RA-08-004, Technical Specification Change Request No. 348 - Relocation of Pressure and Temperature (P-T) Curves to the Pressure and Temperature Limits Report (PTLR)

Technical Specification Change Request No. 348 - Relocation of Pressure and Temperature (P-T) Curves to the Pressure and Temperature Limits Report (PTLR)

ML080740287
Person / Time
Site:	Oyster Creek
Issue date:	03/10/2008
From:	Cowan P AmerGen Energy Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
RA-08-004
Download: ML080740287 (64)

Text

AmerGen Energy Company, LLC www.exeloncorp.com AmerGen. An Exelon Company 200 Exelon Way Kennett Square, PA 19348 10 CFR 50.90 March 10, 2008 RA-08-004 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001

Subject:

Oyster Creek Nuclear Generating Station Facility Operating License No. DPR-16 NRC Docket No. 50-219 Technical Specification Change Request No. 348 - Relocation of Pressure and Temperature (P-T) Curves to the Pressure and Temperature Limits Report (PTLR)

Pursuant to 10 CFR 50.90, AmerGen Energy Company, LLC (AmerGen) hereby requests a change to the Technical Specifications included in Oyster Creek Operating License No. DPR-

16. The proposed change modifies Technical Specifications (TS) Section 1.0 ("Definitions"),

Limiting Conditions for Operation Section 3.3 ("Reactor Coolant"), Surveillance Requirement 4.3

("Reactor Coolant"), and 6.0 ("Administrative Controls") to delete reference to the pressure and temperature curves, and include reference to the Pressure and Temperature Limits Report (PTLR). This change adopts the methodology of SIR-05-044-A, "Pressure-Temperature Limits Report Methodology for Boiling Water Reactors," dated April 2007 for preparation of the pressure and temperature curves, and incorporates the guidance of TSTF-419-A ("Revise PTLR Definition and References in ISTS 5.6.6, RCS PTLR"). The Oyster Creek PTLR has been developed based on the methodology and template provided in SIR-05-044-A, and is enclosed as reference for NRC review.

AmerGen requests approval of the proposed changes by October 15, 2008, in order to support the Oyster Creek fall 2008 refueling outage. Once approved, the amendment shall be implemented within 60 days. The proposed changes have been reviewed by the Oyster Creek Plant Operations Review Committee and approved by the Oyster Creek Nuclear Safety Review Board. No new regulatory commitments are established by this submittal.

We are notifying the State of New Jersey of this application for changes to the Technical Specifications by transmitting a copy of this letter and its attachments to the designated State Official.

If any additional information is needed, please contact Tom Loomis at (610) 765-5510. / ,

RA-08-004 March 10, 2008 Page 2 I declare under penalty of perjury that the foregoing is true and correct. Executed on the 10 th of March, 2008.

Respectfully, Pamela B. Cwan Director - Licensing & Regulatory Affairs AmerGen Energy Company, LLC

Enclosures:

(1) Evaluation of Proposed Change (2) Markup of Proposed Technical Specification and Bases Page Changes (3) Pressure and Temperature Limits Report cc: S. J. Collins, Administrator, USNRC Region I G. E. Miller, USNRC Project Manager, Oyster Creek M. S. Ferdas, USNRC Senior Resident Inspector, Oyster Creek Director, Bureau of Nuclear Engineering, New Jersey Department of Environmental Protection

ENCLOSURE 1 Oyster Creek Technical Specification Change Request No. 348 Evaluation of Proposed Change RA-08-004 Page 1 of 9 ENCLOSURE 1 CONTENTS

SUBJECT:

Relocation of Pressure and Temperature (P-T) Curves to the Pressure and Temperature Limits Report (PTLR)

5.0 REGULATORY ANALYSIS

5.1 No Significant Hazards Consideration 5.2 Applicable Regulatory Requirements/Criteria

6.0 ENVIRONMENTAL CONSIDERATION

7.0 PRECEDENT

8.0 REFERENCES

RA-08-004 Page 2 of 9

1.0 DESCRIPTION

This letter is a request to amend Operating License No. DPR-16 for Oyster Creek Nuclear Generating Station (OCNGS).

The proposed change modifies Technical Specifications (TS) Section 1.0 ("Definitions"),

Limiting Conditions for Operation Section 3.3 ("Reactor Coolant'), Surveillance Requirement 4.3 ("Reactor Coolant"), and 6.0 ("Administrative Controls") to delete reference to the pressure and temperature curves, and include reference to the Pressure and Temperature Limits Report (PTLR). This change adopts the methodology of SIR-05-044-A, "Pressure-Temperature Limits Report Methodology for Boiling Water Reactors," dated April 2007, for preparation of the pressure and temperature curves, and incorporates the guidance of TSTF-419-A ("Revise PTLR Definition and References in ISTS 5.6.6, RCS PTLR"). The PTLR based on the methodology and template provided in SIR-05-044-A is being supplied for review.

2.0 PROPOSED CHANGE

The proposed change modifies:

1) TS Section 1.0 to add a definition of the "Pressure and Temperature Limits Report."

2) TS Limiting Conditions for Operation 3.3 ("Reactor Coolant") to delete reference to the pressure and temperature curves in the TS, and reference the PTLR.

3) TS Surveillance Requirement 4.3 ("Reactor Coolant") to delete reference to the pressure and temperature curves in the TS, and reference the PTLR.

4) TS Section 6.0 to include language from TSTF-419-A concerning: 1) the individual TSs that address reactor coolant system P-T limits; 2) the NRC-approved topical reports that document PTLR methodologies; and 3) the requirements for providing a revised PTLR to the NRC.

Bases pages are supplied for your information only.

3.0 BACKGROUND

A letter from the NRC dated February 6, 2007 (Reference 1) stated in relevant part, "the NRC staff has found that SIR-05-044 is acceptable for referencing in licensing applications for General Electric-designed boiling water reactors to the extent specified and under the limitations delineated in the TR and in the enclosed final SE." This Safety Evaluation Report (SER) permits Boiling Water Reactor (BWR) licensees to relocate their pressure-temperature (P-T) curves from the facility TS to a Pressure and Temperature Limits Report ("PTLR") utilizing the guidance in TS Task Force (TSTF)

Traveler No. 419.

As discussed in the Reference 1 SER, in a letter dated December 20, 2005, the Boiling Water Reactor Owners' Group (BWROG) submitted Licensing Topical Report (LTR)

RA-08-004 Page 3 of 9 SIR-05-044, "Pressure Temperature Limits Report Methodology for Boiling Water Reactors," Revision 0, dated December 2005 (ADAMS Accession No. ML053560336) for review and acceptance for referencing in subsequent licensing actions. The BWROG provided this LTR to support the ability of BWR licensees to relocate their P-T curves and the associated numerical values (such as heatup/cooldown rates) from the facility TS to a PTLR, a licensee-controlled document, using the guidelines provided in Generic Letter (GL) 96-03, "Relocation of Pressure Temperature Limit Curves and Low Temperature Overpressure Protection System Limits" (Reference 2). Proposed revisions to this LTR and responses to NRC staff requests for additional information (RAIs) were provided in a letter from the BWROG dated August 29, 2006 (ADAMS Accession No. ML062440387). The Reference 1 NRC safety evaluation approved the use of this report. The Structural Integrity Associates Report was issued as a final report

(-A) in April 2007 (Reference 3).

TSTF Traveler No. 419 (Reference 4) amended the Standard TS (NUREGs-1430, -

1431, -1432, -1433, and -1434) to: (1) delete references to the TS LCO specifications for the P-T limits in the TS definition for the PTLR, and (2) revise STS 5.6.6 to identify by number and title, NRC-approved topical reports that document PTLR methodologies, or the NRC safety evaluation for a plant-specific methodology by NRC letter and date. A requirement was added to the reviewers note to specify the complete citation of PTLR methodology in the plant specific PTLR, including the report number, title, revision, date, and any supplements. Only the figures, values, and parameters associated with the P-T limits are relocated to the PTLR. The TSTF also specified that the methodology, and any subsequent changes, must be reviewed and approved by the NRC. In this case, the methodology was approved in the Reference 1 letter.

This TS change adopts the methodology of SIR-05-044-A, "Pressure-Temperature Limits Report Methodology for Boiling Water Reactors," dated April 2007, and incorporates the guidance of TSTF-419-A ("Revised PTLR Definition and References in ISTS 5.6.6, RCS PTLR"), dated August 4, 2003.

The Pressure and Temperature Limits Report (PTLR) based on the methodology and template provided in SIR-05-044-A is being supplied for review. The purpose of the Oyster Creek Nuclear Generating Station PTLR is to present operating limits relating to Reactor Coolant System (RCS) pressure versus temperature limits during heatup, cooldown and hydrostatic/class 1 leak testing. The curves, which have been prepared using NRC approved methodology, will allow pressurization at lower temperatures thus saving critical path time. The pressure and temperature curves utilize the methodology of SIR-05-044-A.

4.0 TECHNICAL ANALYSIS

NRC GL 96-03 allows plants to relocate their P-T curves and numerical values of other P-T limits (such as heatup/cooldown rates) from the plant Technical Specifications to a PTLR, which is a licensee-controlled document. As stated in GL 96-03, during the development of the improved standard technical specifications (STS), a change was proposed to relocate the P-T limits currently contained in the plant Technical Specifications to a PTLR. As one of the improvements to the STS, the NRC staff agreed with the industry that the curves may be relocated outside the plant Technical RA-08-004 Page 4 of 9 Specifications to a PTLR so that the licensee could maintain these limits efficiently. One of the prerequisites for having the PTLR option is that all of the methods used to develop the P-T curves and limits be NRC approved, and that the associated Licensing Topical Report (LTR) for such approval is referenced in the plant Technical Specifications.

Based on this prerequisite, the purpose of the Structural Integrity Associates Report is to provide BWRs with an NRC-approved LTR that can be referenced in plant Technical Specifications to establish BWR fracture mechanics methods for generating P-T curves/limits that allow BWR plants to adopt the PTLR option.

Historically, utilities that own BWRs have submitted license amendment requests to update their P-T curves. In addition, the current situation causes both the regulator and licensees to expend resources that could otherwise be devoted to other activities. The objective of the Structural Integrity Associates Report is to avoid these situations by providing P-T curve methods that are generically approved by the NRC so that P-T curves can be documented in a PTLR.

To implement the PTLR, the analytical methods used to develop the P-T limits must be consistent with those previously reviewed and approved by the NRC and must be referenced in the Administrative Controls section of the plant Technical Specifications.

The Structural Integrity Associates Report provides the current Structural Integrity Associates methodology for developing reactor coolant system (RCS) pressure test, core not critical, and core critical P-T curves for BWRs.

As discussed in Section 2.1 of the Reference 1 NRC Safety Evaluation Report, 10 CFR Part 50, Appendix G, requires licensees to establish limits on the pressure and temperature of the Reactor Coolant Pressure Boundary (RCPB) to protect the RCPB against brittle failure (i.e., against brittle 'last-fracture"). These limits are defined by P-T limit curves for normal operations (including heatup and cooldown operations of the RCS, normal operation of the RCS with the reactor being in a critical condition, and transient operating conditions) and during pressure testing conditions (i.e., either inservice leak rate testing and/or hydrostatic testing conditions).

As discussed in the NRC's SER that approves the BWROG LTR SIR-05-044-A (Reference 1), this LTR was prepared by Structural Integrity Associates and has three sections and two appendices. Section 1.0 describes the background and purpose for the LTR. Section 2.0 provides the fracture mechanics methodology and its basis for developing P-T limits. Section 3.0 provides a step-by-step procedure for calculating P-T limits. Appendix A provides guidance for evaluating surveillance data. Appendix B provides a template PTLR.

Section 2.0 provides the fracture mechanics methodology and its basis for developing P-T limits. The NRC staff evaluation of this section is based on the criteria contained in Attachment 1 of GL 96-03. Attachment 1 of GL 96-03 contains seven technical criteria that the contents of proposed methodology should conform to if license amendments requesting PTLRs are to be approved by the NRC staff. The NRC staff's evaluations of the contents of the BWROG methodology against the seven criteria in Attachment 1 of GL 96-03 are provided in Section 3.1 of the SER.

RA-08-004 Page 5 of 9 Section 3.0 of the LTR provides a step-by-step procedure for calculating P-T limit curves. This section indicates that P-T limits may also be developed for other RPV regions to provide additional operating flexibility.

The Pressure and Temperature Limits Report (PTLR) based on the methodology and template provided in SIR-05-044-A is being supplied for review. The pressure and temperature curves utilize the methodology of SIR-05-044-A.

As noted in the SER related to the license renewal of OCGS (Reference 6), commitment 46 discussed submitting revised P-T limits for NRC review and approval for a 60-year operating license prior to the extended period of operation. AmerGen intends to satisfy this commitment at OCNGS unless the approval of this proposed license amendment obviates the need for submitting the curves. Assuming that this amendment request is approved, future changes to the P-T curves will be performed using the approved methodology, and the 10 CFR 50.59 process will be applied to determine whether prior NRC approval of any such changes is needed.

5.0 REGULATORY ANALYSIS

5.1 No Significant Hazards Consideration AmerGen 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 amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No The proposed change modifies Technical Specifications (TS) Section 1.0

("Definitions"), Limiting Conditions for Operation Section 3.3 ("Reactor Coolant"),

Surveillance Requirement 4.3 ("Reactor Coolant"), and 6.0 ("Administrative Controls"), to delete reference to the P-T curves and include reference to the PTLR.

This change adopts the methodology of SIR-05-044-A, "Pressure-Temperature Limits Report Methodology for Boiling Water Reactors," dated April 2007 for preparation of the pressure and temperature curves, and incorporates the guidance of TSTF-419-A ("Revised PTLR Definition and References in ISTS 5.6.6, RCS PTLR"). In an NRC SER dated February 6, 2007, "the NRC staff has found that SIR-05-044 is acceptable for referencing in licensing applications for General Electric-designed boiling water reactors to the extent specified and under the limitations delineated in the TR and in the enclosed final SE." As part of this change, the PTLR based on the methodology and template provided in SIR-05-044-A is being supplied for review. The P-T curves utilize the methodology of SIR-05-044-A.

The NRC has established requirements in Appendix G to 10 CFR 50 to protect the integrity of RCPB in nuclear power plants. Additionally, 10 CFR Part 50, Appendix H, provides the NRC staff's criteria for the design and implementation of RPV material surveillance programs for operating lightwater reactors. Implementing this RA-08-004 Page 6 of 9 NRC-approved methodology does not reduce the ability to protect the RCPB as specified in Appendix G, nor will this change increase the probability of malfunction of plant equipment, or the failure of plant structures, systems, or components.

Incorporation of the new methodology for calculating P-T curves, and the relocation of the P-T curves from the TS to the PTLR provides an equivalent level of assurance that the RCPB is capable of performing its intended safety functions. Thus, the proposed change does not affect the probability or consequences of an accident previously evaluated.

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

Response: No The proposed change does not affect the assumed accident performance of the RCPB, nor any plant structure, system, or component previously evaluated. The proposed change does not involve the installation of new equipment, and installed equipment is not being operated in a new or different manner. The change in methodology ensures that the RCPB remains capable of performing its safety functions. No setpoints are being changed which would alter the dynamic response of plant equipment. Accordingly, no new failure modes are introduced which could introduce the possibility of a new or different kind of accident from any previously evaluated.

This change adopts the methodology of SIR-05-044-A, "Pressure-Temperature Limits Report Methodology for Boiling Water Reactors," dated April 2007 for preparation of the pressure and temperature curves, and incorporates the guidance of TSTF-419-A ("Revise PTLR Definition and References in ISTS 5.6.6, RCS PTLR"). In an NRC SER dated February 6, 2007, the NRC staff has found that SIR-05-044 is acceptable for referencing in licensing applications for General Electric-designed boiling water reactors.

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

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

Response: No The proposed change does not affect the function of the RCPB or its response during plant transients. There are no changes proposed which alter the setpoints at which protective actions are initiated, and there is no change to the operability requirements for equipment assumed to operate for accident mitigation. This change adopts the methodology of SIR-05-044-A, "Pressure-Temperature Limits Report Methodology for Boiling Water Reactors," dated April 2007 for preparation of the P-T curves. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

RA-08-004 Page 7 of 9 Based upon the above, AmerGen 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 NRC has established requirements in Appendix G of Part 50 to Title 10 of the Code of FederalRegulations, to protect the integrity of the RCPB in nuclear power plants.

Appendix G to 10 CFR Part 50, requires that the P-T limits for an operating light-water nuclear reactor be at least as conservative as those that would be generated if the methods of Appendix G to Section Xl of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (ASME Code, Section Xl, Appendix G) were used to generate the P-T limits. Appendix G, also requires that applicable surveillance data from RPV material surveillance programs be incorporated into the calculations of plant-specific P-T limits, and that the P-T limits for operating reactors be generated using a method that accounts for the effects of neutron irradiation on the material properties of the RPV beltline materials.

Table 1 to 10 CFR Part 50, Appendix G provides the NRC staff's criteria for meeting the P-T limit requirements of ASME Code, Section Xl, Appendix G, as well as the minimum temperature requirements of the rule for bolting up the vessel during normal and pressure testing operations. In addition, the NRC staff regulatory guidance related to P-T limit curves is found in Regulatory Guide (RG) 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," and Standard Review Plan Chapter 5.3.2, "Pressure-Temperature Limits and Pressurized Thermal Shock."

Appendix H to 10 CFR Part 50, provides the NRC staff's criteria for the design and implementation of RPV material surveillance programs for operating lightwater reactors.

OCNGS demonstrates its compliance with the Appendix H requirements through participation in the BWRVIP Integrated Surveillance Program (ISP) (Reference 5).

In March 2001 the NRC staff issued RG 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence." Fluence calculations are acceptable if they are done with approved methodologies or with methods which are shown to conform to the guidance in RG 1.190.

Section 182a of the Atomic Energy Act of 1954 requires applicants for nuclear power plant operating licenses to include TS as part of the license. The Commission's regulatory requirements related to the content of TS are set forth in 10 CFR 50.36, which requires that the TS include items in five specific categories: (1) safety limits, limiting safety system settings and limiting control settings; (2) limiting conditions for operation (LCOs); (3) surveillance requirements (SRs); (4) design features; and (5) administrative controls.

Section 50.36(d)(2)(ii) requires that LCOs be established for the P-T limits because the parameters fall within the scope of the Criterion 2 identified in the rule:

RA-08-004 Page 8 of 9 A process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

The P-T limits for BWR-designed light-water reactors fall within the scope of Criterion 2 of 10 CFR 50.36(d)(2)(ii) and are therefore ordinarily required to be included within the TS LCOs for a plant-specific facility operating license. On January 31, 1996, the NRC staff issued GL 96-03 to inform licensees that they may request a license amendment to relocate the P-T limit curves from the TS LCOs into a PTLR or other licensee-controlled document that would be controlled through the Administrative Controls Section of the TS. In GL 96-03, the NRC staff informed licensees that to implement a PTLR, the P-T limits for light-water reactors must be generated in accordance with an NRC-approved methodology and that the methodology to generate the P-T limits must comply with the requirements of 10 CFR Part 50, Appendices G and H, be documented in an NRC-approved topical report or plant-specific submittal, and be incorporated by reference in the Administrative Controls Section of the TS.

This change implements the methodology provided in the Structural Integrity Associates report (Reference 3), which will continue to ensure compliance with Appendices G and H of the Code of Federal Regulations, and the associated regulatory guidance, including TSTF-419-A, which provides TS changes.

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.

6.0 ENVIRONMENTAL CONSIDERATION

AmerGen has determined that 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 surveillance requirement. The proposed amendment, however, 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 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.

7.0 PRECEDENT None.

RA-08-004 Page 9 of 9

8.0 REFERENCES

1. Letter from H. K. Nieh (NRC) to R. C. Bunt (Southern Nuclear Operating Company),

"Final Safety Evaluation for the Boiling Water Reactor Owners' Group (BWROG)

Structural Integrity Associates Topical Report (TR) SIR-05-044, "Pressure Temperature Report Methodology for Boiling Water Reactors" (TAC NO. MC9694),"

dated February 6, 2007.

2. Generic Letter (GL) 96-03, "Relocation of Pressure Temperature Limit Curves and Low Temperature Overpressure Protection System Limits," January 31, 1996.

3. Structural Integrity Associates Topical Report (TR) SIR-05-044-A, "Pressure-Temperature Report Methodology for Boiling Water Reactors," April 2007.

4. TSTF-419-A, "Revised PTLR Definition and References in ISTS 5.6.6, RCS PTLR,"

dated August 4, 2003.

5. Letter from P. S. Tam (NRC) to C. M. Crane (AmerGen Energy Company, LLC),

"Oyster Creek Nuclear Generating Station (OCNGS) - Issuance of Amendment RE:

Use of Integrated Surveillance Program For Reactor Vessel Specimen Surveillance (TAC NO. MB7005)," dated April 27, 2004.

6. "Safety Evaluation Report Related to the License Renewal of Oyster Creek Generating Station," dated April 2007.

ENCLOSURE2 MARKUP OF PROPOSED TECHNICAL SPECIFICATION AND BASES PAGE CHANGES Revised TS Pages ii iii 1.0-8 Insert A 3.3-1 3.3-5 3.3-8a 3.3-9a 3.3-9b 3.3-9c, 4.3-1 4.3-2 6.21 Insert B

TABLE OF CONTENTS (Cont'd)

Page 1.44 Local Linear Heat Generation Rate 1.0-8 1.45 Shutdown Margin (SDM) 1.0-8 1.46 Idle Recirculation Loop 1.0-8 1.47 Isolated Recirculation Loop 1.0-8 1.48 Operational Condition 1.0--8 Section 2 Safety Limits and Limiting Safety System Settings 2.1 Safety Limit - Fuel Cladding Integrity 2.1-4 2.2 Safety Limit - Reactor Coolant System Pressure 2.2-1 2.3 Limiting Safety System Settings 2.2-3 Section 3 Limiting Conditions for Operation 3.0 Limiting Conditions for Operation (General) 3.0-1 3.1 Protective Instrumentation 3.1-1 3.2 Reactivity Control 3.2-1 3.3 Reactor Coolant 3.3-1 3.4 Emergency Cooling 3.4-1 3.5 Containment 3.5-1 3.6 Radioactive Effluents 3.6-1 3.7- Auxiliary Electrical Power 3.7-1

'A Isolation Condenser 3.8-1 3.9 Refueling 3.9-1 3.10 Core Limits 3.10-1 I 3.11 3.12 3.13 3.14 (Not Used)

Alternate Shutdown Monitoring Instrumentation Accident Monitoring Instrumentation DELETED 3.11-1 3.12-1 3.13-1 3.14-1 3.15 Explosive Gas Monitoring Instrumentation 3.15-1 3.16 (Not Used) 3.16-1 3.17 Control Room Heating, Ventilating and Air Conditioning System 3.17-1 Section 4 Surveillance Requirements 4.0 Surveillance Requirement Applicability 4.0-1 4.1 Protective Instrumentation 4.1-1 4.2 Reactivity Control 4.2-1 4.3 Reactor Coolant 4.3-i 4.4 Emergency Cooling 4.4-1 4.5 Containment 4.5- -I 4.6 Radioactive Effluents 4.6- I1 4.7 Auxiliary Electrical Power 4.7- 1 4.8 Isolation Condenser 4.8- 1 4.9 Refueling 4.9- 1 OYSTER CREEK ii Amendment No.: 166-185,1 86, 241 1.49 Pressure and Temperature'Limits Report (PTLR)

TABLE OF COIN TENTS (Cont'd)

Page 4.10 ECCs Related Core Limits 4.10-1 4.11 Sealed Source Contamination 4.11-1 4.12 Alternate Shutdown Monitoring Instrumentation 4.12-1 4.13 Accident Monitoring Instrumentation 4.13-1 4.14 DELETED 4.14-1 4.15 Explosive Gas Monitoring Instrumentation 4.15-1 4.16 (Deleted) 4.16-1 4.17 Control Room Heating, Ventilating and Air Conditioning System 4.17-1 Section 5 Design Features 5.1 Site 5.1-1 5.2 Containment 5.2-1 5.3 Auxiliary Equipment 5.3-1 Section 6 Administrative Controls 6.1 Responsibility 6-1 6.2 Organization 6-1 6.3 Facility Staff Qualifications 6-2a 6.4 DELETED 6-3 6.5 Review and Audit 6-3 6-6 Reportable Event Action 6-9 6-7 Safety Limit Violation 6-9 6-8 Procedures and Programs 6-10 6-9 Reporting Requirements 6-13 6-10 Record Retention 6-17 6-11 Radiation Protection Program 6-18 6-12 (Deleted) 6-18 6-13 High Radiation Area 6-18 6-14 Environmental Qualification 6-19*

6-15 Integrity of Systems Outside Containment 6-19 6-16 Iodine Monitoring 6-19 6-17 Post Accident Sampling 6-20 6-18 Process Control Plan 6-20 6-19 Offsite Dose Calculation Manual 6-20 6-20 DELETED 6-20 6-21 Technical Specification (TS) Bases Control Program 6-21 6-22 Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS 6-2/t

• ~REPORT (PTLR)

• Issued by NRC Order dated 10-24-80 OYSTER CREEK iii Amendment No.:91,. 9-7. 9, 109, 115.

134,I 166 I241 2--2-,2*0-, 24 1

1.42 AVERAGE PLANAR LINEAR HEAT GENERATION RATE The AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) shall be applicable to a specific planar height and is equal to the sum of the heat generation rate per unit length of fuel rod for all the fuel rods in the specified bundle at the specified height divided by the number of fuel rods in the fuel bundle at that height.

1.43 CORE OPERATING LIMITS REPORT The Oyster Creek CORE OPERATING LIMITS REPORT (COLR) is the document that provides core operating limits for the current operating reload cycle. These cycle-specific core operating limits shall be determined for each reload cycle in accordance with Specification 6.9. .f. Plant operation within these operating limits is addressed in individual specifications.

1.44 LOCAL LINEAR HEAT GENERATION RATE The LOCAL LINEAR HEAT GENERATION RATE (LLHGR) shall be applicable to a specific planar height and is equal to the AVERAGE PLANAR LINEAR GENERATION RATE (APLHGR) at the specified height multiplied by the local peaking factor at that height.

1.45 SHUTDOWN MARGIN (SDM)

SHUTDOWN MARGIN is the amount of reactivity by which the reactor would be subcritical when the control rod with the highest reactivity worth is fully withdrawn, all other operable control rods are fully inserted, all inoperable control rods are at their current position, reactor water temperature is 68§F, and the reactor fuel is xenon free. Determination of the control rod with the highest reactivity worth includes consideration of any inoperable control rods which are not fully inserted.

1.46 IDLE RECIRCULATION LOOP A recirculation loop is idle when its discharge valve is in the closed position and its discharge bypass valve and suction valve are in the open position.

1.47 ISOLATED RECIRCULATION LOOP A recirculation loop is fully isolated when the suction valve, discharge valve and discharge bypass valve are in the closed position.

1.48 OPERATIONAL CONDITION The reactor plant operational status as to criticality, reactor mode switch position, reactor coolant temperature, and/or specific system status. These conditions consist of POWER OPERATION, STARTUP MODE, SHUTDOWN CONDITION, COLD SHUTDOWN CONDITION, and REFUEL MODE. A change or entry into an operating condition is signified by movement of the reactor mode switch or a change in reactor coolant temperature from <212°F to >212°F.

OYSTER CREEK 1.0-8 Amendment No.: 147, 178, 191,212, 241

y, PRESSURE AND EM

,tREPRT ATURE LIMrITS (PTLR)

The PTLR Is the unit specific document that provides the reactor vessel pressure and temperature limits, including heatup and cooldown rates, for the current reactor vessel fluence period. These pressure and temperature limits shall be determined for each fluence Denod in accordance with Specification *.1,

3.3 REACTOR COOLANT Aoplicabiliy: Applies to the operating status of the reactor coolant system.

Objective: To assure the structure integrity of the reactor coolant system.

Specification: (A. Pressure Temperature Relationshi s (i) Reactor V essel temaurature minimum reactor vessel to eratuio e at a given pressure shan be in excess of that indicated by t c e-Ps3 o curve in Figures 3.3p1, t3.oand 7 3.3.3 f frectmivef-lp ower ap22, 27.aem32.eff cmi tfrauullepowver years,pr iverly...The maxime, pLimitraiure for Reactoep voritepie Te(iPiTLis (ii) Heatup and Cooldown Operations Reactor , noncritical - the

• -. operations atagiven pressure whntereactor and eminimum reactre h vessel temperature fr heatup s inotcooldown critical shall be he es the Preshere 3eator reactor opereactions up to 22, 27 and 32 effectivefMll power' B. ears, respectivels react(iii)Power operations - the minimum reactor vessel temperature forse-i excess ohall be in excess oft

...Th eavera by heo r co tepar crennorm "i Note: curves a no ex.3ee 3i.2 and 3.3. applyoo

2. T pm hnth e closure head is on the reactor vessel and studs are fully t

' * , *~t e ns ion e d . - _

ressure Tempeaturetepoeratr nd *

• L* I_,ilf-J (ivteppropriature

,*the Tra ofof reactor thesoln tesystem reator atagvesse temperature flane adlsulimeirsuatong (32)r oe l full hedfagssi neratei wt en Report (PTLR) . *. Th* nsiupnto 2has 7ad3 reached thirty fetv two fl effectiveasrset power years of reactor operation.

SB. Reactor mayVessel CloRrEe .020" Hea3(1/3Boltdwn: The reactor studs be elongated design preload) withvessel closure head no restrictions on

• reactor vessel temperature as long as the reactor vessel is at atmospheric he valuessecified in the pressure. Full tensioning of the studs is not permitted unless the*

it_"'*-* ,.-

""*- ........

• temperature of the reactor vessel flange and closure head flange is in Reor "-'L* . C. Thermal Transients

" * " *\"  !.Th~e average rate of reactor coolant temperature change during normal

• * ~~~heatup and cooldown shall not excceed I** nayone hour perio*

2. The pump in an idle recirculation loop shall not be started unless the temperature of the coolant within the idle recirculation loop is within 5.0°F of the reactor coolant temperature.,

. OYSTER CREEK 3.3-1 Amendment No: 42. 120. 151.

[ ... . i 8 . . . -

Pressure and temperature curves are generated in accordance with Reference 15 and are contained in the Pressure and Temperature Limits Report (PTLR).

Section 3.3 Bases:

The reactor coolant system(l) is a primary barrier against the release of fission products to the environs.

In order to provide assurance that this barrier is maintained at a high degree of integrity, restrictions have been placed on the operating conditions to which it can be subjected.

The Oyster Creek reactor vessel was designed and manufactured in accordance with General Electric Specification 21A1 105 and ASME Section I as discussed in Reference 13. The original operating limitations were based upon the requirement that the minimum temperature for pressurization be at least 60'F greater than the nil ductility transformation temperature. The minimum temperature for pressurization at any time in life has to account for the toughness properties in the most limiting regions of the reactor vessel, as well as the effects of fast neutron embrittlement.

urves A, B and C on Figures 3.3.1, 3.3.2 and 3.3.3 are derived from an evaluation of the fracture toughness properties performed on the specimens contained in Reactor Vessel Materials Surveillance Program Capsule No. 2 (Reference 14). The results of dosimeter wire analyses (Reference 14) indicated that the neutron fluence (E>1.0 MeV) at the end of 32 effective full power years of operation is 2.36 x 1018 n/cm 2 at the l/4T (T=vessel wall thickness) location. This value was used in the calculation of the djusted reference nil-ductility temperature which, in turn, was used to generate the pressure-temperature

-urves A, B and C on Figures 3.3.1, 3.3.2 and 3.3.3 (Reference 15). The 250'F maximum pressure test emperature provides ample margin against violation of the minimum required temperature. Secondary containment is not jeopardized by a steam leak during pressure testing, and the Standby Gas Treatment system is adequate to prevent unfiltered release to the stack.

Stud tensioning is considered significant from the standpoint of brittle fracture only when the preload exceed approximately 1/3 of the final design value. No vessel or closure stud minimum temperature requirements are considered necessary for preload values below 1/3 of the design preload with the vessel depressurized since preloads below 1/3 of the design preload result in vessel closure and average bolt stresses which are less than 20% of the yield strengths of the vessel and bolting materials. Extensive service experience with these materials has confirmed that the probability of brittle fracture is extremely remote at these low stress levels, irrespective of the metal temperature.

The reactor vessel head flange and the vessel flange in combination with the double "0" ring type seal are designed to provide a leak tight seal when bolted together. When the vessel head is placed on thee reactor vessel, only that portion of the head flange near the inside of the vessel rests on the vessel flange.

As the head bolts are replaced and tensioned, the vessel head is flexed slightly to bring together the entire S nt to the "0" rings of the head and vessel flain T'-*dnatIoe requirement was that boltup be done at qualication temperatures (T3OL) plus 60'F. Current Code requirements state (Ref. 16) that for application of full bolt preload and reactor pressure up to 20% of hydrostatic test

pressure, derived bythe RPV metal temperature must be at RTNDT or greater. The boltup temperature of 85'F was determining the highest value of (T3OL + 60) and the highest value of RTNDT, and by choosing the more conservative value of the two. Calculated values of (T3OL + 60) and RTNDT of the RPV metal temperature were 85°F and 36°F, respectively (Ref. 15). Therefore, selecting the boltup temperature to be 85°F provides 49°F margin over the current Code requirement based on RTND OYSTER CREEK 3.3-5 Corrected Letter dated 8/7/2000 Amendment No: 15, 42, 120, 151, 188, 203, 212

References:

I. FDSAR, Volume I, Section IV-2

2. Letter to NRC dated May 19, 1979, "Transient of May 2, 1979"

3. General Electric Co. Letter G-EN-9-55, "Revised Natural Circulation Flow Calculation", dated May 29, 1979

4. Licensing Application Amendment 16, Design Requirements Section

5. (Deleted)

6. FDSAR, Volume I, Section IV-2.3.3 and Volume I1, Appendix H

7. FDSAR, Volume 1, Table IV-2-1

8. Licensing Application Amendment 34, Question 14

9. Licensing Application Amendment 28, Item lII-B-2

10. Licensing Application Amendment 32, Question 15

11. (Deleted)

12. (Deleted)

13. Licensing Application Amendment 16, Page 1 i9to )DO) k, -h-dto

14. GPUN TDR 725 Rev. 3: Testing and Evaluation of Irradiated Reactor Vessel Materials Surveillnnri- Prnarnm

15. GENE-1313-01769 B13- (GE Nuclear Energy): Pressure-Temperature Curves Per Regulatory Guide 997 1.99, v:

Revision 2 for Oyster Creek Nuclear Generating Station.

JCG 0"

16. :E aragraph G-2222(C), Appen ix _,ge-c-Mon XI, ASME Boiler and Pressure Vessel Code, 1989 ara apl G 2222 A ýn 'x de, 198 E it gr wit 19 ý 'CLe Edition with ion1989 Addenda 8 A d nýd "Fracture Toughness Criteria for Protection io Vga Against FaC Failure."

'Fýac r

17. GPUN Safety Evaluation SE-000221-004, "Reactor Vessel Thermal s Cycles" al'cP

18. 50.59 Evaluation, OC-2006-E-001, Revised Method For The Determination re of ination nsFatigue rCumulative 0 es 9 Usage Factor for ECR 06-00046 SIR-Or5,-ýOý44-AA,'-'Pre-,s,,siurrre-Temperature Limits Report Methodology for Boiling Water Reactors" OYSTER CREEK 3.3-8a Corrected Letter dated 8/7/2000 Amendment No: 135, 14'0, 151, 198, 203, 21-2, ECR OC 06-00046

(~PAE ONTENTS DELETED FIGURE 3.3.1 OYSTER CREEK P-T CURVES VALID TO 22 EFPY 1400 1200 C

a',

a.

0 1000

-I w

U, U,

w

'p CORE BELT-UNE UMITS WITH ART OF 1456F 0

a.- 800 FOR LOWER SHELL U PLAITE 0--6 w

A.- SYSTEM HY:-RCTEST Lull?

WIT04 FO. lo. VESS!i.

600 BS-N0PN*NUiCJAD H'EATUP/

C0O~OWN- LIM:T U, NUCLEAR ICORE CRITICAL)

I,, LIMIT w

E a.

400 200 0

0 100 200 300 400o MINIMUM REACTOR VESSEL METAL TEMPERATURE CIment NO. 151,183 0OYSTER CREEKEl

- FIGUR* 3.3.2 OYSTER ClEEK P-T CUWRVES VALID TO 27 KFPr

'CORE BILTLINE LIMITS WITH ART OF 1h2.F FOR LOWER SMELL PLATE G-5-6 A

• SYSTEM. IY2ltC°fS- LItAM W*ITH F*;E-;it VIfSSE.

I

• NON.NIJ:LfAR -EAT.J.

COOLDO0o1.%.MzT C - NUCLEAR ICORE CRITICALI UMIT 0 20w 3no 3.3-9b OYSTER CREEK Amendment No. 151, 188

FIGURE 3.3.3 OYSTER CREEK P-T CURVES VALID TO 32 EFPY 1400 1200 I- 1000 CI cc CORE BELTLINE UMITS WITH ART OF i W F FOR LOWER SHELL Lw 800 PLATE G-8-6 0

A - SYSTEM WY.RTEST .MI0 V.1 - JE . IN olS --

6W

- NON-NUCLEAR 'EATUP 600 N, COCOC"' .! W:

cc A CORE CRITICAL, Wn UMNT 400 ,CARE VAUD 4F 0PERAinoN BAEID

) PIlo ON CWM A 0 200 MINIMU cMRrCALJY TEMPI*ATURE a 96OF 0

0 100 200 300 400 500 SOo 3.3-9c

%I- 1c.1. 188

4.3 REACTOR COOLANT Applicability: Applies to the surveillance requirements for the reactor coolant system.

Obiective: To determine the condition of the reactor coolant system and the operation of the safety devices related to it.

Specification: A. Materials surveillance specimens and neutron flux monitors shall be installed in the reactor vessel adjacent to the wall at the midplane of the active core.

Specimens and monitors shall be periodically removed, tested, and evaluated ressure and temperature to determine the effects of neutron fluence on the fracture toughness of the curves are contained in vessel shell materials. "e tto. tese evaluations shall be use the Pressure and ).assess the adequacy of the P-T curves A, B, and C in Figures 3.3.1, 3.3.2 and Temperature Limits 3.3.3. New curves shall be generated as re uired.

Report (PTLR),

B. Inservice inspection of ASME Code Class 1, Class 2 and Class 3 systems and components shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR, Section 50.55a(g), except where specific written relief has been granted by the NRC pursuant to 10 CFR, Section 50.55a(g)(6)(i).

Inservice testing of ASME Code Class 1, Class 2 and Class 3 pumps and valves shall be performed in accordance with Section Xl of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR, Section 50.55a(f), except where specific written relief has been granted by the NRC pursuant to 10 CFR, Section 50.55a(f)(6)(i).

D. A visual examination for leaks shall be made with the reactor coolant system at pressure during each scheduled refueling outage or after major repairs have been made to the reactor coolant system in accordance with Article 5000, Section X1. The requirements of specification 3.3.A shall be met during the test.

E. Each replacement safety valve or valve that has been repaired shall be tested in accordance with Specification C above. Setpoints shall be as follows:

Number of Valves Set Points (psia) 4 1212+/- 36 5 1221 +/- 36 F. A sample of reactor coolant shall be analyzed at least every 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> for the purpose of determining the content of chloride ion and to check the conductivity.

OYSTER CREEK 4.3-1 Amendment No.: 82, 00, 120, 150, 161, 16*, 488, 195, 261

G. Primary Coolant System Pressure Isolation Valves S~ecification:

1. Periodic leakage testing (a)on each valve listed in Table 4.3.1 shall be accomplished prior to exceeding 600 psig reactor pressure every time the plant is placed in the cold shutdown condition for refueling, each time the plant is placed in a cold shutdown condition for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> if testing has not been accomplished in the preceding 9 months, whenever the valve is moved whether by manual actuation or due to flow conditions, and after returning the valve to service after maintenance, repair or replacement work is performed.

H. Reactor Coolant System Leakage

1. Unidentified leakage rate shall be calculated at least once every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

2. Total leakage rate (identified and unidentified) shall be calculated Pressure and temperature at least once every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

curves are contained in the Pressure and Temperature Limits 3. A channel calibration of the primary containment sump flow Report (PTLR), integrator and the primary containment equipment drain tank flow

,"I integrator shall be conducted at least once per 24 months.

I. An inservice inspection program for piping identified in NRC Generic Letter 88-01 shall be performed in accordance with the NRC staff positions on schedule, methods, personnel, and sample expansion included in the generic letter or in accordance with alternate measures approved by the NRC staff.

Data is available relating neutron fluence (E> 1.0MeV) and the change in the Reference Nil-Ductility ransition Temperature (RTNr). The pressure-temperature (P-T) operating curves A,B,and C 'in Fgurese s 3.1,3.3.2,and 3.3..3 wre veloped based on the results of testing and evaluation of specknfis reived f rar, the vessel after 8.38 EFR' of operation. Simllar testing and analysis will be performed throurghN it vesssel Lf .e to trautor the effects of neutro irradiation on the reactor vessel shell mteri*als.

The inspection program will reveal problem areas should they occur, before a leak develops. In addition, extensive visual inspection for leaks will be made on critical systems. Oyster Creek was designed and constructed prior to To satisfy ALARA requirements, leakage may be measured indirectly (as from the performance of pressure indicators) if accomplished in accordance with approved procedures and supported by computations showing that the method is capable of demonstrating valve compliance with the leakage criteria.

NRC Order dated April 20, 1981.

IYSTER CREEK 4.3-2 Amendment No. W,,H8, 120, B!, 4,1&,193

6.20 MAJOR CHANGES TO RADIOACTIVE WASTE TREATMENT SYSTEMS DELETED 6.21 Technical Specifications (TS) Bases Control Prouram This program provides a means for processing changes to the Bases of these Technical Specifications.

a. Changes to the Bases of the TS shall be made under appropriate administrative controls and reviews.

b. Licensees may make changes to Bases without prior NRC approval provided the changes do not require either of the following:

1. A change in the TS incorporated in the license or

2. A change to the updated FSAR (UFSAR) or Bases that requires NRC approval pursuant to 10 CFR 50.59.

c. The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the UFSAR.

d. Proposed changes that meet the criteria of Specification 6.2 1.b. 1 or 6.2 1.b.2 above shall be reviewed and approved by the NRC prior to implementation. Changes to the bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71 (e)

OYSTER CREEK Amendment No. 108, 166*194; 240

Reporting Requirements 5.6

,5.6 Reporting Requirements 5-5 CORE OPERATING LIMITS REPORT (continued "

b. The analytical methods used determine the core9perating limits shall be aYthose previously reviewed'and approved by theNRC, specifically those

... d e s cribe d in th e fo llow ing d oc ume nts :

Identify theTopical Report(s) by nudber and title or identify the)staff Safety Evaluati6'n Report for a plantsiedfic methodology by NRCtfer and date.

The'COLR will contain the-6mplete identification for echi of the TS referenced topical p r oi*'s used to prepare the COLRIR(i.e., report number,

• . / title, revision, date, and any supplements).]

c. The corebop'erating limits shall be det rmined such that all apliable limitsa.

(e.g ruel, thermal mechanicaljir6tis, core thermal hyd raul c limits, Emnergency Core Coolingsy-stems (ECCS) limitsclear limits such as SDM, transient anoiy limits, ss and acciden a alysis limits) of the safety

_ Zanalysis

• ~~d

. T he are t

,OR , in c lu din g a ny ~m tdcyc le re visio n s or su p ple me nts

, sh all bf ,

([~~~ U REPORT ~ vie (PTLR) upo eA

>i *ý r fosuO e ai n S ci

'i) S rSyela te -c CReI ur n3 3 e c o o

a. RCS pressure and temperature R Sm Un sE limits SAheat for ciD up, n4 E cooldown, 3"Reactor PU low EL C°°lant" MT temperature operation, criticality, and hydrostatic testing as 7well as heatup and cooldown rates shall be established and documented in the PTLR for the following:

i)- Limitin-g Conditions for Operation Section 3.3, "Reactor Coolant' ii) Surveillance Requirements Section 4.3, "Reactor Coolant"

b. The analytical methods used to determine the RCS pressure and temperature limits shall be those previously reviewed and aIporoved by the NRC, specifically those described in the following document i) SIR-05-044-A, "Pressure-Temperature Limits Report Methodology for Boiling Water Reactors"

c. The PTLR shall be provided to the NRC upon issuance for each reactor

ENCLOSURE 3 PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

Oyster Creek PTLR Revision Oc Page 1 of 36 Exelon Nuclear Corporation Oyster Creek Generating Station Pressure and Temperature Limits Report (PTLR) for 32 and 36 Effective Full-Power Years (EFPY)

Prepared by: Date: 3/4/08 Independent Review by: RI:CH CI~1kjE&JiCý P6Let I Date: 3/4108 Corporate Asset Managementn%

Approved by: r*g M Date: 3/4/08 Engjfneer ng P Mgr

Oyster Creek PTLR Revision Oc Page 2 of 36 Table of Contents Section Page 1.0 Purpose 3 2.0 Applicability 3 3.0 Methodology 3 4.0 Operating Limits 4 5.0 Discussion 5 6.0 References 8 Figure 1 Oyster Creek Pressure Test (Curve A) P-T Curve (32 EFPY) 10 Figure 2 Oyster Creek Core Not Critical (Curve B) P-T Curve (32 EFPY) 11 Figure 3 Oyster Creek Core Critical (Curve C) P-T Curve (32 EFPY) 12 Figure 4 Oyster Creek Pressure Test (Curve A) P-T Curve (36 EFPY) 13 Figure 5 Oyster Creek Core Not Critical (Curve B) P-T Curve (36 EFPY) 14 Figure 6 Oyster Creek Core Critical (Curve C) P-T Curve (36 EFPY) 15 Table 1 Oyster Creek Pressure Test (Curve A) P-T Curve (32 EFPY) 16 Table 2 Oyster Creek Core Not Critical (Curve B) P-T Curve (32 EFPY) 19 Table 3 Oyster Creek Core Critical (Curve C) P-T Curve (32 EFPY) 22 Table 4 Oyster Creek ART Calculations for 32 EFPY 25 Table 5 Oyster Creek Pressure Test (Curve A) P-T Curve (36 EFPY) 26 Table 6 Oyster Creek Core Not Critical (Curve B) P-T Curve (36 EFPY) 29 Table 7 Oyster Creek Core Critical (Curve C) P-T Curve (36 EFPY) 32 Table 8 Oyster Creek ART Calculations for 36 EFPY 35 Appendix A Oyster Creek Reactor Vessel Material Surveillance Programs 36

Oyster Creek PTLR Revision Oc Page 3 of 36 PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

FOR 32 AND 36 EFFECTIVE FULL- POWER YEARS 1.0 PURPOSE The purpose of the Oyster Creek Generating Station (OCGS) Pressure and Temperature Limits Report (PTLR) is to present operating limits relating to:

• Reactor Coolant System (RCS) Pressure versus Temperature limits during Heatup, Cooldown and Hydrostatic/Class 1 Leak Testing;

• RCS Heatup and Cooldown rates;

• RPV head flange boltup temperature limits.

This report has been prepared in accordance with the requirements of Reference 6.1, Licensing Topical Report SIR-05-044-A, Revision 0. (CM-1) 2.0 APPLICABILITY This report is applicable to the OCGS RPV for 32 and 36 Effective Full-Power Years (EFPY). The following OCGS Technical Specification (TS) is affected by the information contained in this report:

• TS Limiting Conditions for Operation 3.3 ("Reactor Coolant")

" TS Surveillance Requirement 4.3 ("Reactor Coolant")

The Oyster Creek Reactor Vessel Pressure and Temperature Limits for 32 to 50 EFPY has been developed per Reference 6.4. Future revisions of the PTLR to implement the 40 to 50 EFPY must be revised per the 10CFR50.59 review process as applicable.

3.0 METHODOLOGY The limits in this report were derived as follows:

1) The methodology used is in accordance with Reference 6.1, which has been approved for BWR use by the NRC.

2) The neutron fluence is calculated in accordance with NRC Regulatory Guide 1.190 (Reference 6.5) using the RAMA computer code, as documented in Reference 6.2.

Oyster Creek PTLR Revision Oc Page 4 of 36

3) The adjusted reference temperature (ART) values for the limiting beltline materials are calculated in accordance with NRC Regulatory Guide 1.99, Revision 2 (Reference 6.6), as documented in Reference 6.3.

4) The pressure and temperature limits were calculated in accordance with Reference 6.1, as documented in Reference 6.4 and Reference 6.16.

5) This revision of the pressure and temperature limits is to incorporate the following changes:

0 Initial issue of PTLR.

Changes to the curves, limits, or parameters within this PTLR, based upon new irradiation fluence data of the RPV, or other plant design assumptions in the Updated Final Safety Analysis Report (UFSAR), can be made pursuant to 10 CFR 50.59, provided the above methodologies are utilized.

The revised PTLR shall be submitted to the NRC upon issuance.

4.0 OPERATING LIMITS The pressure-temperature (P-T) curves included in this report represent steam dome pressure versus minimum vessel coolant temperature and incorporate the appropriate non-beltline limits and irradiation embrittlement effects in the beltline region.

The operating limits for pressure and temperature are required for three categories of operation: (a) hydrostatic pressure tests and leak tests, referred to as Curve A; (b) core not critical operation, referred to as Curve B; and (c) core critical operation, referred to as Curve C.

Complete P-T curves were developed for 32 and 36 EFPY for Oyster Creek, as documented in Reference 6.4. The OCGS P-T curves for 32 EFPY are provided in Figures 1 through 3, and a tabulation of the curves is included in Tables 1 through 3. The OCGS P-T curves for 36 EFPY are provided in Figures 4 through 6, and a tabulation of the curves is included in Tables 5 through 7 (Reference 6.16). The adjusted reference temperature (ART) tables for the OCGS vessel beltline materials are shown in Table 4 for 32 EFPY and Table 8 for 36 EFPY (References 6.4 and 6.16).

The resulting P-T curves are based on the geometry, design and materials information for the OCGS vessel with the following conditions:

Oyster Creek PTLR Revision Oc Page 5 of 36

• Heatup and Cooldown rate limit during Hydrostatic and Class 1 Leak Testing (Figures 1 and 4: Curve A): _<250F/hourl.

• Normal Operating Heatup and Cooldown rate limit (Figures 2 and 5: Curve B - non-nuclear heating, and Figures 3 and 6: Curve C - nuclear heating): _<100°F/hour2 .

RPV head installation temperature limit (Figures 1 and 4: Curve A - Hydrostatic and Class 1 Leak Testing; Figures 2 and 5: Curve B - non-nuclear heating): _> 60 0F.

5.0 DISCUSSION The adjusted reference temperature (ART) of the limiting beltline material is used to adjust the beltline P-T curves to account for irradiation effects. Regulatory Guide 1.99, Revision 2 (RG 1.99) provides the methods for determining the ART. The RG 1.99 methods for determining the limiting material and adjusting the P-T curves using ART are discussed in this section.

The vessel beltline copper and nickel values were obtained from the evaluation of the OCGS vessel plate and weld materials (Reference 6.3). The copper (Cu) and nickel (Ni) values were used with Tables 1 and 2 of RG 1.99 to determine a chemistry factor (CF) per Paragraph 1.1 of RG 1.99 for welds and plates, respectively.

The peak RPV IDfluence used in the P-T curve evaluation for 32 EFPY is 4.66x1 018 n/cm 2 for OCGS from Reference 6.2, which was calculated using methods that comply with the guidelines of 2

RG 1.190 (Reference 6.5). The peak IDfluence calculated for 36 EFPY is 5.17x1 018 n/cm (Reference 6.2). These fluence values apply to the limiting beltline lower-intermediate shell plate 564-03C for OCGS. The fluence values were adjusted for the lower intermediate plates based upon an attenuation factor of 0.630 for a postulated 1/4t flaw. As a result, the 1/4t 32 EFPY fluence for the limiting lower-intermediate plate is 2.94x1018 n/cm 2 for OCGS. The 1/4t 36 EFPY fluence for the limiting lower-intermediate plate is 3.26x1018 n/cm 2 for OCGS.

The P-T curves for the core not critical and core critical operating conditions at a given EFPY apply for both the 1/4t and 3/4t locations. When combining pressure and thermal stresses, it is usually necessary to evaluate stresses at the 1/4t location (inside surface flaw) and the 3/4t location (outside surface flaw). This is because the thermal gradient tensile stress of interest is in the inner wall during cooldown and is in the outer wall during heatup. However, as a conservative

'Interpreted as the temperature change in any 1-hour period is less than or equal to 25TF.

2 Interpreted as the temperature change in any 1-hour period is less than or equal to 100TF.

Oyster Creek PTLR Revision Oc Page 6 of 36 simplification, the thermal gradient stress at the 1/4t location is assumed to be tensile for both heatup and cooldown. This results in the approach of applying the maximum tensile stress at the 1/4t location. This approach is conservative because irradiation effects cause the allowable toughness at 1/4t to be less than that at 3/4t for a given metal temperature. This approach causes no operational difficulties, since the BWR is at steam saturation conditions during normal operation, which is well within the P-T curve limits.

For the core not critical curve (Curve B) and the core critical curve (Curve C), the P-T curves specify a coolant heatup and cooldown temperature rate of !5100°F/hr for which the curves are applicable.

However, the core not critical and the core critical curves were also developed to bound transients defined on the RPV thermal cycle diagram and the nozzle thermal cycle diagrams. For the hydrostatic pressure and leak test curve (Curve A), a coolant heatup and cooldown temperature rate of _<25 0F/hr must be maintained. The P-T limits and corresponding limits of either Curve A or B may be applied, if necessary, while achieving or recovering from test conditions. So, although Curve A applies during pressure testing, the limits of Curve B may be conservatively used during pressure testing if the pressure test heatup/cooldown rate limits cannot be maintained.

The initial RTNDT, the chemistry (weight-percent copper and nickel) and adjusted reference temperature at the 1/4 thickness location for all RPV beltline materials significantly affected by fluence (i.e., fluence > 10"7 n/cm 2 for E > 1 MeV) are shown in Table 4 for 32 EFPY and in Table 8 for 36 EFPY. The initial RTNDT values shown in Tables 4 and 8 (from Reference 6.8) were developed using the procedures of Branch Technical Position MTEB 5-2 in Standard Review Plan 5.3.2 in NUREG-0800, and they have been previously approved for use by the NRC (Reference 6.13).

Per Reference 6.15 and in accordance with Appendix A of Reference 6.1, the OCGS representative weld and plate surveillance materials data were reviewed from the Boiling Water Reactor Vessel and Internals Project (BWRVIP) Integrated Surveillance Program (ISP). The representative heats of weld and plate material in the ISP are not the same as the limiting weld and plate heats in the vessel.

For the limiting RPV plate 564-03C, BWRVIP "Procedure 2" was utilized since the heat number of this material is different than the heat number of the BWRVIP ISP Representative Material.

Surveillance data was not used in the evaluation procedure since the heat numbers do not match and there are not two or more credible data sets available for this material.

Oyster Creek PTLR Revision Oc Page 7 of 36 For limiting RPV weld heat 86054B, BWRVIP "Procedure 2" was utilized since the heat number of this material is different than the heat number of the BWRVIP ISP Representative Material.

Surveillance data was not used in the evaluation procedure since the heat numbers do not match and there are not yet two or more credible data sets available for this material. Therefore, the chemistry factors (CFs) from the tables in Regulatory Guide 1.99, Revision 2 (Reference 6.6) were used in the determination of the ART values for all materials for the OCGS vessel.

The only computer code used in the determination of the OCGS P-T curves was the ANSYS (Release 8.1 with Service Pack 1) finite element computer program for the feedwater nozzle (non-beltline) stresses. This analysis was performed to determine through-wall thermal and pressure stress distributions for the OCGS feedwater nozzles due to a step-change thermal transient (Reference 6.7). The ANSYS program was controlled under the vendor's 10 CFR 50 Appendix B Quality Assurance Program for nuclear quality-related work. Benchmarking consistent with NRC GL 83-11, Supplement 1 (Reference 6.9) was performed as a part of the computer program verification by comparing the solutions produced by the computer code to hand calculations for several problems. The following inputs were used as input to the finite element analysis:

• With respect to operating conditions, stress distributions were developed for a thermal shock of 450'F, which represents the maximum thermal shock for the feedwater nozzle during normal operating conditions. The stress results for a 4500F shock are appropriate for use in developing the non-beltline P-T curves based on the limiting feedwater nozzle, as a shock of 4500F is representative of the Turbine Roll transient that occurs in the feedwater nozzle as part of the 100°F/hr startup transient. Therefore, these stresses represent the bounding stresses in the feedwater nozzle associated with 100°F/hr heatup/cooldown limits associated with the P-T curves for the upper vessel feedwater nozzle region.

• Heat transfer coefficients were calculated from the governing design basis stress report for the OCGS feedwater nozzle and from a model of the heat transfer coefficient as a function of flow rate. The heat transfer coefficients were evaluated at flow rates that bound the actual operating conditions in the feedwater nozzles at OCGS.

" A two-dimensional, axisymmetric finite element model of the feedwater nozzle was constructed using the same modeling techniques that were employed to evaluate the feedwater nozzle in the governing design basis stress report. In order to properly model

Oyster Creek PTLR Revision Oc Page 8 of 36 the feedwater nozzle, the analysis was performed as a penetration in a sphere and not in a cylinder. To make up for this difference in geometry, a conversion factor of 3.2 times the cylinder radius was used to model the sphere (Reference 6.7). Material properties were evaluated at 3250 F to conservatively bound the 100°F condition where the maximum stresses occurred.

6.0 REFERENCES

6.1 Structural Integrity Associates Report No. SIR-05-044-A, Revision 0, "Pressure-Temperature Limits Report Methodology for Boiling Water Reactors," April 2007, SI File No. GE-1 OQ-401.

(CM-1) 6.2 TransWare Enterprises Inc. Report No. EXL-FLU-001-R-002, Revision 0, "Fluence Evaluation for Oyster Creek Reactor Pressure Vessel," SI File No. OC-05Q-257.

6.3 Structural Integrity Associates, Inc. Calculation No. OC-05Q-301, Revision 1, "Adjusted Reference Temperature Evaluation," May 11, 2006.

6.4 Structural Integrity Associates, Inc. Calculation No. OC-05Q-313, Revision 3, "Revision of the Oyster Creek Generating Station (OCGS) Pressure-Temperature (P-T) Curves to a PTLR Format, November 19, 2007.

6.5 U. S. Nuclear Regulatory Commission Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," March 2001.

6.6 U. S. Nuclear Regulatory Commission Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," May 1988.

6.7 Structural Integrity Associates Calculation No. OC-05Q-307, Revision 0, "Feedwater Nozzle Green's Functions," July 2005.

6.8 General Electric Report GENE-B1i3-01769, "Pressure-Temperature Curves per Regulatory Guide 1.99, Revision 2 for the Oyster Creek Nuclear Generating Station," July 1995, SI File No.

OC-05Q-21 0.

6.9 U. S. Nuclear Regulatory Commission, Generic Letter 83-11, Supplement 1, "Licensee Qualification for Performing Safety Analyses," June 24, 1999.

6.10 Part 50 of Title 10 of the Code of Federal Regulations, Appendix G, "Fracture Toughness Requirements," January 2005.

6.11 Part 50 of Title 10 of the Code of Federal Regulations, Appendix H, "Reactor Vessel Material Surveillance Program Requirements," January 2005.

Oyster Creek PTLR Revision Oc Page 9 of 36 6.12 Manahan, M. P., et. al., Examination, Testing, and Evaluation of Specimens from the 2100 Irradiated Pressure Vessel Surveillance Capsule for the Oyster Creek Nuclear Generating Station, Battelle Columbus Laboratories Report BCL-382-85-1, Rev. 1, October 1985, SI File No. GPUN-27Q-215.

6.13 Letter from Pao-Tsin Kuo (U. S. NRC) to Mr. Timothy Rausch (AmerGen Energy Company, LLC), "Safety Evaluation Report Related To The License Renewal of Oyster Creek Nuclear Generating Station", Docket No. 50-219, dated March 30, 2007.

6.14 Deleted.

6.15 Letter from P. S. Tam (NRC) to C. M.Crane (AmerGen Energy Company, LLC), "Oyster Creek Nuclear Generating Station (OCNGS) - Issuance of Amendment RE: Use of Integrated Surveillance Program For Reactor Vessel Specimen Surveillance (TAC NO. MB7005)", dated April 27, 2004.

6.16 Letter TJG-08-001 from T.J. Griesbach (SI) to Greg Harttraft (Exelon), Revised Calculation of P-T Limit Curves for the Oyster Creek Generating Station, dated February 26, 2008.

Oyster Creek PTLR Revision Oc Page 10 of 36 Figure 1: Oyster Creek Pressure Test (Curve A) P-T Curve (32 EFPY)

(Reference 6.4) 1,400-1,300-1,200 1,100

, 1,000 C. 900 0

I,-

LU

0) 800 0)

LU w 700 Z

0 n- 600 z_

I-500 LU LU.IL 3400__ Bolt-up n- Temp 300 600F 200 100 0 50 100 150 200 250 300 350 MINIMUM REACTOR VESSEL METAL TEMPERATURE (-F)

Oyster Creek PTLR Revision Oc Page 11 of 36 Figure 2: Oyster Creek Core Not Critical (Curve B) P-T Curve (32 FPY)

(Reference 6.4) 1,400_

1,300 1,200 1,100 0.

C' 1,000

m. 900 0

l-LU

0) 800 LU 0 700_ _

LU w 600 zI-2 500 L3 n- 400 Cn

. 300_ _

Bolt-up 200 Temp 60°F 100 /

0 _ _,_II_

0 50 100 150 200 250 300 350 MINIMUM REACTOR VESSEL METAL TEMPERATURE (°F)

Oyster Creek PTLR Revision Oc Page 12 of 36 Figure 3: Oyster Creek Core Critical (Curve C) P-T Curve (32 E FPY)

(Reference 6.4) 1,400_

1,300 1,200 1,100 C- 1,000 z

LU

a. 900 0

I-LU (1) 800 C,,.

0 700_

- 6000 Z

D 600-I-

LU LU CO)

. 300 Minimum Criticality 200 Temp 116°F 100 0

0 50 100 150 200 250 300 350 MINIMUM REACTOR VESSEL METAL TEMPERATURE (OF)

Oyster Creek PTLR Revision Oc Page 13 of 36 Figure 4: Oyster Creek Pressure Test (Curve A) P-T Curves (36 FPY)

(Reference 6.4) 1,400-1,300 1,100 1,000 900 U.

0 I-.

,-I m 800 Cl) w 0- 700 5-Iz W* 400 I n Bolt-UP 300 60F_____

200 100 0

0 50 100 150 200 250 300 350 MINIMUM REACTOR VESSEL METAL TEMPERATURE (OF)

Oyster Creek PTLR Revision Oc Page 14 of 36 Figure 5: Oyster Creek Core Not Critical (Curve B) P-T Curves (36 EFPY)

(Reference 6.4) 1,400 1,300 1,200 1,100 1,000 900 800 0

700 600 C,,

w, 500 400 300 200 100 0

0 50 100 150 200 250 300 350 MINIMUM REACTOR VESSEL METAL TEMPERATURE (OF)

Oyster Creek PTLR Revision Oc Page 15 of 36 Figure 6: Oyster Creek Core Critical (Curve C) P-T Curve (36 EFPY)

(Reference 6.4) 1,400 1,300 1,200 1,100

.53 a.

0 1,000 uJ

a. 900 0

I-.

-J ul U, 800 (0

uJ 0 700 I-U Lu 600 z

500 uJ U, 400 (0

LU

a. 300 200 100 0

0 50 100 150 200 250 300 350 MINIMUM REACTOR VESSEL METAL TEMPERATURE ('F)

Oyster Creek PTLR Revision Oc Page 16 of 36 Table 1: Oyster Creek Pressure Test (Curve A) P-T Curve (32 EFPY)

(Reference 6.4)

Pressure Temperature for for P-T P-T Curve Curve (OF) (psig) 60 0.0 60 375.0 62 375.0 64 375.0 66 375.0 68 375.0 70 375.0 72 375.0 74 375.0 76 375.0 78 375.0 80 375.0 82 375.0 84 375.0 86 375.0 88 375.0 90 375.0 92 375.0 94 375.0 96 375.0 98 375.0 100 375.0 102 375.0 104 375.0 106 375.0 108 375.0 110 375.0 112 375.0 114 375.0 116 375.0 118 375.0 120 375.0 122 375.0 124 375.0 126 375.0 126 412.4 126 419.6 126 427.0

Oyster Creek PTLR Revision Oc Page 17 of 36 Pressure Temperature for for P-T P-T Curve Curve (OF) (psig) 126 434.7 126 442.7 126 451.1 126 459.8 126 468.8 126 478.3 126 488.1 126 498.3 126 508.9 126 520.0 126 531.5 126 543.5 126 555.9 126 568.9 126 582.4 126 596.5 126 611.1 126 626.3 126 642.2 126 658.7 126 675.9 126 693.7 126 712.3 126 731.7 126 748.0 126 748.0 126 748.0 126 748.0 126 748.0 128 755.0 130 762.2 132 769.7 134 777.5 136 785.7 138 794.2 140 803.0 142 812.2 144 821.7 146 831.7 148 842.1 150 852.8 152 864.1 154 875.7

Oyster Creek PTLR Revision Oc Page 18 of 36 Pressure Temperature for for P-T P-T Curve Curve (OF) (psig) 156 887.9 158 900.6 160 913.7 162 927.4 164 941.7 166 956.5 168 972.0 170 988.1 172 1004.8 174 1022.3 176 1040.4 178 1059.3 180 1078.9 182 1099.4 184 1120.6 186 1142.8 188 1165.9 190 1189.8 192 1214.8 194 1240.8 196 1267.9 198 1296.0 200 1325.3 202 1355.9 204 1387.6 206 1420.6 208 1455.0 210 1490.8 212 1528.1 214 1566.9 216 1607.2

Oyster Creek PTLR Revision Oc Page 19 of 36 Table 2: Oyster Creek Core Not Critical (Curve B) P-T Curve (32 EFPY)

(Reference 6.4)

Pressure Temperature for for P-T P-T Curve Curve (OF) (psig) 60 0.0 60 0.0 62 0.0 64 0.0 66 0.0 68 0.0 70 0.0 72 0.0 74 0.0 76 0.0 78 3.2 80 10.2 82 17.6 84 25.2 86 33.2 88 41.5 90 50.1 92 59.1 94 68.5 96 78.2 98 88.3 100 98.9 102 109.9 104 121.3 106 133.2 108 145.5 110 158.4 112 171.8 114 185.8 116 200.3 118 215.4 120 231.2 122 247.5 124 264.6 126 282.3 128 300.8 130 320.0 132 340.0 134 360.8 134 375.0

Oyster Creek PTLR Revision Oc Page 20 of 36 Pressure Temperature for for P-T P-T Curve Curve (OF) (psig) 136 375.0 138 375.0 140 375.0 142 375.0 144 375.0 146 375.0 148 375.0 150 375.0 152 375.0 154 375.0 156 375.0 156 428.5 156 452.9 156 478.3 156 486.5 156 486.5 156 486.5 156 486.5 156 486.5 156 486.5 158 495.9 160 505.8 162 516.1 164 526.8 166 537.9 168 549.5 170 561.6 172 574.2 174 587.2 176 600.8 178 615.0 180 629.7 182 645.1 184 661.0 186 677.6 188 694.9 190 712.9 192 731.7 194 751.2 196 771.4 198 792.6 200 814.5 202 837.4

Oyster Creek PTLR Revision 0c Page 21 of 36 Pressure Temperature for for P-T P-T Curve Curve (OF) (psig) 204 861.2 206 886.0 208 911.8 210 938.7 212 966.6 214 995.7 216 1026.0 218 1057.5 220 1090.3 222 1124.4 224 1159.9 226 1196.9 228 1235.4 230 1275.4 232 1317.1 234 1360.5 236 1405.7 238 1452.7 240 1501.6 242 1552.5 244 1605.5

Oyster Creek PTLR Revision Oc Page 22 of 36 Table 3: Oyster Creek Core Critical (Curve C) P-T Curve (32 EFPY)

(Reference 6.4)

Temperature Pressure for for P-T Curve P-T Curve 0F) (psig) 100 0 100 0 102 0 104 0 106 0 108 0 110 0 112 0 114 0 116 0 118 3 120 10 122 18 124 25 126 33 128 42 130 50 132 59 134 68 136 78 138 88 140 99 142 110 144 121 146 133 148 146 150 158 152 172 154 186 156 200 158 215 160 231 162 248 164 265 166 282 168 301 170 320 172 340 174 375

Oyster Creek PTLR Revision Oc Page 23 of 36 Temperature Pressure for for P-T Curve P-T Curve (OF)

(psig) 176 375 178 375 180 375 182 375 184 375 186 375 188 375 190 375 192 375 194 375 196 375 198 496 200 506 202 516 204 527 206 538 208 550 210 562 212 574 214 587 216 601 218 615 220 630 222 645 224 661 226 678 228 695 230 713 232 732 234 751 236 771 238 793 240 815 242 837 244 861 246 886 248 912 250 939 252 967 254 996 256 1026 258 1057 260 1090 262 1124 264 1160

Oyster Creek PTLR Revision Oc Page 24 of 36 Temperature Pressure for for P-T Curve P-T Curve (OF) (psig) 266 1197 268 1235 270 1275 272 1317 274 1361 276 1406 278 1453 280 1502 282 1553 284 1606

Oyster Creek PTLR Revision Oc Page 25 of 36 Table 4: Oyster Creek ART Calculations for 32 EFPY (References 6.4 and 6.16)

_________ ________ FLATES-

HEAT NO./ ADJUSTMENTS FOR MARGIN AND PIECE CODE FORGING INITIAL CHEMISTRY BELTLINE IRRADIATION SHIFT NO. NO. S/N RTNDT Cu Ni CF ID Fluence Attenuation 1/4-T Fluence FF ARTNDT GD o, Margin ART 2 2 9 2 PART NAME (Note 3) 2F  %  % F 1019 n/cm 10i n/cm' -F F RF 2F gF 564-03A G-8-7 P-2161-1 17 0.21 0.48 139.4 4.66E+18 0.630 2.94E+18 0.665 92.7 17.0 10.7 40.2 149.8 Lower Intermediate Shell Plates 564-03B G-8-8 P-2136-2 8 0.18 0.46 120.7 4.66E+18 0.630 2.94E+18 0.665 80.2 17.0 12.8 42.6 130.8 564-03C G-8-6 P-2150-1 31 0.2 0.51 138.2 4.66E+18 0.630 2.94E+18 0.665 91.9 17.0 12.7 42.4 165.3 564-03D G-307-1 T-1937-2 30 0.17 0.11 79.45 2.51E+18 0.630 1.58E+18 0.515 40.9 17.0 12.6 42.3 113.2 Lower Shell Plates 564-03E G-308-1 T-1937-1 21 0.17 0.11 79.45 2.51E+18 0.630 1.58E+18 0.515 40.9 17.0 14.2 44.3 106.2 564-03F . G-307-5 P-2076-2 3 0.27 0.53 173.9 2.51E+18 0.630 1.58E+18 0.515 89.6 17.0 13.9 43.9 136.5

______________________ *'< K*' 'j 2..........S* 1 SHELLxAXI**)I. KNE I)S$C, . :A . .

Weld Metal ADJUSTMENTS FOR MARGIN AND Weld Type / Heat Flux Lot Number / INITIAL CHEMISTRY BELTLINE IRRADIATION SHIFT AN M NS, No. Number Flux Type RTNDT Cu Ni CF ID Fluence Attenuation 1/4-T Fluence FF ARTNDT OA O Margin ART 2 2 9 PART NAME 2F  %  % F n/cm r/cm2 RF 2F 9F F 2F Lower Intermediate Shel Axial Welds 2-564A, 2-564B, TA5S 0.630 2.59E+18 0.633 61.8 28.0 0.0 56.0 67.8 LwrItreitShlAiaWed 2-6and 2-564C, 860548 ARCOS B-5 -50 0.214 0.05 97.6 4.12E+18 W eld Metal W2 -564MC a ne 86054B ARCOS B-5 -8 0.214 0.05 97.6 2.48E+18 0.630 1.56E+18 0.512 50.0 25.0 0.0 50.0 92.0 Lower Shell Axial Welds 2-564. 2-564E,2 K' ~ Weld Metal K -SHELL ý WELUMFERENTIAK WE'> ., "EAK ADJUSTMENTS FOR MARGIN AND Weld Type / Heat Flux Lot Number/ INITIAL CHEMISTRY BELTLINE IRRADIATION SHIFT No. Number Flux Type RTNDT Cu Ni CF ID Fluence Attenuation 1/4-T Fluence FF ARTNDT O"D HrI Margin ART 2 2 2 PARL NAME SF  % %__F 109 n/cm2 1 0 's n/cm 2F 2F F QFF Lower Shell to Lower Intermediate Shell Circumferential Weld 3-564 1248 ARCOS B-5 -50 0.206 0.07 96.2 2.51E+18 0.630 1.58E+18 0.515 49.5 24.8 0.0 49.5 491

Oyster Creek PTLR Revision Oc Page 26 of 36 Table 5: Oyster Creek Pressure Test (Curve A) P-T Curve (36 EFPY)

(Reference 6.4)

Pressure Temperature for for P-T P-T Curve Curve (OF) (psig) 60 0.0 60 375.0 62 375.0 64 375.0 66 375.0 68 375.0 70 375.0 72 375.0 74 375.0 76 375.0 78 375.0 80 375.0 82 375.0 84 375.0 86 375.0 88 375.0 90 375.0 92 375.0 94 375.0 96 375.0 98 375.0 100 375.0 102 375.0 104 375.0 106 375.0 108 375.0 110 375.0 112 375.0 114 375.0 116 375.0 118 375.0 120 375.0 122 375.0 124 375.0 126 375.0 126 412.4 126 419.6

Oyster Creek PTLR Revision Oc Page 27 of 36 Pressure Temperature for for P-T P-T Curve Curve (OF) (psig) 126 427.0 126 434.7 126 442.7 126 451.1 126 459.8 126 468.8 126 478.3 126 488.1 126 498.3 126 508.9 126 520.0 126 531.5 126 543.5 126 555.9 126 568.9 126 582.4 126 596.5 126 611.1 126 626.3 126 642.2 126 658.7 126 675.9 126 693.7 126 712.3 126 731.7 126 735.9 126 735.9 126 735.9 126 735.9 126 735.9 128 742.3 130 749.0 132 756.0 134 763.3 136 770.9 138 778.7 140 786.9 142 795.5 144 804.3 146 813.6

Oyster Creek PTLR Revision Oc Page 28 of 36 Pressure Temperature for for P-T P-T Curve Curve (OF) (psig) 148 823.2 150 833.2 152 843.6 154 854.5 156 865.8 158 877.5 160 889.8 162 902.5 164 915.7 166 929.5 168 943.9 170 958.8 172 974.4 174 990.6 176 1007.4 178 1024.9 180 1043.2 182 1062.2 184 1081.9 186 1102.5 188 1123.9 190 1146.2 192 1169.4 194 1193.5 196 1218.7 198 1244.8 200 1272.0 202 1300.4 204 1329.8 206 1360.5 208 1392.5 210 1425.7 212 1460.3 214 1496.3 216 1533.8 218 1572.8 220 1613.4

Oyster Creek PTLR Revision Oc Page 29 of 36 Table 6: Oyster Creek Core Not Critical (Curve B) P-T Curve (36 EFPY)

(Reference 6.4)

Pressure Temperature for for P-T P-T Curve Curve (OF) (psig) 60 0.0 60 0.0 62 0.0 64 0.0 66 0.0 68 0.0 70 0.0 72 0.0 74 0.0 76 0.0 78 3.2 80 10.2 82 17.6 84 25.2 86 33.2 88 41.5 90 50.1 92 59.1 94 68.5 96 78.2 98 88.3 100 98.9 102 109.9 104 121.3 106 133.2 108 145.5 110 158.4 112 171.8 114 185.8 116 200.3 118 215.4 120 231.2 122 247.5 124 264.6 126 282.3 128 300.8 130 320.0

Oyster Creek PTLR Revision Oc Page 30 of 36 Pressure Temperature for for P-T P-T Curve Curve (OF) (psig) 132 340.0 134 360.8 134 375.0 136 375.0 138 375.0 140 375.0 142 375.0 144 375.0 146 375.0 148 375.0 150 375.0 152 375.0 154 375.0 156 375.0 156 428.5 156 452.9 156 469.9 156 469.9 156 469.9 156 469.9 156 469.9 156 469.9 156 469.9 158 478.7 160 487.9 162 497.4 164 507.3 166 517.7 168 528.4 170 539.7 172 551.3 174 563.5 176 576.1 178 589.2 180 602.9 182 617.2 184 632.0 186 647.4 188 663.5 190 680.2

Oyster Creek PTLR Revision Oc Page 31 of 36 Pressure Temperature for for P-T P-T Curve Curve (OF) (psig) 192 697.6 194 715.7 196 734.5 198 754.1 200 774.6 202 795.8 204 817.9 206 840.9 208 864.9 210 889.8 212 915.8 214 942.8 216 970.9 218 1000.2 220 1030.6 222 1062.3 224 1095.3 226 1129.6 228 1165.4 230 1202.6 232 1241.3 234 1281.6 236 1323.5 238 1367.2 240 1412.6 242 1459.9 244 1509.1 246 1560.3 248 1613.7

Oyster Creek PTLR Revision Oc Page 32 of 36 Table 7: Oyster Creek Core Critical (Curve C) P-T Curve (36 EFPY)

(Reference 6.4)

Temperature Pressure for for P-T Curve P-T Curve (1F) (psig) 100 0 100 0 102 0 104 0 106 0 108 0 110 0 112 0 114 0 116 0 118 3 120 10 122 18 124 25 126 33 128 42 130 50 132 59 134 68 136 78 138 88 140 99 142 110 144 121 146 133 148 146 150 158 152 172 154 186 156 200 158 215 160 231 162 248 164 265 166 282 168 301 170 320 172 340

Oyster Creek PTLR Revision 0c Page 33 of 36 Temperature Pressure for for P-T Curve P-T Curve (OF) (psig) 174 375 176 375 178 375 180 375 182 375 184 375 186 375 188 375 190 375 192 375 194 375 196 375 198 479 200 488 202 497 204 507 206 518 208 528 210 540 212 551 214 563 216 576 218 589 220 603 222 617 224 632 226 647 228 663 230 680 232 698 234 716 236 735 238 754 240 775 242 796 244 818 246 841 248 865 250 890 252 916 254 943 256 971

Oyster Creek PTLR Revision Oc Page 34 of 36 Temperature Pressure for for P-T Curve P-T Curve (OF) (psig) 258 1000 260 1031 262 1062 264 1095 266 1130 268 1165 270 1203 272 1241 274 1282 276 1324 278 1367 280 1413 282 1460 284 1509

Oyster Creek PTLR Revision Oc Page 35 of 36 Table 8: Oyster Creek ART Calculations for 36 EFPY (References 6.4 and 6.16)

HEAT NO.1 ADJUSTMENTS FOR MARGIN AND PIECE CODE FORGING INITIAL CHEMISTRY BELTLINE IRRADIATION SHIFT NO. NO. S/N RTNDT Cu Ni CF ID Fluence Attenuation 1/4-T Fluence FF ARTNDT OD a1 Margin ART 9 2 2 9 PART NAME (Note 3)  %  % F 1019 n/cm 1019 n/cm OFF OF OF OF OF 564-03A G-8-7 P-2161-1 17 0.21 0.48 139.4 5.17E+18 0.630 3.26E+18 0.692 96.4 17.0 10.7 40.2 153.6 Lower Intermediate Shell Plates 564-03B G-8-8 P-2136-2 8 0.18 0.46 120.7 5.17E+18 0.630 3.26E+18 0.692 83.5 17.0 12.8 42.6 134.0 564-03C G-8-6 P-2150-1 31 0.2 0.51 138.2 5.17E+18 0.630 3.26E+18 0.692 95.6 17.0 12.7 42.4 169.0 564-03D G-307-1 T-1937-2 30 0.17 0.11 79.45 2.79E+18 0.630 1.76E+18 0.539 42.8 17.0 12.6 42.3 115.1 Lower Shell Plates 564-03E G-308-1 T-1937-1 21 0.17 0.11 79.45 2.79E+18 0.630 1.76E+18 0.539 42.8 17.0 14.2 44.3 108.1 564-03F G-307-5 P-2076-2 3 0.27 0.53 173.9 2.79E+18 0.630 1.76E+18 0.539 93.7 17.0 13.9 43.9 140.6 SHELL AXIAL SEA..M.WELD.

Weld Metal ADJUSTMENTS FOR MARGIN AND Weld Type / Heat Flux Lot Number I INITIAL CHEMISTRY BELTLINE IRRADIATION SHIFT No. Number Flux Type RTNDT Cu Ni CF ID Fluence Attenuation 1/4-T Fluence FF ARTNDT oA a, Margin ART 2

PART NAME OF  %  % OF n/cm' n/cm OF OF OF OF OF Lower Intermediate Shell Axial Welds 2-564A, 2-564B, 86054B ARCOS B-5 -50 0.214 0.05 97.6 4.57E+18 0.630 2.88E+18 0.660 64.4 28.0 0.0 56.0 70.4 and 2-564C Lower Shell Axial Welds 2-564D, 2-564E, 2 86054B ARCOS B-5 -8 0.214 0.05 97.6 2.75E+18 0.630 1.73E+18 0.536 52.3 26.1 0.0 52.3 96.6 564F

[1~~~>2K V, '. 'K5V~ - '>SH~ELL CIRCOMFERENTIAL'SEAWVELD~ Q~ ~

Weld Metal ADJUSTMENTS FOR MARGIN AND Weld Type / Heat Flux Lot Number/ INITIAL CHEMISTRY BELTLINE IRRADIATION SHIFT No. Number Flux Type RTNDT Cu Ni CF ID Fluence Attenuation 1/4-T Fluence FF ARTNDT 0D ol Margin ART 2 2 PART NAME _F  %  % OF 1019 n/cm ll0 n/cm OF OF OF OF OF ILower Shell to Lower Intermediate Shell Circumferential Weld 3-564 1248 ARCOS B-5 -50 0.206 0.07 96.2 2.79E+18 0.630 1.76E+18 0.539 51.8 25.9 10.01 51.8 53.7

Oyster Creek PTLR Revision Oc Page 36 of 36 APPENDIX A Oyster Creek Reactor Vessel Material Surveillance Programs Oyster Creek:

In accordance with 10 CFR 50, Appendix H, Reactor Vessel Material Surveillance Program Requirements (Reference 6.11), the first surveillance capsule was removed from the Oyster Creek reactor vessel on February 12, 1983. The surveillance capsule contained flux wires for neutron fluence measurement, Charpy V-Notch impact test specimens and uniaxial tensile test specimens fabricated using materials from the vessel materials within the core beltline region. The flux wires and test specimens removed from the capsule were tested according to ASTM El 85-82. The methods and results of testing are presented in Reference 6.12, as required by 10 CFR 50, Appendices G and H (References 6.10 and 6.11).

Currently, OCGS has made a licensing commitment to replace the existing material surveillance program with the BWRVIP ISP. This program meets the requirements of 10 CFR 50, Appendix H, for Integrated Surveillance Programs, and has been approved by NRC. Under the ISP, there are no further capsules from OCGS to be tested.

Representative surveillance capsule materials for the OCGS limiting beltline plate and weld are in the Cooper and Hatch Unit 2 surveillance capsule programs. The next Cooper surveillance capsule is scheduled to be withdrawn and tested in 2016. The next Hatch Unit 2 surveillance capsule is scheduled for withdrawal and testing in 2017.