IR 05000293/1990013
| ML20055F249 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 06/29/1990 |
| From: | Johnson P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20055F247 | List: |
| References | |
| 50-293-90-13, NUDOCS 9007160142 | |
| Download: ML20055F249 (24) | |
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U.S. NUCLEAR REGULATORY COMMISSION REGION I > Docket No.: 50-293
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Report No.: 50-293/90-13 Licensee: Boston Edison Company 800.Boylstor. Street-Boston, Massachusetts 02199 m Facility: Pilgrim Nuclear Power Station Location: Plymouth, Massachusetts t Dates: May 1 - June 11, '1990 Inspectors: J. Macdonald, Senior Resident Inspector A. Cerne, Resident Inspector C. Carpenter, Resident Inspector W. Olsen, Re:ident Inspector (Temporary Detail) A. Finkel, Senior Reactor Engineer, RI Approved by: [ (A h bb9 9e P. J h son, Acting Chief, Reactor Projects Section 3A ~ late'
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Inspection Summary: Inspection on May 1 - June 11,1990 (Report No. 50-293/90-13) Areas Inspected: Routine safety inspection of actions on previous inspection-findings, plant operations, security, maintenance-and surveillance, engineering support, radiological controls, and safety assessment / quality verificatio :
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9007160142 900629 :n
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PDR ADOCK 05000293-Q PDC i:
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h EXECUTIVE SUMMARY i
< General Conclusions on Adequacy, Strength, or Weakness in Licensee Programs :
, Operations: Operator response to the May 13 reactor scram was well con-l trolled. Conservative actions were taken and appropriate emergency . t operating procedures were utilized. Additionally.. prompt operator response quickly terminated the May 31 reactor coolant system water chemistry transient which resulted from a degraded condensate'demineralizer, l- 1 In contrast, the operations staff failed to ensure the proper completion of a weekly resioual heat removal (RHR) system leakage surveillance proce-
dure.- A licensee-identified violation was noted as a result of this event (Section 4.2).
Radiological Controls:- Management has been very effective in the develop- - ment of a strong station commitment to the ALARA progra Initiation of the source term reduction program represents a noteworthy licensee initiative to ensure continuing long term decline in personnel and site exposures, ,
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Maintenance / Surveillance: The Master Surveillance Tracking Progr:m (MSTP) was well managed and properly implemented. Reviews of surveillance proce-
.dures,have been performance based and have resulted in technical and regu-latory bases content improvement Station response to the repair of the "E" condensate demineralizer was excellent. Interdepartmental communication and coordination of activities necessary to support the repair were strongly eviden '
However,: continuing diesel fire pump component deficiencies indicate weak-nesses in maintenance request implementation and root cause determinations (Section 4.1).
Emergency Preparedness: The internal EP drill conducted on May 30, in preparation for the annual graded exercise, was comprehensively critiqued by.the licensee.. Deficiencies were effectively identified by drill con-trollers and were demonstrated during the conduct of a June 12, 1990 fol-lowup drill to have been effectively correcte Security: Station security performance remains a licensee strength. On-going-security hardware upgrades and programmatic enhancements in progress were indicative of management commitment to the maintenance of a strong security progra S_afety Assessment and Quality Verification:
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During this inspection period the licensee effectively identified several deficient conditions (licen-see-identified: violations), further demonstrating improvement in the abil-ity to be self-critical,
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a Executive Summary 2
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Licensee Event Reports continued to be very well developed', comprehensiv . documents, Notwithstanding improving self-identification awareness and processes, weaknesses have been identified in the licensee's ability to consistently
. ensure that resultant evaluations are routed to appropriate review groups to ensure that a multi-disciplinary review is realize Engineering / Technical Support: Staff personnel performing the engineer :
ing, quality and procurement functions wer knowledgeable of and were per-forming in compliance with site procedure ' Violations A licensee-identified non-cited violation was noted involving the licen-
.see's failure to comply with procedures, in that a surveillance procedure was not completed and was improperly sign off as complete. The surveil-lance was of minor safety significance and prompt corrective actions were taken(50-293/90-13-01, Section 4,2),
A licensee-identified non-cited violation was noted involving the licen-see's failure to perform a required Technical Specification surveillance
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within the specified periodicit The surveillance was of reinor safety significance and prompt corrective actions were taken (50-293/90-13-03, Section 7.7.1' Unresolved Items An unresolved item was identified pending determination by the licensee of the status of the Q parts and materials stored in the B level storage L drawers for. compliance with program criteria (50-293/90-13-02, Section 6,3),
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: TABLE'0F CONTENTS e PAGE . Summa ry o f Fac i l i ty Ac ti vi ti e s . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . - I'
O 2; ' Plant Operations-(IP 71707,93702,92702,90712)*.............-....... I ! 2.15. Control Room.a'nd Station Tours............'..;.................... 1l
.2.2.' Safety Sy ;)rs Review..........................................,- 12?
2.3- Review of c '~ l n g ' 0p e ra t i o n s . . . . . . .. . . . . . . .- . . . . . . . . .. . . . . . . . . . . . . . - 2t a? 2.4 Operatio'ni, Safety Findings.....................................= , 3 2. 5 ' I n op e ra b l e Eq u i pme n t . - . . . . . . . . . . . . . . . . . :. . . . . . . . . . . . . . . . . . . . . . . . . 3 , F< 2,6 Review of. Plant: Events........................................... 3_ l
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2. ; Automatic . Reactor Scram Due to Load Rejection. . . ... . . . . 3-
. 2. : Partial: Reactor Water Cleanup System Isolation. . . .-. . . . e4
?2. . Degraded "E" Condensate 0emineralizer. . . . . . . . . . . . . . . . . 4 c3,- Radiologi cal , Cont rol s ( : P 71707) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. - '5
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' - " 3.1 Installation-offC osed Circuit Television (CCTV) Cameras......... 3.2. SourceLTerm Reduction Program...........:..'.................-...... 5' l 6 1 L 3.3 Access Control Program....................-...................., 6
} ! ' Maintenance / Surveillance:(IP 37828,61726,62703,93702).............. 7
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L Di esel Fi re~ Pump i Batt'e ry Probl em. . . . . . . . . . . . . - . . . . . . . . . . . . . . . . . .
, .21 Missed RHR Piping Temperature Surveillance......................
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7 4. 3 Ma ster ' Surveillance Tracking Program. . . . . . . ... . . . . . . . . . . . . . . . . . . .
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'lg 1 ' Security 1(IP'71707)............................................,..... 10 j .' 5.1 Observations-ofLPhysical Security............................... 10 [ Engineering / Technical Support (IP 37828, 38701, 38702)............... : 10 - 4 4 i ; 6.1~tInoperable Contact on' Loss of Field Relay....................... 10 '(
6.2 l Procurement Programi............................................ 11 ]' 6.'3 ~ Receipt,1 Storage'and Handling of Equipment and Materials '
. Program....................................................... 12-t
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>4> ' Safety Assessment / Quality Verification (IP 35502, 40500, 92700, 92702,
," --92709). .............................................................. 1 l 4 y 7.1 (Closed) Unresolved Item 89-05-01, Safety Evaluations were not'
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, Performed Prior to Di sabling Control Room Annunciators. . . . . . . . 13 ,
'7.2 (Closed) Unreso~1ved Item 87-53-02.4, Certain Procedures ~
Governing Degraded Plant Conditions May Not Be Sufficient to
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Clearly Guide Recovery Sctions f rom Events of. This Nature.. . . . 13 g 7.3 (Closed) Unresolved Item 89-08-02, Inadequate _ Engineering Evaluation and Poor Controls for Application of Furmanite in RCIC System Check Va1ve....................................... 14
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7.4 (Closed). Unresolved Item 87-53-04.1, Develop a Station Procedure , for the Testing of the.Startup Transformer.................... '14 - !
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7.5 (Closed) Unresolved Item 89-07-04, Review Licensee Policy on 1 Rehanging of Danger Tags Without Independent Verification..... 14 l 7.6 Improper.Use of a Maintenance Work Activity..................... ' 15 . ! 7.7 Lice.1see Event Reports.-......................................... 15 j 7. LER 90-07............................................. 16 7.8 Licensee Self Assessment Capability and Quality Assurance Program Implementation......................................., 16 , m Review of Periodic and Special Reports (IP 71707). . . . . . . . ... . . . . . . . . . . 18 , , Management-Maet<nos (IP 30703, 40500)................................ L18
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il , ATTACHMENTS
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Attachment I: c Ions Cr 'ed l
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- *The NRC Inspe . 'ac ' ^assection procedure (IP) that was used as inspection-lt -guidance-is li .: 0 a,- .,plicable report sectio .
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DETAILS 1.0 - Summary of Facility Activities Pilgrim . Nuclear Power Station (Pilgrim, the licensee-or thel plant) was at 84% power at the' beginning of this vaport period, following restart fro the plant's spring surveillance outage. The plant reached 100% power on-May l', 1990c n e r was reduced to approximately 48% on May 2 to perform a main conden e bach a n, and the reactor was returned to full power on May-4. On May'14 at 4:03 p.m., the reactor automatically scrammed from 100% j power due to e tbrbine trip / gen ctor lockout as a result of a fault on 1 one of.the two;345 KV offsite electrical distribution lines (section 2.6.1), -)
(The licensee notified the NRC Operations Center via the Emergency Notifi- Lj cation System (ENS) in accordance with 10 CFR 50.72.) ;
q Fol. lowing troubleshooting into the cause of the turbine trip / reactor ~ scram, I the reactor was made critical at 2:50 a.m. on May 15, 1990. The turbi_ne i generator was synchronized to the grid at 10:30 a.m, on May 15. On May i 16, power: ascension was held to 75% power due to spurious opening of the- j No. 1 bypass valve. On May 20 at 8:00 a.m., the turbine generator was
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removed from the grid to troubleshoot and repair the cause of the bypa's s valve opening. On-May 22 at 4:00 a.m. the turbine generator was synchron-ized to the grid and the plant reached 100% power on May 2 j Ij i On May 30, the licensee conducted a dry run emergency preparedness dril Notification via ENS to the.NRC was also made on June 6 when a failed relay 1 and blown fuse in the Primary Containment Isolation System (PCIS) resulted ~ i in~ a Partial Group 6 isolation (reactor water cleanup system) (section 2.6.2). At the close of this report period, the plant was at 100% powe ]i 2.0L Plant Operations-2.1'LControl Room and Station Tours l The inspector observed plant operations in the following areas during L regular'and.backshift hours: -{ Control Ro'om Fence Line'(Protec.ted Area) Reactor Building Intake Structure <
. Diesel Generator Building Switchgear Rooms Turbine Building i1 ' Control room instruments were observed for correlation between chan- ,nels, proper functioning, and conformance with Tecnnical Specifica- !
tions. Alarms received in the control room were reviewed.and dis- . cussed with the operators. Operator awareness and response to these j conditions:were reviewed. Operators were found cognizant of boaid ' and plant conditions. Control room and shift manning were compared l L with Technical Specification requirements. Posting and control of 1
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L radiation, contaminated and high radiation areas were inspected. Use
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e of1 and compliance'with1 radiation work permits and use of required personnel monitor.ing devices were checked. Plant. housekeeping con-L' t trols, including control of flammable and other hazardous materials, were observed. ' Ouring plant tours,. logs and records were reviewed to
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ensure compliance with station procedures, to determine if entries were correctly made and to verify correct communication of equipment status. -These records included various' operating logs, turnover-sheets, tagout and lifted lead and jumper logs. Inspections were performed-on backshif ts including May 1-3, 7-10, 11, 14-18, 21-24, 29-31, and June 4-7, and 11, 1990. " Deep backshift" inspections were
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conducted as follows: 1 Time Date U 5:45 p.m. - 7:05 /13/90 12:15 :00 /14/90 10:15 p.m. - 11:45 /16/90 ;
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' ' Pre-evolution. briefings were noted to be thorough with' appropriate if questions;and answer The operators appeared to have good knowledge i of plant conoitions. .No unauthorized reading material was observed 'l and food, beverages and hard hats were kept away from control panels.- 1 n, -; Operator response following the reactor scram.on May 13 (section 2.6.1-) was appropriate and conservative; the event' was well handled p _by the control room cre l 2.2 Safety System Review
; ' , Portions of the emergency diesel generators, reactor core isolation !
cooling, core spray, high pressure coolant injection, residual heat
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removal and safety related electrical systems were reviewed to verify.
- . proper alignment and operational status ir'the standby mod The b
" review' included verification that (i) accessiblesmajor flow path
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i valves were correctly positioned,-(ii)' power supplies were energized, ; J(1ii) lubrication and component cooling were proper, and (iv) com- i O' ponents were operable based on a visual inspection of equipment for ! leakage and general conditions. No violations or safety concerns-were identifie ..
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2.3 Review of Tagging Operations-The following tagouts were reviewed with no discrepancies noted: ,
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Tagout Description i 4 90-1-12 Motor Operator M0-220-1 I 90-8-7 Rollamatic Filter 90-33-19 "A" EDG Sprinkler 90-45-69 Tip Drive Mode Switches
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b 12.41 0Jerational Safety Findings Licensee. administrative control of off-normal system configurations, , byL the use of temporary modifications and tagging procedures, was in comp'liance with procedural instructions.and was consistent'with plant j safety. Backshif t inspections- found operators to be alert and atten -
,, tive. Overall plant cleanliness and material condition continued to be good with the exception of several steam leaks.that were still
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presen .5 Inoperable Equipment Actions taken.by plant personnel during periods when equipment wasi ' inoperable were reviewed to verify that: technical specification limits were met; alternate surveillance testing was completed satis-factorily; and, equipment was properly returned to se*vice upon com-pletion.of repair This review was completed for the following-items: Date Out Dats In System
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5/4 5/9 Diesel Fire' Pump 5/9- '5/16 Diesel Fire Pump 1 5/21 5/25 Control Rod 42-07 5/27 "D" SRM .i 5/31 6/5- "E" Condensate Demineralizer a 2.6 Review of Plant Events
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2. Automatic Reactor Scram Due to Load Rejection On May113 at 4:03 p.m., the reactor automatically scrammed-from 100% power due to a turbine-trip / generator lockout as a result of a fault on one of the two 345 KV offsite elec- - trical distribution lines. The fault on.-the grid resulted '! in a loss of generator field and ultimately a load rejec-tion. All systems responded as designed. The licensee dete'rmined that the fault on the.line was caused by a static line (lightning protection) falling across all three phases ; of the 322 canal line from the Canal Electric Station to Carver feeding into the station. This caused the main generator loss of field relay to open, resulting in a load rejectio Details of the generator loss of field are dis-cussed-in section The licensee reported this event to the NRC via ENS at 5:34 i Response to the event by the control room was immediate' and > appropriate. Operators appropriately entered Emergency i Operating Procedure (EOP) No.1 on low reactor vessel water level. On a turbine trip, the closure of the turbine stop ,
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> n, and control valves caused a pressure rise, a. reactor coolant '
y void collapse and subsequently a " shrink" in reactor vessel
,' . ., water level. This was momentary and was the expected re-sponse to a turbine trip. The High Pressure Coolant Injec-
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tion (HPCI) system was placed into full flow test as a pre-
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cautionary measure for pressure and inventory control. Al-though HPCI was not needed, this was a conservative meas-ure. The event was handled well by the control room opera- ' tors and plant response was as expecte '
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2. Partial Reactor Water Cleanup (RWCU) System Isolation t (Group 6)
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On June 6 at 4:55 p.m. with reactor power at 100's, a partial- i Group.6 Primary Containment Isolation System (PCIS) (Reactor Water Cleanup System) actuation occurred when valves MO-1201-5 and M0-1201-80 unexpectedly closed. The operatin , j RWCU pump automatically tripped as designed upon closure of- t g
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these valves. Licensee investigation determined that the- ; isolation of the RWCU system resulted from a failed-CR120 relay and a blown fuse in the PCIS logic. All active com-- ponents responded to the isolation signal as designed. The relay and fuse were replaced and the system was' returned to
.its normal configuration. At the conclusion of this report period, the licensee had not determined a cause for the .
relay failure, The NRC-was notified of this Engineere Safety Feature (ESF) actuation.at 5:30 p.m on June 6 via the EN Licensee actions were expeditious and conserva-
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tive and.the inspector had no further questions, 2 .- 6 . 3 ' Degraded "E" Condensate Demineralizer J
1 On May 31, while increasing-power following backwash of the' [ , . main condenser, a conductivity alarm was received in th ; , control room approximately one to two minutes after placin a the "E" condensate demineralizer into service. The "E" ~
condensate demineralizer was taken out of service and-chemistry personnel were notified to sample the reactor i
% coolant. Conductivity was found to be about 3 umho/cm, :1 sulfate (504) was 235 ppb and nitrates (NO3) were 188 ppb, a confirming a resin intrusion into the reactor coolant system ,
from the condensate demineralizer resin be I
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Appropriate RCS chemistry parameters are maintained to ' d*g , minimize integranular stress corrosion cracking (IGSCC) and '
transgranular stress corrosion cracking (TGSCC). The in-itial levels of sulfate (greater than 200 ppb) placed the ' licensee in EPRI (Electric Power Research Institute) Action i t i
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Level 3. Action Level 3 represents the level above which
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data or judgement indicate that it is inadvisable to con-
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tinue plant operations. In accordance with, station proce-- ; dure 2.4.'148, " Guidance for Recognizing and Responding to Resins, Oil, Air, Glycol, Hydrocarbons, and/or Chloride . Intrusion into the Reactor Vessel," a rod block'was in- t'
* serted due to RCS conductivity greater than 2.0 umho/c Additionally, in accordance with procedure 7.8.1, " Wate .
Quality Limits," the licensee initiated a plant shutdown by- ! reducing recirculation flow until EPRI Action Level 3 was-J, exited approximately a half hour late Conductivity. limits- , remained well below Technical Specification limits through-out this even ^ Licensee. troubleshooting of the "E" condensate demineralizer revealed two damaged laterals; one had a ripped screen and the other had a hole. In addition, two weep holes at th ' bottom of the hub had missing screens. 'The damaged laterals
.were replaced and the two weep holes with missing screens .t were plugge Licensee actions with respect to this event were appropriate, conservative'and in accordance with plant procedures to-ensure reactor water chemistry was expeditiously reduced.to acceptable limits. Technical Specification requirements !
were met at all times. The inspector had no further ques-. tion '
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3.0 Radiological Controls
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3.1 Installation of Closed Circuit Television (CCTV) Cameras' : During the-last two outages, the licensee installed closed circuit .i television (CCTV) cameras in the reactor building, reactor water '
- cleanup (RWCU) pump / heat exchanger rooms and main steam tunnel, and ' 'in the turbine building (recombiner rooms) and condenser bay. The ,
cameras feed into three separate TV monitoring cabinets located out-side the high dose areas. .The purpose of the installation is to re-duce man-rem exposure to reactor operators while performing procedure 2.1.16, " Nuclear Power Plant Operator Tour." Information required by
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' .the procedure on operator tours is visual in nature, such as the overall condition of the equipment in the area, and identification of ,
steam and water leak ' 1-y The cameras are equipped with zoom lenses to perform. visual inspec-tions of the high radiation areas and are rated for high radiation applications. The licensee's long term plans are to continue the
' . installation of CCTVs. During refueling outage #8, the licensee plans to install CCTVs on the turbine deck and additionally in'the drywell to monitor the recirculation pump .
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Installation of the CCTVs in radiation; areas to reduce personnel ex-posure where possible is a good licensee initiative and demonstrates the licensee's commitment to reduce radiation exposur .2' Source Term Reduction Program The licensee has initiated a program to reduce general area radiation dose. levels:substantially in conjunction with their As Low As Reason-ably Achievable (ALARA) program. The source term reduction program includes hydrolasing, hot spot identification and removal of aban- > doned-in place equipmen . Hydrolasing .is an effective method to remove contamination by utiliz-ing water jet sprays under high pressure to remove contamination, f i thereby reducing hot spots and general area dose levels. Hydrolasing ' can be used in tanks, sumps and drain lines. The source of water. is currently demineralized water although the licensee is investigating the use of condensate storage tank water to reduce processin , Three floor drains on the reactor building 51 foot elevation were- f hydrolased during the week of May 14; significant reductions in dose j levels were noted. In one case where the initial readings were g 60-300 mrem / hour, the final readings were 5-20 mrem / hour following 1 hydrolasing,.with no measurable exposure to the nuclear plant attend ' I ants while performing the hydrolasin The licensee formed a source reduction program task force to develop a prioritized list of other contaminated areas to be decontaminated ! using hydrolasing. .The licensee's goal is to perform this operation on all floor drains / equipment drains which have readings greater tnan ^1 5 mrem / hour and to prioritize those. components and drains to ensure the highest dose areas are-performed firs i The source term reduction program also includes a hot spot identifi- f cation program. This program is used to identify, then either flush ' or shield each hot spot to reduce general area dose levels. Removal of abandoned-in place equipment is also included in this progra ' The use of hydrolasing and the hot spot identification program to reduce general. area dose levels and subsequently personnel exposure is an excellent initiative by the licensee and demonstrates manage-ment commitment to. reduce personnel exposure wherever possibl .3 Access Control Program
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j On June 4, the licensee implemented their new Process Building Access Control Program. The -access control. program re-defined the radio- i logical control area (RCA) and reduced the number of egress areas- from fourteen to three. This provides tighter controls over personnel, l equipment and materials leaving a radiological area. The re-defini-tion of RCA means that there are physical areas inside the turbine, , q
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4 radwaste and reactor buildings designated as RCA's. In addition, there are areas where there will no longer be requirements for a . ; self-indicating _ dosimeter. Signs are clearly posted to indicate the radiological control areas and the' requirements for entr The licensee's access control program is a good initiative to provide tighter controls over personnel, equipment and materials leaving an RCA and further illustrates the licensee's commitment to reduce per-sonnel exposur ,0 Maintenance / Surveillance 4.1 Diesel Fire Pump Battery Problem ! During the performance of procedure 8.B.1, " Fire Pump Test," on May 4,1990 the diesel engine failed to start on-the "A" battery bank when the manual start.pushbutton was depressed on the local control panel in the, pump house. The operator noted this in the surveillance procedure, then attempted to' start the engine on the "B" battery bank. ,The engine subsequently started and ran normally. When the Nuclear Watch Engineer (NWE) reviewed the completed surveillance pro-cedure it was noted that the "A" battery bank would not start the : diesel fire pump engine. Failure and malfunction report (F&MR) 90- ; 143 was initiated to address the proble ! The inspector noted that the "A" battery bank had failed to start the engine two weeks previous during performance of the same surveillance , procedure. It was also noted that F&MR 90-124 was initiated at that time to address the problem. Review of F&MR 90-124 by the inspector: revealed that 'even though shift' technical advisor (STA) review indi-cated that the fire pump would not meet its Technical Specification requirements.the NWE determined that.the event was not Technical Specification related end the surveillance was signed-off by the NWE as complet Subsequent followup investigation by the inspector revealed that the j diesel fire pump had failed to start on demand on at least five (5) ; previous occasions dating back'to January 198 Subsequent discus-sion with the fire protection system. engineer revealed that the bat-
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teries- have also had a low specific gravity problem (MR 89-33-5), a charger failure to charge in the automatic position (MR 89-33-17),an-engine starter problem (MR 89-33-200), an inadvertent restart when-secured from running (MR 90-33-39) and a wiring problem with the engine overspeed device (MR 89-33-241). When the licensee was asked ; if a root cause analysis was performed to determine the cause of the unusually large number of electrical problems, the licensee stated that none had been performed to date, i
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/ 8-The Operations-Department, when questioned concerning the' apparent
- , L discrepancy between F&MR's 90-124 and 90-143 in regards- to TS oper-ability determinations, committed to investigate and provide resolu- , tion to the issu s The inspector will continue to monitor licensee -;
actions in this are .2 Missed RHR Piping Temperature Surveillance Technical Specification 6.8.A requires that written procedures and ? administrative policies shall be established and implemented that-meet or exceed the requirements of section 5.1- of ANSI N18.7-197 Additionally, ANSI N18.7-1972, section 5.1.2 states that procedures shall be followed. Contrary to the above, during review of procedure 8.5.2,10, "RHR Piping Temperature and Pressure Monitoring," the.lic-ensee, determined that a required surveillance had not been completed-in its entirety._ However, the Nuclear Watch Engineer signed, in one .' case,: verifying that the acceptance criteria had been met, and noted a's "NP'l (not performed) the other acceptance verification signoff i Additionally, the Master,5urveillance Tracking Program (MSTP) was ' signed as completed als '
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--The~ procedure had not been completed in its entirety because of a
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problem in the screenhouse with trash on the travelling screens. The a temperature and pressure were-taken for the "A" side of RHR but not 'l-for the "B" side of RHR. The "B" side temperature-and pressure read- t ings were "NP'd" and the procedure was subsequently incorrectly signed '! as complete by the Nuclear Watch Engineer (NWE) due to his thinking * that the RHR-system; temperature and pressure surveillance was only to- a be' performed on the "A" side.since the "B" side.had been repaired ?! during the March 1990 outage'(Inspection Report 50-293/90-07, section; 5.3). However, the NWE signed off as complete, incorrectly, that the-saturation temperatures were acceptable, when only the "A" side was' i taken, and "NP'd" the acceptance criteria for the piping pressure when:in' fact piping pressure readings were taken for the "A" sid ;
= Licensee investigation into why the surveillance was incorrectly '
signed'off as complete' determined the~~following reasons: (1) the ,
-procedure was not- '.' user friendly" in -that it was not clear in- its a ' requirements; (2) the mindset of the NWE, who felt that only the "A" a side was required to be performed due to the repair previously com- 1 - pleted on the "B" side and (3) a problern in the screenhouse requiring ,
immediate operator action ! Licensee corrective actions included: (1) revision to procedure 8.5.2.10 to provide more. clarification; and (2) counselling of in-dividuals involved in signing off on the incomplete surveillanc The incorrect signoff of onc acceptance criterion and incorrect "NP" 4 of the other acceptance criterion, although half-of this item was com- l pleted, illustrated inattention-to-detail and procedural noncompli- ! ance. Licensee failure to comply with procedures in accordance with ANSI N18.7-1972 is a violation of NRC requirement However, because '
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the incomplete but signed off surveillance was of minor safety sig-nificance; was identified by the licensee; and because the licensee took prompt ~ corrective actions to correct the problem and to preclude recurrence, a Notice of Violation will not be issued in accordance with the discretion criteria of the Enforcement Policy, Section (10 CFR 2 Appendix C). This is identified as a licensee identified, non-cited violation (NC4 50-293/90-13-01), 4.3 Master-Surveillance Tracking Program
The licensee tracks the surveillance program status as defined in procedure 1,8, " Master Surveillance Tracking Program (MSTP)." . Ele-ments tracked include a listing of all scheduled surveillances,: win-dows of opportunity to perform tests, and methods to identify late end missed surveillance procedures to management for increased visibility and corrective action The inspector reviewed selected portions of the MSTP and the Tech-nical Specifications to evaluate the adequacy of licensee compliance to Technical Specification surveillance requirements. The MSTP and-procedures generally were found to appropriately include the required operability criteria, and surveillance frequencies were noted to be satisfactory .in accordance with TS requirements. The surveil. lance program appeared well managed.in the area of scheduling and in the conduct of surveillances, with only isolated instances of missed sur-veillances, which have been previously identified by the license The. licensee conducts annual audits of a sampling of Technical Speci-
'ications to evaluate the adequacy of compliance / performance t . elected TS sections. A review of the previous audit performed by-the Quality Assurance Department revealed the thoroughness and detail of:their audit with' appropriate recommendations for improvemen However, during a-review of licensee-procedures and drawings-inicon-junction with the MSTP and Technical Specification review, it was:
noted that core spray valves MO-1400-3A/B were not cycled monthly a > required by TS 4.5.A.1.c for motor operated valve operability. A; recommendation for improvement / investigation (RFI 89-324) was written documenting this. The response to the RFI stated that the subject l
' valves were classified as " passive" valves as defined in ASME-Section XI, INV-2000. Since the ASME code does not require exercising of-passive valves, these valves were not included in the valve oper-ability procedure. The RFI further states that the Technical Speci-fication is invoking an operability check of active motor-operated '
valves; the subject valves are normally open and are required to re-
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main open-during an accident; and therefore, their safety criteria-l are met =, Although the technical justification for not cycling these valves was sound and appropriate, the response to the RFI appeared to interpret / clarify the TS and was therefore an inappropriate use of an RFI. No documented justification was found by the licensee for not cycling certain other motor-operated valves as required by TS. The-
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licensee had: considered that since these valves are passive valves and do-not receive an automatic signal to re position, they are not required by TS to be cycled for operability. Discussions 1with the j licensee indicated their agreement that the RFI was inappropriately j used in this case; subsequent to this inspection, the licensee _ issued g Technical Specification Clarification No. 90-03.to document their- l position on this issu l A review of RFI's written by licensee personnel responsible for the , MSTP reviewed no further problems in this area. The recommendations for improvement / investigation were of high qual.ity, detailed and J illustrated the depth of knowledge and professionalism on the part of ( these licensee personne ' The' surveillance program was well managed with knowledgeable, profes- ! sional. individuals administering the program. The MSTP and proce- 4 dures included the required operability criteria to be and surveil-lance frequencies. Quality Assurance audits appeared well conducted,
-thorough and focused. The only area of concern was not cycling all motor operated valves without a documented Technical Specification interpretation of which valves were passive and invoking the ASME interpretatio The inspector had no further question .0 Security 5.1 Observations of physical Security '!
Selected aspects of plant physical security were reviewed during -i regular and backshift hours to verify that controls were in accord-ance with the security plan and approved procedures. This review i included the-following security measures: security officer staffing; vital- and protected. area barrier integrity; maintenance of isolation zones and protected area barrier integrity; and implementation-of access controls including access" authorization and badge issue; searches of personnel, packages, and vehicles; and escorting. The lic-ensee was conducting a well managed and coordinated security program as indicated by a proactive approach to implementation of the fitness-for-duty program and an aggressive security equipment upgrade program.. This will include a new security computer and associated hardware and upgrades to alarm station equipment. No inadequacies j were identifie "
' 6.0 Engineering / Technical Support'
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- 6 .1 - Inoperable Contact on Loss of Field Relay l
On May 13, an offsite electrical fault on the 345 KV distribution 322 canal line caused an actuation of " loss of field" relay (240 LW-KLF-1) which is part of generator lockout scheme GL. This relay operated l~ the generator lockout and auxiliary relays, tripping the generator !
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field breaker and ring bus air' circuit breake'rs (ACB's) 104 and 105, which resulted in a turbine trip and subsequent automatic reacto scram (section 2.6.1).
Investigation.by systen s engineering found that the generator " loss :{ of; field" relay had an inoperable contact, which negated a designed . 1 15-cycle time delay, and allowed the relay to operate instantaneously- l on the transmission line offsite faul i
.In response ~ to the failed relay and in order to conduct troubleshoot-ing on the relay, the licensee initiated temporary modification (TM)
90-11 on May 14 to bypass the failed " loss of field'.' relay contact,- > to provide protection against another inadvertent main generator I trip, until a new relay could be procured, tested and installe ! Following-inspector questioning, the licensee determined that when ithe annual test of the relay was performed by licensee technicians, the: i procedure did not require verification of the 15 cycle time dela j The licensee subsequently initiated a change to the relay-setting j form to include a check of the 15 cycle setpoint during the annual; j relay tes A new relay was received onsite, tested and installed,: i and the temporary modification removed and closed out on May 1 ! Licensee response to this event appeared to be satisfactory, and th l evaluation by systems engineering to determine the root cause of. the ; trip was thorough. The installation of the-temporary modification to ; reduce further challenges to plant safety systems was appropriate and j well planned. The only area of concern was that inspector question- i ing was required as to whether it would be appropriate to check the j set-point of the 15 cycle time delay during performance of the annual .
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j calibration check of'the rela ~! I This area appeared to need further-investigation as to whether other l non-safety equipment trips can cause cha11enges to safety systems, s
'At the end of this inspection period, the licensee was performin !
review of other equipment for similar problems. The inspector con- -i
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cluded that this was appropriate followup actio .2~ procurement Program
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The licensee's program to control procurement activities is described in Quality Control Inspection Manual, sections 7.01, " Receipt Inspec- l tion;" 7.03, " Quality Control Review of Contractor Records;" 7.04- ,
" Surveillance of Contractor QA Programs at PNPS;" 7.08, " Spare Parts l Review;" and 10.01, " Sampling-Inspection." The details-of.the pro-curement operation are described in the Purchasing / Stores and Service , Department Procedures Manua L
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a L: . l A review of.these procedures and a selection of 25 purchase requests' from 1989 - 1990 verified that the following areas were addressed'on~ l the requests:
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10 CFR 21.31 applied >
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Material-identification required 'g
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Engineering signatures required i
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Storage level / shelf life
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In addition to the above, parts, materials and components are pur-chased from the-licensee approved vendor lists. Inspection and audit ; reports of vendors-~on t he approved-vendor lists were on file and in an update configuratio n. The inspector verified that.the 25 purchase requests remaining wers issued to vendors on the licensee-approved - lists. Where-quality :ertifications were required as part of the
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purchase request,- the requirements of ANSI N45.2.13' were satisfie Li In the review of-the 25 purchase requests the-inspector determined
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that the; licensee's procurement program was functioning.as described in controlled procedu.res. The inspector determined that the staff performing the engineering, quality and procurement functions were knowledgeable of and were performing in compliance with site proce- 7 dure .3 Receipt, Storage and-Handling of Equipment and-Materials Program l The receiving and verification of parts and materials is described 4 in procedure 7.01, 'fQuality Control Receipt Inspection," .and proce-dure 7.01~, " Stores-Department". The receiving inspection checklist of procedure 7.01 identifies the attributes of the purchase order and verifies the receipt conditio If a problem is identified, a mate-rial receipt inspection report-(MRIR) is written, which requires
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written disposition of the conditio A' review of 15 MRIRs by the- , inspector established that the concerns were dispositioned in an-acceptable manner and closed in the MRIR tracking syste t During the receiving inspection cycle,. quality control personnel verify-that the items received comply with the purchase order cri-teria, including the type of-protective packaging. =0uring the~in-
, spection of Q stored items located in B level storage drawers, the inspector found parts with protective covers and seals missing and .
noted shelf life requirements which were not identified on.the ., attached equipment tag The MRIR dates for the items identified by the inspector were between - 1982 and 1985. Many of the stored parts had their protective covers removed with no guidance to the store room personnel to repackage the ( items upon issuance or during the post-staging program. The licensee committed to determine the status of the Q parts and materials stored .
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l' . in the B. level storage drawers regarding compliance with program cri- d teri Pending this determination, this is an unresolved ite (50-293/90-13-02).
t 7.0 Safety' Assessment / Quality Verification.
L 7.1 (Closed) Unresolved Item 50-293/89-05-01, Safety Evaluations Were Not-Performed Prior to Disabling Control Room Annunciator Procedure 2.3.1, " General Action for Alarm Response and Annur.ciator Control,"-
:s did not require'a safety evaluation prior to disabling system annun-ciator .The inspector verified that this procedure had been revised to ensure xj that a safety evaluation is performed if required. A disabled annun-- J ciator log index had also been added to the procedure and was main- H tained in. the control room. A review of the annunciator log index 1 -
for 1990 indicated which annunciators required a safety evaluation I and the status of the system. As of this inspection, all disabled 1 annunciators had been returned to service. The analysis reviewed by' [ the' inspector indicated that the requirements of the revised proce - ') dure were complied with and that safety evaluations were documented o; as indicated for selected annunciators removed from service. -This item is close , i b 7.2 (Closed) Unresolved Item 50-233/87-53-02.4, Certain Procedures Governing Degraded Plant Conditions (e.g. Loss of Power, Loss of l Instrument Air) May Not Be Suf feient to Clearly Guide Recovery Actions from Events of This Nature. Also, procedures for restoring offsite power'were not reviewed against past operating experience, especially , events caused-by severe weather, and revised to' reflect, lessons
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j learned and anticipated problems which may need resolution to opti t mize power recovery tim ;i Procedure 5.3.8, " Loss of Instrument Air" had been revised to provide automatic and operator actions at various system pressures and on loss of air pressure. A requirement for component position verifi-a cation had been added to the procedur ' ' Procedure 2.4.16, " Distribution Alignment Electrical System Malfunc-tions" had been revised to prescribe ' operator action for loss of of f-
, site power with both emergency diesel generators available, failure of the "A" diesel to start, and failure of the B" diesel to star ,
The procedure also provided steps for restoring AC power, either via the shutdown transformer or via "back-scuttling" of the unit auxil-iary transforme l l - !.
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Procedure 2.4.144, " Degraded Voltage," is intended to be primarily used when the plant.is at power. It specifies the readying of the
' emergency diesel generators when actual or potential degraded voltage ,
conditions exis l
"- . Procedure 5.3.31,'" Station Blackout," and procedure 2.2.146, " Station Blackout Diesel Generator," provide operator guidance for ad(<essing the loss of all AC power and operation of the station: blackout '
diesel. These procedures also provide steps for restoring AC power E including offsite power. This item is close ' 7.3 (Closed) Unresolved Item 50-293/89-08-02, Inadequate Engineerinj # Evaluation and Poor Controls for Application of Furmanite in Reactor
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. Core Isolation Cooling (RClC) System Check Valve. Procedure 3.M.4- '
T 42, " Valve, Pipe Flange and. Heat Exchanger-Gasket Link Sealing" has ll been: revised to require incorporation of LER 89-014-00 commitments
" . into,the procedure. = Examples of the changes included: (1) Nuclear Engineering Department (NED) to calculate the maximum quantity of ,
Furmanite allowed for the task; (2) NED to ensure that. code compli- a ance is maintained for each modification performed; -(3) Quality Con- 1
. trol hold points identified; and (4) post-maintenance testing per- '
formed and approved by NE The inspector verified that the work performed on RHR valves MO-1001-50 and H0-1001-33A complied with the requirements of procedure 3.M.4-42. Quality control documentation verif.ied that post-maintenance ' testing of the valves indicated no leaks at 930 psig, which is the acceptance criterion for this tes Based on the above review, this item is close d 7.4 (Closed) Unresolved Item 50-293/87-53-04.1, Develop Station Procedur '
.for the Testing of the Startup Transforme The licensee has issued ,
procedure 3.M.3-59, " Power Factor Testing," which provides adequate 'l direction to perform power factor testing of the main, startup, unit auxiliary.and shutdown transformers. Testing of these-transformers j! is scheduled during the next refueling outage. This item is close
' '7. 5 (Closed) Unresolved' Item 50-293/89-07-04, Review Licensee Policy on ;
Rehanging of Danger Tags Without Independent Verification At the time of inspection 50-293/89-07, an inspector noted the rehanging of-
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danger tags after post-work testing without-the independent verifica-tion originally.used-to hang the tag In ' response to the inspector's concern, the licensee corrected the in-consistency in Procedure 1.4.5, "PNPS Tagging Procedure," section [11] and the Tag Change Request sheet (Attachment 4) now require in-dependent verification of all modifications within the provisions of 1.3.34, " Conduct of Operations." Based on the above licensee actions, this concern has been adequately addressed. This item is closed.
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' Improper Use of a Maintenance Work Activit'y
,y ! '~ The licensee recently initiated a new process to control maintenance work activities - the maintenance work activity (MWA). The purpose
' of the maintenance work activity is to provide the direction, scope l , and detail necessary to perform preventive or corrective maintenance -i when there is no impact (based on equipment / hardware use or work j scope) on plant safety, system operability, or personnel safety, or -
the work' scope is sufficiently routine and simple that it is withi the normal skill of the craft. The use of. the MWA program is con-trolled at the discretion of the maintenance department supervisor On May 16 the licensee issued MWA 90-33-223 to replace the diesel ! fire pump bank "B" batteries. On May 31 the licensee declared the i
? diesel fire pump inoperable and issued active.LCO 90-75 to authorize proceeding with the task. The licensee's Quality Assurance Depart-ment (QAD)'promptly identified that performance of the work on the diesel fire pump affected plant operability since an active LCO was entered and was therefore contrary to procedure 1.5.3, " Maintenance l Requests" and procedure 1.5.3.1, " Maintenance Work Plan." The QAD i identified this item as a deficiency; however since immediate cor-- !
rective actions were taken by the maintenance' department, no defi- ) ciency report was issued (this is acceptable in'accordance with lic- i ensee procedures). The licensee attributed the~cause of this defi-ciency to procedural inadequacy. The 'use of an MWA in this instance did not result ir.the work being inappropriately conducted. Licensee
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_ corrective actions include re-briefing of maintenance supervisors on ' the MWA process and revision of procedures'1.5.3 and 1.5.3.1 clarify that accomplishment of work under an MWA may not cause entry- ! into an active LC ;
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o .Self-identification by the quality assurance department of the in-appropriate use of a maintenance work activity in this instance was prompt and appropriate and illustrated the-licensee's willingness and' y
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commitment to self-identify and correct' potential problem areas. The inspector had no further question ! 7;7 Licensee Event Reporting The inspector reviewed'the Licensee Event Report (LER) listed below j to determine that'with respect to the general aspects of the event: .
(1) the report was submitted in a timely _ manner; (2) the description !
of the event was accurate; (3) a root cause analysis was performed; (4) safety implications were considered; and (5) corrective actions implemented or planned appeared sutficient to preclude recurrence of similar event i
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7. LER 90-07
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LER 90-07, "Drywell to Suppression Chamber Vacuum Breaker-Surveillance Not Performed Prior to Startup in 1988," ad- , dresses the April 27,1990 determination that surveillance ! 8. A.2, "Drywell to Suppression Chamber Vacuum Breaker Leak-age Rate Test (1.25)" had not been performed when restart-
<^ ing from Refueling Outage No. 7 (RF0 7). ' The surveillance , had expired during the outage for RF0 7 and should have -
been performed prior to reactor critical.ity in accordance with Technical Specification The licensee determined th cause to be a misunderstanding of TS surveillance require-ments and an. incorrectly scheduled surveillance in the i MSTP, The report provided a comprehensive review of the i details of the missed surveillance and fully addressed the ! reporting criteria above. The licensee's failure to per- j form a required TS surveillance is a violation of NRC re- ' quirement However, because this event was identified ~b I the licensee; is of minor safety significance;'was reported to the NRC as required and corrective actions'were taken to l prevent rec.urrence, a Notice of Violation will not be issued in accordance with the discretion criteria of the Enforce-ment Policy, Section V.G (10 CFR 2, Appendix C). This is . identified as a licensee identified, non-cited violation l (NC4 50-293/90-13-03).
!7.8 Licensee Self-Assessment Capability and Quality Assurance Program Implementation
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The-resident inspector. and regional specialist staff continual _ly {
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evaluate _ the ability of the licensee to perform self-assessments i which contribute to the identification, correction, and ultimately i the prevention of safety-significant operational and technical' ~ issues. .These evaluations include' frequent NRC observation or review
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of the licensee's development of internal and external event reports, 1 communications between both onsite and offsite staff personnel and d
- management, the performance of the oversight functions =of the onsight !
and offsite review committees and appointed subcommittees, and the- l
< performance of the Quality Assurance Department audit, surveillance I and quality control function !
The resident inspectors routinely attend Operations Review Committee j (ORC) meetings. The_ committee members and their alternates are an t experienced, diverse body composed in accordance with TS requirement l The committee convenes weekly and as plant conditions dictate. The i inspectors have observed Committee dialogue to be perceptive and pro-
. fessional. Open discussion is encouraged and direction is well main- ,
tained by the chairman. The committee typically displays a conserva- J
- tive. safety perspective and utilizes design bases documentation in support of conclusion i
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l t The committee is, empowered to charter subcommittees as appropriat to provide oversight of specific issues, The Compliance Division is '- a standing ORC subcommittee to develop and present all LERs as well as p'esent selected F&MRs to ORC for review, Previously, a subcom- , mittee was appointed to provide review of all hydrogen water chemistry ; system procedures prior to full ORC review and approval. The ORC has-been evaluating the use of standing subcommittees to provide similar ; initial procedure reviews station-wid The ORC was noted to_ discharge its duties in accordance with the re-quirements of TS 6.5. A.6. However, the committee is not required by f procedure to review and approve LERs and provide recommendation to-the Station Director before the LER is submitted to the Station Direc-tor for approval. ORC is required (by procedure) to review LERs prior ; to submittal to the NRC. It is noted that, with three exceptions, 4 the ORC has reviewed recent LERs prior to Station Director approva During discussion of ORC practices following completion of the in- ; spection, licensee management stated that they were considering a " procedure change to have ORC review all LERs before they are submitted to the Station Directo ;
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The PNPS offsite review committee is the' Nuclear Safety Review and , Audit Committee (NSRAC). The inspector-reviewed-the committee com-position and the meeting minutes for the January 30, March 9, and April 26, 1990 NSRAC meetings. The TS 6.5.B.2-4 requirements for-committee composition c.nd qualification were met. Review of the meeting minutes indicated that the committee-properly discharged its duties in accordance with the requirements-of TS 6.5.B.7 and 8.- The NSRAC utilizes subcommittees to adciress spscialized plant discipline The quality and de"elopment of the meeting minute tooics was con- :! sistent with the safety' significance of the issue Licensee Event Reports (LERs) have been well -developed, comprehensive
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't reports of consistently high quality which exceed minimum reporting requirements. A review of LER root ~cause analysis and corrective action determinations revealed no programmatic'or repetitive-weak-nesses in any given area. This indicated that the licensee has gene-rally been successful in. assessing deficiencies and implementing ,
effective correct 1ve actions to prevent recurrenc ) Based on the review of ORC and NSRAC activities, the inspector con-cluded that the licensee was providing appropriate management over-sight to promote identification of potential safety issues and that issues reviewed by' these committees were properly dispositione , The Quality Assurance Department (QAD) has the responsibility for all QA/QC activities. The QAD consists of four divisions responsible for i QA audits, QA surveillances, Operational QC, and Quality Engineerin The QAD provided in-depth performance-based reviews which effectively developed observation and deficiency detail. The QAD reports assessed i i
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o b and trended ongoing licensee performance and addressed the implemen-tation of corrective actions for previous issue The' plant staf ' appeared responsive to QAD findings - Additionally, QAD management utilized-established processes to elevate concerns to appropriate l station management when necessary to obtain acceptable resolution ' L Resident inspector review of operational' data, NRC reports, LERs and ; QAD reports indicated that the licensee's identification and assess-
? ment processes are improving. However, it appeared, due to the number of initiatives established by the licensee to ensure that potential safety issues are identified, that the administ' ration of these pro- !
cesses is becoming increasingly cumbersome. The licensee's staff continues to raise concerns, and generally effective evaluation of these concerns has' occurred; however, integration of the evaluations into previously existing processes, including timely ORC reviews of
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issues commensurate with their safety significance, has not neces-sarily been ensured. The inspector met with licensee management to delineate this concern. The licensee acknowledged the inspector's conclusions and subsequently initiated activities. to address this g issue. The inspector will continue to review this area during rou- , tine inspections. The inspector had no further questions, j 8.0 Review of Periodic and Special Reports Upon receipt, the inspector reviewed periodic and special reports sub-mitted pursuant to the Technical Specifications. This review verified, as
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applicable:. (1) that the re;,orted information was valid and included the NRC-required data; (2) th 6 test results and' supporting information were consistent-with design reedictions and performance specifications; and (3) that planned corrr;tive actions were adequate fo'r resolution of the
~ , problem. The inspenor also ascertained whether any reported information should be classif'ed as an abnormal occurrence. The following reports were reviewed:
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'- Monthly Operational Status Summaries for. April, 1990 -- . Operations Review Committee and Nuclear Safety Review and Audit Com- i mittee Meeting Minutes 9.C Management Meetings At periodic intervals during this inspection, meetings were held with
senior plant management to discuss the findings. A summary of findings for the report' period was also discussed at the conclusion of the inspec-tion and prior to report issuance. No proprietary information was identi-fied as being included in the repor i L l! .
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ATTACHMENT 1 Persons Contacted
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Interviews and discussions were conducted with members of the licensee's staff and management during the report period to obtain information' pertinent to the areas inspected. Inspection findings were discussed periodically with the man-agement and supervisory personnel listed belo K. Highfill, Vice President, Nuclear Operations and Station Director E. Kraft, Acting Plant Manager D. Eng, Outage and Planning Manager L. Schmeling, Acting Deputy Plant Manager R. Fairbanks, Nuclear Engineering Department Manager D. Long, Plant Support Department Manager L. Olivier, Operations Section Manager N. DiMascio, Radiological Section Manager J. Seery,. Technical Section Manager G. Stubbs, Maintenance Section Manager T. Sullivan, Chief Operating Engineer J. Neal, Security Division Manager P. Cafarella, Acting Systems Engineering Division Manager B. Sullivan, Fire Protection Division Manager
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