IR 05000461/1997015

Insp Rept 50-461/97-15 on 970707-0825.Violations Noted.Major Areas Inspected:Operations,Maint,Engineering & Plant Support

ML20199A977
Person / Time
Site:	Clinton
Issue date:	11/12/1997
From:	Wright G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20199A962	List: IR 05000461/1997015 ML20199A958 ML20199A968
References
50-461-97-15, NUDOCS 9711180170
Download: ML20199A977 (26)

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U.S. NUCLEAR REGsILATORY COMMISSION R E G D N lli

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Docket No: 50-461 License No: NPF-62

' Report No: 50-461/97015(DRP) -

Licensee: Illinois Power Company Facility: Clinton Power Station

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Location: Route 54 West Clinton,IL 61727

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Dates: July 7 - August 25,1997 Inspectors: T. W. Pruett, Senior Resident inspector K K Stoedter, Resident Inspector R. A. Langstaff, Engineering Inspector M. S.Holmberg, Engineering inspector S. J. Campbell, Senior Resident inspector, Davis-Besse D. E. Zemel, Resident inspector, Illinois Department of Nuclear Safety Approved by: Geoffrey C. Wright, Chief Reactor Projects Branch 4

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9711190170 971112 1 PM ADOCK 05000461 0 PM i

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EXECUTIVE SUMMARY

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- Clinton Power Station NRC inspection Report No. 50-461/97015(DRP)-

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This inspection included aspects of hoensee openstions, engineering, maintenanos, and plant support.- The report covers a Sweek period of resident inspectio ,-

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One violation was identified for the failure to initiate operability detenninations within -

- 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for degraded but operable and inoperable equipment involving intermediate ,

range monitors, emergency diesel generator (EDG) ventilation system, and control room noon lights.- (Section O: 1)

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Operetnisty determinations for the noon indicating light failures were weak in that they did not address the impact on the ability of safety related systems to perform their required functions, the potential generic implications on other safety related systems, and required NRC prompting to ensure that engineering evaluations supported system operabilit (Section 01.1)- f

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Two auxiliary operators were knowledgeable of systems assigned to their watch station and provided good responses to questions asked by the inspectors involving equipment operation. (Section 01.2)

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Control room operators demonstrated good event response during the unexpected flooding of Condenser Water Box B and the failure of the Division i 4160 Volt (V) Bus 1A1 Main Feed Breaker. (Section 01.3)

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Operators demonstrated a poor questioning attitude by-not verifying the condition of the Division lit battery prior to allowing a single cell equalizing charge. (Section M1.3)

- Maintenance

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One violation was identified for the failure to initiate a condition report following the identification of a degraded secondary containment door seal to Residual Heat Removal (RHR) Pump A room. (Section M1.2)

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Two violations were identified for the use of a non-1E Class portable battery charger that '

was not seismically restrained for charging a Division ill battery. (Section M1.3)

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One violation was identified for the failure to select a new pilot battery cell following the completion of a single battery cell equalizing charge on Division lli Battery Pilot Cell 24.

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(Section M1.3).

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A poor maintenance practice was identified for the inability to identify sediment build-up inside Division 11 Battery Cell 28. The reluctance to initiate a condition repost or perform additional inspections of safety related batteries for sediment demonstrated a lack of rigor in implementation of the corrective action program. (Section M1.4)

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EDE 200g ,

. Engineering personnel's identification that a fire in the control room would render '

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equipment used in the safe shutdown analysis inoperable demonstrated good attention to

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detail. The initiation of reviews for other equipmerd relied upon in the safe shutdown -

analysis exhibited improved implementation of the conective action progra (Section E1.1) ,

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Two examples of one violation were identified for the failure to ensure design basis information was adequate and appropriatCy transleted into specifications, drawings, procedures, and instructions for the control room and emergency diesel generator (EDG)

_vontdation systems and the hydrogen mixing compressors. This failure resulted in a non-

_ conservative change to the maximum loading of the EDGs. (Sections E1.2 and E1.3)

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The calculation index data base did not accurately track design calculations used to support changes in the Updated Safety Analysis Report. (Section E1.4)

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=. Engineering personnel were knowledgeable of Generic Letter 96-06. " Assurance of

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Equipment Operability and Containment Integrity During Design Basis Accident

. Conditions." Evaluations of the Generic Letter were easily retrievable and included sufficient information to complete an independent review. However, the inspectors found i ' that the licensee's review of GL 96-06 missed identifying that Containment Penetration MC-86 for the RT system may be impacted during a LOCA. (Section E2.3)

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Plant Support

. The fire brigade responded well to the potential electrical fire on Division i 4160 V Bus 1A1. (Section 01.3)

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A vcJ awareness of radio ogical conditions was*noted,when maintenance requested that radiation protection relocate a posting prior to performing the maintenance on the door

, seal for RHR Pump A room. (Section M1.2)

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Individuals involved with the lifting of a radioactive waste cask were knowledgeable of the activity and the compensatory actions to be taken if the cask dropped. (Section R1.1)

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During a tour of the emerger:cy operations facility (EOF), one instance of poor emergency

planning was identified involvng sampling from the EOF fume hoodc (Section P2.1)

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One violation was identified for the failure to provide adequate lighting for the RCIC ~

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Report Details Summarv of P.3nt l Status

. The facility remained shut down during the inspection period. The failure of two Westinghouse 4160 Volt (V) breakers to open on July 22 and August 5,1997, delayed the startup of the facilit At the end of the inspection period, the licensee was in the process of partially cleaning all

, Westinghouse 4160 V and 6900 V circuit breakers to correct the cause of the failure. The review of Westinghouse breaker failures will be documented in NRC Inspection Report No. 5041/9701 . Operations 01 Conduct of Operations 01.1 Operability Determinations (OD) ' Inspection Scope (71707 and 37551)

The inspectors reviewed operations personnel's response to issues which required the initiation of an operability determination (OD).

E i Observations ano Findinas Intermediate Ranae Monitors (IRMs)

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i On June 25,1997, personnelin maintenance planning determined that an unapproved, conductive, and corrosive soldering flux may have,been used while performing work on several IRMs. On July 30,1997, the inspectors questioned operations personnel to

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determine if an OD had been written to support the continued operability of the IRMs since unapproved flux was used to connect cables to the IRMs. Operations personnel

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initially stated that sn OD for the IRMs had not been initiated because the issue was placed on a Mode 2 restraint list on June 18,1997, requiring that the operability of the IRMs be resolved prior to startup. On June 27,1997, engineering personnel performed

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an evaluation supporting continued operation with the IRMs in a degraded but operable condition. Operations personnel cleared the Mode 2 restraint on July 10,1997, however, they did not further recognize the need to incorporate information from the engineering evaluation into an OD such that the degraded condition of the IRMs could be tracked after .

reaching Mode 2. On August 8,1997,44 days after initial dircovery, operations personnel acknowledged this fact and initiated OD 197-06-200 to document the

, degraded but operable condition of the IRM Emeraency Diesel Generator (EDG) Ventilation System (VD)

On July 2,1997, the inspectors questioned engineering personnel on the operability impacts on the EDGs and the VD system when the outside temperature exceeded the temperature limitations described in Updated Safety Analysis Report (USAR)

Section 2.3.1.1," General Climate" (see Section E1.2 regarding the VD design). Although the potential existed for safety systems to be operable but degraded or inoperable,

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engineering personnel did noi provide operations personnel with informahon regardeng

. the operability of the EDG or the VD systems until July 24,.1997, 22 days after identifying *

the onginal concem. On July 24,1997, engineenng personnel provided operations : "

personnel with OD 1-97-07-250 which steted that the EDGs were operable provided

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outside ambient temperature remained between 5'F and 102.7'F (104*F for Division 111).~

Ambierd temperatures below 5'F or above 102.7'F would require that the EDGs be -

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Following further licensee evaluation of the design of the EDG ventilation system, the licensee issued Licensee Event Report (LER) 97-022-00 on September 22,1997, _

documenting that Divisions I, il and 111 EDGs were inoperable when ambient temperatures

- exceeded 91*F rather than 102.7'F. -To date, the licensee has not completed an evaluation for equipment impact during cold ambierd temperatures (below 5'F). Review

~of this new information and results of EDG impact in colder ambient tempersiures will be_

documented in NRC Inspection Report No. 50-461/9702 Neon Llahts ,

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in 1996, the , licensee decided to replace 592 neon light sockets in the control room. The noon lights were used to provide indication of equipment position (i.e., valves "open" or

"close") and verify equipment operation (i.e., pumps "on" or "off"). Between January 1 ;

and June 2,1997, the licensee identified that 16 neon indicating lights in the control room i

failed after being replaced.- On June 3,1997, the licensee determined that improper use

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of uncontrolled soldering flux caused the neon lights to short circuit and fail. Further, on

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June 5,1997, operations initiated OD 1-97-06-023 to address the effects of using an -

-inappropriate flux on the operability of safety related equipment. After reviewing the final OD, ine inspectors considered it weak in that:- (1) operations recoonized that the failure of the neon indicating light resulted in the inability to operate the specific component, but

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the impact on the ability of the system to perform its safety function was not addressed; (2) operations did not assess the potential generic implications of the failures on other s) tems; and (3) operations utilized electrical maintenance (EM) personnel's determination that the failures were statistically insignificant as a basis for operability instead of using the impact of the failurer on the repeated loss of component and system L

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safety functions. The inspectors concluded that the use of statistics as a basis for operability was inappropriate and, therefore, demonstrated a lack of safety focus by i- operations personne On June 9,1997, the inspectors questioned the licensee to determine if new information obtained during the investigation uf the inappropriate use of flux for the neon lights had been used to revise OD 197-06-023. Operations personnel stated that the OD had not been revised. After prompting by the inspectors, operations personnel realized that the t-OD needed to be updated and subsequently made the revisio .

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The inspectors questioned the use of unapproved flux on other components. Upon further review, the licensee initiated OD 1-97-06-083 to address the use of unapproved flux for Source Range Monitor "B" and also initiated OD 1-97-06-097 to evaluate the operability of the Division I and ll EDGs, Residual Heat Removal (RHR) Pumps A and B,

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the Low Pressure Core Spray Pump, and Shutdown Service Water Pump A. Specifically,

for the components listed on OD 1-97-06-097, if the neon indicating light failed, the component would not have operated or may have tripped, thereby preventing the

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component from performing its intended safety function.LThe need for the NRC to prompt -

operations to ensure that engineering evaluations supported the operability of safety

= related equipment was considered a weakness in the implementation of the OD progra .

The inability to operate specific components and the potential generic implications on other safety related systems as a result of the shorting of the noon light sockets will be documented in NRC inspedian Report No. 50-461/9702 .

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10 CFR Part 50, Appendoc B, Criterion V, " Instructions, Procedures, and Drawings,"

states, in part, that activities affecting quality shall be presortbed by documented ;

instructions or procedures appropriate to the circumstances and shall be accomplished in -

accordance with these instructions, procedures, and drawings. Procedure CPS 1014.06,1

.. " Operability Determinations," Section 2.1.5, specified that the shift supervisor shall povform the OD within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The failure to perform operability determinations for the IRMs EDG VD and noon lights within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is an example of a violation for the failure

to follow procedures as required by 10 CFR Part 50, Appendix B, Criterion VIO 50 441/9701541a). Conclusions Operability determinations for the use of an unapproved soldering flux on the neon light were weak in that they did not address the impact of the socket failures on the ability of

safety relateil . systems to perform their required functions, the potential generic implications of the failures on other safety related systems; and required NRC intervention to ensure thst information in the OD supported system operabilit .2 Auxiliary Operator Watch Station Tours (71707)

On July 16,1997, the inspectors accompanied art operator on tours of the containment, and the fuel, control, and auxiliary buildings. Further,' on July 17,1997, the inspectors o

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toured the site with an outside operator. Both auxiliary operators were knowledgeable of systems assigned to their watch station and provided good responses to questions asked by the inspectors involving equipment operation.

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01.3 Operator Response to Unexpected Events

' Inspection Scope (71707. 71750. and 93702)

The inspectors observed the response of control room operators and the site fire brigade .

. during events involving the flooding of Condenscr Water box B and the failure of the'

l: Division i 4160 V Bus 1 A1 Main Feed Breaker to open during a transfer of the 1 A1 Main Bus from the emergency reserve auxiliary transformer (ERAT) to the reserve auxiliary transformer (RAT).

Observations and Findinas

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Unexpected Floodina of Condenser Water box B On July 16,1997, the inspectors observed the control room operators respond to the -

unexpected flooding of water into Condenser Water box B during maintenance. At the '

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time of the event, maintenance personnel had completed work inside the Water box and had exited the area. No injuries occurred as a result of the event.- During the event, the .

operators demonstrated good response by frequently referencing procedures and diagrams, frequently conducting control room briefs, and using formal (three-way and repeat back) communication within the control room and between the control room and personnel at the Water box. This formal communication contributed to understanding of actions in progress, the evaluation of compensatory measures, the evaluation of impact

, on plant operations, and implementation of industrial safety precaution Unexpected Failure of Division i 4160 Main Feed Br-95 I On July 22,1997, at 1:07 p.m., the control room received a report of smoke in the '

781-foot elevation of the auxiliary bc>lding. At 1:12 p.m., the control room received a --

report of smoke from the Main Feed Breaker to 4160 Bus 1 A1,22 and at 1:14 p.m., the control room dispatched the onsite fire brigade. The inspectors observed the response of ~

the station fire brigade and control room operators following the failure of the Division 14160 V Bus 1 A1 Main Feed Breaker to open during a transfer of the 1 A1 Main Bus from the ERAT to the RA Fire Brigade Response .

Fire brigade members arrived prcmptly at the dress-out area and donned the appropriate fire protective clothing. Once arriving at the Bus 1A1 location, the inspectors noted that fire hoses had already been deployed and that fire suppression foam was available as an altemative extinguishing agent. The fire brigade leader reminded personnel on several occasions that additional safety measures needed to be considered because of the potential hazards involving high-voltage conditions associated with the switchgea Communication between the control room and the fire brigade leader was good, and members of the brigade were frequently briefed. The-brigade was proactive in their preparation to combat the fire and considered the need for additional supplies if the event progressed. For example, additional air packs were located and a phone was obtained to maintain contact with the control room in the event of an unplanned loss of radio commun: cation Control Room Response The inspectors considered the control room response good in that industrial safety precautions were considered and practiced, crew briefs were thorough, and corrective actions to control the situation were evaluated for their impact on plant system Weaknesses observed by the inspectors involving informal communications and operator response to annunciators (i.e., failure to acknowledge annunciators, late in acknowledging annunciators and not reviewing procedures) did not distract the operators from remaining focused on the event.

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The inspectors noted that several procedures were utilized during the event. However, operators did not always review the precautions and limitations associated with the procedures since only specific sections were performed. The licensee assigned an l action item to the operations department to review the use of limited portions of procedures during casualties or event i

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Licensee representatives observing the event also noted areas for improvement including - ,

three-way communications, implementation of annunciator response procedures, maintaining professionalism during events, and notification of site personnel of event

progressio Conclusions

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, , Control room operators demonstrated good event response during the unexpected filling of Condenser Water box B and the failure of the Division i 4160 V Bus 1A1 Main Food

, Breaker. Additionally, the fire brigade responded well to the potential electrical fire at the :

switchgear.

j 02 Operational Status of Facilities and Equipment

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0 Automatic Dooressurization System (ADS) Audit

. . a. . JDAEfetion Scope (71707)

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During an inspection of the ADS, the incoectors reviewed the USAR, the inservice inspection program manual, applicable Aines?can Society of Mechanical Engine (ASME) codes and standards, nitrogen-steag, test correlation data, piping and instrumentation drawings pertinent to ADS and supporting systems, applicable vendor manuals, commitments made pursuant to Generic Letter (GL) 88-14, " Instrument Air

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Supply System Problems Affecting Safety-Related Equipment," and performed field

observations, e Observations and Findinas

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The inspectors identified a broken flexible conduit for the acoustic monitor on Safety Relief Valve (SRV) 1B21 FO41C. Althowh the br'oken-conduit had no impact on monitor

operation, the licensee documented the condition on a CR and corrected the deficienc _

The inspectors noted that the in'strument air amplifiers, which provide the motive force to open the ADS valves, were equipped with a pre-filter. However, the inspectors were

• unable to locate maintenance documents specifying the filters were periodically replace The system engineer stated that he periodically used a filter sight glass to visually monitor the condition of the filters and would request support if the filters appeared degrade The system engineer also stated that periodic monitoring of the filters would be incorporated into routine operator tours following a revision to Procedure CPS 3800.02C001, "C-Area Daily Rounds.'

' Conclusions The ADS and supporting systems were appropriately tested and maintained. The system engineer provided knowledgeable answers involving the testing requirements for the instrument air system and the use of correlation data to support SRV testing with nitrogen

'_instead of process fluids.

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11. Maintenance

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M1 Conduct of Maintenance

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- a. - inanectirn Soone f61726 and 62707)

' The inspectors observed'or reviewed portions'of the following maintenance and surveillance activities:

'.'.- Procedure CPS 8433.01 for 125 VDC battery maintenanc '.

Procedure CPS 9431.08 for reactor protection system main steam line radiation channel calibratio *

Procedure CPS 9071.01 for diesel-driven fire pumps operability test.-

.- Procedure CPS 9059.01 for reactor coolant system leakage test.

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PCICMM523 for containment pressure and suppression pool level transmitter

Preventive Maintenance (PM).

. PMMXDS028 for RHR Pump A room door seal P . PMMXDA034 for RHR Pump A room door seal PM.

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TM 97-052 for installation of temporary modification on Switchgear Ventilation Supply Damper 1VXO3Y ,

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- M1.2 PM on RHR Pumo A Room Door Seals -

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L Inspection Scope (62707)

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The inspectors observed portions of PM activities performed on RHR Pump A room Door 1SD1-22 (PM Numbers PMMXDS028 and PMMXDA034). The scope of the work -

involved an evaluation for indications of wear and replacement of the door seal if necessary. Door 1SD1-22 is one of two airlock doors for RHR Pump A room and forms part of the secondary containment boundar .

Observations and Findinas

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Upon arrival at the work location, mechanical maintenance (MM) personnel noted that a

radiological posting would be blocked while the work was being performed on Door 1S01-22 and requested that radiation protection relocate the posting. The inspectors determined that MM personnel used good judgement in requesting the posting -

be relocated prior to the commencement of maintenance activitie On August 15,1997, MM personnel noted that the seal on Door 1SD1-22 required

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replacement based on the acceptance criteria in PM Task Card PMMXDS028. The

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acceptance criteria in job step 3 of the task card stated to inspect the door seal for signs ;

of wear such as cracks, abrasions, or permanent set ana replace as necessary in . 'l accordance with Procedure CPS 8250.01, " Maintenance of Water Tight Doors." integrity of the door seal ensures that secondary containment is maintained with the door close Due to degrads. con, MM personnel planned to replace the door seal during the next scheduled PM (PMMSOA002) on September 10,1997. . The inspectors questioned *

MM personnel to determine if other actions needed to be taken to compensate for the degraded seal until repairs could be made. The liwnsee roviewed the guidance in .

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Procedure CPS 8250.01 and stated that no further actions were required.

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The inspectors questioned MM personnel to determine if a condition report (CR) would be initiated for the d*0raded condetson. The MM personnel stated that no additional documentation was required because the PM instructions and Procedure CPS _8250.01 contained all of the job steps necessary to replace the deficient seal. The inspectors

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concluded that the licensee's response clearty indicated they were unaware of the i purpose of the corrective action system. While it may be true that all steps necessary to correct the problem existed, the CR and subsequently the CR process would allow trending of similar deficient conditions. Further, the inspectors noted that the PM procedure or Procedure CPS 8250.01 did not require notification of the shift supervisor or

the cognizant engineer in the event degraded or nonconforming conditions were identified during the inspection of door seal Procedure CPS 1016.01, " CPS Condition Reports," Section 8.1.1, specified that any

, person becoming aware of a condition adverse to safe reliable power production initiate a CR. Section 2.2.4 stated that a condition adverse to safe reliable power production included failures or malfunctions, deficiencies, defective items, and nonconformances.

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The failure to initiate a CR is an additional example of a violation for the failure to follow procedures required by 10 CFR Part 50, Appendix B, Criterion V (VIO 50 461/97016-01b).

This violation is significant in that individuals performinhhe maintenance activity did not unders.tand the corrective action process and the significance the CR's role in trending equipment failures. Although the task card provided sufficient instructions to fix the door seal, the licensee agreed that a CR should have been written to document the degraded

' door seal and initiated CR 1-97-08-114 to trend deficiencies identified during the performance of preventive maintenance, Conclusions

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The licensee demonstrated a good awareness of radiological postings in that they requested radiation protection assistance to relocate a posting prior to commencing

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maintenance activities on the door to RHR Pump A room. However, they exhibited a lack of sensitivity toward documenting plant deficiencies through the CR process to allow for trending. As a result, one violation was identified for the failure to initiate a CR following MM personnel's identification of a degraded seal on the door for RHR Pump A roo ..

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l M1.3 Division lli Sinole Cell Battery Charoe inspection Scope (71707 and 62707)

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The inspectors observed the performance of a single cell equalizing charge on the Division ill Battery in accordance with Procedure CPS 8433.01, Revision 14 " Generic Procedure for 125 VDC 01ttery Maintenance."

' Observations and Findinos On July 8,1997, the inspectors noted an entry into the station log which indicated that Division 111 dattery was operable with a single cell equalizing charge on Pilot Cell 24. The inspectors questioned the licensee to determine how the battery could remtin operable during a single cell equaliziag charge. OperaHons believed that Pilot Cell 24 had been disconnected from the battery and that a ceparate portable charger had been connected to the cel The inspectors pe formed a walkdown of the Division 111 Battery and noted that Pilot Cell 24 remained connected to the battery and that the portable charger was not seismically mounted. Further, the inspectors were concemed that the portable charger did not meet the requirements for a Safety Class 1E componen Following the inspectors' observations, operations personnel declared the Division lll Battery inoperable because a non Safety Class IE portable battey chat?ar was ,

connected to a safety related battery and the portable charger was not seismically mounted. The licensee stated that their understanding 01 the configuranon of the battery had not been verified prior to the commencement of the single cell equalizing charg The inspectors requested that the licensee provide documentation that the portable chargers had been either procured or qualified as' Safety Class 1E. The licensee was unable to produce this documentation and on July 9,1997, EM personnel initiated CR 197 07-090 to document the use of the non-Class 1E portable battery charger. The licensee instructed personnel to not use the portable battery chargers until further notic The inspectors questioned the licensee to determine the frequency at which the non Class 1E portable chargers had been used. The licensee stated that these portable chargers had been used on one Division I cell, two Division 11 cells, and four Division 111 cells between June 1989 and July 199 ^

Procedure CPS 8433.01, Revision 14, Section 8.7.3, specified, in part, that a Class 1E portable battery charger be used while equalizing individual cells. The failure to use a Class 1E portable charger is an additional example of a violation for the failure to follow procedures required by 10 CFR Pa t 50, Appendix B, Criterion V (VIO 50 461/97015 01c)

The inspectors reviewed the procedure and found that instructions were also not included to mount the Class 1E chargers when connected to the safety-related battery to satisfy seismic qualifications. Therefore the inspectors concluded that the procedure was inadequr,te from the aspect of ensuring seismic qualification was satisfie __ ___ _

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t Procedure CPS 1019.05, ' Control of Transient Equipment / Materials," Revision 5, Section 8.3, " Securing Equipment from Sliding,* required stable equipment that is not 'in use* should not be set within three feet plus its height of adjacent equipment or ledges unless positively secured. Procedure CPS 1019.05, Section 2.2.6, defined *in use* as an item that is under the control of an individual with maintenance, testing, or other work activity in progress. The failure to adequately securs the portable ba'tery charger was another example of a violation for the failure to follow procedures required by 10 CFR Part 30 Appendix B, Criterion V (VIO 50 461/97015 01d).

The inspectors questioned EM personnel to determine if a different pilot cell had been designated for the Division 111 battery following the completion of the individual cell equalizing charge. Maintenance and engineering personnelinitially stated that a new pilot cell would not be selected since Pilot Cell 24 still reflected actual battery performance. ~

Nevertheless, the inspectors requested that the licensee provide copies of data sheets which specified the voltage and specific gravity measurements during the latest quarteriy surveillance test performed on July 7,1997. Procedure CPS 8433.01, Section 8.4.,

required that the pilot cell be the cell with tb 'owest individual cell voltage with the battery on float charge. The inspectors compared i. quarterty test data for the battery cells to the as len data following completion of the individual equalizing charge on Pilot Cell 2 Based on a review of the data, the licensee concurred that the individual cell voltage on Cell 30 (2.22 V) was lower than the individual cell voitage on Cell 24 (2.23 V).

The licensee stated that Procedure CPS 8433.01 was inadequate in that the procedure did not provide guidance on the selection of a new pilot cell following the completion of an individual cell charge. The procedure did require that a new pilot cell determination be perfonned semi-annually or following the completion of an equalization charge on the entire battery. The inspectors noted that EM personnel demonstrated a poor questioning attitude because the selection of a ne # pilot cell was not considered following completion of an individual charge on Pilot Cell 24. The hi!ure to provide appropriate instructions and procedures regarding the selection of a new pilot cell following the completion of an individual cell charge is i violation for the failure to implement an adequate proceriure as required oy 10 CFR Part 50, Appendix B, Criterion V (VIO 50-461/97015-02). Conclusions Three examples of a violation were identified for (1) tho use of a non-Class 1E portable charger on the Division lll Battery; (2) not seismically restraining the portable charger while connected to a safety related battery; and (3) the failure to select a new pilot cell following the completion of an individual cell equalizing charge. A poor questioning attitude was demonstrated by operations personnelin that they did not verify the condition of the battery prior to a lowing a single cell equalizing charge. Maintenance personnel demonstrated a poor questioning attitude because they did not recognize the potential for a different cell to have a lower individual cell voltage following the charge on Pilot Cell 2 .

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M1A Material Condition of the Division ll Battery ,

i inaceation Soone (62707)  !

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The inspec6 ors examined the battery cells on the Division ll Battery Bu {

- Observations and Findinns f

. 1 The inspectors identlRed the presence of 1/8-inch accumulation of sediment on the -  !

bottom of Cell 28 and that the buildup a,mained well below the inter cell electrodes. An i examination by the inspectors of the remaining battery oells on Division I and lil determined that Cell 28 was the only cell with appreciable sediment buildup.-  :

.

.

l Ahor the inspectors informed the licensee of the sediment buildup, EM personnel visually i inspected Cell 28 and contacted the vendor who manufactured the battery. The vendor

-

stated the sediment bu%g was normal for the battery and that a significant buildup of sedement may contact the intercell electrodes which can short the cell and render the

' entire bus inoperable. However, neither maintsnance nor engineering initiated a CR to  ;

'

document the deficiency or track the sedimen, buildup rate until questioned by the l

, inspectors. Additionally, the licensee did not perform inspections of the remaining safety- .

related batteries (Divisions I, lil, and IV) to determine if the same condition existed for this .

equipment. The reluctance to initiate a CR or inspect the remaining batteries ,

demonstrated a lack of rigor in the implementation of the corrective action progra When questioned by the inspectors, maintenance personnel stated that they had missed identifying the sediment build up during the last visualinspection of the battery. The i failure to recognize the sediment in Cell 28 and its potential to effect the operability of

'

safety- related equipment was considered a poor maintenance work practic '

[ c, '

Conclusions - .- . -

m The reluctance to initiate a condition report or perform inspections to ensure safety  ;

-related batteries did not contain sediment buildup demonstrated a lack of rigor in the

~

, implementation of the corrective action program. The inability of EM personnel to identify  ;

the sediment buildup in the battery during the last visual inspection was considered a poor maintenance practice, lit. Ennineerina E1 .

Conduct of Engineering E1,1 Safe Shutdown Concems identified Durina Review Effort of Breaker Coordination Problems - Insoection Scope (37551) .

'

The inspectors reviewed the licensee's' actions taken after!dentifying that the current

'

configuration of the Division I Switchgear Heat Removal (VX) system did not suppori the safe shutdown analysi i

,

k t--t-- r -e e----r~g w- M m 3 --pr e = + =%=- 7 we^ e-- *-m -en*-- -' e ve w---=s,er*<es-wr e-'v+*--- ~ =-* w w r=v r +- - >-e,-*w=--e -<=-'==*~w=--'

l Observations and Findinor While reviewing the impacts of breaker coordination problerns for an unrelated modification, engineering personnel discovered that if a fire occurred in one of the main control room panels, the operability of the Division i VX Supply Damper 1VXO3YA could potentially be effected. Consequently, engineering personnelinitiated CR 1-97-07-164 to document this findin In response to the CR, engineering personnel developed Temporary Modification 97-052 which lifted a wire between the damper's circuitry and the control room Indication for the '

damper such that a fire within the control room panel would not effect the operation of the damper. A permanent design change to testore the control room indication for the damper and allow the damper to perform its function to open during a fiis in the control

'

room was being developed at the conclusion of this inspection period. Engineering personnel reviewed the circuitry for the remaining dampers needed for safe shutdown and found no other problems. The inspectors considered this finding by engineering to demonstrait a good questioning attitude since it was identified while reviewing an unrelated problem, in addition to the finding by engineering, an independent engineering tecm discovered that the circuitry for the Division l EDG output breaker was not electrically isolated such that a fire in the control room, in conjunction with a loss of offsite power, could result in the output breaker failing to close and the inability to operate the breaker from the remote shutdown panel. Engineering personnel developed a design change which rewired the EDG control circuitry to provide electrical isolation and allow the output breaker to be operated from the remote shutdown panel. As a result of this finding, the remaining circuits routed through the remote shutdown panel were reviewed to ensure that a fire in the main control room would not effect the ability to operate the equipment from the remote shutdown panel. No other problems were identified. The inspectors concluded that the review of remaining circuitry routed through theremote shutdown panel for possible generic implications demonstrated improved implementation of the corrective action progra On August 25,1997, the licensee initiated LER 97-021 because the Division i EDG did not meet the plant's design basis and could prevent power being suppned to Division I safety equipment during a control room fire. The LER stated that an assessment of the safety significance and the implications of this event has potential nuclear safety significance. Although the licensee has corrected the design deficiency, this matter will be tracked as URI 50-461/97015-3 pending further NRC revie , Conclusion The licensee's identification that a fire in the control room could render equipment used in safe shutdown of the unit inoperable demonstrated good attentior. to detail. The initiation of additional reviews for the remaining dampers and other equipment relied upon in the safe shutdown analysis exhibited improved implementation of the corrective action program. One URI was opened to review the EDG design deficiency that had potential nuclear safety significanc _ _ _ _ _ _ _ _ _ . _ - - _ _ - _ .

_ - --

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l E Review of Operaldlity Determination for the VD System Inspection Scope f37551)

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On June 26,1997, e,$::-k,g personnel initiated CR 1-97-06 302 in response to their l

, determination that the maximum calculated load for the VD fans was higher than  !

expected and that USAR Table 8.313, " Plant Loads," incorrectly spoolned the VD fan  !

, load. In response to the CR, operations personr.alinitiated OD 197-06 302 to support continued operability of VD and the EDGs. The inspedors reviewed OD 1-97-06 302 to i ensure that the information used to support continued system operatWilty was consisterd .  !

with the currerd design and licensing basi ! Observations and Findinos .  !

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.The inspootors concluded that this issue was caused by inadequate implementation of . I design control measures in that the original design information was not vedfied as correct in acoca.1ance with the licensee's design control program. Specifically, the original 1 performance curves supplied by the vendor specified that the maximum horsepower (Hp)  !

required by the VD fans occurred when outside air temperature was 96*F and pressure ,

'

was 29.92 inches mercury (Hg). Engineering determined that the temperature of the air at'the inlet to the fan could be cooler than 96'F, that site pressure may be lower than  !

29.92 inches Hg, and that colder air would cause the Hp required by the fan to increas [

Ar, initial calculation using the lowest temperature allowed in the VD fan room per USAR

, Section 9.4.5.1, " Diesel Generator Facilities Ventilation System" (65'F), and i

corresponding site specific pressure (29.12 inches Hg), determined that VD Hp would increase from 105 to 114.

,

Operations and engineering personnel initiated OD 197-06-302 to document the operability of VD and the impact of the increased loading on the EDGs. In OD 197-06 302, the licensee balanced the EDG load increase, as a result of an in.:rease '

,

in VD Hp, against an assumed load decrease from the control room ventilation (VC),

L concluding the EDG load increase was acceptable. Specifically, while VD load increased, engineering personnel believed actual VC load was 205 kilowatts (kW), not the 211 kW load specified in USAR Table 8.313. However, engineering personnel were unable to provide documentation or a calculation supporting the.VC load decrease. The inspectors considered the approached used in the OD inadequate because the VC load  ;

reduction was credited without verifying or checking the adequacy of the desig *

[ 10 CFR Part 50, Appen$x B, Criterion 111. " Design Control," requires, in part, that i measures be watablished to assure that the design basis for structures, systems, and  :

components are correctly translated into specifications, drawings, procedures, and instructions.; The failure to assure that design basis information regardireg the VC and i VD systems were appropriately translated into engineering evaluations was considered an example of a violation of 10 CFR Part 50, Append!x B, Criterion ill  ;

~

(VIO S0 481/97015-04a) .Q:gnclusions *

l_ One example of a 10 CFR Part 50, Appendix B, Criterion til violation was identified for the j failure to assure that design basis information for the VD and VC systems were i

,

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15

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._

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j WCC translated into ensincering evaluations. This failure resulted in a  !'

nonconservative change to the maximum kilowatt loadmg of the EDG i E1.3 Imornoer Reduction of Hvdromen Mudna fHG) Comora**~ Ratina j

, Inanection Scope f37551) [

i

, Engineering personnel initiated safety evaluation (SE) g7-064 to propose changes to

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USAR Table 8.3-13, " Plant Loads? During a review of this SE, the inspectors noted that the maximum loading for HG Compressors 1HG02CA and C8 was dooressed from 60 Hp l

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, to 36 Hp. Since the SE did not provide information to justify the change in compressor ,

'

load, the inspectors questioned the licensee 4 determine if the assumptions used to decrease the Hp reting were consistent witt, the accident analysi !

r ,Qhservations and Pndinas  :

i

,

_

_ The licensee informed the inspectorr that the change in the HG compressor load had

been approved on June 12,19g7, p sr Calculation ig AK-05, " Diesel Generator Load Vonitoring," Revision 4. The inspectors reviewed Calculation ig-AK 05, but were unable _  !

to determine the justification for the change in loads and again requested that engineering j

,

personnel provide ti,e basis for the compressor load reduction. After approximately two i weeks of reviewing information to justify the reduction in Hp, ~ engineering personnel informed the inspectors that the loading change as specified in USAR Table 8.313 was  ;

in erro l

,

!

"

Engineering determined that the original performance curves supplied by the vendor had

not been updated to include information provided by Sargent and Lundy,  !

CPS Architect / Engineer, prior to initial plant startup. Consequently, the licensee used an -

outdated performance curve with a lower HG compressor flow rate (283 cubic feet per ,

minute (CFM)) than that requirew by Technical Specifications 3.6.3.3, L * Containment /Drywell Hydrogen Mixing Systems," (800 CFM) when determining the maximum loading. The licensee later determined that when considering the higher flow ,

rate, the Hp of the HG compressors increased from 35 to 4 r f

i The inspectors were concemed that engineers had not included readily available design information in calculations which were used to propose changes to the facility as

described in the USAR. The inspectors noted that assumptions made in design

,

' calculations were not always verified when proposing changes to the system's desig Through discussions with several engineers, the inspectors determ:ned that the approval i '

of a calculation did not require verification that system design requirements and

! assumptions remained satisfied. The inspectors considered the lack of ensuring previous

assumptions continued to be met when performing new calculations a weakness in the -

l licensee's calculation program, in asponse to the inspectors' concem, the licensee initiated CR 1.g7-07 208 on July 18, 1997, to document the potential for engineering design activities to use obsolete f i information from old desigr calculations because the calculation index database did not l

accurately track design calculations used to support the USAR. The failure to assure that design basis information regarding the Hp required for the hydrogen mixing compressors t was correctly translated into specifications, drawings, procedures, and instructions was *

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+

.-- _

,

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_ . . ~ __ _ _ . - _ ._ _. _ .

t i

considered a second example of a violation of in CFR Part 50, Appendix B, Criterion 111 (VIO 50 441/97015-4b). ,

<

Engineering initiated CR 1 g7 07-105 to document the incorrect Hp rating for the HG compressors. While investigating the CR, it was determined that Sargent and Lundy had not considered the effects of back pressure on the HG compressors due to high suppression pool level or the change in specific gravity due to changes of gas concentrations when determining worse-case Hp requirements. When the licensee considered these additional effects, the required HG compressor Hp ircreased from 48 to 57. In response to this finding, the licensee pedormed another SE to change the HG compressor load as stated on USAR Table 8.313 back to the original 60 H Conclusions -

A second example of a 10 CFR Part 50, Appendix B, Criterion til violation was identified for the failure to assure that design basis information for the hydrogen mixing compressors was appropriately translated into engineering evaluations. This failure resulted in a nonconservative change in the maximum Hp loading for the HG compressors. The calculatior' index data base did not accurately track design calculations used to support changes in the U.SAR.

E2 Engineering Support of Facilities and Equipment E2.1 Secondary Containment Intearity Testina Inspection Scope (37551)

The inspectors reviewed secondary containment testing Procedures CPS 3404.01, " Fuel Building HVAC (VF)," Revision 9, CPS 9065.02, " Secondary Containment Integrity,"

Revision 27, and CPS 9066.01, " Secondary Containment isolation Damper Operability,"

Revision 24, to determine if isolation dampers were adequately tested. Observations and Findinas The inspectors reviewed USAR Section 9.4.2, " Fuel Building HVAC System," and detemlined that the inboard and outboard fuel building ventilation (VF) system isolation dampers were redundant to each other, formed part of the secondary containment boundary, and that " tests are made on the isolation dampers to verify the closure times and leakage chkracteristics,"

The inspectors reviewed Procedures CPS 9065.02, CPS 9066.01, and CPS 3404.01 and determined that damper timing tests were performed. Regarding leakage testing, Procedure CPS 9065.2 tested secondary containment integrity by securing the VF system per Section 8.3 of Procedure CPS 3404.01," initiating the Standby Gas Treatment System," and then measuring the time required to develop 0.25 inches (water gauge) of vacuum. The inspectors noted Procedure CPS 3404.01 directed operators to shut both the inboard and outboard sets of VF isolation dampers during the test. However, the inspectors questioned if measuring the time required to develop the vacuum was an adequate test in independently verifying the leakage characteristics of each isolation damper in the VF system,

in response to questions asked by the inspectors, the licensee initiated CR 197-06151 to address the testing method for secondary containment integrity and verifying leakage

$aracteristics of the dampers. As part of the CR resolution, the licensee performed visual inspections of the dampers, documentad that the dampers were in good condition, and planned to add inspections of the dampers to their surveillance program. In addition, the licensee determined that performing testing with both dampers closed was acceptable. The acceptability of the licensee's testing methodology for VF isolation dampers will be tracked as URI 50461/9701545 pending further NRC review, Conclusions One URI was identified which involved the adequacy of testing secondary containment integrity and damperleakage characteristics of the VF system.

E2.2 Containment Protective Costinas Inspection Scope f37551)

The inspectms reviewed the licensee's actions in response to a June 9,1997, Confirmatory Action Letter Rlil 97-006 for degraded coating Observations and Findinas The inspectors observed testing performed at two test sites selected near welds on the containment liner and one test site on concrete. Testiru performed at one of the test sites near the welds indicated a coating failure because, after instrument tolerances were considered, two of the five pull tests performed at the site had results less than the 200 psi acceptance criteria specified by American Nuclear Standards Institute N5.12,

• Protective Coatings (Paints) for Nuclear Industry.,,

The inspectors interviewed the responsible engineer, rYviewed the licensee's completed engineering evaluat:on for testing, and inspected additional test sites. The inspectors noted that the licensee performed five additional pull tests around the test site where the failure occurred, and performed 15 pull tests at three additional sites near other weld The inspectors considered the licensee's selection of the three additional test sites to be

,epresentative of the one test site which had a failure. The inspectors concluded that the one test failure encountered was an isolated instance, consistent with the licensee's conclusio The inspectors were concemed with past operability on equipment as a result of the impact the peeling containment coatings had on flow through the emergency core cooling system (ECCS) suction strainer during an accident. This matter will be tracked as URI 50461/9701546 pending further NRC review, Conclusions The licensee's pull tests provided assurance that after the degraded coatings were removed, the remaining coating would sufficiently adhere to the containment surface during an accident. One URI was openec to review past operability of equipment as a result of peeling containment coatings which could have effected ECCS strainer flo E Review of Generic Letter (GL) 96-06 Inspection Scope (37551)

The inspectors reviewed selected containment penetrations to determine if the licensee adequately identified penetrations affected by GL 96-06, " Assurance of Equipment Operability and Containment Integrity During Design Basis Accident Conditions." Observations and Findinos On July 2,1997, the inspectors requested that engineering personnel provide the basis for why insulating ihe reactor water cleanup (RT) system piping assured that thermal over-pressurization of the affected containtnent penetrations would not be e conce Engineering personnel re-reviewed the GL 96-06 response and determined that the initial review for RT was insufficient in that the justification used to screen out Containment Penetration MC-86, "RT to Condenser," incorrectly specified that the line was at a high temperature and would cool during a loss-of-coolant-accident (LOCA), hence over-prcssurization due to thermal expansion was not possible. The licensee initiated CR 197-07 015 to document the inadequate justificatio Engineering personnei determined that the piping for Containment Penetration MC-86 is used during start-up to remove (blow-down) excess inventory from the vessel to the condenser and is not in service when RT is in operation. Consequently, the line would be isolated with relativery cool water trapped between the containment inboard and outboard isolation valves. During a LOCA, expansion of the water due to heating would occur and could cause a thermal over pressurization of the piping. Engineering personnel stated they misapplied the assumption that the RT system is operated to allow flow through the piping at a high temperature. The inspectors noted that the omission of Containment Penetration MC-86 may not have been identified had it not been for the inspectors requesting the justification of the RT system penstration The licensee's corrective actions included addir$' g insulation to the piping to prevent over-pressurization from heating, performing an analysis to determine the peak pressure during LOCA conditions and, performing additional reviews of penetrations which credited piping at a high temperature during the initiation of a LOCA. The licensee made a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> notification of a condition outside the design basis per 10 CFR Part 50.72(b)(1)(ii)(B) as a result of this determination. The inspectors noted that the licensee completed the necessary corrective actions to resolve the issu The inspectors' review of other penetrations did not identify any additional concems associated with the justifications for thermal over-pressurization, The inspectors noted i that engineering personnel were knowledgeable of the requirerrents associated with l GL 96-06 and the supporting evaluations were easily retrievable and included sufficient information to complete the revie Qonclusions The inspectors found that the licensee's review of GL 96-06 missed identifying that Containment Penetration MC-86 for the RT system may be impacted during a LOC However, the engineers were knowledgeable of GL 96-06 requirements, and supporting

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t evaluations were easily retrievable and included sufncient information to complete the  !

revie !

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E Review of Reactor Coolant System Lankane Test

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On July 15 and 16,1997, the inspectors reviewed Procedure CPS 9059.01, *Roactor Coolant System Leaka9e Test," which contained instructions for conducting a hydrostatic {

., test on the reactor coolant system piping in accordance with Section XI of the '

A8ME code, Spring 1989. The inspectors compared Procedure CPS 9059.01 to the  ;

A8ME code and concluded that the pmoedure s' ,ee+e,C1 A8ME code requirements for

~

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a visual inspection of the reactor coolant system boundary during the test. The .

quali6 cation records of the individuals responsitde for performing the visual inspection were reviewed by the inspectors, and the laspectors determineo that individuals met ,

A8ME r":M-:=, requirements. The inspectors used drawings to confirm that the valve  !

Ime-un spoolned in Procedure CPS 9059.01 included the required A8ME code class  ;

piping in the test boundar ,

- E7 - Quality Assurance in En98neering Activities -

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E Potentially Defective Fasteners -_ Inspection Scope (36100)

NRC Special Inspection Branch Report No. 99901307/96-01 issued March 4,1997, documented violations and nonconformances of NRC regulations stemming from NRC reviews of vendor corrective actions for defective fasteners supplied by a vendor Accutech (formetty Cardinal industry Products and now a division of B&G Manufacturing Company). The inspectors reviewed licensee corrective actions for Accutech's 10 CFR Part 21 and " courtesy" notification of potentially defective fasteners supplied ,

under Purchase Orders 550170 and 554532 to th'e license t i .- Obstvations and Findinas ,

CR 195-09 002 documented * hat 157 (3/816*1.25") capscrews were purchased from '

Cardinal Industrial Products, which may have received insufficietet heat treatment. The .

Insufficient heat treatment could have reduced the material yield strength to less than half ,

of that of a property heat treated capscrew. Of these capscrews,137 had been removed (scrapped) from stock supplies and 20 had been issued to four maintenance work - ,

requests (MWRs) on nonsafety-related application .

.

CR 196-05-048 documented that 136 (3/816*1") capscrews were purchased from i Cardinal Industrial Products, which may have received insufficient heat treatment. One capscrew had boon removed (scrappeMf) from stock supplies. The remaining capscrews had been issued to 24 jobs on nonsafety-related applications and to three MWRs on safety-related applications (MWR D53720 - Reactor head piping and insulation remova MWR D50488- Replace existing springs and brackets on the EDG room fire damper, and ,

- MWR D60123- Inspect the 1C EDG). Based on review of the affected MWRs, the ,

inspectors concluded that these capscrews had not been installed in a safety-related i application. Additionally, the licensee had concluded that the nonsafety-related i r

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e*w-'y'-'-w- w='-tw+PYv*wtu*-#~TWNW-vemm*h--9W9P-4yma-t--vW ww w ww, pet 'try*wr--demp-v ww-N evwwOww'euFW 9

applications were acceptable based on the minimum expected yield strength being sufficient to hold typicalloads for this size capscre The licensee had placed a hold on ordering additional materials from Accutech, pending resolution of issues raised in NRC Special inspection Branch Report No. 99901307/96-01. The inspectors noted that the scrooning, investigation and corrective actions taken in response to these 10 CFR Past 21 issues complied with the licensee's Procedure UC L4, " Evaluation and Reporting 10 CFR Part 21 Defects and Noncompliance," Revision Qgnclusions Safety-related equipment had not been affected by potentially defective vendor supplied fasteners. The licensee's corrective actions completed in response to this issue were consistent with its administrative procedures and 10 CFR Part 2 IV. Plant Support R1 Radiological Notection and Chemistry (RP&C) Controls

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R1.1 Lift and Weigh of Radioactive Waste Shippina Liner (71750)

The inspectors observed and reviewed the radiologica.1 controls implemented during the preparations for releasing a radioactive waste shipping cask from the site. Radiation protection personnelimplemented proper radiological controls to reduce the dose to workers involved in preparing the cask for shipment and prevented individuals from entering the turbine building truck bay area while the liner was lifted from the cask. The inspectors noted that individuals were familiar with the actions to be takan if the liner fell while being lifted from the cas <-

P2 Status of Emergency Planning (EP) Facilities, Equipment, and Resources P2.1 Tour of Emeraency Operatina Facility Inspection Scope (71750)

On July 7,1997, the inspectors toured the eme ;;ency operations facility (EOF). _ Observations and Findinos The inspectors noted that the main access to the EOF was through a set of airlock door However, an attemate access to the facility for field teams had not been controlled with airiock doors. Emergency planning personnel initiated a work request to install door seals on field team access doors. Procedure RA 17, *lilinois Power Company CPS Emergency Plan implementing Procedure," Section 4.4, * EOF HVAC* specif6d that the EOF area HVAC subsystem provides a positive pressure in relation to the rest of the building and the outside. The inspectors determined that without door seals on the field access team doors, the HVAC system may not have the ability to develop the pressure needed to

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sausfy Pmoodure RA-17. This matter win be tracked as an inspection Follow up item [

(IFl 8044119701547) pending further NRC revie c The inspectors ques 6oned the licensee to determine the last interval for testing of the EOF ventilation system and what soceptance orNorts were utlNaed. Emergency pienning

' personnel performed a review but were unable to determine when the ventilation system  ;

was last tested and could not establish definleve acceptance criterte for testin ~

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CR 1-9747065 was initiated to establish soceptance criteria for testing and to develop j and perform testing of the ventilation syste :

The inspectors noted that a fume hood which may be utilized by fleid teams to perform i rher.2,ws of environmental samples had not been tested to verify the position of the i glass door (sash) to the fume hood assoaisted with minimum and maximum face velocity. -

Emergency planning personnel requested that chemistry personnel perform periodic measurements to ensure the correct sash position would be used during post accident  !

analyses of field sample ' Conclusions Tise inspectors identified one emergency planning weakness involving testing of the I EOF building ventilatio S1 Conduct of Security and Safeguards Activities

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S Protected Area Illumination Inspection Scone 01750)

The inspectors perfumed tours to determine if zones within the protected area were adequately illuminate w i k .

l Observations and Findinos l  ;

On the evening of July 17,1997, the inspectors toured the exterior areas within the protected area. The inspectors observed that temporary lighting was used to provide additional illumination in response to security lighting issues raised by the NRC on l June 18,1997, (NRC Inspection Report No. 50461/97014). The inspectors also

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observed tMt th9 area within the berm for the reactor core isolation cooling (RCIC)

storage tank was poorly illuminated and that tall grass (approximately 3 feet) existed ,

within the berm. The inspectors informed security personnel of the puentiallighting deficiency in the RCIC storage tank berm area at approximately 10.00 p.m. However, no lighting surveys or compensatory measures were taken during the evening of July 17, 1997, or moming of July 18,199 '

_

The CPS Physical Secunty Plan required a minimum illumination level.of 0.2 foot-candles in the protected area. Security personne performed a lighting survey of the RCIC storage '

tank berm area on the evening of July 18,1997, and took compensatory measures when it was determined that 8 of 23 sample locations were between 0.033 to 0.131 foot.

L- candles (VIO 50441/9701548). As a result, the licensee promptly corrected this .

l condition by cutting the grass on July 19,1997. The failure to ensure sufficient  !

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i illumination in the protected area is an addstional example of a violation noted in [

Inspection Report No. 50-461/97014 and therefore, no response is require ! Conclusions '

An addstional example of a violation previously identified in NRC inspection Report No. 5041/97014 involving inadequate illumination of the protect 9d was identified in this

. report. No response to this violation is require V. Mananoment Meetinas X1 Exit Meeting Summary ,

The inspectors presented the inspection results to members of licensee management at the conclusion of the inspection on Au9ust 29,1997. The licensee acknowledged the findings '

presented. The inspectors asked us licensee whether any materials examined during the inspection should be considered proprietary. No proprietary information was identifie X3 Management Meeting Summary On July 31,1997, a meeting was conducted between NRC Region ll1 management and

Clinton Power Station. The meeting was held to discuss the root cause and co Tective actions for the July 22,1907 failure of a safety related circuit breaker to open on deman I On August 19,1997, Chairman Shirley Ann Jackson visited the Clinton site. Chairman Jackson ,

toured the facility and held a brief meeting with licensee management to discuss the results of current personnel performance, the results of procedure improvement efforts and the status of current technical issue .

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PERSONS CONTACTED Licensee

J. Cook, Senior Vice President W. Romberg, Assistant Vice President P Yocum, Manager - Clinton Power Station i

D. Thompson, Manager - Nucioar Station Engineering Department i L Wgley, Assistant Manager - Nuclear Station Engineering Department !

R. Phares, Manager - Nuclear Safety and Performance improvement i J. Pt;chak, Manager Nuclear Training and Support l G. Bnker, Manager - Quality Assurance  !

J. Gruber Director- Corredive Action D. Wood, Director Plant Radiation and Chemistry B. Joyce, Asdstant Plant Manager Maintenance M. Lyon, Assistant Plant Manager- Operations J. Hale, Director - Planning and Scheduling W. Bousquet, Director - P! ant Support and Services M, Stickney, Supervisnr - Regulatory Interface

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INSPECTION PROCEDURL8 USED IP 36100: 10 CFR Pan 21 Inspections at Nuclear Power Reactors IP 37551: Engineering Observations IP 61726:

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Surveillance Observations IP 62707: Maintenance Observation IP 71707: Plant Operations i

. IP 71750: Plant Support and Observations IP 93702: Prompt Onsite Response to Events at Operating Power Reactors

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ITEMS OPENED, CLOSED, AND DISCUSSED Opened 50-461/97015-01a VIO Failure to perform operability determinations within 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> /97015 01b VIO FaiNre to initiate a condition repor /97015-01c VIO Failure to use a 1E portable battery charger during a single cell equalizing charg /97015-01d VIO Failure to secure Class 1E battery charge /97015-02 VIO Inadequate procedure for determining new pilot cel /97015-03 URI Design deficiency of EDG breaker circuit during control room fir /97015-04a VIO Failure to ensure design requirements for VD and HG systems were translated into specifications, 50 461/97015-04b VIO Failure to ensure design requirements for hydrogen mixing compressors were translated into specification /97015-05 URI Review of fuel building ventilation damper testing adequac /97015 06 URI Degrao9d containment coatings effects on ECCS strainer flo /97015-07 IFl Missing seals on EOF field access door Closed 50-461/97015-08 VIO in the RCIC storage tank berm area, it was determined that 8 of 23 sample locations were between 0.033 to 0.131 foot-candles.

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UST OF ACRONYMS ANO INITIALISMS ADS Automatic D$ressurization System ASME American Society of Mechanical Engineers CFR Code of Federal Regulations

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CFM Cubic Feet per Minuto CPS Clinton Power Station CR Condition Report DRP - Division of Reactor Projects

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EDG Emergency Diesel Generator EM Electrical Maintenance 4 EOF Emergency Operations Facility ERAT Emergency Reserve Auxiliary Transformer GL Genedc LeMer HG Hydrogen Mixing Hg Mercury Hp Horsepower IRM Intermediate Range Monitor ISI Inservice inspection LER Licensec Event Report MM Mechanical Maintenance LOCA Loss of Coolant Accident MWR Maintenance Work Request NRC Nucler Regulatory Commission OD Operability Determination PDR Public Document Room PM Preventive Maintenance PMT Post Maintenance Testing RAT Reserve Auxiliary Transformer RCIC Reactor Core isolaticn Cooling 4 RHR Residual Heat Removal -

RT Reactor Water Cleanup SE Safety Evaluation SRV Safety Relief Valve SSTD Solid Sta.e Trip Device SX Shutdown Seivice Water USAR Updated Safety Analysis Report V Volt Vac Volt Attemating Current VC Control Room Ventilation VD Diesel Ventilation Vdc Volt Direct Current VF Fuel Building Ventilation VX Switchgear Ventilation '

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