IR 05000456/1990012

Safety Insp Repts 50-456/90-12 & 50-457/90-15 on 900429- 0616.Violations Noted.Major Areas Inspected:Licensee LER Review,Operational Safety Verification,Radiation Protection, Security & ESF Sys

ML20055G423
Person / Time
Site:	Braidwood
Issue date:	07/16/1990
From:	Farber M NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20055G417	List: IR 05000456/1990012 ML20055G416 ML20055G420
References
50-456-90-12, 50-457-90-15, NUDOCS 9007230161
Download: ML20055G423 (18)

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U.S. NUCLEAR REGULATORY C0 m1SS10N

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t Reports No. 50-456/90012(DRP); 50-457/90015(DRP)  :,

s Docket Nos. 50-456; 50-457 Licenses No. NPF-7?; NPF-77 Licensee: Commonwealth Edison Company Post Office Box 767  !

Chicago, IL 60690 Facility Nanie: Braidwood Station, Units 1 and 2 *

Inspection At: Braidwood Site, Braidwood, Illinois *

i inspection Conducted: April 29 through June 16, 1990  ;

Inspectors: T. E. Tayior J. A. Hopkins T. M. Tongue M. A. Kunowski i D. R. Calhoun j

M. J. Farber, Chief 7-/0 46 ApprovedBy@:

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l Reactor Projects Section IA4 Date inspection Suninary l l

Inspection from April 29 through June 16, 1990 (Reports No. 50-456/90012(DRPi;

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50-457/90015(DRP))

Areas Inspected: Routine, unannounced safety inspection by the resident inspectors, one regional specialist, and one project inspector of Licensee Event Report (LER) review; operational safety verification; radiation protection; security; engineered safety feature systems; monthly maintenance

, observation; monthly surveillance observation; training effectiveness; report review; TMI action plan requirement followup; and meetings and other activitie Results: Of the 10 areas inspected, no violations were identified in In the remaining areas three violations were identified. One concerns the failure to follow the Temporary Lift Procedure for systems / components that are out-of-service, Paragraph 3.f. Another concerns the failure to follow the requirements of Radiation Work Permits, Paragraph 4.c. The third is a non-cited violation concerning unexpected reactor trip _ signals generated during :

performance of a Solid State Protection System (SSPSi surveillance, Paragraph 3.d. One unresolved item was identified concerning the surveillance performed on the 2A auxiliary feedwater pump while the 2B diesel generator was out-of-service, Paragraph Obbd $$8hjj56 PDC

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-l W' ..In:the area of plant operatione and safety assessment / quality verification,

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the licensee's performance coc.inues to improve. The licensee's overall' 1

<.' ' performance in radiological controls, maintenace/ surveillance, emergency  !

preparedness, security, and engineering / technical support was steady overall. . tl;

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I DETAILS i Persons Contacted R Commonwealth Edison Company (CECO) l T. J. Maiman, Vice President, OWR Operations

• R. E. Querio, Station Manager  ;

• D E. O'Brien, Technical' Superintendent p' *K. L. Kofron, Production Superintendent l S C, Hunsader, Nuclear Licensing Administrator L G. R. Masters, Assistant Superintendent - Operations ,

G. E. Groth, Braidwood Project Manager, PWR Projects Department L * J. Legner Services Director

• E. Lohman, Assistant Superintendent - Maintenance P. Smith, Operating Engineer - Unit 1 *

R. .Yungk, Operating Engineer - Unit 2

• W. B. McCue, Operating Engineer - Unit 0 .

R. D. Kyrouac, Quality Assurance Supervisor  ;

D. J. Miller, Regulatory Assurance Supervisor

• D. E. Cooper, Technical Staff Supervisor '

A. D' Antonio, Quality Control Supervisor

• A. Checca, Security Administrator

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R. L. Byers, Assistant Superintendent - Work Planning and Startup

• L. W.~ Raney, Nuclear Safety Supervisor C. Vanderheyden, Training Supervisor P. Maher, Assistant Technical Staff Supervisor

• D. F. Ambler, llealth Physics Supervisor

• E. W. Carroll, Regulatory Assurance '

P. Holland,. Regulatory Assurance l J. Smith, Master, Electrical Maintenance

• T. R. Coslet, Operating Steward

• D. J. Skoza, Engineer

• J. D. Wagner, Regulatory Assurance

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• D. Malone, Chief Steward, Clerical

• S. D..Notter, Nuclear Quality Programs Engineer

• Denotes those attending the exit interview conducted on June 19, 1990, 1990, and at other times throughout the inspection perio .

The inspectors also talked with and interviewed several other licensee employees, including members of the technical and engineering staffs, reactor and auxiliary operators, shif t engineers and foremen, and electrical, nechanical and instrument maintenance personnel, and contract security personnel . LER Review (92700) +

.Through direct observations, discussions with licensee personnel, and review of records, the following event reports were reviewed to determine ,

that reportability requirements were fulfilled, that immediate corrective -

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action was accomplished, and that corrective action to prevent recurrence had been or would be accomplished in accordance with Technical Specifications (TS): (Closed) 457/90002-LL: Inadvertent Reactor Coolant System Depressurization and Pressurizer Cooldown Khile in Cold Shutdown (Mode 5) Due to Personnel Error. This event was reviewed and evaluated by a Special Inspection Team. The staff's concerns and findings are documented in NRC Inspection Report 50-457/9001 the licensee's corrective action for the event include:

engineering evaluation of reactor coolant pump seal damage (0 pen item 50-457/90012-01(DRP)),

cooldown of the pressurizer (0 penengineering Item evaluation of the ra50-457/90012-02(DRP)

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various procedure revisions, implementation of a control room awareness program, manual valve position placards, and training tailgate sessions. The inspector has no further concerns with this event. This LER is considered closed, (Closed) 456/90006-LL: D4esel Generator Slow Start From Crisscrossed Starting Airhnes Due to Programmatic Deficiency. On April 16, 1990, the licensee began troubleshooting the 1B Emergency Diesel Generator (DG) af ter it failed to reach rated speed within 15 time re The cause of the slow start was crossed airlines (quirements.6 to 9 and 5 to 8) for the DG starting air syste Maintenance history identified that the crossed line condition occurred during exhaust manifold replacement during the five year maintenance inspection in November 1989. The airlines were restored to their original configuration and the DG was tested and declared operable on April 18, 1990. The licensee's corrective action included checking the airlines on the remaining DGs (completed June 1, 1990) and revising the DG maintenance procedure. The DG maintenance procedure was revised to require a " time check" when DG maintenance activities require disconnection of the starting air d%tribution syste The licensee identified the root cause of this event to be a programmatic deficiency. The " time check" performed was not capable of detecting the crossed airlines. NRC Region Ill' review based on NUREG-1022 guidelines has characterized the cause of this LER as a personnel error. The airline tubing connections are stamped to identify correct alignment, but were not properly reconnecte Region III management has discussed the root cause discrepancy with the licensee concerning this event. This LER is considered closed, in addition to the foregoing, the inspector reviewed the licensee's Deviation Reports (DVRs) generated during the inspection period. This was done in an effort to monitor the conditions related to plant or personnel performance, potential trends, ctc. DVRs were also reviewed for proper initiation and disposition as required by the applicable procedures and the QA manua No violations or deviations were identifie I

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3. Operational Safety Verification (71707)

During the inspection period, the inspectors verified that the facility was being operated in confomance with the license ( and regulatory requirements and that the licensee's management control system was effectively carrying out its responsibilities for safe operation. This was done on a sampling basis through routine direct observation of :

activities and equipment, tours of the facility, interviews and ,

discussions with licensee personnel, independent verification of safety system status and limiting conditions for operation action requirements ;

(LC0ARs), corrective action, and review of facility records, t

On a sampling basis the inspectors daily verified proper control room staffing and access, operator behavior, and coordination of plant c activities with ongoing control room operations; verified operator ;

1 adherence with the latest revisions of procedures for ongoing activities; verified operation as required by TS; including compliance with LC0ARs, (- with emphasis on engineered safety features (ESF) and ESF electrical '

alignment and valve positions; monitored instrumentation recorder trace and duplicate channels for abnormalities; verified status. of various lit '

annunciators for operator understanding, off-normal condition, and corrective actions being taken; examined nuclear instrumentatior (NI)

and other protection channels for proper operability; reviewed radiation monitors and stack monitors for abnormal conditions; verified that onsite and offsite power was available as required; observed the frequency of 4 - '

plant / control room visits by the station manager, superintendents, assistant operations superintendent, and other managers; end observed the Safety Parameter Display System for operabilit During tours of accessible areas of the plant, the inspectors made note ,

of general plant / equipment conditions, including control of activities in progress (maintenance /surv2111ance), observation of shift turnovers, general safety items, etc. The specific areas observed were: Unit 2 Unidentified Leakage Greater than 1 Gallon Per Hinute (gpm) ,

(71707)

On June 8, 1990, at about 2:40 p.m., while in the process of starting up from the 2A low pressure turbine-outage, unidentified leakage for Unit 2 was identified to be 1.096'gpm.' The TS limit is I gpm. The unit was in Mode 3 (hot standby). The licensee's investigation '

located a minor leak on the 2A seal injection filter vent. The Limiting Conditions for Operations (LCO) for TS 4.6.2-la was entered at 2:40- p.m. and exited within the four hour TS time limit when the leak rate dropped below 1 gpm. Unidentified leakage greater than or equal to 1 gpm has a TS LCO which gives four hours to reduce leakage below 1 gpm or within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> be in cold shutdow ,

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The leak rate was .15 gpm when reactor startup was completed on June 9, 1990. The resident inspectors monitored the licensee's activities during this event and have no further concern *

No violations or deviations were identified.

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! Unexpected Rise in Pressurizer Level While Pele. ing Disc Pressurization on Reactor Coolant Loop Isolation Valves - Unit 2 (92701, 71701)

OnMay3,1990,whileperformingtheappropriatestepsinBw0PRC-9,

" Filling an Isolated Reactor Coolant Loop, thepressurizer(PZR)

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level unexpectedly increased approximately 30% (PZR hot calibration i level channel). Unit 2 was in cold shutdown (Mode 5) maintaining a reactor coolant system (RCS) temperature of approximately 90*F during i

! _the event. The operators were performing Bw0P RC-9 in preparation ,

to open the RCS loop isolation valves by isolating RCS loop vent and drain lines, filling and venting the RCS loops and removing the RCS ,

loop isolation valvo disc pressurization system.-  !

The disc pressurization-system supalies water from the safety  ;

. injection occumulator to the area 3etween the RCS loop isolation '

valve discs. This minimizes RCS leakage from the reactor to the loops. Approximately one week prior to the event, the RCS loop vent lines were isolated to support RLS maintenance activitie :

Dise leakoff slowly filled and pressurized the loops. (Evaluation! l

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of the event by licensee system engineers detemined an approximate 20 psig pressure buildup in the loops.) -

Step 24.f of Bw0P RC-9 opened valve 2RC8042, "RCS Loop Equalization j Line Manual Isolation Valve," to depressurize the area between the loc 9 isolation valve discs. Additionally, this created ; 3/4 inch

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flow path from the loop side of the isolation valve to the reactor sid PZR level increased approximately 30% (3700 gals.) over a :

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three hour period. Unit 2 operators monitored the PZR level increase, lowered it when level reached 70% and maintained PZR level at ,

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approximately 40%.

The licensee evaluated the event and determined that reactor safety was not challenged. The resident inspectors reviewed the licensee's evaluation and had no further concerns, i No violations or m cations were_ identifie ;

c. L High Radiatica Sample System Leak - Unit 1 (71707)

On May 13, 1990, the Unit 1.High Radiation Sample System (HRSS)

L developed a leak. The leak was first identified as a " mist" coming' ,

from the An buildin HRSS drain tank equipment room (EA)

attendant on the 383'

and level of radiation the auxiliary (RP)

_ protection technician identified a leaking sample line pipe' union in the HRSS room on the 401' level of the auxiliary building. The EA and RP m technician _ entered the area in appropriate protective' clothing and L isolated the leak. The area was " roped off" and decontamination efforts commenced, i

The leaking union was on the RCS loop sample line downstream of the containment isolation valve. The licensee's investigation of the event determined that the leak was a result of thermal cycling. The

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i union was tightened and all other HRSS unions for both units were ;

checked. Wtile the RCS loop sample line was being repaired, normal -

and post-accident sampling for the Unit 1 RCS loops was unavailabl " Grab samples" of the RCS loops were taken in the interi j The licensee's response to the event was reviewed by the inspectors and regional RP specialists. No additional concerns were identifie !

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No violations or deviations were identifie l d. Unexpected Generations of Reactor Trip Signals _ During Performance of ]

2BwV5 3.1.2-2, " Reactor Trip Breaker and Gripper Coil Response Time Measurement ' (92701, 61726)

On May 17, 1930, during the performance of 2BwVS 3.1.2-2, two unexpected l reactor trip signals were generated (PZR pressure low and reactor coolant pump low flow). The technicians were performin  !

which returned the Train A logic A switch (24 position)gfromstepposition F.28,.

7 to 0FF. The procedure did not specify the direction, so the technician went clockwise from position 8 to 0FF. This action  ;

generated two reactor trip signals. Additionally, the. source range '

(SR) NI hign volt failure annunciator alarmed and immediately rese ;

Unit 2 was in cold shutdown (Mode 5) with the disconnect switches for i

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the rod control lift coils ope The reactor trip breakers were

" racket!in," but ope '

The. technical staff engineer observing the surveillance discussed-the event with the shift control room engineer (SCRE) and the decision was made to-continue on Train B to identify the cause. At Step F.3.8, the technicians turned the Train B logic A switch clockwise from position 8 to 0FF. At position 21,~the two reactor. trip signals-were generated and SR 2N-32 was de-energized. At position 22, the SR NI high volt failure annunciator alarmed. There were-no other unexpected plant responses during the remainder of the surveillance.-

The licensee's evaluation of the event determined that the SSPS i responded as designed. Position 21 on logic A switch simulates a permissive P-10 signal, which initiated generation of the trip j signals ano de-energization of SR NIs. This was the first time this particular methodology was used to perform the surveillanc The licensee's immediate corrective action was to specify a counter clockwise direction when returning the logic A switch to 0FF. A temporary change was initiated to revise the procedure until a permanent change is completed. Failure to incorporate adequate ,

instructions for the performance of this surveillance is considered

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a violation of 10 CFR 50 Appendix B Criterion V (457/90015-01).

The violation is not being cited because the criteria specified in Section V.G.1 of the enforcement policy were satisfied. .This item '

is considered close The licensee reported this event via the ENS notification system as required by 10 CFR 50.71. Subsequent to the notification, the licensee discussed the possibility of retracting the notification

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, 8 with the NRC. The licensee contended that the trip signal generation and SR de-energization were spurious and per NUREG-1022 i not ' reportable. The licensee was informed by NRC management that ,

the signals generated were considered to be unexpected, not spuriou l Additionally, the licensee was informed that in accordance with i NUREG-1022 guidelines and 10 CFR 50.72 requirements, the event was ,

in fact reportable and le initial notification should not be :

retracte One non-cited violation was identifie ; Reactor Trip Due to Lightning Strike (71707, 92701) i On June 8,1990, at 6:10 a.m., Unit 1 experienced a reactor trip i from a " Power Range Negative Flux Rate H1gn Trip." Prior to the event, Unit I was at 99% power. The suspected cause of the event was lightning strikes from a sever thunderstonn in the area. The, lightning caused five of the primary and two of the backup rod '

control system power supplies to de-energize on overvoltage -

protection, which resulted in releasing control rods into the core and the subsequent negative flux rate trip. The power supplies were reset and the licensee checked the control rod system for any ,

damages, none were found. Also, as designed and expected, a main feedwater isolation occurred in response to the reactor trip, and the auxiliary feedwater pumps auto-started from the Lo-lo steam i generator levels caused by the reactor trip, t On June 9,1990, the unit was brought back on line. The licensee '

and NRC Region 111 management are scheduled to have a meeting to i discuss the licensee's program for lightning protection. Braidwood Station has experienced several reactor trips due to lightning strikes. The licensee has installed the same modifications for -

lightning protection at Braidwood as Byron. It appears to be ,

working at Byron. The licensee's engineering group is evaluating ;

other modifications for lightning protection at Braidwood,

No violations or deviations were identified, Failure to Perform Activities in Accordance With Temporary Lift Program (71707, 92701). l l On May 10, 1990, while performing a fill and vent of the Unit 2 l- containment spray (CS) system, a large portion of the auxiliary building 364' elevation curved wall area and the auxiliary building i L 346' level residual heat removal /CS pump rooms were contaminated.

l The cause of the contamination was an inadvertent opening of a flow p path from the reactor water storage tank (RWST) to an open CS vent

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valve, 2CS014 A. The valve :2CS014A) was listed for closure on the temporary lift (TL) sheet Nr out-of-service (90-2-1349). Due to a a

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miscommunication and failure to ensure completion of the TL sheet requirements, the valve was never closed. When the operators opened the flow path from the RWST, wner began to spill into the

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l above mentioned areas. -In accordance with the procedure for TLs .l

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(BwAP 330-1A2), the operators involved should have taken a copy of i the TL sheet with them, initialed the valve closure when completed, i and returned to the contro. ; aom and initialed the original TL shee ,

The operators failed to comp with these procedure steps. This process could have prevented the contamination.' ,

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Another example occurred on June 3,1990, during activities associated  !

with draining the main steam lines. The lines were drained for a  :

plant shutdown from about 30% power for repairs on the 2A low pressure 1 turbine (LPT). When the operators were directed to commence steam ,

line draining, it was noticed that the TL for 2MS021A thru D

, (steamline drains) had not been hung when power was increased above 20% prior to the problem with the 2A LPT. A review of the TL' sheet  :

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-identified the expiration noted was for " June 30. 1990 or at 20% power  !

on Unit 2." The cause of the events appears to be the lack o awareness by personnel of the date or plant condition required'for t

, the 71 expiration. The failures to perform activities i' accordance ,

with TL' procedure (BwAP 330-1A2) are censidered violati .is of '

10 CFR 50 Appendix B Criterion V (456/90012-01(DPP); 457/90013-02(DRP)),

t One violation was identified, Housekeeping and Plant Cleanliness The inspectors monitored the status of housekeeping and plant t cleanliness for fire protection, protection of safety-related y equipment from intrusion of foreign matter and general protectio The inspectors also monitored various records, such as tagouts, 0 jumpers, shif tly logs and surveillances, daily orders, maintenance items, various chemistry and radiological sampling and analysis, third party review results, overtime records, QA and/or QC audit results and postings required per 10 CFR 19.1 '

4. RP Controls The . inspectors verified that workers were following health physics procedures for dosimetry, protective clothing, frisking, posting, etc.,

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and randomly. examined RP instrumentation for use, operability, and calibration. Two areas of concern were identified:

, Personnel Contamination in Turbine Building

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On May 7 1990, a licensee employee was discovered contaminated 4 _ while attempting to exit the. Radwaste Building (RWB). The contamination was limited to the back of the individual's uniform shirt and the seat of.the pants. Surveys of the areas visited by the employee (RWB, associated trailers and Radwaste Control Room)

did not detect any contaminatio !

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Surveys of the employee's office in the Turbine Tower determined the source of contamination to be a small area of the back and seat of his chair. (100 K dpm/100 so, cm. direct frisk and 5 K dpm/100 sq. c smearable.) The employ.e's skin was contaminated and he received 0.76 mrem to the skin of his whole bod Investigation of the chair material indicated that a liquid contaminant had penetrated the cushion. Isotopic analysis of the contaminated material was inconclusive. The analysis indicated a mix of nuclides similar to Unit 2 reactor coolant. However, the sample would have to be over one year ol (This is older than any samples currently stored in the chemistry lab.)

The licensee's investigation concluded that the contamination was deliberate, the exact source of the contamination could not be pinpointed and the person (s) who contaminated the chair could not be identified. The employee stated to the licensee and regional specialists that he did not know who contaminated the chai Regional and onsite inspectors reviewt.d the licensee's investigative effort and concur with their conclusio Eberline RM-14 Frisker Found Secured On June 5, 1990, the inspector accompanied two equipment attendants (B-men) performing a containment integrity surveillance, 2Bw05 6.1.1. A- The B-men and inspector entered a " roped of f" potentially contaminated area in the Unit 2 euxiliary bu,. ding collection sump pum) room to verify a valve position. When leaving the area, the two bmen performed a whole body frisk using an Eberline RM-14 and exited the area. The resident inspector, who was still in the " roped off" area, noticed that-the RM-14 was turned off and informed the B-men. The RM-14 was turned on, the B-men and inspector frisked themselves and a health physics (HP)

technician was notified. Several swipes of the area were taken with no measurable contamination. The licensee investigation of the event revealed that a HP technician had turned off the RM-14 to " help save the battery."

HP management conducted individual counseling with the technicians on the event. Tailgate training sessions were conducted for HP technicians. The licensee stated that operations personnel will complete the training by June 23, 199 The licensee's corrective action are considered to be adequate and there are no additional concern Outage Radiation Protection Controls (IP B3729)

A regional specialist and the resident inspectors reviewed radiation protection controls during the last half of the first refueling outage for Unit 2. Previous reviews earlier in the outage are documented in Inspection Re 50-456/90010(DRP); 50-457/90011(DRP), and 50-456/9000B(portsDRS); 50-457/9000B(DRS).

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Radiation protection controls were generally good during the outage.

L The dose total for the two month outage was low, approximately 137 person-rem (compared to approximately 235 person-rem for the Unit i first refuel outage in fall of 1989). As with the Unit 1 outage, the licensee used hydrogen peroxide to induce a crud burs Dose rates in the Unit 2 containment after clean up of the crud were significantly lower than those found in Unit 1.. For example, steam

. generator tube sheet dose rates for Unit I ranged from 4-8 R/ hour; whereas, Unit 2 ranged from 3-4 R/ hou In addition to lower dose rates in the Unit 2 containment, several improvements that apparently enhanced exposure control were the use

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of a containment coordinator, use of a strippable cnating for reactor cavity decontamination, and redesigned containment access control facilities.. As with the Unit 1 outage, radiation protection planning for nigh dose-total jobs generally was good. In addition, toe radiation protection group made good use of Unit 1 outage

" lessons-leai:ad."

Several problems with radiation protection controls during the outage, however, were identified. As discussed previously (NRC-Inspection Report 50-456/90010(DRp); 50-457/90011(DRP)), a worker (fuel handler)

received 101 mrem in excess of his 100 mrem administrative limit while cutting an incore detector puide tube (IDGT) in the reactor cavity, Unplanned exposure was romlved when the worker inadvertently raised an irradioted section of the :0GT out of the reactor cavity pool as he made the final' cut for the job. A radiation protection technician

(RpT),'who was monitoring the dose rates with a survey meter, instructed the worker to return the-IDGT to the pool af ter the survey meter went off scale. The worker's electronic dosimeter recorded the highest exposure rate as 60.8 R/ hour. The dose for the RpT from the incident was'42 mre The doses to two other fuel handlers, who were assistin with the cutting, were 44 mrem and 25 mrem, respectivel An investigation of the incident by the licensee and a subsequent review by the regional specialist identified an inadequate prejob briefing as a major factor in the exposure, .The fuel handling supervisor reviewed the job with ALARA personnel prior to the job and then briefed the workers. According to licensee representatives, E the necessity of keeping the IDGT in the pool and below the surface of the water, however, was not adequately stressed by the superviso :in the' informal prejob briefing held with the workers. In addition, ,

a-statement on the Radiation Work Permit (RWP) prohibiting the removal of irradiated components from the cavity was ambiguous in that it did not specifically state that components shall not be pulled above the level of the pool water. A contributing factor *

to the exposure was the use of manual bolt cutters as opposed to underwater hydraulic cutters. . The hydraulic cutters normally used for cutting 10GTs had been misplaced prior to the job.. The a licensee's corrective actions for this event appear adequate to prevent reoccurrence and include a requirement to include all workers at future prejob briefings and development of a written procedure for cutting IDGTs that limits cutting to use of the hydraulic cutters. These actions will be reviewed by regional

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r specialists during future inspections. In eddition, circumstances of the event were discussed with other nenbers of the radiation i protection and fuel handling departments to re-emphasize the  !

potential hazards of IDGT '

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Another problem observed by the inspectors involved two instances  !

of a failure to follow RWP protective clothing requirements. On April 17, 1990, the regional: inspector and a licensee representative ,

observed a worker in the Unit 2 containment without the protective gloves required by RWP No. 90-1227A. The worker had removed the gloves to remove and store his eyeglasses in a case. The licensee representative promptly instructed the worker to re-don the glove '

On April 26, 1990, a Region III senior manager.and a licensee representative observed a worker in the Unit 2 containment without the protective ' gloves required by RWP No. 90-0096. The worker.had removed the gloves and was " blue checking" a valve. The licensee representative promptly instructed the worker to re-don the gloves and leave the containmen Neither worker was contaminated. In addition to the prompt, initial corrective action, the licensee also f discussed the incidents with the work groups, emphasizing the need to follow RWP requirements. The licensee's corrective actions for i this problem were adequat The two instances of the failure to wear protective clothing specified-by the RWP are a violation of Technical Specification (TS) 6.8. ,

(Violation No. 50-456/90012-02(DRSS);50-457/90015-03(DRSS). This ~

TS requires that written procedures be implemented for radiation  ;

protection activities in accordance with Section 7.3 of Appendix A, of Regulatory Guide 1.33. Revision 2, February 1978. Section C.3 of-Braidwood Procedure No. DwRp 1140-1, " Radiation Work Permit Program,"  !

states that the requirements of the RWP must be complied with in all respect One violation was identifie . Security- (81064)

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Each week during reutine activities or tours, the inspectors menitored the licensee's security program to ensure that observed actions were being implemented according to the approved security plan. The inspector', noted that persons within the protected area displayed proper photo-identification badges and those individuals requiring escorts were properly escorted. The inspectors also verified that checked vital areas were locked and alarmed. Additionally, the inspectors also verified that  !

observed personnel and packages entering the protected area were searched by appropriate equipment or by han . ESF Systems'(71710)

During the inspection, the inspectors selected accessible portions of several ESF systems to verify their status. Consideration was given to p the plant mode, applicable TS, LC0ARs, and other applicable requirements.

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-Various observations, where applicable, were made of hangers and  !

supports; housekeeping; valve positions and conditions; potential i ignition sources; major component labeling, lubrication, cooling..etc.;

interior conditions of. electrical breakers and control panels; whether '

instrumentation was properly installed and functioning and significant process parameter values were consistent with expected values; whether ,

instrumentation was calibrated; whether necessary support systems were 1'

operational; and whether locally and remotely indicated breaker and valve positions agree '

During the inspection, the following ESF components were walked down:-

Unit 1  ;

1A and IB Auxiliary Feedwater Syste Unit 2

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2A and 2B Auxiliary Feedwater system, t No violations or deviations were identifie . Monthly Maintenance Observation (62703) [

Station maintenance activities affecting the safety-related systems and components listed below were observed / reviewed to ascertain that they were conducted in accordance with approved procedures, regulatory guides {

and industry codes or standards, and in conformance with Technical ,

Specification i The following items were considered during this review: the LCOs were ,

met while components or systems were removed from and restored to service; *

approvals were obtained prior to initiating the work;' activities were accomplished using approved procedures and were inspected as applicable; ,

functional testing and/or calibrations were performed prior to returning ,

components or systems to service; quality-control records were maintained; .

activities were accomplished by qualified personnel; parts and materials 1used were properly certified; radiological controls were implemented; and .

fire prevention controls were implemented. Work requests were reviewed to determine the status of outstanding jobs and to assure that priority is l assigned to safety-related equipment maintenance which may affect system performanc ;The following maintenance activities were observed and reviewed:

Unit 1 1FWO340, Repair Ground. The licensee replaced solenoid due to water inleakage.

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IBw05 1.3.1.2-1, Moveable Control Assemblies Monthly Surveillanc While perfoming the rod exercises surveillance on Unit 1,  !

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Group 1 rods for shutdown bank A and control bank A and C did

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not move. The licensee entered the applicable LCO and began !

troubleshooting the control rod drive system. Troubleshooting *

indicated two failed circuit cards in slave cycler IAC. The ,

cards were replaced and the surveillance was successfully completed and the LCO exited. The licensee plans to search

for trends by entering the infomation in their local data

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base and comparing this failure with Byro .

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NWR A39578, Reset of Voltage Value for the Power above Permissive P-6 Bistable Reset Setpoint in N-36, ,

While performing surveillance Bwls 3.1.1-230, " Analog Channel Operational Test of NI System Intermediate Range N-36," per NWR .

A39578, licensee technicians adjusted the voltage reset setpoint i

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for the Permissive P-6 Bistable (B/S). The voltage was adjusted

due to an error in the initial calculation of the B/S reset l setpoint. The licensee detemined that the adjustment was minor and did not challenge reactor core safety. The -!

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adjustment changed setpoint from IE-11 amps to SE-11 amps, the licensee reviewed all other N1 B/S setpoints and reset values which required a similar calculation and did not find additional errors. All intermediate range Nls on both units 1 have been adjuste :

The resident inspectors have reviewed the licensee's activities

! and have no further concerns, i Unit 2 +

2A NSIV, Accumulator 0-ring Repai l 2A Mb Pump Casing Drain Valve Replacement (2AF019A).  ;

The original NWR was written to repair the leaking aump casing 3/4" drain valve (2AF019A). Visual inspection of tie valve t internals indicated a cracked valve disc. The work request

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(WR) was revised to replace-the valve disc. The WR was subsequently changed to replace the entire valve when-the replacement disc' obtained was not a "like-for-like" replacemen A new valve of the same style /model as the one removed was :

1, ordered from Byron station and installed. Maintenance Work

Planning, after discussions with the NRC resident inspector, i

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added a memo to the work package to explain the reason for replacing the valve vice the disc.

l The inspectors monitored the licensee's work in progress and verified ,

i that it was being performed in accordance with proper procedures, and ~

approved work packages, that 10 CFR 50.59 and other applicable drawing updates were made and/or planned, and that operator training was conducted in a reasonable period of time. Any discrepancies noted are briefly discussed for each work ite No violations or deviations were identifie . _ _ - _

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8. Monthly Surveillance Observation (61726) '

The inspectors observed surveillance testing required by Technical Specifications during the inspection period and verified that testing was performed in accordance with adequate procedures, that test instrumentation was calibrated, that LC0 were met, that removal and restoration of the affected components were accomplished, that results confonned with TS and )rocedure requirements and were reviewed by personnel other than t1e individual directing the test, and that any deficiencies identified during the testing were properly reviewed and resolved by appropriate management personne The inspectors also witnessed portions of the following test activities: ,

Unit 1 IBw0S 8.1.1.2.a-2, "1B Emergency Diesel Generator Monthly Surveillance."

t BwVS 0.5-3. AF.1-2, "ASPE Surveillance for the Diesel Driven AF Pump i and B-Train AF Valves."

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BwlS 3.1.1-230, " Analog Channel Operational Test of Nuclear Instrumentation System Intermediate Range N-36." '

1Bw05 3.1.1-21. "SSPS, Reactor Trip Breaker, and Reactor Trip Bypass .

BreakerBi-Monthly (Staggered) Surveillance i (TrainA)."  ;

The surveillance was successfully performed. A licensed *

operator made a minor error while resetting the Train B safety '

injection relay. The safety injection recirculation sump

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isolation valve reset button was inadvertently pushed instead

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of the safety injection reset button. This error did not affect satisfactory completion of the surveillance. The i inspector has no additional concern ;

Unit 2

BwHS 4002-066, Periodic Protective Relay Calibration of Overcurrent Relays of the 2A Safety injection Pump." )

l BwoS 6.1.1.A-1, " Unit Two Primary Containment Integrity Verification i of Outside Containment Isolation Devices."

2Bw05 8.1.1.2.a-1, " Unit Two 2A Diesel Generator Operability Monthly (Staggered) and Semi-Annual (Staggered)

Surveillance."

2Bwls 3.2.1-204, " Analog Operational Test / Surveillance Calibration !

of Auxiliary Feedwater Pump Suction Loop P-AF051."

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2BwC o.1.2-2, " Reactor Trip Breaker and Gripper Coil Response Time 1 l Measurement."

2Bwls 3.2.1-204, " Analog Operational Test / Surveillance Calibration l of Auxiliary Feedwater pucp Suction Loop p-AF051." ;

. On June 1, 1990, at 0800, with Unit 2 at 30% p oer, the 2A auxiliary feedwater (AFW) pump was taken out of wrvice (005) ,

to perform 2Bwls 3.2.1-240. The appropriate TS 3.).1.2.a LC0 i for an inoperable AFW pump was entered. . On the previous shift at 5:44 a.m., the 28 energency diesel generator (EDG) ,

was taken 005 to check for crisscrossed airlines. Offsite electrical AC power sources were verified. operable within one hour as required by TS 3.8.1.1. Additionally, equipment that- ,

depended on the 2A EDG'6s an emergency source of power (the 2A EDG is the emergency source of power for the 2A AFW pump) was verified operable within the two hour requirement. If the two hour requirement cannot be met in accordance with TS 3.8. action statement "C," the unit must be in hot standby (Mode 3) '

within six hours and cold shutdown (Mode 5) within the following ,

thirty hours. When the 2A AFW pump was taken 00S for the surveillance activity, the TS action statement "C" for TS 3.8.1.1, had unknowingly been entered and the six hour -;

time period to Mode 3 starte ;

- At-1:46 p.m., the 2A AFW pump was declared operable. The six hour LCO had 14 minutes remaining until entering Mode 3 would ,

have been required. The SCRE on the 7:00 a.m. to 3:00 !

shift (shift 2)didnotrealizethattheyhadenteredthe action statement "C" for TS 3.8.1.1 until the paperwork.for both completed surveillances was reviewed. The shift 2 SCRE ,

is the same individual who authorized the 2A AFW pump 00S and the start'of the 2A AFW pump surveillance. In both instance l he failed to identify that with the 28 DG 005, the 2A AFW 005 3 would place unit 2 in the six hour LC0 for shutdown. The 28 EDG 005 was also noted on the shift turnover sheet of the shift engineer, shift control room engineer, and Unit-2 nuclear -

station operator. The shift turnover sheets are required to

.be signed by both the off-going and on-coming shift. This item is considered unresolved and will be'further evaluated during the next report period (457/90015-04(DRP). The SCRE- i failed to properly implement the requirements-of BwAP 330-1,

" Station Equipment Out-of-Service Procedure," Step D.1, which 1 requires the licensee to review all TS related 005 "to assure opposite train operability and determine any actions necessary to satisfy the TS or surveillance."

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One unresolved item was identifie >

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.  ; TrainingEffectiveness(41400,4,12011 ,

The effectiveness of training rengrams for licensed and non-licensed ,

personnel was. reviewed by the irspectors during the witnessing of the licensee's performance of routine surveillance, maintenance, and operational activities and during the review of the licensee's response to events which occurred during the ins)ection period. Personnel appeared to be knowledgeable of the tas cs being performed, and nothing .

was observed which indicated any ineffectiveness of trainin l i

No violations or deviations were identifie . Report Review

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During the inspection period, the inspector reviewed the licensee's ;

Monthly Performance Report for April and May 1990. The inspector confirmed that the information provided met the requirements of ,

TS 6.9.1.8.and Regulatory Guide 1.1 !

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The inspector also reviewed the licensee's Monthly Plant Status Report i for April 199 No violations or deviations were identifie . . TMI Action Plan' Requirement Followup (?5565) - (Closed) TMI Item 11 B.2.3 (Units 1 and 2): This item required the licensee to review the radiation and shielding design of spaces :

around systems on which personnel occupancy may-be unduly limited i or safety equipment may be unduly degraded by radiation during j operation following an accident resulting in a degraded cor The licensee was also required to perform the plant modifications that will permit access to vital areas and protect safety equipmen NUREG-1002, " Safety Evaluation Report related to the operation of Braidwood Station, Units 1 and 2," dated November _1983, referenced NUREG-0876, * Safety Evaluation Report related to the operation of . Byron Station, Unit 1- and 2," dated February 1982, which documented the acceptability of the radiation and shielding design review-meeting TMI. Item II.B.2 requirements given in NUREG-0737. Based on the inspectors' review of the Safety Evaluation Reports, TMI Action item II.B.2 (Units 12) is closed, (Closed) TMI Item II.B.3.1 (Units 1 and 2): This item required the

. licensee to perform a design and operational review of the reactor coolant and containment atmosphere sampling line systems to determine the capability of personnel to promptly obtain a sample ,

under accident condition ,

NUREG-1002, Supplement No.- 1. " Safety Evaluation Report related to the operation of Braidwood Station, Units 1 and 2," dated September 1986, documented the acceptability of the post-accident- ;

sampling and analysis system in meeting TMl item II.B.3 requirements r

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given in NUREG-0737. Based on the ins 3ectors' review of the Safety Evaluation Report TM1 Action 1 Mm 11.3.3.1 (Units 1 and 2) is closed, (Closed) TM1 Item III.D.3.4.3 (Units 1 and 2): This item required the licensee to review the control room habitability systems guidance of the Standard Review Plan and Regulatory Guides 1.78 and 1.95, to assure that operators in the control room will be adequately protected against exposure to unacceptable levels of radiation during and after a design basis accident and to unacceptable levels of harardous chemicals released on or in the vicinity of the sit NUREG-1002, " Safety Evaluation Report related 4.0 the operation of Braidwood Station, Units 1 and 2 " dated November 1983, referenced NUREG-0876, " Safety Evaluation Report related to t;st operation of Byron Station, Units 1 and 2," dated February 198, ,

concluded that the control room emergency air satisfies the requirements of TMl item 111.D.3.4 given in NUREG-0737 and GDC l Based on the inspectors' review of the Safety Evaluation Reports, TM1 Action item 111.0.3.4 (Units 1 and 2) is close . Meetings and Other Activities (30702)

Visitors from Atomic Energy Council of Taiwan On June 5, 1990, three Radiation Protection Engineers from the Atomic Energy Council of Taiwan toured the Braidwood Station. One of the engineers was participating in a six month exchange program between the NRC in Region 111 and the Atomic Energy Council of Taiwan. The licensee escorted the engineers throughout the statio No violations or deviations were identifie . Unresolved items Unresolved items are matters about which more information is required in order to ascertain whether they are acceptable items, violations, or deviations. An Unresolved item disclosed during the inspection is discussed in Paragraph , ExitInterview(30703)

The inspectors met with the licensee representatives denoted in Paragraph I during the inspection period and at the conclusion of the inspection on June 19, 1990.. The inspectors summarized the scope and results of the inspection and discussed the likely content of this inspection report. The licensee acknowledged the information and did not indicate that any of the information disclosed during the inspection could be considered proprietary in natur .

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