IR 05000482/1993023

Insp Rept 50-482/93-23 on 930913-17.No Violations Noted. Major Areas inspected:10CFR50.59 Safety Evaluation Program & Implementation

ML20057G087
Person / Time
Site:	Wolf Creek
Issue date:	10/06/1993
From:	Westerman T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20057G078	List: IR 05000482/1993023 ML20057G077
References
50-482-93-23, NUDOCS 9310200156
Download: ML20057G087 (11)

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APPENDIX U.S. NUCLEAR REGULA10fiY COMMISSION

REGION IV

Inspection Report:

50-482/93-23 License: NPF-42 Licensee: Wolf Creek Nuclear Operating Corporation

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P.O. Box 411 Burlington, Kansas Facility Name: Wolf Creek Generating Station Inspection At:

Burlington, Kansas Inspection Conducted:

September 13-17, 1993 Inspectors:

W. M. McNeill, Reactor Inspector, Engineering Section i

Division of Reactor Safety Accompanying Personnel:

W. D. Reckley, Project Manager, Nuclear Reactor Regulation S. M. Wittenberg, Intern, Nuclear Reactor Regulation

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/r 6-GP Approved:

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Thomas F. Westerman, Chief, Engineering Section Date Division of Reactor Safety Inspection Summary l

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Areas Inspected; Routine, announced inspection of the 10 CFR 50.59 safety evaluation program and its implementation.

Results:

The licensee has established a well proceduralized program for e

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I performing safety evaluations in accordance with 10CFR50.59 (Section 2.1).

Strong points in the program included the availability of a computerized

word search of the safety analysis report text, a list of accidents in the training material, and the ongoing development of a computerized tool to assist in evaluations (Section 2.1).

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9310200156 931013 PDR ADOCK 05000482 G

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-2-Training was not required for qualification of personnel involved in the

10 CFR 50.59 process as qualification requirements can be satisfied by either training or experience at the supervisor's discretion.

No direct effects on the quality of the safety evaluations reviewed were observed (Section 2.1).

The recently revised training program was found to be thorough and

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complete, but attendance has been limited (Section 2.1).

Screenings and evaluations were observed to be well prepared. Two

examples were identified where further revision to the safety analysis report would have improved text detail (2.2 and 2.3).

Attachments:

Attachment 1 - Persons Contacted and Exit Meeting

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Attachment 2 - Documents Reviewed

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l DETAILS 1 PLANT STATUS During this inspection period the plant was operating at 100 percent power in Mode 1.

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2 SAFETY EVALUATION PROGRAM, 10 CFR 50.59 (37001)

2.1 Proaram The inspectors found that the licensee had established a well proceduralized program (KGP-1220) for reviewing changes, tests, and experiments for changes to the facilities as described by the safety analysis report. The program for safety evaluations included consideration of other licensing basis documents such as the security plan, Technical Specifications, orders, safety evaluation reports, and supplements. Other documents such as the Environmental Protection Program, Quality Assurance Program, Radiological Emergency Response Plan, Licensed Operator Requalification Program, Fire Protection Program, and NRC or INP0 commitments were also considered. However, KGP-1220 required they be considered only to the extent that the information in those documents was in the safety analysis report. This was a narrow definition of licensing basis documents, although when the process was implemented, the inspectors noted that a much broader definition was used.

For example, changes to the fire protection system beyond the description in the safety analysis report were found with safety evaluations.

The licensee considered the program to conform to the Nuclear Safety Analysis Center (NSAC) -125, " Guidelines for 10 CFR 50.59 Safety Evaluations."

NSAC-125 was a reference in the procedure. The training program provided a list of accident events. A computer program, "Zyfind," was capable of performing word searches of the safety analysis report text. A computerized artificial intelligence network, " Safety Review Advisor [ SARA]," is under development by Electrical Power Research Institute and Sargent & Lundy to provide a consistent and comprehensive guide to reviewers of about 1,700 prompting questions. The licensee plans to have " SARA" operational by the end of the year. The above items were considered strengths of the program.

10 CFR 50.59 evaluations were performed in a three-step process. The first step was a screening documented on Form KGF-II5. When changes, experiments, or tests could not be resolved at this step, a second step was performed.

This was called an unreviewed safety question determination and was documented on Form KGF-117. This step determined if the propos d activity involved an unreviewed safety question.

Each evaluation was to have a unique number.

If resolution could not be obtained by this evaluation, then the third step, called an unreviewed safety question, was performed.

The purpose of the third

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step was to assure that the change, experiment, or test was not performed i

without prior NRC approval. All screenings and evaluations were to be approved by qualified personnel. However, there were no requirements for

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personnel to receive licensee training as part of their qualification.

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-4-Qualification requirements of KGP-1220 could be satisfied either by training or based on experience at the supervisor's discretion. There were no provisions for requalification of individuals.

The inspectors, however, found no direct effects on the quality of the safety evaluations performed as a result of this qualification process.

The inspectors found that the licensee had recently revised the training program. The revised training program was found to be very thorough and

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complete. Attendance has been limited to date for this training.

2.2 Imolementation The inspectors reviewed the " Annual Evaluation Report for Wolf Creek Generating Station" for 1992. During 1992, the licensee had made 143 changes of which 57 were design changes, 31 were procedure changes, 21 were safety analysis report changes,15 were temporary modifications, and 19 were miscellaneous. Since the 1992 report, 91 evaluations had been performed consisting of 42 design changes, 22 procedure changes, 7 temporary modifications, and 20 miscellaneous items.

The inspectors attended an "onsite review committee" meeting. The inspectors verified that evaluations were being reviewed and approved by the committee as required by the Technical Specifications and licensee Procedure KGP-1220.

2.2.1 Evaluations The inspectors reviewed the evaluations identified in Attachment 2 for conformance to 10CFR50.59 and licensee Procedure KGP-1220. Additionally the inspectors reviewed the qualification of personnel and that all safety analysis report changes were properly initiated and documented. The evaluations reviewed included 13 design changes, 5 temporary modifications, I safety analysis report change, and 1 miscellaneous change. Overall the inspectors found that evaluations were well prepared and properly documented.

All personnel were qualified to the licensee's qualification requirements.

The following section is the inspectors' comments on one specific evaluation.

Evaluation 92-0075 - Adds Descriotion Of Hydrocen Purae System. May 21. 1992 A safety analysis change request had been issued because Procedure SYS GS-200,

" Containment Pressure Reduction Using the Hydrogen Purge Subsystem," had been q

revised to allow the use of the hydrogen purge subsystem to reduce containment atmospheric pressure. The hydrogen purge was to be used only when normal containment purge systems were out of service. The hydrogen purge had been used once in this manner.

The analysis indicated that during a loss-of-coolant accident, thyroid dose in the exclusion area would be increased by about 14 rem. This was because of the delay in isolating the penetration.

The evaluation concluded that the resulting dose consequences combined with the calculated dose contributions from containment and emergency core cooling system recirculation leakages remained within the guideline values of 10 CFR Part 100. The text for the safety analysis report is to be revised to add

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i I-5-Paragraph 6.2.5.2.2.4, a description of the hydrogen purge system used as a containment pressure reduction system.

The inspectors found this to be an example of where further revision to the safety analysis report could have improved text detail.

For completeness, the safety analysis report could be changed in Table 15.6-8 and Figure 6.3.4-1, Sheet 46 to reflect the additional 14 rem increase under LOCA conditions and changes in valve positions necessary for the purging mode. The licensee did '

not change Section 15 of the safety analysis since the hydrogen purge system is to be used only if the normal containment purge system is unavailable.

2.2.2 Screenings The inspectors reviewed the screenings listed in Attachment 2 for proper identification of evaluations requiring 10CFR 50.59 review. The inspectors

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also reviewed the qualification of personnel. The screenings reviewed included 17 design changes, 7 configuration change packages, and 3 temporary modifications. Overall the inspectors found that the screenings were proper

and well prepared.

All personnel met the licensee qualification requirements of KGP-1220.

The following is the inspectors comments on one specific screening.

q Plant Modification Reauest 04605 - Reactor Coolant Pump Motor Undervoltage Time Dela_y Reduction, April 9, 1993 The licensee prepared this plant modification request as the disposition of a Westinghouse technical bulletin related to reactor protection system (RPS)

response times.

The bulletin alerted licensees to a concern regarding the response time of the reactor coolant pump undervoltage trip function. A discrepancy had been identified regarding assumptions in the safety analyses report and Technical Specification surveillance requirements.

Neither the surveillance requirement nor the safety analyses provided an allowance for the decay of the bus voltage due to the effects of the coastdown of the large pump motors.

The safety analysis assumed a reactor trip on reactor coolant pump undervoltage simultaneously with the loss of electrical power to the bus and coastdown of the reactor coolant pumps. The surveillance procedures ensured that the performance of plant equipment was bounded by the safety analysis i

assumptions, but did not account for the expected voltage decay response of the electrical busses.

A similar problem identified by the bulletin involved a lack of proper accounting for the time delay associated with the mechanical release of the control rod drive grippers.

In response to the bulletin, the licensee calculated the delays associated with the decay of bus voltage and i

the release of the control rod drive grippers.

Procedures were revised to ensure that the total RPS response time was bounded by the safety analysis assumptions.

The inspectors considered the performance of the evaluations and revision of procedures to be a good response to the issues identified in the technical

t-6-bulletin. The evaluations and procedure changes were documented such that the maintenance of design bases and control of the plant configuration were adequately ensured.

The plant modification request received a 10 CFR 50.59 screening that determined that the safety analysis report was not affected by the proposed changes.

The inspectors noted that although the RPS response time was bounded by the safety analysis assumptions, including the delays discussed in the Westinghouse technical bulletin, the licensee had decided not, to include that degree of detail in the safety analysis report.

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ATTACHMENT 1 1 PERSONNEL CONTACTED 1.1 Licensee Personnel

• R. Benedict, Manager Quality Control D. Brown, System Engineer

• N. Carns, President and Chief Executive Officer

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C. Chaney, Engineering Specialist i

• M. Dingler, Manager Nuclear Plant Engineering

• R. Flannigan, Manager Nuclear Safety Engineering

• C. Fowl'r, Manager Maintenance and Modifications

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• R. Har ad, Health Physics D. Hooper, Engineering Specialist J. Hseu, Nuclear Analysis

• W. Lindsay, Manager Quality Assurance

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• R. Logesdon, Manager Chemistry A. Mah, Nuclear Safety Engineer i

P. Martin, Operations Supervisor

• 0. Maynard, Vice President Plant Operations

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• R. Meister, Engineering Specialist

• W. Norton, Manager Technical Support

• J. Pippin, Manager Integrated Plant Scheduling j

• C. Rich, Supervisor Electrical Maintenance

• T. Riley, Supervisor Regulatory Compliance

• F. Rhodes, Vice President Engineering

• R. Sims, Supervisor Results Engineering

• R. Smith, Manager Corporate Communications

• C. Sprout, Managc. System Engineering J. Stamm, Manager Plant Design Engineering

• J. Weeks, Manager Operations

• S. Wideman, Supervisor Licensing

• M. Williams, Manager Plant Support J. Zell, Manager of Design Bases 1.2 NRC Personnel

• G. Pick, Senior Resident Inspector In addition to the personnel listed above, the inspectors contacted other personnel during this inspection period.

• Denotes personnel that attended the exit meeting.

2 EXIT MEETING An exit meeting was conducted on September 17, 1993.

During this meeting, the inspectors reviewed the scope and findings of the report. The licensee did not identify as proprietary any information provided to, or reviewed by, the inspectors.

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ATTACHMENT 2 DOCUMENTS REVIEWED Procedures Procedure No. KGP-1131, " Permanent Modification and Configuration Control,"

Revision 9, March 30, 1993

Procedure No. KGP-1220, " Screening and Evaluating Changes Tests, Experiments,",

Revision 2, January 26, 1992

Procedure No. KGP-1225, " Revisions to the USAR," Revision 4, June 30, 1993

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Procedure No. KPN-C-314, " Plant Modification and Configuration Change Process," Revision 1, March 11, 1993 Procedure No. KP-C2203, " Regulatory Commitment Management System," Revision 0 Lesson Text, ES 16 354 00, "10 CFR 50.59 Evaluations," Revision 0 Lesson Text, ES 16 354 01, "KGP-1220 Screening Evaluations," Revision 0 Evaluations 92-0005, Temporary Modification No. 92-019-BG-0 - Freeze seal technique

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addressed by approved procedure MGM-M00C-12, January 14, 1992 92-0010, Plant Modification Request No. 03434 - Changes setpoints for the nuclear instrumentation system power range channel deviation alarm and adds time delays to the quadrant power tilt ratio alarms, January 29, 1992 92-0016, Plant Modification Request No. 03467 - Constructs four access vaults for the essential service water underground piping, March 18, 1992 92-0020, Temporary Shielding Request No. 92002 - Pressurizer spray valves, February 17, 1992 92-0075, USAR Change Request No.92-065, Adds description of hydrogen purge system, May 21, 1992 92-0103, Temporary Modification No. 92-023-CL-0 - Inject chemicals into service water system, July 22, 1992 92-0150, Plant Modification Request No. 04419 - Transfer from temporary to permanent installation of hose to the spent fuel pool, November 5,1992 92-0213, Temporary Modification No. 92-057-AL-0 - Provides temporary piping to the essential service water system, December 15, 1992 92-0216, Plant Modification Request No. 04380 - Evaluate component cooling water temperature below 60 degrees Fahrenheit, December 22, 1992

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t 93-0048, Plant Modification Request No. 04568 - Clarification in the safety analysis report of spring K-rates of main steam supply valves, March 2,1993 93-0056, Plant Modification Request No. 04550 - Modify braking horse power rates in the safety analysis report of selected 4160 volt motors on emergency diesel generators, March 9, 1993

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93-0058, Plant Modification Request No. 04575 - Change set pressure on nitrogen supply relief valves, March 11, 1993 93-0077, Plant Modification Request No. 04604 - Replacement of under frequency relays for the reactor coolant pumps, March 21, 1993

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93-0100, Plant Modification Request No. 04645 - Diesel starting air tank design pressure, April 28, 1993

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93-0104, Plant Modification Request No. 04629 - Rework of pressurizer safety valve discharge piping to correct misalignment, April 29, 1993 93-0105, Plant Modification Request No. 04642 - Removal of position indication and add of locks to 19 fire protection system valves, May 26, 1993 l

93-0123, Temporary Modification No. 93-029-RM-0 - Route drain from hydrazine l

analyzer to spare connection on steam generator blowdown sample recovery tank, June 2, 1993 l

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93-0125, Plant Modification Request No. 04667 - Correction of line

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identification, June 3, 1993 93-130, Plant Modification Request No. 04674 - Change in panel reference for

pressure indicators, June 4, 1993 93-0146, Temporary Modification No. 93-035-BG-0 - Install temporary blocking device between the valve actuator and the valve body to secure the valve in the full open position, July 16, 1993 Screenings Plant Modification Request No. 03977 - Flow transmitter flow range change, May 7.,

1993 Plant Modification Request No. 04526 - Equipment hatch auxiliary power, February 18, 1993 Plant Modification Request No. 04541 - Relocation of cable terminal locations, February 3,1993 Plant Mcdification Request No. 04542 - Replacement of the secondary fuses for the control power transformers, May 25, 1993

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l Plant Modification Request No. 04544 - Replacement of drip pans on containment coolers, February 18, 1993 i

Plant Modification Request No. 04546 - Modification of limit switch mounting f

on main steam isolation and feedwater isolation valves, February 18, 1993 l

Plant Modification Request No. 04547 - Attach a balancing weight to the j

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auxiliary hoist of the cask handling crane, February 1, 1993 Plant Modification Request No. 04548 - Change wiring locations in safeguards cabinets, February 18, 1993 i

Plant Modification Request No. 04549 - Change termination lugs, February 18, 1993 Plant Modification Request No. 04551 - Rotate motor operator, March 11, 1993 i

Plant Modification Request No. 04553 - Install ground water sealant can for j

heater drain pump, February 18, 1993

Plant Modification Request No. 04554 - Change part number of pressure switches on the auxiliary lube oil pumps, March 5, 1993 Plant Modification Request No. 04555 - Install new diaphragm in reactor make-up tank, February 24, 1993 Plant Modification Request No. 04558 - Remove limit switches from valves on

the heater drain pumps discharge, February 19, 1993 Plant Modification Request No. 04559 - Replace dipstick and tube assembly on fire pump diesel, February 24, 1993 Plant Modification Request No. 04605 - Reactor coolant pump motor undervoltage time delay reduction, April 9, 1993 Plant Modification Request No. 04647 - Change breaker setting in residual heat removal motor control center, April 29, 1993 Configuration Change Package No. 00130 - Site security system intercom loop, December 31, 1992 Configuration Change Package No. 00269 - Replacement of nuclear instrumentation system power ranger meter, March 1, 1993 Configuration Change Package No. 00282 - Replacement of screw and nut on feedback arm assembly for pressurizer spray valve positioned, March 2,1993 Configuration Change Package No. 00372 - Tube plugging of the containment I

coolers, March 27, 1993

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-4-Configuration Change Package No. 00439 - Safety injection pump runout, April 7, 1993

Configuration Change Package No. 00576 - Hammering noise during ALHV-9 motor operator valve test, May 25, 1993 Configuration Change Package No. 00789 - Optional stainless steel disk for

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Velan check valves, September 9,1993 (Revision 1)

Temporary Modification No. 93-037-BM-0 - Install temporary patch to stop water leak from valve body, July 27, 1993 Temporary Modification No. 93-028-EB-0 - Place a freeze seal on line EB-084-HB0-1 downstream of EBV108, May 17, 1993 Temporary Modification No. 93-030-BG-0 - Isolate BGll8 with a freeze seal placed on line BG-161, May 27, 1993

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