ML082390685

Issuance of Amendment 269 Relocation of Pressure and Temperature Curves to the Pressure and Temperature Limits Report

ML082390685
Person / Time
Site:	Oyster Creek
Issue date:	09/30/2008
From:	Geoffrey Miller Plant Licensing Branch 1
To:	Pardee C AmerGen Energy Co
Miller G, NRR/DORL, 415-2481
References
TAC MD8253
Download: ML082390685 (24)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-6001 September 30, 2008 Mr. Charles G. Pardee Chief Nuclear Officer AmerGen Energy Company, LLC 200 Exelon Way Kennett Square, PA 19348

SUBJECT:

OYSTER CREEK NUCLEAR GENERATING STATION -ISSUANCE OF AMENDMENT RE: RELOCATION OF PRESSURE AND TEMPERATURE CURVES TO THE PRESSURE AND TEMPERATURE LIMITS REPORT (TAC NO. MD8253)

Dear Mr. Pardee:

The Commission has issued the enclosed Amendment NO.269 to Facility Operating License No. DPR-16 for the Oyster Creek Nuclear Generating Station (Oyster Creek), in response to your application dated March 10, 2008, as supplemented by letters dated June 30, 2008, and September 29,2008.

The amendment revises the Oyster Creek Technical Specifications (TSs) 3.3, "Reactor Coolant." Specifically, the amendment relocates the pressure and temperature limit curves to the licensee controlled document, "Pressure and Temperature Limits Report," (PTLR).

Additionally, the amendment introduces supporting definitions and adds controls regarding the PTLR to Section 6.0, "Administrative Controls."

A copy of the related Safety Evaluation is also enclosed. Notice of Issuance will be included in the Commission's biweekly Federal Register notice.

Si"Ltu'J 1t('~

G. Edward Miller, Project Manager Plant Licensing Branch 1-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-219

Enclosures:

1. Amendment No. 269 to DPR-16

2. Safety Evaluation CC w/encls: See next page

Oyster Creek Nuclear Generating Station cc:

Mayor of Lacey Township 818 West Lacey Road Forked River, NJ 08731 Additional Distribution via UstServ

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 AMERGEN ENERGY COMPANY, LLC DOCKET NO. 50-219 OYSTER CREEK NUCLEAR GENERATING STATION AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 269 License No. DPR-16

1. The Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment filed by AmerGen Energy Company, LLC (the licensee), dated March 10,2008, as supplemented by letters dated June 30, 2008, and September 29, 2008, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act and the rules and requlations of the Commission; C. There is reasonable assurance: (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

-2

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-16 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.269

• are hereby incorporated in the license. AmerGen Energy Company, LLC shall operate the facility in accordance with the Technical Specifications.

3. This license amendment is effective as of its date of issuance and shall be implemented within 60 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION M~~

Harold K. Chernoff, Chief Plant Licensing Branch 1-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the License and Technical Specifications Date of Issuance: September 30 I 2008

ATIACHMENT TO LICENSE AMENDMENT NO. 269 FACILITY OPERATING LICENSE NO. DPR-16 DOCKET NO. 50-219 Replace the following page of Facility Operating License No. DPR-16 with the attached revised page as indicated. The revised page is identified by amendment number and contains marginal lines indicating the area of change.

Remove Insert 3 3 Replace the following pages of the Appendix A, Technical Specifications, with the attached revised pages as indicated. The revised pages are identified by amendment number and contain marginal lines indicating the area of change.

Remove Insert ii ii iii iii 1.0-9 1.0-9 3.1-1 3.1-1 3.3-8a 3.3-8a 3.3-9a 3.3-9a 3.3-9b 3.3-9b 3.3-9c 3.3-9c 4.3-1 4.3-1 6-22 6-22

-3 (3) Pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess, and use at any time any byproduct, source, or special nuclear materials as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4) Pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess, and use in amounts as required any byproduct, source, or special nuclear materials without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (5) Pursuant to the Act and 10 CFR Parts 30, 40, and 70, to possess, but not separate such byproduct, source, or special nuclear materials as may be produced by the operation of the facility.

C. This license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter 1 and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect and is subject to the additional conditions specified or incorporated below:

(1) Maximum Power Level AmerGen Energy Company, LLC is authorized to operate the facility at steady-state power levels not in excess of 1930 megawatts (thermal) (100 percent rated power) in accordance with the conditions specified herein.

(2) Technical Specifications The Technical Specifications contained in Appendices A and S, as revised through Amendment No. 269, are hereby incorporated in the license. AmerGen Energy Company, LLC shall operate the facility in accordance with the Technical Specifications.

(3) Fire Protection AmerGen Energy Company, LLC shall implement and maintain in effect all provisions of the approved fire protection program as described in the Updated Final Safety Analysis Report for the facility and as approved in the Safety Evaluation Report dated March 3, 1978, and supplements thereto, subject to the following provision:

• Implemented AMENDMENT NO.269

TABLE OF CONTENTS (Cont'd) 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 1.49 Rated Thermal Power 1.0-9 1.50 Thermal Power 1.0-9 1.51 Pressure and Temperature Limits Report (PTLR) 1.0-9 Section 2 Safety Limits and Limiting Safety System Settings 2.1 Safety Limit - Fuel Cladding Integrity 2.1-1 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-[

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 3.8 Isolation Condenser 3.8-1 3.9 Refueling 3.9-1 3.10 Core Limits 3.10-1 3.11 (Not Used) 3.11-1 3.12 Alternate Shutdown Monitoring Instrumentation 3.12-1 3.13 Accident Monitoring Instrumentation 3.13-1 3.14 DELETED 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 4.3-1 4.4 Emergency Cooling 4.4-1 4.5 Containment 4.5-1 4.6 Radioactive Effluents 4.6-1 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, 186,241, 266, 269

TABLE OF CONTENTS (cont'd) 4.10 ECCs Related Core Limits 4.10-1 4.11 Sealed Source Contamination 4.11-1 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 Control Room Envelope Habitability Program 6-21 6-23 Reactor Coolant System (ReS) PRESSURE AND TBvIPERATURE LIMITS REPORT (PTLR) 6-22

• Issued by NRC Order dated 10-24-80 OYSTER CREEK ill Amendment No.: 94,97,98, 108,115, 134, 166, 186, 232, 240, 241, 265t69

1.49 RATED THERMAL POWER (RTP)

RTP shall be a total reactor core heat transfer rate to the reactor coolant of 1930 MWt.

1.50 THERMAL POWER THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.

1.5; PRESSURE AND TEMPERATURE LIMITS REPORT (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 period in accordance with Specification 6.23.

OYSTER CREEK 1.0-9 Amendment No. 2ee,269

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

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

Soecification: A. Pressure Temperature Relationships (i) Reactor Vessel Pressure Tests - the minimum reactor vessel temperature at a given pressure shall be in excess of that indicated by the curves in the Pressure and Temperature Limits Report (PTLR).

(ii) Heatup and Cooldown Operations: Reactor noncritical -. the minimum reactor vessel temperature for heatup and cooldown operations at a given pressure when the reactor is not critical shall be in excess of that indicated by the curves in the Pressure and Temperature Limits Report (PTLR).

(iii) Power operations .- the minimum reactor vessel temperature for power operations at a given pressure shall be in excess of that indicated by the curves in the Pressure and Temperature Limits Report (PTLR).

(iv) Deleted B. Reactor Vessel Closure Head Boltdown: The reactor vessel closure head studs may be elongated ,020" (1/3 design preload) with no restrictions on reactor vessel temperature as long as the reactor vessel is at atmospheric pressure. Full tensioning of the studs is not permitted unless the temperature of the reactor vessel flange and closure head flange is in excess of the values specified in the Pressure and Temperature Limits Report (PTLR).

C. Thermal Transients

1. The average rate of reactor coolant temperature change during normal heatup and cooldown shall not exceed the values specified in the Pressure and Temperature Limits Report (PTLR).

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 50°F of the reactor coolant temperature.

OYSTER CREEK 3.3-1 Amendment No: 42, 120, 161,

+88,269

References:

I. FDSAR, Volume 1, 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 1, Section IV-2.3.3 and Volume II, Appendix H

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

8. Licensing Application Amendment 34, Question 14

9. Licensing Application Amendment 28, Item ill-B-2 to. Licensing Application Amendment 32, Question 15 II. (Deleted)

12. (Deleted)

13. Licensing Application Amendment 16, Page I

14. (Deleted)

15. SIR-05-044-A, "Pressure-Temperature Limits Report Methodology for Boiling Water Reactors,"

April 2007

16. (Deleted)

17. GPUN Safety Evaluation, SE-00022 1-004, "Reactor Vessel Thermal Cycles"

18. 50.59 Evaluation, OC-2006-E-OOI, Revised Method For The Determination of Fatigue Cumulative Usage Factor for ECR 06-00046 OYSTER CREEK 3.3-8a Corrected Letter dated 8P/2000 Amendment No: 135, 140. 151,188.203.212. ECR DC 01') 00046, 269

PAGE CONTENTS DELETED OYSTER CREEK 3.3-9a Amendment No.: 161, 188, 269

PAGE CONTENTS DELETED OYSTER CREEK 3.3-9b Amendment No.:161, 188,269

PAGE CONTENTS DELETED OYSTER CREEK 3.3-9c Amendment No.: 161, 188269

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

Objective: 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 to determine the effects of neutron fluence on the fracture toughness of the vessel shell materials. Pressure and temperature curves are contained in the Pressure and Temperature Limits 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, except where specific written relief has been granted by the NRC pursuant to 10 CFR, Section 50.55a.

C. Inservice testing of ASME Code Class 1, Class 2 and Class 3 pumps and valves shall be performed in accordance with the ASME Code for Operation and Maintenance of Nuclear Power Plants (ASME OM Code) and applicable Addenda as required by 10 CFR, Section 50.55a, except where specific written relief has been granted by the NRC pursuant to 10 CFR, Section 50.55a.

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

OYSTER CREEK 4.3-1 Amendment No.: 82, 90,120,160, 161,164,188,196, 261,268, 269

The following are exceptions to Sections C.1 and C.2 of Regulatory Guide 1.197, Revision 0:

1. The Oyster Creek CRE boundary operability is not dependent on a measured unfiltered air inleakage value (Reference Oyster Creek letter to NRC dated November 17, 2005, Letter No. 2130-05-20218). No inieakage testing for determining the unfiltered air inieakage past the CRE boundary into the CRE is required at the Oyster Creek site.

d. Measurement, at designated locations, of the CRE pressure relative to areas adjacent to the CRE boundary during the pressurization mode of operation by one subsystem (train) of the Control Room Ventilation System operating at the design flow rate, at a Frequency of 24 months. The results shall be trended and used as part of the 24 month assessment of the CRE boundary.

e. The unfiltered air inleakage limit for radiological challenges is the inleakage flow rate assumed in the licensing basis analyses of DBA consequences. Unfiltered air inleakage limits for hazardous chemicals must ensure that exposure of CRE occupants to these hazards will be within the assumptions in the licensing basis.

f. The provisions of Section 1.24 are applicable to the frequencies for assessing CRE habitability measuring CRE pressure and assessing the CRE boundary as required by paragraphs d and c, respectively.

6.23 Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLRl

a. RCS pressure and temperature limits for heat up, cooldown, low temperature operation, criticality, and hydrostatic testing as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:

i) Limiting 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 approved 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 vessel fluence period and for any revision or supplement thereto.

OYSTER CREEK 6-22 Amendment No. ass, 269

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RFLATEn TO AMENDMENT NO.269 TO FACiliTY OPERATiNG LICENSE NO. DPR-16 AMERGEN ENERGY COMPANY, LLC OYSTER CREEK NUCLEAR GENERATING STATION DOCKET NO. 50-219

1.0 INTRODUCTION

By letter dated March 10, 2008 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML080740287), as supplemented by letters dated June 30.2008, and September 29, 2008 (ADAMS Accession No. ML081900082 and ML082740103, respectively),

AmerGen Energy Company, LLC (AmerGen or the licensee) submitted Technical Change Request No. 348, requesting changes to the Technical Specifications (TSs) for the Oyster Creek Nuclear Generating Station (Oyster Creek). The supplemental, letters provided clarifying information within the scope of the original application and did not change the staffs initial proposed no significant hazards determination noticed in the Federal Register on June 17, 200 (73 FR 34339).

The amendment would revise the Oyster Creek TS 3.3, "Reactor Coolant." Specifically, the amendment would relocate the pressure and temperature (P-T) limit curves to the licensee controlled document. "Pressure and Temperature Limits Report," (PTLR). Additionally, the amendment introduces supporting definitions adds controls regarding the PTLR to Section 6.0, "Administrative Controls." Additionally, the NRC staff reviewed the initial limits to be placed in the PTLR which include P-T limit curves for operation up to 50 effective full power years (EFPY) of operation.

2.0 REGULATORY EVALUATION

The Nuclear Regulatory Commission (NRC) has established requirements in Part 50 or Title 10 of the Code of Federal Regulations (10 CFR) to protect the integrity of the reactor coolant pressure boundary in nuclear power plants. The NRC staff evaluates the acceptability of a facility's proposal to relocate the P-T limit curves to the PTLR based on the following NRC regulations and quidance:

Appendix G, "Fracture Toughness Requirements," to 10 CFR Part 50; Appendix H, "Reactor Vessel Material Surveillance Program Requirements," to 10 CFR Part 50; Regulatory Guide (RG) 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials;"

Generic Letter (GL) 92-01, Rev. 1, "Reactor Vessel Structural Integrity;"

-2 Standard Review Plan (SRP) Section 5.3.2, "Pressure-Temperature Limits, Upper-Shelf Energy, and Pressurized Thermal Shock;" and GL 96-03, "Relocation of the Pressure Temperature Limit Curves and Low Temperature Overpressure Protection System Limits."

Appendix G to 10 CFR Part 50 requires that facility P-T iimits for the reactor pressure vessei (RPV) be at least as conservative as those obtained by applying the linear elastic fracture mechanics methodology of Appendix G to Section XI of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code). Appendix H to 10 CFR Part 50 establishes requirements related to facility RPV material surveillance programs. RG 1.99, Rev.

2 contains methodologies for determining the increase in transition temperature and the decrease in upper-shelf energy resulting from neutron radiation. GL 92-01, Rev. 1 requested that licensees submit the RPV data for their plants to the staff for review, and GL 92-01, Rev. 1, Supplement 1 requested that licensees provide and assess data from other licensees that could affect their RPV integrity evaluations. SRP Section 5.3.2 provides an acceptable method for determining the P-T limits for ferritic materials in the beltline of the RPV based on the ASME Code,Section XI, Appendix G methodology.

The most recent version of Appendix G to Section XI of the ASME Code which has been endorsed in 10 CFR 50.55a, and therefore by reference in 10 CFR Part 50, Appendix G, is the 2001 Edition through the 2003 Addenda of the ASME Code. The P-T limit methodology based on this edition of Appendix G to Section XI of the ASME Code (the ASME Code,Section XI, Appendix G methodology) incorporates the provisions of ASME Code Cases N-588 and N-640.

Additionally, Appendix G to 10 CFR Part 50 imposes minimum head flange temperatures when system pressure is at or above 20% of the preservice hydrostatic test pressure.

GL 96-03, "Relocation of the Pressure Temperature Limit Curves and Low Temperature Overpressure Protection System Limits," addresses the technical information necessary for a licensee's implementation of a PTLR. GL 96-03 establishes the information which must be included in: (1) an acceptable PTLR methodology (with the P-T limit methodology as its subset); and, (2) the PTLR itself. Since the licensee requested the initial implementation of a PTLR for the Oyster Creek unit, the staff's review focused on both the implementation of the Oyster Creek PTLR and the appropriate application of SIR-05-044-A, "Pressure Temperature Limits Report Methodology for Boiling Water Reactors," to generate the proposed Oyster Creek P-T limits. The related neutron fluence calculation is reviewed by the NRC's Division of Safety Systems.

3.0 TECHNICAL EVALUATION

3.1 PTLR Implementation The licensee stated in the March 10, 2008, submittal that, "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 staffs evaluations of the contents of the [Boiling Water Reactor Owners Group] methodology against the seven criteria in Attachment 1 of GL 96-03 are provided in Section 3.1 of the [safety evaluation report] [for SIR-05-044-A]." The licensee further stated that, "[t]he Pressure and Temperature Limits Report (PTLR) based on the methodology and template provided in SIR-05 044-A is being supplied for review."

- 3 The staff examined the proposed PTLR and determined that it was developed from the Template PTLR of the SIR-05-044-A report and meets the seven technical criteria:

(1) The PTLR methodology describes the transport calculation methods including computer codes and formulas used to calculate neutron f1uences (Page 3 of the Oyster Creek PTLR). The PTLR incorporates fluence projections using the RAMA fluence methodoloqy. The fiuence calculations were performed in support of the Oyster Creek license renewal and contain fluence calculations assuming 32 and 50 EFPY of operation. NUREG 1875, "Safety Evaluation Report Related to the License Renewal of Oyster Creek Generating Station," documented the NRC staff's acceptance of these fluence projections.

The PTLR provides P-T limits for 32 and 36 EFPY of exposure. The fluence calculation supporting the P-T limits for 32 EFPY are obtained directly from those performed in support of the Oyster Creek license renewal and the P-T limits for 36 EFPY are obtained by linear interpolation of the fluence values at 32 EFPY and 50 EFPY. This is an acceptable analytic approach, because any changes in operating strategy that would affect the interpolated fluence projection for 36 EFPY would also affect the fluence calculation that extends to 50 EFPY, and hence, require an updated fluence calculation.

(2) The PTLR methodology adequately describes the surveillance program (Page 36 of the Oyster Creek PTLR).

(3) The PTLR methodology adequately describes how the low temperature overpressure protection system limits are calculated applying system/thermal hydraulics and fracture mechanics (not applicable to BWRs).

(4) The PTLR methodology adequately describes the method for calculating the adjusted reference temperature (ART) values using RG 1.99, "Radiation Embrittlement of Reactor Vessel Materials," Rev. 2 (Page 4 of the Oyster Creek PTLR).

(5) The PTLR methodology adequately describes the application of fracture mechanics in the construction of P-T limits based on ASME Code,Section XI, Appendix G and the guidance in the NRC's SRP. (Page 4 of the Oyster Creek PTLR stated that the P-T limits were calculated in accordance with the SIR-05-044-A report. This description is sufficient because the SIR-05-044-A report contained detailed information regarding the application of fracture mechanics in the construction of P-T limits based on ASME Code,Section XI, Appendix G and NRC SRP. Further, the Oyster Creek PTLR provided information regarding the finite element analyses performed to generate part of the P-T limits.)

(6) The PTLR methodology adequately describes how the minimum temperature requirements in Appendix G to 10 CFR Part 50 are applied to P-T limits for boltup temperature and hydrotest temperature. (Page 4 of the Oyster Creek PTLR stated that the P-T limits were calculated in accordance with the SIR-05-044-A report. This description is sufficient because the SIR-05-044-A report contained detailed information regarding the minimum temperature requirements for boltup temperature and hydrotest temperature. Additionally, Page 5 of the Oyster Creek PTLR described the application of the minimum temperature requirement for boltup temperature.)

- 4 (7) The PTLR methodology adequately describes how the data from multiple surveillance capsules are used in the ART calculation. (Appendix A of the Oyster Creek PTLR was developed using the PTLR template in the SIR-05-044-A report which does not describe how the data from multiple surveillance capsules are used in the ART calculation.

However, the staff still accepts the Oyster Creek PTLR because this generic information is available in Appendix A of the SIR-05-044-A report.)

Given that the licensee has adequately addressed the criteria from GL 96-03, the NRC staff finds that the licensee has appropriately implemented the PTLR.

3.2 P-T limits The ART values and P-T limits valid for 32 and 36 EFPYs of facility operation using the SIR-05 044-A methodology were documented in the proposed Oyster Creek PTLR. The licensee identified the lower intermediate shell plate G-8-6 as the limiting material for the Oyster Creek RPV. The key parameters in determining the licensee's ART value for the limiting material at the one-quarter of the RPV wall thickness (1/4T) location are shown in Table 4 of the PTLR for 32 EFPYs and Table 8 for 36 EFPYs. Corresponding parameters at the three-quarter of the RPV wall thickness (3/4T) are not provided in the PTLR because the PTLR implied in Section 5, "Discussion," that the P-T limit curves based on the cooldown transient (the relevant critical location is at the 1/4T) are more conservative than the P-T limit curves based on the heatup transient (the relevant critical location is at the 3/4T).

Detailed information regarding the Oyster Creek P-T limits was contained in a calculation package (File No. OC-05Q-313), which was provided for review in the licensee's June 30, 2008, response to the staffs request for additional information. This calculation package documented the detailed thermal analyses and fracture mechanics evaluations for the RPV beltline, the bottom head, and the upper vessel, supporting the proposed "composite" Oyster Creek P-T limits valid for 32 EFPYs and 36 EFPYs. For the RPV beltline and bottom head, the ASME Code,Section XI, Appendix G methodology was used to calculate the pressure and thermal stress intensity factors (Kim and Kit), except that a stress concentration factor of 3 was applied to the Kim value in the bottom head calculation. For the upper vessel, the nozzle corner pressure and thermal hoop stresses were based on plant-specific finite element model (FEM) results for the Oyster Creek feedwater nozzle under the limiting turbine roll event. The formulas in the SIR-05-044-A report were then used to calculate its Kim and Kit values. In the final step, the calculation package utilized the applied Kit values and the plane-strain fracture toughness (Kid values at the crack tip to calculate the allowable pressure stress intensity factor (KIp) at the tip of the postulated flaw at the 1/4T location. Pressure is then obtained by comparing the Kim value based on FEM to the Kip Value.

To evaluate the proposed P-T limits for the Oyster Creek RPV, the staff confirmed the licensee's selection of the lower intermediate shell plate G-8-6 as the limiting beltline material and performed an independent calculation of the ART values for this material using the RG 1.99, Rev. 2 methodology. The staff's ART values for the limiting beltline material at the 1/4T and 3/4T locations are 166.4 of and 138.4 of for 32 EFPYs, which were calculated using materials information for Oyster Creek in the NRC Reactor Vessel Integrity Database (RVID) and the RPV inner diameter (10) fiuence in the Oyster Creek PTLR. The licensee's ART value of 165.3 of at the 1/4T location for 32 EFPYs for the limiting beltline material is close to the staff's value based on identical initial reference temperature (RT NOT) and copper (Cu) and nickel (Ni) values for the limiting material. The licensee did not calculate the ART value at the 3/4T location, which is relevant to the heatup P-T limit calculation, because the SIR-05-044-A report concluded that P

- 5 T limits for the cooldown transient are bounding. The staff will discuss this SIR-05-044-A report conclusion later.

As explained in Section 3.1.2 of this safety evaluation, the proposed Oyster Creek P-T limits are composite, representing the most limiting P-T limits for the RPV beltline, the bottom head, and the upper vessel. The staff has independently verified that, for a cooldown of 100 of per hour, the bottom head P-T limits are the least limiting. Consequentiy, oniy the RPV beltiine P-T limits (upper segment) and the upper vessel P-T limits (lower segment) are represented in the proposed Oyster Creek P-T limits (Curve B). This is also true for the proposed Oyster Creek P T limits (Curve A) for pressure test, which are obtained by setting the thermal contribution to zero. The staff performed independent calculations for both segments of the proposed P-T limits valid for 32 EFPYs. Although the proposed PTLR also contains a set of P-T limits valid for 36 EFPYs, validation of one set of them is sufficient because the methodology used to develop both sets of limits has already been evaluated during the staff's review of the SIR-05-044-A report.

For the RPV beltline P-T limit segment, the staff utilized the ASME Code,Section XI, Appendix G methodology in its independent evaluation, using the K1c curve as resistance and the pressure-dependent Kim formula and the cooldown rate dependent Kit formula as driving forces.

The staff used plant-specific information submitted by the licensee, which included the temperature measurement instrument uncertainty, the pressure measurement instrument uncertainty, and the pressure head for accounting the column of water in the RPV provided in the licensee's calculation package. The NRC staff produced almost identical beltline P-T limits for the two ends of the upper segment. The staff's calculation indicated, however, that the licensee's "Minimum Reactor Vessel Metal Temperature (degrees F)" in the proposed P-T limits is actually the metal temperature at 1/4T.

This represents a deviation from the SIR-05-044-A approach which recommended use of the coolant temperature, instead of the metal temperature at 1/4T, to evaluate K1c

• The SIR-05 044-A report further stated that "[t]he use of the coolant temperature is considered to be a necessary conservatism in P-T curve development to ensure that all design margins and safety factors are maintained." However. since this "necessary conservatism" is not required by the ASME Code,Section XI, Appendix G or by 10 CFR Part 50, Appendix G, the staff considers the licensee's deviation acceptable.

The staff's experience with prior P-T limit reviews suggested that heatup P-T limits could be limiting at certain range of the pressure and temperature. As a result, the staff also performed calculations using the ART value of 138.4 of at the 3/4T location for the limiting beltline material.

This exercise confirmed that for the Oyster Creek RPV, the cooldown P-T limits are more limiting than the heatup P-T limits by about 10°F at high pressure and 20 of at low pressure.

giving additional credibility to the proposed Oyster Creek P-T limits.

For the upper vessel P-T limit segment. the staff utilized the Kit and Kim formulas in the SIR-05 044-A report (also available in the licensee's calculation package) to calculate driving forces and the ASME Code,Section XI, Appendix G Klc curve to calculate resistance. The input nozzle corner pressure and thermal hoop stresses were based on plant-specific FEM results for the Oyster Creek feedwater nozzle under the limiting turbine roll event. The licensee explained in its June 30,2008, response that this transient is an injection of cold feedwater (100 OF) into the hot RPV (550 OF) and represents the most severe event for the feedwater nozzle. The !\IRC staff, therefore, agrees with the licensee that this event is equivalent to the limiting normal/upset design transient for a boiling water reactor (BVlJR) feedwater nozzle. Again, the staff's

-6 calculation produced almost identical P-T values for a randomly selected point along the lower segment of the proposed P-T limits.

The 10 CFR Part 50, Appendix G contains additional requirements for the minimum metal temperature of the closure head flange and vessel flange regions. These considerations were reflected in the "notch" of the upper vessel P-T limits. The staff verified that when the pressure is greater than 20% of the hydro test pressure (375 psig) the temperature for the pressure test P-T limits is greater than the RT NOT of the limiting flange material plus 90 OF (126 OF) and the temperature for the core not critical P-T limits is greater than the RT NOT of the limiting flange material plus 120 OF (156 OF). Oyster Creek pressure test P-T limits also show a 60 of straight line on the low pressure end. This was made to meet the 10 CFR Part 50, Appendix G minimum temperature requirement for pressure test which limits the operating temperature to the highest RTNOT of the closure flange that is highly stressed by the bolt preload. Since this value for the Oyster Creek RPV is 36 of, the licensee's approach is conservative.

Recently, the NRC staff has focused review efforts on whether the P-T limits for BWRs appropriately considered the small bore instrument nozzle located at the beltline region. The licensee's June 30, 2008, response to the staff's RAI concluded that, "the OCNGS instrument nozzles lie outside of the beltline, and are covered by the feedwater nozzle/upper vessel (non beltline) region." Based on this, the NRC staff determined that the beltline P-T limits and the bottom head P-T limits are not affected because the small bore instrument nozzle does not exist in these regions. For the upper head P-T limits, since the small bore instrument nozzle has a much smaller internal diameter than that of a feedwater nozzle, its stress intensity factor would be much less also according to Figure A5-1 of Welding Research Council Bulletin 175, "PVRC Recommendations on Toughness Requirements for Ferritic Materials." Hence, the staff agrees with the licensee that the feedwater nozzle bounds the small bore instrument nozzle at the upper vessel.

Based on the above evaluation, the NRC staff determined that the licensee's proposed P-T limits are in accordance with the SIR-05-044-A report and satisfy the requirements of Appendix G to Section XI of the ASME Code and Appendix G to 10 CFR Part 50. Hence, the licensee's proposed two sets of P-T limit curves are acceptable for operation of the Oyster Creek RPV valid for 32 EFPY and 36 EFPY, respectively.

3.3 Conclusion The NRC staff "finds that the licensee has properly developed a PTLR to control the development of future P-T limits under licensee controlled programs. Additionally, the NRC staff finds that the P-T limits included in the submittal have been appropriately calculated. Therefore, the NRC staff finds the proposed change to be acceptable.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the New Jersey State official were notified of the proposed issuance of the amendment. The State officials had no comments.

-7

5.0 ENVIRONMENTAL CONSIDERATION

The amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration. and there has been no public comment on such finding (73 FR 34339). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.

6.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, that: (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.

Principal Contributors: B. Parks S. Sheng Date: September 30, 2008

September 30, 2008 Mr. Charles G. Pardee Chief Nuclear Officer AmerGen Energy Company, LLC 200 Exelon Way Kennett Square, PA 19348

SUBJECT:

OYSTER CREEK NUCLEAR GENERATING STATION -ISSUANCE OF AMENDMENT RE: RELOCATION OF PRESSURE AND TEMPERATURE CURVES TO THE PRESSURE AND TEMPERATURE LIMITS REPORT (TAC NO. MD8253)

Dear Mr. Pardee:

The Commission has issued the enclosed Amendment No. 269 to Facility Operating License No. DPR-16 for the Oyster Creek Nuclear Generating Station (Oyster Creek), in response to your application dated March 10, 2008, as supplemented by letters dated June 30, 2008, and September 29, 2008.

The amendment revises the Oyster Creek Technical Specifications (TSs) 3.3, "Reactor Coolant." Specifically, the amendment relocates the pressure and temperature limit curves to the licensee controlled document, "Pressure and Temperature Limits Report," (PTLR).

Additionally, the amendment introduces supporting definitions and adds controls regarding the PTLR to Section 6.0, "Administrative Controls."

A copy of the related Safety Evaluation is also enclosed. Notice of Issuance will be included in the Commission's biweekly Federal Register notice.

Sincerely, Ira!

G. Edward Miller, Project Manager Plant Licensing Branch 1-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-219

Enclosures:

1. Amendment No. to DPR-16

2. Safety Evaluation cc w/encls: See next page DISTRIBUTION PUBLIC LPLI-2 RlF RidsAcrsAcnw_MailCTR Resource RidsNrrDirsltsb Resource RidsNrrDorlDpr Resource RidsNrrDorlLpl1-2 Resource RidsNrrDciCvib Resource RidsNrrPMGMiller Resource RidsNrrLAABaxter Resource RidsOgcRp Resource RidsRgn1 MailCenter Resource GHill,OIS RidsNrrDssSrsb Resource

• SE Input ADAMS Accession NO.: ML082390685 OFFICE LPLI-2/PM LPLI-2/LA DSS/SRXB/BC DCI/CVIB/BC DLR/RPB1/BC OGC LPLI-2/BC NAME G. E. Miller A. Baxter G. Cranston" M. Mitchell" D. Pelton MBatv H. Chernoff DATE 9/30108 9/30108 7/18/08 8/8108 9/15/08 9/22/08 9/30108 OFFICIAL RECORD COPY