IR 05000482/1990022

Insp Rept 50-482/90-22 on 900501-31.Noncited Violation Noted.One Unresolved Item Identified.Major Areas Inspected: Plant Status,Operational Safety Verification,Monthly Maint & Surveillance Observation & Refueling Activities

ML20055D958
Person / Time
Site:	Wolf Creek
Issue date:	06/29/1990
From:	Joel Wiebe NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20055D954	List: IR 05000482/1990022 ML20055D953
References
50-482-90-22, NUDOCS 9007100194
Download: ML20055D958 (16)

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APPENDIX U.S. NUCLEAR REGULATORY COMMISSION

REGION IV

NRC Inspection Report: 50-482/90-22 Operating License: NPF-42 l Docket: 50-482 Licensee: Wolf Creek Nuclear Operating Corporation P.O. Box 411 Burlington, Kansas 66839 Facility Name: Wolf Creek Generating Station (WCGS)  ;

Inspection At: WCGS, Coffey County, Burlington, Kansas Inspection Conducted: May 1-31, 1990 Inspectors: M. E. Skow, Senior Resident Inspector Project Section D, Division of Reactor Projects L. L. Gundrum, Resident Inspector Project Section D, Division of Reactor Projects l i

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W. B. Jones, Project Engineer )

Project Section D, Division of Reactor Projects Approved: b

. S. Wiebe, Chief, Project Section D

/ b Atej ivision of Reactor Projects Inspection Summary ,3 Inspection Conducted May 1-31. 1990 (Report 50-482/90-22)

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Areas Inspected: Routine, unannounced inspection including plant status, l operational safety verification, monthly surveillance observation, monthly-i maintenance observation, refueling activities, plant startup from refueling, I onsite followup of events at operating power reactors, and verification of containment integrit Results: A noncited violation of Technical Specification (TS) 3/4.5.1 was !

i identified by the licensee for two reactor coolant system (RCS) safety l injection accumulators. The accumulator discharge valves had not been restored to operable per the TS following surveillance testing. The surveillance procedure did not provide sufficient guidance to restore the ;

accumulators to operable, and the operators did not ensure that the accumulators '

had been properly restored prior to declaring the accumulators operabl The unit experienced an unplanned reactor trip actuation while in Mode 3,-and three unplanned reactor trips (paragraph-8),

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i The reactor trip ectuation occurred because the operators did not restore

, feedwater flow to the steam generators in time to prevent a low level

! following feedwater check valve inservice testing. The operators had not .

considered the feedwater system isolation signal (FWIS). that was presen '

Feedwater was not restored until the low-low steam generator reactor trip c actuation was unavailable. The inservice test procedure which had lead to the ,

initial steam generator high-high levels did not adequately consider the '

i possibility of high steam generator levels.

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, The reactor trip on May 14, 1990, resulted from decreased feedwater heatin l

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The status of feedwater heating was not accounted for by the operators on the surveillance test procedure for main turbine overspeed testing.-

The remaining two reactor trips on May 17 and 19, 1990, resulted from I equipment failures. The first reactor trip occurred when an atmospheric dump

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valve inadvertently opened with the reactor at low power. The second reactor ,

trip occurred from 98 percent power when the main turbine tripped on a high j moisture separator level. In both cases, the licensee's response to the l initial reactor trips, and subsequent plant transients, was vs , good (paragraphs 3 and 8). The licensee's review of the May 19, 1990, reactor trip

did not adequately document the cause of the trip or the corrective action

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take This remains an unresolved item (paragraph 8).

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l DETAIL . Persons Contacted ,

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Principal Licensee Personnel B. D. Withers, President and CEO

• J. A. Bailey, Vice President, Nuclear Operations F. T. Rhodes, Vice President, Engineering and Technical Services -

• G. D. Boyer, Plant Manager

• R. S. Benedict, Manager, Quality Control-(QC) l

• H. K. Chernoff, Supervisor, Licensing ,

A. B. Clason, Supervisor, Maintenance Engineering 4

• M. E. Dingler, Manager, Nuclear Plant Engineering (NPE) Systems -

l R. B. Flannigan, Manager, Nuclear Safety Engineering (NSE) 4

• C. W. Fowler, Manager, Instrumcatation and Control (I&C) ,

• R. C. Hagan, Manager, Nuclear Services

• N. W. Hoadley, Manager, Plant Design, NPE

• R. W. Holloway, Manager, Maintenance and Modification *

• R. K. Lewis, Supervisor, Results Engineering

• W. M. Lindsay, Manager, Quality Assurance (QA)  :

• R. L. Logsdon, Manager, Chemistry i

• T. S. Morrill, Manager, Radiation Protection ,

• D. G. Moseby, Supervisor, Operations

• G. Naylor, Supervisor, Operations Support-

• B. Norton, Manager, Technical Support

• E. Parry, Director, Quality

• J. M. Pippin, Manager, NPE C.'M Sprout, Section Manager, NPE, i

• L. W. Stevens, Supervising Engineer, NSE

• J. D. Weeks, Manager, Operations

'S. G. Wideman, Senior Licensing Specialist ,

• M. G. Williams, Manager, Plant Support *

The inspectors also contacted other members of the licensee's' staff during_ {

the inspection period to discuss identified issue '

' Denotes personnel in attendance at the exit meeting held on May 31, 1990, i Plant Status I

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A refueling outage was in progress and the plant was in Mode 5 (cold shutdown) at-the beginning of the inspection period. . On May 7, 1990, the plant was placed in Mode 4 (hot shutdown), and on May 10 the plant was placed in Mode 3 (hot standby). On May 12 the licensee placed the plant ;

in Mode 2 (startup), and on May 13 the reactor reached criticality and .

Mode 1. A reactor trip occurred on May 14.from about 18 percent power because of a loss of feedwate_r heating.. On May 15, following the licensee's determination that all safety systems; functioned as designed following the reactor trip, the reactor was restarted and placed in

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Mode Refueling Outage IV was completed on May 16 when the main, .

generator was synchronized to the grid. The reactor was returned to

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Mode 2 on May 17 to repair a leak on the "A" main feedwater pump seal ;

return line. While at the reduced power level, a reactor trip occurred when an atmospheric relief valve failed open. On May 18, the reactor was .

restarted and returned to Mode.l. On May 19, the reactor tripped from about-98 percent power when a dump valve on a moisture separator reheater failed to open resulting in a main turbine trip. The reactor was restarted and returned to Mode 1 on May 20, 199 The plant was operating at 100 percent power at the end of the inspection period.-

In addition to the three unplanned reactor trips.-a reactor trip actuation occurred on May 10, 1990, because of a. low-low steam generator level .

following a surveillance test'while the reactor was shut dow . Operational Safety Verification (71707} [

The purpose of this inspection was to ensure:that the facility was being operated safely and in conformance with license and regulatory requirement It also was to ensure that the licensee's management contro11 systems were ,

effectively promoting continued safe operation. The methods used t .!

perform this inspection included direct observation of activities and equipment, tours of the facility. -interviews and discussions with licensee personnel, independent verification of safety system status and limiting conditions for operation (LCOs), corrective actions, and review of facility ;

record Areas reviewed during this inspection included, but were not limited to, control room activities, routine surveillances, engineered safety feature operability, radiation protection controls, fire protection, security, !

, plant cleanliness, instrumentation and alarms, deficiency reports, and i I

corrective actions. Selected inspector observations are discussed below:

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o On May 2, 1990, the positive displacement charging pump speed controller failed high causing the pump to gain speed and increase flow and pressure. The relief valve downstream of-the pump lifted'as designed and limited pressure in the system piping to less than the '

hydrostatic test pressure. However, a welded flange connection joint upstream of the relief valve began-to leak. .The licensee removed and replaced the valve (BG V8118) and a short run of piping that included the 1 leaking joint. The joint was sent.to a lab for analysis and  ;

preliminary results indicated that the weld was already weakened by '

fatigue when the pressure surge occurred. The licensee began an ,

evaluation to determine if modifications are required to prevent +

recurrence of the fatigue. The pump, valve, piping, and weld are ,

nonsafety relate ,

o On May 2, 1990, BG HCV-83578, the reactor coolant pump (RCP) seal..

. injection valve, failed open. The "A" valve (BG HCV-8357A) had previously failed open and the licensee was preparing to replace it when the "B" valve failed. The valves, solenoid-operated and

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pilot-assisted, are used to throttle alternate seal injection to the RCPs. The valves also isolate on a safety injection signal. The licensee has determined that, if these valves fail open following a  ;

safety injection signal, the plant remains within the design basis -

and the event bounded by the Updated Safety Analysis Report (USAR).

The licensee previously received information concerning the potential ,

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failure mode'of the valves as well as possible means to preclude that failure mode. The potential failure mode involved air or-vapor

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displacing the water above the valve disk in the pilot portion of- the valve. If a high differential pressure then developed across the ,

valve, it could be driven open by the differential pressur *

Potential failure mode conditions were present during the outag i Water may have drained out from above the disk when the system was

. drained during the outage. A high differential pressure was present  !

because the charging system was supplying water to the reacto i coolant system (RCS), which was at a relatively low pressur *

The licensee replaced the valves and installed the solenoids below the system pipe instead of above it as originally installed. This f

. modified installation was expected to prevent water from draining out t of the pilot portion of the valve operator when the system is draine While the licensee knew that a possibility existed for this type of failure and that the valves were known to leak by, the licensee was

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planning to wait until the next outage to modify the valve orientation, i There are eight similar valves installed in the plant. The licensee -

has determined that these valves failing open 1s-bounded within the USAR accident analysi The licensee is performing engineering-design work on the reorientation of those valves and stated that the eight .

other valves would also be reoriented in Refueling Outage V in 199 j

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o Operators responded well to the various events discussed above and in paragraph 8. They quickly and accurately recognized the transients and took prompt corrective actior.s using appropriate procedure ,

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o On May 10, 1990, with the reactor in Mode 3, the licensee concluded that two of the four main steam isolation valves (MSIVs) were inoperable. This conclusion was reached after two valves-were identified as partially open, although the respective valve switche were in the closed positio l At 4:30 p.m. (CDT) the shift supervisor (SS) noted that MSIV AB HV-17 indicated partially open with its switch in the closed position. This indication was determined from the main control

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i board valve position status lights. An operator was then dispatched l to the valve to determine its actual position. When the operator- f reported that the valve appeared closed, the licensee initiated Work-

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Request (WR) 02897-90 to troubleshoot the valve position indicatio a At 5:30 p.m. the SS noted that MSIV AB HC-11 also appeared to be ,

partially open, as indicated by its respective status light b WR 02905-90 was then initiated to troubleshoot both MSIVs with dual -

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position indication, and WR 02897-90 was subseq0ently cancelled. At I 8:30 p.m. (CDT) the operators slow-closed both AB HV-17 and -1 Small valve movement was observed d; ring the close signal which !

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verified that the valves had been slightly ope The licensee determined that the valves had come off their seats and 1 were, therefore, inoperable because they were not positioned in accordance with their switch position TS 3.7.1.5 requires that,:in Modes 2 and 3, an inoperable MSIV be maintained closed or the reactor.be placed in HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.' With two .

MSIVs inoperable, the licensee _is required to meet the.cime-restraints ,

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specified in TS 3.0.3. Although the licensee had_ remained in Mode 3 with two MSIVs. inoperable, the LCO time restraints were not exceeded i for either TS 3.7.1.5 or 3.0.3. The requirements of TS 3.7.1.5 and 3.0.3 became applicable at approximately 4:30 p.m.-and 5:30 p.m.,

respectively, as a result of the indication that the valves vere partially ope The. licensee subsequently concluded that the valve disks he.t come off l their seats as'a result of.the valves being previously slow closed, c which does not securely seat the disks. The thermal expansion of the ;

valves during initial heatup apparently caused the' disks to lift off ;

their seats. A fast close signal, however, aligns the valves'

hydraulic accumulatur to the valve actuator position to drive the !

valve closed. This action ensures that the disks are securely engaged with the seats and, as long as the fast ci re signal is present, hydraulic pressure is maintained over the piston from the

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accumulator. Based on this finding, the licensee concluded that all ,

the MSIVs were operable and would have properly isolated their

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respective steam line if given a fast close. signa +

The inspector verified that the licensee's response to this event did not result in a violation of the TS. This instance demonstrates the

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importance.in believing plant indications until proven otherwis The inspectors found that the licensee responded appropriately to the above events. In each case, the facility was operated safely and in ,

conformance with license and regulatory requirement .

No violations or deviations were identifie . Monthly Surveillance Observation (61726)

The purpose of this inspection was to ascertain whether surveillance of *

safety-significant systems and components was being. conducted in accordance with TS. Methods used to perform this inspection included direct 3

. observation of licensee activities and review of record ;

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J Items inspected in this area included, but were not limited to, l verification that:

o Testing was accomplished by qualified personnel in accordance with an approved test procedure, 1 o The surveillance procedure was.in conformance with TS requirements, o The operating system and test instrumentation was within its current calibration cycle, o Required administrative approvals and clearances were obtained prior to initiating the tes o LCOs were met and the system was properly returned to service, and l r

o The test data were accurate and complete and the test results met TS <

requirement Surve111ances witnessed and/or reviewed by the inspectors are listed below:

o STS KJ-001B, Revision 9 " Integrated 0/G and Safeguards Actuations [

Test - Train B," performed May 1, 1290; o STS RE-014, Revision 0, " Cross Calibration of Wide and Narrow Range  :

RTDs " performed May 8,1990;

, o STS AL-103, Revision 11. " Turbine Driven Auxiliary Feedwater Pump-In l Service Pump Test," performed May 10, 1990; I

o STS AL-211, Revision 2, " Turbine Driven Auxiliary Feedwater System- ,

Flow Path Verification and In Service Check Valve Test," performed May 10, 1990; o STS PE-013, Revision 7 " Personnel Air Lock Seal Test," performed May 12, 1990; L

o STS SF-003, Revision 2, " Digital Rod Position Indication Test," i performed May 12, 1990; and <

o STS RE-013, Revision 3, "Incore-Excore Detector Calibration,"~

performed May 25, 199 A selected inspector observation is discussed below:

During the performance of STS AL-211, the procedure directed the operator  ;

' to line up discharge from the. turbine driven auxiliary teedwater pump to the four steam generators. The procedure indicated that 140,000 pounds-mass per hour was required to each steam generato The procedure then directed that flow be stopped and that operators verify that a check valve in each line shuts by positioning an operator to listen to the valve

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close. > :The operators performed .the flow isolation on one steam 3 generator at a time. At the time the test of the first check valve was

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accomplished, the "A" and "C" stear generators developed high-high water

'r ' level and an engineered safety features actuation occurred in the form of .

a feedwater, isolation system actuation. The operators subsequently allowed the iater_ level in the. steam generators to decrease by'sJeam rernovai When the licensee went to initiate feedwater.to the steam

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generators, a P-4 feedwater isolation signal- (FWIS) prevented using the ,

. normal feedwater system. Although the operators were aware of the P-4 j

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interlock, they failed to consider the- effect the interlock would-have' on

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-the feedwater system until it was too late to fill the-steam' generators in l;

@ an orderly manner. The delay in the addition of feedwater through the; e ' auxiliary feedwater system resulted in the low low-steam generator reactor.-

trip setpoint being reached. The reactor trip breakers were all ready

= open so ro' rod movement occurred. The operators implemented a: step-from-General: Procedure GEN 00-002, Revision 18, which allowed bypassing the P-4-FWIS and thus_ allowing the use of the normal feedwater system to rais steam generator level to the appropriate band._ The licensri, subsequently '

-revised the procedure to feed one steam generator ~at a timi n the-required = flow rate while performing the test. The test was then completed successfull The inspectors found that, with the exception of STS-AL-211', the-

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surveillance test procedure provic%d specific guidance on proper 't performance of the tests. However, STS-AL-211-did not provide sufficient ;

guidance on testing each check valve individually.-:The senior operators 1

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did- not provide sufficient' oversight of the= surveillance activity to

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-ensure that the high-high steam generator isolation did not occu Following the feedwater isolation, proper consideration was not..given to- !

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restoring feedwater to the steam generator once the : level had decrease .

This lack of prior planning resulted in'the low-low steam generator l reactor protection system sctuatio = . Monthly Maintenance Obseration - (62703)

a The purpose 'of . inspections in this area was to- ascertain that maintenance

.I activities on- safety-related - systems and components.were conducted in accordance with. approved procedures and TS. Methods used in this :!

4;. inspection included direct observation, personnel interviews, and reccrds

revie l This inspection included verification that

t o Activitie4 did not violate LCOs and that redundant components were i operable;

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Required administrative approvals and clearances'were obtained before initiating work; I k

o Radiological controls were properly implemented;

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o Fire prevention controls were implemented; i i

o Required-alignments and surveillances to verify postmaintenance operability were performed;' '

o Replacement parts and materials used were proper 1'y certified;

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o Craftsmen were qualified to accomplish the designated task and additional technical expertise was made available when needed; o QC hold points and/or checklists were used and QC personnel observed'

designated work activities; and , Procedures ut,ed sa m adequate, approved, and.up to date.-

Portions of selected maintenance activities regarding the WRs were observed. The following WRs-and related documents were reviewed:

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WR 02040-90 -Repair Fire Door No.15031

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WR 02662-90 Remove =and-replace Valve BG HCV-8357A'

WR 02687-90 Remove and replace' Valve BG V8118 and upstream oipe, perform hydro-test'after weld:

a WR 02720-90 Remove and replace: Valve BG HCV-8357Bc

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- - . . 1 WR 02926-90 Correct condition causing' cooler housing drain line- 1 to drip water WR 50642-90 Spent fuel pool cooling pump room cooler' pipe 3 replacement-- '_

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o A discussion of BG V8118 and the upstream pipe sectionl replacement appears in paragraph 3 of this repor !

o A discussion of the replacement of Valves BG HVC-8757A=and -B 1 appears in paragraph:3 of this repor *

l o WR 02926-90 concernedLwater found dripping from the "A motor driven

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aux 111ary feedwater pump (MDAFP) toom air cooler; housing drain.- Th i licentee found that-the leak was due to.a tube leak in the cooler. .

The tube was plugged and the cooler was returned to service. The "A" .

safety injection (SI) pump room cooler was also found to need repair- 1 earlier during the' inspection. That' cooler was used as a base (line inspection by the>1icensee with the expectation'that it may have'been !

the worst of the. room coolers. The licensee is developing a program t

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to open and inspect all of the room coolers in 'the future. A detailed inspection of the "A":MDAFP room air cooler has not yet been-performed by the license . The licensee evaluated' the -leakage that occured and determined that - 1 the leakage did not affect operability of the service-water system (loss of water) nor did it affect the operability of the equipment intheaffectedroom(coolingability). The licensee also evaluated'

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the effects of plugging the leaking.tubescand determined that the~

cooler was still'able,to transfer the required heat load. ' Since-the leakage was minor and the number of tubes plugged was not large.: the ,

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NRC has reasonable assurance that-the' cooler would have and will~now-remove the required heat loa The NRC had previously documented the review of'the licensee's-leakage and erosion problems in NRC. Inspection Report 50-482/90-2 ;

Review of the licensee's evaluation described above will be tracked by Inspector Followup Item 482/9020-01. ' Refueling Activities ='(60710)' '

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During Refueling Outage IV, which was completed:on May 16; 1990, .;

several significant items were scheduled for wor These items 1 included:

q o Reactor refueling; ,

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o Fuel assembly inspection;

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o Replacement of the seals in thel"B and,"C" RCPs;-

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o Performance of. eddy current tests on., tubes ~in'"B"1and "0" steam generators;  !

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o Performance of eddy! current tests on'the rod control!

cluster assemblies;

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o Reparation / modification of polar crane snubbe'rs and' wheels;- 1

o Performance of sludge lancing and search and' removal of foreign  !

objects- from the steam generator; 1 o

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o Performance of containment'loca111eak rate tests;-  ;

o o Installation of two independent RCS level indications in the control room- for midloop operation; and:

i o Fit testing of a new type of reactor cavity sealering, l

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All significant work was completed during the outage with the exception of-the fit test of the new reactor cavity-seal ring. The new seal ring did not arrive at.the site in time. The licensee was-found to have completed '

the refueling outage in a controlled and systematic manne Improved l management involvement was. evident,  ;

i No violations or deviations were identified, q Plant Startup From' Refueling (71711)

The purpose.of this inspection area was to ascertain whether systems .;

! maintained or: tested during Refueling Outage IV were returned to an- ),

operable status before plant startup- and to determine whether plan ,

startup. heatup, approach to criticality, and core physics tests following the refueling outage were conducted in'accordance with approved- ,

procedures, i

Prior to plant startup after Refueling Outage IV, a walkdown of the high .  ;

pressure SI system was performe The system was walked down to a independently verify that it had been returned to service in accordanc with approved procedure In= addition,:the' inspectors observed the-digital-rod position indication tes Startup and power- ascension . tests -

were still being reviewed by the inspector and will be discussed-in a future inspection. repor As the licensee's outage was drawing to a close, the inspectors verified that mode changes were performed in accordance with technically adequate and approved procedures. The startup test' program was. conducted in accordance with these procedure ,

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No violations or deviations were identifie ,

i = Onsite Followup of Events at Operating Power Reactors (93702)

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l The purpose of this inspection activity was.to provide onsite inspection of events at operating power reactors. Specific inspection activities included:

,, o Observing plant status; L o Evaluating the-significance of the events, performance of safet systems, and actions taken by the licensee; o Confirming that the licensee had made proper notification of the . '

I events and of any new developments'o* significant changes.in plan I conditions; and

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The following items were considered during the followupi j

o= Details regarding the cause of'the event,. i

" Event chronology,-

o Functioning of safety, systems as' required by plant conditions,,  ; Radiological. consequences and' personnel exposure, t

o Proposed licensee. actions =to~ correct the cause of.the event, and o Corrective actions taken.or planned prior to resumption of-facility' . ,

operation ,

Selected events requiring licensee . event reports (LERs)' that' occurred i during this report period are listed-in the table below:

Date Event * Plant Status Cause $

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5/04/90 FBVIS/CRVIS Mode Blown'. fuse

.(Cold Shutdown)J 5/10/90 FWIS Mode 3 .

High steam generator'

-(Hot. Standby), level 5/10/90 Reactor Trip- Mode 3 . Low-steam generator-  ;

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5/10/90 SI breakers Mode 3 Personne1' error and not locked procedural deficiency ,

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5/14/90 Reactor Trip . Mode 1 Low steam (18 Percent Power) generator level 1 5/17/90 Reactor Trip Mode 2 "C"1 steam generator '

(Startup). atmospheric relief I valve failed'open 5/19/90 Reactor Trip Mode 1 HighLlevel in the:"A" (98 PercentLPower) moisture separator reheater, ,

• Event CRVIS - Control room ventilation isolation system actuation FBVIS - Fuel  !

building ventilation isolation. system' actuation FWIS - Feed water isolation > system actuation SI - Safety injection

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-Selected inspector observations regarding the events are discussed below: ,

. o On May 10,-'1990, while the operators were performing tests on. check valves on the auxiliary feedwater system . lines to the steam , =

generators, an FWIS and, later, a reactor' trip occurred. During performance of the test, "A" and "C" steam. generators were.over .

filled'which caused the feedwater isolation. During recovery of the '

steam generator water ~1evel, the licensee was not able to reset the isohtion signal. The operator was unable.to control the water level ,

decro se c.nd a reactor trip actuation occurred (in Mode 3)_on lo steam generator water leve ,

o The inspector reviewed a licensee identified TS violation concerning:

the SI accumulators with the reactor in Mode 3 and the-primary system ,

pressurized to greater than 1000 pounds per-square-inch: gauge (psig). On May 10, 1990, the licensee identified that.the p~ower, supply t breakers to the "C" and "D" accumulator isolation valves were left closed following a surveillance tes This is contrary to TS 3. .?

which requires. each accumulator isolation . valve to.be.open- and power ~.

removed. This TS is applict.ble for Modes 1, 2,:and 3 with_. primary i system pressure greater than 1000 psig. The-licensee had performe an. isolation valve leak test on each of:the accumulator' check valves earlier in the day. This test was-conducted-in accordance with STS PE-019E, "RCS Isolation Check Valve Leak Test,"_-Revision This procedure required that the operator restore power to.the applicable '

isolation valve and then close the valve'to allow' leak 1 testing'of the check valve. . This was performed for each of the four check valve _

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The procedure required that each isolation' valve. be opened before; '

proceeding with the procedure; however,' no. requirements were established for opening the applicable breaker.

L An operator performing a procedure to seat the SI. check valves '

identified that the breakers for'the "C".andL"D", accumulator 1 i isolation valve operators wereTclosed. The. breakers were opened to meet the-TS requirements. Although the-breakers for;two.of the~-

l accumulator isolation valve operators were closed,_the accumulators would,have functioned if needed since' the ' valves were ope .,

The failure _to restore the breakers-to the open position'was .  !

identified by the inspectors as a potential violation of TS 3 51' as

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described above (482/9022-01). However, because.the licensee-identified the occurrence, immediately restored the breakers to their 1 proper lineup, reported the event, and took: prompt ccrrective action- ;

by revising the procedure to require opening of each +of the. breakers, no violation citation will be issued. LThe criteria specified Lin! -

'Section V.G.1 of the Enforcement Policy was mett, o On May 14, 1990, at 11:16 p.mf (CDT), during the performance.of the main turbine overspeed surveillance. test,.a high-high steam generator level was received. This resulted in an FWIS. Because of the'

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fesdwater isolation, a low-low steam generator. level was reached

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causing a reactor trip from 18-percent thermal power. All safet systems performed per this desig Earlier lon May 14, 1990,1the licensee had isolated the main steam feedwater preheating when a main.fe$ater pump was placed in service. When the main ~ turbine v '

? c'#line in-accordance with the procedure, the extraction fe 'ing.was alsr los 'Feedwater. temperature subsequen' .J from 400*F .o ..

approximately 150*F. This caused ' .a," then " swell" effect in the steam generators, which res;' .a high-high steam l generato j

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level, sWith the feedwater syste: .. ,d, the steam generator levels quickly. decreased to the la.-low level setpoint. - This

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resulted in the reactor power system actuation and initiation'of the .

auxiliary feedwater system. All safety . systems performed per thi 3 desig '

The' licensee had; experienced similar' problems in the past with controlling steam generator levels without adequate feedwater-heating.. Procedures for plant startup;were revised, however, STN-AC-007, " Turbine,0verspeed Trip Test," was not identified as? ,!

requiring: revision. 'Although STN-AC-007 did not provide caut,ons.on 1

- the loss of feedwater heating, the operators should-have been . . .

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cognizant of thejplant status and the effect removing equipment'fro i service would have on the. plan ,

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o i On May.17, the licensee found a failed gasket on the 'A" main feed 'l pump. seal water return orifice...The leak =was not isolable from the i main condenser. To repair.the leak, the licensee performed controlled power reduction to remove the main'; generator from the grid q and shut down the secondary side of the olant. 'At'-10:52 p.m.', the

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MSIVs were shut and at 10:58 a.m. the MS1V bypass valves were shu ,

Temperature was being controlled in the reactor by. modulating the 1 l steam generator atmospheric relief valves. At=11 p.m.' the "C" steam- '

j 3 generator atmospheric: relief valve failed open.and onecof three

, Loop 3 main steam line low-pressure bistable annuciators began to j

, come in and out. At:this time, the plant wasJn an uncontrolled j cooldown. - Operators promptly recognizedithe failed open atmospheric- i relief valve... Computer records show that actionLto shut'the other j three atmospheric reliefs took' place within 15 seconds and that they ,

were all' shut within 1 minute. While operators 1tried unsuccessfully 1

- to close the "C"4 atmospheric relief valve by: cycling:its'setpoint; j

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control in the control room, the shift. supervisor dispatched foU operators to shut the manually operated block valve. An'I&C i technician was also available in the control' room ~and was dispatched to attempt to shut the-relief. valve; During the event, level and ,

pressure.in the "C" steam. generator were decreasing,'as were RCS .

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Tavg, pressure, and pressurizer level. .At approximately 25 percent l ' level on the "C" steam generator' level narrow range , the shift- .. ,

supervisoruordered a manual- reactor trip. At 11:03 p.m., before thec j order'could' be carried out by the reactor operator, the reactor- -l

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tripped on "C" steam generator low-low level (two of four channels). i

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The atmospheric relief-valve'was still open-'and the uncontrolled  !

cooldown continued as operators,tried.to shut the block valve. -At- t 11:09 p.m. they reported to the control room that the block valve was .;

. shut as far as they could shut i The I&C technician =found that the'

atmospheric' relief valve I.to P'(current to pneumatic) converter had-

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failed, wnich provided an air signal of approximately.12 pounds per.-

square inch _-(PSI) to the. valve' positioner.- This-resulted in an approximate 75 percent open demand signal. The I&C. technician =

manually' failed the_ atmospheric relief valve closed by holding the-air relay linkage to provide a closed. signal. ,The atmospheric relief-valve responded and shut._ At:11:10 p.m. operators reported to'the '

control room that with steam'. flow: stoppeds they were able to shut the ,

block valve another sixEturns; When the technician' released the ':!

linkage,'the relief valve returned open.' However, the cooldown was '

stopped with the block valve shut and operators. returned the reactor: -l coolant system. temperature to normal for shutdown conditions.' s Minimum conditions during the event were; .

RCS-Tavg 529 F; a

RCS pressure 2230 psig; [

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Pressurizer. level _11 percent;_ i

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"C" steam generator-level.56 percent byLwide range, narrow range-level wa.; off scale low; and '

Incore thermocouples read as low:as-525 o On.May 19, 1990, the reactor tripped on:"A"' moisture separator reheater (MSR) high-high level . The: licensee'found that a high_ level ,

switch in the "A" MSR drain tank had failed. The switch was

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t replaced. The switch providedlan open' signal-for the drain tank dump

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valve to the-condenser and provided an-alarm: signal _for an alarm q annunciator, rThe licensee found that a 'similar switch oni"B"'and "D" -

MSR drain-tanks required adjustment. The licensee inspected the MSR-

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drain tank level control valve"and upstream check valve. That check- a valve and other similar check valves were'known.to the licensee to e occasionally stic The check valve was partially disassembled and *

the hinge and hinge pin.were inspected. No. problems-were'found by: ,

the licensee during those inspections. The licensee's posttrip  !

review did'not-fully address the root cause of the trip. . The i posttrip review-stat C that'the cause of the' trip was AF.LV-28 an ,

AF LV-29. The posttrip review did not~ discuss.the: failed switch or- y the results of their inspection of. the level control evalve: and j upstream check valve. The posttH) review is the documentation upon -

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which the-licensee bases a restart decision and it should contain l

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complete ract^cause ano corrective action. . pending completion of inspector review of the posttrip' reviews, this is an unresolved item (482/9022 4 2). 1

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The operators response to the reactor trips ensured that the reactor plant had responded'as designed and took appropriate measures to place the plant ,

into a stable configuration. The reactor trip actuation and first reactor ;

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-trip.resulted from both insufficient guidance in procedures and operators ,

not remaining completely cognizant of the plant teatus and the effect I their actions would not have on the reactor plan l-l The inspectors will-review the LERs:fortthefe events and'will report any~

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D findings in a~ subsequent inspection repor j

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i l Verification of Containment Integrity (1715

.The purpose of this inspection was to verify that:the licensee has established containment integrity prior to commencing startup of-the~

reacto R The inspectors participated in- the final walk down .of containment with the:

. licensee. The containment and-penetrations appeared'to be in proper < j condition:for startu From outside of containment, penetration-

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. conditions appeared-proper. -'The inspector witnessed the air lock local-leak rate test that was performed after final containment closure and it~ "

l appeared satisfactor Verification of penetration isolation valves.was performed by Region:IV_ j inspectors and was documented in NRC Inspection Report'50-482/90-1 ;

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10. Unresolved item j Unresolved items are matters about which more'information is. required in- ;

I order to ascertainLwhether they are acceptable,' items'of noncompliance, or -)

deviations. One unresolved item disclosed during' the -inspection is ;

discussed in-paragraph j 1 Exit Meetino (30703)

.Theinspectorsmetwitklicenseepersonnel:(denotedin: paragraph 1)on '

May 31, 1990. The-inspectors summarized the scope and findings offthe ;

l- inspection. The licensee.did not identi.fy as proprietary any:of the a informationlprovided to, or reviewed by, the inspector i l

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