IR 05000226/2003025

Insp Rept 50-259/85-15,50-260/85-15 & 50-296/85-15 on 850226-0325.Violations Noted:Failure to Have All Three Trains of Standby Gas Treatment Operable & to Adhere to Tech Spec 6.3.A.6

ML20127K754
Person / Time
Site:	Browns Ferry, 05000226
Issue date:	04/23/1985
From:	Cantrell F, Lenahan J, Patterson C, Paulk G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20127K710	List: IR 05000226/2003025 ML20127K705 ML20127K745
References
50-259-85-15, 50-260-85-15, 50-296-85-15, NUDOCS 8505220179
Download: ML20127K754 (12)

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. ce Moo U21TED STATES o NUCLEAR REGULATORY COMMISSION 5

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101 MARIETTA STREET, ATLANTA, GEORGI A 30323

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Report Nos.: 50-259/85-15, 50-260/85-15, and 50-296/85-15 ~

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Licensee: Tennessee Valley Authority 2a 500A Chestnut Street y Chattanooga, TN 37401 g Docket Nos.: 50-259, 50-260 and 50-296

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i License Nos.: DPR-33, DPR-52, i and DPR-68 *

Facility Name: Browns Ferry 1, 2, and 3

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Inspection Conducted: February 26 - March 25, 1985 ]

Inspectors: ,

G. L. 'Paulk b

U Y STr ! I4

'Date ' Signed _

LJ. F. /LL/ An ~

C. A. Pitter' son i V vla/re Date' Signed MD.LL/L J.J. ~Lenaht!uP(piragraph Obnly)

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'Datd Signed

r Accompanying Personnel: C. R. Brooks -

Approved by: aiz F. S. Cantrell, Sectibry'pM4f t//O/r5 Date S'igned

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Division of Reactor Profd(ts a-SUMMARY f

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Scope: This routine, unannounced inspection involved 180 resident inspector-  ;

hours in the areas of operational safety, maintenance observation, reportable "

occurrences, unit three water level testing, and IEB 81-0 _ Results: Two violations were identified: y i

(1) Violation of Technical Specification (TS) 6.3.A.6 for adherence to and I adequacy of Surveillance Instruction (SI) 2, Instrument Checks and _

Observation; and adherence to SI 4.2.B-4, Drywell High Pressur ;

(2) Violation of TS 3.7.B.1 for failure to have all three trains of Standby  ;

Gas Treatment operable, j s

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8505220179 850424 i PDR ADOCK 05000259 3 G PDR

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REPORT DETAILS Persons Contacted Licensee Employees J. A. Coffey, Site Director G. T. Jones, Plant Manager

'J. E. Swindell, Superintendent - Operations / Engineering

~J. R. Pittman, Superintendent - Maintenance J. H. Rinne, Modifications Manager J. D. Carlson, Quality Engineering Supervisor D. C. Mims, Engineering Group Supervisor Ray Hunkapillar, Operations Group Supervisor C. G. Wages,-Mechanical Maintenance Supervisor T. D. Cosby, Electrical Maintenance Supervisor R. E. Burns, Instrument Maintenance Supervisor A. W. Sorrel, Health Physics Supervisor R. E. Jackson, Chief Public Safety T.' L. Chinn, Technical Services Manager

.T. F. Ziegler, Site Services Manager J. R. Clark, Chemical Unit Supervisor B. C. Morris, plant Compliance Supervisor A. L. Burnette, Assistant Operations Group Supervisor R. R.'Smallwood, Assistant Operations Group Supervisor T. W. Jordan, Assistant Operations Group Supervisor S. R. Meahr, Planning / Scheduling Supervisor

~ R. Hall, Design Services Manager W. C. Thomison, Engineering Section Supervisor A. L. Clement, Radwaste Group Controller Other licensee employees contacted included licensed reactor operators senior reactor operators, auxiliary operators, craftsmen, technicians, public safety officers, Quality Assurance, Quality Control and engineering personne ' Exit Interview (30703)

The inspection scope findings were summarized on March 22, 1985, with_the Plant Manager and/or Assistant Plant Managers and other members of his staff. The licensee acknowledged the findings and took no exceptions. The licensee did not identify as proprietary any of the materials provided to or reviewed by.the inspectors during this inspectio . Licensee Action on Previous Enforcement Matters (92702)

This-subject was not addressed in the inspectio . . - .- - -- - . . - -. . -. - - . - - _- - . -

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[ Unresolved Items * (92701) [

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' NOT: Curves,' Technical Specification-(TS) Figure 3.6-1, Out of Dat .

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During a routine - review of Unit 1 ' Technical . Specifications, the inspector noted_that the -figure- which specifies reactor ' vessel

,; . temperature -limitations for vessel . temperature limitations fo'r critical

operations - was outdated. TS 3.6. A.2 i requires that during al operations with a critical core, except when the vessel is vented, the -

reactor vessel shell and fluid temperatures shall- be at or above the temperature of~ curve #3 of Figure 3.6-1. Figure:3.6-1 contains a note

" stating.that the curves are allowed to be used thru 4.0 EFPY (effective

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, full power years). Since the exposure:on the Unit 1 reactor vessel'has L' ' exceeded 5.1 EFPY, the inspector questioned the _ applicability of Figure 3.6-1.to' the existing condition A licensee representative informed the inspector that the limits in Figure 3.~6-1 were conser-

_vative based upon:TVA's analysis for exposure through 4.0 EFPY-and were

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being ? administrative 1y imposed : until the Technical Specification

. amendment ~ request (NO.191), which contains this analysis, is approved.

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The' inspector further noted that technical specifications did not;

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contain a corresponding surveillance for this - LCO (T.S. 3.6.A.2).

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A licensee representative stated that the minimum temperature requirement

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Lis _ checked during the performance of GOI-100-l '" Integrated Plant Operations". This procedure was reviewed and found to contain a check.

, that primary water temperature is greater than 180*F. in -Step 6.b of

.Section II, '" Preparations of Startup'_'; however, no. step could be found '

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that would verify that reactor vessel -shell temperatures were at: or

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above the temperature. of curve #3'of Figure 3.6-1 as required by TS 3.6.A.2 i

A discrepancy also existed between the Units 1 and 2 TS_and the Unit 3 TS. . Units 'I and 2 TS 3.6.A.1 contain an exception on the minimum

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(- temperature for critical operations provided that the reactor vessel is-i vente Unit 3 ' TS 3.6. A.1 does not contain the . exception. .The L

licensee was. shown this inconsistency and. is reviewing this require-

ment. These three items (out of date curves, lack of surveillance, and l- discrepancy between units) 'are unresolved pending further investigation

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p Reversed motor. pinion gear in limitorque operator I

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_ Local. Leak Rate Testing (LLRT)

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. Unit--I was: shutdown on March 19 as required by TS 4.7.A.2.h after the .

H -- licensee identified that the High Pressure Coolant Injection (HPCI)

turbine exhaust valve (73-23) had failed a LLRT. These valves have a C

• An Unresolved Item is a matter. about which more information is required to

[ _ determine whether it is acceptable or may involve a violation or deviation.

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double gasket that' allows . testing of the bonnet seal. . ^ However, these

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valves were never previously given a LLRT but were only part of an integrated-leak rate tes ~

-Unit three RCIC ' turbine exhaust valve (71-15) failed its LLRT on

March-19,-1985. 'Also, the RCIC vacuum pump discharge valve (71-32).and

.HPCI turbine: exhaust drain: valve (73-24) for all three units cannot be

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-local leak rate tested since the valves; do not have a double gaske The licensee plans to modify these valves and perform the required

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-testing prior to restarting of .the applicable unit. . This will remain -

an unresolved item pending a review of the test requirements, past history, and valve modifications.(259/260/296/85-15-02). .Unmonitored Stack Release *

On ' March L18', 1985, the licensee discovered that the stack radiation ,

monitor had been left in _the purge cycle after replacement of a filte The stack release was. unmonitored for two hours; no change in release rate was noted. The licensee's evaluation of the release and associated calculations. will be reviewed by the inspector when available. This item will remain unresolved pending a review of the licensee evaluatio '(259/260/296/85-15-03). Operational Safety (71707, 71710)'

-The --inspectors were kept informed on a daily' basis of _the overall plant status and :any significant safety matters related to plant operation ' Daily discussions were -held each _ morning with plant management and various members of the' plant operating staf .

The inspectors made frequent visits to the control rooms such that each wa visited at least daily when an inspector was.on site. Observations included instrument readings,; setpoints -and recordings; status of operating - systems;

. status and alignments 1 of - emergency standby ' systems; onsite :and- offsite

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emergency _ power sources available for automatic operation; purpose - of temporary tags on equipment controls and switches; annunciator alarm status;

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adherence to procedures; adherence to limiting conditions for coerations; nuclear ~ instruments. operable; temporary - alterations in effect; daily journals and logs; stack. monitor recorder traces; and control rocm mannin This inspection activity also included. numerous informal discussions with

' operators and their supervisor General plant tours were' conducted on at least a weekly basis. Fortions of'

. the turbine building, each reactor building and outside areas were visite Observations included valve positions and system alignment; snubber and hanger . conditions; containment isolation alignments; instrument readings;.

housekeeping; proper ' power ' supply and breaker alignments; radiation area controls;= tag controls on equipment; work activities in progress; radiation

, protection' controls; adequate vital area controls; personnel search and -

escort; and vehicle search and escort. Informal discussions were held with oe

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selected plant personnel in their functional areas during these tour Weekly verifications of system status which included major flow path valve alignment, instrument alignment, and switch position alignments were

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performed on the Standby Gas Trettment (SBGT) system A complete walkdown of the accessible portions of the SBGT, Off-Gas, and Air Dilution-systems was conducted to verify-system operability. Typical of the items checked during the walkdown were: lineup procedures match plant

~ drawings - and the as-built configuration; hangers and supports operable; housekeeping adequate; electrical- panel interior conditions; calibratio date appropriate; system instrumentation on-line; valve position alignment correct; -valves locked as appropriate and system indicators functioning properl SBGT 'C' Train Inoperable During a routine tour of the normally locked SBGT room on March 8, 1985, at 1407, the 480 volt circuit breaker (2A) for the humidity control heater for the 'C' train of SBGT was found in the tripped condition.- The heater maintains the relative humidity less than 70*4 to the activated charcoal bed in the filter train of SBG The. Shift Engineer and Plant Manager were notified of this finding at 1425 on March 8, 198 At the back panel of Unit 2 control room, the inspector noted that bo_th

- the "0FF" (green) and "0N" (red) indicating lights on the relative humidity heater control handswitch (HS-65-60) were not illuminated and a maintenance request sticker, MR A-312188, dated October 16, 1984, was still in' place. A review of MR A-312188 revealed that the request was opened and closed on October 16,- 1984; however, the sticker had not been remove The previous trouble was also due to the breaker tripping but no cause could be. identified. The breaker was replaced and the 'C' train returned to service on March 9, 198 Technical Specification 3.7.B.1 requires that all three trains of SBGT be operable. The Plant -Manager was informed in an exit meeting on March 22, 1985, of this violatio (259/260/296/85-15-04).

l Additionally, a complete eeluation of the circuit breaker has not been completed since the licensee does not have appropriate breaker curves from the manufacturer. This will remain an Inspector Followup Item pending-the results of the breaker evaluation (259/260/296/85-15-05). High Airborne Activity in Reactor Building On the weekend of March 2 and 3, 1985, high airborne activity was discovered in the reactor building, elevation 62 The source was i

isolated to the Unit 1 Reactor Water Cleanup (RWCU) system precoat

tank. . Water vapor was seen rising from the tank from a standpipe inside the tank. The exact source of the airborne activity has not

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been determined by the licensee.

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The inspector observed in the Unit 1 control room on March 5,1985, that the drywell equipment drain sump pump was pumping 200 F. water to the clean radwaste system. Normally the water is less than 160*F. to prevent possible airborne problems but a failure of the recirculation valve prevented recirculation of the water by the associated cooler to reduce the temperature below 160 F. A maintenance request sticker dated February 16, 1985, (MR 1800040) was beside the recirculation valve. The inspectors discussed the concern and possible connection with the Health Physics Supervisor and Plant Management. No source for the airborne activity had been identified by the licensee when the unit was shutdown on March 19, 1985. An investigation is planned by the licensee after. the unit returns to service. This will remain

.an open item pending the results of the licensee's investigatio (259/260/296/85-15-06). Water Level Problems Unit three was shutdown on March 9, 1985, to investigate problems previously identified with the reactor vessel water level instrument This was the subject of a special inspection conducted by the regional office staff and resulted in an Enforcement Conference on March 14, 1985. -(Reference IE Report 85-19).

The_ licensee discovered a leak in the reference leg connection to water level instruments LT 3-206 and LT 3-53. The leak was reported as one drop every two minutes at zero psig and was located inside the reactor water cleanup heat exchanger room at penetration X-28A. The licensee is not certain whether the leak had started during or prior to pipin inspection. The faulty section of one-inch piping has been replace Initially the crack was thought to be fatigue failure. Presently, the crack is attributed .to trans granular stress corrosion cracking resulting from chloride contaminatio The source of the chloride contamination has not been determined. This will remain an Inspector Followup Item pending the complete evaluation and subsequent review by the resident inspector and Region II (259/260/296/85-15-07).

6. Maintenance Observation (62703)

Plant maintenance activities of selected safety-related systems and components were observed /reveiwed to ascertain that they were conducted in accordance with requirements. The following items were considered during this review: the limiting conditions for operations were met; activities were accomplished using approved procedures; functional testing and/or calibrations were - performed prior to returning components or systems -to service; quality control records were maintained; activities were accom-plished by qualified personnel; parts and material used were properly certified; proper tagout clearance procedures were adhered to; Technical Specification adherence; and radiological controls were implemented as require ._, _ _ _

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Maintenance requests were reviewed to determine status of outstanding jobs and to ; assure that priority was assigned to safety-related equipment maintenance which might affect plant safety. The inspectors observed the below listed maintenance activities during this report period: ' Reversed pinion gear in limitorque operator On March 13, 1985, during the performance of a Surveillance Instruc-

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tion,. the ~ Unit 3 Residual Heat Removal (RHR) pump 'C' suction valve

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(FCV 3-74-12 Limitorque Model SMB-1) failed to close. Initially, the failure was attributed to a loose. set screw on the motor pinion gea Later, on. March 14, 1985, the valve failed again and the failure was

. attributed to the motor pinion gear -being installed in the reverse direction. The lockwire holding the pinion gear setscrew in place was found broken into several pieces which allowed the pinion gear, which slid on the shaft, to become unmeshed with the driven gea The licensee found that a personnel error was made during the initial troubleshooting of the valv Plant mechanical maintenance procedure MMI-87, Preventative and Corrective Maintenance of Limitorque Operator, had been previously revised to include a check for the proper pinion gear orientation anytime maintenance was performed on the valve. The check included double verification of the proper orientatio The check was. performed on March 14, 1985, and the incorrect orientation was not identified. When the pinion gear was incorrectly installed has not been determine Due to this error and previous errors (IE Report 84-52), a valve inspection program was initiate This will remain an open . item pending completion of the inspection program. (259/260/296/85-15-08). HPCI Minimum Flow Orifice

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During a routine tour of the Unit 3 Reactor Building on March 19, 1985, l the inspector noted loose fasteners on the High Pressure Coolant

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Injection System (HPCI) pump minimum flow orifice flange. One stud bolt was visibly loose in that it was about 1/8-inch away from the flange bearing surface. Also, several stud bolts had insufficient thread engagement (only 3 or 4 threads appeared to be engaged by the bolts). The inspector informed licensee management on March 19, 198 On March 20, 1985, a license representative stated that a maintenance i history review was being performed to identify when the minimum flow orifice was last worked and that minimum ' flow orifices on other units

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and ECCS systems would be inspecte This item will be followed up l when the licensee's investigation is complete (296/85-15-09).

t-L Mechanical Maintenance Instruction, MMI-87 Preventive and Corrective Maintenance of Limitorque Operator Special Test 8503. Investigation of Unit 3 Water Level Instrumentation l Problems.

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, RHR Pump Motor Changeou : Special training of plant personnel after water level instrumentation

. problem Low Pressure Coolant Injection'(LPCI) Motor-Generator Changeou No violations or deviations were identified in this are . Surveillance Testing Observation (61726)

The inspestors observed and/or reviewed the- bel'ow listed surveillance procedures. The inspection consisted of _ a review of .the . procedures. for technical adequacy, conformance_to technical specifications, verification of

. test instrument calibration, observation of the test, removal from service and return to service of the' system, a review of test data, limiting condition 'for operation met, . testing accomplished by_ qualified personnel, and that the' surveillance was completed at the required frequenc ' Operating Instruction 0I-65 - Standby Gas Treatment System .SI 4.7.8-1 - SBGT System Operability Test SI-4.7.B-10 - SBGT System Train Operation with Heaters on l' SI 2 - Instrument Checks and Observations During a routine tour of the' Units 1 and. 2 control rooms on March 25,

.1985, the inspector reviewed ~ reactor water level data . which - was recorded during the daily instrument checks required by Technical Specification 4.2. A, 4.2.B,- and 4.2.F. This requirement is imple-mented by Surveillance Instruction 2 (SI-2), Instrumentation Checks and'

f 0bservation Section 2.1' of SI-2 states that certain water level

instruments (which are listed in discrete groups in the instruction)

should provide appror.imately the same level indication and that, as a

part lof the instrument check, readings from instruments within each group' are compared. The inspector learned from licensee' . represen-l- .tatives that - no comparison between

the instruments is routinely performed and - that no criteria is - available to determine what ,

constitutes a satisfactory compariso The SI-2 data sheet only requires that water level be greater that 10 inches (the' Technical F Specification required scram value) . The inspector compared the water ll _ level indications as specified in SI-2 and found the_ instructions to be.

t' inappropriate since a comparison of LI-3-62 and LI-3-185, as required-l by the-instruction, yielded a difference of about'170 inches. LI-3-62-is the shroud level range indicator and normally reads greater than 200 inches at full power (due to jet- pump , flow effects at the sensing

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L lines) whereas - LI-3-185 is a narrow range GEMAC instrument which The instructions

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normally reads about 33 ' inches at full power.

,. contained in SI-2 also did not fully implement the Technical Specifi-L cation instrument check requirement regarding instrument comparisons

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'with independent-instruments. . Tech'nical Specifications 1.V.4 and SI-2

Section 2.0-both require instrument' checks to included a comparisor of the instrument' with other independent instruments.which measures the same va-iable, where possible. . Although other, independent instruments are,av6tlable, .Section 2.1 of SI-2 erroneously requires comparison of reactor water level instruments which are not independent in that they .

have common sensing . lines. Failure to provide 'an adequate procedure -

and to adhere to the procedure for the performa'nce of reactor water

[l level instrument checks is a violation of Technical Specification 1 6.3.A.6. This- violation was discussed in an exit meeting with .the l

. Plant Manager on March 22, 198 (259/260/296/85-15-10). ' SI 4.2.B-4-Drywell-High Pressure  !

On March 14,1985, . the inspector observed the performance of Surveil-

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lance Instruction.4.2.B-4, Instrumentation that Initiate or Control the iCSCS - Drywell High Pressure (PS-64-58E-H). SI 4.2.B-4 satisfies the monthly channel -functional. test and the quarterly sensor calibration on pressure switches PS-64-58E through PS-64-58H. These pressure switches are utilized in annunciation and containment ' spray logi The inspector noted : that instrument valve manipulations and calibrator-connections.were not in strict compliance with the. procedure. Step 4.2-of SI 4.2.8-4 specifies that the instrument isolation valve is to be closed and the pneumatic calibrator is to be connected to the test te . The technicians had closed the isolation valve, disconnected a fitting

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on the downstream side of. the -drain valve and connected the pneumati . calibrator to that fitting and then opened the drain valv This variance with the. procedure had no' effect on the results of the surveillance. The inspector also noted a weakness in ' Step 4.1 of SI 4.2.B-4. .In . the step, voltmeters are connected to terminals. in order to monitor the pressure switch contacts; contacts are checked open (250 VDC).. The inspector observed that one switch was found with about-150 t VDC and the other with about 290 VDC present. The instrument mechanics stated that this was expected since one switch was.used in a 125 VDC -

annunciator circuit and the other was used in the 250 VDC logic circuit. During the performance of step 4.5 of the surveillance, the instrument mechanics failed to decrease pressure below 1.2 psi as specified; instead, the pressure was' decreased until the relay droppe out which was about 1.7 ps These two items (calibrator hookup and pressure decrease) are examples of a failure to adhere to procedure ~as required by Technical Speci-fication 6.3.A.6. This violation was discussed in an exit meeting with the Plant Manager on March 22, 1985 (259/260/296/85-15-10). Reportable Occurrences (90712, 92700)

The below ' listed Licensee Events Reports (LERs) were reviewed to determine

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. included: adequacy' of event description, verification of compliance with

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technical specifications and regulatory requirements, corrective action taken, existence .of potential- generic problems, reporting requirements satisfied, and the relative safety significance of each even LER No._ 296/85-06, ' dated March 15, 1985, Mismatch of Reactor Water

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Level . Indicators, failed to include the following information required by 10 CFR 50.73(b):

(1) The method of discovery of the component or system failure was not discussed. An engineering review conducted several days after the event actually discovered the failur (2) Operator actions which affected the course of the event 'were not discussed. After a mismatch between water level indicators-developed, operators discounted level instruments "B" and controlled water level based upon the "A" and "C" instrument (3) Personnel errors were not discussed and the fact that a cognitive error (e.g., failure to recognize the actual plant condition) had occurred was not included. Operators had controlled reactor water level- based upon erroneous indication and as a result, actual water level drifted to a limiting safety system setpoint without the operator's knowledge. In addition, operators failed to recognize that Technical Specification required instruments were inoperabl (4) The discussion of secondary functions which were lost as a result

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of the failure was inadequate. The report did not indicate that the minimum number of operable- Rector Protection System (Scram)

Instruments was no longer available as a result of the event and that operators failed to take the action required by Technical Specification The effect on level switches 'used in High Pressure ~ Coolant Injection (HPCI) and Reactor Core Isolation Cooling (RCIC) systems control circuitry was also not include (5) The discussi.on on the effects of the failure mode was inadequate in that the' LER merely indicated that the loss of water in the reference leg caused level indicators to indicate erroneousl The loss of water in the reference leg actually causes an indication error to occur in the non-conservative directio (6) .The LER did not reference a previous similar event at the same plant which occurred on November 19, 1984.

t LER No. 296/85-04, dated February 21, 1985, Late Performance of Surveillance Test, was found deficient in the following areas:

(1). The method of discovery of the error was not discusse (2) Dates and times for then the surveillance was due, completed, and last performed were not include E

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'(3) No assessment of the safety consequences was provide (4) The discussion :of- the personnel error did ~not state whether th error was contrary to approved procedures, resulted from.an error-in. approved procedures or was not covered by approved procedure (5) The. discussion of the corrective ' action. planned to prevent recurrence was' inadequate in that it.merely stated that a minor procedural revision would. be - initiated to . preclude recurrenc ' The procedure wasJ not1 identified nor was the apparent Linadequacy :

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- in the procedure discusse , (6): Previous similar events; known to the licensee were not liste The following -LERs contain problems with surveillance not being completed within the required frequency:

BFR0 259/84-17,.259/83-55,-259/83-65, 259/83-27, 259/82-97, 260/83-79 LLER No. '259/85-01, dated February 15, 1985, Unidentified Leakage in-

~Drywell, was found deficient in that:

'(1) = A'n . assessment of the safety consequences resulting from the'

installation of a temporary hose which effectively rendered- a

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containment isolation valve inoperable was not include (2)- All personnel errors _were not discussed and , attributed ' to procedural _ or ' cognitive deficiencies. Many p'ersonnel errors contributed to the- event : including a failure of the operator to -

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recognize the actual plant condition (e.g., .four test valves out of position with a temporary jumper hose installed around. an i isolation valve) prior to .startup as well as inadequate shift L turnovers between maintenance supervisors.

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l' These problems were' discussed with the licensee 'during an ' exit meeting on i

March 22, 1985, and will be tracked as an Inspector Followup-Item (259/260/-

t =296/85-15-11).- A licensee representative stated that a checklist-would be L - . developed to ensure all reporting requirements of 10 CFR 50.73(b) would be

? included in future LERs.

- (C1csed) IEB 81-01, Surveillance of Mechanical Snubbers

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.-The licensee submitted. its initial response to IEB 81-01 for Browns Ferry

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Units -1, 2 and 3 to NRC, -Region II in a letter dated March 13, 1981. The

' licensee reported 14 INC ' snubbers were installed inside the drywell con-tainments-in each unit. A summary of the .INC snubber inspection program

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,' -which was performed during a refueling outage to verify snubber operability

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.was provided. . The letter also summarized the program for inspection of i

mechanical- snubbers _ manufactured by other vendors, such as Pacific Scien-tific or PSA, and provided results of previous inspection A schedule for additional snubber inspections to be performed during ~ refueling outages

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was given to comply with the Bulletin's requirements. Implementation of the c . proposed inspection program discussed in the licensee's March 13, 1981, letter was' concurred with by letter.from NRC, Region II dated June 12, 198 The licensee submitted the results of the IEB 81-01 inspection performed on mechanical snubbers .to NRC ' Region II by letters dated July 23, 1981 and February 10, 1982. The licensee's final . response pertaining to IEB 81-01 was -submitted to NRC Region II by letters dated February 16, 1983, which summarized. the inspection results performed on . mechanical snubber Subsequent to issuance of . Bulletin 81-01, the licensee replaced the -INC

snubbers with PSA mechanical snubber In addition, the Units 1,. 2 and 3 Technical Specifications were amended to require routine surveillance and testing'of mechanical snubbers. Prior to issuance of IEB 81-01, Technical Specifications requirements pertaining to inspection and testing of-mechanical snubbers, did not exist; however, the licensee had an in-depth

. inspection program in effect ' prior to this tim NRC Region II has inspected the implementation of the Technical Specification requirements

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during inspections performed December 13 - 16, 1983, Report Number 50-259, 260, 296/83-56 and found. a satisfactory program. Followup of the ' snubber

- program was also covered in Inspection Reports 50-259, 260, 296/82-09 and 83-4 IEB 81-01 is closed.

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