IR 05000285/1990030

Insp Rept 50-285/90-30 on 900513-0616.No Violations Noted. Major Areas Inspected:Monthly Maint,Surveillance,Security, Complex Surveillance Observation,Review of Licensee Outage Plans & Verification of Containment Integrity

ML20055J154
Person / Time
Site:	Fort Calhoun
Issue date:	07/24/1990
From:	Constable G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20055J147	List: IR 05000285/1990030 ML20055J146 ML20055J150
References
50-285-90-30, NUDOCS 9008010132
Download: ML20055J154 (20)

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.V,S, ! NUCLEAR REGULATORY; COMMISSION

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f NRC< Inspection. Report: '50-285/90-30., 9 License:

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Docket:

50-285.

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Licensee:

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, Omaha Public Power District _(OPPD)

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444: South 16th Street Mall',

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_ Omaha, Nebraska 68102,2247

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7Fac151ty Name:

FortCalhoun<Stafin[FCS)

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41nspeclion At:

FCS, Blair, Nebraska

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' Inspection Conductsd:.May13tbroughJune16,1990-

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" nspectors:

P'. Harrell,-Senior Resident Inspec' tor

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T. Reis, Resident Inspector >

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Approved:

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Diyision of Reactor Projects

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' Inspection Summary "

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-i Insoeciion Cdnducted May 13-throuah June 16-1990'2(Report 50-285/90-30Y

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Areas Inspected:

Routine, unannounced inspection of previously identified'

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items;licenseeeventreportfollowup; operational?safetyverification;monthlys

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maintenance, surveillance, security, and radiological protection' observations;

onsite followup'of events; complex surveillance observation;Treview of licensee

outage plans;' plant startup from refueling; and verification of containment.

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Resultsi,Three violations were identified'as described below:

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Inadequate postmaintenance' testing wakperformed after replacement of a I

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l solenoid-operated valve in the safety-related instrument air accumulator

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ass'embly for Valve YCV-1045A (steam supply valve for the turbine-driven'

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auxiliary feedwater pump). 'The inadequate tesi,ing resulted in

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Valve YCV-1045A not being able to' meet its design-basis requirements for l

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approximately 3 weeks (par'agraph"7, AppendixiB).

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ENCLOSURE CONTAINS o

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S!JE80ARDS INFORMAtl0N,

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UPON SEPARAil0N THl3

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PACC 10 DEColffR0ulD

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a An inadequate radiological survey in the con'tainment" lower cavity resulted

in higher than anticipated personnel exposure (paragraph 9, Appendix B).

However, in accordance with Section V.G.1 of the NRC's Enforcement policy, this violation is not being cited.

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An ' inadequate security vital area barrier was identified (paragraph 1,

Appendix C).

No response to this violation is required.

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The licensee identified an instance where an individual violated radiological work permit instructions.. This event will be reviewed by RIV radiological protection specialists during a future inspection and is being tracked as an unresolved item (285/9030-02),

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i Itoappearbd that the" licensee may have f ailed to comply with the procedural

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requirements for mixing boric acid. This issue is being tracked as an unresolved item (285/9030-03) and will.be addressed subsequent to an, internal

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licensee investigation, scheduled to be completed in the near future.

Within,the remaining areas, it appeared the licensees actions met the appropriate regulatory requirements.

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SAfi

..;3 INFORMAil0N

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UP0h MPARATl0N TNis

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-3 PACC 10 DCC0hTR0utD

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DETAILS

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Persons Contacted OPPD

• R. Andrews, Division Manager, Nuclear Services

• J. Bobbt., Supervisor, Radiation Protection

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• J. Chase, Manager, Nuclear Licensing and' Industry Affairs M. Core, Supervisor, Maintenance

• S. Gambhir, Division Manager, Production Engineering

• J. Gasper, Manager, Training

• R. Jaworski, Manager, Station Engineering

'J. Kecy, Supervisor, Systems Engineering

• L. Kusek, Manager, Nuclear Safety Review Group

• D. Lovett, Supervisor, Radiological Protection Operations

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• 0. Matthews, Supervisor, Station Licensing

• W. Orr, Manager, Quality Assurance and Quality Control

• R. Patel, Nuclear Safety Review Group Specialist

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• T. Patterson, Assistant Manager, Fort Calhoun Station

• G. Peterson, Manager, Fort Calhoun Station

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• R. Phelps, Manager, Design Engineering

• J. Sefick, Manager, Security Services

• R. Short, Supervisor, Special Services Engineering.

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• C. Simmons, Station Licensing Engineer

"J. Tills, Assistant Manager, Fort Calhoun Station D. Trausch, Supervisor, Operations-

• Denotes attendance at the monthly exit interview.

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The inspectors also contacted other plant personnel.

2.

Plant Status At the beginning of this inspection period, FCS was in 'the latter stages of its twelfth refueling outage.

The plant exited Mode 5 (refueling shutdown) on May 20, 1990, and entered Mode 3 (hot shutdown) on May 22,

1990.- Criticality was achieved on May 25, 1990.

Low power physics testing was performed May 25-27, 1990, without incident.

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The ger,erator was synchronized to the grid on May 29, 1990. The plant remained at 30 percent power from May 29-31, 1990, to perform I

a boric-acid soak of the steam generators.

On May 31, 1990, power was briefly reduced to 10 percent to facilitate repair of a leaky fitting in the main turbine electrohydraulic control system and from May 31 to

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June 4, 1990, boric acid soaking resumed.

Power ascension began on June 4,1990, and 98 percent power was achieved on June 7, 1990. A rapid power reduction to 77 percent was performed on

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SMEGUARDS INFORMAil0N

UPON SEPARAil0N THl3 PACC 13 DG00mt0LLID

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June 13, 1990, due to an inadvertent lifting of-a relief valve on the discharge of a heater drain pump.

The inspection period concluded with the plant at 94 percent power for moderator temperature coefficient testing.

High vibration was exhibitec auring startup on the Number 8 bearing of the main turbine generator.

The licensee plans to reduce power to

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approximately 10 percent in the near future to examine and rebalance the p

bearing.

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Review of Previously Identified Items (92701 and 92702)

a.

(Closed) Open Item 285/8846-04:

Repair of Seals on Gammametrics Cable Connectors.

This item 'avolved leaking seals on the connectors installed in

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containment for the Gammametrics cables.

The cables are used to

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provide indication of wide-range nuclear power to the control room.

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The licensee committed to repair the connector seals on all four cables.

Subsequent to this commitment, the licensee, in a letter dated February 9, 1990, *,tated that two of the four cables would be

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replaced in lieu of repairing the connector seals. The licensee j

replaced two cables in accordance with the instructions provided by l

Maintenance Work Order (MWO) 894427 The Office of Nuclear Reactor Regulation (NRR) reviewed the'

licensee's decision to replace two cables instead of repairing all four cables.

NRR stated that the action was satisfactory since the environmental qualification of two of four channels met the requirements of Regulatory Caide 1.97.

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(Closed) Violation 285/88201-13:

Failure to Provide Postmaintenance Test Instructions.

This violation involved a specific instance of failing to provide postmaintenance testing instructions.

Specifically, for preventive maintenance of Pressurizer Relief Iso 4 tion Valve HCV-150, performed

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in accordance with Proceoure PM-EE-4.0, "7700 Line Motor Control Centers," no postmaintenance testing instructions were provided.

Although this was a single finding, it appeared to be indicative of weaknesses in the licensee's conduct of maintenance.

In response to this violation and other maintenance weaknesses identified by the Operational Safety Team Inspection, the licensee revised Procedures 50-M-100, " Conduct of Maintenance"; S0-M-101,

" Maintenance Work Control"; and S0-M-102, "Postmaintenance fest."

Since the implementation of the revised procedures, the NRC has performed two inspections in the area of maintenance work control as

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[NCLOSURE CONTAINS SAFEGUARDS INFORIAAfl0N

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UPON SEPARAfl0N TH18 5-Pact IS DEC0f0 ROLLED documented in NRC Inspection Reports 50-285/89-01 and 50-285/90-04.

No violations of the failure to provide postmaintenance test instructions were identified.

In NRC Inspection Report 50-285/90-04 e

it was concluded that the licensee's program for postmaintenance testing appeared to meet the appropriate requirements. A related violation concerning the adequacy of postmaintenance test instructions is discussed in paragraph 7 of this report.

For the specific violation addressed, the inspector verified that Procedure EM-PM-EX-1100, "480 Volt Motor Control Center Maintenance,"

which supersedes Procedure PM-EE-4,0, contained the appropriate postmaintenance testing instructions.

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(Clos,.') Violation F.2 (OSTI Unresolved Item 285/88201-24):

Inadequate Corrective. Action Program.

This violation was related to the licensee's failure to establish a corrective action program that provided the appropriate requirements to ensure that conditions adverse to quality were promptly identified and resolved.

To address this violation, the licensee completed the following actions:

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Revised Procedure QAM-20, " Control of Internal Deficiencies and

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Corrective Action," to specify the method to be used for i

documentation and evaluation of deficiencies.

• Revised Section 10.4, " Deficiency Control and Corrective Action," of the quality assurance plan to define a significant deficiency and to specify the level of management review.that will be performed for this class of deficiencies.

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(Closed) Open Item 285/8909-02:

Seismic Qualification of Valve YCV-1045B.

This item was related to Valve YCV-1045B not being seismically qualified by the licensee.

In the licensee's response to NRC Inspection Report 50-285/85-22, the licensee stated that a calculation had been completed to verify that the as-built conditions of the restraints were adequate for seismic qualification of the valve.

The licensee was addressing the generic aspects of.the seismic qualification of small-bore piping and valves by implementation of the actions specified in Generic Letter 87-02, " Verification of Seismic Adequacy of Plant Equipment (VSI A-46)."

The generic aspects of this item will be reviewed during a future inspection performed to verify that the licensee has taken actions to meet the requirements

of Generic Letter 87-02.

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SAFEGUARDS INFORIAAtl0N UPON SEPARAfl0N THit '

PAEls DE3NmBAD-6-

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(0 pen) Open Item 285/8922-08:

Replacement of Nonconforming Breakers in Safety-Related Inverters.

NRC Bulletin 88-10 " Nonconforming Molded-Case Circuit Breakers," was issued to alert the licensee of the potential existence of inadequate breakers. As a result of the licensee's evaluation, nonconforming breakers were identified as being installed in the safety-related inverters.

The licensee committed to replacing the breakers during the 1990 refueling outage; however, due to supplier problems, new breakers were not available.

In a letter dated April 11, 1990, the licensee notified NRR that the breakers would not be replaced until the 1991 outage.

NRR concurred with the licensee's proposal.

This item will be resolved during a future inspection that will address the specific and generic concerns with nonconforming breakers when the licensee's actions taken in response to NRC Bulletin 88-10 l

are reviewed, i

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(Closed) Violation 285/8938-03:

Failure to Meet the Technical i

Specification (TS) Limiting Condition for Operation (LCO) for Two Inoperable Reactor Protection System (RPS) Channels.

This item was related to the licensee's failure to reduce reactor power to less than 70 percent when two RPS channels were declared inoperable.

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In a letter dated October 31, 1989, the licensee addressed the actions that would be taken in response to this violation, as discussed below:

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Revise procedures to include a requirement that senior plant

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management be provided timely notification of safety-significant i

occurrences.

The licensee revised Procedure 50-G-15 "Backshif t, Weekend, Holiday Visit and Duty Assignments," to establish a duty supervisor program.

This program will ensure _that a member of

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senior management is continuously available to provide the appropriate resources to support the shift supervisor.

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Evaluate the schedule for upgrading the alarm procedures provided to the operations staff.

In NRC Inspection Report 50-285/90-14, a Notice of Violation (285/9014-01) was issued that addressed the adequacy of alarm procedures. This item will be reviewed during followup of the violation.

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• Review procedures to clarify on-shift responsibilitie.

ENCLOSURE CONTAINS

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SAFEGUARDS INFORMAil0N i

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UPON SEPARAT10N THIS PACC iS DECONTROLLED

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Procedure 50-0-5, " Shift Supervisor Duties," was revised to clarify the responsibilities of the shift supervisor.

• Implement a TS interpretation procedure.

Procedure N00-QP-32, " Technical Specification Interpretations,"

was issued to provide the method by which formal interpretations of the TS shall be requested, processed, and maintained.

Investigate proceduralizing actions for operability

determinations.

Procedure 50-0-1,." Conduct of Operations," was revised to incorporate detailed guidance concerning operability determinations.

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Improve the screening process for-industry information.

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Procedure N00-QP-21. " Operating Experience Review Program," was revised to improve the screening of industry information by

establishment of a priority system that classes an event based on the safety significance.

• Review TS for actions needing time clarification.

The licensee completed a review to determine which TS LCO actions need clarification with respect to the amount of time available prior to initiation of LCO action statements.

The operations staff was reviewing the listing.

TS amendment requests

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will be submitted in the future, as appropriate.

• Review procedures to lower the threshold for initiation of an investigation.

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Procedure NSRG-3, " Reviews and Investigations, was revised to

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establish a lower threshold at which an investigation would be initiated.

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(Closed) Open Item 285/8944-03:

Install a Linkage Retainer on Valves HCV-840A and HCV-8408.

L This item involved a review performed by the licensee during followup on a problem where the linkages for nine valves were not properly secured.

The nine valves have dual linkage mechanisms installed.

One mechanism is used to connect the valve stem to the valve operator and the other mechanism is used to connect the valve stem to the handwheel.

Should the linkage for the handwheel become inadvertently engaged while operating the valve with the valve operator, the valve-will not operate because the valve operator can not supply sufficient torque to rotate the handwheel and its installed gear box.

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ENCLOSURE CONIMNb -

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SAFEGUARDS INFORMAil0N

UPON $[PARATION THl$

PAGE IS DECONTROLLID

.g The licensee installed retaining devices, used to ensure that the handwheel mechanism does not become engaged, on all valves outside containment prior to the refueling shutdown. A review of documentation indicated that Valves HCV-840A and HCV-840B required

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the installation of retaining devices. The licensee stated the devices would be installed during the refueling outage.

During the 1990 outage, the licensee performed a visual inspection of the valves located inside containment, including Valves HCV-840A and HCV-840B.

Based on this inspection, the licensee did not identify any valves that required the installation of a. retaining device, h.

(Closed) Unresolved Item 285/8950-02: Concerns Identified with the Raw Water (RW) System.

This item was related to concerns identified with the RW system that involved the licensee's ability to place the plant in bot shutdown without RW system flow; potential for common-mode failure of all four of the RW pumps; and the licensee's ability to place the plant in cold shutdown, as required by TS 2.0.1, in the event that all RW pumps are lost.

To address this item, the licensee issued Licensee Event Report (LER)89-017 to fully document the status of the concerns and to provide a description of the actions to be taken to resolve the concerns. This item is considered closed as followup reviews will be performed during closecut of LER 89-017.

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(Closed) Unresolved Item 285/8950-05: Use of the Containment Spray (CS) Pumps for the Shutdown Cooling (SDC) Mode of Operation.

This item was related to the use of the CS pumps for the SDC mode of operation. The licensee identified that the suction piping for the CS pumps was nat designed to meet the entry conditions (i.e., pressure and temperature) for the SDC mode of operation.

To address this issue, the licensee issued LER 89-024 to fully document the concerns and to provi'e corrective actions for resolution of the problems.

This item is considered closed because a review of the licensee's actions specified by LER 89-024, will be performed.

Items related to this issue are discussed in paragraph 4 of this inspection report.

Overall,.the actions taken by the licensee in response to previously identified items appeared to be conservative and provided adequate controls to prevent recurrence.

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SAFECUARDS INFORMATiON UPON SEPARATION THl3

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PACE IS DECONTROLLED

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4.

LER Followup (92700)

The following licensee event reports were reviewed to verify that reportability requirements were fulfilled and corrective actions were accomplished to prevent recurrence, a.

(Closed) LER 29-022 documented a situation where, on August 5, 1989, a change to Procedure ST-CEA-1, " Control Element Assemblies," was approved by the plant review committee (PRC) and would have made both emergency diesel generators inoperable during performance of a portion of the test.

Prior to performance of the test, the procedure problem was identified by a shift supervisor and was never performea.

The root causes of the incident were determined to be inadequate preparation of a 10 CFR Part 50.59 evaluation relative to the test and an administrative error made by the PRC during review of the procedure change. The event was fully documented in NRC Inspection Report 50-285/89-48. The event resulted in the documentation of a noncited violation pursuant to Section V.G.1 of Appendix C to 10 CFR Part 2.

In LER 89-022, the licensee committed to perform the following carrective actions:

Train all regular and alternate PRC members on the event in a case-study format.

• Incorporate lessons learned from this event into the Part.50.59 i

safety evaluation training program.

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Revise the Part 50.59 safety evaluation training program to include periodic recertification of personnel qualified to

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perform safety evaluations.

• Reissue the Part 50.59 safety evaluation for the procedure revision to ST-CEA-1.

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(0 pen) LER 89-024 described an issue, identified by the licensee,

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where the suction piping of the CS pumps did not meet the pressure and temperature requirements for entry into the SDC mode of operations.

In the LER, the licensee committed to take certain actions to address this issue prior to startup from the 1990 refueling outage.

The inspector reviewed the licensee's actions, as discussed below:

Procedures 01-5C-3, " Alternate Shutdown Cooling Utilizing Containment Spray Pumps," and 01-SC-4, " Termination of Shutdown Cooling," were revised to specify the conditions when the CS i

pumps could be used for SDC operations.

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Operations Memorandum 89-05, used to provide instructions to operations while procedures were being updated, was cancelled when the revised procedures were issued.

• Emergency Procedures AOP-19, " Loss of Shutdown Cooling"; AOP-6,

_." Fire Emergency"; E0P-03, " Loss of Coolant Accident"; E0P-06,

" Loss of All Feedwater"; and E0P-20, " Functional Recovery Procedure";;were revised to specify the conditions that the CS pumps could be used for SDC.

• Attachment 12. " Emergency Repair of LPSI Pump SI-1B," to Procedure AOP-6 was issued to provide instructions for connecting temporary electrical power to a low pressure safety

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injection (LPSI) pump in the event that the normal electrical supply was damaged in a fire.

This LER remains open pending licensee completion of the long-term actions specified in the LER.

c.

(Closed) LER 90-001 documented an event where the licensee failed to-provide a firewatch for a nonfunctional fire barrier in accordance with TS 2.19(7).

To address this event, the licensee:

Revised Procedure 50-G-58, " Control of Fire Protection System Impairments," to require that the fire protection system engineer coordinate planned fire barrier impairments.

• Reemphasized to shift supervisors that a firewatch must be established prior to authorizing a fire barrier impairment.

In addition to the above, the licensee established, in March 1990, a group dedicated to implementation of hourly firewatch patrols.

The firewatch patrol function was previously performed by security

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i guards.

Since this dedicated group has been established, no problems have been experienced with missed firewatch patrols.

Based on the reviews performed by the inspectors, it appeared that the licensee had taken timely actions to implement controls to prevent recurrence of the identified events, i

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5.

Followup on TMI Item 1.D.2.3 (TI 2515/65)

In NRC Inspection Report 50-285/90-07, a review was performed to verify that the licensee had installed a safety parameter display system (SPDS)

in accordance with TMI Item I.D.2.3, as specified by NUREG-0737. The review concluded that all items related to-the SPDS had been completed, except for providing a continuous display of the critical safety function boxes on an SPDS terminal in the control room.

During the 1990 refueling outage, the licensee installed a critical safety function box display in accordance with Modification Request FC-87-055,

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PACE IS DECONTR0t1ED-11-

"ERF Computer Terminal _ Upgrade." Based on this action, this item is considered closed.

6.

Om rational Safety Verification (71707)

The inspectors conducted reviews and observations of selected activities to verify that facility operations were performed in compliance with the appropriate regulatcry requirements. The inspectors identified the following items:

a.

During a tour of the control room on May 17, 1990, the inspector noted that it did not appear that the licensee was complying with the requirements specified in TS 2.1.4(5).

This TS states that, to determine leakage to the secondary' system, one of the following must be operable at all times:

(a) steam generator blowdown radiation sample instrument, (b) condenser offgas radiation monitor, or (c) periodic secondary samples analyzed for activity. At the time the inspector identified this apparent problem, the plant was in Mode 5 (refueling shutdown) and ".ue of the actions specified by the TS were being performed.

To resolvo this issue, the inspector contacted NRR for a clarification of TS 2.1.4(5).

NRR stated that operable, as related to this particular TS, was intended to mean:

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In Mode 5, the licensee shall have the capability of performing one of the actions specified by the TS.

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In Modes 1 through 4, the licensee was required to perform one of the actions specified by the TS.

This clarification of the TS was discussed with the licensee.

Prior to leaving Mode 5, the licensee initiated sampling of the steam generators for activity on a daily basis.

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On June 7, 1990, severe weather conditions, including the development of a wall cloud just west of the plant and the siting of tornadoes in the Omaha area, developed with little warning.

The plant Manager evacuated all nonessential personnel to their designated emergency response locations. The Plant Manager; Supervisor,-Operations; and i

Manager, Station Engineering reported to the control room in j

preparation for the declaration of a notification of unusual event (NOVE).

The inspector also reported to the control room.

The conditions did not develop into a tornado at the site and the storms at the site were of moderate intensity.

The storm quickly

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passed and a NOVE was not declared.

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The inspector noted that evacuation of plant personnel was prudent and effectively executed. The communications established in the control room and the preparations made in anticipation of a NOVE i

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appeared to be actions that would.have enabled the licensee to transcend smoothly into an emergency classification, had it been necessary.

No violations or deviations were identified in this program area.

7.

Monthly Maintenance Observations (62703)

The inspectors reviewed maintenance activities, as discussed below:

a.'

On May 29, 199P, the licensee informed the inspector that Valve YCV-104',A (steam supply valve to the turbine-driven auxiliary

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feedwater p'..np) had failed its quarterly surveillance test. The safety-re bted accumulator assembly, designed to hold the Valve YCV-1045A closed in the event of a_ steam generator tube rupture

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coincident with a loss of instrument air, did not meet the test '

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acceptance criteria.

System air leaks prevented the accumulator assembly from fulfilling its design function of holding the valve closed for at least 30 minutes.

Valve-YCV-1045A fails open on a loss of instrument air pressure.

The licensee's review revealed that the accumulator assembly had

. undergone preventive maintenance (PM), on May 5,1990, to replace the ASCO solenoid-operated. valve used.to control the position of Valve YCV-1045A.

Fasteners,_~used to secure;the diaphragm for the solenoid-valve, were left untorqued after maintenance was completed.

The fasteners ensure the integrity of the instrument air system pressure boundary. The postmaintenance testing (PMT) instructions provided by Procedure MP-50V-1,."ASCO-Solenoid Valve Preventive Maintenance to Maintain 79-01B Qualification," were inadequate.

The PMT instructions required only that'the replaced _ solenoid be cycled to verify operation, no instructions for leak testing the assembly

were provided.

Thus, the pressure integrity of the accumulator assembly was not verified following maintenance.

Section 5.1.6.1 of ANSI 18./-1972 requires that maintenance affecting

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the function of safety-related components be-performed in a manner that ensures that quality is at least equivalent to that specified in l

the design requirements.

The design requirements for Valve YCV-1045A, as documented in Section 6,2 of Design Basis Document SDBD-FW-AcW-117 and l

Section 5.9.5 of the Updated Safety Analysis Report, state that i

Valve YCV-1045A has an air accumulator and ren.ains operable for 30 minutes following a loss of instrument air.

The PMT instructions for Maintenance Procedure MP-SOV-1 failed to assure quality at least equivalent'to design requirements in that no instructions were provided to pressure test the accumulato'.' assembly following replacement of the solenoid-operated valve.

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nonconforming condition was not discovered until the quarterly

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scheduled surveillance test was performed approximately 3 weeks after the solenoid replacement.

This is a violation (285/9030-01).

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The licensee repaired the solenoid-operated valve diaphragm, on May 29, 1990, and verified that the accumulator assambly for Valve YCV-1045A functioned properly, including no pressure boundary air leaks.

The licensee investigated the potential for inadequate PMT

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instructions affecting the integrity of other safety-related air operated valves.

The investigation. revealed that the ASCO solenoids for Valves HCV-4838 and HCV-4380 (component cool.ing water inlet and

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outlet valves for reactor cooling pump bearing oil cooler) had underwent PM in accordance with Procedure MP-50V-1 on April 26, 1990.

However, surveillance testing, in accordance with Procedure OP-ST-CCW-3004, " Component Cooling-Water Category A and B Valve Exercise Test,'! was pa: rormed on May. 12, 1990.

The licensee also verified that non t iety-related valves affected by the PM had t

been properly tes+ w.

t During review of this problem, the licensee noted that the personnel assigned the responsibility to review the adequacy of PM procedures i

lacned the plant specific and design basis knowledge to verify that i

adequate PMT instructions were included in the procedures.

Therafore, as short-term corrective action, the licensee reassigned the procedure r

review responsibility to the respective system engineers who understand the design basis requireme.nts of the individua' components.

Long-term corrective actions were initiated by the licensee and included training for specia1 services engineers,_the PM administrator, and PM planners to. strengthen their PMT knowledge level. -In addition, all outage and nonoutage PM procedures will be reviewed to verify that adequate PMT instructions have been included.

During review of this event, the irspector noted that the licensee aggressively investigated the root cause of the incident and

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implemented a short-term corrective action plan.

b.

During review of surveillance tests, the inspector noted that some of the equipment failed to meet the acceptance criteria.

The inspector verified that the licensee had issued MW0s for the-repair of the equipment.

The inspector reviewed 12 completed MW0s to verify that the equipment was repaired before its system was required to be operable, and that appropriate PMTs had been performed.

No problems were identified during the review of the completed MW0s.

During observation of the maintenance activities performed by_

licensee personnel, the inspectors observed that the maintenance evolutions were performed in accordance with the appropriate regulatory requirements.

No deviations were identified in this program area; however, one violation was identified.

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UPON SEPARAil0N THIS PACE IS DECONTROLLED-14 8.

Monthly Surve111anet Observations (61726)

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The inspectors observed TS-required surveillance testing on safety-related systems and components, as discussed below:

a.

On May 13, 1990, the inspector observed testing of the remote control panel (Al-179) for the auxiliary feedwater (AFW) system. The AI-179 controls are used to operate AFW components, for plant cooldown, in the event that the contrei room has to be evacuated.

The testing was performed in accordance with Procedure OP-ST-AFW-3005, "AFW Operational Verification From AI-179."

The test adequately verified that the controls on AI-179 functioned properly and that control of the components could be switched from the control room to AI-179.

No problems were identified during performance of the testing, b.

During review of surveillance test requirements contained in the TS, the inspector noted that TS 3.6(3)(a) stated that alternate manual starting between the control room console and the local panel for the safety injection, shutdown cooling, and containment spray pumps shall be practiced during refueling outages.

The inspector was aware that the plant does not have local control panels installed for operation of these pumps.

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The inspector notified the licensee of the apparent discrepancy.

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licensee reviewed the origin of this requirement and could not determine why the TS addressed local control panels since the panels have never been installed in the plant. The Acting Division Manager, i

Nuclear Operations, stated that an amendment request, with proper justification, would be submitted to NRR in the near future to remove the requirement from the TS.

c.

The inspector reviewed a sampling of 16 completed surveillance tests that were performed during the 1990 refueling cutage. The inspector identified the following items

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In ten test procedures, the inspector noted that procedure steps were initialled by the individual completing the work; however, the

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individual's signature was not recorded in the test, as required by the testing instructions, to identify who performed the work.

Identification of the individuals performing tha test is required by ANSI N18.7-1976.

The licensee reviewed the ten test procedures and was able to identify each individual's initials and verify that each step was performed by a qualified individual.

For this reason, it did not appear that this problem was safety significant.

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The Acting Division Manager, Nuclear Operations, stated that a review of the requirements for identifying the individuals that perform testing would be completed to verify that an acceptable program was in place and changes to the program would be made, if required.

In addition, the inspector reviewed Procedure 50-G-23, " Surveillance Test Program," which provides direction to personnel as to when an

"N/A" (not applicable) can be entered in the signature block of a procedure step in lieu of performing the step.

The inspector revlewed Procedure 50-G-23 and noted that the guidance provided appeared to be too unrestrictive.

Procedure 50-G-23 states that, during the performance of the test, any step / data entry that is not applicable shall be identified by "N/A," initialled, dated, and an explanation provided in the remarks or on the comment sheet /

chronological log. Based on this guidance, an unapproved procedure

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change could be made due to an individual's interpretation of the requirements.

The inspector did not identify any examples wPere an "N/A" was used

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inappropriately.

The Acting Division Manager, Nuclear Operations, stated that a review of the guidance for the ut.e,of "N/A" would be performed to verify that the instructions prov ded to' plant personnel-were clear.-

The inspector will review these areas of concern in the future during routine inspections of the surveillance testing program.

No violations or deviations were identified in this program area.

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9.

Security and Radiological Protection (RP)' Observations (71707)

During this inspection period, the_ inspectors identified the following items:

a.

On May 24, 1990, the licensee identified a situation where an l

individual violated radiological work permit (RWP) instructions by l

entering a high radiation area without an integrating alarming I

dosimeter. The individual promptly realized'the error,. exited the area, and received minimal exposure.

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The licensee maintained it exercised positive control over the situation in accordance with TS 5.11.1(c) and therefore no violation

of TS occurred.

RP specialists from Region IV will review the event during a future inspection.

Pending completion of this review, this-issue will be tracked as an Unresolved Item (285/9030-02),

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i b.

On May 5, 1990, two licensee employees received higher than anticipated personnel exposures while reinstalling.the fuel transfer tube flange.

The event was documented in NRC Inspection Report 50-285/90-26 and focused on the licensee's compliance with 10 CFR Part 20.101 and a concern involving the failure of integrated slarming dosimeters in humid environment _ - _ _ _ _ _ _ _ _ _ _ _.

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During this inspection period, the inspector continued followup of the event and noted that RP personnel were aware that the radiation levels in the area of the transfer tube were significantly higher l

than the survey which was used for prejob planning indicated and

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The licensee performed a self-initiated investigation and found:

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Surveys used for planning the job did not reflect actual doses l

and were not appropriate'to the task.

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The RWP and associated as-low-as-reasonbly-achieveable (ALARA)

briefing did not accurately communicate radiological hazards and applicable protective measures.

• The RP technician providing. coverage did not take measures to

adequately control exposure and exercised improper judgement in allowing the job to proceed.

• RP supervision needed to play a more active role in the planning l

of the transfer tube flange installation.

l The licensee prepared a comprehensive root cause analysis document

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and identified the following corrective actions,

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Documented prejob surveys, in advance of the actual task being performed, will be required for those tasks that involve potential high individual exposure.

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A discussion of thie event will be incorporated into the RP technician training

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• Detailed instructions w 11 be incorporated into the RWP for RP i

technician coverage.

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l Current survey data will be used in the ALARA briefing process.

• Weaknesses identified in the ALARA briefing process (e.g.,

I detailed guidance on conduct of ALARA briefing) will be I

corrected.

" Sacrificial dosimetry" and "Teledose" capabilities will be utilized where appropriate, and will be specified on applicable i

L RWPs.

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• The use of self-reading dosimeters having a range that exceeds the specified alarm setpoint of the companion Xetex, will be.

required.

Section 20.201(b)(2) of 10 CFR Part 20 requires that radiological surveys be performed, that are reasonable under the circumstance, to evaluate the extent of radiation hazards that may be present.

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-17-use of the lower cavity postdecontamination survey for job preplanning

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for installation of the flange was inappropriate for the radiological hazards associated with work in close proximity to the upender

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assembly.

Normally, a Notice of Violation would be issued for noncompliance with 10 CFR Part 20,201. However, the violation is not being cited because the criteria specified in Section V.G 1 of the Enforcement Policy were satisfied.

c.

A violation involving an inadequate vital area barrier was identified.

The details are provided in Appendix C of this inspection report.

No deviations were identified in this program area; however, one violation and one unresolved item were identified.

10. Verification of Containment Integrity (61715)

The inspectors performed a walkdown of selected containment penetrations to verify that the licensee had established containment integrity prior to

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commencing heatup of the reactor coolant system above 210*F.

The review included verification of containment isolation valve positions, that electrical and mechanical penetration assemblies were tested and properly restored to service, and that local leak-rate tests were completed on the equipment hatch, fuel transfer tube, and personnel

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airlock.

No violations or deviations were identified in this program area.

11. Onsite Followup of Events (93702)

On May 7,1990, the licensee notified the inspector that a problem was identified with the low-temperature werpressurization (LTOP) analysis performed by Combustion Engineering (CE).

In discussions between CE and the licensee, it was determined that the LTOP analysis incorrectly assumed that the power-operated relief valve (PORV) opened. instantaneously, an ir. correct assumption.

l CE reperformed the analysis using an assumption that the PORV opened in 0.8 seconds.

Based on the reanalysis, the pressure and temperature operating limits specified in the TS will require that an amendment request be submitted to specify new values.

The TS operating parameters

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The primary-to-secondary delta.T limit will have to be reduced from 50 to 30 F for starting a reactor coolant pump during a solid-water i

condition in the reactor coolant system (RCS).

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Plant cooldown will be restricted to less than or equal to 20 F i

per-hour when the RCS temperature reaches 300*F.

The 20 F per-hour limit was previously applicable at an RCS temperature of 280'F.

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The high pressure safety injection (HPSI) pumps will have to be

disabled at 325'F for the first pump, 3200F for the second pump, and 290'F for the third pump. The previous ten.teratures at which the HPSI pumps were required to be disabled were 320, 312, and 271'F, respectively.

The licensee reviewed the results of the reanalysis and noted that the data was more conservative than the values presently provided in the TS.

Therefore, the licensee implemented administrative controls to ensure that the plant was operated in compliance with the newly established limits, f

In a telephone conference, on May 10, 1990, between NRR, Region IV, and the licensee, NRR concurred that the licensee could implement administrative controls for the new limits.

This concurrence was based on the licensee providing a letter to NRR to specify the types of administrative controls that would be implemented, and a commitment from the licensee to submit an amendment request for the appropriate TS by July 31, 1990.

To document the details related to this issue, the licensee issued LER 90-015.

Routine review of this issue will be performed during verification of the completion of the licensee actions specified in the LER.

No violations or otviations were identified in this program area.

12.

Complex Surveillance 161701)

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l The inspector observed tasting of the alternate shutdown panel (AI-185)

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that was performed to verify that the controls functioned adequately.

Panel AI-185 controls are used to place the plant in cold shutdown in the event that the control room must be evacuated.

The testing was performed in accordance with Procedure OP-ST-ASP-002, " Alternate Shutdown Capability Control Circuitry Verification."

The testing adequately verified that all the controls on AI-185 functioned as designed to ensure that plant shutdown could be performed from the remote location.

No violations or deviations were identified in this program area.

13.

Review of Licensee Outage Plans (60705 and 60710)

The inspector reviewed the licensee's outage plans to establish a baseline of significant activities scheduled to be completed during the outage.

The baseline was established to provide a method of determining, after completion of the outage, the significant scheduled activities that had been deferred by the licensee.

Of the modifications scheduled to be completed during the outage, completion of two was deferred until-the plant resumes power operation.

One item involved upgrading the emergency lighting system.

Although not

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UPON SEPARATION MS Pact IS DEC0lmt0LLED 3g, completed during the outage, work was in progress and it was anticipated that the modfication will be completed by July 31, 1990, the licensee's commitment date to the NRC.

In 1986 the licensee issued a 10 CFR Part 21 report that documented problems with Valves HCV-247 and HCV-248, used for long-term core cooling.

The springs in the valves failed' catastrophically due to hydrogen embrittlement.

During the licensee's review, it was also noted that Valves HCV-249 and HCV-2988, long-term core cooiing valves, were also susceptible to failure by embrittlement of the valve spring.

To address this apparent problem, the licensee committed to replace the valves during the 1987 refueling outage.

Valves HCV-247 and HCV-248 were

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replaced; however, Valves HCV-249 and HCV-2988 had not yet been replaced due'to problems encountered with obtaining replacement valves, and were currently scheduled to be replaced during the 1991 outage.

The licensee stated, in internal Memorandum PED-FC-90-323, that the valves had been stroka tested each outage and there had been no indication of degradation;

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therefore, immediate replacement was not required.

The licensee's position was reviewed by Region IV personnel. As a result

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of the review, it was concluded that the licensee's position of replacing the valves during the 1991 outage was satisfactory.

No violations or deviations were identified in this program area.

14.

Plant Startup from Refueling (71711)

On May 17, 1990, the licensee completed its major outage work and commenced RCS heatup.

The transition from Mode 5 (refueling shutdown) to Mode 3 l

(hot shutdown) was made on May 22, 1990.

The inspector observed, on an intermittent basis, control room activities during this interval and found them well coordinated. Tho mode transition was observed to have been performed in accordance with Procedure Op-2, " Plant Startup From Refueling or Cold Shutdown to Hot Shutdown."

Between Modes 5 and 3, the l'.censee passed several plateaus:

150'F, 210$F, 300'F, steam bubble in pressurizer, and Mode 3 itself.

For each plateau there were numerous. equipment lineups and verifications of operability that j

were required to be satisfied.

These were controlled by Procedure OP-1,

" Master Checklist for Startup." The inspector observed that the procedure was followed.

The reactor achieveo criticality on May 25, 1990, at approximately 6:30 p.m.

Between May 22 and May 25, 1990, the major work prerequisites to achieving criticality were verifying operability of the turbine-driven auxiliary feedwater pump, verifying operability of the RPS channels, restoring Raw Water Pump AC-20B or AC-10A to an operable status, and achieving proper boric acid concentration in the boric acid storage tanks and the safety injection and refueling water tank.

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-20-EE N The' inspector observed that the restoration of FW-10, AC-10A, and the j

calibration of the RPS channel were performed effectively with the use of procedures. With respect to mixing boric acid in the storage tanks, it

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appeared the licensee might have violated procedural requirements.

The issue will be tracked as an Unresolved Item (285/9030-03).and will be

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addressed subsequent to an internal, licensee investigation, scheduled to j

be completed in the'near future.

Prior to achieving criticality, the inspector independently verified

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proper system lineup of the auxiliary feedwater and high pressure safety.

injection systems.

Procedures 01-51-1, "High Pressure Safety Injection (HPSI)," and 01-FW-4, " Auxiliary Feedwater Pump Operation and Testing Auxiliary Feed Operation," were used for the walkdown.

No anomalies were identified,

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After achieving criticality, the licensee implemented Proce-dure RE-0PT-RX-0001, " Post Refueling Core Physics Testing and Power Ascension." The inspector observed selected portions of the performance of Attachments IA "CEA Withdrawal Base Count Rate Determination"; IB,

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"CEA Withdrawal '

erse' Count Rate Determination"; 1E, " Initial Criticality and Low b er '

.:s Test Data Record"; 3, " Full Length and Non-Trippable CEA Coupling

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,:hn i q ue. " During the portions of the tests witnessed, no problems were noted. The low power physics testing team of engineers executed the procedure in a professional manner.

On May 29, 1990, the inspector witnessed power ascension from 2 to 12 percent reactor power in accordance with Procedure OP-3, " Plant Startup -

From Hot Standby to Minimum Load." Between 2 and 12 percent, the turbine was reset, the chest warmed, and the turbine rolled. At 12 percent, the

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turbine generator was loaded and commercial operation began.

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operations staff executed the transition from hot standby tc minimum load in a cautious, professional manner.

In summary, it appeared that the licensee conducted the' plant startup from refueling in a proceduralized, well-controlled, and effective manner.

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apparent procedural noncompliance issue noted above appeared to be isolated.

No violations or deviations were identified in this program area; however, one unresolved item was identified.

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Exit Interview The inspectors met with Dr. J. K, Gasper, Manager, Training, and other members of the licensee staff on June 19, 1990. The meeting attendees are listed in paragraph 1 of this inspection report.

At this meeting, the inspectors summarized the scope of the inspection and the findings.

During the exit interview the licensee did not identify any proprietary information.

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