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

En tergy 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.

A DC

Entergy Nuclear Operations, Inc.

Pilgrim Nuclear Power Station Letter Number: 2.06.018 Page 2 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 Office of Nuclear Reactor Regulation Mail Stop: 0-8C-2 U.S. Nuclear Regulatory Commission 1 White Flint North 11555 Rockville Pike Rockville, MD 20852 U.S. Nuclear Regulatory Commission Region 1 475 Allendale Road King of Prussia, PA 19406 Senior Resident Inspector Pilgrim Nuclear Power Station Ms. Cristine McCombs, Director Mass. Emergency Management Agency 400 Worcester Road Framingham, MA 01702 Mr. Robert Walker, Director Radiation Control Program Commonwealth of Massachusetts 90 Washington Street Dorchester, MA 02121

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 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 in the 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 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 1 Mev, 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 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 of 1.14E+18 n/cm2 at weld 1-338AIC location and 1.28E+18 n/cm2 at the lower intermediate shell location, respectively. The N2 nozzle peak fluence at 54 EFPY was calculated to be 2.81 E+1 7 n/cm2. 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 in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed License Amendment (LA) does not involve a significant increase in the 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."

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 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 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.

(.

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

(

ThIs page Is intenfonadly left blank I&

X

(

/

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

=r t

-=

L 4 zd

=w~z R=4r X34~

PILGRIM REACTOR VESSEL PRESSURE-TEMPERATURE LIMITS HYDROSTATIC AND LEAK TESTS

/

.1--+

A --- I --

I -

i - " I I

- t--- -t - -I

- 1,0' t-- -I 1,200 1,100 1,000 900 800 I2; 700 w

U) a:

500 a.

400 300 200 100

.1 n -I--- I-IBOTTOM _- =

HEAD

^ C~-

---_=

zL~-I-ztj T

I 1

T t-I -I*

I :

I

*rrrr

+

,_s

'F41-: 1'

-I-4e

-I--

I.-.

-- I-I IH7-

-f-4iE -,L -ti z]..

_. I'--

7.

I I

I

-I-a 7-A+ -- I

-I --, _---I I-- t1-44 I -

I I - I I -I I

I

- I -

y 1

11 1

1 1 -

tH-1 1+ I-1_

r_

- I r=

FIGURE 3.6.1 Pilgrim Reactor Pressure Vessel Pressure-Temperature Umits for Hydrostatic Leak Tests The curve applies through Operating Cycle 16 b=,=fWH~~-7--

I n

C- _...,-.

w J:-l]=-

C 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 UMITS SUBCR~mCAL HEATUP AND COOLDOWN 12100_.-...-

=

.:=

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

L Et__

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

/

~TEMPERATURE ( F)

Amendment No. 2 8r82,04 4, 40, 197 314.6-iS

1,0 1,100 1,000 900

... oo 800 oa 2: 700 a:

i 600 Ula:

500 a.

C pv'?S F7c A

.(

PILGRIM REACTOR VESSEL PRESSURE-TEMPERAMRE UMITS CRMCAL CORE OPERAMON I

I I

400 I

300 200 100 FIGURE 3.63 Plfgrim Reactor Pressure Vessel Pressure-Temperature Umits for Critical Core Operation The curve applies through Operating Cycle 16 I

.I I

.I I

I I

I 1-50 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'm 00 E

-BHM 1,100 900 800-I I

LuI 6o0 S00

4.4--

300 200 100 0

FIGURE 3.6-1 P9IndAn RAe -

or Pressure Vesel Pressure-Temperaturenfts for HydrstIc Loek Tes Curve Appes Through 34 EFPYs

- l pm.

_~~~

4F-...

I, I

i

_Z_

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 1,000 900 owl lml HM w

0.

600 700 BOO 400 200 100 FIGURE 3.02 Pilri Reacho Pru Vsel PressureTempertu Umlts for Subcrical Hea ad Cooldown Cudve AppIles Throuh 34 EFPYS V.

0 anew_-

fl~

i i

I O t

+ -

- -a i.-.+-

r-4 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 I

8 2I44,149h

/

3/4.6-10

PILGRIM REACTOR VESSEL PRESSURE-TMPEAMRE UMIM CRITICAL CORE OPERATION

• 1 i20O 1,100 1,000 80 700 U'

e 600.

400 300 200 100 0

t'- F-+ -

-I 4 FIGURE 36-3 I -

t -- I -4 Plftfrlm I

B.No-Raetor Pressur Vona P.

se-Temerature U nfts for C~ricul Core Opeaton Ca" Applles Through 34 EFPYs L--I Y JI--47 -L I-VT-

+i I

I iI-- i I4 I I

1-4--+-4-*--F-4---I----- -s --

--,4

-Y

q -

I -

I m9 4II It 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.-1 1

C BASES:

314.6 PRIMARY SYSTEM BOUNDARY (Cont)

A.

Thermal and Pressurization Limitations (Cont)

For critical core operation when the water level Is within 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 Is RT,, + 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 in the limits for hydrostatic and leak tests and for non-critical operation, as shown on Figures 3.6-1 and 3.6-2, respectively, In order to be consistent with the limits for critical operation.

The closure region Is more 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 Is the 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 Is defined 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 Is chosen 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 ul The adjusted reference temperature shft Is based on Regulatory Guide 1.99, Resdef 2;

-0 U,

the analytical resufts of

__t E

_-M61 MC+ Ae

.on. 1,datsdJan&Rt'2.M5. regarding projected neutron fluence; A Bc as I 1fun-W1 ofGI t Smperaure and p-esro l

2 11ca6}

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: lz ee Gary L. Stevens, P. E.

H. L. Gustin, 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 512-59191 70497-554 IL Stonington, Cr San Jose, CA Slier Spdng, ED Sundse, FL 950-5994050 40697880 3014450 954-572-2902 UnwontnOH Witter, CA 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 is based 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:

'Section 7 of Reference [71 also reports fluence values for the recirculation inlet (N2) nozzles as I.30x 1017 n/cm2 for 20.7 EFPY and 2.8lxlO7 n/cm2 for 54 EFPY. Both of these values are greater than the l.Ox lol n/cm2 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]:

For Pressure 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 Integrity Associates, Inc.

For Core Not Critical Conditions (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 +

1200F.

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 + 600F. This limit has been a standard recommendation for the BWR industry for non-ductile failure protection.

For Core Critical Conditions (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 + 400F, 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

Structural lntegrity 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 %essel at 70OF)

Hydro Test Pressure =

1,565 psig Flange RTNDT =

10.0 OF Fluid Temperature T

(F) 70.0 70.0 75.0 80.0 85.0 90.0 95.0 100.0 105.0 110.0 115.0 120.0 125.0 130.0 135.0 140.0 145.0 114t Temperature 70F0 70.0 70.0 75.0 80.0 85.0 90.0 95.0 100.0 105.0 110.0 115.0 120.0 125.0 130.0 135.0 140.0 145.0 Kic (ksrinch1 2) 54.74 54.74 57.00 59.51 62.27 65.33 68.71 72.44 76.57 81.13 86.17 91.74 97.90 104.71 112.23 120.54 129.72 Kip (ksrinch1 2) 36.49 36.49 38.00 39.67 41.51 43.55 45.81 48.30 51.05 54.09 57.45 61.16 65.27 69.80 74.82 80.36 86.48 Calculated Pressure P

psls) 0 814 847 885 926 971 1021 1077 1138 1206 1281 1364 1455 1556 1668 1792 1928 Adjusted Temperature for P-T Curve

('F) 70.0 70.0 75.0 80.0 85.0 90.0 95.0 100.0 105.0 110.0 115.0 120.0 125.0 130.0 135.0 140.0 145.0 Adjusted Pressure for P-T Curve (PSg) 0 795 829 866 907 953 1,003 1,058 1,120 1,188 1,263 1,345 1,437 1,538 1,650 1,773 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 %essel at 70'F)

Hydro Test Pressure =

1,565 psig Flange RTNDT =

10.0 OF Fluid Temperature T

rF) 70.0 70.0 75.0 80.0 85.0 90.0 95.0 100.0 105.0 110.0 115.0 120.0 125.0 130.0 135.0 140.0 145.0 150.0 155.0 114t Temperature 7.F) 70.0 70.0 75.0 80.0 85.0 90.0 95.0 100.0 105.0 110.0 115.0 120.0 125.0 130.0 135.0 140.0 145.0 150.0 155.0 Kic (ksi*lnch¶1 2) 49.84 49.84 51.59 53.52 55.66 58.02 60.63 63.52 66.71 70.23 74.13 78.43 83.19 88.44 94.25 100.68 107.77 115.62 124.28 Kgp (ksi-inch1 2) 33.23 33.23 34.39 35.68 37.11 38.68 40.42 42.35 44.47 46.82 49.42 52.29 55.46 58.96 62.84 67.12 71.85 77.08 82.86 Calculated Pressure P

J sg) 0 741 767 796 827 862 901 944 992 1044 1102 1166 1237 1315 1401 1496 1602 1719 1847 Adjusted Temperature for P-T Curve (F) 70.0 70.0 75.0 80.0 85.0 90.0 95.0 100.0 105.0 110.0 115.0 120.0 125.0 130.0 135.0 140.0 145.0 150.0 155.0 Adjusted Pressure for P-T Curve 0

723 749 777 809 844 883 926 973 1,026 1,084 1,147 1,218 1,296 1.383 1,478 1,584 1,700 1,829 Attachment to Si R 1 08, Rev. 2 6

0 Structural Integrity Associates, 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 Temperature T

(F) 70.0 70.0 75.0 80.0 85.0 90.0 95.0 100.0 105.0 110.0 115.0 120.0 125.0 130.0 135.0 140.0 145.0 150.0 155.0 160.0 165.0 114t Temperature

( F) 70.0 70.0 75.0 80.0 85.0 90.0 95.0 100.0 105.0 110.0 115.0 120.0 125.0 130.0 135.0 140.0 145.0 150.0 155.0 160.0 165.0 Kc (ksl*lnchl')

47.67 47.67 49.19 50.87 52.73 54.78 57.05 59.56 62.33 65.39 68.78 72.52 76.66 81.23 86.28 91.86 98.03 104.85 112.38 120.71 129.92 Kip (ksi~inch"2) 31.78 31.78 32.79 33.91 35.15 36.52 38.03 39.71 41.55 43.60 45.85 48.35 51.10 54.15 57.52 61.24 65.35 69.90 74.92 80.47 86.61 Calculated Pressure p

(psig) 0 709 731 756 784 814 848 885 926 972 1022 1078 1139 1207 1282 1365 1457 1559 1671 1794 1931 Adjusted Temperature for P-T Curve rF) 70.0 70.0 75.0 80.0 85.0 90.0 95.0 100.0 105.0 110.0 115.0 120.0 125.0 130.0 135.0 140.0 145.0 150.0 155.0 160.0 165.0 Adjusted Pressure for P-T Curve SW) 0 690 713 738 765 796 830 867 908 954 1,004 1,060 1,121 1,189 1,264 1,347 1,439 1,540 1,652 1,776 1,913 Attachment to SIR-00-108, Rev. 2 7

h Structural Integrity 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)

AT114t =

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 Temperature T

(F) 70.0 70.0 75.0 80.0 85.0 90.0 95.0 100.0 105.0 110.0 115.0 120.0 125.0 130.0 135.0 140.0 145.0 150.0 155.0 160.0 165.0 170.0 114t Temperature (OF) 70.0 70.0 75.0 80.0 85.0 90.0 95.0 100.0 105.0 110.0 115.0 120.0 125.0 130.0 135.0 140.0 145.0 150.0 155.0 160.0 165.0 170.0 Kc (ksl*Inchl 2) 45.70 45.70 47.02 48.47 50.07 51.85 53.81 55.98 58.37 61.02 63.95 67.18 70.75 74.70 79.07 83.89 89.22 95.12 101.63 108.82 116.78 125.57 Kip (ksd~lnch"2) 30.47 30.47 31.34 32.31 33.38 34.57 35.87 37.32 38.92 40.68 42.63 44.79 47.17 49.80 52.71 55.93 59.48 63.41 67.75 72.55 77.85 83.71 Calculated Pressure P

(psig) 0 679 699 720 744 771 800 832 868 907 951 999 1052 1110 1175 1247 1326 1414 1511 1618 1736 1866 Adjusted Temperature for P-T Curve

(-F) 70.0 70.0 75.0 80.0 85.0 90.0 95.0 100.0 105.0 110.0 115.0 120.0 125.0 130.0 135.0 140.0 145.0 150.0 155.0 160.0 165.0 170.0 Adjusted Pressure for P-T Curve (PSlg) 0 661 681 702 726 752 782 814 849 889 932 980 1,033 1.092 1,157 1,229 1,308 1,396 1,492 1,599 1,717 1.848 Attachment to SIR-00-108, Rev. 2 8

A m tStructural Integrity 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 Temperature T

rF) 70.0 70.0 100.0 100.0 105.0 1 10.0 1/4t Temperature rF) 70.0 70.0 100.0 100.0 105.0 110.0 Kic (ksl~inchlrl) 102.04 102.04 158.63 158.63 171.83 186.40 K~p (ksilinch 112) 68.03 68.03 105.76 105.76 114.55 124.27 Calculated Pressure P

08d0) 0 313 313 1369 1483 1609 Adjusted Temperature for P-T Curve

('IF) 70.0 70.0 100.0 100.0 105.0 110.0 AdJusted Pressure for P-T Curve (P8i9) 313 313 1369 1483 1609 Attachment to SIR-00-108, Rev. 2 9

V Structulral Integrity Associates, 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 ksilinch12 (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 Temperature T

7.F) 70.0 70.0 75.0 114t Temperature

(`F) 70.0 70.0 75.0 Kic (kslinch1 2) 80.28 80.28 85.23 1lp (khl*inchl 2) 53.52 53.52 56.82 Calculated Pressure P

(psig) 0 1595 1693 Adjusted Temperature for P-T Curve (OF) 70 70 75 Adjusted Pressure for P-T Curve (psig) 0 1595 1693 Attachment to SIR-00-108, Rev. 2 10 V

Structural Integrity Associates, 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 5 507.5 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 Temperature 114t T

te perature K'c 00 5

10 15 20 25 30 35 40 45 50 55 60 65 70.0 70.0 75.0 80.0 85.0 00.0 95.0 100.0 105.0 110.0 115.0 120.0 125.0 130.0 130.0 131.7 135.0 140.0 144.6 151.6 155.0 158.9 160.0 165.0 0F.

0.0 5.0 10.0 15.0 20.0 25.0 30.0 35.0 40.0 45.0 50.0 55.0 60.0 65.0 70.0 70.0 75.0 80.0 85.0 90.0 95.0 100.0 105.0 110.0 115.0 120.0 125.0 130.0 130.0 131.7 135.0 140.0 144.6 151.6 155.0 158.9 160.0 165.0 ksltlct

.l Ilia)

(kai~nch'21 0

38.51 39.07 39.69 40.37 41.12 41.96 42.88 43.89 45.02 46.26 47.64 49.16 50.83 52.69 54.74 64.74 57.00 59.51 62.27 65.33 68.71 72.44 76.57 81.13 88.17 91.74 97.90 104.71 104.72 107.18 11Z23 120.54 128.95 143.34 151.09 160.66 163.49 17T.19 0

15.83 16.11 16.41 16.76 17.13 17.55 18.01 18.52 19.08 19.70 20.39 21.15 21.99 22.92 23.94 23.94 25.07 26.32 27.71 29.24 30.93 32.79 34.86 37.14 39.66 42.44 45.52 48.92 48.93 50.16 52.68 56.84 61.05 68.24 72.12 76.90 78.32 85.17 Calculated Adjusted Pressure Temperature P

for P-T Curve (psig)

(F) 0 70.0 353 70.0 359 70.0 366 70.0 374 70.0 382 70.0 391 70.0 402 70.0 413 70.0 425 70.0 439 70.0 455 70.0 472 70.0 490 70.0 511 70.0 534 70.0 534 70.0 559 750 587 80.0 618 85.0 652 90.0 690 95.0 731 100.0 777 105.0 828 110.0 884 115.0 946 120.0 1015 125.0 1091 130.0 1091 130.0 1118 131.7 1175

, 135.0 1267 140.0 1361 144.6 1522 151.6 1608 155.0 1715 158.9 1746 160.0 1899 165.0 Adjusted Pressure for P-T Curve (ps~q)

-18 335 341 348 355 364 373 383 395 407 421 436 453 472 493 515 515 541 569 599 634 671 713 759 810 866 928 997 1,073 1,073 1,100 1.156 1,249 1,343 1.503 1,590 1.696 1,728 1.881 Attachment to SIR 108, Rev. 2 1 i Structural Integrity Associates, Inc

Table 8 Tabulated Values for Beltline Core Not Critical Curve (Curve B) for 34 EFPY Pressure-Temgerature Curve 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 6.6 U kxsiinchV (Fronm Appendix G. br coddown rate above)

Am 1.4 0

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 fre aiM vessel at 70'F)

Hydro Test Pressure =

1,565 pslg Flange RTNDT 10.0

• F Fluid Temperature 114t T

Temperature Nc (IF) ff) kincl 0

0.0 0

a 0.0 37.3C 5

5.0 37.73 10 10.0 38.21 15 15.0 38.74 20 20.0 39.32 25 25.0 39.97 30 30.0 40.61 35 35.0 41.4e 40 40.0 42.33 45 45.0 43.29 50 50.0 44.3t 55 55.0 45.5&

60 60.0 46.8.

65 65.0 48.21 70.0 70.0 49.U 70.0 70.0 49.8 75.0 75.0 51.5' 80.0 80.0 53.5:

85.0 85.0 55.6 90.0 90.0 58.0:

95.0 95.0 60.8 100.0 100.0 63.5 105.0 105.0 66.7 110.0 110.0 70.2 115A0 115.0 74.1 120.0 120.0 76.4 125.0 125.0 83.1 130.0 130.0 884M 130.0 130.0 88.4 131.7 131.7 90.3 135.0 135.0 94.2 140.0 140.0 100.4 144.6 144.6 107.1 151.6 151.6 118.:

155.0 155.0 124.:

15B.9 1U.9 131.1 160.0 160.0 133.1 165.0 165.0 144..

170.0 170.0 156.

175.0 175.0 169.1

~IDt)

(kslu'nch'0 )

I I

2 I

2 3

2 1

3 I

3 9

18 I3 30 67 16

'5 15 D8 a

1622 15.44 15.68 15.94 16.23 16.55 16.91 17.30 17.74 1822 1a75 19.33 19.96 20.70 21.49 21.49 22.37 23.33 24.40 25.58 28.89 28.33 29.93 31.69 33.63 35.79 38.17 40.79 40.80 41.75 4370 46.91 50.16 55.72 5871 6241 63.50 68.80 74.65 81.11 Calculated Adjusted Pressure Temperature P

for P.T Curve fpstg)

VF) 0 70.0 339 70.0 344 70.0 350 70.0 355 70.0 362 70.0 369 70.0 377 70.0 386 70.0 395 70.0 406 70.0 418 70.0 431 70.0 446 70.0 462 70.0 479 70.0 479 70.0 499 75.0 520 80.0 544 85.0 570 90.0 600 95.0 632 100.0 667 105.0 707 110.0 750 115.0 798 120.0 851 125.0 910 130.0 910 130.0 931 131.7 974 135.0 1046 140.0 1118 144.6 1242 151.6 1309 155.0 1391 158.9 1416 160.0 1534 165.0 1664 170.0 1809 175.0 Adjusted Pressure for P-T Curve Epsig)

-18 321 326 331 337 344 351 359 367 377 388 400 413 427 443 461 461 480 502 526 552 581 613 649 688 732 760 833 891 891 913 956 1.028 1.100 1,224 1291 1,373 1.398 1,516 1,646 1,790 Attachment to SIR-00-108, Rev. 2 12 5

Structural Integrity 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 6.6U18 ksl hnch 1' (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 -

10.

f F

Fluid Temperature IMt T

Temperature Kc iF) 0 0

5 10 15 20 25 30 35 40 45 50 55 60 65 70.0 70.0 75.0 60.0 65.0 90.0 95.0 100.0 105.0 110.0 115.0 120.0 125.0 130.0 130.0 131.7 135.0 140.0 144.6 151.6 155.0 158.9 160.0 185.0 170.0 175.0 180.0 185.0 rF) 0.0 0.0 50 10.0 15.0 20.0 25.0 30.0 35.0 40.0 45.0 50.0 55.0 60.0 650 70.0 70.0 75.0 so.o 65.0 90.0 95.0 100.0 105.0 110.0 115.0 120.0 125.0 130.0 130.0 131.7 135.0 140.0 144.6 151.6 155.0 158.9 160.0 165.0 170.0 175.0 180.0 165.0 tkilncl K:p I111) Di~ranch112) 0 36.77 37.14 37.56 38.02 38.52 39.08 39.70 40.38 41.14 41.97 42.90 43.92 45.04 46.29 47.67 47.67 49.19 50.87 52.73 64.78 57.05 59.56 62.33 65.39 68.78 72.52 76.86 61.23 61.24 82.69 86.28 91.86 91.51 107.18 112.36 118.81 120.71 129.92 140.09 151.33 163.75 177.46 0

14.96 15.14 15.35 15.58 15.83 16.11 16.42 16.76 17.14 17.56 18.02 18.53 19.09 19.72 20.40 20.40 21.16 22.01 22.93 23.9s 25.10 26.35 27.74 29.27 30.96 32.83 34.90 37.19 37.19 38.02 39.71 42.50 4533 50.16 52.76 55.98 56.93 61.53 66.62 72.24 78.45 85.31 Calculated Adjused Preaure Temperature P

for P-T Curve fpsig)

MP 0

70.0 333 70.0 338 70.0 342 70.0 347 70.0 353 70.0 359 70.0 36e 70.0 374 70.0 382 70.0 391 70.0 402 70.0 413 70.0 426 70.0 440 70.0 455 70.0 455 70.0 472 75.0 491 80.0 511 85.0 534 90.0 560 95.0 588 100.0 61g 105.0 653 110.0 690 115.0 732 120.0 778 125.0 629 130.0 829 130.0 848 131.7 885 135.0 946 140.0 1011 144.6 1118 151.6 1176 155.0 1248 158.9 1269 160.0 1372 165.0 1485 170.0 1611 175.0 1749 160.0 1902 185.0 Adjusted Preesure for P-T Curve J1d9e

-18 315 319 324 329 335 341 348 355 364 373 383 395 407 421 437 437 454 472 493 516 541 569 600 634 672 714 760 811 811 829 867 929 992 1,100 1.158 1.230 1,251 1.354 1.467 1.592 1,731 1.58 Attachment to SIR-00-108, Rev. 2 13 V

Structural Integrity Associates, 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"

6.

i 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 FdLl Vessel

• 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 Temperature 114t T

Temperature K,0 Calculated Adjusted Presure Temperature lnch*O (F) 00 5

10 15 20 25 30 35 40 45 50 55 60 65 70.0 70.0 750 80.0 65.0 90.0 95.0 100.0 105.0 110.0 115.0 120.0 125.0 130.0 130.0 131.7 135.0 140.0 144.6 151.6 155.0 158.9 160.0 165.0 170.0 175.0 180.0 185.0 190.0 fF) 0.0 0.0 5.0 10.0 15.0 20.0 25.0 30.0 35.0 40.0 45.0 50.0 55.0 0o.0 65.0 70.0 70.0 75.0 80.0 85.0 90.0 95.0 100.0 105.0 110.0 115.0 120.0 125.0 130.0 130.0 131.7 135.0 140.0 144.8 151.6 155.0 158.9 160.0 165.0 170.0 175.0 180.0 185.0 190.0 (kitlnch"") (ksl 0

0 36.28 14.71 36.61 14.87 36.97 15.05 37.36 15.25 37.80 15.47 38.28 15.71 38.82 15.98 39.41 16.28 40.06 16.60 40.78 16.96 41.58 17.36 42.U4 17.80 43.43 18.29 44.51 18.83 45.70 19.42 45.70 19.42 47.02 20.08 48.47 20.81 50.07 21.61 51.85 22.50 53.81 2348 55.98 24.56 58.37 25.76 61.02 27.08 63.95 28.54 87.18 30.16 70.75 31.95 74.70 33.92 74.71 3393 76.14 34.64 79.07 36.11 83.89 38.52 88.78 40.96 97.13 45.14 101.63 47.39 107.18 50.16 108.82 50.98 116,78 54.96 125.57 59.36 135.28 64.21 146.02 69.58 157.88 75.51 171.00 82.07 p

n)

(pstg)

(Plo 0

328 332 336 340 345 350 356 363 370 378 387 397 408 420 433 433 448 464 482 502 523 548 574 804 636 872 712 756 756 772 805 859 913 1008 1057 1118 1137 1225 1323 1432 1551 1684 1830 for P-T Curve (IF 70.0 70.0 70.0 70.0 70.0 70.0 70.0 70.0 70.0 70.0 70.0 70.0 70.0 70.0 70.0 70.0 70.0 75.0 e0.0 85.0 90.0 95.0 100.0 105.0 110.0 115.0 120.0 125.0 130.0 130.0 131.7 135.0 140.0 144.6 151.6 1565.0 158.9 160.0 165.0 170.0 175.0 180.0 185.0 190.0 Adjused Presmure for P-T Curve (palg)

-18 310 313 317 322 327 332 338 345 352 360 369 379 389 401 415 415 429 446 463 483 505 529 556 586 818 654 694 738 73 754 787 840 895 988 1,038 1.100 1,118 1.207 1.305 1,413 1,533 1.665 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 Temperature T

r*F) 0 0

5 10 15 20 25 30 35 40 45 50 55 60 65 70.0 70.0 75.0 80.0 85.0 90.0 95.0 100.0 105.0 110.0 115.0 114t Temperature 0F) 0.0 0.0 5.0 10.0 15.0 20.0 25.0 30.0 35.0 40.0 45.0 50.0 55.0 60.0 65.0 70.0 70.0 75.0 80.0 85.0 90.0 95.0 100.0 105.0 110.0 115.0 Kc (ksi'1nch112) 0 44.81 46.03 47.38 48.87 50.52 52.34 54.35 56.58 59.04 61.75 64.76 68.08 71.74 75.80 80.28 80.28 85.23 90.70 96.75 103.43 110.82 118.98 128.00 137.97 148.99 NP (Itsi'linch"1) 0 10.44 11.05 11.73 12.48 13.30 14.21 15.22 16.33 17.56 18.92 20.42 22.08 23.91 25.94 28.18 28.18 30.65 33.39 36.41 39.75 43.45 47.53 52.04 57.03 62.53 Calculated Pressure P

(psig) 0 311 329 350 372 396 423 453 487 523 564 608 658 713 773 840 840 913 995 1085 1185 1295 1416 1551 1699 1863 Adjusted Temperature for P-T Curve 7.F) 70.0 70.0 70.0 70.0 70.0 70.0 70.0 70.0 70.0 70.0 70.0 70.0 70.0 70.0 70.0 70.0 70.0 75.0 80.0 85.0 90.0 95.0 100.0 105.0 110.0 115.0 Adjusted Pressure for P-T Curve (psig) 0 311 329 350 372 396 423 453 487 523 564 608 658 713 773 840 840 913 995 1.085 1,185 1,295 1,416 1,551 1,699 1,863 Attachment to SlR 108, Rev. 2 16 V

Structural Integrity Associates, 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.

Z800 a-0 L

700 Co LU o

600 z

500 tz m

400 co z

300 200 100 0

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

PNPS Pressure Test Curve (Curve A), 24 EFPY 250 Attachment to SI R 1 08, Rev. 2 21 Structural Integrity Associates, Inc.

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

a.

900 a.

I 800 0

IL R

e 700 U) o 600 C-z 500 U) a:

300 200 100 co 0

50 100 150 200 MINIMUM REACTOR VESSEL METAL TEMPERATURE ff)

PNPS Pressure Test Curve (Curve A), 34 EFPY 250 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 03 I

800 I-OR U

700 0

to w

o 600 n

400 0

0 w

300 200 100 it 0

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

PNPS Pressure Test Curve (Curve A), 44 EFPY 250 Attachment to SIR-00-108, Rev. 2 23 V

Structural integrity Associates, Inc.

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

900-v 800 a.

L 0'

700 (00 z

500

=

goo E

300 200 100 co 0

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

PNPS Pressure Test Curve (Curve A), 54 EFPY 250 Attachment to SIR-00-1 08, Rev. 2 24 3 Structural Integrity Associates, Inc.

Figure 5 Core Not Critical Curve (Curve B) for 24 EFPY 1,200 1,100 co U

z o-0 U-C:

0 0p 0W a.

1,000 900 700 600 500 400 300 200 100 0

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

PNPS HeatuplCooldown, Core Not Critical Curve (Curve B), 24 EFPY 250 Attachmnent to SIR-00-I108, Rev. 2 25 5

Structural Integrity Associates, Inc.

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

900 3

a.

a I

800-IL OR W

700 o

600-Z 500 w

$ 300 a.

200 0.

100 0o 0

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

PNPS HeatuplCooldown, Core Not Critical Curve (Curve B), 34 EFPY 250 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.

a.

0 I-to

-J "I

I-z t=

3 900 8o0 700 600 SW0 400 300 200 100 0

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

PNPS HeatuplCooldown, Core Not Critical Curve (Curve B), 44 EFPY 250 Attachment to SJR-00-108, Rev. 2 27 V

Structiral Integrity Associates, Inc.

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

900 a-=

800 IL X

700 n

'Ix o

600 I-z Soo i

, 300 20 100 0

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

PNPS HeatuplCooldown, Core Not Critical Curve (Curve B), 54 EFPY 250 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 1,100 1,000

. ~ ~.

r (I

a 01 w

CL 0

ul Go CD as 0

z en A

0 w

a.

900 800 700 600 500 400 300 200 100

/ /

/4' a

_i 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 co Ora 0.f2 z

mu Ir 0o us a-1,100 1,000 900 800g 700 600

' 500 400 -

300 200

/ /

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 1,100 1,000

.q U

a.

a of 0

U-w 0a, w

0 e"

Uj z

Mi a) 900 800 700 600 500 400 300 200 1

/

/

- I

-I~

I

& I l

I 100 0

Pl..

H

.I.

w n C o r C r i t l

C u r v 0

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

-i a0a w

a-0 I-

-J 0:

0.

0 I-e-

'U W

0

'U W.

1,100 1,000 900 800 700 600 500 400 300 I

I I

I I

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

Structural Integrity 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 I

if i

Z Fka I

I i

I A tT "

j Add Itf 34 44 54 A A, 1,K'UU 1,100I 1,000:

900 DUI mum HEAD CD, w

a.

g to 800 700 600 500 400 300 200 100 0

FIGURE X.X.X Pilgrim Reactor Pressure Vessel Pressure-Temperature for Hydrostatic Leak Tests Limits 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. 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 1,100 1,000 900 800 a

700

'U n

600 0I'U W.

500, 0.

10TTOM 400 300 200 100 FIGURE X.XX Pilgrim Reactor Pressure Vessel Pressure-Temperature Limits ~

for Subcritical Heatup and Cooldown 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:

LU 8

on 800 -

700 600 500 -

400 300 -

200:

100 02 FIGURE X.X.X Pilgrim Reactor Pressure Vessel Pressure-Temperature Limits for Critical Core Operation 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. 2 A-4 Stuctura ntegrity Associates, Inc.