IR 05000267/1988013

Ack Receipt of Informing NRC of Steps Taken to Correct Violations Noted in Insp Rept 50-267/88-13

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)

Text

_ ___ __ . _ _ . _ _ _ _ _ - _ _ _ - - _ _

. P L 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 neede 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 procedure We have noted your exception to Item 3 of the violation and agree with your respons Please provide the supplemental information within 30 days of the date of this lette Sincerel . Original SV ' SP U,3; CALLAN i

L. J. Callan Directo*

! Division of Reactor Projects l cc:

'

Fort St. Vrain Nuclear Station ATTN: C. Fuller, Manager, Nuclear Production Division 16805 WCR 191 Platteville, Colorado 80651

/

• C:PSB 0:DR *RIV:PSB RPMullikin;df 11/ /88 TFWesterman 11/ /68 LJCallan 11 /88

/h ,

• Previously concurred bbk$$.bbb$$7

< PDC

_ _ _

• e

'

.

. .

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) DRS RIV File DRP

-- ------ ---- _ _- _-

,- ,

!

.

.. .

,, ' * .

[

hPubl c Service ~ =%  :

P.O. Box 840 :

Denver, CO 80201 0840 '

16805 WCR 19 1/2, Platteville, Colorado 80651

!

R.o. wulA448, J '

,

VICE PRESIDENT i

'

NUCLEAR OPERATIONS September 23, 1988 t Fort St. Vrain Unit No I {

t P-88341 l

'

{N3{ .

'

V. S. Nuclear Regulatory Commission

f I ATTN: Document Control Oesk Washington .

E 2 IE j

[ y ;

.

t Docket No. 50-267 SUBJECT: NRC Inspection f

'

Report 88-13 '

l REFERENCE: 1) NRC Letter, Callan i to Williams dated I August 19, 1958 l (G-88335) i

i 2) PSC Letter, Williams to NRC dated }

September 15, 1988

' (P-88337)

t Gentlemen:

[

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 submitte Identification of Violation:

Technical Specification 7.4.a requires that written protecures shall be established, imolementec, anc maintained for activities including surveillance testin Technical Specification 7.4.b requires that these procedures ce periccically reviewe U lb G U() Q h k'

F

. ,

. .

.' ,

,

P-88341 -2- September 20, 1988 Contrary to the above: 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, 198 The out-of-date procedure resulted in the erroneously placing licensee themselves in the Technical Specification Limiting Conditions for Operation 4.2.9 action statement that could have resulted in a plant shutdow . The licensee failed to establish and implement procecuralized loads controls for the sheading of electrical DC in following a turbine trip with a loss of offsite power accordance with the plant Reference Design M*,nual, Station Batteries Design Criteria DC-92-1, 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- 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 admitte 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 pressur operated at a pressure slightly abova cold reheat 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 interscac The surveillance is performed on a quarterly basis or in response to an unanticipatec alarm for

• nterspace pressurization gas flo 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 of and for identifying the exact location and pathway a leak. If conservative acceptance criteria scecified in the test tre exceeced, the test conductor is directed to utilize

. .

.

'

,

.

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 leakage whendefinitive method for determining PCRV penetration exceede the acceptance criteria of SR-5.2.16.a-Q have been When used, SR-RE-151-X is attached to the SR-5.2.16.a-Q being performed and becomes a part of that surveillanc During the incident noted, it was determined that typing errors and errors in transposition had occurred during the preparation of SR-RE-151- SR-RE-151-X had only been performed on one occasion previous to the noted inciden' SR-RE-151-X had been ceveloped in early 1987 as part of an overall revision to Station Betterment Engineering acministrative controls and implementing procedure SR-RE-151-X replaced a previously used procecure, primarily for the purpose of reformatting information. During this teformatting, errors were .

introduced into the procedure which overly conservative calculation and resulted in the unnecessarily entering the Action Statement of LCO 4. 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 Section 8.2.3.4, states that either Station Battery 1A orAnalysis adequate to 1B is Report (F 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. further states "Battery 1A or IB is adequate to supply the required AC power."shutdown DC leads for four hours following a loss of all (attached), Station assumes Battery Design Criteria DC-92-1, Table 1 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 which are required of providing an adequate power supply to the loads to function for four hours. These loacs are assumed to function in CC-92- A Loss of All AC Power emergency proteoure would be tne logical cocument Fort St. toVrain's direct manual sheoding of the DC loads. At present, 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 4 Generator Section 10.3.2, to Start (Emergency Procecure EP F-4). The FSAR, assesses the latter event. Fort St. Vrain's

_

____

.

. ,

,

.

-

.

,

P-58341 -4- September 20, 1988 Emergency loss of all ACProcedures power event.do not address, nor does the FSAR assess, 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 FSA PSC Preliminary Safety Analysis Report (PSAR) submitted Amencment 5 to the FS October 13, 1967 in a letter cated Question IX.! from the Atomic Energy Commission's was (AEC) Director of the Division of Reactor Licensing addressed in that lette PSC to "Analyze the consequences ofQuestion IX I asked, in part, for power, a total loss of electric including the standby diesel generators but not the stationis power battery."

not PSC responded that "the total loss of electric 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 ecuipmen PSC recognizes that 10CFR50.63, issued June 21, 1938, requires nuclear station power plant licensees to develop wreating procedures for blackout events for specified duration 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 availabl The procedure is discussed further in the correctiva steps which will be taken portion of this respons 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 073 These procedures will address credible as well as non-credible emergency events at FS PSC will ensure that those accidents assessed in the FSAR, which are not considered to ce credible, will be addressec by the new E0P' Exception is taken Special Instruction (SI)totoitem 3 noted in N tre Operators the Notice 85-15 ofincorporated Violatio 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, provisions for revising OC-92-1, whichincluded in CN-2672 were revised tne time requirement for securing tne CC loao for Ccmouter Inverter N-9234 from 30 minutes to two nour CN-2672 was implerentec in accordance with change process at FS administrative controls which govern the design

_ _ _ _ _ _ _ _ _ _ _ _ _ _

. . .

. ,

,

• . .

,

r P-88341 -5- September 20, 1988 (2) The Corrective Steps Which Have Geen Taken And The Results Achieveo: 1 For item 1, Non-Conformance Report (NCR)88-172 was initiated against rat SR-RE-151-X to recheck the calculations for the leak SR-5.2.16.a-Q and SR-RE-151-X were reviewed for validity and revisions to the procedures were initiate Previous documentation of completed SR-RE-151-X was reviewed, recalculated and found to be within Technical Specification NCR 85-172 was close Limit For item 2, SI 85-15, Issue 4, was issued effective June 23, 1938. This SI provided interim instructions for the operating i staff to secure the necessary loads. The development of a new p

'

Emergency Procecure to address a loss of all AC power event was also initiated at this tim No corrective actions are necessary for item (3) The Corrective Steps Which Will Be Taken To Avoid Further Violations:

For item 1, i errors identified in SR-RE-151-X and SR-5.2.16.a-Q will be correcte Calculations included in SR-5.2.16.a-Q will l, be updated to reflect the current Technical Specification 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, is being prepared. An Engineering Evaluation, analyzing the DC .

load profiles t l required to support plant operations for one hour 1

'

without AC power being available, is being prepare A letter, from R. O. Williams (Reference 2), has been to submitted the NRC dated Septecer 15, 1988 i justification for which describes PSC's operating with only DC power available for up l to one hour if AC power is unavailable. The station battery load <

profile technical specification surveillance will be revised to  ;

reflect a one hour battery load profil i

"

l No additional corrective steps will be taken for item '

(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- The revised procecure will ce i j

inolemented by October 4,193 t

1 T

, *. . i

, - i

, ,

• *

.

w I

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 FS Should you have any further questions, please contact Mr. M. Holmes at (303) 430-696 $1ncerely, anP R. O. Williams, J Vice President, Nuclear Operations ROW:0LW/dje ec: Aegional Acministrator, Region IV ATTN: Mr. T. F. Westerman, Chief Projects Section 8 Mr. Robert Farrell Senior Resident Inspector Fort St. Vrain