Text

For Operating License DPR-43 License Amendment Request 185 to the Technical Specifications, Core Operating Limits Report Implementation
	ML022200080
Person / Time
Site:	Kewaunee
Issue date:	07/26/2002
From:	Warner M; Nuclear Management Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	GL-88-016, NRC-02-064
	Download: ML022200080 (170)
	v • d • e

Kewaunee Nuclear Power Plant Point Beach Nuclear Plant N490 Highway 42 6610 Nuclear Road Kewaunee, Wl 54216-9511 Two Rivers, Wl 54241 NMC 920.388.2560 920.755.2321 Committed to Nuclear Excell Kewaunee / Point Beach Nuclear Operated by Nuclear Management Company, LLC NRC-02-064 1 OCFR50.90 July 26, 2002 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555 Ladies/Gentlemen:

Docket 50-305 Operating License DPR-43 Kewaunee Nuclear Power Plant License Amendment Request 185 To The Kewaunee Nuclear Power Plant Technical Specifications, "Core Operating Limits Report Implementation."

In accordance with the requirements of 10 CFR 50.4 and 10 CFR 50.90, Nuclear Management Company, LLC.,

hereby requests amendments to Facility Operating License DPR-43 for Kewaunee Nuclear Plant. The purpose of this license amendment request (LAR) is to implement a Core Operating Limits Report (COLR) at the Kewaunee Nuclear Plant (KNPP). NMC request approval of this LAR in February of 2003 with 60 days to implement.

NRC Generic Letter 88-16 provides guidance for licensees allowing relocation of cycle dependent variables from the Technical Specifications (TS) to a formal report, COLR.

This approval requires that the values of these variables included in a COLR, are determined with NRC-approved methodologies referenced in the TS, and that changes to these variable values are reported to the NRC as they are made. The variables removed from the TS and relocated to the COLR can then be changed via the appropriate regulatory change mechanisms, principally 10 CFR 50.59. This avoids the need to frequently change TS, reducing unnecessary burden on the licensee and NRC staff.

Cycle specific parameters approved for Westinghouse units for relocation to the COLR include:

(1) moderator temperature coefficient, (2) shutdown bank insertion limits, (3) control bank insertion limits, (4) axial flux difference limits, (5) nuclear heat flux hot channel factor limit (FQ),

(6) nuclear enthalpy rise hot channel factor limit (FNM), (7) refueling boron concentration limit, and (8) shutdown margin.

Docket 50-305 NRC-02-064 July 26, 2002 Page 2 In addition, as described in WCAP-14483-A, "Generic Methodology For Expanded Core Operation Limits Report," the reactor core safety limit curves and cycle specific DNB related parameter limits (RCS flow, temperature and pressure) may also be relocated. To relocate these parameters requires the addition of the Departure From Nucleate Boiling Ratio (DNBR) design limit, the fuel centerline melt temperature limit, and the retention of the minimum reactor coolant flow design limit in the TS.

Provided with this license amendment request is a description and justification of changes, safety evaluation, and no significant hazards determination. Also provided, for information only, is the COLR for the Kewaunee unit. The parameter limits contained within the COLR include limits applicable to present operation.

Several editorial changes are also included in this submittal. These editorial changes include capitalization of defined words, proper outline numbering, adding proper "if - then" action statements, and correcting punctuation errors. to this letter contains a description of the changes, a safety evaluation, a significant hazards determination, and environmental considerations for the proposed changes, attachment 2 contains the strike-out Technical Specification pages, attachment 3 contains the affected Technical Specification pages as revised, attachment 4 contains the bases pages, and attachment 5 contains the core operating limits report.

To the best of my knowledge and belief, the statements contained in this document are true and correct. In some respects, these statements are not based entirely on my personal knowledge, but on information furnished by cognizant NMC employees and consultants. Such information has been reviewed in accordance with company practice and I believe it to be reliable.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on July 26, 2002.

Sincerely, Mark E. Warner Site Vice President GOR Attachments cc -

US NRC Region III US NRC Senior Resident Inspector State of Wisconsin

ATTACHMENT 1 Letter from Mark E. Warner (NMC)

To Document Control Desk (NRC)

Dated July 26, 2002 License Amendment Request 185 Description of Proposed Changes Safety Evaluation Significant Hazards Determination Environmental Considerations

Docket 50-305 NRC-02-064 July 26, 2002, Page 1 Description of Proposed Changes to Technical Specification Generic Letter (GL) 88-16, "Removal of Cycle-Specific Parameter Limits from Technical Specifications," endorsed the removal of cycle-specific parameter limits from the Technical Specifications (TS). This removal is allowed provided the parameters are determined using NRC approved methodologies and changes to the parameters are reported to the NRC. This alternative consists of three separate actions to allow modifying KNPP's TS. These actions, which have been completed as shown in this submittal, are:

"* the addition of the definition of a named formal report that includes the values of cycle-specific parameter limits that have been established using an NRC-approved methodology and consistent with all applicable limits of the safety analysis,

"* the addition of an administrative reporting requirement to submit the formal report on cycle-specific parameter limits to the Commission for information, and

"* the modification of individual TS to note that cycle-specific parameters shall be maintained within the limits provided in the defined formal report.

Cycle specific parameters previously approved for Westinghouse units for relocation to the COLR include:

1) moderator temperature coefficient,

2) shutdown bank insertion limits,

3) control bank insertion limits,

4) axial flux difference limits,

5) nuclear heat flux hot channel factor limit (FQ),

6) nuclear enthalpy rise hot channel factor limit (FNAH),

7) refueling boron concentration limit, and

8) shutdown margin.

In addition, as described in WCAP-14483-A, "Generic Methodology For Expanded Core Operation Limits Report," the reactor core safety limit curves and cycle specific departure from nucleate boiling (DNB) parameter limits (RCS flow, pressure and temperature) as well as the overtemperature AT and overpower AT trip setpoint parameter values may also be relocated.

This relocation requires the addition of the departure from nucleate boiling ratio (DNBR) design limit, the fuel centerline melt temperature limit, and the retention of the minimum reactor coolant flow design limit in the TS.

Docket 50-305 NRC-02-064 July 26, 2002, Page 2 Therefore, to allow for the relocation of those parameter allowed by GL 88-16 and WCAP-14483-A, the following changes are made:

1) An addition of a definition of the formal report that includes the values of cycle-specific parameter limits is added to the definition section of the TS as TS 1.0.q, "Core Operating Limits Report (COLR)

2) An addition of the requirements for the COLR is added to TS Section 6.9, "Reporting Requirements," as TS 6.9.a.4, "Core Operating Limits Report (COLR)."

3) Various individual TS have been modified to state that their cycle-specific parameters shall be maintained within the limits provided in the COLR.

4) Relocation to the COLR of the departure-from-nucleate-boiling (DNB) parameters of the reactor coolant system (RCS) average coolant temperature (Tavg), RCS flow rate, and pressurizer pressure, as well as the overtemperature AT (OTAT) and overpower AT (OPAT) trip setpoint parameter values. Also, relocate the reactor core safety limit and replace it with the fuel departure-from-nucleate-boiling ratio (DNBR) limit and the fuel centerline temperature limit.

The definition of Shutdown Margin is being moved from TS 3.10.a to the definition section. To move more towards the industry standards, KNPP TS are transitioning towards improved standard technical specification (ISTS). As a part of this transition the definition of shutdown margin is changed to that of ISTS. The ISTS definition is modified to conform with KNPP specific modes and terminology and to add clarity to the phrase "independent means" stated in ISTS. This relocation is considered administrative.

Several editorial changes are also included in this submittal. These editorial changes include capitalization of defined words, proper outline numbering, adding proper "if - then" action statements, and correcting punctuation errors.

Basis for Changes Addition of the definition of the Core Operating Limits Report is added to the TS definition section as TS 1.0.q.

The addition of the COLR definition is in accordance with the guidance and requirements provided in GL 88-16. The definition is as provided in NUREG 1431, "Standard Technical Specifications, Westinghouse Plants," Revision 2.

This change is administrative only, as it defines the COLR and its use. The definition of the COLR does not, in itself, impose new requirements on the operation of the Kewaunee Nuclear Plant. Therefore, this change is considered administrative. It is in accordance with and required by the guidance of GL 88-16.

Docket 50-305 NRC-02-064 July 26, 2002, Page 3 TS Section 6.9. "Reporting Requirements." Addition of COLR reporting requirements contained in TS 6.9.a.4.

The addition of an administrative reporting requirement which defines the approved methodologies to be used for Kewaunee is proposed. This change includes the requirements from enclosure 2 of GL 88-16. On page 2 it states:

A new administrative reporting requirement shall be added to existing reporting requirements as follows.

[CORE] OPERATING LIMITS REPORT

[6.9.x] [Core] operating limits shall be established and documented in the [CORE] OPERATING LIMITS REPORT before each reload cycle or any remaining part of a reload cycle. (If desired, the individual specifications that address [core] operating limits may be referenced.)

The analytical methods used to determine the [core] operating limits shall be those previously reviewed and approved by the NRC in

[identify the Topical Report(s) by number, title, and date, or identify the staff's safety evaluation report for a plant-specific methodology by NRC letter and date]. The [core] operating limits shall be determined so that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met.

The [CORE] OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements thereto, shall be provided upon issuance, for each reload cycle, to the NRC Document Control Deck with copies to the Regional Administrator and the Resident Inspector.

KNPP TS 6.9.a.4, was designed to meet the intent of this requirement. TS 6.9.a.4.A states the requirements for establishing and documenting the core operating limits for each reload or partial reload cycle. TS 6.9.a.4.A also lists those TS parameters that are included and relocated to the COLR.

TS 6.9.a.4.B states the requirements for the methods used to determines those parameters or the methods that are used to verify the use of those parameter meets KNPP's safety analysis requirements. TS 6.9.a.4.C states that the core operating limits shall be determined such that all applicable limits of the safety analysis are met. In addition, TS 6.9.a.4.D, states that the COLR shall be provide to the NRC upon issuance.

The reporting condition proposed for addition to the TS is in accordance with the guidance in GL 88-16. The addition of this requirement ensures that the cycle-specific operating parameters are determined in accordance with NRC approved methodologies and continue to meet the acceptance criteria of the safety analyses as described in the updated Final Safety Analysis Report.

Thus, reactor operation will continue to meet all regulatory and design basis requirements.

Docket 50-305 NRC-02-064 July 26, 2002, Page 4 Application of each of the approved methodologies and the corresponding cycle specific parameter and are shown in Table 1. These methodologies have been approved for the use at KNPP. The addition of these methodologies to the Technical Specifications therefore, does not impose any new requirements on the operation of KNPP and is administrative.

Modify Individual TS to Specify that their limits are contained in the COLR This change is part of the requirements for implementation of a COLR in accordance with GL 88-16. TS 6.9.a.4.A lists the TS in which the limits have been replaced by a statement directing the user to the COLR.

Relocation of the Safety Limit and DNB parameters Relocation of the Reactor Core Safety Limit curves was approved and is subject to the conditions defined in WCAP-14483-P-A. In accordance with the conditions of approval for the WCAP, it is proposed to add the DNBR design limits and fuel centerline melt temperature limits to the Technical Specifications and relocate the reactor core safety limit curves to the COLR.

Operation within the acceptable region of the reactor core safety limit curves ensures these safety limits are met. The following DNBR limits will be added to the Technical Specifications in section 2.0:

The departure from nucleate boiling ratio (DNBR) shall be maintained > 1.14 for the HTP DNB correlation In addition, the following peak fuel centerline temperature limit will be added:

The peak fuel centerline temperature shall be maintained < 4700 *F.

In accordance WCAP-14483-A and associated conditions of approval, the cycle specific DNB parameters related to reactor coolant system temperature, pressure and flow will be relocated to the COLR. The following reactor coolant system design minimum flow limits will be retained in the Technical Specifications:

During steady-state power operation, reactor coolant flow rate shall be >_ 93,000 gallons per minute average per loop.

Relocation of the DNB parameter limits to the COLR is in accordance with WCAP-14483-A.

The methodology accepted for reference is as described in the table previously. The flow limits are as approved with Amendment 157 on September 5, 2001.

Docket 50-305 NRC-02-064 July 26, 2002, Page 5 Relocating the definition of Shutdown Margin from TS 3. 10.a to definition TS 1.0.r This change is done to place the definition in the section which the user would first look to find its meaning.

Additional changes are being made to conform to the description used in the improved standard technical specification (ISTS). This definition provides a clearer description of which rods are included in the calculation. The ISTS definition is also modified to provide additional clarity and KNPP mode terminology.

This change is consistent with current industry standards (NUREG 1431, Rev 2), modified to be plant specific for the KNPP.

The clearer description comes from ISTS use of the term RCCA. Rod control cluster assembles (RCCA) refers to all reactivity control rods in the reactor. In KNPP's current definition, the term used is control rods. This use of the term control rods could lead to some confusion as to whether the shutdown rods are included. Although the shutdown rods were included, this change provides clarity. An additional modification when using the ISTS definition is a substep which is similar to a condition in the KNPP TS but is worded different. In KNPP TS the definition refers to calculating the reactivity at the HOT SHUTDOWN condition. In ISTS, this conditions wording is changed to state the condition at which the fuel and moderator temperature conditions are at the nominal zero power design level, which is KNPP's HOT SHUTDOWN conditions or the zero power program level.

An additional reference is being added to the ISTS definition modified for the KNPP. In the definition for shutdown margin a statement is made allowing the licensee to include the reactivity of the highest worth RCCA if all rods are verified fully inserted by two independent means. In KNPP TS 3.10.e, reference is made to indication available for the operator to verify rod position relative to rod bank position. To avoid confusion of the operator, this definition is modified to reference TS 3.10.e. By referencing this TS the operators know which independent means of rod indications can be used.

This change is considered administrative, as it is consistent with KNPP's current licensing basis, stated in TS 3.1O.e.

Editorial Changes Several editorial changes are also included in this submittal. These editorial changes include capitalization of defined words, proper outline numbering, adding proper "if - then" action statements, and correcting punctuation errors.

Docket 50-305 NRC-02-064 July 26, 2002, Page 6 Safety Evaluation NRC Generic Letter 88-16 provides guidance for licensees, allowing relocation of cycle dependent variables from the Technical Specifications (TS) to a Core Operating Limits Report (COLR). Relocation of these variables is allowed if they are determined with NRC-approved methodologies and the TS contain reporting requirements described in the GL. The variables removed from the TS and relocated to the COLR can then be changed via the appropriate regulatory change mechanisms, principally 10 CFR 50.59. This avoids the need to frequently change TS. Appropriate safety limits are maintained in the Specifications.

Changes to the COLR parameters are reported to the NRC so that they can be monitored and trended.

Cycle specific parameters previously approved for Westinghouse units for relocation to the COLR include: (1) moderator temperature coefficient, (2) shutdown bank insertion limits, (3) control bank insertion limits, (4) axial flux difference limits, (5) nuclear heat flux hot channel factor limit (FQ), (6) nuclear enthalpy rise hot channel factor limit (FNJ), (7) refueling boron concentration limit, and (8) shutdown margin. In addition, as described in WCAP-14483-A, "Generic Methodology For Expanded Core Operation Limits Report," the reactor core safety limit curves and specific DNB related parameters may also be relocated with the addition of the departure from nucleate boiling ratio (DNBR) design limit, the fuel centerline melt temperature limit, and the retention of the minimum reactor coolant flow design limit in the TS.

This amendment request adds the definition of the COLR, and the reporting requirement as recommended by NUREG-1431, meeting the guidance in Generic Letter 88-16.

The cycle specific parameters that are relocated will be determined by analyses performed in accordance with NRC approved methodologies.

The methodologies are listed in the added reporting requirement section and are approved for use at KNPP and/or the KNPP Updated Safety Analysis Report (USAR).

Relocation of the specific DNB parameter limits and Reactor Core Safety Limit Curves is in accordance with approved WCAP-14483-A.

Specific safety limits related to DNBR, fuel centerline melt temperature and reactor coolant system flow are added or retained in the Specifications with this change.

Therefore, the Specifications continue to meet the criteria of 10 CFR 50.36, Criterion 2, in that appropriate safety limits are maintained in the Technical Specifications.

The reporting requirement ensures that the limits on the safety analyses continue to be met and that all changes to the design are performed in accordance with approved methodologies and reviewed under existing regulatory change mechanisms, principally 10 CFR 50.59. As such, these changes are essentially administrative and continue to ensure that KNPP is operated within the limits of its design and analyses.

Operation of KNPP in accordance with these proposed changes continue to ensure safe operation of the plant and will not be inimical to the health and safety of the public.

Docket 50-305 NRC-02-064 July 26, 2002, Page 7 No Significant Hazards Determination In accordance with the requirements of 10 CFR 50.4 and 10 CFR 50.90, Nuclear Management Company, LLC, requests amendments to Facility Operating Licenses DPR-43 for Kewaunee Nuclear Plant. The purpose of the proposed amendments is to implement a Core Operating Limits Report (COLR) at the Kewaunee Nuclear Plant (KNPP).

In accordance with the requirements of 10 CFR 50.91, NMC has evaluated operation of the Kewaunee Nuclear Plant in accordance with the proposed changes against the standards of 10 CFR 50.92. Operation of the Kewaunee Nuclear Plant in accordance with the proposed changes results in no significant hazards consideration. Our evaluation and basis for this conclusion follows.

1. Operation of the Kewaunee Nuclear Plant in accordance with the proposed amendments does not result in a significant increase in the probability or consequences of any accident previously evaluated.

The proposed changes relocate certain cycle specific parameters from the Technical Specifications to a Core Operating Limits Report (COLR) or are administrative in nature.

Appropriate design and safety limits are retained or added to the Specifications thereby meeting the requirements of 10 CFR 50.36.

Specific, approved methodologies used to determine and evaluate the parameter requirements are added to the Specifications and a reporting requirement is added to ensure the NRC is apprised of all changes. Approved methodologies are required to be used to evaluate and change parameters, and appropriate safety and design limits are maintained in the Technical Specifications. Thus, operation of KNPP will continue to meet all design and safety analysis requirements. Therefore, neither the probability nor consequences of an accident previously evaluated can be increased.

2. Operation of the Kewaunee Nuclear Plant in accordance with the proposed amendment does not create a new or different kind of accident from any accident previously evaluated.

Operation of KNPP, in accordance with the proposed changes, will continue to meet all design and safety limits. Appropriate design and safety limits continue to be controlled within the Technical Specifications. These changes will not result in a change to the design and safety limits under which KNPP operation has been determined to be acceptable.

Therefore, these changes cannot result in a new or different kind of accident from any accident previously evaluated.

Docket 50-305 NRC-02-064 July 26, 2002, Page 8

3. Operation of the Kewaunee Nuclear Plant in accordance with the proposed amendment does not result in a significant reduction in a margin of safety.

Appropriate safety limits continue to be controlled by the Specifications. Changes to cycle specific parameters related to these limits will be accomplished using NRC approved methodologies, thereby ensuring operation will continue within the bounds of the existing safety analyses including all applicable margins of safety.

Therefore, operation in accordance with the proposed changes cannot result in a significant reduction in a margin of safety.

Operation of the Kewaunee Nuclear Plant in accordance with the proposed amendments does not result in a significant increase in the probability or consequences of any accident previously evaluated; does not create a new or different kind of accident from any accident previously evaluated; and, does not result in a significant reduction in a margin of the safety. Therefore, operation in accordance with the proposed amendments involves no significant hazards consideration Environmental Considerations This proposed amendment involves a change to the Technical Specifications. It does not modify any facility components located within the restricted area, as defined in 10 CFR 20. NMC has determined that the proposed amendment involves no significant hazards considerations and no significant change in the types of effluents that may be released offsite and that there is no significant increase in the individual or cumulative occupational radiation exposure.

This proposed amendment accordingly 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 this proposed amendment.

Docket 50-305 NRC-02-064 July 26, 2002, Page 9 TABLE 1 Relocated Parameter and Approved Methods Parameter NRC Approved Methodology Reactor Core Safety Limits SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION ON "QUALIFICATIONS OF REACTOR PHYSICS METHODS FOR APPLICATION TO KEWAUNEE" REPORT, dated August 21, 1979, report date September 29, 1978 KEWAUNEE NUCLEAR POWER PLANT - REVIEW FOR KEWAUNEE RELOAD SAFETY EVALUATION METHODS TOPICAL REPORT WPSRSEM-NP, REVISION 3 (TAC NO MB0306) dated September 10 2001.

Shutdown Margin SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION ON "QUALIFICATIONS OF REACTOR PHYSICS METHODS FOR APPLICATION TO KEWAUNEE" REPORT, dated August 21, 1979, report date September 29, 1978 Moderator Temperature Coefficient SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION ON "QUALIFICATIONS OF REACTOR PHYSICS METHODS FOR APPLICATION TO KEWAUNEE" REPORT, dated August 21, 1979, report date September 29, 1978 Shutdown Bank Insertion Limits SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION ON "QUALIFICATIONS OF REACTOR PHYSICS METHODS FOR APPLICATION TO KEWAUNEE" REPORT, dated August 21, 1979, report date September 29, 1978 KEWAUNEE NUCLEAR POWER PLANT - REVIEW FOR KEWAUNEE RELOAD SAFETY EVALUATION METHODS TOPICAL REPORT WPSRSEM-NP, REVISION 3 (TAC NO MB0306) dated September 10 2001.

Docket 50-305 NRC-02-064 July 26, 2002, Page 10 TABLE 1 Relocated Parameter and Approved Methods Parameter NRC Approved Methodology Control Bank Insertion Limits SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION ON "QUALIFICATIONS OF REACTOR PHYSICS METHODS FOR APPLICATION TO KEWAUNEE" REPORT, dated August 21, 1979, report date September 29, 1978 KEWAUNEE NUCLEAR POWER PLANT - REVIEW FOR KEWAUNEE RELOAD SAFETY EVALUATION METHODS TOPICAL REPORT WPSRSEM-NP, REVISION 3 (TAC NO MB0306) dated September 10 2001.

Height Dependent Heat Flux Hot SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION Channel Factor (FQ)

ON "QUALIFICATIONS OF REACTOR PHYSICS METHODS FOR APPLICATION TO KEWAUNEE" REPORT, dated August 21, 1979, report date September 29, 1978 Nissley, M.E. et, al., "Westinghouse Large-Break LOCA Best-Estimate Methodology,"

WCAP-10924-P-A, Volume 1, Revision 1, Addendum 4, March 1991, Volume 1: Model Description and Validation; Addendum 4: Model Revisions.

N. Lee et al., "Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code,"

WCAP-10054-P-A (Proprietary) and WCAP-10081 -NP-A (Non-Proprietary), dated August 1985.

C.M. Thompson, et al., "Addendum to the Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code: Safety Injection into the Broken Loop and COSI Condesation Model," WCAP-10054-P-A, Addendum 2, Revision 1 (Proprietary) and WCAP-10081-NP (Non-Proprietary), dated July 1997.

Docket 50-305 NRC-02-064 July 26, 2002, Page 11 TABLE 1 Relocated Parameter and Approved Methods Parameter

[NRC Approved Methodology ANF-88-133 (P)(A) and Supplement 1, "Qualification of Advanced Nuclear Fuels' PWR Design Meghodology for Rod Burnups of 62 GWd/MTU," Advanced Nuclear Fuels Corporation, dated December 1991 XN-NF-77-57, Exxon Nuclear Power Distribution Control for Pressurized Water Reactors, Phase II, dated January 1978, and Supplement 2, dated October 1981.

Nuclear Enthalpy Rise Hot Channel SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION Factor (FNAH)

ON "QUALIFICATIONS OF REACTOR PHYSICS METHODS FOR APPLICATION TO KEWAUNEE" REPORT, Dated August 21, 1979, Report Date September 29, 1978 KEWAUNEE NUCLEAR POWER PLANT - REVIEW FOR KEWAUNEE RELOAD SAFETY EVALUATION METHODS TOPICAL REPORT WPSRSEM-NP, REVISION 3 (TAC NO MB0306) Dated September 10 2001.

XN-NF-82-06 (P)(A) Revision 1 and Supplements 2, 4, and 5, "Qualification of Exxon Nuclear Fuel for Extended Burnup, Exxon Nuclear Company, dated October 1986.

ANF-88-133 (P)(A) and Supplement 1, "Qualification of Advanced Nuclear Fuels' PWR Design Meghodology for Rod Burnups of 62 GWd/MTU," Advanced Nuclear Fuels Corporation, dated December 1991.

EMF-92-116 (P)(A) Revision 0, "Generic Mechanical Design Criteria for PWR Fuel Designs," Siemens Power Corporation, dated February 1999

Docket 50-305 NRC-02-064 July 26, 2002, Page 12 TABLE 1 Relocated Parameter and Approved Methods Parameter NRC Approved Methodology Axial Flux Difference SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION ON "QUALIFICATIONS OF REACTOR PHYSICS METHODS FOR APPLICATION TO KEWAUNEE" REPORT, dated August 21, 1979, report date September 29, 1978 KEWAUNEE NUCLEAR POWER PLANT - REVIEW FOR KEWAUNEE RELOAD SAFETY EVALUATION METHODS TOPICAL REPORT WPSRSEM-NP, REVISION 3 (TAC NO MB0306) dated September 10 2001.

Overtemperature AT KEWAUNEE NUCLEAR POWER PLANT - REVIEW FOR KEWAUNEE RELOAD SAFETY EVALUATION METHODS TOPICAL REPORT WPSRSEM-NP, REVISION 3 (TAC NO MB0306) dated September 10 2001.

Overpower AT KEWAUNEE NUCLEAR POWER PLANT - REVIEW FOR KEWAUNEE RELOAD SAFETY EVALUATION METHODS TOPICAL REPORT WPSRSEM-NP, REVISION 3 (TAC NO MB0306) dated September 10 2001.

DNB Limits KEWAUNEE NUCLEAR POWER PLANT - REVIEW FOR KEWAUNEE RELOAD SAFETY EVALUATION METHODS TOPICAL REPORT WPSRSEM-NP, REVISION 3 (TAC NO MB0306) dated September 10 2001.

Refueling Boron Concentration SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION ON "QUALIFICATIONS OF REACTOR PHYSICS METHODS FOR APPLICATION TO KEWAUNEE" REPORT, dated August 21, 1979, report date September 29, 1978

ATTACHMENT 2 Letter from Mark E. Warner (NMC)

To Document Control Desk (NRC)

Dated July 26, 2002 License Amendment Request 185 Strikeout TS Pages:

TS i, TS ii, TS iv, TS vi TS 1.0-4 TS 1.0-6 TS 2.1-1 TS Figure 2.1-1 TS 2.3-2 through TS 2.3 TS 3.1-1 through TS 3.1 TS 3.8-1 TS 3.10-1 through TS 3.1C TS Figure 3.10-1 through TS Figt TS 6.9-3 through TS 6.9-3 9

)-7 ire 3.10-6 5

TABLE OF CONTENTS TECHNICAL SPECIFICATIONS APPENDIX A Section Title PAae 1.0 Definitions...............................................................................................................

1.0-1 1.0.a Quadrant-to-Average Power Tilt Ratio........................................................

1.0-1 1.0.b Safety lim its...............................................................................................

1.0-1 1.0.c Lim iting Safety System Settings..................................................................

1.0-1 1.0.d Lim iting Conditions for Operation................................................................

1.0-1 1.0.e Operable - O perability..............................................................................

1.0-21 1.0.f Operating................................................................................................

1.0-4 1 1.0.g Containm ent System Integrity.....................................................................

1.0-2 1.0.h Protective Instrum entation Logic...............................................................

1.0-32 1.0.i Instrumentation Surveillance.......................................................................

1.0-3 1.0.j Modes.......................................................................................................

1.0-4 1.0.k Reactor Critical..........................................................................................

1.0-4 1.0.1 Refueling O peration...................................................................................

1.0-4 1.0.m Rated Power............................................................................................

1.0-64 1.0.n Reportable Event.....................................................................................

1.0-6=4 1.0.0 Radiological Effluents................................................................................

1.0-5 1.0.p Dose Equivalent 1-131................................................................................

1.0-6 1.0.q Core Operatina Lim its Report....................................................................

1.0-6 1.0.r Shutdown Margin.......................................................................................

1.0-6 2.0 Safety Lim its and Lim iting Safety System Settings...................................................

2.1-1 2.1 Safety Lim its, Reactor Core........................................................................

2.1-1 2.2 Safety Lim it, Reactor Coolant System Pressure.........................................

2.2-1 2.3 Limiting Safety System Settings, Protective Instrumentation..........................................................................................

2.3-1 2.3.a Reactor Trip Settings................................................................

2.3-1 2.3.a.1 Nuclear Flux.......................................................

2.3-1 2.3.a.2 Pressurizer.........................................................

2.3-1 2.3.a.3 Reactor Coolant Tem perature............................ 2.3-2 2.3.a.4 Reactor Coolant Flow.........................................

2.3-3 2.3.a.5 Steam Generators..............................................

2.3-3 2.3.a.6 Reactor Trip Interlocks......................................

2.3-34=

2.3.a.7 Other Trips..........................................................

2.3-4 3.0 Lim iting Conditions for Operation.............................................................................

3.0-1 3.1 Reactor Coolant System............................................................................

3.1-1 3.1.a O perational Com ponents..........................................................

3.1-1 3.1.a.1 Reactor Coolant Pum ps......................................

3.1-1 3.1.a.2 Decay Heat Removal Capability......................... 3.1-1 3.1.a.3 Pressurizer Safety Valves...................................

3.1-2 3.1.a.4 Pressure Isolation Valves.................................

3.142 3.1.a.5 Pressurizer PORV and PORV Block Valves....... 3.1-3 3.1.a.6 Pressurizer Heaters..........................................

3.1-43 3.1.a.7 Reactor Coolant Vent System...........................

3.1-44 3.1.b Heatup & Cooldown Limit Curves for Normal Operation...............................................................................

3.1-64 3.1.c Maxim um Coolant Activity.......................................................

3.1-6 3.1.d Leakage of Reactor Coolant...................................................

3.1-97 3.1.e Maximum Reactor Coolant Oxygen, Chloride and Fluoride Concentration.........................................................

3.1-4--8 3.1.f Minim um Conditions for Criticality.........................................

3.1-148 LAR 185 TS i 07/26/2002

Section Title 3.2 Chemical and Volume Control System..........................................................

3.2-1 3.3 Engineered Safety Features and Auxiliary Systems......................................

3.3-1 3.3.a Accumulators...........................................................................

3.3-1 3.3.b Emergency Core Cooling System.............................................

3.3-2 3.3.c Containment Cooling Systems..................................................

3.3-4 3.3.d Component Cooling System.....................................................

3.3-6 3.3.e Service W ater System..............................................................

3.3-7 3.4 Steam and Power Conversion System........................................................

3.4-1 3.4.a Main Steam Safety Valves........................................................

3.4-1 3.4.b Auxiliary Feedwater System.....................................................

3.4-2 3.4.c Condensate Storage Tank........................................................

3.4-4 3.4.d Secondary Activity Limits..........................................................

3.4-5 3.5 Instrumentation System.............................................................................

3.5-1 3.6 Containment System.................................................................................

3.6-1 3.7 Auxiliary Electrical Systems........................................................................

3.7-1 3.8 Refueling Operations.................................................................................

3.8-1 3.9 Deleted 3.10 Control Rod and Power Distribution Limits................................................

3.10-1 3.10.a Shutdown Reactivity...............................................................

3.10-1 3.10.b Power Distribution Limits......................................................

3.10-21 3.10.c Quadrant Power Tilt Limits....................................................

3.10-64 3.10.d Rod Insertion Limits..............................................................

3.10-64 3.10.e Rod Misalignment Limitations...............................................

3.104 =5 3.10.f Inoperable Rod Position Indicator Channels..................... 3.1040-85 3.10.g Inoperable Rod Limitations...................................................

3.10-86=

3.10.h Rod Drop Time.....................................................................

3.10-06 3.10.i Rod Position Deviation Monitor.............................................

3.10-496 3.10.j Quadrant Power Tilt Monitor.................................................

3.10-96 3.10.k Core Average Temperature..................................................

3.10-96=

3.10.1 Reactor Coolant System Pressure........................................

3.10-96 3.10.m Reactor Coolant Flow........................................................

3.10-407 3.10.n DNBR Parameters................................

3.10-407 3.11 Core Surveillance Instrumentation............................................................

3.11-1 3.12 Control Room Post-Accident Recirculation System..................................

3.12-1 3.14 Shock Suppressors (Snubbers)................................................................

3.14-1 4.0 Surveillance Requirements......................................................................................

4.0-1 4.1 Operational Safety Review.........................................................................

4.1-1 4.2 ASME Code Class In-service Inspection and Testing.................................

4.2-1 4.2.a ASME Code Class 1, 2, 3, and MC Components and Supports..................................................................................

4.2-1 4.2.b Steam Generator Tubes...........................................................

4.2-2 4.2.b.1 Steam Generator Sample Selection and Inspection....................................................

4.2-3 4.2.b.2 Steam Generator Tube Sample Selection and Inspection....................................................

4.2-3 4.2.b.3 Inspection Frequency.........................................

4.2-4 4.2.b.4 Plugging Limit Criteria.........................................

4.2-5 4.2.b.5 Deleted 4.2.b.6 Deleted 4.2.b.7 Reports...............................................................

4.2-5 4.3 Deleted LAR 185 TS ii 07/26/2002 Rne

Section 4.4 4.5 4.6 4.7 4.8 4.9 4.10 4.11 4.12 4.13 4.14 4.15 4.16 4.17 5.0 Design 5.1 5.2 5.3 5.4 Containment Tests 4.4-1 4.4.a Integrated Leak Rate Tests (Type A)........................................

4.4-1 4.4.b Local Leak Rate Tests (Type B and C).....................................

4.4-1 4.4.c Shield Building Ventilation System............................................

4.4-1 4.4.d Auxiliary Building Special Ventilation System............................ 4.4-3 4.4.e Containment Vacuum Breaker System.....................................

4.4-3 Emergency Core Cooling System and Containment Air Cooling System Tests................................................................................

4.5-1 4.5.a System Tests...........................................................................

4.5-1 4.5.a.1 Safety Injection System......................................

4.5-1 4.5.a.2 Containment Vessel Internal Spray System...............................................................

4.5-1 4.5.a.3 Containment Fan Coil Units................................ 4.5-2 4.5.b Com ponent Tests.....................................................................

4.5-2 4.5.b. 1 Pum ps................................................................

4.5-2 4.5.b.2 Valves.................................................................

4.5-2 Periodic Testing of Em ergency Power System...........................................

4.6-1 4.6.a Diesel Generators.....................................................................

4.6-1 4.6.b Station Batteries.......................................................................

4.6-2 Main Steam Isolation Valves.......................................................................

4.7-1 Auxiliary Feedwater System.......................................................................

4.8-1 Reactivity Anom alies..................................................................................

4.9-1 Deleted Deleted Spent Fuel Pool Sweep System................................................................

4.12-1 Radioactive Materials Sources..................................................................

4.13-1 Testing and Surveillance of Shock Suppressors (Snubbers)..................... 4.14-1 Deleted Reactor Coolant Vent System Tests.........................................................

4.16-1 Control Room Postaccident Recirculation System....................................

4.17-1 Features.....................................................................................................

5.1-1 Site...........................................................................................................

5.1-1 Containment..............................................................................................

5.2-1 5.2.a Containment System................................................................

5.2-1 5.2.b Reactor Containment Vessel...................................................

5.2-2 5.2.c Shield Building..........................................................................

5.2-2 5.2.d Shield Building Ventilation System............................................

5.2-2 5.2.e Auxiliary Building Special Ventilation Zone and Special Ventilation System.......................................................

5.2-3 Reactor Core.............................................................................................

5.3-1 5.3.a Fuel Assem blies.......................................................................

5.3-1 5.3.b Control Rod Assem blies...........................................................

5.3-1 Fuel Storage..............................................................................................

5.4-1 5.4.a Criticality..................................................................................

5.4-1 5.4.b Capacity..................................................................................

5.4-1 5.4.c Canal Rack Storage..................................................................

5.4-2 LAR 185 TS iii 07/26/2002 Title

?me

Section Title PAge 6.0 Administrative Controls............................................................................................

6.1-1 6.1 Responsibility.............................................................................................

6.1-1 6.2 Organization..............................................................................................

6.2-1 6.2.a Off-Site Staff............................................................................

6.2-1 6.2.b Facility Staff.............................................................................

6.2-1 6.2.c Organizational Changes.........................................................

6.2-21 6.3 Plant Staff Qualifications............................................................................

6.3-1 6.4 Training.....................................................................................................

6.4-1 6.5 Deleted.........................................................................................

6.5 6.5-6 6.6 Deleted.....................................................................................................

6.6-1 6.7 Safety Limit Violation.................................................................................

6.7-1 6.8 Procedures................................................................................................

6.8-1 6.9 Reporting Requirements............................................................................

6.9-1 6.9.a Routine Reports........................................................................

6.9-1 6.9.a. 1 Startup Report....................................................

6.9-1 6.9.a.2 Annual Reporting Requirements......................... 6.9-1 6.9.a.3 Monthly Operating Report...................................

6.9-3 6.9.a.4 Core Ooeratina Limits Report.............................

6.9.3 6.9.b Unique Reporting Requirements.............................................

6.9-35 6.9.b. 1 Annual Radiological Environmental Monitoring Report.............................................

6.9-35 6.9.b.2 Radioactive Effluent Release Report................ 6.9-35 6.9.b.3 Special Reports................................................

6.9-3=5 6.10 Record Retention.....................................................................................

6.10-1 6.11 Radiation Protection Program...................................................................

6.11-1 6.12 System Integrity.......................................................................................

6.12-1 6.13 High Radiation Area.................................................................................

6.13-1 6.14 Deleted...................................................................................................

6.14-1 6.15 Secondary W ater Chemistry.....................................................................

6.15-1 6.16 Radiological Effluents..............................................................................

6.16-1 6.17 Process Control Program (PCP)...............................................................

6.17-1 6.18 Offsite Dose Calculation Manual (ODCM).................................................

6.18-1 6.19 Major Changes to Radioactive Liquid, Gaseous and Solid W aste Treatment Systems........................................................

6.19-1 6.20 Containment Leakage Rate Testing Program...........................................

6.20-1 7/8.0 Deleted LAR 185 TS iv 07/26/2002

LIST OF FIGURES 2.1-1................. Safety Limits Roactor Coro, TIiorm3al and HydraulicDeleted 3.1-1................. Heatup Limitation Curves Applicable for Periods Up to 3311]

Effective Full-Power Years 3.1-2................. Cooldown Limitation Curves Applicable for Periods Up to 33(11 Effective Full-Power Years 3.1-3................. Dose Equivalent 1-131 Reactor Coolant Specific Activity Limit Versus Percent of Rated Thermal Power 3.1-4................. Deleted 3.10-1 Rquid Sh,utd.o.n ReactiAiy' "Ractor Boron Con-AR-t at-nDeleted 3.10-2...............

Hot Channol Factor Normalizod

.p.rating Ein.okpo*eleted 3.10-3............... Co trol Bank IncrA.to. LmiteDeleted 3.104........

1.......

Deleted Poermiiblo Oporating B on Indwic.atod Flux Difforonco as a F

ti Of BuFrup (Typical 3.10-5............... DeletedTargot o

n.'di c

".od Flux D on.

.c.

.Function of Oporating P, Wor Lovol (Typi,, al) 3.10-6

....... Deleted V(Z) a" a Function of oro Height 4.2-1................. Deleted 5.4-1................. Minimum Required Fuel Assembly Bumup as a Function of Nominal Initial Enrichment to Permit Storage in the Tranfer Canal Note:

(1]

Although the curves were developed for 33 EFPY, they are limited to 28 EFPY (corresponding to the end of cycle 28) by WPSC Letter NRC-99-017.

LAR 185 07/26/2002 TS vi FIGURE TITLE

TECHNICAL SPECIFICATIONS AND BASES 1.0 DEFINITIONS The following terms are defined for uniform interpretation of the specifications.

a. QUADRANT-TO-AVERAGE POWER TILT RATIO The QUADRANT-TO-AVERAGE POWER TILT RATIO is defined as the ratio of maximum-to-average of the upper excore detector currents or that of the lower excore detector currents, whichever is greater. If one excore detector is out of service, the three in-service units are used in computing the average.

b.

SAFETY LIMITS SAFETY LIMITS are the necessary quantitative restrictions placed upon those process variables that must be controlled in order to reasonably protect the integrity of certain of the physical barriers which guard against the uncontrolled release of radioactivity.

c.

LIMITING SAFETY SYSTEM SETTINGS LIMITING SAFETY SYSTEM SETTINGS are setpoints for automatic protective devices responsive to the variables on which SAFETY LIMITS have been placed. These setpoints are so chosen that automatic protective actions will correct the most severe, anticipated abnormal situation so that a SAFETY LIMIT is not exceeded.

d.

LIMITING CONDITIONS FOR OPERATION LIMITING CONDITIONS FOR OPERATION are those restrictions on reactor operation, resulting from equipment performance capability, that must be enforced to ensure safe operation of the facility.

e. OPERABLE-OPERABILITY A system or component is OPERABLE or has OPERABILITY when it is capable of performing its intended function within the required range. The system or component shall be considered to have this capability when: (1) it satisfies the LIMITING CONDITIONS FOR OPERATION defined in TS 3.0; and (2) it has been tested periodically in accordance with TS 4.0 and has met its performance requirements.

Implicit in this definition shall be the assumption that all necessary attendant instrumentation, controls, normal and emergency electrical power sources, cooling or seal water, lubrication or other auxiliary equipment that is required for the system or component to perform its intended function is also capable of performing their related support functions.

f.

OPERATING A system or component is considered to be OPERATING when it is performing the intended function in the intended manner.

LAR 185 TS 1.0-1 07/26/2002

g. CONTAINMENT SYSTEM INTEGRITY CONTAINMENT SYSTEM INTEGRITY is defined to exist when:

1.

The nonautomatic Containment System isolation valves and blind flanges are closed, except as provided in TS 3.6.b.

2.

The Reactor Containment Vessel and Shield Building equipment hatches are properly closed.

3.

At least ONE door in both the personnel and the emergency airlocks is properly closed.

4.

The required automatic Containment System isolation valves are OPERABLE, except as provided in TS 3.6.b.

5.

All requirements of TS 4.4 with regard to Containment System leakage and test frequency are satisfied.

6.

The Shield Building Ventilation System and the Auxiliary Building Special Ventilation System satisfy the requirements of TS 3.6.c.

h.

PROTECTIVE INSTRUMENTATION LOGIC

1.

PROTECTION SYSTEM CHANNEL A PROTECTION SYSTEM CHANNEL is an arrangement of components and modules as required to generate a single protective action signal when required by a plant condition. The channel loses its identity where single action signals are combined.

2.

LOGIC CHANNEL A LOGIC CHANNEL is a matrix of relay contacts which operate in response to PROTECTIVE SYSTEM CHANNEL signals to generate a protective action signal.

3.

DEGREE OF REDUNDANCY DEGREE OF REDUNDANCY is defined as the difference between the number of OPERATING channels and the minimum number of channels which, when tripped, will cause an automatic shutdown.

4.

PROTECTION SYSTEM The PROTECTION SYSTEM consists of both the Reactor PROTECTION SYSTEM and the Engineered Safety Features System.

The PROTECTION SYSTEM encompasses all electric and mechanical devices and circuitry (from sensors through actuated device) which are required to operate in order to produce the required protective function. Tests of PROTECTION SYSTEM will be considered acceptable when tests are run in part and it can be shown that all parts satisfy the requirements of the system.

LAR 185 TS 1.0-2 07/26/2002

i.

INSTRUMENTATION SURVEILLANCE

1.

CHANNEL CHECK CHANNEL CHECK is a qualitative determination of acceptable OPERABILITY by observation of channel behavior during operation. This determination shall include, where possible, comparison of the channel indication with other indications derived from independent channels measuring the same variable.

2.

CHANNEL FUNCTIONAL TEST A CHANNEL FUNCTIONAL TEST consists of injecting a simulated signal into the channel as close to the primary sensor as practicable to verify that it is OPERABLE, including alarm and/or trip initiating action.

3.

CHANNEL CALIBRATION CHANNEL CALIBRATION consists of the adjustment of channel output as necessary, such that it responds, with acceptable range and accuracy, to known values of the parameter which the channel monitors. Calibration shall encompass the entire channel, including alarm and/or trip, and shall be deemed to include the CHANNEL FUNCTIONAL TEST.

4.

SOURCE CHECK A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a source of increased radioactivity.

5.

FREQUENCY NOTATION The FREQUENCY NOTATION specified for the performance of surveillance requirements shall correspond to the intervals in Table TS 1.0-1.

LAR 185 TS 1.0-3 07/26/2002

j.

MODES

k.

REACTOR CRITICAL The reactor is said to be critical when the neutron chain reaction is self-sustaining.

I.

REFUELING OPERATION REFUELING OPERATION is any operation involving movement of reactor vessel internal components (those that could affect the reactivity of the core) within the containment when the vessel head is unbolted or removed.

m. RATED POWER RATED POWER is the steady-state reactor core output of 1,650 MWt.

n.

REPORTABLE EVENT A REPORTABLE EVENT is defined as any of those conditions specified in 10 CFR 50.73.

LAR 185 TS 1.0-4 07/26/2002 COOLANT TEMP FISSION MODE REACTIVITY A*k/k T

0 OE Tag OF POWER %

REFUELING

<-5%

_ 140

-0 COLD SHUTDOWN

___-1%

<*200

-0 INTERMEDIATE (1)

> 200 < 540

-0 SHUTDOWN HOT SHUTDOWN (1)

>540

-0 HOT STANDBY

< 0.25%

-Tope,

< 2 OPERATING

< 0.25%

-Toper 2

LOW POWER PHYSICS (To be specified by specific tests)

TESTING (1) Refer to Fi...F..-O.

.4the required SHUTDOWN MARGIN as specified in the Core Operatinq Limits Report.

o.

RADIOLOGICAL EFFLUENTS

1.

MEMBER(S) OF THE PUBLIC MEMBER(S) OF THE PUBLIC shall include all persons who are not occupationally associated with the plant. This category does not include employees of the utility, its contractors or vendors. Also excluded from this category are persons who enter the site to service equipment or to make deliveries. This category does include persons who use portions of the site for recreational, occupational or other purposes not associated with the plant.

2.

OFF-SITE DOSE CALCULATION MANUAL (ODCM)

The ODCM shall contain the current methodology and parameters used in the calculation of off-site doses due to radioactive gaseous and liquid effluents, and in the calculation of gaseous and liquid effluent monitoring alarm/trip setpoints, and in the conduct of the Radiological Environmental Monitoring Program. The ODCM shall also contain (1) the Radioactive Effluent Controls and Radiological Environmental Monitoring Programs required by TS 6.16.b, and (2) descriptions of the information that should be included in the Annual Radiological Environmental Operating and Radioactive Effluent Release Reports required by TS 6.9.b.1 and TS 6.9.b.2.

3.

PROCESS CONTROL PROGRAM (PCP)

The PCP shall contain the current formulae, sampling, analyses, tests, and determinations to be made to ensure that the processing and packaging of solid radioactive wastes, based on demonstrated processing of actual or simulated wet solid wastes, will be accomplished in such a way as to assure compliance with 10 CFR Part 20, 10 CFR Part 61, 10 CFR Part 71, federal and state regulations, burial ground requirements, and other requirements governing the disposal of the radioactive waste.

4.

SITE BOUNDARY The SITE BOUNDARY shall be that line beyond which the land is neither owned, nor leased, nor otherwise controlled by the licensee.

5.

UNRESTRICTED AREA An UNRESTRICTED AREA shall be any area at or beyond the SITE BOUNDARY access to which is not controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials, or any area within the SITE BOUNDARY used for residential quarters or for industrial, commercial, institutional, and/or recreational purposes.

LAR 185 TS 1.0-5 07/26/2002

p.

DOSE EQUIVALENT 1-131 DOSE EQUIVALENT 1-131 is that concentration of 1-131 (,u Ci/gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of 1-131, 1-132, 1-133, 1-134 and 1-135 actually present. The thyroid dose conversion factors used for this calculation shall be as listed and calculated with the methodology established in Table III of TID-14844, "Calculation of Distance Factors for Power and Test Reactor Sites."

DOSE CONVERSION FACTOR J

ISOTOPE 1.0000 1-131 0.0361 1-132 0.2703 1-133 0.0169 1-134 0.0838 1-135

a. CORE OPERATING LIMITS REPORT (COLR)

The COLR is the unit specific document that provides cycle-specific parameter limits for the current reload cycle. These cycle specific parameter limits shall be determined for each reload cycle in accordance with Specification 6.9.a.4. Plant operation within these limits is addressed in individual Specifications.

r.

SHUTDOWN MARGIN (SDM)

SDM shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuminq:

a. All rod cluster control assemblies (RCCAs) are fully inserted except for the single RCCA of highest reactivity worth, which is assumed to be fully withdrawn. However.,

with all RCCAs verified fully inserted by two independent means (TS 3.10.e), it is not necessary to account for a stuck RCCA in the SDM calculation.

With any RCCA not capable of being fully inserted, the reactivity worth of the RCCA must be accounted for in the determination of SDM. and

b. In the OPERATING and HOT STANDBY MODES, the fuel and moderator temperatures are changed to the program temperature.

LAR 185 07/26/2002 TS 1.0-6

2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS, REACTOR CORE APPLICABILITY Applies to the limiting combination of thermal power, Reactor Coolant System pressure and coolant temperature during the OPERATING and HOT STANDBY MODESepertfiOR.

OBJECTIVE To maintain the integrity of the fuel cladding.

SPECIFICATION

a. The combination of rated power level, coolant pressure, and coolant temperature shall not exceed the limits.hoWn in i*gur, TS 2.1 ispecified in the COLR. The safety limit is exceeded if the point defined by the combination of Reactor Coolant System average temperature and power level is at any time above the appropriate pressure line.

b. The departure from nucleate boiling ratio (DNBR) shall be maintained _> 1.14 for the HTP DNB correlation

c. The peak fuel centerline temperature shall be maintained < 4700 OF LAR 185 0772B12=2 TS 2.1-1

2.3 LIMITING SAFETY SYSTEM SETTINGS, PROTECTIVE INSTRUMENTATION APPLICABILITY Applies to trip settings for instruments monitoring reactor power and reactor coolant pressure, temperature, flow, pressurizer level, and permissives related to reactor protection.

OBJECTIVE To prevent the principal process variables from exceeding a SAFETY LIMIT.

SPECIFICATION

a. Reactor trip settings shall be as follows:

1. Nuclear Flux A. Source Range (high setpoint) inspan of source range B. Intermediate range (high setpoint)

40% of RATED POWER C. Power range (low setpoint)

525% of RATED POWER D. Power range (high setpoint)

< 109% of RATED POWER E. Power range fast flux rate trip 15%Aq/5 sec (positive)

F. Power range fast flux rate trip 1 %Aq/5 sec (negative)

2. Pressurizer A. High pressurizer pressure

<2385 psig B. Low pressurizer pressure

> 1875 psig C. High pressurizer water level

<190% of full scale LAR 185 07726/20=

TS 2.3-1

3.

Reactor Coolant Temperature A.

Overtemperature AT <ATo [KI-K 2 (T-T')l+v1 s

K 3 (P-P')-f(AI)]

1+ Z2 S where AT0

=

Indicated AT at RATED POWER, %*F T

=

Average temperature, OF Tv

=

66-73[*]° F P

=

Pressurizer pressure, psig P'

=

2236-psig K,

=

4--[j K2

=

G[*j]

K(3

=

"ONODO6[

Ti

=

aOC-sec.

"T2

=

4-J sec.

f(AI)

=

An even function of the indicated difference between top and bottom detectors of the power-range nuclear ion chambers. Selected gains are based on measured instrument response during plant startup tests, where qt and qb are the percent power in the top and bottom halves of the core respectively, and qt + qb is total core power in percent of RATED POWER, such that:

(-)1.

For q,- qb within 4-2[*,

4QJ %, f (AI) = 0.

(Mb2.

For each percent that the magnitude of qt - qb exceeds,49[*] % the

=

T trip setpoint shall be automatically reduced by an equiAlent of 2-.6] % of RATED POWER.

(-e3.

For each percent that the magnitude of qt - qb exceed -42[1 % the A

T trip setpoint shall be automatically reduced by an equi-'alent of 4-*[] % of RATED POWER.

Note: [*] As specified in the COLR LAR 185 TS 2.3-2 07760

B.

Overpower AT< ATo[ K 4 -K 5

T-3 S T K6 (T-T')-f(AI)1 where AT0

= Indicated AT at RATED POWER, *%r T

= Average Temperature, *F T'

= 6674[*jF K4

<440UJ K5

_ 0.0275 f*]for increasing T; O-fljfor decreasing T K6

> O-O02-[]for T > T; O-[*]for T < T' T3

= 40-lsec.

f(Al)

= As defined above Note: [] As specified in the COLR

4.

Reactor Coolant Flow A.

Low reactor coolant flow per loop > 90% of normal indicated flow as measured by elbow taps.

B.

Reactor coolant pump motor breaker open

1.

Low frequency setpoint > 55.0 Hz

2.

Low voltage setpoint > 75% of normal voltage

5.

Steam Generators Low-low steam generator water level > 5% of narrow range instrument span.

LAR 185 TS 2.3-3 07726=200

6.

Reactor Trip Interlocks Protective instrumentation settings for reactor trip interlocks shall be as follows:

A.

Above 10% of RATED POWER, the low pressurizer pressure trip, high pressurizer level trip, the low reactor coolant flow trips (for both loops), and the turbine trip-reactor trip are made functional.

B.

Above 10% of RATED POWER, the single-loop loss-of-flow trip is made functional.

7.

Other Trips Undervoltage __ 75% of normal voltage Turbine trip Manual trip Safety injection trip (Refer to Table TS 3.5-1 for trip settings)

LAR 185 TS 2.3-4 07/2672U=

F'IGUR TS.

xC 2.1 1 TS FIGURE 2.1-1 DELETED LAR 185 07/26/2002 PAGE 1 OF 1

3.1 REACTOR COOLANT SYSTEM APPLICABILITY Applies to the OPERATING status of the Reactor Coolant System (RCS).

OBJECTIVE To specify those LIMITING CONDITIONS FOR OPERATION of the Reactor Coolant System which must be met to ensure safe reactor operation.

SPECIFICATIONS

a. Operational Components

1. Reactor Coolant Pumps A. At least one reactor coolant pump or one residual heat removal pump shall be in operation when a reduction is made in the boron concentration of the reactor coolant.

B.

When the reactor is in the OPERATING mode, except for low power tests, both reactor coolant pumps shall be in operation.

C. A reactor coolant pump shall not be started with one or more of the RCS cold leg temperatures < 200OF unless the secondary water temperature of each steam generator is < 100OF above each of the RCS cold leg temperatures.

2. Decay Heat Removal Capability A. At least -ltWwOo of the following FfJ;ýoJuRr heat sinks shall be OPERABLE whenever the average reactor coolant temperature is < 350°F but > 2000F.

1. Steam Generator 1A

2. Steam Generator 1 B

3.

Residual Heat Removal Train A

4. Residual Heat Removal Train B If less than the above number of required heat sinks are OPERABLE, then corrective action shall be taken immediately to restore the minimum number to the OPERABLE status.

LAR 185 TS 3.1-1 07/26/2002

B. TWwQo residual heat removal trains shall be OPERABLE whenever the average reactor coolant temperature is < 200OF and irradiated fuel is in the reactor, except when in the REFUELING MODE with the minimum water level above the top of the vessel flange >_ 23 feet, one train may be inoperable for maintenance.

1. Each residual heat removal train shall be comprised of:

a) ONnEe OPERABLE residual heat removal pump b) ONnfe OPERABLE residual heat removal heat exchanger c) An OPERABLE flow path consisting of all valves and piping associated with the above train of components and required to remove decay heat from the core during normal shutdown situations. This flow path shall be capable of taking suction from the appropriate Reactor Coolant System hot leg and returning to the Reactor Coolant System.

2.

If one residual heat removal train is inoperable, then corrective action shall be taken immediately to return it to the OPERABLE status.

3. Pressurizer Safety Valves A. At least one pressurizer safety valve shall be OPERABLE whenever the reactor head is on the reactor pressure vessel, except for a hydro test of the RCS the pressurizer safety valves may be blanked provided the power-operated relief valves and the safety valve on the discharge of the charging pump are set for test pressure plus 35 psi to protect the system.

B.

Both pressurizer safety valves shall be OPERABLE whenever the reactor is critical.

4. Pressure Isolation Valves A. All pressure isolation valves listed in Table TS 3.1-2 shall be functional as a pressure isolation device during OPERATING and HOT STANDBY MODES, except as specified in 3.1.a.4.B. Valve leakage shall not exceed the amounts indicated.

B.

In the event that integrity of any pressure isolation valve as specified in Table TS 3.1-2 cannot be demonstrated, reactor operation may continue, provided that at least two valves in each high pressure line having a non-functional valve are in, and remain in, the mode corresponding to the isolated condition.()

C. If TS 3.1.a.4.A and TS 3.1.a.4.B cannot be met, then an orderly shutdown shall be initiated and the reactor shall be in the HOT SHUTDOWN condition within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , the INTERMEDIATE SHUTDOWN condition in the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and the COLD SHUTDOWN condition within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

(1) Manual valves shall be locked in the closed position. Motor operated valves shall be placed in the closed position with their power breakers locked out.

LAR 185 TS 3.1-2 07/26/2002

5. Pressurizer Power-Operated Relief Valves (PORV) and PORV Block Valves A. Two PORVs and their associated block valves shall be OPERABLE during HOT STANDBY and OPERATING modes.

1. With one or both PORVs inoperable because of excessive seat leakage, within one hour either restore the PORV(s) to OPERABLE status or close the associated block valve(s) with power maintained to the block valve(s);

otherwise, action shall be initiated to:

Achieve HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months

- Achieve HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months

2. With one PORV inoperable due to causes other than excessive seat leakage, within 1-one hour either restore the PORV to OPERABLE status or dose its associated block valve and remove power from the block valve_..

Restore the PORV to OPERABLE status within the following 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or action shall be initiated to:

Achieve HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months

- Achieve HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months

3. With both PORVs inoperable due to causes other than excessive seat leakage, within 1-one hour either restore at least one PORV to OPERABLE status or close its associated block valve and remove power from the block valve and Achieve HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months

- Achieve HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months

4. With one block valve inoperable, within 1-one hour restore the block valve to OPERABLE status or place its associated PORV in manual control.

Restore the block valve to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; otherwise action shall be initiated to:

Achieve HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Achieve HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months

5. With both block valves INPERABL-inoperable, within onehour restore the block valves to OPERABLE status or place their associated PORVs in manual control. Restore at least one block valve to OPERABLE status within the next hour; otherwise, action shall be initiated to:

Achieve HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Achieve HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months

6. Pressurizer Heaters A. At least one group of pressurizer heaters shall have an emergency power supply available when the average RCS temperature is > 350°F.

LAR 185 TS 3.1-3 07/26/2002

7. Reactor Coolant Vent System A. A reactor coolant vent path from both the reactor vessel head and pressurizer steam space shall be OPERABLE and closed prior to the average RCS temperature being heated > 200OF except as specified in TS 3.1.a.7.B and TS 3.1.a.7.C below.

B. When the average RCS temperature is > 2000F, any one of the following conditions of inoperability may exist:

1. Both of the parallel vent valves in the reactor vessel vent path are inoperable.

2.

Both of the parallel vent valves in the pressurizer vent path are inoperable.

If OPERABILITY is not restored within 30 days, then within one hour action shall be initiated to:

Achieve HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Achieve HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Achieve COLD SHUTDOWN within an additional 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months C. If no Reactor Coolant System vent paths are OPERABLE, then restore at least one vent path to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . If OPERABILITY is not restored within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , then within 4-one hour action shall be initiated to:

Achieve HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Achieve HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Achieve COLD SHUTDOWN within an additional 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months

b. Heatup and Cooldown Limit Curves for Normal Operation

1. The reactor coolant temperature and pressure and system heatup and cooldown rates (with the exception of the pressurizer) shall be limited in accordance with Figures TS 3.1-1 and TS 3.1-2. Figures TS 3.1-1 and TS 3.1-2 are applicable for the service period of up to 33(1) effective full-power years.

A. Allowable combinations of pressure and temperature for specific temperature change rates are below and to the right of the limit lines shown. Limit lines for cooldown rates between those presented may be obtained by interpolation.

B. Figures TS 3.1-1 and TS 3.1-2 define limits to assure prevention of non-ductile failure only. For normal operation other inherent plant characteristics, e.g.,

pump heat addition and pressurizer heater capacity may limit the heatup and cooldown rates that can be achieved over certain pressure-temperature ranges.

(1) Although the curves were developed for 33 EFPY, they are limited to 28 EFPY (corresponding to the end of cycle 28) by WPSC Letter NRC-99-017.

LAR 185 TS 3.1-4 07/26/2002

C. The isothermal curve in Figure TS 3.1-2 defines limits to assure prevention of non-ductile failure applicable to low temperature overpressurization events only.

Application of this curve is limited to evaluation of LTOP events whenever one or more of the RCS cold leg temperatures are less than or equal to the LTOP enabling temperature of 2000F.

2. The secondary side of the steam generator must not be pressurized > 200 psig if the temperature of the steam generator is < 700F.

3. The pressurizer cooldown and heatup rates shall not exceed 200°F/hr and 100°F/hr, respectively. The spray shall not be used if the temperature difference between the pressurizer and the spray fluid is > 3200F.

4. The overpressure protection system for low temperature operation shall be OPERABLE whenever one or more of the RCS cold leg temperatures are < 2000F, and the reactor vessel head is installed.

The system shall be considered OPERABLE when at least one of the following conditions is satisfied:

A. The overpressure relief valve on the Residual Heat Removal System (RHR 33-1) shall have a set pressure of < 500 psig and shall be aligned to the RCS by maintaining valves RHR 1A, 1B, 2A, and 2B open.

1. With one flow path inoperable, the valves in the parallel flow path shall be verified open with the associated motor breakers for the valves locked in the off position. Restore the inoperable flow path within 6-five days or complete depressurization and venting of the RCS through a > 6.4 square inch vent within an additional 8-eight hours.

2. With both flow paths or RHR 33-1 inoperable, complete depressurization and venting of the RCS through at least a 6.4 square inch vent pathway within 8-eight hours.

B. A vent pathway shall be provided with an effective flow cross section 2! 6.4 square inches.

1. When low temperature overpressure protection is provided via a vent pathway, verify the vent pathway at least once per 31 days when the pathway is provided by a valve(s) that is locked, sealed, or otherwise secured in the open position.

If the vent path is provided by any other means, then verify the vent pathway every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

LAR 185 TS 3.1-5 07/26/2002

c. Maximum Coolant Activity

1. The specific activity of the reactor coolant shall be limited to:

A. < -20.20 pCi/gram DOSE EQUIVALENT 1-131, and 91 'Uci B. < --

gross radioactivity due to nuclides with half-lives > 30 minutes E

cc excluding tritium ( Eis the average sum of the beta and gamma energies in Mev per disintegration) whenever the reactor is critical or the average coolant temperature is > 5000F.

2. If the reactor is critical or the average temperature is > 500°F:

A. With the specific activity of the reactor coolant > 0.20 pCi/gram DOSE EQUIVALENT 1-131 for more than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months during one continuous time interval, or exceeding the limit shown on Figure TS 3.1-3, be in at least INTERMEDIATE SHUTDOWN with an average coolant temperature of < 500OF within 6-six hours.

91 g"Ci B. With the specific activity of the reactor coolant > -=-

of gross radioactivity, E

cc be in at least INTERMEDIATE SHUTDOWN with an average coolant temperature < 500OF within 6-six hours.

C. With the specific activity of the reactor coolant > 0.20 pCi/gram DOSE 91.iCi EQUIVALENT 1-131 o r >--- -

perform the sample and analysis requirements E

cc of Table TS 4.1-2, item 1.f, once every 4-four hours until restored to within its limits.

3. Annual reporting requirements are identified in TS 6.9.a.2.D.

LAR 185 TS 3.1-6 07/26/2002

d. Leakage of Reactor Coolant

1. Any Reactor Coolant System leakage indication in excess of 1 gpm shall be the subject of an investigation and evaluation initiated within 4-four hours of the indication. Any indicated leak shall be considered to be a real leak until it is determined that no unsafe condition exists. If the Reactor Coolant System leakage exceeds 1 gpm and the source of leakage is not identified within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , then the reactor shall be placed in the HOT SHUTDOWN condition utilizing normal operating procedures. If the source of leakage exceeds 1 gpm and is not identified within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , then the reactor shall be placed in the COLD SHUTDOWN condition utilizing normal operating procedures.

2. Reactor coolant-to-secondary leakage through the steam generator tubes shall be limited to 150 gallons per day through any one steam generator. With tube leakage greater than the above limit, reduce the leakage rate within 4-four hours or be in COLD SHUTDOWN within the next 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

3. If the sources of leakage other than that in 3.1.d.2 have been identified and it is evaluated that continued operation is safe, then operation of the reactor with a total Reactor Coolant System leakage rate not exceeding 10 gpm shall be permitted. If leakage exceeds 10 gpm, then the reactor shall be placed in the HOT SHUTDOWN condition within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months utilizing normal operating procedures.

If the leakage exceeds 10 gpm for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , the reactor shall be placed in the COLD SHUTDOWN condition utilizing normal operating procedures.

4. If any reactor coolant leakage exists through a non-isolable fault in a Reactor Coolant System component (exterior wall of the reactor vessel, piping, valve body, relief valve leaks, pressurizer, steam generator head, or pump seal leakoff), then the reactor shall be shut down; and cooldown to the COLD SHUTDOWN condition shall be initiated within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of detection.

5. When the reactor is critical and above 2% power, two reactor coolant leak detection systems of different operating principles shall be in operation with one of the two systems sensitive to radioactivity. Either system may be out of operation for up to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months provided at least one system is OPERABLE.

LAR 185 TS 3.1-7 07/26/2002

e. Maximum Reactor Coolant Oxygen, Chloride and Fluoride Concentration

1. Concentrations of contaminants in the reactor coolant shall not exceed the following limits when the reactor coolant temperature is > 2501F.

NORMAL STEADY-STATE TRANSIENT OPERATION (ppm)

LIMITS (ppm)

A.

Oxygen 0.10 1.00 B.

Chloride 0.15 1.50 C.

Fluoride 0.15 1.50

2. If any of the normal steady-state operating limits as specified in TS 3.1.e.1 above are exceeded, or if it is anticipated that they may be exceeded, then corrective action shall be taken immediately.

3. If the concentrations of any of the contaminants cannot be controlled within the transient limits of TS 3.1.e. 1 above or returned to the normal steady-state limit within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , then the reactor shall be brought to the COLD SHUTDOWN condition, utilizing normal operating procedures, and the cause shall be ascertained and corrected. The reactor may be restarted and operation resumed if the maximum concentration of any of the contaminants did not exceed the permitted transient values-. Otherwise a safety review by the Plant Operations Review Committee shall be made before starting.

4. Concentrations of contaminants in the reactor coolant shall not exceed the following maximum limits when the reactor coolant temperature is _ 2500F.

NORMAL TRANSIENT CONTAMINANT CONCENTRATION LIMITS (ppm)

(ppm LIMITS___(ppm)_

A.

Oxygen Saturated Saturated B.

Chloride 0.15 1.50 C.

Fluoride 0.15 1.50

5. If the transient limits of TS 3.1.e.4 are exceeded or the concentrations cannot be returned to normal values within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , then the reactor shall be brought to the COLD SHUTDOWN condition and the cause shall be ascertained and corrected.

6. To meet TS 3.1.e.1 and TS 3.1.e.4 above, reactor coolant pump operation shall be permitted for short periods, provided the coolant temperature does not exceed 2500F.

LAR 185 TS 3.1-8 07/26/2002

f.

Minimum Conditions for Criticality

1. The reactor shall not be brought to a critical condition until the pressure-temperature state is to the right of the criticality limit line shown in Figure TS 3.1-1.

2. The reactor shall be maintained subcritical by at least 1% Ak/k until normal water level is established in the pressurizer.

3. When the reactor is critical ard rag% R*A*TED POWER, the moderator temperature coefficient shall be-Q as specified in the COLR, except during LOW POWER PHYSICS TESTING. When th* reactor i" 60% RATED POWER, tho moedorator tomporaturo cooffticint shall be Zoro or negati.o

4. Th neractor Will naye a moeieaora iormpora!uro coonicionil no less neg I-wethan

-J

4. If the limits of 3.1.f.3 cannot be met, then power operation may continue provided the following actions are taken:

A. Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , develop and maintain administrative control rod withdrawal limits sufficient to restore the moderator temperature coefficient to within the limits specified in TS 3.1.f.3. These withdrawal limits shall be in addition to the insertion limits specified in TS 3.10.d.

B. If the actions specified in TS 3.1.f.64.A are not satisfied, then be in HOT STANDBY within the next 6-six hours.

LAR 185 TS 3.1-9 07/26/2002

-A;ýý

hAI*P J*

_! = =l

I......

iý

3.8 REFUELING OPERATIONS APPLICABILITY Applies to operating limitations during REFUELING OPERATIONS.

OBJECTIVE To ensure that no incident occurs during REFUELING OPERATIONS that would affect public health and safety.

SPECIFICATION

a. During REFUELING OPERATIONS:

1. Containment Closure

a. The equipment hatch shall be closed and at least one door in each personnel air lock shall be capable of being closed'1 ) in 30 minutes or less. In addition, at least one door in each personnel air lock shall be closed when the reactor vessel head or upper internals are lifted.

b. Each line that penetrates containment and which provides a direct air path from containment atmosphere to the outside atmosphere shall have a closed isolation valve or an operable automatic isolation valve.

2. Radiation levels in fuel handling areas, the containment and the spent fuel storage pool shall be monitored continuously.

3. The reactor will be subcritical for 148 hours0.00171 days 0.0411 hours 2.44709e-4 weeks 5.6314e-5 months prior to movement of its irradiated fuel assemblies. Core subcritical neutron flux shall be continuously monitored by at least ltWw&o neutron monitors, each with continuous visual indication in the control room and QoNnfe with audible indication in the containment whenever core geometry is being changed. When core geometry is not being changed at least OoNnae neutron flux monitor shall be in service.

4. At least QoNnEe residual heat removal pump shall be OPERABLEepeFable.

5. When there is fuel in the reactor, a minimum boron concentration of 22

...,d ANchtdo'n margin of_>5%.k/k as specified in the COLR shall be maintained in the Reactor Coolant System during reactor vessel head removal or while loading and unloading fuel from the reactor. The required boron concentration shall be verified by chemical analysis daily.

(1) Administrative controls ensure that:

"* Appropriate personnel are aware that both personnel air lock doors are open,

"* A specified individual(s) is designated and available to close the air lock following a required evacuation of containment, and

"* Any obstruction(s) (e.g., cables and hoses) that could prevent closure of an open air lock can be quickly removed.

LAR 185 TS 3.8-1 07/26/2002

6. Direct communication between the control room and the operating floor of the containment shall be available whenever changes in core geometry are taking place.

7. Heavy loads, greater than the weight of a fuel assembly, will not be transported over or placed in either spent fuel pool when spent fuel is stored in that pool. Placement of additional fuel storage racks is permitted, however, these racks may not traverse directly above spent fuel stored in the pools.

8. The containment ventilation and purge system, including the capability to initiate automatic containment ventilation isolation, shall be tested and verified to be operable immediately prior to and daily during REFUELING OPERATIONS.

9. a. The spent fuel pool sweep system, including the charcoal adsorbers, shall be operating during fuel handling and when any load is carried over the pool if irradiated fuel in the pool has decayed less than 30 days. If the spent fuel pool sweep system, including the charcoal adsorber, is not operating when required, fuel movement shall not be started (any fuel assembly movement in progress may be completed).

b. Performance Requirements

1. The results of the in-place cold DOP and halogenated hydrocarbon tests at design flows on HEPA filters and charcoal adsorber banks shall show

>99% DOP removal and >99% halogenated hydrocarbon removal.

2. The results of laboratory carbon sample analysis from spent fuel pool sweep system carbon shall show >95% radioactive methyl iodide removal when tested in accordance with ASTM D3803-89 at conditions of 300C and 95% RH.

3. Fans shall operate within +10% of design flow when tested.

10. The minimum water level above the vessel flange shall be maintained at 23 feet.

11. A dead-load test shall be successfully performed on both the fuel handling and manipulator cranes before fuel movement begins. The load assumed by the cranes for this test must be equal to or greater than the maximum load to be assumed by the cranes during the REFUELING OPERATIONS. A thorough visual inspection of the cranes shall be made after the dead-load test and prior to fuel handling.

12. A licensed senior reactor operator will be on-site and designated in charge of the REFUELING OPERATIONS.

b.

If any of the specified limiting conditions for REFUELING OPERATIONS are not met, refueling of the reactor shall cease. Work shall be initiated to correct the violated conditions so that the specified limits are met, and no operations which may increase the reactivity of the core shall be performed.

LAR 185 TS 3.8-2 07/26/2002

3.10 CONTROL ROD AND POWER DISTRIBUTION LIMITS APPLICABILITY Applies to the limits on core fission power distributions and to the limits on control rod operations.

OBJECTIVE To ensure: 1) core subcriticality after reactor trip, 2) acceptable core power distribution during power operation in order to maintain fuel integrity in normal operation transients associated with faults of moderate frequency, supplemented by automatic protection and by administrative procedures, and to maintain the design basis initial conditions for limiting faults, and 3) limited potential reactivity insertions caused by hypothetical control rod ejection.

SPECIFICATION

a.

Shutdown Reactivity When the reactor is subcritical prior to reactor startup, the HOT--SHUTDOWN MARGINr-I shall be at least that chown in Figur. T-8 3. 1 ai Sho mgd hoo i do d a the amount by wheih th rFoaedr thre wOuld the highocst worth Gentrol mod-ro-m-anod fully withdrawn, and acuignehangoc in Xonon eF b9FRe.

b.

Power Distribution Limits

1. At all times, except during Low Power Physics Tests, the hot channel factors defined in the basis must meet the following limits:

A. FQN(Z) Limits shall be fo' So*mA.. PoWor COLR.

-(2.35)'P x K(Z) for P >.*I IeFart*.GF'-eI as specified in the 114%yA FQ(-E--4-7O.4(--1-7Q§--*-

(1.70) x K(Z) for P -*-.5 0FQ(Z) x 2

1 X0 1.05* (2.28)/P* K(.7) for-P > A FQ()-4O3.*47Q--*-

(1.56) K(Z) for P -*-.5

[6Sd1

[Std]

Wh8F8+

Id 1F; 4RA ir~ffiin ()f flul DOWOF at WIG ther eau l5r %dc mrw%~

is the f-nction giYn iAn Figuro TS 3.10.2 ic tho cwre height location for tho F,_cf L.tcret B. F N Limits shall befcr bomncv ro-vWozr opera"ion -Fuel as specified in the COLR.

LAR 185 07/26/2002 TS 3.10-1 L

t

.1_

t

,t..IN

&L=

L-- I..=

._z --.1Q4__--*.70 [1I., 0.(

P)]!

[Hl u'y.

_* -- *4

-_4'*-5

[1 r+., 0.2(1 P),]

WhieFe+

P ic tho fracionie of fudl poere at W:hich tho coro i OPER'\\TING

2. If, for any measured hot channel factor, the relationships of F.N (Z) and FHN specified in the COLR',S 3.0.b.4 are not true, then reactor power shall be reduced by a fractional amount of the design power to a value for which the relationships are true, and the high neutron flux trip setpoint shall be reduced by the same fractional amount.

If subsequent incore mapping cannot, within a 24-hour period, demonstrate that the hot channel factors are met, then the overpower AT and overtemperature AT trip setpoints shall be similarly reduced.

3. Following initial loading and at regular effective full-power monthly intervals thereafter, power distribution maps using the movable detection system shall be made to confirm that the hot channel factor limits of TS 3.10.b.1 are satisfied.

4. The measured FQEQ(Z) hot channel factors under equilibrium conditions shall satisfy the fellewi..

-relationship for the central axial 80% of the core for Som...

Po.We CoFpeFatieR. fe-as specified in the COLR.

FQ*(Z). 1.03

1. x '.,(,)

-*-,2.35).'P., K(Z)

ENVY]

i(Z) x 1.03 1.05 x,.'(

-,-e( 2.2.- P x K(Z)

fMtd, P

i the fraction of full poWer at which the c1 rI rc OPERA.TING Y(7Z) ic defined in Figure T-S 3.10 6 E'Q(.Z) ic a moeasurod FQ dictribution obtainod duOrig the targlet fluM detoRmRinatien

5. Power distribution maps using the movable detector system shall be made to confirm the relationship of SF 3.b.4 FEQ(Z) specified in the COLR according to the following schedules with allowances for a 25% grace period:

A. During the target flux difference determination or once per effective full-power monthly interval, whichever occurs first.

B. Upon achieving equilibrium conditions after reaching a thermal power level > 10%

higher than the power level at which the last power distribution measurement was performed in accordance with TS 3.10.b.5.A.

C. If a power distribution map indicates an increase in peak pin power, FIHN, of 2% or more, due to exposure, when compared to the last power distribution map, then either of the following actions shall be taken:

LAR 185 TS 3.10-2 07/26/2002

FQEQ(Z) shall be increased by Rn additional 2% the penaltv factor specified in the COLR for comparison to the relationship of c....fiod in TS 3.10.-b. 4, OQR FEu(Z) specified in the COLR, or ii.

FQE'(Z) shall be measured by power distribution maps using the incore movable detector system at least once every 7seven effective full-power days until a power distribution map indicates that the peak pin power, FAN, is not increasing with exposure when compared to the last power distribution map.

6. If, for a measured FoEQ, the relationships of TS 3.10*b*. 4F EQ(Z) specified in the COLR are not satisfied and the relationships of T-9 3.10..I._FMZ and F6HN specified in the COLR are satisfied, then within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months take one of the following actions:

A. Take corrective actions to improve the power distribution and upon achieving equilibrium conditions measure the target flux difference and verify that the relationships O_.f

-po-.f,'d in TS 3.10.b.1. F.FlE(Z) soecified in the COLR are satisfied, or B. Reduce reactor power and the high neutron flux trip setpoint by 1% for each percent that the left hand sides of the relationships Qf pof.ifi.d in TS 3.10.1..

specified in the COLR exceed the limits specified in the right hand sides. Reactor power may subsequently be increased provided that a power distribution map verifies that the relationships of T-S 3.1.bMA FQEQ(Z) specified in the COLR are satisfied with at least 1% of margin for each percent of power level to be increased.

7. The reference equilibrium indicated axial flux difference as a function of power level (called the target flux difference) shall be measured at least once per full-power month.

8. The indicated axial flux difference shall be considered outside of the limits of TS 3.10.b.9 through TS 3.10.b.12 when more than one of the OPERABLE excore channels are indicating the axial flux difference to be outside a limit.

9. Except during physics tests, during excore detector calibration and except as modified by TS 3.10.b.10 through TS 3.10.b.12, the indicated axial flux difference shall be maintained within the targeta---61% band about the target flux difference as specified in the COLR.

10. At a power level > 90% of rated power, if the indicated axial flux difference deviates from its target band, the flux difference shall be returned to the target band within 15 minutes or reactor power shall be reduced to a level no greater than 90% of rated power.

11. At power levels > 50% and

90% of rated power:

A. The indicated axial flux difference may deviate from its-the _-5%-target band, specified in the COLR. for a maximum of 4-one hour (cumulative) in any 24-hour period provided the flux difference does not exceed an envelope. =bou..dod by 0 0%

and +10% frm tho taFrgt axial flu d-ifforonc at 99% Feted poWor and inreoacing by 1% and '1% frm the targot amial flu* diefforonc for oAch 247% doc~eroaoi LAR 185 TS 3.10-3 07/26/2002

Fatod poew* r 0% and > 50%n specified in the COLR.

If the cumulative time exceeds 4-one hour, then the reactor power shall be reduced to

50% of rated thermal power within 30 minutes and the high neutron flux setpoint reduced to <

55% of rated power.

If the indicated axial flux difference exceeds the outer envelope specified in the

COLR_,

then the reactor power shall be reduced to

50% of rated thermal power within 30 minutes and the high neutron flux setpoint reduced to

<*55% of rated power.

B. A power increase to a level > 90% of rated power is contingent upon the indicated axial flux difference being within its target band.

12. At a power level no greater than 50% of rated power:

A. The indicated axial flux difference may deviate from its target band.

B. A power increase to a level > 50% of rated power is contingent upon the indicated axial flux difference not being outside its target band for more than two hours (cumulative) of the preceding 24-hour period.

One half of the time the indicated axial flux difference is out of its target band, up to 50% of rated power is to be counted as contributing to the 4-one hour cumulative maximum the flux difference may deviate from its target band at a power level

90% of rated power.

13. Alarms shall normally be used to indicate nonconformance with the flux difference requirement of TS 3.10.b.10 or the flux difference time requirement of TS 3.10.b.1 1.A.

If the alarms are temporarily out of service, then the axial flux difference shall be logged, and conformance with the limits assessed, every hour for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , and half-hourly thereafter.

c.

Quadrant Power Tilt Limits

1. Except for physics tests, whenever the indicated quadrant power tilt ratio > 1.02, one of the following actions shall be taken within 2-two_ hours:

A.

Eliminate the tilt.

B.

Restrict maximum core power level 2% for every 1% of indicated power tilt ratio

> 1.0.

2. If the tilt condition is not eliminated after 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , then reduce power to 50% or lower.

3. Except for Low Power Physics Tests, if the indicated quadrant tilt is > 1.09 and there is simultaneous indication of a misaligned rod:

A. Restrict maximum core power level by 2% of rated values for every 1% of indicated power tilt ratio > 1.0.

LAR 185 TS 3.10-4 07/26/2002

B. If the tilt condition is not eliminated within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , then the reactor shall be brought to a minimum load condition (* 30 Mwe).

4. If the indicated quadrant tilt is > 1.09 and there is no simultaneous indication of rod misalignment, then the reactor shall immediately be brought to a no load condition

(* 5% reactor power).

d.

Rod Insertion Limits

1. The shutdown rods shall be fk4ly-withdrawn to within the limits, specified in the COLR.

when the reactor is critical or approaching criticality.

2. The control banks shall be limited in physical insertion; insertion limits are shown-if specified in the COLR' Fg.e T-8O4. If any one of the control bank insertion limits ch.wn in Figu.ro TS 3.10- is not met:

A. Within 1-one hour, initiate boration to restore control bank insertion to within the limits o -ig TS 3.10 specified in the COLR, and B. Restore control bank insertion to within the limits of Figuro TS 3.1 0 3 specified in the COLR within 2-two hours of exceeding the insertion limits.

C. If any one of the conditions of TS 3.10.d.2.A or TS 3.10.d.2.B cannot be met, then within --one hour action shall be initiated to:

- Achieve HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months

- Achieve HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months

3. Insertion limit does not apply during physics tests or during periodic exercise of individual rods. However, the shutdown margin., n.dcod in Fig...ro T-3.40 las specified in the COLR, must be maintained except for the Low Power Physics Test to measure control rod worth and shutdown margin. For this test, the reactor may be critical with all but one high worth rod inserted.

e.

Rod Misalignment Limitations This specification defines allowable limits for misaligned rod cluster control assemblies. In TS 3.10.e.1 and TS 3.10.e.2, the magnitude, in steps, of an indicated rod misalignment may be determined by comparison of the respective bank demand step counter to the analog individual rod position indicator, the rod position as noted on the plant process computer, or through the conditioning module output voltage via a correlation of rod position vs.

voltage. Rod misalignment limitations do not apply during physics testing.

1. When reactor power is Ž_85% of rating, the rod cluster control assembly shall be maintained within +/- 12 steps from their respective banks. If a rod cluster control assembly is misaligned from its bank by more than +/- 12 steps when reactor power is

Ž!85%, then the rod will be realigned or the core power peaking factors shall be determined within 4-four hours, and TS 3.10.b applied. If peaking factors are not determined within 4-four hours, the reactor power shall be reduced to < 85% of rating.

LAR 185 TS 3.10-5 07/26/2002

2. When reactor power is < 85% but > 50% of rating, the rod cluster control assemblies shall be maintained within +/- 24 steps from their respective banks. If a rod cluster control assembly is misaligned from its bank by more than +/- 24 steps when reactor power is < 85% but > 50%, the rod will be realigned or the core power peaking factors shall be determined within 4-four hours, and TS 3.10.b applied. If the peaking factors are not determined within 4-four hours, the reactor power shall be reduced to < 50% of rating.

3. And, in addition to TS 3.10.e.1 and TS 3.10.e.2, if the misaligned rod cluster control assembly is not realigned within 8-ei"ht hours, the rod shall be declared inoperable.

f.

Inoperable Rod Position Indicator Channels

1. If a rod position indicator channel is out of service, then:

A. For operation between 50% and 100% of rating, the position of the rod cluster control shall be checked indirectly by core instrumentation (excore detector and/or thermocouples and/or movable incore detectors) at least once per 8-eight hours, or subsequent to rod motion exceeding a total displacement of 24 steps, whichever occurs first.

B. During operation < 50% of rating, no special monitoring is required.

2. Not more than one rod position indicator channel per group nor two rod position indicator channels per bank shall be permitted to be inoperable at any time.

3. If a rod cluster control assembly having a rod position indicator channel out of service is found to be misaligned from TS 3.10.f.1.A, then TS 3.10.e will be applied.

g.

Inoperable Rod Limitations

1. An inoperable rod is a rod which does not trip or which is declared inoperable under TS 3.10.e or TS 3.10.h.

2. Not more than one inoperable full length rod shall be allowed at any time.

3. If reactor operation is continued with one inoperable full length rod, the potential ejected rod worth and associated transient power distribution peaking factors shall be determined by analysis within 30 days unless the rod is made OPERABLE earlier. The analysis shall include due allowance for nonuniform fuel depletion in the neighborhood of the inoperable rod. If the analysis results in a more limiting hypothetical transient than the cases reported in the safety analysis, the plant power level shall be reduced to an analytically determined part power level which is consistent with the safety analysis.

h.

Rod Drop Time At OPERATING temperature and full flow, the drop time of each full length rod cluster control shall be no greater than 1.8 seconds from loss of stationary gripper coil voltage to dashpot entry. If drop time is > 1.8 seconds, the rod shall be declared inoperable.

LAR 185 TS 3.10-6 07/26/2002

i.

Rod Position Deviation Monitor If the rod position deviation monitor is inoperable, individual rod positions shall be logged at least once per 8-eight hours after a load change > 10% of rated power or after

> 24 steps of control rod motion.

j.

Quadrant Power Tilt Monitor If one or both of the quadrant power tilt monitors is inoperable, individual upper and lower excore detector calibrated outputs and the quadrant tilt shall be logged once per shift and after a load change > 10% of rated power or after > 24 steps of control rod motion. The monitors shall be set to alarm at 2% tilt ratio.

k.

Core Average Temperature During steady-state power operation, Tav, shall be maintained 4 4BOFwithin the limits specified in the COLR, except as provided by TS 3.10.n.

Reactor Coolant System Pressure During steady-state power operation, Reactor Coolant System pressure shall be maintained within the limits specified in the COLR> 2205 psig, except as provided by TS 3.10.n.

m.

Reactor Coolant Flow

1. During steady-state power operation, reactor coolant flow rate shall be > 93,000 gallons per minute average per loop and greater than or equal to the limit specified in the COLR. If reactor coolant flow rate is 4 03,000 gil*,, nl po. miuto por lop

,-not within the limits as specifled in the COLR, action shall be taken in accordance with TS 3.1 0.n.

2. Compliance with this flow requirement shall be demonstrated by verifying the reactor coolant flow during initial power escalation following each REFUELING, between 70%

and 95% power with plant parameters as constant as practical.

n.

DNBR Parameters If, during power operation any of the conditions of TS 3.10.k, TS 3.10.1, or TS 3.10.m.1 are not met, restore the parameter in-2 two hours or less to within limits or reduce power to < 5% of thermal rated power within an additional-6 six hours. Following analysis, thermal power may be raised not to exceed a power level analyzed to maintain a DNBR greater than the minimum DNBR limit.

LAR 185 TS 3.10-7 07/26/2002

FIGURE TS 3.0* 4 TS FIGURE 3.10-1 DELETED LAR 185 07/26/2002

FIGURE T4.41-2 2

TS FIGURE 3.10-2 DELETED LAR 185 07/26/2002 Page 1 efA

FIGURE T-S 3.10-3 TS FIGURE 3.10-3 DELETED LAR 185 07/26/2002

TS FIGURE TS 3.10-4 DELETED LAR 185 07/26/2002 Page 1 9f

TS FIGURE TS 3.10-5 DELETED LAR 185 07/26/2002 Page 1ef 4

TS FIGURE 3.10-6 DELETED LAR 185 07/26/2002

3. Monthly OPERATING Report Routine reports of OPERATING statistics and shutdown experience shall be submitted on a monthly basis to the Document Control Desk, U.S. Nuclear Regulatory Commission, Washington, D.C., 20555, with a copy to the appropriate Regional Office, to be submitted by the fifteenth of each month following the calendar month covered by the report.

4. Core Operating Limits Report (COLR)

A. Core operatinq limits shall be established prior to each reload cycle, or prior OT a reioaa CVCie. ana snaii De aocumeniea in me COLR tor the following:

(1)

TS 21

(")

Reactor CL (21 (3)

(41 1K

{(()

Reactor Core Safety Limit uverremoerature A I ;*eTDOInr

-S 2.3.a.3.B TR

1 f M" Over~ower AT Setpoint ModrlAtor Temner~ture Coefficient (MTC)

I M

[Al Moderator TemnPr-qF1j-rP Cnpffiniant ( TC)

(5)

TS 3.1.f.4 Moderator I eml2erature Coefficie

!..

4

S 3.1u.a I./~I tllflflrl I-crnrnn ( :~n~ntrntl/nn

-nuauown Marain (8)

TS 3.10.b.1.A FQ"Z) Limits (9)

TS 310.b.1.B FH'" Limits T10) TS 3.1.b.4.

F imits (11) TS 3.10O.b.5.C.i FQ"Z pnly "S 3.10.b.9 Axial Flux Difterence Tarqet Band

13) TS 3.10.b.11.A Axial Flux Difference Envelope

14) TS 3.10.d.1 Shutdown Bank Insertion Limits T15) TS 3.10.d.2 Control Bank Insertion Limits

ý16) TS 3.10.k Core Averaae Temperature

17) TS 3.10.1 Reactor Coolant System Pressure t53.ui.

Reactor LCoolant I-low B.

The analytical methods used to determine the core operatina limits shall be those previously reviewed and approved by the NRL(,

specifically those described in the following documents:

(1)

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION ON "QUALIFICATIONS OF REACTOR PHYSICS METHODS FOR APPLICATION TO KEWAUNEE" REPORT, dated August 21, 1979, report date September 29, 1978 (2)

KEWAUNEE NUCLEAR POWER PLANT REVIEW FOR KEWAUNEE RELOAD SAFETY EVALUATION METHODS TOPICAL REPORT WPSRSEM-NP, REVISION 3

MB03062 dated September 10, 2001.

LAR 185 07726=2002 TS 6.9-3 to any remalnlna DorrIon (Z

)

(..

ArI..

.l--

(3)

Overpower AT SetDoint 6-ý 13of"Aalin LZrrnn 1-nnrantr-ntinn I--

s

_:_,L S

.3.a.3.A

(3)

Nissley, M.E. et, al., "Westinghouse Large-Break LOCA Best Estimate Methodology," WCAP-10924-P-A, Volume 1, Revision 1, Addendum 4, March 1991, Volume 1:

Model Description and Validation; Addendum 4: Model Revisions.

(4)

N. Lee et al., 'Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code," WCAP-10054-P-A

-Proprietary) and WCAP-10081-NP-A (Non-Proprietary), dated August 1985.

(5)

C.M. Thompson, et al., "Addendum to the Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code: Safety Injection into the Broken Loop and COSI Condesation Model,"

WCAP-10054-P-A, Addendum 2, Revision 1 (Proprietary) and WCAP-10081-NP (Non-Proprietary), dated July 1997.

(6)

XN-NF-82-06 (P)(A) Revision 1 and Supplements 2, 4, and 5, "Qualification of Exxon Nuclear Fuel for Extended Burnup, Exxon Nuclear Company, dated October 1986.

(7)

ANF-88-133 (P)(A) and Supplement 1, "Qualification of Advanced Nuclear Fuels' PWR Design Meghodology for Rod Burnups of 62 GWd/MTU,"

Advanced Nuclear Fuels Corporation, dated December 1991.

(8)

EMF-92-116 (P)(A)

Revision 0, "Generic Mechanical Design Criteria for PWR Fuel Designs," Siemens Power Corporation, dated February 1999.

(9)

XN-NF-77-57, Exxon Nuclear Power Distribution Control for Pressurized Water Reactors, Phase II, dated January 1978, and Supplement 2, dated October 1981.

C. The core operatinq limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal nydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits suc as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

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

LAR 185 TS 6.9-4 0772B=

b.

Unique Reporting Requirements

1. Annual Radiological Environmental Monitoring Report A. Routine Radiological Environmental Monitoring Reports covering the operation of the unit during the previous calendar year shall be submitted prior to May 1 of each year. The report shall include summaries, interpretations, and analysis of trends of the results of the Radiological Environmental Monitoring Program for the reporting period. The material provided shall be consistent with the OFF-SITE DOSE CALCULATION MANUAL (ODCM) and Sections IV.B.2, IV.B.3, and IV.C of Appendix I to 10 CFR Part 50.

2. Radioactive Effluent Release Report Routine Radioactive Effluent Release Reports covedng the operation of the unit for the previous calendar year shall be submitted by May 1 of each year. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit. The material provided shall be consistent with the objectives outlined in the ODCM and the PCP, and in conformance with 10 CFR 50.36a and Section IV.B.1 of Appendix I to 10 CFR Part 50.

3. Special Reports A. Special reports may be required covering inspections, test and maintenance activities. These special reports are determined on an individual basis for each unit and their preparation and submittal are designated in the Technical Specifications.

(1) Special reports shall be submitted to the Director of the NRC Regional Office listed in Appendix D, 10 CFR Part 20, with a copy to the Director, Office of Inspection and Enforcement, U.S.

Nuclear Regulatory Commission, Washington, D.C. 20555 within the time period specified for each report.

LAR 185 TS 6.9-5 07726=2

ATTACHMENT 3 Letter from Mark E. Warner (NMC)

To Document Control Desk (NRC)

Dated July 26, 2002 License Amendment Request 185 Affected TS Pages:

TS i, TS ii, TS iv, TS vi TS 1.0-1 through TS 1.0-6 TS 2.1-1 TS Figure 2.1-1 TS 2.3-1 through TS 2.3-4 TS 3.1-1 through TS 3.1-9 TS 3.8-1 through TS 3.8-2 TS 3.10-1 through TS 3.10-7 TS Figure 3.10-1 through TS Figure 3.10-6 TS 6.9-3 through TS 6.9-5

TABLE OF CONTENTS TECHNICAL SPECIFICATIONS APPENDIX A Section Title Pae 1.0 Definitions...............................................................................................................

1.0-1 1.0.a Q uadrant-to-Average Power Tilt Ratio........................................................

1.0-1 1.0.b Safety lim its...............................................................................................

1.0-1 1.0.c Lim iting Safety System Settings..................................................................

1.0-1 1.0.d Lim iting Conditions for Operation................................................................

1.0-1 1.0.e Operable - Operability................................................................................

1.0-1 1.0.f Operating..................................................................................................

1.0-1 1.0.g Containment System Integrity.....................................................................

1.0-2 1.0.h Protective Instrumentation Logic.................................................................

1.0-2 1.0.i Instrumentation Surveillance.......................................................................

1.0-3 1.0.j Modes.......................................................................................................

1.0-4 1.0.k Reactor Critical..........................................................................................

1.0-4 1.0.1 Refueling Operation...................................................................................

1.0-4 1.0.m Rated Power..............................................................................................

1.0-4 1.0.n Reportable Event.......................................................................................

1.0-4 1.0.0 Radiological Effl uents................................................................................

1.0-5 1.0.p Dose Equivalent 1-131................................................................................

1.0-6 1.0.q Core Operating Lim its Report.....................................................................

1.0-6 1.0.r Shutdown M argin.......................................................................................

1.0-6 2.0 Safety Lim its and Lim iting Safety System Settings...................................................

2.1-1 2.1 Safety Lim its, Reactor Core........................................................................

2.1-1 2.2 Safety Lim it, Reactor Coolant System Pressure.........................................

2.2-1 2.3 Limiting Safety System Settings, Protective Instrumentation..........................................................................................

2.3-1 2.3.a Reactor Trip Settings................................................................

2.3-1 2.3.a.1 Nuclear Flux.......................................................

2.3-1 2.3.a.2 Pressurizer.........................................................

2.3-1 2.3.a.3 Reactor Coolant Tem perature............................ 2.3-2 2.3.a.4 Reactor Coolant Flow........................................

2.3-3 2.3.a.5 Steam Generators..............................................

2.3-3 2.3.a.6 Reactor Trip Interlocks........................................

2.3-4 2.3.a.7 Other Trips..........................................................

2.3-4 3.0 Lim iting Conditions for Operation.............................................................................

3.0-1 3.1 Reactor Coolant System............................................................................

3.1-1 3.1.a Operational Com ponents..........................................................

3.1-1 3.1.a.1 Reactor Coolant Pum ps......................................

3.1-1 3.1.a.2 Decay Heat Removal Capability......................... 3.1-1 3.1.a.3 Pressurizer Safety Valves...................................

3.1-2 3.1.a.4 Pressure Isolation Valves...................................

3.1-2 3.1.a.5 Pressurizer PORV and PORV Block Valves....... 3.1-3 3.1.a.6 Pressurizer Heaters............................................

3.1-3 3.1.a.7 Reactor Coolant Vent System.............................

3.1-4 3.1.b Heatup & Cooldown Limit Curves for Normal Operation.................................................................................

3.1-4 3.1.c M axim um Coolant Activity.........................................................

3.1-6 3.1.d Leakage of Reactor Coolant.....................................................

3.1-7 3.1.e Maximum Reactor Coolant Oxygen, Chloride and Fluoride Concentration.............................................................

3.1-8 3.1.f M inim um Conditions for Criticality.............................................

3.1-8 TS i

Section Title Paae 3.2 Chemical and Volume Control System..........................................................