IR 05000482/1985023

Insp Rept 50-482/85-23 on 850501-31.Violation Noted:Slide Pole for Containment Water Level Indicator LF-LI-9 Observed to Be Inoperable Due to Unauthorized Mod in Form of Tape Wrapped Around & Obstructing Level Indicator Slide Pole

ML20136J290
Person / Time
Site:	Wolf Creek
Issue date:	07/31/1985
From:	Bruce Bartlett, Bundy H, Cummins J, Martin L NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20136J220	List: IR 05000482/1985023 ML20136J215 ML20136J235
References
50-482-85-23, NUDOCS 8508200678
Download: ML20136J290 (18)

Text

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I APPENDIX B

!- US~ NUCLEAR REGULATORY COMMISSION

REGION IV

~ NRC~ Inspection' Report: -50-482/85-23 LP: NPF-32 Docket: 50-482 Licensee: Kansas Gas and Electric Company (KG&E)

Post Office Box 208 Wichita, Kansas 67201 Facility Name: Wolf Creek Generating Station (WCGS)

Inspection At: Wolf Creek Site, Coffey County, Burlington, Kansas Inspection Conducted: May 1 to 31, 1985 ps Inspectors- '/. , [//d J///4 p J.9 7Cummins, ahior Reactor Inspector, / Datb Operations (pars. 2, 3 4, 5, 6, 7, 8, 9, 11,.12, 13, 14, and 15)

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/ , V//df/5 Y /A6 B W Bartlett Resi~ dent Reactor Inspector, Operations

/Dat6 (pars. 2, 3, , 7, 8, 9, 11, 13, 14, and 15)

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. 7-3l-75 H.~F. Bundy, Resident React %f Inspector, Date Operations, (pars. 2, 3, 4, 6, 7, 9, 10, 11, 13, 14, and 15)

Approved: /

LV E. Martin, Chief,' Project Section A D //////4 patg ReactorProjegBranch2 8500200678 B5000942 PDR ADOCK 050

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Inspection Summary Inspection Conducted May 1 to 31, 1985 (Report 50-482/85-23)

Areas Inspected: Routine, unannounced inspection including licensee actions on previous inspection findings; independent inspection; engineered safety '

features system walkdown; initial criticality witnessing; followup ~ on I&E ,

headquarters request for an update of inspections related to dam safety; l operational safety verification; event followup; plant tours; plant status; L startup test witnessing; and Quality First program review. The inspection l involved 396 inspector-hours onsite by three NRC inspectors including 82 -

l inspector-hours onsite during offshift I Results: Within the 9 areas inspected, one violation was identified (failure to follow plant procedures, para. 15).

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I l DETAILS l

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Persons Contacted Principal Licensee Personnel l

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+ L. Koester, Vice President-Nuclear '

( + C. Mason, Director-Nuclear Operations

• H. K. Chernoff, Licensing

+J. Bailey, Director of Engineering and Technical Services  !

+G. Rathbun, Manager.of Licensing & Radiological Services l

+0. L. Maynard, Licensing Supervisor

+W. J. Rudolph, Manager of Quality Assurance, Site  !

+* G. Williams, Supt. of Regulatory, Quality, and Administrative l Services

• K. R. Peterson, Licensing

+ M. Grant, Director-Quality J. J. Johnson, Chief of Security

• C. J. Hoch, QA Technologist

+J. A. Zell, Operations Superintendent

+F. T. Rhodes, Plant Superintendent

• M. Lindsay, Quality Systems Supervisor f R. Hoyt, Emergency Plan Supervisor ,

G. Wedd, Environmental Management Supervisor i

+G. Boyer, Superintendent of Technical Support j

+B. Norton, Reactor Engineer Supervisor {

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NRC Representativies

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+ D. Martin, Regional Administrator, RIV .

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+ P. Denise, Director, Division of Reactor Safety and Projects, RIV

)' +L. E. Martin, Chief, Project Section A (RPB2), RIV

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+J. Youngblood, Chief, Licensing Branch 1, NRR  !

+*J. Cummins, Senior Resident Inspector, WCGS

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• B. L. Bartlett, Resident Inspector, WCGS

• H. F. Bundy, Resident Inspector, WCGS j

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• Denotes those personnel in attendance at the exit mcating held on

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+ Denotes those personnel in attendance at the management meeting held on

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May 30, 1985.

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The NRC inspectors also contacted other members of the licensee's staff i

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-4-2. Plant Status At the time of the NRC inspection, WCGS had completed the required low power (5%) testing. Key events that have taken place since KG&E was issued a low power operating license are listed below:

Date Event March 11, 1985 NRC issued a low power operating license to KG&E for WCGS March 12, 1985 Started initial core load March 17, 1985 Completed initial core load March 21, 1985 Reactor vessel head installed -

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Entered Mode 5 April 17, 1985 Entered Mode 4 April 26, 1985 Entered Mode 3 May 21, 1985 Entered Mode 2 May 22, 1985 Initial criticality May 30, 1985 Completed required low power tests 3. Licensee Actions on Previous Inspection Findings (Closed)SERItem(50-482/84-00-144): This item tracked valve and system lineups required to support plant startup. The NRC inspectors have verified completion of these lineup (Closed)SERItem(50-482/84-00-26): This item tracked a safety ava!uation report commitment for the training supersisor and two training instructors to obtain senior reactor operator licenses. These licenses have been issue (Closed) SER Item (50-482/84-00-44): Postaccident Sampling Syste This item was inspected and closed by a Region IV NRC inspector in NRC Inspection Report 50-482/85-21. This item was tracked and closed as Open Item 482/8404-0 (Closed) Open Items (482/8456-01 and 482/8508-06): These items tracked unresolved NRC comments on WCGS procedures. The NRC inspector reviewed resolutions of comments for the following procedures:

- STS CR-001, Shift Log for Mode 1 and 2

- STS IC-500B, Channel Calibration DT/T avg Instrumentation The NRC inspector verified that resolutions were adequat . _ .. .-_ -. -

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(Closed) Open. Item (482/8456-02): This item tracked a concern that sufficient guidance for marking a step in operating procedures "nct applicable (N/A)" was not provided. Administrative Procedure AOM 02-021, Rev. 6, was revised to provide additional guidanc (Closed) Open Item (482/8323-03): This item tracked various NRC concerns associated with the liquid radioactive waste systems including satisfactory completion of the preoperational tests. The licensee satisfactorily completed Tests SPE RE-001 and SPE RE-002 and the NRC inspector found the data acceptable to close this item. Other concerns associated with this item have been previously satisfied. License condition Attachment 1, Item 1.a is satisfie (Closed) Open Item (482/8509-03): This item tracked retesting and analysis of data for certain snubbers and springs for which data did not meet acceptance criteria or data was not obtained during initial testing. It is also discussed as license condition 1.e, but incorrectly referenced in Attachment 1 to the license as (50-482/8509-04). These components were retested per Power Ascension Test SU7-0015 and the data was evaluated per Engineering Evaluation Request EER-85-XX-3 Engineering acceptance of all data was documented in Engineering Disposition 0-E-980-XX which was approved on May 12, 1985. This item is closed and the license condition is satisfie (Closed) Open Items (482/8519-03 and 482/8323-04): These items tracked completion and satisfactory preoperational testing of the gaseous radwaste system. They involved Deficiency Deferral RD #61 which concerned inability to perform Section 7.8 of Test SU3-HA01 due to equipment problems with the hydrogen and oxygen analyzers in the purge

line from the volume control tank to the gaseous radwaste system. The

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licensee satisfactorily completed Test SPE K0T-006, " Gaseous Radwaste i System."

(Closed) Open Item (482/8511-15): This item references Deficiency Deferral RD #67 which documents numerous minor deficiencies noted during performance of the solid radwaste system preoperational test. These deficiencies were appropriately resolved and the resolutions were

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documented and approved.

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(Closed) Open Item (482/8508-08): This item tracked issuance of

, Surveillance Test Procedures STS 1C-830 and STS IC-840. They were l

issued on April 11 and 18, 1985, respectivel (Closed) Open Item (482/8511-23): This item references Deficiency

, Deferral RD #74 which documents failure of the radwaste decant level

! controller during preoperational testing. The controller was repaired and satisfactorily retested per Procedure SU4-HCO3, Rev. 1, Retest 1.

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-6-(Closed) Open Item (t32/8511-22) This item tracked a problem in obtaining satisfactory conctete solidification during waste resin solidification. It was also documented as Deficiency Deferral RD #7 Satisfactory solidification was demonstrated during performance of Preoperational Test SU4-HC01, Rev. 1, Retest 1. Similar Open Item 482/8323-06 has also been closed in NRC Inspection Report 50-482/85-2 (Closed) Open Items (482/8443-01 and 482/8511-20): These items tracked a requirement resulting from a test discrepancy in Preoperational Test SU3-AB03 to retest the main steam isolation valves to determine fast closure times under steam pressure. All valves were tested per Procedure STS AB-201, Rev. 4 on May 5, 1985, and satisfied acceptance criteri This was also documented as Deficiency Deferral RD #6 (Closed) Open Item (482/8511-12): This item tracked a test deficiency on two floor drain sump pumps for the containment instrument tunnel. It was also documented as Deficiency Deferral RD #37. The pumps were satisfactorily tested as required by engineering disposition to Work Request (WR) 02977-85 per WR 06929-8 (Closed) Open Item (482/8511-19): This item references Deficiency Deferral RD #29 which involves failure of data in Preoperational Test Procedure SU3-EC01 to satisfy acceptance criteri Contrary to test acceptance criteria, the fuel pool filters allowed some particulate of greater than 25 micron size to pass. The disposition to Engineering Evaluation Request 085-EC-25 recommends acceptance of the filters and test data "as-is" based on meeting plant chemistry and manufacturer's standards. The test data was from these tests. Acceptance "as-is" is documented in Test Discrepancy Report 04 to SU3-EC01. Data from additional tests that were performed supports this dispositio (Closed) Open Item (482/8511-11): This item references Deficiency Deferral RD #53 which involves malfunctioning Meters NKII-1, NKII-2, and NKII-6 in the 125 volt DC vital electrical system. All three meters were replaced per KG&E WR 90148-85 and calibrated per Procedure RNM-C-1301, Rev. (Closed) Open Item (482/8511-18): This item involved verifying the licensee had procedures in place to ensure that in the event of long term low power operation, the safety analysis was not exceeded. The NRC inspector verified that the needed char,ges were made to the appropriate operating instructions and surveillance procedure (Closed) Open Item (482/8459-06 and license condition, Attachment 1, Item 2.d): This item concerned emergency core cooling system pump flow with degraded bus frequency. The NRC inspector reviewed the following documents SFR-1-NE-55, EER-85-NE-01, and EER-85-NE-02 which verified acceptable pump operation:

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-7-(Closed) Open Items (482/8509-01 and 482/8511-09 and license condition Attachment 1, Item 1.d): This item concerned Deficiency Deferral RD #27 on auxiliary feedwater pump turbine vibration. The NRC inspector reviewed SFR-AL-107 and witnessed selected portions of the retes (Closed) Open Item (482/8455-04 and license condition Attachment 1, Item 1.c): This item concerned Deficiency Deferral RD #03 on retesting the power operated relief valves (PORVs) at normal operating pressur The NRC inspector observed all required retesting and reviewed the applicable documentatio (Closed) Open Item (482/8455-06): This item concerned completion of the reactor coolant system resistance temperature detector (RTO)/ core exit thermocouple cross calibration prior to initial criticalit This item was not a license condition. The NRC inspector reviewed documents SFR-1-BB-150, SFR-1-88-158, and TOR-00 The NRC inspector has no further questions on this item. In NRC Inspection Report 50-482/85-08, this item was inadvertently referenced as open item 50-482/8455-0 (Closed) Violation (482/8451-01): This violation involved the separation criteria between nonsafety conduits and safety cable trays and cables exiting cable trays specified in the Final Safety Analysis Report (FSAR).

The licensee performed an evaluation to justify their construction pract ces, and submitted a propnsed FSAR change to the NRC's Office of Nuclear Reactor Regulation (NRR) on January 14, 1985. This change was approved by NRR on February 14, 1985. This item was inadvertently closed as Open Item 50-482/8452-01 in NRC Inspection Report 50-482/85-1 (Closed) Unresolved Item No. 2B (482/84-51): Lack of procedural incorp-oration of RCI #1-1357-E. The subject RCI from Daniel International Corporation (DIC) to the A/E essentially requested the A/E's concurrence that references in DIC Procedure QCP-X-302 to A/E design drawings was sufficient criteria to preclude unacceptable spreading of Unistrut side wall The engineer agreed and thus, the procedure required no supple-mental criteria to achieve a satisfactory installatio In essence, the noted sidewall bowing is acceptable and is the result of installation of proper Unistrut hardware and proper tightening of attaching bolting. This item was inadvertently closed as Unresolved Item No. 2B (50-482/84-15) in NRC Inspection Report 50-482/85-1 (Closed) Open Items (482/8508-03 and 04): These items tracked inspector concerns on the manner in which certain valves were cycled during the performance of Preoperational Test SU3-0009 and retesting of a portion of the same test. The licensee has adequately addressed the NRC inspector's concerns and has completed the recuired retesting. This item was inadver-tently closed as 482/85-03 and 04 in Inspection Report 50-482/85-1 *

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-8-(Closed)OpenItem(482/8427-02): This cpen item tracked inspector review of reactor coolant pump (RCP) seal leakage data. This data was reviewed following the last RCP seal replacement and was found acceptable. It will continue to be monitored as part of the routine inspection program. This item was inadvertently closed as 482/8427-06 in NRC Inspection Report 50-482/85-1 . Independent Inspection During this inspection period, the licensee's response to IE Information Notice (IEN) 83-75, " Improper Control Rod Manipulation" and IEN 84-06,

" Steam Binding of Auxiliary Feedwater Pumps," was reviewe IEN 83-75: This notice described an event in which improper control rod manipulations were performed due to inadequate consnunications from and controls by Plant Management. The NRC inspector verified that the licensee has procedures which define the steps necessary for recovery from a mispositioned rod, that procedures are implemented for verifying rod position when one form of normal indication is lost and that training has been provided for operators in the proper movement of control rods, the consequences of improper movement, and the consequences of operating with a mispositioned ro IEN 84-06: This notice addressed an event pertaining to steam binding in the auxiliary feedwater pumps due to back leakage from the main feedwater system. The NRC inspector verified the discharge and suction piping of the auxiliary feedwater pumps was not hot, the licensee has a temperature element tied to the plant computer on the discharge piping which will alarm in the control room, the operators have received training in identifying air binding of pumps and periodic leakage tests of the check valves in the auxiliary feedwater system are performed in accordance with regulation _

No violations or deviations were identifie . Engineered Safety Features (ESF) System Walkdown The NRC inspectors verified the operability of the ESF systems by walking down selected accessible portions of the systems. The NRC inspectors verified that valves and electrical circuit breakers were in

the required position, power was available, and valves were locked where 3 required. The NRC inspectors also inspected system components for damage

! or other conditions that might degrade system performance. The ESF j systems listed belew were walked down during this inspection report

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i - Auxiliary Feedwater System

- Emergency Diesel Generators

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120 Volt Vital AC System No violations or deviations were identifie *

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-g-6. Public Meeting on State and Local Emergency Plans On May 21, 1985, the NRC inspectors attended a public meeting on the state and local emergency plan for the WCGS. The meeting ,as held in the Coffey County Courthouse in Burlington, Kansas. Arrangements for the meeting had been coordinated by Mr. Richard Leonard of the Federal Emergency Management Agency (FEMA). The purpose of the meeting was to acquaint the public with the content of the off site emergency plan and the people responsible for implementing it. Representatives from Kansas Gas and Electric Company, Coffey County, and the State of Kansas described the function that the group they represented performed in responding to an emergency at WCG Each of the representatives also described the facilities, equipment,~ and personnel that would be available to the/n for responding to an emergency at WCG . Initial Criticality Witnessing The licensee took the plant critical for the first time on May 22, 1985, at 7:47 a.m. CO The NRC resident inspectors observed the licensee's performance of initial criticality and subsequently observed selected parts of low power testing (5% power).

The NRC inspector ascertained conformance of the licensee to license and procedural requirements, observed the performance of the operating staff and ascertained the adequacy of test program records. During initial criticality, the NRC inspectors witnessed the performance of selected portions of the following licensee procedures:

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SU7-S011 - Initial Criticality and Low Power Test Sequence

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STS-RE-002 - Determination of Estimated Critical Position

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STS-BB-004 - RCS Leak Rate

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STS-IC-244 - Analog Channel Operability Test, NIS PR N44

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STS-IC-235 - Analog Channel Operability Test, IR N35 During initial criticality witnessing, the NRC inspectors verified that:

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Crew requirements were being met as defined in plant procedures, Technical Specifications (TS) and license condition The appropriate procedures were in use and properly reviewe On a sample basis that the prerequisites have been satisfie Inverse count rate ratio (ICRR) plots were properly calculated and verifie Technical support was available and consulted when require Boron concentration was determined by sampling and analysis at the frequency required by TS and procedure __

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Shift turnovers conformed to administrative procedure Adequate corrective action was taken for deficiencies encountere Logs contained accurate, pertinent, and up-to-date informatio NRC inspector observations noted during the performance of initial criticality are discussed below:

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The NRC inspector observed the sampling and analysis of the RCS by licensee chemists and verified the boron concentration was properly determined, recorded, and transmitted to the control room in a timely manner. The NRC inspector verified the proper procedure was available (CHM 02-050, Rev. 0) and was followe Using the licensee's procedure, the NRC ins,pector independently

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calculated an estimated critical concentration (ECC) of 1272 ppm approximately 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> prior to criticality. Approximately 30 minutes prior to criticality, the NRC inspector performed another ECC calculation with the results showing 1273 ppm. Both results agreed with the licensee's calculations. The reactor achieved criticality at a boron concentration of 1343 ppm. This concentration was outside the limits established by the test but was within +1% delta K/K. The licensee's reactor engineering group was requested to respond to NRC inspector questions concerning this difference between the calculated and actual baron concentration Their written response is quoted below:

"As required by Tech Spec 4.1.1.1.1.c an ECP was performed within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of criticality. Upon establishing critical conditions it was noted that predicted conditions were outside the +500 pcm band allowed by procedure. Actual conditions were in Tact greater than-500 pcm from predicted conditions. Per Step 2.3 of STS-RE-002, the following actiggs were taken: 1) the reactor was held stable and critical at 10 amps on Intermediate Range instrumentation; 2)

ORPI and step counters were checked to verify agreement between the two, 3) boron concentration which had been logged during dilution to criticality was examined to determine if any anomalies existed, and 4) the Core Design Report was examined to insure an error had not occurred in the calculation. A discussion was then held with the Shift Supervisor explaining the actions that had been take The recommendation was made to the Shift Supervisor to remain critical and continue testing based on the fact that additional physics data would be required to determine if the additional core reactivity was real and how it would affect core physics parameter A similiar discussion to that above is entered in Section 10 of SU7-5011."

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During subsequent low power physics testing, the NRC inspector observed the power range nuclear instrumentation reading 5% of full power at a time when the license only allowed the reactor to be at 5%. In discussions with the licensee, it was determined that the power range instrumentation is calibrated to indicate conservatively until higher power levels have been reached and it has been properly calibrated. The licensee's written response to NRC inspector questions is quoted below:

"To obtain flux map data requires the reactor to be at 3% 1 1%

power per procedur The procedure also states that Power Range

, instrumentation and core delta T be used for indication of core power. Prior to increasing power a discussion was held with Operations personnel discussing the power increase and that core delta T is the best indication of power at these power level A digital volt meter (DVM) was connected to the most conservative delta T signal to more accurately measure delta T. During the power increase the most conservative Power Range channel indicated approximately 5.3% power at which time power was reduced to less than 5% as indicated by the channel. It should be noted that the delta T signal indicated power of approximately 2%. Having the Power Range channel exceed 5% power is not considered to be a problem for the following reasons: 1) delta T is the best indication of core power at this time, 2) the Power Range instrumentation is conservatively set at 2-3 times below the normal expected full power current (this is evidenced by delta T at approximately 2% with Power Range at 5.3% - a factor of difference), and 3) the normal accuracy of Power Range instrumentation is approximately 1 2%."

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During the performance of low power physics testing the " worth" of certain control banks was required to be performed. During the measurement of control bank "C" (CBC), it was calculated that the reactor would not go critical under the test core configuration and boron concentration; however, as CBC was being pulled to its full out position, the core instrumentation indicated the reactor would go critical sooner than expected. The pulling of CBC was secured until the RCS boron concentration was increased. The licensee's written response to NRC inspector questions is quoted below:

" Tech Spec 4.10.1.2 required, in brief, that control rods not fully inserted, with less than the required Shutdown Margin, be demonstrated capable of trip insertion 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to reduction in Shutdown Margin. This surveillance would be required prior to measurement of all rods in minus one (ARI-1) rod wort With critical conditions established at SBA and SBB full out, the reactor was tripped. CBC contained the single rod to be withdrawn to 228 steps, thus CBC was to be pulled to 115 steps to demonstrate trip insertio Prior to this an evaluation had been made that o

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-12-criticality c old not be expected because SBA/SBB worth was approximately 1700 pcm and CBC worth at 115 steps was approximately 400 pc This was also discussed with Westinghouse startup personnel who agreed that criticality would not be expecte Pulling CBC to 115 steps was commenced and at approximately 80-90 steps a flux doubling alarm was received. This also resulted in addition of a small amount of 2000 ppm water to the RCS. At this point the Reactor Engineer, Westinghouse Engineer, Shift Supervisor, and Supervising Operator discussed subsequent action It was decided to approach CBC at 115 steps slowly in anticipation of a possible criticality. Again, the core design parameters were reviewed with the same results as previously determined. With CBC at 102 steps it was determined that the reactor was, or soon would be, critica The rods were inserted back into the core. Several discussions were held between Westinghouse, Reactor Engineering, and Operations as to the course of action to be followe It was decided that approximately 700 gallons of boric acid would be added to the RCS based on the observed conditions. After allowing for mixing, CBC was pulled to 115 steps and tripped back into the cor Subsequent discussions and evaluations have resulted in the following conclusions: 1) while fully aware that the rod worths used in evaluations did not meet actual conditions, the magnitude of the difference was much larger than expected, and 2' the core is more reactive than predicted by the Core Design Report (preliminary data reductions from testing indicate approximately 500-600 pcm).

As a followup to this situation, it has been requested that the KG&E Nuclear Fuels Group provide computer runs for the following cases: 1) the worth of CBC with all other rod banks inserted in the core and boron at the critical condition of SBA and SBB withdrawn, and 2) a criticality search on CBC with all other banks inserted and boron at the critical condition of SBA and SBB withdraw This situation and actions are documented in Section 10 of SU7-5011."

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Themoderatortemperaturecoefffcient(MTC)isrequiredbyTStobe less positive than 0 delta K/K/ F. During low testingtheMTCwasmeasuredtobe0deltaK/K/gowerphysics F as documented in Special Report 50-482/85-007:

"The moderator temperature coefficient for the all rods withdrawn, beginning of cycle life (BOL), bgt zero thergal power condition has been measured to be + 1.03 x 10 delta K/K/ F at a critical Boron concentration of 1353 ppm. This is more positive than the limit of Technical Specification 3.1.1.3 !

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-In accordance with Technical Specification 3.1.1.3 Action Statement, control rod withdrawal limits have been established and incorporated into tne appropriate plant operating procedures to assure that the moderator tempegature coefficient is maintained less positive than 0 delta K/K/ The predicted average core burnup necessary for restoring the positive moderator temperature coefficient to within its limit for the all rods withdrawn condition is 5400 MWD /MTU. This is based on calculations for the all rods withdrawn, hot zero thermal power, Xenon and Samarium free condition at a critical Boron concentration

.of 1215 ppm."

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During the approach to critical, the NRC inspector observed selected portions of STS-IC-244, Analog Channel Operability Test, NIS PR N44 Protection Set 4;" and STS-IC-235, " Analog Channel Operability Test IR N35." The personnel performing the surveillance were observed to follow plant procedures and requirement The NRC inspector observed the performance of STS-BB-004, "RCS Water Inventory Balance," verified all data as it was taken and following the procedure performed his own leak rate calculatio Except for the use of too many significant digits, which did not affect the final answer in any significant way, the NRC inspector had no comment During the performance of S07-S011, the NRC inspector independently verified the licensee was meeting 33 separate Technical Specifications and 13 prerequisites. The NRC inspectors had no comment ,

No violations or deviations were identifie . Followup on I&E Headquarters Request for an Update of Inspections Related to Dam Safety By memorandum dated March 14, 1985, the Director of the Division of Emergency Preparedness and Engineering Response of the Office of Inspection and Enforcement requested that Region IV provide information on inspections performed on the Wolf Creek Generating Station cooling water reservoir main and saddle dams. Information on inspections performed during the period from September 30, 1981, to Decenber 31, 1984, by the NRC, the dam owner, or contractors to the dam owner, was requested. The NRC inspectors toured the main and saddle dams, discussed the construction and maintenance of the dams with licensee personnel, and reviewed inspection reports related to the dams. The following information was provided by the NRC inspectors in response to the above request:

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-14-The Wolf Creek reservoir provides cooling water to the WCG The f

reservoir retaining structure consists of a main dam and five saddle dam KG&E owns and maintains the dam f l Key events related to the WCGS reservoir da j October 1980 Construction of reservoir completed i November 1980 Reservoir fill started July 1982 Full pool of 1087 feet MSL reached NRC inspections of the da Determined from a review of Region IV inspection reports !

issued from September 30, 1981, to December 31, 1984, that no Region IV inspections of the dam had been performed during that time interva The WCGS SER Section 2.5.6 discussed an inspection performed by the NRC staff. Minor seepage from the main dam was noticed during this inspection. NRC staff conclusions concerning this seepage were presented in WCGS SSER No. 5. The staff concluded, based on a study conducted by the applicant that:

1) the seepage was below estimated design seepage levels; 2) the seepage does not pose any threat to the safety of the ,

dam; and 3) the dam will remain functional during both static j and dynamic (operating basis earthquake) loading condition l The study conducted by the applicant included construction of I a wier for monitoring main dam seepage, j

' KG&E and contractor (Dames and Moore) inspections of the WCGS

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reservoi KG&E and/or Dames and Moore personnel have performed inspections of activities related to the construction, filling, and operations of

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the reservoi . Maintenance activities related to the reservoir da ,

So far, no items of major significance have been identified that would be detrimental to the functioning of the dam. Maintenance has been performed to repair minor erosion, clean drainage ditches, and clear brus ,

No violations or deviations were identified.

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-15-9. Operational Safety Verification

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The NRC inspectors verified that the facility is 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 and limiting conditions for operations, and reviewing facility records. The NRC inspectors, by observation and direct interview, verified the physical security plan was being implemented in accordance with the security pla NRC inspector observations noted during this inspection period are discussed below:

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On May 21, 1985, during a control room tour, the NRC inspector observed the handswitch to Valve PCV-456A (PORV) in the closed position. Atthetimeofthisobsegvation,theplantwasinMode 3, RCS cold leg temperature was 368 F and the block valve to PORV 456A was closed due to seat leakage through the valve. TS 3.4.4, Action A allows the block valve to a PORV to be closed indefinitely; however, TS 3.4.4, Action B states that if a PORV is inoperable due to causes other than excessive seat leakage, the licensee has 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to restore the PORV to operable status or perform the actions required by the action statement. The NRC inspector questioned the reactor operator, supervising operator, and shift supervisor as to why TS 3.4.4, Action B had not been entered. They stated that in their opinion having the handswitch in the closed position did not make the valve inoperable and thus Action B was not appropriate. The NRC inspector pointed out that, with the handswitch in the close position, the valve would not automatically open if it was called upon and this made the PORV inoperable. After some discussion, it was agreed that the handswitch in the close position made the valve inoperable and the reactor operator removed the handswitch from the close positio On May 21, 1985, during a control room tour, the NRC inspector observed the following event. Shortly after starting the spent fuel pumps in order to check the proper operation of the spent fuel pool cleanup system, Annunciator 47E (refueling water storage tank (RWST) level HI-LO) energized. After checking the level indicator, the operators determined the alarm was due to high level. By the time the pump was stopped approximately 8000 to 12,500 gallons of pure water had been added (3%). The supervising operator (50) took steps to determine what the addition of this amount of pure water had done to the baron concentration of the RWST. The original calculation using control room level indicators and conservatively assumed original RWST concentration indicated the RWST had dropped L

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-16-below the minimum allowed by T The 50 declared the RWST inopt rable and entered the appropriated TS action statement. The 50 then sent operators out to read the more accurate local level indicators, verify the RWST was not overflowing, and to get a more accurate pre-event RWST boron concentration from chemistry. While the operators were getting the above information, the 50 had calculated the amount of boric acid he needed to inject to the RWST in order to bring it back within TS and had dispatched an operator to initiate the injection. When the operators returned with the new more accurate numbers, a new calculation was performed and this one showed the RWST concentration was still above the TS minimu The S0 exited the TS action statement but continued the addition of boric acid in order to return the RWST to its previous concentratio The cause of the event was determined to be an open mini-flow valve to the RWST from the discharge of the pump. At the present time, the spent fuel pool (SPF) does not have any spent fuel. The water in the SPF pool does rot have any boron. When the SPF pump was started it was taking su; tion from the SFP and discharging back to the SFP after going throigh the cleanup system; however, with the mini-flow valve open, pure water was being pumped into the RWS During the last performance of SYS EC-200, a temporary procedure change (TPC) had been written to allow the mini-flow valve to be opened; however, the TPC did not call for the valve to be closed again and it was missed. As allowed by procedure when they were preparing to start the SFP pumps, a field verification was not performed, only a documentation review was performed and thus the out of position valve was misse In discussions with Plant Management concerning this event the NRC inspector was informed that operations personnel would be cautioned on the proper way to write a temporary procedure change. In addition, the licensee's write-up of this event was required reading for the operator No violations or deviations were identifie . Event Followup The NRC inspector discussed Wolf Creek Event (WCE) Report 85-69 with the licensee's Chief of Security. The event involved a contractor craftsman taping the strike open on a normally locked but nonvital doo Disciplinary action was taken againt the craftsman per a memorandum from the Plant Manager to all personnel. All plant personnel receive training in security prior to being badge No violations or deviations were identifie _ _ _ _ _ _ _ .-

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-17-11. Information Meeting with Local Officials On May 29, 1985, the WCGS NRC resident inspectors attended a meeting with local government officials and other citizens of the surrounding communitie The purpose of the meeting was to:

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Briefly describe the NRC's responsibilities.

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To introduce the resident inspectors and the cognizant NRC Section l Chief to the local public.

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Briefly describe the Public Document Room including its location.

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Provide telephone numbers of appropriate NRC contacts.

I The meeting was coordinated by the NRC Region IV State Liaison Officer

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who also presided over i . Management Meeting l On May 30,-1985, a meeting between Region IV and KG&E was held at the l WCGS sit At the meeting KG&E representatives apprised the Region IV representatives of the status of the WCGS plant and its readiness for a full power licens . Startup Test Witnessing I

During the inspection period, the NRC inspectors witnessed selected portions of startup testing activitie By discussions with startup test engineers, plant operations personnel, and other test support personnel; by periodic review of operations control room logs and

! completed parts of test procedures; and by observations of test activities,.the NRC inspectors verified that startup testing was conducted in accordance with requirements. The following are specific items that were verified by the NRC inspectors:

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Test was conducted in accordance with an approved procedure of the appropriate revisio Test equipment observed being used for testing had a current calibratio Startup engineers coordinated test activities with reactor plant operator Test results were within specificted tolerance of the test procedure or justification was documented for out of specification

' dat Selected parts of the following test procedures were witnessed by the NRC inspectors:

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SU7-SF03.3 - Hot Full Flow Rod Drop Test

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SU7-SF03.4 - Hot No Flow Rod Drop Test

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SU7-BB04 - RCS Flow Coastdown Measurement

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SU7-B805 - Pressurizer Continuous Spray Flow Setting

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SU7-SR04 - Incore Instrumentation Operability Test

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5U7-0016 - Biological Shield Testing

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SU7-SR01 - Incore Movable Detector Flux Mapping at Low Power

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The NRC inspectors discussed conduct of testing, test results, and test l

deficiencies with startup engineers and other cognizant licensee

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personnel during all phases of test performance. The NRC inspectors modified their inspection schedule as necessary to ensure all shifts involved in testing were observe No violations or deviations were identifie ,

14. Review of Quality First Files During this inspection period, the NRC inspectors participated in an NRC team inspection of the licensee's Quality First program. The details I and findings of this inspection will be reported in NRC Inspection Report

[. 50-482/85-28.

I 15. Plant Tours At various times during the course of the inspection period, the NRC l inspectors conducted general tours of the reactor building, auxiliary

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building, radwaste building, fuel handling building, control building, turbine building, and the secured area surrounding the buildings, t During the tours, the NRC inspector observed housekeeping practices,

, fire protection barriers and equipment, maintenance on equipment, and g

discussed various subjects with licensee personnel.

l NRC inspector findings during the tours are discussed below:

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During a tour of lower containment on May 24, 1985, with the plant in Mode 2 at 3% power, the NRC inspector observed an object secured k with strapping tape around the slide pole to LF-LI-9, " Containment Sump Water Level." The tape rendered the level indicator

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inoperable eliminating the redundancy in the Class 1E level l monitoring in the containment normal sump The NRC inspector L reported his observation to the shift supervisor who dispatched

, personnel to remove the tap This is a violation.

(50-482/8523-01)

l t 1 Exit Meetina l

h The NRC inspectors met with licensee personnel to discuss the scope and findings of this inspection on June 3, 1985. They also attended exit meetings conducted by other NRC inspectors.

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