ML20035C271

Notice of Opportunity for Public Comment on Proposed Generic Communication (GL 93-XX), Line-Item TS Improvements to Reduce Testing During Power Operation

ML20035C271
Person / Time
Issue date:	03/25/1993
From:	Marcus G Office of Nuclear Reactor Regulation
To:
Shared Package
ML20035B467	List: ML20035B466 ML20035B468 ML20035B470 ML20035C271
References
GL-93-05, GL-93-5, NUDOCS 9304060425
Download: ML20035C271 (54)

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{{#Wiki_filter:N e IP .~ FEDERAL REGISTER NOTICE (FRN) SOLICITING PUBLIC COMMENTS NUCLEAR REGULATORY COMMISSION Proposed Generic Communication "Line-Item Technical Specification Improvements to Reduce Testing During Power Operation" AGENCY: Nuclear Regulatory Commission. ACTION: Notice of opportunity for public comment.

SUMMARY

The Nuclear Regulatory Commission (NRC) is proposing to issue a generic letter. A generic letter is an NRC document that 1) requests licensees to submit analyses or descriptions of proposed corrective actions, or both, regarding matters of safety, safeguards, or environmental significance, or 2) requests licensees to submit infonaation to the NRC on other technical or administrative matters, r.r, 3) transmits information to licensees regarding approved changes to rules or regulations, the issuance of reports or evaluations of interest to the industry, or changes to NRC administrative procedures.

This draft generic letter presents the results of the NRC staff examination of technical specification surveillance requirements that require testing during power operation. This draft generic letter also provides guidance to assist licensees in preparing a license ameadment request to implement the recommendations that resulted from the examination. The NRC j is seeking comment from interested parties regarding both the technical and regulatory aspects of the proposed generic letter presented under the 9304060425 930325 PDR I&E MISC PDR

N. -- 2 -Supplementary Inferaation heading. This proposed generic letter and. supporting documentation were discussed in meeting number 234 of the Committee { to Review Generic Requirements (CRGR). The proposed generic letter, as I approved by the CRGR, and the supporting documentation are available in the the Public Document Rooms under accession number 9303190122. The NRC-will consider comments received from interested parties in the final evaluation of the proposed generic letter. The NRC final evaluation will include a review i of the technical position and, when appropriate, an analysis of the j value/ impact on licensees. Should this generic letter be issued by the NRC, i it will become available for public inspection in the Public Document Rooms. Comment period expires [ FILL IN DATE 30 DAYS AFTER NOTICE ISSUE). Comments i submitted after this date will be considered if it is practical to do so, but j j assurance of consideration cannot be given except for comments received on or l r before this date. ? ADDRESSES: Submit written comments to Chief, Rules Review and Directives Branch, U.S., Nuclear Regulatoey Commission, Washington, DC 20555. Written comments may also be delivered to Room P-223, Phillips Building, 7920 Norfolk Avenue, Bethesda, Ihryland, from 7:30 am to 4:15 pm, Federal workdays. Copies j of written comments received may be examined at the NRC Public Document Room, l 2120 L Street, NW. (Lower Level), Washington, DC. l t i 1 4 r L.- --- --

s.- ) 3 i i FOR FURTHER INFORMATION CONTACT: Thomas Dunning, (301) 504-1189 SUPPLEMENTARY INFORMATION: Draft Generic Letter, "Line-Item Technical Specification Improvements to Reduce Surveillance Requirements for Testing During Power Operation" Dated at Rockville, Maryland, this 25' day of /ksn 4-1992. ? FOR THE NUCLEAR REGULATORY COMMISSION H/L-Gail H. Marcus, Chief Generic Communications Branch i Division of Operating Reactor. Support Office of Nuclear Reactor Regulation 2 i ) -1

~ ). s:- -TO: ALL HOLDERS OF OPERATING LICENSES OR CONSTRUCTION PERMITS FOR NUCLEAR ] i POWER REACTORS j r

SUBJECT:

LINE-lTEM TECHNICAL SPECIFICATION IMPROVEMENTS TO REDUCE SURVEILLANCE j REQUIREMENTS FOR TESTING DURING POWER OPERATION (Generic Letter 93- ). j The staff of the U.S. Nuclear Regulatory Commission (NRC) has completed a com-l prehensive examination of technical specification (TS) surveillance require-i ments that require testing during power operation. This effort is a part of the NRC Technical Specification Improvement Program (TSIP). The results of 7 this work are reported in NUREG-1366, " Improvements to Technical Specifications Surveillance Requirements," December 1992. NUREG-1366 is available for } i examination in the NRC Public Document Room, 2120 L street, NW, Lower Level, j Washington, DC 20S55 and for purchase from the GPO Sales Program by writing to i the Superintendent of Documents, U.S. Government Printing Office, P.O. Box 37082, Washington, DC 20013-7082. In performing this study, the ii staff found that while the majority of the testing at power is important, s safety can be improved, equipment degradation decreased, and an unnecessary .[ t burden on t.ersonnel resources eliminated by reducing the amount 'of testing that i i' the TS require at power operating conditions. However, only a small fraction. i of the TS surveillance intervals was considered to warrant relaxation. The staff has prepared the enclosed guidance to assist licensees in preparing a ' i l license amendment request to implement these recommendations as line-item TS i improvements. The NRC issued improved standard technical specifications in i September 1992 that incorporated the recommendations of NUREG-1366. [ i; l 1 r i

^ ~ .r h Licensees and applicants are encouraged to propose TS changes that are con-l C sistent with the enclosed guidance. The NRC project managers will review l 1 f requests for license amendments to verify that they conform to this guidance. Please contact the project manager or the contact indicated below if you have f questions on this matter. ) l Any response to the suggestion that licensees or applicants propose these TS j changes is voluntary. Therefore, any action taken in response to the guidance 'f i provided in this generic letter is not a backfit under Section 50.109 of Title 10 to the Code of Federal Reaulations (10 CFR 50.109). The following i information, although not requested under the provisions of 10 CFR 50.54(f), i would be helpful to the NRC in evaluating the cost of complying with the j suggestion to propose TS changes addressed by this generic letter: i 1. The licensee staff time and costs to prepare the amendment request. l 2. An estimate of the long-term costs that would be incurred or saved in the l future as a result of implementing this TS change. I

Contact:

T. G. Dunning, NRR (301) 504-1189 f i

L This request is covered by Office of Management and Budget Clearance Number 3150-0011, which expires June 30, 1994..The estimated average number of burden hours is 40 person hours per licensee response,. including those needed to assess the new recommendations, search data sources, gather and analyze the data, and prepare the required letters. Send comments regarding this burden estimate or any other aspect of this collection of information, including suggestions for reducing this burden, to the Information and Records Management Branch (MNBB 7714), Division of Information Support Services, Office of Information and Resource Management, U.S. Nucl~ ear Regulatory Commission, Washington, D.C. 20555 and to Ronald Minsk, Office of Information and Regulatory Affairs (3150-0011), NE0B-3019, Office of Management and Budget, Washington, D.C. 20503. Sincerely, i l ) James G. Partlow 9 Associate Director for Projects Office of Nuclear Reactor Regulation !l

Enclosure:

i As stated l

/ L.

l GUIDANCE FOR IMPLEMENTING LINE-ITEM TECHNICAL SPECIFICATI0N' IMPROVEMENTS TO REDUCE TESTING DURING POWER OPERATION i

INTRODUCTION This enclosure provides guidance for preparing a license amendment request to l change the technical specifications (TS) to reduce testing during power opera-i tion. These line-item TS improvements are based on the recommendations of an NRC study that included a comprehensive examination of surveillance require-j ments and is reported in NUREG-1366, " Improvements' to Technical Snecifications Surveillance Requirements." i Each of the applicable NUREG recommendations is addressed herein with examples i of TS changes based upon standard technical specification (STS) requirements f that were used as model TS when many plants obtained their operating license. The title and number of each of these line-item improvements corresponds to.the section title and number in NUREG-1366 in which the staff recommended the l change The staff is providing the NUREG recommendation for each item, but the ); NUREG finding is provided only where it is necessary to clarify the intent of l l the NUREG recommendation. The staff is providing the wording for the changes } to specific TS sections using the noted model STS requirements, with the l.i reactor vendor identified in brackets and noted as "(Typ)" where it.is typical l of the change that applies to the TS for reactors of more than one type or l vendor. The staff is providing the wording for a few of the recommendations from TS changes that have been approved for a specific plant. In this case, t I l I

> the plant is identified in brackets as the source of the guidance. The proposed TS changes for plants that have TS in a format that is different than the STS should be consistent with the intent of the NUREG recommendation, the enclosed guidance, and the format of individual plant TS. COMPATIBILITY WITH OPERATING EXPERIENCE Licensees should not propose changes to extend any surveillance interval if the recommendations of NUREG-1366 are not compatible with plant operating exper-ience. Therefore, each licensee should include a statement in the license amendment request that all proposed TS changes are compatible with plant operating experience and are consistent with this guidance. LINE-ITEM TS IMPROVEMENTS 4.1 Moderator Temoerature Coefficient Measurements (pWR) Findings: (1) Technical Specifications require a determination of moderator temperature coefficient at 300 ppm boron concentration. (2) If measured moderator temperature coefficient is more negative (less conservative than the TS value), the licensee must measure the moderator temperature coefficient every 14 EFPDs until the end of the cycle. (3) Measuring the moderator temperature coefficient at low boron concentrations is difficult. (4) VEPC0 [ Virginia Electric Power Company] proposed a method for eliminating this requirement

~ t- - T -l below 60 ppm. (5) Method is plant-specific. l l t Recommendation: Other licensees may wish to use the VEPC0 approach. li i i The following condition must be met and addressed to justify the use' of the 'l VEPC0 approach: 1 Results of plant-specific analysis are required that show that the maximum l; rassible change in moderator temperature coefficient (MTC) from 60 ppm to j the end of the operating cycle (EOC) is less than the difference:in the values of MTC from 60 ppm to EOC HTC that are specified in this Technical Specification. } 3/4.1 Reactivity Control Systems - Moderator Temperature Coefficient, l [W STS (Typ)] TS 4.1.1.3: j The MTC shall be determined to be within its limits during each fuel cycle f as follows:

]

t The MTC shall be measured and compared to the BOL limit specification a. 3.'..l.3a., above, prior to initial operation above 5% of RATED i THERMAL POWER, after each fuel loading; and b. The MTC shall be measured at any THERMAL POWER and compared to -{3.0]. l ? x 10-4 delta-k/k/ degree-F (all rods withdrawn, RATED THERMAL POWER [ t condition) within 7 EFPD after reaching an equilibrium boron concen-I f i r ~

1 tration of 300 ppm.*, In the event this comparison indicates the MTC i is more negative than -[3.0] x 10-4 delta-k/k/ degree-F, the MTC shall be measured, and compared to the E0C MTC lir,it.of Specification i l 3.1.1.3b., at least once per 14 EFPD during the remainder of the fuel j t

cycle, j

i

• Once the eouilibrium boron concentration (all rods withdrawn. RATED i

THERMAL POWER condition) is 60 com or less. further measurement of the MTC I may be suspended if the measured MTC at an eouilibrium boron concentration ) of 60 ppm or less is less neaative than Ithe credicted value of MTC at ? a 60 onml. i (Footnote added to be consistent with recommendation.) l 4.2 Control Rod Movement Test ~ 4.2.1 Pressurized Water Reactors t I Recommendation: Change frequency of the PWR control rod movement test to quarterly. 1 l1 3/4.1.3 Movable Control Assemblies, [W STS (Typ)] TS 4.1.3.1.2: I' Each full-length rod not fully inserted in the core shall be determined to be OPERABLE by movement of at least 10 steps in any one direction at least { once per 92 days. f ? 4 'i

f ~ i 3 h _.5-(Replaced "31" with "92" days.) l 4.2.2 Boilina Water Reactors -l a ' i Recommendation: The TS should be changed to require that.if a con - i trol rod is immovable because of' friction or mechanical interference, f the other control rods should be tested within 24 hours and every l 7 days thereafter. (NOTE: Existing TS requirements include testing control rods every 7 days. Therefore, the recommendation to change j t the frequency for tests that apply when a control rod is immovable'to j i include "once every 7 days thereafter" is already covered by the i existing requirements that apply before the occurrence of an -[ immovable rod as noted in item a below.)' 3/4.1.3 Control Rods, [BWR/6 STS (Typ)] TS 4.1.3.I'2: When above the low power setpoint of the RPCS, all withdrawn control rods l not required to have their directional-control valves disarmed electric-: ll ally or hydraulically shall be demonstrated OPERABLE by moving each control rod at least one notch: l a. At least once per 7 days, and t b. Within 24 hours when any control rod is immovable as a result of i excessive friction or mechanical interference. [ i i I (Replaced "At least once per" with "Within.") .i ~ e 9 p 4

't .._o t 4.3 Standby Liouid Control System BWR 1 Recommendation: (1) Explosive valves should be tested once each j refueling interval for fuel cycles up to 24 months duration. (2) The i .J SBLC system pump test should be required by technical specifications quarterly, in agreement with the ASME Code. i 3/4.1.5 Standby Liquid Control System. [BWR/5 STS] TS 4.1.5: The standby liquid control system shall be demonstrated OPERABLE: a. At least once per 24 hours by verifying that: j (No change to items a.1, a.2, and a.3.) 3 b. At least once per 31 days by: f 1. (Unused) l (Item b.1 is noted as " Unused" since it is relocated to item c.I, below No change to items b.2, b.3, and b.4.) l c. At least once per 92 days by: l i (New item c. The current item c is renumbered-as item d, below.) l 1. Starting both pumps and recirculating demineralized water to the l I test tank. i i (Item c.1 is relocated from b.1, above.) e I

t .y* e i i i - 7'- a i -d. At' least once each refuelino interval by: Ij 'i (Replaced "per 18 months during shutdown" with "each refueling interval.") j 1. Initiating one of the standby liquid control system loops, including an explosive valve, and verifying that a flow path from the pumps to the reactor pressure vessel is available by l i pumping demineralized water into the reactor vessel. The .i i replacement charge for the explosive valve shall be from the ] same manufactured batch as the one fired or from another batch -i which has been certified by having one of the batch successfully fired. Both injection loops shall be tested in any two ) consecutive refuelino intervals. i (Item c.1 was relocated-from item b.1, above, and replaced "36 months" ] with "any two consecutive refueling. intervals." No change to items d.2 [. through d.5 that were renumbered as items c.2 through c.5.) 1 4 3/4.1.5 Standby Liquid Control System, [BWR/4 STS] TS 4.1.5: l i i i The standby liquid control system shall be demonstrated OPERABLE by: I Demonstrating that when tested (pursuant to Specification 4.0.5) (at-c. least once per 92 days), the minimum flow requirement of (41.2) gpm

l at a pressure of greater than or equal to (1220) psig 'is met.

4 1 4

P j ~ (Item c is consistent with the recommended change. No change to item c or-l to items a and b is required.) f t i d. At least once each refuelina interval by. 'I (Replaced "per 18 months during shutdown" with "each refueling interval.") 1. Initiating one of the standby liquid control. system loops, j t' including an explosive valve, and verifying.that a flow path from the pumps to the reactor pressure vessel is available by r pumping demineralized water into the reactor vessel. The-replacement charge for the explosive valve shall be from the same manufactured batch as the one fired or.from another batch i which has been certified by having one of the batch successfully -l fired. Both injection loops shall be tested in any two consecutive refuelina intervals. 1 (Replaced "36 months" with "any two consecutive refueling intervals." No l l change to items d.2 through d.4.) l I 4.4 Closure Time Testina of Scram Discharae Volume Vent and Drain Valves (BWR) t i i Recommendation: Other BWR licensees may wish to use the Georgia-Power Co./GE method on a plant-specific basis to extend the SDV vent and drain valve closure l 'I time requirement. f i b

.^-

i _g. I The following' condition must be met and addressed to justify the use.of the t Georgia Power Co./GE method: j i I Results of plant-specific analysis are required using approved methods, i for example, MDE 103 1184, to derive a new vent and drain valve closure time. The analysis must take into account assumptions about the value of each of the following factors: (1) scram time, (2) displacement volume of l water per individual control rod drive, (3) average expected post-scram leakage flow per individual control rod drive, (4) SDV drain flow before { isolation, and (5) minimum scram discharge volume. -l t I 3/4.1.3 Control Rods, [BWR/6 STS) TS 4.1.3.1.4: Plant-specific valve closure times should be provided in item a.1 of TS1 i 4.1.3.1.4 that is addressrd under the recommendations for Sec" on 4.5, l below. 9 4.5 Reactor Scram Testino to Demonstrate Operability of Scram Discharae Volume (SDV) Vent and Drain Valves (BWR) Recommendations: (1) Remove the requirement for a scram check of SDV vent and drain valve operability at 50% rod density or less. } i (2) Require an evaluation of SDV system response after each scram to i 1 verify that no abnormalities exist prior to plant restart. (3) Require vent and drain valve operability testing during a scram l from shutdown conditions. 3 L t i

l ; l 3 '3/4.1.3 Control Rods, [BWR/6.STS] TS 4.1.3.1.1: -j The scram discharge volume drain and vent valves shall be demonstrated OPERABLE by: i 'l t a. At least once per 31 days verifying each valve to be open, and I 't b. Evaluatina SDV system response crior to olant startuo after each scram to verify that no abnormalities exist. i i .I (This change to Item b replaces the 92-day cycling test for each valve.) The scram discharge volume shall be determined OPERABLE by demonstrating: i I The scram discharge volume drain and vent valves OPERABLE, where a. 3 control rods are scram-tested from a shutdown condition at least .I once per'18 months, by verifying that the drain and vent valves: ~ l (Replaced "a 50% rod density or less" with "a shutdown condition.") j 3. Close within (30) seconds after receipt of a signal for l control rods to scram, and j l t 2. Open when the scram signal is reset. l l b. (No change.) -l i i

g i

4: l - l l 5.1 Nuclear Instrumentation Surveillance (PWR1 c l ) Recommendation: Change surveillance intervals of analog channel l functional tests of nuclear instrumentation to quarterly. ] Plant-specific requirements have been established based upon the staff's review-and approval of topical reports for extending the surveillance intervals for j reactor protection system cilannels from monthly to quarterly-as follows: e Letter from C. O. Thomas (NRC) to J. J. Sheppard (WOG - CP&L), of f February 21, 1985,

Subject:

Acceptance for Referencing of Licensing l Topical Report WCAP-10271, " Evaluation of Surveillance frequencies and Out -l Of Service Time for the Reactor Protection Instrumentation Systems." Also i [ see Westinghouse Owners-Group Guidelines for Preparing Submittals I Requesting Revision of Reactor Protection System Technical Specification, 1 Revision 1, per letter 0G-158, L. D. Butterfield (WOG - CECO) to Harold R. t Denton (NRC), of September 3, 1985. i [ l Letter from A. C. Thadani (NRC) to T. A. Pickens (BWROG - NSPC), of -i 7 July 15,1987,

Subject:

General Electric Company (GE) Topical Reports [l NEDC-30844, "BWR Owners Group Response to NRC Generic ~ Letter 83-28," and } l NEDC-30851P, " Technical Specification Improvement Analysis for BWR RPS." t Letter from A. C. Thadani (NRC) to C. W. Smythe (BWOG - GPU), of December I 5, 1988,

Subject:

NRC Evaluation of BWOG Topical Report BAW 10167 and l Supplement 1, " Justification for Increasing the Reactor Trip System On-I L b

u ~ r i l l l Line Test Interval." I For CE plants, there is no generic evaluation for increasing RPS surveillance intervals. Therefore, guidance on the recommended TS change is as follows: I i i 3/4.3.1 Reactor Protective Instrumentation, [CE STS] TS Table 4.3-1: i TABLE 4.3-1 i REACTOR PROTECTIVE INSTRUMENTATION SURVEILLANCE REQUIREMENTS ) CHANNEL MODES FOR WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE i FUNCTIONAL UNIT CHECK CALIBRATION TEST IS REOUIRED-l 'I t i i

2. Linear Power Level j

- High S D(2,4)M(3,4), R I, 2 ~ c CHANNEL MODES FOR WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE j FUNCTIONAL UNIT CHECK CALIBRATION TEST IS RE0VIRED

3. Logarithmic Power Level - High S

R(4)(10) Q and S/U(1) 1,2,3,4,5 (Changed Channel Functional Test frequency f,om "M" to "Q.")

_~-_ P-I 5.2 S1 ave Relav Yestina (PWR.-BWR) Recommendation: Perform relay testing on a staggered test basis over a cycle and leave the tests carrying highest risk to a refueling outage or other cold shutdown. The following condition must be met and addressed to justify this approach: i Plant-specific analysis is required to identify those slave relays that should be tested only during a refueling outage or other cold shutdown- ~ i because of a high risk associated with such testing. j i 3/4.3.2 Engineered Safety Feature Actuation System (ESFAS) Instrumentation, [W STS (Typ)] TS Table 4.3-2: TABLE.4.3-2' r ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION l t -l SURVEILLANCE REQUIREMENTS SLAVE RELAY FUNCTIONAL UNIT TEST [High risk items] R [Non-high risk items) 111

L l l (1) Every ( ') days on a STAGGERED TEST BASIS. l l (Add " SLAVE RELAY TEST" column to TS tables that do not _ have it and add j footnote (1). The test frequency for high risk items is "R," and the test. I i frequency for the remaining items is to be specified in the footno'te at. the current TS frequency for slave relays tests, but on a staggered test basis.) i r l l 5.3 sest Intervals for RPS and ESFAS (PWR; BWR) l Recommendation: Test three-channel systems on the four-channel l schedule. Do not test one of the thne channels during a four-l channel test interval. Thus, the sequence of testing would be: Three channels Four channels A A I L B B I C C i D 1 A A 3/4.3.1 Reactor Trip System Instrumentation, [W STS (Typ)] TS Table 4.3-1: 1 i TABLE 4.3-1 l TABLE NOTATION l a W J

~- e . (11) Each channel shall be tested at least every 92 days on a STAGGERED TEST BASIS. Individual channels in three-channel systems may be tested on the same schedule for the correspondino channel of four-channel systemst (The addition to Note (11), which specifies staggered testing of RPS' channels, allows testing of three-channel systems on the same schedule for the corresponding channel of four-channel systems. The same addition f should be made to the corresponding note in TS Table 4.3-1 that requires staggered testing of ESFAS channels.) ^ 5.4 Hydroaen Monitor Surveillance (PWR. BWR) e Recommendation: Change frequency of calibration to once each 9 refueling interval and analog channel operational test to quarterly. 4-3/4.6.5 Combustible Gas Contro' - Hydrogen Monitors, [W STS (Typ)] TS 4.6.5.1: Each hydrogen monitor snall be demonstrated OPERABLE by the performance of a CHANNEL CHECK at least once per 12 hours, an ANALOG CHANNEL OPERATIONAL TEST at least once oer 91 days, and at least once each refuelino interval by performing a CHANNEL CALIBRATION using sample gas containing: (Replaced "31" with "92" days and "92 days on a STAGGERED TEST BASIS" with - each refueling interval.")

i e,- p .:. a. One volume percent hydrogen, balance nitrogen,. and b. Four volume percent hydrogen, balance nitrogen. 5.5 Reactor Trio Breaker Testina (PWR) A TS change was not recommended for this item. 5.6 Power Rance Instrument Calibration PWR) A TS change was not recommended for this item. 5.7 Control Element Assembly calculator Surveillance (CE CPC PWR). Recommendation: Extend the surveillance interval from monthly to; quarterly. 5 t 3/4.3.1 Reactor Protective Instrumentation, [CE STS] TS Table 4.3-1: TABLE 4.3-1 REACTOR PROTECTIVE INSTRUMENTATION SURVEILLANCE REQUIREMENTS i I CHANNEL MODES FOR WHICH l t CHANNEL CHANNEL FUNCTIONAL. SURVEILLANCE-l FUNCTIONAL UNIT CHECK CALIBRATION TEST IS REOUIRED j I

15. CEAC Calculators S

R Q,R(6) 1,-2 l '{

l y t . j l (Channel Functional Test frequency changed from "M" to "Q.") 5.8 Incore Detector Surveillance (CE and B&W PWRs)- Recommendation: The B&W surveillance requirement for incore detectors should be used for CE plants. [ -l i 3/4.3 Instrumentation - Incore Detectors, [B&W STS) TS 4.3.3.2: The incore detector system shall be demonstrated OPERABLE. a. By performance of a CHANNEL CHECK within 7. days prior to its;use:for l measurement of the AXIAL POWER IMBALANCE or the QUADRANT POWER TILT. j i b. At least once per 18 months by performance of a CHANNEL CALIBRATION 'l which does not include the neutron detectors. i 'I 5.9 Resoonse Time Testina of Isolation Instrumentation (PWR. BWR) a ~! i Recommendation: Delete requirement from both BWR and PWR technical specifications to perform response time testing where the required l response time corresponds to the diesel start time. 3/4.3.2 ESFAS Instrumentation, {W STS '(Typ)] TS Table 3.3-5: } t TABLE 3.3-S ENGINEERED SAFETY FEATURES RESPONSE TIMES s 1

~, i ] i INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS (Identify item) IB j (Replaced specified response time with "NA" for those Initiating Signal' and Function entries where the response time [ excluding the response time of valves that is confirmed under the inservice testing program] corresponds to the diesel start time.) l 5.10 Source Ranae Monitor and Intermediate Ranoe Monitor Surveillances (BWR1 f Recommendation: The calibration interval for the.BWR CRMs and IRMs should be changed to once each refueling interval. 3/4.0.6 Control Rod Block Instrumentation, [BWR/6 STS (Typ)] Table 4.3.6-1: TABLE'4.3.6-I l CONTROL R0D BLOCK INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL OPERATIONAL 1 CHANNEL FUNCTIONAL CHANNEL CONDITIONS-IN WHICH TRIP FUNCTION CHECK TEST CALIBRATION SURVEILLANCE RE0UIRED I 3. SOURCE RANGE MONITORS i f

a. Detector not

e 1 I. full in NA S/U(b),W NA 2, 5

b. Upscale NA S/U(b),W B

2, 5

c. Inoperative NA S/U(b),W NA'

_ 2, 5 1 t-

d. Downscale NA S/U(b),W B

2, 5 [ 'I 4. INTERMEDIATE RANGE MONITORS f

a. Detector not full in NA S/U(b),W NA 2, 5

b. Upscale NA S/U(b),W H

2, 5 i

c. Inoperative NA S/U(b),W NA 2, 5

d. Downstale NA S/U(b),W B

2, 5 I (Changed Channel Calibration frequency from "Q" to "R.") i 5.11 Calibration of Recirculation Flow Transmitters (BWR)' f i A TS change was not recommended for this item. l 5.12 Autoclosure Interlocks (PWR. BWR) l i A TS change was not recommended for this item. l 5.13 Turbine Overspeed Protection System Testina (PWR. BWR) J ~

i. . l l Recommendation: Where the turbine manufacturer agrees, the turbine i valve testing frequency should be changed ta quarterly. The following condition must be met and addressed to justify the use of this approach-A statement is required confirming the turbine manufacturer's 1 concurrence with the proposed change. 3/4.3.4 Turbine Overspeed Protection, [W STS (Typ)] TS 4.3.4.2: The above required Turbine Overspeed Protection System shall be f demonstrated OPERABLE: i a. At least once per 92 days by direct observation of the movement of i each of the following valves through at least one complete cycle from I the running position-i (No change to the listing of turbine valves. Replaced "7" with "92" days I and " cycling" with " direct observation of the movement" of each valve.) i b. (Unused) (Item b is noted as " Unused" since surveillance for direct observation of valve movement is included in item a above.) i 't

i ) i 5.14 Radiation Monitors (PWR. BWR) 1 ( I Recommendation: In order to decrease licensee burden and increase j the availability of radiation monitors, change the monthly channel functional test to quarterly. 3 i I l 3/4.3.2 Engineered Safety Feature Actuation System Instrumentation, j i [CE STS (Typ)] TS Table 4.3-2: l t i Table 4.3-2 . [ ENGINEERED SAFETY FEATURE ACTUATION SYSTEMS INSTRUMENTATION SURVEILLANCE REQUIREl'ENTS CHANNEL MODES FOR WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST IS REOUIRED i

5. SHIELD BUILDING FILTRATION (SBFAS) l i

!a

e. Containment i

Radiation - High Gaseous Monitor S R D I, 2, 3, 4 . 1 Particulate Monitor S R Q I, 2, 3, 4 i t 4 -%+ c e

i h- .l .' + 1 Area Monitor S R D.- 1, 2, 3, 4 q r f i CHANNEL MODES FOR WHICH CHANNEL CHANNEL JUNCTIONAL SURVEILLANCE i FUNCTIONAL UNIT CHECK CALIBRATION TEST 'IS'REOUIRED

7. CONTAINMENT PURGE b

VALVES ISOLATION 'l

e. Containment i

Radiation - High Gaseous Monitor S R H. 1, 2, 3, 4 Particulate Monitor S R Q 1,2,3,4 Area Monitor S R H 1, 2, 3, 4 (Channel Functional Test frequency changed-from "M" to "Q.") 3/4.3.3-Monitoring Instrumentation - Radiation Monitoring' Instrumentation, [CE STS) TS Table 4.3-3: TABLE 4.3-3 . RADIATION MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS

C CHANNEL MODES FOR WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST IS RECUIRED b (All items) (No change) (No change) Q (No change) i (Channel Functional Test frequency changed from "M" to "Q.") { i 3/4.3.3 Monitoring Instrumentation - Radioactive Liquid Effluent Monitoring _ Instrumentation, - Radioactive Gaseous Effluent Monitoring Instru-mentation, [W STS(Typ)] TS Table 4.3-8 and Table 4.3-9: [ No change in existing STS guidance is required. The surveillance interval for an Analog Channel Operational Test (equivalent of a { Channel Functional Test for other reactor vendors) is'specified as "Q" (quarterly). Plants having a monthly test interval for this l surveillance may request a change in the test interval to quarterly. t 5.15 Radioactive Gas Effluent Monitor Calibration Standard (PWR. BWR)' -l A TS change was not recommended for this item. 6.1 Reactor Coolant System isolation Valves (PWR) { Recommendation: Increase the 72-hour time for remaining in cold ) h

- t y , i shutdown without leak testing the RCS isolation valves to 7 days. l 3/4.4.6 Reactor Coolant System Leakage - Leakage Detection Systems, [W STS (Typ)] TS 4.4.6.2.2: i Each Reactor Coolant System Pressure Isolation Valve specified in r Table 3.4-1 shall be demonstrated OPERABLE by verifying leakage to be i within its limit: i a. At least once per 18 months, l 4 b. Prior to entering MODE 2 whenever the plant has been in COLD SHUTDOWN + for 7 days or more and if leakage testing has not-been performed in + the previous 9 months. . t (Replaced "72 hours" with "7 days." _No change to items c, d and ei) i 6.2 Power-for Pilot-) Operated Relief Valves (PORVs) and Block Valves (PWR) Recommendation: Direction concerning PORV and block valves { surveillances will be provided in the resolution of GI-70 and GI-94. i This guidance was provided by Generic Letter 90-06 of June 25, 1990. 6.3 Hioh Point Vent Surveillance Testino (PWR1 e n ,-_-_w .m i.-., m

i _ 25 - Recommendation: Licensees to evaluate applicability of Catawba Technical Specification Ba.ses with respect to high point vent 1 surveillance testing and revise the frequency of testing of RCS vent i J valves to cold shutdown or refueling if appropriate. l l ~ Catawba TS Bases 3/4.4.11, Reactor Coolant System Vents, states the following: Reactor Coolant System vents are provided to exhaust noncondensible gases l and/or steam from the primary system that could inhibit natural circula-tion core cooling. The OPERABILITY of at least one Reactor Coolant System l l vent path from the reactor vessel head, and the pressurizer steam space ensures that the capability exists to perform this function. { i The valve redundancy of the Reactor Coolant System vent paths serves to i minimize the probability of inadvertent or irreversible actuation while ensuring that a single failure of a vent valve, power supply or control l system does not prevent isolation of the vent path. The function, capabilities, and testing requirements of the Reactor { Coolant System vent systems are consistent with the requirements of 'f i Item II.B.1 of NUREG-0737, " Clarification of TMI Action Plan Require-l I ments," November 1980. Licensees should confirm and incorporate the applicable portions of the above I Catawba TS Bases into the Bases Section for Reactor Coolant System Vent TS to implement the following TS change. -l ~! e -e,~

. l 3/4.4.11 Reactor Coolant System Vents, [W STS (Typ)] TS 4.4.11.1: Each Reactor Coolant System vent path block valve not required to be l closed by ACTION a. or b., above, shall be demonstrated.0PERABLE at least once per COLD SHUTDOWN. i f no+ oerformed within the previous 92 days, by. operating the valve through one complete cycle of full travel from the l control room. 1 (Added " COLD SHUTOOWN, if not performed within the previous" 92 days.) 6.4 Low-Temperature Overpressure Protection (PWR) l A TS change was not recommended for this item. - i, 6.5 Specific Activity of the Reactor Coolant 100/E (PWR. BWR) A TS change was not recommended for this item. ] i i 6.6 Pressurizer Heaters (PWR) Recommendation: The capacity of pressurizer heaters should be tested once each refueling interval for those plants without dedicated j safety-related heaters. The capacity of pressurizer heaters should be tested every 92 days for plants with dedicated safety-related heaters. For those PWRs which have pressurizer heaters tied to a vital bus, no testing of switching between power supplies should be -,n-n -p q .y .ee--.,e a,+ w

,,9

. {

required. { 3/4.4.3 Pressurizer, [W STS.(Typ)] TS 4.4.3.2: i 1 The capacity of each of the above required groups'of pressurizer heaters f shall.be verified by energizing the heaters and measuring circuit-current at least once per 92 days. i (No change. This TS guidance is applicable for plants with dedicated - safety-related heaters.) The capacity of each of the above required groups of pressurizer heaters j shall be verified by energizing the. heaters and measuring circuit current - at least once each refuelina interval. I (Replaced "per 92 days" with "each refueling interval." Applicable for ] i plants without dedicated safety-related heaters.) b i 3/4.4.3 Pressurizer, [W STS (Typ)] TS 4.4.4.3: The emergency power supply for the pressurizer ' heaters shall be demonstra-- ted OPERABLE at least once per 18 months by manually transferring power from the normal to the emergency power supply and energizing the heaters. f (No change, but this TS is not applicable for plants with some pressurized j heaters permanently tied to a vital bus and it may be removed.) l ) e t ~

... ~ v - 28 7.1 Surveillance of Boron Concentration in the Accumulator / Safety in_iection/ Core Flood Tank (PWR) Recommendation: It should not be necessary to verify boron concen-tration of accumulator inventory after a volume increase of 1% or more if the makeup water is from the RWST and the minimum concentra-tion of boron in the RWST is greater than or equal to the minimum boron concentration in the accumulator, the recent RWST sample was within specifications, and the RWST has not been diluted. 3/4.5.1 Accumulators - Cold Leg Injection, [W STS (Typ)] TS 4.5.1.1.1: Each cold leg injection accumulator shall be demonstrated OPERABLE: a. (No change.) b. At least once per 31 days and within 6 hours of.each solution volume increase of greater than or equal _ to [1%' of tank volume) by verifying the boron concentration in the water-filled accumu-lator. This surveillance is not reouired when the~ volume increase makeuo source is the RWST and the RWST has not been diluted since verifyina that the RWST baron concentration is-eoual to or areater than the accumulator boron-concentration limit. (Added clarification to note when surveillance is not required. For B&W

l i i h-I t i 1 -l and CE plants, the term " cold leg injection accumulator" is replaced with " core flooding tank" or " safety injection tank," respectively, and "RWST" l is replaced with " borated water storage tank" or " refueling water tank,;" respectively.) 7.2 Verification That ECCS lines Are Full of Water (Contain No Air) (PWR) f A TS change was not recommended for this item. ] i ( 7.3 Verification of Proper Valve Lineuos of ECCS and Containment Isolation ] Valves (PWR. BWR) . i i A TS change was not recommended for this item. r i 7.4 Accumulator Water level and Pressure Channel Surveillance Reouirements' (PWR) Recommendation: (1) Licensees to examine channel checks surveillance j f and operational history to determine if.there is a basis for justify-ing the extension of frequency for analog channel operational tests-l l for pressure and level channels. (2) Add a condition to the ECCS l accumulator LCO for the case where "One accumulator is inoperable due j l to the inoperability of water level and pressure channels," in which l the completion time to restore the accumulator to operable status l I will be 72 hours. f i-f [ 4

. i .. ~ i The f4RC staff and industry effort to develop new STS recognized that accumula-l l tar instrumentation operability is not directly related to the capability of l the accumulators to perform their safety function. Therefore, surveillance. requirements for this instrumentation are being relocated from the new STS and the only surveillance that is being retained is that required to confirm that i the parameters defining accumulator operability are within their specified i limits. } l 3/4.5.1 Accumulators - Cold leg Injection, [W STS (Typ)] TS 4.5.1.1.1: j i Each cold leg injection accumulator shall be demonstrated OPERABLE: l a. At least once per 12 hours by: i 1. Verifying that the contained borated water volume and nitrogen j cover-pressure in the tanks are within their limits, and-i i 3 l (Removed the reference to verifying operability "by the absence of alarias" i j i' consistent with the removal of the surveillance requirements for this i instrumentation. Added clarification to verifying that the noted j parameters are within their limits.) f i 2. Verifying that each cold leg injection accumulator isolation 4 valve is open. i r 4 i i (flo change for item a.2.) l l 3/4.5.1 Accumulators - Cold Leg Injection, [W STS (Typ)] TS 4.5.1.1.2: l [ t p.,,.

f 0 ~ i j n-I Each accumulator water level and pressure channel shall_ be demonstrated l OPERABLE: a. At least once per 31 days by the performance of an analog 5 channel operational test, and t b. At least once per 18 months by:the performance of a CHANNEL t CALIBRATION. i Specification 4.5.1.1.2 above may be removed from TS but should be retained as e an existing plant procedure requirement that may be subsequently modified under plant change control procedures and the related requirements of the Administra-tive Controls Section of the TS. 7.5 Visual Insoection of the Containment Sumo (PWR) i i Recommendation: Inspection of the containment at least once daily if' l l the containment has been entered that day, and during.the final. entry to ensure that there is no loose debris that would clog the sump. l i i ~ i 3/4.5.1.2 ECCS Subsystems - Tavg Greater Than or Equal to (350) degrees-F, [CE STS (Typ)] TS 4.5.2: 't i Each ECCS subsystem shall be demonstrated OPERABLE by: l (No change to items a and b.) t

n l i t ! o c. By visual inspection which verifies that no loose debris (rags, j i trash, clothing, etc.) is present in the containment.which could be ia transported to the containment sump and cause restriction of the pump j i suctions during LOCA conditions. This visual inspection shall be performed: I a 1. For all accessible areas of the containment prior to. establishing CONTAINMENT INTEGRITY, and i e .I 2. At least once daily of the areas affected within contain-q ment by containment entry and durina the final entry when l l . CONTAINMENT INTEGRITY is established. .{ (The underlined additions were made, and "at the completion of containment entry" was removed as it implied an inspection separate from that activity-for which the containment entry was made.) j ~ i 7.6 Verification of Boron Concentration in the Boron Iniection Tank (Westinahouse PWR) Recommendation: Measure concentration of boron in'the boric acid l 1 storage tank rather than in the BIT if it'can be justified that the j concentrations are the same. I s The following condition must be met and addressed to justify the use of this j approach-l ]

e.- - $_- i A justification is required that the measurement of the boron concentra-tion in the boric acid storage tank verifies the boron concentration in the BIT. i 'l b i 3/4.5.4 Baron Injection System - Baron Injection Tank, (W STS] TS 4.5.4.1: i The boron injection tank shall be demonstrated OPERABLE by: a. Verifying the contained borated water volume at least once per 7 days, 1 i b. Verifying the boron concentration of the water in the tank by measurina the boron concentration in the boric acid storace tank once per 7 days, and (Added clarification of where measurement is made.) l i c. Verifying the water temperature at least once per 24 hours. l (No change for item c.) 8.1 Containment Sorav System (PWR) 1 Recomnendation: The surveillance interval [ air or smoke flow test) j i shoultl.be extended to 10 years. Recent Experience: On June 11, 1991, the Southern California Edison Company 1 I

l l ol' .l I 'l v-. ! (SCE) reported that a containment spray system (CSS) air flow test for San Onofre Unit 1 indicated that several nozzles were blocked. SCE investi-l gated and found that seven nozzles were clogged with sodium silicate, a coating i i material that was applied to the carbon steel CSS piping in 1977. The licensee [ conducted air flow tests in 1980,1983, and 1988 and obtained acceptable j results. 3 This event does not alter the recommendation for an extension of the air flow test surveillance interval for plants with the more commonly used stainless steel piping system. However, licensees for plants using carbon steel piping l must justify any change in the surveillance interval because of the San Onofre i experience. i i t 3/4.6.2 Depressurizing and Cooling Systems - Containment Spray System, l [CE STS (Typ)] TS 4.6.2.1: I Each Containment Spray 3ystem shall be demonstrated OPERABLE: l i i i d. .At least once per 10 years by performing an air or smoke flow test. [ through each spray header and verifying each spray nozzle is unobstructed. I (Changed the surveillance interval from "5" to "10" years.) t 8.2 Containment Purae Supply ~and Exhaust Isolation Valves (PWR) l i i

3,- -t ).

• A TS change was not recommended for this item.

8.3 Ice Condenser Inlet Doors (PWR) Finding: Duke Power Co. justified'a surveillance interval' for-containment inlet door testing that eliminated the need for a shutdown. [ Duke Power Co. had 6 years of testing experience for-McGuire Units 1 and 2 without a failure and the design does not allow water condensation to freeze, a common cause of stuck doors.] Recommendation: The Duke proposal may be used by other utilities if it can be justified on a plant-specific basis. 3/4.6.7 Ice Condenser - Ice Condenser Doors, [McGuire TS (Typ)] TS 4.6.5.3.1: P Inlet Doors - Ice condenser inlet doors shall be: l a. (No change) t b. Demonstrated OPERABLE during shutdown at least once each refuelina interval by: (Replaced "per 9 months" with "each refueling interval.") i 1) (No change.) .l _~

. -~. a;~~ i i 2) (No change.) h 3) Testing all doors and verifying that the torque required to_ open each door is less than [195] inch-pounds when the door is } 40 degrees open. This torque is defined as the " door opening i torque" and is equal to the nominal door torque plus a frictional torque component. l i (Replaced "a sample of at least 25% of the" with "all" and removed the ] last sentence of this section relating to selecting door samples such that all doors are tested at least once during four test intervals.) l ? 8.4 Testina Suppression Chamber to Drywell Vacuum Breakers (BWR) t Recommendation: (1) The monthly surveillance test should be retained. l. (2) The time each vacuum breaker shall be tested following any j i discharge of steam to the suppression chamber should be changed to 12 j hours. ] 3/4.6.4 Vacuum Relief, Suppression Chamber - Drywell Vacuum Breakers, [BWR/5 STS] TS 4.6.4.1: l l i Each suppression chamber - dryv: ell vacuum breaker shall be: i a. Verified closed at least once per 7 days. b. Demonstrated OPERABLE: 1. At least once per 31 days and within 12 hours after any dis-- f l 4

y m, , i e charge of steam to the suppression chamber from the safety-j relief valves, by cycling each vacuum breaker through at least [ one complete cycle of full travel. I l (Replaced "2" with "12" hours. No change to items 2 and 3.) [

i P

8.5 Hydrogen Recombiner (PWR) Recommendation: Change the surveillance test interval for the hydrogen recombiner functional test to once each refueling. interval. a [The test interval is 6 months for some plants.] l I 3/4.6.5 Combustible Gas Control - Electric Hydrogen Recombiners, l t [B&W STS' (Typ)] TS 4.6.5.2: 'l i Each hydrogen recombiner system shall be demonstrated OPERABLE: j a. (No change) b. At least once each refuelino interval by: (Replaced "18 months," which is the current STS requirement for PWRs, with [ .t "each refueling interval." No change to items b.1, b.2, and b.3.) 3/4.6.7 Atmospheric Control - Containment and Drywell Hydrogen Recombiner [ Systems, [BWR/6] TS 4.6.7.1: i r Each containment and drywell hydrogen recombiner system shall be i r -..

i" II i. i ; demonstrated OPERABLE: ) -l a. (No change.) i b. At least once each refuelina interval by:- j (Replaced "per 18 months," which is the current BWR/6 STS' requirement, [ with "each refueling interval." No change to items b.1 through b.4.)' ) 8.6 Sodium Tetraborate Concentration'in Ice Condenser Containment Ice Recommendation: Change the analysis interval to once each refueling j interval. 3/4.6.7 Ice Condenser - Ice Bed, [W STS] TS 4.6.7.1 ,J The ice condenser shall be determined OPERABLE: i I 'I a. (No change.) 't b. Once each refuelina interval by chemical analyses which verify that i at least nine representative samples of stored ice have a. boron con-j I centration of-at least 1800 ppm as sodium tetraborate and a pH of 9.0. to 9.5 at 20 degrees-C. ) l f (Combined item b and b.1, with the surveillance interval being "Once each '}1 ~ l I 1..

~ - _. I \\ 5.i' l ; i i \\ ' { l refueling interval" rather than "At least once per 9 months.") c1 At least once per 9 months by: (Renumbered item b as item c.) l ~ (No change to items c.1 and c.2. Renumbered items'b.2 and b.3 as items I c.1 and c.2.) I ~! d (No change to this item. Renumbered item c as item d) j 2 i 9.1 Auxiliary Feedwater Pumn and System Testino-(PWR) l -I Recommendation: Change frequency of testing AFW pumps to quarterly on a staggered test basis. 3/4.7 Plant Systems - Auxiliary Feedwater, [CE STS (Typ)] TS 4.7.1.2: i l Each auxiliary feedwater pump shall be demonstrated OPERABLE: t I I a. At least once per 31 days by: j l 1. Verifying that each valve (manual, power-operated or automatic) in the flow path that is not locked, sealed, or otherwise .l secured in position, is in its correct position. i i l { Renumbered items a.1 and a.2 as items b.1 and b.2 below, and renumbered i item a.3 as a.l ) l .?

. b. At least once per 92 days on a STAGGERED TEST BASIS by: 1. Verifying that each motor-driven pump develops a discharge pressure of greater than or equal to psig at a flow of greater than or equal to gpm. 2. Verifying that the turbine-driven pump develops a discharge pressure of greater than or equal to psig at a flow of greater than or equal to gpm when th'e secondary steam supply pressure is greater than psig. The provisions of Specification 4.0.4 are not applicable for entry into MODE 3. (Added item b. Renumbered items a.1 and a.2 as-items b.1 and b.2.) 9.2 Main Steam line Isolation Valve (MSIV) Surveillance Testina l A TS change was not recommended for this item. l I 9.3 Control Room Emeraency Ventilation System (pWR. BWR) l Findings: (1) The surveillance requirements for the control room l f emergency ventilation system contain a requirement ~that the control l room temperature be verified every 12 hours to assure that it is less i than a temperature typically in excess of 100 degrees-F. (2) This j temperature limit is to ensure equipment operability and human f habitability. It does not appear to be effective for either purpose. 1

S*: (.. - Recommendation: Replace this requirement with a more useful surveillance or delete it if a more effective limit cannot be - established. Because the burden for verifying that the control room temperature is within its limit is not believed to be significant, no change to existing TS are pro-posed in response to this recommendation. However, changes to temperature limits may be proposed on a plant-specific basis to reflect the initial-temper-i ? ature useo cn calculate the control room peak temperature.during a station-j black out event. 10.1 Emercency Diesel Generator Surveillance Recuirements (PWR. BWR) Recommendation: (1) When a EDG itself is inoperable (not including a. l support system or independently testable component), the other EDG(s) i should be tested only once (not every 8 hours) and within 8 hours-unless the absence of any potential common mode failure can be demon-strated. (2) EDGs should be loaded in accordance with the vendor i recommendations for all test purposes other than the refueling outage LOOP tests. (3) The hot-start test following the 24-hour EDG' test should be a -simple EDG start test. If the hot-start-test is not j performed within the required 5 minutes following the 24-hour EDG test, it should not be necessary to repeat the 24-hour EDG test. The only requirement should be that the hot-start test is performed f within 5 minutes of operating the diesel generator at its continuous i i .i

.. rating for 2 hours or until operating temperatures have stabilized. (4) Delete the requirement for alternate testing that requires testing of EDG and other unrel;ted systems not associated with an inoperable train or subsystem (other than an inoperable EDG). 3/4.8.1 A.C. Sources - Operatir.g, [ Typical STS Requirements, non-vendor specific] TS 3.8.1.1, ACTIONS: a. With an offsite circuit of the above required A.C. electrical power sources inoperable.... (Delete the following requirement to test EDGs: "If either diesel generator has not been successfully tested within the past 24 hours, demonstrate its OPERABILITY by performing Surveillance Requirements 4.8.1.1.2.a.5 and 4.8.1.1.2.a.6 for each such diesel generator, separately, within 24 hours.") b. .. If the diesel generator became inoperable due to any cause other than an inoperable support system or preplanned preventive maintenance or testing, demonstrate the OPERABILITY of the remaining OPERABLE diesel generator by performing Surveillance Requirements 4.8 .l.2.a.5 and 4.8.1.1.2.a.6 within 8 hours, unless testino of an independently testable component has demonstrated the absence of any potential common mode failure for the remainina diesel cenerator. (Added the noted conditions under which testing of an EDG is not required

.' and replaced "24 hours" with "8 hours." Remove any other requirement to perform the specified surveillances every 8 hours thereafter or to perform testing of alternate trains of other systems.) d. With two of the above required offsite A.C. circuits inoperable, restore.... (Deleted the following requirement to test EDGs: " demonstrate the OPERABILITY of two diesel generators separately by performing the requirements of Specifications 4.8.1.1.2.a.5 and 4.8.1.1.2.a.6 within 1 hour and at least once per 8 hours thereafter, unless the diesel generators are already operating;") TS 4.8.1.1.2: a. In accordance with the frequency specified in Table 4.8-1 on a STAGGERED TEST BASIS by: 6) Verifying the generator is synchronized, loaded to greater than or equal to [ continuous rating] kW in accordance with the manufacturer's recommendations, and operates with a load greater than or equal to [ continuous rating] for at least 60 minutes, and (Replaced "less than or equal to [60] seconds" with "accordance with the manufacturer's recommendations.")

A ! e. At least once per 18 months, during shutdown, by: 7) Verifying the diesel generator operates for at least 24 hours. ... Within 5 minutes after completing this 24-hour test, perform Specification 4.8.1.1.2.a.5);*.... (Replaced TS "4.8.1.1.2.e.6).b)" [ simulated loss-of-offsite power start and load test] with "4.8.1.1.2.a.5)" [EDG start test).)

• If Specification 4.8.1.1.2.a.5) is not satisfactorily completed, it is not necessary to repeat the preceding 24-hour test.

Instead, the diesel generator may be operated at [ continuous rating] kW for 2 hours or until operating temperature has stabilized. (Replaced the reference to TS "4.8.1.1.1.e.6).b)" with "4.8.1.1.2.a.5)" and replaced "I hour" with "2 hours." This footnote may be added if it does not exist in plant TS.) TS (Plant-specific): Where plant TS require the testing of the one train (EDG, system, or sub-system) when an alternate train, system, or subsystem (other than an EDG) is inoperable, such requirements may be removed from plant TS. 10.2 Battery Surveillance Reouirements (PWR, BWR) I l

4* .* A TS change was not recommended for this item. t 11 REFUELING A TS change was not recommended in this area. 12 SPECIAL TEST EXCEPTIONS Suspendina Shutdown Marain Reauirements (pWR) Recommendation: All PWR licensees may select the Florida Power and l Light Co. (FP&L) proposal to eliminate one rod drop test if they. I satisfy the condition of performing a rod drop test no more than 7 days before reducing shutdown margin. If a rod drop test has been performed within this time, another test is not necessary. 3/4.10 Special Test Exceptions - Shutdown Margin, [FP&L TS (Typ)] TS 4.10.1.2: i Each CEA not fully inserted shall be demonstrated capable of full insertion when tripped from at least the 50% witi: drawn position within 7 days prior to reducing the SHUTDOWN MARGIN to less that the limits of ~ Specification 3.1.1.1. (Replaced "24 hours" with "7 days.") t 13 RADI0 ACTIVE EFFLUENTS

.e s% 4 Waste Gas Storace Tanks (PWR) Recommendation: The surveillance requirement for the limit on the number of curies in the waste gas tank should be changed to: "The quantity of radioactive material contained in each waste gas decay tank shall be determined to be within the limit at least once every 7 days whenever radioactive materials are added to the tank, and at least once every 24 hours during primary coolant system degassing operations." 3.11 Radioactive Effluents - Gas Storage Tanks, [W STS (Typ)] TS 4.11.2.6: The quantity of radioactive material contained in each gas storage tank shall be determined to be within the above limit at least once per 7 days when radioactive materials are added to the tank and at least once per 24 hours durinq orimary coolant system decassino operations. (Replaced "24 hours" with "7 days" and added the new requirement for performing surveillance "at least once per 24 hours during primary coolant system degassing operations.") 14 CONCLUSIONS General Recommendations Items (1) through (3) of the General Recommendations did not include any

a . recommendations for changes to technical specifications. (4) Section 4.0.2 of the Technical Specifications, which allows the extension of a surveillance test interval, should be made applicable to Section 4.0.5 concerning the ASME Code testing in those Technical Specifications which presently do not allow Section 4.0.2 to be applied. 3/4.0 APPLICABILITY [All STS (Typ)] TS 4.0.5 (c): 9 (c) The provisions of Specification 4.0.2 are applicable to the above required frequencies for performing inservice inspection and testing 4 activities. For plants with custom TS, the reference to TS 4.0.2 should be replaced with the applicable TS section that allows surveillance intervals to be extended by 25 percent of the specified interval. In addition, the term "above" may be deleted from the reference to the " required frequencies for performing inser-vice inspection and testing activities." Finally, if plant TS do not include a general specification (TS 4.0.5) on inservice inspection and testing, a new numbered general specification requirement should be proposed based on the STS model specification (TS 4.0.5), or the following statement should be proposed for addition to the specification that allows surveillance intervals to be extended by 25 percent of the specified interval: i 1 i This provision is applicable to the required frequencies for performing inservice inspection and testing of ASME Code Class 1, 2, and 3 a

l L s components, pumps, and valves in accordance with Section XI of the AS Boiler and Pressure Vessel Code and applicable addenda as required by 10 CFR Part 50, Section 50.55a(g). - END - .}}