IR 05000267/1987014

Insp Rept 50-267/87-14 on 870601-30.No Violations Noted. Major Areas Inspected:Operational Safety Verification, Followup on Previous Insp Findings,Design Control,Followup of Plant Trip Maint & Surveillance

ML20235M045
Person / Time
Site:	Fort Saint Vrain
Issue date:	07/13/1987
From:	Farrell R, Jaudon J, Michaud P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20235M021	List: IR 05000267/1987014 ML20235M017
References
50-267-87-14, NUDOCS 8707170033
Download: ML20235M045 (9)

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APPENDIX U. S. NUCLEAR REGULATORY COMMISSION- 1 REGION IV j NRC Inspection Report: 50-267/87-14 License: DPR-34 Docket: 50-267 .

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Licensee: Public Service Company of Colorado (PSC) )

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Facility Name: Fort St. Vrain Nuclear Generating Station Inspection At: Fort St. Vrain (FSV) Nuclear Generating' Station, Platteville, Colorado and PSC Offices, Denver, Colorado (

Inspection Conducted: June 1-30, 1987 Inspectors: E . F'fi r r e l l , Senior Resident Inspector (SRI) Date b8[

M ... Date hM d P. W ." Mi c h a'u d ,"R q$ i d e n't Inspector (RI)

Approved: *+ /M

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V3; P. Jaudon, Chief, Project Section A Date Reactor Projects Branch

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8707170033 870713 PDR ADOCK 05000267 0 PDR l

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-2-Inspection Summary Inspection Conducted June 1-30, 1987 (Report 50-267/87-14)

Areas Inspected: Routine, unannounced inspection of operational safety verification, followup on previous inspection findings, design control, followup of plant trip, maintenance, and surveillanc Results: Within the six areas inspected, no violations or deviations were identified.

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-3-DETAILS Persons Contacted Principal Licensee Employees D. Alps, Supervisor, Security i

• F. Borst, Manager, Support Services / Radiation Protection i *L. Brey, Manager, Nuclear Licensing and Fuels R. Craun, Manager, Nuclear Site Engineering D. Evans, Superintendent, Operations

• M. Ferris, Manager, QA Operations W. Franek, Superintendent, Planning, Scheduling, & Stores l

• D. Frye, Senior Specialist, Nuclear Licensing j

• C. Fuller, Manager, Station J. Gramling, Supervisor, Nuclear Licensing Operations ,

• M. Holmes, Manager, Nuclear Licensing 1

• R. Husted, Engineer, Licensing

• M. Niehoff, Manager, Nuclear Design

• F. Novachek, Manager, Technical / Administrative Services

• T. Prenger, Manager, QA Services

• J. Reesy, Manager, Special Projects

• P. Tomlinson, Manager, QA R. Walker, Chairman of the Board and CEO

• D. Warembourg, Manager, Nuclear Engineering

• R. Williams Jr., Vice President, Nuclear Operations The NRC inspectors also contacted other licensee and contractor personnel during the inspectio * Denotes those attending the exit interview conducted June 30, 198 l Followup on Previous Inspection Findings (CLOSED) Violation (267/8422-01): Inadequate Weld Control. This violation involved weld data sheets not attached to the control work procedure (CWP) and not containing the required testing and inspection information, and weld rod control. The NRC inspector examined recent CWPs, which included welding activities, and found no evidence of a recurrence.of these problems. Specific items contained in the original violation were corrected in early 1985. This item is close (CLOSED) Violation (267/8422-02): Inadequate Quality Control (QC) Design Controls. Hold points were not inserted in CWP deviation reports, and a review of completed CWPs had not been performe Procedures QICM-5 and i

G-5 were revised and specific items identified in the violation were corrected in December 1984. This item is close (CLOSED) Violation (267/8422-03): Systems Returned to Service Without the Shift Supervisor Performing the Required Verifications. A revision was made to Administrative Procedure G-9, " Controlled Work Procedures,"

requiring the shift supervisor to have a CWP in hand before a clearance

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J can be placed or removed, thus ensuring work completion prior:to returning a system to service. Training was held on the new G-9 procedure during the first quarter of 1985,'and .there has' not been an apparent. recurrence of this problem since. This item is close (CLOSED) Violation (267/8422-06): Failure'to Submit Adequate Informatio This violation concerned submittals to IE Bu11etin'80-11 not made under

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oath or affirmatio Procedure ENG-7 was revised. to. prevent- a ' recurrenc The NRC inspector found nothing to suggest this problem has occurred since. The technical issue of IE Bulletin 80 11 will be' addressed during-closure of the bulletin. This item is' close (CLOSED) UnresolvedItem'(267/8429-05): Inadequate High Range Area Monitor Calibratio RT-93250-14 was not being.. calibrated as required by NUREG 0737. SR 5.4.9-A4, "High Range Area Monitors Calibration," dated :

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May 20, 1985, checks both RT-93250-14 and RT-93251-1 with a 1 R/hr sourc This item is close ,

i (CLOSED)OpenItem(267/8430-01): Inadequate Control of Nonconforming i Material . Nonconforming materials did not have a " HOLD"l tag placed on them; therefore, they'could have been used without verifying disposition of the nonconformanc Procedure Q15, " Control.of Nonconforming Items," ,

was revised and now requires " HOLD" tags to be attached to the material I and to the nonconformance report (NCR) unti1~it'is dispositioned. This- ;

item is close I i

(CLOSED) Violation (267/8705-01): NCR Hold Tag Placed on Installed Plant l Equipment Without Shift Supervisor Concurrence. Procedure Q-15 was- '

revised, removing any ambiguity about obtaining shift supervisor. approval of hold tags. No repetitions have'been identified. This~ item is closed.- Operational Safety Verification The NRC inspectors reviewed licensee activities to ascertain that the  ;

facility is being operated safely and in conformance with regulatory j requirements and that the licensee's management control system is  :

i effectively discharging its responsibilities for continued safe operation.

l l The review was conducted by direct observation of activities, tours of the facility, interviews and discussions with licensee personnel, independen verifications of safety system status and limiting conditions for operation, and review of facility record ;

Logs and records reviewed included:

. Shift supervisor logs

. Reactor operator' logs j

. Equipment operator logs

. Auxiliary operator logs

. Technical Specification compliance legs l

. Operations order book

. Operations deviations reports i l

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. Clearance log

. Temporary configuration reports

. Station service requests (SSR)

During ' tours of accessible areas, particular attention was directed to the .

following:

. Monitoring instrumentation

. Radiation controls l . Housekeeping

. Fluid leaks i . Piping vibrations  ;

. Hanger / seismic restraints j

. Clearance tags i

. Fire hazards

. Control room manning  ;

. Annunciators j No violations or deviations were identifie . Design Control Independent Design Verifications The NRC resident inspectors, in response to a headquarters request for assistance, reviewed the licensee's design verification activities for calculations done by contractor Special emphasis was placed on calculations of firewater flow for safe shutdown from 82 percent powe Documents reviewed included: i

. PSC Memorandum NDS87-0648, Niehoff to Holmes, June 12, 1987, regarding " Independent Verification of Secondary Coolant Flows During Safe Shutdown Cooling" ]

l . Proto Power Letter, Geaney to Tilson, June 11, 1987, regarding l independent verification 0f Proto Power Calculations 82-03, 82-09, 82-12, and 82-18

. Proto Power Letter, Geaney to Tilson, June 9,1987, regarding independent verification of inputs to Proto Power i

Calculations 82-03 and 82-12 l

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. Proto Power Letter, Geaney to Tilson, June 9,1987, regarding  ;

I Proto Power independent design ~ verifications

. Proto Power Letter, Geaney to Tilson, June 4,1987, regarding pressure drop program independent. verification

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. Proto Pcwer Report EE-EQ0057, Revision A, May 1, 1987,

" Evaluation of Test Data for. the Confirmation of Firewater Flow

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Rate to the Circulator Water Turbine During EES' Cooldown for i Safe Shutdown Cooling" 1'

. Proto Power Calculation 82-32, April 30, 1987, " Determination of Circulator Water Turbine Flow Rate for Safe Shutdown Cooling"

. PSC Letter P-87122, Brey to Berkow, April 10, 1987, " Additional Information for Analysis of Firewater Cooldown for 82 percent Power Operation"  ;

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. PSC Letter P-87110, Brey to Berkow, March 20, 1987, " Additional Information for Analysis of Firewater Cooldown for 82 percent Power Operation" )

. PSC Letter P-87055, Brey to Berkow, February 17, 1987,

" Additional Information for Analysis of Firewater Cooldown for i 82 percent Power Operation" l

. Proto Power Calculation 82-18, February 19, 1987, "Effect of l Varying Pelton Wheel Flow Rate on EES Secondary Cooling Flow Rate for PSC - Fort St. Vrain EQ Program" )

. GA Technologies Report 909269, Issue A, February 4,1987, "EES Cooldowns for EQ and Appendix R Events with Vent Lines (1.5H Delay)"

. Proto Power Calculation 82-12, Revision C, January 28, 1987,

" Appendix R Safe Shutdown Cooling at X2 Power Level for PSC -

Fort St. Vrain"

. Proto Power Report EE-EQ-0023, Revision A January 23, 1987,

" Engineering Evaluation of the Procedure to Recover From an Actuation of the Steam Line Rupture Detection / Isolation System for Power Levels Through P2"

. PSC Letter P-87002, Williams to Berkow, January 15, 1987, proposed Technical Specification change eliminating reliance on reheater !;ection of steam generator for safe shutdown cooling

. PSC Letter P-86683, Warembourg to Berkow, December 30, 1986, analysis of firewater cooldown for 82 percent power operation

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Proto Power Calculation 82-03, Revision B, December 15, 1986,

"EES Shafe (sic) Shutdown Cooling for PSC - Fort St. Vrain ( Main Steam Vent Flow Path)"

. GA Technologies Report 909204, Issue A, December 4,1986,

"Effect of Firewater Cooldown Using EES Bundle on Steam Generator Structural Integrity" l .

Proto Power Calculation 82-10,' November 25, 1986, "EES Flooding l Time With Fire Water"

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. Proto Power Calculation 82-09, November 17,1986, " Pressure Drop Program"

. Letter Vishwanath Prasad to Geaney, October 10, 1986, regarding review of the technical basis of pressure drop program

. GA Technologies Report 909052, October 6,1986, " Firewater Cooldown with 300'F EES Exit Temperature (1 1/2 Hour Delay)"

. GA Technologies Report 909041, September 30, 1986, "The Effect of Delayed Firewater Cooldown with Loss of Liner Cooling on PCRV Temperatures"

. GA Technologies Report 909030, September 16, 1986, " Study of Firewater Cooldown After 1 1/2 Hour Interruption of Forced Cooling"

. Proto Power Calculation 82-01, September 11, 1986, "EES Safe Shutdown Cooling for PSC - Fort St. Vrain"

. GA Technologies Report 907935, Issue A, February 20, 1986, "FSV:

Delayed Firewater Cooldown; Effect of Liner Cooling on Orifice Valve Temperatures"

. PSC QA Audit of Proto Power performed in 1986 The licensee used several means to accomplish independent design verification (IDV). In all cases, an IDV was performed by the contractor responsible for the calculations in accordance with the contractor's 10 CFR 50, Appendix B, QA progra In many cases, the calculations of one contractor were verified by another contractor through alternate calculations. Computer codes were validated and input data independently verifie b. Design Change Control The wording in Procedure Q-15, Issue 6, " Control of Nonconforming Items," as modified by Procedure Deviation Reports (PDRs) 86-1942, 87-1065, 87-1299, and 87-1654 appears to allow design changes to be accomplished via NCRs. This is not the Mcensee's practice. The wording of paragraph 3.3.1.a) states, " Major design (system design / component design) are not altered."

The licensee does not appear to be using NCRs as a design change vehicle; however, the current wording of Procedure Q-16, appears to allow minor changes by this method without defining a break point between major changes and minor change Procedure clarification to preclude use of NCRs to change design is considered to be an open item (267/8714-01).

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-8- Temporary Configuration Reports (TCRs)

The licensee has over 100. TCRs installed. : Many of these TCRs are several years old. There is no definition.of." temporary" in the l

licensee's procedure SMAP-28 Issue 3, modified by PDR 86-1952,

" Processing of Temporary Configuration Reports." This is considered an open item (267/8714-02).

TCRs will be inspected in depth during a future NRC inspection.

No violations or deviations were identified .in 'this. inspection are . Followup of Plant Trip On June 12,1987, at 2:21 p.m. (MDT), the interlock selector switch (ISS).

was placed to the " POWER" position from the " LOW POWER" position'.- This

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was done to allow reactor operators to place the plant in an " automatic" mode to verify operation of control-systems. This requires a power level in excess of 30 percent reactor power, which is inhibited in the " LOW

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POWER" position of the ISS. When the ISS Was turned to the." POWER" position, a Loop 1 shutdown occurred. The ISS'was then returned to the

" LOW POWER" position. The cause of the loop shutdown was later determined l to be two improperly terminated wires for HV-2253, Loop 1 hot reheat stop check valve. This caused a false " CLOSED" signal to the plant protective system (PPS), which in turn caused the loop shutdown as soon as the ISS was placed in the " POWER" position. The'two wires involved were grounded during the EQ outage as part.of the electrostatic noise reduction work, which grounded spare conductors. The wires were labeled and shown as spares on the associated drawings, but were terminated incorrectly. . All systems initially responded properly and control room operators were taking actions to recover the loop, which included transferring house power, remoying the turbine generator from service, and reducing reactor powe Shortly after tripping the turbine, the reactor operator noticed Loop 2 feedwater flow and circulator speeds decreasing, and inserted a manual scram. All systems performed as expected following the scram. The cause of the decreasing primary and secondary flows in Loop 2 was a loss of steam, which provides the motive power to the circulators and the feed pump in service at the tim Normal operation on a turbine trip is for the main steam bypass valves to open, supplying steam to the bypass flash tank via desuperheaters, then onto the circulator (s) or feed pump (s). As the main steam pressure decreases, the startup bypass system automatically takes. control from the main steam bypass system to send steam to the bypass flash tank without desuperheating. On this occasion, the startup bypass system did not operate due to a valve position limit switch out of adjustment, which prevented the system isolation valve from opening. This resulted in the main steam bypass system continuing to operate with the desuperheater quenching the steam, which then resulted in a loss of steam. The elapsed time until primary coolant flow was restorec' was 16 minutes; secondary flow was restored in 5 minutes via the emergency condensate syste .The incorrectly terminated wires for HV-2253 were determinate and the entire PPS was functionally checked with the ISS in the " POWER" positio The improperly set limit switch on the Loop 2 startup bypass control valve

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was adjusted. The plant returned to' criticality on June 15, 1987, at 5:55 a.m. (MDT). The NRC resident inspectors will continue to monitor further corrective actions by the licensee as a result of this tri No violations or deviations were identified in this inspection are . Maintenance A. number of maintenance items were accomplished during the two day outage following the plant trip on June .12,1987. The NRC inspectors attended maintenance planning meetings and monitored the maintenance activitie ;

Items accomplished included the repair of the improperly grounded wires on Valve HV-2253, replacement of two welded valves, V-2237 and V-22431, replacement of bonnet gaskets on Valves TV-2228-2 and TV-2228-6, and a test of the entire PPS with the interlock selector switch in the " POWER" position. The NRC inspectors observed portions of these activities,.all of which were completed satisfactorily and on schedule. The NRC inspectors noted the effectiveness and efficiency of the PSC " war room" approach to the outage maintenance management. The " war room" is staffed L with a team composed of individuals from each group (planning and scheduling, engineering, licensing, operations, etc.,) who are dedicated to supporting the outage activities. Problems are categorized, evaluated, and resolved as appropriate to support the outage schedule. This concept was instituted near the end of the EQ outage and expedited its completio This approach, again proved valuable during this shorter outag No violations or deviations were identified in this inspection are ;

l 7. Surveillance j During the inspection period, the NRC resident inspectors observed performance of the weekly 10-inch control rod scram tests. Additionally, numerous plant protection system surveillance were observed as part of the scram investigation and recovery documented in paragraph Reactor operators were observed prior to reactor criticality verifying surveillance status and performing control room precriticality i surveillance in preparation for startu No violations or deviations were identified in this inspection are . Exit Meeting The NRC senior resident inspector conducted an exit meeting on June 30, 1987, attended by those indicated in paragraph 1. At this time, the inspector reviewed the scope and findings of the inspection.

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