IR 05000317/1990013
| ML20055G138 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 07/11/1990 |
| From: | Roxanne Summers NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20055G134 | List: |
| References | |
| 50-317-90-13, 50-318-90-13, NUDOCS 9007200167 | |
| Download: ML20055G138 (30) | |
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n -<* . U.S. NUCLEAR REGULATORY COMMISSION '
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REGION I ! 1 Report Nos.:! - 50-317/90-13; 50-318/90-13 I
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License Nos'.: 'DPR-53/DPR-69 ! 1)
, ' . -Licensee: Baltimore Gas-and Electric Compan ; -Post Office Box 1475 t[
g Baltimore, Maryland ~21203 ; V . I h l Facility: Calvert Cliffs Nuclear Power Plant, Units 1 and 2 0
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? -Locationi Lusby, Maryland- '
d Conducted: June 3.-1990, through June 30, 1990 l . . .
.+, JInspectors:-- Larry E. Nicholson,'3enior Resident Irspector 'I Allen G. Howe, Resident Inspecter s 'Tae J. Kim. Resident Inspector .l <
Victt,r M. McCree,. Operations Engineer 1 A Herbert J. Kaplan, Senior Reactor Engineer- i
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Approved by':
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' "" *M ~ 9 ~ l( ~ b Robert J. Sumnirs, Acting CITel .
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Date' I t- Reactor Projects Section No. IA 1 Division of Reactor Projects j
>t Inspection Summary: I
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This: inspection report documents routine and- reactive ' inspections during da ' and'backshift ' hours of : station activities including: plant operations;. radio- ' logical protection; surveillance and maintenance; emergency ' preparedness;' l security; engineering: and ' technical support;s and safety' assessment / quality: .
: verificatio '
Results:- j No violations were identified. -Two unresolved items were identified regarding - ; resolution'of the' emergency switchgear room air condi.tioning system operability
+ ' issue -(Section 7.3) and the errors found in the calculations for low tempera-ture overpressure protection (Section 7.5). An Executive ' Summary follows.- .,
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p s . . , EXECUTIVE SUMMARY Calvert Cliffs Inspection Report Nos. 50-317/90-13 and 50-318/90-13 Plant Operations: (Modules 71707,93702) The operations staff was observed to i-be adequately aware and in control of plant conditions. Problems that occurred during testing were handled in a deliberate and conservative manne , Radiological Protection: (Module 71707) A region-based inspection of this area was conducted during the inspection period with the results contained in Inspection Report 50-317/90-14 and 50-318/90-12. Routine review by the resident inspectors revealed no noteworthy finding Surveillance and Maintenance:- (Modules 61726, 62703) Various surveillance ' tests were observed and reviewed with no deficiencies noted. A review of the Unit 2 pressurizer heater sleeve project concluded that adequate management oversite and direction .was being applied, in addition, work on this - project p" wss found to be accomplished in a controlled andeffectivo manner. A weakness was identified. regarding the inefficient scheduling of repair activities that results in equipment being routinely being removed from servic Emergency Fregredness: (Module 71707) Various aspects of the drill conducted during the inspection period was observed with no noteworthy findings identifie ' I See,urity: (Module 71707) Routine review in this area identified no noteworthy finding Engineering and Techn_ical Support (Modules 71707, 90712, 92700) Significant System and Design Engineering effort was expended to resolve the saltwater system beat exchanger fouling and piping flange weld failure problems, as well as the resolution of damage to piping supports caused by water hammer. A good safety perspective was observed in addressing. these issues, although the large workload in engineering remains a concern. Various other engineering projects were reviewed with no deficiencies note Safety Assessment / Quality' Verification: (Modules 71707,30703) The Unit 1 Startup Assessment Report represented a forthright self-assessment
, of _ personnel and equipment performance during the April 4-23, 1990 startup and power operation of Unit 1. Licensee investigation and corrective action to date, in regard to a weakness in conditional release of material program, has been adequately focused to address the root causes, A weakness was identified in QA audit report response timeliness in regard to a previous safety sy: tem functional inspection. Implementation of the Problem Report system continues to be a concer l
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r TABLE OF CONTENTS i L .t v s
1. , Summa ry. o f. Fa c i l i ty Ac ti v i t i e s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . - 1-
! Operations (IP 71707,71710)................................ 1 l
2.1 -Operational Safety Verification........................ 'l j 2.2 , Engineered Safety Features System Walkdown. . . . . . . . . . . . . ! p , Radiologica Controls..................................'...... -3 5 Maintenance and Surve111ance................................ 3-o I 4.1: Maintenance Observation................................ 4.2 Surveillance Observation ....................,.......... 3 l
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5.: Eme rg e n cy - p r e p a red n e s s , . . . . . . . . . . . . . . . . . . . . -. . . . . . . . . . . . . . . . . 11 j Security................................... .................. 11 o p, Engineering.ano Technical Support........................... 1.2 -
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'; l7.1. Inspection and Repair of Pipe Supports................. 12: !
4 7.2' 5elt Water System Performance..............~.............- 13
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7:3 Twitchgear Room Air Conditioning....... 3...,.......... 15 T 7'4 LPotential Problems in Circuit Breaker Coordination.....
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16 - 7.5 ~ Low Temperature Overpressure Protection. (LTOP). . . . . . . . .- 17 j
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Safety. Assessment'and Quality' Verification................. -
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6.1 Plant Operations'and Safety Review Committee........... '17 j 8.2 . Review of Written Reports...................-............- 17
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8.3 Temporary Waivers of Comp 11ance........................ .18
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8.4l Programmatic Review of Low Pressure Safety Injection j
- System Sa fety ~ System Functional Inspection. . . . . . . . . . . 18 4 8.5 Root Cause Analysis on.. Reactor Coolant' Pump Vibratio .6 Review of Licensee ~ Unit'1.Startup Assessment Reports... .
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'8.7-' Licensee Investigation ~on Conditional-Release of Ma t e ri a l P ro g ram . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21 l; t
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TableofContents.(Continued)- t i
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: Followup _of Previous Inspection Findings.................... 21
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-i 9;) JSignificant Issues Management System........... ....... 22' i 9.2 '(Open STI-15) Alternate Safe Shutdown Control Room- i
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Evacuation........................................... 12 3 ' , i 9.-3 -(Closed) UNR; 50-317/88-19-02. ......................... 23 l 9.4-(Closed) Violation 50-317/87-09..-....................... 23 l e 9.5:-(Closed)1UnresolvedItem 50-317/88-05-01...............- 12 4 i
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10. Management; Meeting...........................................
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: 10.1 Prelimi na ry Inspection Findi ng s . . . . . . . . . . . . . . . . . . . . . . . . ' 24 l .
L10.2 Attendance'at Management Meetings Conducted by Region ; [n Based Inspectors..................................... 25 ,
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DETAILS , 1. ' Summary of-Facility Actidities E ! Unit I remained in' cold shutdown for the duration of the inspection perio for a planned maintenance outage. Major outage activities completed -dur- L ing- this period included steam generator eddy current ' inspections and reactor coolant pump seal replacemen , L Unit 2 remained defueled for the extended Cycle 8 refueling outage with
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the. fuel in the spent fuel pool. Repair of the pressurizer heater pene-
.trations' continued throughout the. inspection period.
r-I' A management meeting was conducted between the licensee and the NRC Calvert Cliffs Assessment Panel on June 7,1990, to provide-an update of the status-for the Calvert Cliffs Performance Improvement Plan. The meet-ing was ' held at the_ NRC Region I Office. The results of the licensee startup self- assessment effort was also presented by the licensee at the ' meetin On June 27, 1990, J..Wiggins, Deputy Director of NRC Region I Reactor Projects, and C. Cowgill, Acting Brar,ch Chief for NRC Rcgion 1 Reactor Projects, tourea the site. The tour was conducted by the Senior Resident Inspector.- A' public meeting was held by the NRC on June 27, 1990, at the Holiday Inn in Solomons, M:ryland.- The purpose of the meeting was to solicit public comment on the-Calvert Cliffs Performance Improvement Plan. A transcript of. the meeting will be distributed to the Local Public Document Room ano various local libraries.
L Plant Operations 2.1 . Operational Safety Verification The inspectors routinely observed plant operation and' verified that the facility was operated safely and in accordance with licensee pro-cedures. and regulatory requirements. Regular tours were. conducted of the following plant areas: ,.. -- control room -- security access point ? - . primary auxiliary building -- protected area fence
-- radiological-control point -- intake structure n -- electrical switchgear rooms -- diesel generator rooms -- auxiliary feedwater pump rooms -- turbine building L !
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: Control room instruments and plant computer indications were observed p for correlation between channels and for conformance with ' technical l b specification (TS) requirements. Operability of engineered safety ;
L features, other safety related systems and onsite and offsite power t sources was verified. The inspectors observed various alarm condi- : r tions and confirmed 'that operator response was. in accordance with , plant operating procedures. Ro_utine operational surveillance testing ! was also observe Compliance with TS and implementation 'of appro-priate action statements for equipment out of. service was verified, c
Plant radiation monitoring system indications and plant stack traces . were reviewed for unexpected changes. Logs and records were reviewed' ! to determine if entries were accurate and identified equipment status *
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or deficiencies. These records included operating logs, turnove , e ' sheets, system Lsafety tags, and the jumper and lifted lead boo ; Plant housekeeping controls were monitored, including control and i storage of flammable material and other potential safety hazard , The inspector also examined the condition of various fire protection, '
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meteorological, and seismic monitoring systems. Control room and ! shif t manning were r.ompared to regulatory requirements and portions F of shif t . turnovers w<re observed. The inspectors found.that control , room tecess was broperly controlled and a professional atmosphere was , maintaine ) In- addition to normal utility working hours, the review of plant operations was routir.ely conducted during evening shifts and weekend J and midnight shifts. Extended coverage was provided for 8 hours dur- * ing' backshif ts and 9 hours during deep backshif ts. Operators were alert and t.ttentive to duties, r i Engineered Safety Features System Walkdown l In addition to routine observations msde during regular plant tours, 1 the inspectors conducted walkdowns of the accessible portions of .: selected safety related systems. The inspectors verified system ' operability through reviews of valve lineups, control room system prints, equipment conditions, instrument' calibrations, surveillance L test frequencies and .results, and control room indication The inspectors performed a walkdown of the following . system during this , inspection perio ,
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Auxiliary Feedwater System (AFW)
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The inspectors concluded that the Unit 1 AFV system was in a general state of readiness, although numerous deficient condition cards were hanging in the room. While performing this inspection, the inspector noted .that- the plant paging system could not be heard in the APA roo This observation was relayed to the Shift Supervisor who
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m-e s e i ,- i !+ h b ; notified the onsite telecommunications group and entered the discrep-ancy into the telecommunications log that is maintained in the shift L ' supervisor's office. The inspector discussed this method with var-ious station staff members and expressed concern that this is an example of: dealing with a condition adverse to quality without
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utilizing the formal problem report syste A review of the log ! indicated approximately twenty entries over the past year pertaining n to ' problems with the plant paging system. These problems were
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handled through an arrangement with a contractor that circumvented the problem report, maintenance request, and maintenance order sys-tems. The licensee was evaluating this situation as the inspection k period' ended. Followup of this item will be included during subse- [ < quent inspections regarding the implementation of the Significant Issues Management Syste . Radiological Controls During routine. tours of the accessible plant areas, the inspectors observed the implementation of selected portions of the licensee's Radio-logical Controls Program. The utilization and compliance with special l work permits (SWPs) were reviewed to ensure detailed descriptions of L radiolegical conditions were provided and that personnel adhereo to SWP requit ements. - The inspectors observed controls of access to various i radiologicelly controlled areas and use of personnel monitors and frishing methods ' upon exit frnm these area Posting and control of radiation areas, contaminated areas and hot spots, and labelling and control of con-tainers holding radicactive materials were verified to be in accor6nce with:licenee procedures.. Health Physics technician control and monitor-ing of these activities were determined to be . adequate. No noteworthy concerns in thie, area were identified.
i-4. Maintenance and Surveillance r 4.1 Maintenance Observation L The inspectors observed maintenance activities, interviewed person-n nel, 'and reviewed records to verify- that work was conducted in ac-cordance with - approved procedures, Technical Specifications, and-
. applicable industry codes and standards. The inspectors also verif-1ed that: redundant components ' were operable; administrative' con -
u 'trols were followed; tagouts were adequate; personnel were qualified; correct replacement parts were used; radiological controls were t l
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e proper; fire protection was adequate; quality control hold. points l ' were adequate and observed; adequate post-maintenance testing was performed; and independent verification requirements were implemen-ted, The inspectors independently -verified that selected equipment p 7 was prop 9rly returned to service.
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Outstanding work requests were reviewed to ensure that the licensee gave priority to safety-related maintenanc The inspectors observed / reviewed portions of the following maintenance activities, h Fire Pump Maintenance [ ' The inspectors reviewed the maintenance performed during this period on the diesel driven fire pump and associated fire pro-L tection equipment. The diesel driven fire pump was removed from service on May 7 for a routine overhaul in accordance with main-tenance order (MO) No. 209-0 '. ? -349A. The pump casing was dis-F assembled and refurbished as :.ecessary. No discrepancies were identified regarding the actual pump overhaul. The pump was [ returned to service on June While establishing isciation for the above overhaul, an operator incorrectly assumed that turning a handwheel on a relief valve downstream of the pump would gag the valve shut for pump isola-tion. . This handwheel in fact adjusted the setpoint of the ' relief valv The licensee discovered this discrepancy and maintained the pump as inoperable until the relief' valve could be removed and tested to ensure the correct lift setpoint. This error was documented to the NRC via a special report, dated June 13, as required by technical specifications since the pump was not returned to service within seven days. An internal problem report was generated that identified the error and developed long term corrective actions. Short term corrective actions included placing a sign on the valve indicating that the handwheel alters the lift setpoin On June 19, the inspector witnessed the performance of annual preventive maintenance (PM) on the diesel engine for the above
. pum This work was being _ performed in accordance with M0 No.
' 200-141-443 Although no discrepancies were identified, the inspector questioned why the PM for the engine was not performed' e while the pump was out for overhaul. In addition, planners were J
' preparing a package to replace a leaking sightglass that had "
been identified as broken for almost a year. This sightgla s is located just upstream of the relief valve discussed abov ! e _,
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c, .. , ; D l a ; m- The inspector discussed the methods used for. scheduling work and J' agreed _with the licensee that this area is weak and in need of l improvement. - Although the inspectors did not identify specific ;
. instances where safety was significantly compromised, it became evident _through discussions with various station personnel that '
the scheduling of maintenance is frequently not well coordinated ;
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to.effect the maximum amount of repair while minimizing the time l F important equipment is removed from service. This was reflected 3 , in a similar finding _ by the licensee's independent assessment P
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' team during the recent Unit 1 startup. The inspector concluded ! that the licensee is aware of this weakness and is implementin l Steps for correctio i Cleaning the Service Water Heat Exchanger f L- ^ On June 20, 1990, the inspectors witnessed the cleaning of she I No. 12 service water heat exchanger performed in accordance with : operations procedure 01-29. This evolution was necessary due .to ! a reduction of syste;a flow to 10,007. gallons per . minute -(gpm). ! 01-29 requires that flow be maintained greater than 10,500 gpm ; in the current plant conditions. Approximately 12 gallons of shells and grass were removed from the heat exchanger tube ! sheet. This .particular heat exchanger was last cleaned on , ' June 5, 199 No discrepancies -were identified during this ' observatio ,
, Power Operated Reli,ef Valve Circuit Repair s .On June 25, 1990, the inspectors witnessed the job preparation g and calibration of instrumentatien associated with the Unit 1-power operated relief valve This work was performed in ac- v cordance .with maintenance order No. 200-176-569 No discrep--
ancies were identified, Pressurizer Heater Sleeve Replacement
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Ouring the present inspection period the inspector reviewed the . ongoing activities covering the removal and replacement of one - F hundred nineteen (119) Unit 2 pressurizer heater sleeves. The . L . replacement program was initiated by the licensee because of ; reactor coolant leakage caused by primary water stress corrosion t cracking (PWSCC) of_ the existing alloy 600 (Inconel) heater i sleeves. The licensee concluded that a principal cause of the : L cracking was the excessive residual stress and cold work k imparted to the surface of the sleeves during a reaming J l
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[ operation employed by the vessel manufacturer (Combustion Engi-neering) to accommodate the heaters. To reduce the potential for. leakage, the licensee opted to utilize a new Babcock and
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Wilcox -(B&W) heater sleeve design which featured the use of a !(4 more corrosion resistant sleeve material. Installation was per-
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formed by B&W using special manufacturing techniques to minimize c the need to remove excessive metal so as to allow for smooth insertion of the heater Figure 1 shows sketches. of the old and new heater desig The new sleeve material is nickel base alloy _ 690 which contains r
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nominally 29% by weight Cr as compared to the original nickel '
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alloy 600 sleeve material with 15.5% by weight Cr. Use of alloy 690 for this specific application was approved by NRC in' a i letter dated, April 5,1990.- NRC Regulatory Guide 1.85 had pre- )' viously approved the use of alloy 690 for heat exchanger tubin The design, fabrication, and examination of the replacement heater sleeves were required to be performed in accordance with Class I requirements of the 1986 Edition of ASME Section III and X As shown in Figure 1 the new design consisted of.two sleeves (an inner and outer sleeve) installed from the outside of the pres-surizer, The inspector reviewed three key travelers covering the detailed instructions for the installation and inspection of the heater sleeve and heater assemblies. : The B&W travelers were ;
-identified - as D-50-1175163-00,- G1-50-1176166-00 and H-50- 1 1176167-0 After enlarging _the original penetrations to 1 11/32" diameter to - accommodate the new sle6ves the records 1: .showed that the bores were visually and liquid penetrant inspec-te Except for minor indications in the original Inconel 8 weld deposited I.D. cladding, liquid penetrant inspection of the' '
new machined surfaces did not reveal any- rejectable indication The original low alloy steel surfaces which had been exposed to , leakage were found to- be free of significant pitting. Indica-tions that were .found in the original cladding were reportedly ) removed by slight buffin ; Prior to insta11ation' of the new sleeves, the 0,D surfaces of ! the- head surrounding the penetrations were clad with inconel filler. metal ERNI-Cr-3(82) to facilitate the welding of the
, outer Inconel sleeve to the alloy steel head without having to , -stress relieve the vessel. The records showed that the cladding *
operation was performed by Welding Services Inc. (WSI) under the direction of .B&W using an automatic Tungsten Inert Gas (TIG) Temper Bead Procedure WPS/BGE-1 qualified in accordance with the requirements of Code Case N-432. The use of the subject proced- ' ure which is based on controlled heat input allowed the cladding i I
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of.the head.to be accomplished without performing a high' temper- I L_ T . ature (1200'F - 1200 F) post weld heat treatment. The procedure i as required by the Code Case employs a 300'F preheat and a 500 F ', s post. weld heat treatment. The inspector verified that the key
;, criteria for procedure qualification had been met - i.e., th i i Procedure Qualification Record (PQR) 6356 for WPS/BGE-1 showed !
that the heat affected zone (HAZ) fracture toughness test values ; [ were better than the base meta P ' ' The inspector's review of the aforementioned travelers indicated ! that after the 0.Ds of the head was cladded and inspected by-p liquid penetrant and ultrasonic examinations, B&W proceeded to P install the heater assemblies, as' follows: (a) weld outer e
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3 sleeve to new cladding on 0.D. of head (structural weld) in ! E accordance with manual- TIG Procedure WPS/BGE-3, (b) weld outer
!j sleeve to 1.D. of head as a non structural weld in accordance i with automatic autogenous (without filler metal) TIG Procedure '
WPS/BGE-2, (c) weld outer sleeve to inner sleeve as a-structural weld usint an automatic TIG procedure in accordance with Proced , ! ure WPS/BGE-4, and (d) insert alloy 600 heater and weld it to inner sleeve as a structural weld using automatic TIG Procedure . WPS/BGE- The inspector's review of the above Inconel . welding ! procedures and procedure qualification records indicated con- l formance. to Section IX and Code Case requirement Review of appropriate travelers indicated that the subject welds had been '
, liquid! penetrant . inspected. On June 12,1990, the inspector- I '
visually inspected the finished welds, all of which were com-pleted except for eighty heater to sleeve seal weld The
~. , welds, which were for the most part in the as-welded condition, ,
were typica'i of TIG welds, relatively smooth and free of gross ~ L surface irregularitie . The inspector.also reviewed the following documents and found no j
' deficiencies or discrepancies: . (a) certified material test i reports (CMTR) for SB-166 (bar)-alloy 690 outer. sleeves (Valinox - heat number NX6753), and 58-163 tubing alloy 690 inner sleeve , (SANDVIK - heat number 764336), (b) CMTR for Inconel.. ERNI Cr-3 .
filler wire (INCO heat number NX380300), and (c) Welder Perform- l ance Qualification Records for Welder 556 ~ It is noted that the inspector reviewed reports that indicated B&W performed metallurgical checks of the sleeve material to assure -optimum corrosion resistanc On the basis of these checks, the tubing material for the inner sleeves was reannealed at 2000'F to produce a yield strength of 44,702 psi. The bar material used for the outer sleeve was reported to exhibit a T yield strength of 40,405 - 41,416 psi . (Inconel alloys with
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tv .: yield strength values in excess of 50,000 psi are considered to n . be susceptible to 1WSCC). The licensee provided the inspector - with two significant metallurgical reports of B&W studies that l~ l provide additional assurance that the installed heater assem-F blies ' were of optimum quality. . These reports were (1) welda-o
[ bility studies of alloy 690 and (2) metallurgical evaluation of temper bead procedure qualification test assemblies. The licen-(
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L . see also provided several sectional test assemblies representing the- various sleeve welds. All of the samples exhibited good fusion and were free of cracks and porosity. In order to mini- i mize fit-up problems during the insertion of the heaters into '
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the sleeves, B&W employed a cold shrinking technique during p installation of the outer sleeve into the head penetration.; ' The inspector reviewed the nonconformance list and found no ; W significant problems except for the indications noted previously i during liquid penetrant inspe: tion of the existing I.D. clad-din Seventy-one (71) of one hundred nineteen outer sleeves failed to pass a free path gaging test prior to installation of , the heaters. Correction of this condition was accomplished by r , controlled grinding using a flapper wheel. Also, instances were i F noted where temperature deviations were observed during preheat and post weld heat treatments involving the temperbead cladding L operation? In these instances the temperature deviated from the more stringent B&W requirements, but not the Code Case require- ; ment The inspector reviewed several licensee's audit reports covering the replacement program. The reports were found to be ' explicit and covered various phases of the program including ~ administrative, QA, engineering and work activities. No signif-
- icant audit findings were reporte '
L , ' One of thelone hundred twenty (120) penetrations is not intended to be sleeved because of excessive metal removed during removal of an existing sleeve used for metallurgical investigation. It will be plugge Four instrumentation nozzles were removed and E replaced with new materia Only one of the- four nozzles exhibited crack The ' cracked portion is presently being , examined in the laborator Conclusion - " The inspector's review of the heater sleeve project indicated r
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that the licensee's contractor (B&W) was accomplishing the work I
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in a controlled and effective manner under an acceptable QA pro- ; L, gram. The licensee was maintaining close scrutiny of all oper- i ations. Most of the welding was performed using automatic pro-cedures. Extreme care was being exercised to avoid damaging the i sleeves during installation to minimize the potential for future stress corrosion cracking. No unacceptable conditions were observe '
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i - n ?.. 4 4.2 Surveillance Observation L The inspectors witnessed selected survd11ance tests to determine whether properly approved procedures. were in use; technical specifi-cation frequency and action statement requirements were _ satisfied; necessary equipment tagging was performed; test instrumentation . was in' calibration and properly used; testing was performed by qualified personnel; and test results satisfied acceptance criteria or- were-properly dispositioned, Portions of - the following- activities were reviewed, Low Pressure Safety Injection System (LPSI) Pump Performance '- Test, p The inspector witnessed satisfactory performance of STP-073J-1
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on June 14, 1990. The plant was shut down in Mode 5 with reduced RCS inventory. The purpose of the test was to verify operability of LPSI pumps by meeting ASME Code Section XI requirements. The quarterly! test required operators to measure
. specific pump parameters and compare the measurements to pre '
deternc..ied limits to verify acceptable pump performanc Overall licensee performance in preparation, test, coordination and results. analysis was satisfactory, The Control Room Super-visor conducted a deliberate and complete pre-test briefing which -included. a review of the procedure, noting initial condi-tions and precautions. The -inspector observed succinct. and pro-
.fessional communications bete en control room and auxiliary building personnel through the Wration of the tes In addi-tion, .the inspector noted c. 'ervative and safety conscious
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. discussions between operators curing test preparatio While preparing for the. test,: -thet operators questioned the accuracy of indicated RCS level, since'non-condensible gases had not been vented from the reactor vessel head following an
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earlier evolution to raise level. The-operators concluded that,
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after venting, RCS level.would lower thereby reducing the safety margi Prior to initiating the test, the operators further raised RCS level and vented the reactor vessel head to ensure-minimum RCS~1evel requirement > The inspector also observed appropriate response when they tem-porarily~ halted the test af ter operators noted deficiency tags on each LPSI -and HPSI suction pressure gage. The gages are nor-mally isolated and used for maintenance purposes only. The licensee modified the procedure to remove the installed gages .. i l _ . . - . .
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110 b . e a and L replace;them with temporary test gages to facilitate con-' ! tinued -testing. The inspector' determined that - the licensee-- ,~ first identified the ' deficient gages in February 199 The licensee initiated a maintenance request (MR) to document the t , gage deficiency and has initiated an' engineering-review to' ad- -,! dress a perceived design problem. Licensee personnel indicated that short term corrective actions include procedure changes to i require installation of temporary test gages during testing and l long.--term ~ corrective actions may require a design change for-
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O installation' of gages that are less sensitive to vibration ; ,' induced performance degradatio The inspectors noted - uncer - !
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tainty among the station staff -involved over the need to 'initi-ate a problem report to docume'nt the above proble ; 1 High Pressure Safety Injection (HPSI) Pump Performance Tes The inspector . witnessed the satisfactory performance of STP- I 0731-1 on June 15, 1990. The purpose of the test was to verify '[
'the operability of HPSI pumps by meeting ASME Code Section XI j : requirements. The quarterly test required operators to measure I specific pump parameters and compare the measurements to pre- 7 \
determined limits to verify acceptable pump performanc The inspector noted changes in the test procedure to install temporary test gage This was in response to the deficient gages found the previous day when performing STP-073J-1. ihe-licensee temporarily halted the performance of this test _ when one of the test gages - failed to function properl As was - 1
~o bserved during the LPSI test, the inspector noted appropriate '
operator response to halt the procedure, and thoughtfully pursu ; corrective action prior to continuing the test. No unacceptable .i L conditions were noted, j Pressurizer Relief Valve (ERV) Channel Functional Test (M-6728-2) ' On June 26, 1990, the inspector witnessed testing of the Unit 1 pressurizer power operated' relief valve actuation circuitry as- 3 required by: technical specification 4.4.9.3.1.a. . This test was i performed in accordance with-test procedure-STP-M-672B-1,'" Pres-surizer Relief Valve Channel Functicnal Test". The test was conducted in accordance with the procedure with repairs per- I formed as require ;
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While' attempting to.-initiate this test,. the craf t experienced difficulty in establishing a closed test loop. The test equip- 1 ment was checked _ and found to be _ working properly. Test: equip- , ment connections and leads were _ checked and found acceptable, ' yet the problem eventually corrected itself. for no apparent-The craft then proceeded with the performance of th ' reaso s tes The inspector commented that this anomaly was not anno-tated on1the test cover sheet for future reference in case the r-problem should reoccu The licensee agreed that the problem l
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should have been documente The inspector concluded that the absence of this documentation is an example of ai weakness ,i regarding the failure to aggressively capture all available test data, Review of Surveillance Results ; The inspectors reviewed the following sample of 'recently com- ! pleted safety-related system surveillances to verify that the * tests . had been conducted in accordance with approved' procedure This review included review of the workplan, pre-test briefing,. authorized signature, tagging, acceptance criteria _and review by ,
,cognizart personne '
STP M-0-103-1, "HPSI MOV Maximum Differential Pressure Stroke Test" STP M-520, " Calibration of ESFAS Transmitters" Licensee efforts during the observed surveillance tests 'were ,
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conservative .and exhibited safety-consciousnes No unaccept-able. conditions were note . Emergency Preparedness
' .The -inspectors observed portions of the emergency preparedness exercise conducted by the licensee on June 25,-1990. The purpose of this exercise '
was- to demonstrate -an adequate readiness to respond to health physics
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scenarios by L testing all . major elements of the emergency response plan c with <the exception of fire and personnel injury. The inspector observed activity in both the technical support and operations support centers. No noteworthy items.were identifie . Security '
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During routine inspection tours, the inspectors observed implementation of portions of .the security pla Areas observed included access point searchi equipment operation, condition of physical barriers, site access s-control, security force staffing, and response to system alarms and de-graded conditions, These areas of program implementation were determined to be adequate. No notaworthy findings were observe J y y ..
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, 7. Engineering an'd Technical Support ~
The- inspector reviewed selected design changes and ' modifications made to the. facility which the licensee determined were not unreviewed safety questions and did not require prior NRC approval as described by 10 CFR 1 50.59. Particular attention was given to safety evaluations, Plant Oper-ations Review Committee approval, procedural controls, post-modification _l testing, procedure changes resulting from this modification, . operator j training,. and UFSAR and drawing revision The - following activities were i reviewed: i
! Inspection and Repair of pipe Supports 'l In March 1989, the licensee identified damage to a Unit I low pres- 'sure safety injection (LPSI) system piping restraint, R-2, and deter-mined- that the cause of the damage was due to piping movement from . system water hamme The root cause of the water hammer has been traced to check valve sla This issue was reported in Licensee Event Report 89-0 ]'
As a result of thi.; event, the licensee initiated a check valve slam investigation proje: This project set . out- to review and analyze-
'this problem. and develop corrective actions. A part.of this project i was a . plant history review for previous similar damage. This-review
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identified one additienal support, R-16, on the LPSI system, that had { previous damage which was repaired in 1986.
1 The licensee contracted for a dynamic analysis of the effects of - check valve slam events on the LPSI and component cooling water (CCW) ! system These systems were selected for detailed analysis. because : ! the design and operational >aracteristics of.these systems. indicated- l a likelihood for check Gye slam events and actual events were 1 observed by operations personne This analysis- provided conserva- i tive estimates of the loads transferred to che system piping sup- ! port From that analysis, the licensee determined several -piping j supports in the LPSI and CCW systems that had high estimated loads i and needed detailed inspections to determine pctential damag Inspections of the selected supporte were initiated by the licensee on June 11, 1990. The initial results of these. inspections identi- a fied support discrepancies on the .LPSI and CCW systems that were attributed either to check valve slam damage, possible check valve ! slam damage or construction defects, or clearly construction defect Additionally, a few supports in the high pressure safety injection 4 , (HPSI) system were identified as having construction defects with one ! of these potentially damaged by check valve sla The inspector walked down several of the affected LPSI and HPSI system supports f with the system engineer to review the discrepancie i
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The inspector- reviewed licensee corrective actions for the a'ffected supports. These actions included design. analysis of piping stresses
- ~ to determine system operability and an. evaluation to determine the : appropriate repairs and modifications to the affected hangers.- The '
results of the piping stress analysis showed that the piping-integ-s rity:Was not compromised. Design actions to address damage caused by water hammer include strengthening of these supports to withstand the k : re-evaluated forces from water hammer. Other discrepancies are also ' L 'being evaluated and corrected.
- To address the root cause of check valve slam events, the licensee is-
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evaluating methods to reduce these events via operational controls' < and design changes. An operational control on the LPSI system has been put in place to reduce system flow to a minimal rate prior to I shifting system pump A design change under consideration is- to L install new check valves that are designed to reduce check valve sla The inspector. concluded that a significant amount of engineering resources have been invoived in this issue and that as the inspection period closed, actions by the licensee to identify and correct- sup-port discrepancies have been' appropriate. The inspector.also'recog-nizes that additional resources will be required to resolve the root -- _ causes of this 1.ssue.
w= kc 7 2~ Salt Water System Performance $ W The inspectors continued to monitor the extensive engineering and field work involving the Unit 1 salt. water system._ This effort was D '
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; initiated following the discovery'of a. pinhole leak in the discharge
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piping of Unit I salt water pump No.13. The location of the failure <
' .was in arslip-on flange weld of a 30 inch concrete lined pipe 1 The-cause of the ' leak has been determined to be from localized; corrosio ~
resulting from- a failure of the internal ~ grout pipe lining. This' hand grout was applied to protect - the field weld from corrosion following the installation of slip-on flanges.
- The leak that occurred was from an elliptical hole of approximately 0.25 square' centimeters through an inner and outer weld used to con-
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" nect a pipe to a pipe flange. This particular flange arrangement s uses two wolds. The pressure boundary is established by an internal circumferential weld which is . not accessible after the pipe . is - installed. The second weld is an externa _l circumferential weld that bridges the outside gap between the pipe and flange and does not assist as a pressure boundary. The original seismic design assumed - that the internal pipe lining would remain intact to prevent any - degradation of the internal weld.
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encies. Of these six, .four had intact but degraded internal welds c [. and the remaining two flanges had through wall f ai l ure s .- In each :I' k case, the hand grout. was removed and replaced with an epoxy-tyra lining. The remaining inspection effort was ongoing -as the period 3 ende ,
'A significant concern'regarding-these failures is that the inner weld
would fail without detection for a period of time until the outer weld. faile This would constitute an- undectable failure of the pressure boundary, The licensee had not finalized their resolution - , of this concern .when the - inspection period ended, ' Efforts were 4 underway to allow use of the outer weld as a pressure boundary. In , addition, the licensee is considering drilling holes in the oute welds to. enable inspection- and detection of a f ailure of the inner 4 weld, The inspectors will continue to perform independent visual inspections :of the welds and monitor the licensee's correctiv . action A.second' concern was raised by the resident inspector staff--regarding the increased biological fouling rate of the service water heat exchangers, As discussed in Inspection Report 317/90-09 and 318/ 90-09,' the licensee had to place the No, 11 salt water header back in , service with an unisolable defect to facilitate cleaning-of a fouled No,12 service _ water heat exchanger. The heat exchangers were - required to be cleaned every 48 hours to remove a buildup.of shells and: grass from the tube-sheet face. Salt water passes - through -the , tubes 'and serves as the ultimate' heat sink -to the bay. 'Sho rt; -te rm
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r corrective actions included cleaning the ' intake structures and per-formance of high velocity flushes with chlorinated wate These actions were being implemented as the inspection period ende j U The inspectors concluded that adequate safety perspective was being ; applied by the station engineering staff in dealing with the salt i water-problems. - Engineering management was persistent in emphasizing the need to resolve the issues prior to either unit restart. No dis-1.' crepancies were identified during the review of this i s s ue '. The unresolved item associated with these issues (317/90-09-01 and 318/90-09-01) will remain open pending final resolution of the , problems.
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7.3 -Switchgear Room Air Conditioning L
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This issue was identified by the licensee and pertains to the ability of the equipment in the electrical switchgear rooms to perform under . certain_ temperature and air conditioning (A/C) conditions. The elec- '" trical equipment in the switchgear rooms supplies the power for the- i S : operation of safety related mechanical equipment required for safe j shutdown. This equipment was designed to- conform to the service
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conditions specified .in ANSI C37.04 and ANSI C37.20. -These standards establish the maximum air space temperature of 104 degrees F for which the- equipment will operate with a high degree of reliabilit . Plant operating experience and calculations indicate that with no y cooling, the - temperature in the rooms will exceed 104 degrees i The switchgear room HVAC system was designed as a safety _ related sys- ' f,
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tem with two . redundant cooling units. Each unit is powered from a separate .and redundant emergency bus to ensure operation of at least one cooling 1 unit given any-single active failure. The original plant-technical-specifications contained a requirement to maintain the room
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temperature less than 104 degrees F.- This requirement was deleted during the transition to the standardized technical specification, The licensee . can -not' locate any ' supporting engineering analysis t support.the deletion-of this requiremen '
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l - Operation of the switchgear room HVAC system is prescribed in oper-ating procedure OI-22H. Section VI of this procedure specifies the ', requirements -for operation without either A/C compressor availabl This section previously' contained provisions to place both compressor-b units to off and use instrument air to force the system dampers - to remain in the outside air mode. This mode was used to take advantage of-the cool spring and fall outside air and prevent excessive cycling , of - the compressor- units. .This condition was also utilized during s _ periods when-both cooling units were inoperable. The switchgear room temperatures are alarmed at 104 degrees F in the control room with an 'i
. annunciator response requirem'ent -to reduce temperature to less than'
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104 degrees or be in hot shutdown within 24 hour The. licensee realized during a period last summer that the switchgear
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rooms could not be maintained at a temperature less than 104 degrees F with both cooling units out of service and the outside dampers blocked open. Both reactors were in cold shutdown at the tim A N , non-conformance report was initiated and it was later discovered that to the method used to gag the dampers open would be ineffective follow-ing a loss of turbine building instrument air. This is significant since a loss of turbine building instrument air would occur following a safety injection actuation signal. This problem was addressed dur-
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ing the April power operations by mechanically blocking the outside dampers open in lieu of using instrument air, i i> l'~
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= F would be exceeded within 5 minutes in a worst case situation. The t licensee has contracted for additional analysis to-support an accept- !!
able' room temperature of 150 degrees Preliminary results indicate .; '
.that this would allow. operators approximately 10 minutes to restore room cooling following a safety injection. .This would also . suggest -
that both cooling units should be maintained operable to allow for a - single failure and still provide cooling to the switchgear room , r E> The inspectors reviewed the engineering work ongoing tc, resolve ~ this- j issue and discussed the scenarios with various members of the station i staf The licensee has identified this problem as_ a startup issue l
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that must be resolved prior to Unit 1 entry into mode 4. 'This issue
'is- identified as an unresolved item for review of the final analysis 4-and corrective actions (317/90-13-01 and 318/90-13-01).
-7.4. Potential: Problems in Circuit Breaker Coordination , On- June 13, 1990, the licensee identified five thermal magnetic' j p" breakers which may not coordinate with their respective- safety- i'
~ related 480V Motor Control Center (MCC) feeder breakers for a certain range of ground fault current Ground faults in these circuits could . lead to the deenergization of the associated MCCs. As ~ an .
immediate safety measure, the licensee opened these breakers and , applied , " Danger" . tag Installation .of replacement breakers with
' ground fault sensors and relays are Lplanned as the replacement '
, breakers are availabl ,
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V Problems in . protective relay and - circuit breaker coordination had b~een identified as a- potential generic safety concern by the NRC via-NRC. Information Notice-(IN 88-45) which was issued to. all utilities on. July 7, 1988. The licensee initiated a b eaker coordination study
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in'early 1989. To date, the licensee has completed studies on the -13 ( KV system, the 4160 V- system, and the safety related 480 V load cen- , ters. A preliminary review of' the safety related 480 V MCCs also has ? been completed during - which the above problems were found. The -[ licensee plans to complete a detailed study of the safety-related : 480 V system by the' end of August. The inspector concluded that : ; appropriate engineering resources are being applied in this area.- 1 The inspector will review the results of the detailed study on the ! 480 V safety related system in a future inspectio *
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., '7.5; Low Temperature Overpressure Protection (LTOP) '
The licensee continued to struggle with the implementation of_a sound program to- provide LTOP protectio The discovery of severa analytical deficiencies in the previous licensee calculations caused a significant delay in the Unit 1 outage. - The initial problem that caused the reanalysis pertained to the discovery that the- original LTOP calculations had assumed instantaneous' actuation of -the power operated relief valves to mitigate an LTOP event. This problem was I discovered after a question was raised by the NRR Project Manage l The licensee had'not previously verified the total response time of a
,% PORV, In addition, subsequent errors were found regarding.the model-
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ing of the' reactor coolant system pressure and _ the actual ' energy
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input into the reactor -coolant system following an LTOP event. Unit i 1 was being maintained -in a vented condition with the pressurizer manway removed until these issues are resolve , J The inspectors monitored -the 1icensee progress towards resolution of the 'above issues and participated in several conference calls with " the NRC on. the subject. This item is identified as an unresolved item (317/90-13-02 and 318/90-13-02) pending -a review of the -final resolution. Of particular concern is that the issue of not testing-
'the PORVs to verify stoke times within the LTOP assumptions was iden-tified to the licensee via NRC Information Notice 89-32, dated March 23, 198 The licensee has acknowledged an inappropriate -1 i
initial' response to the Information Notic , ' Safety Assessment and Quality Verification
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8.1. Plant Operations and-Safety Review Committee The inspector attended several Plant Operations and Safety Review '
' Committee-(POSRC). meeting TS 6.5 requirements- for required member
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attendance were verified. The meeting agendas included' procedural
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changes, proposed changes to the- TS, Facility Change Requests, - and J
: minutes from previous meetings. Items for which adequate review tim ~
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was 'not available' were postponed to allow committee members time for l m further review and. comment. Overall, the level of' review and member- J participation was adequate in fulfilling the POSRC responsibilitie ! I _ Review of Written Reports j d;
.J Periodic 'and Special Reports, Licensee Event Reports (LERs), and-Safeguards Event Reports-(SERs) were reviewed for clarity, validity, accuracy of the root cause and safety significance description, and adequacy of corrective actio The inspector determined whether t further information was require The inspector also verified that I the reporting requirements of 10 CFR 50.73, 10 CFR 73.71, Station Administrative and Operating, and Security Procedures, and Technical Specification 6.9 had been met. The following reports were reviewed: l j
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LER 317/90-17- Leaking Weld and- Bio-Fouling in Saltwater System Special Report concerning the inoperability of the diesel ' fire pump ' in excess of seven days, dated June 13, 199 .3- Temporary Waivers of Compliance ! The- inspectors discussed .the procedures for requesting a temporary i waiver of compliance from- the Technical Specifications with the i appropriate members of the licensee staff. A copy of the NRC memor-andum on the subject, dated- February 22, 1990, was provided to the licensee for incorporation into the applicable station.. procedure This memorandum is also available in the public document -' rooms .
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8.4 Programmatic Review of Low pressure Safety Injection System Safety-System Functional Inspection ' The inspector- performed- a programmatic review of licensee. activities toi address findings from the Low Pressure Safety Injecti on (LPSI) j system Safety System Functional Inspection (SSFI).Tne purpose of the- ! review was 'to identify: responsible licensee organizations and' assess the : completeness and adequacy ~ of their activities to ensure that 1
. inspection findings were tracked and resolve An SSFI-is a performance based inspection to assess safety system and safety support system functionalit The licensee,- through its , quality ' assurance audit program, employed- six contractors and six ;
licensee personnel to perform the LPSI. system SSFI between June 26, . andi October 6,1989. The team found that the LPSI=' system in the , sinjection. and shutdown cooling modes of: operation was acceptabl l However, the team observed thirty-eight weaknesses that they con-- 4
.cluded could degrade LPSI system performance in the . injection or-:
shutdown cooling modes or cause the-system to operate in an unantici- - pated ~or unexpected manne The inspector found that the licensee's 1 Quality Assurance Audit Section developed forty-two . audit findings 3
- from the team's thirty-eight observations. The licensee tasked 4 d engineers to review all of.the inspection findings and recommend to cthe Plant Operations and Safety Review Committee (POSRC)' those that- 1 should be considered restart 1tems. The engineers presented to POSRC the sresults of their review including 12 items requiring corrective action . prior to restart. The committee concluded that only 6 find-ings warranted resolution prior to restar The inspector observed .
that only 3 :of the 6 restart items were included in the 12 recom- 1 mended by the task engineer _ -
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.The; inspector determined that the Quality Assurance-(QA) Audit Sec .
tion . is responsible for tracking audit findings and that a~ recent change to QA audit procedures (QAP 21) required that the audited organization be responsible for finding resolution. After a review of QA audit records and discussions with licensee personnel, thel _ inspector concluded that LPSI system SSFI restart- items had been promptly resolved and that the remaining open items were effectively tracked. A sLbstantial number of the remaining findings were- found
'open awaiting QA audit assessment for final closur The inspector .specifically reviewed the status of the 12_ safety significant find-ings -recommended by engineers to the Plant Operations and Safety Review Committee (POSRC) and found thct 5' findings remained .open awaiting formal technical. resolution from the- audited organizations, i
Based- upon the review performed, the inspector- concluded that the l licensee's program is adequate to ensure tracking and resolution of -{ LPSI system SSFI findings. However, the inspector noted weakness in '
. licensee implementation of the QAP 21 Section 8.2 requirements for the audited organization's response to findings-within 30 drys of the audit report date and Section 8.3 requirements 'for request' ng exten-
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sions of responso due date l 8.5 Ro_ot Cause Analysis on Reactor Coolant Pump-Vibration j During a' previous inspection (50-317/90-08 and 50-318/90-08), the ;j ' inspectors had reviewed the licensee's engineering analysis to sup-- port- continued operation of Unit 1 Reactor Coolant Pump (RCP) 128 j following an ' alarm that indicated high vibration on April. 8,199 The- analysis _ had:cencluded-that the most probable cause of the vibra-
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t' ion anomaly was a slight misalignment of the pump-to-motor ~ couplin This - conclusion : was _ supported by a further vibration -data analysis l q 4 conducted during the April startup of Unit During the current Unit 1 maintenance ou'tage, the licensee inspected the coupling faces and the pump shaft as part of the RCP seal package I
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)i f replacement project. It was noted during disasse'mbly of the pump }
half-coupling thut the lock nut was removed with less than the ! 1 required torque of 500 ft-lbs. The licensee also identified that the i i s'pli t ring and' the spacer ring were slightly warped. The licensee system engineers. concluded that these observed discrepancies were a j ,, result of the coupling misalignmen The inspector reviewed the
, records from the previous pump coupling reassembly work conducted in April 28, 1988 and . verified that the records indicate the lock nut q
was torqued to 500 ft-lbs with a QC hold point, ,
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To achieve' b'etter alignment of ' the pump drive half coupling, ' the e Llicensee- required. additional reference measurements on- the clearance
- ' of the coupling and the pump shaf t. -The new procedure' also requires a " Blue" check of the pump coupling to ensure proper alignment. The system engineers plan to monitor RCP vibrations . closely during the next Unit 1 startu The inspector determined that the licensee's- . inspection was thorough and concurred with the system engineer's ;
conclusion on the root cause analysi .6 Review of Licensee Unit 1 Startup Assessment Report On May 29,1990, the licensee submitted Unit 1 Startup Assessment u
. Report (SAR): documenting thei r. self-assessment of personnel and equipment performance during April 4-23, 1990, startup and power i operation of. Unit The SAR, including the line organization's }
self-evaluation and an Independent Assessment Team Evaluation, was .i reviewed by the inspectors. On June 7,1990, the licensee management participated in a management meeting with the NRC staff at the NRC 7 Region I Offices in King of Prussia, Pennsylvania, to present the assessment results contained in-the SA The inspectors- concluded that the SAR represents a forthright self-assessment and indicates successful completion of the licensee's Unit
! .1 Startup Plan, dated March 19, 1990. The SAR provided a slightly :
more positive view of overall performance than the NRC inspection
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report (50-317/89-08: and 50-318/90-08); however, the difference is , not significan The-licensee assessment agrees closely with the.-
' inspectors' assessment regarding: safety .significant areas to be tar- t geted for further _ improvement and continued management attentio . These include interdepartmental communications, attention to detail, , ,
procedure compliance _, root .cause. analysis, and maintenance order _ backlo The' inspectors agree that a number of benefits were derived in the q areas of preplanning, prioritizing of activities, teamwork, and self - ' assessment as a direct result of the use of the Unit 1 Startup Pla ; Especially, utilization of the Start-up Review Board (SURB) as char-- i . ;1
,. .tered . in the startup plan, focused management and staff efforts by conveying expected results and instilling a sense of accountability .,
and ownership. 'The licensee management plans to utilize major ele- -i ments of the startup plan-, including the SURB, for the next~ Unit 1 startup and other major startups in the futur ,, i I 't (-.
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8.7= Licensee Investigation on Conditional Release of Material Program r During a Joint Utility Management Audit in May; 1990,- the licensee ; identified that inddegoate control of conditionally released lubrica-ting. oil-.resulted in lack of traceability of the: consumed oil. - When- > the lubricating oil . (Texaco Regal R&0 46) was received in January,
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1989, it was placed on hold pending receipt inspection testing. per the commercial dedication program, Some oil however was condition- .
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ally 1 released for plant:use and a non-conformance report (NCR 5863) was issued' to control the oi However,' due to: an inadequate- -l requirament for controlling conditionally released material, espec-
-ially for a commercial- grade consumable item such as oil, the sub- -
ject lube oil was not traced within the plant by specific batch traceabilit As an immediate corrective action, the licensee is in the process of replacing _ lube oil in plant aquipment which uses Texaco Regal R&0.46 oil. The licensee also has taken oil samples from all reactor cool- ,
-ant. pump motor' bearing reservoirs to determine if any ~ are contam-
-i inated, L At Lthe end of the inspection period, the licensee has not- a
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, ' completed 1 reviews on the results of the oil. sample analysis. The
' licensee. is also conducting operability analysis- on plant equipment l
, which.uses the subject oil, As. a ..long term corrective action to improve the conditional release
. program, . the Calvert Cliffs Instruction - (CCI-116), " Identification and . Control of Nonconforming Conditions," was revised to require
. technical review by design engineering and concurrence of the General '
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Supervisor-Quality Assurance for each conditional - release request,- -!
' The Llicensee ~ is also in the . process of. upgrading the Procurement '
Quality Unit procedures to require better control and traceability of' ", l . commercial' quality consumable ' items, During the inspection, there was.one open conditional release item (opened in January,.1990).which was in a paperwork close-out process. The inspector's review- of the - m . package indicated no discrepancies. The inspector determined Lthat the licensee's investigation effort and corrective actions to date
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' tors will' review the results of the licensee's' investigation upon'its
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completio
-- 9, _ Followup of Previous Inspection Findings Licensee actions taken in response to open items and findings: from pre- .vious inspections were reviewed. The inspectors determined if corrective a actions were appropriate and thorough and previous ~ concerns were resolved, ' . Items were closed where the inspector determined that corrective' actions -
would - prevent recurrence, Those items for which additional licensee
', action was warranted remain open. The following items were reviewed, ,
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.\ - r 9.1- Significant Issues Management System
.t (0 pen) RAT Item Paragraph 3. (0 pen) Unresolved Item (50-317/: !
89-03-01 and 50-318/89-03-01). j On April 27, 1990,. the . licensee issued. a complete revision to CCI-
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116, " Identification and- Control of Nonconforming Conditions" This ,
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revision implemented the Problem Report (PR) as an initial . step in 1 the consolidation of the licensee's various tracking system's for: problems, maintenance and other issues into a: single system.
A The PR is used to identify potential conditions that are- adverse to
,. quality and replaces the Nonconformance Report (NCR) as he initial document to identify these -condition The PR differs frcm the NCR in that it allows--for the screening and resolution of problems that: l
. ^ are not conditions adverse to quality as well as those that ar If i the issue is determined to be a condition adverse to quality, it is- <" 1 transformed-into a NC If initial screening determines that !t is not a quality issue, the PR provides for resolution via other . avail-B
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able methods. In addition, the PR implements- criteria to- determine , the applicability of a root cause analysi I The inspector reviewed CCI-116 and discussed the PR system with the
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licensee. The PR is designed to provide screening for reportability- , by .first . line super 51sion yet the inspector noted - that some of- the s
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criteria are general in nature and could be misinterprete The sinspector. noted that there .is some uncertainty; among plant- :i 0 workers regarding PR's. Examples observed during this inspection 4 i period include a hesitancy -to : issue a PR when gages for a; surveil- J lance test were found to have exceeded their calibration dates and i is the test. was stopped to' correct the proble In this case, the- : General Supervisor,made the decision to . issue the PR. .The inspector ? , also received feedback from. a few workers who questioned why there was a need for another document to identify issues adverse to qualit since-many PR's are transformed into NCR's,
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The inspector concluded that the PR system constitutes a small modif-
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ication to the previously existing NCR-system and is an initial step 4 to consolidate the various tracking systems. The licensee currently has a schedule to continue this consolidation-with action to be com-plete by the end of _1990. - Some of the observations identified above 1 m were assessed by the inspector to be, in part, the result of- a pro-
* . gram undergoing change. Since the PR is an initial step toward con-solidation of the licensee's multiple tracking sy stems , this issue .will ' remain open pending NRC review of future licensee action ! ~
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5149 h[ > 2 . 9.2 (0 pen STI-15) Alternate Safe Shutdown Control Room Evacuation Procedure
! .This issue was reviewed in NRC Inspection Report 50-317/90-05 and 50- i 318/90-0 During that inspection, the licensee initiated analysis for Auxiliary Feedwater (AFW) . initiation time. Timely initiation is-requiredito ensure that Appendix R criteria are met. The analysis showedLthat the criteria would be met if AFWLwere initiated within 45' ,
minutes. The previous inspection also noted weaknesses .in the vali-dation process. The licensee has placed validation of the: Alternate - Safe: Shutdown,-Control Room Evacuation. Procedure AOP-9A on-the-Unit 21 startup issues lis This item remains' open pending resolution by the licensee regarding the possible need- for a Technical Specifica--
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tion change for' shift staffing, completion of work for Unit 2, final procedure. validation by the licensee, and a review of the technical basis documen .3 (Closed) UNR 50-317/88-19-02
This issue ^ involved the effect of a higher service water temperature-on- Emergency Diesel Generator operability. Using the 10 CFR 50.59 process, the licensee increased the allowable bay temperature to 8 degrees F. 'This corresponded to an allowable service water tempera- , ture of.105 degrees F. The licensee had not evaluated- the ' impact of-
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the higher service water temperature on generator loading and dura-tion of loading, and therefore agreed to resolve therissue in con-junction .with an overall plant cooling study being conducted by
, : Design-Engineerin .The . inspector reviewed associated correspondence and the licensee's '
actions to dat From the review and- discussions with. cognizant 1 slicensee personnel, the inspector concluded that the licensee's plant cooling studynto resolve Salt Water Heat Exchanger' unresolved items-- 150-317/90-09-01 and 50-318/90-09-01 would .also; address the 1 generator
' operability issu The generator operability issue will. be' tracke by the above unresolved item and reviewed by the inspector ' prior to; unit-startu This item is considered close ,
9.4 (Closed) Violation 50-317/87-09 (his violation involved the isolation of all four level sensor chan- J nels for the refueling water tank that occurred on February 17, 198 ,
. An enforcement conference was held on April 28, 1987 to discuss the '
specifics of this event. The cause of the error was determined to be an inadequate communication between the control room operator and an outside operator. As stated in LER 317/87-005, the level transmit-ters were inoperable for approximately 40 minute . f _ _ _ _ _ _ _ _ _ _ _ . __________._._________________._________._8
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:The inspector' reviewed the proposed actions contained in the licensee response to ' the Notice of Violation, dated- July 2,1987, and ' spot -
checked' several 'of the corrective - actions to verify they have-
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remained in place. No additional problems were identified regarding this issue' or the implementation of the corrective action This item is considered close .5- (Closed) Unresolved Item 50-317/88-05-01
.This issue concerned aspects of the reactor coolant system - (RCS) I pressure-temperature -limit curves (P-T curves). One-issue regards an analysis of the effects of a Unit 2 reactor pressure vessel head bolt ,
up with temperatures as low as 70 degrees F as allowed by the tech- ; nical specification curve but below' the 90 degree F limit provided-in g the Reactor Vessel Assembly Technical Manual, The second issue > involved the-need'for a clearer basis for why the 0-10 year P-T curve allowed a minimum RCS temperature of-70 degrees F and the 10-40 year P-T curve was: lower with a minimum RCS temperature of 40 degrees for !
. pressures below 20 psi The inspector reviewed licensee correspondence and documents relative j
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to these issues, From this review the inspector agreed with- the licensee that the effects of reactor vessel head bolt up at tempera-tures would produce no adverse effect The inspector also agreed 4 that' operation at temperatures below 70 degrees F at pressures below
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20 psia would produce no adverse effects. The bases for'these curves appear to be for refueling operations, q
;No > additional problems were identified regarding this issue, This item is considered close . ' Management-Meeting a
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During this inspection, periodic meetings 'were held with' station manage-ment to discuss inspection observations and findings, At the close of the inspection period, an exit meeting was held. to summarize the conclusions of the inspection-
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10.1 Preliminary Inspection Findings
' Unresolved Item 50-317/90-13-01 and 50-318/90-13-01, Review Resolu--
p tion of.Switchgear Room HVAC Operability Concerns Unresolved Item 50-317/90-13-02 and 50-318/90-13-02, Review Resolu-tion of LTOP Issues.
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FIGURE 1 - SHOWS OLD AND NEW HEATER SLEEVE ASSEMBLIES
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