Text

License Amendment Request 204 to Technical Specifications; Extend Reactor Trip System and Engineered Safety Features Actuation System Surveillance Test Intervals as Evaluated in WCAP-15376-P-A
	ML041610046
Person / Time
Site:	Kewaunee
Issue date:	06/01/2004
From:	Coutu T; Nuclear Management Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	NRC-04-064, WCAP-15376-P-A
	Download: ML041610046 (66)
	v • d • e

Commted to NudearExcerl . Kewaunee Nuclear Power Plant Operated by Nuclear Management Company, LLC June 1, 2004 NRC-04-064 10 CFR 50.90 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 Kewaunee Nuclear Power Plant Docket 50-305 License No. DPR-43 License Amendment Request 204 to the Kewaunee Nuclear Power Plant Technical Specifications: Extend Reactor Trip System and Engineered Safety Features Actuation System Surveillance Test Intervals as Evaluated in WCAP-15376-P-A References 1) WCAP-15376-P-A, Revision 1, 'Risk-informed Assessment of the RTS and ESFAS Surveillance Test Intervals and Reactor Trip Breaker Test and Completion Times," dated March 2003.

2) NRC Generic Letter 93-05, "Line-item Technical Specification Improvements to Reduce Surveillance Requirements for Testing During Power Operation," dated September 27,1993.

3) Industry/Technical Specification Task Force (TSTF) Standard TS (STS) Change Traveler 411, Revision 1, 'Surveillance Test Interval Extensions for Components of the Reactor Protection System (WCAP-1 5376)," dated August 7, 2002.

4) NRC Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment In Risk-informed Decisions on Plant-Specific Changes to the Licensing Basis," dated July 1998.

5) NRC Regulatory Guide 1.177, "An Approach for Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications," dated August 1998.

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iDocket 50-305 NRC-04-064 June 1, 2004 Page 2 Pursuant to 10 CFR 50.90, the Nuclear Management Company, LLC, (NMC) is submitting this License Amendment Request (LAR) to the Kewaunee Nuclear Power Plant (KNPP) Technical Specifications (TS) to revise TS 1.0, "Definitions,"

Table TS 3.5-2, 'Instrument Operation Conditions for Reactor Trip," and Table TS 4.1-1,

'Minimum Frequencies for Checks, Calibrations, and Test of Instrument Channels."

The proposed changes include:

adding a definition of 'staggered test basis,"

increasing the surveillance testing interval for the Reactor Protection System (RPS) and Engineered Safety Features Actuation System (ESFAS) analog channels from monthly to semiannually,

increasing the surveillance testing interval for the RPS and ESFAS logic cabinets from monthly to quarterly on a staggered test basis, and

adding a completion time (CT) for the RTBs of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

The above changes are generically evaluated in WCAP-1 5376-P-A (reference 1). All conditions stipulated in the Nuclear Regulatory Commission (NRC) Safety Evaluation Report (SER) are addressed in Enclosure 1 of this letter. The approach used in WCAP-1 5376 is consistent with the NRC's approach for evaluating risk-informed changes to a plant's current licensing basis as presented in Regulatory Guide (RG) 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-informed Decisions on Plant-Specific Changes to the Current Licensing Basis," (reference 4) and RG 1.177, "An Approach for Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications," (reference 5). The proposed changes also implement recommendations from NRC Generic Letter 93-05 (reference 2) and are consistent with applicable sections of NRC-approved traveler Industry/Technical Specification Task Force (TSTF) Standard TS (STS) Change Traveler 411, Revision 1, "Surveillance Test Interval Extensions for Components of the Reactor Protection System (WCAP-15376),"

(reference 3).

Enclosure 1 to this letter contains a description of the proposed changes, historical background, a technical analysis, and a regulatory analysis including a no significant hazards determination and an environmental considerations review. Enclosure 2 contains a proprietary analysis showing the applicability of WCAP-1 5376 to the KNPP. contains the non-proprietary version of the applicability analysis. Enclosure 4 contains the strikeout Technical Specification pages and enclosure 5 contains the affected Technical Specification pages as revised. Enclosure 6 contains a Westinghouse affidavit, proprietary information notice, and copyright notice for . Enclosure 7 contains a list of commitments resulting from this correspondence.

- 'Docket 50-305 NRC-04-064 June 1, 2004 Page 3 As enclosure 2 contains information proprietary to Westinghouse, it is supported by an affidavit (Enclosure 6) signed by Westinghouse, the owner of the information. The affidavit sets forth the basis on which the information may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in paragraph (b) (4) of 10 CFR 2.790 of the Commission's regulations. Accordingly, it is respectfully requested that the information, which is proprietary to Westinghouse, be withheld from public disclosure in accordance with 10 CFR 2.790. Correspondence with respect to the copyright or proprietary aspects of the items listed above or supporting the Westinghouse Affidavit, should reference the appropriate authorization letter and be addressed to J.A. Gresham, Manager of Regulatory Compliance and Plant Licensing, Westinghouse Electric Company, P.O. Box 355, Pittsburgh, Pennsylvania 15230-0355.

The NMC requests approval of the proposed amendment by June 2, 2005. Once approved, the amendment shall be implemented within 90 days. If you have any questions or require additional information, please contact Mr. Gerald Riste at (920)388-8424. A complete copy of this submittal has been transmitted to the State of Wisconsin as required by 10 CFR 50.91(b)(1).

Summary of Commitments This letter makes three commitments as detailed in enclosure 7.

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

Executed on June 1, 2004.

Thomas Coutu Site Vice President Kewaunee Nuclear Power Plant Nuclear Management Company, LLC Enclosures (7) cc: Administrator, Region 1II, USNRC Senior Resident Inspector, Kewaunee, USNRC Project Manager, Kewaunee, USNRC Public Service Commission of Wisconsin

ENCLOSURE I LICENSE AMENDMENT REQUEST 204: DESCRIPTION, PROPOSED CHANGES, BACKGROUND, TECHNICAL ANALYSIS, REGULATORY SAFETY ANALYSIS, ENVIRONMENTAL CONSIDERATIONS, AND REFERENCES

1.0 DESCRIPTION

The Nuclear Management Company (NMC), LLC, proposes to amend operating license DPR-43, Appendix A, "Technical Specifications." The amendment will revise the Kewaunee Nuclear Power Plant (KNPP) Technical Specifications (TS) TS 1.0, "Definitions," Table TS 3.5-2, "Instrument Operation Conditions for Reactor Trip," and Table TS 4.1-1, "Minimum Frequencies for Checks, Calibrations and Test of Instrument Channels." The TS revisions will add a definition for "staggered test basis," increase surveillance test intervals (STIs) for Reactor Protection System (RPS) and Engineered Safety Features Actuation System (ESFAS) analog channels and logic cabinets and add a completion time (CT) for the reactor trip breakers (RTBs).

The above proposed changes have been generically evaluated and approved by the Nuclear Regulatory Commission (NRC) in WCAP-1 5376-P-A, Revision 1, "Risk-informed Assessment of the RTS and ESFAS Surveillance Test Intervals and Reactor Trip Breaker Test and Completion Times," March 2003 (reference 1). All conditions stipulated in the NRC Safety Evaluation Report (SER) for WCAP-15376-P-A are addressed in section 4.0 of this enclosure.

The STI changes requested in this proposed license amendment will reduce the required testing on the RPS reactor trip system (RTS) and ESFAS components without significantly impacting the systems' reliability. The proposed change also will reduce the potential for reactor trips and actuation of engineered safety features associated with the testing of these components. The CT for the reactor trip breakers will provide additional time to complete test and maintenance activities while at power, potentially reducing the number of forced outages related to compliance with reactor trip breaker TSs.

2.0 PROPOSED CHANGE

S The proposed changes in STIs (referred to as "test frequencies" at KNPP) and the RTB CT are based on WCAP-1 5376-P-A. The following changes are proposed specifically for the implementation of the revised STIs and RTB CT of WCAP 15376-P-A at the KNPP.

Markup and clean copies of the proposed changes are located in Enclosures 4 and 5.

The TS bases will be updated appropriately to reflect the changes listed below:

1. Add the definition for staggered test basis from standard TS as TS 1.0.i.6.

2. Change Table TS 3.5-2, "Instrument Operation Conditions for Reactor Trip,"

Item 17, "Reactor Trip Breakers," to include a 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months CT for the RTBs. This will provide additional time to complete test and maintenance activities while at power.

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3. Change the test frequencies or STIs from "monthly" to "semiannual" for the following protective instrumentation channels listed in Table TS 4.1-1, "Minimum Frequencies for Checks, Calibrations and Test of Instrument Channels":

a. Item 4, Reactor Coolant Temperature

b. Item 5, Reactor Coolant Flow

c. Item 6, Pressurizer Water Level

d. Item 7, Pressurizer Pressure

e. Item 8a, 4-KV Voltage and Frequency

f. Item 1I a, Steam Generator Low Level

g. Item I b, Steam Generator High Level

h. Item 12, Steam Generator Flow Mismatch

i. Item 18a, Containment Pressure (SIS signal)

j. Item 18b, Containment Pressure (Steamline Isolation)

k. Item 18c, Containment Pressure (Containment Spray Actuation)

I. Item 23, Steam Generator Pressure

4. Change the test frequency or STI from "monthly" to "quarterly" for Item 1, Nuclear Power Range, listed in Table TS 4.1-1, "Minimum Frequencies for Checks, Calibrations and Test of Instrument Channels."

5. Change the test frequency or STI from "monthly" to "quarterly" and add a note "on a staggered test basis" in the remarks column for Table TS 4.1-1, Item 26, Protective System Logic Channel Testing.

KNPP is not adopting the following TS changes that were included in the generic evaluation of WCAP-15376:

1. The two hour bypass time for RTBs from WCAP-1 5376 will not be amended into KNPP TSs. Kewaunee's current licensing basis is for an eight hour bypass time for the RTBs for test or maintenance (reference 6).

2. The change to the RTB test frequency from "monthly" to "bimonthly on a staggered test basis" will not be amended into the KNPP TSs. In the WCAP analysis, the RTB test frequency change was from two months to four months.

Kewaunee's current RTB test frequency is one month and, therefore, the WCAP analysis is not applicable (i.e., two months as the starting point for the risk assessment in comparison to one month).

The proposed TS changes do not involve changes to actuation setpoints, setpoint tolerance, testing acceptance criteria, or channel response times. No hardware changes are proposed nor required to implement these changes at the KNPP.

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3.0 BACKGROUND

In the early 1980s, in response to growing concerns of the impact of the current testing and maintenance requirements on plant operation, the Westinghouse Owners Group (WOG) initiated a program to develop a justification to be used to revise generic and plant specific instrumentation TS (part of the Technical Specification Optimization Program (TOPS)). Operating plants experienced many inadvertent reactor trips and safeguards actuations during performance of instrumentation surveillances, causing unnecessary transients and challenges to safety systems. Significant time and effort on the part of the operating staff was devoted to performing, reviewing, documenting, and tracking the various surveillance activities, which in many instances seemed unwarranted based on the high reliability of the equipment. Significant benefits for operating plants appeared to be achievable through revision of instrumentation test and maintenance requirements. The results of the WOG studies, and the recommended changes to the testing of reactor protection and engineered safeguards instrumentation, were documented in WCAP-10271-P-A and WCAP-10271-P-A, Supplement 1, both titled, "Evaluation of Surveillance Frequencies and Out of Service Times for the Reactor Protection Instrumentation System," (references 2 and 3), and in WCAP-10271-P-A, Supplement 2, Revision 1, "Evaluation of Surveillance Frequencies and Out of Service Times for the Engineered Safety Features Actuation System," (reference 4).

In February 1985, the NRC issued the SER (reference 5) for WCAP-10271 and WCAP-10271, Supplement 1. The SER approved quarterly testing, six hours to place a failed channel in a tripped mode, increased CTs (also referred to as allowed outage times (AOTs)) for test and maintenance, and testing in bypass for analog channels of the RTS. The quarterly testing had to be conducted on a staggered basis. The SER specifically stated that for analog channels shared by the RTS and the ESFAS, the approved relaxations applied only to the RTS function.

On March 20, 1986, the WOG submitted WCAP-1 0271, Supplement 2, "Evaluation of Surveillance Frequencies and Out of Service Times for the Engineered Safety Systems Actuation System." On May 12,1987, the WOG submitted WCAP-10271, Supplement 2, Revision 1. Supplement 2 and Supplement 2, Revision 1 specifically demonstrated the applicability of the justification contained in WCAP-10271 to the ESFAS for two, three, and four loop plants with either relay or solid state protection systems. In February 1989, the NRC issued the SER for WCAP-1 0271, Supplement 2 and WCAP-1 0271, Supplement 2, Revision 1 (reference 7). The SER approved quarterly testing, six hours to place a failed channel in a tripped mode, increased CTs for test and maintenance and testing in bypass for analog channels of the ESFAS. Staggered testing was not required for ESFAS analog channels and the requirement was removed from the RTS analog channels.

The NRC issued a Supplemental SER (SSER) for WCAP-1 0271, Supplement 2 and Supplement 2, Revision 1(reference 8) on April 30, 1990. With the issuance of the SER and SSER, the relaxations for the analog channels of the RTS and ESFAS were the same. Additionally, the CTs for test and maintenance of the RTS and ESFAS actuation logic were also the same.

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In 1992, the NRC completed an evaluation of surveillance testing at power, which indicated that testing in many areas could be reduced without any significant decrease in safety. These findings and recommendations are documented in NUREG-1 366, "Improvement to Technical Specifications Surveillance Requirements," (reference 9) and Generic Letter 93-05, "Line-Item Technical Specifications Improvements to Reduce Surveillance Requirements for Testing During Power Operation" (reference 10).

Reduced surveillance testing of the RPS and ESFAS analog instrumentation was recommended in both of these documents.

In June 1995, the WOG submitted WCAP-14333, "Probabilistic Risk Analysis of the RTS and ESFAS Test Times and Completion Times," Revision 0. The report proposed further relaxation of the WCAP-10271 approved TS requirements by increasing the test bypass times (BTs) and the CTs for both the solid state protection system and relay protection system RTS and ESFAS designs.

In WCAP-14333, the WOG evaluated the impact on core damage frequency (CDF) and public risk of additional time for testing and extended CTs for the RPS and ESFAS. The additional time allowance was an extension from 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months for the test CT for analog channels, an extension from 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for the allowed bypass time for analog channels, a CT change from 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for logic cabinet, and a CT change from 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for the master relays. This WCAP did not propose addition extensions of any STIs. On July 15, 1998, the NRC completed a review of the WCAP-14333 (reference 13) for reference in license applications based on meeting conditions listed in the NRC SER (reference 11). The WOG issued implementation guidance for WCAP-14333 on December 2, 1998 (reference 12).

By letter dated November 8, 2000, the WOG submitted WCAP-1 5376, Revision 0, "Risk-informed Assessment of the RTS and ESFAS Surveillance Test Intervals and Reactor Trip Breaker Test and Completion Times." Topical report WCAP-1 5376 provides the technical justification for increasing the CT and BT for the RTB. It also provides the justification for increasing the surveillance test interval for the RTB, analog channels, and logic cabinets for components of the RTS and ESFAS. The approach used in this program is consistent with the NRC approach for using probabilistic risk assessment in risk-informed decisions on plant specific changes to the current licensing basis as presented in Regulatory Guide (RG) 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Current Licensing Basis," (reference 14) and RG 1.177, "An Approach for Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications," (reference 15).

The NRC approved WCAP-15376 for referencing in license change submittals on December 20, 2002 (reference 16). The WOG issued implementation guidance for WCAP-1 5376 on April 3, 2003 (reference 17). The WOG reissued guidance with slight editorial changes on May 6, 2004 (reference 22).

The NRC required that applicants for TS amendments concerning all of the above WCAPs meet certain plant specific conditions. The plant specific conditions are stipulated in the SERs and SSER for all WCAP and supplements discussed above.

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None of the generically approved changes to CTs, BTs, or STIs of WCAP-10271 and its supplements, or WCAP-1 4333 have specifically been incorporated into the KNPP TS.

The NMC will address the stipulated conditions of the WCAP-1 0271 and WCAP-14333 SERs and SSER for KNPP in the technical analysis below since the stipulated plant specific conditions of the previous reports constitute a basis for the changes evaluated in WCAP-1 5376. The evaluations are presented in Section 4.0 under the appropriate WCAP heading.

4.0 TECHNICAL ANALYSIS

Current Design Basis The RPS continuously monitors selected process variables associated with fission product design barrier limits that define the boundaries for safe reactor power operation.

The RPS trips the reactor and returns the core to a subcritical condition if the value of any single monitored process variable approaches its associated barrier design limits.

The system consists of electronic equipment (circuitry, cables, relays, etc.) necessary to monitor the selected process variables and to generate the reactor trip signals when a design limit is challenged. If a reactor trip is required, two reactor trip breakers are actuated by two separate logic matrices that interrupt power to the control rod drive mechanisms (CRDMs). Opening either reactor trip breaker interrupts power to all CRDMs, allowing free fall into the core. The RPS also monitors other plant systems for conditions or events that could cause barrier design limits to be challenged.

The RPS is designed for high functional reliability and in-service testability to avoid undue risk to the health and safety of the public. Protection channels required for power operation above permissive P-1 0 are designed with sufficient redundancy to allow individual channel calibration. Tests are made by use of signal substitution techniques during power operation without negating reactor protection. Removal of one trip channel is accomplished by placing that channel in the tripped mode. Therefore, a two-out-of-three channel becomes a one-out-of-two channel when testing. Redundancy and independence designed into the RPS must be sufficient to ensure that no single failure or removal from service of any component or channel of the system results in a loss of the protection function. The redundancy provided must include, as a minimum, two channels of protection for each protection function to be served.

Current Licensing Basis The current STI or test frequency for RPS and ESFAS instrumentation, including the analog channels and actuation logic, is monthly. There currently is no CT specified in the KNPP TS for the RTBs. RTB test and maintenance activities are currently completed within the eight hour bypass time allowed by TS.

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Basis for the ChanQes The technical basis for the proposed changes to the STIs and the RTB CT include the WOG studies documented in the originals, supplements, and revisions of WCAP-10271, WCAP-14333, and WCAP-1 5376 (references 1, 2, 3, 4, and 13), which have all been reviewed and approved by the NRC staff. Additionally, the NRC staff recommended the TSs be changed in NUREG-1366, "Improvements to Technical Specifications Surveillance Requirements," (reference 9) and GL 93-05, "Line-item Technical Specifications Improvements to Reduce Surveillance Requirements for Testing During Power Operation," (reference 10). Standard TS Change Travelers (TSTF-41 1 and TSTF-418), which incorporate the changes from WCAP-14333 and WCAP-1 5376 into Standard Technical Specifications (STS), have been issued by NEI and reviewed and approved by the NRC (references 20 and 21).

Deviations from the surveillance testing frequencies of WCAP-1 5376 and the TSTFs require justification. The NMC has requested that Item 1, Nuclear Power Range, of Table TS 4.1-1 be changed to a quarterly frequency rather than the NRC accepted frequency of 184 days. This change is based on the current KNPP TS requirement for the quarterly calibration of Item 1, Nuclear Power Range. The KNPP TS definition, TS 1.0.i.3, "Channel Calibration," specifies that, "Calibration ... shall be deemed to include the Channel Function Test." Channel calibrations are not within the scope of WCAP-1 5376, and therefore, the channel calibration for this item will continue to be performed on a quarterly frequency. The channel functional test required by Item 1 was proposed as quarterly to be consistent with the required quarterly channel calibration surveillance. This will allow both the channel calibration and functional test to be performed at the same time. There is no net increase in risk associated with the proposed deviation since the channel functional test is already required as part of the channel calibration and the frequency of the channel calibration will not be changed.

The STI, CT, and bypass time changes in WCAP-10271 were justified for a large number of RPS and ESFAS signals that are common to most plants. The STIs, CTs, and bypass times for signals not specifically addressed in WCAP-1 0271 cannot be changed based on WCAP-1 0271 without a plant specific technical justification. The NMC reviewed WCAP-1 0271 to ensure the signal applicability and found that two logic configurations required plant specific justification to apply WCAP-10271. The following technical justifications allow applying the changes justified in WCAP-1 0271, and subsequently WCAP-14333 and WCAP-1 5376, to the two configurations described below:

1. Low Flow Both Loops (KNPP Table TS 3.5-2, item number 10)

WCAP-1 0271, Supplement 1, Table 3.2-3, "Results of Fault Tree Analysis for A Relay Logic Reactor Protection System," does not address the loss of flow in both loops trip for a two loop plant. For KNPP's configuration, the logic is 2 of 3 in 2 of 2 loops, a failure in either loop would cause a failure to trip. Thus, the failure to trip probability can be calculated by doubling the value for an individual loop, which is documented in Table 3.2-3 of the WCAP. The following shows the results of this justification:

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Table 4.1 Failure to Trip Probability for Comparison Base Case, no CC(1) Case 1, no CC l Base Case, CC l Case 1, CC 1.5x10-4 3.8x10 l 3.2x10 l 6.4x10-4 (1) CC = common cause These values are still within the range of the signals in Table 3.2-3 of the WCAP.

The bases for the conclusions of WCAP-10271, Supplement 1 (less than a factor of six increase with no common cause and less than a factor of four increase with common cause) apply to this signal as well. Therefore, the conclusions of WCAP-1 0271, Supplement 1, that an increase on surveillance interval is acceptable, apply to the KNPP low flow both loops trip as well.

2. Containment Spray (Table TS 3.5-3, item number 3.b)

The containment spray logic in the WCAP-10271, Supplement 2 analysis is two-out-of-four, whereas the logic at KNPP is one-out-of-two three times. An increase in the failure probability of the containment spray signal is of very little consequence at KNPP due to robust design of Kewaunee's large dry containment. In the current Kewaunee PRA model, a complete failure of the containment spray system results in a CDF and LERF increase of 0.06 percent and 0.6 percent, respectively. As a result, the failure of the initiating signal for containment spray is of very low risk significance and an increased surveillance frequency is of very little consequence at Kewaunee.

Based on the above, the two KNPP specific configurations are justified to be of low risk significance. The changes justified in WCAP-1 0271, and subsequently in WCAP-1 4333 and WCAP-1 5376, can be applied to KNPP.

Implementation of the proposed changes does not reduce safety. The proposed STI changes will reduce the required testing on the RTS and ESFAS components without significantly impacting their reliability, and reduce the potential for reactor trips and actuation of engineered safety features associated with more frequent testing of these components. The CT extension for the RTBs provides additional time to complete test and maintenance activities while at power, potential reducing the number of forced outages related to compliance with the RTB TS. Additionally, NUREG-1 366 and GL 93-05 both support this proposed change by stating that safety can be improved, equipment degradation decreased, and unnecessary burden on personnel resources eliminated by reducing the amount of testing that the TSs require during power operation.

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Technical Evaluation for Changes in WCAP-10271 -P-A In February 1985, the NRC issued the SER (reference 5) for WCAP-10271 and WCAP-1 0271, Supplement 1. In February 1989, the NRC issued the SER for WCAP-1 0271, Supplement 2 and WCAP-1 0271, Supplement 2, Revision 1 (reference 7). The following table provides a summary of the changes from WCAP-10271 and its supplements.

Table 4.2 Summary of WCAP-10271 (TOP) RTS and ESFAS Surveillance Test Interval, Completion Ti e, and Bypass Time Changes - Relay Protection Systems Component STI Completion Time Bypass Time Analog Channel 1 month to 3 months 1 hours1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Logic Cabinet No change 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Master Relay No change No change 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Slave Relay No change No change 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Reactor Trip No change No change No change Breakers In WCAP-10271, and its supplements (references 2, 3, and 4), the WOG evaluated the impact of increased STIs and CTs and their effect on core damage frequency (CDF) and public risk. The NRC staff concluded in its evaluation that an overall upper bound of CDF increase due to the proposed STI and CT changes is less than six percent for Westinghouse pressurized water reactor (PWR) plants. The NRC also concluded actual CDF increases for individual plants were expected to be substantially less than six percent. The NRC Staff considered this CDF increase to be small compared to the range of uncertainty in the CDF analyses and, therefore, was acceptable.

To incorporate the extended times from WCAP-1 0271 and its supplements, the NRC required that an applicant for a TS amendment meet certain plant specific conditions stipulated in the SERs and SSER. The five conditions in the NRC SER dated February 21, 1985 (reference 5) were to be applied to RPS instrumentation. The two conditions in the NRC SER and SSER dated February 22, 1989 (reference 7), and April 30, 1990 (reference 8), were to be applied to the ESFAS instrumentation.

The generically approved changes to the analog channel STIs, CTs, and BTs specified in WCAP-10271, including its supplements, have not been amended into the KNPP TS.

Additionally, the NMC does not plan to adopt the changes for CT and BT from WCAP-10271 since these items are not in KNPP TS. The NMC has addressed each SER condition below since these stipulated plant specific conditions constitute part of the basis for the changes evaluated in WCAP-1 5376 (specifically, the analog channel STI from one month to three months is adopted with KNPP's proposed change from one month to six months).

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1. The RPS SER requires the use of a staggered test plan for the RPS channels changed to the quarterly testing frequency.

NMC Response: The NRC Staff subsequently concluded that a staggered test strategy need not be implemented for ESFAS analog channel testing and is no longer required for RPS analog channel testing (reference 7). This NRC conclusion was based on the small relative contribution of the analog channels to RPS/ESFAS unavailability, process parameter signal diversity, and normal operational channel testing spacing.

2. The RPS SER requires that plant procedures require a common cause evaluation for failures in the RPS analog channels changed to quarterly testing frequency and additional testing for plausible common cause failures.

NMC Response: In accordance with KNPP's Action Request (AR) Process, Corrective Action Program (CAP), and existing plant procedures, all equipment operability concerns (including equipment in the RPS and ESFAS) are immediately reported to the Shift Manager and entered into the AR Process.

Initiation of an AR results in entry into the CAP. In the CAP AR process, the request is screened for operability and reportability as well as whether or not a root cause, apparent cause, condition, or maintenance rule evaluation is required. An apparent cause evaluation looks at the "extent of condition" to determine if the failure mechanism could be common to other plant equipment.

KNPP will incorporate an administrative control into the CAP AR process requiring an apparent cause evaluation for RPS and ESFAS failures to determine the extent of condition.

3. The RPS SER requires installed hardware capability for testing in the bypass mode. That approval of routine channel testing in a bypassed condition is contingent on the capability of the RPS design to allow such testing without lifting leads or installing temporary jumpers.

NMC Response: KNPP's RPS is designed to test analog channels in the trip mode. Testing is performed by placing the channel being tested in the tripped mode rather than bypassing the channel. The result is the logic being reduced to one-out-of-two logic for a two-out-of-three logic channel and one-out-of-three logic for a two-out-of-four logic channel. If a channel failure occurs and the failed channel is not the one being tested, the logic is reduced to one-out-of-one. This logic results in permitting the remaining operable channel to trip the reactor if necessary. A channel failing to a trip condition during testing of another channel results in a reactor trip.

Testing in bypass is allowed by TS 3.5.b and TS 3.5.d for a short period of time (approximately 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ) for on-line testing or in the event of failure of a subsystem instrumentation channel where the failed channel must be blocked to prevent unnecessary reactor trip. The TS would apply during a maintenance Page 9 of 23

activity. Therefore, it is acceptable to use temporary jumpers as required to complete testing. The TS is not routinely entered for surveillance testing.

4. The RPS SER indicated that, for channels that provide input to both the RPS and the ESFAS, the more stringent ESFAS requirements still apply.

NMC Response: The extensions generically approved in the SER and SSER for the ESFAS analog channels (references 7 and 8) were the same as those approved for the RPS analog channels. Therefore, this condition from the RTS SER is no longer applicable.

5. The RPS SER requires confirmation that the instrument setpoint methodology includes sufficient margin to offset the drift anticipated as a result of less frequent surveillance.

NMC Response: This condition of WCAP-10271 is also a stipulated condition of WCAP-1 5376 and will be addressed in the section of the technical analysis of WCAP-15376.

6. The ESFAS SER and SSER required confirmation of the applicability of the generic analyses to the plant.

NMC Response: The generic analyses used in WCAP-10271 and its supplements are applicable to KNPP. Kewaunee Nuclear Power Plant uses the Foxboro H-Line Process Control System and the Westinghouse Relay Protection System for both the Engineered Safety Features and Reactor Protection System.

Both of these systems were modeled in the generic analyses. For logic configurations not modeled in the generic analysis, the NMC has provided a technical justification for applying the generic changes to KNPP. Additionally, information provided in enclosures 2 and 3 demonstrates the applicability of the generic WCAP analysis to KNPP's ESFAS and RTS. These tables are based on implementation guidelines that were issued by the WOG for licensees implementing the TS CT and BT changes that were justified in WCAP-14333 and the STI and CT changes justified in WCAP-1 5376.

7. The ESFAS SER and SSER required confirmation that any increase in instrument drift due to the extended STIs is properly accounted for in the setpoint calculation methodology.

NMC Response: This condition of WCAP-10271 is also a stipulated condition of WCAP-1 5376 and will be addressed in the technical analysis of WCAP-1 5376.

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Technical Evaluation for Changes in WCAP-14333-P-A In WCAP-14333, the WOG proposed and justified further extensions of the RTS and ESFAS CTs and BTs. This WCAP did not propose additional extensions of any STIs.

The NRC approved these generic CT and BT extensions in an SER dated July 15, 1998 (reference 11). The table below summarizes the changes evaluated in WCAP-14333.

Table 4.3 Summary of WCAP-14333 RTS and ESFAS Completion Time and Bypass Test Time Changes - Relay Protection Systems Component STI Completion Time Bypass Time Analog Channel No change 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 4 hours to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Logic Cabinet No change 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months No change Master Relay No change 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months No change Slave Relay No change 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months No change Reactor Trip No change No change No change Breakers The SER for WCAP-14333 indicated that the increase in CDF and Large Early Release Frequency (LERF) for those plants that have not implemented the changes evaluated in WCAP-10271, and its supplements, is small. Specifically, for two-out-of-three and two-out-of-four logic the increases in CDF are approximately 3.1 percent and 2.3 percent, respectively. The LERF would increase by only four percent for both two-out-of-three and two-out-of-four logic schemes. The NRC staff concluded the implementation of the changes specified in WCAP-14333 would result in a very small quantitative impact on plant risk.

The NRC staff required that an applicant for a proposed amendment incorporating the extended times into their TS must meet certain plant specific conditions stipulated in the SER. The generically approved changes to CTs and BTs specified in WCAP-14333 have not been amended into the KNPP TS because KNPP currently does not contain any CTs or BTs for these components in TS. Additionally, KNPP will not be amending these component CTs and BTs into TS with this change. However, since these plant specific conditions share a commonality with the changes evaluated in WCAP-1 5376, and the WOG guidance is similar for both WCAPs, the NMC has addressed each of the conditions stipulated in the WCAP-14333 SER below.

1. Confirm the applicability of the WCAP-14333 analyses for the plant.

NMC Response: The information provided in enclosures 2 (proprietary) and 3 (non-proprietary) demonstrates the applicability of the generic WCAP-14333 and WCAP-15376 analysis to KNPP. The tables in the enclosures are from the implementation guidelines issued by the WOG for licensees implementing the TS changes supported by the WCAPs.

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2. Address the Tier 2 and 3 analyses including the Configuration Risk Management Program (CRMP) insights which confirm that these insights are incorporated into the decision making process before taking equipment out of service.

NMC Response: This stipulated condition is addressed below in the technical analysis section for WCAP-1 5376.

Technical Analysis for Changes in WCAP-15376-P-A In WCAP-1 5376, the WOG provided the technical basis to justify extending the STIs for the analog channels, the logic cabinets, and RTBs, and for extending the CT and BT for the RTBs. The NRC approved WCAP-1 5376 on December 20, 2002 (reference 16).

The table below summarizes the changes evaluated in WCAP-1 5376.

Table 4.4 Summary of WCAP-15376 RTS and ESFAS Surveillance Test Frequency, Completion Time, and Bypass Test Time Changes - Relay Protection Systems Component STI Completion Time Bypass Time Analog Channel 3 months to 6 No change No change months Logic Cabinet 1 month to 6 months No change No change Master Relay No change No change No change Slave Relay No change No change No change Reactor Trip 2 months to 4 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Breakers months The KNPP STIs or test frequencies are closely aligned with the pre-technical specification optimization program (pre-TOP) values since the extensions generically approved in WCAP-10271, and its supplements, and WCAP-14333 have not been incorporated into the KNPP TS. The tables below compare the current KNPP TS times with the pre-TOP times, and the WCAP-1 5376 times (i.e., the proposed TS times).

Table 4.5 STI Comparison for KNPP Component Current STI Pre-TOP STI WCAP-15376 (Proposed TS)

Analog Channels 1 month 1 month 6 months a)

Logic Cabinets I month 2 months 6 months Reactor Trip 1 month 2 months 4 monthsbT)

Breakers Notes:

(a) With the exception of the Nuclear Power Range at KNPP, which will be changed to a quarterly frequency as described earlier in this document.

(b) NMC will not include the RTB STI change because the WCAP analysis for this change is not applicable to KNPP.

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Table 4.6 CT Comparison for KNPP Component Current TS Pre-TOP WCAP-15376

_(Proposed TS)

Reactor Trip Breakers 0 hours0 days 0 hours 0 weeks 0 months a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Notes:

(a) The current KNPP TS does not provide for a restoration time.

The risk-informed approach used in WCAP-1 5376 is consistent with the NRC approach for using probabilistic risk assessment in risk-informed decisions on plant-specific changes to the current licensing basis. The NRC's approach is presented in RG 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Current Licensing Basis," (reference 14) and RG 1.177, "An Approach for Plant-Specific, Risk-Informed Decision making: Technical Specifications," (reference 15). The risk evaluation considers the three-tiered approach as presented in RG 1.177. Tier 1, "PRA Capabilities and Insights," assesses the impact of the proposed CT (AOT) change on CDF, incremental conditional core damage probability (ICCDP), LERF, and incremental conditional large early release probability (ICLERP). Tier 2, "Avoidance of Risk-Significant Plant Configurations," considers potential risk-significant plant operating configurations. Tier 3, "Risk-Informed Plant configuration Control and Management," considers risk evaluations of configurations when entered on a plant-specific basis.

Tier 1, Core Damage Frequency Assessment WCAP-1 5376 compares the cumulative impact of the proposed STI changes and the RTB CT changes on CDF using the pre-TOP values as the basis. This comparison credits the expected reduction in reactor trips due to the reduced analog channel testing resulting from extending the analog channel STIs from monthly to quarterly, as evaluated in WCAP-10271 and its supplements. The comparison indicates that the cumulative impact on CDF when using the pre-TOP values as the base case is 5.7E-07 per year for two-out-of-four logic and 1.1 E-06 per year for two-out-of-three logic. These are small increases to the CDF per the acceptance criteria of 1.OE-06 per year given in RG 1.174.

Tier 1. Incremental Conditional Core Damage Probability The ICCDP calculations are a direct function of the duration of a CT, and therefore, only apply to the CT change for the reactor trip breakers. Calculations were performed considering two CTs; 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months for maintenance and 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for a test. The resulting calculated values were below 5E-07 for both CTs. The WCAP-1 5376 calculation of the ICCDP increase for the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months CT (i.e., the proposed CT for KNPP) was 6.92E-08, which is below the RG 1.174 value of 5.OE-07. This is considered very small for a single TS CT per RG 1.177.

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Tier 1, Large Early Release Frequency Assessment The WCAP-1 5376 base case used LERF values of 2.38E-06 per year for two-out-of-four logic and 2.44E-06 per year for two-out-of-three logic. WCAP-1 5376 documented the combined impact on LERF due to the changes proposed by the WCAP. The LERF impact was an increase of 3.09E-08 per year for two-out-of-four logic and 5.68E-08 for two-out-of-three logic. These increases in LERF are small based on the RG 1.174 guidance of I.OE-07 per year. Therefore, the proposed changes are acceptable.

Tier 1, Incremental Conditional Large Early Release Probability Assessment Detailed calculations to determine the impact on ICLERP for the proposed changes are not required. For the proposed changes, ICLERP calculations only apply to the RTBs because they are the only components for which the CT is being extended. Reactor trip breakers are used to mitigate core damage, not containment failure. Reactor trip breakers success or failure has no direct impact on the functioning of containment systems. Large releases are related to containment bypass events, containment isolation failures, and containment failures. Reactor trip breaker success or failure has no direct bearing on these functions. The extended RTB CT will result in a slight increase in frequency of some core damage sequences. The LERF will increase only in direct proportion to the increased frequency of core damage sequences involving RTB failures because the success or failure of the containment systems is independent of the reactor trip breakers. Therefore, because the impact of the reactor trip breaker CT increase on CDF and LERF is small and the ICCDP is acceptable, the ICLERP will also be acceptable.

WCAP-1 5376, Revision 1, does contain calculated values for ICLERP for an RTB out of service. The calculation was used to answer an NRC request for additional information (RAI) (reference 8) during the review of the WCAP. The ICLERP for an RTB out of service for a completion time was incorporated into the WCAP report.

Tier 2, Avoidance of Risk-Significant Plant Configurations The Tier 2 requirements of RG 1.177 state that the licensee should provide reasonable assurance that risk-significant plant equipment outage configuration will not occur when specific plant equipment is out of service. Tier 2 requires an examination of the need to impose additional restrictions when operating under the proposed RTB CT such that risk-significant equipment outage configurations are avoided.

The Tier 2 requirements of RG 1.177 have been addressed at KNPP. The KNPP currently has in place a risk-informed on-line risk management process, which supports the requirements of 10 CFR 50.65(a)(4). This risk-informed assessment process is governed and implemented by plant procedures. These procedures assure that the risk associated with the various plant configurations planned during power conditions are assessed and appropriately managed. Additionally, KNPP performs a shutdown safety assessment during scheduled and unscheduled outages that includes an independent review of the plant outage schedule and performance of a safety assessment checklist.

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The WCAP-1 5376 identified the following restrictions on concurrent removal of certain equipment when an RTB is out of service. The recommended Tier 2 restrictions are provided in Section 8.5 of WCAP-1 5376.

Activities that could degrade the availability of the auxiliary feedwater system, reactor coolant system pressure relief (pressurizer PORVs and safety valves), AMSAC (ATWS (Anticipated Transient Without Scram)

Mitigating System Actuation Circuitry), or turbine trip should not be scheduled when a RTB is out of service.

Activities that could degrade other components of the RPS, including master relays or slave relays and activities that cause analog channels to be unavailable should not be scheduled when a logic cabinet is unavailable.

Activities on electrical systems that support the systems or functions listed in the first two bullets should not be scheduled when a RTB is unavailable.

KNPP will implement administrative controls to include the above restrictions when an RTB or a logic cabinet is removed from service or the diverse scram system (DSS) will be operable. The DDS is initiated on a signal from the existing AMSAC system and de-energizes the rod drive MG set exciter field. Removing the rod drive MG set exciter field will interrupt power to the control rod grippers, allowing the control rods to free fall into the core, ending the ATWS event.

Tier 3, Risk-Informed Plant Configuration Control and Management The objective of the third-tier requirements is to ensure that the risk impact of out-of-service equipment is evaluated prior to performing any maintenance activity. The third-tier requirement is an extension of the second-tier requirement, but addresses the limitation of being able to identify all possible risk-significant plant configurations in the second-tier evaluation. As with Tier 2, Tier 3 requirements of RG 1.177 have been addressed at KNPP through administrative controls (procedures and guidelines) used to support the Maintenance Rule requirements specified by the NRC in 10 CFR 50.65(a)(4).

WCAP-15376 SER Conditions Kewaunee is implementing the changes in the analog channel and logic cabinet STIs and the CT for the RTBs. Kewaunee is not implementing the changes to the RTB bypass time or the RTB test frequency (STI). As with the other WCAP SERs, the NRC SER approving WCAP-1 5376 contained several plant specific conditions that require evaluation prior to implementation. The conditions of the WCAP-1 5376 SER are evaluated by the NMC for the KNPP below.

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1. Confirm the applicability of the topical report to the plant and perform a plant-specific assessment of containment failures and address any design or performance differences that may affect the proposed changes.

NMC Response: In order to address condition 1 for WCAP-14333 and WCAP-1 5376, the WOG issued implementation guidelines to help licensees confirm the WCAP analyses were applicable to their plants. Tables 1 through 5 in Enclosure 2 list the important parameters and assumptions made in the generic analyses that are relevant to the requested changes. The information presented in Tables I through 5 confirms the applicability of both WCAP-14333 and WCAP-1 5376 analyses to KNPP.

Component Failure Probability Component failure probability data used in WCAP-1 5376 report was reviewed against KNPP specific data. The KNPP specific component failure probability was 7.33E-06 for input logic relays. This calculated probability is less than that calculated and reported in Section 8.2, Table 8.6, of the WCAP. Therefore, it was determined that the WCAP data is representative of KNPP.

Containment Failure Assessment The LERF analysis completed to support WCAP-1 5376 was based on a large dry containment with LERF contributions from containment isolation failure and containment bypasses from an Interfacing Systems LOCA (ISLOCA) and steam generator tube rupture (SGTR) events, excluding steam generator (SG) tube creep rupture. KNPP's large dry containment is similar to that of the reference plant, and therefore, the WCAP results are applicable. Additionally, in the June 2002 Kewaunee PRA Peer Review, the Kewaunee Level 2 PRA model was evaluated against the Nuclear Energy Institute's (NEI) Peer Review Guidance (NEI-00-02). Two category B facts and observations were generated. These items have been resolved and with these enhancements to the KNPP LERF model, the KNPP Level 2 analysis supports risk significance evaluations with deterministic inputs.

2. Address the Tier 2 and Tier 3 analyses including risk significant configuration insights and confirm that these insights are incorporated into the plant-specific configuration risk management program.

NMC Response: See the discussion regarding Tier 2 and Tier 3 requirements on pages 14 and 15 of this enclosure. KNPP commits to incorporating the recommended restrictions from WCAP-1 5376 into the appropriate administrative controls.

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3. The risk impact of concurrent testing of one logic cabinet and associated reactor trip breaker needs to be evaluated on a plant-specific basis to ensure conformance with the WCAP-15376-P, Rev. 0 evaluation, and RGs 1.174 and 1.177.

NMC Response: The response to NRC RAI 4 (reference 18) provided the ICCDP for this configuration (both the logic cabinet and associated RTB out of service) for preventive maintenance for a total time of 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months , which is comprised of a CT of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , plus 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to reach Mode 3 (Hot shutdown).

The ICCDP for a duration of 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months in this configuration is 3.2E-07, which meets the RG 1.177 (reference 15) acceptance criteria of 5E-07. Bypassing one logic cabinet and associated RTB for 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months of testing will also meet the RG 1.177 ICCDP guideline since the ICCDP value above is based on the logic cabinet and RTB being out of service for 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months at the same time.

This condition is addressed by the above discussion and by demonstration that the WCAP-15376 is applicable to the KNPP. Enclosure 2 contains the evaluation showing that the generic analysis of the WCAP covers KNPP and that the generic risk measures calculated are a good representation for the KNPP.

4. To ensure consistency with the reference plant, the model assumptions for human reliability in WCAP-15376-P, Rev. 0 should be confirmed to be applicable to the plant-specific configuration.

NMC Response: See Enclosures 2 (proprietary) and 3 (non-proprietary),

Table 5.

5. For future digital upgrades with increased scope, integration and architectural differences beyond that of Eagle 21, the staff finds the generic applicability of WCAP-1 5376-P, Rev. 0 to future digital systems not clear and should be considered on a plant-specific basis.

NMC Response: The applicability of the changes justified in WCAP-1 5376 to future digital systems is not addressed in the WCAP and will need to be addressed separately for new designs. Condition 5 does not apply to the KNPP at this time.

Additional commitment from NRC RAI:

WOG guidelines (reference 17) for implementation of WCAP-15376 impose an additional commitment from the response to NRC RAI Question 18 (reference 19) requires that each plant will review their setpoint calculation methodology to ascertain the impact of extending the COT Surveillance Interval from 92 days to 184 days.

NMC Response: The WOG response to this NRC RAI (reference 19) noted that plant-specific RPS and ESFAS setpoint uncertainty calculations and assumptions, Page 17 of 23

including instrument drift, will be reviewed to determine the impact of extending the surveillance interval of the channel functional test from 92 days to 184 days.

An analysis of the drift characteristics of the KNPP plant specific data indicated that extending the bistable testing from one month to the requested interval of six months would remain acceptable (i.e., within the assumptions of the setpoint study). However, this expectation will be validated using future surveillance results subsequent to changing the interval to six months.

5.0 REGULATORY SAFETY ANALYSIS 5.1 No Significant Hazards Consideration The Nuclear Management Company (NMC), LLC, proposes to amend Appendix A of the operating license DPR-43, "Technical Specifications." The amendment will revise the Kewaunee Nuclear Power Plant (KNPP) Technical Specifications (TS)

TS 1.0, "Definitions," Table TS 3.5-2, "Instrument Operation Conditions for Reactor Trip," and Table TS 4.1-1, "Minimum Frequencies for Checks, Calibrations and Test of Instrument Channels." The TS revisions will increase channel operation test (COT) surveillance test intervals (STls) (referred to as "test frequencies" at KNPP) for the Reactor Protection System (RPS) and Engineered Safety Features Actuation System (ESFAS) analog channels and logic cabinets and will add a completion time (CT) for the reactor trip breakers (RTBs). The proposed changes have been generically evaluated and approved by the Nuclear Regulatory Commission (NRC) in WCAP-1 5376-P-A, Revision 1, 'Risk-informed Assessment of the RTS and ESFAS Surveillance Test Intervals and Reactor Trip Breaker Test and Completion Times," March 2003.

The NMC has evaluated whether or not a significant hazards consideration is involved with this proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, Issuance of Amendment," as discussed below:

1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

No. The proposed changes to the STIs and the RTB CT reduce the potential for inadvertent reactor trips and spurious actuations, and therefore, do not increase the probability of an accident previously evaluated.

The proposed changes will not result in a significant increase in the risk of plant operation as demonstrated in WCAP-1 5376-P-A. The impact of plant safety as measured by core damage frequency (CDF) is less than 1.0E-06 per year and the impact of large early release frequency (LERF) is less than 1.OE-07 per year. For the addition of the RTB CT, the incremental conditional core damage probabilities (ICCDP) and incremental conditional large early release probabilities (ICLERP) are less than 5.OE-08. These changes meet the acceptance criteria in Regulatory Page 18 of 23

Guides 1.174 and 1.177. Therefore, there will not be a significant increase in the probability of an accident.

The proposed changes did not include any hardware changes, and therefore, all structures, systems, and components will continue to perform their intended function to mitigate the consequences of an event within the assumed acceptance limits. The proposed changes do not affect source term, containment isolation, or the radiological release assumptions used in evaluating radiological consequences of previously analyzed accidents.

Therefore, the proposed changes do not increase the consequences of an accident previously evaluated.

Based on the above paragraphs, it is concluded the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

No. The proposed changes do not involve any hardware changes, any setpoint changes, any addition of safety related equipment, or any changes in the manner in which the systems provide plant protection.

Additionally, all operator actions credited in accident analyses remain the same. There are no new or different accident initiators or new accidents scenarios created by the proposed changes. Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed amendment involve a significant reduction in a margin of safety?

No. The safety analyses acceptance criteria in the Updated Safety Analysis Report (USAR) are not impacted by these changes. The proposed changes do not alter the manner in which safety limits, limiting safety system settings, or limiting conditions for operation are determined.

All signals and operator actions credited in the USAR accident analyses will remain the same. Redundant RPS and ESFAS trains are maintained and diversity with regard to the signals that provide reactor trip and engineered safety features actuation is also maintained. The calculated impact on risk continues to meet the acceptance criteria contained in Regulatory Guides 1.174 and 1.177. Therefore, the proposed changes do not involve a significant reduction in the margin of safety.

Conclusion Based on the above, it is concluded the proposed changes present no significant hazards consideration under the standards set forth in 10 CFR 50.92(c) and, accordingly, a finding of 'no significant hazards consideration" is justified.

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5.2 Applicable Regulatory RequirementslCriteria The regulatory bases and guidance documents associated with the systems discussed in this amendment application are listed below. It is important to note that the KNPP was designed, constructed, and is being operated to comply with the owner's understanding of the intent of the Atomic Energy Commission (AEC) General Design Criteria (GDC) for Nuclear Power Plant Construction Permits as proposed on July 10, 1967. Since the construction of the plant was about 50 percent completed prior to the issuance of the February 20, 1971, 10 CFR 50 Appendix A General Design Criteria, the plant was not required to be reanalyzed and the FSAR was not required to be revised to reflect these later criteria. However, the AEC SER, issued July 24, 1972, acknowledged that the AEC staff assessed the plant, as described in the FSAR, against the Appendix A design criteria and "...are satisfied that the plant design generally conforms to the intent of these criteria." Therefore, the GDC numbering below reflects that of the proposed AEC GDC (July 1967) and not that of the current GDC in 10 CFR 50 Appendix A.

GDC 14 requires core protection systems, together with associated equipment, be designed to prevent or to suppress conditions that could result in exceeding acceptable fuel damage limits.

GDC 15 requires protection systems to be provided for sensing accident situations and initiating the operation of necessary Engineered Safety Features.

GDC 19 requires protection systems to be designed for high functional reliability and in-service testability necessary to avoid undue risk to the health and safety of the public.

GDC 25 requires that means shall be included for suitable testing of the active components of protection systems while the reactor is in operation to determine if failure or loss of redundancy has occurred.

GDC 20 requires the redundancy and independence designed into protection systems to be sufficient to assure that no single failure or removal from service of any component or channel of such a system will result in loss of the protection function. The redundancy provided shall include as a minimum, two channels of protection for each protection function to be served.

GDC 23 requires that the effects of adverse conditions to which redundant channels or protection systems might be exposed in common, either under normal conditions or those of an accident, shall not result in loss of the protection function or shall be tolerable on some other basis.

GDC 26 requires the protection systems be designed to fail into a safe state or into a state established as tolerable on a defined basis if conditions such as disconnection of the system, loss of energy (e.g., electrical power, instrument air), or adverse environments (e.g., extreme heat or cold, fire, steam, or water) are experienced.

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GDC 27 requires that two independent control systems, preferably of different principles, shall be provided for reactivity control.

GDC 31 requires that the reactor protection system shall be capable of protecting against any single malfunction of the reactivity control system by limiting reactivity transients to avoid exceeding acceptable fuel damage limits.

KNPP original design implemented the principles of IEEE 279, "Standard, Nuclear Power Plant Protection Systems," August 1968.

There will be no changes to the RTS and ESFAS instrumentation design such that compliance with any of the regulatory requirements and guidance documents above would come into question. The above evaluations confirm that the plant will continue to comply with all applicable regulatory requirements.

In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security of the health and safety of the public.

6.0 ENVIRONMENTAL CONSIDERATION

S This proposed amendment involves a change to a requirement with respect to the installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or a change to a surveillance requirement. NMC has determined that the proposed amendment involves no significant hazards considerations and no significant change in the types of any effluents that may be released offsite and that there is no significant increase in the individual or cumulative occupational radiation exposure.

Accordingly, this proposed amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with this proposed amendment.

7.0 REFERENCES

1. WCAP-15376-P-A, Revision 1, "Risk-Informed Assessment of the RTS and ESFAS Surveillance Test Intervals and Reactor Trip Breaker Test and Completion Times," March 2003.

2. WCAP-1 0271 -P-A, "Evaluation of Surveillance Frequencies and Out of Service Times for the Reactor Protection Instrumentation System," dated May 1985.

3. WCAP-10271-P-A, Supplement 1, "Evaluation of Surveillance Frequencies and Out of Service Times for the Reactor Protection Instrumentation System," dated July 1985.

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4. WCAP-1 0271-P-A, Supplement 2, Revision 1, "Evaluation of Surveillance Frequencies and Out of Service Times for the Reactor Protection Instrumentation System," dated June 1990.

5. Letter from C. 0. Thomas (NRC) to J. J. Sheppard (WOG), "Acceptance for Referencing of Licensing Topical Report WCAP-1 0271, 'Evaluation of Surveillance Frequencies and Out of Service Times for the Reactor Protection Instrumentation Systems,"' dated February 21, 1985.

6. Letter from TR Quay (NRC) to DC Hintz (WPS) dated July 8, 1987.

7. Letter from Charles E. Rossi (NRC) to Roger A. Newton (WOG), "Safety Evaluation by the Office of Nuclear Reactor Regulation Review of Westinghouse Report WCAP-1 0271 Supplement 2 and WCAP-1 0271 Supplement 2, Revision 1, 'Evaluation of Surveillance Frequencies and Out of Service Times for the Engineered Safety Features Actuation Systems,"' dated February 22, 1989.

8. Letter from Charles E. Rossi (NRC) to Gerard T. Goering (WOG), 'Westinghouse Topical Report WCAP-1 0271 Supplement 2, Revision 1, 'Evaluation of Surveillance Frequencies and Out of Service Times for the Engineered Safety Features Actuation System,"' (NRC Supplemental Safety Evaluation), dated April 30,1990.

9. NUREG-1 366 "Improvement to Technical Specifications Surveillance Requirements," NRC Division of Operational Events Assessment, Office of Nuclear Reactor Regulation, December 1992.

10. NRC Generic Letter 93-05 "Line-Item Technical Specifications Improvements to Reduce Surveillance Requirements for Testing During Power Operation,"

September 27, 1993.

11. Letter from T.H. Essig (NRC) to L.F. Liberatori Jr. (WOG), "Review of Westinghouse Owners Group Topical Reports WCAP-14333P and WCAP-14334NP, Dated May 1995, 'Probabilistic Risk Analysis of the RPS and ESFAS Test Times and Completion Times,' (TAC No. M92782)," dated July 15, 1998.

12. WOG-98-245, Letter to WOG Primary Representatives and Licensing Subcommittee Representatives, "Implementation Guideline for WCAP-1 4333-P-A, Rev. I (Proprietary), 'Probabilistic Risk Analysis of the RPS and ESFAS Tests Times and Completion Times,' (MUHP-03054)," dated December 2, 1998.

13. WCAP-14333-P-A, Revision 1, "Probabilistic Risk Analysis of the RPS and ESFAS Test Times and Completion Times," October 1998.

Page 22 of 23

14. NRC Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment In Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," dated July 1998.

15. NRC Regulatory Guide 1.177, "An Approach for Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications," dated August 1998.

16. Letter from W.H. Ruland (NRC) to R.H. Byran (WOG), "Acceptance for Referencing of Topical Report WCAP-1 5376-P, Revision 0, 'Risk-Informed Assessment of the RTS and ESFAS Surveillance Test Intervals and Reactor Trip Breaker Test and Completion Times,' (TAC. No. MB0983)," dated December 20, 2002.

17. WOG-03-202, letter to WOG Management Committee and Licensing Subcommittee, "Transmittal of Approved Topical Report: WCAP-1 5376-P-A, Rev. 1, (Proprietary) "Risk-Informed Assessment of the RTS and ESFAS Surveillance Test Intervals and Reactor Trip Breaker Test and Completion Times," and Implementation Guidelines (MUHP-3046)," dated April 3, 2003.

18. WOG letter OG-02-002, from R.H. Bryan (WOG) to Document Control Desk (NRC), "Transmittal of Response to Request for Additional Information (RAI)

Regarding WCAP-1 5376-P, Rev. 0, "Risk-informed Assessment of the RTS and ESFAS Surveillance Test Intervals and Reactor Trip Breaker Test and Completion Times," (MUHP-3046), January 8, 2002.

19. WOG letter OG-01-058, from R.H. Bryan (WOG) to Document Control Desk (NRC), "Transmittal of Response to Request for Additional Information (RAI)

Regarding WCAP-1 5376-P, Rev. 0, "Risk-informed Assessment of the RTS and ESFAS Surveillance Test Intervals and Reactor Trip Breaker Test and Completion Times," (MUHP-3046), September 28, 2001.

20. Industry/Technical Specification Task Force (TSTF) Standard TS (STS) Change Traveler 411, Revision 1, 'Surveillance Test Interval Extensions for Components of the Reactor Protection System (WCAP-1 5376)," dated August 7, 2002.

21. Industry/Technical Specification Task Force (TSTF) Standard TS (STS) Change Traveler 418, Revision 2, "Surveillance Test Interval Extensions for Components of the Reactor Protection System (WCAP-1 5376)," dated February 21, 2003.

22. WOG-04-233, letter to WOG Management Committee and Licensing Subcommittee, "Transmittal of Revised Implementation Guidelines for WCAP-1 5376-P-A, Rev. 1, 'Risk-Informed Assessment of the RTS and ESFAS Surveillance Test Intervals and Reactor Trip Breaker Test and Completion Times'(MUHP-3046)," dated May 6, 2004.

Page 23 of 23

ENCLOSURE 3 APPLICABILITY OF WCAP-14333 AND WCAP-1 5376 TO KEWAUNEE NON-PROPRIETARY The following tables demonstrate the applicability of WCAP-14333 and WCAP-1 5376 to the Kewaunee Nuclear Power Plant (KNPP). The tables were taken directly from the Westinghouse Owner's Group (WOG) implementation guidance letters and fulfill part of condition 1 from both NRC SERs approving the above WCAPs. Table 1 and 2 show applicability of the general analysis parameters from the WCAPs. Table 3 shows the applicability of the reactor trip actuation signals while Table 4 shows the applicability of the ESFAS actuation signals. Table 5 demonstrates the applicability of the human reliability analysis and fulfills condition 4 of the WCAP-1 5376 NRC SER.

Table 1 WCAP-14333 Implementation Guidelines:

Applicability of the Analysis General Parameters Parameter WCAP-14333 Analysis Plant Specific Assumptions Parameter Logic Cabinet Type (1) SSPS or Relay Relay Component Test Intervals (2)

Analog channels 3 months 1 month (12)

Logic cabinets (SSPS) 2 months NA

. Logic cabinets (Relay) 1 month 1 month

. Master Relays (SSPS) 2 months NA

Master Relays (Relay) 1 month 1 month

Slave Relays 3 months 18 months

. Reactor trip breakers 2 months 1 month (13)

Analog Channel Calibrations (3)

Done at-power Yes Yes

Interval 18 months 18 months (3)

Page 1 of 12

NON-PROPRIETARY Table 1 (continued)

WCAP-14333 Implementation Guidelines:

Applicability of the Analysis General Parameters Parameter WCAP-14333 Analysis Plant Specific Assumptions Parameter Typical At-Power Maintenance Intervals (4)

Analog channels 24 months Equal to or greater than 24 months

Logic cabinets (SSPS) 18 months NA

. Logic cabinets (Relay) 12 months Equal to or greater than 12 months

Master relays (SSPS) Infrequent (5) NA

Master relays (Relay) Infrequent (5) Infrequent

. Slave relays Infrequent (5) Infrequent

. Reactor trip breakers 12 months 18 months AMSAC (6) Credited for AFW pump AMSAC initiates AFW start pump start Total Transient Event Frequency (7) 3.6 1.4 ATWS Contribution to CDF (current PRA model) (8) 8.4E-06 7.9E-07 Total CDF from Internal Events (current PRA model) (9) 5.8E-05 3.4E-05 Total CDF from Internal Events (IPE) (10) Not Applicable 6.6E-05 NOTES FOR TABLE 1:

1. Both types of logic cabinets, SSPS and Relay, are included in WCAP-14333. The analysis is applicable to KNPP.

2. Test intervals are equal to or greater than those used in WCAP-14333 or are justified in notes 12 and 13.

Therefore, the WCAP-14333 analysis is applicable to KNPP.

3. Channel calibrations can be performed at power and most calibration intervals are equal to or greater than that used in WCAP-14333. The exception is the Nuclear Power Range, which is calibrated quarterly at power. The frequency of this calibration will not change with this proposed amendment. This difference in calibration frequency does not effect the conclusions of the WCAP. Therefore, the WCAP analysis is applicable to KNPP.

Page 2 of 12

NON-PROPRIETARY NOTES FOR TABLE 1 (CONTINUED):

4. KNPP's maintenance intervals are equal to or greater than those used in WCAP-14333, and therefore, the analysis is applicable to KNPP.

5. Only corrective maintenance is performed on master and slave relays. The maintenance interval on typical relays is relatively long, that is, experience has shown they do not typically fail completely. Failure of slave relays usually involve failure of individual contacts. This is consistent with KNPP's experience and WCAP-14333 applies to KNPP.

6. AMSAC will initiate AFW pump start and WCAP-14333 is applicable to Kewaunee.

7. This entry includes the total frequency for initiators requiring a reactor trip signal to be generated for event mitigation to assess the importance of ATWS events to CDF. Events initiated by a reactor trip are not included.

Since the plant specific value is less than the WCAP-14333 value, this analysis is applicable to KNPP.

8. This entry includes the ATWS contribution to core damage frequency (from at-power, internal events). This is required to determine if the ATWS event is a large contributor to CDF.

9. This entry indicates the total CDF from internal events (including internal flooding) for the most recent PRA update.

This is required for comparison to the NRC's risk-informed CDF acceptance guidelines.

10. This entry indicates the total CDF from internal events from the IPE model submitted to the NRC in response to Generic Letter 88-20, "Individual Plant Examination for Severe Accident Vulnerabilities - 10 CFR 50.54(f). See Note 11 for differences between the most recent KNPP PRA model update and that included in the KNPP GL 88-20 response.

Page 3 of 12

NON-PROPRIETARY NOTES FOR TABLE 1 (CONTINUED):

11. The Kewaunee PRA has been continuously updated since the final IPE submittal in December 1992, and incorporates plant design changes and upgraded methods to be consistent with the current state of the art. This list is not an exhaustive list, but a list of the major changes.

Removed operator action to stop RHR pumps running on miniflow

Took credit for RWST refill Took credit for air accumulators on certain AOV's

Modeled alternate means of cooling air compressors

Changed reactor cavity configuration from dry to wet

Test and maintenance modeled for both trains instead of just one Loss of DC bus modeled for each train instead of most conservative

Loss of AC Bus modeled for each train instead of most conservative

Component cooling modeled so each train has a 0.5 probability of being in standby

LOCA's, SGTR's, and SLB's modeled so each loop has a 0.5 probability of being the broken loop

Charging pump relief valve model corrected

Service water strainers removed based on analysis Pressurizer PORV block valves no longer assumed open

Converted from Grafter to WinNUPRA

Updated all failure rate data, including initiating events Incorporated approved WOG seal LOCA model

Recalculate all important HEPs

Based on revised TH analysis, removed credit for LPI in Medium LOCA Redesigned ISL removing credit for RWST refill and MOV closure against high pressure Redesigned steam line break, modeling PTS concerns

Resolved numerous Peer review Facts & Observations Reperformed all thermal hydraulic computer code calculations incorporating new steam generators and power uprate.

12. Analog channel test is at the original licensed interval of one month. The STI increase to three months was justified and approved by the NRC in WCAP-1 0271-P-A, which has not been implemented at KNPP. However, this analysis remains applicable.

13. RTB testing is at the original licensed interval of one month. This interval was not evaluated in the WCAP.

Therefore, the WCAP RTB STI is not applicable and KNPP will not implement any change to RTB test frequency.

Page 4 of 12

NON-PROPRIETARY Table 2 WCAP-15376 Implementation Guidelines:

Applicability of the Analysis General Parameters a,c WCAP-15376 Analysis Plant Specific Parameter Assumptions Parameter 4 4 4 4 I

4 4 4 4 4 4 1 1 I I Page 5 of 12

NON-PROPRIETARY Table 2 WCAP-15376 Implementation Guidelines:

Applicability of the Analysis General Parameters a,c I ¶ 1- 1

+ +

.4- 4

1. -r NOTES FOR TABLE 2:

Page 6 of 12

NON-PROPRIETARY NOTES FOR TABLE 2: a.c Page 7 of 12

NON-PROPRIETARY Table 3 WCAP-14333 and WCAP-15376 Implementation Guidelines:

ac Applicability of Analysis Reactor Trip Actuation Signals Page 8 of 12

NON-PROPRIETARY NOTES FOR TABLE 3: a,c Page 9 of 12

NON-PROPRIETARY Table 4 WCAP-14333 and WCAP-15376 Implementation Guidelines:

Applicability of Analysis Engineered Safety Features Actuation Signals a,c Page 10 of 12

NON-PROPRIETARY NOTES FOR TABLE 4 (CONTINUED): a.c Page 11 of 12

NON-PROPRIETARY Table 5 WCAP-15376 Implementation Guidelines:

Applicability of the Human Reliability Analysis a,c

.1-

+

1-1*

1-

1-Page 12 of 12

ENCLOSURE 4 NUCLEAR MANAGEMENT COMPANY, LLC KEWAUNEE NUCLEAR PLANT DOCKET 50-305 May 31, 2004 Letter from Thomas Coutu (NMC)

To Document Control Desk (NRC)

License Amendment Request 204 Strike Out TS Pages:

TS 1.0-3 TS B3.5-1 Table TS 3.5-2, page 3 of 4 Table TS 4.1-1, page 1 of 7 Table TS 4.1-1, page 2 of 7 Table TS 4.1-1, page 3 of 7 Table TS 4.1-1, page 4 of 7 Table TS 4.1-1, page 5 of 7 8 PAGES TO FOLLOW

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 that 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.

6. STAGGERED TEST BASIS A STAGGERED TEST BASIS shall consist of the testing of one of the systems.

subsystems, channels, or other designated components during the interval specified by the Surveillance Frequency. so that all systems, subsystems, channels, or other designated components are tested during n Surveillance Frequency intervals, where n is the total number of systems, subsystems.

channels, or other designated components in the associated function.

Amendment No. 462 TS 1.0-3 049/29O02

BASIS - Instrumentation System (TS 3.5)

Instrumentation has been provided to sense accident conditions and to initiate operation of the engineered safety features.(') Section 2.3 of these specifications describes the LIMITING SAFETY SYSTEM SETTINGS for the protective instrumentation.

Safety Iniection Safety Injection can be activated automatically or manually to provide additional water to the Reactor Coolant System or to increase the concentration of boron in the coolant.

Safety Injection is initiated automatically by (1) low pressurizer pressure, (2) low main steam line pressure in either loop and (3) high containment pressure. Protection against a loss-of-coolant accident is primarily through signals (1)and (3). Protection against a steam line break is primarily by means of signal (2).

Manual actuation is always possible. Safety Injection signals can be blocked during those OPERATING MODES where they are not required" for safety and where their presence might inhibit operating flexibility; they are generally restored automatically on return to the "required" OPERATING MODE.

Reactor Trio Breakers With the addition of the automatic actuation of the shunt trip attachment, diverse features exist to effect a reactor trip for each reactor trip breaker. Since either trip feature being OPERABLE would initiate a reactor trip on demand, the flexibility is provided to allow plant operation on a reactor trip breaker (with either trip feature inoperable) for up to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . This specification also requires the plant to proceed to the HOT SHUTDOWN condition in accordance with the Kewaunee STANDARD SHUTDOWN SEQUENCE if a reactor trip breaker is bypassed for greater than 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

One RTB train may be inoperable for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months from the time of discovery. This time is based on evaluations of WCAP-1 5376-P-A. This specification also requires the plant to proceed to HOT SHUTDOWN condition if the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months completion time is not met.

Containment Sprav Containment sprays are also actuated by a high containment pressure signal (Hi-Hi) to reduce containment pressure in the event of a loss-of-coolant or steam line break accident inside the containment.

The containment sprays are actuated at a higher containment pressure (approximately 50% of design containment pressure) than is Safety Injection (10% of design). Since spurious actuation of containment spray is to be avoided, it is initiated only on coincidence of high containment pressure (Hi-Hi) sensed by three sets of one-out-of-two containment pressure signals provided for its actuation.

1)USAR Section 7.5 Amendment No. 401-TS B3.5-1 o9130/93

TABLE TS 3.5-2 INSTRUMENT OPERATION CONDITIONS FOR REACTOR TRIP 2 3 4 5 NO. OF MINIMUM MINIMUM PERMISSIBLE OPERATOR ACTION IF NO. OF CHANNELS TO OPERABLE DEGREE OF BYPASS CONDITIONS OF COLUMN 3 NO. FUNCTIONAL UNIT CHANNELS TRIP CHANNELS REDUNDANCY CONDITIONS OR 4 CANNOT BE MET 17 Reactor Trip Breaker 2 1 2 The RTBs may Restore train to OPERABLE (RTB) be bypassed for status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . If train up to 8 hrs. for cannot be restored to surveillance OPERABLE status in 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

testing or be in HOT SHUTDOWN in an maintenance additional 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Maintain HQTl SHUTDOWN and open the RTBs (Independently Test Shunt 2/bkr 1 2 A fter 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months maintain HOT and Undervoltage Trip SHUTDOWN and open the Attachments) _ _ RTBs Amendment No. 94 Page 3 of 4 11/12/91

TABLE TS 4.1-1 MINIMUM FREQUENCIES FOR CHECKS, CALIBRATIONS AND TEST OF INSTRUMENT CHANNELS CHANNEL DESCRIPTION CHECK CALIBRATE TEST l REMARKS

1. Nuclear Power Range Each shift(a) Daily(a) Quarterl t (a) Heat balance h4y(b) (b) Signal to AT; bistable action (permissive, rod stop, trips)

Effective Full Effective Full Power (c) Upper and lower chambers for axial off-set Power Month(c) Quarter(c) Quarterly(d) using incore detectors.

The check and calibration for axial offset shall also be performed prior to > 75% power following any core alteration.

(d) Permissives P8 and P10 and the 25% reactor trip are tested quarterly.

2. Nuclear Intermediate Each shift(a,c) Not applicable Prior to each (a) Once/shift when in service Range startup if not (b) Log level; bistable action done previous (permissive, rod stop, trips) week(b) (c) Channel check required in all plant modes

3. Nuclear Source Range Each shift(a,c) Not applicable Prior to each (a) Once/shift when in service startup if not ()Oc/hf hni evc donetu peifous (b) Bistable action (alarm, trips) done previous (c) Channel check required in all plant modes

4. Reactor Coolant Each shift (c) Each refueling cycle Menthly(a) (a) Overtemperature AT Temperature Semiannual(a) (b) Overpower AT Monthly(b) (c) Channel check not required below HOT Semiannual(b) SHUTDOWN

5. Reactor Coolant Flow Each shift Each refueling cycle Mont Amendment No. 161-Page 1 of 7 O2/4004

TABLE TS 4.1-1 MINIMUM FREQUENCIES FOR CHECKS, CALIBRATIONS AND TEST OF INSTRUMENT CHANNELS CHANNEL DESCRIPTION CHECK CALIBRATE TEST REMARKS

6. Pressurizer Water Level Each shift Each refueling cycle Menthly~lmi

7. Pressurizer Pressure Each shift Each refueling cycle Menthlygiaq

8. a. 4-KV Voltage and Not applicable Each refueling cycle MetySemiA Reactor protection circuits only Frequency nnual

b. 4-KV Voltage Not applicable Each refueling cycle Monthly Safeguards buses only (Loss of Voltage) ._l

c. 4-KV Voltage Not applicable Each refueling cycle Monthly Safeguards buses only (Degraded Grid)

9. Analog Rod Position Each shift(a,b) Each refueling cycle Each refueling (a) With step counters cycle (b) Following rod motion in excess of 24 steps when computer is out of service

10. Rod Position Bank Each shift(a,b) Not applicable Each refueling (a) With analog rod position Counters cycle (b) Following rod motion in excess of 24 steps when computer is out of service Amendment No. 1-61 Page 2 of 7 2/1Ae/2004

TABLE TS 4.1-1 MINIMUM FREQUENCIES FOR CHECKS, CALIBRATIONS AND TEST OF INSTRUMENT CHANNELS CHANNEL DESCRIPTION CHECK [ CALIBRATE TEST REMARKS

11. a. Steam Generator Low Each shift Each refueling cycle Menhy~lmia Level noiu

b. Steam Generator Each shift Each refueling cycle MORthlySemia High Level nnual

12. Steam Generator Flow Each shift Each refueling cycle MenithlySmia Mismatch nnual

13. Deleted

14. Residual Heat Removal Each shift (when Each refueling cycle Not applicable Pump Flow in operation)

15. Deleted

16. Refueling Water Storage Weekly Annually Not applicable Tank Level

17. Deleted Amendment No. 16 Page 3 of 7 02/12/2001

TABLE TS 4.1-1 MINIMUM FREQUENCIES FOR CHECKS, CALIBRATIONS AND TEST OF INSTRUMENT CHANNELS CHANNEL DESCRIPTION CHECK CALIBRATE TEST REMARKS

18. a. Containment Each shift Each refueling cycle MeIgthly (a) Isolation Valve Signal Pressure Semiannual (SIS signal) (a)

b. Containment Each shift(a) Each refueling cycle(a) Menthl (a) Narrow range containment pressure Pressure Semiannual (-3.0, +3.0 psig excluded)

(Steamline Isolation) (a)

c. Containment Each shift Each refueling cycle Menthy Pressure Semiannual (Containment Spray Act)

d. Annulus Pressure Not applicable Each refueling cycle Each refueling (Vacuum Breaker) cycle

19. Radiation Monitoring Daily (a,b) Each refueling cycle (a) Quarterly (a) (a) Includes only channels R1 1 thru R1 5, R1 9, System R21, and R23 (b) Channel check required in all plant modes

20. Deleted

21. Containment Sump Level Not applicable Not applicable Each refueling cycle

22. Accumulator Level and Each shift Each refueling cycle Not applicable Pressure

23. Steam Generator Each shift Each refueling cycle Monthly Pressure Amendment No. 1 Page 4 of 7 021- 2/2001

TABLE TS 4.1-1 MINIMUM FREQUENCIES FOR CHECKS, CALIBRATIONS AND TEST OF INSTRUMENT CHANNELS CHANNEL DESCRIPTION CHECK CALIBRATE TEST REMARKS

24. Turbine First Stage Each shift Each refueling cycle Monthly Pressure

25. Portable Radiation Monthly (a) Annually Quarterly (a) Channel check required in all plant modes Survey Instruments

26. Protective System Logic Not applicable Not applicable Meng*y~uaft (a) On a staggered test basis Channel Testing JaLlncludes auto load sequencer

27. Deleted

28. Deleted

29. Seismic Monitoring Each refueling Each refueling cycle Not applicable System cycle l

30. Fore Bay Water Level Not applicable Each refueling cycle Each refueling cycle

31. AFW Flow Rate (a) Each refueling cycle Not applicable (a) Flow rate indication will be checked at each unit startup and shutdown

32. PORV Position Indication Monthly Each refueling cycle Not applicable

a. Back-up Monthly Each refueling cycle Not applicable (Temperature)

33. PORV Block Valve Monthly Each refueling cycle Not applicable Position Indicator Amendment No. 1-Page 5 of 7 02/2/2001

ENCLOSURE 5 NUCLEAR MANAGEMENT COMPANY, LLC KEWAUNEE NUCLEAR PLANT DOCKET 50-305 June 1,2004 Letter from Thomas Coutu (NMC)

To Document Control Desk (NRC)

License Amendment Request 204 Affected TS Pages:

TS 1.0-3 TS B3.5-1 Table TS 3.5-2, page 3 of 4 Table TS 4.1-1, page 1 of 7 Table TS 4.1-1, page 2 of 7 Table TS 4.1-1, page 3 of 7 Table TS 4.1-1, page 4 of 7 Table TS 4.1-1, page 5 of 7 8 PAGES TO FOLLOW

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 that 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.

6. STAGGERED TEST BASIS A STAGGERED TEST BASIS shall consist of the testing of one of the systems, subsystems, channels, or other designated components during the interval specified by the Surveillance Frequency, so that all systems, subsystems, channels, or other designated components are tested during n Surveillance Frequency intervals, where n is the total number of systems, subsystems, channels, or other designated components in the associated function.

TS 1.0-3 Amendment No.

TABLE TS 3.5-2 INSTRUMENT OPERATION CONDmONS FOR REACTOR TRIP 7 _ 1 2 3 _ 4 5 6 NO. OF MINIMUM MINIMUM PERMISSIBLE OPERATOR ACTION IF NO. OF CHANNELS TO OPERABLE DEGREE OF BYPASS CONDmONS OF COLUMN 3 NO. FUNCTIONAL UNIT CHANNELS TRIP CHANNELS REDUNDANCY CONDmONS OR 4 CANNOT BE MET__

17 Reactor Trip Breaker 2 1 2 The RTBs may Restore train to OPERABLE (RTB) be bypassed for status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . If train up to 8 hrs. for cannot be restored to surveillance OPERABLE status in 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , testing or be in HOT SHUTDOWN in an maintenance additional 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and open the RTBs (Independently Test Shunt 2/bkr 1 2 - l After 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months maintain HOT and Undervoltage Trip SHUTDOWN and open the Attachments) _ ___ _ RTBs Amendment No. I Page 3 of 4

BASIS - Instrumentation System (TS 3.5)

Instrumentation has been provided to sense accident conditions and to initiate operation of the engineered safety features.(' Section 2.3 of these specifications describes the LIMITING SAFETY SYSTEM SETTINGS for the protective instrumentation.

Safety Injection Safety Injection can be activated automaticallyormanuallyto provide additional waterto the Reactor Coolant System or to increase the concentration of boron in the coolant.

Safety Injection is initiated automatically by (1) low pressurizer pressure, (2) low main steam line pressure in either loop and (3) high containment pressure. Protection against a loss-of-coolant accident is primarily through signals (1) and (3). Protection against a steam line break is primarily by means of signal (2).

a.,

Manual actuation is always possible. Safety Injection signals can be blocked during those OPERATING MODES where they are not "required for safety and where their presence might inhibit operating flexibility; they are generally restored automatically on return to the 'required" OPERATING MODE.

Reactor Trip Breakers With the addition of the automatic actuation of the shunt trip attachment, diverse features exist to effect a reactor trip for each reactor trip breaker. Since either trip feature being OPERABLE would initiate a reactor trip on demand, the flexibility is provided to allow plant operation on a reactor trip breaker (with either trip feature inoperable) for up to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . This specification also requires the plant to proceed to the HOT SHUTDOWN condition in accordance with the Kewaunee STANDARD SHUTDOWN SEQUENCE if a reactor trip breaker is bypassed for greater than 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

One RTB train may be inoperable for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months from the time of discovery. This time is based on evaluations of WCAP-15376-P-A. This specification also requires the plant to proceed to HOT SHUTDOWN condition if the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months completion time is not met.

Containment Sprav Containment sprays are also actuated by a high containment pressure signal (Hi-Hi) to reduce containment pressure in the event of a loss-of-coolant or steam line break accident inside the containment.

The containment sprays are actuated at a higher containment pressure (approximately 50% of design containment pressure) than is Safety Injection (10% of design). Since spurious actuation of containment spray is to be avoided, it is initiated only on coincidence of high containment pressure (Hi-Hi) sensed by three sets of one-out-of-two containment pressure signals provided for its actuation.

USAR Section 7.5 TS B3.5-1 Amendment No.

.?

TABLE TS 4.1-1 MINIMUM FREQUENCIES FOR CHECKS, CALIBRATIONS AND TEST OF INSTRUMENT CHANNELS CHANNEL DESCRIPTION CHECK CALIBRATE TEST REMARKS

1. Nuclear Power Range Each shift(a) Daily(a) Quarterly(b) (a) Heat balance (b) Signal to AT; bistable action (permissive, rod stop, trips)

Effective Full Effective Full Power (c) Upper and lower chambers for axial off-set Power Month(c) Quarter(c) Quarterly(d) using incore detectors.

The check and calibration for axial offset shall also be performed prior to > 75% power following any core alteration.

(d) Permissives P8 and P10 and the 25% reactor trip are tested quarterly.

2. Nuclear Intermediate Each shift(a,c) Not applicable Prior to each (a) Once/shift when in service Range startup if not (b) Log level; bistable action done previous (permissive, rod stop, trips) week(b) (c) Channel check required in all plant modes

3. Nuclear Source Range Each shift(ac) Not applicable Prior to each (a) Once/shift when in service donetu peifous (b) Sistable action (alarm, trips) done previous (c) Channel check required in all plant modes

4. Reactor Coolant Each shift (c) Each refueling cycle Semiannual(a) (a) Overtemperature AT Temperature Semiannual(b) (b) Overpower AT (c) Channel check not required below HOT SHUTDOWN

5. Reactor Coolant Flow Each shift Each refueling cycle Semiannual Amendment No. I Pagel1of 7l I..

TABLE TS 4.1-1 MINIMUM FREQUENCIES FOR CHECKS, CALIBRATIONS AND TEST OF INSTRUMENT CHANNELS CHANNEL DESCRIPTION CHECK CALIBRATE TEST REMARKS

6. Pressurizer Water Level Each shift Each refueling cycle Semiannual

7. Pressurizer Pressure Each shift Each refueling cycle Semiannual

8. a. 4-KV Voltage and Not applicable Each refueling cycle Semiannual Reactor protection circuits only Frequency

b. 4-KV Voltage Not applicable Each refueling cycle Monthly Safeguards buses only (Loss of Voltage)

c. 4-KV Voltage Not applicable Each refueling cycle Monthly Safeguards buses only (Degraded Grid)

9. Analog Rod Position Each shift(a,b) Each refueling cycle Each refueling (a) With step counters cycle (b) Following rod motion in excess of 24 steps when computer is out of service

10. Rod Position Bank Each shift(a,b) Not applicable Each refueling (a) With analog rod position Counters cycle (b) Following rod motion in excess of 24 steps when computer is out of service Amendment No. I Page 2 of 7

TABLE TS 4.1-1 MINIMUM FREQUENCIES FOR CHECKS, CALIBRATIONS AND TEST OF INSTRUMENT CHANNELS CHANNEL DESCRIPTION CHECK CALIBRATE TEST REMARKS

11. a. Steam Generator Low Each shift Each refueling cycle Semiannual Level I

b. Steam Generator Each shift Each refueling cycle Semiannual High Level I

12. Steam Generator Flow Each shift Each refueling cycle Semiannual Mismatch I

13. Deleted

14. Residual Heat Removal Each shift (when Each refueling cycle Not applicable Pump Flow in operation)

15. Deleted

16. Refueling Water Storage Weekly Annually Not applicable Tank Level

17. Deleted Amendment No. I Page 3of 7

.. ..- ~ ;.~ *

. . t

TABLE TS 4.1-1 MINIMUM FREQUENCIES FOR CHECKS, CALIBRATIONS AND TEST OF INSTRUMENT CHANNELS CHANNEL DESCRIPTION CHECK CALIBRATE TEST REMARKS

18. a. Containment Each shift Each refueling cycle Semiannual (a) Isolation Valve Signal I Pressure (a)

(SIS signal)

b. Containment Each shift(a) Each refueling cycle(a) Semiannual (a) Narrow range containment pressure I Pressure (a) (-3.0, +3.0 psig excluded)

(Steamline Isolation)

C. Containment Each shift Each refueling cycle Semiannual Pressure I (Containment Spray Act)

d. Annulus Pressure Not applicable Each refueling cycle Each refueling (Vacuum Breaker) cycle l

19. Radiation Monitoring Daily (a,b) Each refueling cycle (a) Quarterly (a) (a) Includes only channels R1 1 thru R1 5, R1 9, System R21, and R23 (b) Channel check required in all plant modes

20. Deleted

21. Containment Sump Level Not applicable Not applicable Each refueling cycle

22. Accumulator Level and Each shift Each refueling cycle Not applicable Pressure

23. Steam Generator Each shift Each refueling cycle Pressure Semiannual Amendment No. I Page 4 of 7

TABLE TS 4.1-1 MINIMUM FREQUENCIES FOR CHECKS, CALIBRATIONS AND TEST OF INSTRUMENT CHANNELS CHANNEL DESCRIPTION CHECK CALIBRATE TEST REMARKS

24. Turbine First Stage Each shift Each refueling cycle Monthly Pressure

25. Portable Radiation Monthly (a) Annually Quarterly (a) Channel check required in all plant modes Survey Instruments

26. Protective System Logic Not applicable Not applicable Quarterly(a,b) (a) On a staggered test basis Channel Testing (b) Includes auto load sequencer

27. Deleted

28. Deleted

29. Seismic Monitoring Each refueling Each refueling cycle Not applicable System cycle

30. Fore Bay Water Level Not applicable Each refueling cycle Each refueling cycle

31. AFW Flow Rate (a) Each refueling cycle Not applicable (a) Flow rate indication will be checked at each unit startup and shutdown

32. PORV Position Indication Monthly Each refueling cycle Not applicable

a. Back-up Monthly Each refueling cycle Not applicable (Temperature)

33. PORV Block Valve Monthly Each refueling cycle Not applicable Position Indicator Amendment No. I Page 5 of 7

ENCLOSURE 6 NUCLEAR MANAGEMENT COMPANY, LLC KEWAUNEE NUCLEAR PLANT DOCKET 50-305 June 1, 2004 Letter from Thomas Coutu (NMC)

To Document Control Desk (NRC)

License Amendment Request 204:

Westinghouse Affidavit for Withholding, Proprietary Information Notice, and Copy Right Notice 8 PAGES TO FOLLOW

Westinghouse Westinghouse Electric Company Nuclear Services P.O. Box 355 Pittsburgh, Pennsylvania 15230-0355 USA U.S. Nuclear Regulatory Commission Direct tel: (412) 374-4643 Document Control Desk Direct fax: (412) 374-4011 Washington, DC 20555-0001 e-mail: greshaja@westinghouse.com Our ref: CAW-04-1794 February 24, 2004 APPLICATION FOR WITHHOLDING PROPRIETARY INFORMATION FROM PUBLIC DISCLOSURE

Subject:

WCAP-15376 Implementation Guideline Approach to Address the Conditions and Limitations in the NRC's Safety Evaluation (Proprietary)

The proprietary information for which withholding is being requested in the above-referenced report is further identified in Affidavit CAW-04-1794 signed by the owner of the proprietary information, Westinghouse Electric Company LLC. The affidavit, which accompanies this letter, sets forth the basis on which the information may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in paragraph (b)(4) of 10 CFR Section 2.790 of the Commission's regulations.

Accordingly, this letter authorizes the utilization of the accompanying affidavit by Nuclear Management Company.

Correspondence with respect to the proprietary aspects of the application for withholding or the Westinghouse affidavit should reference this letter, CAW-04-1794, and should be addressed to the undersigned.

Very t, A. Gresham, Manager I Regulatory Compliance and Plant Licensing Enclosures cc: D. Holland B. Benney E. Peyton A BNFL Group company

I CAW-04-1794 AFFIDAVIT COMMONWEALTH OF PENNSYLVANIA:

ss COUNTY OF ALLEGHENY:

Before me, the undersigned authority, personally appeared J. A. Gresham, who, being by me duly sworn according to law, deposes and says that he is authorized to execute this Affidavit on behalf of Westinghouse Electric Company LLC (Westinghouse), and that the averments of fact set forth in this Affidavit are true and correct to the best of his knowledge, information, and belief:

,~A. Gresham, Manager Regulatory Compliance and Plant Licensing Sworn to and subscribed before me this . day of e2004 I

M CMss Notary Public SmmuSLWNot4P L Faior, Not&i iA Bores inur 292007 mber, PenaAso f NoMdes

2 CAW-04-1794 (1) I am Manager, Regulatory Compliance and Plant Licensing, in Nuclear Services, Westinghouse Electric Company LLC (Westinghouse), and as such, I have been specifically delegated the function of reviewing the proprietary information sought to be withheld from public disclosure in connection with nuclear power plant licensing and rule making proceedings, and am authorized to apply for its withholding on behalf of Westinghouse.

(2) I am making this Affidavit in conformance with the provisions of 10 CFR Section 2.790 of the Commission's regulations and in conjunction with the Westinghouse "Application for Withholding" accompanying this Affidavit.

(3) I have personal knowledge of the criteria and procedures utilized by Westinghouse in designating information as a trade secret, privileged or as confidential commercial or financial information.

(4) Pursuant to the provisions of paragraph (b)(4) of Section 2.790 of the Commission's regulations, the following is furnished for consideration by the Commission in determining whether the information sought to be withheld from public disclosure should be withheld.

(i) The information sought to be withheld from public disclosure is owned and has been held in confidence by Westinghouse.

(ii) The information is of a type customarily held in confidence by Westinghouse and not customarily disclosed to the public. Westinghouse has a rational basis for determining the types of information customarily held in confidence by it and, in that connection, utilizes a system to determine when and whether to hold certain types of information in confidence. The application of that system and the substance of that system constitutes Westinghouse policy and provides the rational basis required.

Under that system, information is held in confidence if it falls in one or more of several types, the release of which might result in the loss of an existing or potential competitive advantage, as follows:

(a) The information reveals the distinguishing aspects of a process (or component, structure, tool, method, etc.) where prevention of its use by any of

3 CAW-04-1794 Westinghouse's competitors without license from Westinghouse constitutes a competitive economic advantage over other companies.

(b) It consists of supporting data, including test data, relative to a process (or component, structure, tool, method, etc.), the application of which data secures a competitive economic advantage, e.g., by optimization or improved marketability.

(c) Its use by a competitor would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing a similar product.

(d) It reveals cost or price information, production capacities, budget levels, or commercial strategies of Westinghouse, its customers or suppliers.

(e) It reveals aspects of past, present, or future Westinghouse or customer funded development plans and programs of potential commercial value to Westinghouse.

(f) It contains patentable ideas, for which patent protection may be desirable.

There are sound policy reasons behind the Westinghouse system which include the following:

(a) The use of such information by Westinghouse gives Westinghouse a competitive advantage over its competitors. It is, therefore, withheld from disclosure to protect the Westinghouse competitive position.

(b) It is information that is marketable in many ways. The extent to which such information is available to competitors diminishes the Westinghouse ability to sell products and services involving the use of the information.

(c) Use by our competitor would put Westinghouse at a competitive disadvantage by reducing his expenditure of resources at our expense.

Ir - V , 4 CAW-04-1794 (d) Each component of proprietary information pertinent to a particular competitive advantage is potentially as valuable as the total competitive advantage. If competitors acquire components of proprietary information, any one component may be the key to the entire puzzle, thereby depriving Westinghouse of a competitive advantage.

(e) Unrestricted disclosure would jeopardize the position of prominence of Westinghouse in the world market, and thereby give a market advantage to the competition of those countries.

(f) The Westinghouse capacity to invest corporate assets in research and development depends upon the success in obtaining and maintaining a competitive advantage.

(iii) The information is being transmitted to the Commission in confidence and, under the provisions of 10 CFR Section 2.790, it is to be received in confidence by the Commission.

(iv) The information sought to be protected is not available in public sources or available information has not been previously employed in the same original manner or method to the best of our knowledge and belief.

(v) The proprietary information sought to be withheld in this submittal is that which is appropriately marked in "WCAP-15376 Implementation Guideline Approach to Address the Conditions and Limitations in the NRC's Safety Evaluation" (Proprietary) on behalf of the Westinghouse Owners Group by Westinghouse, being transmitted by the Westinghouse Owners Group letter and Application for Withholding Proprietary Information from Public Disclosure to the Document Control Desk. The proprietary information as submitted for use by the Westinghouse Owners Group is applicable to other licensee submittals.

This information is part of that which will enable Westinghouse to:

5 CAW-04-1794 (a) Provide risk-informed assessment of the RTS and ESFAS to extend the interval for surveillance testing.

(b) Provide licensing defense services.

Further this information has substantial commercial value as follows:

(a) Westinghouse plans to sell the use of similar information to its customers for purposes of extending surveillance testing intervals (b) Westinghouse can sell support and defense of extending surveillance testing intervals.

Public disclosure of this proprietary information is likely to cause substantial harm to the competitive position of Westinghouse because it would enhance the ability of competitors to provide similar assessments and licensing defense services for commercial power reactors without commensurate expenses. Also, public disclosure of the information would enable others to use the information to meet NRC requirements for licensing documentation without purchasing the right to use the information.

The development of the technology described in part by the information is the result of applying the results of many years of experience in an intensive Westinghouse effort and the expenditure of a considerable sum of money.

In order for competitors of Westinghouse to duplicate this information, similar technical programs would have to be performed and a significant manpower effort, having the requisite talent and experience, would have to be expended.

Further the deponent sayeth not.

" P J PROPRIETARY INFORMATION NOTICE Transmitted herewith are proprietary and/or non-proprietary versions of documents furnished to the NRC in connection with requests for generic and/or plant-specific review and approval.

In order to conform to the requirements of 10 CFR 2.790 of the Commission's regulations concerning the protection of proprietary information so submitted to the NRC, the information which is proprietary in the proprietary versions is contained within brackets, and where the proprietary information has been deleted in the non-proprietary versions, only the brackets remain (the information that was contained within the brackets in the proprietary versions having been deleted). The justification for claiming the information so designated as proprietary is indicated in both versions by means of lower case letters (a) through (f) located as a superscript immediately following the brackets enclosing each item of information being identified as proprietary or in the margin opposite such information. These lower case letters refer to the types of information Westinghouse customarily holds in confidence identified in Sections (4)(ii)(a) through (4)(ii)(f) of the affidavit accompanying this transmittal pursuant to 10 CFR 2.790(b)(1).

COPYRIGHT NOTICE The reports transmitted herewith each bear a Westinghouse copyright notice. The NRC is permitted to make the number of copies of the information contained in these reports which are necessary for its internal use in connection with generic and plant-specific reviews and approvals as well as the issuance, denial, amendment, transfer, renewal, modification, suspension, revocation, or violation of a license, permit, order, or regulation subject to the requirements of 10 CFR 2.790 regarding restrictions on public disclosure to the extent such information has been identified as proprietary by Westinghouse, copyright protection notwithstanding. With respect to the non-proprietary versions of these reports, the NRC is permitted to make the number of copies beyond those necessary for its internal use which are necessary in order to have one copy available for public viewing in the appropriate docket files in the public document room in Washington, DC and in local public document rooms as may be required by NRC regulations if the number of copies submitted is insufficient for this purpose. Copies made by the NRC must include the copyright notice in all instances and the proprietary notice if the original was identified as proprietary.

- \

ENCLOSURE 7 LIST OF REGULATORY COMMITMENTS The following table identifies those actions committed to by NMC in this document. Any other statements in this submittal are provided for information purposes and are not considered to be regulatory commitments.

COMMITMENT Due Date/Event

1. KNPP will implement administrative controls to include 1. Prior to the following restrictions when an RTB or a logic cabinet is implementation.

removed from service:

Activities that could degrade the availability of the auxiliary feedwater system, reactor coolant system pressure relief (pressurizer PORVs and safety valves), AMSAC, or turbine trip should not be scheduled when a RTB is out of service or the diverse scram system should be operable.

Activities that could degrade other components of the RPS, including master relays or slave relays and activities that cause analog channels to be unavailable should not be scheduled when a logic cabinet is unavailable.

Activities on electrical systems that support the systems or functions listed in the first two bullets should not be scheduled when a RTB is unavailable or the diverse scram system should be operable.

2. Instrument drift characteristics will be validated using 2. Two years from future test data subsequent to changing the frequency to a implementation.

six month interval.

3. KNPP will incorporate an administrative control into the 3. Prior to CAP AR process requiring an apparent cause evaluation for implementation RPS and ESFAS failures to determine the extent of condition.

Page 1 of I

ML041610046

Contents

Text

1.0 DESCRIPTION

2.0 PROPOSED CHANGE

3.0 BACKGROUND

4.0 TECHNICAL ANALYSIS

6.0 ENVIRONMENTAL CONSIDERATION

7.0 REFERENCES

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ML041610046

Text

1.0 DESCRIPTION

2.0 PROPOSED CHANGE

3.0 BACKGROUND

4.0 TECHNICAL ANALYSIS

6.0 ENVIRONMENTAL CONSIDERATION

7.0 REFERENCES

Subject:

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