IR 05000334/1987011

Insp Rept 50-334/87-11 on 870605-0708.No Violations Noted. Major Areas Inspected:Physical Security,Radiological Controls & Housekeeping & Fire Protection.Unresolved Items Noted Re Storage of Transient Equipment

ML20236J716
Person / Time
Site:	Beaver Valley
Issue date:	07/29/1987
From:	Lester Tripp NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20236J704	List: IR 05000334/1987011 ML20236J701
References
50-334-87-11, IEB-86-002, IEB-86-2, IEIN-80-21, IEIN-86-047, IEIN-86-47, IEIN-87-016, IEIN-87-021, IEIN-87-16, IEIN-87-21, NUDOCS 8708060260
Download: ML20236J716 (18)

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U. S. NUCLEAR REGULATORY COMMISSIO REGION I Report N /87-11

' Docket'N Licensee: Duquesne Light Company One Oxford Center 301 Grant Street Pittsburgh, PA 157.79-Facility Name: Beaver Valley Power Station, Unit 1 Location: Shippingport, Pennsylvania

. Dates: June 5 - July 8, 1987 Inspectors: F. I. Young, Senior Resident Inspector, BV-1 l M. P ndale, Resident Inspector, BV-1 Approved by: 8 .-

'L. E. Tffpp, Chief, Reactor Projects Section 3A

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Inspection Summary: Inspection No. 50-334/87-11 on June 5 - July 8, 1987 Areas Inspected: Routine inspections'by.the resident inspectors (148 hours0.00171 days <br />0.0411 hours <br />2.44709e-4 weeks <br />5.6314e-5 months <br />) of licensee actions on previous inspection findings, plant operations, physical security, radiological controls, housekeeping and fire protection, maintenance, surveillance testing, in-office review of LERs and. licensee event Results: No violations were identified. Two unresolved items were identified con-cerning storage of transient equipment in safety-related areas (Detail 4.1.2) and monitoring of containment liner bulges (Detail E). Four previously open NRC un-resolved items were reviewed and closed during this inspection, one remained ope !

870B060260 870730 t PDR ADOCK 05000334.... I G PD l

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TABLE OF CONTENTS Pate Persons Contacted.................................................... 1 Summa ry of Faci l i ty Acti vi ti es. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 Followup on Outstanding

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Items........................................ 1 Plant Operations..................................................... 4 4.1 Genera 1......................................................... 4 4.2 Operations...................................................... 6 4.3 Plant Security / Physical Protection.............................. 9 4.4 Radiation Controls.............................................. 10

'4. 5 Plant Housekeeping and Fire Protection.......................... 10 Maintenance Activities............................................... 10 Surveillance Testing................................................. 11 Rod Control System Anomalies......................................... 11 Containment Liner Bu1ges............................................. 13 In-Office Review of LERs............................................. 14 1 Followup on Information Notices (ins)................................ 15 1 Review of Periodic Reports........................................... 16 1 Unresolved Items..................................................... 16 1 Exit Interview....................................................... 16

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DETAILS i

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! During the report period, interviews and discussions'were conducted with mem-bers of licensee management and staff as necessary to support inspection i activitie . Summary of Facility Activities At the beginning of the inspection on June 5,.1987, the plant was operating at 100% power. On June 8, a power reduction to 30 percent was initiated to ';

complete repairs to the "C" feedwater regulating valve. On June 9, while '

attempting to' increase power from about 33 percent, a reactor / turbine trip occurred apparently caused by dropped control rod (s). Full power operation resumed on June 10. On June 11, a plant shutdown was initiated due to the

, misalignments of two control rods. However, when 70 percent power was reached, plant operators.were able to realign the control rods. The plant was subse- i quently returned to full power late on ' June 11. With the exception of power  !

reductions to approximately 50 percent over the periods of June 26-29 and  !

July 4-6 per load' dispatcher requests, plant operation continued at full power '

through'the end of the inspection perio . Followup on Outstanding Items The NRC Outstanding Items (01) List was reviewed with cognizant licensee per-sonnel. Items selected by the inspector were subsequently reviewed through discussions'with licensee personnel, documentation reviews and field-inspec-tion to determine whether licensee actions specified in the OIs had been

.' satisfactorily completed. The overal1' status of previously identified in-spection' findings were r.sviewed, and planned / completed licensee actions were discussed for those items reported below:

3.1 (Closed) Unresolved' item (84-08-01): Misidentification of a primary sys-tem flange leak on March 10, 198 The licensee was to develop a long term corrective action plan to prevent recurrence of this and similar events. 'In this case, the licensee stated that some contributing fac-tors to the misidentification were: (1) the lack of independent verifi-cation of the source of the leaks (2) the large amount of reactor coolant system (RCS) temperature monitoring bypass piping in the loop area and (3) the lack of access platforms inside the RCS loop cubicle areas to simplify component identificatio Station Administrative Procedure No. 28, Reactor Containment Entries was

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revised to require that all personnel making a containment entry shall be assembled for a pre-work briefing during which the essential informa-tion for the entry will be note Additional controls ensure that when the work activity to be performed is of a non-routine nature, a plant supervisor is required to be present fo* the initial job assessment, i

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Following the completion of the containment entry, all job members are debriefed by the work party supervisor. Both the pre-work and post-entry ,

briefings are formally documented by the license Additional corrective actions which have been proposed by the licensee include (1) removal of the RCS temperature monitoring bypass piping and the installation of in-line RCS temperature detectors and (2) the addi-tion of access platforms inside the RCS loop cubicle areas. The merits of the above modifications are currently under review by the license The effectiveness of the licensee's corrective actions will be routinely reviewed by the resident inspector Although not a contributing factor to the March 10, 1984 misidentifica-tion of the primary system flange leak, NRC inspectors noted that some plant equipment had been informally labeled by plant personnel. The licensee is taking corrective action to assure that appropriate plant equipment is correctly identified in a controlled manner through an approved identification system. A memorandum, dated 1/29/87, from the plant manager, was issued to plant personnel emphasizing the station policy that prohibits the use of magic markers or markers of any type (i.e. , pens, pencils, etc.) for the purpose of labelin The policy further states that for markings in the plant which are beneficial and/or necessary, requests may be made to develop permanent labels. This pro-gram has been initiated; however, many temporary and uncontrolled mark-ings still exist throughout the plan Licensee management acknowledged this concer The implementation of the plant labeling project, cur-  !

rently ongoing, will routinely be reviewed by the inspector. There are no further questions at this time. Unresolved Item (84-08-01) is close .2 (Closed) Unresolved Item (85-06-03): A noted increase in the number of incident reports (IR) caused a concern that the increased volume could possibly lead to a delay in completion of the irs, especially for those of greater importance to plant safety. Significant plant problems and associated corrective actions were to be prioritized and resolved in a timely manne The licensee currently implements a revised system of reporting potential incidents involving station equipment or system ,

Station Administrative Procedure (SAP) Chapter 13, Preparation of Draft Incident Reports (DIR), Unit Off Normal Reports (UONR) and Conduct of Critiques, is the governing procedure. SAP 13 specifies that if the potential incident is not an operational event (as defined in Appendix B of the procedure), then a 00NR shall be initiated. If the potential incident is an operational event (Appendix A of SAP 13), then a DIR is to be written. The process inherently prioritizes events due to the type of report file SAP 13 also specifies the type of review required for each event. Standard UONR and DIR report forms are used for the associated reports, which include information such as event description, technical specification compliance, plant conditions, immediate and long term corrective actions and the failure data (as applicable to the type of report). A recent initiative generated from the Offsite Review Com-mittee recommends that DIRs be completed within 30 days. The latest DIRs

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generally conform to this initiative, however, the Unit 1 staff personnel who currently initiate and process UONRs and DIRs will also assume the same responsibilities for Unit 2 without any current plans to increase the size of the staff. However, the inspector reviewed the completion time for current DIRs and UONRs and found them to be reviewed in a timely manne Implementation of SAP 13, including turnaround time and backlog of UONRs and DIRs for both Units 1 and 2 will be reviewed through routine resident inspections. .This item is close .3 (0 pen) Unresolved Item (85-18-02): Failure of 0 rings and gaskets on low head safety injection (LHSI) pump control rods. The licensee performed a system test to monitor LHSI surge pressure and system leakage per OST 1.11.14 on August 6, 1986 (documented in NRC Inspection Report N /87-06). The licensee felt that a contributing cause of the fail-ure was the loosening of securing nuts which keep the wedge control rods in place. The plant maintenance department recommended that the securing nuts be checked for tightness during the performance of the system's monthly surveillance test. Pending resolution / implementation of the proposed recommendation, this item will remain ope .4 (Closed) Unresolved Item (86-08-02): Evaluate cable routing to demon-strate compliance with UFSAR criteria concerning cont Ni of 9 ble routin The licensee determined that the cable routing was found to conform to UFSAR Section 8.5 commitments, as implemented by BV-1 Cable Installation Specification No. BVS-368 and the current BV-1 specification BVS-3001, The inspector reviewed licensing documents to verify compliance with tne applicable requirements. These are consistent with the design basis requirements for BV-1. This item is close .5 (Closed) Unresolved Item (87-06-05): A 10 CFR 50.55(e) report was filed for Unit 2 concerning the design deficiency in the control room safety related HVAC control circuits on loss of electrical power. A subsequent Unit I review (requested by the resident inspector) found that after actuation of the containment isolation Phase B 60 minute timer, any loss of power and subsequent restoration of power would result in a timer reset which would delay the startup of the control room emergency pres-surization fans. Design Change Package (DCP) No. 611 removed the auto-matic time delay start feature of Unit 1 control room emergency pres-surization (CREP) fans and added an uninterruptible (DC) power supply for the Unit 2 CREP fan timers. This modification also installed an automatic time delay for Unit 2 CREP fans on a Unit 1 containment isola-tion phase B or control room high radiation signals. Because of this DCP to the plant, Unit 1 CREP fans do not start automatically and require manual start by the control room operators. With the new DC power supply on the Unit 2 CREP fans and removal of automatic start feature of Unit 1 CREP fans, the inspector determined the design portion of the unre-solved item had been properly addresse _ _ _ _ _ __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _-

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The second part of this unresolved item was to address the apparent com-munications problem between units 1 and 2. The licensee determined that all Unit 2 Significant Deficiency Reports may not have been reviewed for Unit 1. applicability since no formal documentation from that review was required. The licensee subsequently forwarded all SDRs from Unit 2 to '

the Unit 1 Licensing and Compliance Group for Unit 1 applicability re-views. Licensing and Compliance has determined that at least 50% of the SDRs do not apply. The remaining SDRs will be further reviewed for applicability. A long term plan for Licensing and Compliance has been to consolidate both the Unit 1 and Unit 2 groups. This consolidation l has begun and will continue gradually so that the transition period will ;

have little or no adverse impact on the proper functioning of the grou '

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The adequacy of Unit 1/2 communications will routinely be reviewed by the inspectors. This item is close . Plant Operations 4.1 General 4. Inspector Tours i

Inspection tours of the following accessible plant areas were conducted during both day and ght shifts w:th respect to Technical Specification (TS) compliance, housekeeping and cleanliness, fire protection, radiation control, physical security / plant protection and operational / maintenance admini- '

strative control Control Room -- Safeguards Area

-- Auxiliary Building -- Service Building

-- Switchgear Area -- Diesel Cenerator Buildings

-- Access Control Points -- Containment

-- Fence Line (Protected Area) -- Yards Area

-- Turbine Building -- Intake Structure 4. Storage of Transient Equipment In Safety-Related Arcas NRC Information Notice No. 80-21 (Anchorage and Support of Safety-Related Electrical Equipment), dated May 16, 1980, identified deficiencies with the anchorage of non-seismic Category I ancillary items. The location of this type of equipment may be such that they could potentially dislodge, impact and damage safety-related equipment during an earthquak The inspector reviewed the licensee's storage of transient equipment having the potential to adversely affect safety-related equipment in accordance with the guidance provided in NRC Region I Temporary Instruction No. RI-87-0 I

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.. . 5 Information Notice No. 80-21 prompted the licensee to generate a procedure to address the storage of transient equipmen The plant Maintenance Manual provides general guidance.to

tenance personnel on the subject. The inspector conducted a complete plant tour and found several items of concern:

-(1) many unrestrained, uoright ladders were noted throughout the plant, (2)'large tool cabir:ets (on rollers) .were found unrestrained in some safety-related areas, (3) unrestrained temporary equipment was. stored adjacent to the alternate shut-down panel below the control room, (4) some compressed ai bottles were not adequately restrained in the plant, and (5) free standing equipment in the control room was not a suf -

ficient distance from safety-related equipment such that safety-related equipment could potentially be impacted if the equip-ment were to overturn. These concerns were brought to 13cen-see's management attention. The licensee recognized the need to broaden the scope of their current procedure. A copy of RI-87-03 was provided to the licensee (Attachment-1 as enclosed with this report) so that procedure upgrades can be made to reflect all current NRC concerns. The licensee conducts plant housekeeping tours (with representatives from various station groups) on a monthly frequency, and stated that the guidance provided to them by RI-87-03 u uld be considered as guidelines for the tours. The development and implementation of the necessary administrative controls or other mechaaisms to con-trol the storage of transient equipment in safety-related areas and the effectiveness of the plant housekeeping tours will be reviewed at a future time. This is an Unresolved Item (50-334/87-11-01).

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4. Onsite Safety Committee (0SC) Changes During this inspection period, the licensee received Amendment 110 to the Technical Specifications (TS). This TS change affected the OSC in two ways; first a new member was added with expertise in the instrumentation and controls area, and second, the OSC is now only responsible for review of intent changes to procedures (TS 6.5.1.6) and procedure temporary. changes (TS 6.8.3). TS 6.8.2 now includes a caveat wnich states that non-intent changes shall receive an independent review by a qualified individual and shall be approved by the designated manager or director. This .hange omits the OSC's burden of reviewing administrative changes to procedures and changes that do not affect the purpose of the procedur ]

On April 7, the OSC held a special meeting to determine an acceptable method for implementing the change concerning the review of non-intent changes to procedures. During this meet-ing the c. embers discussed what constitutes an independent re-view, what the minimum qualifications of the reviewer should I

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be, and criteria for determination of intent or non-intent change to a procedur It was recommended that an independent'

reviewer be an individual who is not directly responsible for the work function under review or involved in performance or implementation of the change. The OSC noted that Station Ad-ministrative Procedure'(SAP) 11, Procedure Changes, defines intent and non-intent changes and that the reviewers should adhere to tiis guidance. It was determined that OSC members and Alternates have the qualifkations to perform these review Other personnel will require training prior to making these decisions. An example checklist for these reviews was dis-cussed by the OSC. They found it acceptable and suggested that it be incorporated into SAP 10, OSC, and referenced by SAP 1 The changes associated with implementation of this TS amendment will be reviewed in future inspection Tie inspector noted that during the interim period the OSC is continuing to review all procedure change .2 Operations During the course of the inspection, discussions were conducted with ,

operators concerning knowledge of recent changes to procedures, facility i configuration and plant conditions. During plant tours, logs and records I were reviewed to determine if entries were properly made, and that equip-ment status / deficiencies were identified and communicated. These records included operating logs, turnover sheets, tagout and jumper logs, process computer printouts, unit off-normal and draft incident reports. The inspector verified adherence to approved procedures for ongoing activi-ties observed. Shift turnovers were witnessed and staffing requirements confirmed. In general, inspector comments or questions resulting from these reviews were resolved by licensee personnel. In addition to normal working hours, plant operations reviews were conducted during midnight shifts and weekends on the following dates and times: June 20, 1987, 9:00 am - 11:00 am; June 27, 1987, 4:00 am - 6:00 am; June 30, 1987, 3:30 am - 6:00 am. The inspectors verified that plant operators were alert and displayed no signs of inattention to duty or fatigu . Feedwater Regulating Valve Failure On June 8, 1987, while at 100% power, plant operators noted that the "C" feedwater regulating valve (FRV) was experiencing slightly erratic operations, such that swing 3 in feedwater system flow were evident. An operator was immediately dis-patched to the FRV room to locally observe valve operatio The valve was found to have two (of four total) broken yoke studs, and the actuator's " top hat" was cocked toward one sid The plant subsequently initiated a power reduction to 30% to commence repair work on the valv All FRVs adequately con-trolled feedwater flow and steam generator levels during the l

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load reduction. When 30% power was reached, the FRV bypass valve was placed in service and the "C" FRV was isolate While at 30% power to complete the FRV repairs, control rod misalignments were identified by plant operators (discussed in Detail 7). FRV repairs were completed on June 9, 198 The licensee had experienced numerous, continued problems with their FRVs, for which several attempts (through various FRV modifications) had been implemented. While some improvements have been noted, operational problems continue to occur. Ad-ditional feedwater system modifications are planned for the next refueling / maintenance outage, currently scheduled for November, 1987. The inspector will continue to monitor acti-vities associated with the continuing feedwater system problem . Unlocked Valve On June 14, during a routine tour, a security guard pulled on the lock for safety injection valve SI-30, and found that the lock was not engaged. SI-30 is an isolation valve for the supply line from the refueling water storage tank to the low head safety injection pumps. The Operations Group was informed of the open lock and immediately properly locked the valve and initiated a walkdown of all locked safety related valves. No additional discrepancies were ncted during the walkdown. The I

lock for SI-30 appeared to have been fully closed, however, was not engaged in the locking mechanis Locks and chains are used at BV-1 as administrative controls for important com-ponents required for safety. In response to this event, plant operators were instructed by the Operations supervisor on the importance of ensuring that locks are adequately closa1 and controlled. This appeared to be an isolated incident for which immediate corrective action and investigation was taken by the licensee. The inspector had no further questions at this tim .2.3. ESF Actuation

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On June 16, 1987, with the plant at 100% power, the Train "B" control room emergency bottled air pressurization system l l

(CREBAPS) was actuated. The bottles were audibly determined l to have been discharging into the control room, and control I room emergency air pressure high alarms were received and i l ac. knowledged by plant operators. Plant operations personnel i determined that an actual initiation signal requiring CREBAPS  !

operation was not present, and subsequently attempted to reset  !

the system; however, reset attempts were unsuccessful. An i operator was subsequently instructed to manually isolate the CREBAPS bottles and to remain in the area so that the bottles i could be promptly unisolated if required. Per the Action

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Statement of Technical Specification 3.7.7.1, Control Room l

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Habitability Systems, isolation of the CREBAPS bottles requires compliance with TS 3.0.3 (within one hour, action shall be initiated to place the unit in a mode in which the specifica-tions do not apply).

The licensee suspected a faulty Train "B" reset pushbutto However, several subsequent attempts to reset (including attempts to jiggle pushbutton wires) were unsuccessful. A plant operator was then directed to open the Train "B" control breaker that controls the Train "B" Unit 1/2 interfacing con-tac Reset of the system was then successful using the reset pushbutto The Train "A" control breaker that controls the Train "A" interfacing contact was also opened to prevent a similar occurrence on Train "A". The licensee notified the NRC via ENS phone of the ESF actuation within the required 4-hour time perio The licensee determined that the actuation signal was spurious and was caused by a Unit 2 control room radiation monito An event critique was conducted on June 16 by the licensee which found: (1) that no personnel were actively involved with any procedure or test evolution which may have caused the spurious actuation, and (2) any portion of the associated circuit may have malfunctioned to cause tne actuation. The critique attempted to reconstruct the events to establish the cause for the actuation. The licensee is continuing an inves-tigation to bette. determine the nature of the actuation signal and tt3 benefit of a system modification such that the Unit 2 control room radiation monitors will not cause an actuation on a loss of power (this feature is included in the Unit 1 control room radiation monitor). The status of the licensee's investigation will be reviewed when the licensee event report is submitte .2.4 Inadvertent Gaseous Discharge On June 30, a reactor operator (RO) commenced a gaseous waste 1 system discharge at 3:00 a.m. for radioactive waste discharge authorization gas (RWDA-G) No. 0864 for the "B" waste decay tank (GW-TX-1B). However, about one hour later (3:57 a.m.),

the R0 noted that the strip chart indicated tank pressure was dropping on the "C" waste decay tank (GW-TK-10) instead of GW-TK-18. The discharge was immediately terminated, and the Nuclear Shift Operations Foreman and Radiological Control De- I partment were notified. GW-TK-10 pressure decreased about 8 i psig (from 53 to 45 psig). Abnormal Gas Release Record

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(Authorization No. 0865) was subsequently generated, which j determined that no discharge / dose action levels or technical '

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, -9 Prior to initiating the' discharge, the R0 was talking with the-Radiological Control Department technician. The technician'-

was talking about a sample. evolution for GW-TK-1C.. The licen-see stated that this created'a " mind-set" for the RO, thereby-rpening the valves which resulted in the discharge of GW-TK-1 Licensee immediate corrective actions included terminating the release and generating an abnormal release record. The licen-see plans'to cover this event by the required operator readin . progra Additionally. prior.to any liquid or gaseous waste-discharge in the near. future, the Nuclear Station'0perating Sepervisor is to be contacted so that the details of the dis-charge can be discusse This event is an example of. personnel error resulting in' gaseous release. At 1 east ten similar type of events hav occurred involving gaseous or liquid releases or ESF actuation NRC Inspection Report No. 50-334/87-07 requested that this type of event be addressed since a relative 1, large number of simi-lar events have recently occurred. Thes. events, collectively, are cause for increased concern.' The licensee submittal is expected to also address this particular event, therefore, the associated corrective' actions will be reviewed during NRC re-view of the licensee's respons .3 P_1_ ant Security / Physical Protection Implementation of the Physical Security Plan was observed in various plant areas with regard to the following:

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Protected area barriers were not degraded;

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Isolation zones were clear;

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Persons and packages were checked prior to allowing entry into the Protected Area;

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Vehicles were properly searched and vehicle access to the Protected Area was in accordance with approved procedures;

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Security access controls to Vital Areas were being maintained and that persons in Vital Areas were properly authorize Security posts were adequctely staficd and equipped, security per-sonnel were alert and knowledgeable regarding position requirements, and that written precedures were available; and

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Adequate lighting was maintaine No deficiencies were identifie . _ - _ _ _ _ _ - _ _____-_-__a

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4.4. Radiation Controls 4

.l Posting and control of radiation and high radiation areas were inspecte Radiation Work Permit compliance and use of personnel monitoring devices were checke Conditions of step-off pads, disposal of protactive cloth-ing, cleanliness of work areas, radiation control' job coverage', area monitor operability and calibration (portable and permanent) and person-nel frisking were observed on a sampling basi During a routine plant tour, it was noted that plant personnel were smoking in a posted radiation area. Signs were posted in that area 1 which specifica'ly permits smoking. The area involved is the Radiation  ;

Work Permit Sign-In/ Anti-Contamination Clothing Storage Are The area i is posted as a Radiation Area _and a full body frisk is required upon exiting the area. Since this is a posted radiation area, there is~a potential for area contamination. The inspector brought this concern i to the licensee's management' attention. The licensee acknowledged the inspector's concern and stated that a review of the issue would be in-

-itiated. The licensee was. informed that this area will be reviewed by NRC Region I Health Physics specialists during a future inspectio There were no additional concern .5 Plant Housekeeping and Fire Protection Plant housekeeping conditions including general cleanliness conditions and control and storage of flammable material and other potential safety hr. ards were observed in various areas during plant tours. Maintenance of fire barriers, fire barrier penetrations, and verification of posted-fire watches in these areas were also observed. The inspector.noted plant housekeeping in several areas was declining. The decline was also noted by licensee management who re-emphasized the need to upgrade cleanliness of the plant to station personne . Maintenance Activities The inspector observed / reviewed various maintenance and problem identification activities for compliance with requirements and applicable codes and standards, QA/QC involvement, safety tags, equipment alignment /use of jumpers, personnel qualifications, radiological controls, fire protection, retest and report-abilit The following activities were reviewed:

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MWR No. 860391, HS-P-28 Discharge Isolation Valve Leak

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MWR No. 86125'>, EDG No. 2 Oil Pressure Alarm Contact Failure

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MWR No. 870859, Broken Studs on "C" Feedwater Regulating Valve No significant problems were noted.

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" Surveillance Testing The inspector witnessed / reviewed selected surveillance tetts to determine l whether properly approved procedures were in use, details were adequate, test i instrumentation was properly calibrated and used, technical specifications were satisfied, testing was performed by qualified personnel and test results satisfied acceptance criteria or vere properly dispositioned. The following surveillance testing activities were reviewed:

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OST 1.1.1, Control Rod Assembly Partial Movement Test

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OST 1.6.2, Reactor Coolant System Water Inventory Balance OST 1.6.4, Measurement of Controlled Leakage

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OST 1.6.5, RCS Loop Stop Valve Breaker Alignment Verification

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OST 1.6.7, Accident Monitoring Instrumentation Channel Test

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OST 1.11.1, Safety Injection Pump Test No discrepancies were rote . Rod Control System Anomalies 7.1 Misaligned Control Rods on 6/9/87 On June 9, while at about 30% power (for repair work to the "C" feedwater regulating valve, Detail 4.2.1), operational anomalies were noted in the rod control system during control rod manipulation activities. While moving Control Bank D control rods, it was noted that a total of six control rods were misaligned, four in Control Bank D and two in Shutdown Bank B. Plant operators declared the six control rods inoperable due to rod misalignment at 4:30 a.m. on June 9, and therefore entered the Action Statement of Technical Specification (TS) 3.1.3.1 (Reactivity Control Systems). TS 3.1.3.1 requires that when more than one control rod is misaligned by more tl.an +/- 12 steps, the unit is required to be placed in a Hot Standby condition within six hours. Abnormal Operating Procedure (A0P) No. 8, RCCA Misalignment, was entered and used to recover the six control rod All control rods were recovered by 9:15 a.m. , and OST 1.1.1, Control Rod Assembly Partial Movement Test, was subsequently performed successfully to verify the operability of all control rod The licensee determined that the cause for the misalignment and rod con-trol problems was due to overheating of logic circuitry in the control rod drive system cabinets. Additional ventilation through the use of temporary fan units has been provided to assure localized heat remova Logic circuit overheating has been a recurring problem at BV-1, resulting from high ambient room temperatures. The licensee's action has typically been to open the cabinet doors and provide the temporary fans to improve

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air circulation, as was done in this case. The licensee is continuing their root cause investigation and long term corrective action plan for the continuing rod control system proble .2 Dropped Control Rod (s)  !

Following the installation of temporary fans in the Rod Control System cabinet room and satisfactory test results of OST 1.1.1, the licensee initiated a power ascension to full power. As the reactor operator in-itiated a demand signal to withdraw control rods on Control Bank 0 to maintain primary system temperature, control rod (s) apparently dropped into the reactor core, and the resulting high negative neutron flux rate (2 out of 4 logic for actuation) caused a reactor / turbine trip from 33%

power at 11:45 a.m. Plant operators immediately entered Emergency Operat-ing Procedure No. E-0. Recovery actions from the plant trip were ob-served by the resident inspectors. Two unexpected events occurred during the transient: (1) a momentary undervoltage condition on the 4160 Volt DF emergency bus resulted in an automatic start of the No. 2 emergency diesel generator, however, the bus voltage did not degrade to the point where the emergency diesel generator was required to assume loads on the DF emergency bus, (2) plant loads supplied from the 480 Volt emergency stub bus No. 8N were found to be de-energized due to an undervoltage condition on the stub bus. The loads were subsequently restarted suc-cessfully. The licensee stated that the cause of the two undervoltage conditions following the plant transient was that the buses involved were maintained at the low end of their respective allowable voltage valuet, and when the trip (and subsequent automatic bus supply transfer) occurred, a slight decrease in bus voltage was experienced. However, as noted above, the degraded bus voltages were recovered immediately upon auto-matic transfer of the buses. The plant was returned to full power opera-tion following the trip on June 11, 198 The licensee's immediate corrective action for the rod control system problems was to replace 4 damaged circuit cards in the rod control cabinets (one logic cabinet card and three power cabinet cards). Further investigation into the circuit cards, which were removed and sent to an outside vendor, indicated that one of the firing circuit cards from the power cabinet was damaged. The remaining three circuit cards were un-damage Since the licensee had determined that the problems were a direct result of excessive temperatures in the rod control room (greater than 90 F), portable air conditioning units were purchased and installed in the individual cabinets, creating localized cool areas. Additionally, a temporary air conditioning unit was ordered by the licensee to cool the entire Rod Control Cabinet Room. There is currently a proposed design change package, No. 750 - Switchgear Ventilation Modification, which is planned to include modifications to provide permanent cooling to the rod control system cabinet room. The effectiveness of the short term cor-rective action and implementation of the long term corrective action will be reviewed through routine resident inspections. At the end of this

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l inspection period, the control rod cabinet room temperatures were being maintained in the low'80s. The licensee event report associated with this event will be reviewed by the inspector when issue . 3 Misaligned Control Rods on 6/11/87 On June 11, 1987, at 12:55 a.m., the licensee determined that two control l rods were misaligned, both in Control Bank D. The Rod Position Indica--

tion (RPI) system display read higher than that allowed by plant Techni-cal Specifications (misaligned by more than 12 steps above the group demand counter). Further investigation by taking direct primary RPI voltage readings showed that the control rod positions were actually

, lower than that allowed by plant TS (indicating a misalignment of more than 12 steps below the group demand counter). At 12: 55 a.m. , when the two associated control rods were declared inoperable due to rod mis-alignment, the Action Statement of TS 3.1.3.1 (Reactivity Control Sys-tems) was entered which required that the unit be placed in a Hot Standby condition within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. A plant shutdown was initiated at 4:15 Plant shutdown activities were observed by the resident inspector. When 70% power was reached, readings were again taken on both the control room RPI displays and the primary RPI voltages for control bank D. All read-ings were found to be within r: ecifications and compared favorably with the group demand counter. The TS limiting condition for operation was then satisfied, and the plant shutdown was terminated at 5:25 a.m. while at 70%' power. Full power operation was resumed late on June 11. Based on experience over the past several years, the RPI system is known to be temperature sensitive during power increases and the licensee has been trying to develop more representative primary voltaga curve (s) for de-termining position to preclude occurrence of this type of problem. The inspector will continue to review licensee activities in this are . Containment Liner Bulges ,

Unresolved Item No. 82-11-03 was opened on June 4, 1982, pertaining to licen-see investigation of containment liner plate bulges. Licensee Event Report 82-13/99T identified the discovery of five bulges in the containment line The licensee subsequently initiated a surveillance test, BVT 1.3-1.47.6 (Con-tainment Liner Bulge Monitoring), to measure the size and growth of the bulge The initial performance of the test (in July, 1983) documented 52 bulges throughout the containment building. Of those documented, 14 were eithe*

fully or partially measured, and used as baseline data for comparison purpose The others were inaccessible. A second test was performed in December, 198 That test documented 101 bulges. The licensee attributed the increase in number to a more complete survey of the containment. Twenty-three (23).of the 101 were either fully or partially measured. The third performance of BVT 1.3-1.47.6 was in July,1986, which documented the existing 101 bulges (no new ones), also fully or partially measured 23 bulges. Unresolved Item 82-11-03 was closed when this BVT was included in the surveillance schedule for the fourth refueling outage by NRC Inspection Report 50-334/83-2 _-_ - ________ -

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BVT 1.3-1.47.6 measures the size of the bulges with the use of an aluminum angle reference tool anchored to the containment steel liner with magnet The measurements are taken by depth gauge. The estimated accuracy of this type of measurement is 1/16". The total of all comparisons made during the retest yielded 2.2 oercent with greater than 1/16" difference in the size of the bulges as cor red to the last measuremen The licensee therefore con-cluded that the test results did not ref! :t a trend of increased or decreased bulge size, and that the monitoring of the bulges should be discontinue The inspector reviewed the test results comparisons for BVT 1.3-1.47.6. Fol-lowing discussions with NRC Containment Systems personnel, a preliminary de-termination was made that the BVT should not be discontinued. The bases for the determination was: (1) the operating condition for the BV-1 containment is sub-atmospheric, thus placing additional loads on the containment steel liner and (2) degradation of plant components should continue to be monitored such that imminent failure of the component could be prevented. Additionally, NRC personnel questioned whether ultrasonic testing has been performed on any of the bulges to verify that the containment liner has positive anchorag No such testing has been performed or documented to date. The licensee should either perform engineering evaluations to determine whether the BVT can safely be discontinued or continue these measurements. Pending either completion of the licensee evaluation or a decision to continue the g rformance of this test, this is an Unresolved Item (50-334/87-11-02).

9. Inoffice Review of Licensee Event Reports (LERs)

The inspector reviewed LERs submitted to the NRC Region I Office to verify that the details of the event were clearly reported, including the accuracy of the description of cause and adequacy of corrective action. The inspector determined whether further information was required from the licensee, whether generic implications were indicated, and whether the event warranted onsite followu The following LERs were reviewed:

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LER 87-09, Automatic Flow Diversion of Auxiliary Building Exhaust (De-tailed in NRC Inspection Report 50-334/87-07)

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LER 87-10, Inoperable Chlorine Detection System (Detailed in NRC Inspec-tion Report 50-334/87-10)

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LER 87-11, Automatic Flow Diversion of Auxiliary Building Exhaust (De-tailed in NRC Inspection Report 50-334/87-07)

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LER 87-12, Reactor / Turbine Trip due t> Low Auto-Stop Oil Pressure (De-tailed in NRC Inspection Report 50-334/87-07)

It should be noted that the licensee submitted LER 87-11 as a voluntary LE NRC Inspection Report No. 50-334/87-07 cited the failure to make the required 10 CFR 50.72 report as a violation of NRC requirements (87-07-04). LERs are used to collect, collate, store, retrieve and evaluate information concerning licensee event Providina accurate information is necessary for proper

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, 15 classification of events. Therefore, NRC followup on violation 50-334/87-l 07-04 will include a review of licensee resolution and corrective action for the information provided for LER 87-1 . Followup on Information Notices (ins)

10.1 IN 87-16, Decraaation of Static "0" Ring Pressure Switches IN 87-16 was issued to alert licensees of the potential for degradation of certain static "0" rings (SOR) pressure switches with Kapton diaphragms caused by exposure to ammonia and other chemicals that may be present in the process media. The licensee reviewed the IN for applicability and initiated an engineering memorandum (EM) to determine whether the concern was applicable to Unit IN 86-47 and IE Bulletin No. 86-02 also addressed problems associated with SOR pressure switches. The lic-ensee developed a list of 13 SOR pressure switches used at BV-1, none of which used Kapton as the diaphragm material. Additionally, of the 13 pressure switches, the licensee identified that 4 of them were not listed in the material equipment list (MEL) under the 50R, Inc. vendor code and therefore, were not identified during the IN 86-47 review or the response to IE Bulletin No. 86-02. However, the licensee determined that the switches are not used in safety-related applications, nor have SOR model numbers 102 or 103 (as noted in IN 86-47 and IE Bulletin N ). Therefore, the conclusions of the previous reviews were not affecte Based on the above, the concerns ident :ied in IN 87-16 are not applicable to BV-1 and no further action is require .2 IN 87-21, Shutdown Order Issued Because Licensed Operators Asleep While On Duty This information notice was issued to licensees to reaffirm the prin- 1 ciples of high standards of control room professionalism and operator awareness that are essential to ensure that a nuclear facility is oper-ated safely, and in a manner in which will protect the health and safety of the public. The inspector discussed this item with Operations De-partment personnel and reviewed their actions in regard to this matte The contents of IN 87-21 were reviewed by the licensee and made available to all BV-1 assigned licensed shift personnel. A copy of the daily training roster (which reflected operator attendance of the review of IN 87-21) was reviewed by the inspector. Further, the Nuclear Station Operating Supervisor reviewed the BV-1 administrative procedures and determined that Operating Manual Chapter 48, Conduct of Operations, ade-quately defines the shift workers responsibilitie Licensee management also conducted recent spot checks which have indicated that similar problems, as noted in IN 87-21, do not exist at BV-1. DLC has consist-ently held a strong position regarding this topic. An announcement to all station personnel, signed by the Senior Vice President - Nuclear, was posted on April 14, 1987, reaffirming its policy with respect to sleeping on duty. The licensee intends to vigorously enforce this policy s - ---

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, 16 l at'BVP The inspector performed various backshift and weekend inspec-i tions to verify that this policy has been implemented (see Section 4.2).

No concerns were identifie . Review of Periodic Reports Upon receipt, periodic reports submitted pursuant to Technical Specification 6.9 (Reporting Requirements) are reviewed. The review assessed whether the reported information was valid, including the NRC required data, and whether the results and supporting information were consistent with design predictions and performance specifications. The inspector also ascertained whether any reported information should be classified as an abnormal occurrence. The following periodic report was reviewed:

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Monthly Operating Report for Plant Operations from May 1-31, 1987 No deficiencies were note '12. Unresolved Items Unresolved items are matters for which more information is required in order to determine whether they are acceptable items or violations. Unresolved items identified during this inspection are discussed in paragraphs 4.1.2 a,d . Exit Interview Meetings were held with senior facility management periodically during the course of this inspection to discuss the inspection scope and findings. A summary of inspection findings was further discussed with the licensee at the conclusion of the report perio I

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Region 'I Temporary Instruction No. RI 87-03 , Revision 0 i STORAGE OF TRANSIENT EQUIPMENT IN SAFETY-RELATED AREAS l

Purpose The purpose of this temporary instruction (TI) is to provide guid-ance to Region I Resident Inspectors for reviewing the storage of transient equipment having the potential to adversely affect; safety- ,

related equipment. This is applicable for reactor sites with operating licenses (OL's) or sites expected to get OL's within the next yea Objectives The following objectives are to be met by this review: Ascertain the status of licensee administrative controls (or other type of facility procedures) in the subject are . Determine the proper' implementation of administrative controls (or other type of facility procedures) for the 50bject are . Identify if deficiencies exist independent of whether or not facility procedures cover the subject are . Where deficiencies exist, assess the ifcensee corrective action process with respect to the applicable NRC Information Notice (IN)

Backgrouna The General Design Criteria (GDC) (10 CFR 50 Appendix A) coupled with the updated Safety Analysis Reports (USAR's) provide a basis to assure that transient equipment does not adversely affect the safety function of struc-tures, systems, and components that are safety related. Transient equip-ment includes: dollies; block and tackle; filled gas bottles; heavy equip-ment, (stationary or on rollers) such as welding machines or tool cabinets; scaffolding,'(stationary or on rollers); and, temporary office spaces along with housed furniture, cabinets, etc. GDC Criterion 2 states

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that: " structures, systems, and components ... shall be designed to with-stand the effects of natural phenomena... The design basis for these structures, systems, and components 'shall reflect: ... appropriate com-binations.of the effects of normal.or accider.t conditions with the effects of the' natural phenomena....". Criterion 4 states that: " structures,

. systems, and components ... shall be designed to accommodate the effects E of and be compatible with the environmental conditions associated with

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normal: operation ... and postulated accidents ....These structures, systems, .

and components shall be appropriately protected against dynamic effects, including the effects of r.if ssiles, pipe whipping, and discharging fluids that may result from equipment failure and from events and conditions out-side the. nuclear power plant...."

To meet the criteria, the USAR's generally consider major structure design initiatives such as separating redundant safety-related pumps and valves between trains with seismic Category I walls. However, for the more subtle issue, storage of transient equipment, the USAR's or facility pro-cedures may not provide control of transient equipment (or ' temporarily stored equipment) that could become a missile as a result of. a natural phenomenon or industrial acciden The NF.C's Information Notice No. 80-21, dated May 16, 1980, " Anchorage and Support of Safety-Related Electrical Equipment" identified deficiencies with the anchorage of non-seismic Category I. ancillary items (transient equipment 'plus installed systems such as non-safety related ductwork).

Further, their locations may be that they could potentially dislodge, im-pact, and damage safety-related equipment during an earthquake. This TI focuses-on the transient equipment, not the non-seismic Category I installed system Recent team inspections conducted by Region I have identified that some licensees appear to have a general disregard for, and lack of formal controls on, proper anchorage of transient equipment temporarily stored in safety-related area INSPECTION REQUIREMENTS During the next full (one month) inspection period after the date of this TI, the resident inspectors will review the below listed items during their facility tour Projects Section Chiefs may authortre substitution of the review for some elements of inspection procedures 71707 and 4270 .

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' Identify and' review for adequacy estabitshed administrative controls <

ir other facility procedures that control the storage of transien !

aquipment in safety-related area l Verify proper implementation of applicable procedures established as noted in 2 above. If no procedures were estabitshed, identify if i equipment is properly stored and assess management awareness of the issue and question why procedures do not exist to control such .

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The enforceability of the above-referenced CDC traditionally has been to use 10 CFR SC Append!x.50 Criterton 3. " Design Control " and the NRC-approved Quality Assurance Plan, which assures that applicable regulatory requirements and design-bases are correctly translated into drawings, procedures, and instruction However, the main focus of this review is to survey Itcensee status and to iden-  :

tify deficient.ltcenseesp ' .

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Inadequate corrective actions in this~ area could also be enforceable through 10 CFR 50 Appendix B Criterton XVI. Spectfic guidance for each of the above-noted inspection items is addressed belo The general guidance ~ to be' applied is that all equipment / objects in safety-related areas shall be securely anchored or removed when permanent plant equipment in that area is required to be operable. ' . Licensee review of IN 80-21 should have confirmed that procedural controls were estabitshed for the control of transtent equipment in safety-related areas. Further, Itcensee review should have vertfled proper implementation ri these control . Little formal guidance is delineated on the adequacy of control for the storage-of transtent equipment. Considerable inspector judgement is war-rante Restraints should be' reasonable and substantial for equipment or, J rollers; f.e., heavy chain or wire rope. Household string would obylously l be unacceptable. Free standing equipment (restrained by friction) should be a sufficient distance from safety-related equipment such thaf tf it overturns, it would. not impact safety-related equipment; 1.e., twice the ,

height of the free-standing objec : ,

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Aditional initiatives would be to designate storage areas; 1.e, for Reactor Building refueling equipment.or house store equipment in ~3 seismic Category I walls. Ideally, all transient equipment should  !

be removed from safety-related areas. Scaffolding should not be - . erected over safety-related equipment required to be operable.

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and documented in the attachmen ,. .

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L Licensee failure to' follow eshbitshed procedure controls should 'be  :

discussed with licensee management. For plants that are .in outages, I inspector findings should be. tempered by safety significance while in

. cold shutdown. Inspector-noted negative observations from the prio power operation period should be discussed with Itcensee management with forewarning on the upcoming power operation period. Readiness

. Assessment Team inspections should focus on this area in preparation for plant startup. Enforcement should be consistent with safety significanc .m : :. .

4. .The main focus of this inspectio,n is to survey Itcensee status in controlling this. area. It is also to identify Itcensees with significant deficiencies in the interest of enhancing safety. As a minimum, outstanding issues should be followed by an unresolved item and appropriately documented in the inspection report. _

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