View source for 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)

{{Adams
| number = ML080740287
| issue date = 03/10/2008
| title = Technical Specification Change Request No. 348 - Relocation of Pressure and Temperature (P-T) Curves to the Pressure and Temperature Limits Report (PTLR)
| author name = Cowan P
| author affiliation = AmerGen Energy Co, LLC
| addressee name = 
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000219
| license number = DPR-016
| contact person = 
| case reference number = RA-08-004
| document type = Letter type:RA, License-Application for Facility Operating License (Amend/Renewal) DKT 50, Technical Specifications
| page count = 64
| project = 
| stage = Other
}}

=Text=
{{#Wiki_filter: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)

==1.0    DESCRIPTION==

==2.0    PROPOSED CHANGE==

==3.0    BACKGROUND==

==4.0    TECHNICAL ANALYSIS==

==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&sect;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&deg;F to >212&deg;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&deg;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&deg;F and 36&deg;F, respectively (Ref. 15). Therefore, selecting the boltup temperature to be 85&deg;F provides 49&deg;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 &#xfd;n 'x                                                                 de, 198 E it gr wit 19 &#xfd; 'CLe Edition with ion1989     Addenda 8 A d n&#xfd;d         "Fracture Toughness Criteria for  Protection io      Vga Against   FaC Failure."
                                                 'F&#xfd;ac 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,-&#xfd;O&#xfd;44-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&deg;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 hours 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 hours 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 hours.
: 2.         Total leakage rate (identified and unidentified) shall be calculated Pressure and temperature                        at least once every 8 hours.
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      *&#xfd;                                                         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&deg;&deg;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&JiC&#xfd; 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&deg;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&deg;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&deg;F/hr startup transient. Therefore, these stresses represent the bounding stresses in the feedwater nozzle associated with 100&deg;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&deg;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&deg;F 100                  /
0                                _ _,_II_
0         50       100      150        200     250       300       350 MINIMUM REACTOR VESSEL METAL TEMPERATURE (&deg;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&deg;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         &#xfd;                                  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.}}