110.0 110.0 81.13 54.09 1206 110.0 1,188 115.0 115.0 86.17 57.45 1281 115.0 1,263 120.0 120.0 91.74 61.16 1364 120.0 1,345 125.0 125.0 97.90 65.27 1455 125.0 1,437 130.0 130.0 104.71 69.80 1556 130.0 1,538 135.0 135.0 112.23 74.82 1668 135.0 1,650 140.0 140.0 120.54 80.36 1792 140.0 1,773 145.0 145.0 129.72 86.48 1928 145.0 1,910 Attachment to SIR-00-108, Rev. 2 S A0 StruCtUral integrity Associates, Inc.

Table 2 Tabulated Values for Beltline Pressure Test Curve (Curve A) for 34 EFPY Pressure-Temperature Curve Calculation (Pressure Test = Curve A) hiputs Plant = -Pilgrim Component = Beltlne Vessel thickness, t = :5.5312 inches, so '4t = 2.352 4inch Vessel Radius, R = - 113.91 Inches ARTNDT = 81.0 F= 34 EFPY Cooldown Rate, CR = 0 F/hr KIT = 0.00 ksI*inch11 2 (From Appendix G, for cooldown rate above)

AT1 /4t = 0.0 *F (no thermal for pressure test)

Safety Factor = 1.50 (for pressure test)

Mm = 2.178 (From Appendix G, for Inside surface axial 11aw)

Temperature Adjustment = 0.0 -F Height of Water for a Full Vessel = 507.5 Inches Pressure Adjustment = 18.3 psig (hydrostatic pressure for a full %esselat 70'F)

Hydro Test Pressure = 1,565 psig Flange RTNDT = 10.0 OF Fluid Calculated Adjusted Adjusted Temperature 114t Pressure Temperature Pressure for T Temperature Kic Kgp P for P-T Curve P-T Curve rF) 7.F) (ksi*lnch¶ ) (ksi-inch1 2) 12 J sg) (F) 70.0 70.0 49.84 33.23 0 70.0 0 70.0 70.0 49.84 33.23 741 70.0 723 75.0 75.0 51.59 34.39 767 75.0 749 80.0 80.0 53.52 35.68 796 80.0 777 85.0 85.0 55.66 37.11 827 85.0 809 90.0 90.0 58.02 38.68 862 90.0 844 95.0 95.0 60.63 40.42 901 95.0 883 100.0 100.0 63.52 42.35 944 100.0 926 105.0 105.0 66.71 44.47 992 105.0 973 110.0 110.0 70.23 46.82 1044 110.0 1,026 115.0 115.0 74.13 49.42 1102 115.0 1,084 120.0 120.0 78.43 52.29 1166 120.0 1,147 125.0 125.0 83.19 55.46 1237 125.0 1,218 130.0 130.0 88.44 58.96 1315 130.0 1,296 135.0 135.0 94.25 62.84 1401 135.0 1.383 140.0 140.0 100.68 67.12 1496 140.0 1,478 145.0 145.0 107.77 71.85 1602 145.0 1,584 150.0 150.0 115.62 77.08 1719 150.0 1,700 155.0 155.0 124.28 82.86 1847 155.0 1,829 Attachment to Si R 108, Rev. 2 6 0 StructuralIntegrityAssociates, Inc.

Table 3 Tabulated Values for Beltline Pressure Test Curve (Curve A) for 44 EFPY Pressure-Temnerature Curve Calculation (Pressure Test = Curve A)

Inputs Plant= Pllgrim Component = BeltUne Vessel thickness, t= 5.5312 inches, so 41= 2.352 41inch Vessel Radius, R= 113.91 inches ARTNDT = 88.0 OF ===> 44 EFPY Cooldown Rate. CR = 0 - F/hr KIT = 0.00 kslinch1't2 (From Appendix G. for cooldown rate above) bTI=;4t 0.0 OF (no thermal for pressure test)

Safety Factor = 1.50 (for pressure test)

Mm = 2.178 (From Appendix G, for inside surface axial law)

Temperature Adjustment = 0.0 OF Height of Water for a Full Vessel = 507.5 inches Pressure Adjustment = 18.3 psig (hydrostatic pressure for a full vessel at 70'F)

Hydro Test Pressure = 1,565 psig Flange RTNDT = 10.0 *F Fluid Calculated Adjusted Adjusted Temperature 114t Pressure Temperature Pressure for T Temperature Kc Kip p for P-T Curve P-T Curve (F) ( F) (ksl*lnchl') (ksi~inch"2) (psig) rF) SW) 70.0 70.0 47.67 31.78 0 70.0 0 70.0 70.0 47.67 31.78 709 70.0 690 75.0 75.0 49.19 32.79 731 75.0 713 80.0 80.0 50.87 33.91 756 80.0 738 85.0 85.0 52.73 35.15 784 85.0 765 90.0 90.0 54.78 36.52 814 90.0 796 95.0 95.0 57.05 38.03 848 95.0 830 100.0 100.0 59.56 39.71 885 100.0 867 105.0 105.0 62.33 41.55 926 105.0 908 110.0 110.0 65.39 43.60 972 110.0 954 115.0 115.0 68.78 45.85 1022 115.0 1,004 120.0 120.0 72.52 48.35 1078 120.0 1,060 125.0 125.0 76.66 51.10 1139 125.0 1,121 130.0 130.0 81.23 54.15 1207 130.0 1,189 135.0 135.0 86.28 57.52 1282 135.0 1,264 140.0 140.0 91.86 61.24 1365 140.0 1,347 145.0 145.0 98.03 65.35 1457 145.0 1,439 150.0 150.0 104.85 69.90 1559 150.0 1,540 155.0 155.0 112.38 74.92 1671 155.0 1,652 160.0 160.0 120.71 80.47 1794 160.0 1,776 165.0 165.0 129.92 86.61 1931 165.0 1,913 Attachment to SIR-00-108, Rev. 2 7 h StructuralIntegrity Associates, Inc.

Table 4 Tabulated Values for Beltline Pressure Test Curve (Curve A) for 54 EFPY Pressure-Temnerature Curve Calculation (Pressure Test = Curve A)

Inputs: Plant = Pilgrim Component = Beltine Vessel thickness, t = 5.5312- Inches, so qt = 2.352 4inch Vessel Radius, R = 113.91 Inches ARTNOT 05.3 - F 54 EFPY Cooldown Rate, CR = 0 *Fhr KIT = 0.00 ksi*inch't 2 (From Appendix G, for cooldown rate above)

AT114 t = 0.0 -F (no thermal for pressure test)

Safety Factor = 1.50 -(for pressure test)

M, = 2.178 (From Appendix G, for inside surface axial flaw)

Temperature Adjustment = 0.0 F Height of Water for a Full Vessel = 507.5 Inches Pressure Adjustment = 18.3 psig (hydrostatic pressure for a full vessel at 700F)

Hydro Test Pressure = 1,565 psig Flange RTNDT = 10.0 F Fluid Calculated Adjusted Adjusted Temperature 114t Pressure Temperature Pressure for T Temperature Kc Kip P for P-T Curve P-T Curve 2

(F) (OF) (ksl*Inchl ) (ksd~lnch"2) (psig) (-F) (PSlg) 70.0 0 70.0 70.0 45.70 30.47 0 70.0 70.0 45.70 30.47 679 70.0 661 75.0 75.0 47.02 31.34 699 75.0 681 80.0 80.0 48.47 32.31 720 80.0 702 85.0 85.0 50.07 33.38 744 85.0 726 90.0 90.0 51.85 34.57 771 90.0 752 95.0 95.0 53.81 35.87 800 95.0 782 100.0 100.0 55.98 37.32 832 100.0 814 105.0 105.0 58.37 38.92 868 105.0 849 110.0 110.0 61.02 40.68 907 110.0 889 115.0 115.0 63.95 42.63 951 115.0 932 120.0 120.0 67.18 44.79 999 120.0 980 125.0 125.0 70.75 47.17 1052 125.0 1,033 130.0 130.0 74.70 49.80 1110 130.0 1.092 135.0 135.0 79.07 52.71 1175 135.0 1,157 140.0 140.0 83.89 55.93 1247 140.0 1,229 145.0 145.0 89.22 59.48 1326 145.0 1,308 150.0 150.0 95.12 63.41 1414 150.0 1,396 155.0 155.0 101.63 67.75 1511 155.0 1,492 160.0 160.0 108.82 72.55 1618 160.0 1,599 165.0 165.0 116.78 77.85 1736 165.0 1,717 170.0 170.0 125.57 83.71 1866 170.0 1.848 Attachment to SIR-00-108, Rev. 2 8 Am tStructuralIntegrity Associates, Inc.

Table 5 Tabulated Values for Feedwater Nozzle/Upper Vessel Region Pressure Test Curve (Curve A)

Pressure-Ternnerature Curve Calculation (Pressure Test = Curve A)

Inputs: Plant = Pilgrlim Component =Upper Vessell (based on FW nozzle)

Nozzle thickness, t= 7 .0 inches, so ~t= 2.646 -4inch ARTNDT= 10,0 OF === -All EFPYs: Conserwatiely useflange.

K, 120.9 ksi~inch1B2 for a pressure of 1,565 psig Safety Factor= 1.50 --

Temperature Adjustment = 0.0 -- ' F Pressure Adjustment = -; 0.0 -psig Unit Pressure = fV1',565 psig Flange RTNDT =  :. 10.0 *F Fluld Calculated Adjusted AdJusted Temperature 1/4t Pressure Temperature Pressure for T Temperature Kic K~p P for P-T Curve P-T Curve rF) rF) (ksl~inchlrl) (ksilinch 112) 08d0) -

('IF) (P8i9) 70.0 70.0 102.04 68.03 0 70.0 70.0 70.0 102.04 68.03 313 70.0 313 100.0 100.0 158.63 105.76 313 100.0 313 100.0 100.0 158.63 105.76 1369 100.0 1369 105.0 105.0 171.83 114.55 1483 105.0 1483 1 10.0 110.0 186.40 124.27 1609 110.0 1609 Attachment to SIR-00-108, Rev. 2 9 V Structulral IntegrityAssociates, Inc.

Table 6 Tabulated Values for Bottom Head Pressure Test Curve (Curve A)

Pressure-Temnerature Curve Calculation (Pressure Test = Curve A)

Inputs. Plant = . .TPliprim Component = Bottom Head (Penetrations Portion)

Vessel thickness, t = 7.250 inches, so At = 2.693 4inch ARTNDT = 29.0 . F > All EFPYs Klp= 39.6 - ksilinch1 2 (for a pressure of 1,180 psig)

Safety Factor= 1:.0.

Temperature Adjustment = 0.0 F Height of Water for a Full Vessel = 0.0 inches (FEM stresses Include weight of water)

Pressure Adjustment = 0.0. psig Unit Pressure = 1,180 psig Flange RTNDT = i 10.0 OF Fluid Calculated Adjusted Adjusted Temperature 114t Pressure Temperature Pressure for T Temperature Kic 1lp P for P-T Curve P-T Curve 7.F) (`F) (kslinch1 2) (khl*inchl 2 ) (psig) (OF) (psig) 70.0 70.0 80.28 53.52 0 70 0 70.0 70.0 80.28 53.52 1595 70 1595 75.0 75.0 85.23 56.82 1693 75 1693 Attachment to SIR-00-108, Rev. 2 10 V StructuralIntegrityAssociates,Inc.

Table 7 Tabulated Values for Beltline Core Not Critical Curve (Curve B) for 24 EFPY Pressure-Tremperature Curve Calculation (HeatupVooldowi, Core Not Crifical = Curve B)

InPdsl lant a Pilgrim Component - :Belaine Vessel thickness, t a -5.5312 inches. so 11- 2.352 Jinch Vessel Radius. R a 113.91 Inches ARTNDr w 68.1 -*F =2a3 24 EFPY Coddown Rate, CR = 100 *F/hr K, T 6.86 ksilnch'0 (From Appendix G, br cooldown rsate abo)

. 0.0 . F (consermtlity neglect)

Safety Factor 2.00 (for heatuplcooldown)

U, a 2.178 (Frrn Appendix G, fr inside surface axial taw)

Temperature A4ustment a 0.0 *F Height of Water fbr a Full Vessel 5507.5 inches Pressure Adjustment a 18.S psig (hydrostatic pressure for a lul wesse at 70F)

Hydro Test Pressure a 1,565 psig Flange RTNDT 10.6 - F Fluid Calculated Adjusted Adjusted Temperature 114t Pressure Temperature Pressure for T te perature K'c  : .l P for P-T Curve P-T Curve 0F. ksltlctIlia) (kai~nch'21 (psig) (F) (ps~q) 0 0.0 0 0 0 70.0 -18 0

38.51 15.83 353 70.0 335 5 5.0 39.07 16.11 359 70.0 341 10 10.0 39.69 16.41 366 70.0 348 15 15.0 40.37 16.76 374 70.0 355 20 20.0 41.12 17.13 382 70.0 364 25 25.0 41.96 17.55 391 70.0 373 30 30.0 42.88 18.01 402 70.0 383 35 35.0 43.89 18.52 413 70.0 395 40 40.0 45.02 19.08 425 70.0 407 45 45.0 46.26 19.70 439 70.0 421 50 50.0 47.64 20.39 455 70.0 436 55 55.0 49.16 21.15 472 70.0 453 60 60.0 50.83 21.99 490 70.0 472 65 65.0 52.69 22.92 511 70.0 493 70.0 70.0 54.74 23.94 534 70.0 515 70.0 70.0 64.74 23.94 534 70.0 515 75.0 75.0 57.00 25.07 559 750 541 80.0 80.0 59.51 26.32 587 80.0 569 85.0 85.0 62.27 27.71 618 85.0 599 00.0 90.0 65.33 29.24 652 90.0 634 95.0 95.0 68.71 30.93 690 95.0 671 100.0 100.0 72.44 32.79 731 100.0 713 105.0 105.0 76.57 34.86 777 105.0 759 110.0 110.0 81.13 37.14 828 110.0 810 115.0 115.0 88.17 39.66 884 115.0 866 120.0 120.0 91.74 42.44 946 120.0 928 125.0 125.0 97.90 45.52 1015 125.0 997 130.0 130.0 104.71 48.92 1091 130.0 1,073 130.0 130.0 104.72 48.93 1091 130.0 1,073 131.7 131.7 107.18 50.16 1118 131.7 1,100 135.0 135.0 11Z23 52.68 1175 , 135.0 1.156 140.0 140.0 120.54 56.84 1267 140.0 1,249 144.6 144.6 128.95 61.05 1361 144.6 1,343 151.6 151.6 143.34 68.24 1522 151.6 1.503 155.0 155.0 151.09 72.12 1608 155.0 1,590 158.9 158.9 160.66 76.90 1715 158.9 1.696 160.0 160.0 163.49 78.32 1746 160.0 1,728 165.0 165.0 17T.19 85.17 1899 165.0 1.881 Attachment to SIR 108, Rev. 2 1i Structural Integrity Associates, Inc

Table 8 Tabulated Values for Beltline Core Not Critical Curve (Curve B) for 34 EFPY Pressure-TemgeratureCurve Calculation (Heatup/Cooldow7, Core Not Criical = Curve B) nputs Ptant Pilgrim Component = Belilne Vessel thickness, t a 5.U312 Inches. so ht

• 2.352 Atnch Vessel Radius, R - 1t13.1 Inches ARTNOT -61.0 *F -===- 34 EFPY Cooldown Rate, CR 100 *Fthr Kn 6.6U kxsiinchV (Fronm Appendix G. br coddown rate above)

Am 0 1.4 F (consevtively neglect)

Safety Factor

• 2.00 (for heatupiocooidown)

M, 2.178 (From Appendix G, for inside surface axial law)

Temperature A4ustment  : 0.L *F Heigt of Water br a Full Vessel= 507.5 Inches Pressure Adjustment . 16.3 psig (hydrostatic pressure freaiM vessel at 70'F)

Hydro Test Pressure = 1,565 pslg Flange RTNDT 10.0 *F Fluid Calculated Adjusted Adjusted Temperature 114t Pressure Temperature Pressure for T Temperature Nc P for P.T Curve P-T Curve 0

(IF) ff) kincl ~IDt)(kslu'nch' ) t fps g) VF) Epsig) 0 0.0 0 a 0 70.0 -18 a 0.0 37.3C 1622 339 70.0 321 5 5.0 37.73 15.44 344 70.0 326 10 10.0 38.21 15.68 350 70.0 331 15 15.0 38.74 15.94 355 70.0 337 20 20.0 39.32 16.23 362 70.0 344 25 25.0 39.97 16.55 369 70.0 351 30 30.0 40.61 I 16.91 377 70.0 359 35 35.0 41.4e 17.30 386 70.0 367 40 40.0 42.33 I 17.74 395 70.0 377 45 45.0 43.29 2 1822 406 70.0 388 50 50.0 44.3t 1a75 418 70.0 400 I

55 55.0 45.5&2 19.33 431 70.0 413 60 60.0 46.8. 19.96 446 70.0 427 3

65 65.0 48.21 20.70 462 70.0 443 70.0 70.0 49.U 21.49 479 70.0 461 70.0 70.0 49.8 21.49 479 70.0 461 75.0 75.0 51.5' 22.37 499 75.0 480 2

80.0 80.0 53.5: 23.33 520 80.0 502 1

85.0 85.0 55.6 24.40 544 85.0 526 3 90.0 552 90.0 90.0 58.0: 25.58 570 I

95.0 95.0 60.8 28.89 600 95.0 581 100.0 100.0 63.5 28.33 632 100.0 613 105.0 105.0 66.7 29.93 667 105.0 649 3

110.0 110.0 70.2 9 31.69 707 110.0 688 115A0 115.0 74.1 33.63 750 115.0 732 120.0 120.0 76.4 35.79 798 120.0 760 125.0 125.0 83.1 38.17 851 125.0 833 18 130.0 891 130.0 130.0 884M 40.79 910 I3 891 130.0 130.0 88.4 40.80 910 130.0 30 131.7 131.7 90.3 41.75 931 131.7 913 135.0 135.0 94.2 4370 974 135.0 956 140.0 140.0 100.4 46.91 1046 140.0 1.028 67 144.6 144.6 107.1 50.16 1118 144.6 1.100 16 151.6 151.6 118.: 55.72 1242 151.6 1,224

'5 155.0 155.0 124.: 5871 1309 155.0 1291 15B.9 1U.9 131.115 6241 1391 158.9 1,373 160.0 160.0 133.1 63.50 1416 160.0 1.398 D8 1534 165.0 1,516 165.0 165.0 144.. 68.80 170.0 170.0 156. 74.65 1664 170.0 1,646 175.0 175.0 169.1 81.11 1809 175.0 1,790 Attachment to SIR-00-108, Rev. 2 12 5 StructuralIntegrity Associates, Inc.

Table 9 Tabulated Values for Beltline Core Not Critical Curve (Curve B) for 44 EFPY Pressure-Temperature Curve Calculation (HeatuplCooldowi, Core Not Critical = Curve B) fnUDsW Planit- Plgrim Ccmponent - *1ttllne Vessel thickness. I - 5.5312 inches, soA - 2.352 'Anch Vessel Radius, R = 113M inches ART~or

• U.0 F =..= 44 EFPY CoddownRate CR*= 100. -F/r 1

• 6.6U18 ksl hnch ' (From Appendix G. fr cooldown rate abow)

Tua

• 6.0 *F(consrvatsiely neglect)

Saety Factor 2.A (for heatup/cooddown)

- 2.178 (From Appendix G for inside surface axial law)

Temperature A4ustment 0 .: F Height ofWaterfr a FuiI Vessel 5 07.5 Inches Pressure A4usm ergt 16.3 psig (hydrostatic pressure for a fui esse at 70rF)

Hydro Test Pressure

• 1165t psig Flange iRTwer - f

10. F Fluid Calculated Adjused Adjusted Temperature IMt Preaure Temperature Preesure for T Temperature Kc K:p P for P-T Curve P-T Curve iF) rF) tkilncl I111) Di~ranch112) fpsig) MP J1d9e 0 0.0 0 0 0 70.0 -18 0 0.0 36.77 14.96 333 70.0 315 5 50 37.14 15.14 338 70.0 319 10 10.0 37.56 15.35 342 70.0 324 15 15.0 38.02 15.58 347 70.0 329 20 20.0 38.52 15.83 353 70.0 335 25 25.0 39.08 16.11 359 70.0 341 30 30.0 39.70 16.42 36e 70.0 348 35 35.0 40.38 16.76 374 70.0 355 40 40.0 41.14 17.14 382 70.0 364 45 45.0 41.97 17.56 391 70.0 373 50 50.0 42.90 18.02 402 70.0 383 55 55.0 43.92 18.53 413 70.0 395 60 60.0 45.04 19.09 426 70.0 407 65 650 46.29 19.72 440 70.0 421 70.0 70.0 47.67 20.40 455 70.0 437 70.0 70.0 47.67 20.40 455 70.0 437 75.0 75.0 49.19 21.16 472 75.0 454 60.0 so.o 50.87 22.01 491 80.0 472 65.0 65.0 52.73 22.93 511 85.0 493 90.0 90.0 64.78 23.9s 534 90.0 516 95.0 95.0 57.05 25.10 560 95.0 541 100.0 100.0 59.56 26.35 588 100.0 569 105.0 105.0 62.33 27.74 61g 105.0 600 110.0 110.0 65.39 29.27 653 110.0 634 115.0 115.0 68.78 30.96 690 115.0 672 120.0 120.0 72.52 32.83 732 120.0 714 125.0 125.0 76.86 34.90 778 125.0 760 130.0 130.0 61.23 37.19 629 130.0 811 130.0 130.0 61.24 37.19 829 130.0 811 131.7 131.7 82.69 38.02 848 131.7 829 135.0 135.0 86.28 39.71 885 135.0 867 140.0 140.0 91.86 42.50 946 140.0 929 144.6 144.6 91.51 4533 1011 144.6 992 151.6 151.6 107.18 50.16 1118 151.6 1,100 155.0 155.0 112.36 52.76 1176 155.0 1.158 158.9 158.9 118.81 55.98 1248 158.9 1.230 160.0 160.0 120.71 56.93 1269 160.0 1,251 185.0 165.0 129.92 61.53 1372 165.0 1.354 170.0 170.0 140.09 66.62 1485 170.0 1.467 175.0 175.0 151.33 72.24 1611 175.0 1.592 180.0 180.0 163.75 78.45 1749 160.0 1,731 185.0 165.0 177.46 85.31 1902 185.0 1.58 Attachment to SIR-00-108, Rev. 2 13 V Structural IntegrityAssociates, Inc.

Table 10 Tabulated Values for Beltline Core Not Critical Curve (Curve B) for 54 EFPY Pressure-Temperature Curve Calculation HeatupfColdow, Core Not Critical = Curve B)

__ts: Plant - Plgrimn Component

• Bo wine Vessel thickness 1=- .5312 inches. so -A 2.352 Anch Vessel Radlus, R* 113.91 Inches ARTNOT U 15.3 *F .=.=.> 54 EFPY Cooldown Rate. CR 100 :F~hr K., = 6.56 klislncht0 (From Appendix G. for cockdown rate above)

AT" i 6. F (conserntlely neglect)

Safety Factor

• 2.00 (for heatup/cooldown)

M. 2.178 (From Appendix G. for Inside surface axis law)

Temperature Aqustment - e0.0 IF Heighl of Water for a FdLlVessel

• 507.5 Inches Pressure Adjustment

• 1-6.3 psig (hydrostatic pressure for a full vessel at 70F)

Hydro Test Pressure - 1.585 psig Flange RTKNT- 10., *F Fluid Calculated Adjusted Adjused Temperature 114t Presure Temperature Presmure for T Temperature K, 0 p for P-T Curve P-T Curve (F) fF) (kitlnch"") (ksl lnch*On) (pstg)

(Plo (IF (palg) 0.0 0 0 0 70.0 -18 00 0.0 36.28 14.71 328 70.0 310 5 5.0 36.61 14.87 332 70.0 313 10 10.0 36.97 15.05 336 70.0 317 15 15.0 37.36 15.25 340 70.0 322 20 20.0 37.80 15.47 345 70.0 327 25 25.0 38.28 15.71 350 70.0 332 30 30.0 38.82 15.98 356 70.0 338 35 35.0 39.41 16.28 363 70.0 345 40 40.0 40.06 16.60 370 70.0 352 45 45.0 40.78 16.96 378 70.0 360 50 50.0 41.58 17.36 387 70.0 369 55 55.0 42.U4 17.80 397 70.0 379 60 0o.0 43.43 18.29 408 70.0 389 65 65.0 44.51 18.83 420 70.0 401 70.0 70.0 45.70 19.42 433 70.0 415 70.0 70.0 45.70 19.42 433 70.0 415 750 75.0 47.02 20.08 448 75.0 429 80.0 80.0 48.47 20.81 464 e0.0 446 65.0 85.0 50.07 21.61 482 85.0 463 90.0 90.0 51.85 22.50 502 90.0 483 95.0 95.0 53.81 2348 523 95.0 505 100.0 100.0 55.98 24.56 548 100.0 529 105.0 105.0 58.37 25.76 574 105.0 556 110.0 110.0 61.02 27.08 804 110.0 586 115.0 115.0 63.95 28.54 636 115.0 818 120.0 120.0 87.18 30.16 872 120.0 654 125.0 125.0 70.75 31.95 712 125.0 694 130.0 130.0 74.70 33.92 756 130.0 738 130.0 130.0 74.71 3393 756 130.0 73 131.7 131.7 76.14 34.64 772 131.7 754 135.0 135.0 79.07 36.11 805 135.0 787 140.0 140.0 83.89 38.52 859 140.0 840 144.6 144.8 88.78 40.96 913 144.6 895 151.6 151.6 97.13 45.14 1008 151.6 988 155.0 155.0 101.63 47.39 1057 1565.0 1,038 158.9 158.9 107.18 50.16 1118 158.9 1.100 160.0 160.0 108.82 50.98 1137 160.0 1,118 165.0 165.0 116,78 54.96 1225 165.0 1.207 170.0 170.0 125.57 59.36 1323 170.0 1.305 175.0 175.0 135.28 64.21 1432 175.0 1,413 180.0 180.0 146.02 69.58 1551 180.0 1,533 185.0 185.0 157.88 75.51 1684 185.0 1.665 190.0 190.0 171.00 82.07 1830 190.0 t.812 Attachment to SIR-00-108, Rev. 2 14 V Sfrucfural Integrity Associates, Inc.

Table 11 Tabulated Values for Feedwater Nozzle/Upper Vessel Region Core Not Critical Curve (Curve B)

Pressure-Temperature Curve Calculation (Heatup/Cooldoun, Core Not Critical = Curve B) tius Plant=: Pilgrlm-Component - Upper Vessel (based on FW nozzle)

Nozzle Whickness, t = 7.0 Inches, so 41- 2.646 *Jinch ARTNOT = 10. F= All EFPYs Consenoativly use flange.

K.= 120.9 ksiinch'la br a pressure of 1,565 psig KIT 46.2 kslinch"2 (Includes TPO effects)

Safety Factor = 2.00 (for heatuplcooldown)

Temperature Adjustment = 0.o F Pressure Adjustment = 0.0 psig Unit Pressure = 1,565 psig Flange RT"oT= 10.0 -F Fluid Calculated Adjusted Adjusted Temperature 114t Pressure Temperature Pressure for T Temperature 4 K4N P for P-T Curve P-T Curve ED} CF) (ksdlnchl:2) (ks'lnchlj2) (psig) (IF) WpSWg 0 0.0 0 0 0 70.0 0 0 0.0 50.18 1.99 26 70.0 26 5 5.0 51.96 2.U8 37 70.0 37 10 10.0 53.93 3.87 50 70.0 50 15 15.0 56.11 4.96 64 70.0 64 20 20.0 58.52 6.16 80 70.0 80 25 25.0 61.19 7.49 97 70.0 97 30 30.0 64.13 8.97 116 70.0 116 35 35.0 67.38 10.59 137 70.0 137 40 40.0 70.98 12.39 160 70.0 160 45 45.0 74.95 14.38 186 70.0 186 50 50.0 79.34 16.57 215 70.0 215 55 55.0 84.20 19.00 246 70.0 246 60 60.0 89.56 21.68 281 70.0 281 65 65.0 95.49 24.64 319 70.0 313 70.0 70.0 102.04 27.92 361 70.0 313 70.0 70.0 102.04 27.92 361 70.0 313 75.0 75.0 109.28 31.54 408 130.0 313 80.0 80.0 117.28 35.54 460 130.0 313 85.0 85.0 126.12 39.96 517 130.0 313 90.0 90.0 135.90 44.85 581 130.0 313 95.0 95.0 146.70 50.25 650 130.0 313 100.0 100.0 158.63 56.22 728 130.0 313 105.0 105.0 171.83 62.81 813 130.0 313 110.0 110.0 186.40 70.10 907 130.0 313 115.0 115.0 202.52 78.16 1012 130.0 313 120.0 120.0 220.32 87.06 1127 130.0 313 125.0 125.0 240.00 96.90 1254 130.0 313 130.0 130.0 261.75 107.78 1395 130.0 313 130.0 130.0 261.80 107.80 1395 130.0 1,395 131.7 131.7 269.66 111.73 1446 131.7 1.446 135.0 135.0 285.79 119.80 1551 135.0 1,551 140.0 140.0 312.36 133.08 1723 140.0 1,723 144.6 144.6 339.26 146.53 1897 144.6 1,897 Attachment to SIR 108, Rev. 2 15 V Structural integrity Associates, Inc.

Table 12 Tabulated Values for Bottom Head Core Not Critical Curve (Curve B)

Pressure-Temnerature Curve Calculation (Heatup/Cooldoiwi, Core Not Critical = Curve B)

Inputs Plant = Pilgrim'.

Component = Bottom Head (Penetrations Portion)

Vessel thickness, t = 7.250 Inches, so 4t = 2.693 4inch ARTNOT = 29.0: ' OF - > All EFPYs K=- 39.6 ksi*inchlr2 (for a pressure of 1,180 psig)

KIT = '2392 ksiPnch1t 2 (includes TPO effects)

Safety Factor - 2.00 (for heatuplcooldown)

Temperature Adjustment = 0.0 OF (neglect = conservative)

Height of Water for a Full Vessel 0.0 inches (FEM stresses include weight of water)

Pressure Adjustment = 0.0 :psig Unit Pressure = 1,180 psig Flange RTNOT = 10.0 *F Fluid Calculated Adjusted Adjusted Temperature 114t Pressure Temperature Pressure for T Temperature Kc NP P for P-T Curve P-T Curve r*F) 0F) -

(ksi'1nch 11 2 ) (Itsi'linch"1) (psig) -

7.F) (psig) -

0 0.0 0 0 0 70.0 0 0 0.0 44.81 10.44 311 70.0 311 5 5.0 46.03 11.05 329 70.0 329 10 10.0 47.38 11.73 350 70.0 350 15 15.0 48.87 12.48 372 70.0 372 20 20.0 50.52 13.30 396 70.0 396 25 25.0 52.34 14.21 423 70.0 423 30 30.0 54.35 15.22 453 70.0 453 35 35.0 56.58 16.33 487 70.0 487 40 40.0 59.04 17.56 523 70.0 523 45 45.0 61.75 18.92 564 70.0 564 50 50.0 64.76 20.42 608 70.0 608 55 55.0 68.08 22.08 658 70.0 658 60 60.0 71.74 23.91 713 70.0 713 65 65.0 75.80 25.94 773 70.0 773 70.0 70.0 80.28 28.18 840 70.0 840 70.0 70.0 80.28 28.18 840 70.0 840 75.0 75.0 85.23 30.65 913 75.0 913 80.0 80.0 90.70 33.39 995 80.0 995 85.0 85.0 96.75 36.41 1085 85.0 1.085 90.0 90.0 103.43 39.75 1185 90.0 1,185 95.0 95.0 110.82 43.45 1295 95.0 1,295 100.0 100.0 118.98 47.53 1416 100.0 1,416 105.0 105.0 128.00 52.04 1551 105.0 1,551 110.0 110.0 137.97 57.03 1699 110.0 1,699 115.0 115.0 148.99 62.53 1863 115.0 1,863 Attachment to SlR 108, Rev. 2 16 V Structural IntegrityAssociates, Inc.

Table 13 Tabulated Values for Core Critical Curve (Curve C) for 24 EFPY Pressure-Temnerature Curve Calculation (Core Crifical = Curve C) nput Plant = Pilgrim EFPY r 24 Cure A Leak Test Temperature = 131.7 *F(at 1.100 psig)

Hydro Test Pressure = 1,565 psig Flange RTNDT = 1.0 .-F Curve B Curve B Curve B Curve B Curve B Curve B Curve B Curve B Temperature Pressure for Temperature Pressure for Temperature Pressure for Minimum Minimum Curve C Curve C Beitline Beltine Bottom Head Bottom Head UpperVessel Upper Vessel Temperature Pressure Temperature Pressure (F) (psg) (IF) (psg) ('F) (psg) (OF) (psig) (IF) (psig) 0 -18 0 0 0 0 0.0 -18 70.0 -18 0.0 335 0.0 311 0.0 26 0.0 26 70.0 26 5.0 341 5.0 329 5.0 37 5.0 37 70.0 37 10.0 348 10.0 350 10.0 50 10.0 50 70.0 50 15.0 355 15.0 372 15.0 64 15.0 64 70.0 64 20.0 364 20.0 396 20.0 80 20.0 8o 70.0 8o 25.0 373 25.0 423 25.0 97 25.0 97 70.0 97 30.0 383 30.0 453 30.0 116 30.0 116 70.0 116 35.0 395 35.0 487 35.0 137 35.0 137 75.0 137 40.0 407 40.0 523 40.0 160 40.0 160 80.0 160 45.0 421 45.0 564 45.0 186 45.0 186 85.0 186 50.0 436 50.0 608 50.0 215 50.0 215 90.0 215 55.0 453 55.0 658 55.0 246 55.0 246 95.0 246 60.0 472 60.0 713 60.0 281 60.0 281 100.0 281 65.0 493 65.0 773 65.0 313 65.0 313 105.0 313 70.0 515 70.0 840 70.0 313 70.0 313 110.0 313 70.0 615 70.0 840 70.0 313 70.0 313 110.0 313 75.0 541 75.0 913 75.0 313 75.0 313 115.0 313 80.0 569 80.0 995 80.0 313 80.0 313 120.0 313 85.0 599 85.0 1,085 85.0 313 85.0 313 125.0 313 90.0 634 90.0 1,185 90.0 313 90.0 313 130.0 313 95.0 671 95.0 1,295 95.0 313 95.0 313 135.0 313 100.0 712.8 100.0 1416 100.0 313 100.0 313 140.0 313 105.0 759 105.0 1551 105.0 313 105.0 313 145.0 313 110.0 810 110.0 1699 110.0 313 110.0 313 150.0 313 115.0 866 115.0 1863 115.0 313 115.0 313 155.0 313 120.0 928 120.0 313 120.0 313 160.0 313 125.0 997 125.0 313 125.0 313 165.0 313 130.0 1073 130.0 313 130.0 313 170.0 313 130.0 1073 130.0 1,395 130.0 1,073 170.0 1,073 131.7 1100 131.7 1,446 131.7 1.100 171.7 1,100 135.0 1156 135.0 1.551 135.0 1.156 175.0 1.156 140.0 1249 140.0 1,723 140.0 1.249 180.0 1.249 144.6 1343 144.6 1.897 144.6 1,343 184.6 1,343 151.6 1503 151.6 1.503 191.6 1,503 155.0 1590 155.0 1,590 195.0 1,590 158.9 1696 158.9 1,696 198.9 1,696 160.0 1728 160.0 1,728 200.0 1,728 165.0 1881 165.0 1,881 205.0 1,881 Attachment to SIR-00-108, Rev. 2 17 Structural Integrity Associates, Inc.

Table 14 Tabulated Values for Core Critical Curve (Curve C) for 34 EFPY Pressure-Temnierature Curve Calculation (Core Critical = Curve C)

Wpus Plant- Pllgrim EFPY= 34 Curve A Leak Test Temperature 144.6 *F(at 1,100 psig)

Hydro Test Pressure 1--,55 pslg Flange RTD-r U 10.0 *F Curve B Curve B Curve B Curve B Curve B Curve B Curve B Curve B Temperature Pressure for Temperature Pressure for Temperature Pressure for Minimum Minimum Curve C Curve C Be1ltlne Beitline Bottom Head Bottom Head Upper Veml Upper Vessel Temperature Pressure Temperature Pressure (IF) (psig) (OF) (psig) (F) (psig) (OF) (psg) (OF) (psig) 0.0 -18 0 0 0 0 0.0 -18 70.0 -18 0.0 321 0.0 311 0.0 26 0.0 26 70.0 26 5.0 326 5.0 329 5.0 37 5.0 37 70.0 37 10.0 331 10.0 350 10.0 50 10.0 50 70.0 50 15.0 337 15.0 372 15.0 64 15.0 64 70.0 64 20.0 344 20.0 396 20.0 80 20.0 80 70.0 80 25.0 351 25.0 423 25.0 97 25.0 97 70.0 97 30.0 359 30.0 453 30.0 116 30.0 116 70.0 116 35.0 367 35.0 487 35.0 137 35.0 137 75.0 137 40.0 377 40.0 523 40.0 160 40.0 160 80.0 160 45.0 388 45.0 564 45.0 186 45.0 186 85.0 186 50.0 400 50.0 608 50.0 215 50.0 215 90.0 215 55.0 413 55.0 658 55.0 246 55.0 246 95.0 246 60.0 427 60.0 713 60.0 281 60.0 281 100.0 281 65.0 443 65.0 773 65.0 313 65.0 313 105.0 313 70.0 461 70.0 840 70.0 313 70.0 313 110.0 313 70.0 461 70.0 840 70.0 313 70.0 313 110.0 313 75.0 480 75.0 913 75.0 313 75.0 313 115.0 313 80.0 502 80.0 995 80.0 313 80.0 313 120.0 313 85.0 526 85.0 1,085 85.0 313 85.0 313 125.0 313 90.0 552 90.0 1.185 90.0 313 90.0 313 130.0 313 95.0 581 95.0 1,295 95.0 313 95.0 313 135.0 313 100.0 613 100.0 1416 100.0 313 100.0 313 140.0 313 105.0 649 105.0 1551 105.0 313 105.0 313 145.0 313 110.0 688 110.0 1699 110.0 313. 110.0 313 150.0 313 115.0 732 115.0 1863 115.0 313 115.0 313 155.0 313 120.0 780 120.0 313 120.0 313 160.0 313 12&0 833 125.0 313 125.0 313 165.0 313 130.0 891 130.0 313 130.0 313 170.0 313 130.0 891 130.0 1.395 130.0 891 170.0 891 131.7 913 131.7 1.446 131.7 913 171.7 913 135.0 956 135.0 1,551 135.0 956 175.0 956 140.0 1,028 140.0 1,723 140.0 1,028 180.0 1,028 144.6 1.100 144.6 1,897 144.6 1,100 184.6 1.100 151.6 1,224 151.6 1.224 191.6 1.224 155.0 1,291 155.0 1,291 195.0 1,291 158.9 1,373 158.9 1,373 198.9 1,373 160.0 1,398 160.0 1,398 200.0 1.398 165.0 1,516 165.0 1.516 205.0 1,516 170.0 1.646 170.0 1.646 210.0 1,646 175.0 1,790 175.0 1,790 215.0 1.790 Attachment to SIR-OO-108, Rev. 2 18 Structural Integrity Associates, Inc.

Table 15 Tabulated Values for Core Critical Curve (Curve C) for 44 EFPY Pressure-Temnerature Curve Calculation (Core Critical =Curve C)

Inuts Pblnt = Pilgrim EFPY 44 Cure A Leak Test Temperature = 151.6 IF (at 1,100 psig)

Hydro Test Pressure :1,565 psig Flange RTNOT 10.0 F Curve B Curve B Curve B Curve B Curve 8 Curve B Curve B Curve 8 Temperature Pressure for Temperature Pressure for Temperature Pressure for Minimum Minimum Curve C Curve C Bettllne BeoMine Bottom Head Bottom Head UpperVessel UpperVessel Temperature Pressure Temperature Pressure rF) (poig) CF) (psig) (IF) (psg) (IF) (peig) rF) (psig) 0.0 -18 0 0 0 0 0.0 -18 70.0 -18 0.0 315 0.0 311 0.0 26 0.0 26 70.0 26 5.0 319 5.0 329 5.0 37 5.0 37 70.0 37 10.0 324 10.0 350 10.0 50 10.0 50 70.0 50 15.0 329 15.0 372 15.0 64 15.0 64 70.0 64 20.0 335 20.0 396 20.0 80 20.0 80 70.0 80 25.0 341 25.0 423 25.0 97 25.0 97 70.0 97 30.0 348 30.0 453 30.0 116 30.0 116 70.0 116 35.0 355 35.0 487 35.0 137 35.0 137 75.0 137 40.0 364 40.0 523 40.0 160 40.0 160 80.0 160 45.0 373 45.0 564 45.0 186 45.0 186 85.0 186 50.0 383 50.0 608 50.0 215 50.0 215 90.0 215 55.0 395 55.0 658 55.0 246 55.0 246 95.0 246 60.0 407 60.0 713 60.0 281 60.0 281 100.0 281 65.0 421 65.0 773 65.0 313 65.0 313 105.0 313 70.0 437 70.0 840 70.0 313 70.0 313 110.0 313 70.0 437 70.0 840 70.0 313 70.0 313 110.0 313 75.0 454 75.0 913 75.0 313 75.0 313 115.0 313 80.0 472 80.0 995 80.0 313 80.0 313 120.0 313 85.0 493 85.0 1.085 85.0 313 85.0 313 125.0 313 90.0 516 90.0 1.185 90.0 313 90.0 313 130.0 313 95.0 541 9S.0 1.295 95.0 313 95.0 313 135.0 313 100.0 569 100.0 1416 100.0 313 100.0 313 140.0 313 105.0 600 105.0 1551 105.0 313 105.0 313 145.0 313 110.0 634 110.0 1699 110.0 313 110.0 313 150.0 313 115.0 672 115.0 1863 115.0 313 115.0 313 155.0 313 120.0 714 120.0 313 120.0 313 160.0 313 125.0 760 125.0 313 125.0 313 165.0 313 130.0 811 130.0 313 130.0 313 170.0 313 130.0 811 130.0 1,395 130.0 811 170.0 811 131.7 829 131.7 1,446 131.7 829 171.7 829 135.0 867 135.0 1,551 135.0 867 175.0 867 140.0 929 140.0 1,723 140.0 929 180.0 929 144.6 992 144.6 1.897 144.6 992 184.6 992 151.6 1,100 151.6 1,100 191.6 1,100 155.0 1.158 155.0 1,158 195.0 1.158 158.9 1.230 158.9 1,230 198.9 1,230 160.0 1,251 160.0 1,251 200.0 1.251 165.0 1.354 165.0 1.354 205.0 1.354 170.0 1.467 170.0 1,467 210.0 1.467 175.0 1.592 175.0 1,592 215.0 1.592 180.0 1.731 180.0 1,731 220.0 1.731 185.0 1.884 185.0 1.884 225.0 1,884 Attachment to SIR-00-108, Rev. 2 19 Structural Integry Associates, Inc.

Table 16 Tabulated Values for Core Critical Curve (Curve C) for 54 EFPY Pressure-Temrnerature Curve Calculation (Core Critical = Curve C) bnuts Plant = Pllgrim EFPY= -4 Curm A Leak Test Temperatue = 158.9 *F(at 1.100 pslg)

Hydr Test Pressure = 1,585 pslg Flange RTD= 10.0 -F Curve B Curve B Curve B Curve B Curve B Curve B Curve B Curve B Temperature Pressure for Temperature Pr ssure for Temperature Pressure for Minimum Minimum Curve C Curve C Baeline Bettline Bottom Head Bottom Head UpperVessel UpperVessel Temperature Pressure Temperature Pressure

(*F) (pds) (OF (psg) rF) (psg) (OF) (psg) CF) Ipdg) 0.0 -18 0 0 0 0 0.0 -18 70.0 -18 0.0 310 0.0 311 0.0 26 0.0 26 70.0 26 5.0 313 5.0 329 5.0 37 5.0 37 70.0 37 10.0 317 10.0 350 10.0 50 10.0 50 70.0 50 15.0 322 15.0 372 15.0 64 15.0 64 70.0 64 20.0 327 20.0 396 20.0 80 20.0 80 70.0 80 25.0 332 25.0 423 25.0 97 25.0 97 70.0 97 30.0 338 30.0 453 30.0 116 30.0 116 70.0 116 35.0 345 35.0 487 35.0 137 35.0 137 75.0 137 40.0 352 40.0 523 40.0 160 40.0 160 80.0 160 45.0 360 45.0 564 45.0 186 45.0 186 85.0 186 50.0 369 50.0 608 50.0 215 50.0 215 90.0 215 55.0 379 55.0 658 55.0 246 55.0 246 95.0 246 60.0 389 60.0 713 60.0 281 60.0 281 100.0 281 65.0 401 65.0 773 65.0 313 65.0 313 105.0 313 70.0 415 70.0 840 70.0 313 70.0 313 110.0 313 70.0 415 70.0 840 70.0 313 70.0 313 110.0 313 75.0 429 75.0 913 75.0 313 75.0 313 115.0 313 80.0 446 80.0 995 80.0 313 80.0 313 120.0 313 85.0 463 85.0 1,085 85.0 313 85.0 313 125.0 313 90.0 483 90.0 1,185 90.0 313 90.0 313 130.0 313 95.0 505 95.0 1,295 95.0 313 95.0 313 135.0 313 100.0 529 100.0 1416 100.0 313 100.0 313 140.0 313 105.0 556 105.0 1551 105.0 313 105.0 313 145.0 313 110.0 586 110.0 1699 110.0 313 110.0 313 150.0 313 115.0 618 115.0 1863 115.0 313 115.0 313 155.0 313 120.0 654 120.0 313 120.0 313 160.0 313 125.0 694 125.0 313 125.0 313 165.0 313 130.0 738 130.0 313 130.0 313 170.0 313 130.0 738 130.0 1.395 130.0 738 170.0 738 131.7 754 131.7 1.446 131.7 754 171.7 754 135.0 787 135.0 1.551 135.0 787 175.0 787 140.0 840 140.0 1,723 140.0 840 180.0 840 144.6 895 144.6 1.897 144.6 895 184.6 895 151.6 988 151.6 988 191.6 988 155.0 1,038 155.0 1.038 195.0 1.038 158.9 1.100 158.9 1,100 198.9 1.100 160.0 1,118 160.0 1,118 200.0 1,118 165.0 1.207 165.0 1.207 205.0 1,207 170.0 1,305 170.0 1,305 210.0 1,305 175.0 1,413 175.0 1,413 215.0 1,413 180.0 1,533 180.0 1,533 220.0 1,533 185.0 1,665 185.0 1.665 225.0 1,665 190.0 1,812 190.0 1,812 230.0 1,812 Attachment to SIR-00-108, Rev. 2 20 Structural Integrity Associates, Inc.

Figure 1 Pressure Test P-T Curve (Curve A) for 24 EFPY 1,200 1,100 -

1,000

• i 900-c.

a-0 Z800 L 700 Co LU o 600 z 500 tz m 400 co z 300 200 100 0

0 50 100 150 200 250 MINIMUM REACTOR VESSEL METAL TEMPERATURE ('F)

PNPS Pressure Test Curve (Curve A), 24 EFPY Attachment to SI R 108, Rev. 2 21 Structural IntegrityAssociates, Inc.

Figure 2 Pressure Test P-T Curve (Curve A) for 34 EFPY 1,200 1,100 1,000 a.

a. 900 0

I 800 IL R

e 700 U) o 600 C-z 500 U) a: 300 200 100 co 0 50 100 150 200 250 MINIMUM REACTOR VESSEL METAL TEMPERATURE ff)

PNPS Pressure Test Curve (Curve A), 34 EFPY Attachment to SIR-00-108, Rev. 2 22 V Structural Integrity Associates, Inc.

Figure 3 Pressure Test P-T Curve (Curve A) for 44 EFPY 1,200 1,100 1,000

• i 900 0

3I 800 I-OR U 700 0

w o 600 to 0

n 400 0

w 300 200 100 it 0 50 100 150 200 250 MINIMUM REACTOR VESSEL METAL TEMPERATURE (F)

PNPS Pressure Test Curve (Curve A), 44 EFPY Attachment to SIR-00-108, Rev. 2 23 V Structural integrityAssociates, Inc.

Figure 4 Pressure Test P-T Curve (Curve A) for 54 EFPY 1,200 1,100 1,000 a; 900-a.

v 800 L

0' (0

700 0

= goo z 500 E 300 200 100 co 0 50 100 150 200 250 MINIMUM REACTOR VESSEL METAL TEMPERATURE (f)

PNPS Pressure Test Curve (Curve A), 54 EFPY Attachment to SIR-00-1 08, Rev. 2 24 3 StructuralIntegrityAssociates, Inc.

Figure 5 Core Not Critical Curve (Curve B) for 24 EFPY 1,200 1,100 1,000 co U 900 0z C:

700 0

600 o-U-

500 0p 0

W 400 a.

300 200 100 0

0 50 100 150 200 250 MINIMUM REACTOR VESSEL METAL TEMPERATURE (F)

PNPS HeatuplCooldown, Core Not Critical Curve (Curve B), 24 EFPY Attachmnent to SIR-00-I108, Rev. 2 25 5 Structural IntegrityAssociates, Inc.

Figure 6 Core Not Critical Curve (Curve B) for 34 EFPY 1,200 1,100 1,000 Q. 900 a.

3a I 800-IL OR W 700 o 600-Z 500 w

a.

$ 300 200 0.

100 0o 0 50 100 150 200 250 MINIMUM REACTOR VESSEL METAL TEMPERATURE (F)

PNPS HeatuplCooldown, Core Not Critical Curve (Curve B), 34 EFPY Attachment to SIR-00-108, Rev. 2 26 Structural Integrity Associates, Inc.

Figure 7 Core Not Critical Curve (Curve B) for 44 EFPY 1,200 1,100 1,000 S

0. 900 a.

0 8o0 I-to

-J 700 "I

I-600 z

SW0 t=

400 3

300 200 100 0

0 50 100 150 200 250 MINIMUM REACTOR VESSEL METAL TEMPERATURE (f)

PNPS HeatuplCooldown, Core Not Critical Curve (Curve B), 44 EFPY Attachment to SJR-00-108, Rev. 2 27 V StructiralIntegrityAssociates, Inc.

Figure 8 Core Not Critical Curve (Curve B) for 54 EFPY 1,200 1,100 1,000 3cm 900 a-

= 800 IL X 700 n

'Ix o 600 I-z Soo i

, 300 20 100 0

0 50 100 150 200 250 MINIMUM REACTOR VESSEL METAL TEMPERATURE (F)

PNPS HeatuplCooldown, Core Not Critical Curve (Curve B), 54 EFPY Attachmnent to SIR-00-108, Rev. 2 28 V Structural integrity Associates, Inc.

Figure 9 Core Critical Curve (Curve C) for 24 EFPY 1,200 .. ~ ~. . . . . . . . . . . r 1,100 (I

1,000 a 900 01 CL w 800 0

700 ul Go CD as 600 0

500 z

400 0

en w

A

a. 300

/4' 200

/

100

/

. . . . . . . ._ _ _i a . . . . . . . . . .

0 50 100 150 200 250 MINIMUM REACTOR VESSEL METAL TEMPERATURE (F)

PNPS HeatuplCooldown, Core Critical Curve (Curve C), 24 EFPY Attachment to SIR-00-108, Rev. 2 29 V Structural Integrity Associates, Inc.

Figure 10 Core Critical Curve (Curve C) for 34 EFPY 1.200 . . . . . . . . . . . .

1,100 1,000 900 co a

Or 800g 0.

f2 700 600

'500 z

Ir 400 -

mu 0o us 300

/

a-200

/

100 a -

0 50 100 150 200 250 MINIMUM REACTOR VESSEL METAL TEMPERATURE (OF)

PNPS HeatuplCooldown, Core Critical Curve (Curve C), 34 EFPY Attachment to SIR-00-108, Rev. 2 30 5 Structural Integrity Associates, Inc.

Figure 11 Core Critical Curve (Curve C) for 44 EFPY 1,200 . . . . . . . . . . . . - . .q .

1,100 1,000 U 900 a.

a 800 of 0

U-700 0

a, w

w 600 0

Uj 500 z

e" 400 Mi

- I -I~ I &I l I a) 300 200 1

/

/

100 0 Pl.. H rit w n . C o. r . C.I. ..l C ur v 0 50 100 150 200 250 MINIMUM REACTOR VESSEL METAL TEMPERATURE (F)

PNPS HeatuplCooldown, Core Critical Curve (Curve C), 44 EFPY Attachment to SIR-00-108, Rev. 2 31 V Structural Integrity Associates, Inc.

Figure 12 Core Critical Curve (Curve C) for 54 EFPY i inn 1,100 1,000

-i 900 a

0a w

800 0

0.

a- 700 0

I-600 I-

-J 0:

0 500 -

'U e-W.

400

'U W

I I II I 300  :

/

200

/

100 0 I a 50 100 150 200 250 MINIMUM REACTOR VESSEL METAL TEMPERATURE rF)

PNPS HeatupICooldown, Core Critical Curve (Curve C), 54 EFPY Attachment to SIR-00-108, Rev. 2 32 V StructuralIntegrity Associates, inc.

Appendix A P-T Curves Using Technical Specification Format Requirements Attachment to SIR-06-108, Rev. 2 A-1 Structural Integrity Associates, Inc.

Figure A-1 PILGRIM REACTOR VESSEL PRESSURE-TEMPERATURE LIMITS HYDROSTATIC AND LEAK TESTS EFFECTIVE FULL POWER YEARS (EFPY)w 24 34 44 54

,! ' I if i Z

i I A tT" " j AddItf A A, Fka I I 1,K'UU DUI mum HEAD 1,100I 1,000:

900 800 700 gCD, w

a. 600 to 500 400 300 FIGURE X.X.X 200 Pilgrim Reactor Pressure Vessel Pressure-Temperature Limits for Hydrostatic Leak Tests 100 0

140 150 160 170 180 190 200 210 220 230 240 250 260 270 50 60 70 80 90 100 110 120 130 TEMPERATURE (F)

Attachment to SIR-00-108, Rev. 2 A-2 r Structural Integrity Associates, Inc.

Figure A-2 PILGRIM REACTOR VESSEL PRESSURE-TEMPERATURE LIMITS SUBCRITICAL HEATUP AND COOLDOWN EFFECTIVE FULL POWER YEARSIMEPY) u 24 34 44 54 1,200 10TTOM 1,100 1,000 900 800 a 700

'U n 600 0

I'U W.

0.

500, 400 300

- FIGURE X.XX 200 Pilgrim Reactor Pressure Vessel Pressure-Temperature Limits ~

for Subcritical Heatup and Cooldown 100 0 .- - ~ . . I  ; .

50 60 70 80 90 100 110 120 130 140 150 160 170 180 190 200 210 220 230 240 250 260 270 TEMPERATURE ('F)

Attachment to SIR-00-108, Rev. 2A- A-3 VStructurl Integrity Associates, Inc.

Figure A-3 PILGRIM REACTOR VESSEL PRESSURE-TEMPERATURE LIMITS CRITICAL CORE OPERATION EFFECTIVEFULLPOWERYEARSIEPY)a 24 34 44 54 1,200 1,100 1,000 900:

800 -

700 8

LU on 600 500 -

400 300 -

FIGURE X.X.X 200: Pilgrim Reactor Pressure Vessel Pressure-Temperature Limits for Critical Core Operation 100 02 270 120 130 140 150 160 170 180 190 200 210 220 230 240 250 260 50 60 70 80 90 100 110 TEMPERATURE (¶F)

Attachment to SIR-00-108, Rev. 2 A-4 Stuctura ntegrityAssociates, Inc.