L-02-112, Supplement to License Amendment Request Nos. 295 and 167

Supplement to License Amendment Request Nos. 295 and 167

ML023430150
Person / Time
Site:	Beaver Valley
Issue date:	12/02/2002
From:	Bezilla M FirstEnergy Nuclear Operating Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
L-02-112
Download: ML023430150 (7)

Text

"I F IOC Beaver Valley Power Station Route 168

"" PO Box 4 FirstEnergyNuclear OperatingCompany Shippingport, PA 15077-0004 Mark B. Bezilla 724-682-5234 Site Vice President Fax: 724-643-8069 December 2, 2002 L-02-112 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555-0001

Subject:

Beaver Valley Power Station, Unit No. 1 and No. 2 BV-1 Docket No. 50-334, License No. DPR-66 BV-2 Docket No. 50-412, License No. NPF-73 Supplement to License Amendment Request Nos. 295 and 167 Pursuant to 10 CFR 50.90, FirstEnergy Nuclear Operating Company (FENOC) requested an amendment to the above licenses in the form of changes to the Technical Specifications. FENOC letter L-01-135, dated October 31, 2001 submitted these changes. The License Amendment Requests propose the creation of a Pressure and Temperature Limits Report (PTLR) for each unit based on the guidance provided by Generic Letter 96-03 "Relocation of the Pressure Temperature Limit Curves and Low Temperature Overpressure Protection System Limits."

The original transmittal was supplemented by FENOC letters dated December 21, 2001 (L-01-157), February 4, 2002 (L-02-008) and May 31, 2002 (L-02-058). On October 8, 2002 the NRC issued a safety evaluation approving the methodology for referencing the PTLR in the administrative section of the Beaver Valley Power Station Units 1 and 2 technical specifications. As a result of the October 8, 2002 NRC letter, FENOC has added a reference to the subject safety evaluation to the administrative section of the proposed Beaver Valley Power Station Units 1 and 2 technical specifications.

Attachment A of this letter contains the revised markups for both Beaver Valley Power Station units that incorporate the reference to the October 8, 2002 NRC safety evaluation as proposed Specification 6.9.6.b.1. Proposed Technical Specification 6.9.6.b for both BVPS units is further revised by replacing Specifications 6.9.6.b.1 and 6.9.6.b.2 submitted in L-01-135 with a single entry (Specification 6.9.6.b.2). This editorial change simplifies the proposed Technical Specification wording and is consistent with the guidance of NUREG-1431. "Standard Technical Specifications Westinghouse Plants,"

Revision 2. The pages contained in the attachment replace those submitted by FENOC letter L 135. IQN

" Beaver Valley Power Station, Unit No. 1 and No. 2 License Amendment Request Nos. 295 and 167 L-02-112 Page 2 The attached changes do not change the safety analysis or no significant hazard evaluation contained in L-01-135.

There are no new commitments made in this letter.

If there are any questions concerning this matter, please contact Mr. Larry R. Freeland, Manager, Regulatory Affairs/Performance Improvement at 724-682-5284.

I declare under penalty of perjury that the foregoing is true and correct. Executed on December c* , 2002.

Sincerely, Mark B. Bezilt*

Attachments:

A. Revised ProposedTechnical Specification Changes c: Mr. D. S. Collins, NRR Project Manager Mr. D. M. Kern, NRC Sr. Resident Inspector Mr. H. J. Miller, NRC Region I Administrator Mr. D. A. Allard, Director BRP/DEP Mr. L. E. Ryan (BRP/DEP)

L-02-112 Attachment A Beaver Valley Power Station, Unit Nos. 1 and 2 Revised Proposed Technical Specification Changes License Amendment Request Nos. 295 and 167 The following are page replacement instructions.

Beaver Valley Power Station Unit 1

• Replace page 6-19 transmitted by L-01-135 (consisting of two sheets) with page 6-20 transmitted by L-02-112 (also consisting of two sheets).

Beaver Valley Power Station Unit 2

• Replace page 6-20 transmitted by L-01-135 (consisting of two sheets) with page 6-21 transmitted by L-02-112 (also consisting of two sheets).

ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT (Continued)

Caldon, Inc. Engineering Report-160P, "Supplement to Topical Report ER-80P: Basis for a Power Uprate With the LEFM- System" Revision 0, May 2000.

c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as shutdown margin, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

6.9.6 PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR) a, Reactor Coolant System oressure and temoerature limits for heatup. cooldown, low temperature operation, criticality.

hydrostatic testing, Overpressure Protection System (OPPS) enable temperature, and Power Operated Relief Valve (PORV) lift settings as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:

1. Specification 3.4.9.1. "Reactor Coolant System Pressure/Temperature Limits". and

2. Specification 3.4.9.3. "Reactor Coolant System Overpressure Protection Systems".

b. The analytical methods used to determine the RCS pressure and temperature limits shall be those previously reviewed

• nc mrv*9 by thez NRC. sncfica11v' tbos described in the followina documents:

1. NRC letter. BEAVER VALLEY POWER STATION, UNIT NOS. 1 AND 2 - ACCEPTANCE OF METHODOLOGY FOR REFERENCING PRESSURE AND TEMPERATURE LIMITS REPORT (TAC NOS.

MB3319 AND MB3320). dated October 8. 2002.

2. WCAP-14040-NP-A. "Methodoloav Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves". Revision 2.

The methodology listed in WCAP-14040-NP-A was used with two exceptions:

a) Use of ASME Code Case N-640, "Alternative Reference Fracture Toughness for Development of P-T Limits for Section XI, Division 1". and

b) Use of methodology of the 1996 version of ASME Section XI. ADpendix G, "Fracture Toughness Criteria for Protection Against Failure".

C. The PTLR shall be Drovided to the NRC upon issuance for each reactor fluence period and for any revision or supplement thereto.

6.10 DELETED 6.11 RADIATION PROTECTION PROGRAM Procedures for personnel radiation protection shall be prepared consistent with the requirements of 10 CFR Part 20 and shall be approved, maintained and adhered to for all operations involving personnel radiation exposure.

BEAVER VALLEY - UNIT 1 6-20 Amendment No. 2-5-.

(next page is 6-23)

ADMINISTRATIVE CONTROLS REPORTING REOUIREMENTS (Continued)

c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as shutdown margin, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

6.9.6 PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

a. Reactor Coolant System pressure and temperature limits for heatup. cooldown. low temperature operation, criticality.

hydrostatic testing. Overpressure Protection System (OPPS) enable temperature, and Power Operated Relief Valve (PORV) lift settings as well as heatup and cooldown rates shall be

1. Specification 3.4.9.1, "Reactor Coolant System Pressure/Temperature Limits"- and

2. Specification 3.4.9.3, "Reactor Coolant System Overoressure Protection Systems".

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 documents:

1. NRC letter. BEAVER VALLEY- POWER STATION. UNIT NOS. 1 AND 2 - ACCEPTANCE OF METHODOLOGY FOR REFERENCING PRESSURE AND TEMPERATURE LIMITS REPORT (TAC NOS.

MB3319 AND MB3320). dated October 8. 2002.

2. WCAP-14040-NP-A. "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves", Revision 2.

The methodolocry listed in WCAP-14040-NP-A was used with two exceptions:

a) Use of ASME Code Case N-640, "Alternative Reference Fracture Touahness for Development of P-T Limits for Section XI. Division I". and b) Use of methodoloay of the 1996 version of ASME Section XI. Appendix G. "Fracture Toughness Criteria for Protection Against Failure".

c. The PTLR shall be provided to the NRC upDon issuance for each reactor fluence period and for any revision or supplement thereto.

6.10 DELETED 6.11 RADIATION PROTECTION PROGRAM Procedures for personnel radiation protection shall be prepared consistent with the requirements of 10 CFR Part 20 and shall be approved, maintained and adhered to for all operations involving personnel radiation exposure.

6.12 HIGH RADIATION AREA 6.12.1 In lieu of the "control device" or "alarm signal" required by paragraph 20.1601 of 10 CFR 20, each high radiation area in which the intensity of radiation is greater than 100 mrem/hr but less than 1000 mrem/hr shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring BEAVER VALLEY - UNIT 2 6-21 , Amendment No. 43