LR-N04-0136, Request for Change to Technical Specifications Pressure/Temperature Limits

Request for Change to Technical Specifications Pressure/Temperature Limits

ML040990522
Person / Time
Site:	Hope Creek
Issue date:	03/31/2004
From:	Brothers M Public Service Enterprise Group
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
LCR H03-09, LR-N04-0136
Download: ML040990522 (50)

Text

PSEG Nuclear LLC

, - AP.O. Box 236, Hancocks Bridge, New Jersey 08038-0236 MAR 3 1 2004 0 PSEG NuclearIlLC LR-N04-0136 LCR H03-09 United States Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 REQUEST FOR CHANGE TO TECHNICAL SPECIFICATIONS PRESSUREITEMPERATURE LIMITS HOPE CREEK GENERATING STATION FACILITY OPERATING LICENSE NPF-57 DOCKET NO. 50-354 Pursuant to 10 CFR 50.90,-PSEG Nuclear LLC (PSEG) hereby requests a revision to the Technical Specifications for the Hope Creek Generating Station. In accordance with 10 CFR 50.91(b)(1), a copy of this submittal has been sent to the State of New Jersey.

The proposed amendment would revise the reactor pressure vessel (RPV) pressure-temperature (P-T) limits and extend the validity of the limits to 32 effective full power years (EFPY). The current P-T limits expire at the end of the current operating cycle (defined as the end of the next refueling outage). The updated P-T limits are based on an RPV neutron fluence calculated using an NRC staff-accepted neutron fluence methodology for boiling water reactors.

PSEG has evaluated the proposed changes in accordance with 10 CFR 50.91(a)(1),

using the criteria in 10 CFR 50.92(c), and has determined this request involves no significant hazards considerations. An evaluation of the requested changes is provided in Attachment 1 to this letter. The marked up Technical Specification pages affected by the proposed changes are provided in Attachment 2. Attachment 3 provides a description of inputs, methodology and results for the revised pressure-temperature curves.

0o0 95-2168 REV. 7/99

.- Document Control Desk LR-N04-0136 MA/R 3 1 2004 PSEG requests approval of the proposed License Amendment by September 30, 2004 to be implemented within 60 days of the issue date. The License Amendment is required to permit restart after completion of the Hope Creek Fall 2004 refueling outage.

Should you have any questions regarding this request, please contact Mr. Paul Duke at 856-339-1466.

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

Executed on 71 4,1'#Zgoo-/

(date) Michael H. Brothers Vice President - Site Operations Attachments (3)

C Mr. H. Miller, Administrator - Region I U. S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 U. S. Nuclear Regulatory Commission ATTN: Mr. D. Collins, Licensing Project Manager - Hope Creek Mail Stop 08C2 Washington, DC 20555-0001 USNRC Senior Resident Inspector - HC (X24)

Mr. K. Tosch, Manager IV Bureau of Nuclear Engineering P.O. Box 415 Trenton, NJ 08625 LR-N04-0136 LCR H03-09 REQUEST FOR CHANGE TO TECHNICAL SPECIFICATIONS PRESSUREITEMPERATURE LIMITS Table of Contents

1. DESCRIPTION .................................... .1

2. PROPOSED CHANGE ..................................... 1

3. BACKGROUND .................................... .1

4. TECHNICAL ANALYSIS ................................. .. 2

5. REGULATORY SAFETY ANALYSIS ................................... 5 5.1 No Significant Hazards Consideration ..................................... 5 5.2 Applicable Regulatory Requirements/Criteria ..................................... 6

6. ENVIRONMENTAL CONSIDERATION ................................... 7

7. REFERENCES ..................................... 7 LR-N04-0136 LCR H03-09

1. DESCRIPTION This letter is a request to amend Operating License NPF-57 for the Hope Creek Generating Station (HCGS). The proposed amendment would revise the reactor pressure vessel (RPV) pressure-temperature (P-T) limits in Technical Specification (TS) Figures 3.4.6.1-1, 3.4.6.1-2 and 3.4.6.1-.3 and extend the validity of the limits to 32 effective full power years (EFPY). The updated P-T limits are based on an RPV neutron fluence calculated using an NRC staff-accepted neutron fluence methodology for boiling water reactors. The revised P-T limit curves satisfy the requirements of Appendix G to 10CFR Part 50.

2. PROPOSED CHANGE The marked up pages for the proposed changes to the Technical Specifications are included in Attachment 2 of this submittal.

The current TS Figures 3.4.6.1-1, 3.4.6.1-2 and 3.4.6.1-3, which establish P-T limitations for the reactor coolant system, will be replaced by the figures in Attachment 2. The updated P-T curves are valid through the end of the 40-year operating license and are based on a re-calculation of neutron fluence using an NRC staff-accepted neutron fluence methodology for boiling water reactors (BWRs) which is consistent with the requirements of Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," (Reference 1).

Changes to the TS Bases would also be made to reflect the changes to the P-T limits. The marked up Bases pages are also included in Attachment 2 of this submittal.

PSEG plans to implement the proposed change to support the next refueling outage (i.e., Fall 2004) and subsequent restart. Because the current set of P-T curves expire at the end of the current operating cycle (defined as the end of the next refueling outage), a license amendment is required before the end of the refueling outage. The next refueling outage is currently scheduled to begin in October 2004.

3. BACKGROUND 10 CFR 50.60 requires that light-water nuclear power reactors meet the fracture toughness requirements for the reactor coolant pressure boundary set forth in Appendix G to 10 CFR 50. Appendix G is the regulatory basis for P-T curves for light water reactors. Appendix G specifies fracture toughness requirements for ferritic materials of pressure-retaining components of the reactor coolant pressure boundary to provide adequate margins of safety during any condition of LR-N04-0136 LCR H03-09 normal operation, including anticipated operational occurrences and system hydrostatic tests, to which the pressure boundary may be subjected over its service lifetime. Appendix G also requires that the reference temperature and Charpy upper-shelf energy for reactor vessel beltline materials account for the embrittlement caused by neutron fluence over the life of the vessel.

RG 1.190 describes methods and assumptions acceptable to the NRC staff for determining the pressure vessel neutron fluence. RG 1.99, 'Radiation Embrittlement of Reactor Vessel Materials" (Reference 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.

The P-T curves approved in Amendment No. 131 (Reference 3) were developed using the methodology specified in American Society of Mechanical Engineers (ASME) Code Cases N-588 and N-640, as well as the 1989 ASME Code, Section Xl, Appendix G, and 10 CFR Part 50, Appendix G. Adjusted reference temperatures at the nil ductility transition values were developed for the reactor pressure vessel materials in accordance with RG 1.99, Revision 2. In the Safety Evaluation for Amendment No. 131 the NRC staff concluded that, while there was ample margin to allow use of the revised P-T curves through Cycle 11, the revised P-T curves could not be approved for the 32 EFPY for which they were intended until the fluence values were recalculated using the guidance of RG 1.190.

InAmendment 139 (Reference 4), the NRC staff approved use of the existing P-T limit curves for one additional cycle (through the end of Cycle 12). The estimated fluence at the end of Cycle 12 is less than half of the calculated fluence for 32 EFPY upon which the current P-T limits are based.

10 CFR 50 Appendix G requires reactor vessel beltline materials to be tested in accordance with the surveillance program requirements of 10 CFR 50 Appendix H. In References 5 and 6, PSEG requested a change the HCGS reactor vessel material surveillance program to incorporate the Boiling Water Reactor Vessel and Internals Project (BWRVIP) Integrated Surveillance Program (ISP) into the HCGS licensing basis. The updated fluence calculations described in Section 4 of this Attachment meet the condition for use of compatible neutron fluence methodologies acceptable to the NRC staff.

4. TECHNICAL ANALYSIS Revised pressure-temperature (P-T) curves were developed for pressure test, core not critical, and core critical conditions. A report describing the inputs, methodology and results for the revised curves is provided in Attachment 3. The LR-N04-0136 LCR H03-09 revised curves are applicable for 32 effective full power years (EFPY). Curves applicable for 48 EFPY are included in the report for information only.

The curves were developed using the methodology specified in American Society of Mechanical Engineers (ASME) Code Cases N-588 and N-640 as well as the 1989 ASME Code, Section Xl, Appendix G, and 10 CFR Part 50, Appendix G.

ASME Code Case N-640 permits application of the lower bound static initiation fracture toughness value equation (Kic equation) as the basis for establishing the P-T curves in lieu of the lower bound crack arrest fracture toughness value equation (i.e., the KiA equation). The KIA equation is based on conditions needed to arrest a dynamically propagating crack and is the method invoked by Appendix G to Section Xl of the 1989 ASME Code. Use of the Kic equation in determining the lower bound fracture toughness in the development of the P-T operating

-limits curve is more technically correct than the use of the KIA equation because the rate of loading during a heatup or cooldown is slow and is more representative of a static condition than a dynamic condition. RG 1.147, "Inservice Inspection Code Case Acceptability, ASME Section Xl, Division 1,"

Revision 13, January 2004 (Reference 7) states that Code Case N-640 is acceptable to the NRC for application in licensees' Section Xl inservice inspection programs.

ASME Code Case N-588 permits postulation of a circumferentially oriented flaw (in lieu of an axially-oriented flaw) for the evaluation of the circumferential welds in RPV P-T limit curves. RG 1.147 states that Code Case N-588 is acceptable to the NRC for application in licensees' Section Xl inservice inspection programs.

Neutron Fluence Calculations The neutron fluence calculations were updated using the NRC-approved General Electric Nuclear Energy (GENE) methodology documented in GENE's Licensing Topical Report NEDC-32983P-A (Reference 8). The NRC-accepted proprietary methodology is fully described in NEDC-32983P-A and is not repeated herein. In general, GENE's methodology is consistent with the guidance in RG 1.190 for neutron flux calculations and is based on a two-dimensional discrete ordinates code.

The fluence is based upon operation for 32 EFPY, with 12 EFPY at 3293 MWt, the original licensed thermal power, 3 EFPY at 3339 MWt, the current licensed thermal power, and the remaining 17 EFPY at 3952 MWt which bounds operation at the current licensed thermal power through the end of the 40-year operating license. The updated fluence evaluation (at 3952 MWt) was performed using the GENE methodology. The calculated RPV flux value for 3293 MWt is based on the neutron transport calculation described in Reference 9 in which the jet pumps were not modeled. Results from the updated fluence evaluation demonstrated that the neutron transport calculation in Reference 9 provided sufficient Attachment I LR-N04-0136 LCR H03-09 conservatism to permit its use for operation at 3293 MWt. The peak flux for 3339 MWt includes a 1.4% uprate from 3293 MWt.

The peak calculated RPV inside surface fluxes are provided below:

Power Level (MWt) Flux (n/cm'-s) 3293 9.3x104 3952 1.22x10" The calculated fast neutron fluences at the end of plant life (32 EFPY) are provided below:

Parameter Fluence (n/cm')

Peak Surface 1.1x10' Peak 1/4 T 7.6x101' Limiting Beltline Material Peak Surface 5.3x101" Limiting Beltline Material Peak 1/4 T 3.7x1 0" Regulatory Guide 1.99 and Adiusted Reference Temperature Adjusted reference temperature (ARTNDT) values were recalculated in accordance with Regulatory Guide 1.99, Revision 2 based on the new calculated fluence values.

Upper shelf energy (USE) calculations were performed and confirmed that all USE values are greater than 50 ft-lb throughout RPV life as required by 10 CFR 50 Appendix G. For the limiting USE material, the USE at 32 EFPY decreased from 61 to 60 ft-lb.

Pressure-Temperature Curve Evaluation Three regions of the reactor pressure vessel (RPV) were evaluated to develop the revised P-T curves: (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.

The methodology used to generate the P-T curves in this submittal is similar to the methodology used to generate the curves approved in HC TS Amendment No. 131. In this update, however, the estimate of the RPV neutron fluence was based on a new fluence methodology that follows the guidance of Regulatory Guide 1.190.

Non-Beltline Regions Non-beltline regions are defined as the vessel locations that are remote from the active fuel and where the end-of-life (EOL) neutron fluence is not sufficient (i.e., <

LR-N04-0136 LCR H03-09 1017 n/cm2) to cause any significant embrifflement. Non-beltline components include nozzles, closure flanges, some shell plates, the top and bottom head plates, and the control rod drive penetrations.

For the feedwater nozzles, an updated ASME Code stress and fatigue analysis demonstrated acceptable fatigue life through the end of the 40-year operating license, including the effects of operation at 3952 MWt which bounds operation at the current licensed thermal power. Calculated stresses from the updated analysis were used to develop pressure-temperature limits for the upper vessel.

The methodologies used to develop the proposed P-T limit curves satisfy the requirements of the regulations (as modified by application of ASME Code Cases N-588 and N-640). The revised P-T curves and outputs from the Integrated Surveillance Program (which when approved by NRC for use at HCGS, will be used, as appropriate, for future adjustments to P-T limits) ensure that adequate RPV safety margins against non-ductile failure will continue to be maintained during normal operations, anticipated operational occurrences, and hydrostatic testing. Together, these measures ensure that the integrity of the reactor coolant system will be maintained for the life of the plant.

5. REGULATORY SAFETY ANALYSIS 5.1 No Significant Hazards Consideration PSEG Nuclear (PSEG) has evaluated whether or not a significant hazards consideration is involved with the proposed amendment 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 revised curves are based on uprated fluence projections and are applicable for the service period up to 32 effective full power years (EFPY). There are no changes being made to the reactor coolant system (RCS) pressure boundary or to RCS material, design or construction standards. The proposed heatup and cooldown curves define limits that continue to ensure the prevention of nonductile failure of the RCS pressure boundary. The design-basis events that were evaluated have not changed. The modification of the heatup and cooldown curves does not alter any assumptions previously made in the radiological consequence evaluations since the integrity of the RCS pressure boundary is LR-N04-0136 LCR H03-09 unaffected. Therefore, the proposed changes will not significantly increase the probability or consequences of an accident previously evaluated.

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

Response: No.

Revisions to the heatup and cooldown curves do not involve any new components or plant procedures. The proposed changes do not create any new single failure or cause any systems, structures or components to be operated beyond their design bases.

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

3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No.

The proposed figures define the limits for ensuring prevention of nonductile failure for the reactor coolant system based on the methods described in 1989 ASME Code Section Xl Appendix G, 10CFR 50 Appendix G, and ASME Code Cases N-640 and N-588.

The effect of the change is to permit plant operation within different pressure-temperature limits, but still with adequate margin to assure the integrity of the reactor coolant system pressure boundary. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, PSEG concludes that the proposed changes present 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 RequirementslCriteria 10 CFR 50.60 requires that light-water nuclear power reactors meet the fracture toughness requirements for the reactor coolant pressure boundary set forth in Appendix G to 10 CFR 50. Appendix G is the regulatory basis for P-T curves for light water reactors. Appendix G specifies fracture toughness requirements for ferritic materials of pressure-retaining components of the reactor coolant pressure boundary to provide adequate margins of safety during any condition of normal operation, LR-N04-0136 LCR H03-09 including anticipated operational occurrences and system hydrostatic tests, to which the pressure boundary may be subjected over its service lifetime. Appendix G also requires that the reference temperature and Charpy upper-shelf energy for reactor vessel beltline materials account for the embrittlement caused by neutron fluence over the life of the vessel.

Regulatory Guide (RG) 1.190, uCalculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," dated March 2001 contains the U.S. Nuclear Regulatory Commission (NRC) staffs guidance on how to determine neutron fluence. RG 1.99, 'Radiation Embrittlement of Reactor Vessel Materials" (Reference 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.

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.

6. ENVIRONMENTAL CONSIDERATION PSEG has determined the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or a surveillance requirement. 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 effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51 .22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), an environmental assessment of the proposed change is not required.

7. REFERENCES

1. NRC Regulatory Guide 1.190, 'Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," Revision 0. March 2001

2. NRC Regulatory Guide 1.99, "Radiation Embrittlement of Reactor Vessel Materials," Revision 2, May 1988

3. Hope Creek Generating Station, Issuance of Amendment Re: 1.4%

Increase In Licensed Power Level (TAC No. MB0644), July 30, 2001

- Attachment 1 LR-N04-0136 LCR H03-09

4. "Hope Creek Generating Station - Issuance of Amendment Re: Use of Existing Pressure-Temperature Curves through Cycle 12 (TAC No.

MB4685)," August 13, 2002.

5. LR-N02-0406, "Request for Change to Reactor Material Surveillance Program," dated December 23, 2002

6. LRN-03-0344, "Request for Additional Information on Reactor Pressure Vessel Material Surveillance Program," dated August 14, 2003

7. Regulatory Guide 1.147, "Inservice Inspection Code Case Acceptability, ASME Section Xl, Division 1," Revision 13, January 2004

8. NEDC-32983P-A, 'Licensing Topical Report, General Electric Methodology for Reactor Pressure Vessel Fast Neutron Flux Evaluations,"

Rev. 1, December 2001

9. GE-NE-523-A164-1294R1, Hope Creek 1 Generating Station, RPV Surveillance Materials Testing and Fracture Toughness Analysis,"

December 1997.

LR-N04-0136 LCR H03-09 HOPE CREEK GENERATING STATION FACILITY OPERATING LICENSE NPF-57 DOCKET NO. 50-354 REQUEST FOR LICENSE AMENDMENT TECHNICAL SPECIFICATION PAGES WITH PROPOSED CHANGES The following Technical Specifications for Facility Operating License No. NPF-57 are affected by this change request:

Technical Specification Page Index xviii Figure 3.4.6.1-1 3/4 4-23 Figure 3.4.6.1-2 3/4 4-23a Figure 3.4.6.1-3 3/4 4-23b Bases 3/4.4.6 B 3/4 4-5 Bases Table B 3/4.4.6-1 B 3/4 4-7 Bases Figure B 3/4.4.6-1 B 3/4 4-8 Bases Table B 3/4.4.6-2 B 3/4 4-9 (New Page)

B 3/4 4-10 (New Page)

BASES SECTION PAGE INSTRUMENTATION (Continued)

Remote Shutdown Monitoring Instrumentation and Controls ........................................... B 3/4 3-5 AccidontMonitoring Instrumentation ......... B 3/4 3-5 Source Rang Monio tors .......... . 3/4 3-5 T raversing In-Core ProbeSyten .. B 3/4 3-5 3/4.3.8 DELETED ................................................ B 3/4 3-7 3/4.3.9 FEEDWATER/MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION . ...... . 3/4 3-7 Figure B3/4 3-1 Reactor Vesasl Water Level ... .... B 3/4 3-8 3/4.3.10 MECHANICAL VACUUM PUMP TRIP INSTRUMENTATION.. B 3 3-9 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 RECIRCULtTION SYSTEM . .3 3/4 4-1 3/4.4.2 SAPEU4Y/RELIEF VALVES ........ .... B 3/4 4-2 3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection systemsy . . 3/4 4-3 Operational Leakage ....... ......... .; 3/4 4-3 3/4.4.4 CHEMISTRY .. .. B 3/4 4-3 3/4.4.5 SPECIFIC ACTIVITY .................. .. .. .. B 3/4 4-4 3/4.4.6 PRESSURE/TEPERAT1URZ LnaTS ......... B 3/4 4-5 Table E3/4.4.6-1 Reactor Vessel Toughness . .....

9 3/4 4-7 Figure E3/4.4.6-lFast Neutron Fluence (E>4Mev) at (1/4)T as a Function of Service life ......... 3/4 4-8

= oes'=C ~tsI 71 L It-I ,V 3/44

< eesl SS (9' - te- -ev ol HOPE CRSEER xvi 11 Amendment No. 143

Figure 3.4.6.1-1 Hydrostati Pressure and Leak Tests PressurTemperature Limits - Curve A 1,200 I1,0All ,akagI s-tgt-c syst, and h

/100 ;press Stbs p ertoa uring the I _. 's; z ,J,$ SO^ life of the ure boundar lZ _ Z LZ_ = Scompliance wl SME Code 90XI g

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HOPE CREEK 3/4 4-23a Amendment No. 139 0

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/ -1,000  ; performed whenthe reacgfrt cl_  ; '  ;

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HOPE CREEK 314 4-23b Amendment No. 139 t

REACTOR COOLANT SYSTEM BASES 3/4.4.6 PRESSURE/TEMPERATURE LIMITS All components in the reactor coolant system are designed to withstand the effects of cyclic loads due to system temperature and pressure changes.

These cyclic loads are introduced by normal'load transients, reactor trips, and startup and shutdown operations. The various categories of load cycles used for design purposes are provided in Section (3.9) of the UFSAR. During startup and shutdown, the rates of temperature and pressure changes are limited so that the maximum specified heatup and cooldown rates are consistent with the design assumptions and satisfy the stress limits for cyclic operation. Specifically the average rate of change of reactor coolant temperature during normal heatup and cooldown shall not exceed 100°F during any 1-hour period.

The operating limit curves of Figures 3.4.6.1-1, 3.4.6.1-2, and 3.4.6.1-3 are derived from the fracture toughness requirements of 10 CFR 50 Appendix G and ASME Code Section XI, Appendix G and ASME Code Cases N-588 and N-640.. The curves are based on the RTmm and stress intensity factor information for the reactor vessel components. Fracture toughness limits and the basis for compliance are more fully discussed in SAR Chapter 5, Paragraph 5.3.1.5, "Fracture Toughness."

The reactor vessel materials have been tested to determine their initial RTNOT. The results of some of these tests are shown in Table B 3/4.4.6-1. Reactor operation and resultant fast neutron, E greater than 1 MeV, irradiation will cause an increase in the RT=,. Therefore, an adjusted reference temperature, based upon the fluence, nickel content and copper content of the material in question, can be predicted using Bases Figure B 3/4.4.6-1 and the recommendations of Regulatory Guide 1.99, Rev. 2, "Radiation Embrittlement of Reactor Vessel Material". The pressure/

temperature limit curves, Figures 3.4.6.1-1, 3.4.6.1-2, and 3.4.6.1-3, includes an assumed shift in RTmT for the end of life fluence.

The actual shift in RTIS of the vessel material will be established periodically during operation by removing and evaluating, irradiated flux wires installed near the inside wall of the reactor vessel in the core area.

Since the neutron spectra at the flux wires and vessel inside radius are essentially identical, the irradiated flux wires can be used with confidence in predicting reactor vessel material transition temperature shift. The operating limit curves of Figures 3.4.6.1-1, 3.4.6.1-2, and 3.4.6.1-3 shall be adjusted, as required, on the basis of the flux wire data and recommendations-of Regulatory Guide 1.99, Rev. 2.

[OPE CREEKB 3/4 4-5 Revised bv letter dated June 10, 2003

13C C A ) F SES TABLE B 3/4.4.6-1

( ~~1 s1TAtt~nor bJ O^ oF3XJ~c' REACTOR VESS. OUGRNESS HEAT/SLAB (t'A N - o4o0J ,,,: PREDICTED EOL BELTLINE WELD SEAM 1.0.ES UPPER SHELF IMAX. EOL COMPONENT OR MAT'L TYPE HEAT/LOT C / Ni() F R , (FT-LBS) RiwT/ F)

Plate SA-533 GR B CL.1 5K3025-1 .15 .71 +19

_ _ _ _ _ 1 .jwL_

Weld Vert. seams for D53040/ -30 shells 4&5 1125-02205 0 O.t NOTE:

• These values are given only for the benefit of calculating the end-of-life (EOL) RT=.

I HEAT/SLAB HIGHEST REFERENCE NON-BELTLINE MT'L TYPE OR OR TEMPERATURE COMPONENT WELD SEAM I.D. HEAT/LOT w (°F)

Shell Ring Connected to SA 533,.GR.B, C1.1 All Heats +19 Vessel Flange Bottom Head Dome SA 533, GR.B, Cl.l All Heats +30 Bottom Head Torus SA 533, GR.B, C1.l All Heats +30 LPCI Nozzlest' SA 508, Cl. 2, All Heats -20 Top Head Torus SA 533, GR.B, C1.1 All Heats +19 Top Head Flange SA 508, C1.2 All Heats +10 Vessel Flange SA 508, Cl.2 All Heats +10 Feedwater Nozzle SA 508, Cl.2 All Heats -20 Weld Metal All RPV Welds All Heats 0' Closure Studs SA 540, GR.B, 24 All Heats Meet 45 ft-lbs & 25 mils lateral expansion at +10'F

11) The design of the Ho -Cz'Rek vessel results in these nozzles experiencing a predicted EOL fluence at M4T of the vessel thickness ofc1Bx 10" n/cm2 . .I 513 Therefore, these nozzles are predicted to have an EOL RTN= of R o0f HOPE CREEK B 3/4 4-7 Amendment No. 131

CI 3/4E 4.6- LOWER - INTERMEDIATE SHELL FAST NEUTRON FLUENCE (E>1 MeV)

'-~ AT 1/4 T AS A FUNCTION OF SERVICE LIFE*

Bases Figure B 3/4.4.6-1

• At 80% capacity factor (40 years 32 EFPY)

HOPE CREEK B 3/4 4-8 I-f a s-i rc -3 3/. 4. b tI

BASES TABLE B 3/4.4.6-2 Numeric Values for Pressure/Temperature Limits Figure 3.4.6.1-1, Curve A Bottom Head Upper Vessel Beltline Temperature Pressure Temperature Pressure Temperature Pressure (OF) (psig) (°F) (psig) (OF) (psig) 79 0 79.0 0 79.0 0 79 929 79.0 292 79.0 691 88 1040 118.0 292 88.0 743 90 1068 118.0 925 93.0 777 92 1097 123.0 996 - 98.0 814 94 1126 128.0 1074 103.0 855 96 1157 133.0 1161 108.0 900 98 1190 138.0 1257 113.0 950 100 1223 118.0 1,005 123.0 1,065 128.0 1,133 133.0 1,207 Figure 3.4.6.1-2, Curve B Bottom Head Upper Vessel Beltline Temperature Pressure Temperature Pressure Temperature Pressure

(°F) (psig) (°F) (psig) (OF) (psig) 79 0 79.0 0 79.0 0 79 606 79.0 50 79.0 416 88 690 79.0 75 88.0 455 92 732 79.0 90 93.0 480 96 778 79.0 100 98.0 508 100 827 79.0 125 103.0 538 104 881 79.8 175 108.0 572 108 939 86.6 202 113.0 610 112 1002 90.6 220 118.0 651 116 1070 96.6 250 123.0 697 120 1144 98.4 260 128.0 747 124 1224 102.6 285 133.0 803 103.7 292 138.0 864 148.0 292 143.0 932 148.0 740 148.0 1,008 148.0 745 153.0 1,091 148.0 750 158.0 1,183 151.6 830 163.0 1,284 155.8 910 159.7 990 163.3 1070 165.5 1150 167.5 1230 HOPE CREEK B 3/4 4-9 Amendment No.

BASES TABLE B 3/4.4.6-2 (continued)

Numeric Values for Pressure/Temperature Limits Figure 3.4.6.1-3, Curve C Temperature Pressure

("F) (psig) 88.0 0 88.0 50 88.0 75 88.0 90 92.0 100 103.4 125 119.8 175 126.6 202 130.6 220 136.6 250 138.4 260 142.6 285 143.7 292 188.0 292 188.0 740 188.0 745 188.0 750 191.6 830 195.8 910 199.7 990 203.3 1070 205.5 1150 207.5 1230 HOPE CREEK B 3/4 4-10 Amendment No.

LR-N04-01 36 LCR H03-09 INSERT CURVE 6 1,200 rr 11-i--

1,100 II' I I 1,000 cm

0. 900 0

'U 800 0ax u-

-J

'U 700 U, l In

'U I Co 0 600 _1 F--

C, ]

'U 500 z

I- I 400

'u -I (0

co 300 I0-0~

200 - Beitline Boltup 100 790F 1IT I I - - - Bottom Head

- - Upper Vessel 0 F I v . . I . . _ . . I 0 50 100 150 200 250 MINIMUM REACTOR VESSEL METAL TEMPERATURE (°F)

All system leakage and hydrostatic pressure tests performed during the service life of the pressure boundary in compliance with ASME Code Section Xl.

This figure is valid for 32 EFPY of operation.

LR-N04-01 36 LCR H03-09 JINSERT CURVE B 1,200 1,100 1,000 cn 900 a ' 1 1 1 1_ I I .

a.

0 w

800 IL 0

I-

-j 700 w

0 600 I--

C.,

a:

500 z

I--

w 400 n

C,

'U a:

0. 300

-Beltline 200 Boltup- - - Bottom HeadE 100 790 F I

-Upper Vessel I I I I I I I I I 0 -

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

All heatup and cooldowns that are performed when the reactor is not critical at the normal heatup and cooldown rate.

This figure is valid for 32 EFPY of operation.

LR-N04-01 36 LCR H03-09 lINSERT CURVE C 1,200 I

I 1,100 I

1,000 oi a 900

=:Bo 800 CL 0

I--

-j W 700 co W

LU o 600 a

z 3 400 0.

Co LU 300 I I

. _ Miniinur n 200 : Criticality 100 I

with

_- .Nornal Water Level 1 1I II I I I 88 0F 37X~

I I I1-0 _

-- _ I---r-t-s-t 0 50 100 150 200 250 MINIMUM REACTOR VESSEL METAL TEMPERATURE (°F)

All heatup and cooldowns that are performed when the reactor is critical at the normal heatup and cooldown rate.

This figure is valid for 32 EFPY of operation.

- Attachment 2 LR-N04-01 36 LCR H03-09 INSERT TS BASES 314.4.61 The fluence in Bases Figure B 3/4.4.6-1 was determined using methodology described in NRC-approved General Electric Nuclear Energy Licensing Topical Report NEDC-32983P-A. This methodology is consistent with the guidance in Regulatory Guide 1.190, Rev. 0, 'Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence."

LR-N04-01 36 LCR H03-09 INSERT FIGURE B 3a4.4.6-1 I-0 x

0 A

NE 0

0 0

a; 1 I z

14I 0 5 10 15 20 25 30 35 40 Service Life, Years*

- Attachment 3 LR-N04-01 36 LCR H03-09 HOPE CREEK GENERATING STATION FACILITY OPERATING LICENSE NPF-57 DOCKET NO. 50-354 Revised Pressure-Temperature Curves for Hope Creek SIR-00-136, Rev. 1

Structural Integrity Associates, Inc.

6855 S.Havana Street Suite 350 Centennial, CO 80112-3868 Phone: 303-792-0077 Fax: 303-792-2158 whww.structintcom btemplet@structintcom March 23, 2004 BPT-04-005 SIR-00-136, Rev. I Mr. Randal J. Schmidt PSEG Nuclear Hope Creek Generating Station P.O. Box 236 Hancocks Bridge, NJ 08038

Subject:

Revised Pressure-Temperature (P-T) Curves for Hope Creek

Reference:

PSEG Purchase Order No. 4500204466 dated 6/27/03

Dear Randy:

The attachment to this letter documents the revised set of pressure-temperature (P-T) curves developed for the Hope Creek Generating Station, 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 32 and 48 effective full power years (EFPY). The curves were developed in accordance with 1989 ASME Code Section XI Appendix G, U.S. IOCFR 50 Appendix G, and ASME Code Cases N-588 and N-640.

The inputs, methodology, and results for this effort are summarized in the attachment. The calculations for this work (PSEG-IOQ-301 and -302) are also attached.

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

Prepared By: I Reviewed By:

P. Templq R. Li us':r E gr Engineer Approved By:

Gary lY. Stevens, P. E.

Associate Attachments cc: PSEG-IOQ401 Austjn, TX Charlote, NC N.Stonington, CT Rockville, MD San Jose, CA Sunrise, FL Uniontown, OH 512.533-9191 704.573.1369 860.5994050 301-231-7746 408&978-200 954.572.2902 330-899.9753

• ATTACHMENT Revised P-T Curves for Hope Creck 1.0 Introduction This attachment documents the revised set of pressure-temperature (P-T) curves developed for the Hope Creelk Generating Station. This work includes a full set of updated P-T curves (i.e.,

pressure test, core not critical, and core critical conditions) for 32 and 48 effective full power years (EFPY). The curves were developed using the methodology specified in ASME Code Cases N-588 [2] and N-640 [3], as well as 1989 ASME Code Section XI Appendix G [4],

IOCFR50 Appendix G [5], and WRC-175 [6]. The improvement realized from the Code Case methodology is as much as 60'F, and is primarily obtained from using the critical fracture toughness, KIc, in accordance with Code Case N-640. This revision of the previously issued report incorporates revised RTNDT values [17], effects of a potential Extended Power Uprate (EPU) of 20% increase [18], and an updated feedwater nozzle thermal stress and fatigue evaluation [16, 19] (which supersedes the work in Reference 20). The new fluence estimates for projected EPU conditions [17] only affect beltline components which are exposed to high fluence, and therefore [1] is still the controlling document for the initial RTNDT values for non-beltline components.

2.0 RTNDT Values Adjusted reference temperature (ARTNDT) values were developed for the Hope Creek reactor pressure vessel (RPV) materials in accordance with NRC Regulatory Guide 1.99, Revision 2

[13] based on the revised fluence data contained in Reference [17]. An EXCEL spreadsheet was set up to perform the RTNDT calculations for the different EFPY levels required for this work, and is shown in Table 1 for 32 EFPY and Table 2 for 48 EFPY. The ARTNDT values calculated in Table 1 match the values from Reference [17], with only minor variation due to differences in the number of significant figures used in the calculations. To calculate the 48 EFPY ARTNDT values shown in Table 2, the fluence values from [17] were conservatively multiplied by a factor of 1.5 corresponding to a linear increase with time. The most limiting beltline material is the Intermediate Plate, Heat No. 5K3025/1.

Attachment to SIR-00-136, Rev. 1 Structural Integrity Associates, Inc.

Table 1: HopeCreek RPVMaterdaIARTNDT 32 EFPY Calculations Chemistry Chemistry Adju Iments For 1l1t Part Name& Host Initial RTo,r Factor ARTuor Margin Terms ARTw,er Material No. (F) C (wt %) NI (wt %) (T) EFPY (T) o ff) orF) (TV)

(Lower Plates) 5K(323011 -10 0.07 0.58 44 320 16 8.0 0.0 22 6C3511 -II 0.09 0.54 58 32.0 21 10.6 0.0 31 8C4511 I 0.08 0.57 51 32.0 19 0.3 0.0 38 (Lower Intermediate Plates) 51(2963M1 -10 0.07 0.58 4432.0 16 8.0 0.0 22 51(253011 19 0.08 0.58 511 32.0 19 0.3 0.0 58 5K323811 7 0.09 0.64 58 32.0 21 10.6 0.0 49 (I1ntemediateFPlates) 5(1" S302511 - .19 .. 0.15'.~ 031 113 20 28 .** 13.9 .00 75 51(26081 - 19 0.09 0.58 58 32.0 14 7.2 0.0 48 51(269811 19 0.10 0.58 85 32.0 1s 81.0 0.0 51 (LPCI Nozzle) 19488)1 -20 0.12 0.80 88 32.0 20 9.0 0.0 20 1002411 .20 0.14 0.82 105 32.0 24 12.1 0.0 28 VetcalWelds 3:

SheI3:SMAWIW13 510-0205 -40 0.09 0.54 109 32.0 27 13.5 0.0 14 She3: SAW IW13 D5304011125-02205 -30 0.051 0.811 107 32.0 28 13.2 0.0 23 GrithWelds 314:

She3J4: SMAW IW 519-01205 .49 0.01 0.53 20 32.0 5 2.5 0.0 -39 Ues 314. SMAW IWS 504-01205 -31 0.01 0.51 20 32.0 5 2.5 0.0 -21 Shells3/4. SMAWI WS 510-01205 -40 0.09 0.54 109 32.0 27 13.5 0.0 14 Shelts314: SAWJ`WS D5304011810-02205 -49 0.081 0.811 107 32.0 26 13.2 0.0 4 Shells3)4: S.AW IW D5573311810-=205 -40 0.10 0.88 128 32.0 31 15.6 0.0 22 LPINozzle Welds:

5MAW1W179 001-01205 -40 0.02 0.51 27 32.0 8 3.1 0.0 -28 SMAWIW179 519-01205 -49 0.01 0.53 20 32.0 5 2.3 0.0 -40 1MWWI79 504-01205 -31 0.01 0.51 20 32.0 5 2.3 0.0) -22 ertical Welds 4&5:

Shehs4&5: SMlAWIWI4&l5 510-01205 .40 0.09 0.54 109 32.0 40 19.9 0.0 40 Shefs4&5: SMAWIWI14&15 D530-4011125-02.205 -30 0.081 0.811 107 32.0 39 19.5 0.0 48 61th1 Welds 415:

Shels415: SMAW I W7 510-01205 -40 0.09 0.54 109 32.0 40 19.9 0.0 40 helres4)5: SAW I W D5304011125-02205 -30 0.081 0811l 107 32.0 39 19.5 00 4 WalThicknes Ince Fluenee at lolAttenuation @1141 4

Fluence ~ 11I4ltFluence Factor. F Location Full I 11t tFPY (nkme') J U. Intern') I lS4Ml5is" (Lower Plates) 6.100O 1.525 32 0 j .lOEtB15 0.894 7.63E+17 0.365 I(Lower Intermediale Plates) 32.0 tlO112418 0.694 7.83E+17 I 0.365 (Intermediate Plates) 32.0 5.30812-7 0.694 3.68E4+17 I 0.247 (LPCI Nozzle) J 320 £ 4701+17 j 0694 3 288.17 j 0 230 Table 2: Hope Creek RPV Material ARTNOT 48 EFPY Calculations Cemistry Chemistry Adjstments For 11141 ___

Part Name & Heat InItial RT,erT Factor SRTwes Margin erms ARTeny Material No. F) Cu (wt % NI (wtI)~ ) EFPY (F) o, (T) o(ff) (F (Lower Plates) 5K(323011 . -10 0.07 0.58 44 32.0 20 9.8 0.0 29 6C35I1 -11 0.09 0.54 58 32.0 26 12.9 0.0 41 6C45(1l 1 0.08 0.57 51 32.0 23 11.3 0.0 46 (Lower Intermediate Plates) 51(963/I -10 0.07 0.58 - 44 32.0 20 9.8 0.0 29 51(353011 19 0.08 0.56 SI 32.0 23 11.3 0.0 64 31<3238111 7 0.09 0.84 58 32.0 28 12.9 0.0 59 (Intermediate Plates) 51K3025) .1 . 1517< .7 . l37.30 9 51(26081 19 0.09 0.58 58 32.0 18 8.9 0.0 55 51K2898)1 19 0.10 0.58 65 32.0 20 10.0 0.0 59 (LPCI Nozzle) 19488)1 -20 0.12 0.80 88 32.0 25 12A4 0.0 30 10D024)1 -20 0.14 0.82 105 32.0 30 152 0.0 41 Vrtical Welds 3:

e3: SJAW IW13 el 510-01205 -40 0.09 0.54 109 32.0 34 18.8 0.0 27 SIel3:SAW IW13 D5304011125-022D5 -30 0.081 0.811 107 32.0 33 18.5 0.0 38 Grth Welds 314:

Slefls314:SMAWI'At 519-01205 -49 0.01 0.53 20 32.0 8 3.1 0.0 -37 hells3f4: SMAWIW8 504-01205 -31 0.01 0.51 20 32.0 6 3.1 0.0 -19 Sels314:SMAW IWS 510-01205 -40 0.09 0.54 109 32.0 34 18.8 0.0 27 Shis3J4:SAW/M W8 530401181".2205 -49 0.082 0.8111 107 32.0 33 16.5 0.0 17 Sels314:SAW IW6 D5573311810-02205 -40 0.10 0.88 126 32.0 39 19.4 0.0 38 LPINozzle Welds:

SAIW179 001.01205 -40 0.02 0.51 27 32.0 a 3.9 0.0 -24 S.AIW1179 51901205 -49 0.01 0.53 20 32.0 6 2.9 0.0 -37 1MWIW7V 504-01205 -31 0.01 0.51 20 32.0 6 2.9 0.0 -19 ertical Welds 4.8S:

Slels4&5: SJAAWIW14&15 510-01205 -40 0.09 0.54 109 32.0 48 20.0 0.0 48 hlelis4&5: WAAWIW14&15 D5304011125.02205 -30 0.081 0.811 107 32.0 48 20.0 0.0 58 61rthWelds 4/5:

Slefls415:8MAWIW7 510.1205 -40 0.0 0.54 109 32.0 48 20.0 0.0 48 Shetls415: SAWIW? 05304011125-02205 -30 0.6081 0611l 107 320 4 20 0 0.0 58

[ Wall Thickness inchest -Fkuence altD Aitenuatior, @1) Filueica Q 1)41 Fluence Factor, FFE (Lower Plates) 6.100 1.525 320 1165SE.18 0.694 1.14E+1 044 (Lowear ntrmnediale Plales) 30 1.658.8 0.9 118+8 0.4 (Intermeduate Plates)

(LPCl Nozzle) ________ ____

32.0 32.0 7.95E+17 7.051+17 0.694 0.694 5.51E+17 4.89E417 0.305 0.289 3.0 P-T Curve Methodology The P-T curve methodology is based on the requirements of References [2] through [6]. The supporting calculations for the curves are contained in References [7] and [8]. There are three regions of the reactor pressure vessel (RPV) that are evaluated: (1) the beltline region, (2) the Attachment to SIR-00-136, Rev. 1 V Structural IntegrityAssociates, Inc.

bottom head region, and (3) the feedwater nozzle/upper vessel region. These regions bound all other regions with respect to brittle fracture.

The approach used for the beltline and bottom head (all curves), and upper vessel (Curve A only) includes the following steps:

a. Assume a fluid temperature, T. The temperature of the metal at the assumed flaw tip, TI,4t (i.e., 1/4t into the vessel wall) is conservatively assumed equal to fluid temperature. The assumed temperature also must account for an instrument uncertainty of 90 F [14].

b. Calculate the allowable stress intensity factor, K1C,based on T,/4t using the relationship from Code Case N-640 [3], as follows:

KIC = 20.734 e0.02(TI4IARTN + 33.2 (eqn. from Ref. [9])

where: T1,4t = metal temperature at assumed flaw tip (fF)

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

Kic = allowable stress intensity factor (ksiNinch)

c. Calculate the thermal stress intensity factor, Krr from Code Case N-588 [2] for the beltline and bottom head regions, or from finite element results for the feedwater nozzle/upper vessel region.

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

KIp (KIc-Krr)/SF where: Klp = 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, K1p.

f. Subtract any applicable adjustments for pressure from P. The beltline and bottom head include a pressure adjustment of 20.3 psig to account for the static pressure head of a full vessel. An instrument error of 20.5 psig was also assumed [15].

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

The approach used for the upper vessel (Curves B & C) includes the following steps:

Attachment to SIR-00-136, Rev. 1 StructuralIntegrityAssociates,Inc.

a. Assume a fluid pressure, P. The assumed pressure includes an instrument uncertainty of 20.5 psig [15].

b. Calculate the thermal stress intensity factor, Krr, based on finite element stresses.

The feedwater nozzle stresses were obtained from the finite element analysis results contained in Reference [16]. The highest linearized (membrane and membrane + bending) thermal stresses for all of the design basis transients were selected to encompass all expected operating conditions.

ysy= 43.975 ksi @ 575 0F for SA-508 Cl. 2 [1l, 10]

Calculate tn"2 . The resulting Mm value is obtained from G-2214.1 [2].

Kim is calculated from the equation in Paragraph G-2214.1 [4]:

Kim = Mm* rsm Kmb is calculated from the equation in Paragraph G-2214.2 [4]:

Klb = (2/3) Mm* asb The total Krr is therefore:

KIT R*SF*(Klm+ Kjb) where: R = correction factor, calculated to consider the nonlinear effects in the plastic region based on the assumptions and recommendations of WRC Bulletin 175 [6].

= [fay, - cpm + ((total - ays) / 30)] / (atotal - Gpm)

SF = Safety Factor for KIT

= 1.3 (conservatively used based on the recommendation in WRC-175 [6])

c. Compute the allowable pressure stress intensity factor, Kjp, is as follows:

K. =F(a/r.)Vcaypt where: ri = actual inner radius of nozzle rc nozzle corner radius [7]

rn = apparent radius of nozzle = ri + 0.29r, t = nozzle corner thickness a = crack depth (inches)

= 1/4t' F(ar1) = nozzle stress factor, from Figure A5-1 of [6]

Attachment to SIR-00-136, Rev. 1 Structural Integrity Associates, Inc.

Kip = allowable pressure stress intensity factor (ksiNinch).

Upm = primary membrane stress, PR/t (primary bending stresses are conservatively treated as membrane stresses, so apb = 0)

d. Calculate the allowable stress intensity factor, Kic, using the following relationship for a heatup/cooldown P-T curve:

KIP= 1c-K 2.0 thus: KIC = 2.0Kgp + KrT

e. Calculate the temperature, T, using the relationship from Code Case N-640 [3],

as follows:

Kic = 20.734 e[0 02(TJ14 1 RTH) + 33.2 (eqn. from Ref. [9])

where: Tr/4t = metal temperature at assumed flaw tip (IF),

assumed equal to T, the temperature at the inner vessel wall ARTNDT = adjusted reference temperature for location under consideration and desired EFPY (0F)

K1c = allowable stress intensity factor (ks Winch) thus: T=50*LNFKc- 33.21 L ~20. 7 34 + ARNDT

f. The curve was generated by scaling the stresses used to determine the pressure and thermal stress intensity factors. The primary stresses were scaled based on pressure, while the secondary stresses were scaled based on temperature difference.

g. Repeat steps (a) through (f) for other pressures 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 [5]:

ForPressure Test Conditions(Curve A):

• If the pressure is greater than 20% of the pre-service hydro test pressure (312.6 psig), the temperature must be greater than ARTNDT of the limiting flange material + 90'F.

• 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 + 60'F. The instrument uncertainty of 90F was not Attachment to SIR-00-136, Rev. 1 V StructuralIntegrityAssociates, Inc.

applied since the 600 F is an additional margin above that recommended in Reference [10], and has been a standard recommendation for the BWR industry for non-ductile failure protection. Therefore, the 60'F is considered to adequately encompass instrument uncertainty.

For Core Not CriticalConditions (Curve B):

• If the pressure is greater than 20% of the pre-service hydro test pressure (312.6 psig), the temperature must be greater than RTNDT 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. The instrument uncertainty of 90 F was not applied since the 600 F is an additional margin above that recommended in Reference [10], and has been a standard recommendation for the BWR industry for non-ductile failure protection. Therefore, the 60'F is considered to adequately encompass instrument uncertainty.

For Core CriticalConditions (Curve C):

• Per the requirements of Table 1 of Reference [5], the core critical P-T limits 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 0F.

• Another requirement of Table 1 of Reference [5] (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 + 600 F) if the pressure is below 20% of the pre-service hydro test pressure.

Also per Table 1 of Reference [5], 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 + 40'F, whichever is greater.

After accounting for instrument uncertainties, 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 Curvcs Tabulated values for the P-T curves are shown in Tables 3 through 11. The resulting P-T curves are shown in Figures 1 through 5. Note that since the upper vessel (non-beltline) curve is limiting for core not critical conditions for both 32 and 48 EFPY, Curve C is identical for both EFPY levels (i.e., no fluence effects).

Attachment to SIR-00-136, Rev. 1 StructuralIntegrityAssociates, Inc.

6.0 References

1. GE-NE-523-A164-1294R1, "Hope Creek 1 Generating Station RPV Surveillance Materials Testing and Fracture Toughness Analysis," December 1997, SI File No. PSEG-1OQ-201 (PSEG VTD 323326 Rev. 1, including outstanding change 80010289, CD M-548).

2. ASME Boiler and Pressure Vessel Code, Code Case N-588, "Alternative Reference Flaw Orientation of Appendix G for Circumferential Welds in Reactor Vessels,"Section XI, Division 1, Approved December 12, 1997.

3. ASME Boiler and Pressure Vessel Code, Code Case N-640, "Alternative Reference Fracture Toughness for Development of P-T Limit Curves,"Section XI, Division 1, Approved February 26, 1999.

4. 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," 1989 Edition.

5. U. S. Code of Federal Regulations, Title 10, Part 50, Appendix G, "Fracture Toughness Requirements," 1-1-98 Edition.

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

7. Structural IntegrityAssociates CalculationNo. PSEG-lOQ-301, Revision 1, "Development of Pressure Test (Curve A) P-T Curves," 03/23/2004.

8. Structural Integrity Associates Calculation No. PSEG-1OQ-3 02, Revision 1, "Development.

of Heatup/Cooldown (Curves B & C) P-T Curves," 03/23/2004.

9. ASME Boiler and Pressure Vessel Code,Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, Nonmandatory Appendix A, "Analysis of Flaws," 1989 Edition.

10. Hitachi Stress Report No. RS-9130 Rev.4, VPF 2883-459-5, "Detail Analysis For Feed Water Nozzle," PSEG Vendor Technical Document 320118, SI File No. PSEG-1 OQ-203.

11. ASME Boiler and Pressure Vessel Code, Section m, Rules for Construction of Nuclear Power Plant Components, Appendices, 1989 Edition.

12. Deleted.

13. USNRC Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," U. S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, (Task ME 305-4), May 1988.

14. Temperature Loop Uncertainty, PSEG Nuclear Calculation SC-SCOOOI, Revision 0, S1 File No. PSEG-lOQ-103 (11/1/00 e-mail).

Attachment to SIR-00-136, Rev. 1 StructuralIntegrityAssociates, Inc.

15. Pressure Loop Uncertainty, PSEG Nuclear Calculation SC-BB-0355, SI File No. PSEG-IOQ-103 (11/1/00 e-mail).

16. Structural Integrity Associates Calculation No. HC-05Q-305, Revision 0, "Feedwater Nozzle Stress and Fatigue Analysis," 3/22/04.

17. GE-NE-000-0005-3385-01 RO,"Project Task Report PSEG Nuclear LLC Hope Creek Generating Station Extended Power Uprate Task T0301: RPV Fracture Toughness Evaluation," July 2003, SI File No. PSEG-IOQ-21 IP (PSEG VTD 326202).

18. GE-NE-26A5958R2, "Reactor Vessel Extended Power Uprate Certified Design Specification," November 25, 2004, SI File No. PSEG-lOQ-209P (PSEG VTD 326205).

19. Structural Integrity Associates Calculation No. HC-05Q-304, Revision 0, "Feedwater Nozzle Thermal and Stress Analysis," 3/22/04.

20. GE-NE-000-0006-0156-OIRO, "Project Task Report PSEG Nuclear LLC Hope Creek Generating Station Extended Power Uprate Task T0302: Reactor Vessel Integrity - Stress Evaluation," November 2003, SI File No. PSEG-IOQ-210P (PSEG VTD 326203).

Attachment to SIR-00-136, Rev.1 Structural Integrity Associates, Inc.

Table 3 Tabulated Values for Beltline Pressure Test Curve (Curve A) for 32 EFPY Pressure-Temperature Curve Calculation (Pressure Test = Curve A)

Inputs: Plant = 'Hope Creek Component = Beltline Vessel thickness, t = s6.1000 inches, so nJt = 2.470 C4inch Vessel Radius, R = 126.5 inches ARTNDT= 75 OF =====> ' 32 EFPY Cooldown Rate, CR = 0 iF/hr KrT X 0.00' ksi*inchlr2 (From N-588, for cooldown rate above)

AT14t 0.0 .F (no thermal for pressure test)

Safety Factor = 1.50 ;. (for pressure test)

M 2.287 (From N-588, for inside surface axial flaw)

Temperature Adjustment = 9.0 , OF Height of Water for a Full Vessel = 562.5- inches Pressure Adjustment = 20.3 psig (hydrostatic pressure for a full vessel at 70*F)

Pressure Adjustment 20.5 psig (Instrument Uncertainty)

Conversion factor from ksi to psi = 1000-Hydro Test Pressure = 1,563 - psig Flange RTNDT 19.0 F Gauge Calculated Adjusted Temperature Adjusted Pressure Temperature Pressure for T Temperature Kic Kip P for P-T Curve P-T Curve 2

(7F) (*F) (ksl*lnchlI ) (ksl*inch'l2) (psig) (7F) (psig) 79.0 79.0 55.80 37.20 0 79.0 0 79.0 70.0 52.07 34.72 732 79.0 691 88.0 79.0 55.80 37.20 784 88.0 743 93.0 84.0 58.17 38.78 818 93.0 777 98.0 89.0 60.80 40.53 855 98.0 814 103.0 94.0 63.70 42.47 895 103.0 855 108.0 99.0 66.91 44.61 941 108.0 900 113.0 104.0 70.45 46.97 990 113.0 950 118.0 109.0 74.37 49.58 1045 118.0 1,005 123.0 114.0 78.70 52.47 1106 123.0 1,065 128.0 119.0 83.49 55.66 1174 128.0 1,133 133.0 124.0 88.78 59.18 1248 133.0 1,207 138.0 129.0 94.62 63.08 1330 138.0 1,289 143.0 134.0 101.08 67.39 1421 143.0 1,380 148.0 139.0 10822 72.15 1521 148.0 1,480 153.0 144.0 116.11 77.41 1632 153.0 1,591 158.0 149.0 124.83 83.22 1755 158.0 1,714 163.0 154.0 134.47 89.65 1890 163.0 1,849 168.0 159.0 145.12 96.75 2040 168.0 1,999 Attachment to SIR-00-136, Rev. 1 V Structural integrityAssociates, Inc.

Table 4 Tabulated Values for Beltline Pressure Test Curve (Curve A) for 48 EFPY Pressure-Temperature Curve Calculation' (Pressure Test = Curve A)

Inputs: Plant = Hope Creek Component =' Beltline:;

Vessel thickness, t 6.1000 inches, so St = 2.470 'qinch Vessel Radius, R = 126.5 '; inches ARTNDT = 89' F F 8EFPY Cooldown Rate, CR = 0 F/hr KIT =. 0.00 ksiinch' 2 (From N-588, for cooldown rate above)

AT114t--= 0.0'

• F(no thermal for pressure test)

'Safety Factor= 1.50- (for pressure test)

Mm= 2.287,: (From N-588, for inside surface axial flaw)

Temperature Adjustment = 9.0. 'F Height of Water for a Full Vessel 562.5 : inches Pressure Adjustment = 3s -'20.3 psig (hydrostatic pressure for a full vessel at 70'F)

Pressure Adjustment = 20.5 psig (Instrument Uncertainty)

Conversion factor from ksi to psi = 1000 Hydro Test Pressure =;. 1,563 i,.'. psig Flange RTNDT='- 19.0 OF Gauge Calculated Adjusted Temperature Adjusted Pressure Temperature Pressure for T Temperature Kic Kip P for P-T Curve P-T Curve 2 2 (7F) ("F) (ksi*inchlt ) (ksl*inch' ) (psig) (7F) (Psig) 79.0 79.0 50.31 33.54 0 79.0 0 79.0 70.0 47.49 31.66 668 79.0 627 88.0 79.0 50.31 33.54 707 88.0 666 93.0 84.0 52.11 34.74 732 93.0 692 98.0 89.0 54.10 36.07 760 98.0 720 103.0 94.0 56.30 37.53 791 103.0 751 108.0 99.0 58.73 39.15 826 108.0 785 113.0 104.0 61.41 40.94 863 113.0 822 118.0 109.0 64.38 42.92 905 118.0 864 123.0 114.0 67.66 45.11 951 123.0 910 128.0 119.0 71.28 47.52 1002 128.0 961 133.0 124.0 75.29 50.19 1058 133.0 1,017 138.0 129.0 79.72' 53.14 1121 138.0 1,080 143.0 134.0 84.61 - 56.40 1189 143.0 1,148 148.0 139.0 90.01 60.01 1265 148.0 1,224 153.0 144.0 95.99 63.99 1349 153.0 1,308 158.0 149.0 102.59 68.39 1442 158.0 1,401 163.0 154.0 109.89 73.26 1545 163.0 1,504 168.0 159.0 117.96 78.64 1658 168.0 1,617 173.0 164.0 126.87 84.58 1783 173.0 1,743 178.0 169.0 136.72 91.15 1922 178.0 1,881 183.0 174.0 147.61 98.41 2075 183.0 2,034 Attachment to SIR-00-136, Rev. 1 V StructurallntegrityAssociates, Inc.

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

Pressure-TemperatureCurve Calculation (Pressure Test = Curve A)

Inputs: Plant = 5Hope Creek Component Upper Vessel (based on FW nozzle)

ARTNDT 40.0 .F ======> i. All EFPYs Vessel thickness, t 6.169 inches, so Jt = 2.484 Finch Vessel Radius, R - 126.5: inches Nozzle corner thickness, t' 9.7 inches, approximate F(a/rf) 1.44.-: nozzle stress factor Crack Depth, a= 2.425 .'inches Safety Factor= 1.50 Temperature Adjustment 9.0 F Pressure Adjustment 20.5 ' psig (Instrument Uncertainty)

Conversion factor from ksi to psi 1000 .,

Unit Pressure :1,563.::. psig Flange RTNDT =I 19.0 OF Gauge Calculated Adjusted Temperature Adjusted Pressure Temperature Pressure for T Temperature Kic Kip P for P-T Curve P-T Curve (7F) (79) (ksl*inch1 2 ) (ksi*inchlt 2) (psig) (7F) (psig) 79.0 79.0 78.43 52.29 0 79.0 0 79.0 79.0 78.43 52.29 313 79.0 292 118.0 109.0 115.62 77.08 313 118.0 292 118.0 109.0 115.62 77.08 946 118.0 925 123.0 114.0 124.28 82.86 1017 123.0 996 .

128.0 119.0 133.86 89.24 1095 128.0 1074 133.0 124.0 144.45 96.30 1182 133.0 1161 138.0 129.0 156.15 104.10 1277 138.0 1257 143.0 134.0 169.08 112.72 1383 143.0 1363 148.0 139.0 183.37 122.25 1500 148.0 1479 Attachment to SIR-00-136, Rev. I Structural Integrity Associates, Inc.

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

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

Inputs: Plant = -7Hope Creek Component = Bottom Head (Penetrations Portion)

Vessel thickness, t = - 6.100 inches, so nlt = 2.470 -linch Vessel Radius' R = 126.5 inches ARTNDT=- 30-0 F All EFPYs Safety Factor= 1.50 Safety Factor = I' 2.30 Bottom Head Penetrations Mm = 2.287. (From N-588, for inside surface axial flaw)

Temperature Adjustment = , 9.0 . F (Instrument Uncertainty)

Height of Water for a Full Vessel = 562.5 inches Pressure Adjustment =. 20.3 psig (full vessel at 70 0F)

Pressure Adjustment= 20.5 psig (Instrument Uncertainty)

Conversion factor from ksi to psi = 1000 Unit Pressure = .1,563 psig Flange RTNDT 19.0 I- -0F Gauge Calculated Adjusted Temperature Adjusted Pressure Temperature Pressure for T Temperature Kic Kip P for P-T Curve P-T Curve (7F) (OF) (ksi*inchlf2 ) (ksi*inchl 2) (psig) (79) (psig) 79.0 79.0 88.44 58.96 0 79 0 79.0 70.0 79.34 52.90 970 79 929 88.0 79.0 88.44 58.96 1081 88 1040 90.0 81.0 90.70 60.47 1109 90 1068 92.0 83.0 93.05 62.03 1137 92 1097 94.0 85.0 95.49 63.66 1167 94 1126 96.0 87.0 98.03 65.35 1198 96 1157 98.0 89.0 100.68 67.12 1231 98 1190 100.0 91.0 103.43 68.95 1264 100 1223 102.0 93.0 106.30 70.86 1299 102 1258 104.0 95.0 109.28 72.85 1336 104 1295 106.0 97.0 112.38 74.92 1374 106 1333 108.0 99.0 115.62 77.08 1413 108 1372 110.0 101.0 118.98 79.32 1454 110 1413 112.0 103.0 122.48 81.65 1497 112 1456 114.0 105.0 126.12 84.08 1542 114 1501 116.0 107.0 129.92 86.61 1588 116 1547 118.0 109.0 133.86 89.24 1636 118 1595 Attachment to SIR-00-136, Rev. I V Structural IntegrityAssociates, Inc.

Table 7 Tabulated Values for Beitline Core Not Critical Curve (Curve B) for 32 EFPY Pressure-Temperature Curve Calculation (Heatup/Cooldown, Core Not Critical = Curve B)

Inputs: Plant= Hope Creek Component = Beitline Vessel thickness, t = 6.1000 inches, so q1t = 2.470 flinch Vessel Radius, R 126.5: inches ARTNDT= -. 75 . F  : 3=EFPY 32 F' Cooldown Rate, CR =, 100;:' : °F/hr Kn  ; 8.76 ksilinch' t2 (From N-588, for cooldown rate)

MT 0.285 - (From Figure G-2214-2)

ATI4t -0.0. F = Conservatively assumed zero Safety Factor = 2.00 Mm .' 2.287- (From N-588, for inside surface axial flaw)

Temperature Adjustment 9.0' ,F (Instrument Uncertainty)

Height of Water for a Full Vessel 562.5 :':,'inches Pressure Adjustment = 20.3: psig (hydrostatic pressure for a full vessel at 700 F)

Pressure Adjustment = 20.5 psig (Instrument Uncertainty)

Conversion factor from ksi to psi =.. 1000 Hydro Test Pressure =; 1,563 psig Flange RTNDT = 19.0 : OF Gauge Calculated Adjusted Temperature Adjusted Pressure Temperature Pressure for T Temperature Kic Kip P for P-T Curve P-T Curve 2

(OF) (OF) (ksi*inchl ) (ksi'inchl'2) (Psig) (7F) (psig) 79.0 79.0 55.80 23.52 0 79.0 0 79.0 70.0 52.07 21.66 457 79.0 416 88.0 79.0 55.80 23.52 496 88.0 455 93.0 84.0 58.17 24.71 521 93.0 480 98.0 89.0 60.80 26.02 549 98.0 508 103.0 94.0 63.70 27.47 579 103.0 538 108.0 99.0 66.91 29.08 613 108.0 572 113.0 104.0 70.45 30.85 650 113.0 610 118.0 109.0 74.37 32.81 692 118.0 651 123.0 114.0 78.70 34.97 737 123.0 697 128.0 119.0 83.49 37.37 788 128.0 747 133.0 124.0 88.78 40.01 844 133.0 803 138.0 129.0 94.62 42.93 905 138.0 864 143.0 134.0 101.08 46.16 973 143.0 932 148.0 139.0 108.22 49.73 1049 148.0 1,008 153.0 144.0 116.11 53.68 1132 153.0 1,091 158.0 149.0 124.83 58.04 1224 158.0 1,183 163.0 154.0 134.47 62.86 1325 163.0 1,284 V Structural IntegrityAssociates, Inc.

Attachment to SIR-00-136, Rev. 1

Table 8 Tabulated Values for Bcitline Core Not Critical Curve (Curve B) for 48 EFPY Pressure-Temperature Curve Calculation (Heatup/Cooldown, Core Not Critical = Curve B)

Inputs: Plant = .Hope Creek Component Belti'ne Vessel thickness, t 6.1000 inches, so 'It = 2.470 'inch Vessel Radius, R 126.5 inches ARTNDT FF89 48 EFPY.,

Cooldown Rate, CR 100 °F/hr K = 876 ksilinch"' (From N-588, for cooldown rate)

MT=' 0.285 (From Figure G-2214-2)

ATi1 4 = 0.0 - °F = Conservatively assumed zero Safety Factor = . 2.00.-

Mm 2.287, (From N-588, for inside surface axial flaw)

Temperature Adjustment =: 9.0  :'OF Height of Water for a Full Vessel = '562.5 inches Pressure Adjustment = 20.3 psig (hydrostatic pressure for a full vessel at 700F)

Pressure Adjustment = 20.5 psig (Instrument Uncertainty)

Conversion factor from ksi to psi = 1000 Hydro Test Pressure = 1,563 psig Flange RTNDoTr 190 :0 OF Gauge Calculated Adjusted Temperature Adjusted Pressure Temperature Pressure for T Temperature Kic KP p for P-T Curve P-T Curve (7F) (°F) (ksi*inch )t2 (ksi*inchl' 2) (psig) (°F) (psig) 79.0 79.0 50.31 20.78 0 79.0 0 79.0 70.0 47.49 19.37 408 79.0 368 88.0 79.0 50.31 20.78 438 88.0 397 96.0 87.0 53.28 22.26 469 96.0 429 104.0 95.0 56.77 24.00 506 104.0 465 112.0 103.0 60.85 26.05 549 112.0 508 120.0 111.0 65.65 28.45 600 120.0 559 128.0 119.0 71.28 31.26 659 128.0 618 136.0 127.0 77.89 34.57 729 136.0 688 144.0 135.0 85.65 38.44 811 144.0 770 152.0 143.0 94.75 42.99 906 152.0 866 160.0 151.0 105.42 48.33 1019 160.0 978 168.0 159.0 117.96 54.60 1151 168.0 1,110 176.0 167.0 132.66 61.95 1306 176.0 1,265 184.0 175.0 149.92 70.58 1488 184.0 1,447 192.0 183.0 170.17 80.71 1702 192.0 1,661 Attachment to SIR-00-136, Rev. I X Structural IntegrityAssociates, Inc.

Table 9 Tabulated Values for Fcedwater Nozzle/Upper Vcssel Region Core Not Critical Curve (Curve B)

Pressure-Temperature Curve Calculation (Heatup/Cooldown, Core Not Critical = Curve B)

Inputs: Plant = Hope Creek Component = Upper Vessel ARTNDT = 40.0 apm = 20.61 ksi for a pressure of 1,005 psig aypb = 0.00 ksi for a pressure of 1,005 psig Gsm (onha) = 4.68 ksi for a temperature of 547°F [16]

0 sb (oigmal) = 16.29 ksi for a temperature of 547°F 116]

airS = . 44.0 ksi Mm = 2.88 F(afrn) = 1.44 Temperature, Adjustment = 9.0 °F (Instrument Uncertainty)

Pressure AAdjustment = 20.5 psig (Instrument Uncertainty)

Hydro TeF t Pressure = 1563 psig Rl ange RTNDT = 19.0 CF Adjusted Calculated Adjusted Pressure for Saturation Temperature Temperature Pressure for Calculation Temperature Krr Kip Kic for P-T Curve P-T Curve (psig) (OF) (ksiinch 12

) ksl*inch"I2 (ksi*inch 1 2 ) (°F) (°F) (psig) 0 212.1 19.7 0.0 19.7 79.0 0 70.5 316.1 31.7 5.7 43.2 3.4 79.0 50 95.5 334.7 33.8 7.8 49.4 27.6 79.0 75 110.5 344.3 34.9 9.0 52.9 37.5 79.0 90 120.5 350.2 35.6 9.8 55.2 43.0 79.0 100 145.5 363.6 37.1 11.9 60.8 54.4 79.0 125 195.5 386.1 39.7 15.9 71.6 70.8 79.8 175 222.5 396.5 40.9 18.1 77.2 77.6 86.6 202 240.5 403.0 41.6 19.6 80.9 81.6 90.6 220 270.5 413.0 42.8 22.0 86.9 87.6 96.6 250 280.5 416.1 43.2 22.9 88.9 89.4 98.4 260 305.5 423.7 44.0 24.9 93.8 93.6 102.6 285 312.4 425.7 44.3 25.5 95.2 94.7 103.7 292 312.5 425.7 44.3 25.5 95.2 94.8 148.0 292 760.5 514.8 54.5 62.0 178.4 137.3 148.0 740 765.5 515.5 54.6 62.4 179.3 137.6 148.0 745 770.5 516.2 54.6 62.8 180.2 137.9 148.0 750 850.5 527.4 55.9 69.3 194.6. 142.6 151.6 830 930.5 537.7 57.1 75.8 208.8 146.8 155.8 910 1010.5 547.4 58.2 82.4 222.9 150.7 159.7 990 1090.5 556.5 59.3 88.9 237.0 154.3 163.3 1070 1170.5 565.2 55.6 95.4 246.4 156.5 165.5 1150 1250.5 573.3 51.2 101.9 255.0 158.5 167.5 1230 Attachment to SIR-00-136, Rev. 1 V Structural Integrity Associates, Inc.

Table 10 Tabulated Values for Bottom Head Core Not Critical Cunrc (Curvc B)

Pressure-Temperature Curve Calculation (Heatup/Cooldown, Core Not Critical = Curve B)

Inputs: Plant = Hope Creekv Component = Bottom Head (Penetrations Portion)

Vessel thickness, t = 6.100 inches, so nt = 2.470 'linch Vessel Radius, R => 126.: inches ARTNDT = 30.0  ; = => FA EPPYs Safety Factor = js 2.00 Stress Concentration Factor = 2.30 Bottom Head Penetrations Cooldown Rate, CR = 100 -F/hr Mm, = ., 2.287: (From N-588, for inside surface axial flaw)

KIT = 8.76 ksi*inchl 2 (From N-588, for cooldown rate)

MT =. 0.285 . (From Figure G-2214-2)

Temperature Adjustment = 9.0 OF, Instrument Uncertainty Height of Water for a Full Vessel = K 562.5 inches (FEM stresses include deadweight)

Pressure Adjustment = 20.3- psig (full vessel at 70'F)

Pressure Adjustment = 220.5- psig (Instrument Uncertainty)

Conversion factor from ksi to psi = 10 1000::

Unit Pressure = 1,563 psig Flange RTNDT = ' 19.0 OF Gauge Calculated Adjusted Temperature Adjusted Pressure Temperature Pressure for T Temperature Kic Kip P for P-T Curve P-T Curve (7F9 (°F) (ksl*inch'7) (ksi*inch11) (psIg) (7F) (psig) 79.0 79.0 88.44 39.84 0 79 0 79.0 70.0 79.34 35.29 647 79 606 88.0 79.0 88.44 39.84 731 88 690 92.0 83.0 93.05 42.14 773 92 732 96.0 87.0 98.03 44.64 818 96 778 100.0 91.0 103.43 47.34 868 100 827 104.0 95.0 109.28 50.26 921 104 881 108.0 99.0 115.62 53.43 980 108 939 112.0 103.0 122.48 56.86 1043 112 1002 116.0 107.0 129.92 60.58 1111 116 1070 120.0 111.0 137.97 64.61 1185 120 1144 124.0 115.0 146.70 68.97 1265 124 1224 128.0 119.0 156.15 73.70 1351 128 1310 132.0 123.0 166.39 78.82 1445 132 1404 136.0 127.0 177.48 84.36 1547 136 1506 Attachment to SIR-00-136, Rev. 1 V Structural IntegrityAssociates, Inc.

Table 11 Tabulated Values for Core Critical Curve (Curve C) for 32 & 48 EFPY Pressure-TemperatureCurve Calculation (Core Critical = Curve C)

InPuts: Plant = Hope Ceek EFPY=  ; 32 &48 Curve A Leak Test Temperature = - 130.0 OF (at 32 EFPY and 1,1 00 psig)

Curve A Leak Test Temperature = , 138.0 °F (at 48 EFPY and 1,1 00 psig)

Hydro Test Pressure= 1,563 psig Flange RTNDT=.-.:-.19.0.- . F Curve B Curve B Temperature Pressure for Curve C Curve C Upper Vessel Upper Vessel Temperature Pressure OF (psig) (°F) (psig) 0.0 88.0 0 3.4 50.0 88.0 50 27.6 75.0 88.0 75 37.5 90.0 88.0 90 43.0 100.0 92.0 100 54.4 125.0 103.4 125 70.8 175.0 119.8 175 77.6 202.0 126.6 202 90.6 220.0 130.6 220 96.6 250.0 136.6 250 98.4 260.0 138.4 260 102.6 285.0 142.6 285 103.7 291.9 143.7 292 148.0 292.0 188.0 292 148.0 740.0 188.0 740 148.0 745.0 188.0 745 148.0 750.0 188.0 750 151.6 830.0 191.6 830 155.8 910.0 195.8 910 159.7 990.0 199.7 990 163.3 1070.0 203.3 1070 165.5 1150.0 205.5 1150 167.5 1230.0 207.5 1230 Attachment to SIR-00-136, Rev. I V Structural IntegrityAssociates, Inc.

Figure 1 Pressure Test P-T Curve (Curve A) for 32 EFPY Hope Creek Pressure Test Curve (Curve A), 32 EFPY 1,200 1,100 1,000 la 0 900 X I 0~

g 800 -

7 .

0 -I I I-

-j LU 1 I7 I W

U) 700 I o

0 0 I 600 J

z 500 f-

_j 400 I w

U)

U) 300 I -F ILl 200 - -Beltline H B0ltup

- - - Bottom Head 100 - 790 F I

- - Upper Vessel O' I 1 1 1 1 1 i+ I i i i iI I 1 0 50 100 150 200 250 MINIMUM REACTOR VESSEL METAL TEMPERATURE (°F)

Attachment to SIR-00-136, Rev. I V Structural Integrity Associates, Inc.

Figure 2 Pressure Test P-T Curve (Curvc A) for 48 EFPY Hope Creek Pressure Test Curve (Curve A), 48 EFPY 1,200 I I 1,100 1,000 900 T. 800 0

Il

-a, I

V0 700 I.

W W A

• 1 0

o-600 :I I

]

z 500 I-L) I

-1

_; 400 W.

I I

an w 300 -

sr:

200 -

-Beltline 0oltup I I - - - Boftom Head 100 -

790 F

- - Upper Vessel T

-- i -- ,

I I I I I I 0 . h. . . . . . . . I . , I , . I I . I.

0 50 100 150 200 250 0

MINIMUM REACTOR VESSEL METAL TEMPERATURE ( F)

Attachment to SIR-00-136, Rev. 1 V StructuralIntegrityAssociates, Inc.

Figure 3 Core Not Critical Curve (Curve B) for 32 EFPY Hope Creek HeatuplCooldown, Core Not Critical Curve (Curve B), 32 EFPY 1,200 1,100 1,000

-a 900 a j _

0 w I

= 800 I I EL 0

_j w

a: 700

'TI co a ]

'U J

f I-0 600 i I -1 I

I-500 I

_z I

0~

400

I 300

- Beltline 200 l Boltup : - - - Bottom Head 100 -

- 79°F

- -Upper Vessel

=

i I I I -I -I-TE1 1IT1 1 0

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

Attachment to SIR-00-136, Rev. 1 V StructuralIntegrityAssociates, Inc.

Figure 4 Core Not Critical Curve (Curve B) for 48 EFPY Hope Creek Heatup/Cooldown, Core Not Critical Curve (Curve B), 48 EFPY 1,200 II I I I I I I I I I I ,- I I I I II I I

3:

1,100 - I I I I I I I I I I I I1 I I I ' . I I I~~~~ I ~~~~

_IIII

.t 1,000-ZU) 0..

. I

< 900 I LU a- -- AI -

I o 800 I-j , g

-J w l lI I I I I' _1 cn 700-I I o 600-

_ I LU w 500 I I 1 I1 z

I E 400- IF I Lu Il 300

_ _ l Co - 13eltline 200

- - - Bottom Head 100 Bo tup - - Upper Vessel I I I 0

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

Attachment to SIR-00-136, Rev. I Structural Integrity Associates, Inc.

Figure 5 Core Critical Curve (Curve C) for 32 EFPY Hope Creek HeatuplCooldown, Core Critical Curve (Curve C), 32 & 48 EFPY 1,200 - _I I I I I I I I i I I I I II - I I

-1 1 II I I 1 1 I1 1 I 1 I 1I I T I I I I

_I

a. i-_iI 900 I a II Lu 800 0

I-j Lu co 700 co Lu 600 0

Lu z

500 I-j 400-Lu Cn Lu 300 a-

= . Minimum 200 -_ Criticality with

=- = .Normal Water 100 Level 88°F 0

-I 1 I I-- =4 =_ i i i i4 i 1 i i4 Er-_

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

Attachment to SIR-00-136, Rev. I r Structural IntegrityAssociates, Inc.