ML013320464

From kanterella
Jump to navigation Jump to search

License Amendment, Phase 1 Conversion to Improved Standard Technical Specifications (TAC Nos. MB1577 and MB1579)
ML013320464
Person / Time
Site: Beaver Valley
Issue date: 01/17/2002
From: Lawrence Burkhart
NRC/NRR/DLPM/LPD1
To: Myers L
FirstEnergy Nuclear Operating Co
Burkhart L, NRR/DLPM, 415-3053
References
TAC MB1577, TAC MB1579
Download: ML013320464 (18)
v • d • e


Text

January 24, 2002 Mr. L. W. Myers Senior Vice President FirstEnergy Nuclear Operating Company Beaver Valley Power Station Post Office Box 4 Shippingport, PA 15077

SUBJECT:

BEAVER VALLEY POWER STATION, UNIT NOS. 1 AND 2 - ISSUANCE OF AMENDMENT RE: PHASE 1 CONVERSION TO IMPROVED STANDARD TECHNICAL SPECIFICATIONS (TAC NOS. MB1577 AND MB1579)

Dear Mr. Myers:

The Commission has issued the enclosed Amendment No. 246 to Facility Operating License No. DPR-66 and Amendment No. 124 to Facility Operating License No. NPF-73 for the Beaver Valley Power Station, Unit Nos. 1 and 2. These amendments consist of changes to the Technical Specifications (TSs) in response to your application dated March 28, 2001, as supplemented May 1, 2001, and June 13, 2001.

These amendments relocate certain Beaver Valley technical specifications to the Licensing Requirements Manual or to the Offsite Dosage Calculation Manual. The major change proposed in this request involves the application of the TS screening criteria of 10 CFR 50.36.

A copy of our safety evaluation is also enclosed. The Notice of Issuance will be included in the Commission's biweekly Federal Register notice.

Sincerely,

/RA GWunder for/

Lawrence J. Burkhart, Project Manager, Section 1 Project Directorate I Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket Nos. 50-334 and 50-412

Enclosures:

1. Amendment No. 246 to DPR-66
2. Amendment No. 124 to NPF-73
3. Safety Evaluation cc w/encls: See next page

January 24, 2002 Mr. L. W. Myers Senior Vice President Beaver Valley Power Station Post Office Box 4 Shippingport, PA 15077

SUBJECT:

BEAVER VALLEY POWER STATION, UNIT NOS. 1 AND 2 - ISSUANCE OF AMENDMENT RE: PHASE 1 CONVERSION TO IMPROVED STANDARD TECHNICAL SPECIFICATIONS (TAC NOS. MB1577 AND MB1579)

Dear Mr. Myers:

The Commission has issued the enclosed Amendment No. 246 to Facility Operating License No. DPR-66 and Amendment No. 124 to Facility Operating License No. NPF-73 for the Beaver Valley Power Station, Unit Nos. 1 and 2. These amendments consist of changes to the Technical Specifications (TSs) in response to your application dated March 28, 2001, as supplemented May 1, 2001, and June 13, 2001.

These amendments relocate certain Beaver Valley technical specifications to the Licensing Requirements Manual or to the Offsite Dosage Calculation Manual. The major change proposed in this request involves the application of the TS screening criteria of 10 CFR 50.36.

A copy of our safety evaluation is also enclosed. The Notice of Issuance will be included in the Commission's biweekly Federal Register notice.

Sincerely,

/RA GWunder for/

Lawrence J. Burkhart, Project Manager, Section 1 Project Directorate I Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket Nos. 50-334 and 50-412

Enclosures:

1. Amendment No. 246 to DPR-66
2. Amendment No. 124 to NPF-73
3. Safety Evaluation cc w/encls: See next page DISTRIBUTION:

PUBLIC MO'Brien ACRS PDI-1 R/F LBurkhart BPlatchek, RI EAdensam OGC PHearn JMunday GHill (4) WBeckner

  • See Previous Concurrence ACCESSION NO.: ML013320464 *SE provided 8/30/01 with no major changes OFFICE PDI-1/PM PDI-1/LA RTSB
  • OGC*w/changes PDI-1/SC(A)

NAME LBurkhart SLittle for MO'Brien WBeckner AHodgdon JMunday DATE 12/31/01 12/31/01 8/30/01 1/17/02 1/24/02 OFFICIAL RECORD COPY

Beaver Valley Power Station, Units 1 and 2 Mary OReilly, Attorney Bureau of Radiation Protection FirstEnergy Nuclear Operating Company ATTN: Larry Ryan FirstEnergy Corporation P O Box 2063 76 South Main Street Harrisburg, PA 17120 Akron, OH 44308 Mayor of the Borough of FirstEnergy Nuclear Operating Company Shippingport Regulatory Affairs Section P O Box 3 Thomas S. Cosgrove, Manager (2 Copies) Shippingport, PA 15077 Beaver Valley Power Station Post Office Box 4, BV-A Regional Administrator, Region I Shippingport, PA 15077 U.S. Nuclear Regulatory Commission 475 Allendale Road Commissioner James R. Lewis King of Prussia, PA 19406 West Virginia Division of Labor 749-B, Building No. 6 Resident Inspector Capitol Complex U.S. Nuclear Regulatory Commission Charleston, WV 25305 Post Office Box 298 Shippingport, PA 15077 Director, Utilities Department Public Utilities Commission FirstEnergy Nuclear Operating Company 180 East Broad Street Beaver Valley Power Station Columbus, OH 43266-0573 ATTN: R. E. Donnellon, Director Projects and Scheduling (BV-IPAB)

Director, Pennsylvania Emergency Post Office Box 4 Management Agency Shippingport, PA 15077 2605 Interstate Dr.

Harrisburg, PA 17110-9364 Mr. J. A. Hultz, Manager Projects & Support Services Ohio EPA-DERR FirstEnergy Corporation ATTN: Zack A. Clayton 76 South Main Street Post Office Box 1049 Akron, OH 44308 Columbus, OH 43266-0149 FirstEnergy Nuclear Operating Company Dr. Judith Johnsrud Beaver Valley Power Station National Energy Committee Mr. B. F. Sepelak Sierra Club Post Office Box 4, BV-A 433 Orlando Avenue Shippingport, PA 15077 State College, PA 16803 L. W. Pearce, Plant Manager (BV-IPAB)

FirstEnergy Nuclear Operating Company Beaver Valley Power Station Post Office Box 4 Shippingport, PA 15077

PENNSYLVANIA POWER COMPANY OHIO EDISON COMPANY FIRSTENERGY NUCLEAR OPERATING COMPANY DOCKET NO. 50-334 BEAVER VALLEY POWER STATION, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 246 License No. DPR-66

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

A. The application for amendment by FirstEnergy Nuclear Operating Company, et al. (the licensee) dated March 28, 2001, as supplemented May 1, 2001, and June 13, 2001, 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 regulations 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. 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-66 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 246, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3. This license amendment is effective as of the date of its issuance and shall be implemented within 30 days.

FOR THE NUCLEAR REGULATORY COMMISSION

/RA/

Joel T. Munday, Acting Chief, Section 1 Project Directorate I Division of Licensing Project Management Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: January 24, 2002

ATTACHMENT TO LICENSE AMENDMENT NO. 246 FACILITY OPERATING LICENSE NO. DPR-66 DOCKET NO. 50-334 Replace the following pages of Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Insert Remove Insert III III 3/4 3-36a ----

IV IV 3/4 4-3a ----

V V 3/4 4-5 3/4 4-5 VII VII 3/4 4-6 3/4 4-6 VIII VIII 3/4 4-15 3/4 4-15 X X 3/4 4-16 ----

XI XI 3/4 4-17 ----

XII XII 3/4 4-27 3/4 4-27 XIII XIII 3/4 4-28 3/4 4-28 XVII XVII 3/4 4-29a 3/4 4-29a 3/4 1-7 3/4 1-7 3/4 4-32 ----

3/4 1-8 ---- 3/4 4-33 ----

3/4 1-9 ---- 3/4 6-11 3/4 6-11 3/4 1-10 ---- 3/4 7-11 3/4 7-11 3/4 1-11 ---- 3/4 7-15 3/4 7-15 3/4 1-12 ---- 3/4 7-22 ----

3/4 1-13 ---- 3/4 7-23 ----

3/4 1-14 ---- 3/4 7-26 ----

3/4 1-15 ---- 3/4 7-27 ----

3/4 1-16 3/4 1-16 3/4 7-28 ----

3/4 1-17 ---- 3/4 7-29 ----

3/4 1-20c 3/4 1-20c 3/4 7-30 ----

3/4 1-21 3/4 1-21 3/4 7-31 ----

3/4 3-34 3/4 3-34 3/4 7-32 ----

3/4 3-34a ---- 3/4 7-34 ----

3/4 3-35 3/4 3-35 3/4 9-5 3/4 9-5 3/4 3-36 3/4 3-36 3/4 9-6 ----

3/4 9-7

PENNSYLVANIA POWER COMPANY OHIO EDISON COMPANY THE CLEVELAND ELECTRIC ILLUMINATING COMPANY THE TOLEDO EDISON COMPANY FIRSTENERGY NUCLEAR OPERATING COMPANY DOCKET NO. 50-412 BEAVER VALLEY POWER STATION, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 124 License No. NPF-73

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

A. The application for amendment by FirstEnergy Nuclear Operating Company, et al. (the licensee) dated March 28, 2001, as supplemented May 1, 2001, and June 13, 2001, 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 regulations 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. 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. NPF-73 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 124, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto are hereby incorporated in the license. FENOC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3. This license amendment is effective as of the date of its issuance and shall be implemented within 30 days.

FOR THE NUCLEAR REGULATORY COMMISSION

/RA/

Joel T. Munday, Acting Chief, Section 1 Project Directorate I Division of Licensing Project Management Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: January 24, 2002

ATTACHMENT TO LICENSE AMENDMENT NO. 124 FACILITY OPERATING LICENSE NO. NPF-73 DOCKET NO. 50-412 Replace the following pages of Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Insert Remove Insert III III 3/4 3-44 3/4 3-44 IV IV 3/4 4-5a ----

V V 3/4 4-8 3/4 4-8 VI VI 3/4 4-9 3/4 4-9 VII VII 3/4 4-24 3/4 4-24 VIII VIII 3/4 4-25 ----

IX IX 3/4 4-26 ----

X X 3/4 4-34 3/4 4-34 XI XI 3/4 4-38 3/4 4-38 XII XII 3/4 4-40 ----

XIII XIII 3/4 6-10 3/4 6-10 3/4 1-7 3/4 1-7 3/4 7-10 3/4 7-10 3/4 1-8 ---- 3/4 7-14 3/4 7-14 3/4 1-9 ---- 3/4 7-20 ----

3/4 1-10 ---- 3/4 7-21 ----

3/4 1-11 ---- 3/4 7-22 ----

3/4 1-12 ---- 3/4 7-23 ----

3/4 1-13 ---- 3/4 7-24 ----

3/4 1-14 ---- 3/4 7-25 ----

3/4 1-15 3/4 1-15 3/4 7-26 ----

3/4 1-16 ---- 3/4 7-27 ----

3/4 1-21 3/4 1-21 3/4 7-28 ----

3/4 1-22 3/4 1-22 3/4 7-29 ----

3/4 3-40 3/4 3-40 3/4 7-30 ----

3/4 3-41 3/4 3-41 3/4 9-5 3/4 9-5 3/4 3-42 3/4 3-42 3/4 9-6 ----

3/4 3-43 3/4 3-43 3/4 9-7 ----

3/4 10-5 ----

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NOS. 246 AND 124 TO FACILITY OPERATING LICENSE NOS. DPR-66 AND NPF-73 PENNSYLVANIA POWER COMPANY OHIO EDISON COMPANY THE CLEVELAND ELECTRIC ILLUMINATING COMPANY THE TOLEDO EDISON COMPANY FIRSTENERGY NUCLEAR OPERATING COMPANY BEAVER VALLEY POWER STATION, UNIT NOS. 1 AND 2 DOCKET NOS. 50-334 AND 50-412

1.0 INTRODUCTION

By letter dated March 28, 2001, as supplemented by letters dated May 1 and June 13, 2001, the FirstEnergy Nuclear Operating Company (FENOC; the licensee) submitted a request for changes to the Beaver Valley Power Station, Unit Nos. 1 and 2 (BVPS-1 and 2), Technical Specifications (TSs). The requested changes would relocate various TS requirements that do not meet the four criteria of Title 10 of the Code of Federal Regulations (10 CFR) Section 50.36 to licensee-controlled documents. This license amendment request represents the start of the BVPS-1 and 2 TS conversion to the improved standard TSs (ISTS).

The May 1 and June 13, 2001, letters provided clarifying information that did not change the initial proposed no significant hazards consideration determination or expand the scope of the initial Federal Register notice.

2.0 BACKGROUND

This amendment request proposes to revise the BVPS-1 and 2 TSs in order to implement the improvements endorsed in the Nuclear Regulatory Commission's (NRCs) "Final Policy Statement on Technical Specification Improvement for Nuclear Power Reactors," published in the Federal Register on July 22, 1993 (58 FR 39132), (the policy statement). The policy statement addresses the benefits to be derived from the ISTS and encourages licensees to use the ISTS as the basis for plant-specific TS amendments. The policy statement not only encourages licensees to use the ISTS as the basis for complete conversions, but also states that licensees may adopt portions of the STS without fully implementing all STS improvements.

The NRC policy statement endorsed the criteria for evaluating the content of the TSs. The criteria were used by the NRC and industry working groups to develop the content of the ISTS for each of the industry owner groups. The resulting ISTS applicable to Westinghouse plants are found in NUREG-1431, Title 10 of the Code of Federal Regulations Standard Technical Specifications - Westinghouse Plants.

The policy statement defines four screening criteria for determining which of the TS requirements should be retained and which could be relocated from the license. The four TS screening criteria contained in the policy statement were subsequently incorporated into 10 CFR 50.36 on July 19, 1995 (60 CFR 36953). The policy statement describes the advantages of adopting the ISTS and applying the application criteria for screening the existing TSs. The advantages noted in the policy statement include the clarification of the scope and purpose of the TSs and the enhancement of safe operation by focusing the licensees and plant operators attention on those plant conditions that are most important to safety. In addition, the policy statement also noted that these improvements to the TSs should result in more efficient use of NRC and industry resources.

The policy statement also explains that TS requirements that do not meet any of the four screening criteria may be proposed for removal from the TS and relocated to licensee-controlled documents. In NRC Administrative Letter (AL) 96-04, Efficient Adoption of Improved Standard Technical Specifications, the NRC recommends acceptable methods for relocating TS requirements that do not meet the policy statement screening criteria. AL 96-04 recommends that relocated TS requirements be moved to ....licensee-controlled documents for which there is an applicable regulatory process for future changes. Documents that are referenced in the Updated Final Safety Analysis Report (UFSAR) are identified in AL 96-04 as acceptable locations for TS requirements that do not meet the policy statement screening criteria.

This amendment request also proposed to add a TS Bases control program (Section 5.5.14 of NUREG-1431); however, the licensee requested in their June 13, 2001, letter that this part of the amendment be approved early to support another licensing request. The NRC approved this requested change on July 20, 2001, in Amendment Nos. 239 and 120 for BVPS-1 and 2, respectively.

3.0 EVALUATION The NRC staffs review of the proposed changes to the BVPS-1 and 2 TSs evaluates the compliance of these changes with the criteria of the NRC policy statement, as codified in 10 CFR 50.36, and with the ISTS for Westinghouse plants. The Final Policy Statement criteria are as follows:

Criterion 1 Installed instrumentation that is used to detect and indicate in the control room a significant abnormal degradation of the reactor coolant pressure boundary.

Criterion 2 A process variable, design feature, or operating restriction that is an initial condition of a design-basis accident (DBA) or transient analysis that either assumes the failure of or presents a challenge to fission product barrier integrity.

Criterion 3 A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to fission product barrier integrity.

Criterion 4 A structure, system, or component which operating experience or probabilistic safety assessment has shown to be significant to public health and safety.

As a result of its review, the NRC staff finds that the proposed changes to relocate the screened TSs are consistent with the screening criteria and are, therefore, acceptable. A discussion of each of these changes is included in Section 4.

In conducting the review, the NRC staff evaluated the proposed TS relocations and determined that they are also consistent with the guidance of NRC AL 96-04. FENOC proposes to relocate TSs and associated Bases that do not meet any of the four screening criteria to either the Licensing Requirements Manual (LRM) or the Offsite Dosage Calculation Manual (ODCM).

Both of these manuals are referenced in the BVPS-1 and 2 UFSARs. The NRC staff has determined that the ODCM and the LRM are acceptable documents for relocating these TS requirements since they are referenced in the UFSAR and, therefore, any changes to these relocated TSs will be controlled under 10 CFR 50.59.

3.1 Relocated TS Requirements The licensee proposes to relocate certain TS requirements from the TSs to licensee-controlled documents as noted in Table 1 of Attachment B to the March 28, 2001, letter. This table lists all specifications and specific TS details that are relocated, based on the Final Policy Statement.

The table provides the following: a TS reference number, the document to which the TS requirements will be relocated, and the method for controlling future changes to these requirements. The NRC staff evaluation of each relocated specification and specific current technical specification (CTS) detail listed in Table 1 is provided below:

3.1.1 Boration Systems TSs TS 3/4.1.2.1 Flow Paths - SHUTDOWN TS 3/4.1.2.2 FLOW PATHS - OPERATING TS 3/4.1.2.3 CHARGING PUMP - SHUTDOWN TS 3/4.1.2.4 CHARGING PUMPS - OPERATING TS 3/4.1.2.5 BORIC ACID TRANSFER PUMPS - SHUTDOWN TS 3/4.1.2.6 BORIC ACID TRANSFER PUMPS - OPERATING TS 3/4.1.2.7 BORATED WATER SOURCES - SHUTDOWN TS 3/4.1.2.8 BORATED WATER SOURCES - OPERATING The licensee proposes to relocate requirements to maintain a source of borated water, the requirements for one or more flow paths to inject this borated water into the reactor coolant system (RCS), the requirements for boric acid transfer pumps, and the requirements for appropriate charging pumps to provide the necessary charging head to overcome reactor pressure for boron injection to the LRM. The relocation of the eight specifications addressing the boration system is addressed as a group because each represents an element of the boration system and as such there are common functional requirements as well as similar

relationships to DBAs. The CTS contains redundant requirements for some components in the boration system section and the emergency core cooling system (ECCS) section. The licensee proposes to relocate the boration system requirements for the charging pumps and the refueling water storage tank (RWST) (both operating and shutdown) from the CTS and to retain the corresponding ECCS requirements in the ISTS ECCS section consistent with the applicable criteria; however, the requirements for the RWST are only contained in the BVPS boration systems TS 3.1.2.8, Borated Water Sources - Operating. This BVPS TS contains the requirements for the boric acid storage system and the RWST applicable in Modes 1-4. Since this is the only BVPS TS to contain requirements for the RWST applicable in Modes 1-4, TS 3.1.2.8 has an additional design basis ECCS that is separate from the boration systems design basis. The ECCS design basis of the RWST meets the policy statement Criterion 3 for retention in the TS. The RWST portion of TS.3.1.2.8 is not proposed for relocation and will be retained in the TS due to the RWST ECCS design-basis applicability. The RWST requirements retained in BVPS TS 3.1.2.8 effectively address the ECCS function of the RWST in a manner similar to the ISTS requirements for the RWST. The NRC staff finds the proposed relocation to be in accordance with 10 CFR 50.36 and to be acceptable.

Reactivity Control System CTS 3/4.1.3.3, "Position Indication System - Shutdown" (Units 1 and 2)

The position indication system consists of individual rod position indicators and demand indicators. The individual rod position indication system in Unit 1 is based on an analog design and the system in Unit 2 utilizes a digital design. In Modes 1 and 2, both the individual and demand position indicators are required to be operable.

The position indication system provides indication of rod position to the operator which is used by the operator to verify that the rods are correctly positioned. In operating Modes (1 and 2),

this indication is used during reactor startup and operation to monitor rod position in order to verify insertion and alignment limits are met (initial conditions of DBAs) and to verify that the rods are fully inserted into the core immediately following a reactor trip. However, in the shutdown Modes addressed by TS 3.1.3.3, "Position Indication System - Shutdown," the position indication provides information only and is not relied on by the operators to verify rod insertion or alignment or reactor trip since these functions are required only in Modes 1 and 2.

Since applicable criteria are only met in Modes 1 and 2, CTS 3/4.1.3.3, "Position Indication System - Shutdown," will be relocated from the TSs to the LRM. The NRC staff finds the relocation acceptable.

Special Test Exceptions CTS 3/4.10.5, "Position Indication System - Shutdown" (Unit 2 only)

CTS 3.10.5 allows for an exception to the operability requirements of the Unit 2 digital rod position indication system during shutdown conditions. The exception is required during rod drop time measurements since the data necessary to determine the rod drop time are derived from the induced voltage in the position indicator coils as the rod is dropped. This induced voltage is minuscule compared to the normal voltage and cannot be observed if the position indication systems remain operable. Since CTS 3/4.10.5 does not contain requirements that

are assumed in a DBA or transient, this CTS does not satisfy any of the screening criteria and is being relocated to the LRM. The NRC staff finds the proposed relocation acceptable.

CTS 3/4.3.3.1, "Monitoring Instrumentation Radiation Monitoring" The monitors proposed for relocation from the BVPS TSs are listed in Table 2 of the licensees amendment application. The selected radiation monitors consist of area and noble gas effluent monitor types. Area radiation monitors provide continuous surveillance of radiation levels in the selected areas of the plant outside the control room. The areas monitored include locations that may be occupied by operating personnel and that may be exposed to significant radiation levels. The alarms associated with the area monitors provide sufficient warning of high radiation levels and/or abnormal conditions to operating personnel. The area radiation monitors do not meet the 10 CFR 50.36 screening criteria since they are not relied upon to detect and indicate in the control room significant abnormal degradation of the reactor coolant pressure boundary. Since neither the area nor the gaseous effluent monitor TS satisfy the criteria of 10 CFR 50.36, the requirements for area monitors will be relocated to the LRM and those for effluent monitors will be relocated to the ODCM. The relocation of the radiation monitors listed in Table 2 of the licensees application is consistent with the requirements of 10 CFR 50.36 and is therefore acceptable.

CTS 3/4.4.1.4.2, "RCS LOOP ISOLATION VALVES - SHUTDOWN" (MODES 5 AND 6)

CTS 3/4.4.1.4.2 requires power to be removed from the RCS loop isolation valve operators when an RCS loop has been isolated in Modes 5 and 6. This CTS functions with two other CTSs to assure that an RCS isolation valve is not opened inadvertently. Inadvertent opening of an RCS loop isolation valve could result in a positive reactivity addition due to adding a diluted boron solution from the previously isolated RCS loop to the rector vessel. The RCS Loop Isolation Valves - Operating CTS requires power to be removed from the RCS loop isolation valves; this is similar to CTS 3/4.4.1.4.2, only the operating CTS is for Modes 1-4. The RCS Isolated Loop - Startup CTS (3.4.1.5) requires controlling the opening of RCS isolation valves that have been isolated for more than 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The RCS Isolated Loop - Startup CTS requires verification of the boron concentration in the isolated loop prior to opening the RCS isolation valve. Since these TSs overlap, resulting in redundant TSs, CTS 3/4.4.1.4.2 does not satisfy the criteria for retention in TSs and has been identified for relocation. The Startup and Operating CTS will be retained in the CTS. This arrangement results in the content of the RCS Loop CTS being consistent with the ISTS and is therefore acceptable.

CTS 3/4.4.2, "RCS Safety Valves - Shutdown" (MODES 5 AND 6)

The licensee proposes to relocate the CTS requirements for the RCS Safety Valves - Shutdown (Modes 4 and 5) to the LRM. The pressurizer safety valves protect the RCS from being pressurized above the RCS pressure safety limit. In the CTS, the pressurizer safety valves are required to be operable in order to provide overpressure protection from operating conditions (Modes 1-3) down to the RCS temperature at which the overpressure protection system residual heat removal (RHR relief valves) is required to be operable. Therefore, the requirements of CTS 3.4.3, Safety Valves - Operating (in Modes 1-3), and CTS 3.4.9.3, Overpressure Protection System, provide continuous RCS overpressure protection. As such, the CTS 3/4.4.2, Safety Valve - Shutdown, requirement for a single pressurizer safety valve to

be operable during all of Modes 4 and 5 is not required for RCS overpressure protection. In addition, the operability of a single safety valve in Modes 4 and 5 is not an assumption of any safety analysis for the mitigation of a DBA or transient in Modes 4 and 5. For these reasons, CTS 3/4.4.2 is being relocated to the LRM. The NRC staff finds the proposed relocation to be in accordance with 10 CFR 50.36 and to be acceptable.

CTS 3/4.4.7, "RCS Chemistry" The reactor coolant system chemistry limits of CTS 3/4.4.8 are proposed to be relocated to the LRM. The reactor coolant chemistry program limits primary coolant oxygen, chloride and fluoride content. This program also provides surveillance practices to monitor those elements in order to ensure that degradation of the reactor coolant pressure boundary is not exacerbated by poor chemistry conditions. Degradation of the reactor coolant pressure boundary is a long-term process; therefore, more direct means, which are controlled by regulations and TSs, are employed for this purpose. In-service inspection and primary coolant leakage limits are two more direct means that are provided to prevent long-term degradation of the reactor coolant pressure boundary materials, and provide long-term maintenance of acceptable structural conditions of the system.

CTS 3/4.4.7 does not satisfy applicable criteria since the Reactor Chemistry limitations are not of immediate importance to the operator, and are not required to ensure operability of the reactor coolant system pressure boundary. For these reasons, CTS 3/4.4.7, "RCS Chemistry,"

is being relocated to the LRM. The NRC staff finds the proposed relocation acceptable.

CTS 3/4.4.9.2, "Pressurizer Temperature Limits" The pressurizer temperature limits are proposed for relocation to the LRM. Limits are placed on pressurizer operation to prevent non-ductile failure of piping. These limits are consistent with the accepted structural analysis. Since the pressurizer normally operates in temperature ranges above those for which there is a reason for concern of non-ductile failure, temperature limits are placed on the pressurizer to assure compatibility of operation with the fatigue analysis performed in accordance with the American Society of Mechanical Engineers (ASME) Code requirements.

While these limits represent operating restrictions, Criterion 2 does not apply since the CTS operating restrictions are not restrictions that are required to preclude unanalyzed accidents and transients. A failure of pressurizer integrity would result in an analyzed event (loss-of-coolant accident (LOCA)) for which numerous systems and components are required for mitigation and are retained in the TSs. The pressurizer temperature limits are not relied on to prevent or mitigate a DBA or transient, nor are these limits an operating restriction that is required to preclude an unanalyzed accident or transient (Criterion 2). For these reasons, CTS 3/4.9.2, Pressurizer Temperature Limits, is being relocated to the LRM. The NRC staff finds the proposed relocation acceptable.

CTS 3/4.4.10, "RCS Structural Integrity" The licensee proposes to relocate the RCS structural integrity TS to the LRM. The structural Integrity CTS addresses inservice inspection and testing programs for ASME Code Class 1, 2

and 3 components that ensure the structural integrity and operational readiness of these components will be maintained at an acceptable level throughout the life of the plant. These programs provide means to prevent and detect system degradation rather than actions to mitigate unplanned events or transients. The degradation of structural integrity is a long-term process and not of immediate importance to the operator. For these reasons, CTS 3/4.4.10 does not satisfy Criterion 2 and is being relocated to the LRM. The NRC staff finds the proposed relocation acceptable.

CTS 3/4.4.12, "Reactor Coolant System Vents" (Unit 1)

Reactor Coolant System Head Vents (Unit 2)

The licensee proposes to relocate CTS 3/4.4.12 to the LRM. The vents covered by this TS exhaust non-condensable gases and/or steam from the RCS. These gases may inhibit natural circulation core cooling following any event such as a LOCA that involves a loss of offsite power and requires long-term cooling. The functional capabilities and testing requirements for reactor vessel vents are consistent with the requirements of Item II.B.l of NUREG-0737, Clarification of TMI Action Plan Requirements. The operation of these vents is an operator action after the event has occurred, and is only required when there is indication that natural circulation is not occurring; however, the operation of these vents is not assumed in any safety analysis since operation of the vents is not part of the primary success path for any design basis event. For these reasons, CTS 3/4.4.12 is being relocated from the TSs to the LRM. The NRC staff finds the proposed relocation to be in accordance with 10 CFR 50.36 and to be acceptable.

CTS 3/4.7.2, "Steam Generator Pressure/Temperature Limitation" The licensee is proposing that requirements for the steam generator pressure/temperature limits in CTS 3/4.7.2 be relocated to the LRM. These pressure and temperature limits ensure that the pressure induced stresses are within the maximum allowable fracture toughness stress limits. The values of the pressure and temperature limits are based on maintaining steam generator reference transition nil ductility temperature at a level sufficient to prevent brittle fracture. However, if failure of steam generator integrity occurs, the plant condition that results is bounded by the analysis of a steam generator tube rupture or other LOCA events for which adequate mitigation systems and components are provided. The systems and components provided to mitigate analyzed events resulting from a failure of steam generator integrity are retained in the TSs. The Steam Generator Pressure/Temperature Limitation is not an initial condition of a DBA or transient, nor is this limitation an operating restriction that is required to preclude an unanalyzed accident or transient; thus, it may be relocated to the LRM.

CTS 3/4.7.6.1, "Flood Protection" The limit on the Ohio River flood level specified in the CTS ensures that the facility operation will terminate before flood conditions threaten safety-related equipment. The specified TS limit was selected as an appropriate water level to allow adequate time to terminate plant operation and initiate flood protection measures for safety-related equipment.

Although the flood protection CTS specifies an operating condition, the flood protection operating condition does not satisfy Criterion 2 since it is not a process variable or an active

design feature (under the control of the operator). This justifies relocating CTS 3/4.7.6.1 to the LRM.

CTS 3/4.7.9.1, "Sealed Source Contamination" Existing TS 3/4.7.9.1, "Sealed Source Contamination," requires that sealed sources containing radioactive material are free of specified removable contamination, thereby ensuring that leakage from byproduct, source and special nuclear material sources will not exceed the allowable values specified in 10 CFR Part 20. The associated action statement requires that, if the removable contamination exceeds limitations, the sealed sources be either decontaminated or discarded. The limitations expressed in this TS do not impact reactor operation or the safety of reactor operations; therefore, CTS 3/4.7.9.1 does not satisfy any of the four criteria; accordingly, the requirements specified in this CTS are being relocated to the LRM. The NRC staff finds the relocation acceptable.

CTS 3/4.7.12, "Snubbers" CTS 3/4.7.12, "Snubbers," requires that all snubbers be operable. Snubbers are passive devices that are designed to prevent unrestrained pipe motion under dynamic loads and allow normal thermal expansion of piping and nozzles to eliminate excessive thermal stresses during heatup or cooldown. The TS action statement for snubbers only requires that an inoperable snubber be replaced or repaired. The surveillance requirements for snubbers are that they be periodically examined under the inservice inspection program. Since the requirements of CTS 3.7.12 that all snubbers are operable, are requirements that do not satisfy the four criteria, the CTS requirements for snubber operability are being relocated to the LRM. The NRC staff finds the relocation acceptable.

CTS 3/4.7.13.1, "Auxiliary River Water System" (Unit 1)

CTS 3/4.7.13.1, "Standby Service Water System (SWE)" (Unit 2)

CTS 3/4.7.13.1 requires the operability of Auxiliary River Water System (ARWS) for Unit 1 and the Standby Service Water System for Unit 2. These two systems are nonsafety systems and are employed as backup cooling water systems. These systems were not assumed to be available for the DBA Analysis. Since these systems do not need to be in the TS pursuant to 10 CFR 50.36, CTS 3/4.7.13.1 is being relocated to the LRM.

CTS 3/4.9.5, Communications The licensee proposes to relocate the requirements of CTS 3/4.9.5, Communications, to the LRM. This specification establishes requirements to maintain communication between the control room and the refueling station during refueling operations to ensure that refueling personnel can be promptly informed of significant changes in the plant status or core reactivity conditions observed on the control room instrumentation. This requirement represents good operational practice and is designed to ensure safe refueling operations; however, the refueling system design accident or transient response does not take credit for communications. Since these communication requirements do not satisfy the criteria of 10 CFR 50.36 for retention in TS, CTS 3/4.9.5 is being relocated to the LRM. The NRC staff finds the relocation acceptable.

CTS 3/4.9.6, "Manipulator Crane" The requirements of CTS 3/4.9.6 ensure that during refueling operations the manipulator crane and auxiliary hoist will have sufficient load handling capacity, the manipulator crane will have an appropriate overload cut off limit and the auxiliary hoist will have load indicators. Additionally, this specification ensures that the core internals and reactor vessel are protected from excessive lifting force during refueling operations in the event they are inadvertently engaged during lifting operations. Although this specification contains requirements designed to prevent damage to fuel assemblies, core internals, and the reactor vessel, these requirements are not relied upon to prevent or mitigate the consequences of a DBA. The limitations of this specification only apply to design requirements. Design control requirements are adequately governed by regulation and the required quality assurance plan; therefore, CTS 3/4.9.6 is relocated to the LRM. The NRC staff finds the relocation to be in accordance with 10 CFR 50.36 and to be acceptable.

The proposed changes to the BVPS TS that are described in the LAR provide clearer, more readily understandable requirements to ensure safe operation of the plant. The NRC staff concludes that the proposed changes satisfy 10 CFR 50.36 with regard to the content of the TS and conform to the model provided in NUREG-1431. On this basis, the NRC staff concludes that the proposed changes to the BVPS-1and 2 TSs are acceptable.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Pennsylvania State official was notified of the proposed issuance of the amendments. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendments change a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and changes surveillance requirements. The NRC staff has determined that the amendments involve 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 amendments involve no significant hazards consideration, and there has been no public comment on such finding (66 FR 33111). Accordingly, the amendments meet 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 amendments.

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 amendments will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributor: P. Hearn Date: January 24, 2002

Retrieved from "https://kanterella.com/w/index.php?title=ML013320464&oldid=2716019"