IR 05000352/1997001

Insp Repts 50-352/97-01 & 50-353/97-01 on 970204-0329. Violations Noted.Major Areas Inspected:Operations, Engineering,Plant Support & Maintenance

ML20148C166
Person / Time
Site:	Limerick
Issue date:	05/07/1997
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20148C149	List: IR 05000352/1997001 ML20148C143 ML20148C151
References
50-352-97-01, 50-352-97-1, 50-353-97-01, 50-353-97-1, NUDOCS 9705150136
Download: ML20148C166 (49)

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U.S. NUCLEAR REGULATORY COMMISSION

, REGION I i

License Nos. NPF-39 NPF-85

Report Nos. 97-01 97-01 Docket Nos. 50-352 50-353

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. Licensee

PECO Energy

. Correspondence Control Desk P.O. Box 195 j' Wayne, PA 19087-0195 l l

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. Facility Name: Limerick Generating Station, Units 1 and 2 i

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Inspection Period: February 4, through March 29,1997

Inspectors: N. S. Perry, Senior Resident inspector F. P. Bonnett, Resident inspector R. L. Nimitz, Senior Radiation Specialist, DRS A. Lohmeier, Senior Reactor Engineer, DRS L. A. Dudes, Reactor Engineer, DRS L. M. Harrison, Fire Protection Specialist, DRS G. C. Smith, Senior Physical Security inspector, DRS R. J. Summers, Senior Resident inspector M. J. Buckley, Resident inspector Approved by: Richard R. Keimig, Chief Projects Branch 4 9705150136 970507 "

PDR ADOCK 05000352 "

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EXECUTIVE SUMMARY Limerick Generating Station, Units 1 & 2 NRC Inspection Report 50-352/97-01, 50-353/97-01 This integrated inspection included NRC review of PECO Energy operations, engineering, maintenance, and plant support. The report covers an 8-week period of resident inspection; in addition, it includes the results of announced inspections by a regional senior radiation specialist, a senior physical security inspector, two engineering inspectors and a fire protection program inspector.

Operations e Conduct of operations was professional and safety-conscious (Section 01.1).

• Operator conduct during the recent refueling outage at Unit 2 was very good with one minor exception. Good coordination between field personnel and control room operators was demonstrated during refueling cavity draindown. The foreign material controls implemented during suppression pool activities were very effective; however, the drywell condition was unsatisfactory during the close out ;

inspection, requiring additional cleaning. System restorations and startup activities l were well managed and controlled, with very good interfaces between the operators, and engineering and maintenance personnel (Section 01.2).

• Several deficiencies including inadequate procedural guidance regarding applicable technical specifications, inadequate review of plant impact for a reactor protection system (RPS) modification, and madequate control panel walkdown after the RPS channel was de-energized contributed to removing control rod blades and drives without the required source range rnonitor indication and alarm function operable (Section 01.3).

• Unit 1 operators responded promptly and effectively to the failure in the reactor manual control system (Section O2.2).

• Plant management made several conservative decisions associated with the Unit 2 refueling outage, which expanded the scope of the outage. The basis for these decisions was to ensure component reliability and qua'!ty, and demonstrated that there was no unreasonable schedule pressure (Section 06.1).

• Self-assessment activities were effective. Oversight review committees reviewed several activities related to safe station operation. The Nuclear Review Board meeting was devoted to review of the February 1997 Unit 2 refueling outage and discussed the higher nun.ar of issues that were associated with emergent work or schedule changes (Section 07.1).

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Maintenance

• The Limerick Unit 2 inservice inspection (ISI) program documentation for the first ten year interval, including examination schedules, examination samples and component safety classifications, met all requirements in the 1986 Edition of the ASME Boiler and Pressure Vessel Code. (Section M1.1)

• The inspectors found less .har, tdaquate procedural adherence while performing o

nondestructive surface exams on ASME Code Class 1 components. (Section M1.2)

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• The oversight of the work performed by the ISI vendor included only n ten percent '

data review and the licensee does not require direct observation of the actual exams i being performed in the field by its qualified personnel. (Section M1.2)

• The Internal Vessel Visual Inspection was a thorouch, clear inspection of the reactor internal piping. (Section M1.3)

• Maintenance activities at both units were performed very well. The safeguards auxiliary switch replacement at Unit 1, as well as the many jobs observed during <

the Unit 2 refueling outage, were well planned and executed. Communications between operators and maintenance technicians were very good. System manager j oversight was very good, and procedure compliance was exceptional. The  ;

inspectors found the controls and standards of the Foreign Material Exclusion i program for the Unit 2 refueling floor to be wellimplemanted. (Section M1.4)

• Surveillance tests conducted by operators during the Unit 2 plant startup from the !

refueling outage were performed very well. Excellent oversight by the system !

managers and shift supervision, as well as thorough prebriefings and l communications during the surveillance, were noted. (Section M1.5) l l

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• The scheduling of simultaneous outages for two systems important to safety demonstrate an inadequacy in management's oversight of work week planning.

Shift operators placed high confidence on an attachment to an administrative guideline without fully understanding the cautions within the body of the procedure.

(Section M2.1)

Enaineerina

• PECO engineering and maintenance organizations exhibited strong performance in the root cause evaluation and replacement of a cracked safe end nozzle (N12B). l (Section E1.1)

• In January 1997, a modification to the Unit 2 fuel preparation machines was made, which made the information in the Updated Final Safety Analysis Report (UFSAR)

incomplete. A test fixture was added that caused the fuel to be raised closer to the water surface. This resulted in raising two spent fuel assemblies closer to the surface of the spent fuel pool than allowed by the UFSAR. Th;s modification was ;

not evaluated to determine if it activity would make informe. tion in the SAR l iii

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$ inaccurate or incomplete. Failure to perform a 10 CFR 50.59 review, as required, is

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a violation (Section E1.2).

e Actions taken to ensure that all the steam flooding dampers are operar,b were very 3 good. The testing of the dampers was carefully planned and systematically i

implemented so that the deficient dampers were correctly identified and corrective i actions were immediately taken to restore these dampers to an operable status. A ,

1 Non-Cited Violation resulted due to the failure to conduct a preventive maintenance I

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$. activity to clean and lubricate the steam flooding dampers, which resulted in J

operation outside the design basis of the plant (Section E1.3).

3- Plant Sucoort e

e Overall effective applied radiological controls were established and implemented for

. the Unit 2 outage. ALARA controls for ongoing work activities were effective.

j- Some indications of lax worker attention to radiation protection procedures were

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noted (Section R).

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o Access authorization (AA) records reviewed con'.ained the appropriate information

required by procedures and regulatory requiremrnts, upon which to base a decision

! to grant unescorted access. All the individuals who were working for a specific short-term contractor onsite, and whose AA records were reviewed, had been

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processed through the AA program for unescorted access except for a replacement f crane operator who was onsite for one day to teplace the regular crane operator j who was off on.the day the records were reviewed. The replacement crane operator was on the site as a visitor and was under proper escort. (Section S1)

j e Fire protection equipment conditions were good and housekeeping was very good.

l Licensee efforts for improving fire protection program procedures were found to be

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appropriate. (Section F.2)

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! * Weaknesses were identified regarding PECO's process for controlling hotwork ,

j activities, including the failure to involve the industrial Risk Management Group in

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l i conducting independent reviews and tu Jse this group for program oversight. This i reflected negatively on the effectiveness of the fire protection program. (Section i F.1 )

I j 'o- - Recent actions taken to improve the reliability of emergency lighting were good,

.despite the longstanding history of problems in the plants. (Section F.2)

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.* The process for tracking satisfactory completion of fire brigade program training

!- requirements was weak. Performance measures for assessing the effectiveness of the fire protection program were narrowly focused. (Sections F.5, F.6)

e Good initiatives have been taken recently to identify and resolve fire protection i issues. QA Audits and surveillances were focused appropriately. Good problem y identification was noted. (Section F.6, F.7)

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ReDort Details Summarv of Plant Status Unit 1 began this inspection period operating at 100 percent power and remained essentially at full power for the remainder of the period, with minor exceptions for turbine control valve testing and control rod pattern adjustments. 1 Unit 2 began this inspection period shut down for the fourth refueling outage which began on January 30. At the conclusion of the outage, the operators made the reactor critical on February 26, and synchronized the main turbine generator to the grid on February 28. On I

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March 27, and until the end of the inspection oriod, operators reduced reactor power to 70 percent due to high vibrations and shaft alignment problems at the Alterex bearing of the main generator.

1. Operations 01 Conduct of Operations'  !

01.1 General Comments (71707)

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Using Inspection Procedure 71707, the inspectors conducted frequent reviews of ongoing l plant operations. In general, PECO Energy's conduct of operations was professional and !

safety-conscious. Activities associated with the Unit 2 outage were generally well ,

controlled and conducted. Several minor operation errors occurred early in the outage, but l were adequately addressed so that the remainder of the outage was error free. j Communications were excellent. Activities at Unit 1 were responded to promptly and j effectively. '

01.2 Unit 2 Refuelina Activities and Start-uo a. Scoce (60710,71707)

The fourth refueling outage at Unit 2 began on January 30 (see inspection Report 50- j 352/353 96-10) and was completed on February 28. The inspectors monitored daily activities during the outage and observed several special activities including: fuel movement activities on the refueling bridge, the refueling cavity draindown, the closeout 4 inspections of the drywell and suppression pool, and the unit start-up. The inspectors attended daily meetings and briefs and discussed issues with the appropriate plant representatives when necessary.

b. Observations and Findinas j l

The inspectors observed that fuel movement activities on the refueling bridge were accomplished well with no major problems encountered. In general, activities progressed

' Topical headings such as 01, M8, etc., are used in accordance with the NRC standardized reactor inspection report outline. Individual reports are not expected to address all outline topics. l

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smoothly, with the bridge operator, fuel handling supervisor, and spotter exhibitmg !

excellent three part communications. Activities were coordinated very well with the 1 l control room operators, and all activities observed were conducted safely and effectively. l The inspector found the fuel hand!!ng personnel to be very knowledgeable and  !

professional. '

On February 3, an equipment operator (EO) identified two valves not properly locked as l required by administrative procedure A-C-8, Control of Locked Valves and Devices.

Specifically, A-C-8 requires that the handwheel or other operating mechanism be locked to restrict operation of the valve. The EO found the valves closed as required, but the valves were inadequately locked, in that the valves could be manually operated without opening the locks. Results of the investigation into this event indicated that the valves were inadequately locked and inadequately independently verified due to concerns with dose levels in the area. The valves are located in the drywell under vessel area where elevated dose levels were present. Self induced pressure to exit the area quickly led to inadequate checks by both individuals. Corrective actions included: the valves were immediately l locked, as required, and independently verified; all operation shifts were briefed on the event; and the valves were removed from the locked valve list, since it was determined that it is not necessary or required to lock these valves. This failure to properly lock two valves and independently verify them as required by procedures constitutes a violation of i minor significance and is being treated as a non-cited violation, consistent with Section IV of the enforcement policy.

On February 20, the inspectors observed portions of the reactor coolant system (RCS)/ Refueling Cavity draining evolution in support of returning the unit to operations following refueling. The overall evolution was conducted in a controlled and safe manner, although the draining evolution encountered some minor delay due to equipment malfunction associated with one of the flow path valves in rejecting the water from the cavity. Good command and control by shift management and good coordination between field personnel and the control room operators were observed by the inspectors.

As a result of PECO Energy obtaining deferral from the NRC for upgrading suppression pool suction strainers until January 1,1999 for Unit 2, the inspectors observed activities in the suppression pool, including cleaning activities, maintenance activities, and closecut activities. In general, foreign material exclusion (FME) controls were found to be very good throughout the outage. Vacuuming of accessible areas along the catwalk, desludging of the suppression pool floor, and filtering the water were a primary focus. Additionally, tF.e downcomers (inside and outside) were inspected and three minor strainer repairs were completed. Near the end of the refueling outage, the inspector conducted a pre-closeobi inspection of the suppression pool; this inspection occurred just prior to the formal closecut inspection by plant management. The inspector concluded that the conditions in the suppression pool were extremely clean, that extensive cleaning and inspection had been done, and the FME controls had been effective during the outage.

Near the end of the Unit 2 iefueling outage, the inspector conduc,tsd a pre-closecut inspection of thrs Unit 2 drywell; this inspection occurred just prior to the formal closeout inspection by plant management. The inspector concluded that the conditions in the drywell were adequate, but there were a number of housekeeping deficiencies noted.

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l Tools and trash, such as paper, wire, and other materials, were inappropriately left in the I drywell. After the plant management inspection, the inspector confirmed that management I independently found the conditions in the drywell unacceptable. Plant management had additional cleaning performed prior to final drywell closecut. The inspector concluded that the management standards were appropriate to ensure that allinappropriate materials were ;

properly removed from the drywell.

Operators conducted the Unit 2 plant startup very well on February 26, following the refueling outage. Shift management exercised positive control of control room traffic l permitting only necessary personnel to be in the control room during the startup. The l reactor operator's (RO) approach to criticality, control of plant heat-up, and generator i synchronization were performed very well without incident. A second licensed operator !

double verified rod withdrawal activities and a reactor engineer was present for reactivity concerns. Additional control room operators ensured that the RO withdrawing the control rods was not disturbed by any other control room activities. Interfaces between the operators, and engineering and maintenance personnel were very good.

c. Conclusions Operator conduct during the recent refueling outage at Unit 2 was very good with one minor occurrence. Good coordination between field personnel and control room operators was demonstrated during refueling cavity draindown. The FME controls implemented during suppression pool activities were very effective; however, the drywell condition was unsatisfactory during the close out inspection, requiring additional cleaning. System restorations and startup activities were well managed and controlled, with very good interface between the operators, and engineering and maintenance personnel.

01.3 Control Rod Blade Removal with Inocerable Source Ranae Monitors - Unit 2 l

a. Scoce (71707)

On February 7, control room operators identified that the two inservice source range monitor (SRM) channels were inoperable during control rod drive (CRD) mechanism and l control rod blade removal activities from the reactor pressure vessci (RPV). This activity !

did not conform with Technical Specifications (TS) 3.9.10.2, Multiple Control Rod )

Removal, in that, the continuous visual indication and audible alarm function was de- !

energized. The inspector reviewed and discussed this event with the appropriate members !

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b. Observations and Findinas j The control room staff was not aware of the TS requirement that two continuous visual SRM indications must be operable as specified in TS 3.9.10.2, prior to beginning CRD !

maintenance activities. Further, the applicable procedures in effect did not give sufficient )

guidance to inform the operators of this requirement. A modification to the B reactor protection systems (RPS) uninterruptable power supply (UPS) was started during the CRD activities which, unknown to the shift operators, de-energized all of the SRM indication / recorders in the control room.  ;

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The unit operators were monitoring in vessel control rod (CR) blade and drive replacement activities on the core component transfer authorization sheet (CCTAS), and were complying with GP-13, Control Rod Drive / Control Rod Blade Outage Maintenance Coordination Procedure. Surveillance ST-6-107-360-2, Verification of Ter.h Spec Compliance Prior to Removal of Single / Multiple Control Rod Drives in OPCON 4 or 5, which verifies that at least two SRMs aie operable and that adequate shutdown margin (SDM)

exists, had been completed on a previous shift. The ST was required to be completed once every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> during CRD removal activities and was due to be completed prior to 3:30 p.m., on February 7.

At approximately 9:00 a.m. on February 7, the B RPS/UPS (Panel 2B-Y160) was de-energized to install a planned modification. The control room supervisor (CRS) briefed the operators and informed them that the 9/D SRMs would be inoperable and de-energized. It was assumed, however, that the A/C SRMs would remain operable since the A RPS/UPS was operable and the surveillance verifying TS compliance (ST-6-107-360-2) was

<;atisfactory. The CRS believod that the unit needed only to comply with TS 3.9.2, which required at least two SRMs to be operable during core alterations. However, at this time, no core alterations were in progress since CR blade and CRD maintenance activities are not core alterations.

At 10:50 a.m., the RO performing a required channel check of SRM counts for a CR blade move noted that all four SRM recorders were de-energized. He immediately informed the CRS who reviewed TS 3.9.2, and determined that the limited condition for operation (LCO)

was satisfied since no core alterations were in progress. The CRS then reviewed surveillance ST-6-107-360-2, end determined that TS 3.9.10.2, Multiple Control Rod Removal was, in fact, applicable. TS 3.9.10.2 required that two SRMs be operable, which included continuous visual indication in the control room and at least or'e audible alarm.

The CRS immediately suspended allin-vesuel work.

Operations management initiated an investigation which determined that the de-energization of the 2B-Y160 v:nel for the B RPS/UPS modification also de-energized the A/C SRM recorders. An ina%quate review of the plant impact of the RPS modification had been performed during the planning phase. Further, several procedures used during the CRD maintenance did not refer to TS 3.9.10.2, including GP-13. It was not until the CRS reviewed surveillance ST-6-107-360-2 that the control room staff became aware of the applicable TS.

The inspector determined that the control room staff had not been aware of the TS requirements of TS 3.9.10.2, prior to beginning CRD maintenance activities and that the applicable procedures had not given sufficient guidance to the operators. Further, operators did not perform an adequate review of vital control room parameters after de-energizing the B RPS/UPS panel. The surveillance, ST-6-107-360-2, which ultimately brought the appropriate TS to the attention of the CRS, was not re-performed immediately after de-energizing the RPS panel to assure proper TS compliance. The inspector concluded that this activity was in violation of TS 3.9.10.2.

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The inspector determined that the safety consequences of this event were minor. A 1994 ;

TS change to the definition of Core Alteration eliminated CR blade and drive replacement l activities from the definition, since this is accomplished when there are no fuel assemblies in the associated core cell. The safety evaluation performed for the TS change stated that CR movement in a control cell without fuelinstalled had no significant reactivity effect, therefore SDM would not be affected. Operators had SRM indication on the process l

computer (credit is not taken for this indication for TS) and period meters, a CR withdrawal block was active with the B/D SRM de-energized, and no other core alterations were being performed. Management is currently evaluating a TS change to TS 3.9.10.2 to make it consistent with TS 3.9.2 and the definition of core alteration. This violation does not suggest a safety or regulatory concerr'. This licensee-identified and corrected violation is a

Non-Cited Violation, consistent with Section Vll.B.1 of the NRC Enforcement Policy. )

c. Conclusion

Several deficiencies including inadequate procedural guidance regarding applicable TS, inadequate review of plant impact for a RPS modification, and inadequate control panel J

walkdown after the RPS channel was de-energized contributed to removing CR blades and drives without the required SRM indication and alarm function operable.

02 Operational Status of Facilities and Equipment j

02.1 Routine Plant Tours (71707)

The inspectors used Procedure 71707 to perform routine tours of the facility and also to-walk down accessible portions of engineered safety feature (ESF) systems:

• Residual Heat Removal (RH3) System - Unit 2

• Emergency Diesel Generators (EDG) - Unit 1 and 2

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Equipment operability, material condition, and housekeeping were acceptable in all cases.

Several minor discrepancies were brought to management's attention and were corrected.

The inspectors identified no substantive concerns as a result of these walkdowns.

O2.2 Reactor Manual Control System Failure - Unit 1 a. Insoection Scope (71707)

On February 20, the inspector observed the control room operator response to a failure in the reactor manual control system (RMCS) on Unit 1. The inspector observed and discussed the follow-up actions with the appropriate control room staff.

b. Observations and Findinas The inspector determined that the control room operator responded well to a failure in the RMCS on Unit 1. The unit was at approximately 100 percent power when an overhead annunciator alarmed, indicating that a Rod Withdrawal Block had occurred, in addition, the reactor manual controls panel indicated that rod blocks were present and that the "A" and

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"B" control rod drive flow equalizing solenoid valves were performing erratically (rapid on -

off indication). Operators referred to the appropriate alarm response procedures and system operating procedures, and dispatched an EO to the equalizing valves. The EO determined that the solenoids were chattering. Maintenance personnel, and system and reactor engineering personnel were contacted to assist the control room staff in determining the cause. Operators verified that the RMCS was, in fact, failed, which limited their response to any power transients. They also referred to the appropriate technical i specifications to determine if the control rods were still operable in this condition. I Although the rods could not be moved manually, they were still capable of being l

scrammed. Therefore, the operators concluded that the control rods were still operable.

Based on an independent review of the associated technical specifications and discussions i with engineering personnel, the inspector also concluded that the control rods were still l operable. I l

Maintenance personnel isolated the failure to a faulty power supply. When the associated fuses were lifted, the solenoid valve chattering ceased. Since a replacement power supply was not readily available, the operators and the reactor engineers assessed what actions the operators should take if a plant power transient occurred. It appeared that if, for ,

example, a recirculation pump tripped from these conditions, it was likely that the reactor j core would enter an unstable region of the power-to-flow map. Without the capability to ,

change power using the control rods, operators would have to scrrm the plant manually if

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such a transient were to occur. After the inspector observed the discussions between the ,

reactor engineer and the shift supervisor regarding operator response to postulated plant '

transients without RMCS, the shift supervisor was asked if any controls had been j established to minimize the probability of such events. At the time, the shift had not yet !

considered the need for such controls; however, upon recognizing that a number of j maintenance activities associated with the ongoing outage at Unit 2 were in close proximity to the Unit 1 Recirculation Pump Motor-Generator Sets and local control panels, the control room immediately dispatched an operator to set up a restricted work zone around the equipment. The inspector verified that the restricted work zone was established and that contracted maintenance employees in the area were adhering to the i newly established controls. 1 c. Conclusion The inspector concluded that operators responded promptly and effectively to the failure in I the Unit 1 reactor manual control system.

06 Operations Organization and Administration 06.1 Manaaement Decisions Durina Refuelina Outaae a. Scope (71707)

Plant management made several conservative decisions associated with the Unit 2 refueling outage. )

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b. Observations and Findinas

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Several emergent work activities arose during the Unit 2 refueling outage. Plant management addressed these items conservatively, incorporating the work into the outage.

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Management decided to perform permanent repairs to plant components where temperary repairs might have been acceptable. The focus of these decisions was on quality to ensure that the plant operated safely and reliably during the next fuel cycle. The inspector noted no additional schedule pressure as a result of these items, which included: l i  !

• A permanent repair (code repair vice a weld overlay) on the N12B nozzle (see i Section E1.1).

• Delaying natural circulation operations until after the N12B nozzle repair was completed.

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• During natural circulation operations, main steamline backfill or separator installation would not occur until the RHR mode of shutdown cooling was restored. '

• High pressure coolant injection (HPCI) and reactor core isolation cooling (RCIC) test runs were conducted on main steam rather than auxiliary steam.

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• Activities were stopped on the refuel floor several times, due to fuel preparation  !

machine problems, when SRMs were found inoperable, and for blown refueling I bridge fuses.

• During startup activities, a condensate check valve was dismantled and repaired rather than leak seal repaired.

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Plant management made several conservative decisions associated with the Unit 2 ,

refueling outage, which expanded the scope of the outage. The basis for these decisions  !

! was to ensure component reliability and quality, and demonstrated that there was no unreasonable schedule pressure.

07 Quality Assurance in Operations

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07.1 Self-Assessment Activities (717071

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During the inspection period, the inspectors reviewed or attended multiple self-assessment j activities, including:

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e the quarterly Nuclear Review Board (NRB) meeting on March 6;

• various Plant Operational Review Committee (PORC) meetings and meeting minutes;

• various quality verification (QV) and independent safety engineering group (ISEG)

reports.

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The NRB and PORC reviewed several activities related to safe station operation. The members of NRB and PORC actively participated in the meeting with open discussions on the plant issues while maintaining a focus on safety. A major portion of the NRB meeting was devoted to review of the February 1997 Unit 2 refueling outage NRB members appropriately challenged allissues that were raised; noteworthy was the discussion which centered around the higher number of issues that were associated with emergent work or schedule changes. The NRB concluded that more information was needed and that the Peach Bottom facility experiences should be considered also. The inspectors concluded that the self-assessment activities were effective.

08 Miscellaneous Operations issues 08.1 (Closed) VIO 50-352/96-10-01, and VIO 50-353/96-10-02 These violations were reviewed in NRC Integrated Inspection Report 50-352/96-10, 50-353/96-10 and Notices of Violation. No responses were required to these violations based on NRC review of the root causes and corrective actions taken. These violations are administratively closed.

08.2 (Closed) LER, 2-97-001. Condition Prohibited by Technical Specifications in that Control Rods were Removed from the Core with Source Ranae Monitors Inoperable (90712)

This event is reviewed in section 01.3 of this inspection report. The LER met the requirements of 10 CFR 50.73, and the inspectors had no further questions regarding the event.

II. Maintenance M1 Conduct of Maintenance M 1.1 Inservice insocction (ISil Proaram Review a. Inspection Scope (73753)

The inspectors reviewed the results of past ISI programs for Unit 1 and Unit 2 to ascertain compliance of the Inservice Inspection program, the Nondestructive Examination (NDE)

procedures and the certifications of NDE personnel with the requirements of the American Society of Mechanical Engineers (ASME), Boiler and Pressure Vessel Code,Section XI.

Included in the review were the results of Unit 2 ISI inspections for the first interval, second period, first outage (2RO3) through the first interval, thiru period, first outage (2RO4). Also reviewed were the results of Unit 1 ISI for the first interval, third period, second outage (1RO6).

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) 9 b. Observations and Findinas I

Unit 2 First Interval - Second Period (3/17/93 - 2/19/95) - First Outage (2RO3)

The inspectors reviewed disposition of severalindications recorded during this period, including-I

• Liquid penetrant examination (PT) of weld EBB-208-1 FW16 that identified l porosity in the weld, which was evaluated to be within acceptable ASME Code requirements;

• PT of weld 28BP-201 HL1 also identified weld porosity within acceptable ASME Code requirements; l

• PT examination of RPV stabilizer bracket welds with rounded base metal !

indications were deemed acceptable within ASME Code requirements; and

• Visual examination (VT) of RPV steam dryer drain channel welds which identified linear indications in the support ring that were accepted by j evaluation (the examination was extended to include 100% of the steam )'

dryer in accordance with IWB-2430 and examinations will be performed for the next three inspection periods in accordance with IWB-2420 (B)).

i Unit 1 First interval --Third Period (3/12/94 - 3/1/96) - Second Outage (1RO6)

The inspectors found three conditions reported by the licensee during this period, including:

• Flaws in weld H-3 detected in the core shroud barrel were evalueted in l accordance with "BWR Core Shroud inspection and Flaw Evaluation  ;

Guidelines" and it was concluded there was sufficient margin to continue l operation;  !

l The examination of the steam dryer drain channel welds which revealed

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three cracks (3 inch) in the support ring, which were evaluated as inconsequential to the design life of the plant; and

• An ultrasonic test (UT) of nozzle to safe end weld VRR-1RD-1 A N2H l performed to detect any changes in a previously identified flaw in accordance with Generic Letter (GL) 88-01 revealed no change in the size of the flaw.

Unit 2 First interval - Third Period (First Outage (2RO4,1997))

The inspectors reviewed the ongoing ISI activities during this refueling outage. Selected for review were the ISI exam? nations in process during the inspection week. These included the following activitiec:

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Summary No. 680400 (Code Category B-F, item 85.150.) " Reactor Water Cleanup  !

System Component DCA-201-E2-W9, bimetallic half-coupling to thermowell weld i ultrasonic examination performed by GERIS." A circumferrential PT was also performed on j this weld. The inspection Check List was signed by General Electric and the Authorized l Nuclear Insurance Inspector (ANil), the examination summary sheet was signed by a NDE  ;

level ll, level lll and the ANil. No indications were reported. Documentation was still in process.  !

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Summary No. 445500 (Code Category C-F-2, item C5.51.) - RHR System Component GBB-  !

218-3 FW2O elbow to pipe wdd Magnaflux Test (MT), PT, and UT (hand held).  !

Engineering review of inspection results was not completed. The indication was found in  :

heat affected zone area at inner diameter of pipe. The disposition was not reviewed at this {

time. i i

Summary No. 340100/200 (Code Category BF, item B5.130.) - RHR System Component j DCA-418-1 FW1 12 inch pipe to safe-end weld UT (manual and using SMART system). I Examinations revealed several non-relevant indications (NRis), metallurgical and geometric.  !'

None of the NRIs met the reporting criteria outlined in the ASME Section XI 1980 Edition with-1981 addenda. Performance and review of results were signed by Level lils. All 3 signatures required on the examination results were not completed at that time. 1 Summary No. 712600/700 (Code Category B-D, item 3.100.) - Low Pressure Coolant

• lnjection loop C system Component N17D-1R nozzle to shell weld UT using GERIS 2000 i

' system and manual UT. The disposition was not completed and documented at time of, i inspection. l l'

Summary No. 702300/400 (Code Category B-A, item B1.12.) - Shell right and left side vertical seam weld UT inspection using GERIS 2000 system and manual examination. The j documentation of this examination was not completed at the end of the inspection. 1 In addition to the data review, the inspectors reviewed the personnel records of several of -

the ISI vendor's Level I and il examiners, and found that the personnel performing and j reviewing the NDE exams were all qualified in accordance with the Code.  ?

i The inspectors discussed the ISI program with the ANil to determine the level of *

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~ involvement and the degree to which the station will involve the ANil in the case of programmatic changes. The ANil at Limerick is assigned there full time, and he is -

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knowledgeable and involved in the work at the station.  !

c. Conclusions

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inspection results reviewed adhered to the Code requirements, and no regulatory issues  ;

were identified. The data packages that were not complete at the time of the inspection j will be reviewed by the inspectors when the 2RO4 ISI report is issued this spring. The  !

inspectors also reviewed the state of completion of weld examination for Class 1 y

. components for the first ten year interval. PECO Energy is on schedule for the completion i of the ISI first ten-year interval and should complete allinspections during 2RO5. l

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M1.2 NDE Observation and Data Review a. Insoection Scone (73753)

The inspector observed several NDE activities in progress and reviewed the NDE examination data.

b. Observations and Findinas The inspector observed NDE subcontractors performing liquid penetrant, magnetic particle and ultrasonic testing on various ASME Class 1 components. The inspector witnessed portions of the liquid penetrant tests for the following welds: RPV-21N-N12B, RPV-21N-N12D, and a bracket to vessel weld (270 azimuth). The contractor examiners did not have the appropriate tools required to perform a penetrant test when they arrived at the specific job site. Once the inspector arrived to observe the exams, the appropriate tools were obtained and, in some cases, the exam was started over to ensure proper procedural adherence. Specifically, while performing a penetrant test on the RPV-2lN-N128 weld, the Level 11 examiner applied the penetrant without determining the temperature of the material or verifying the time at which he applied the penetrant. Both of these parameters have significance in the accuracy of liquid penetrant exams. Once the examiner was aware of the inspector observing the performance of the exam, a second coat of penetrant was -

applied, this time noting the time and the temperature of the part.

On another occasion, the inspector arrived at a job site to observe a liquid penetrant exam.

Upon the arrival of the inspector, one of the Level 11 examiners had to return to the drywell entry point to obtain a watch. Again, this action was deemed a weakness by the inspector because the penetrant exam has several time dependent holdpoints and timers and thermometers should have been obtained as part of the preparation.

Also, while performing exam observations in the drywell, the inspector asked the examiners to verify that they could establish a magnetic field in a weld that was slated for a magnetic particle exam that morning (weld id APE-2MS-LD N3D). This particular weld had two containment air cooling units blowing full force on the weld. The examiners were not able to verify the magnetic field within the surface of the weld because the fans were blowing the particulate across the pipe itself. The inspector questioned the adequacy of the weld coverage and was assured the examiners would address the fan problem in some way. Upon review of the data package for this weld, the inspector noted that the examiners credited their bodies as shielding from the fans. When this issue was brought to senior management's attention, an examiner re-performed the examination.

After observing questionable exams being performed on ASME Class 1 components, the inspector questioned the PECO NDE Level lli regarding the oversight of the NDE vendor.

During followup to this question, the inspector determined that the licensee has no requirement or formal program in place for PECO personnel to oversee the vendor NDE activities and document the results.

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r The inspector obcerved several automated ultrasonic exams being performed on the safe

end-to-nozzia md nozzle-to-vessel welds. The SMART and GERIS automated exams performed by General Electric were of high quality with personnel exhibiting good

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communications and redundant control of the activity with additional monitors outside of the drywell to support the activity, in addition, the inspector observed manual ultrasonic

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exams performed on ASME Class 1 welds and verified that the calibrations were in accordance with the Code. The UT exams observed were thorough, and no procedural or regulatory issues were identified.

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c. Conclusions

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While the PECO NDE Levelllis do review ten percent of the data after the field exam has been performed, the NRC staff is concerned that this will not preclude identifying poorly

} performed exams. The inspector concluded that the lack of random over-check (re-

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performance of the examination) will potentially yield missed indications and/or defects in Limerick's ASME Code components. Immediate actions taken by management in response to the inspector concerns included; re-performance of nine NDE examinations by the PECO NDE personnel, PECO NDE oversight on 100% of all remaining ISI examinations and i evaluation of the effectiveness of the NDE vendor oversight program.

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M1.3 In-Vessel Visual Insoections a. inspection Scone (73753)

4 The inspector observed the in-progress visual inspection of the reactor internals, j specifically, the visual examinations of the core spray spargers and t-box junction, which-

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has historically exhibited cracking in older BWRs. Prior to observing the exams, the

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inspector reviewed PECO's procedure (MAG-CG-408, Rev. 6, " Visual Examination of Reactor VesselInternals") for compliance with the ASME Code, i

b. Observations and Findinas The inspector noted that, during the in-Vessel Visual Inspection (IVVI), a PECO NDE Level ill was present on the refuel bridge at all times. The contractor also provided NDE Level lli oversight for all aspects of the IVVI. The NDE Level its performing the visual exam and recording data exhibited excellent communication with the camera operators. The camera

, resolution of the vessel internal piping was excellent and the one mil and one-half mil calibrations required by the Code were easily obtained. PECO Nuclear has aggressively pursued excellence in IVVI, which was demonstrated through the use of two independent

, color cameras for critical vessel inspections. Overall, the inspector determined that the

IVVI, performed during the Limerick Unit 2 outage, was of high quality.

c. Conclusion The Internal Vessel Visual Inspection was a thorough and clear inspection of the reactor e internal piping.

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M1.4 Maintenance Activities a. insoection Scoce (62707)

The inspectors observed all or portions of the following work activities:

• Safeguards Transformer Breaker Auxiliary Switch Replacement - Unit 1

• D14 EDG Heat Exchanger Cleaning - Unit 1 e 2A Recirculation Pump Seal Replacement - Unit 2

• 2A Residual Heat Removal Pump Replacement - Unit 2

• A-C-130 Refuel Floor Foreign Material Controls - Unit 2 b. Observations and Findinas Control and oversight of maintenance technicians replacing an auxiliary switch on the D11- j bus safeguards transformer breaker were very good. The technicians complied with the l procedure and used self-checking techniques beyond those required by procedure. The inspectors noted good engineering support as well as very good communications among the control room operators, technicians, and system engineers in the field. Good technical assistance was present during the maintenance and post maintenance testing. The system ;

manager conducted a brief in the control room that resolved the shift manager's concerns, including the impact to the plant during the switch replacement and post-maintenance testing. A good questioning attitude by the technicians was noted when an unexpected voltage was found on the switch. The system manager determined that a space heater in a parallel path around the switch caused the unexpected voltage.

Unit 2 Outaae Activities Maintenance technicians worked safely and used good work practices while replacing the 1 A recirculation pump seal. The inspectors observed the seal package removal from the pump and transport to the upper drywell elevation. Technicians rigged several chain falls which supported the seal package to an opened floor grate where the seal was lifted to the upper drywell elevation. Appropriate safety barriers were in place and radiological controls were used to prevent the spread of contamination.

The activities associated with the replacement of the 2A RHR pump motor were conducted very well. The inspectors observed the removal of the original pump motor and the setting of the new motor in place. The original motor was experiencing a small oil leak in the upper bearing. The oil from the leak had been drawn into the motor by normal cooling air flow when the motor was running, resulting in the windings becoming oil soaked. The new motor was smaller in physical size and required several modifications of existing support systems to be performed. Technicians aligned the motor and conducted an extensive acceptance test plan which included, meggering the motor, running the motor for three hours while monitoring winding and feeder cable temperatures, and motor vibration testing. The new motor performed very well.

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The inspectors found the controls and standards of the FME program for the Unit 2 refueling floor to be well implemented. The inspectors reviewed procedure A-C-130, Fuel Floor Foreign Material Exclusion, and independently observed the work practices of individuals on the refueling floor. The FME areas were clearly marked with signs delineating what level of housekeeping was required. Appropriate material accountability logs were located outside the access to the controlled area and were being actively used by personnel. Clear materials were not used or were appropriately marked with yellow tape. Lanyards were used for smallitems, such as pens, on the refueling bridge.

Personnel were aware of FME controls and requirements when questioned by the inspectors. No major deficiencies were noted during the outage.

c. Conclusion Maintenance activities at both units were performed very well. The safeguards auxiliary switch replacement at Unit 1, as well as the many jobs observed during the Unit 2 refueling outage were well planned and executed. Communications between operators and maintenance technicians were very good. System manager oversight was very good, and procedure compliance was excellent. The inspectors found the controls and standards of the Foreign Material Exclusion (FME) program for the Unit 2 refueling floor to be well implemented.

M1.5 General Comments on Surveillance Activities (61726)

The inspectors observed selected surveillance tests to determine whether approved procedures were in use, details were adequate, test instrumentation was properly calibrated and used, technical specifications were satisfied, testing was performed by knowledgeable personnel, and test results satisfied ar.ceptance criteria or were properly dispositioned.

The inspectors observed portions of the following surveillance activities:

• Unit 1 - 18 RHR Quarterly Pump, Valve and Flow Test

• D24 EDG Monthly Test

• Unit 2 - HPCI Operability Verification (Low Power Test)

• Unit 2 - RCIC Operability Verification (Low Power Test)

• Unit 2 - HPCI Operability Verification (High Power Test)

• Unit 2 - Main Turbine Stop Valves and Overspeed Testing

• D21 EDG Fast Start Test

• D11 EDG Slow Start Test

• D11 Bus Undervoltage Relay Functional Test

• Unit 1 - RCIC Quarterly Pump, Valve and Flow Test

• Unit 2 - HPCI Quarterly Pump, Valve and Flow Test

• Unit 1 - HPCI Quarterly Pump, Valve and Flow Test Surveillance tests observed by the inspectors were conducted well using approved procedures, and were completed with satisfactory results. Communications between the various work and support groups were good, and supervisor oversight was good.

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The inspectors observed several surveillance testsd' uring the Unit 2 plant startup from the f

refueling outage. During the low power operability tests for the HPCI and RCIC systems, the inspectors noted excellent performance by the operators and oversight by the system managers and shift supervision. Shift management conducted detailed pre-test briefings and communications during the surveillance tests were noteworthy.

M2 Maintenance and Material Condition of Facilities and Equipment M2.1 Work Control Removal of Two Safety Systems a. Scope (62707)

The inspector identified that two Unit 1 systems important to safety were removed from ,

service and tagged for planned maintenance simultaneously. The inspector raised concerns  !

to shift management as to why the HPCI system and the D14 EDG were both out of I service. The inspector also discussed this issue with representatives from the Operations Support Group and Work Week Management.

b. _Qbservations and Findinas Maintenance Planning originally scheduled a HPCI system outage for work week 12 (March j 24 through 29). Due to schedule problems from the previous week, D14 EDG work was  !

carried over and sponsored into the same work week. The operations clearance and I tagging (OPCAT) cc.ordinator ran a probabilistic safety assessment (PSA) for both of these systems out of service together to determine the risk impact for core damage frequency..

(CDF), in the Sentinel computer. The result was an orange code which means that a reassessment of the schedule should be performed (Sentinel CDF codes are, in increasing significance order, Green, Yellow, Orange, and Red). The HPCI and D14 EDG work windows were then scheduled such that one would follow the other. The OPCAT and  !

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Work Week Manager reviewed an attachment to the administrative guideline procedure AG-43, Guideline for the Performance of System Outages, which indicated that the HPCI system and the D14 EDG could be removed from service simultaneously. The OPCAT and Work Week Manager discussed the Sentinel analysis results with the PSA Branch at Chesterbrook, who ran the same model in its Sentinel computer. The results were a yellow code which indicated that both systems could be removed from service together. The 1 OPC/(T and Work Week Manager, therefore, decided to change the work week schedule to i

~ allow both systems to be unavailable simultaneously based on the Chesterbrook Sentinel  !

re';ults and the AG-43 attachment.

Administrative guideline procedure AG-43 provides decision making and planning guidance for the execution of system outages. based on PSA insights and operating judgement. The HPCI and diesel systems are considered to be systems important to PSA safety in AG 43.

-The attachment in AG-43 was the reference given the inspector by the shift manager as i

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~ justification for allowing the HPCI system and D14 EDG being removed from service simultaneously. The inspector concurred that the attachment did allow the two systems to be out of service simultaneously, but also noted that a step in the body of AG-43 stated that only one system important to PSA safety per unit should be uneveitable at a time. 'j i

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i Further, if two or more systems were unavailable, then Operations Management should I review and approve the plan.

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reviewed the circumstances of the event. The OPCAT and Work Week Manager believing i that they were in compliance with AG-43, made their decision to change the work week j schedule, but failed to consult and obtain Operations Management's approval. Further, the i differing results between the Limerick and Chesterbrook Sentinel computers was attributed

to the large number of case runs available in each computer. The Limerick Sentinel

computer has not activated all of the potential case studies models available, whereas  ;

Chesterbrook has. Therefore, the Limerick Sentinel computer was more conservative.

Operations management initiated a Performance Enhancement Program (PEP) review of the  !

event and discussed the event with all OPCATS and Work Week Managers. The inspector l

discussed providing shift management with more insight and background for system j outages in the daily OPCAT Shift Briefing Sheet. Corrective actions regarding improving i the accuracy of the Sentinel computer at Limerick and AG-43 guidance will be determined during th3 PEP evaluation.

c. Conclusions The scheduling of simultaneous outages for two systems important to PSA safety demonstrated an inadequacy in managements oversight of work week planning. Shift operators placed a high confidence in an attachment of an administrative guideline without fully understanding the cautions within the body of the procedure.

M8 Miscellaneous Maintenance issues M8.1 (Closed) Unresolved item 50-352/353 95-10-01. Generic Concerns With Storaos of Soare Plant Components (92902)

The inspector reviewed PECO Energy's corrective actions addressing several generic concerns dealing with the storage of spare safety-related components. Component engineering personnel initiated a PEP review and formed a committee to evaluate station ,

compliance with the applicable sections of the American National Standard Institute (ANSI)

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N45.2.2, Packaging, Shipping, Receiving, Storage, and Handling of items for Nuclear

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Power Plants. Several procedures were revised to bring Limerick warehouse and storage practices into conformance with the ANSI standard. Site engineering completed an evaluation for a reduction of commitment in accordance with 10 CFR 50.54(a) and -  ;

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determined that the change in practices implemented did not decrease the level of commitment already stated in the UFSAR. The inspector reviewed the evaluation and-concurred with its conclusion. This item is closed.

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E1 Conduct of Engineering E1.1 N12B Instrument Nozzle Reoair 1 a. Insoection Scoce (37551)

The inspector reviewed engineering documents related to the cause and resolution of the through wall axial linear crack on the RPV N128 nozzle.

b. Observations and Findinas in preparation for an ISI nondestructive examination on the N12B safe end-to-pipe weld, DCA-294-E02-W1, leakage was observed from the safe end side of the weld. A through i wall crack was identified in the heat affected zone of the weld, approximately 1/4" from

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the toe of the fillet weld and 1/4" from top dead center.

The root cause investigation following the identification of the defect in the N12B 1 instrument nozzle was comprehensive, with engineering and maintenance personnel providing excellent support to the entire effort. The inspector reviewed the final report issued on the metallurgical evaluation of the crack (" White Paper on the Cause and Implications of the Indication on N128 Safe End," 2/28/97). The final metallurgical I evaluation indicated that the fracture surface of the axial crack was inter-granular. The j report also concluded that this flaw was in the presence of unusually high localized '

stresses due to a misalignment in the original fit-up.

I The inspector reviewed the work control and implementation process for the replacement l activity. The General Electric traveler outlining the steps of the activity was reviewed and i determined to have the appropriate level of detail and signatures of responsible personnel.

The pre-job reviews and briefs were extensive and addressed all technical aspects of the replacement. The work orders were specific and highly detailed so as to clearly define expectations throughout the job process. Overall, the replacement activity was performed at a high level of professionalism and technical proficiency. j The inspector reviewed Nonconformance Report (NCR) No. 97-00484, which addressed the operability of the instrument line, included the 10 CFR 50.59 review and outlined the replacement activity. The NCR was thorough and addressed all technical and regulatory concerns associated with this complex activity.

c. Conclusion i

The inspector determined that the engineering and maintenance organizations exhibited strong performance in the planning as well as the implementation of this complex !

replacement activity. Also, the depth of the root cause evaluation was good, exhibiting a questioning attitude by the engineering management, l

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j E1.2 Fuel Preparation Machine

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a. Scope (37551) <

i l l* On February.4, personnel identified that the Unit 2 fuel preparation machines did not have l the uptravel stops set properly, which resulted in a condition where two spent fuel - '

assemblies were raised closer to the water surface in the spent fuel pool than allowed by 1 the UFSAR. The inspectors reviewed the event by discussing it with appropriate

! personnel, inspected the preparation enachines, and reviewed the root causes and

} corrective actions.

i i. b. Observations and Findinas

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During fuel movements, early in the Unit 2 refueling outage, personnel on the refuel floor

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- noted that the spent fuel in the fuel preparation machines was closer to the water surface i than allowed by the UFSAR. The fuel preparation machines are mounted on the wall of the -

! fuel storage pool and are used for providing underwater fuel inspection capability. The UFSAR states that the fuel preparation machine carriage has an uptravel stop to prevent i raising irradiated fuel above the safe water shield level. . Additionally,- the UFSAR states l that mechanical stops prevent the carriage from lifting the irradiated fuel bundle or.

assembly to a height where water shielding is less than 7 feet. Test fixtures were installed

} on the fuel preparation machines prior to February 4, to perform fuelinspections, which

raised the fuel closer to the water surface by several inches. These test fixtures were a installed without the required.10 CFR 50.59 reviews being performed. Plant personnel l concluded that the required reviews were not performed due to inadequate review and .

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oversight of an activity conducted by a vendor, in part due to a lack of clarity regarding j which work group has responsibility for the associated work. Immediate corrective actions j taken were:- fuel preparation machine activities were stopped for the outage, the test l fixtures were removed, and the machines were administratively tagged out.

i j 10 CFR 50.59 states, in part, that the holder of a license authorizing operation of a  ;

l. utilization facility may make changes in the facility as described in the safety analysis -'

l report, without prior Commission approval, unless the proposed change involves a change

in the tschnical specifications incorporated in the license or an unreviewed safety question.

! Limerick Generating Station procedure LR-C-13,10 CFR 50.59 Reviews, Revision 6, the

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procedure implementing 10 CFR 50.59, requires,-in part, that. activities shall be evaluated j J . to' determine if the activity will or doet, make information in the SAR inaccurate or -

incomplete.

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In January 1997, a modification to the Unit 2 fuel preparation machines was made, which

, made the information in the SAR incomplete in that a test fixture was added, which raised i .the fuel closer to the water surface. This resulted in the raising of two spent fuel assemblies closer to the surface of the spent fuel pool than allowed by the UFSAR. This

activity was not evaluated to determine if the activity would make information in the SAR i inaccurate or incomplete. Failure to perform a 10 CFR 50.59 review as required is a

violation. (VIO 50-353/97-01-01)

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At the time the spent fuel was closest to the spent fuel pool water surface, health physics

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personnel measured dose rates at the water surface directly above the snent fuel; no readings above normal background were measured. An analysis performed by health physics (HP) personnel demonstrated that assuming the worst possible case, i.e. with spent fuel having the most severe history assumed in the UFSAR, and having been removed from the vessel at the earliest permissible point, dose rates for the assembly at the depth which was found on February 4, would have remained below the UFSAR design basis limit, c. Conclusions .

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The inspectors concluded that PECO Energy provided inadequate oversight and review of a vendor activity that resulted in operation of a refueling activity outside that assumed and reviewed in the UFSAR. For this instance, safety consequences were low.

i E1.3 Steam Floodina Damoers inocerable - Unit 2 a. Scoce (37551)

During the Unit 2 refueling outage, plant personnel identified that eight Unit 2 pairs, one Unit 1 pair, and several single back pressure dampers failed to actuate following a test signal. These dampers are required to mitigate the consequences of a postulated high energy pipe rupture. The inspectors reviewed the results of the testing, and discussed the causes and corrective actions with engineering personnel.

b. Observations and Firdinas On February 6, plant personnel identified that six Unit 2 steam flooding heating, ventilation, and air conditioning (HVAC) dampers would not function. The cause of the failures to actuate was determined to be excessive friction in the bushing located between the solenoid anu the damper linkage. The dampers were immediately repaired by cleaning and lightly lubricating the bushing and solenoid stem. Fourteen additional dampers were tested to address the potential generic implications of the failures; ten dampers failed to actuate during this testing. Based on these results, all Unit 2 dampers, all common control enclosure dampers, and all accessible Unit 1 dampers were tested;in all,102 damper pairs were tested, and only three Unit 1 pairs were not, due to the dampers being inaccessible

' with the plant operating. This additional testing identified eight pair and ten single damper failures on Unit 2, two single control enclosure failures, and one pair and five single failures on Unit 1. In all cases, the failures were the result of excessive friction in the bushing located between the solenoid and the damper linkage. Each failed damper was immediately repaired by cleaning and lightly lubricating the bushing and solenoid stem, and retesting satisfactorily prior to testing the next damper. Additionally, all dampers tested were properly cleaned and lubricated. All testing was completed prior to the restart of Unit 2 after the refueling outage. The inaccessible Unit 1 dampers were determined to be operable based on their maintenance history and engineering judgement, and will be tested at the next opportunity when plant conditions permit.

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An analysis of each of the failed dampers was performed, which concluded that operation occurred outside the design basis of the plant; this condition was properly reported. The analysis concluded that although, during a high energy line break event, conditions could exist exceeding the environmental qualification limits for certain components required to mitigate the consequences of the event, the components would have performed their

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safety function. This conclusion was based on the short duration of the elevated environmental conditions and environmental qualification analysis of the same equipment currently approved for Peach Bottom Atomic Power Station.

The root cause of the damper failures was determined to be the lack of a proper preventive maintenance (PM) activity to periodically cycle, clean, and lubricate the mechanisms.

Corrective actions included cycling, cleaning and lubricating all accessible dampers, with plans to address the inaccessible dampers at the first opportunity, and the appropriate PM activities for each damper will be developed including the appropriate frequency for actuation and scope of maintenance to be performed. The frequency and scope of PM activities to be performed will be further evaluated based on the results of future testing of the dampers. Additionally, appropriate responsible personnel will review the potential generic concerns.

Technical Specification 6.8.1 states in part, that written procedures shall be established, f

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implemented, and maintained covering the applicable procedures recommended in Appendix A of Regulatory Guide 1.33, Revision 2, February 1978. Appendix A, Procedures for Performing Maintenance, Regulatory Guide 1.33, Revision 2, February 1978, states in part, that PM schedules should be developed to specify lubrication schedules, and inspection of equipment.

Administrative Procedure A-C-028, Preventive Maintenance Program, Revision 0, written to comply with TS 6.8.1, requires that personnel ensure each PM task is scoped and performed. Failure to have a PM activity to clean and lubricate the steam flooding dampers constitutes a vio'ation of this requirement. This licensee-identified and corrected violation is a Non-Cited Violation, consistent with Section Vll.B.1 of the NRC Enforcement Policy.

c. Conclusions Actions taken to ensure that all the steam flooding dampers are operable were very good.

The testing of the dampers was carefully planned and systematically implemented so that the deficient dampers were correctly identified and corrective actions were immediately taken to restore these dampers to an operable status. A Non-Cited Violation resulted due to the f ailure to have a PM activity to clean and lubricate the steam flooding dampers, which resulted in operation outside the design basis of the plant.

l E8 Miscellaneous Engineering issues (90712)

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E8.1 (Closed) LER 1-96-019. Capability to Reiect the Electrical Load of an RHR Pomo Not Fully Verified This event was discussed in NRC Integrated Inspection Report 50-352/96-07, and 50-353/96-07; and Notice of Violation, and resulted in a non-cited violation. The LER met the

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requirements of 10 CFR 50.73, and the inspectors had no further questions regarding the event.

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E8.2 ! Closed) LER 1-97-001. Safetv-Related Loaic Circuits Not Fullv Tested During an engineering review in response to Generic Letter 96-01, Testing of Safety- ;

Related Logic Systems, the engineering staff identified three issues involving contacts

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within a EDG/4kV subsystem logic channel that had not been tested per the surveillance requirements of TS. The requirements had not been met since the original plant start-up.

The untested contact circuits involved the EDG output breaker trip while in test mode of operation, the offsite power voltage monitoring logic, and the 4kV load breaker connected position trip logic. An inadequate review of the logic circuits and misinterpretation of the i TS requirements was the cause of the event.

I Management promptly corrected this surveillance inadequacy, tested all of the contacts, and restored them to an operable status. The safety consequence of this event was minimal because all of the contacts, except one, operated correctly when tested. The inspector reviewed the corrective actions and found them to be adequate. The licensee-identified and corrected violation of Technical Specification Surveillance Requirements is a Non-Cited Violation, consistent with Section Vll.B.1 of the NRC Enforcement Policy.

E8.3 (Closed) LER 1-97-002. Failure to Imolement Technical Soecification Reauired Visual Insoections of Blind Flanaes i

On January 23,1997, a task team reviewing a recently modified primary containment penetration for the 10 CFR 50.54(f) review, determined that a number of blind flanges on both units had not been periodically inspected per the surveillance requirements of TSs. -

This condition existed since start-up of each unit and was caused by a misinterpretation of the surveillance requirements for blind flanges on containment penetrations.

Management promptly corrected this surveillance inadequacy, inspected all of the blind flanges, and restored them to an operable status. The safety consequence of this event was minimal because previous integrated Leak Rate Tests and Local Leak Rate Tests had proven the integrity of the primary containment. Had one of the flanges been removed or become degraded, station personnel would have identified the condition through unusual nitrogen make-up to the containment or inability to maintain appropriate containment pressure. The inspector reviewed the corrective actions and found them to be adequate.

The licensee-identified and corrected violation of Technical Specification Surveillance -

Requirements is a Non-Cited Violation, consistent with Section Vll.B.1 of the NRC Enforcement Policy.

E8.4 (Closed) LER 1-97-003. Dearaded Back Pressure Damoers Needed for Pioe Ruoture Mitiaation Result in Ooeration Outside Desian Basis This event is reviewed in section E1.3 of this inspection report. The LER met the requirements of 10 CFR 50.73, and the inspectors had no further questions regarding the event.

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4 E8.5 (Closed) LER 2-97-002. N128, 2" RPV Instrument Nozzle Safe End Leak This event is reviewed in section E1.1 of this inspection report. The LER met the requirements of 10 CFR 50.73, and the inspectors had no further questions regarding the event.

IV. Plant Suonort R1 Radiological Protection and Chemistry (RP&C) Controls R1.1 Unit 2 Refuelino Outaae Radiolonical Controls (External Exoosure Controls)

a. Inspection Scope (837501 The inspector selectively reviewed external exposure controls for the Unit 2 outage. The inspector reviewed numerous ongoing radiologically significant work activities including:

work activities within the drywell, diving operations within the suppression pool, main l steam isolation valve work activities, initial control rod drive removal, reactor water

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cleanup work activities, and main condenser work activities. The inspector selectively 1 reviewed radiation surveys including radiation and contamination surveys; use of dosimetry; and calibration and use of radiation survey instrumentation. I The inspector also reviewed the Radiological Protection Organization's efforts to reduce occupational radiation exposure to as low-as-is-reasonably-achievable (ALARA). The ;

inspector selectively reviewed implementation of pre-job ALARA plans including lessons

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learned, as appropriate, from previous outages. The inspector also reviewed the Unit 1 :

Health Physics Outage Report for the 1996 Unit 1 refueling outage. l

b. Observations and Findinos I

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PECO Energy provided overall effective applied radiological controls for ongoing radiological work activities. The inspector observed overall effective external exposure controls, including: high radiation area controls; very good control and posting and/or labeling, as appropriate, of contaminated areas and contaminated or radioactive materials; and overall very good controls for survey and removal of material from the radiological controlled areas.

The following matter regarding high radiation area controls was brought to the radiation I protection manager's (RPM) attention:

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Technical Specification 6.12.2 required that, if reasonable, an enclosure be constructed to prevent unauthorized access to those high radiation areas with radiation levels such that a major potion of the body could receive, in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, a dose greater than 1,000 millirem /hr (measured at 18 inches) (i.e., locked high radiation l areas). The inspector noted that a catwalk access way in the Unit 2 suppression I pool leading to an area under the drywell, was posted as a locked high radiation area and was provided with a flashing light to alert personnel. However, the inspector observed that it appeared that the gate could be bypassed and personnel

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could gain unauthorized access by climbing over the walkway hand rails. The {

inspector noted that a reasonable enclosure / gate could be constructed to prevent ;

such a bypass and that NRC Information Notice No. 88-79, " Misuse of Flashing !

Lights for High Radiation Area Controls," dated October 7,1988, provided details on the TS requirements for high radiation area controls and use of flashing {

lights /bemers. '

The RPWI subsequently indicated that an inaccessible hot spot on a pipe was the reason for the posting and that the area did not meet the radiation dose rate criteria to be considered a locked area. The area was conservatively posted as a locked ,

high radiation area. However, because of the potential transient nature of the hot l spot, the personnel constructed an enclosure to prevent unauthorized access. The i licensee also indicated there were no other similar type areas, but that high radiation area controls would be reviewed.

l The inspector's review of ongoing work activities indicated appropriate radiological surveys were made to support ongoing work activities. Radiation dosimetry was provided and properly worn, including additional dosimetry, as appropriate. PECO Energy continued to effectively implement use of its real time electronic dosimetry system to track and control radiation exposures. Alarm setpoints were adjusted downward in light of the observations !

identified during NRC Combined Integrated Inspection 50-352;353/96-01. Dosimetry i records were updated and maintained, including exposures sustained due to personnel contamination and/or airborne radioactivity.

The licensee performed a study of its gamma whole body friskers and determined that they could also serve as passive whole body counters with a minimum detectable activity on -

the order of 1 % of an annual limit of intake assuming a clearance time, of radioactivity from the body, of seven hours. The inspector noted that the licensee sustained 68 personnel contaminations during the outage as compared to the previous Unit 1 outage where 189 personnel contaminations were sustained, indicating significantly improved controls for personnel contamination prevention.

The licensee implemented pre-job ALARA plans, including self-assessment / areas for improvement and lessons learned outlined in the 1996 Unit 1 Health Physics Outage Report. Of particular note was the enhanced shielding efforts within the drywell and routine hydro-lazing of reactor vessel nozzles prior to inspections. Exposure goals were reasonable and accumulated radiation exposure was evaluated relative to pre-established values. The licensee met established ALARA goals for the outage. The licensee had <

estimated an outage aggregate exposure for the Unit 2 outage of 140 person-rem (based on expected TLD readings). The licensee sustained an outage exposure of 138.5 person-rem (based on TLD readings). The outage lasted from January 30, until approximately February 29.

The inspector noted an area for additional review relative to exposure reduction. Although estimated exposure was generally low, these observations indicated additional opportunitia for occupational dose reduction for a routinely performed outage task.

Specifically, the following observations were made:

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The inspector noted that, due to a short pendant for the CRD crane outside the -

drywell CRD hatch, personnel operating the crane were required to be in close proximity to the elevated dose rates associated with the spud and/or flange end of ,

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the CRD. Further, because of the small arm of the crane, personnel needed to be in close proximity to the CRD to push it over the contaminated area while the CRD was suspended in air in order to place it into the non-contaminated area. The

, licensee indicated these matters would be reviewed.

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The inspecar followed the initial CRD as it was being transported from the CRD hatch to the C".T.; storage area. The inspector noted the wagon-like handle to pull the CRD was missing from the CRD cart, requiring personnel to be in close proximity to the cart while transporting it to the CRD storage location. Further, the

, inspector noted that the ramp had not been installed at the doorway to the CRD storage area, and due to limited work space, personnel were required to install the

[ ramp in elevated radiation dose rates emancting from the CRD.

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Notwithstanding these observations, the average aDDrogate dose sustained for ,

j changeout of CRDs was generally low.

The inspector also noted, during review of removal of pipe whip restraints on the 295'

elevation of the drywell, that personnel were using manual wrenches to remove tne bolts

, from the restraints. The inspector questioned this activity relative to exposure reduct!on

purposes and the job supervisor indicated a high torque wrench had been tried, but malfunctioned. The inspector noted that, although the ALARA review for the task I

continued to be valid, improved tooling may result in dose reduction for this task. The licensee indicated this matter would be reviewed.

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PECO Energy provided special controls and training of personnel for those areas known to contain hard-to-detect radionuclides (e.g., Zn-65). The inspector also reviewed selected personnel skin contamination dose estimates and noted calculations to be reasonable.

The licensee was providing generally effective controls for suppression pool diving

operations, including provision of breathing air whose quality had been tested for use. The

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inspector brought the following observations to the licensee's attention:

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The procedure for control of underwater diving (HP-C-320, Revision 3) specified the

use of two independent pre-dive surveys (with different survey meters) to be

performed in dive locations. The inspector's review of survey data sheets identified only one survey meter was apparently used. Subsequent inspector discussions with radiation protection (RP) personnel covering dive activities indicated two survey meters were used, but summary radiation survey data was provided. The inspector indicated survey data sheets should reflect use of two survey meters in that, had the technicians who performed the surveys been unavailable, it would not have

been apparent that two properly calibrated radiation survey meters were used for

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RP personnel, while wearing potentially contaminated gloves, were observed to be handling the insides of dive helmets. An RP superviscr coached the technicians '

regarding handling of dive suits with potentially contaminated gloves.

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As specified by procedure, divers exit;ng the water were hosed off with clean  !

water. The procedure did not provide for survey (for dose estimation purposes) of divers as they were exiting to identif / any highly radioactive material (e.g., hot

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particles) that may adhere to their suits / equipment prior to its dislodge by water sprays. RP personnel at the dive location indicated no hot particles had been I identified during numerous routine surveys. The licensee indicated this matter l would be reviewed.

The inspector noted that on February 3, at about 10:00 a.m., a radiological engineer came !

upon a contractor worker in the upper elevations of the drywell whose electronic pocket ;

dosimeter was alarming. The RPM indicated the worker was apparently not able to hear the dosimeter because of the positioning of the body (in close proximity to a pipe) and the !

worker's use of a headset. The worker's electronic dosimeter was set to alarm at 200 i millirem, the radiation work permit (RWP) limit (RWP NO. 8, Revision 0) was 300 millirem, i and the worker received 309 millirem based on the electronic dosimeter reading. i i

The inspector noted that the individual was immediately removed from the drywell, and a  !

work suspension was implemented for the involved contractor work force and work ectivity (non-destructive examination of pipe nozzles). The licensee also issued an event ;

report for this matter (i.e., performance enhancement process report) and revised HP job l coverage standards to provide additional guidance to RP personnel for coverage of work in )

high noise environments or environments where personnel were tu wear head sets. The.

revised standards were issued to RP personnel via a group information notice (issued February 10,1997), and by training following the event.

The inspector noted that failure to adhere to the RWP limit was a violation of TS 6.11, which required that procedures for personnel radiation protection be adhered to for all operations involving personnel radiation exposure. Specifically, procedure A-C-100, i Revision 1, " Radiation Protection Program," required, in part, in Section 5.4.2, that written i radiological control instructions on RWPs be obeyed. The inspector noted that due to the inability to hear the dosimeter alarm, the RWP limits were exceeded.

The inspector reviewed this violation with respect to the criteria for exercise of discretion outlined in Section Vll.B.1 of the NRC Enforcement Policy.

Regarding the criteria for exercise of discretion, the inspector noted the following:

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The violation was identified by the licensee.

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The violation did not appear to be a violation that could reasonably be expected to have been prevented by the corrective actions for a previous violation or licensee finding that occurred within the past 2 years.

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Specific corrective actions were taken, as discussed above, to prevent recurrence.

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The violation did not appear to be willful. l l

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i The individual was preparing to leave the area. Consequently, no significant additional l exposure was expected.

. Based on the above, this licensee-identified and corrected violation is a Non-Cited

} Violation, consistent with Section Vll.B.1 of the NRC Enforcement Policy.

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! c. Conclusions

} PECO Energy established and implemented overall very good applied external radiological

} controls and procedures for the Unit 2 outage. Overall contamination controls were

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effective. The licensee exhibited, overall, a very good safety focus on outage radiological j controls as exemplified by the planning provided and the radiological controls implemented.

i Overall, exposure reduction initiatives observed in the field were effective. The licensee

promptly responded to indications of weaknesses in high radiation area controls.

No safety concerns were identified.

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j R1.2 Unit 2 Refuelino Outaae Radioloaical Controls (Internal Exoosure Controls)

f a. Insoection Scope (83750)

i j The inspector selectively reviewed internal exposure controls for the Unit 2 outage. The 1 inspector reviewed numerous ongoing radiologically significant work activities including -

I work activities within the drywell, diving operations within the suppression pool, outboard

!' main steam isolation valve work activities, initial control rod drive removal, reactor water-l cleanup work activities, and main condenser work activities. The inspector selectively <

reviewed airborne radioactivity surveys and calibration and use of airborne radioactivity j sampling instrumentation.

i The inspector reviewed records, discussed the program with cognizant personnel and l observed internal exposure control practices during tours of the radiological control area.

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{ b. Observations and Findinas

) The inspector noted airborne radioactivity sampling and analysis to be commensurate with

! the radiological hazard present, and that as of the end of the inspection, no individual had )

sustained any significant intake of airborne radioactivity. The licensee appropriately

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implemented use of engineering controls to reduce ambient and peak concentrations of -

i airborne radioactivity.

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Individuals were observed to properly use respiratory protective equipment with the l exception of the below discussed observation.

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During review of reactor water clean-up (RWCU) work activities on February 6, at about l l 1:30 p.m., the inspector observed an individual don a fullface respirator in preparation for j

entry into a portion of the RWCU area. The inspector noted that a HP technician taped the

, respirator to the hood in preparation for entry into the room. The inspector observed that )

the individual's respirator had tape wrapped around the filter / mask connection. Subsequent l 1  ;

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licensee and inspector review indicated that the individual had performed a self-repair of a cracked neck connector on the respirator. The inspector noted the repair was not in accordance with procedure HP-C-512, " Issue and Control of Respiratory Protection Equipment," which required, in Section 5, that workers inspect each respirator prior to use and immediately report equipment malfunctions to health physics. The individual did not report the malfunction. This matter was considered a violation of TS 6.11, which required that procedures for personnel radiation protection be adhered to for all operations involving personnel radiation exposure.

When identified to RP supervisory personnel the licensee took the following actions.

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The worker was counseled and was provided another respirator that was determined to be in proper mechanical condition.

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The technician who taped the respirator to the worker's hood was coached on the event and the need to have a questioning attitude for problems.

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The licensee initiated a safety stand down for all jobs requiring use of respirators to discuss the event.

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All respirators available for issuance were inspected. No other similar problems were identified.

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All control point supervisors were informed of the event and instructed to inform workers of procedure requirements.

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The evaluation indicated that, notwithstanding the cracked and taped respirator, the worker had performed a satisfactory negative pressure check on the respirator.

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The licensee performed an exposure assessment assuming that the v arker wore a malfunctioning respirator into his work area. Based on air sample activity measured, no significant resonnel exposure would have occurred.

Based on the above, the inspector reviewed this matter relative to Section IV of the ,

" General Statement of Policy and Procedure for NRC Enforcement Actions," (60 FR 34381; June 30,1995). The inspector concluded that this failure constitutes a violation of minor safety significance and is being treated as a non-cited violation, consistent with Section IV of the enforcement policy.

c. Conclusions PECO Energy implemented an effective internal exposure control program.

No safety concerns were identified.

R2 Status of RP&C Facilities and Equipment a. Insocction Scooe (83750)

The inspector selectively reviewed the calibration and testing of various radiation survey instrumentation including personnel contamination monitors, portable radiation survey instrumentation, and refueling floor area radiation monitors (i.e,. installed criticality monitors and supplemental area radiation monitors used to support reactor cavity work activities). The review was with respect to applicable national standards and TSs.

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b. Observations and Findinas The inspector's selective review indicated radiological monitoring instrumentation reviewed, including air sampling instrumentation, was calibrated and periodically tested as appropriate. Further, the refueling floor area radiation monitors were calibrated and tested in accordance with TS requirements. Supplemental area radiation monitors, ringing the reactor cavity, were noted to be calibrated and periodically checked for operability.

The inspector's tours and discussions with the radiation control personnel indicated no plant areas had become unusable as a result of operational occurrences or spills. The inspector noted overall station housekeeping to be very good in addition, due to effective identification and repair of valves, the drywell exhibited generally very low contamination levels, which supported ease of access and minimal use of extensive protective clothing.

c. Conclusions Radiation monitoring instrumentation was calibrated and tested, as appropriate. Installed area radiation monitors were calibrated and tested in accordance with TS requirements.

Supplemental area radiation monitors were calibrated and periodically tested for operability.

No safety concerns or violations were identified.

R4 RP&C Staff Knowledge and Performance in RP&C a. Insoection Scone (83750)

The inspector evaluated knowledge and performance of radiological controls staff and radiation workers during ongoing work activities including drywell work, diving operations within the suppression pool, outboard main steam isolation valve work activities, initial control rod drive removal activities, reactor water cleanup work activities, and main I condenser work activities. The inspector interviewed radiological controls staff and radiation workers to evaluate their overall knowledge of radiological work conditions and radiological controls. I b. Observations and Findinas l

Radiological controls personnel and radiation workers exhibited, overall, a very good knowledge of radiological conditions and controls at their work locations. Overall, radiological controls personnel and radiation worker performance was very good. i

c. Conclusions Overall, radiological controls personnel and worker performance was very good.

No safety concerns or violations were identified.

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R5 RP&C Staff Training and Qualification in RP&C a. Inspection Scone (837501 The inspector selectively reviewed the training and qualifications of radiological controls staff and radiation workers. The review was against criteria contained within program procedures,10 CFR 19.12 and 10 CFR 50.120. The inspector selected contractor RP technicians observed to be providing radiological controls for radiologically significant work activities and reviewed their training and qualification records. The inspector also reviewed ,

the status of training of selected radiation workers.

b. Observations and Findinas PECO Energy upgraded the general employee training (GET) program to increase practical factors training. The GET training was increased from 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. Additional training was included on contamination controls including use of contamination monitoring equipment. Mockups were used to train personnel on the radiation work permit program

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and the proper use of the electronic dosimetry system. The licensee implemented a separate 8-hour course for contractor personnel. Contractor radiological controls personnel were provided training and qualifications for their assigned duties in accordance with program procedures. The licensee effectively used control point manuals and radiological job / task guides to instruct HP technicians in expected radiological hazards and job control measures.

The following observation was brought to the licensee's attention:

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Radiological survey information. posted near the Unit 2 drywell entrance, was well above head level and difficult to read. The licensee initiated a review of this matter, i

c. Conclusions PECO Energy provided appropriate training and qualification of contractor radiological controls personnel and radiation workers.

No safety concerns or violations were identified.

R6 RP&C Organization and Administration a. Inspection Scoce (83750)

The inspector selectively reviewed recent changes in the RP organization. The inspector also reviewed the radiological controls organization established to support Unit 2 outage activities. The review was against criteria contained in TSs and the UFSAR.

b. Observations and Findinas t

There were no sigriificant changes within the radiological controls organization since the previous inspection in this area (Reference NRC Combined Inspection No. 50-352;50-

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353/96-09, dateo January 9,1997). No apparent staffing deficiencies were noted during independent in-field observations of work activities including observation of backshift activities. The radiological controls organization was appropriately defined. The inspector noted administrative controls had been established for work activities of a recently hired radiological controls engineer pending that individual's training and qualification. l c. Conclusions PECO Energy established and implemented an overall effective organization to support the Unit 2 outage. Administrative controls were established for individuals not fully qualified.

R7 Quality Assurance in Radiological Protection and Chemistry Activities (83750)

a. Inspection Scooe (83750)

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The inspector selectively reviewed audits, assessments, and surveillance of the radiological I controls program. The review was against criteria contained in TS 6.5.

I b. Observations and Findinas The licensee augmented the staff to review and audit outage activities. The surveillances I of ongoing activities were performance-based. Appropriately qualified auditors were used j to perform the audits, surveillances, and assessments.

As part of the inspection, the inspector reviewed act5ns on self-identified events. The following event was reviewed:

> On February 3, at about 9:00 a.m., six individuals entered the radiological controlled area'- l (RCA) via the chemistry lab without logging into the realtime dose tracking system (RADOS) and wearing an electronic dosimeter as required by station RP procedures and postings on the RCA access door from the chemistry lab area. The posting on the access j door indicated electronic dosimeter required. The individuals had been provided access by ;

a recently hired staff chemist who opened a key locked door to grant access. The error {

was identified when the individuals were performing personnel contamination monitoring upon exit and recognized that they did not have their electronic dosimeters. The individuals were escorted out of the area and the licensee counseled the individuals regardirig the event.

A dose assessment and exposure evaluation report was completed for each individual which indicated no radiation exposure was received and no significant exposure was likely.

The individuals were not contaminated. The staff chemist was counseled regarding the need to be diligent to ensure that persons entering the RCA wore proper dosimetry. The licensee reviewed the signs posted at the area and enhanced the posting to highlight the !

need to self-check for proper dosimetry prior to entry to the RCA. Additional signs were placed on the wall of the access point to preclude inability to see signs when the access door was opened. The licensee enhanced postings to increase awareness that an RCA boundary is to be crossed and proper dosimetry is needed. The licensee also initiated action to provide a night order to chemistry staff to reinforce the requirements for entry

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into the RCA through the chemistry access point. The licensee also initiated a review of the general employee training program.  !

The inspector noted that failure of the six individuals to adhere to the door posting was an ,

apparent violation of TS 6.11, which requires that procedures for personnel radiation l protection be adhered to for all operations involving personnel radiation exposure. ;

Specifically, procedure A-C-100, Revision 1, " Radiation Protection Program," requires, in !

part, in Section 5.4.2 that written radiological control instructions be obeyed. The +

inspector noted that the workers did not adhere to the posting on the door and obtain an ;

electronic dosimeter.

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The inspector reviewed this violation with respect to the criteria for exercise of discretion :

outlined in Section Vil of the NHC Enforcement Policy.

Regarding the criteria for exercise of discretion, the inspector noted the following: :

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The violation was id6atified by the licensee.

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The violation did not appear to be a violation that could reasonably be expected to have been prevented by the corrective actions for a previous violation or licensee -

finding that occurred within the past 2 years.

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Specific corrective actions were taken, as discussed above, to prevent recurrence.

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The violation did not appear to be willful.

Based on the above, this licensee-identified and corrected violation is a Non-Cited Violation, consistent with Section Vil.B.1 of the NRC Enforcement Policy.

c. Conclusion Overall, surveillance and audits were of good quality. The licensee took appropriate actions on self-identified issues. .

No safety concerns were identified. I R8 Miscellaneous issues R8.1 (Closed) Unresolved item 50-352, 353/96-04-02. Verification of UFSAR.

Descriotions a. Insoection Scope (837501 A recent discovery of a licensee operating their facility in a manner contiary to the UFSAR description highlighted the need for a special focused review that compares plant practices, procedures and/or parameters to the UFSAR description. While performing the inspections discussed in this report, the inspectors reviewed the applicable portions of the UFSAR that related to the areas inspected. Identified changes were reviewed relative to 10 CFR Parts 50.59 and 50.71(e) requirements regarding UFSAR changes.

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b. Observations and Findinas During NRC Combined inspection No. 50-352/96-04, 50-353/96-04 (conducted May 7, 1996, through July 1,1996) the inspector reviewed the conformance of the radioactive waste storage and processing facilities relative to descriptions within the UFSAR. The actions on the identified discrepancies were discussed in NRC Combined Inspection Report No. 50-352/353/96-09, dated January 9,1997.

During the inspection, the inspector met with cognizant personnel and discussed the j actions taken on the identified discrepancies as described below. l

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UFSAR Section 11.4.2 did not reflect the current operating practices relative to the equipment and floor drain filters and fuel pool filter /demineralizer. l

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l The inspector found that centrifuges were not used for dewatering waste sludge l tank contents and had not been used for approximately 5 years. Plant personnel l initiated an engineering change request to abandon the equipment. An action '

request was issued in December 1994 to abandon the equipment.

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UFSAR Section 11.4.2 did not reflect the current operating practices relative to use I of the intermediate spent resin tanks.

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The inspector identified that the intermediate spent resin tanks have not been used for collection of waste before pumping it to the waste sludge tank. The waste has been pumped directly to the waste sludge tank even though the plant had passed its initial operation phase.

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UFSAR Section 11.2.2.5 did not reflect the current operating practices relative to-operation of the evaporator and the other components of the system.

The inspector identified that although the initial phase of plant operations had passed, the evaporator was only partially installed and had not been used. Further, piping had not been connected and the other components of the system had not been used.

Relative to the above matters, the inspector's review indicated that personnel initiated a UFSAR engineering change request (ECR) on May 21,1996, to update the UFSAR. The ECR was completed on October 16,1996, to reflect the current operating practices relative to operation of the evaporator and the other components of the system.

Further, in December 1995, PECO Energy determined that UFSAR Section 12.2.1.7,

" Stored Radioactivity," did not specifically authorize or discuss storage of radioactive materialin the yard areas around the station (i.e., outside process buildings). PECO Energy had stored, and continued to store, radioactive materialin selected portions of the yard areas. The areas included the new fuel storage area and the area directly in front of the radwaste enclosure truck bay. A 10 CFR 50.59 review and UFSAR change request were initiated. The 10 CFR 50.59 and UFSAR change request was approved by the PORC and the Plant Manager in May 1996. The safety evaluation concluded that no unreviewed

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safety question existed and that station procedures provided appropriate control for the stored materials.

The inspector noted that each of the above discrepancies existed for an extended time duration and that the UFSAR was not updated in accordance with 10 CFR 50.71 (e)(4) to reflect the changes.

The inspector's review indicated the discrepancies were identified by the licensee, the licensee performed appropriate safety evaluations once the matters were identified. The safety evaluations did not identify any significant safety concerns associated with the minor discrepancies, in accordance with 10 CFR 50.71(e)(4) and 10 CFR 50.4 (b)(6), on January 31, the licensee submitted Revision 6 of the UFSAR to update it to reflect current i

practices for storage and staging of low level radioactive / contaminated materialin the yard areas of the station and operation of radioactive waste systems.

Based on the above, the inspector reviewed this matter relative to Section IV of the

" General Statement of Policy and Procedure for NRC Enforcement Actions," (60 FR 34381; June 30,1995). The inspector concluded that the above matter constitutes a violation of minor safety significance and is being treated as a non-cited violation, consistent with Section IV of the enforcement policy.

c. Conclusions 1

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l The licensee updated the UFSAR to reflect current practices for storage and staging of low level radioactive / contaminated material in the yard areas of the station and operation of radioactive waste systems. The unresolved item (URI 50-352,353/96-04-02) associated with updating of the UFSAR in accordance with 10 CFR 50.71 (e) is closed.

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R8.2 (Closed) VIO 352. 353/96-09-03. Failure to orovide comotete descriotion of the

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orocosed manner and conditions of waste disposal (relative to the presence of the concrete slab (92904)

This violation concerned an instance where PECO Energy failed to provide a complete and accurate description of the proposed manner and conditions of waste disposal and an analysis and evaluation of pertinent information on the nature of the environment (relative to the presence of an undisclosed concrete slab). This inaccuracy was material, in that, the presence of the concrete storage pad, when identified, required further technical review by the NRC staff to determine if any new consequence was introduced. Corrective actions included: supplemental information was provided to the NRC by letter with the revised calculation based on the presence of the concrete stab; a memorandum was issued to appropriate Limerick and PECO Nuclear Headquarters personnel outlining the basis for the violation and reinforcing the requirement to provide complete and accurate information to the NRC in accordance with 10 CFR 50.9; the oppropriate procedures will be revised to reinforce the requirements of 10 CFR 50.9; and a Traini1g Bulletin will be distributed to appropriate personnel describing the violation, reinforc;ng the requirement to provide ccc..plete and accurate information to the NRC, and indicating that information provided to the NRC should include not only information required by any atsociated regulations or regulatory requirements, but also information such as engineenng judgements or assumptions used to support the development of the conclusioas.

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R8.3 (Uodated) Unresolved item 50-352, 353/96-10-03, Loss of Control of Master Kevs a. Inspection Scoce (83750)

l The inspector reviewed the actions relative to the identification of uncontrolled locked high radiation area keys identified by the licensee.

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b. Observations and Findinas TS 6.12.2 requires, in part, that areas accessible to personnel with radiation levels such that a major portion of the body could receive, in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, a dose greater than 1000 millirems shall be provided with locked doors and the keys shall be maintained under the administrative control of shift supervision on duty and/or HP supervision.

On January 30, the RPM became aware that master keys, that could be used to open locked high radiation area doors at the Limerick and Peach Bottom stations, were improperly controlled and in the possession of unauthorized personnel between mid-1993 and November 1996. The keys had been improperly made and distributed to fire protection personnel by the corporate locksmith. In addition, the RPM was unaware of the existence of the master keys maintained by the corporate locksmith. Consequently, the inspector concluded that the administrative key control program for locked high radiation.

areas was not effective.

The inspector noted that the keys possessed by the fire protection personnel were unauthorized and were not under the administrative control of shift supervision on duty and/or HP supervision. As a result of the identification of the unauthorized keys, the licensee took the following actions:

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The security access authorization was removed for the individuals known to possess the keys.

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The locksmith who provided the master keys was disciplined.

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The licensee initiated tours every 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of locked high radiation area doors to review for unauthorized entries. Based on the belief that all keys were recovered, on the evening of January 30, the licensee suspended the tours. On January 31, the licensee became aware that additional keys were potentially outstanding and re-initiated the tours every two hours to check on locked high radiation area access doors.

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Due to questions on the number of outstanding keys, the RPM initiated actions to change out all high radiation area access door lock sets with special proprietary lock sets. The installation of the new key lock sets was completed on February 2, and the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> tours were suspended.

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The licensee initiated reviews for unplanned / unexplained radiation exposures for the affected areas using a combination of key card data and knowledge of the work areas of individuals known to posses the unauthorized keys. No unplanned or

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unusual exposures were noted based on reviews back through January 1994. The l licensee was continuing to review dosimetry records for potential unplanned / unexplained radiation exposures of personnel.

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The licensee placed the new keys (master keys and lock set changing keys) under I the administrative control of the RPM. )

l In addition, the licensee continued the investigation to determine the circumstances

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surrounding the matter. The inspector will review the investigation of this matter when it i is complete, including the security aspects of this issue, since the keys also had the potential to allow unauthorized access to plant security vital areas. The adequacy and effectiveness of control of keys to locked high radiation areas relative to the requirements outlined in TS 6.1'2.2 will remain unresolved pending review of the investigation. (URI 50-352/353 96-10-04) 4 c. Conclusions I

The licensee demonstrated a very good safety focus in response to the high radiation area key event.

R8.4 Plant Tour Observations During the inspection, the inspector made various tours of the radiological controlled area.

The inspector's review indicated generally very good housekeeping.

F1 CONTROL OF FIRE PROTECTION ACTIVITIES F1.1 Fire Risk Evolutions a. Insoection Scope (64704)

The inspectors reviewed the established administrative processes for controlling and evaluating fire hazards including limiting the interaction of combustible and flammable materials with ignition sources. This review was conducted to verify that adequate guidance and proper authorization requirements existed for identifying and limiting fire risk.

b. Observations and Findinas The inspectors found that the administrative process for controlling ignition sources included the use of a permit system for authorization to perform hotwork activities. This authorization was granted by the work group supervisor in the field overseeing the job task. The inspector found that the process began with the job planner making an initial determination of whether other areas or equipment would be affected by the hotwork activity, and subsequently, whether the fire protection group, Industrial Risk Management (IRM), should be involved. The inspector found that the majority of job tasks involving hotwork were planned several weeks and months prior to work beginning and noted that such work areas in the plant could change significantly within hours.

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in addition, the inspector found that IRM does not nor is it procedurally required to perform independent fire protection reviews to ensure procedural adherence for combustibles, hotwork, or assuring the fire protection program is effective from an oversight perspective.

Also, no procedure existed to direct scheduled plant walkdowns by the fire protection staff. Instead, reliance was placed on workers and management housekeeping tours to I provide assurance that the fire protection program was maintained appropriately. l l

The inspector determined that the hotwork process implemented at Limerick failed to have '

independent fire protection reviews by iRM who are knowledgeable of fire protection. This weakness may have contributed to a recent fire that occurred on February 11,1997, resulting from hotwork. Specifically, four separate work group supervisors f ailed to i identify combustibles within a radius of 35 ft of the hotwork activity. Because the job was I planned on May 14,1996, a determination was made that no additional measures were l necessary based on the plant conditions at (nat time. This included IRM's involvement. In addition, as documented in Performance Enhancement Program (PEP) issue No. 10006631, the job foreman assigned a firewatch for this job without verifying the firewatch's l qualifications. Further PECO review found that the firewatch assigned was not aware of his responsibilities, c. Conclusions The inspector identified several weaknesses with PECO's process for controlling hotwork l activities, including little oversight from the IRM group.

F2 STATUS OF FIRE PROTECTION FACILITIES AND EQUIPMENT l l

F2.2 Facility Tour a. Inspection Scoce (64704)

The inspector toured accessible vital and non-vital areas of the site and inspected the fire protection water suppression systems, tire pumps, piping and distribution systems, post indicator valves, contents of indoor fire protection storage facilities and outdoor hose houses, emergency lighting patterns for access and egress routes for selected safety-related plant cquipment areas, and the condition of fire brigade equipment.

b. b Ob.servations and Findinas The insrector found that fire protection equipment material conditions were good and fireloading was properly maintained in those areas selected for review. Fire brigade members' protective clothing and gear were found in good condition and were organized satisfactorily in the brigade room and fire brigade cabinets throughout the plants.

The inspector determined that the housekeeping conditions in plant areas containing safety-related equipment or components were very good No examples were noted involving the improper control of combustibles, improper storage of radioactive materials, or improper control of hazardous chemicals. However, the inspector found that no-smoking controls in the Unit 1 emergency diesel generator (EDG) corridor and along the

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Unit 1 EDG "A" exterior wall areas need to be improved. Several crushed cigarette butts were found outside the Unit 1 EDG structure adjacent to the "A" EDC cubicle and a butt was found in a plastic-lined trash receptacle inside the Unit 1 EDG corridor. Both areas were no-smoking areas.

Regarding emergency lights, the inspector found that recent actions have been implemented to improve their effectiveness. These actions included the creation of a database in February for tracking recurring light failures and performance monitoring of parameters, including light location, area temperature, and charging voltage. In addition, the licensee participated in the Electric Power Research Institute (EPRI)-sponsored conductance testing program established to eliminate destructive drain down testing and increase component reliability of emergency lights by performing ohmic measurements instead. Another recent acuon taken by PECO since February included identifying each emergency light unit associated with safe shutdown with a placard to enhance assurance of proper alignment and no impairment of the lighting lamp heads. The inspector determined that the initiatives taken by PECO were good considering the longstanding history of emergency light failures at the plants, as evidenced by numerous corrective maintenance activities for emergency lights. In addition, the inspector found that the implemented "Fix-It-Now (FIN)" program for correcting emergency light failures within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for a maintenance turnaround time reflected an appropriate heightened awareness of the importance of emergency hghts. The inspector noted that the licensee intends to provide position identification marks on emergency light lamp heads to provide for easy position indication validation without the need for a drawing in the field. No emergency light deficiencies were identified in the plants during the tour, c. Conclusions The inspectors concluded that fire protection equipment conditions were good and housekeeping was very good. No-smoking controls need to be improved in the vicinity of the Unit 1 EDGs. Recent initiatives taken by PECO to improve emergency lighting were good considering the longstanding history of emergency light failures in the plants.

F3 FIRE PROTECTION PROCEDURES AND DOCUMENTATION F3.1 Procedure and Documentation Review a. Inspection Scope (64704)

The inspector reviewed a sample of fire protection procedures, completed surveillances, and PEP issues that capture fire protection issues for appropriate resolution.

b. Observations and Findinos The inspector found that the licensee had initiated a Project Plan to improve fire protection surveillances. Action Reauest A1078758 assigned the task to review the AG-CG- type surveillances for fire protection equipment and resulted from a self-assessment performed by PECO. IRM plans to transfer these surveillances to other departments for performance to enable IRM to perform better performance monitoring of systems and oversight of the

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38 fire protection program. The inspector found that this action request did not include the recurring task (RT-) type fire protection procedures that are important to an effective fire protection program. Recurring task procedures included operability tests for sprinklers, foam system, fire hose inspections, and back-up diesel fire pump flow tests.

The inspector noted that administrative procedures governing fire protection did not clearly define responsibilities within the IRM group. The licensee stated that it would consider addressing this issue in the Project Plan. Problem identification by the licensee was good via the PEP process. j The inspector found that pre-fire plans used by fire brigade members needed revision to reduce verbiage and enhance the drawings by removing unnecessary information. During a brigade drill observed by the inspector, as discussed in report section F4.1, the brigade l staff stated that the pre-fire plans were too cumbersome for timely use of the drawings. I Additionally, the inspector noted that procedure AG-CG-12.1, Revision 2, " Actions For Fire i Protection impairments," was ambiguous, it allowed continuous firewatches to assist in I providing work activities. The inspector discussed this statement with IRM personnel and i verified that firewatches were not allowed to perform work that would prohibit prompt l extinguishment of fires that might occur. '

c. Conclusions The inspector concluded that the licensee's efforts for improving the fire protection program procedures were appropriate. Good problem identification by the licensee was noted.

F4 FIRE PROTECTION STAFF KNOWLEDGE AND PERFORMANCE F4.2 Fire Briaade Drills a. Inspection Scoce (64704)

The inspector observed an unannounced fire drill to evaluate the fire brigade's effectiveness and understanding of fire attack strategies. The drill was conducted to demonstrate the following:

• an understanding of the fire attack strategy; e the ability to assess the fire properly;

• an awareness of vital equipment in the area; e effective communication with other fire brigade members; and

• an awareness of additional hazards in the fire area.

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b. Observations and Findinas The inspector observed a fire drill on March 12,1997. A Class A fire was simulated in the Snubber Rebuild Facility. The inspector determined, based on drill observations and post-drill discussions with respondin0 brigade members, that the performance of the drill participants was acceptable. However, the fire brigade was over-cautious in applying extinguishant on the fire causing a longer delay than expected by the inspector. This determination of acceptability was based on the following:

o use of an appropriate suppressant type on the fire;

• command and control demonstrated by the fire brigade leader;

• teamwork displayed by fire brigade members; and e communications among brigade members.

The inspector found the quality of the critique following the drill effective for providing constructive feedback to the brigade regarding individual performance. Comments made by the drill critiquer/ evaluator were accurate and straightforward, c. Conclusion The inspector determined that the performance of the fire brigade during the drill was acceptable. The critique oy the fire protection engineer was accurate and straightforward.

F5 FIRE PROTECTION STAFF TRAINING AND QUALIFICATION F5.1 Fire Briaade a. Insoection Scope (64704)

The inspector reviewed the program requirements, medical approvals, and training provided for fire brigade members. Completed training records of selected personnel were reviewed to verify their qualification for duty.

b. Observations and Findinas The inspector verified that seven fire brigade members selected for review had successfully completed the required training courses, drills, and respirator training and passed their annual medical physicals. The inspector found that the responsibility for assuring all fire brigade members were qualified was shared between the Operations and IRM departments.

Discussions the inspector held with the responsible individuals revealed that neither individual was aware of the complete expectations for this ta;k In addition, the database shared by both departments was incapable of presenting necescary information for brigade members other than for the current year. Although the tracking system was poor for tracking fire brigade member qualification, no deficiencies were identified where a non-qualified member was assigned brigade duty. The inspector concluded that the lack of well-defined responsibilities for and tracking of fire brigade members' qualification was a program weakness.

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c. Conclusion i i

The inspector concluded that the licensee's process for tracking satisfactory completion of fire brigade member's program requirements was weak. No deficiencies were identified where a non-qualified member was assigned brigade duty.

F6 FIRE PROTECTION ORGANIZATION AND ADMINISTRATION F6.1 Program Reviews a. Inspection Scoce (64704)

The inspector performed a review of PECO's established performance indicators, process for resolving identified issues, and resolution of identified deficiencies.

b. Observations and Findinos I l

The inspector found that the licensee had only one specific performance indicator used to assess the fire protection program. This indicator trended the corrective maintenance I backlog. The IRM manager stated that they were considering additional performance j indicators at the time of this inspection, but a decision had not been finalized. The licensee l further stated that the Fire Protection Council, established on February 11,1997, was l created to improve the fire protection program by establishing a process for resolving l identified issues and, when fully implemented, will include the following actions: -

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e identification of strengths and weaknesses including those that require ,

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management sponsorship I

e exchange of information from several avenues including other plants, industry i forums, NRC, and the Quality Assurance department; e application of lessons learned between organizations; and e identification of consistent approaches in resolving regulatory issues. l Although the indication used to assess program performance was minimal, the inspector noted that the licensee appropriately had assigned a PEP issue to evaluate the collection of fire protection deficiencies being reported under the PEP program.

c. Conclusions The inspector concluded that PECO's performance measures for assessing the effectiveness of the fire protection program were narrowly focused. Good initiatives were taken recently instituted to identify and resolve program issues, however, their effectiveness could not be assessed because the Fire Protection Council and the above actions were very recent activities.

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F7 QUALITY ASSURANCEIN FIRE PROTECTION ACTIVITIES F7.1 Audits and Surveillances a. Inspection Scope (64704)

The inspector reviewed the three most recent audits completed to satisfy the technical specification requirements and 21 surveillances that evaluated the effectiveness of fire protection measures, equipment, program implementation, and problem identification and resolution.

b. Observations and Findinas The inspector found that audits:

e demonstrated good problem identification; I e had been appropriately completed,

• were effectively conducted, and l

• clearly communicated findings in reports.

c. Conclusion The inspector concluded that QA audits and surveillances were focused appropriately and verified selected fire program attributes for compliance with program requirements. Good problem identification was demonstrated for which resolution of issues was found to be tracked appropriately for closure.

F8 Miscellaneous Fire Protection issues (90712)

F8.1 (Undate) Unresolved item No. 50-352, 353/96-06-01 reaardina Inadeauate Emeraency Liahtina and Missina Safe Shutdown Plant Eauioment and (Closed) LER 50-352/96-015, Failure to Maintain Eouioment Needed to Assure Fire Safe Shutdown Caoability This issue involved the licensee's failure to prestage a 150 ft electrical jumper used as support equipment to assure safe shutdown of the plant in the event of a fire. This cable l would be used to re-energize the controls for the automatic depressurization system valves for depressurization control for the reactor during shutdown following a fire in the remote shutdown panel room. This necessary cable was not staged since August 11,1989, when Unit 2 began operations, contrary to operating license condition 2.C.(3) for Units 1 and 2.

In addition, on January 9,1995, revisions were made to special event procedures 8-2 and 8-4 that provided more simple pathways for operators to use for the installation of the two electiical jumpers needed for safe shutdown. However, these new pathways were found not to be illuminated by 8-hour emergency lighting, as required by UFSAR sections 9A.3.1.2 and 9A.3.2.2. Corrective actions implemented by the licensee included the f abrication and placement of the second necessary electrical jumper and resolution of the

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emergency lighting issue. PECo performed an engineering review of the fire safe shutdown analysis to determine the equipment and procedure required to be available and revised special event procedures for clarity, if the fire of interest had occurred, there were portable lights available for use in the plant.

One of the two required jumpers was pre-staged and could have been used on either unit as needed until a second jumper could be fabricated. The materials necessary to make the second jumper were available at the site. Therefore, missing equipment would have had only a smallimpact on the plant's ability to reach the cold shutdown condition. However, j the Limerick licenses require all equipment and procedures necessary to reach and maintain !

the cold shutdown condition following analyzed fires to be pre-staged. Therefore, the l failure to prestage the necessary safe shutdown support equipment including emergency lighting for all operator pathways used to implement special event procedures for a fire in the remote shutdown panel room is an apparent violation of the facility operating license )

and UFSAR. Other discrepancies that affect the ability to reach and maintain fire safe

. shutdown conditions have been recently reported in LERs 50-352/96-012 and 021 and  !

NRC Inspection Report 50-352/96-10. This issue will remain unresolved pending further l NRC review of the collective impact of these recently reported fire protection concerns.

F8.2 (Closed) LER 1-96-023 Fire Protection System Surveillance Tests not Performed Due l

to Personal Error i l

This event was reviewed in NRC Integrated Inspection Report 50-352/96-07, and 50- i 353/96-07; and Notice of Violation, and resulted in an Unresolved item. The LER met the i requirements of 10 CFR 50.73, and the inspectors had no further questions regarding the j event.  ;

F8.3 Documents Reviewed ,

A list of fire protection documents reviewed follows:

FIRE PROTECTION DOCUMENTS REVIEWED Procedures A-C-920, Rev. O Nuclear Generation Group Fire Protection

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Program RT-6-078-322-0, Rev. 2 Appendix R Mobile D/G Operability Test RT-6-000-900-0, Rev. 6 Inspection of Safe Shutdown Equipment AG-CG-012.02, Rev.1 Control of Combustible and Flammable Materials AG-CG-012, Rev. O Hot Work Guideline A-C-030, Rev. O Plant Material Condition and Housekeeping Controls G 0000G 71 (Action Request) Inventory of Fire Brigade Equipment NE-C-250, Rev.1 Fire Protection Review NE-C-250-2, Rev. O Fire Protection Review Checklist NE-C-250-3, Rev.1 Maximum Allowed incremental Combustible Load increase ST-7-022-953-0, Rev. 7 Technical Requirement Hose Cart Visual Inspection

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AG-CG-12.2, Rev.1 Control of Combustible and Flammable Materials AG-CG-12.1, Rev. 2 Actions for Fire Protection impairments ST-7-022-551-0, Rev. 4 Fire Drill RT-7-108-300-2, Rw. 2 Safe Shutdown Eight (8) Hour Self-Contained Battery Pack Discharge Test (4/15/94,7/1/94, 6/14/96)

RT-6-108 300-1, Rev. 4 Safe Shutdown Eight (8) Hour Self-Contained Battery Pack Operation Verification (11/26/95, 3/10/96, 8/13/96)

SE-8, Rev.17 Fire 1

QA Surveillances LSR-97-0009 (1/8-15/97) LSR-95-0160 (3/22/95) I LSR-96-0019 (1/23/96) LSR-9F,-0216 (5/8/95) I LSR-96-0155 (3/8-14/96) LSR 95-0255 (6/15/95)

LSR-96-0218 (6/17/96) LSR-95-0316 (8/10/95)

LSR-96-0183 (5/7/96) LSP.-95-0328 (8/22/95)

LSR-96-0219 (6/17/96) LSR-95-0357 (9/19/95)

LSR-96-0222 (6/17/96) LSR-94-0089 (4/13/94)

LSR-95-0043 (1/27/95) LSR-94-0064 (3/8/94) i LSR-95-0065 (2/3/95) LSR-94-0060 (3/3-4/94)

LSR-95-0080 (2/10/95) LSR-94-0041 (2/17/94)

LSR-95-0124 (2/28/95)

Audits A0839320 (9/9/94): LGS Fire Protection Plan & Independent Fire Protection, Loss Prevention Inspection and

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Assessment A0938345 (8/8/95): Fire Protection Loss Prevention Program LAR-96-007 (10/2/96): LGS Triennial Fire Protection Assessment

Lesson Plans MCTR-1075, Rev. 2: Maintenance Continuing Training GETCM-10304, Rev. OA: General Employee Training Self-Assessment (industrial Risk Management)

A1078758 (1/23/97)

F8.4 Review of UFSAR Commitments Following the discovery of a licensee operating their facility in a manner contrary to the UFSAR description, the NRC has highlighted the need for a special focused review that compares plant practices, procedures and/or parameters to the UFSAR descriptions. While performing the inspections discussed in this report, the inspectors reviewed the applicable portions of the UFSAR that related to the areas inspected. This included portions of

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Appendix 9A, " Fire Protection Evaluation Report" pertaining to the fire protection program.

Except for the emergency lighting issues discussed in Section F8.1 above, the inspector verified that the UFSAR wording was consistent with the observed plant practices, l procedures and/or parameters. l

F8.5 (Closed) LER 1-97-004 Previous Condition Prohibited by Technical Soecifications in l that 10 Fire Rated Assemblies were not Insoected Since Issuance of the Facility I Operatina Licenses For Each Unit  !

This LER reports an instance where 10 Unit 1 and 2 fire rated assemblies were found to be omitted from surveillance testing, and had not been inspected as required by a previous Technical Specification and the current Technical Requirements Manual. The appropriate surveillance procedures were revised, and inspections of the subject fire rated assemblies l were performed; no degradations were identified. The 10 fire rated assemblies were '

omitted due to inadequate reviews of procedures against the air, steam, fire and water boundary architectural drawing. Additional corrective actions include performing reviews with verifications of procedures against the drawing. This licensee-identified and corrected violation is a Non-Cited Violation, consistent with Section Vll.B.1 of the NRC Enforcement Policy. The LER met the requirements of 10 CFR 50.73, and the inspectors had no further questions regarding the event.

S1 Conduct of Security and Safeguards Activities a. Inspection Scope (81700)

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On March 12, the inspector reviewed selected access authorization (AA) records to verify the AA program provided high assurance that individuals granted unescorted access were trustworthy and reliable, and did not constitute an unreasonable risk to the health and safety of the public. The inspector selected AA records for short-term contractors that were on site the day of the inspection for review.

b. Observation and Findinas The inspector reviewed the AA records for 15 short-term contractors, that were performing roofing work. The inspector reviewed information in the Plant Information Management System (PIMS). The PIMS contains security profiles for allindividuals granted unescorted access. The inspector also reviewed selected background investigations for the contractors and reviewed visitor logs to assure that the contractors had been processed through the AA process and were not working onsite as visitors.

c. Conclusions The inspector determined that all the AA records reviewed contained the appropriate information required by the procedures and regulatory requirements, upon which to base a

dec;sion to grant unescorted access. The inspector also determined that all the individuals that were working onsite for that contractor and whose AA records were reviewed, had been processed through the AA program for unescorted access except for a replacement crane operator that was onsite for one day to replace the regular crane operator who was

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off on the day the records were reviewed. The replacement crane operator was on the site as a visitor and was being properly escorted.

S8 Miscellaneous Security and Safeguards issues S8.1 (Closed) VIO 352. 353/E96-144, 243 Safeauards information not orocerly controlled (9290M These violations concerned instances where safeguards information was not properly controlled in both cases, the follow-up investigations were comprehensive and in-depth. ;

Corrective actions were prompt and comprehensive, and included: the safeguards information was immediately properly controlled; generation and revisions to safeguards information maintained by the security section are performed on a designated stand-alone PC, controlled by the security section; individuals involved with safeguards information were retrained; the quantity of safeguards information, the number of individuals who !

handle it, and the number of locations where it is stored were reduced; and appropriate l procedures were revised to upgrade the expectations for handling safeguards information i and a new position of Safeguards Administrator was established, who will be responsible for providing safeguards training, conducting periodic audits, performing self-assessments, and recommending any necessary programmatic changes.

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The inspectors reviewed the corrective actions associated with the violations, and :

concluded that the concerns were adequately addressed. This violation is closed.

V. Manaaement Meetinas X1 Exit Meeting Summary The inspector presented the inspection results to members of plant n anagement at the conclusion of the inspection on April 3,1997. The plant manager acknowledged the inspectors' findings. The inspectors asked whether any materials examined during the inspection should be considered proprietary. No proprietary information was identified.

X2 Review of UFSAR Commitments l A recent discovery of a licensee operating their facility in a manner contrary to the UFSAR '

description highlighted the need for a special focused review that compares plant practices, procedures and/or parameters to the UFSAR description. While performing the inspections discussed in this report, the inspector reviewed the applicable portions of the UFSAR that related to the areas inspected. The inspector verified that the UFSAR wording was consistent with the observed plant practices, procedure and/or parameters.

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a INSPECTION PROCEDURES USED IP 37551: Onsite Engineering IP 60710: Refueling Activities IP 61726: Surveillance Observation IP 62707: Maintenance Observation IP 71707: Plant Operations IP 71750: Plant Support Activities IP 73753: Inservice Inspection IP 83750: Occupational Radiation Exposure IP 90712: In-office Review of Written Reports IP 90713: Review of Periodic and Special Reports IP 92902: Follow-up - Maintenance IP 92904: Follow-up - Plant Support IP 93702: Prompt Onsite Response to Events at Operating Power Reactors Tem:mrary Instruction 2515/134: Licensee On-Shift Dose Assessment Capabilities ITEMS OPENED, CLOSED, AND DISCUSSED Ooened 353/97-01-01 ViO Failure to Perform a 10 CFR 50.59 Review of a Modification to the Fuel Preparation Machine (E1.2)

Closed 352,353/1-96-019 LER Capability to Reject the Electrical Load of an RHR Pump Not Fully Verified (E8.1)

352,353/1-96-023 LER Fire Protection System Surveillance Tests Not Performed Due to Personal Error (F8.2)

352, 353/1-97-001 LER Safety Related Logic Circuits Not Fully Tested (E8.2)

352, 353/1-97-002 LER Failure to implement Technical Specification Required Visual inspections of Blind Flanges (E8.3)

352, 353/1-97-003 LER Degraded Back pressure Dampers Needed for Pipe Rupture Mitigation Result in Operation Outside Design Basis (E8.4)

352, 353/1-97-004 LER Previous Condition Prohibited by Technical Specifications in that 10 Fire Rated Assemblies were not inspected Since issuance of the Facility Operating Licenses For Each Unit (F8.3)

353/2-97-001 LER Condition Prohibited by Technical Specifications in that Control Rods were Removed from the Core with Source Range Monitors inoperable (08.2)

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j i 353/2-97-002 LER N128,2" RPV Instrument Nozzle Safe End Leak (E8.5)

! 352,353/95-10-01 URI Generic Concern - Storage of Plant Components (M8.1)

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" 352, 353/96-04-02 URI UFSAR not updated to reflect current practices relative to processing or storage of radioactive waste (R8.1)

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j 352,353/96-09-03 VIO 50.9 Violation (R8.2) i l 352,353/96-10-01 VIO Low EDG Fuel Oil Storage Tank Level (08.1)  !

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i 352, 353/96-10-02 VIO . PCIVs not properly sealed closed (08.1)  ;

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352,353/96-144 VIO - Safeguards Information Not Properly Controlled (S8.1) )

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] 352, 353!96-243 VIO Safeguards information Not Properly Controlled (S8.1) i p

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i 352,353/96-10-04 URI Loss of Control of Master Keys (R3)

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b LIST OF ACRONYMS USED

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AA Access Authorization ALARA As Low As Reasonably Achievable ANil Authorized Nuclear insurance inspector ANSI American National Standard Institute ASME American Society of Mechanical Engineers BWR Boiling Water Reactor CCTAS Core Component Transfer Authorization Sheet CDF Core Damage Frequency CFR Code of Federal Regulations CR Control Rod i CRD Control Rod Drive i CRS Control Room Supervisor ECR Engineering Change Request EDG Emergency Diesel Generator EO Equipment Operator ESF Engineered Safety Feature FME Foreign Material Exclusion GET General Employee Training GL Generic Letter HP Health Physics HPCI High Pressure Coolant Injection HVAC Heating,. Ventilation and Air Conditioning IFl Inspection Fo!!cw-up item ISEG Independent Safety Engineering Group ISI Inservice Inspection IVVI In-Vessel Visual Inspection l&C Instrumentation and Control

. LCO Limiting Condition For Operation LER Licensee Event Report MT Magnaflux Test NCV Non-Cited Violation NCR Nonconformance Report NDE Nondestructive Examination NMD Nuclear Maintenance Division NRB Nuclear Review Board NRC Nuclear Regulatory Commission NRI Non-relevant Indications OPCAT Operations Clearance and Tagging PEP Performance Enhancement Process PIMS Plant Information Management System PM Preventive Maintenance PORC Plant Operations Review Committee PSA Probabilistic Safety Analysis PT Penetrant Examination QV Quality Verification RCA Radiological Controlled Area RCIC Reactor Core Isolation Cooling RHR Residual Heat Removal

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RCS Reactor Cociant System

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) RMCS Reactor Manual Control System '

RO Reactor Operator  :

j RP&C Radiological Protection and Chemistry j '

RP Radiation Protection RPM Radiation Protection Manger i RPS Reactor Protection System ,

RPV Reactor Pressure Vessel -1 RWCU Reactor Water Clean-up ) '

RWP Radiation Work Permit i SDM Shutdown Margin J

SRM Source Range Monitor

. TS Technical Specification j UFSAR Updated Final Safety Analysis Report i UPS Uninterruptable Power Supply  ;

URI Unresolved item

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, VIO Violation j VT Visual Examination

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