ML021540337

Issuance of License Amendment Control Rod Block Instrumentation

ML021540337
Person / Time
Site:	Vermont Yankee File:NorthStar Vermont Yankee icon.png
Issue date:	08/27/2002
From:	Pulsifer R NRC/NRR/DLPM/LPD1
To:	Thayer J Vermont Yankee
References
TAC MB2227
Download: ML021540337 (23)

Text

August 27, 2002 Mr. Jay K. Thayer Site Vice President - Vermont Yankee P.O. Box 0500 185 Old Ferry Road Brattleboro, VT 05302-0500

SUBJECT:

VERMONT YANKEE NUCLEAR POWER STATION - ISSUANCE OF AMENDMENT RE: CONTROL ROD BLOCK INSTRUMENTATION (TAC NO. MB2227)

Dear Mr. Thayer:

The Commission has issued the enclosed Amendment No. 211 to Facility Operating License DPR-28 for the Vermont Yankee Nuclear Power Station, in response to Vermont Yankee Nuclear Power Corporations (VYNPC) application dated June 21, 2001, as supplemented on February 8, 2002. On July 31, 2002, VYNPCs interest in the license was transferred to Entergy Nuclear Vermont Yankee, LLC (ENVY) and Entergy Nuclear Operations, Inc. (ENO).

On August 6, 2002, ENO requested that the U.S. Nuclear Regulatory Commission (NRC) continue to review and act on all requests before the Commission which had been submitted by VYNPC. Accordingly the NRC staff has acted upon the request.

The amendment revises the control rod block instrumentation requirements contained in Technical Specifications (TSs) 2.1.B, Figure 2.1.1, and Tables 3.2.5 and 4.2.5. Some of the control rod block trip functions are being relocated to the Vermont Yankee Technical Requirements Manual and some of the requirements for the retained trip functions are clarified.

Two trip functions are being added to the TSs.

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

Sincerely,

/RA/

Robert M. Pulsifer, Project Manager, Section 2 Project Directorate I Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-271

Enclosures:

1. Amendment No. 211 to License No. DPR-28

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

Vermont Yankee Nuclear Power Station cc:

Regional Administrator, Region I Mr. Raymond N. McCandless U. S. Nuclear Regulatory Commission Vermont Department of Health 475 Allendale Road Division of Occupational King of Prussia, PA 19406 and Radiological Health 108 Cherry Street Mr. David R. Lewis Burlington, VT 05402 Shaw, Pittman, Potts & Trowbridge 2300 N Street, N.W. Mr. Gautam Sen Washington, DC 20037-1128 Manager, Licensing Entergy Nuclear Vermont Yankee, LCC Ms. Christine S. Salembier, Commissioner P.O. Box 0500 Vermont Department of Public Service 185 Old Ferry Road 112 State Street Brattleboro, VT 05302-0500 Montpelier, VT 05620-2601 Resident Inspector Mr. Michael H. Dworkin, Chairman Vermont Yankee Nuclear Power Station Public Service Board U. S. Nuclear Regulatory Commission State of Vermont P.O. Box 176 112 State Street Vernon, VT 05354 Montpelier, VT 05620-2701 Director, Massachusetts Emergency Chairman, Board of Selectmen Management Agency Town of Vernon ATTN: James Muckerheide P.O. Box 116 400 Worcester Rd.

Vernon, VT 05354-0116 Framingham, MA 01702-5399 Mr. Michael Hamer Jonathan M. Block, Esq.

Operating Experience Coordinator Main Street Entergy Nuclear Vermont Yankee, LCC P. O. Box 566 P.O. Box250 Putney, VT 05346-0566 320 Governor Hunt Road Vernon, VT 05354 Mr. Michael R. Kansler Sr. Vice President and Chief Operating G. Dana Bisbee, Esq. Officer Deputy Attorney General Entergy Nuclear Operations, Inc.

33 Capitol Street Mail Stop 12A Concord, NH 03301-6937 440 Hamilton Ave.

White Plains, NY 10601 Chief, Safety Unit Office of the Attorney General One Ashburton Place, 19th Floor Boston, MA 02108 Ms. Deborah B. Katz Box 83 Shelburne Falls, MA 01370

August 27, 2002 Mr. Jay K. Thayer Site Vice President - Vermont Yankee P.O. Box 0500 185 Old Ferry Road Brattleboro, VT 05302-0500

SUBJECT:

VERMONT YANKEE NUCLEAR POWER STATION - ISSUANCE OF AMENDMENT RE: CONTROL ROD BLOCK INSTRUMENTATION (TAC NO. MB2227)

Dear Mr. Thayer:

The Commission has issued the enclosed Amendment No. 211 to Facility Operating License DPR-28 for the Vermont Yankee Nuclear Power Station, in response to Vermont Yankee Nuclear Power Corporations (VYNPC) application dated June 21, 2001, as supplemented on February 8, 2002. On July 31, 2002, VYNPCs interest in the license was transferred to Entergy Nuclear Vermont Yankee, LLC (ENVY) and Entergy Nuclear Operations, Inc. (ENO).

On August 6, 2002, ENO requested that the U.S. Nuclear Regulatory Commission (NRC) continue to review and act on all requests before the Commission which had been submitted by VYNPC. Accordingly the NRC staff has acted upon the request.

The amendment revises the control rod block instrumentation requirements contained in Technical Specifications (TSs) 2.1.B, Figure 2.1.1, and Tables 3.2.5 and 4.2.5. Some of the control rod block trip functions are being relocated to the Vermont Yankee Technical Requirements Manual and some of the requirements for the retained trip functions are clarified.

Two trip functions are being added to the TSs.

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

Sincerely,

/RA/

Robert M. Pulsifer, Project Manager, Section 2 Project Directorate I Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-271

Enclosures:

1. Amendment No. 211 to License No. DPR-28

2. Safety Evaluation cc w/encls: See next page DISTRIBUTION:

PUBLIC J. Zimmerman T. Clark R. Dennig E. Marinos PDI-2 R/F R. Pulsifer OGC ACRS S. Richards R. Caruso G. Hill (2) C. Anderson, RI Package No.: TSs:

Accession Number: ML021540337 *See Previous Concurrence OFFICE PDI-2/PM PDI-2/LA EEIB* SRXB* OGC* PDI-2/SC(A)

NAME RPulsifer TClark EMarinos RCaruso RHoefling JZimmmerman DATE 8/26/02 8/26/02 6/20/02 6/21/02 7/2/02 8/27/02 OFFICIAL RECORD COPY

ENTERGY NUCLEAR VERMONT YANKEE, LLC AND ENTERGY NUCLEAR OPERATIONS, INC.

DOCKET NO. 50-271 VERMONT YANKEE NUCLEAR POWER STATION AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 211 License No. DPR-28

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

A. The application for amendment filed by Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, Inc. (the licensees) dated June 21, 2001, as supplemented by letter dated February 8, 2002, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions 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 Commissions 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 Commissions 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 3.B of Facility Operating License No. DPR-28 is hereby amended to read as follows:

(B) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 211, 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 issuance, and shall be implemented within 90 days of issuance. The implementation of this amendment shall include the relocation of certain technical specification requirements to the Vermont Yankee Nuclear Power Station, Technical Requirements Manual as described in the licensees application dated June 21, 2001, and supplemented by letter dated February 8, 2002, and evaluated in the staffs Safety Evaluation attached to this amendment.

FOR THE NUCLEAR REGULATORY COMMISSION

/RA/

Jacob I. Zimmerman, Acting Chief, Section 2 Project Directorate I Division of Licensing Project Management Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: August 27, 2002

ATTACHMENT TO LICENSE AMENDMENT NO. 211 FACILITY OPERATING LICENSE NO. DPR-28 DOCKET NO. 50-271 Replace the following pages of the 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 8 8 9 9 11 11 16 16 20 20 51 51 52 52 69 69 74 74 77 77 78 78 80a 80a 90 90 122 122 259 259

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 211 TO FACILITY OPERATING LICENSE NO. DPR-28 ENTERGY NUCLEAR VERMONT YANKEE, LLC AND ENTERGY NUCLEAR OPERATIONS, INC.

VERMONT YANKEE NUCLEAR POWER STATION DOCKET NO. 50-271

1.0 INTRODUCTION

On June 21, 2001, Vermont Yankee Nuclear Power Corporation (VYNPC or the licensee) requested an amendment to Facility Operating License DPR-28 for the Vermont Yankee Nuclear Power Station (VY) (Reference 1). The licensee subsequently supplemented its amendment request with another submittal on February 8, 2002 (Reference 2). On July 31, 2002, VYNPCs interest in the license was transferred to Entergy Nuclear Vermont Yankee, LLC (ENVY) and Entergy Nuclear Operations, Inc. (ENO). On August 6, 2002, ENO requested that the U.S. Nuclear Regulatory Commission (NRC) continue to review and act on all requests before the Commission which had been submitted by VYNPC before the transfer. Accordingly the NRC staff has acted upon the request.

The licensee proposed to relocate Section 2.1.B, APRM Rod Block Trip Setting, of the current technical specification (CTS) to the Technical Requirements Manual (TRM) and revise the associated Figure 2.1-1, ?APRM Flow Reference Scram and APRM Rod Block Settings. The licensee also proposed to revise TS 3.1.B.a, 4.1.B, 3.6.G.1.a, and 6.6.C.4, which refer to TS 2.1.B or the flow-biased APRM rod block trip setting. In addition, the licensee proposed to: (1) change a column heading in Table 3.2.5; (2) relocate several control rod block (CRB) instrumentation trip functions from Table 3.2.5 and the associated surveillance Table 4.2.5; (3) add action statements for two rod block monitor (RBM) channels inoperable to Table 3.2.5; (4) add RBM inoperable surveillance (SR) requirements to Table 4.2.5; (5) add the reactor mode switch in shutdown position trip function to Tables 3.2.5 and 4.2.5; and (6) change and/or relocate several Bases sections to reflect the corresponding TS changes.

2.0 BACKGROUND

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

controls. However, the regulation does not specify the particular requirements to be included in a plant's TSs.

The Commission has provided guidance for the contents of TSs in its "Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors" (FPS), (58 FR 39132)

July 22, 1993, in which the Commission indicated that compliance with the FPS satisfies § 182a of the Act. In particular, the Commission indicated that certain items could be relocated from the TSs to licensee-controlled documents, consistent with the standard enunciated in Portland General Electric Co. (Trojan Nuclear Plant), ALAB-531, 9 NRC 263, 273 (1979). In that case, the Atomic Safety and Licensing Appeal Board indicated that "technical specifications are to be reserved for those matters as to which the imposition of rigid conditions or limitations upon reactor operation is deemed necessary to obviate the possibility of an abnormal situation or event giving rise to an immediate threat to the public health and safety."

Consistent with this approach, the FPS identified four criteria to be used in determining whether particular safety functions are required to be included in the TSs, 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 or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

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

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

These criteria have been codified in 10 CFR 50.36 (specifically in 10 CFR 50.36(c)(2)(ii)). See Final Rule, "Technical Specifications," 60 FR 36593, July 19, 1995. As a result, TS requirements which fall within or satisfy any of the criteria in the FPS must be retained in the TSs, while those TS requirements which do not fall within or satisfy these criteria may be relocated to licensee-controlled documents.

The Final Policy Statement provides that those existing TS LCOs which do not satisfy these four specified criteria may be relocated to the Final Safety Analysis Report (FSAR), such that future changes could be made to these provisions pursuant to 10 CFR 50.59. Other requirements may be relocated to more appropriate documents (e.g. Security Plan and Quality Assurance (QA) Plan) and controlled by the applicable regulatory requirement.

3.0 EVALUATION The FPS states that LCOs and associated requirements that do not satisfy or fall within any of the four specified criteria presently contained in 10 CFR 50.36 may be relocated from existing TSs (an NRC-controlled document) to appropriate licensee-controlled documents. The staff reviewed the control rod block instrumentation related TSs proposed for relocation from the VY CTSs against these criteria, as described below. These specifications include the LCOs, action statements, and associated surveillance requirements. The TRM is an acceptable location for these requirements because the TRM is incorporated by reference into the VY Updated Final Safety Analysis Report (UFSAR). Therefore, changes to these relocated requirements will be adequately controlled by the licensee under 10 CFR 50.59. In addition, these requirements will continue to be implemented by appropriate station procedures (i.e., operating procedures, maintenance procedures, surveillance and testing procedures, and work control procedures).

3.1 Change a Column Heading in Table 3.2.5 to Required Channels The licensee proposed to change the column heading in Table 3.2.5 from ?Minimum Number of Operable Instrument Channels per Trip System to ?Required Channels. Under this heading, Table 3.2.5 specifies the minimum number of required operable channels for each trip function.

The control rod block (CRB) logic arrangement differs from the reactor protection system (RPS) trip system arrangement. The VY CRB logic is based on ?a one!out!of!n, taken once logic.

That is, any one trip from a CRB channel results in a rod block. The licensee stated that since the logic used to generate the CRB is actually a single-trip system containing two CRB circuits, the operability requirement for the CRB instrumentation is better described as Required Channels. This is consistent with the limiting condition for operation (LCO) 3.3.2.1 and Table 3.3.2.1-1 of NUREG-1433, Standard Technical Specifications [STS], General Electric Plants, BWR/4, (Reference 3), in which the CRB instrumentation trip functions are specified in terms of required channels.

The licensee further stated that since Table 3.2.5 will specify the required channels, the required number of channels for the rod block monitor (RBM) trip function changes from one to two. Although a single channel would provide a rod block and satisfy the safety function, requiring two operable channels will provide redundancy and satisfy the single-failure-proof criterion.

The licensee also proposed changing the Trip System Logic trip function in Table 3.2.5 to Trip System, to better define the system operability, stating that there is only one required channel associated with the trip function. The licensee revised the associated Action Note 8 to state that with the number of operable channels less than the required number, place the inoperable channel in the tripped condition within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

The licensee concluded that the net effect of these proposed changes is to redefine the minimum operability requirements for the CRB instrumentation. The changes provide clarity and the operability requirements are technically equivalent before and after the change.

A trip system is an interconnected arrangement of components using instrument channel outputs, trip logics, and trip actuators to accomplish a trip function when the logic is satisfied.

NUREG-1433 refers to the corresponding column heading as Required Channels. The

objective of the column heading is to specify the minimum number of channels required to be operable for each trip function. While the RPS has two trip systems, there is only one trip system for each CRB function. Therefore, it is acceptable to change the column heading to Required Channels, as the licensee proposed. The staff also finds that the proposed changes to the Trip System Logic trip function and the associated Note 8 are acceptable, since the changes add clarity.

3.2 Relocate Some Control Rod Block Requirements from the TSs to the Technical Requirement Manual The licensee selected a number of CRB instrumentation trip functions to be relocated from the TSs to the licensee-controlled TRM, which is part of the UFSAR. The licensee stated that these CRB requirements do not meet the screening criteria of Title 10, Section 50.36(c)(2)ii of the Code of Federal Regulations (CFR) and do not need to be included in the TSs. The licensee further stated that relocating the CRB instrumentation trip functions to the TRM will not relax the basic requirements, because any subsequent revisions to the relocated requirements will be administratively governed by the 10 CFR 50.59 process. The licensee also noted that the proposed changes are consistent with the STS for boiling water reactors (BWR/4), NUREG-1433 (Reference 3).

Section 3.2.1 of this document provides the background information for each CRB trip function in Tables 3.2.5 and 4.2.5, and presents the licensees basis for relocating these trip functions from the TSs to the TRM. Section 3.2.2 discusses the licensees proposal to relocate the CRB requirements in CTS 2.1.B. Section 3.2.3 provides the staffs overall evaluation of the proposal to relocate CRB instrumentation requirements. This section also discusses the regulatory guidelines for determining which requirements need to be included in the TSs.

3.2.1 Relocate CRB Trip Function Requirements in Tables 3.2.5 and 4.2.5 Table 1.0 of this document shows the CRB trip functions that the licensee proposes to relocate from Table 3.2.5 of the CTS. In the VY TSs, the Notes specify the actions that are required if the operability requirements cannot be met.

TABLE 1.0 Control Rod Block Instrumentation Requirements to be Relocated to the Technical Requirement Manual Affected CRB Minimum # of Trip Function Modes in Which Trip Setting Proposed Instrumentation Operable Instr. Function Must Be Change Channels Per Operable (Notes 10 & 11 Trip System apply) (To Be Revised)

Source Range 2 a. Upscale (Note 2) Refuel Startup # 5 x 10 5 cps Relocate to Monitor (Note 3) TRM (SRM)

SRM 2 b. Detector Not Fully Refuel Startup Relocate to Inserted TRM Intermediate 2 a. Upscale Refuel Startup 108/125 Full Scale Relocate to Range Monitor TRM (IRM)

IRM 2 b. Downscale (Note 4) Refuel Startup 5/125 Full Scale Relocate to TRM IRM 2 c. Detector Not Fully Refuel Startup Relocate to Inserted TRM Average Power 2 a Upscale (Flow- Run ------- 0.66 (W-W) + 42% (Note 5)

Range Monitor Biased) Relocate to (APRM A-F) TRM APRM 2 b. Downscale Run -------- 2/125 Full Scale Relocate to TRM Scram Discharge 1 (Per Volume) Level High Refuel/ Run 12 gallons Relocate to Volume (SDV) Startup TRM

3.2.1.A. Background of CRB Trip Functions Source Range Monitor (SRM) Upscale and Detector-Not-Fully-Inserted Control Rod Block Trip Functions The SRMs are considered to be non-safety-related, but were retained in the TSs because the SRMs provide the sole on-scale neutron monitoring during shutdown, refueling, and startup conditions. The only SRM scram trip function is SRM high-high, which occurs only if the shorting links (which place the SRMs into the RPS circuitry) are removed. The SRM high-high trip function also provide indications and alarms to alert the operators.

The licensee proposed to relocate the SRM upscale and detector-not-fully-inserted CRB trip functions. The SRM upscale rod block ensures that SRM detectors are retracted in order to minimize their exposure and prolong the life of the SRMs. The rod block setting is selected at the upper end of the SRM detection range. With an SRM count rate of less than 100 counts per second (cps), the SRM detector-not-fully-inserted rod block ensures that no control rods are withdrawn unless all SRM detectors are properly inserted (to provide neutron flux monitoring and reactor period indications until the SRM/Intermediate Range Monitor (IRM) overlap range is reached). The alarm function warns the operators, and the control rod withdrawal blocks prevent further increase in the core reactivity when adequate on-scale neutron monitoring is not available. Action Note 7 in Table 4.1.2, Scram Instrument Calibration, of the VY TSs states that IRM and SRM channels shall be determined to overlap during startup after entering the STARTUP/HOT STANDBY MODE and the IRM and Average Power Range Monitor (APRM) channels shall be determined to overlap during each controlled shutdown, if not performed within the previous 7 days. Thus, Action Note 7 ensures that sufficient SRM/IRM and IRM/APRM overlap is achieved before the licensee retracts the SRMs or IRMs. Therefore, even if the detector-not-fully!inserted CRB is relocated to the UFSAR (the TRM), the TS will retain the requirement that there will be sufficient overlap.

Intermediate Range Monitor Upscale, Downscale, and Detector-Not-Fully-Inserted CRB Trip Functions The IRMs are safety-related and provide neutron flux monitoring between the source range and power range during reactor startup and shutdown. They also provide protective trips during refueling. The IRM upscale rod blocks ensure that core reactivity does not increase when the neutron flux is low, unless the IRMs are properly ranged up or the IRM upscale high flux level will result in half scram. Therefore, during low neutron flux operation, the IRM upscale rod block also stops a further rod withdrawal error in time to prevent a scram. The IRM downscale rod block function (in effect at the lower power range) also prevents control rod withdrawal and the associated increase in reactivity when the IRMs are ranged above the existing reactor neutron flux. The IRM downscale rod block prevents continuation of the reactor startup if the operator ranges the IRM above the appropriate monitoring scale. If the reactor mode switch is in the startup or refueling position and the IRMs are not fully inserted, the detector-not-fully-inserted rod block ensures that no control rod can be withdrawn without proper neutron monitoring capability. Again, Note 7 of Table 4.1.2 ensures sufficient IRM/APRM overlap protection.

Average Power Range Monitor Upscale (Flow-Biased) and Downscale CRB Trip Functions The APRMs provide continuous indications of core average power, and the APRM trip setpoints are flow-biased up to the rated thermal power, with a fixed maximum APRM scram at 120 percent of rated thermal power. The licensee proposed to relocate the APRM upscale and downscale rod blocks to the TRM. The flow-biased APRM upscale rod block inhibits control rod withdrawals at off-rated conditions that may lead to operation above the upper boundary of the licensed rod line. While the flow-biased APRM rod blocks limit the degree to which power can increase above the upper boundary of the licensed rod line with control rod movements, the flow-biased APRM scram protects against core-wide transients that may endanger the fuel cladding integrity or the reactor coolant pressure boundary.

With the mode switch set to run, the APRM downscale rod block stops control rod withdrawal during power range operation unless the APRM channels are operating properly or are bypassed correctly. During reactor operation in the run mode, all APRMs that are not bypassed must be on scale. VY Bases 3.1, Reactor Protection System, states that each protection trip system has one more APRM than is necessary to meet the minimum number required. This allows the bypassing of one APRM per protection trip system for maintenance, testing, or calibration without changing the minimum number of channels required for inputs to each trip system. In addition, the APRM inoperable trip function in Table 3.1.1, Reactor Protection System (SCRAM) Instrumentation Requirements, requires two operable instrument channels per trip system.

Scram Discharge Volume Level High CRB Trip Function The scram discharge volume (SDV) water level CRB function prevents control rod withdrawal and indicates to the reactor operator that water is accumulating in the SDV. The SDV water level is monitored to ensure that the available volume can contain the water discharged by the control rod drives during a scram. Further water accumulation in the SDV water level up to the scram setpoint will result in a scram and the SDV water-level-high CRB trip function gives operators an opportunity to take appropriate action before a scram.

3.2.1.B. Licensees Justification The licensee proposed to remove the SRM (upscale, detector-not-fully-inserted), IRM (upscale, downscale and detector-not-fully inserted), and the APRM (upscale and downscale) rod block trip functions from Tables 3.2.5 and 4.2.5, and relocate CRB instrumentation trip functions, with their respective operability and surveillance requirements to the VY TRM. The Action Notes (1, 2, 3, 4, and 11 for Table 3.2.5 and 6 for Table 4.2.5) for these CRB instrumentation trip functions would also be relocated to the TRM. However, Action Note 10 of Table 3.2.5 will be retained in the TSs, since the action statement also applies to the RBM. The licensee would also (1) delete from Bases 3.2 discussions on the relocated CRB trip functions, and (2) replace the phrase has no scram function, in the SRM discussion in Note 5 of Bases 3.3 with provides a scram function in noncoincident configuration.

The licensee evaluated the safety function of the CRB instrumentation against the screening criteria 1, 2, 3, and 4 of 10 CFR 50.36(c)(2)ii. The licensee stated that the SRM, IRM, and APRM CRB instrumentation (1) are neither used for, nor capable of, detecting a significant abnormal degradation of the reactor coolant pressure boundary (RCPB) before a design-basis

accident (DBA), (2) do not monitor a process variable that is an initial condition of a DBA or transient, and (3) are not part of a primary success path in the mitigation of a DBA or a transient. The licensee stated that these CRB trip functions are also not a significant contributor to core damage frequency (CDF) and offsite dose, according to NEDO-31466, Technical Specification Screening Criteria Application and Risk Assessment, (Reference 4). Therefore, these CRB trip functions also do not qualify under Criteria 4 and can be removed from the TSs without affecting safety.

Similarly, the licensee stated that the SDV water-level-high CRB trip function is not credited in any DBA or transient analysis and that does not meet criteria 1, 2, or 3 of 10 CFR 50.36(c)(2)ii. This trip function is also not a significant contributor to core damage frequency and offsite dose, according to Reference 4. Therefore, the SDV water-level-high CRB trip function does not qualify under Criteria 4 and can be removed from the TSs without affecting safety.

3.2.2 Relocate Limiting CTS 2.1.B, APRM Rod Block Trip Setting Since the licensee proposed relocating the flow-biased APRM instrumentation requirements from Table 3.2.5 to the TRM, the licensee also proposed to relocate the corresponding APRM trip setting specified in CTS 2.1.B from the TS to the TRM.

3.2.2.A. Background for APRM Flow Reference Scram and APRM Rod Block Settings CTS 2.1.B and Figure 2.1.1, APRM Flow Reference Scram and APRM Rod Block Settings, establish the flow-biased APRM rod block setting as a percent of the rated thermal power (shown in Table 1.0 of this safety evaluation). The licensee proposed to relocating the APRM rod block requirement in CTS 2.1.B to the TRM and deleting the flow-biased APRM trip setpoint values shown in Figure 2.1-1. CTS 2.1.B specifies the APRM gain for the flow-biased rod block setting during off-rated power and flow conditions for operation at certain combinations of maximum fraction of limiting power density (MFLPD) and core thermal power. The corresponding VY TS Bases 2.1.B states that the slope of the flow-biased APRM rod block trip response setting is adjusted to track the required trip setting with changes in the recirculation flow. This allows an effective rod block if the core average power increases above the power level specified for a given flow rate. Therefore, the APRM flow-biased rod blocks limit inadvertent operation above the licensed rod line during plant maneuvers or transients.

However, the APRM flow-biased scram trip setpoint is also adjusted, which ensures that a scram will occur before the thermal-mechanical overpower conditions are reached. The RBM is designed to prevent an erroneous rod withdrawal leading to local fuel damage, and the APRM flow-biased CRB provides additional protection.

3.2.2.B Licensees Justification The licensee stated that the APRM rod block requirements in CTS 2.1.B do not meet the screening criteria of 10 CFR 50.36(c)(2)ii for inclusion in the TSs. The submittal compared the CRB function to the four screening criteria and concluded that the APRM CRB requirements can be relocated to the TRM with no negative impact on plant safety.

The licensee also proposed deleting references to CTS 2.1.B and Table 3.2.5 in CTS 3.1.B.a, 4.1.B, 3.6.G.1.a, and 6.6.C.4. The discussions on the APRM flow-biased rod block in Bases 2.1.B would also be deleted.

3.2.3 Staffs Evaluation of the Relocation of the CRB Requirements The proposed amendment would not eliminate the CRB requirements or disable the CRB instrumentation. Instead, the CRB requirements will be relocated to the licensee-controlled TRM, where changes are governed by the 10 CFR 50.59 process. Therefore, the proposed changes would be consistent with the Commissions policy of streamlining the TSs and retaining only those requirements that are necessary to maintain the safety of operating nuclear reactors.

The Commissions Interim Policy Statement on Technical Specification Improvements for Nuclear Power Reactors, (published in the Federal Register (52 FR 3788) on February 6, 1987) (Reference 5), established the screening criteria for deciding the content of TSs. The focus of the screening criteria was operating parameters and associated surveillance requirements that (1) are significant to safety, (2) directly contribute to the prevention of accidents, or (3) mitigate the consequences of an accident. In the May 1988 Split Report (Reference 6), the staff reviewed the Westinghouse, BWR, and Combustion Engineering owners groups recommendations. Appendix D to that report listed the regulatory requirements and the operating restrictions that should be maintained in the TSs and those that could be relocated into licensee-controlled documents. The staffs review focused on those LCOs that the owners groups recommended to be relocated and verified that none included parameters or functions that were important to safety. BWR-Table 2, General Electric Standard Technical Specification LCOs Which May Be Relocated, of Reference 6 listed the following CRB instrumentation for possible relocation.

Table 2 LCO Report Control Rod Block Instrumentation Plant**

Item 3.3.6 135 APRM H, GG 137 SRM H 138 IRM H,GG 139 SDV Water Level GG

• • These are the plants that applied the screening criteria to their TS and proposed the changes. H stands for Hatch (BWR/4) and GG stands for Grand Gulf (BWR/6).

In Note 1 (associated with the BWR-Table 2), the NRC staff stated that the LCOs listed in this table may be relocated to other licensee-controlled document contingent upon NRC staff approval of the relocation of and controls over [the] relocated requirement.

The FPS (Reference 7), published on July 22, 1993, gave further guidance on applying screening criteria 1, 2, and 3. The FPS considered experiences and lessons learned since the publication of the interim policy, and developed further on criteria 4 (using plant-specific PRA, risk survey, and any available risk insights or PRA literature.) The Commission also stated that the screening criteria may be applied to both the STSs and custom TSs. In July 1995, the screening criteria were codified in the 10 CFR 50.36 rulemaking, but the Commission kept the FPS explaining that it contains additional insights and discussions that would be useful to the staff and industry in applying the TS screening criteria.

The licensee does not intend to convert to the STS at this time, but may, as the Final Policy Statement states, apply the screening criteria to amend the custom TSs. According to the 10 CFR 50.36 rulemaking, the requirements that do not qualify under the 10 CFR 50.36 screening criteria may be relocated to the (1) FSAR, (2) operating procedures, or (3) quality assurance plan. Subsequent changes to these requirements would not require a license amendment or prior NRC approval, but can be made by the licensee under the control of the 10 CFR 50.59 process. Therefore, VYNPC may relocate any requirements that do not meet the TS screening criteria to the TRM. In addition, the CRB instrumentation functions that VYNPC proposes to relocate are included in the Split Report list of CRB instrumentation approved for relocation (with certain provisions).

The staff evaluated the licensees proposal to relocate the SRM (upscale, detector-not-fully-inserted), IRM (upscale, downscale, detector-not-fully-inserted), APRM (flow-biased upscale, downscale), and SDV (water-level-high) CRB trip functions and the requirements in CTS 2.1.B to the TRM. These CRB functions are not used to detect, or indicate in the control room, abnormal degradation of the reactor coolant pressure boundary. Although these CRB functions help the operators avoid scrams, they are not used as an initial condition to any design-basis analysis that postulates the failure of, or challenge to, the integrity of a fission product barrier.

The associated scram functions are a part of the primary success path which functions or actuates to mitigate a DBA analysis. Lastly, there are no available risk insights that indicate that these CRB functions and the requirements in CTS 2.1.B limit the likelihood or severity of an accident. Therefore, the staff accepts that the proposed CRB trip functions and the associated requirements do not meet the requirements in 10 CFR 50.36(c)(2)ii.

The staff does consider the CRB functions to be important operating features in the operation of BWRs. These CRB functions prevent operator actions that might lead to an adverse operating condition. In addition, the SRM, IRM, and APRM rod blocks alert the operator by preventing control rod withdrawals during startup when the appropriate neutron monitoring is not available. However, it is the staffs understanding that these CRB trip functions will remain operable and the licensee is relocating to the TRM to streamline the TSs. During the development of the STS, these CRB trip functions and requirements were not included in the STS. Therefore, the proposed changes are consistent with the STS. The staff accepts the relocation of the CRB functions listed in Table 2 of this document and the requirements in CTS 2.1.B from the VY TSs to the licensee-controlled TRM.

The staff also approves the relocation of (1) SR requirements in Table 4.2.5; (2) Action Notes 1, 2, 3, 4, and 11 in Table 3.2.5; and (3) Action Notes 6 in Table 4.2.5 and Figure 2.1-1. Since the CRB trip functions would be relocated, it is acceptable to relocate the corresponding Action Notes, SR requirements and trip setpoint plot. In addition, the staff finds the proposed changes to CTS 3.1.B.a, 4.1.B, 3.6.G.1.a, and 6.6.C.4 acceptable, because the licensee would delete

references to the relocated CTS 2.1.B requirements. The staff also has no objection to the proposed changes to Bases 2.1 and 3.2 (deleting discussions on the relocated CRB requirements) and item 5 in Bases 3.3, which adds a clarifying phrase.

3.3 Add Rod Block Monitor Channels Inoperable Requirements to Table 3.2.5 and 4.2.5 3.3.1. Background for RBM Inoperable Requirements The rod block monitor is designed to automatically prevent fuel damage in the event of an erroneous rod withdrawal. The rod withdrawal error (RWE) analysis assumes the operator makes a procedural error and withdraws the maximum-worth control rod to its rod block position. Because of the resulting reactivity addition, an RWE causes an increase in the core average power and local power (power near the vicinity of the control rod). The RBM monitor initially sounds an alarm, and the operator must acknowledge the alarm and take corrective action. If rod withdrawal is from a limiting rod pattern, the RBM system blocks further rod withdrawals before the fuel cladding is damaged. The RBM uses readings from the local power range monitors (LPRM) near the control rod selected for withdrawal to compute the trip setpoints. The VY RBM trip setpoint is flow-biased with the trip setpoint clamped at 100 percent of the rated core flow. The clamp setpoint is cycle-specific and reported in the core operating limit report (COLR).

The RBM must be operable above 30 percent of rated thermal power. VY TS Bases 3.3.B.6 defines the limiting rod pattern, and the licensee proposed to add the following clarifications to the Bases, During reactor operation with certain limiting control rod patterns, the continuous withdrawal of a designated single control rod could result in a violation of the MCPR [minimum critical power ratio] safety limit or the 1-percent plastic strain limit. A limiting control rod pattern is a pattern which results in the core being on a thermal limit (i.e., operating on a limiting value for APLHGR [average planar LHGR) , LHGR [linear heat generation rate], or MCPR).

3.3.2. Licensees Justification The licensee proposed to add an action statement for the condition where two RBM channels are inoperable in Table 3.2.5. The licensee would also add an RBM inoperable trip function surveillance requirement to Table 4.2.5, which would require a channel function test of the RBM trip function every 3 months. The licensee stated that calibration is not applicable to the RBM inoperable function. The existing action statements in Notes 7, 9, and 10 of Table 3.2.5 would continue to apply to the RBM inoperable trip function.

In addition, the licensee would add the clarifying statement, rated thermal power, to Note 7 of Table 3.2.5. The current TS does not specify any requirements to ensure that the RBM upscale (flow-biased) is not bypassed when power is greater than 30 percent. Consequently, the licensee would add Note 13 under the calibration heading of Table 4.2.5, stating, Includes calibration of the RBM Reference Downscale function (i.e., RBM upscale function is not bypassed when >30% Rated Thermal Power.) The licensee stated that calibrating the RBM upscale trip function every 3 months will verify that the trip function is not bypassed for operation above 30 percent of rated thermal power. Note 9 in the CTS specifies the required action statement when one RBM channel is inoperable. The licensee proposed to revise Note 9 to specify the action required when two RBM channels are inoperable. Since the licensee

proposed to change the column heading of Table 3.2.5 from Minimum Number of Operable Instrument Channels per Trip System to the Required Channels, the number of required channels for the RBM has changed from 1 to 2 channels. The following table shows the proposed RBM trip function changes.

Table 3.2.5 Affected Minimum # Trip Function Modes in Trip Setting Proposed CRB Operable Instr. Which Change Function Channels Per Function Trip System Must Be (revised)1 Operable Rod Block 1 (change to 2) a. Upscale run # 0.66(W-W) + change Monitor (Flow! Biased) N - with max as Note 9 (Note 9) (Note 7) defined in COLR RBM (A/B) (Note 5)

( Notes 10 1 (change to 2) b. Downscale run $ 2/125 Full- change and 9) (Note 7) Scale Note 9 2 c. Inop. run N/A Add to the (Note 7) table The existing Action Note 9 requires that with one RBM channel inoperable, the licensee must take the following actions:

a. Verify that the reactor is not operating on a limiting control rod pattern, and

b. Restore the inoperable RBM channel to operable status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Otherwise, place the inoperable rod block monitor channel in the tripped condition within the next hour.

In the supplement submitted on February 8, 2002, the licensee proposed to revise required actions in Note 9 as follows:

With one or two RBM channels inoperable,

a. Verify that the reactor is not operating on a limiting control rod pattern (as described in the Bases of specification 3.3.B.6); and

b. If one RBM is inoperable, restore the inoperable channel to operable status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />; and 1

Heading changed to Required Channels, as discussed in change 3.1 in this document.

c. If the required actions and associated completion times of Notes 9.a and 9.b (above) are not met, or if two RBM channels are inoperable, place one RBM channel in the tripped condition within the next hour.

The licensee stated that when one RBM channel is inoperable, the required action statements and conditions remain unchanged, except for Action 9.c. The proposed Action 9.c. allows either of the two inoperable RBM channels to be placed in the tripped condition within the next hour, instead of requiring the inoperable channel to be tripped within the next hour. Placing the inoperable channel in the tripped condition causes a control rod block, thus providing equivalent protection to the RBM.

Since the VY RBM trip function consists of two channels, either of which will provide a rod block, the current requirements in Table 3.2.5 do not address the condition when two RBM channels are inoperable. The licensee stated that since there is no TS-allowed action for two inoperable RBM channels, the following requirement in 10 CFR 50.36 become applicable:

When a limiting condition for operation of a nuclear reactor is not met, the license shall shutdown the reactor or follow any remedial action permitted by the Technical Specifications until the condition can be met.

Since no remedial action is specified in the TS for two inoperable RBM channels, the unit must exit the applicable condition specified in the TS for the RBM channels. Therefore, the licensee would have to reduce the reactor power to no more than 30 percent of the rated thermal power (the power level at which the RBM is not required to be operable). The licensee stated that placing one of the two inoperable RBM channels in the tripped condition within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> would provide adequate protection since it would result in a control rod block.

The licensee stated that referencing Bases 3.3.B.6 in Action 9.a alerts the control room operator to the meaning of a limiting control rod pattern, increasing the operators understanding of the action statement. The licensee proposed to also revise Bases 3.3.B.6 to define a limiting control rod pattern. This is discussed under the RBM trip function background, above.

3.3.3 Staffs Evaluation of the RBM Inoperable Requirements As discussed in the background, the RBM prevents a continuous RWE at power resulting in fuel damage. The CTS, Note 9a and b allow one RBM channel to be inoperable for up to 25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br />, even if the reactor is operating at a limiting control rod pattern. However, the redundancy and logic of the RBM systems ensure that the operable RBM channel will monitor the local reactivity changes. With two RBM channels inoperable, the proposed action statement would allow no more than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of operation with no RBM protection.

LCO 3.3.2.1, Control Rod Block Instrumentation, of the STS in NUREG-1433, requires that if one RBM channel is inoperable, that inoperable channel must be restored to operable condition in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. If the required action cannot be met or two RBM channels are inoperable, one RBM channel must be placed in its tripped position in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. Therefore, the proposed Action Note 9.c parallels the requirements in the STS. However, VY Action Note 9a requires the additional step of verifying whether the reactor is operating with a limiting control rod pattern.

This action ensures that the operator understands the impact of the reduced margin to the

thermal limits with one or two RBM channels inoperable. Although Note 9.a does not have specific operator action statements associated with it, if the reactor is operating at a limited rod pattern (close to the thermal limits), the operator would be aware of the potential for exceeding the thermal limits with further control rod withdrawals. In addition, the STS Bases B 3.3.2.1 explains the reasoning behind the 1-hour completion time, stating that it is intended to give the operator time to evaluate and repair the inoperable channels. The 1-hour completion time minimizes risk while allowing time to restore the inoperable channel(s) to operable status or trip the inoperable channel(s).

The staff accepts the licensees proposal to add a two RBM inoperable function to Action Note 9 in Table 3.2.5, because the proposed action statements are more conservative than the CTS and are consistent with the requirements specified in LCO 3.3.2.1 of the STS. These changes add further SR requirements and some clarifying statements to the TSs. The staff also accepts the proposed SR changes in Table 4.2.5 and has no objection to the corresponding Bases changes.

3.4 Add Reactor Mode Switch in Shutdown Position Requirement in Table 3.2.5 The licensee proposed to add Reactor Mode Switch-Shutdown Position operability and surveillance requirements to Tables 3.2.5 and 4.2.5, as shown below.

Table 3.2.5 Required Channels Trip Function 2 (Note 13) Reactor Modes Switch - Shutdown Position (Note 12)

Note 12: Required to be operable when the reactor mode switch is in shutdown position.

Note 13: With one or more Reactor Mode Switch-Shutdown Position channels inoperable, immediately suspend control rod withdrawal and immediately initiate action to fully insert all insertable control rods in core cells containing one or more fuel assemblies.

Table 4.2.5 Trip Function Functional Test Reactor Modes Switch - Shutdown Position Every Refueling (Note 12)

Note 12: Trip system logic testing is not applicable to this function. If the required surveillance frequency (every Refueling Outage) is not met, functional testing of the Reactor Mode Switch-Shutdown Position function shall be initiated within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after the reactor mode switch is placed in Shutdown for the purpose of commencing a scheduled Refueling Outage.

The licensee stated that the CTS does not specify any requirements for the reactor mode switch in the shutdown position. No reactivity initiated events are analyzed, with the reactor mode switch in the shutdown position, because it is assumed that the CRB would prevent

control rod withdrawals. The proposed reactor mode switch in shutdown position CRB trip function prevents an inadvertent criticality and makes the shutdown CRB assumptions valid.

When the reactor mode switch is placed in the shutdown position, it initiates a reactor scram, enables the IRM high-high flux, and selects the APRM high-high 15-percent flux setpoint. The VY TS defines shutdown as, the reactor is in a shutdown condition when the reactor mode switch is in the shutdown mode position and no core alterations are being performed. When the mode switch is placed in the shutdown position a reactor scram is initiated, power to the control rod drives is removed, and the reactor protection system trip systems are de-energized.

Since the Reactor Mode Switch in Shutdown Position initiates a CRB, Note 13 of Table 3.2.5 ensures that control rods are not withdrawn with the channels inoperable. In addition, the licensee proposed to revise Bases 3.2 to add a discussion concerning the Reactor Mode Switch-Shutdown Position and the staff finds it acceptable. Since the proposed changes to Tables 3.2.5 are more restrictive changes and provide adequate protection, the staff finds the proposed changes acceptable.

The licensee also proposed to perform reactor mode switch-shutdown position functional testing every refueling outage. Note 12 of Table 4.2.5 states that the trip system logic testing is not applicable to this function. This is because there is no trip setting associated with the reactor mode switch in the shutdown position, since the channels are mechanically actuated by the reactor mode switch position. Note 12 of Table 4.2.5 also specifies the required action if the SR frequency is not met. In Bases 4.2, the licensee provided the justifications for the proposed SR requirements and frequency, stating that operating experience shows that surveillance once per operating cycle during refueling is adequate to ensure functional operability. The frequency is based on the need to perform this SR under the applicable condition. In general, the channel functional testing for the reactor mode switch in the shutdown position is performed by attempting to withdraw any control rod and verifying that a CRB occurs. Note 12 states that if the frequency is not met, functional testing can be performed 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after placing the Reactor Mode Switch in the Shutdown Position. The 1-hour timeframe permits entry into the Mode Switch-Shutdown Position condition in order to perform the functional test. Performing the channel functional test in any other Mode would require using jumpers, or other means to bypass some of the active trip functions. The proposed Reactor Mode Switch-Shutdown Position SR requirement is consistent with the STS (NUREG-1433), and the staff finds it acceptable.

3.4 Staff Review Results The staff evaluated the licensees proposal to (1) relocate the CTS 2.1.B, APRM Rod Block Trip Setting, (2) revise the associated Figure 2.1-1, ?APRM Flow Reference Scram and APRM Rod Block Settings; (3) delete references to CTS 2.1.B in CTS 3.1.B.A, 4.1.B, 3.6.G.1.a, and 6.6.C.4, (4) relocate CRB instrumentation trip functions and the associated SR requirements from Tables 3.2.5 and Table 4.2.5; (5) add required actions for two inoperable RBM channels to Tables 3.2.5 and 4.2.5; (6) add operability and SR requirements for Reactor Mode Switch-Shutdown Position; and (7) change and/or relocate several Bases sections to reflect the corresponding TS changes. On the basis of its review, the staff finds the proposed changes acceptable.

4.0 STATE CONSULTATION

In accordance with the Commissions regulations, the Vermont State official was notified of the proposed issuance of the amendment. The State official had no comments.

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 and changes surveillance requirements. The NRC staff has determined that the amendment involves no significant increase in 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 (67 FR 12608). 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 Commissions 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.

7.0 REFERENCES

1. Vermont Yankee Nuclear Power Corporation, Technical Specification - Proposed Change No. 247, Control Rod Block Instrumentation, June 21, 2001.

2. Vermont Yankee Nuclear Power Corporation, Technical Specification - Proposed Change No. 247, Control Rod Block Instrumentation-Supplement No. 1, February 8, 2002.

3. Nuclear Regulatory Commission, Standard Technical Specifications, General Electric Plants, BWR/4, NUREG-1433.

4. General Electric, Technical Specification Screening Criteria Application and Risk Assessment, NEDO 31466, November 1987.

5. Nuclear Regulatory Commission, Interim Policy Statement on Technical Specification Improvements for Nuclear Power Reactors, 52 FR 3788, February 6, 1987.

6. Nuclear Regulatory Commission, NRC Staff Review of Nuclear Steam Supply System Vendor Owners Groups Application of The Commissions Interim Policy Statement Criteria to Standard Technical Specifications, (Split Report), May 9, 1988.

7. Nuclear Regulatory Commission, Final Policy Statement on Technical Specification Improvement for Nuclear Power Reactors, July 22, 1993 (57 FR 39132)

Principal Contributor: Z. Abdullahi Date: August 27, 2002