ML20127P719
| ML20127P719 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 01/25/1993 |
| From: | Tuckman M DUKE POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20127P723 | List: |
| References | |
| NUDOCS 9302020058 | |
| Download: ML20127P719 (16) | |
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m 7x Catarba Nwirar Gewgior Depetment na l' resident W10 Cuncord Road
' (803)&113:05 Office YA, Se 2976-
"(803)C11442G fax DUKEPOWEft January 25,1993 U.S. Nuclear Regulatory Commission ATTN:
Document Control Desk Washington, D.C. 20555
Subject:
Catawba Nuclear Station, Units 1 and 2 Docket Nos. 50-413 and 50 414 Proposed Changes to Technical Specifications 3/4.3.1 and 3/4.3.2-Increased Surveillance Test Intervals and Allowed Outage Times for Reactor. -
Trip System (RTS) and Engineered Safety Features Actuation ~ System'(ESPAS).
Instrumentation 1
Gentlemen:
Please find attached an application for amendment to Facility' Operating License Nos. NPF-35 and NPF-52 for Catawba Units 1 and 2, respectively. This proposed change involves; increasing the surveillance test intervals and allowed outage times for the RTS and ESFAS equipment based upon analyses performed by Westinghouse. for the Westinghouse Owners Group (WOG) and approved by the NRC.
1 In February 1985, the NRC issued the Safety Evaluation Report (SER) for WCAP-10271 and ~
i WCAP-10271, Supplement 1. The SER approved quarterly testing, six hours to place a :
failed channel in the tripped mode, increased allowed ; outage. times for test and maintenance,-
and testing la bypass for analog channels of the RTS.: By letter dated September 8,1986, the NRC issued Amendment Nos. 9 and 2 to the facility operating licenses for Catawba Units 1 and 2, respectively, which implemented most of the items approved in the SER for the RTS.
t In February 1989, the NRC issued the SER for WCAP-10271, Supplement 2 and Supplement'
- 2. Revision 1 and supplemented the SER in their letter dated April 30,1990. The SER and-SSER' approved relaxations similar to the RTS for the ESFAS. The three' SERs together -
4 approved all relaxations requested by the WOG in the WCAP-10271 program except the extensions requested for the reactor trip breakers and the main feedwater isolation on lowL H
T-cold coincident with high feedwater flow (this functional unit does not apply to Catawba).
The NRC review of the WCAP-10271 program is therefore considered complete, as the final
.WCAP in the series was issued in June 1990.
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9302020058 930125 PDR ADOCK'05000413 I
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Document Control Desk Page 2 -
January 25,1993 The proposed changes to the RTS and ESFAS instrumentation are based upon WCAP-10271; its supplements, and the NRC SERs described above.
Additional details pertaining to the description of the amendment request and background are provided in Attachment 1 of this submittal.- Attachment 2 contains the justification and safety?
evaluation and includes Duke Power Company's responses to the conditions imposed by the -
NRC in the SERs. Attachment 3 provides a discussion on the basis for a no significant hazards consideration determination. The proposed Technical Specification amendments fo~r' Catawba are presented in Attachment 4. A copy _of this amendment ' request is being provided to the appropriate South Carolina state official.
If you have any questions regarding this amendment request, please call L.J. Rudy at (803);
831-3084.
H Very truly yours, kbt yr.w-M.S. Tuckman IJR/s Attachments l
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Document Control Desk Page 3 January 25,1993 xc (W/ Attachments):
S.D. Ebneter Regional Administrator, Region Il R.E. Martin ONRR W.T. Orders Senior Resident Inspector Heyward Shealy Chief, Bureau of Radiological Health, SC American Nuclear Insurers M&M Nuclear Consultants INPO Records Center l
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.; I Document Control Desk-Page 41 January 25,1993 M.S. Tuckman, being duly sworn, states that he is Vice President of Duke Power Company;.
that he is authorized on the part of said Company to sign and file with the Nuclear Regulatory Commission' this revision to the Catawba Nuclear Station License Nos. NPF-35 and NPF-52 and that all statements and matters set forth therein are true and correct to the best of his knowledge.
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.)' \\0Ws M.S. Tuckman, Vice President 1
Subscribed and sworn to before me this 25th day of January,1993.
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LL otary Public My commission expires:
Notary Public. ScWh Carolina stes at Lars.
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ATTAClIMENT 1 IIACKGROUND AND DESCRII' TION OF AMENDMENT REQUEST Q
s llackcround The purpose of this Technical Specification amendment request is to obtain relaxation regarding the conduct of surveillance testing of the Reactor Trip System (RTS) and -
Engineered Safety Features Actuation System (ESFAS), As a result of concern of the impact of existing testing and maintenance requirements on plant operation, particularly in th.e area of instrumentation, the Westinghouse Owners Group (WOG) initiated a program to develop justification to be utilized in revising individual plant Technical Specifications. Operating j
plants have experienced many inadvertent reactor trips and safeguards actuations during.
j performance of instrumentation surveillance, causing unnecessary transients and challenges to plant safety systems. Signi5 cant time and effort on the part of the plant staff was devoted to performing, reviewing, documenting, and tracking various surveillance activities, which in many instances appeared unwarranted based on the high reliability of the equipment.
Signi0 cant benents for operating plants appeared to be achievable through revision of instrumentation test and maintenance requirements. A complete chronology of the WOG efforts and interactions with the NRC is contained in a document titled " Westinghouse Owners Group Guidelines for Preparing Submittals Requesting Revision of Reactor Protection System Technical SpeciRcations Based on Generic Approval of WCAP-10271 and Supplements" (TOPS Guidelines - August 1990).
Description olomeitdment Reqvnt The list of Technical Specification changes included in this amendment request is as follows:
(a)
Changes as described in the marked-up copy of Technical Specincation 3/4.3.1 (Attachment 4), These changes include:
(i)
The surveillance test interval in Table 4.3-1 for functional unit 18, Reactor Trip System Interlocks, Analog Channel Operational Test, is changed from monthly to "R" (at least once per 18 months) for each of the interlocks.
(ii)
Increase in surveillance intervals for Reactor Trip System (RTS) analog channel operational tests from once per month to once per quarter, (iii)
In Table 3.3-1, new ACTION 7 is added to allow 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> to restore an inoperable channel to operable status before requiring shutdown to HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to allow bypass of a channel for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for surveillance testing, provided the other channel is OPERABLE.
Make new ACTION 7 applicable to functional units 17 (Safety injection input from ESF) and 20 (Automatic Trip and Interlock Logic), rather than ACTION 9.
(b)
Changes as described in the marked-up copy of Technical Specification 3/4.3.2 (Attachment 4), These changes include:
(i)
Increase in surveillance intervals for Ergineered Safety Features Actuation System (ESFAS) analog channel operational tests from once per month to once.
per quarter.
(ii)
Increase in the time that an inoperable ESFAS channel may be maintained in an untripped condition from I hour to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
(iii)
Increase in the time that an inopernble ESFAS channel may be bypassed to allow testing of another channel in the same function from ~2 hours to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
(iv) in Table 3.3-3, revise the following ACTIONS in accordance with the Westinghouse Owners Group guidelines as follows:
- ACTION 14 is changed to allow 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> before placing the unit in HOT STANDBY and increases from 2 to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> the time that a channel may be bypassed.
- ACTION 15 is changed to increase the time that an inoperable channel may be untripped from 1 to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
- ACTION 16 is changed to increase the time that an additional channel may be bypassed from 2 to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
- ACTION 19 is changed to allow the inoperable channel to remain untripped for 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to allow an additional channel to be bypassed for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
- ACTION 21 is changed to allow 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> to restore an inoperable channel prior to placing the unit in HOT STANDBY and increases the time that a channel may be bypassed from 2 to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. Make new ACTION 21 applicable to functional units 5 (Feedwater Isolation) and 6 (Turbine Trip),
rather than ACTION 27.
- ACTION 28 is deleted, as the 1 out of 2 logic is no longer employed.
- Create new ACTIONS 14a,14b,15a,16a,16b, and 21a to be inserted in Table 3.3-3 following ACTIONS 14,15,16, and 21, respectively. ACTION 14a allows 8 hot rs prior to placing the unit in HOT STANDBY The remainder of these newly-created ACTIONS are made to apply to those functional units not addressed in the WOG generic submittal and for which -
NRC approval for relief has not been given.
(c)
Revisions to the 3/4.3.1 and 3/4.3.2 REACTOR TRIP SYSTEM AND ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION BASF1 (d)
Revisions to Technical Specifications 3/4.3.1 and 3/4.3.2 to delete obsolete footnotes
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ATTACllMENT 2 JUSTIFICATION AND SAFETY EVALUATION
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i Jm110tallnp and Snfety Evnhtallml in WCAP-10271 and its supplements, the WOG evaluated the impact of the proposed surveillance test interval (STI) and allowed outage time (AOT) changes on core damage frequency and public risk. The NRC staff concluded in its evaluation of the WOG evaluation that an overall upper bound increase of the core damage frequency due to the proposed STl/AOT changes is less than 6 percent for Westinghouse Pressurized Water Reactor (PWR).
plants. The NRC staff also concluded that actual core damage frequency increases for individual plants are expected to be substantially less than 6 percent. The NRC staff considered this core damage frequency increase to be small compared to the range of uncertainty in the core damage frequency analyses and therefore acceptable.
The NRC staff concluded in addition that a staggered test strategy need not be implemented for ESFAS analog channel testing and is no longer required for RTS analog channel testing.
(Since Duke Power Company has never applied for an increased surveillance test interval for the Catawba RTS, the staggered test strategy was never implemented.) This conclusion was based on the small relative contribution of the analog channels to RTS/ESFAS unavailability, process parameter signal diversity, and normal operational testing sequencing.
The NRC determined that the requirement to routinely verify permissive status is a different consideration than the availability of trip or actuation channels which are required to change state on the occurrence of an event and for which the function availability is more dependent on the surveillance interval. The definition of the channel check includes comparison of the channel status with other channels for the same parameter. For the RTS interlocks, the change from a monthly surveillance requirement to at least once every 18 months is therefore justified.
The proposed changes are consistent with the NRC staft's letters dated February 21,1985, February 22,1989, and April 30,1990, to the WOG regarding evaluation of WCAP-10271, WCAP-10271 Supplement 1, WCAP-10271 Supplement 2, and WCAP-10271 Supplement 2, Revision 1. The staff has stated that approval of these changes is contingent upon confirmation that certain conditions are met. It is the interpretation of Duke Power Company-that conditions imposed in the SER for WCAP-10271 and WCAP-10271 Supplement I for -
the RTS instrumentation shall also be applied to the ESFAS where appropriate. Duke Power Company's response to these conditions is provided below:
The first condition in the RTS SER required the use of a stnggered test plan for the RTS channels changed to the quarterly test frequency.
Response
The NRC did not impose this requirement for ESFAS channels and it was subsequently removed for the RTS channels. Duke Power Company never applied for an amendment to change the surveillance interval for RTS channels in the past; therefore, the staggered test plan was never utilized.
The second condition in the RTS SER required that plant procedures require a common i
cause evaluation for failure in RTS channels changed to the quarterly test frequency and ndditional testing for plausible common cause failures.
Response
Station directives, maintenance management procedures, and section guidelines have been established to allocate responsibilities for the administration of station equipment reliability programs. Identification of a significant or recurring failure in an RTS/ESFAS channel requires the origination of a Problem Investigation Report (PIR) to document the deficiency and mandate a review for additional investigative action. This assessment includes, but is not limited to, evaluations for common cause failure mechanisms, excessivt calibration drift, and subsequent testing of comparable equipment when deemed necessary.
Several programs are employed by station personnel to perform periodic engineetng review of failure data. These include the Component Engineering Section's Failure and Analysis Trending Program, as well as the Nuclear Plant Reliability Data System (NPRDS), which are used to accomplish this objective.
The third condition in the RTS SER required installed hardware capability for testing in the hypass mode.
Resnonse Catawba does not have installed bypass capability within the 7300 Protection and Control System at this time. Currently, Catawba has no plans to pursue that portion of the WOG effort which would allow for routine channel testing in the bypass condition. Routine channel testing will continue to be performed in the tripped condition.
The fourth condition in the RTS SER involved channels that provide input to both the RTS and the ESFAS. As stated by NRC in the safety evaluation for WCAP-10271:
"In order to avoid confusion in plant Technical Specifications regarding such dual function channels, the staff concludes that either (1) the channels should not be changed in the RTS tables until the ESFAS review is finished or (2) cautionary notes in the RTS tables should refer to the more stringent ESFAS requirements."
Response
Now that the ESFAS SER has been issued and all of the relaxations for the RTS analog channels are applicable to the ESFAS analog channels, this condition does not apply.
Cautionary notes as described above have been deleted.
The fifth condition in the RTS SER, and second in the ESFAS SER, addresses setpoint drift Confinnation is needed to show that the instrument setpoint methodology includes sufficient adjustments to offset the drift anticipated as a result of less frequent surveillance.
Response
Catawba engineering personnel have reviewed "as found" and "as left" data for the RTS and
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- ESFAS setpoints for a.l_2-month period _and concluded that sufficient margin is'present to offset the drift anticipated as a. result of quarterly surveillanceCThe allo,wable' margin present ~
uin the setpoints is more than adequate to offset any_ drift observed based upon review'of the' l
data." This information is availablelfor NRC inspection._.
The first conditloit in the ESFAS SER required that the plant'-specific applications niust ;
confirrn the appl'cability of the generic ~ analyses to' the plant,:
Resnonse The'WCAP methodology addresses two-loop, three-loopiand four-loop plants with relap. or:
solid state systems..The RTS and ESFAS functions for which increased surveillance::
intervals and allowed outage times are being requested in this amendment reque'st are thosei for which NRC approval has already been granted through issuance of the SERs and-supplements for the basis WCAP series. No additional changes are being requested in this -
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amendment request beyond those given approval by the NRC.
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ATTACllMENT 3 NO SIGNIFICANT II AZARDS CONSIDERATION DETERMINATION AND ENVIRONMENTAL IMPACT ANALYSIS
No SignificantlhzardLConsidervlon Detennination The standards used to arrive at a proposed determination that the changes described involve no significant hazards consideration are included in 10 CFR 50.92. The regulations state that if operation of the facility in accordance with the proposed amendment would not: (1) involve a significant increase in the probability or consequences of an accident previously evaluated, or (2) create the possibility of a new or different kind of accident from any accident previously evaluated, or (3) involve a signiGeant reduction in a margin of safety then a no significant hazards determination can be made.
Duke Power Company has reviewed the requirements of 10 CFR 50.92 as they relate to the proposed RTS and ESFAS Technical Speci6 cation changes for Catawba and determined that a signi6 cant hazards consideration is not involved. In support of this conclusion, the following analysis is provided.
Criterion 1 - Operation of Catawba in accordance with the proposed license amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.
The determination that the results of the proposed changes are within all acceptable criteria was established in the SERs prepared for WCAP-10271, WCAP-10271 Supplement 1, WCAP-10271 Supplement 2, and WCAP-10271 Supplement 2, Revision 1 issued by letters dated February 21,1985, February 22,1989, and April 30,1990. Implementation of the proposed changes is expected to result in an acceptable increase in total RTS yearly unavailability. This increase, which is primarily due to less frequent surveillance, results in an increase of similar magnitude in the probability of an Anticipated Transient Without Scram (ATWS) and in the probability of core melt resulting from an ATWS and also results in a small increase in core damage frequency (CDF) due to ESFAS unavailability, Implementation of the proposed changes is expected to result in a signi6 cant reduction in the probability of core melt from inadvertent reactor trips. This is a result of a reduction in the number of inadvertent reactor trips occuring during testing of RTS instrumentation. For Catawba, this reduction is primarily attributable to less frequent surveillance.
The reduction in core melt frequency from inadvertent reactor trips is sufficiently large to counter the increase in ATWS core melt probability resulting in an overall reduction in total core melt probability.
The values determined by the WOG and presented in the WCAP for the increase in CDF were verified by Brookhaven National Laboratory (BNL) as part of an audit and sensitivity ana!ysis foi the NRC staff. Based on the small value of the increase compared to the range of uncertainty in the CDF, the increase is considered acceptable.
Changes to surveillance test frequencies for the RTS interlocks do not represent a signi0 cant reduction in testing. The currently specified test interval for interlock channels allows the surveillance requirement to be satisfied by verifying that the permissive logic is in its l
required state using the annunciator status light. The surveillance as currently _ required only verifies the status of the permissive logic and does not address verification of channel setpoint or operability. The setpoint verification and channel operability are verified after a refueling shutdown. The definition of the channel check includes comparison of the channel status with other channels for the same parameter. The requirement to routinely verify permissive status is a different consideration than the availability of trip or actuation channels which are required to change state on the occurrence of an event and for which the function availability is more dependent on the surveillance interval. The change in surveillance requirement to at least once every refueling does not therefore represent a signilicant change in channel surveillance and does not involve a significant increase in unavailability of the RTS.
The proposed changes do not result in an increase in the severity or consequences of an accident previously evaluated. Implementation of the proposed changes affects the probability of failure of the RTS but does not alter the manner in which protection is afforded nor the manner in which limiting criteria are established.
Criterion 2 - The proposed license amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.
The proposed changes do not result in a change in the. anner in which the RTS provides m
plant protection. No change is being made which alters the functioning of the RTS. Rather, the likelihood or probability of the RTS functioning properly is affected as described above.
Therefore, the proposed changes do not create the possibility of a new or different kind of accident.
Since Duke Power Company is not proposing to conduct routine channel testing in bypass for Catawba, no hardware changes are necessary to support testing in the bypass mode.
Criterion 3 - The proposed license amendment does not involve a significant reduction in a margin of safety.
The proposed changes do not alter the manner in which safety limits, limiting safety system setpoints, or limiting conditions for operation are determined. The impact of reduced testing other than as addressed above is to allow a longer time interval over which instrument uncertainties (e.g., drift) may act. Experience has shown tot the initial uncertainty assumptions are valid for reduced testing.
Implementation of the proposed changes is expected to result in an everall improvement in safety by:
1)
Less frequent testing will result in fewer inadvertent reactor trips and actuation of Engineered Safety Features Actuation Systein components.
2)
Higher quality repairs leading to improved equipment reliability due to longer allowable repair times.
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Improvements in the effectiveness of the operating staff in monitoring and controlling plant operation. This is due to less frequent distraction of the operator and shift supervisor to attend to instrumentation testing.
The foregoing analysis demonstrates that the proposed amendment to Catawba's Technical Specifications does not involve a significant increase in the probability or consequences of a previously evaluated accident, does not create the possibility of a new or different kind of accident, and does not involve a significant reduction in a margin of safety.
Based upon the preceding analysis, Duke Power Company concludes that the proposed amendment does not involve a significant hazards consideration.
Environmental Impa&hlis The proposed Technical Specification amendment has been reviewed against the criteria of 10 CFR 51.22 for environmental considerations. The proposed amendment does not involve a significant hazards consideration, nor increase the types and amounts of effluents that may be released offsite, nor increase individual or cumulative occupational radiation exposures.
Therefore, the proposed amendment meets the criteria given in 10 CFR 51.22(c)(9) for a categorical exclusion from the requirement for an Environmental Impact Statement.
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