IR 05000482/1987001
| ML20211Q172 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 02/18/1987 |
| From: | Bruce Bartlett, Cummins J, Hunter D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20211Q138 | List: |
| References | |
| 50-482-87-01, 50-482-87-1, NUDOCS 8703030031 | |
| Download: ML20211Q172 (13) | |
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APPENDIX B U.S. NUCLEAR REGULATORY C0lHISSION
REGION IV
NRC Inspection Report:
50-482/87-01 License: NPF-42 Docket: 50-482 Licensee: WolfCreekNuclearOperatingCorporation(WCNOC)
Post Office Box 411 Burlington, Kansas 66839 Facility Name: Wolf Creek Generating Station (WCGS)
Inspection At: Wolf Creek Site, Coffey County, Burlington, Kansas Inspection Conducted: January 1-31, 1987 2 [lif[7?
Inspectors:
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- g d. E. Cummins, Senior Resident Inspector, Date Operations
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f B. L. Bartlett, Resident Reactor Inspector, Date Operations
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Approved:
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D. R. Hunter, Chief, Reactor Project Date Section B, Reactor Projects Branch
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I G703030031 870219
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{DR ADOCK 050004G2 t
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Inspection Summary Inspection Conducted January 1-31, 1987 (Report 50-482/87-01)
Areas Inspected: Routine, unannounced inspection including plant status; followup of previously identified NRC items; operational safety verification; monthly surveillance observation; review of LERs; 10 CFR Part 21 Report followup; onsite event followup; startup testing after refueling; and emergency preparedness drill.
Results: Within the nine areas inspected, one violation was identified (fire damper not operable as required by TS, paragraph 4). One unresolved item is identified in paragraph 10.
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DETAILS 1.
Persons Contacted -
Principal Licensee Personnel B. D. Withers, President, Wolf Creek Nuclear Operating Corporation
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J. A. Bailey, Vice President, Engineering and Technical Services
- F. T. Rhodes, Vice President, Nuclear Operations
- R. M. Grant, Vice President, Quality
- G. D. Boyer, Plant Manager 0. L. Maynard, Manager of Licensing C. M. Estes, Superintendent of Operations M. D. Rich, Superintendent of Maintenance M. G. Williams,' Superintendent of Regulatory, Quality, and Administrative Services W. J. Rudolph, QA Manager, WCGS
- A. A. Freitag, Nuclear Plant Engineering Manager, WCGS
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M. Nichols, Plant Support Superintendent-K. Peterson, Licensing
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- G. Pendergrass, Licensing
- W. M. Lindsay, Supervisor, Quality Systems
- C. J. Hoch, QA Technologist
- R. Flannigan, Supervisor of Compliance Engineering W. Norton, Supervisor of Reactor Engineering
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The NRC inspectors also contacted other members of the licensee's staff l
during the inspection period to discuss identified issues.
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- Denotes those personnel in attendance at the exit meeting held on February 3,1987.
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Plant Status i
Except during the periods described below, the plant operated in Mode 1 during this inspection period, o
On January 8,1987, a safety injection and reactor trip occurred
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while the plant was in Mode 1 at 100 percent power. The event was inadvertently initiated by technicians performing a surveillance on
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the main steam line pressure transmitters. The plant was returned to Mode 1 operation on January 12, 1987.
o On January 17, 1987, the plant was taken to Mode 2 and the turbine
shut down for turbine bearing balancing. The plant was returned to
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Mode 1 operation on January 18, 1987.
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o On January 20, 1987, a reactor trip due to high turbine vibration occurred. The plant was in Mode 1 at 100 percent power at the time of the trip. The licensee changed balancing weights on the turbine and on January 22, 1987, the plant was rett :ed to Mode 1.
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o On January 31, 1987, the plant was taken to Mode 2 and.the turbine was shut down to correct a vibration problem in the turbine. The plant was returned to Mode 1 operation on February 1,1987.
To correct the vibration problems discussed above, the licensee changed balancing weights, reworked a flexible coupling to the generator exciter, and realigned the generator exciter to the main-turbine shaft.
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3.
Followup On Previously Identified NRC Items
(Closed)OpenItem(482/8530-04): Cutler-Hammer Switches. This open item tracked activities related to a problem tne 11censee identified with
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Cutler-Hammer Type E-30 switches binding and potentially making safety-related components inoperable. Through testing and analysis the licensee and vendor determined.that the problem experienced at WCGS was
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probably due to the sharp pointed pushrod in this. type switch locking with a molded plastic part and binding the switch rather than allowing it to return to its normal position. The problem was corrected by replacing the sharp pointed pushrod with a rounded point pushrod in all safety-related switches of this type. The work was accomplished in accordance with Plant
Modification Request (PMR) 1316. Licensee Memorandum KWOLKWO 86-589 listed
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the work requests that performed the work delineated in PMR 1316. This item is closed.
(Closed)OpenItem(482/8621-01):
Instructor Certification Procedure not Referenced in Requalification Trainint Program Procedure. The licensee changed Administrative Procedure ADM 06-224, Revision 5, " Licensed Operator e
Requalification Training Program," to reference ADM 06-230. " Instructor
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Training, Qualification, Continuing Training, and Certification," as committed to. This item is closed.
(Closed)OpenItem(482/8621-02): Clarification of Training Tine Interval.
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The licensee changed Administrative Procedure ADM 06-224, Revision 4,
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" Licensed Operator Requalification Training Program," Section 6.1.4.4 to read " training cycle" rather than "requalification cycle," as committed.
This item is closed.
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4.
Operational Safety Verification The NRC inspectors verified that the facility was being operated safely and in conformance with regulatory requirements by direct observation of licensee facilities, tours of the facility, interviews and discussions with licensee personnel, independent verification of safety system status l
and limiting conditions for operations, and reviewing facility records.
l The NRC inspectors, by observation of randomly selected activities and
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interview of personnel verified that physical security, radiation l
protection, and fire protection activities were controlled.
By observing accessible components for correct valve position and electrical breaker position, and by observing control room indication, the
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.5 NRC inspectors confirmed the operability of the containment spray system and the auxiliary feedwater system. The NRC inspectors also visually inspected safety components for leakage, physical damage, and other impairments that could prevent them from performing their designed functions.
Selected NRC inspector observations are discussed below:
o During a plant tour on January 29, 1987, the NRC inspector observed that the flexible conduit to the electro-thermal link in Fire Damper GK-GD050 (located on the 2016 ft. level of the control building between Switchboard Room No. 4 and Eattery Room No. 4) was routed under the damper fire curtain and could possibly interfere with the operation of the fire curtain and prevent it from fully closing thereby rendering the damper inoperable. This observation was discussed with-licensee personnel and on February 3,1987, the licensee performed an operational test of the fire damper by electrically melting the fire curtain retaining thermal link. When tested, the electrical conduit
, jammed the fire curtain approximately 2 to 3 inches from the fully closed position. The test verified that the fire damper in the installed configuration observed on January 29, 1987, was inoperable.
TS 3.7.11 requires that this fire damper be operable at all times.
The failure to maintain Fire Damper GK GD050 operable is a violation (482/8701-01).
5.
Monthly Surveillance Observation The NRC inspectors observed selected portions of the performance of surveillance testing and/or reviewed completed surveillance test procedures to verify that surveillance. activities were performed in accordance with TS requirements and administrative procedures. The NRC inspectors considered the following elements while inspecting surveillance activities:
o Testing was being accomplished by qualified personnel in accordance with an approved procedure, o
The surveillance procedure confomed to TS requirements.
o Required test instrumentation was calibrated.
o Technical Specification limiting conditions for operation (LCO) were satisfied.
o Test data was accurate and complete. Where appropriate, the NRC
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inspectors performed independent calculations of selected test data to verify their accuracy.
o The performance of the surveillance procedure conformed to applicable administrative procedures.
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o The surveillance was performed within the required frequency and the~
test results met the required limits.
Surve111ances witnessed and/or reviewed by the NRC inspectors are listed below:
o STS IC-230, Revision 6, " Analog Channel Op Test of the Balance of Plant Emg. Safety Features Act. System," performed on January 15, 1987.
o STS NB-005, Revision 3, " Breaker Alignment Verification,"
(Section 5.1 only), performed on January 20, 1987.
o STS EC-1008, Revision 0, " Spent Fuel Pool Cooling Pump 'B' Inservice Pump Test," performed on January 9, 1987.
o STS PE-19A, Revision 0, " Safety Injection Test Header Fill and Vent,"
performed on January 9, 1987.
o STS PE-19C, Revision 0, "RCS Isolation Check Valve Leak Test From BIT to RCS Cold Legs," performed on January 9,1987.
o STS RE-004, Revision 6, " Shutdown Margin Determination," perfonned on January 8,1987, for four hours after reactor trip.
o STS RE-004, Revision 6 " Shutdown Margin Determination," performed on January 9,1987, for a hot, Xenon free reactor coolant system (RCS).
No violations or deviations were identified.
6.
Review of Licensee Event Reports (LER)
During this inspection period, the NRC inspectors performed followup on a Wolf Creek LER. The LER was reviewed to ensure:
o Corrective action stated in the report has been properly completed or work is in progress, o
Response to the event was adequate.
o Response to the event met license conditions, connitments, or other applicable regulatory requirements.
o The information contained in the report satisfied applicable reporting requirements.
o Generic issues were identified.
The LER discussed below was reviewed:
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i LER 482/86-34: Technical Specification Violation-Inoperable Containment Isolation Valve.
Inis LER reported an event where an exterior containment
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isolation valve (LF FV-96) in the floor and equipment drains' system failed a leak rate test, because the lock wiring of its hand wheel operator had
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been installed in such a way that it prevented the. valve from stroking fully closed. The licensee determined that the inner containment isolation valve for this penetration was operable during the time LF FV-96 would not
have fully closed. Since no other occurrences of this type have been identified, the licensee considers it to be an isolated occurrence. The LER was made required reading for appropriate licensee personnel.
No violations or deviations were identified.
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10 CFR Part 21 Report Followup The NRC inspector by review of documents, discussions with licensee personnel, and where applicable inspection of hardware verified that the 10 CFR Part 21 Reports discussed below had been reviewed and appropriately acted on by the licensee.
(Closed) Fuel Oil Filter Differential Pressure Switch for Diesels, Dated
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March 28, 1984, and Reported By Colt Industries.
This 10 CFR Part 21 Report stated a potential problem with a fuel oil filter differential pressure switch manufactured by United Electric Controls Company (UECC)
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Stock No. J27KB, Model No. 232. The switches were used in the fuel oil system of some Colt-Pielstick emergency diesel generators. The licensee determined that UECC switches Stock No. J27KB, Model No. 254 were used at WCGS rather than the Model No. 232, and therefore, the 10 CFR Part 21 Report did not apply to WCGS.
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(Closed) Deficient Gimpel Trip and Throttle (T&T) Valve For the Turbine
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Driven Auxiliary Feed Water Pump (TDAFWP) Dated February ZZ,1984, and Reported By Terry Corporation. This 10 CFR Part 21 Report stated that the 4-inch Gimpel T&T Valve (FC HV312) may not fully close with a condition of high inlet pressure and low steam flow. To correct this problem the vendor recomended that the closing spring be replaced with one of sufficient strength to close the valve. As recommended by the vendor, the licensee
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This work was documented on Wolf Creek Work Request (WR) y the vendor.
replaced the spring in Valve FC HV312 with one supplied b
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No. 13534-84.
(Closed) Unqualified Terminal Blocks For Hydraulic Actuators, Dated
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September 5,1985, and Reported By Anchor Darling Valve Company. This 10 CFR Part 21 Report reported an incident where unqualified Type NU 2
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terminal blocks were supplied to the licensee by the vendor. The licensee subsequently installed the unqualified terminal blocks in safety-related main steam and feedwater isolation valves. The licensee identified the
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problem, removed the unqualified terminal blocks from safety-related spare parts, and replaced the terminal blocks in the main steam and feedwater isolation valves with qualified terminal blocks. This item was previously discussed in NRC Inspection Report 50-482/85-30, paragraph 13.a.
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Emergency Preparedness Exercise On January 28, 1987, the WCGS NRC resident inspectors and the NRC resident inspector from Callaway participated in the licensee's annual emergency preparedness exercise. The WCGS and the Callaway NRC resident inspectors back each other up in the event of an emergency.
The county and state also participated in the exercise and the NRC Region IV Incident Response Center was activated.
Players performed the activities during the exercise scenario that they would perfonn during an actual emergency at WCGS.
The purpose of this exercise was to perform the annual NRC and Federal Emergency Management Agency (FEMA) evaluation of the capability of the licensee, the county, and the state to respond during a radiological energency at WCGS. Onsite activities were evaluated by the NRC and offsite (county and state) activities were evaluated by FEMA.
Region IV NRC inspectors evaluated licensee activities during the exercise and their observations and findings will be reported in NRC Inspection Report 50-482/87-04.
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Onsite Event Followup The NRC inspector performed onsite followup of nonemergency events that occurred during this report period. The NRC inspector (when available)
observed control room personnel response, observed instrumentation indicators of reactor plant parameters, reviewed logs and computer printouts, and discussed the event with cognizant personnel. The NRC inspector verified the licensee had responded to the event in accordance with procedures and had notified the NRC and other agencies as required in a timely fashion.
Engineered safety feature actuations that occurred during the report period are listed in the table below. Where applicable, the NRC inspector will review the LER for each of these events and will report any findings in subsequent NRC inspection reports.
Date Event Plant Status Cause 1/08/87 Reactor Trip /
Mode 1 Personnel error in Safety Injection 100 percent isolating wrong steamline piessure transmitter 1/09/87 Reactor Trip /
Mode 3 I&C surveillance FWIS*
1/17/87 AFAS*
Mode 2 Breaking condenser vacuum to reduce turbine speed 1/20/87 Reactor Trip Mode 1 Hi-Hi turbine vibration 100 percent
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4 1/21/87 AFAS*
Mode 2 Breaking condenser vacuum to reduce turbine speed
- FWIS - Feedwater Isolation Signal
- AFAS - Auxiliary Feedwater Actuation Signal Selected NRC inspector observations are discussed below:
The NRC inspector performed onsite followup of the reactor trip / safety injection which occurred on January 8,1987, at 6:18 The event was caused when an instrumentation and control-(I&C) p.m. CST.
technician who was calibrating the main steam line pressure transmitters isolated the wrong transmitter. With one pressure transmitter in bypass per the surveillance procedure and another accidentally valved out of the service, the required 2 out of 3 logic for a low steamline pressure safety injection (SI) was completed. As required the licensee declared a notification of unusual
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event (NOVE) which was terminated at 6:50 p.m. CST.= All systems perfonned as designed, except that one of the two power operated relief valves (PORV) position indicator lights indicated that the valve had stuck open, and the block valve' breakers tripped on instantaneous overcurrent. The block valve breakers were reset and the block valve for the indicated open PORY was closed. The PORY indicator was repaired prior to returning the
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unit to service. The licensee, based on engineering evaluation, raised the block valve circuit breaker instantaneous trip settings to prevent a
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reoccurrence of them tripping.
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10. Startup Testing After Refueling The NRC inspector reviewed data from core physics tests associated with the first refueling to verify the results met acceptance criteria and that any identified deficiencies were resolved in a timely manner.
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Surveillance of Core Power Distribution Limits The NRC iqspector verified that the TS for heat flux hot channel factor (F (Z)), the combingtion of RCS flow rate and nuclear enthalpy rise hot channel factor (F delta H), quadrant power tilt ratio
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j (QPTR), and axial flux difference (delta I) had been complied with I
by:
(1) Choosing at random a flux map performed on December 29, 1986, at 100 percent power, after it had been processed by the data analysis code "Incore," and verifying the appropriate control rod positions, core power level, burnup and other required data
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were properly input and that all incore detectors were properly normalized.
(2) Examining the printout in Item 1 above and verifying the predicted versus measured reaction rates were within expected l
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limits, the hot channel factors were within TS limits and.
reflected the ' applicable engineering tolerances and nuclear
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penalties.
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(3)' Verifying the,QPTR based on reaction rateA fr " 5Y""'tric thimbles and, the QPTR based on assetly if delta H from all;
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assemblies were in close agreesnnt.
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(4) Verifying the computer code "Incbre" used to. process the inure data, which is on a tinu share, mhinframe, offsite computerAs -
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controlled by.the licensee's vendt.r quality assurance progem
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and is periodically reverified through the use o,f duimy data.
(5) Verifying that Delta'I and QPTR lialits had been observed ands properly recorded.,
(6) Reviewing the surveillance test pIocedures listed below to verify that surveillan'ce activities were perfonned in accordance with 15 requirements and/or were technically adequate:
o STS CR-001, Revision 5, " Shift Log for Modes 1, ;2, fad 3,"
perfonned on September 2 and Octouer C,1986.
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o STS IC-504A,; Revision 2, " Reactor Coolant 410w-Full' Flow
. Calibration,"' pe'eformed on September 15, 1986.
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,o i STS 1C-504B, Revision 3. " Reactor Coolant Flow Transmitter Calibration," performed on Septenbar 15, 1986.
O STS IC-504C, Revision 2, " Channel Calibration of 7300
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Process Reactor Coolant Flow," performed on
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October 18, 1986.
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'STS MT-054, Revision 1, "Feedwater Venturi Inspection,"
performed on November 18, 1986.
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o STS RE-009, Revision 1, " Heat Flux Hot Channel Factor,
Measurement," performed on September 5 and October 1,;1986.
o STS RE-010, Revision 1, "RCS R Calculation," perfonned on
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September 5 and October l',' 1986.
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o STS RE-012 Revision 1, QP,TR Detertainat. ion," performed on December 23 and 30,1F86. '
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o STS SF-002, Revisfori 2,'"Cdre Axisl Flux Difference,"
performed on October 7, 8, and 11,1986.
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STS WL-001, Revision 7, " Weekly Surveillance Logs,"
performed on December 14, 21., 28, 1986, January 4 and 10;
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o RXE 03-001, Revision 0, "Incore Data Reduction and
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Analyyis."
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Core Therr.a1 Power Evaluation lg
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The NRC inspec1.6* verified that the licensee's procedure for the calculation of c.bre thennal power is technically adequate and
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perfonned in accordance with plant TS by:
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In' dependently calculating the core thermal power using s
licensee procedure STS SE-002, Revision 0, " Manual o
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Calculation of Reactor Thermal Power," on December 30, 1986, with the following results:
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NRC Inspector Calculation = 98.48 percent
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Computer Calculation
= 98.81 percent Nuclear Instrument Reading = 99.0 percent
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('2). Reviewing STS SE-002, Revision 0, performed on December 21, s
23, 27, and 29, 1986; on December 26, 1986, at 3:00 a.m.,
4:40 p.m., and 11:00 p.m. and verifying the data was reasonable, consistent with previous data, and properly recorded. The 3'
' calculations were correct and if required the power range nuclear instruments were adjusted to agree with the results of the heat balance.
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(3) Verifying that the instruments used to obtain data were
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within calibration due dates.
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(4) Verifying that the surveillance procedures listed in (2)
above were reviewed, approved, and documented in accordance
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with administrative procedures.
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Determiration of Reactor Shutdown Margin
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The NRC' inspector verified that adequate shutdown margin was
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niaintained by:
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(1) Hanually perfonning STS RE-004, Revision 6, " Shutdown Margin Determination," on January 9,1987, and verifying the licensee's computer determined shutdown margin agreed with the NRC inspector's calculation and that the procedure appeared to be technically adequate.
(2) Reviewing the completed surveillances listed below to verify they had been perfonned in accordance with TS.
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o STS CR-001, Revision 5, " Shift Logs for Modes 1, 2, and 3," performed daily from December 18-31, 1986, and
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o STS RE-002, Revision 5, "Detennination of Estimated Critical Position," perfonned twice on December 18, and once on December 20, 1986.
o STS RE-003, Revision 1, " Shutdown Margin By Minimum Bank Height," performed on December 20, 1986.
o STS RE-004, Revision 6. " Shutdown Margin Determination," performed on December 17 and 20,1986.
(3) Reviewing STS RE-005, Revision 1, " Core' Reactivity Balance,"
performed on September 2 and October 3,1986, to verify that the licensee's procedure for demonstrating agreement
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between the overall core reactivity balance and the
,j predicted values was performed at the frequency required by TS, and was made using the parameters required by TS.
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(4) Reviewing WCRX-04, Revision 2, " Control Room Operating Curves and Tables Cycle 2," and verifying the operating curves in the control room were current and up-to-date.
WCRX-04 is controlled by Administrative Procedure ADM 05-500, Revision 1, " Control Room Operating Curves and Tables Reference Manual-Formatting and Revision Guidelines." ADM 05-500 requires that changes to WCRX-04 be approved by the submitter, an independent reviewer and the reactor engineering supervisor.
If WCRX-04 is a minor procedure (as defined by TS) then it would be covered under TS 6.8.2.b which requires minor procedures, prior to initial use, to be approved by the plant safety review comittee (PSRC) or a subcommittee thereof and TS 6.8.3.b
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which requires all temporary or pennanent changes to minor
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procedures under the jurisdiction of reactor engineering to
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be approved by a cognizant group leader and subsequently reviewed and approved by the appropriate PSRC subcomittee.
Discussions with licensee management revealed that they believed that all information going into WCRX-04 and all
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revisions to WCRX-04 were made through PSRC reviewed and approved procedures and thus WCRX-04 was merely a l
controlled summary of data and did not require specific l
PSRC approval.
Pending the results of the licensee's verification that the above statement is fully accurate, this will remain an unresolved item (482/8701-02).
11. Unresolved Item i
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Unresolved items are matters about which more information is required in order to ascertain whether they are acceptable items, items of
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noncompliance, or deviations. An unresolved item disclosed during the l
inspection is discussed in paragraph 10.
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Exit Meeting The NRC inspectors met with licensee personnel to discuss the scope and findings of this inspection on February 3,1987. The NRC inspectors also attended entrance / exit meetings of the NRC region-based inspectors identified below:
Inspection Lead Area Inspection Period Inspector
' Inspected Report No.
01/12-16/87 D. Hunnicutt Document Control /
Startup Data Review 87-02 01/26-30/87 C Hackney Emergency Prepared-87-04 ness Drill
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