IR 05000267/1988013
| ML20195H909 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 11/28/1988 |
| From: | Callan L NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | Robert Williams PUBLIC SERVICE CO. OF COLORADO |
| References | |
| NUDOCS 8812010111 | |
| Download: ML20195H909 (2) | |
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E I$ W In Reply Refer To:
l Docket:
50-267/88-13 Public Service Company of Colorado ATTh:
Robert O. Williams, Jr., Senior Vice President, Nuclear Operations P.O. Box 840 Denver, Colorado 80201-0840 Gentlemen:
<Thank you for your letter of September 20, 1988, in response to our letter and Notice of Violation dated August 19, 1988. As a result of our review, we find that additional information, as discussed with your Mr. Holmes (during a telephone call on November 22,1988), is needed.
With regard to item 1. Surveillance Procedure SR-5.2.16.a-Q and SR-RE-151-X are plant procedures that are reviewed and approved by the PORC.
Your response does not address corrective steps to avoid further violations which may have been prevented by your procedure review process.
In addition, yone response does not address corrective steps to assure that changes to your 'sechnical Specifications are appropriately incorporated in affected plant procedures.
We have noted your exception to Item 3 of the violation and agree with your response.
Please provide the supplemental information within 30 days of the date of this letter.
Sincerely
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. Original SV ' SP U,3; CALLAN L. J. Callan Directo*
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Division of Reactor Projects l
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Fort St. Vrain Nuclear Station
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ATTN:
C. Fuller, Manager, Nuclear Production Division 16805 WCR 191 Platteville, Colorado 80651
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Public Service Coinpany of Colorado-2-Fort St. Vrain Nuclear Station ATTN:
P. Tomlinson, Manager Quality Assurance Division 16805 WCR 19)
Platteville, Colorado 80651 Colorado Radiation Control Program Director Colorado Public Utilities Connission bectoDMB(IE01)
bec distrib, by RIV:
RRI R. D. Martin, RA Section Chief (DRP/B)
Project Engineer. DRP/B RPB-DRSS Lisa Shea, RM/ALF MIS System RSTS Operator K. Heitner, NRR Project Manager (MS:
13-D-18)
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P.O. Box 840
Denver, CO 80201 0840 16805 WCR 19 1/2, Platteville, Colorado 80651
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R.o. wulA448, JR.
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VICE PRESIDENT i
NUCLEAR OPERATIONS
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September 23, 1988 t
Fort St. Vrain
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Unit No I t
P-88341 l
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V. S. Nuclear Regulatory Commission
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Washington 0.C.
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t Docket No. 50-267 SUBJECT: NRC Inspection f
Report 88-13
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l REFERENCE: 1) NRC Letter, Callan i
to Williams dated I
August 19, 1958 l
(G-88335)
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i 2) PSC Letter, Williams to NRC dated
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September 15, 1988 (P-88337)
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Gentlemen:
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This letter is in response to the Notice of Violation received as a result of the inspection conducted by Messrs. R. E. Farrell and R. P. Mullikin durin9 June 1-30,1988 (Reference 1).
The following response to the items contained in the Notice of Violation is hereby submitted.
Identification of Violation:
Technical Specification 7.4.a requires that written protecures shall be established, imolementec, anc maintained for activities including surveillance testing.
Technical Specification 7.4.b requires that these procedures ce periccically reviewed.
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P-88341-2-September 20, 1988 Contrary to the above:
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The licensee failed to maintain Procedure SR-5.2.16 a-Q, Issue 34, "PCRV Closure Leakage Determination." cur rent with a Technical Specification change issued on March 18, 1982.
The out-of-date procedure resulted in the licensee erroneously placing themselves in the Technical Specification Limiting Conditions for Operation 4.2.9 action statement that could have resulted in a plant shutdown.
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The licensee failed to establish and implement procecuralized controls for the sheading of electrical DC loads following a turbine trip with a loss of offsite power in accordance with the plant Reference Design M*,nual, Station Batteries Design Criteria DC-92-1, 3.
Special Instruction to Operators No. 85-15 Issue 4, dated June 23, 1938, which was subsequently issued on an interim basis to establish procedural controls for shedding of DC loads, failed to require shedding of Computer Inverter N-9234 consistent with DC-92-1.
This is a
Severity Level IV violation, (Supplement I)(267/8813-01)
(1) The Reason For The Violation If Admitted:
The violation for items 1 and 2 identified above is admitted; the violation for item number 3 is not admitted.
Item 1 noted in the Notice of Violation was cue to the fact that calculations included in SR-5.2.16.a-Q had not been updated to reflect the Maren 18, 1982 change to Fort St. Vrain's Technical Specifications. This revision was intenced to accommodate an operating mode where the Loop II Steam Generator penetration could ce operated at a pressure slightly abova cold reheat pressure.
SR-5.2.16.a-Q,
"PCRV Closure Leakage Determination," is used to cetermine tne leakage rate from the Prestressed Concrete Reactor Vessel (PCRV) closure interscace.
The surveillance is performed on a quarterly basis or in response to an unanticipatec alarm for
- nterspace pressurization gas flow.
The methocniogy of SR-5.T.16.a-Q is a sequential process for determining if the leak rite is in violation of Tecnnical Specifications and for identifying the exact location and pathway of a leak.
If conservative acceptance criteria scecified in the test tre exceeced, the test conductor is directed to utilize
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P-88341-3-September 20, 1988 procedure SR-RE-151-X, "Penetration Interspace Decay Tests " for final assessment of the leak rate. Leakage Pressure SR-RE-151-X is the more definitive method for determining PCRV penetration leakage when the acceptance criteria of SR-5.2.16.a-Q have been exceeded.
When used, SR-RE-151-X is attached to the SR-5.2.16.a-Q being performed and becomes a part of that surveillance.
During the incident noted, it was determined that typing errors and errors in transposition had occurred during the preparation of SR-RE-151-X.
SR-RE-151-X had only been performed on one occasion previous to the noted inciden't.
SR-RE-151-X had been ceveloped in early 1987 as part of an overall revision to Station Betterment Engineering acministrative controls and implementing procedures.
SR-RE-151-X replaced a previously used procecure, primarily for the purpose of reformatting information.
During this teformatting, errors were
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introduced into the procedure which resulted in the overly conservative calculation and Action Statement of LCO 4.2.4.
unnecessarily entering the Item 2 noted in the Notice of Violation was due to the fact that Public Service Company of Colorado (PSC) has not developed an operating procedure to address a loss of all AC pcwer event (Station Blackout).
Fort St.
Vrain's (FSV) Final Safety
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Section 8.2.3.4, states that either Station Battery 1A orAnalysis Report (F 1B is adequate to supply its service lead requirements, as defined in Station Battery Design Criteria DC-92-1, to a minimum voltage of 105 Volts DC for a four hrur discharge.
The FSAR Section 8.2.3.4 further states "Battery 1A or IB is adequate to supply the required shutdown DC leads for four hours following a loss of all AC power."
Station Battery Design Criteria DC-92-1, Table 1 (attached),
assumes that several loads, which are n e, t automatically shed from the DC buses, are manually shed uen their function is no longer required.
If these loads were not shed at or before the times assumed in the load profile analysis during a loss of all AC power. Station Batteries 1A anc 1B may not be capable of providing an adequate power supply to the loads which are required to function for four hours. These loacs are assumed to function in CC-92-1.
A Loss of All AC Power emergency proteoure would be tne logical cocument to direct manual sheoding of the DC loads. At present, Fort St.
Vrain's Emergency Procedures address Loss of Outside Power anc Turbine Trip (Emergen:y Procedure EP F-3), anc Loss of Outside Power and Turoine Trip with Failure of One Stancey Diesel Generator to Start (Emergency Procecure EP F-4).
The FSAR,
Section 10.3.2, assesses the latter event.
Fort St. Vrain's
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P-58341-4-September 20, 1988 Emergency Procedures do not address, nor does the FSAR assess, loss of all AC power event.
a PSC has not previously developed an emergency procecure for a loss of all AC power event as it was not considered to be a credible event in the FSAR. PSC submitted Amencment 5 to the F Preliminary Safety Analysis Report (PSAR)
in a letter cated October 13, 1967 Question IX.!
from the Atomic Energy Commission's (AEC) Director of the Division of Reactor addressed in that letter.
Licensing was PSC to "Analyze the consequences ofQuestion IX I asked, in part, for a total loss of electric power, including the standby diesel generators but not the station battery." PSC responded that "the total loss of electric power is not consicered credible, the reasoning for this being summarized in Section A below."
PSC then proceeded to discuss the effects of a loss of all AC power on reactor plant ecuipment.
PSC recognizes that 10CFR50.63, issued June 21, 1938, requires nuclear power plant licensees to develop wreating procedures for station blackout events for specified durations.
Although 10CFR50.63 is not specifically applicable to FSV, PSC has committed to develop an interim procedure for station blackout (Reference 2).
Tais procecure will address the requirements for manual shedding of DC loads during an interval when AC power is not available.
The procedure is discussed further in the correctiva steps which will be taken portion of this response.
FSV's current Emergency Precedures do not address, for the most part, events which are not considered to be credible. PSC is in the process of ceveloping new, symptom oriented, Emergency Operating Prececures (EOP) which will be implemented prior to startup following the fourth refueling.
The new E0P's are being developed in accordance with the requirements of Supplement 1 of NUREG 0737.
These procedures will address credible as well as non-credible emergency events at FSV.
PSC will ensure that those accidents assessed in the FSAR, which are not considered to ce credible, will be addressec by the new E0P's.
Exception is taken to item 3 noted in the Notice of Violation.
Special Instruction (SI) to tre Operators No. 85-15 incorporated the load shedding requirements of page 8.2.9 of Change Notice (CN) 2672, "Replacement of Batteries IA,1B and 1C ano recuired shedding of Co puter Inverter N-9234, included in CN-2672 were provisions for revising OC-92-1, which revised tne time requirement for securing tne CC loao for Ccmouter Inverter N-9234 from 30 minutes to two nours.
CN-2672 was implerentec in accordance with administrative controls which govern the design change process at FS _ _ _ _ _ _ _ _ _ _ _ _ _ _
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r P-88341-5-September 20, 1988 (2) The Corrective Steps Which Have Geen Taken And The Results Achieveo:
For item 1,
Non-Conformance Report (NCR)88-172 was initiated against SR-RE-151-X to recheck the calculations for the leak rate.
SR-5.2.16.a-Q and SR-RE-151-X were reviewed for validity and revisions to the procedures were initiated.
Previous documentation of completed SR-RE-151-X was reviewed, recalculated and found to be within Technical Specification Limits.
NCR 85-172 was closed.
For item 2,
SI 85-15, Issue 4, was issued effective June 23, i
1938. This SI provided interim instructions for the operating staff to secure the necessary loads. The development of a new p
Emergency Procecure to address a loss of all AC power
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also initiated at this time.
event was No corrective actions are necessary for item 3.
(3) The Corrective Steps Which Will Be Taken To Avoid Further Violations:
For item 1,
errors identified in SR-RE-151-X and SR-5.2.16.a-Q l
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Calculations included in SR-5.2.16.a-Q will be updated to reflect the current Technical Specifications.
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l For item 2, corrective steps have been initiated in several areas to avoid further violations.
An interim procedure, addressing plant operations with only DC power available for up to one hour,
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is being prepared. An Engineering Evaluation, analyzing the DC t
load profiles required to support plant operations for one hour l
without AC power being available, is being prepared.
A letter,
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from R. O. Williams to the NRC dated Septecer 15, 1988 i
(Reference 2), has been submitted which describes PSC's l
justification for operating with only DC power available for up to one hour if AC power is unavailable.
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The station battery load profile technical specification surveillance will be revised to
reflect a one hour battery load profile.
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l No additional corrective steps will be taken for item 3.
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(4) The Date Wnen Full Compliance Will Be Achieved:
For item 1 full corr 911ance will os achieved following tne ucdating of SR-5.2.16.a-0.
The revised procecure will ce j
i inolemented by October 4,1938.
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P-88341-6-September 20, 1988 For item 2 full compliance will be achieved following the issuance of an interim procedure which addresses plant operations with only 0; power available.
The new procedure will be issued prior to restart following the current circulater outage at FSV.
Should you have any further questions, please contact Mr. M. H.
Holmes at (303) 430-6960.
$1ncerely, anP R. O. Williams, Jr.
Vice President, Nuclear Operations ROW:0LW/dje Aegional Acministrator, Region IV ec:
ATTN: Mr. T. F. Westerman, Chief Projects Section 8 Mr. Robert Farrell Senior Resident Inspector Fort St. Vrain