ML20080S930

Affidavit of Ck Mccoy Re Events Surrounding Util Statements to NRC in Respecting Plant DG Instrument Air Quality

ML20080S930
Person / Time
Site:	Vogtle
Issue date:	03/03/1995
From:	Mccoy C GEORGIA POWER CO.
To:
Shared Package
ML20080S898	List: ML20080S895 ML20080S915 ML20080S930
References
93-671-OLA-3, OLA-3, NUDOCS 9503130104
Download: ML20080S930 (195)

Text

{{#Wiki_filter:c i i UNITED STATES OF AMERICA ' NUCLEAR REGULATORY COMMISSION BEFORE.THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of a Docket Mos. 50-424-OLA-3 GEORGIA POWER COMPANY, 31 31 t 50-425-OLA-3 t Re: License Amendment r (Vogtle Electric Generating' Plant, s .(Transfer to Units 1 and 2) southern Nuclear) t ASLBP NO. 93-671-OLA-3 AFFIDAVIT OF C. KENNETH McCOY Personally appeared before the undersigned attesting officer, duly authorized to administer oaths in the state and county aforesaid, C. Kenneth McCoy, who, being duly sworn, states and deposes on oath as follows: 1. My name is C. Kenneth McCoy. I am of legal majority, suffer from no disabilities and am otherwise competent to make this affidavit. All statements made herein are based on my own personal knowledge. This affidavit is offered in support of " Georgia Power Company's Motion for Summary Disposition of Intervenor's Air Quality Statements Allegation." 2. I am employed by Georgia Power Company (" Georgia Power" or " Company") as Vice President.- Vogtle Project. I was elected to that position by the Georgia Power Board of Directors on May 18, 1988 and I still hold that position today. I also hold the position of Vice President - Vogtle Project of the Southern Nuclear 9503130104 950303~ PDR ADOCK 05000424 0 PDR;

, Operating Company (" Southern Nuclear") to'which I was elected by the Board of Directors of Southern Nuclear when it was incorporated in December 1990. -My business address is 40 Inverness Center Parkway, Birmingham, Alabama 35242. A summary of my professional qualifications is' attached hereto as Exhibit 1. This Affidavit describes the events surrounding Georgia Power's statements to the Nuclear Regulatory Commission ("NRC") in its April 9, 1990 letter respecting the Plant Vogtle diesel generator instrument air quality. a Summarv of Events Followina the March 20, 1990 Site Area Emeraency. 3. On March 20, 1990, a Plant Vogtle worker accidentally ' backed a truck into a switchyard support column causing a loss of 3 off site power to Unit 1. Georgia Power personnel attempted to restore power to the plant by starting the only available emergency diesel generator (1A). Twice the diesel started and ran for about' one minute before tripping. On the third attempt, the diesel l started and ran, restoring power to the plant 36 minutes after the initial loss of off site power. The 1B diesel was out of service at the time for a maintenance overhaul. 4. Following the March 20, 1990 site area emergency at Plant Vogtle, an NRC Augmented Inspection Team ("AIT"), including, i i among others, Messrs. Ken Brockman from Region II and Rick Kendall from NRC headquarters, arrived at Plant Vogtle on March 22, 1990. i I ) t

2 i t i ~ t

5. - On March 23, 1990,.the NRC issued'a Confirmation of

'i ..' Action (" COA") letter to Georgia Power which, among_other things, provided -that Georgia Power. was 'not to restart Vogtle Unit - 1 without NRC approval.- + 2 6. The AIT was replaced with an Incident Inspection Team l .("IIT") on March 25-26,'1990. Mr. Kendall. carried over from the i AIT-to work on the IIT. Mr..Brockman,_while not an IIT member,- became the NRC Region II_ Point of Contact for the IIT. Mr.-Ali Chaffee was.the.IIT team leader. 7. On March 27 and 28, 1990,' NRC inspector Milt' Hunt i witnessed special testing-for. a determination of diesel 1B [ ] operability and Mike Horton, the Vogtle Engineering' Support i Manager, said the NRC was happy with the testing result'.- Egg NRC ) s Inspection Report No. 90-05, dated April 26, 1990,-attached as Exhibit 2, report " Details" at 2, and handwritten note from Mike Horton to George Bockhold, the Vogtle General Manager (Exhibit 3). Milt Hunt was assisting the IIT in observation ~of the testing. He later returned to the site with NRC inspector Peter Taylor and. witnessed additional special testing.. Hunt Aff. 1 14.- ji 8. Air quality, including the possibility ' of small j . debris or moisture in the diesel air system,-was discussed _at an ( IIT meeting'on March 28, 1990. In response to a question from the IIT, Georgia Power committed to review the last historic dew point on the 1A diesel' prior to March 20, 1990, and, in addition, take f a new dew point readings. Both the IIT and Georgia Power' were -l i l l

r 1 a t.i i:i S s s' Lattempting to identify the cause of the 1A diesel spurious trips-on g q March 20,. 1990. Jing IIT ' Document.145, Tr. 95-97, attached as ~ t Exhibit'4. O 9.- Between March'28 and April 3, as a-l follow up to.the - IIT J request, Georgia Power tested the diese1~.. air system for moisture and conducted a review of the' control air filters.. Georgia Power' stated that, based upon tests done, the air quality'

was satisfactory, and was not consideredeto be the root cause of the 1A diesel trips'on March 20,'1990.

jigg IIT. Document 257,,Tr. 59-60, attached as. Exhibit.5. 10. .I participated in an April 4,l1990 meeting on diesel testing with - George Bockhold, Mike Horton, Skip ' Kitchens; '(the Vogtle Assistant. General Manager of Operations), and others which was taped by Mr. Mosbaugh. During the meeting, George Bockhold and ^ I said that NRC's Ken Brockman would be briefing the NRC Region II Administrator on Friday- (April 6) about releasing Georgia Power . from the Confirmation of Action letter. Tape ~No'.:32, attached.as Exhibit 6, TJ. 5-8. Later-that day, Messrs.-Mosbaugh and Horton-praised Engineering personnel for the work they did on the diesels-l i and how well they interfaced with the IIT. Mr. Mosbaugh-said'that .j i at some point he thought the IIT would be satisfied on the diesel issues sufficiently to release the hold'on startup. Isb. at 31-33. i 11. By April 6, reports.of higher than expected dew points were made to Georgia Power management and, in turn, to the IIT. Jing IIT Document 203, attached as Exhibit 7, Tr. 4. The IIT' l !j ' !.l l o

) ' team'leaderf ndicated that the IIT.may have.been informed'of the 3 I m' situation prior to the morning of. April. 6. ' In any event,.the - } Vogtle General: Manager explained that on. April 5 he had learned ~ that the dew point test results on March 29 were unsatisfactory for - the 1A' diesel. - He further stated that preliminary indications were --that the high readings were due to a bad dew point ~ses.or'instru-ment.I' The' basis for the General Manager's belief that the test instrumentation was suspect ' ncluded additional recent -" bad" i ~[ readings.F Representatives of the diesel generator vendor 't (Cooper):had been contacted to verify Georgia Power's belief that i any immediate problem associated with the controls-of-the-diesel i did not call into question the' operability of the engines. L& at j Tr. 5-7. A new dew point instrument or equivalent was being sought j on the morning of April'6. & at 7. I 12. Between April 6 and April 9, Georgia Power had. performed additional dew point readings. On April 9, NRC representatives were informed of the dew point readings obtained by new instrumentation. One dew. point reading at 60.9

• F on the j

i l' As a precaution, a bleed and feed on the air storage tanks- ~ ..had been started to lower the air's dew point. 1 F These readings, which were among those-taken with three j -different instruments, indicated that - the dew points of eight j separate air systems were either out of specification high or were physically impossible readings. As a result, engineering personnel reasonably concluded that the readings were not ' accurate. 133 Georgia Power Company's Response'to Intervenor's Seventh-Request. j for Interrogatories - (August 8, 1994) at 4-5. The readings were i recorded by hand on a single sheet of paper, which was shared with-NRC inspector Hunt. Hunt Aff. 1 27, Exhibit 4. i i f a,. ~ I =

i i . y !!l ^ i i diesel: 2A = receiver,1was Lattributed to the associated' air dryer ~ . b e i n g -- t u r n e d off on Friday (April 6).. Egg.IIT. Document ' 206, ~ attached'as Exhibit'8,:Tr. 4-5.. The IIT team leader indicated to Plant ~ Vogtle. personnel t h a t ' U n i t _- 2 d i e s e l - r e l a t e d air. quality' history, was 'not of substantial interest; "we just need the. i o information that shows us to what extent air poor-(sic) qualityj l might have had ' an. impact' on the operation of unit 1A' diesel." &~- . at 6. Plant ' Vogtle personnel had also inspected control air a filters in March ~and the NRC was' informed that the filters looked I new, and did not appear to have been subjected to " dirty" air.. &. at 9. s 13. 'Also, on April 9, based upon a review of historic preventative maintenance ("PM") documentation,- the NRC was informed of PM results which showed unacceptable dew points. Egg IIT Document 206 (Exhibit 8) at Tr. 7-8. Georgia. Power offered the actual numbers from the PM packages, including the-1B-train diesel package from March 1989 when the dryer was replaced. & at 9. The NRC requested a table of historic measurements. & at 7. 14. Early on April 10, 1990 dew point measurements on .j the 1A diesel were reportedly telefaxed to the NRC. Egg IIT Document 05-202-90, attached as Exhibit 9 (the facsimile copy bears information indicating.it was transt " ted on April 11); gan also IIT Document 233, attached as Exhibi',10, Tr. 6. { 15. On April 10, 1990, Mr. Mosbaugh provided a note to George Bockhold which stated, on the basis of an April 10

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b-gj $3'- p., m -.g gi ~ .-r-a.- a w -m e -e e

= memorandum from Georgia Power engineer Tim Steele, that ' the diesels' dew point control and air-quality have D9.t always been l satisfactory. Egg Exhibit 11. ' Tim Steele's memorandum, for the j . most part, discussed the maintenance history (prior to mid-1989) of diesel air dryers. The memorandum also noted that on March 9,1990 and March 31, 1990,2/ dew point measurements were high for both air i dryers of the Unit 1A diesel, but that it.was. strange that both t - dryers would be high at the same time because they were in separate - systems. Because of this, Mr. Steele suggested that the method of j dew point measurement was suspect and might need investigation. A table of dew point measurements for the Unit-1A diesel was attached to Mr. Steele's memorandum. 16. During an internal meeting of Plant Vogtle personnel on the morning of April 11 which was taped by Mr. Mosbaugh, the Vogtle General Manager questioned a number of engineers on his staff, including Ken Stokes, Paul Kochery, Paul Burwinkel, and Tim Steele about the accuracy of the air quality statement in the April 9 letter. The General Manager confirmed that the reference to " initial reports" in the April 9 letter specifically referred to the work order (MWO 19001513) associated with instruments later i l' These were the same readings which Mr. Bockhold told the NRC I about on April 6. (Ess 1 11 above). Mr. - Steele's memorandum i referred to March 31, 1990 as the date the readings were taken, however, the readings were actually taken on March 29 as reported in Maintenance Work Order ("MWO") 19001513, attached as Exhibit 12. Mr. Steele's confusion regarding the date of the readings likely stems from the fact that the final signature indicating that the j work was performed was not executed until March 31. l w , =

no a. l .e m ' determined to. be faulty. He; explained that the NRC had'been told- .about the air receiver inspection, that air filters were. clean, that air receiver blowdowns showed no.significant water discharge,. and that the Cooper representatives' opinion was that air. quality at Vogtle was not a problem.- He specifically-asked.whether the-period of May 1988 to May 1989, when one'or more dryers may have been out of service,. would affect the air quality statement. in thel l April 9 letter. There was a consensus that the dew point acceptance criteria of 50 'F had not always been met in the past, but that based on their engineering judgment the air quality was acceptable. This conclusion was based on the factors discussed a above as well as the expected thirty-degree dew point depression resulting from the change in system pressure from about 250 psi to q l 60 psi. At the end of the meeting, the General Manager told the participants that he intended to inform the IIT personnel-that the preventative maintenance program in_1988 was "not as good" as the current program but that, - based on engineering. judgment, they: f 4 believed they had satisfactory air quality. Rag-Tape No. 41, Tr. l s 40-48, attached as Exhibit 13.F { 17. During a discussion with the IIT on April 11, Plant f Vogtle personnel informed the NRC that they could not obtain good, -{ F Georgia Power does'not believe that the transcript version j of Tape 41, which was reviewed and hand-marked by Intervenor and attached to the OI report as Exhibit.66, is completely accurate.- However, for purposes of this motion for summary disposition only, j ! Georgia Power has included' that version of the transcript as Exhibit 13 hereto. consistent data earlier than the data they had transmitted to the ~ NRC (Exhibit 9). Georgia Power's belief remained as it had been on April 9 that the current air quality of the diesel was satisfactory, although "during that period of time" in 1988, one of the air dryers was out of service for maintenance. The Company also explained that daily blowdown of air receivers helped assure t freedom from moisture in the control system air and that inspection of the control air filters at each overhaul period indicated no rust or corrosion products. NRC personnel were further informed that the April 6 inspection of one air receiver showed slight corrosion around welds and a minor oil film on the bottom of the receiver, none of which affected the control air quality. S.gg IIT Document 233 (Exhibit 10) at Tr. 6-7. 18. The IIT completed their report of the March 20, 1990 Site Area Emergency in June 1990. jileg Loss of. Vital AC Power and the Residual Heat Removal System during Mid-loop Operations at Vogtle Unit 1 on March 20, 1990 ("NUREG-1410"). The report, at S 3.2.2, describes the starting and control air system design and operation and states, at p. 3-10: The dew point has generally been kept at close to 40

• F.

The dryers on occasion have been out of service for short periods;

however, no evidence has been found of significant moisture or its effects in the instrument air lines or sensors.

The 5-micron filter has always been clean when replaced; no significant amount of contaminants have been found in the instrument air system. Thus, air quality does not appear to have been a factor in the emergency diesel generator response during the incident.

.. *W 1 ]g q y I D19. On April'112, 1990, --NRC Region II released Georgia + Power to restart Unit 1. On April 15,1990, Unit 1 entered Mode 2. - The Anril 9; 1990 Letter and NRC's-Anoroval to Restart Unit 1. I 20. , Parallel to.the preparation-for the April 9, 1990 i meeting with the - NRC, Georgia Power : prepared.a " Status -of: Corrective ' Action"- write-up-which addressed the.' site Area' I Emergency. By April 5, early in the morning, a draft had been j prepared by site'personnelJ(attached as Exhibit 14) which included .j the following language: The following actions are being implemented to ensure a. f - high state of diesel reliability.... '5. Since March 20,: l 1990, GPC has performed numerous. sensor calibrations' (including jacket water temperatures),.' extensive logic. l testing, special pneumatic leak testing'and air quality. reverification, and multiple engine starts and runs under various conditions.... 's On April 5th and 6th, the company telefaxed copies of this document. S l to Ken Brockman (Region II, Atlanta) and David Matthews (NRR, Wash. J D.C.), respectively. On April 5, this document was also routed for 'l review to Plant Vogtle personnel, including Mr. Mosbaugh, and the NRC resident inspector.F 21. By early April 7,1990, the draft letter was revised i (133 attached Exhibit 15) and included the air quality statement which appears in the final signed letter: F i The upper right-hand corner of Exhibit 14 contains a list of distributees, including the NRC Residet.t Inspector denoted by "NRC Re(s)." i

n. "

p I[_. i y T %] GPC 'has i reviewed airl quality l of, the. D/G air ~ system including dewpoint - control and.has. concluded that a :- 4 4 quality.is satisfactory.. Initial reports-of higher than' expected dowpoints. were later attributed ;to faulty j w . instrumentation. .This was -confirmed-by internal" j

inspection of one air receiver on April 6,a1990, periodicL l

replacement of,the control air' filters'which showed no' a ? indication of corrosion and daily air receiver blowdowns l .with no significant water discharge. 1 y 22. The final, signed-April 9, 1990 letter,jttached as l Exhibitk 16,. discussed only the current.. status of ' the. diesels' ' instrument air quality; it was not intended to describe all-past' The' letter conveyed Georgia Power's judgment. - maintenance issues. on April.9, 1990, that.the diesel control air ~ quality relative to moisture or " humidity" was satisfactory at. that time. Although higher than expected dew points had, :in fact, been recorded during. 'i the Plant's recovery from the March 20 Site Area Emergency (i.e., I f the "initia1 reports"), these post-event measurements: were erroneous, due to faulty instrumentation.' ? i NRC Staff Review of the April 9 Air Ouality" Statement. !l 23. During the August -1990 '.NRC Operational Safety. ] Inspection,F the NRC inspectors reviewed documentation associated I J with'.the Unit i diesel' starting air system. They noted that a j mejority of the dew point measurements taken were within .y specifications and that the reasons for out of specification f i W This ' inspection was conducted, in part, as the result of i allegations submitted to the NRC by Mr. Mosbaugh,:which included an allegation that the April'9 air quality statement was false. r r \\ I

.-m n- ' lj,- p f, y 'readi' gs included problems.with measurement instruments, air dryers - n T-tbeing-out'~of'. service.for : extended periods.of.. time, and~ repressurization of, the air system following, maintenance. The t. inspectors reviewed' records associated with the inspection of.the-air' receiver, inspection and replacement-of an. air filter,. and a replacement of: the' dew' point measurement instrument. The ' inspectors spoke with system engineers.who said there was no evidence of internal moisture or corrosion. in :.the control air D sensing lines when-disconnected for maintenance troubleshooting. The inspection team concluded that Georgia ' Power did have - an adequate basis ' to assess the. quality of the diesel starting air: - system, which was primarily the visual' _ inspection of. the system : components for degradation. Thus,, contrary to' Intervenor's allegation, they did not find that Georgia Power misrepresented the diesel air quality in the April 9 letter.. 333 NRC: -Inspection ' Report No. 90-19, Supplement 1, attached as Exhibit.17,-at 18-19. 24. The NRC Staff performed another inspection of'the L .vogtle diesel starting air system in 1994. NRC inspectors reviewed diesel maintenance history records to determine if out-of-tolerance dew point conditions resulted in detectable water formation or ~- adverse operation of the control air system. They specifically reviewed maintenance work orders related to the March 20, 1990 1A diesel-failure. They concluded that the maintenance documentation provided no indication that water ha'd been detected in the control and protection portion of the system at any ' time. This was.

a m I ,L -(' r e confirmed. L by discussions.with craft and engineering. personnel s ' involved'in.the 1990 troubleshooting; activities. The inspectors further' concluded that water formation in the air system was - unlikely because of'the decrease in dew point associated'with the pressure reduction in' the system-from 250 psi to 60 pai. They i -found. that-even with the highest-dew. point conditions 'ever measured, the probability of condensation within' the 60' psi system air was' not significant. Egg NRC Inspection -Report ' No. 94-12, attached as Exhibit 18, at 8-9. \\ 25. On May 9, 1994, the NRC cited Georgia. Power :for violation of - 10 C.F.R. S 50.9, finding that Georgia Power - had [ provided incomplete information'about the diesel air quality in its April 9, 1990 letter to NRC.2' Georgia Power submitted a Reply tof .the NOV on July 31, 1994. Following a review of Georgia Power's Reply, the NRC withdrew the violation of incomplete air. quality information on the basis that the air quality statements in Georgia Power's April 9 letter "were sufficient in scope' and GPC had an adequate technical basis to support a finding that' air quality was acceptable." Modified Notice of Violation and Proposed Imposition of Civil Penalties, dated-February 13, 1995,-attached as Exhibit 19, Appendix at 2. f 2/ The NOV stated that the Company failed to state that actual-high dew points had occurred and that the causes of those high dew-points included. failure to'use air dryers for extended periods of time and repressurization of the diesel air. receivers following. l -maintenance. Egg' Notice of Violation and Proposed Imposition-of' Civil Penalties (May 9, 1994) (the "NOV") at 3-4. I

-~ g M 'd C. Kenneth McCoy , /)1slore;pt this $f^pribed Sworn to and subs 3 , day of N'NQcQ.13!'5. 9 / .b- .9l O, \\ ~ i W,g.;.uygue v- 'Mtcog[gnexpires: } l l.

'~ ~~ R g Statement of Qualifications I 4 of..- KEN MCCOY w Ken McCoy is Vice President - Nuclear of Georgia Power'Companyl and is responsible for the oversight of Alvin W.-Vogtle Nuclear Plant..He- - also serves as Vice President - Vogtle Project of Southern Nuclear Operating i . Company. McCoy served in _the United States Navy from 1964.- 1974 where he J held positions as a Nuclear Submarine Officer, Instructor, Construction' . Engineer Officer and Squadron Engineer Officer. - In 1974 McCoy went to. work for Mississippi Power and Light - - Company at the Grand Gulf Nuclear Station. While at Grand Gulf, McCoy served in various positions including Engineer, Assistant to the Plant Manager, Plant Manager, Assistant to the Senior Vice President and Imaned Employee to t the Institute of Nuclear Power Operations. ~ In 1985, McCoy joined the permanent staff of the Institute 'of Nuclear Power Operations holding various positions including Division Director, Plant Operations Division. Mr. McCoy was employed by Georgia Power Company in 1988 as Vice President - Vogtle Project. i A native of Mississippi, McCoy received his bachelor's degne in Electrical Engineering from the University of Mississippi in 1%5. He received an MBA from Mississippi College in 1977. 1; I 9 L i i ww -, .4 ...w,...

)f ' ' ~ 7_ I 'l i /pg aseg* - usetT80 STATES. o** NUCLEAR REGULATORY COMMisslON Qd Lss y Y-S s t nooiO= u - 101 MARIETTA STREET, N.W. ATLANTA, GEORGI A 30323 \\.....- - APR 2 61990 ~ t . Docket Nost 50-424, 50-425 Licerse Nos. NPF-68, NPF;81 4 Georgla Power Company I -ATTN: Mr. W. G. Hairston, III Senior Vice President:- Nuclear Operations. P. 0. Box 1295 ' Birmingham, AL. 35201 l Gentlemen: I

SUBJECT:

NOTICE OF VIOLATION (INSPECTION REPORT N05.-50-424/90-05 AND 50-425/90-05) 'l This' refers to the Nuclear Regulatory Commission (NRC) inspection conducted by i Messrs. R. F. Aiello and R. D. Starkey. on February 17 - March 30,1990 The i inspection' included a review of activities authorized for your. Vogtle. facility.. At the conclusion of the. inspection, the findings were discus;ed with those i members of your staff identified in the enclosed inspection report. Areas examined during the inspection are identified in the report..Within these areas, the inspection consisted of selective examinations of procedures -and representative records,- interviews with personnel, and. observation of activities in progress. i The inspection findings indicate that certain activities. violated NRC 1 requirements.. The violation, references to pertinent requirements, and elements to be included in your response are presented in the enclosed Notice I of Violation. i i In accordance with 10 CFR 2.790(a), a copy of this letter and the enclosure will be placed in the NRC Public Document Room. The response directed by this letter and the enclosures are not subject to the clearance procedures of the Office of Management and Budget issued-under the Paperwork Reduction Act of 1980, PL 96-511. l eQ3 J N b a 'Ygv h d 0 W k g f. l 4 a e 4 c, %aase a

} p a: i 'V i Georgia Power Company 2-APR 2 61990 Should you have any questions concerning this. letter, please contact us., Sincerely, m'. Alan R. Herdt, Chief Reactor Projects Branch 3-Division of Reactor Projects

Enclosures:

1. Notice'of Violation 2. Inspection Report cc w/encis: R. P. Mcdonald 'f Executive Vice President-Nuclear Operations Georgia Power Company P. O. Box 1295 Birmingham, AL 35201 C. K. McCoy j Vice President-Nuclear Georgia' Power Company P. O.'1295 Birmingham, AL 35201 G. Bockhold, Jr. General Manager, Nuclear Operations . Georgia Power Company P. O. 1600 Waynesboro, GA' 30830 i J. A. Bailey i Manager-Licensing 4 Georgia Power Company i P. O. Box 1295 Birmingham, AL 35201 Ernest L. Blake, Esquire Shaw, Pittman, Potts and Trowbridge 2300 N Street, NW Washington, D. C. 20037 (_ cc w/encls cont'd - see page 3) i

g .i Georgia Power Company 3 ApR 2 6 1990 - i cc w/encls: (Cont'd) J..E. Joiner, Esquire .l Troutman,' Sanders Lockerman, and Ashmore i 1400 Candler Building 127 Peachtree Street, NE Atlanta, GA ' 30303 D; Kirkland..III, Counsel Office of the Consumer's Utility Council Suite 225, 32.Peachtree Street, NE Atlanta, GA 30302 t Office of Planning and Budget Room 615B 270 Washington Street, SW Atlarta, GA 30334 Office of the County Commissioner Burke County Commission Waynesboro, GA 30830 J. Leonard Ledbetter, Director Environmental Protection Division + Department of Natural Resources 205 Butler Street, SE, Suite 1252 Atlanta, GA 30334 Attorney General l Law Department 132 Judicial Building Atlanta, GA 30334 Star.e of Georgia I l l 3 \\ -1

o ,y-g unmisosvaras g' NUClaAR RaOULATORY C0hAR$19slON 4 t p asesom u g-tes essaierra stasar.m.w. 1* Avi.Anta.osomosa asass Report Nos.: 50-424/90-05 and 50-425/90-05 t , Licensee: Georgia Power Company P.O. Box 1295 Birmingham, AL 35201 Docket Nos.: 50-424'and 50-425 License Nos.: NPF-68 and NPF-81' Facility Name: Vogtle Nuclear Station Units 1 and 2 Inspection Conducted: February 17 - March 30, 1990 i Inspectors: M ry//dr/ 8/4/96 R.T. K1ello Actingfenior Resident Inspector Aate/51gned j WAnwuAJ Mska i " R.1.~5terkey, Re ent. Inspector . Matv51gned j Accompanied By: Milt Hunt and Leigh Trocine i Approved By: M r_d r - /.7C - 7O K.jt. Broctpfh,JGction Chief Date 51gned I Dfvision of ReaEtor Projects j SupetARY Scope: This routine _ inspection entailed resident inspection in the - following areas:. plant operations, radiological

controls, maintenance, surveillance, security, and quality progress and 1

administrative controls affecting quality, j Results: One cited violation and three non-cited violations were identified.. I The cited violation was in the area of operations for failure to. mechanically secure valve 1-1208-U4-176 during Mode 5 (Cold Shutdown). i as required by TS 3.4.1.4.2.C (paragraph 2.a). Two of the non-cited. 1 violations were in the area of operations for failure to properly - 1 review and approve - a revision to refueling procedure; 93271-C j (paragraph 3.b.(1)(g)) and failure' to incorporate adequate cautions a q in SSPS procedures regarding' simultaneous loss of both SRN!s when { placing both SRNIs in inhibit error inhibit (paragraph 2.a). The _i third non-cited violation was in the area of maintenance for failure of persons performing maintenance activities to notify QC as required by administrative procedure 00201-C para holdpoints were reached (paragraph 3.b.(1)(graph 4.5.2 when QC e)). One weakness was identified in the area of refueling-concerning ) inattention to detail. See paragraph 2.b.(8) for details. ~ t

I~ m ] e 9[ f_ -DETAILS.

1.. ' Persons Contacted Licensee Employees-

• J. Aufdenkampe, Manager Technical Support H
• G. Bockhold, Jr.. General Manager Nuclear Plant C. Coursey, Maintenance Superinten'ent-d
• G.= Frederick, Safety Audit and Engineering Group Supervisor j
• H. Handfinger, Manager Mainteunce
• W. Kitchens, Assistant General Manager Plant Operations j
• R. LeGrand, Manager Health Physics and Chemistry i

G. McCarley, Independent Safety Engineering Group Supervisor

• A. Mosbaugh, Assistant General Manager Plant Support R. Odom, Nuclear Safety and Compliance Manager '

j

• J. Swartzwelder, Manager Operations i

Other licensee employees contacted included technicians, supervisors, i engineers, operators, maintenance personnel, quality control inspectors, and office personnel.

• Attended Exit Interview

] An alphabetical list of acronyms and initialisms is located in th'e last paragraph of this inspection report. 2. Operational Safety Verification - (71707)(93702) The facility began this inspectio'n period with Unit 1 at' 961 power and 1 coasting down in preparation for 1R2 and Unit 2 at 1001 power. Unit 1: On February 23, 1990, at 5:55 p.m. EST, with the unit at 881 power, a NUE i was declared due.to the discovery by the licensee of missing core clamp bolts on seismically qualified switchgear and the subsequent deenergizing l of a contaimeent isolation valve (paragraph 3.b.(1)(c)).. To comply. with .i the TS action statement, the licensee began a shutdown of the unit and although the bolts were replaced and the CIV was. reenertized before the shutdown was completed, plant management elected to camp :ete the shutdown and enter into a planned refueling outage. The reactor was manually tripped from approximately' 151-powee on February 23, 1990, at 8:58 p.m. EST, and the unit entered refueling outage 1R2. On March 13, 1990, at approximately 12:00 a.m. EST, an ESF actuation occurred when the standby train of the Fuel Handling Building Post accident HVAC auto started. One train was already in service to support refueling activities. No alam of the actuation was received in the control room. The cause of the actuation is under investigation. )

n .. n, 3 x i 2~ z On March-20', 1990, at 9:20.a.m. EST, with the unit'in Mode 6'(Refueling).. d a truck backed into an insulator support forf the "A" Reserve Auxiliary Transformer subsequently causing a ~ loss.~of power to ~ the "A" 4160. VAC amergency_ bus._ Thirty-six' minutes later, on the thir6 start attempt, the ~1A DG was ~ started and. supplied power - to the "A" emergency bus. During. this event..the_."B" Reserve Auxiliary Transfomer and the 18 DG were down i for maintenance. Power was still beingl supplied to the non-vital' buses through the main-transformers backfeeding ' to ;the Unit. Auxiliary Transformers. The"B"emergencybus-wasbeing.fedfromthe"A"? RAT.throughanalternate' l supply breaker. _ When the undervoltage was sensed'at the A". emergency bus, DG.1A started and sequenced the loads to the "A" Bus. Eighty. seconds i after the DG output break _er closed, DG 1A tripped. DG 1A did not restart due' to a starting logic lock up_ which required the sequencer _ to 'be manually reset before a restart could be attempted. Operators were dispatched to DG'1A and' the sequencer. When the sequencer 1 was: reset the engine started and the required -loads sequenced onto the bus. After 70 seconds, the engine tripped again and did,not restart due to another starting logic lock up. Fifteen minutes after the second trip, the OG was started fro:a the engine control panel using the' emergency start push button. It was subsequently manually loaded and continued.to run until the "B" RAT was-energized to supply ~ power to the 4160 volt 1E bus. Because there was a. loss of power to both Unit 1 vital buses for more than - 15 minutes, a Site. Area Emergency was-declared at. 9:40 a.m., EST on March 20,1990.. On March 21,1990, ~an AIT from NRC was dispatched to the site to review the. events surrounding the SAE. 0n March 25, 1990, the AIT was upgraded to an IIT. The,IIT will_ issue a report.. NUREG-1410, upon t completion of their investigation. A region based inspector arrived on site on March 27,.1990, to assist the IIT in observation of the Unit 1 DG testing. The inspector witnessed the air leakage testing of the sensors for DG 18. ~ The purpose of this test was perfomed to verify the operability of the pneunttic controls.for _the engine. All tests-were successfully completed. A UV test was perfomed, the engine started, the loads sequenced onto the generator, and 'the - generator remained loaded to ensure that' the controls were functioning properly. An operational surveillance was then. performed and DG18-was. ' declared operational. This pemitted DG 1A to be removed from service for testing. The IIT requested that a UV test be perfomed on DG 1A to 'detemine the cause for its failure on March 20, 1990. The pneumatic logic for the engine control system was then reexamined by the licensee and representatives of the diesel manufacturer. This examination also included an air leakage test of the sensing elements which measure the various operating parameters of the engine. During this test, two jacket water temperature sensors were found to be either out of . =. a.

~ .{ .m' j n i, s l l n 3 i 1 l calibration or defective and were replaced..On March 20, 1990, following. restoration of "B" RAT, the licensec replaced the-three lube oil pressure-1 sensors' after finding one defective or out of calibration.. The engine was ~ l then successfully started and loaded three times. The logic'- testingL

witnessed by the: inspector; included five starts 'and.was concluded with" another UV start.

J The licensee 1then had the sensor. vendor. representative review the~ calfbration methods used by the licensee to detemine if the cause of-the { sensor failures was.due to either calibration practices or a problem with; i 'the sensors themselves. The licensee also contracted with an independenti testing firm to conduct test on the defective sensors. I Details.of the testing ' program and its' results will be incorporated in the - j IIT inspection report. 4 f Unit 2: 2 ~ On March 20,1990, _the unit tripped and entered Mode 3 (Hot Standby). i This was due to an electrical transient being sensed during the event on Unit 1 - (see L above). Troubleshooting and repairs continued for the following two days. The. final resolution of the trip was an improperly j set differential overcurrent. relay. On March 22, the unit entered Mode 2 (Startup), tied to the grid, and entered Mode 1 (Power Operations). The l unit remained at 1001 power until' the end of this inspection period, a. Control Room Activities Control Room tours and observations were performed to verify that i facility operations were being safely-conducted within regulatory-requirements. These inspections. consisted of one or more of,the following attributes as appropriate at the time of the inspection. - Proper Control Room staffing - Control Room access and operator behavior i - Adherence to approved procedures for activities in progress - Adherence to Technical Specification Limiting Conditions for a Operations - Observance of instruments and recorder traces of safety related and important to safety systems for abnormalities l - Review of annunciators alarmed and action in progress to correct l - Control Board walkdowns i - Safety parameter display and the plant safety monitoring system operability status - Discussions and interviews with the Shift Superintendent, i Shift Supervisor, Reactor Operators, and the Shift Technical Advisor (when stationed) to determine the plant status, plans, and to assess operator knowledge - Review of the operator logs, unit logs, and shift turnover sheets i .v

i

R a

t ~ l N a 4 1 i On February 26, 1990, while Unit I was in Mede 5 withiRCS level at-1 195' feet, 5 inches, the inspector discovered that RftfST discharge. valve, 1-1208-U4-176, was closed but was not mechanically secured, as required by TS 3.4.1.4.2.c. Instead of a chain ~and lock, the valve had ~ a clearance hold tag which provided only administrative control "N to preclude. valve operation. The-licensee stated that procedure 10019-C Control of Safety Related j Locked Valves, Rev. 5, step 5.1.4, permits use of a hold tag in cases a where.it is not feasible to physically lock an apparatus.. Valve i 1-1208-U4-176.has a small diameter, solid wheel type valve handle and i cannot be mechanically secured with a typical-chain and lock. However, the valve handle does have two siaall holes" drilled into it through which a wire or cable can be routed to secure the valve. Following notification that the valve was unsecured, the licensee ' i routed and crimped a steel cable through the drilled holes.which mechanically secured the valve as required by TS. The licensee was encouraged to reevaluate their locked valve program and determine if there ' are other required locked valves. that fit in this same. category. Failure to mechanically secure valve 1-1208-U4-176 is a violation of TS 3.4.1.4.2.c. This item is~ identified as: J 3 VIO 50~424/90-05-01, " Failure To Mechanically Secure ~Yalve l 1-1208-U4-176 During Mode 5 As Required By TS 3.4.1.4.2.c." i On March 22, 1990, with Unit 1 in Mode 5, during performance of procedure 24831-1, Reactor Trip to ESF Logic Response Time Test, the. I source range'NIs, N!-31 and NI-32, were rendered. inoperable when both trains of SSPS were selected to the Inhibit Error Inhibit position. i A similar event occurred on March 27, 1990, while. performing T-ENG-90-12, 8-Train Undervoltage Test. In both cases, the operators quickly identified the problem and the SRNIs were restored to service within approximately 30 seconds. Neither procedure 24831-1 nor i [ T-ENG-90-12 contained a caution to alert operators that placing both SSPS switches to Inhibit Error Inhibit would cause both NIs to be i inoperable. Furthermore, T-ENG-90-12 did. not contain any steps to i restore SSPS to its normal configuration. Failure to establish, implement and maintain an adequate engineering procedure for nuclear instrumentation is a violation of TS 6.7.1.a. The licensee has initiated corrective action by requiring that all SSPS procedures which use the Inhibit Error Inhibit switches be i reviewed for adequacy. A memorandum concerning these events was placed in the Operations Required Reading Book n the-control. room.- l Licensed operator requalification training on SSPS will be updated to reflect these events concerning SSPS. This licensee identified ~ violatiori is not being cited because criteria specified in Section j ^ V.G.1 of the NRC Enforcement Policy were satisfied. In order to track this item, the following is established. = 1

~ l 1 5 l NCY 50-424/90-05-02, " Failure To Incorporate Adequate. Cautions In SSPS Procedures Regarding Simultaneous Loss Of Both Source Range NIs When Placing Both SRNIs In Inhibit Error Inhibit." l b. Facil'ity Activities I Facility tours and. observations were performed to - assess the effectiveness of the administrative controls established by' direct observation -of plant activities, interviews -and discussions with. i licensee personnel, independent verification of safety systems.ttatus and LCOs, -licensee meetings and facility records.- During these inspections the following objectives were achieved: (1) Safety System Status (71710) Confirmation of system l operability was obtained by. verification that1flowpath valve alignment, control and power supply alignments, component i conditions, and support systems for the accessible portions of the ESF trains were proper. The inaccessible portions are i confirmed as availability permits. (2) Plant Housekeepine Conditions - Storage of material and 'l components and cleanliness conditions of various areas l throughout the facility were observed to determine whether 1 safety and/or fire hazards existed. i On March 15.-1990, an inspector toured the Unit I containment building with the Manager-Health Physics and Chemistry and.the Manager-Maintenance. Topies ~ of diseussion included { housekeeping HP practices, and maintenance activities. In p&rticular, the method by which HP will decontaminate the l containment pool when the pool water level is lowered in i preparation for reinstallation of the reactor vessel head was discussed. Also observed was the installation of the reactor vessel level sight gages which are to replace the existing tygon tube and will be used for reactor vessel indication in Mode 5 and Mode 6 during RCS drain down to mid-loop operation. The j Manager - Maintenance also answered questions concerning the .i snubber reduction effort and, in particular, the seismic snubbers which have been removed from the SGs during the current refueling outage. No deficiencies were noted by the inspector. (3) Fire Protectine - Fire protection activities, staffing, and equipment were observed to verify that fire brigade staffing was appropriate and that fire alarms, extinguishing equipment, ~ actuating controls, fire fighting equipment, emergency equipment, and fire barriers were operable. l l

7 '6 l 1 On February 22.1990, ' the inspectors observed an' announced ' fire drill. The simulated fire occurred in the Unit 2' AFW sump pump Fire team members responded quickly and appropriately room. J during the drill. 0ther plant. staff were on hand to assist the fire team in laying out hoses and staging other support l equipment. The inspectors noted that, as in previo;s fire drills, the fire team was not permitted to charge the fire hoses: i to simulate actual hose handling conditions. Consequently, the-hoses, once inside the butiding, were looped and bent-into positions which would.not have been possible if the hoses had' been fully charged with water. The inspectors were informed by the fire protection system engineer that plant management has

(

forbidden the charging of fire hoses during drills. ) -( Plant management has subsequently revised its position and has - directed that the training objectives of drills be rewritten to l include grading criteria to evaluate the fire team's placement and simulated charging of fire hoses. Additionally, a fire i drill scenario has been developed for use in one of the. site i support buildings which will include-actual charging of the i hoses. The licensee's. response adequately addressed the j inspector's concern. (4) Radiation Protection Radiation protection' activities, staffing, and equipment were observed to verify proper program implementation. The inspection included review of the plant program effectiveness. Radiation work pemits and personnel compliance were reviewed during the daily plant tours. Radiation Control Areas were observed to verify proper identification and implementation. (5) Security -- Security controls were ' observed to verify that j security barriers were intact, guard forces were on duty, and access to the Protected Atta was controlled in accordance with the facility security plan. Personnel were observed 'to. verify proper display of badges and that personnel' requiring escort were properly escorted. ' Personnel within Vital Areas were J observed to ensure proper authorization for the area. Equipment operability or proper compensatory activities were verified on a periodic basis. 1 After a recent housekeeping tour of Unit 1 and Unit 2 auxiliary buildings, the inspector' observed that signs posted on inactive a card readers can be confusing to the user and cause unnecessary phone calls to.either Security or Health Physics. For example, j the posting on a Unf t 1 charging pump room door inactive card i L reader stated, " Card Reader Lnoperable - Call Security / Health i j Physics." Numerous other inactive card readers had signs which i l. l I . ~..

~._ ~ _.. a ] o ] } q s; 7: ]

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l read,-" Card Reader Not In 'Use - Call,HP. (4016) for Access." The. Manager'. Health Physics and Chemistry stated _ that he was"not. j aware of a need or requirement for. these postings'. He ' stated j that he would work l with the Manager-Security ; to ' properly identify ~ inactive card readers.- As : a ' result of their initiative, inactive card readers. in the auxiliary buildings and control building have been reposted.with signs which.' simply.' state "NOT IN SERVICE." (6) Surveillance (61726)(61700)' - Surveillance tests were observed.. l to verify that approved procedures were being-used.. qualified _ .i personnel were conducting the tests; tests were adequate to l verify equipment operability, calibrated equipment was utilized, and TS requirements were. followed. The inspectors observed portions of the following surveillances and/or. reviewed. completed data against acceptance criteria: Surveillance No. Title i h 14805-1,:Rev. 9' RHR Pump And Check Valve IST 14825-2, Rev. 4 Quarterly Inservice Valve Test l 24805-1, Rev. 4 Steam Pressure Loop 4-(Protection IV) IP-546 ~ ACOT and Channel Calibration 24831-1, Rev. 5T Reactor Trip And ESF Logic Response Time Test-54065-1, Rev. 5 Train "B" DG And ESFAS Test T-ENG-90-11/12, Rev. 1/1 .A/8 Train Undervoltage Test On March 29, 1990, the Resident inspector examined the integrated leak ' rate test data acquisition process under the guidance of the Manager - Maintenance. The inspector noted that the electrical test equipment was supplied by non safety related-125V inverters and all process equipment (precision manometers and data acquisition. systems) were connected to mitigate single point failures from rendering the-ILRT invalid. The inspector had no further comments. (7) Maintenance Activities (62703) An inspector observed maintenance activities to verify that correct equipment clearances were in effect, work requests and fire prevention work permits, as required, were issued and being followed, quality control personnel were available for inspection activities as required, ratesting and return of systems to service was prompt and correct, and TS requirements were being

pg ? Es I i t y l 8 I 'i followed. lThe Maintenance Work Order backlog was reviewed.- 1 Maintenance was observed and/or work packages were reviewed for' the following maintenance activities: l MWO No. . Work Description-1 'I 18801635 Repair SG Blowdown HX Flange Leak And: Pressure Test Tubes For Leaks 18903586 Inspect Wom Gear For Casting Porosity 1 On 1-HV-11605 18905202 Perform Motor Control Center (MCC INBJ) 1 Maintenance 19000222 Installation Of DCP 89-VCN0115 Which 1 Installs And-Feeds Disconnect Switches l' In Containment 19000840 Main Steam Supply To TDAFW Pump HV-3019 Exceeded Its Maximum Stroke Time 19001511 DG 18 Calibration Of Lube 011 High 1' Temperature Trip Switch (8) Refueline Activities (60705) (60710):. New~ fuel receipt, core alterations, and fuel shuffle evolutions'were observed to verify. program' effectiveness, approved procedures were being used, and personnel were qualified._ The inspector-observed portions of. l the following evolutions: l 93300-C. Rev. 5, Conduct of Refueling Operations-1 93330-C, Rev. 4. Development and Implementation of the Fuel .l Shuffle Sequence Plan _ l 93010-C, Rev. 5. Unioading. Inspection' and Storage of New Fuel-l 93020-C, Rev. 4, Technical Inspection :f New Fuel While observing core alterations in the containment-building and fuel shuffling in the spent fuel pool, one weakness, inattention-to detail, was identified due to the following incidents: On March 2,1990, the fuel handling system transfer tube access plug clearance. required per procedure 93300-C. was not hung prior to spent fuel movement through the transfer tube. On March 3,1990, spent fuel storage rack location U-3 was damaged due to a misalignment while conducting core alterations. _._~

e m 9 On.' arch 4. -1990, fuel bund 1'e SC36 was Ioaded. into spent-M fuel pit location Y8 instead of. location.Y9. On. March 6,1990, ~ new fuel. assembly G-7, in lieu of G-5.. a was placed in the spent fuel pool by error. Prompt corrective action addressing these fuel. handling problems

was noted by the inspector. They. included the following

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All fuel handling crews were counseled on the importance of procedural comp iance associated with. fuel handling activities. Three of the individuals involved were removed from fuel handling activities. Additional Quality Assurance coverage has been added.

Four, hours coverage will be provided in each 12 hours.

Additional Supervisory surveillance has been added to. ensure procedural compliance. Additionally. Outage Management attention has been increased.. To reduce fatigue as is' contributing factor, shifts have been changed from 10, 10,'13, to three 9's. A fuel pool map will be made prior to. commencing fuel load to ensure the pool is in 'accordance ~with the shuffle sheets. As the fuel is transferred from the Fuel Handling Building to the Containment,- a serial number. check will be made while.the fuel is in.the upender as a final verification. The bundles associated with the bent fuel rack have been inspected by camera. Results were reviewed by onsite personnel and sent to the Westinghouse fuels group for review. 110 problems have been identified. The licensee's preparation and execution of placing the unit into mid-loop operation was accomplished in a safe and pre-planned manner. The inservice testing of the steam generators proceeded in an effective manner. The plugging of four tubes was indicative of good chemistry practices. Only one of these tubes actually. exceeded the plugging limit (40% of nominal tube wall thickness) and was required to be plugged. The other three tubes did not exceed the plugging limit, but were plugged as a precautionary measure. The Unit 1 snubber

~ H M. 10. O inspections went satisfactorily. Of.the 188-snubbers tested, i only 10 failed. All.10. failures were previous failures from 1R1 and did not' require a. scope. increase. Scheduling and coordination meetings were conducted on a frequent: basis 'with - appropriate levels of management in attendance. J On March 20, Unit.1 expe'rienced a loss of all'AC to the safety _ related 4.16kv buses which resulted in elevated temperatures occurring in the RCS while in a mid-loop status. See paragraph 2 for details. i (9) Calibration ~ (56700) - The inspector reviewed the. licensee's implementation of the Analog Channel Operational Test and 3 Calibration surveillance program to ensure conformance with license requirements, technical specifications. licensee. j' comitments, and industry guides and standards. The inspector examined selected surveillance procedures for' technical content, i verified that ~ calibration frequency met TS requirements. - reviewed completed surveillances, and witnessed the perfomance of two surveillances. The inspector also reviewed the'. licensee's program-for surveilling non-technical specification i components associated with safety-related systems or functions. A The licensee utilized the Surveillance Tracking System for tracking both the TS required surveillances and those non-technical specification surveillances associated with. safety-related systems or functions. j The inspector reviewed the following. surveillance procedures for technical content and verified that their calibration frequency meets TS requirements. The inspector also reviewed the most recently completed of each of these surveillances to verify that i the acceptance criteria had been. net, that the proper approved ~ test procedure had been used, and that procedural steps had been i signed off and all necessary values entered. 24493-1, Rev. 2 Pressurizer Level Control L-459 Channel l Calibration i 24571-1, Rev. 3 Containment Wide Range Pressure IP-10942' i Channel Calibration 24750-1, Rev. 4 SG Level (Narrow Range) Protection Channel i II, IL-519 Analog Channel Operability Test l And Channel Calibration i 24782-1, Rev. 8 Reactor Coolant Flow Loop 1 Protection Channel I, IF-414 Analog Channel Operability i Test And Channel Calibration

j t 11 l Eight additional procedures in the areas of reactor protection, ' ECCS, and plant auxiliary systems were reviewed to ensure TS. t required testing frequency was correctly stated. The inspector also observed performance of _ portions of the following 1 surveillances. 24805-1, Rev. 4 SteamPressureLoop4(ProtectionIV)1P-546 ACOT And Channel Calibration. i 24831-1, Rev. ST Reactor Trip And ESF_ Logic Response Time { Test i During these observations, the inspector questioned the technicians concerning their experience and qualifications and - was satisfied that they met industry standards. No violations or deviations were identified. 3. ReviewofLicenseeReports(90712)(90713)(92700) In-Office Review of Periodic and Special Reports a. This inspection consisted of reviewing the below listed reports to l determine whether the information reported by the licensee was i technically adequate and consistent with the inspector knowledge of l the material contained within the report. Selected material within i the reports was questioned randomly to verify accuracy and to provide i a reasonable assurance that other NRC personnel have an appropriate document for their activities. Monthly Operating Report - The report dated March '12,1990, was. reviewed. The inspector had no comments. Annual Report - The 1989 annual report dated Febrary 26. 1990, was I reviewed. Part 2 of this report will be submitted by May 1,1990.' The inspector had no consents. Special Report - The following special reports were reviewed. t (a) 1-90-02, "SG Tubes Plugged During 1R2." This special report dated March 22, 1990, regarding the number of SG j tubes plugged during 1R2 was reviewed. The inspector had no connants. (b) 2-90-02, " Valid Diesel Generator Failures." The inspector questioned the licensee regarding a sentence in this report i which stated that both diesel generators were out of j service simultaneously for a period of 1 hour and 56 minutes. After a review by the inspector and the licensee, i it was determined that this statement was totally in error, J i i l

og y o nu '12 This Special Report was revised by the licensee on. March 12,1990 to state that "at no time.were both diesels 'out of service' simultaneously. D b.. Deficiency Cards.and' Licensee Event Reports'- Deficiency Cards 'and. Licensee Event Reports were reviewed for.. potential generic impact, to detect trends, and to determine whether corrective actions appeared appropriate. Events which were reported - pursuant to 10, CFR 50.72, were reviewed following occurrence to - determine if. the technical specifications and other regulatory requirements were satisfied. In-cffice review of LERs may resultLin further followup to verify that the stated corrective actions have-been completed.. or to identify violations.'in addition to those described in the LER. Each LER was reviewed for enforcement action in accordance with 10 CFR Part 2. Appendix C, and where the violation was not cited the criteria specified in Section V.G of the Enforce-ment Policy were satisfied. Review of DCs was performed to maintain a realtime status of deficiencies,- determine regulatory compliance. follow the licensee corrective actions, and' assist as a basis for closure of. the LER when reviewed. Due to the numerous DCs processed only those OCs - which result in enforcement action ' or further inspector followup with the licensee:at the end of.the inspection are. 1 listed belcw. The DCs and LERs denoted with an asterisk indicates l that reactive inspection ' occurred following the event and prior to l receipt of the written report. (1) The following Deficiency Cards were reviewed: (a) DC 1-90-0030,." Train A And 8 Sequencer Loss' 0f Power Relay ,j Was Not Properly Tested." .i On February 15,1990, ' the licensee identified that the ) Train A And 8 Sequencer Loss of Power Relay was not ) properly tested in accordance' with TS. No surveillance i test has verified that thel relay operation will result in a Train "C" AFW - actuation. This item will be further i followed up when submitted as an LER._ 1 (b) DC 1-90-0031 "DG Surve111ance' Requirement Was Not ) Completely Satisfied During 1R1." q i On February 16, 1990, the licensee discovered that a DG surveillance requirement had,not been completely satisfied during the first Unit I refueling outage. The DG electrical trips that are automatically bypassed upon loss-af voltage on the emergency bus concurrent with an SI signal were not verified to actual 1J be bypassed. This item will be further followed up when submitted as an LER. -.. =. - - -

4 A 13 (c) *DC'1-90-0034, " Missing Seismic'Solts On Transformers _ Leads To TS Required Unit Shutdown." On February ' 23, 1990, a system engineer found core clamp' bolts missing on seismically-qualified switchgear. The a switchgear was deenergized as was one 'of its loads, 'a l Containment Isolation Valve. After the. four hour time j period had expired for reenergizing' the valve, unit shutdown was initiated as required by Technical j Specifications. Although the bolts were replaced and the- ) CIV was reenergized before shutdown was completed, plant management elected to complete. the shutdown and enter into i a planned refueling outage approximately four hours early.- Two related DC's, 1-90-0035-and 2-90-0021. concerning steel hold down wedges on seismically qualified switchgear were j also written. Based on GE-type test results, the licensee l concluded that the transformers without the upper support i wedges meet the operability requirements at Vogtle and are safe for. continued operation. ) (d) DC 1-90-0050, " Source Range Monitor Inoperable At Time Of l Entry Into Mode 6." l. I On March 1, 1990, the licensee was in a refueling outage on Unit 1.- Mode 6 was re-entered with the comencement of - ) fuel reload. At the time of the Mode 6 entry, one of the j required two $RNIs was under an LCO for performance of an !&C surveillance..The-SRNI was in test and a channel calibration was in progress. This ites will' be further? followed up when submitted as an LER.. (e) DC 1-90-0081, " Missed QC Holdpoints." 1 During the perfomance of itf019001152, a QC holdpoint'to inspect the "B" RHR pump ' motor rotor was inadvertently 1 missed. The cause was due to QC and maintenance personnel i not being cognizant of the holdpoints as work was being performed. A similar event occurred on March 11, 1990, when safety related leads on MCC 1880(67) were relanded ] without QC notification (DC 1-90-0094). This licensee identified violation is not being cited because the criteria specified in section V.G.1 of the NRC enforcement I policy were' satisfied. In order to track this item, the following is established. NCV 50-424/90-05-03, " Failure Of Persons Performing Maintenance Activities To Notify QC' As Required By Administrative Procedure 00201-C Paragraph 4.5.2 When QC Holdpoints Were Reached." t e--

.j e + ( !14' i j 'l

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t .(f) *0C'l-90-0102, " Inadvertent Actuation Of Fuel Handling 'i Suilding Post Accident.HVAC Train _B." j i On March 13,.1990, the' standby train of the Fuel Handling j Building Post Accident HVAC' auto started. One train was already in service to support refueling activities.: No alars of the actuation was received:in the control room. 7 The cause of:the actuation uas believed to have been caused a by a low negative pressure signal. This 11 tem: will be. j further followed up when submitted as an LER. i

i (g) DC 1-90-0103, " Failure To Properly. Review And Approve A j

Revision To Refueling Procedure 93271-C.." j On March 14, 1990, the licensee discovered;that' procedure 93271-C Sigma Refueling Machine Programming-Instructions, was revised from Rev.-O to Rev. I without:the approval of the General Manager, or review by the PRB, as required by j procedure 00051-C, Procedure Review and Approval, Rev. 12.- 1 Technical Specification 6.4.1.6.a. requires that the PR8 be - l responsible for review of fuel handling ' procedures. Furthermore. ' Tables 1 ~ and 2 of procedure 00051-C,; require General Manager approval PRB review of alll fuel, handling 'i procedure revisions. The licensee initiated prompt corrective action.on March - 14, 1990, by having procedure 93271-C' properly reviewed and approved.: Therefore, this 1 i licensee identified violation-is not being cited because i criteria specified in Section V.G.1 on the NRC. Enforcement Policy were satisfied. In order to. track this item.:the following is established. i ' NCV 50-424/90-05-04 and 50-425/90-05-01, ' " Failure,To l Properly Review And Approve a Revision To Refueling Procedure 93271-C." i th) *DC 1-90-0123. " Loss Of. All Offsite And Onsite A.C. Power To i The Unit 1 Vital Buses For More Than 15 Minutes." l This event, which occurred on March 20, 1990, is discussed under paragraph 2 and will be followed up when the LER is _ issued. (1) DC 1-90-0126. " Liquid Waste Discharge Made While Radiation .l Monitor (1RE-0018) Inoperable." l 1 i On March 17, 1990, with the liquid radweste effluent line radiation monitor (IRE-0018) isolated under a work order 1 clearance, a' liquid waste release was made. The release i was authorized.under a release permit without complying with TS ' 3.3.3.9, Action 37. This item will be further i followed up when submitted as an LER. i

1 ~ FD ~ 9 -15 i

hs (j).DC 2-90-0022. " Surveillance Not Completely Perfomed On Containment Integrity Valves Outside Containment."

, - On January 3, 1990, and February 1, ~ 1990, a : partial surveillance was perfomed on containment integrity valves - outside containment. _ All but two of the required valves ~ were "NA'd" ~ on the surveillance data sheets. There was no

{

record in the survei_11ance that the remaining valves were j verified closed as required' by the surveillance H requirement. This item will be further followed up when submitted as an LER. I (k) *DC 2-90-0026, " Unplanned Reactor Trip Due To Electrical Transient As A Result Of A Loss Of Power Event-On Unit 1." This event, which occurred on March 20, 1990, is discussed under paragraph 2 and will be followed up when.the LER is issued. (2) The following LER was reviewed'and closed. (a) *S0-424/90-01, Rev. O. " Reactor Trip Due To Inadvertent I Closure Of Main Steam Isolation Valve." 1 On' January 24, 1990, partial stroke testing of-a Main' Steam Isolation Valve was in progress. During a previous test, i the valve had. failed'to reopen automatically at the 105. closed position as designed. As a result, plant personne1' were prepared to install a jumper.to reopen the valve if it' 1 failed to reopen automatically. The test began.and an indicator illuminated at approximately.105 closed; however. ? unknown to the. personnel involved there were two limit switches which were not. adjusted to actuate concurrently. Consequently, when the indicator illuminated, the. other limit switch had not yet actuated ~and.it appeared that the valve would not reopen automatically. The jumper was i installed to initiate valve reopening; however, position indication was lost and the MSIV 'went fully closed. MSIV closure resulted in a rapid decrease in water level:in Steam Generator #4 to the low-low level setpoint and an j automatic reactor trip occurred. The MSIV closed when its r actuator fuses blow. Although a simulation of the event failed to duplicate the blown fuses and MS!V closure, an i ~ engineering judgement has determined that the Georgia Power Company electricians inadvertently created a momentary electrical short which led to the fuses blowing.. This 1 apparent cognitive personnel error was not the result of j failing to follow approved procedures or the result of any 4 4 I E 1

f ^ i s .C 1 16 j unusual characteristics ' of the Lwork location. Corrective -actions include: a) fuse replacement,.b) procedure revision i to include.a caution that the indicator _may light prior to the valve receiving the reopen signal c) limit switch-j adjustment to -obtain concurrent actuation -and d). ? counselling of the electricians involved: regarding the-necessity of exercising caution when. testing. circuits having the; potential for causing reactor. trip. The l inspector has no further connents.- 4. Actions on Previous Inspection Findings - (92701)(92702) ) i o a. Part 21 Reports (1) (Closed) 50-424/P21-89-03, " Deficiencies In Control-Room. Emergency Filtration System And Isolation Of The Normal-Control 1 Room HVAC System." Corrective actions taken included the addition of'backdraft dampers to -eliminate the potential for system backflow l identified on July 2,1987, and the deactivation of two outside 1 air intake dampers to preclude postulated spurious ' damper-actuation on July 4,1987. The inspector has no further comments. i (2) (Closed) 50-424 & 50/425/P21-89-04, "American Air Filter Seismic. Door Tabs Found To Be Missing _ From ESF Unit Coolers. Without 1 Tabs Access Doors May Not Operate During Seismic Event And { Could Negate Function Of Coolers."- 1 Without the seismic retaining tabs, the access doors of the unit coolers may fail open during a seismic event. Under this circumstance, the return air may bypass 'the cooling coils and, depending on which access door opened, could negate the cooling j function of the coolers. The resulting increase in the roon j temperature could adversely' affect the safe shutdown of the plant. The lack of retaining tabs on the access doors could lead to the inoperability of the coolers in the event of an earthquake. The licensee has reinstalled the seismic tabs (or used a lock and hasp) on both units. The inspector has no further comments. 1 (3) (Closed) 50-424 & 50-425/P2189-16. " Cooper-Bessemer Standby DG-l At Susquehanna Had A Crankcase Explosion Which Originated From j The Thrust Side Of The Number Seven Left Piston Skirt." i J The engine in question is a KSV-16-T. KSV DGs are not used at j Plant Vogtle. Therefore, this part 21 is not applicable. The inspector has no further questions. 2 -.,

g - ~ 1 'l j .j i -17 .(4): (Closed). 50-424 & 50-425/P21-89-18. "PT21 From Limitorque RE SM < Actuators Found To Have Melamine Torque Switches That Undergo - 5 Post Mold Shrinkage And Causes Can Binding. Melamine Torque . Switch Found Not To Be Qualified." 4 The licensee's review regarding Limitorgue SMS actuations for: Unit 2 has now been completed and identified.22.. effected. l motor-operated valves. These valves are part of the MOV Test-Program and will require Movat tests to establish new baseline data after the required maintenance.- The licensee's review for 5 the effected valves on Unit 1 is still in process. The inspector has no further connents. i. (5) (Closed) 50-424 & 50/425/P21-89-19 "PT21 From Dresser l Industries RE Pressure Reducing Sleeves Manufactured By Pacific i Pumps. Part Of The Dresser Pump Division., May Have A Brittle Crack Failure Upon Start Due To Sleeves Being Through Hardened i Vice Surface Hardened." i Pacific Pumps stated that some pressure reducing sleeves were I "through hardened "vice" surface hardened" which could result in a - brittle crack failure within' one hour after operation. Pacific j Pumps has identified the following three Georgia Power Company j orders on which these "through hardened" sleeves may have been provided: j 1 G.O. AT-70093,

Custanc, Order No; -PAV-27380,. CN2, l

Charging / Safety Injection Pump. Three sleeves were provided, j two each on Item 036 and one as part of an internal assembly, Item 057. l G.O. AT-70135, Customer Order No. PAV-27380, CN13 Safety l Injection Pump. Two sleeves were provided on Item 083. l G.O. AT-70345, Customer Order No. PAV-28100 CN69, Safety Injection Pump Sleeve provided as part of an internal assembly. Pacific Pumps has stated that there is no concern if the sleeves have been installed on an operating pump, since failure, if it-was to occur, would happen within the first hour of operation. l Pacific Pumps later advised Westinghouse that the above l identified sleeves provided to Georgia Power Company are acceptable and there are no further actions required. The j inspector has no further comments. j 3 '(6) (Closed) 50-424 & 50-425/P21-89-20. "PT21 From Cooper-8essemer Concerning The EDG Intake Rocker Arm Assembly. Potential Interference Between The Connector Push-Rod And The End Socket i Of The Rocker Arm. Submittal References A Similar Notification From Gulf States Utilities On October 31, 1989." ~

_ _ ~ 9: i 18-0 'i h Cooper-Bessemer stated that there is _ a ~ potentia 11 interference - .between the connector push rod and.the end socket'of;the rocker-am. Cooper's investigationLshowed results similar' to those { identified in the investigation by GSU. Any interference would J show up in assembly or during maintenance start-up runs. This 1 is _ significant when the engine is still in a maintenance or assembly mode and not ye^ operational.. This is a. replacement j parts concern, since equipment installed on engines that have - i been operated or tested (site or factory)'have demonstrated that no interference exists and are, therefore, not affected. g Vogtle's parts issues history has been reviewed and it was i detamined that the warehouse has not issued these items for i maintenance.- The assemblies in question were placed on j warehouse hold January 25, 1990, pending QC inspection. The inspector has no further connents. 5. Release from CAL 1 On April 9,1990 GPC management briefed the Regional Administrator and. 1 the regional staff concerning the-event review which ' the licensee had conducted after the March 20, 1990, Site Area Emergency event. The short-term corrective actions which the site had implemented were considered to be adequate to allow the plant to start up. This released them from Item #1 of CAL-50-424/90-01. Long-tern corrective actions will i be presented to the Region no later than May 15, 1990. 6. ExitInterviews-(30703) i The inspection scope and findings were suunarized on March 29.-1990, with J those persons indicated in paragraph 1 above. The inspectors described 3 the areas inspected and discussed in detail the inspection results. No dissenting comments were received from the licensee. The licensee did not i identify as proprietary any of the materials provided to or reviewed by ) the inspector during this inspection. Region. based NRC exit interviews were attended during the inspection period by a resident inspector. This inspection closed six 10 CFR Part 21 Reports, and one Licensee Event Report. The it m identified during this inspection were: V10 50-424/90-05-01 " Failure To Mechanically Secure Valve i 1-1208-U4-176 During Mode 5 As Required By TS 3.4.1.4.2.c" - paragraph 2.a. I i NCV 50-424/90-05-02, " Failure To Incorporate Adequate Cautions In SSPS Procedures Regard 1H Simultaneous Loss Of Both' Source Range NIs 1 When Placing Both SRNIs En Inhibit Error Inhibit" - paragraph 2.a..

.l 4 19 1 NCY 50-424/90-05-03, " Failure-Of Persons Performing Maintenance Activities To Notify QC As; Required By Administrative Procedure 00201-C Paragraph 4.5.2 When QC Holdpoints Were Reached" - paragraph 3.b.(1)(e). i NCV 50-424/90-05-04 and 50-425/90-05-01,. " Failure To Properly Review And App (rove a Revision To Refueling ' Procedure 93271-C" - paragraph 3.b.(1)g). 7. Acronyms And Initialisms ACOT Analog Channel Operability Test AFW Auxiliary Feedwater System 'AIT Augmented Inspection Team CFR Code of Federal. Regulations CIV Containment Isolation Valve DC Deficiency Cards DCP Design Change Package DG Diesel Generator ECCS Emergency Core Cooling System l EDG Emergency Diesel Generator ESF Engineered Safety Features ESFAS Engineering Safety Features Actuation System i EST Eastern Standard Time GE General Electric GSU Gulf Station Utilities I HP Health Physics l HV High Voltage i HX Neat Exchanger HVAC Heating Ventilation and Air Conditioning IIT Incident Investigation Team ILRT Integrated Leak Aate Test s IST Inservice Testing KSV (tradename) LCO Limiting Conditions for Operations LER Licensee Event Report MCC Notor Control Center .l MOV Notor Operated Valve MSIV Main Steam Isolation Valve i W0 Maintenance Work Order i NCV Non-cited Violation l NI Nuclear Instrumentation NPF Nuclear Power Facility NRC Nuclear Regulatory Commission NUE Notice of Unusual Event PR8 Plant Review Board QC Quality Control RAT Reserve Auxiliary Transfonner

.,a- .a. 4 20 RCS. Reactor Coolant System Rev Revision j RHR Residual Heat Removal System i RMWST Reactor Makeup Water Storage Tank i SAE Site ~ Area Emergency SG Steam Generator SI Safety Injection System SM8 (prefix to melanine torque switches) SRNI Source Range Nuclear Instrumentation SSPS Solid State Protection System I TDAFW Turbine Driven AFW Pump TS Technical Specification UAT Unit Auxiliary. Transformer ' UV Under Voltage VAC Voltage-Alternating Current VIO Violation 1R1 Unit 1 First Refueling Outage IR2 Unit 1 Second Refueling Outage . i

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cc~ jy [G..:. - OFFICIAL TRANSCRIPT OF PROCEEDINGS ~ ti -- Agericy! U.S. NUCLEAR REGULATORY COMMISSION

Title:

BRIEFING MEETING Docket No\\ 9 toCATion Waynesboro Georgia DAlt March 28,1990 1-105 pga i ANN RILEY& ASSOCIATES, LTD. 1612 K St. N.W. Suke 300 Mshington, D.C 20006 idPROKCT o63877

y. _ ~ 1 'y P y ,d' 'Page;95> w. L 1 ~- -'look at'.it asia group.-- as those groups-when they come in. j _2' - the logic board, we'll' vent-that group;ofLsensors offLto- '3-watch the diesel-trip.- 7 In this case ~we're looking at the smaller pieces, ] 4 r 5-here we're looking at the' larger-combination of~tho' trips. i 6 for an' actual running diesel and watch the diesel trip.. }; , So'it sounds like-you're satisfied' m! 7 MR. CHAFFEE 8 that this'is.a comprehensive test'and if the problem is-9 still there -- that's a bad question -- this:is.as 10 comprehensive as you can make it.. 11 MR. HOLMES: Yes, we believe it is as comprehensivel - 12 a test as we can make it to try to identify the root cause, ~ 13 of the problem. We'll be lookingL - first-of. all. we'll try 14 to recreate the situation, then' we'll go.and look at the 15' sensors, we'll look at the logic, we'll look at.the lines, 16 we'll look at the integrated tests of starting and stopping-l~ 17' and tripping and then we'll.go.back and do another UV, test l' ..l - 18 to see if we created another problem for ourselves. 19 I hope I don't have to come back up here and~tell-i 20 you we found nothing. i 21 MR. CHAFFEE: As far as the air quality that's in 22 the pneumatics, is there any way to get a handle on whether 23 or not that contributed to this intermittent problem? I 24 understand that you do tests to make sure it's not a 25 problem, but like was there a test taken on the air before L. 92 PROJECT 063971 l l lL w M e r---- r -m w--- w-

a c L ,1 e (' Pago 96 I 'l the: event,ihas one;been takenfafter the' event, is there any

f.,,

'23 difference in those?_.Is thereganything like~that that'.can ,p 3-m 3 bedone. to' try to. determine cif 'somehow you> had a, fluke, some f ; I 4 debrisfor moisture.or something that maybe was there at'the-s 5 time'that is gone now and that's-why_-- you'know:what I'm- -- n 6 saying? 7 MR. BURR: We can.probably get dew point readings. '8 MR. CHAFFEE: -Idon't-knowwhenyourlastdewpoint[ 9 was, I don't know if it was - - q 10 MR. BOCKHOLD: We'll go ahead and take the action'.to-5 11 .go ahead and go to our logs and find out what the last dow _. 12 point was,- when it was and we'll take a dew point,.we'llcadd lM i. 13 that into some parallel path in the testing modes. We_ve j; u. 14 got plenty of time to get a dew point. ,i 15 MR. HOLMES: You're also asking about air quality. l 16 MR. CHAFFEE: From what I can tell,.it sounds like j l.j 17 you're come up with about every. test possible if the thing y' 18 .is there to find'it. So now the only thing I.can think of_ I:, l 19 is how can _you figure out 'if there's something intermittent. Ii 20 The only thing I can think of that's intermittentLis maybe I' 21 some sort.of poor air quality or debris that caused one of l 22 these sensors to act irregularly for a period of time.and e; 23 then somehow in testing the diesel it's gone away. So maybe: i 24 there's some way to go back and find out that maybe that was 25 the case. But I don't know how to do that. 92 PROJECT 063972 I i I l l

~ ^ t .g1 7' Page 97l }-'! That's-the only thing.I can think'of. ij o 1 2 3 T .2 MR. BOCKHOLD: We'll go looklat it..We can'get out _j ! r 3; the'INPO guidelines on instrument air systems ~and'seelwhat yj 1_ i 4 'other' kind'of tests that we haveLbeen' running on' instrument: 4l; 5 air systems or if we can come up with the same testingithat-0 6 we do for normal plant instrument air systems,:we may be (i l7, 7' .able to do it for a section.of the diesel.- l 8 MR.-KENDALL: I think we'd like to know what test'on'1 j '9 the air system is routinely done for the diesel start-l l 10 system. 11 MR. BOCKHOLD: We'll take that action, we'll' go do. f [!' 12 that and then we'll go run a battery of tests on the air. i .13 system for the diesel. I.]; 14 MR. KENDALL: Ken, I:have one more. question. On'thelj 15 logic board test and the sensor testing, is the extent ofl l' - 13 16 the testing that's going to be done on the A diesel

l. l 1

17 generator now in this plan, is it essentially-identicalto 18 what was done during the-36 month inspection and testing? 19 MR. KOCHERY: No. 1 20 MR. HOLMES: I believe that the logic board testing' ~I l [ 21 would be very similar to what was done post-maintenance. 1 22 MR. KOCHERY: If you look at the procedure we marked 23 there, you'll see some of them on the schedule. The marked 24 up version of the procedure. Basically all the trips will I' 1t 1 25 be looked at. l 92 PROKCT 063973

...Jnvo, J .71T DOCUMENT NO. ~ 2 57 as'7-F t p OFFICIALTRANSCRIPTOF PROCEEDINGS i t h Agericy: suci..r m.sul.cory commi.. ton

Title:

i ur ret.conr.r. ace i Docket No. L. 4 LOCATION: t um April 3, 1990 'l - 74 pgg i i ANN RILEY& ASSOCIATES, EID. 1612 K Sr. N.W., Suhe 300 Mahington, D.C 20006 L (202) 295-3950 i l I

'j a i 59-1! diesel ( bac)t to service -and anything that.was suspect, if-1t l ~ -2 ilinauq461e) at all, we replaced it with a new one.- 3 VOTCE:, Okay. : Just:a-second, George. l 4 (Pause.) c,! } 5 VOICE: Is'there Cooper there, also?~ - f ' 'y i 6 VOICE: No, Cooper is not. j 7-VOICE: So Cooper is not on-site? 8 VOICE: No. They've gone back. 9 VOICE:- Is Cooper -- are you going to use Cooper l 10 to get involved in this test you're proposing for looking at j 11 the [ inaudible) cycle water? 12 VOICE: We've talked to them about the theory and 13 we'll collect data and counsel with them as necessary. j v 14 VOICE: George is now here from Region II. Wh'ich 15 sensors are remaining in the diesel? I'm looking for why-do 16 you believe that it would be okay to place that diesel back i 17 in service in its present condition? Are there any. suspect la sensors remaining installed? 19 VOICE: There are no suspect sensors remaining at i 20 all in the diesel. i 21 VOICE: How have you ruled out the possibility j H22 that air quality, poor air quality may have caused the j 23 problem? l 24 VOICE: [ Inaudible) air quality both for normally j F f 25 and when the team was here, they asked us to test the air j t i

5 i 60 I ~1' -quality and we tested it. i 2 VOICE:.Okay. That test includes oil,' moisture 3 and (inaudible?) lj 4 VOICE:.That includes moisture and'looking at the ~5 . filter. ) l 6 VOICE: How often are those filters changed? i 17 VOICE: The overhaul period. 8 ' VOICE: Which is what? 9 VOICE: Eighteen months in our particular case. 5 i 10 VOICE: So they're changed out just before they 11 start the diesel back, when you start (inaudible) the l 12 testing? 13 VOICE: Well, you go ahead and you -- as-part of 14 the tear-down, as part of that procedure, you change it out, 15 you pull it and you look and see if you've accumulated any 16 dirt and grime.and stuff on that filter, and the filters { 17 would come out very clean. 18 VOICE: So you were confident, based upon tests l 19 done, that the quality of the air is now satisfactory and i 20 you do not believe that was the root cause of the problem j 21 before. I 22 VOICE: That is correct. 5 23 VOICE: This is Al Chaffee. We have not yet j 24 reviewed the data that the licensee has given on the air 25 quality. i

p y: .t ) 1.

1-

'2-l '~

. 3

-4 .5 a i 7 '8 9 10 Transcript' of audiotape No. 32, 11 . transcribed by Danette L. Holbrook, Certified Court 12 Reporter and. Notary Public. "13 14 15-t 16 .i 17 4 18 t 19 20 21 i 22. 23-BROWN REPORTING, INC. 24 1100 SPRING STREET, SUITE 750 ATLANTA, GEORGIA 30309 ~ 25 (404) 876-8979 l ) 922CMECT 056515

g,

.7 F

gy' 75 1 N 11 _the --i(inaudible) --- we better'do'that,-and'we will m L2 ' collect: no data until all of' this is . inaudible). - - I will obviously -- (inaudible).. ( 3 4' Okay.. '5 (Inaudible). 6-G. Bockhold: Okay. So we're dangling 7 ahead of the -- (inaudible) 10:00 o' clock NRC 8 (inaudible). b 9 J.A.: NRC wants a teleconference-10 everyday at 10:00 o' clock. 11 Everyday at 10:00 o' clock.. 12. Okay. .f- .13' I want to talk about that so we can do 14 that later. 15 Well, the 10:00 o' clock meeting that we 16 had for department heads the overall critique 17 summary -- we've got to have -- obviously. .They 18 haven't had a meeting for less than three hoursLfor 19 whatever reasons (inaudible). .20 McCoy: George, we might want to let 21 everybody in here know the strategy and the action 22 plan for getting the confirmatory action 23 (inaudible) who's going to be the contact pcint 24 so if anybody gets phone calls or whatever, then we I' 25 will make sure they get the directed to ~ 92 PROJECT 056519

t> 6. 1 .(inaudible) - .If you get phone. calls'the way that 2. the confirmatory action letter would be. released 3 which-isL the only thing.now restricting the start-up 4 is by our project' chief ~ Ken Brockman, and he willLbe 5 , calling-me on that to set up that (inaudible) 6 somebody else had already started giving direction 7' and so forth. So he's going to try and release'that 8 in the region, and we will try and release it down 9 here.to make sure we got one official chain; and he r 10. request that the information, he reads brief the 11 administrator region and administrative. I 12 believe George passed out a paragraph (inaudible). 13 (Inaudible). 14 G. Bockhold: (Inaudible) along 15 outline we need to include Ken Brockman in the 16 telephone conference will be that he had an I 17 ' opportunity to (inaudible) as far as confusion-18 (inaudible) number I will give it to you 19 (inaudible). 20 McCOY: We don't have anybody from the .21 NRC in here do we? 22 VOICE: No (Inaudible). 23 JA: Is that we made the comment that i a 24 actions were made to start releasing the .(' 25 confirmatory action letter, that would be Al 92 PROECT 056520

E a 7 1. -Gibson?. 2 McCOY: No that would be.a lead of Ken-3 Brockman and-Ken Brockman went and talked to Al -4 Gibson and cleared that up, so there's no '5 outstandino 6 JA: Right. I guess my point is that o -7 .there wasn't any of my people quoted in it. 8 No. What it talked about was 9 Al'Gibson got me identified -- 10 (inaudible) meeting that we had and we started 11 to (inaudible). 12 G. BOCKHOLD: Major activity that I think 13 we need to continue to do is laid out here There 14 are some peripheral activities, such as the ENN and 15 Jim Roberts are trying to get the back up ENN -- 16 (inaudible) And I can put out a memo on ENN 17 communications that will clear that.- The other '18 thing was the control of vehicles and such in the 19 safety area, we already put out a memooon that. The 20 final thing on the list is associated with the 21 critique (inaudible) together; and I've 22 already had Jim Roberts and George Frederick put 23 together a critique summary for the NRC 24 (inaudible) and Ken Brockman is prepared to I. 25 talk. Talk to the regional adminis? n tor on Friday 92 PROJECT 056521

8 to 9et our release (inaudible). Is that what we're saying? (Inaudible). (Inaudible). BOCKHOLD: Well, yeah, I think we need to finish our meeting at the same time. I think, you know, the events that are (inaudible). McCOY: Well, also we need to be sure that everybody understands that those recommendations are not what we're going to do i (inaudible) those are just the initial recommendations. They're not even (inaudible). SKIP: I know that, but does the NRC (inaudible). (Inaudible). (Inaudible). HORTON: This is just a proposed meeting (inaudible) on but what I resolved in my mind is the segregation of (inaudible) overlap where it's always got the deisel inoperable (inaudible) well, anyway on the top line item there as far as B goes (inaudible) get it rolling, keep an eye on it, keep it from -- (inaudible) that issue here. The bottom line is i 92 IHCklECT 056522

31 1 painting a gloomy picture; and he took much delight 2 in telling them that we had.it under control and i 3. (inaudible) exit, and we don't need their 4 services. Now, we did it with our existing staff r 5 and that is a big compliment that is extremely -- we 6 don't realize how serious this is being taken by the 7' utilities external to us. I think most of utilities i 8 would have stumbled for quite some time; and he i 9 reiterated that he was amazed that we more or less 10 kept the outage on schedule despite this and that 11 despite the fact the outage people are upset at our 12 schedule being late they don't really care about IIT 7 13 event, we have really done an amazing thing that we 14 should be proud of what we've done; and I really l 15 want you to pass that along to your guys because I 16 think their wholeheartedly. Now, I have the other 17 comments he said yesterday. 18 MOSBAUGH: No. I think as Mike says 19 that, you know, the engineering group is the one t 20 that has worked the troubleshooting, 'orked with the f 21 team, and worked with Rick Kendle, you know; and 22 their focus of the team has been the diesel, you ( 23 know, the two big impacts of the emergency that we t .4 got ourselves into is the failure of the diesel to i f' i 25 perform and the failure of the plant to carry out p p ECT O M

F 32 l' the emergency plan. Those are the two big things 2 that really went wrong and made this the. magnitude. 3 event that it is.. NRC has yet to really dig in real 4 hard in the emergency plan area. They spent most'of 1 i '5 their time digging into the diesel area,.and'they're. i 6 ~ slowly -- they're certainly not satisfied yet, but 7 they're continuing to ask for information; but 8 slowly we are resolving issues with the diesel and 9 satisfying them, and I think at some point here,~you 10 know, they will be satisfied on the diesel issues I 11 sufficiently to release the hold; and that's 12 particularly significant because in those two areas, 13 the emergency plan and the diesel, they not only 14 have a hold on Unit 1 which they have in the i 15 confirmatory action letter; but if those areas went i 16 sour or went sour sufficiently, they have the l 17 potential for shutting down Unit 2. Because both of 18 those issues are generic and so this thing could 19 have expanded. It hasn't; and, you know, it'could i 20 have gotten a whole lot worse because it could have 21 ended up being a Unit 2 impact as well; and I think 22 it's a tribute to everybody that was on the team t 23 with the diesel troubleshooting and all those that r 24 supported that team and everybody that's helped the i 25 corrective actions in the other areas. There were l h

e i i I 33 i 1-numerous other areas like some of.the things with t ~2 the met tower and the ERF (inaudible) - and, you 3 know, PERMS and ENN and, you know, there's a lot of 4. different things where I think we've~ explained to '5 'the team's satisfaction the behavior of the plant, 6 you know; the team comes in and if you can't really 7 explain the behavior of the plant, you know, it was 8 unexpected but, you know, if you can explain the 9 behavior of the plant and if you have an explanation 10 and it seems reasonable, you know; they're i 11 satisfied, you know. If they come in and you really 12 can't demonstrate that you understand where it was 13 fired from and why it was designed that way, they 14 start getting in to your design and configvtation 15 control areas and say, my God, they don't even know 16 what they got down there; and so forth; and then, 17 you know, they start turning that into to generic 18 issues. And, you know, we've effectively stopped 19 that from happening by having the information, 20 having the expertise and providing it to the team 21 properly, you know. There are some inadequacies, 22 you know. All these events reveal some of those, 23 you know; we learned a lot, the design of the (, 24 diesel, trips that we really probably don't all need 25 in there in a UV type run, you know, and functioning g pgojECT 0%M7

in os.uo,,s " N i- '~ TIT DOCUMENT NO. 203. ci-.lS) -b - 9L- ~ 't. OFFICIALTRANSCRIFTOF PROCEEDINGS i Nuclear Regulatory Comission ggg Tide: Tel Phone Conference:

IIT, l

Licensee, Region II (CLOSED) Docket No. l L LOCA g Bethesda, Maryland DATg Friday, April 6, 1990 Pacts 1 - 34 i P i l. ANN RII1Y& ASSOCIATES, LTD. 1612 K St. N.W. Suite 300 l Mahington, D.C 20006 l (202) 293-3950

1 4: I o when the work order was closed, in comparison to when the 'l calibration itself was done. -So, I'm sure that caused.some 2 And some of the other words there were the job-3 confusion. got changed and-assigned to a different person for'a period l 4 of time, and he started to use different words. 5' So, why don't you telecopy what.you want us-to. 6 We'll give it to Mark Briney, and Mark Briney will '7 fill in? i fill that in and supply any information that way. R 8 MR. KENDALL: Okay. That sounds great, and we t 9 j realize it's going to take a couple of days, probably, to do 10 i 11 it. MR. BOCKHOLD: Okay? 12 MR. KENDALL: Fantastic. r l 13 i MN. CHAFFEE: Okay. Then let's go on to the l 14 15 diesel generators themselves. f Maybe the first thing we should do is talk about i 16 this dew point situation and what you guys believe with 17 regard to that, and then I guess -- I thought we'd go in and 18 a talk a little bit about what you found on the testing and 19 where you're going with the testing. 20 On the dew point situation, MR. BOCKHOLD: Okay. 21 i yesterday afternoon it came to my attention that on the 29th 22 of March we had run a test, and the test on the dew point 23 So, you know, we had some concern about 24 was unsatisfactory. 1 why the test on the A Diesel was unsatisfactory on the 29th, 25 )

W T' ~ ! f '_ f i I5 ^\\ gci l' and we're pulling in together a bunch of information. I 2-At this point -.and this is speculation on my it -3 part -- the evidence is tending to point to a bad ^ 3-g- instrument, a bad dow-point sensor instrument,'and we only i 4 5 have one onsite,.and we're getting another one, and-other 6 than that, you can speculate seven different dozen ways on j this thing, but that's what the evidence is starting to ] 7 8 point to, because when we test air at similar conditions, it ) 9' all appears to be higher right now. Okay?- J 10 And.it's at a significantly different condition, 11 like our instrument air in the turbine building. The 12 instrument does appear to work correctly, but at the diesel f l I. 13 temperature pressure dew point, the instrument may not be l 14 working correctly. 15 MR. KENDALL:' This is a test instrument. j 16 MR. BOCKHOLD: So,~ basically, what happened is wo ( 17 got this information; put the jacket water test, basically, 5 18 on hold until we could determine what we had; and what we ') J 19 did in the meantime is that the appropriate procedure that 20 the vendors and our experts tell us to use if you have a 21 higher dew point in the diesel storage tanks is basically to 22 do a feed-and-bleed on the tank, and over a day or so, the air will clean up to -- the dew point will clean up to the 23 24 required quality. 25 We started that. We checked the instrument lines

6 te 1. at one of the low points-on the A~ Diesel. We also checked 2 the.receivir by blowing it down. We haven't really gotten anyrealwaderoutofthereceiver.inblowing.itdown. The 3 1 n= 4 comment was that we haven't seen any water coming out of the-5 bottom of the receiver, and there's n' drain valve right*-- l-6 there's a drain pipe right on the bottom. Further, the lI 7 diesel system engineer blew one of the drain points down'on

t.

8 -- and this is the A Diesel -- on the control air system, l 1 l. 9 and he didn't see any moisture come out-of that line. 10 And we've run some other tests. Like we ran one 11 test quickly on the B Diesel. That showed bad. We're off j to run a test in'a few minutes on one of Unit 2's diesels. 12 I expect that's going to show bad, because'right at this I 13 y point, what I believe is that the instrument is bad. 14 In parallel with this, we're going to buy -- we're 15 going to find another instrument,.so we can do this. test l 16 with a different instrument and see what that tells us. j 17 In parallel with this, when the cooper people get =; 18 j in in the morning, which I guess is about 11 o' clock or so, l 19 i 20 we'll give them a call. Given the indication that we have on the air and the dew point that this instrument is 21 ') reading, we believe we can probably do the jacket-water test 22 without doing any damage to the control or instrument air 23 24 system. I We believe that even at an elevated dew point, 25 \\

I l u -i 7' i 1 this is a long-term problem and.not an immediate problem for 2 -- associated with the controls on the diesel. We believe' 3 .the diesels ~are operable right now, for example, and we a 4 believe.this is -- you wouldn't want-to run like this for- ,:i a 5-months, if you had an_ elevated dew point.- l !. !c p 6 so, we want to verify our belief;with the Cooper -l 'y

7 people..

If we do verify our belief with the Cooper people, 8 we will go-ahead and run the jacket-water test. 9 MR. CHAFFEE: When do you expect to have the new i i 10 instrument onsite to do the dew point? I 11 MR. BOCKHOLD: Don't know. Maintenance was off-l 12 this morning to go find one from one of our fossil plants or j I-13 maybe even buy one in Augusta. 14 MR.' CHAFFEE: Okay. 15 MR. BOCKHOLD: I'm not sure we can get exactly the 16 same instrument that we'have. The one that we have has a 17 radioactive source in it, and you have to be, you know, .l 18 appropriately licensed to have this instrument. 19 so, we'll get something that's equivalent, but it i 20 probably won't be exactly the same instrument. 1 21 MR. CHAFFEE: But you'll get one that meets 22 whatever the standards are for its readings being -- felt to 23 be correct, one that's calibrated and that's -- I don't know-24 if there's any industry standards in that area for that type L ( 25 of test instrument or not.

8 1 You will ensure that your test instrument is i 2 properly pedigreed. 3 MR. BOCKHOLD: Yes. 4 MR. CHAFFEE: Okay. 5 Well, okay. I guess as far as 6oing the test 6 before you have satisfied yourself, through a test, that the 7 dew point of the air is within spec or not, I guess you're 8 probably right that it wouldn't cause any damage to the 9 diesel. It obviously would be prefer 2ble that you get that 10 thing all resolved and clean up before you ran the diesel. i 11 As far as that goes, George, I think what we'll do 12 is, after the call, Region II and we will talk about Cox a t 13 little bit and decide what our feelings are on that. I 14 understand where you're coming from. 15 MR. BOCKHOLD: We've basically put the test on 16 hold until we could resolve the issues with dew point. ) 17 Okay? ) 18 And one of the parallel paths was that our i 19 engineers, our folks believe that the air quality that we 20 are actually seeing in the instrument controls and in the 21 receiver, even if the dew point was a little bit high, would 22 not be of concern for operability of the diesel engines, 23 would not affect the control systems. We are verifying that 24 with Cooper. If Cooper agrees with those opinions, we would i 25 be prepared to go ahead and run the test. I

b l, 9 l' In the meantime, in parallel, we.are doing the '2 ' appropriate' procedure with the air receiver and the air-s 3' dryer that if the dow point'is not-correct, we'll go ahead' 4-and lower the dew point in the air tank.. But we're not i 5 seeing the~ dew point in the air tank getting any better. j 6 So, we're starting to believe, more and more, we have a' bad' i instrument, and the instrument has somehow failed. 8 MR.'CHAFFEE: I see. Okay. 'I understand. l 9 [ Pause.) 10 MR. CHAFFEE: George, Rick is going to talk to:you 'll a little bit at Catawba. 12 MR. KENDALL: It's our understanding that Catawba-I 13 is the only other plant with TDI diesels that has a 14 refrigerant-type dryer, and there were some problems at 15 Catawba with their dew point and moisture affecting their 16 Calcon pressure switches, and we understand'that you don't 17 think you have a dow-point problem, and we. understand that is you've also got a different model of pressure switch that 19 may not be subject to the same types of-problems that they 20 had at Catawba. 21 However, when we go back and look at this thing, 22 one of the tests that was run was on jacket-water pressure 23 disconnecting the sense line, and the diesel tripped after 24 80 seconds, which was one of the timeframes during the i 25 event, and just putting everything together, one of the C _

/ 1 3 0 W. 0 / '6 4 W -TIT DOCUMENT NO.-206 CT-a o - F-9d c I OFFICIALTRANSCRIPTOFPROCEEDINGS f i i J I Agertcy: nucle., mesul.cory Co..i..i n

Title:

IIT Tel* Phone Conference with Vogtle (CLOSED) Docket No. i i mg Bethesda, Maryland i 1 i mm Monday, April 9, 1990 pg,a 1 - 15 i i ( ANN RILEY& ASSOCIATES, LTD. i 1612 Kst KW,suke300 Mihington,D.C 20006 (202) 29F3950

4 h 1 day was.that you had gotten a'new instrument,'but whenlyou j L .2 did testing with it, you got negative. numbers, which didn't 3 make.anys sense. so, you were going lto go get another. 4 instrument for measuring the. air quality from Hatch, and I-1 'S-don't know -- have you gotten that instrument and used it, 1 6. or are you still waiting for it? 7 MR. WARD: 'Getting one from V.C.-Summer,:and this-8 traceable instrument, I think,.is identical or similar to-9 the one that we originally had and all of the numbers.that 10 were reported Sunday were in the range of 36 to 45 degrees. 11 MR. CHAFFEE: With.the' exceptions criteria being 12 32 to-52, I think? .i 13 MR. WARD: Thirty-two to 50, I believe. \\ 14 MR..CHAFFEE: Okay. 15 Did Region II just join us? Who just came on the 16 line? 17 MR. KITCHENS: Hey, let me give you the status 18 today on the dew points for all the diesel generators. 19 MR. CHAFFEE: Okay. 20 MR. KITCHENS: All eight of the receivers are in~ 21 spec except for one. The number two receiver on the Unit 2-22 A diesel, dew point, the last taken,'was 60.9 degrees 23 fahrenheit, which is about 11 degrees higher than 24-recommended by the manufacturer. So, we're blowing that 25 receiver down to get it in spec. All the others are within l

-5 1 the 32 to 50~ range. In fact, they're all like 30 -- in the 2 30s and low 40s,.except for one of the ones'on Unit 2-A 3

diesel.

One of the receivers on 2-A is'43.9 degrees 4 fahrenheit. dew point, and the other was 60.9. ~ SL 12un readings all look' reasonable,' and we believe 6. them, you know, with this new instrumentation that we're 1 i 7 using. 8 MR. CHAFFEE: Okay. Well, the following question 9 to that is if you believe that 60.9-degree reading you have ] 10 on-one of the Unit 2-A receivers, why is it different than j ~ 11 the others? ~! 12 MR. KITCHENS: That's the one that was turned off? L 13 MR. WARD: Yes,.that's the one we turned'off. 14 Yes. + '15 MR. KITCHENS: Yes. I think that's the one that l 16 we had the air -- we had the air dryer turned'off on Friday? l 17 MR. WARD: Yes. 18 MR. KITCHENS: Yes. We went out to look at these ~19 on Friday, and one of them had its air dryer turned off. I 20 don't know if that's the reason, Al, but that would be -- 21 possibly be a reason. 22 MR. CHAFFEE: Okay. 23 Well, then the follow-up question would be if the 24 air dryer was turned off for that one and that accounts for I 25 why it was high, what's the history of these air dryers, you .i ~

y l i '6-l' know,' stemming from the event? I' realize I'm being real. 2-inquisitive here. Is there any chance that you had your air J 3' dryers off or any ' evidence to substantiate whether or not 0 4 there is any reoccurring problem there, or fis this -- or do ] 5. you know that this air dryer being off:is, basically, a one- {

6 time occurrence?

7 MR. KITCHENS: I don't know. Our history -- you a know, we dug up the history -- I-assumed they had shared it 9 with you -- from the~1ast year, where we do these PMs. It i 10 pretty much had passed on -- I don't have the history in 11 front of me for the Unit 2 one. a 12 MR. CHAFFEE: Okay. I am going to make'a -i 13 statement here. I'm not sure if it's true, but I don't 14 think we've seen or.have been given the information on the 15 dryer performance. If I'm wrong,'then disregard the l 16 following: 1 1 17 Please provide us that information that addresses j 18 the air-dryer performance on the, particularly, Unit 2 -- 19 oh, I see. That's right. That was Unit 2 that had the 20 problem, not Unit 1-A. We really wanted the Unit 1. l 21-MR. KITCHENS: Is what you need the actual dew l 22 points measured during the PMs for, say, the past year? a 23-MR. CHAFFEE: We just need the information that l 24 shows us to what extent air poor quality might have had an f 25 impact on the operation of the Unit 1-A diesel.. l l t

7 1 MR. KITCHENS: That's a pretty ger - ric thing to ,f 2-ask for, A1, and I don't know how to provide that to you. I 3 can just give you the PM results that show the dew points, 4 when we've taken them during the PMs, and we do a monthly PM 5 for them, and they basically have all -.maybe one or two I 6 passed in the last -- since october of '88. 7 MR. CHAFFEE: You meant to say only one or two 3 8 failed. Right? Not passed. All but one or two passed? I 9 didn't hear you correctly. 10 MR. WARD: We had a couple of them fail'is what'he 11 intended to say. 12 MR. CHAFFEE: Okay. Well, maybe you can just give 13 us that table then. Give us a table of these surveillance 14 results over the past couple of years, and we can go from 15 there. In those cases where there has been a failure, then 16 what we would be interested in there is to know how long did 17 'that condition exist. I guess it sounds like the answer 7 18 would be between surveillances. 19 And the other thing that we would be interested 20 in, as well, if you did have poor air quality for that 21 period of time, what impact, if any, would that have on the l I 22 diesels trip circuitry? 23 MR. KITCHENS: We haven't had poor' air quality. j I I It's been within the -- you know, the recommended -- the 24 3 s 25 vendor's recommendation, pretty much for the last year. Up i 4 i

g . _ : s a..... ~ E( t c 8 1 '[ 1 until March 31st, when we reported'a failure, which now.we 2-

don't really believe that it was a failure because of the.

3 . instrument, I only see.one. failure.over the last 12 months '4 during.a PM. 5 MR. CHAFFEE: -You've got the datalin front;of?you? 6 MR.. KITCHENS:. As.a matter of facti. that one - 7 failure, if I'm reading this right, was March 16, '89. I 8 don't have the data. I have a list of all the work orders 9 where we did-it and which ones passed and which ones' failed,. 10 and you know, that's why we were thinking it'aight have been-11 an instrumentation problem. We have not had'a problem, o 'h 12 really, with meeting the vendor recommendations for 32 to I i 13 50. You know, that's a vendor-recosmonded' number that our-j- 14 dryer should be able to meet. I i 15 MR. CHAFFEE:. I see what you1just said. You also. .L 16 brought up another good point, which is that, you know, the lt' 17 way you got into this thing here recently was you thought j 18 you had bad air, but the instrument.was bad. e 19 MR.. KITCHENS: Right. I guess I'n'saying, as an 20 overview, I just went back a little over a year's worth and 21 asked for that history, and we have been doing the PMs and 22 we have routinely been. 23 MR. CHAFFEE: Okay. 24 MR. KITCHENS: I do not believe we have an air-25 quality problem, unless there's one associated with the

~. - 2 9 1 vendor's recommendation. . MR. CHAFFEE: Okay. 3 'MR. KITCHENS: 'I'd be glad to give you the -- I 4 can just have somebody-look up the actual numbers from all 5 of.these -- you know, what the actual dew-point numbers were 6 that were obtained for back in that period and, you know,- a 7 furnish that-for you for the A-train diesel or for both of 8 . them, whichever. 9 The only problem -- we did have one problem on a 10 B-train diesel. One of the air dryers -- we got a: fail, a 11 78-degree number on March the 18th of '89, and we replaced 12 the dryer and fixed it. You know, we actually replaced it-i 13 with a new dryer, and it's passed since then, every time. 14 That's the B-train.' 15 MR. KENDALL: Of Unit 17 16 MR. KITCHENS: Yes. 17 MR. WARD: The other fact,that ties in with that, 18 A1, is the filters on the inlet to the controlled air system 19 are replaced during the refueling overhauls, and those were-20 pulled this March -- early March, and tho' reports-are they 21 were all in a as-new condition, did not show signs of having 22 been susceptible to any kind of dirty air. 23 MR. LAZARUS: This is Bill Lazarus. Al had to go 24 to another conference call. I'll fill in for him. 25 Rick, did you get what you needed? L l Me 4 ww -+ ii+e.- -we=,m-y -+e*---'% w --rw-C e ---*M?--1+--siw@w gw-- t

$' 5

• h. h)" i J,

3-u-M 03:a1 , :'m :: acese u M :o2 / UNIT 1 A TRAIN O!ESEL GENERA 708 ' AIR RECEIVER DEW POINT MA5UREENTS . M NO. DATE RECEIVER E01 RECE!VER E02 19001651 4/8/90 34*F 33'F 19001513 3/31/90 80*F 60*F 19000899 3/12/90 48'F 45'F 3/9/90 61*F 66*F 19000465 2/11/90 37'F 37'r i 18906'445 1/18/90 44*F 44*F 18906199 12/19/89 40*F 37'r 18905007 11/20/89 40*F 47'r 18904442 10/20/89 38'T 45'F 18903652 9/27/89 45'F 45'F 18903214 8/24/89 37'F 35'F 18902798 7/30/89 45'F 49'F 18902453 6/28/89 48'F X02 was tagged out for natntenance 18900984 3/16/89 22.6*F 20.1*F 4 i l 4 _ _. _ _ _ _ _ _ _ _ _ _ _ _. - - - - - - - - - - - - - - - - - - - - - ^ - - - - - - - - - ~ ~

~71589.67449 . TIT DOCUMENT NO. 233' pf-f ~ 1 L OFFICIALTRANSCRIPT OF PROCEEDINGS Agen@ wue1..r a. gut.cory commissto. l Tide: 11r r.1.contor.ne. wien Licensee and Vogtle Docket No. I i gg Bethesda, Maryland I DATE: Wednesday, April 11, 1990 1-8 { PAGES: -t i t 1 (. MW MUW& ASSOCMTES, UD. 1612 K St. N.W, Suite 300 'Ashington, D.C 20006 (202) 295-3950

i 1 1 UNITED' STATES OF AMERICA '2' NUCLEAR REGUIATORY COMMISSION 3 4

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X 5

In the Matter of: 6 IIT Teleconference with t i 7 Licensee and Vogtle 8 - - - - - - - - - - - - - - - -X 9 10 Nuclear Regulatory Commission 11 Operations Center 12 7735 Old Georgetown Road 13 Bethesda, Maryland 14 Wednesday,' April 11, 1990 15 16 The above-entitled matter commenced at 10:03 f 17 o' clock a.m., when were present: 18 19 Alfred Chaffee, IIT Team Leader 20 Rick Kendall, NRC 21 Ken Burr, Vogtle '22 George Bockhold, Vogtle 23 John Aufdenkampe, Vogtle 24 25 1

l h - s qp l 2 s. 1 P R O C.'E E D I N G S i' t 2. [10:03 a.m.) 3' .MR. AUFDENKAMPE:-. I think 12 videotapes that we' j i 4-had.to' duplicate, bu'.. those were supposed to be done. l 5 yesterday and last night, the duplications, and those will j i 6 be' Fed Ex'd, along with the other data'that we have 7 accumulate, tonight. So, you should have the' videotape and-8 an additional box of data tomorrow morning. j 9 MR. CHAFFEE: Okay. Great. j 10 .How about the diesel generator alarm and trip i'l 11 setpoints? l 12 MR. AUFDENKAMPE: It should be in the box j - f 13 tomorrow. l 14 MR. CHAFFEE: So I guess we'll probably get the { 15 box on Thursday? t 16 MR. AUFDENKAMPE: Yes. You should have it .i 17 Thursday morning by 9 o' clock or so. [

I 18 MR. BOCKHOLD:

Ten o' clock is when Federal Express j i 19 arrives. I 20 MR. CHAFFEE: Okay. l i 21 We just got the dew point information on the air -i 22 receiver. That was handed to me about 10 minutes ago. i i 23 I guess the accumulator air pressure information - l 24 - where does that stand? And by that I mean -- l 25 MR. AUFDENKAMPE: I don't know that we understand i G -

i r 3 <1 your question. -t i 2 MR. CHAFFEE: The question'was the air pressures 3 that we observed to exist on the diesel start-air system 4 during'the event. 5 MR. AUFDENKAMPE: This is the conversation with s 6 the operators? I 7 MR. CHAFFEE: Yes. 8 MR. AUFDENKAMPE: We have that information, and we t 9 were going to call Gene directly with that. 10 MR. CHAFFEE: Okay. Great. So, you'll probably 11 talk to him this morning? 12 MR. AUFDENKAMPE: Yes. I will make sure that you u 13 know he physically has the information. I will make sure 14 that happens before noon, A1. 15 MR. CHAFFEE: Okay. 16 Then the next question I had is -- you know, 17 yesterday we mentioned the fact that we'saw something in a - t 18 - I guess it was an outage schedule on the diesel, two 19 items. One had to do with an entry that said that the 20 alarms will not come in on the 1-A diesel generator. That i 21 was supposedly some work that was done on March 22nd. And I .22 was curious -- have you been able to find out what that was j 23 all about? 24 ,MR. AUFDENKAMPE: I have not found anything on {l 25 that. 4

l f

1

-Ken, do.you want to talk ~about the thing.on.the ) I 2 fuel' rack, explain that? 3 MR.-BURR I thought you were going to talk to Ray ~ '4 Howard. I did not talk to him' yesterday. 5 MR. AUFDENKAMPE: Well, what you' told.me is what I l l 6 understand to be correct. .I did not talk to Ray Howard i 7 ~directly, but'I understand that to be correct, that all they 8 did was to take care of an interference. They cut off part j } 9 of the -- what's it called, Ken, onLthe rack? 3 10 MR. BURR: Yes. It's.the shaft that extends out. j 11 It has no effect on the engine whatsoever. 12 MR.-KENDALL: Was this like something to keep; d' 13 somebody from stubbing their toe on it? q 14 MR. BURR:. I think it was. ] 15 Again, I should have talked to Ray _Howard, and'if 3 16 you don't mind, I'd like to do that. -1 17 MR. CHAFFEE: Okay. 18 We're still interested.- We'd like to know what- '19 those two activities were. We're trying to understand what -l .20 they mean, particularly the entry about the alarms will not 21 come in on the 1-A diesel generator, because apparently that 22 work was done after the event, and'we want to know if it-23 sheds any light on what happened in the event. 24 ,MR. BURR: You're seeing this in a work order, 25 right? .,.1, .. ~ a

x i 5 1~ MR. CHAFFEE: No. ~ 2-MR. KENDALL: Ken,Jwhat it is -- it's a big chart, .3 .and it looks like -- across the top, it's got rows,.and the p 4 rows are listed by the date, and it looks like it's for the 5-month of March. It goes from the 1st through the 31st. And 6' the' columns down the side are all the stuff that you did on 7 the diesel. So, it's a diesel outage planning chart or. 8_ something. I don't think it's a planning chart as much as 9 it is an actual chart of what happeaed.- And it's got blocks 10 in there for when certain tests were run and when the' diesel l 11 was started and what surveillances were performed, andLit's 12 got a list -- it must have 60 or 70 entries in it. 21 13 MR. CHAFFEE: It's like a bar graph that goes 14 across, sort of like critical path, and it had all the-1 15 testing that was done up through the 31st in there. So, 16 apparently, that would tend to indicate that this' entry on 17-the 22nd about alarms will not come in on the 1-A dieaal= 18 generator is being some sort of work that was really done. 19 MR. KENDALL: Ken? Is Ken Burr there? 20 MR. BURR:- Yes, I'm here. 21 MR. KENDALL: If you give me your number, I can 22 give you a call back and give you a reference number or 23 whatever it is off the top we're talking about, so you can 24 see where it is we got the information. 25 MR. BURR: Okay. My number is area code 205-677-

.s. - 6 ??' ["11 -7836.. ,g-2' MR. KENDALL: Okay. I'll give you a call Nhen { 3' this is. finished.. 4- .MR. CHAFFEE: Okay. 5' MR. BOCKHOLD:- Hey, A1,'this is George Bockhold. . p 6 MR."CHAFFEE: Hi, George. 7 MR. BOCKHOLD:- On one of your questions yesterday 8 and what we faxed uplto you and'you said you just received 9 on the air receiver dew point measurements -- 10 MR. CHAFFEE: 'Right. 4 11 MR. BOCKHOLD: I really don't have. good data prior ~ 12 to the last'date shown on that chart there, and'we had'been 13 working on our PM program, and we really don't have i 14 consistent data earlier on on the diesel. We've looked at 15 that, but that does not -- we believe, in fact, the air 16 quality of the diesel was basically' satisfactory. 17 We did have one of the air dryers out for some . 1 18 maintenance during that period of time, also, earlier on, in 19- '88, and that kind of stuff, but the kayfthings about the 20 satisfactory. quality of air is associated with the fact that q l 21 the normal receiver is at roughly 250 pounds, and the air 22 pressure is reduced to 60 pounds. That reduces the dew 33 point about 30 degrees, and -- going to the control system. 24 So, given the fact that the room is heated and the roon y l-25 pretty much stays at a constant temperature -- it will vary . ~. s.I

a s 7 'l . somes it's not' air. conditioned -- and the fact that we blow , e dotnt the receiver on,. basically, a daily basis, even if the 2 3 air in~ the receivers was saturated with' water, we'd get a - 4 30-degree decrease in dew points for the control air, would 5 not be moist air. 6-Also, we, at each ofLthe overall periods,-have 7 . inspected the' control ~ air filters, and they have been' 8 essentially like new. We didn't'see any rusting,or I I 9 . corrosion products 'in those filters..And also, we inspected

1

-10 the.one air receiver, and we.only saw a very light corrosion' l 11 around the wells and some minor oil in the botton, and none I 12 of that really got to the control air. ) P-13 So, we think -- we believe we still have 14 satisfactory air ever since startup on thase machines, but 15 our PM history is not as good in the '88 timeframe, because 16 we added some PM program at that particular point in time. .i i 17 MR. CHAFFEE: Okay. .j 18 Okay. I understand. l 19-Okay. I understand, from talking to Warren, that H -20 the discussion he-had yesterday with several' people at your 21 site was very helpful. 'l 22 MR. AUFDENKAMPE: -You have to pass on to Warren 23 that it was educational both ways. 24 MR. CHAFFEE: Okay. f y 25 MR. AUFDENKAMPE: But it was lengthy. 4 i

g .n 4 s .) 'l MR'. CHAFFEE Yes. .I_ guess _you must have lasted,. j u. 2 _ what, 2 or-3 hours, 4 hours? 1 ? .i 3 MR. AUFDENKAMPE: .I sat in'on about 3 hours of it, 1 4 at least 3 hours. l .t t 5 MR. CHAFFEE: Who is talking right now?- Is that j 6 John Aufdenkampe?- } 7-Okay. So, that takes care of documents. It ' l a sounds like we're getting a package'of stuff tomorrow. 1 1 9 Talked about the test program. j t i 10 Anything else we have? l 11 That's all the questions I have today. That was a .\\ 12 nice,.short call. r 13 Mk. AUFDENKAMPE: I guess we'll talk to~you j 14 tomorrow at 10:00. I 15 MR. CHAFFEE: Right. i 16 Thank you very much. I i 17 [Whereupon, at 10:14 a.m. the Interviey was j -l la concluded.] 19 i I 20 i f 21 22 23 l 24 p o 25 1 I

v ~! t a. i l-REPORTER'S CERTIFICATE l ,s This is to' certify that the, attached. proceed-j ings before.the United States Nuclear Regulatory Commission in the matter oft i Teleconference j NAME OF PROCEEDING: j u DOCKET NUMgER: t PLACE OF FROCEEDING: Bethesda, Maryland were held as herein appears, end that this.is the. original transcript thereof for the file of the United States. Nuclear Regulatory Co4 mission { taken by me and thereafter reduced to typewriting ~by me or under the direction of the court report-i ing company, and that the transcript is a true and accurate record of the foregoing proceedings. j L4-6 i P.ossie Sutton Official Reporter Ann Riley & Associates, Ltd. I i l + ) l l t i f 1 l' l ) l

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'[ G wpf 41 l 1 quality'of the diesel generator focusing'on /4 [c,wf Giu M4fm see,(inaudiblegdewpoint control d has-2. 3 concluded that air. quality is satisfactory. ' i 4 Initial. reports e 5 expected' dowpoints later (i=-ad i M a L. 6 That specific was in reference. 7 to the March 29, March 30, and March 31 work 8 order associated with the instruments-that we 9 later determine,d w a bad, and we got to l VC. fv* >W I'#TV :_h:. + and we 10 (inaudible) figured out (inau$C S 11 how to work the dible) instrumentation. 12 This was confirmed by an internal ) afieyge W 13 -inspection of.one.(inaudible) on April 6, Hg, l 14 1990. (Ir.eudibl+) looked in there and found a \\ 15 light (I n t '_' fm _ible) of oil an a found some ~csdd minor corrosion on.(i>naudible) also. M up 1s. as 16 M0 17 We periodically aced in a control is air filters were done in March, 1990. Fr'on 19 all reports that I've heard on that is that 20 those air filters were always clean. Even in 21 the previous outage they were very clean and 22 practically brand new. 23 We showed no indication of corrosion 24 and air receiver and daily air receiver 4Q 25 (ipmadible) showed no indication of corrosion r L_--______-_-_-__-.____----___------__-_-_______-_-_--_----____--_-__-____

e 42 1. -and' daily air receiver blow down with:no 2 significant water discharge.- 3 We believe, although Schwartzwalder -4 is checking, is ever since we started up j i 5: we've'been doing daily air receiver blowdowns 6 as part of operations. 4gQ3 i 7 Further, I guess, (in rfh-! when S the air quality came up, spoke..to the cooper (owpper-i 9 people. And the (hrrfiLir) reaction 10 ( ) if you do these daily-blowdowns,- 11 and you don't have a air receiver full of. 44We M 12 water, because of the (in::ditic) marine 13 engines and because you have a pressure 14 reduction going to the controls, you really 15 don't have to worry about air quality. -t '16. i ud m kThy/Afued.the word that was use qg,, y e w+jgysej. p(in a us aniw)j .-ps.< 17 you know, 11Ty (inct_i' ^). 18 That was the flavor that came to me, okay. 19 Ne've done I ouess additional g 4 some + .vg e 20 research. We have (inaudible) '88 ff1 timejM~P pame-y, 9e 21 -well, we had the (incut.ibl+) program et-4 fM 22 working out the (inaudible) program W 4 (inaudiile).Mb$d>j 23 ? 24 The '88 time frame from like 5/10/88 l 25 to 5/2/89, somewhere in that time frame, I

m i i 43 I 1 approximately a year ago, the year.before 2 that, we may have had one or more of the 3 dryers'out.of service for several months, 5 4 okay, and that's indicated here. i 5 We probably were'not'doing good PM's l 6 on k g ryer quality at that particular 7 (irendible), okay. i a I guess my question to the group is I 9 t h a t a>w(-i n e.a i s i-i the statement that we l 10 made in our letter at all? 11 Is the other facts, the fact that i brY HR 12 (inaudible) blow do,wi tje air dryer, the fact- -j un w -~ 13 thei.. den ' t- (inaudible) the air dryer, the 14 fact that essentially a/swI yepr or approximately

1. As

.l 15 a year before the 'finaudible), you know fp.at % & wk2 M /rea Mrs, w n*b h. l ~r 16 did have k & pfaudible)O did ;;t fully (innefikl-) 4 sAs 17 et terent (1 _ _ d i b i; }, (gh M* M *I VOICE: The 18 -fin kwgeh4 enu.astion I he$<d to& g g [ m n 19 answer first, (inaudible), the (inaudible)Entsh%#[ t 20 you know,'the right way. (Inaudible) j .21 seventeen, twelve, twenty-two, you have 22 (inaudible). I 23

Remember, (inaudible).

/h4} 24 VOICE: That's theoretically not 25 possible. 1 I F 1

o 1 ~ l h .i 44 i [69 VOICE: I mean, what we've got to 1 -- 2 basically.say.is our PN program.before, I 3 3 don't know -- /d7, VOICE: June,~ July, '89; right? 4 5 69 VOICE: Yeah, somewhe.re -- .i jkM4WY" VOICE: (Inaudible) '89. l 5 t fhh VOICE: Y e'a h, somewhere in that 7 8 time frame our PN program was suspect.

Okay, l

9 the readings were suspSct. You know, that i 10 doesn't mean that you did have good air, but 11 it doesn't mean that you dian't'have good air. 12 We don't know.if we had good ~ air or bad air, 13 okay. t 14 But in the meantime we did pull the 15 filter, okay, and we had been doing the 16 blowdowns and all of that impJies thpt pga 1 My PTJ f*I9K W Te%d* \\ 17 air, although it may not demonstrate j 18 Fia udiile) the best quality, w$s I 19 satisfactory. 20 N# CNY VOICE: (Inaudible)n.oucan'tFin;.a.,7aeen y 21 that we have a bad air and m i t) because 22 anytime you've got a minor corrosion, you can 23 see anytime (inaudible). i l hjf VOICE: You know, I guess I would 24 25 tend to believe that we had good air based t l f

h l 45 1 'l upon two things; one, pulling the ai.. ilter. be l 2 and inspecting it, which was (in&aud bl ) new.. b e). -3 We didn't'see':any buildups of ( a 4 The second thing'I guess I would 5 tend to believe, you know, two.hundred and AM i s' fifty pounds (inaudible) or thereabouts and j 7 kind of room' temperature, okay,'~and your l 8 dryer (inaudible) an expansion process, even r 9 if you didn't have a dryer s through an 10 expansion process, the (iwanrettrtw) in my .I 11 opinion. M M 12 You know, so some two'hundred and. 13 fifty pounds or two hundred.and ten: pounds 14 (inaudible) sixty peunds- (inaudible) dry air. l 15 So I would then conclude,'if my J 16 logic i co rect, I would conclude.that l pp jef kvr un owedt e <. 17 (in ),,my-a ir qua lity i s still a valid 18 (inaudible). ~' ' (Inaud ble) d' 19 VOICE: h7 /q Vp {C E.: Yes. You know, 20 y.r NSWit th g 21 (inaudible) ??v== (in rdibl-) _.. this a 3 22 generic letter and stated what our air

23 quality was.

Vw. h$ 1

• m+..,4dlE*ENN'lA*h*E*L*,'jfjt***

25 quality (inaudible)jvogel naudible) system. I f

f 1 i tt f '4" g fifty de resNlC Y- [ I W been established q] > inaudible).- J1 (inausible)y" g & f) - *p f p ys R H 2 .hi-wwmpreneur ( .3 ~That.'s what we said.our requirements are., l e ~ Dewpoint-criteria we t 1 ed 4-5 based on a' design.capabilityi(inaudib ) and a' '6 ' minimum diesel generator-(4-l 7 so basically we sa 'th t our s criteria is fifty degrees (inau I) Q.RW.4WYt' h VOICE: Ye ah, W(p ssbe+.w pli is /Ar fChg 9 inaudible).- 6% VOICE:yk & lw %d J n,sd k;f ed S **Y**5 10 6, demonst %c ^ ^ ' - e n. An aet4 L)A.sa m s. rate it-il per

ically, y question really focuses on n#v) r, t

~ this'( b 12 le) letter and how'long -- you 13 know, I believe from what I've heard from all 14 the experts that. Item-4 here is still valid. h 15 I mean, we believe that we've had 16 satisfactory quality.a'ir going to the control 17 system. 18 OICE: 'Yes, I think that because tw* (inaudible).A ///4. 19 the - 20 @~ VOICE: Given the fact that you 21 have an expansion process in the dryer. /M VOICE: Again, George, I say, you 22 23 know, it depends on what you're going to call 24 satisfactory. 25 If you're going to say satisfactory l

L 47 1 is what we said-in. response to the generic 2 letter, I'm not'sure that we can show that-3 we've met that criteria. 4 The-problem is that you can't tell 5 what any of these numbers are because the way 6 the PM's done, you just can't tell. ) 7 You know, there's a number down i 8 here, but there's no calculation. You're 9 doing the measurements at atmospheric -{ 10 pressure, yet they need to be corrected back 11 to a system pressure, you know, and we're 12 getting high numbers. % *) he dogmes bS"*

• i;\\

k VOICE ( in e u u t s A.u ) 13 m :, :.4 % 4~(' *+ N... 4. enh u... a. u. q y 4, + back M. l 15 h VOICE: We've made engineering 16 judgments, okay, on this particular statement. 17 I would go ahead and [M VOICE: You know, we're saying if-18 19 there's internal corrosion that we're 20 observing, therefore air quality is met and 21 that may or may not be the (inaudible). 74s Ah' hvb W>M 22 Y Vo E: (Inaudible) you can see i. 23 the (i dible). / Jfe gg / ph VOICE: ( IT.a o d i L 1 -) based on our 24 25 j udgment when we pull it down three times a I p n r -~

r- .v;g c t 48 + 1 day'and o tions says based on*what we've. eraj/OrM fvvvyd SIM t s e e r,.. ( iNa ible)gno corr sion,'we have,to'say 2 kb8 b .(in $d ~ 3 b u i l a' g(g 4)) D W ds~/sh. 4 OICEs-Ti h VOICE: 5 Insudible recent-6 requests, G e o r g e', for all these work' orders', 7 okay. They've asked for all these hh VOICE: 8 I think at ten o' clock 1T% 9 I'll talk to (inaudible) about going back on 10 our past' work orders in.our PM program.in '88 11 was not as good as our PM program has.been in '12 the past year, basically,. in '89 going to90, 13 and we'll provide that information to our 14 engineering judgment that-we had satisfactory 1 15 air quality. I think that's 16 bb ~ expansion of air from VOICE: . The 17 receiver pressure.to eighty pounds is going to 18 result in about an eighty degree depression of 19 dewpoint pardon me, thirty degree 20 depression at dowpoint.. 21 VOICE: (Inaudible). 22 VOICE: Yeah. 23 VOICE: Yeah. 24 VOICE So the receiver -- I hh VOICE: The-absolute worst case 25 i 'l i

f.eR-c5 *90 11837 ID SONOPCO-UQGTLE TEL NO:1-295-e77-7se5 , a695 PS2 WO kW N fs^ f I4ntd v l I ST/,TU$ OF CORRICTIVE ACTIONS F0LLOW!NG 764 V MARCH to, 1990 SITE AREA ENER6ENCY fQg.) v On March 20, 1990, a site area emergency was declared due to a loss of offsite d5 - power concurrent with a loss of ons< to emergency diesel generater. capability, gpg $ In accordance with VESP procedures, an event review team has investicated the events leading uppgfpjlowing'the site area emergency. While the review team results are m.... pending fina managecont review and approval, the investigation Ts:M M ?y coup etc. Those actions considered t important for continued safe plant operation have been tuplemented. These g_ include establishment of a management policy on control and operation of voniclesI(soaattachedletterfromGeorgeBoekholdtoallsitepersonnel); upgrading of emergency notifi41on network communications (see attached letter from George Sockhold to all Emergency Directors and Commentcators); complete retesting and calibration of both Unit 1 emergency diesel generator control systems; barricades to prevent unnecessary entry into plant quitchyard areas;. and communications of immediate corrective aqtions related to operations to licensed operators. p [g.M q aea S In addition, the event report also everal nyertermrecommendations which require additional managementh.it and e These include the mid. loop operations: sequencing of outage activities plant -- " ^ ' - post maintenance d'esel functional test ification system upgrades; changing diesel generator cent c; and, aluating the duties and responsibilities of the Emergency Director. + pnM l The most significant occurrence during the eve h20,1990, involved the I i failure of Olesel Generator (DG) 1AJgaamain to support shutdown cooling. The event critique team, rr-':; -f utility and vendor technical experts has investigated the 06 failure nd provided the following i facts ourtne bench seest all three skes was r k rature switches were foundtobesethigl,duringthe maintenance inspection in early March a. 1990(byapprezimately410degreesFabovethesetpoint). All three were adjusteddownwerdusingecalibrationtechniquethatmayhavedifferedfrom I that previously used. b. Following the March 20 event, all three switches were again bench tested. Switch T5 19110 was found to have a setpoint of 197 degr' e F which was approximately 6 dsDrees F below its previous setting. Inh T519111 was founc to have 4.Jstpoint of 199 degrees F which was appm :aately the same as th inal setting, Switch TS 19112 was found to have a setpoin 4 f-) sett which was approximately 17 degrees F below the previ 186 as read. justed. Switch T5 1911Z also had e ses11 leak a g j acceptable to argport diagnostic engine tests and was o$ M re notalled. S$ v c. During the subsequent test run of the 06 on h $0, one of the switches a.g (T519111) tripped and would not rose peered to be an c4* i intermittent failure because it subsequently et. This switch and the

men-assothseto:SONQPCO-VOGTLE TEL tot 1-2g[htG-Je95 a695 P03 5_ t leaking switch (T519112)d with n9 additional problems,were replaced with new s testing has been conducte d. The Unit 1 jacket water temperature switches have been reca11brated with the manufacturer's assistance to ensure a consistent calibration technique, Subsequent testing indicated that the diesel annunicator indication of s. March 20, 1990 is reproduced on a high jacket water temperature trip. Based on the above facts, the event review team concluded that the jacket water high temperature switches were the most probable cause of both trips on March 20,1990. The following actions are being implemented to ensure a high state of diesel 1 reliability. 1. A test of the jacket water system temperature transient during engine starts is m progress. The purpose of this test is to determine the actual jacket weMr ?seperature at the switch locations with the engine in a normal st wih lineup, and then followed by a series of starts without air rolling the engine to replicate the starts of March 20. 2. Operators are being trained prior to their next shift te ensure that they y understand that an emergency reset will override the high jacket wate - Q rature trip. 3. The undervoltage start feature of the Unit 1 DG has been modified sue Qt non-essential engine trips are bypassed. Qhis change will be ,,,,,QA.o mi + i* - it.ementedonUnit2priortoApril 30,1990. egn.P. mw,., 4. GPC is evaluating the possibility of a design change and Technical Specificationchangetodeletethejacketwaterhightemperaturetripasan i essential engine trip. 5. Since March 20 1990, SPC has performed numerous sensor calibrations (including jacket water temperatures), extensive logic testing, special pneumatic leak testing and air quality referification_, and me tiple engine a9< 'I etion er sne.se corrective >* starta med runs undee warlenn -dttiansMC are 0, ora ie. gns,. stir,-sdese,,ynatant,atthe GPC will continue to work with the Transamerica DeLaval Incorporated,6wners Group to iPprove 08 reliability. SpC'will also review possible improvements to protective instrumentation and controls and any additional engine enhancements will be scheduled for refueling overhaul periods, fd.M4y GpC will continue to work with the IIT and an independent lab to tL. = the cause of failure of the temperature and pressure switches currently under quarantine. ~ 92 PROJECT 001630 u____________________________________

TEL ' NO: 1-295-E77-79 eat f;717 sqgp g ie=====m M.-J.59. 96.97,.lp.1,Smm-WGTLE I ", DM;;;L"*,,"' f*xygg- /1' Cie*. Tone % 6.23 93 $OMtf6p'i~J ~ M1*c" cafd = =. =s E.o%dne*" April 8, 1980 ,,. - c.,,,

w. e. w.:

a seem vee e.m.sen wi w two.n. ELV-01816 00!! 8 Docket No. 60-424 l U. 5. Nuclear Regulatery Commission l Region 11 i 101 Marietta Street, N. W. Atlanta. 4A 30323 ATTN: Mr. $. D. Ebneter l I

Dear Mr. Ebneter:

i I l. V0GTLE ELECTRIC GENERATING PLANT CONFIRMATION OF ACTION ttTTER !^ On March 20, 1990, a site area emergency was declared due to a less of offsite power concurrent with a loss of onsite emergency diesel generator capability. . l Following the event, GPC received a Confinnation of Action Letter dated March 23, 1990 con:erning certain actions we were taking. We have reviewed the event team report and the a>propriate corrective actions necessary for entry into Mode I have been accompliswd. Therefore, we are requesting approval to return Unit I to Mode 2 and subsequent power operation. The fellowing discussion provides justification for this request. In accordance with Vogtle Electric Generating plant procedures en event review team has investigated the events leading up to and following t$e site area emergency. The event review team has presented the results of their review to management and those recommendations considered taportant for continued safe plant operation have been taplemented. These inclede establishment of a Banagement policy on control and operation of Vehicles (see attached letter free George Beckhold to all site personnel); upgrading of emergency notification see attached letter from George Beckhold to all networkcommunications(Communicators); complets ratesting and calibration of Emergency Directors and both Unit 1 emergency diesel generator control systems; barricades to prevent l. unnecessary entry into plant switchyard areas; and communications of ismediate i corrective actions related to operations to licensed operators. 1 In todition, the event report also contains a number of longer-ters recommendations which require additional management review and evaluttion. 1-These include the sequencing of outage activities; plant condition's permitted during mid-loop operations; post-maintenance diesel functional testirbe1 logic; emergency notification system upgrades; changing diesel generator con and re-evaluating the duties and responsibilities of the Emergency Directer. Bl%lT A$ PAGE/#4OFff7fAGSS) i I i M

I U. 5. Nuclear. Regulatory Commission Region 11 ELV-01516 faggTuo 20, 1990 involved the The most si9nificant occurrence during the event 'of March failure of Diesel Generator (0G) 1A to remain running to support shutdeun cooling. The event critique team, utilizing utility and vender technical I esperts has investigated the DG failure and provided the following facts: a. During bench testi , all three jacket water temperature switches were 1-found to be set hi during the DG maintenance inspection in early March I L ! 1990 (by approxima oly 6-10 degrees F above the setpeint). All three were adjusted downward using a calibration technique that may have differed from that previously used. b. Following the March 20 event, all three switches were again bench tested. l $ witch TS 19110 was found to have a setpoint of 197 degrees F which was approximately 6 degrees F below its previous setting. Switch TS 19111 was found to have a setpoint of 199 degrees F which was approximately the same as the original setting. Switch TS 19112 was found to have a setpoint of 186 degrees F which was approximately 17 degrees F below the previous l setting and was readjusted. Switch TS 19112 aise had a small leak which l was judged to be acceptable to support diagnostic engine tests and was reinstalled. c. During the subsequent test run of the DG en March 30, one of the switches (TS lelll) tripped and would not reset. This appeared to be an intermittent failure because it subsequently reset. This switch and the 1eaking switch (TS 19112) were replaced with new switches. All subsequent l l testing has been conducted with no additional problems. d. The Unit 1 jacket water temperature switches have been recalibrated with I the manufacturer's assistance to ensure a consistent calibratten technique. e. Subsequent testing indicated that the diesel annunicator indication of March 20, 1990 is reproduced on a high Jacket water temperature trip. Based on the above facts, the event review team concluded that the jacket water i high temperature switches were the most probable cause of both trips on March l-

20. 1990.

l l l EXIElII # 8 l PAGEf#f0F/f7]AGRS) M l

g TEL NO 1-205-877-7665 ne.< e-o 8PR-97 '98 Seide ' ID SONOPC04CETLE I U. S. Nuclear Regulatory Commission Region II ELV-01516 7-Th = The following actions are being implemented to ensure a high state of diesel reliability.

1. A test of the jacket water system temperature transient during engine starts.

was conducted. The purpose of this test was to determine the actual Jacket water temperature at the switch locations with the enaine la a normal standby lineup, and then followed by a series of staris without air rolltag the engine to replicate the starts of March 20. The test showed that Jacket water temperature at the switch location decreased from a standby temperature of 163 degrees F to approximately 156 degrees F and remained steady. l 2. Operators are being trained prior to their next shift to ensure that they understand that an emergency reset will override the high jacket water temperature trip. 4(ww % w i// 44 4 w A +. D auf.A.ZL, 4 :G Se /ns 3. The unarvoltage start feature f the Unit 1 DGs has been modified such that i the non-essential engine trips are bypassed. ver, alarms are still provided to inform the operators of off normal itions. This chance will be implemented on Unit 2 prior to April 30, 1990 , h %(JJ i > eM (

d. salt 4.

GPC is evaluating the possibility of a design change and Tech al fa n Specification change to delete the jacket water high temperature trip as an essential engine trip. j E. GPC has reviewed air quality of the D/g air system includlag dowpoint control and has concluded that air quality is satisfactory.. Initial reports of higher than expected dowpoints were later attributed to/ faulty + instrumat,. This was confirmed by internal inspection of one air receiver gon Aprl' 6,1990, periodic replacement of the centel air filters which showed.no indication of corrosion and daily air receiver blowdowns with no j g significant water discharge. P 5. Since March 20, 1990, SpC has performed numerous sensor calibrations (including jacket water temperatures), extensive logic testing, special pneumatic leak testing, and multiple engine starts and runs under various conditions. Since March 20, the 1A DG tas been start #G 18 times, and the 18 DG has been started 19 times. No failures or probless have occurred during any of these starts. In addition, an undervoltage start test without air roll was conducted on April 6,1990 and the 1A 0/G started and loaded properly. CompletionofthesecorrectiveactionsjustifySpC's determination that the DG's are operable. [ t l amar ze \\ PI25 /46DF M73.sE(s) M 1 ,x+- ,.c --.-,--n-

W -07 '90 08:4 ID 500PCO-90GTLE 1:a. tus a cos-o,,-,es,2 ..,em l* i l l U. 5. Nuclear Regulatory Cosmission Region 11 ELV-01515 l Paam Faur EPC will continue to work with the Transamerica DeLaval Inco rated Owners Group to improve OG reliability. GPC will also review possib improvements to protective instrumentation and controls and any additional engine enhancements will be scheduled for refueling overhaul periods. SPC will continue to work with the !!T and an independent lab to determine the cause of failure of the temperature and pressure switches currently under quarantine. l Based on the above discussion, we believe we have completed the appro>riate l corrective actions necessary to safely operate the unit. We request (RC approval to enter Mode 2 by close of business on Monday, April 9,1990. Should you have any questions, please inquire. Sincerely, i W. 8. Hairsten, !!! { WGH,111/NJ5/gm Attachment xc: Conrota Power Camaany Mr. C. K. McCoy Mr. G. Bockhold, Jr. Mr. R. M. Odom Mr. P. D. Rushton NORMS I U.1. Nuclear Rea jlatory fennittian Document Control iksk Mr. T. A. Reed, Licensing Project Manager, NRR Mr. R. F. Atello, senior Resident Inspector, Vogtle i 1 I s u err E 8 PAGE/470F//7F/SE(S) i \\ t 1

.~. .g. Mbro:e P:nr Cc~ an,. 233 Poomer%ese Mama Gecma303C8. '

• tonorte 404 526 3195 e
• Aatng Accress'

~ f e 95 1 Birmingna n A;acaraa 35201 ' Teteonor'e 205 868 5581 - . April-9,l1990 .rrm.-. c we,., . w. a. Haitsion. ini i Senior V.ce President. l , Nuclear Operations ELV-01516 0012-Docket No. 50-424 j U. S. Nuclear Regulatory Commission I Region 11 101 Marietta Street, N. W. i -Atlanta, GA 30323 4 ATTN: Mr. S. D. Ebneter l

Dear Mr. Ebneter:

i V0GTLE ELECTRIC GENERATING PLANT i CONFIRMATION OF ACTION LETTER

l 1

On March 20,.1990, a site area emergency was declared due to'a loss of offsite i power concurrent with a loss of onsite emergency diesel generator capability. Following the event, GPC received a Confirmation of Action Letter dated March 23, 1990 concerning certain actions we were taking. W . e have reviewed the March 20th event and the appropriate corrective actions necessary for entry into i Mode 2 have been accomplished. Therefore, we are requesting approval to return Unit I to Mode 2 and subsequent power. operation. The following discussion provides justification for this request. j In accordance with Vogtle Electric Generating Plant procedures, an event review l team has investigated the events leading up to and following the site area-emergency. The event review team has presented the results of it's review to < j management and those recommendations considered important for continued safe plant operation have been implemented. These include establishment-of a j management policy on control and operation of. vehicles (see attached letter from j George Bockhold to site personnel);. upgrading of emergency notification network l communications (see attached letter from George Bockhold to Emergency. Directors j and Communicators); ratesting and calibration of both Unit I emergency. diesel . generator control systems; temporary barricades to prevent unnecessary entry into low voltage switchyard areas; and communications of immediate corrective H actions related to operations to licensed operators. In addition, the event review team report also contains a number of longer-tern 1 recommendations which require additional management review and evaluation.- i These include the sequencing of outage activities; plant conditions during i mid-loop operations; post-maintenance diesel functional testing; emergency 1 notification system upgrades; changing diesel generator control logic; and. i re-evaluating the duties and responsibilities of the Emergency Director. 92 PfitOJECT 001623

v a q j y LGeorgia Power 1 u '( U._ S. Nuclear. Regulatory Commission n Region II- 'ELV-01516 Paae Two ' The' most significant occurrence during the. event of March 20, 1990, involved the-j -failure of Diesel Generator (DG) 1A to remain running to support shutdown U cooling. Georgia Power Company,- utilizing utility and vendor technical experts - 'has investigated the DG failure and has determined the following: During bench testing, all three jacket water temperature switches were a. l found to be set high during the DG maintenance' inspection in early March. ? 1990-(by approximately 6-10 degrees F above theLsetpo bt). All.'three were - adjusted downward using a calibration technique that may have differed from 'l that previously used. /! b. Following the March 20 event,- all three switches were again bench tested. Switch TS 19110 was found to have a setpoint of 197 degrees F which was approximately 6 degrees F below its previous setting. Switch TS 19111 was i found to have a setpoint of 199 degrees F which was approximately the same as the original setting. Switch TS 19112 was found to have a setpoint of i 186 degrees F which was approximately 17 degrees F below the previous 'l setting and was readjusted. Switch TS 19112 also had a small leak which i was judged to be acceptable to support diagnostic engine tests and was. J reinstalled. t During the subsequent test run of the DG on March 30, one of the switches c. (TS 19111) tripped and would not reset. This appeared to be an 1 intermittent failure because it subsequently reset. This switch and the leaking switch (TS 19112) were replaced with new switches. All subsequent -testing has been conducted with no additional problems. d. The Unit 1 jacket water temperature switches'have been recalibrated with-the manufacturer's assistance to ensure a consistent. calibration technique. Subsequent testing indicated that the diesel annunicator indication of e. March 20, 1990 is reproduced on a high_ jacket water temperature trip. f. A test of the jacket water system temperature transient during engine starts was conducted. The purpose of this test was to determine the actual jacket water temperature at the switch locations with the engine in a i normal standby lineup, and then followed by a series of starts without air rolling the engine to replicate the starts of March 20. The test showed that jacket water temperature at the switch location decreased from a i standby temperature of 163 degrees F to approximately 156 degrees F_ and L remained steady. { e l l l 00i M - . ~-. ~..

h M l

Georgia Poher1

.G U.: S._ Nuclear Regulatory Commission i Region IIL ELV-01516 i Pace Three l g. LSince March 20, 1990, GPC has performed' numerous sensor calibrations '(including jacket water temperatures),. extensive logic: testing,'special i pneumatic leak testing, and multiple engine starts and runs under various l conditions. 'Since March 20,.the 1A DG has~been started 18 times, and the 18 DG has been started 19 times. No failures or problems.have occurred during any of these starts. In addition, an undervoltage start test without-air roll was conducted on April 6,1990 and the 1A D/G started and loaded properly. . Based on the above facts, we have concluded that the jacket water high temperature switches were the most probable cause of both trips on March 20, 1990. 4 in addition, the following actions have been or are being implemented-to ensure ] a high state of diesel reliability. 1. Operators are being trained prior to their next shift to ensure that they j i - understand that an emergency reset will override the high Jacket water-temperature trip. Alarm response procedures will be revised ~to address-emergency reset functions prior to April 30, 1990. 1 2. The undervoltage start feature of the Unit 1 DGs has been modified such that 11 the non-essee.tial engine trips are bypassed. 'However, alarms are still- ) provided to inform the operators of off normal conditions. (This change 1 l-will be implemented on Unit 2 prior to April 30,.1990.) l 3. GPC is evaluating the possibility of a design. change and Technical Specification change to delete the jacket water high temperature trip as an a essential engine trip. p 4. GPC has reviewed air quality of the D/G air system including dewpoint'- control and has concluded that air quality-is satisfactory. Initial reports-of higher than expected dowpoints were later. attributed to faulty instrumention. This was confirmed by internal inspection of one' air-receiver on April 6,1990, the periodic replacement of the contal air-filters last done in March,1990 which showed no indication of corrosion _and daily air receiver blowdowns with no significant water. discharge. 5. Based on discussions with the NRC in Atlanta on April 9, 1990, GPC will finish reviewing the event review team's long term recommendations and will transmit a summary and schedule of the actions taken or to be taken to the NRC by May 15. 1990. The administrative procedures that specify control of vehicles in the perimeter area will also be revised by May 15. j 92 PROJECT i 001625 E

Georgia Powerd

U. S. Nuclear Regulatory Commission Region II ELV-01516-Paae Four

~ 6. GPC will ~ continue to work with the IIT and an independent lab.to evaluate the instruments currently under quarantine. Upon completion of the the lab test,_ calibration procedures will be revised as necessary to ensure consistent performance.. Completion of these' investigations, reviews, tests and corrective actions justify GPC's determination that the DG's are operable. GPC will continue to work with the Transamerica DeLaval Incorporated Owners Group to improve DG reliability. GPC will also review possible improvements to protective instrumentation and controls. Based on-the above discussion, we have completed.the appropriate corrective actions necessary to safely operate the unit. We request NRC' approval to allow Unit I to return to operation. Should you have any questions, please inquire. Sincerely, g),b lI = ? W x W. G. Hairston, III WGH,III/NJS/gm Attachment xc: Georoia Power Company Mr. C. K. McCoy Mr. G. Bockhold, Jr. Mr. R. M. Odom Mr. P. D. Rushton NORMS U. S. Nuclear Reoulatory Comission Document Control Desk Mr. T. A. Reed, Licensing Project Manager, NRR Mr. R. F. Aiello, Senior Resident Inspector, Vogtle 92 PROJECT 001626 .2

_~ _ ~ q 't? a*w vairs tsoo/11 v1Hv119-a n w unus wurtseeirates' souctsAn neeUSMORYecasemastoss ' h.....) nuesom u istanameena einest w. m. avi.aw?A.esonesa sessa NOV 011931 ' Docket Nos. 50-424, 50-425 License Nos. NpF-68, NPF.81 Georgia power Company. ATTN: ~ Mr. W.E. Nairston !!! Senior Vice president - Nuclear Operations j

p. 0.. Box 1295 i

Birmingham. AL 35201 1 Gentlemen:

SUBJECT:

V0GTLE SPECIAL TEAM INSPECTION rep 0RT NO3. 50-424,425/g0-19 SUppLENENT 1 This refers to the inspection conducted by a Special Inspection Team on August 6. through 17, 1990. inspection was transmitted to you on Januaryprevious correspondence asso 11, 1991. As discussed in the team would be the subject of separate correspondence. Insp . part, the results of that followup team. This report includes, in The inspection included a review of activities authorized for your Vogtle facility. Wtified in the enclosed inspection report. inspection, these fin At the conc 19sion of the a Areas examined during the inspection are identified in the report. these areas Ifithin and represen,tative recthe inspection consisted of selective examinations of pri activities in progress.ords,-interviews with personnel, and observation of weaknesses in operational polices and practices.The inspecti ' inspection summary of the enclosed inspection report.These are identified in the The inspection findings indicate that certain activities appeared to violate NRC requirements. accurate information to the NRC during the insThe apparent violation as - !i , for escalated enforcement action.- Accordingly,pection is under consideration issue is not being issued at this time, and a response to this subject is nota required. However, please be advised that the number and characteritation of j violations described in the enclosed Inspection Report associated with this subject may change as a result of further NRC review. by separate correspondence of the results of our deliberations on thisYou wil matter (. We will contact you at a later date to arrange en enforcement conference to discuss this issue. The additional violation described in this report, references to pertinent requirements, and elements to be included in your response are described in the Notice of Violation. w wt -%+ +- we +-r---i a-d --e

.~, 4 L ew sairs is ro ti v1Huilu-a cam uous u .-Georgia Power Company 2 g gg You are required to respond to this letter and Notice and should follow the instructions specified in the enclosed Nottce.when prepartrig your response to the violations. In your response, you should document the specific actions taken and any. additional actions you plan to prevent recurrence. _ Af ter -. reviewing your response to this Notice, including your proposed corrective actions and the results of future inspections, the NAC will determine whether further MRC enforcement action is necessary to ensure compliance with _ NRC regulatory requirements. In accordance with 10 CFR 2.790(a), a copy of this letter and its enclosures will be placed in the NRC Public Document Room. The responses directed by this letter and the enclosed Notice are not subject to the clearance procedures of the office of Management and Budget as retutred by the Paperwork Reduction Act of 1980, Pub. L. No. 96.511. Should you have any questions concerning this letter, please contact us. Sincerely. Ellis W. Nerschoff eting Director Division of React Projects

Enclosures:

1. Notice of Violation 2. NRC Inspection Report 50-424,425/90-19, Supplement 1 cc w/ enc 1s: R. P. Mcdonald Executive Vice President-Nuclear Operations Georgia Power Company P. O. Box 1295 tirsingham, AL 35201 I 'C. K. McCoy .Vfce President-Nuclear Georgia Power Company i P. O. 1235 Sirmingham, AL 35201 W. B. Shipman General Manager, Nuclear Operations Soorgia Power Company .P. O. 1600-Waynesboro, GA 30830 (cc w/ enc 1s cont'd - see page 3) i

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Georgia Power Company 3 g g gj .ec.w/encia: - J.- A. Bailey (Continued)- .f Nanager-Licensing Georgia Power Company P. 0. ' toa 32g5 - Strafagham. AL 35201 4, D. Kirkland, !!!, Counsel Office of the consumer's Utility Council i - Suite 225, 32 Peachtree $treet. NE R . Atlanta, GA 30302-5 Office of Planning and Sudget l i Room 6158 t 270 Washington Street, SW Atlanta, GA 30334 Office of the County Commissioner Surke WaynesCounty Commtss1on boro, GA 30830 ~ t Joe D. Tanner, Commtssioner Department of Natural Resources .205 Sutler Street. SE. Sutte 1252 Atlanta, GA 30334 Theems Mill, Manager Radioactive Materials Prograa Department of Natural Resources 878 Peachtree $t., NE., Roos 600 ' Atlanta. AA 30309 Attorney General - Law Department 132 Judfcial 8vildfng Atlanta, GA 30334 ' Dan Salth. Program Director of Power Production Oglethorpe Power Corporation 2100 East Enchange Place P. O. Box 134g . Tuckpr. CA-30085-1349 Charfes A. Patrtria', Est. Pcul, Mastings, Janofsky & Walker L 12th Floor 1050 tonnecticut Avenue 8Af Washington, D. C. 20036 4

j K, i s *w _ sait Isero/lI w1NW"l15-2*E0s wow 4 l ] .7 DCt05URE 1 N0TJr.C 0F VIOLATION -Georgia power Company Vogtle Units 1 and 2 Docket Nos. 80-424 and 50-425 License Nos. NPF-88 and NPF 41 During an NRC inspection conducted on August 6 through.17,1990, a violation of WRC requirements was identified. In accordance with the " General Statement of 1 policy and Procedure for NRC Enforcement Actions " 10 CFR Part 2, Appendix C (1990), the violation is listed below. Technical Specification 6.7.1.4 requires that written procedures be established or implemented for those activities delineated in Appendix A of Regulatory Guide 1.33. Revision 2. February 1978. Contrary to the above, during the inspection conducted on August 6-17, I 1990,. two examples were identified in 'which the licenses failed to i establish or implement the procedures for these required activities as follows: i 1. Administrative procedure 00150-C. " Deficiency Control." states that a deficiency card must be written if the deficiency involves safety-related com are to be dispositioned "use as-is/ repair "ponents which or other conditions in~volving safety-related components which require engineering support or other technical t assistance to detersine if the component is deficient. On August 17 1990, the NRC identified that a deficiency c. 1 was not written on r,esidual heat removal (RNR) pump fit (a safet) elated component) to document the pump's degraded conditions which were dispositioned "use as-is". (Discussed in Section 2.2 of this inspection report) 2. Administrative Procedure 00100-C, " Quality Assurance Records Administration," Paragraph 4.1.1.8, specifies that quality assurance (QA) records will exhibit necessary and appropriate signatures or initials and dates. On August 17, 1990, the NRC identified that the Unit Superintendent incorrectly initialed, dated, and signed a QA record which voided Temporary Change Procedure (TCP) 1802-C-7-90-1 to Abnormal Operating procedure 18028-C, " Loss of Instrument Air." with the date of L June 12,1990 in lieu of the actual date (June 15,1990) on which the document was signed. (Discussed in Section 2.3 of this inspectionreport) This is a Severity Level IV violation (Supplement I).

l s 3 I sd taitt is/FO/II w1NW71W.3*g3W WoWd Georgia Power Company 2 Docket Nos. 50-424 and 10 425 Vogtle Units 1 and 2 License Nos. NPF-68 and NPF-81 Pursuant to the provisions of 10 CFR 2.201. Georgia Power Company is hereby i requirod to submit a written statement or explanation to the U.S. Nuclear Regulatory Commission. ATTN: Document Control Desk, Washington. DC 20555 with a copy to the Regional Administrator, Region !!, and, if applicable, a, copy to the NRC Resident Inspector within 30 days of the date of the letter transmitting this Notice. This reply should be clearly marked as a " Reply-to a Notice of Violation" and should include for each violation: (1) the reason for the violation, or. if contested, the basis for disputing the violaties, (2) the corrective steps that have been taken and the results achieved. (3) the corrective steps that will be taken to avoid further violations, and (4) the date when full. compliance will be achieved. If an adequate reply is not received within the time specified in this Notice, an order may be issued to show cause why the license should not be modified, suspended, or revoked, or i why such other action as may be proper should not be taken. Where good cause is shown, consideration will be given to extending the response time.' i FOR THE NUCLEAR REGULATORY COMMIS$!0N' Ellis W. Merschoff, ing Director Division of Reacto rejects 4 Dated at Atlanta, Georgia this 01 day of Nov. 1991 i i I

O' 4 d-stset tseto/ll WANW*llW-E'o3W WoWA 1 i ENCLOSURE 2 l l Report Nos.: 50-424.425/90-19 Supplement 1-Licensee: Georgia Power Company P.O. Box 1295 Birmingham, AL 35201 t Docket Nos.: 50-424 and 50-425 License Nos.: NPF-68' and NPF-81 s Facility Name: Vogtle Electric Generating Plant. Units 1 and 2 Inspection Conducted: August 6-17, 1990 Team Leader: Chris A. VanDenburgh. Section Chief. Otvision of Reactor Inspections and Safeguards. Office of Nuclear Reactor Regulation 4 Team Members: Ron Aiello - Resident Inspector Vogtle i Norris Branch - Senter Resident Inspector. Watts Sarr i i Robert E. Carroll, Jr. - Project Engineer. DRP. Region !! Larry Garner - Senior Resident Inspector Robinson Neal K. Hunes.ller - Licensing Examiner. NRR Larry L. Robinson - Investigator. 01. Region !! Robert D. Starkey - Resident Inspector. Vogtle Craig T. Tate - Investigator. 0!. Regfon II Peter A. Taylor - Reactor Inspector. DRS. Region 11 NcKenate Thomas - Reactor Inspector. DRS. Reglen II John D 11ces. Jr. - attens Engineer. NRR $vbmitted by: 4M O = - M T Mf( Pierce M. 5ktaper. Section Chief 35 Region II. Division of Reactor Projects Date 51'gned Approved by: M 7N 3 A. R. Merdt Catef Branch 3 Region II Division of Reactor Projects Date 51gned ) l

s.* d erstI IseroetI - r WANW'MW-3*03g WOWW 1 i TABLE OF CONTENTS 9 ' !NSPECTION 5UMARY................................................ 1 INSPECTIONS DETAIL5................................................ 5-1.0 INSPECTION 08JECTIVE5........................................ 5

2. 0 ALLEGAT ION F0LLOWP..........................................

5 2.1 Improper Installation of FAVA 5ystem.................... 6 2.2 Operability of Residual Heat Removal Pump............... 11 2.3 Backdating of $1gnatures................................ 13 2.4 Reportability of Previous Engineered Safety Features Actuation System Load Se 0uta ge s.......................... q uence r 15 2.5 Air Quality of Emergency Diesel Generator S ta rt t ng A t r Sy stem..................................... 18 1 2.6 Reportability of Previous Sys tem Outages................ Ig 2.7 Intimidation of Plant Review Board Members.............. 19 2.8 Personnel Accountability................................ 21 3.0 EXIT INTERVIEW 5.............................................. 22 APPENDIX 1 - LIST OF TRANSCRIBE 0 INTERVIEWEES........ 23 APPENDIX 2 - PERSONS C0NTACTED.................................... 24 APPENDIX 3 - LIST OF ACR0NYMS........................... 26 a .f* .6' D ' m. -~ --e

s, w saitt is roet: w m" w -* uw unwa' INSPECTION SUMMRY Activities which occurred in early 1990 at the Vogtle Electric Generating P (VEOP) raised concerns within the Nuclear Regulatory Commission (NAC abilit/ and the determination of the Itcensee to operate the facility in a safe and conservative manner. To andress these concerns the NRC performed a accordance with approved procedures and within the the facility's operating license. events. NRC concerns regarding ' the safe operation of the facility i heightened with the receipt activities at VEGP. of several allegations relating to operational i with the operational events and the allegations was viewed a 1 indicator of a non-conservative attitude on the part of the. facility's operating staff. This warranted the immediate initiation of special inspection activities. i Specifically, the inspection objectives were to: i 1) Assess' the operational philosophy, policy, procedures and practices of the facility's operating staff and management regarding operational safety. 2) Determine the technical validity and sa fety significance ' of the allegations and thstr impact on the safe and conservative operation of the faci lity. These inspection objectives were accomplished by the~ use of two inspection teams-an operations followup team and an allegations followup team. efforts of these two inspection teams were closely coordinated; however, they The independently pursued the objectives outlined above. The operations followup team monitored control room activities on a 24-hour basis. in order to: (1) evaluate the operational philosophy, policies, procedures, and practices of the operating staff and management and (2) determine if the plant was being operated in a safe and conservative manner in accordance with the facility's operating license. The allegations followup team verified the technical validity and safety significance of the allegations. staff, this team interviewed members of the plant staff in order to determ (2))their (1 their personal involvement and knowledge of the specific allegattens and These interviews were transcribed. practice and understanding of the station operatio Although an O! investigator was assigned to the laspection team to assist during the transcribed interviews, this NRC investigations may be implemented to further review th

e ci g ocirt 1 sero /II WANW71w=2* eau yogg 2 described in the allegations.The inspection substantiated the occurrenc* However, most of the allegations were not and one apparent violation (50-424,425/90-19-12) substantiated discussed in part in this inspection report supplement and two violations were idenstfled 425/90-19-01 and 50-424.425/90-19-02).'in-the initial part of this inspection identified as non-cited violations (50-424/90-10-03 and 50-425/90-0 i The operations followup team identified several occasions where responsible managers and supervisors verbally inspection team during the inspection. supplied inaccurate information to the Additional observations and conclusions of the inspection team are detailed in NRC-Inspection Report 50-424,425/90-19 issued January ll,1991. The bases fo these previous conclusions are summartzed below. Doerational Policies and Practices NRC Inspection Report 50-424,425/90-19 affect the operation of the facility. licensee's operational policie The allegation followup team's review of the allegations identified the following additional exam affect the safe operation of the facility: 1) The Itcensee's method of conducting Plant Review Board (PR8) acetings h the potential for adversely affecting open discussions among the PRB members. asaber felt intimidated and feared retribution during a because of the presence of the general manager and the absence 'of dissenting opinions in the PR8 neeting minutes. is necessary to ensure that PRS members freely and openly empress t technical opintons and safety concerns. (Section2.7) 2) The licensee's practice of signing and dating quality assurance records was contro11eo by administrative procedures; however, there was a date of performance. confirmed example in which a signature was backdated t j The backdating issue was verified and is identified i as an example of Violation 50-424,425/90-19-13: Implement Procedures for Required Activities." (Section 2.3)" Failure to Est 3) The licensee's practice of not initiating a deficiency card (DC) during troubleshooting activities involving the questioned operabtitty of the residual heat removal (RHR pump prevented a documented en evaluttfon for either the nu) lear service cooling water (N5CW) gine c i or the excessive vibration on the RHR pump motor. outlet leak implement this administrative procedure was toentified as an esseple of The failure to i Violation 50-424,425/90-19-13: " Failure to Estabitsh or Implement Procedures for Required Activities." (Section 2.2)

~ s.1 l 1 11 d 'scit Iseroe s i. -wAww ilw-r pag uon, j g.- 3 el 4) The licensee's method of appraising the performance ~ of the licensed operators resulted in a potential disincentive for identifying items which say result in LERs.or violations. (Section 2.8)- Accuracy of Information The inspection concluded that during the inspection inaccurate information was received on several occasions, from responsible managers and operators en topics well within' the scope of their specific responsibility. In four instan'ces the initial information supplied wa's clearly incorrect or inadequately researched. The inspection team. concluded that in each of these examples, licensee = officials provided inaccurate, unsworn,, oral statements concerning information which concerned topics well within their responsibili-ties. In two cases, the inaccurate information was clearly significant to the inspection process. Specifically. (1) if the containment isolation valves received an automatic closure signal, the valves could remain open without a violation of T5 3.6.3, and (2) if the snubber modifications had been performed in conjunction with other preplanned preventive and corrective maintenance, then the voluntary entries into LCO 3.7.8 would not have been required. The inspection team identified that the failure to provide accurate information is e violation of the requirements of 10 CFA 50.9 concerning accuracy and completeness of information. This is identified as an apparent Violation 50-424, 425/90-19-12: " Failure to provide Accurate Information to the 81RC as Required by 10 CFR 50.9*, as noted by the following examples: 1) Containment 1 solation Valves: During a Unit 1 surveillance procedure, the unit shift supervisor (U33) stated, and the ' operations manager later confiteed, that the containment isolation valves for the hydrogen eenitor system were allowed to. be opened without entering the LC0 action requirements for T5 3.6.3 because the - valves received an avtamatic isolation signal. The inspection identified that these containment 1 solation valves were remotely-operated, manual valves without automatic isolation signals. (Discussed in Section 2.2.1.1 of Inspection Report 50-424,425/90-19 issued January ll,1991). 2) Snubbor Redvetion: The operations manager stated that, after Unit 1 refue'ing outage 1R2. the modifications to the snubbers were done in conjunction with preplanned system outages which were required for other preventive or corrective maintenance or testing. The inspection e identified that few of the snubber modifications were done jointly with pre-planned system outages. (Discussed in Section 2.1.1.4 of Inspection

1. Report 50-424.425/90-19 issued January ll,1991).

i l' 3) personnel Accountability: The operations manager stated that the shift superintendents 55 he personally p(rep)ared their performance appraisals. reported directly to t The inspection . ~- i i

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1 4 \\ i revealed that thel 55 reported to the unit superintenden't-(U5), and that the US personally prepared the performance appraisals of - the 55. (Discussed in Section 2.8 of this inspection report) 4. T5 3.0.3 Actions: The Unit Superintendent indicated that there were no Operations-Department actions which were. anticipated or reeutred within the first three hours of entering the action statement of T5 3.0.3. i The inspection identified that the VEGP management policy and statement practice reeutred preparations for a power reduction, including informing the load dispatcher within the first hour. (Discussed in Section 2.1.1.3 of Inspection Report 50-424,425/90-19 dated January ll 1991). In summary, this supplement of the inspection identified one violation, one apparent violation, and two inspector followup items. The violations include: . t (1) a violation of T5 6.7.1.a in that, two examples were identified of the licensee failing to implement actions in accordance with administrative procedures and (2) the apparent violation of 10 CFR 50.9 which relate to four examples in which responsible licensee officials provided inaccurate information to the NRC during the inspection. The two inspector followup items include: (1) an unreviewed safety question concerning the use of the alternate radwaste building, and (2) the lack of operator guidance concerning the acclicable limiting conditions of operation during engineered safety features actuation system sequencer outages. i P l e 4 l l .-----------------_________--_-__________,--e-v--- we-.- q n,- v-g m---w c--rs

l m ct d test Isero/tI w1 Nee M w.y.g 3,, g, j. L . /. y g I n IN5pECTION DETAlt.$ 1.0 INSPECTION OBJECTIVES p L Recent activities which have occurred at VEGP have raised. concerns within l' the NRC as to the ability and the determination of the licenses to operate the l facility in a safe and conservative manner. To address this concern, the NRC performed a special team inspection to determine if the licensee operates the facility in accordance with approved procedures and within.the redvirements of the facility's operating license. In addition to the occurrence of specific events, NRC concerns regarding the safe operation of the facility were heightened with the, receipt' of several allegations relating to operational activities at VEGp. The aggregation of the facts and circumstances associated with the operational events and the allegations was viewed as a possible ' indicator of a non-conservative attitude on the part of the facility's operating staff which warranted the immediate initiation of special inspection activities. Secause a non-conservative attitude or operating philosophy may represent a hazard to the health and safety of the public, a special inspection team comprising staff from the Region !! Office end the Office of Nuclear Reactor Regulation (WAR), assisted by staff from the Office of Investigations (01), was formed to determine the individual validity and cc11ective moact of these allegations on the safe operation of the facility. The purpose of the inspection was te determine if the Itcensee operates the facility in a conservative and safe manner in accordance with approved precedures, and the requirements of the facility's operating ifcense. Specifically, the inspection objectives were to: 1) Assess the operational philosophy, policy, procedures, and practices of the facility's operating staff and management regarding operational safety, and 2) Determine the technical validity and safety significance of each of the allegations and their impact on the safe and conservative operation of the fac111ty. These inspection objectives were accomp11shed by the use cf two inspection teams--an operations followup team and an allegations followup team. The efforts of these two inspection teams were closely coordinated; however, they independently pursued the objectives outlined above. The operations followup team monitored control room activities on a 24-hour basis in order to: (1) evaluate the operational philosophy, policies, precedures, and practices of the operating staff and management and (2) determine if the plant was being operated in a safe and conservative manner in accordance with the facility's operating license. -A

't : d cciel tseroe i www9 w-s m u n., g The specific inspection activities of the operations team was described in Inspeetfon Report 50-424.425/90-191ssued January 11, 1.991. The efforts and conclusions.of the allegations followup team are described in this supplement to that inspection report. In addition, this supplement identifies several violations and potential weaknesses in the Itcensee's operational polices and practices. Specific details are contained in the sections that follow and in .the Inspectier. Summary. i 2.0 ALLEGAT]ON FOLLOWUP The inspection team reviewed several allegations for their technical valtetty and interviewed 1(censed and non-Itcensed personnel to determine their personal knowledge and experience regarding these issues. This portion of the inspection was performed to determine the validity and significance of the allegations. The inspection of the allegations included technical reviews of the Ifeensee's records. logs, and interviews of the personnel involved in the alleged violations. Although a transcribed record was not reoutred for every discussion with the Itcensee's staff, the inspection team conoscted sworn. transcribed interviews with selected f adividuals in order to document (1) the individual's personal knowledge and involvement in the alleged violations and (2) the circumstances and ratf or. ale for tneir incividual actions. Although an O! investigator was assigned to the inspection team to assist during the transcribed interviews, this inspection was not an invest 1pation into the i intent of the alleged violations. The intent aspect of the a leged activities may require further NRC investigations. l The interviews were transcribed after the technical evaluations of the ellegations la order to permit a focused interview and to sintof te the legth and scope of the transcribed proceedings. The transcribed. f aterviewees are listed in Appendia 1 in the order in which they were conducted. The sworn testimony was a factor on which the inspection team reached its conclusion on each of the allegations. These conclusions are presented in the material that l i follows (Sections 2.1 through 2.8). 2.1 Laprooer Installation of FAVA System I An allegation indicated that VEGP installed and operated a radweste microfiltration system, known as the FAVA system, without performing an adequate engineering and safety evaluation (i.e.,10 CFR 50.59). Furthereore, i the material configuration. fabrication and quality of the system did not sett L the guidance of Regulatory Guide (RG) 1.143 and the requirements of the i American Society of Mechanical Engineer's (ASME) Code. ~' The FAVA system was temporarily installed for'removin was later determined to be better suited for' as-low g Nfobtun-95. The systen l as-reasonably-achievable considerations during refueltag estage 3R2 particularly for removing Cobalt *l9 e and Cobalt-60. VIGp planned to replace this temporary modtfication with a i permanent, high-quality, steel system in the future; however, the health and j + \\ l

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+ safety of the public may be jeopardized if a break in the system (resulting in a radioactive release to an unrestricted area) occurred in the interin. i Discussion i t In February 1988 VEGp expertenced difficulty in removing colloidat Niobium-95 following a reactor shutdown for maintenance work. FAVA Control Systems because o(FAVA) was hired to help rectify this problem. FAVA was selected f its experience in filtration and denfperalization. The situation i was corrected by installing a 0.35-micron filter system downstream of the existing vendor-supplied pre-filters. However, a large volume of reewaste was generated as the 0.35-eferon filters rapidly exhibited high differential pressure and were required to be changed frequently. The need to change filters frequently also resulted in additional radiation exposure to Radwaste i Department personnel. Upon evaluation of the performance of the 0.35 micron filter system, the Radweste Department felt that the best approach to the problem was a back-flush, pre-coat filter system. However, no operational data was available for a system of this type in this specific application. FAVA supplied a proprietary Ultra Filtration System (Model No. 5FD/E) for testing purposes in l order to evaluate whether or not this was a viable and economic solution to the I problem. The FAVA syster. was irstalled before the Unit ) refueling outage and was operated under Test Procedure T-OPER-8803. The test system kept liquid i effluent releases well below TS Itatts. On the basis of an evaluation of test results by the Radwaste Chemistry, and Engineering Departments, a general work i order was initiated to purchase a permanent system. t In the early part of 198g, a Quality Assurance (OA) Department audit identified a significant audit finding involving a programmatic breakdown in the procurement of the FAVA system and the failure to meet commitments of the Final Safety Analysis Report (F5AR). Because of that finding, the FAVA system was removed from service. In late 1989, the licensee sought to reinstall the FAVA l system under a temporary modification because colloidal Cobalt-59 and Cobalt-60 i had to be removed. The Plant Review Board (pA8) reviewed this temporary modification and several members espressed strong objections to it based on the previous OA audit finding. 1 subsequently, a request for engineering assistance (REA) was submitted and a 10 CFR 50.59 safety evaluation was performed in late 1989. This safety evaluation did not properly address the guidance of RG 1.143 regarding the use of polyvinyl chloride (PVC) piping. Therefore, another safety evaluation.was performed in February 1990 to address this issue-parttevlarly with respect to radt,ation degradation. ThEFebruary 1990 safety evaluation specifically stated that the FAVA system did not confors to the criteria of RG 1.143. This deviation was found to be acceptable for the following reasons: I u. - ^ ^ ^ ~

i st*w sciri iseroetI wmwww-r ens unus t 1) The design of the FAVA system had beert previously evkluated and-found to be adequate in the response to REA VG-9057 dated November 28,1983 (log 3G-8592). 2) The location of the FAVA microfiltration system inside a shielded waterttght vault proviGed adequate assurance that any system failures w,111 be contained and would not create the potential for offsite releases of radioactivity. 3) The presence of PVC pipe in the FAVA system, although prohibited by RG 1.143, was acceptable because the radiation esposure to the plastic was l within acceptatsle limits for up to 6 months based on the following: l 4) The amount of PVC piping used was not extensive and was contained on the FAVA filter skid. i b) There were no reported leaks or malfunctions during the approximately 6 months that the FAVA system filter was previously in use. 1 c) Since the FAVA system filter skid was located within the dominera112er vault, it would be protected from being damaged. c) On the basis of the assumed length of time that the PVC piping would be used in a radioactive environment and the activity levels of the effluent at this stage in the liquid radwaste process, the integrated dose to the PVC piping would be well below the radiation damage threshold for PVC pipe as reported. in Electric Power Research Institute (EpAI) Report NP-2129, dated November 1981 (i.e., 6.5 rad over a 6 month period versus the radiation damage threshold of 5.0 a 5 10 rad). e) The PVC pipe would not be subjected to excessive pressure conditions 4 since the maximum available inlet pressure to the filter was 80 to 100 pounds per square inch gauge maximum allowable working pressure o(psig) which is 'well below the f 120 psig for the PVC pipe. f) The system could be operated at design-basis conditions for 182 days before it would exceed the radiation damage threshold.

However, under conditions currently existing at the plant, the expected dose to the PVC piping will be less than 0.1 percent of the design basis.

Although the testimony of one of the PR8 members indicated that the temperature effects on the use of PVC in the FAVA System were not adequately evaluated before the system was installed, the testimony of the corporate system engineer indicated that this was considered prior to installation although not specifically documented in the safety evaluation. 1 r-,

e i At*w scart is/rO/11 U1NW71W-3 *Bla yoy g y 'V!GP sanagement substovently consulted the NRC resident Inspector to seek 'an .NRC position with regard to placing this system back in service.. Supplemental I;; formation was also provided documenting reasons why it should not be placed ta service. Reactor Regulation (NRR) for review.This package was forwarded to Region 11 and -i In March 1990, following Region !! and NRR concurrence via a telephone conference, the licensee placed the FAVA system i in service with the following NRC stipulations: 1) procedures for operating the FAVA system required an operator to be in attendance for the entire length of time the system would be in operation. 2) All hoses going to and coming from the FAVA systea required verification that they met the requirements of RG 1.143. i 3) The cover over the FAVA system was required to be securely fastened when the system was in operation to ensure that if a spraying leak developed, it would be contained in the concrete vault. 4) The design of the walls of the alternate radwaste building (AR8) was required to be evaluated to determine whether or not a design modification should be made to reduce the potential of wall leakage in the event that a j hose leak developed and sprayed its contents on the walls. i In June 1990, in response to item 4 (above), the licensee revised Part G of the safety evaluation for the FAVA system. part G of the safety evaluation addressed the effect that operation of the FAVA system would have on the probability of occurrence or consequences of accidents described in the F5AR. Although there was no camparable accident analysis in the FSAR that addressed t l the ARB accidents or the consequences of accidents in the ARS, 'the FSAR accident analyses (Chapters 15.7.2 and 15.7.3) did describe worst-case releases of the contents of the recycle holdup tank (HUT).

The first bounding analysis in Chapter 15.7.2 addressed the release of the entire gaseous radioactive contents of the HUT to the environment at ground level and the second bounding analysis addressed the release of the entire._

liquid contents of the HUT through an assumed crack in the ARS floor directly into the ground water supply. part 20 limits were not exceeded.In both cases, the 10 CFR part 100 and 10 CFR These criteria were consistent with criteria provided in NRC Circular 80-18, "10 CFR 50.59 Safety Evaluations for Changes to L Radioactive Waste Treatment System." However, neither of these analyses i L addressed the potential for wall spray down and leakage tnrough the ARB walls L and the subsequent release path to the environment. Therefore, the licensee i revised the safety evaluation in June 1990 to address the consequences of 'a hose $reak on the FAVA system which would result in well spray down and poten,tial leakage to the environment. The inspection team's review of the revised Part G of the safety evaluation y identified several erroneous assumptions with respect to the release path and I -the ettution volumes that could be used in the analysis of a hose break and resultant wall spray down. However, the inspection team also fevnd that the design of the FAVA system (i.e., the use of a system cover) would prevent well [;

.~ i 'e. 0 *e Iw101 sd As: Fl la/poptj w1Nw"llw.3 g g p 10 s 9 4 spray down and that the only potential source for wall spray down and subsequent leakage was from a hose break in another radwaste ; system in the AR$. Therefore, the inspection tese concluded that the FAVA systes safety evaluation dated June 1990, adequately addressed the temporary modif'ication for the -installation of the FAVA systes; however, the inspection team's review identified an unreviewed safety question concerning the release paths and consequences of a failure of the other radweste systems in the ARS. In_ addition, the team noted that. in $upplements 3 and 4 of the Safety ' Evaluation Report ($ER), the NRC -staff reviewed and accepted the design of the ARS and specifically addressed the consequences of a hose break on a radweste i system in the AR8. However, the 5ER supplemnts addressed the offects of high airborne activities and puddiing and did not address the potential for wall spray down and leakage. The AR8 was installed before the plant was licensed; therefore, the NRC approved the design and use of the AR8 in Supplements 3 and 4 of the SER. Thus, there was no requirement to perform another evaluation of the potential effects of hose breaks on systems other than the system being installed by the temporary modification (i.e., the FAVA system). Because the design of the FAVA system ef fectively prevented a wall spray down, this was not-a ' concern that was required to be addressed by the FAVA system safety -cvaluation. Nevertheless, now that it ' w been identified, the consequences of a hose break and wall spray down in the other AR8 radwaste systems must be resolved. Therefore, this issue will be followed as an inspector followup item - pe ndi rig further review and evaluation and is identified as IFI 1 50-424,425/90-19-14:

• Potential unreviewed Safety Question Regarding $ pray Down of the Alternate Radwaste Building."

Cone 19sion Although the FAVA system was ' originally installed without an adequate safety .i evaluation and did not meet the regulatory guidance, the inspection team confirmed that the subsequent safety evaluations were acceptable for the l system's use. As a result of QA Department's significant audit finding in early 198g involving 'a breakdown in procurement and failure to meet FSAR commitments, the system was removed from service. Subsequently, the FAVA system was returned to service following two' safety evaluations which adequately addressed the use of PVC piping with respect to radiation degradation and pipe rupture. Therefore, these safety evaluations Justified the use of the FAVA systes, even though the recommendations of RG 1.143 and ASME Code requirements were not met..Although the safety evaluations did not specifically address high-temperatun effects, the testimony indicated that these effects had been considered before the l system was installed. l Although the safety evaluation performed in June 1990 at the request of the WRC P.egion !! Office did not adequately evaluate the effects of a wall spray down and. wall leakage to an unrestricted area, this evaluation was not retvired because the FAVA system has a protective cover and the use of hoses and effects l of hose breaks (i.e., airborne activity and puddling) were addressed in SER Supplements 3 and 4. 1

1 1 - 3 'd or8ri Isero<ti v29,31,,7 Il I i Regardless of whether the safety evaluation was required to address'the effects of a bresh in the hoses (which could result in wall spray down or leakage), the f inspection team identified a new concern involving the use o'f the ARS because i the safety evaluation inadequately addressed the potential. effects of wall spray down from any other source in the ARB owing to erroneous assumptions concerning the release path and the dilution volumes. This issue associated with the potential effects of wall spray down in the ARS should be reviewed by the licensee under 10CFR50.59 requirements. 2.2 Oscreo111ty of the Residual Heat Removal Pues i An allegation indicated that during Unit I refueling outage IR2 with residual i heat removal (RHR) Train A out of service for maintenance, the Train B RHR pump i esperienced excessive vibration and a nuclear service cooling water (N$CW) i noter cooler outlet leak. In addition, TS 3.9.8.1, "RHR and Coolant l Circulation " was allegedly violated bacause the Operations Department chose not to declare RHR pump 18 inoperable in an effort to mitigate the impact on [ the critical work path. lj 3 i! gijgssion TS 3.9.8.1 requires at least one RHR train to be operable and in operation during Mode 6 (refueling) when the water level above the top of the reactor vessel flange is greater than or equal to 23 feet or more. Otherwise. .i Suspend all operations involving an increase in the reactor decay heat load or a reduction in boron concentration of the reactor coolant system and immediately initiate corrective action to return the required RHR train to operable and operating status as soon as possible and close all containment penetrations providing direct access from the containment atmosphere to the outside atmosphere within 4 hours. The inspection team verified that during Unit I refueling outage 1R2 with higher than normal vibration measurements on the RHR pump 18 and a leak on the NSCW outlet of the RHR motor cooler, Operations Department personnel did not declare the pump inoperable. This determination was made after consulting with the on-shift duty engineer from the Engineering Department and was based on the determination that the pump would fulfill its intended safety function in Mode i 6. Specifically, the RNR pump was capable of removing decay heat from the partially defueled reactor core. The testimony of the individuals involved indicateo that this operability determination was based on the fact that the vibration readings taken at the inservice test (!$T) surveillance points did not reach the IST Alert levels and were therefore acceptable for continued service. Although the high vibration readings on the top end of the RHR pump were later determined tty the vendor (Westinghouse) to be excessive, at the time of the operability evaluation, the licensee accepted these values, regardless of 'their magnitude, because the readings at 15T test points were below the Alert levels. The testimony also 7

3' N IP8PI 15/roeit v2seig,g. q l 1 12 i indicated that, even with a leak on the NSCW outlet of thei RHR motor cooler, i the actor was receiving full cooling water flow and cooling would not have been immediately compromised following a complete N5CW discharge pipe break. Furt'nermore, the testimony indicated that the Operctions Department had impleinented compensatory actions - to monitor the vibration levels and N5CW leakage and ensure the continued operability of the pump by stationing an operator at the RHR pump to monitor the vibratton levels and notify the centrol i room if the vibetion levels increased, thus allowing the control room to implement the actions of the Limiting Conditions for Operation (LCO). The inspection team alsa noted that in event of a catastrophic fatture of the RHR pump, all the requirad actions of TS 3.9.8.1 (i.e., closing all containment penetrations) could have been completed within the required 4 hour time persed of the LCO because the LLO for TS 3.9.4, " Containment Building Pene.trations," was in effect during this time period. This LC0 was implemented due to the l movement of irradiated fuel f rom the core to the spent fuel pool. The LC0 required that, The equipment door be closed and held in place by at least four bolts; at l 1 east one door in each airlock be closed; and each penetration providing t direct access from the containment atmosphere to the outside atmosphere shall be either closed by an isolation valve, blind flange, or sanual valve, or be capabl6 of being closed by an operable automatic containment ventilation isolation valve. As a result of the implementation of T5 3.9.4, the only remaining action for the LC0 of T5 3.9.8.1 would have been to close the containment purge valve I which receives an automatic closure signal and could have been isolated within the LC0 action times. During the course of this review, the inspection team found that the licenses failed to initiate a deficiency card for either the N5CW 1eak or the excessive vibration as required by Operations procedure 00150-C. " Deficiency control." i This procedure requires that a deficiency card be written if the deficiency involves safety-related components which are to be dispositioned i "use-as-is/ repair," or other conditions involving safety-related components l which require engineering support or other technical assistance to determine if the component is deficicnt. Failure to establish, implement, and maintain adequate operating procedures represents a violation of 15 6.7.1.a. This ites 1s identified as an example of Violation 50-424/90-19-13: " Failure.To Establish I or Implement Procedures for Required Activities." ~ Conclusion i The inspe: tion team confirmed that the Operations Department had an edesoste engineering basis for accepting the operability of the RHR pump in spite of the pump's deficiencic1 In addition, the team concluded that declaring the pump l inoperable uov1d not have impacted the critical work path: the LCO actions would not have been restrictive because contaireent (escluding ventilation) had been isolated as required by T5 3.9.4. The LC0 actions would not have l

.~ n. T 'd 2*8et.tseroeIt wAny,4y.g. g 13 . prevented the continuation of refueling activities because the actions to close all containment penetrations providing direct access from the containment atmosphere to the outside atmosphere would only have reeutred closing the containment purge valve which has an automatic closure signal. l In addition, the inspection team identified that the licensee violated the i station's administrative procedures by failing to initiate a deficiency card for either the N5CW outlet leak or the excessive vibretten on the RNR motor as. .I required by Operations Procedure 00150-c. ] 2.3 Backdatino of $1onatures An allegation indicated that a temporary change to Abnormal Operating Procedure l (A0P) 18028-C, " Loss of Instrument Air," was not approved within the 14-day requirement of TS 6.7.3.c; and that the unit superintendent intentionally incorrectly signed and dated the temporary change to indicate-that the T5 requirement was satisfied. Discussion 1 i TS 6.7.3.c requires that temporary changes to ADPs which do not involve changes l to the intent of the original procedure be documented and reviewed in accordance with TS 6.7.2 and approved within 14 days of implementation. TS 5.7.2 requires stat changes to A0Ps be reviewed as stated in administrative I procedures and approved by the PR8 and general manager. Administrative procedure 00100-C, " Quality Assurance Records Administration," Paragraphs 4.1.1.4 and 4.1.1.8, require that corrections to Quality Assurance records eshibit necessary and appropriate signatures, initials, and dates. Operations Procedure 18028-C, Revision 7, provided operator actions in the I event of a loss of the instrument air system. A temporary thange to the procedure was initiated on May 29, 1990, to delete the references to the header i isolation at 70 psig and the associated actions,. This change was processed in accordance with Administrative Procedure 00052-C, " Temporary ' Changes to o Procedures," which allowed the temporary implementation of minor changes to l procedures as long as the change was approved by the PR8 and signed by the general manager within 14 days of the temporary change. Therefore, Temporary i Change Procedure (TCP) 1802-C-7-90-1 was required to be approved by the ptB and signed by the general manager by June 12, 1990. The PRB tabled the TCP on June 8,1990, (pRB meeting 90-81) and assigned action i to the Operation's Department to void the TCP or revise the TCP to incorporate the PRS comments. Revision 8 to Operations Procedure 18028-C was developed to modify valve numbers and descriptions reflected in Temporary Modifications i 1-90-006 and 2-90-002. This revision superseded the changes of the TCp. On June 12,1990, the PR8 approved kcvision 8 (pR9 meeting 90-82) and the TCP was removed from the control room copies of the procedure. On June ll,1990, the unit superintendent lined out the operations manager's previous approval of the TCP and marked the TCP form as disapproved by the Operations Department. The date entered on the form was June 12, 1990.

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p 14 On June 22, 1990, the PRS secretary initiated PC 1-90-242 wnich indicated that the unit superintendent incorrectly dated the TCP with the.date of June 12, 1990, rather than the actual date of June 15, 1990, and DC 1-90-283 which indicated that the TCP was not processed within the required 14 days (i.e., by June 12,1990). The resolution of these DCs, the associated PA8 meeting minutes, and discussions with the operations manager and Nuclear Safety and Compliance Department staff indicated that described deficiencies were ) acknowled9ed and confirmed by the Operations Department on July 3,1990, and i attributed to personnel error. The TCP form was dated with the date on which

J the Operations Department decided to void the TCP and not the date on which the original was actually signed.

As part of the corrective actions for DC 1-90-282, a TCP record correction notice was initiated to correctly indicate the case on which the TCP form was processed; however, the TCP record correction notice could not be produced--one was subsequently written on August 14, 1990. In addition, the operations manager counselled the unit superintendent and assigned him to invest 19 ate both DCs because he was the most knowledgeable of the deficiencies and the assignment served to reinforce the reprimand. The subseeuent PRB meeting of June 28,1990, (PRB meeting 90-90) determined that the 14-day T5 violation addressed in DC 1-90-283 was reportable to the VEGP vice president, but not to I the NRC. However, the inspection team found that the report to the VEGP vice l president was not made. On August 9,1990, the PR8 (PRB meeting 90-104) confirmed that the report was required. As of August 17, 1990, the licensee had not issued tne required report to the VEGP vice president; however, the itsensee intended to issue the report. With respect to the rationale for the unit superintendent's actions, the inspection team learned (during discussions with the Technical Support flanager) that the PR8 secretary told the unit superintendent on June 15,1990, that the TCP needed to be voided and a DC written for violatin9 the 14-day reeutrement of TS 6.7.3. As discussed in Section 2.8 of this inspection report. Operations Department personnel are held personally accountable for violations and LERs (i.e., there is a direct impact on their bonus pay) therefore, a reportable i i occurrence based on this event could have adversely impacted the unit superintendent's salary. The testimony 6f the unit superintendent indicated that he dated the TCP with the date (June 12,1990) on which the PRS disapproved it and not the date on which it was actually signed (June 15, 1990). Additionally, the unit superintendent had no recollection of any discussions on June 15, 1990, l i regarding violation of the 14-day TS requirement. He indicated that he never considered the 14-day requirement despite his previous knowledge and training concerning this requirement and the June 12, 1990, expiration date indicated on the,'TCP form. j I The testimony of the PR8 secretary indicated that during a discussion with the { unit superintendent on June 15, 1990, she identified the need to void the TCP, as well as the need to write a DC for vtoisting the 14-day T5 requirement. i

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l 15 ~ Therefore, the inspection team was concerned about whether :the TCP was voided before er after the PR8 secretary identified the need to void the TCP and initiate a DC. In order to resolve this discrepancy, the inspection team discussed the discrepancy with the PRS secretary on August 16, 1990. In l addition to earlier testimony, the PR8 secretary indicated that during her i discussions concerning the TCP with the unit superintendent on June 15, 1990, the unit superintendent had indicated that the 1CP had already been veided earlier in the day. Conclusion On the basis of the statements of the US that he had dated the TCP based on the PR8 disapproval date and not the date which he signed it, the inspection team concluded that backdating to avoid a violation of the 14-day T5 require-ment was not substantiated. In addition, the concern that this practice was a plant wide problem was not substantiated. However, the inspection team did-confirm that TCP 1802-C-7-90-1 had been dated incorrectly; this is a violation of Administrative Procedure 00100-C, " Quality Assurance Records Administra-tion " Paragraphs 4.1.1.4 and 4.1.1.8 and will be identified as an example of Violation 50-424.425/90-19-13: " Failure to Establish or Implement Procedures for Raquired Activities." 2.4 Reportacility of Previous Enaineered $sfety Features Actuation System II5FA5) Load leavencer Dutanes An allegation indicated that the Operations Department incorrectly used a 72-hour shutdown requirement when one of the two E5FAS load seguencers was previously inoperable. It was also indicated that VESP had taken no action to ensure that the past occurrences were identified and reported to the NRC as required by 10 CFR 50.73, despite newly acquired information that doenergining an E5FA5 sequencer required entry into the 1 hour LCO action requirements of T5 3.0.3. In addition, the possibility existed that the LC0 for T5 3.0.3 (i.e., 7 hours to hot standby) were exceeded when the sequencers were previously deenergized for maintenance and testing. This concern was based on (1) the lack of a specific T5 for the sequencers. (2) the Operations Department historically linking the sequencer outages to the emergency diesel generator (EDG) LC0 of T5 3.8.1.1.b (78 hours to hot standby). (3) a Ilmited review of past maintenance work orders (MW0s) indicated possible sequencer deenergina-tion; and (4) comments by the engineering staff that the sequencers had been previously deenergised. Discussion There are two E5FAS sequencers for each unit--one for each 4.16-kilovelt (kV) bus. Each sequencer is activated by one of two conditions, emergency undervoltage (UV) on the associated emergency bus or a respective train's safety injection (51) signal. Upon receipt of either or both of the initiating signals, each sequencer will perform a11 or part of the following functions: Start the associated EDS Stop any test seque' ace in progress

- _~ >c l w I FP8FI IS/P0/II ' w1NW"Ily-3.p3W WOW A .e 1s ~ Close the associated EDG breaker (UV only) Energire the associated train's engineered safejy features (E5F) loads as determined by the initiating renal. Each ESFAS sequencer contains three levels of UV detection and system response, as well as the power supply for this UV circuitry. Four potential transfereers monitor the emergency bus voltage for these three levels of degraded bus l < 70 percent; Level 2, t 86 percent; and Level 3,

• 88.5 voltage (Level 1, l an analog signal to three sets of four bistables located percent) and furnie in one of the five sequencer cabinets.

.l Level 1 is the " loss of voltage" and Level 2 is the " degraded voltage" which is i referred to in T5 Table 3.3-2, Items 6.d. 8.a, and 8.b. As these T5 items i (applicable in Modes 1 through 4) do not address the loss of all four channels in Level 1 or in Level 2 (as would be the case when the sequencer is i deenergized), T5 3.0.3 would apply if such a loss were to occur. It should be 3 noted, however, that if the sequencer were deenergized, it could not respond to a safety injection signal either. Therefore, there would be only one automatic safety injection actuation channel (i.e., associated with the unit's unaffected sequencer) and Item 1.b of T5 Table 3.3-2 (6 hours to het standby) would be the i most limiting LCO. Cis:ussions with the operat'icns manager, the assistant general manager plant support, and system engineers for the ESFAS and sequencers confirmed that the Operations Department historically linked the seovencer outages to the EDG LC0 of T5 3.8.1.1.b (78 hours to hot standby). Although the applicability of TS Table 3.3-2 and 75 3.0.3 to sequencer outages had been recently identified, past sequencer outages were not reviewed. Therefore, with the assistance of the licensee, the inspection team reviewed the completed led 0s which were performed on the sequencers en Units 1 and 2, as well as the related Instrumentation and Control (!&C), Engineering, and Operations Department surveillance tests. The. review of completed MW0s did identify several instances where the work performed would most likely require the sequencers to be 6eenergized; however, the associated unit was found to have not been in Modes 1, 2, 3, or 4 at the time-the work was performed. Somewhat related to this concern, the review did identify two occurrences (March 4 and June 17,1987) where the Unit 1 Train 8 sequencer was inoperable during the change of soevencer controller card A (SLOT A4-3). Specifically, when the controller card was removed, both the automatic $1 function and UV function for the sequencer were rendered inoperable. Because the unit was in Mode 3 (het standby) during these two occurrences, the soevencers and the E5FA5 were required to be operable per T5 3.3.2. News'er, the assectated LC0 status sheets (1-87-356, sated March 4,1987 and 1-87 44, dated June 17,1987) only recognised T5 LC0 3.8.1.1.b as being applicabi to the outage. Despite the fact that LCOs associated with T5 Table 3.3-2 (!tes 1.b) and T5 3.0.3 were not recognized, these T5 were not violated since the system was restored within 30 minutes and 10 minutes, respectively. I --a

j'd 8*8'I 18/roetI way uw m % 17 51stlar to the MWO review, the inspection team's review of related 18,C. Engtseering, and Operations Department's surveillance tests.did not find any esemples of the sequencers or the ESFA5 being deenergized in Modes 1 through 4. Completed 18-month ESFAS channel calibrations. EDG tests, and E5FA5 tests were i verified as having been done in Modes 5 and 6. Completed quarterly testing of the ESFA$ Auto 51 K610 slave relay, which removed the automatic $1 signal to J the sequencer, were verified to be performed within time limits allowed by T5 3.3.2. All other sequencer testing that used installed test circuitry is i automatically bypassed on en $1 or UV signal. In addition to the inspection team's review of MW0s and surveillance test procedures, the system en0'neers for the sequencers and E5FAS [as well as the cuclear steam supply system (NS$5) supervisor) were asked if they knew of any time in which the sequencers were deenergized in Modes 1 through 4. None of these engineers resembered any such occurrences. A review of applicable operator training material (System Description 8b for l Engineered Safety Features System Sequencers) revealed that there was no reference to E5FA5 T5 3.3.2 Just those for the diesel and other power sources and distributions (i.e., TS 3.8.1.1. T5 3.8.3.2, T5 3.8.2.1. TS 3.8.3.1, and T5 f 3.5.3.2.). This finding, along with the March 4 and June 17, 1987, occurrences discussed above, indicates that the Operations Department historically has not linked sequencer outages to the LCOs of TS 3.3.2 or TS 3.0.3. Nevertheless, 1 t discussions with the operations manager and the licenced operators on shif t indicated that although no written guidance or TS interpretation existed for j the sequencers, the Operations Department staff would currently consider all i applicable TS requirements, including T5 3.3.2 and 3.0.3. Conclusion The LCO actions of T5 Table 3.3-2 "E5FAS Instrumentatten,' are applicable for determining the operability of E5FA5 components; however. if a lead sequencer is not operable, the more restrictive requirement of 75 Table 3.3-2, TS 3.0.3, t' or the affected system LE0 should be considered. Although the EDG LC0 of T5 I 3.8.1.1.b had been used for sequencer outates in the past, the allegation's concern of possibly exceeding the LC0 for 15 3.0.3 when the sequencers were previously etenergized were not confirmed. Because there is no specific T5 for the sequencers and considering (1) their unique interaction with numerous other systems and equipment, and (2) the varying degrees in which related failures, maintenance work, and surveillances can affect the sequencers' associated functions, the inspection team concluded that additional guidance for the operators is warranted. Therefore, this issue will;be followed as an inspector followup item pending further review and l evaluation and is identified as IFI 50-424,425/90-19-15: " Lack of Operator Guidance Concerning the LCO Actions Applicable During E5FA5 Sequencer Outages." t I

i i J j sd spies ignoen v4wwize.risse uovg.. j 18 2.5. Air Quality of Emersency Diese1 Generator Startino Air System An allegation indicated that VEGP had -no basts for its conclusions regardl t,he air quaitty ef. the EDG starting air system and misrepresented the air quality in the Itcensee's written response to. the Confirmation. of Action Letter (CAL) dated March 23. 1990. i S Discussion i The inspection team reviewed the maintenance records and deficiency cards associated with Unit 1 EDG starting air system. The team noted that the maximum dewpoint reading of 50 degrees Fahrenheit was established when preoperational tests were initially performed on Unit 1 in November 1986. Dewpoint measurements were taken after this date but not en a scheduled frequency. During the latter part of 1988, a monthly preventive maintenance i (pM) schedule was estab11shed to measure the EDG starting air systes dewpoint. The current PM program required checking the dowpoint monthly, cleaning the air dryer condensing units, and cleaning the fan motors. i In addition, Operating i Procedure 11882-1 "Dutside Area Rounds," required that the EDG starting air system air receivers and air dryers be blown down on a daily basis until they were free of moisture. j The inspection team verified that the plant equipment - l operators blew down the air systems on each shift during the performance of their rounds. i A review of the Unit 1 EDG maintenance history records indicated that the i majority of the dowpoint measurements taken were within specifications. There were instances, however, when the dowpoint measurements were above specifica-tions. These conditions were primarily attributed to problems with-(1) the i dowpoint measuring instruments. (2) system air dryers bei eut of service for i extended periods of time, and (3) repressurfsing the air start system l following maintenance. .t The insoection team reviewed maintenance records associated with an int i inspection of the EDG air start system air receiver. 5-micron control att system filter inspection and replacement, and the replacement of the dowpoint measuring instrument with an EG4,G' analyzer. i Following the loss of offsite i power event of March 20, 1990, the control air system instrument Itnes were disconnected for maintenance troubleshooting and functional tests of Calcon a The systes engineers associated with this work stated that ne j sensors. evidence of internal soisture or corrosion was noted during inspection and calibration of the Calcon sensors or the control air system instrument lines 1 when this equipment was disconnected for maintenance troubleshooting and testing. Conclusion The inspection team concluded that the Itcensee did have an adequate basis to essess the quality of the EDG starting atr system. This was based primarily upon the records of the visual inspection of EDS atr start system components for degradatten. In addition, the PM program Gewpoint readings have shown more coasistency since the licenses changed over to an E04,G analyzer. The allegatten +r* -m# eWwr= =L-e e

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4 19 ~ that GPC did not have a basis for their statements and misrepresented the air quality in the licensees written response to the CAL, was not confirmed, i 2.6 Resortablitty of previous System Outanes i t An allegation indicate'd that VEGP failed to immediately notify the NRC as required by 10 CFR 50.72 when VEGP identified that both trains of the 1 containment fan coolers (CFCs) had been previously inoperable at the same time on Unit 1. Discussion i The inspection team's review of plant records indicated that this condition occurred when EDG #1A was declared inoperable when tape (used when the EDG was being painted) was found on the (DG fuel rack. piston from moving and injecting fuel into the EDG.The tape kept the fuel injector With EDG #1A inoperable,. the equipment associated with the Train A was also inoperable. In the process of investigating the installation of the tape,.VEGP identified that this condition existed during a period when the Train 8 containment fan coolers were also in a degraded condition for maintenance. During the performance of Surve111aue Procedure 14623-1, Train 8 containment i fan cooler (CFC) 1-1501-A7-003 failed to start in slow speed. LCO 1-90-560 was initiated at 1:15 a.a. on June Ig,1990, and maintenance en the CFC was-initiated. The CFC was returned to operable status en June 19 1990, at 2:15 p.m. Approximately 9 hours later (on June 19,1990, at 11:59p.m..(LC0 1-90-562)), EOG #1A was enormined to be inoperable because the tape had been i installed on the fuel rack. On July 17, 1990, VESP issued LER 90-014 to identify the previously unrecognized violation of the LC0 in accordaece with l 10 CFR 50.73. Conclusion Based upon the fact that VEGP did not become aware that both trains of CFCs were sinvitaneously inoperable until after the Train 8 CFC fan had been returned to service, the immediate notification reevirements of 10 CFR 50.72 i were not appitcable. The allegation that VEGP fatied to immediately notify the 3 NRC upon discovery of the previously degraded condition of the CFCs was not confirmed. 2.7 Intimidation of Plant Review Board Members An allegation indicated that PR8 members were allegedly intimidated and tressured by the general manager in a PRS meeting. The meeting occurred in February 1990, to det6rmine the acceptability of the safety analysis for the installation of the FAVA microfiltration system. J

~.. I 'd 388Pt'is/ Poet y2ny.32,,,,, 20 Diseussien As discussed in Section 2.1 of this inspection report, sieveral safety cvaluations were performed. for the installation of a temporary modification which installed the FAVA microft1tration system. Discussions with PRB members Indicated that during the review of these safety evaluations, varteus PRB sombers

• had expressed reservations on several occasions concerning the acceptability of the installation of the FAVA system.

j Despite these reservations, the inspection team's review of the PR8 Meeting l cinutes associated with this temporary modification identified few instances of the PR8 members documenting their dissenting opinions. Specifically, PR8 meeting 90-15 (dated February 8,1990) documented one PRS member's negative vote and dissenting opinions regarding the acceptability of exempting the temporary modification from regulatory requirements and the adequacy of the system's safety evaluation. PR8 Meeting 90-28 (dated March 1,1990) indicated that information and issues regarding the FAVA system's safety analysis were presented to the PRS and that the general manager solicited written comments i and euestions from other members for resolution. The only other example was in PRS meeting 90-32 (dated March 6,1990) which identified a dissenting cpinion related to the acceptability of voting on the FAVA system installation t when the PR8 member who raised the initial questions and concerns on the' cperation of the FAVA system was not present. l Discussions with the PRS members indicated that during the various p48 meetings } co cerning the installation of the FAVA system, the PRS members felt intimidated and pressured by the presence of the general manager at the PRS meeting. The sworn testimony confirmed that en one occasion an alternate voting member felt intimidated and feared retribution er reta11stien because the general manager was present at the meeting and the PRS member know the i ge:eral manager wanted to have t,he temporary modification approved. Meuever, the testimony also indicated that the PRS member did not alter his vote and felt comfortable with how he had voted. In addition, the PRS member was not i aware of any occasions on which he or any other PRS member had succumbed to intimidation or feared retribution. l The inspection tese verified that the general manager was informed following this meeting that several PR8 members viewed his presence as intimidating. As a result, on March 1,1990, the general manager met with all PRS members to reiterate the member's duties and responsibilities. He specifically told the members that his presence at PA8 meetings must not influence them and that alternates should be selected who would feel comfortable with this responst-bility. He also addressed the difference between professional differences of 1 cpinten and safety or quality concerns, and their respective methods for ] resektion. I

~ \\ q 2

• d essy. i g,,,,, 3.

,wg 21 Conclusion The inspection team concluded that in one case a PRS voting member felt intimidated and feared retribution because the general manager was present-at the PRS meeting. However, this member stated that he did not change his vote r in response to this pressure and the general manager met with the PR8 to allay fears. Based on the testimony, the inspection team concluded that retribution did not occur. Nevertheless, this confirmed event and the absence of + dissenting opinions in the PR8 aceting minutes indicate that there was a potential for an adverse affect on.open discussions at the meeting. The licensee needs to ensure that PRS members freely and openly empress their ~ technical opinions and safety concerns. 2.8 personnel Accountability As a result of several comments and questions by the Ifeensed operators to the inspection team, the team reviewed the method used to rate the performance of the shift superintendents (55) and unit shift supervisors. l Discussion The operations manager stated that the 55 reported directly to the operations manager and that he personally preparea their performance appraisals. The l inspection identified that the $$ reported to the Unit Superintendent (US), and that the U$ personally prepared the performance appraisals of the $5. The personnel accountability system, first used in 1989, was a pay-for-performance methodology. Annual pay increases and a percentage of the Operations Department bonus were dependent on their ratings in accountability categories. Each accountability category was sedivided into performance categories. Most of the performance categories were based upon group i performance. Once these are eliminated. any differential in pay will result free eight performance categories. Implementation of the plan in 1989 could result in up to an 88,000-a year difference in bonus pay to a 55. The performance categories and their relative weights are: Personnel safety

4. X Regulatory compliance 10.2%

) ESFA5 actuation 12.2% i Reactor trips 10.2% MWO performance 4.15 Special projects 8.2% Personnel development 30.65 f, - Training 20.45 l Therefore, 51 percent will be associated with personnel development and training and 32.6 percent will be associated with the number of LERs. and violations [i.e.. regulatory compliance (30.2 percent). E5FA5 actuation (12.2 percent) and reactor trips (10.2 percent)]. .l

8 'd PE'PI Isero I v2w,11W-2 en um 22 Conclusion The inspection team concluded that there was a potential disincentive for identifying items which may result in t.ERs or violations. In addition, the inspection team concluded that the operations manager provided incorrect or inadequately researched information to the inspection team. The inaccurate information concerned whether the operations manager personally performed the performance appraisals of shift superintendents. The inspection team identified that this fativre to provide accurate information is an example of an apparent violation of the 10 CFR 50.9 requirements to provide accurate information to the NRC and will be identified as an example of Violation 50-424,425/90-19-12: " Failure to Previde Accurate Information as Required by 10 CFR 50.9 to the NRC." 3.0 EXIT INTERVIEWS The inspection scope and findings were summarized on August 17, 1990, with those persons indicated in Appendix 2. The inspection team described the areas 'i inspected and discussed in detail the inspection results. The licensee made numerous dissenting comments. The licensee did not identify as proprietary any of the materials provided to or review 6d by the inspector during this inspection. 1 W s

P 'd FS8Pl IS/to/tI ts1N'8i1W-2 *e3td wog3 e 23 APPENDIX 1 LIST OF TRANSCRIBED INTERVIEWS DATE llM.3 [ERSM 8/14/90 904 hours George Bockhold Jim Swarttwelder 911 hours 1023 hours Harvey Handfinger 1026 hours Bill Ofehl 1109 hours Mike Horton 1335 hours Mike Cnance 1136 hours Jtsmy Paul Cash 1338 hours Dudley Carter 1529 hours Bruce Kaplan 1625 hours Greg Lee 1800 hours Jeff Gasser 8/15/90 906 hours Allen Mosbaugh 937 hours Ernie Thornton 1009 hours John Gwin 1048 hours Steve Waldrup 1335 hours Jerry Bowden 1452 hours John Williams 1637 hours Carolyn Tynan 1730 hours John Williams

P s 7,. 'st N 'aserr'iseroeti d v1w,,3,,,,., ~ .g4 s APPENDIX 2 PERSONS CONTACT'E0 Licensee Employees 'J. Aufdenkampe, Manager Technical Support "G. Bockhold, Jr., General Manager. Nelear Plant

• D. Carter, Shif t Superintendent J. Bowden, Work Planning 1'

J. Cash, Unit Superintendent M. Chance, Senior Engineer. Engineering Support

• S. Chesnut, Technical Support C. Coursey, Maintenance Superintendent W. Diehl, Shift Supervisor Operations "G. Frederick, Safety Audit and Engineering Group Supervisor J. Gasser, Shift Superintencent. Operations "L. Glenn, Manager - Corporate Concerns
• D. Gustafson, Maintenance Engineering Superviser J. Gwin, Corporate System Engineer i
• H. Handfinger, Manager Maintenance J
• K. Holmes,. Manager Training and Emergency Preparedness
• M. Norton, Manager Engineering Support S. Kaplan, Santor Engineer, Engineering Support G. Lee, Plant Engineering Supervisor, Operations
• R. LeGrand, Manager Health Physics and Chemistry
• G. McCarley,ality Concerns CoordinatorIndependent safety Engineering G W. Lyons Qu Supervisor
• C. McCoy, Vice-President, Georgia Power Company
• R. Mcdonald Executive Vice-President, Georgia Power Company
• D. Moncus, Outage and Planning
• A. Moskaugh, VEGP Staff R. Odom, Nuclear Safety and Compliance Mar.ager
• A. Rickman, Senior Engineer - Nuclear Safety and Compliance i
• L. Russell, Independent Safety Engineering Group, 50N0PC0 l
• M. $heibant, Senior Engineer i
• C. Stinespring, Manager Plant Administration

"$. Swanson, Outage and Planning Supervisor 'J. Swart welder, Manager Operations ) E. Thorton, Shift Supervisor, Operations

• E. Toupin. Oglethorpe Power Corporation 4

~ C. Tynan, PRB Secretary

5. Waldrup, Planning and Scheduling Supervisor J. Williams, shift Superintendent, Operations

)

• Attended exit interview, August 16, Igg 0.

} ,d..

~ h z...- - ' I '~ d 8E8'I IS/Po/1i WANwhy.2* m q I s' 25 APPENDIX 2: -t PERSONS CDNTACTED (continued) i NRC Empleyees Who Attended Exit Interview R. Aiello, Resident Inspector - Vogtle j

8. Sonser, Senior Resident Inspector - Vogtle M. Branch, Senior Resident Inspector - Watts Bar K. Brockman, Chief, Reactor Projects section 38 - R!!

R. Carroll, Project Engineer - RI! L. Garner, Senior Resident Inspector - Robinson N. Huneauller, Reactor Engineer - NRR D. Matthews, Project Director - NRR J. Milhoan, Deputy Regional Administrator - R!! L. Reyes. Director Division of Reactor Projects -' RI! R. Starkey. Resident Inspector - Vogtle P. Taylor, Reactor Inspector - RI! M. Thomas, Reactor Inspector - R!! C. VanDenburgh, Section Chief - NRR J. Wticox, Operation Engineer - NRR ii ( 5 '- m a -m u .w

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,gg,.llw-s ien um. ..o' 26 s APPENDIX 3 l LI5T OF ACRONYMS A0P Abnormal Operating Procedure ARS Alternate radwaste building A5ME American Society of Mechanical Engineers CAL Confirmation of action letter CFC Containment Fan Cooler CFR Code of Federal Regulations DC Deficiency card i DRP Division of Reactor Projects EDG Emergency diesel generator EPRI Electric Power Research Institute E5F Engineered safety features E5FAS ' Engineered safety features actuation system FAVA FAVA Control Systems FSAR Final Safety Analysis Report HUT Holdup tank !&C Instrumentation and controis IFI Inspector followup (sen !$T Inservice test l kV Kilovolt LC0 Limiting condition for operation LER Licensee Event Report led 0 Maintenance work order NRC Nucleer Regulatory Commission i NRR Nuclear Reactor Regulation NSCW Nucioar service cooling water NS$5 Nuclear steam supply system 01 Office of Investigations PM Preventative maintenance PRS Plant Review Board plig Pounds per square inch gauge PVC Polyvinyl chloride QA Quality Assurance RI! Region !! Office RCS Reactor coolant system REA Request for engineering assistance r RG Regulatory Guide RHR Residual heat removal - SER Safety Evaluation Report r' 51 Safety injection 50N0PC0 Southern Nuclear Operating Company 55 Shift superintendent E

e'* g, m ti.,o i, i7.'t i 27 L, o I.PPENDIX 3 Lilf 0F ACRONYM 5 (continued) TCP Temporary ch'ange to procedure T5 Technical $pecification US Unit Superintendent Ull, Unit shift superintendent UV Undervoltage i VEGP Vogtle Electric Generating Plant l 9 9 i ~ 3 i i 1 l W* F l

ENCLOSURE l1-1! M '- 9 lood Docket Nos. 50 424, 50-425 License.Nos. NPF-68, NPF-81 ) i -Georgia Power Company [ ATTN: Mr. C. K. McCoy ~ Vice President-Yogtle Electric-Generating Plant P. O. Bcx 1295 l Sirmingham, AL 35201 ~ t Gentlemen: l

SUBJECT:

NOTICE OF VIOLATION (NRC INSPECTION REPORT N05. 50-424/94-12 AND 50-425/94-12) This refers to the inspection conducted by R. Noore of this office on l May 9-20, 1994. The inspection included a review of activities authorized for i your Vogtle facility. At the conclusion of the inspection, the findings were discussed with those members of your staff identified in the enclosed report. Areas examined during the inspection are identified.in the report. Within these areas, the inspection consisted of selective examinations of procedures q r and representative records, interviews with personnel, and observation of i activities in progress. l Based on the results of this inspection, certain of your activities appeared a to be in violation of NRC requirements, as specified in the enclosed' Notice of i Violation (Notice). The violation.is of concern because in one case an l equipment protection function on safety related equipment was disabled. In the other, required interim measures were not accomplished.for Emergency -i Diesel Generator air system parameters identified outside their acceptance criteria. ~ I h You are required to respond to this letter and should follow the instructions specified in the enclosed Notice when preparing your response. In your response, you should document the specific actions taken and any additional actions you plan to prevent recurrence. After reviewing your response to this Notice, including your proposed corrective actions and the results of future inspections, the NRC will determine whether further NRC enforcement action is necessary to ensure compliance with NRC regulatory requirements. In accordance with 10 CFR 2.790 of the NRC's " Rule of Practice," a copy of this letter, its enclosure and any reply will be placed in the NRC Public-Document Room. If you wish to withhold information contained therein, please notify this office by telephone within ten. days of the date of this letter and promptly thereafter submit a written application to withhold information contained therein. Such application must be consistent with the requirement ) ,* n - L.L \\\\ \\ AM 3%.

~. .~. - M,,. y.g y _ Gxrgia Power _ Company - >2 - of 10 CFR 2.790(b)(1). - If we do not here: from you in this_ regard within the

period specified above,. this letter, its-enclosure and any reply will' be placed in the NRC Public Document Room.

' Should you' have -any questions concerning.this letter,-please contact. us. _ Sincerely, ORIGINAL $1GNED BY CHARLES A. CASTO Charles A.- Casto, Acting Chief Engineering Branch Division of Reactor Safety

Enclosures:

1. Notice of Violation 2. NRC Inspection Report cc w/encls: J. D. Woodard Senior Vice President-Nuclear Georgia Power Company y P. 0..' Box 1295 ( Birmingham, AL 35201 e J. B. Beasley Gen 6ral Manager, Plant Vogtle Georgia Power Company P. O. Box 1600 l Waynesboro, GA-30830 J. A. Bailey Manager-Licensing Georgia Power Company P. O. Box 1295 Birmingham, AL 35201 i Nancy G. Cowles, Counsel Office of the Consumer's Utility Council 84 Peachtree Street, NW, Suite 201 AtIanta, GA 30303-2318 i Office of Planning and Budget Room 6158 270 Washington Street, SW Atlanta, GA 30334 (ce w/encls cont'd - See page 3)

4 Georgia Power Company 3 M - 91994 (cc w/encls cont'd) Office of the County Commissioner Burke County Commission Waynesboro, GA 30830 Harold Reheis Director Department of Natural Resources 205 Butler Street, SE, Suite 1252 Atlanta, GA 30334 Thomas Hill, Manager Radioactive Materials Program Department of Natural Resources 4244 International Parkway Suite 114 Atlanta, GA 30354 Attorney General Law Department 132 Judicial Building Atlanta, GA 30334 [' Ernie Toupin Manager of Nuclear Operations Oglethorpe Power Corporation 2100 E. Exchange Place Tucker, GA 30085-1349 Charles A. Patrizia, Esq. Paul, Hastings, Janofsky & Walker 12th Floor 1050 Connecticut Avenue, NW Washington, D. C. 20036 (bec w/encls - See page 4)

+ 2 ,m 1 'GeorgiaLPower Company-4 y ~ . bcc w/encls: D. Seymour, RI! l G.: Hallstros. RII O. Hood. NRR- . P. Skinner, RII - . M. V. Sinkule, RI! ' Document Control Oesk-NRC Senior Resident inspector U.S. Nuclear Regulatory Commission P. O. Box 572 Waynesboro, GA 30830 5 e ? .e. A t I,. T

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+ P' ENCLOSURE I 4 NOTICE OF VIOLATION Georgia Power Company Docket Nos. 50-424 and 50 425 Vogtle Nuclear Plant License Nos. NPF-68 and NPF-81 During an NRC inspection conducted on May 9-20, 1994, a violation of NRC requirements was identified. In accordance with the " General Statement of. Policy and Procedure for NRC' Enforcement Actions," 10 CFR 2,-Appendix C, the violation is listed below: 10 CFR 50 Appendix 8, Criterion V, Instructions, Procedures,' and Drawings, as implemented by the Vogtle Electric Generating Plant Operations Quality Assurance Policy Manual, revision 12, requires that activities affecting quality shall be prescribed by documented procedures and activities shall be accomplished in accordance with these procedures. Contrary to the above, on May 9-20, 1994, two examples were identified in which activities affecting quality were not accomplished in accordance with prescribed procedures. Example 1: Procedure SCL 00166, Diesel Generator Air Start Dryer Maintenance, revision 5, step 4.E, required that moisture checks be accomplished every 12 hours if dew point analysis indicated air system dew point was not within the acceptance criteria of 32*F to 508 'F. On-January 19, 1994, dew point analysis indicated that the dew point exceeded the acceptance criteria for six of eight air receivers. These results were documented on maintenance work orders 19303293, 29303950, and 19303290. No moisture checks were performed and the. actual air quality was not verified until February 5, 1994. This analysis verified that EDG 1A receiver K02 exceeded the acceptance criteria. 4 Example 2: Procedure 27563 C, Generator and Engine Control Panel Functional Test, revision 8, step 4.2.57, required that tubing E-14'to the jacket water pressure switch (1 PSL 19114) be re-connected following completion of the test. During the April 1, 1993, performance of this procedure on Emergency Diesel Generator (EDG) 1A, the E-14 tubing was not re-connected. This resulted in the jacket water low ) pressure trip being disabled for approximately one year. This is a Severity Level IV violation (Supplement I). i 0

1 [. g-l Pursuant-to the provisions of 10 CFR 2.201, Georgia Power Company is hereby required to submit a written' statement or explanation to the U.S. Nuclear - Regulatory Commission, ATTN: Document Control Desk, Washington, D.C. 20555 with a copy to the Regional Administrator, Region II, and a copy to the NRC Resident Inspector at the Vogtle facility, within 30 days of the date of the letter transmitting this Notice of Violation (Notice). This reply should be clearly marked as a " Reply to a Notice of Violation" and should include for each violation: (1) the reason for the violation, or, if contested, the basis for disputing the violation, (2) the corrective steps that have been taken and the results achieved, (3) the corrective steps that will be taken to avoid further violations, and (4) the date when full compliance will be achieved. If an adequate reply is not received within the time specified in this Notice, an order or Demand for Information may be issued as to why the license should. q not be modified, suspended, or revoked, or why such other action as may be proper should not-be taken. Where good cause is shown, consideration will be i given to extending the response time. Dated at Atlanta, Georgia this 9th day of June 1994 I S O O e e a e E

noe UNfft0 STATES NUCLEAR REOULATORY COMMIS$40N ,f neoen s ist MAAleTTA STREEr.GW. SUM 200 3 ArtAurA. osonoiA sossee Report No.: 50-424/94-12 and 50-425/94-12 Licensee: Georgia Power Company P.O. Box 1295 Birmingham, AL 35201 Docket Nos.: 50-424 and 50-425 License Nos.: NPF-68 and NPF-81 i Facility Name: Vogtle 1 and 2 Inspection Conducted: Nay 9-20, 1994 Inspectors. Aser f-3/- # R. Moore, Region II Date Signed l' Yh J~ 3! - 94! A. MacDonald, Region II Date Signed I hl))CGtL 1/ f V K. Waterman, NRR p+ 4 eta.1 Dath Signe'd Approved by: O7v?-( Ic C!4!9 M. Shymlock, Chier (' Date 51gnkd Plant Systems Section Division of Reactor Safety

SUMMARY

Scope: This routine electrical maintenance inspection focussed on the effectiveness of the licensee's corrective actions for emergency diesel generator (EDG) 1A and 18 failures which occurred in 1990. The failures were related to malfunctions of the EDG pneumatic control and protection system. Results: The licensee's corrective actions effectively resolved problems with the pneumatic control system which contributed to EDG failures in 1990. One violation was identified during this inspection. The violation identified two examples of the licensee's failure to follow maintenance procedures (paragraph 2.5.1 and 2.7).

i REPORT DETAILS .i 1. Persons Contacted: Licensee Employees

• 8. Beasley, General Manager
• W. Burweister, Engineering Support Manager
• R. Burns, Engineering Support

]

• S. Chesnut Technica support Manager i
• W. Copeland, Materials Supervisor 1
• C. Coursey, Maintenance Superintendent j
• R. Dorman, P1 ant-Training Manager
• C. Eckert, Senior _ Technical Specialist j
• W. Gabbard, Nuclear Specialist
• J. Gasser, Operation Unit Superintendent -

i

• W. Kitchens, Assistant General Manager - Support j
• R. Moye, Plant Engineering-Supervisor i
• M. Shelbani, Nuclear Safety Supervisor i
• K. Stokes, Senior Engineer
• J. Swartzwelder, Outage and Planning Manager
• K. Burr, Senior Project Engineer l

Other licensee employees contacted included technicians, supervisors, engineers maintenance personnel and office personnel. i NRC Resident Inspectors -l

• B. Bonser, Senior Resident Inspector

(

• D. Starkey, Resident inspector
• M. Shymlock. RII, Plaat Systems Section Chief j
• Attended exit meeting l

Abbreviations and Acronyms are listed in paragraph 5.0 l 2.0 Electrical Maintenance (62705) I i

2.1 Background

t In March 1990, EDG 1A experienced failures attributable to malfunctions ) of the pneumatic protection and control systes. Investigations in 1990 l concluded that the primary root cause was improper intermittent operation of the Calcon jacket water temperature sensors. Additional contributing causes were identified as pneumatic control system leaks and inconsistent Calcon instrument. calibration techniques. Similar i trips were experienced on EDG 1B in May of 1990, i In this inspection, the inspectors assessed the effectiveness of the I licensee's corrective actions for pneumatic protection and control { system problems identified in the 1990 EDG failure evaluations. Additionally, the inspectors reviewed other factors which could potentially impact the pneumatic control systes function such as air i quality, system configuration, modifications, and equipment history of f [ i f i i

l y 2. l i critical system components such as the P-3 pressure switch and the. 6 pneumatic logic boards. Equipment history was assessed by review of approximately 780 maintenance work orders (MW0s) and 165 deficiency cards (DCs) from 1988 to the present. j 2.2. Calcon Instrument Malfunctions i The' inspectors assessed the effectiveness of the licensee's' actions to 'I address Calcon instrument malfunctions which were identified as 'a i primary contributor to the March 20,1990, EDG 1A failures. The corrective action addressed improvement of procedures-and methodology for calibration of the instrumentation. A review of instrument equipment history indicated the effectiveness of upgrading calibration procedures and methodology. j 2.2.1 Calibration Procedures Calibration procedures prior to April 1990, were generic procedures I supplemented.with written instructions from a Request for Engineering 1 Review (RER). The following procedures provided guidance for instrument i calibrations in that time period-i e Procedure 22721-C, Pressure Switch Calibration, revision 3 e Procedure 22332-C, Temperature Switch Calibration, revision 2 l I e Procedure 23820-C, Generic Instrumentation Calibration, revision 3 j e RER 88 0707, Instrument Tolerances for EDG Calcon Switches, dated l November 10,1988 ] After the March 20,1990, EDG 1A failures, the licensee evaluated calibration procedures and methodology and developed new procedures in l conjunction with Wyle Labs. The new procedures provided more specific instructions and improved calibration methodology. The inspectors-i reviewed the following procedures which were being used for instrument ~ calibrations at the date of this' inspection:

• Procedure 22981-C, Calcon Pneumatic Temperature Sensor Calibration, Equipment Nos.1(2)TSH-19112,1(2)TSH-19119, y

1(2)TSH 19146,1(2)TSH-19153, revision 6 l'

• Procedure 22983-C, Calcon Pressure Switch Calibration, Equipment.

1 Nos.1(2)PSL-4749 A, 8 C, D, E,1(2)PSL-4859 A, 8, C, D, E, i 1(2)PSL-19114 and 19121, revision 2

• Procedure 22982-C, Calcon Pneumatic Vibration Sensor Model E-4600 i

Functional Test, revision 2 i The current procedures provided specific instructions for calibrating the Calcon sensors and provided a more systematic and well-defined calibration process. For example, these calibration procedures provided ( _ ~,.

1 M 1 3J m ~' detailed requirements for test equipment'and cleaning materials incorporating' vendor recommendations. - Additional instructions were 1 included to address sensor venting problems and expected sensor-j performance. The. instructions more clearly defined the procedure for. adjusting the sensor to achieve the correct-instrument' response. A. sensor preheat period not specified in the previous procedures was addressed in the new procedures. The new procedures also address

verification of. sensor tube connection tightness. Analysis of calibration processes' in 1990 indicated that loose sensor tubes impacted sensor setpoints.

The current calibration procedures required that a calibration be. performed three times to verify that the trip and reseti ~ values were'within the specified limits. The inspectors concluded that the changes in calibration procedures provided for more consistent - reliable calibration of Calcon instruments. i The new instructions also addressed isolation of problem sensors for-analysis and_. installation practices. Sensors with excessive drift or calibration problems were to be tagged and stored for engineering analysis and Engineering was required to be-notified-.if any prob es was i encountered during calibration of a sensor, including any sensor found out of calibration. Specific guidance was incorporated in the procedures to minimize the presence of foreign material in the sensor body.- For example, the use of locktite was specified as a thread sealant' as opposed to " pipe dope" which had been found on sensor - internals and contributed to improper sensor venting. Additionally, c specific instructions were provided for the application of thread f-sealant after the sensor fitting was screwed into the sensor. body approximately two full turns. The inspectors concluded that these . additional instructions contributed to reduction _in sensor failures. 1 2.2.2 Instrument Failure Experience The inspectors reviewed MW0s to assess the Calcon instrument equipment ) history at Vogtle to determine if the instruments'- reliability had improved as-a result of the corrective ' actions discussed above. NUREG 1410 listed 67 Calcon instrument failures at Vogtle between'1985 and 1990. This included 48 temperature sensors, 13 pressure sensors, 3 vibration sensors and 3 air trip valve (p3) failures. The inspectors reviewed the following MW0s which identified Calcon Instrument failures since April,1990: TEMPERATURE SENSOR MW0s 19002711 (3 Failures) 19203584 2 Failures 19103008 19203585 3 Failures 19103009 29003403 19104772 29200295 19104829 -i ,I l

s i a L 9 .4' s su PRESSURE SENSOR NW0s VIBRATION SENSOR MW0s 19203577 19104783 '29201061-29102850 29102840 (3 Failures) ~19101227

The MW0s listed'above indicate that 22 sensor failures occurred since.

April 1990., These included,14 temperature, 6 pressure and 2 vibration - sensors that either failed to properly function or calibrate. Several --of'the temperature malfunctions occurred on EDG 18 on'May 23,.1990,. following initial' use of the new calibration procedures..No instrument' malfunctions have occurred since April 1993. The inspectors concluded that the reliability of the sensors had impreved'since the Vogtle Loss. of Vital AC Power event on March 20,1990,(67 sensor problems prior to. April 1990, versus 22 sensor problems after April 1990, with-no. malfunctions since April 1993). 2.3 Critical Components The inspectors reviewed the equipment history for selected critical components of the EDG pneumatic control system to determine if past performance of these components impacted EDG reliability. The components reviewed were the P3 shutdown pressure switches and the pneumatic logic boards. 2.3.1 P3 Shutdown Pressure Switches. The purpose of the P3 shutdown Calcon pressure switch was to trip the EDG when a trip parameter, such as high crankcase pressure, reached its setpoint. This switch ensured a shutdown following establishment of a-trip condition. Setpoint errors-could result in inappropriate initiation of trips from non-emergency trip parameters. The inspectors reviewed the MW0s and DCs initiated from 1990 to the present. There have been three incidents in which' the P3 pressure switch was thought to have failed (MW0s 19001537 and 19001542, and DCP 90-VIN 0164). The failure addressed by MWO 19001542 was reported on March 25, 1990, and required replacement of the P3 switch. The pressure switch failure addressed by MWO 19001537 occurred when the P3 switch failed to reset after tripping. The switch was replaced. MWD 19001511 dated March 28, 1990, tested the EDG 1A'P3 switches at various air. pressures and with different orifices stres on the test stand. The test conclusion was that repeatability throughout the variations was consistent. The following MW0s during 1990 included-P3 switch replacements and calibrations: 19000068,19002711, and 19000016. These MW0s did not identify problems with setpoint repeatability. DCR 90-VIN 0164 was initiated to lower the set point on P3 pressure switches. This DCR was cancelled and the set points were not changed. The basis for cancellation stated that the P3 set / reset set point values were not the cause of EDG 1A failures being investigated. P3 switch 4

,4 i k

.L 5-

' operation was impacted by normally charged lines being blod down during m maintenance. These lines'had not.been sufficiently' recharged prior to attempted.EDG starts. The inspectors concluded that P3. switch malfunctions have not impacted EDG reliability. The Maintenance history indicated few failures and the calibration documents did not identify occurrences of setpoint ' repeatability problems. 'i 2.3-2 Pneumatic System. Logic The inspectors reviewed MW0s and DCs from 1990 to the present to evaluate the failure history of EDG pneumatic logic boards. The following MW0s were identified which documented pneumatic logic board j replacements and repair of components on pneumatic logic boards. 29004795 -Pneumatic logic board replacement 19001219 Pneumatic logic board replacement 19001409 Pneumatic logic board component (OR gate) cleaned 19001537. Pneumatic logic board replacement' l 29303314 Pneumatic logic board component replacement The inspectors reviewed these MW0s in detail and confirmed that pneumatic logic boards were replaced during pneumatic control system troubleshooting. The original logic boards that were replaced were later inspected and tested by the vendor and determined to be - i acceptable. The logic boards had not failed, but were replaced during troubleshooting.as a potential failure cause. ~ The inspectors reviewed vendor letter, dated June 5,1990, which documented the pneumatic control component testing of the IB shutdown logic boards and verified that the shutdown logic board did not fail. l The pneumatic control system for each EDG is functionally tested during each refueling outage using 27563-C, Generator And Engine Control Panel Functional Test Procedure. Revision 1 of this procedure, approved February 20, 1990, was the version used prior to the March 1990 EDG. failures. The inspectors reviewed 27563-C, Revision 1 and verified that the procedure performed a functional check of pneumatic control system start functions and engine protective trip functions. The functional test of the pneumatic logic boards at refueling outage intervals met the requirements included in the'Transamerica Delaval Incorporated Diesel Generator Owners Group Maintenance Matrix, Revision i 3. The pneumatic control system functional tests were documented on MW0s. The inspectors reviewed completed functional tests documented;in the following MW0s to determine if the testing identified poor performance of the pneumatic control logic. t i A l a

3 - 6 i MW0 Nos. y. l 29002105 29002102 19000095 19000094 19203296 19203299-Review'of the six functional test MW0s identified one pneumatic logic or-gate which required clear,ing. This repair MWO was 19001409 i previously reviewed as part of pneumatic control system failure history, i The only other problem discovered was in MWO 25002102 which identified a Timer /Not element on a logic board which required adjustment. Based on the_ review of failure history and functional test results, the inspectors concluded that pneumatic logic board reliability has been. acceptable. The licensee testing met vendor owners group requirements and verified pneumatic control system. performance. 2.4 Pneumatic System Leakage Pneumatic control system leakage was identified as a contributing factor { to the EDG 1A aM EDG 18 failures which occurred in 1990. The inspectors screened approximately 780 MW0s and 165 DCs from 1990 to the j present to identify the MW0s associated with EDG pneumatic control system leakage. The following MW0s were reviewed: 29004795 29004733 29000182 19001185 i 19001404 19001433 19001435 19001537 19001576 19002289 19002711 19003164 19003510 19104783 19104997 19105032-19105050 29201061 19301705 29303314 -i 19000016 19104772 19001629 19001511 19001683 191-303 190-154 191-293 The MWO review indicated that 59 percent of the leakage MW0s occurred in 1990 with 31 percent in 1991, 3 percent in 1992 and 7 percent in 1993. Fifty-five percent of the leakage was attributed to venting pneumatic-l trip switches. Twenty-four percent of the leakage was caused by i component leaks and 21 percent caused by fitting leaks. i During 1990, the licensee performed functional testing including soap bubble leak checks of all fittings which were disassembled during i testing. The functional testing was performed on a refueling outage interval in accordance with procedure 27563-C, Generator And Engine l Control Panel Functional Test Procedure, Revision 1. No specific leakage acceptance criteria was utilized. Many of the pneumatic control system connections utilized Swagelock compression fittings. No j procedure was used for these fitting connections during 1990.- Trainir.g on proper Swagelock compression fitting installation was provided to plant personnel as part of job position training. l

i I -Revision eight of Procedere 27563 C was. reviewed by the inspectors to assess present leakage control practices. This revision included- ') detailed leakage measurement of the pneumatic system. A modification ) was implemented to add test valves to facilitate the leakage testing. The inspectors verified that test valves had been added to pneumatic c trip sensor lines. Procedure 27563-C, revision eight was reviewed and J the inspectors noted the detailed leakage checks and the specific leakage acceptance criteria. Procedure 20440-C, Swagelock Fittings Replacement / Instruction, revision 1, was. approved September 25, 1992, to control replacement and installation of Swagelock compression fittings. The reduction in the.. number of EDG control system leakap related MW0s indic:ted that.the detailed leakage -testing and Swage ock compression fitting procedure had improved the pneumatic control system pressure integrity. The inspectors witnessed testing of EDG 2A on Nay 13, 1994, and EDG 18 on May 18, 1994 During both tests, the inspectors checked the pneumatic trip switches and none were found to be venting. Pneumatic tubing fittings inside the engine control panels and bulkhead fittings at the engines were checked for leaks. No leakage was detected at the engine bulkhead fittings or inside the engine control panels. { The inspectors conc *1uded the pneumatic control system leakage occurred in 1990 and contributed to the EDG 1A and 18 failures in 1990. This leakage was discussed in NUREG 1410. Licensee actions have significantly improved pneumatic control system pressure integrity. An adequate program was established to routinely monitor and control system leakage. Present EDG reliability was not impacted by pneumatic system leakage. 2.5 System Air Quality j 1 The inspectors reviewed the licensae's activities to maintain the air quality of the pneumatic control and protection syst.es. The potential impact of air system moisture on EDG reliability was also reviewed. The EDG Vendor provided no specific criteria for moisture content. ) Acceptance criteria for air moisture content was. provided by the j licensee's response, dated February 17, 1989, to NRC Generic Letter i 88-14 Instrument Air Supply System Problems Affecting Safety Related Equipment. The acceptance criteria was for a 50 'F dew point at 250 pounds (psig) air pressure. 2.5.1 Maintenance of Air Quality Refrigerant compressor air dryers were used to remove moisture from compressed air and dew points were periodically monitored to verify the dew point criteria was maintained. EDG air start receiver dew points were measured every 28 days. The inspectors reviewed maintenance procedure SCL-00166. EDG Air Start Dryer Maintenance, revision 6. When a dew point was not within the acceptance criteria (32'F to 50*F), the procedure directed that the system engineer and operations be notified. I

~ j 8 The receiver.was-not isolated unless periodic moisture checks indicated water in the control air.. system.- Moisture checks were to be conducted . every 12 hours at-a control air. test, connection in the EDG control ! cabinet. - The procedure specified opening the test' connection valve "for a few seconds" to check for moisture. The inspectors noted that.this-i blow down' time may not be sufficient to determine if water was in the I 250 psig piping outside the cabinet. The licensee. initiated actions to i revise the-procedure to extend the blowdown time. With the exception i noted, the inspectors. concluded that the procedure provided adequate guidance for monitoring. system dew point. j

The inspectors reviewed the following MW0s which documented dew points outside the acceptance criteria between 1988 and 1994:

i EDG 1A EDG 1B EDG 2A EDG 28 I 18806224. 18905009 29104594 29200789 18809080 18808711 29200210-29303950 18900984 18906446 29200783 i 19000899 19001770 29200951 l 19001513 19002901 29201404 t 19001651 19003585 19102066 19102064 i 19202414 19102968 19303293 19103401 l / 19303295 19103676 t i 19400830 19104653 19300472 i 19303290 l Dew points outside the' acceptance criteria indicated that the air dryers were not functioning correctly and the interim actions previously discussed were required to assure moisture was not introduced into the t air system. Corrective actions were to repair the dryer and perfor1s a feed and bleed on the receiver to reduce the dew point. The inspectors l noted that the occurrences of dew points outside.the acceptance criteria decreased after 1990, indicating improved performance in maintaining air j dryer equipment. In reviewing dew point analysis results, the inspectors noted that on i January 19, 1994, six of eight air receiver dew point analyses indicated j dew points outside the acceptance criteria. These results wer. documented on MW0s 19303293, 29303950, and 19303290. The interim i actions required by the maintenance procedure, SCL-00166, performance of moisture checks every 12 hours, were not accomplished.. Additionally, no i further dew point analysis was accomplished until February 5,1994.- The February 5,1994, analysis indicated that all _ receivers'except receiver i K02 on EDG 1A were within the acceptance criteria. In. addressing this j issue with the inspectors, the licensee stated that the dew point L measuring and test equipment validity was suspect because the results were inconsistent with previous analysis and the analysis method used differed from previous methodology. The inspectors concluded that i

9 although ~there was a basis to question the dew' point analys.is results,- 'the interim actions of 12 hour moisture checks were required until-the ' dew point conditions were verified within the acceptance criteria. This

issue was identified as one example of NRC Violation 94-12-01,~ Failure to Follow Maintenance Procedures. An additiona.1 example is discussed in paragraph 2.7 of this report. -

The inspectors reviewed EDG maintenance history to determine if.out-of-tolerance dew point conditions resulted in detectable water formation er. adverse operation of the pneumatic control and protection system. The

i troubleshooting MW0s related to the March 20,1990, EDG 1A and May 23, i

1990, EDG 18 failures were specifically reviewed. The maintenance.' documentation provided no indication that water had been detected in the cor. trol and protection portion of the air start system at any time. Discussions with the craft and engineering staff involved in the 1990 trouble shooting activities and current EDG maintenance also provided no indication that water had been detected in the air system. During the inspection, the inspectors observed dew point measurement on four air-receivers. The analysis on a receiver on EDG 2B indicated a dew' point ) which exceeded the acceptance criteria. The inspectors observed blowdowns on the 2B receivers and control air system. No detectable l moisture was observed. The inspectors concluded that the out-of-tolerance condition did not result in detectable water formation in the i control air system. j An additional factor which indicated that water formation in the control l air system was unlikely was that dew point values decrease when the j system pressure is reduced. The dew point of the 250-psig supply air 1 will significantly decrease when the pressure is reduced to 60 psig by the control cabinet pressure regulator.. This was confirmed by review of 1 a psychrometric chart that plots dew point' temperatures as a function of l pressure. Using the chart, and _ assuming a worst casa dew point of 85'F-i (29'C) at 250 psig, the equivalent dew point at 60 psig is approximately l 50*F (10*C). - The control cabinets are heated with resistance-type heaters and shield the control components from outside air drafts. All i system orifices are located in the control cabinet. The minimum design temperature for the control cabinets is the same as. for the EDS, 50*F 1 (10*C). Consequently, even with the highest dew point conditions that have been measured to date, the dew point of the air in the control: 1 cabinets was only equal to the control. cabinet ambient temperature. _ The inspectors concluded that probability of condensation within the 60 psig air supply in the control cabinets was not significant. 2.5.2 Potential Moisture Impact j The inspectors conducted a detailed review of the pneumatic control and i protection system operation to assess the potential impact on EDG reliability from water in the system. The control logic component l design is such that the presence of moisture in the air supply will not 4 cause EDG trips during the startup phase of operation. The critical components for'this condition would be the AND module (AND-14) and a 1 i Timer /NOT module (Timer /NOT-ll), which were in the logic board. These l

10 elements were included on Engine Control Panel Schematic 09 500-76021, sheet 1 Of. 9, revision 9. If there were enough water to cause.the Timer /NOT element to sense a false pressure signal, there would.be a similar response at the AND module, which'would result in pressurization . of the' B port of the P3 OR module.. This would result in either the EDG tripping before 60 seconds -or the EDG not tripping at all. The timing of the EDG' trips reviewed did not indicate this occurrence. The. inspectors concluded that water had not been a contributor to these EDG trips. There was a 5-micron air filter in the 250-psig air lines immediately before the 60-psig pressure regulator in the control cabinet. The purpose of the filter was to remove particulate from the air before it was admitted into the. pneumatic control modules. If water were present in-the 250 psig air supply line, the 5-micron filter in the control. cabinet would atomize the water droplets into a fine mist. Assuming the water droplets were approximately 5 microns in diameter, the smallest j orifice in the control system is 0.006 inch (152 microns).. approximately l 30 times larger than the atomized water droplets. Consequently, even if all of the air flowed through the 0.006-in orifice, the probability of choked flow is insignificant. Additionally, the majority of the control-air bypasses the 0.006-inch orifice and pressurizes the A port in the P3 i upstream OR gate. Consequently, the effect of moisture on the pressurization of the P3 switch OR gate ports was insignificant.. ] / Water inside the control modules could cause corrosion of the metal I parts inside the logic modules and inside the EDG instrumentation.. This could affect the sensitivity of the instruments, and thereby affect the startup of the EDG..However, review of. MW0s and DCs for the two units did not reveal any cases of corrosion caused by unknown sources of water. One MWO, 19104783, did state that the vendor introduced water into a sensor during a pneumatic leak test with a bubbler. The-inspectors conclude that the presence of water in the control system air lines can not be confirmed by evidence of corrosion. l 2.6 Modifications The inspectors reviewed modifications to the pneumatic control and protection portion of the air start system to verify the completion of corrective actions and assess the impact on EDG reliability. Corrective i actions for the March 20,1990, EDG 1A failure included establishing the loss of Offsite Power (LOSP) start as an emergency start and deleting the jacket water high temperature trip as an' emergency mode trip. 1 Additional modifications included changes.to various orifice components. i The following modifications were related to corrective actions for the [ EDG 1A failures. Design Change Packages (DCPs) 90-V2N0137 and 90-i L VIN 0133 were completed in August,1990 and established the LOSP EDG ~ start as an emergency mode start, i.e. EDG non-emergency trips disabled during LOSP start. The jacket water high temperature. trips were disabled by installation of isolation valves in the sensor instrument lines on DCPs 90-VIN 0138 and 90-V2N0166 in November,1990. The l 'i i i

- =. ~ 11~ modifications to the pneumatic logic to delete the jacket water high temperature' trips,' DCPs 91-VIN 0ll3 and 91-VIN 0ll4, were completed in 1991 for Unit 1 and 1992 for Unit 2. l The following modifications were related to changes' in orifice components in the pneumatic control system. Minor Design Deviations (MDDs) 89-VlM194 installed 0.014.. inch orifices' in the lube oil pressure sensing lines where no orifice was ~previously installed. This was.to. assure establishing low lube oil. protection for an emergency start. 1 following a normal shutdown and was. completed in March 1990. In October 1990, MD0s 90 V2M193 and 90-VIM 194 decreased the orifice size in the shutdown logic board from 0.028 to 0.020 inches. DCPs 91-VIN 0ll3 and 91-V2N0114 discussed above also installed 0.006 inch orifices in the Jacket water temperature sensor air supply lines similar. to other non-l emergency. trip sensors. The inspectors' configuration walkdowns discussed in paragraph 2.7 of this report' verified installed orifices were consistent with as-built drawings for the sample reviewed. The inspectors concluded that changes to the pneumatic control system i appropriately implemented the design control process and contributed to increased EDG reliability. y 2.7 EDG Pneumatic Control System Configuration - I The inspectors reviewed MW0s, Deficiency Cards, and performed system walkdowns to determine if the EDG pneumatic control system configuration .j p was maintained in accordance with system design drawings-The MWO review identified a tagging concern related to the high temperature jacket water pneumatic trip switches and their respective test _ valves. MW0s 29004795, 19004621, 19004622, 29005610 and deficiency card 290-225 documented and resolved the tagging concern for all four ' l' EDGs. i i MWO 19001219 documented problems with the pneumatic control system of I EDG 1A noted during functional testing on March 9,1990. The tubing to vibration try switches was left disconnected which prevented the system l from pressurizing properly. Once the tubing was connected, the EDG l operated satisfactorily. The MWO review did not identify any instances i of pneumatic tubing being connected to the incorrect sensor or component. l The inspe: tors performed a walkdown of portions of the EDG 1A and IB pneumatir, control systems. Plant configuration was checked against .( system design drawing, Engine Control Panel Schematic, drawing FW-700-i 7602, shest 1 of 13, revision A, and Engine' Pneumatic Schematic, drawing FW-700-7602, sheet 10 of 13, revision A. h On May 10, - 1994, during system walkdown, the inspectors identified i tubing connection E-14 capped and disconnected from the EDG 1A low I pressure jacket water Calcon trip sensor, IPSL19114. The trip sensor line was disconnected and capped during the performance of system l functional testing which was performed each refueling in accordance with i i l A

13 ' Vibration Sensor Vibration Sensor i Left Bank Turbocharger left Bank' Turbocharger Left Bank Engine Left Bank Engine Right Bank Engine Right Bank Engine 2.8. EDG Reliability The inspectors reviewed the EDG demand and failure hist'ory to determine whether corrective' actions for the 1990 EDG 1A instrument failures impacted EDG reliability. Additionally,.the failures were reviewed to i determine if the' licensee's categorization of valid and invalid failures was consistent with Regulatory Guide 1.108. Periodic Testing of Diesel 'i Generator Units Used as Onsite Electric Power Systems at Nuc ear Power 1 Plants,. revision 1. i In 1990 there were a total of 32 EDG failures on the four Vogtle EDGs, nine were valid failures. Unit I reliability was 0.95 with an unavailability of 11.06 percent on EDG 1A and 2.67 percent on EDG 18. Unit 2 reliability was 0.96 with an unavailability of 3.63 percent on 1 EDG 2A and 2.51 percent on EDG 28. For 1993, There was a total.of 7 EDG failures, none were valid failures. Unit 1 reliability was 0.98 with an unavailability of zero percent for both EDG 1A and 18. Unit 2 i reliability was 0.99. with unavailability of 0.26 percent for EDG 2A and 0.38 percent for EDG 28. These statistical values indicate an r improvement in ~EDG reliability and availability since 1990. An' i additional indicator of the effectiveness of the corrective actions was that no EDG. failures were attributable to' pneumatic control system i malfunctions after 1990. Review of EDG failures since 1990 indicated that the failures had been categorized in accordance with AG 1.108. The i inspectors concluded that EDG performance history demonstrated that j corrective actions from the March 1990, EDG 1A failures were effective J in resolving pneumatic control system problems and improving EDG reliability. i 3.0 Follow-up of Previous Enforcement Items (92702) 3.1 Violation 50-424,425/92-3001, Failure To Identify Conditions Adverse To Quality For EDG 1A Failure Of November 18, 1992 j This itw addressed the licensee's failure to identify and investigate a valid Eb6 failure caused by an air start system component deficiency. l The inspectors reviewed the corrective actions for this violation. The a corrective actions included training, Operations policy changes, and procedure revisions. Policy changes documented on Licensee Interoffice Correspondence dated i December 30, 1992, specified that an extra plant equipment operator J should be present at the EDG for testing. This correspondence also i established the policy that the EDG testing be performed early on day shift. Licensee Interoffice Correspondence dated November 30, 1992, established policy that the Operations Manager be notified when i 1 j 1

., y '14 .questionstabout operability or reliability of safety'related equipment - arise. Policy. changes.also included assignment of responsibility for EDG failure. classification to engineering. The_ operating shift is? z required to notify. licensee management and engineering for an operabillty evaluatton when an-EDG~does not start.- Training included' adding the EDG 1A failure issue to operatcr & qualification training and an event review for shift supervisoi.. -The inspectors. reviewed Training Lesson Plan RQ-LP-63123-01, Revision 2 Licensed Operator. Re-qualification,' Current Events. The lesson plan included;a description of the EDG 1A Failure Event of November 18, 1992, and.a description' of the air start system and the EDG control circuit. start push-button and starting relays. The licensee's EDG operability test l procedure was revised. The inspectors reviewed procedure 14980-1/2, Diesel Generator OperabilityL Test, Revision 31/17. The inspectors witnessed EDG tests of EDG 2A on May 13, 1994, and EDG 18 on May.18,1994. The tests were conducted in accordance with the revised cperations policy.. Testing was performed early on day shift and-included an additional plant equipment operator stationed at the EDG. The licensee's corrective action for Violation 50-424,425/92-30 01 was adequate. This item is closed.- 3.2. Violation 50-424,425/92-30-02,-Inadequate Procedural Acceptance Criteria For EDG Air Start Valve Maintenance e This item addressed the use of incorrect acceptance criteria for safet'y. related maintenance troubleshooting. Two maintenance procedures were used on one MW0. each providing different acceptance criteria for the air start valve cap to piston clearance. The inspectors reviewed the corrective action for this violation. The corrective action consisted of procedure revision and a review of maintenance procedures..The licensee's review of maintenance procedures identified no other cases of inconsistent acceptance criteria. The inspectors reviewed procedure 27562-C, Emergency Diesel Generator Maintenance, Revision 15 and procedure 27598-C, Emergency Diesel Generator Air Start Valve Maintenance, Revision 5. The inspectors. verified that the two procedures incorporated the correct air start valve cap to piston clearance of.002 to.004 inches when new and a wear limit of.0055 inches. The corrective action for violation 50-424,425/92-30-02 was acceptable. This item is closed.. 4.0 Followup on Previously Identified Inspection Findings (92701) Inspector Followup Item (IFI) 50-424,425/92-30 03, EDG Local Load Monitoring This item addressed the lack of procedural guidance to prevent EDG overload when operating in the local mode. The inspectors reviewed the licensee's actions to address this ites. The licensee revised the p.rocedure for EDG local operation to include a note directing the

y:

e 15 . operator to' monitor the EDG phase ammeters during local operation. The note provided a maximum steady state ampere limit to prevent EDG. overload. The inspectors reviewed abnormal operating procedure 18038-1, 0peration From Remote Shutdown Panels, Revision 18 and verified the EDG overload guidance.was included. The licensee's. actions on. IFI 50-424,425/92-30-03 were adequate. This item is closed.. i - 5.0, ' Exit Meeting. The inspection scope and findings were summarized on May 20, 1994, with those persons. indicated in paragraph 1. The inspector described the areas inspected and discussed in detail. the inspection findings listed 3 below. No dissenting comments were received from the licensee. The ifcensee did not identify as proprietary any of the material provided to ' l the inspectors. (0 pen)~ Violation 94-12-01 Failure.to Follow Maintenance Procedures (Closed) Violation 92-30-01, Failure To Identify Conditions Adverse To ~ Quality For EDG 1A Failure of November 18 1992 (Closed) Violation 92-30-02, Inadequate Procedural Acceptance Criteria For EDG Air Start Valve Maintenance - l (Closed) IFI 92-30-03 EDG Local Load Monitoring .( 5.1 Acronyms and Abbreviations ^ Calcon California Controls (company) DC Deficiency Carc DCP Design Change Package j DCR Design Change Request EDG Emergency Diesel Generator LOSP Loss of Offsite Power -i MDD Minor Design Deviation MWO Maintenance Work Order RER Request for Engineering Review VEGP-Vogtle Electric Generating Plant 1 i .i 'I l l 1 I i .a..-

). l UNITSO eTATe8 ~ NUCLEAR REGULATORY COMMISSION l WAeNHeeTON D.C. 300sedett w.... h bruary 13, 1995 EA 93-3041 Georgia Power C ATTN: Mr. C. K. Vice President Nuclear Vogtle Project a 40 Inverness Center Parkway Post-Office Sex 12g5 Birmingham, Alabama 38201

SUBJECT:

N00!FIED NOTICE OF VIOLATION A S PROPOSED IMP PENALTIES $200,000 (Vogtle Generating Electric Plant) Dear Mr. McCoys This refers to your letter dated July 31,19M in response to the Notice of Violation and Prowsed Iagosition of Civil Penalties (Notice) sent to you by our letter dated it 9, le94, and your supplemental response to the Notice dated February 1, 1 5. The Notice described violations identified as a-L(NRC's) Office of Investigations COIresult of an invest December)17,gtle Electric Generating) Plant icensee Vo at Georgia Power Campany s (GPC or 1993. VISP that was completed on received by Region !! in June 1990 alleging, in partThe investig 4 that material falso reliability of the Diesel Generators (Des). statements were m The Notice was based on five instances where the'NRC believ to provide information to the IRC that was complete and accurate in all material respects. regulatory concern, The violations collectively represented a very significant and as such, were cat imod in accordance with the Enforcement Pol as a Severity Level 11 les and a civil penalty in the-amount of $200, was proposed. Three Gemands for Information also issued to GPC regarding the performance failures of six ind'CDFIs)ls were involved in the circumstances of the violations. vidua and E and requested that the staff withdraw, Violation 9.In y You also wit row an t earlier request that the NRC reconsider the assessment of the materiality of l the submittals involved with certain of the violations.In adeitti Violation D,. also requested that the staff recognize your strong belief that, held the April 19 Licensee Event Reportinning with the April 9 presentation an and continuin ust, 1990 recordkeeping practices were a cont ing factor of acreas ng in GPC's inability to provide accurate and couplete data to the NRC.gnifici i j E 'd II'dI E00I I I

j i j Soorgia Power Company The NRC has reviewed your responses and found them to be detailed and h l 2 in providing additional information for consideration in,this metter and the extent of the staff's knowledge accurate. After consideration o,f you,rto responses and the statements of fact explanations,. and a unents for mitigation contained therein, the NRb concludes that Viola ions A, C, D, and E occurred as stated in the Notice and that es set forth in the enclosed 1 Appendix, Violatten 8 should be w,ithdrawn., Although the staff continues to believe that Violation 0 occurred, it does recognize that recordkeepine practices agy have contributed to the violations as events unfolded. The futC also concludes that a Severity Level !! dest'instion continues te be $ppropriate for the problem represented by the remaining violations and that a a 100.000 civil penalty remains appropriate. Accontingly I have decided to i issue the enclosed Modified Notice of Violatten and propo, sed leposition of Civil Penalties in the amount of $200,000. 4' Because you have previously responded to the original Notice under the provisions of 10 CFR t.201, you are not required to submit a written response to the modified Notice. - As provided in the instructions of the enclosed i

Notice, i

Notice, you may pay the civil penalty within 30 days of the date of this by letter addressed to James Lieberman, 01 rector Office of Enforcement' U.S. Nuclear Regulatory Commission; One Whlke Flint North,- 11555 Rockville Pike; Rockv1 le 10 20852-1734, with a check, draft, money order, or electronic transfer, p,ayable to the Treasurer of the United States in the amount of $200,000. The NRC has reviewed your corrective actions and has concluded that the actions taken anc comeitted to !n your initial and supplemental responses are sufficient to provide assurance that events such as those that formed the basis for the lotice should not recur. We will review the effectiveness of your corrective actions during subsequent inspections. This letter also addresses the responses to the three DFIs that we:a issued on May 9 1994. The three DFis discussed the performance failures of six individuals involved in the circumstsecos of the violations a enable the MitC to determine whether additional enforcement actions were necessary. By separate correspondence, the individuals who were the subject of the OFis were given the opportunity to submit separate responses to the DFis. The NRC has reviewed your responses to the DFIs and the additional comments in your supplemental response to the Notice as well as the six individuals' responses to the DFIs and Mr. George Bockhold, Jr.'s February 1 1995-supplemental response to the DFI regarding his performance failu,res. After evaluating the responses the NRC maintains that four of the five originally cited vio ations and the,associsted performance failures occurred as stated in the Notice and DFis. Your supplemental response to the Notice and Mr. Bockhold's supplemental response to the DFI acknowledge Mr. Rockhold's role and responsibility in the events underlying the enforcement action. In an effort to provide the IstC additional assurance that Mr. Bockhold will provide the NRC complete and 1 t 4 DSitt S66t'11*Et

Georgia power Company .3-accurate infomation in.the future current employer Southern Nuclear Mr. Beckhold has requested, and his-Operating Company, has agreed, to laplement-apersonaltrainIngo abilities te perfom pportunity whica focuses uponle in licens and develops, his any future line management ro commensurate with the standard of care reflected in the enforcement Southern Nuclear and SpC have committed to maintain present positten in the Southern Ceepany system, and to prohibit him from Southern Nuclear untti the satisfactory completion o committed to provide the INIC with 60 days notice prior to his assumption of. You also - such a position. Although GpC has identified a variety of corrective actions in an effort to ensure the accuracy and completeness of information provided to the NRC in the future, except for Mr. Sockhold, the DFI responses did not identify individualized corrective actions taken or planned by Spc to address the specific performance failures of.the remaining individual.. careful censideration as to whether additions actions.should be taken withThe i i regard to these individuals, as well as Mr. Beckhold.. to ensure future compliance. The NRC has considered the effect that SpC's general corrective actions have had on these individuals as well as the effect that the 0 had on these employees. These six 6pC esplayees have been publicly identified by MRC as having performed poorly and have had to ceamit time and energy this matter ine uding providing responses to the NRC. This setter has received wide public exposure and has also received wide exposure within the GPC organization. The MC also notes your acknowledgement that all individuals associated with this enforcement action aave learned a great deal about the attention to detail required when making communications to the NRC. In addition, the NRC recognizes that the performance failures of four of the individuals were limited to the submittal of a single letter and in the case of one of the otlar individuals, his performance failures were 11.iited to two.- submittals, In the case of the sixth individual namely Mr. Beckhold acknowledged his role and responsib111ttes with r,essect le four of the, he has submittals and has consitted to the actions noted above. Based on these censiderations, the NRC believes that these individuals will likely conform their conduct to avoid being the subject e,f stellar letc enforcement action. Therefore, no further action (other than that described above regarding Mr. Sockhold) util be taken regardtng these individuals. separate correspondence, the NRC is issuing letters to the six individualsBy stating that the NRC reaffirms its assessment of inadequate individual performance displayed during these events, and stressing the importance of individual accountability in providing complete and accurate information to the NRC. The NRC will provide the individuals with a copy of the Modified Notice of Violation and Proposed Impositten of Civil Penalties to emphasize the seriousness with which the NRC views the violations and associate performance failures on the part of these individuals. will aise confirm its understanding of Mr. Sockhold's cosmitments in ourIn add correspondence to him. isus soi n n

3 Georgia Power Company -4 Subject to SPC's and Mr. Bockhold's consiteents, the NRC staff conclud the. involved individuals' actions do not warrant any addittenal enforcemen sanctions,-including letters of reprimand. penalty, we will consider this. enforcement action fully resolved.Upon subject to BPC's and Mr. tockhold's commitments, the lac staff has no ere Also concerns with the character and intes ty of the individuals or the SPC-arising out of the events that were t subject of the Notice and DFis. this letter and the enciesures will be placed in the Room. $1ncerely, [, ~a N M11hoan heuty Executive ofrector for Nuclear Reactor Resulation Regional Operations and Researc,h Docket No. 50-424 License No. NPF-44 4 EA 93-304 Enclosures 1. Modified Notice of Violation and Proposed Imposition of Civil Penalties - $200,000 2. Appendix 1 r ) 4 4 ? .a 4 i a C

• d 15:24 5661*tt*te 08d-

M00! Fit 0 NOTICE OF VIOLATION -AND PROP 0$tD. IMPOSITION OF CIVIL PENALTIES Georgia Power Company Vogtle Electric Generating Plant ' Docket No. 424 License No. NPF-68 .EA 931304 Duri an NA1 an NRC inspection conducted from August 6,1990 to August 17, 1990 and investigation completed on December 17 1993, violations.of NRC-requirements were identified. In accordance w,ith the " General Statement of i Policy and Procedure for NRC Enforcement Actions, 10 CFR Part 2, Ap the Nuclear Regulatory Commission. proposes to im i 42 U.S.C. 2282. and 10 CFR 2.205.gy Act of 1954, pose civil pe to Section 234 of the Atomic Ener i as amended The particulcr violations (Act).. civil penalties are set forth below: and associated j licenseeshal)becompleteandaccurateinallmaterial requires that infomation provided to the NRC by a { A. Contrary to the above, information provided to the NRC. Region II'0ffice by Georgia Power Company (GPC) in an April 9,1990 letter and in an'. i April 9, 1990 oral presentation to the NRC was inaccurate in a material respect. $pecifically, the letter states that: "Stace March 20, the IA.- 00 has been started la times ~* No failures or problems have, occurred during any of these start These statements are inaccurate in that they represent'that 19 t i consecutive successful starts without problems or failures had cccurred' i on the IB Diesel Generator (DG) for the Vogtle facility as'of April 9,1990, when, in fact, of the 19 starts referred to in the letter ) associated with the 18 08 at the Voqtle facility, three of those starts had eroblems. Specifically,-Start,32 tripped on high temperature lube. i oil, Start 134 tripped on low pressure jactot water and Start 134 had a high temperature jacket water trip alarm.. As of April 9.11990 the 18 1 OG had only 12 consecutive successful starts without problems e,r failures rather than the 19 represented by GPC. The same inaccuracy was presented to the NRC at its Region !! Office during-an oral pres >mtation by GPC on April 9, 1990. l i The inaccuracy was material. In considering a restart decision, the NRC was especially Interested in the reliability of the OGs and specifically. } asked that GPC address the matter in its presentation on restart. NRC relied, in part, upon this information presented by GPC on-The 4 4 April 9,1990 in the oral presentation and in the GPC ;etter in reaching the NRC decision to allow Vogtle Unit I to return to power operation. i 8. Contrary to the above, information provided to the NRC by SPC in a Licensed Event Report (LER), dated April 19, 1990, was inaccurate in a material respect. Specifically, the LER states: " Numerous sensor ) calibrations (including jacket water temperatures), special pneumatic i i is.4 seset rze

i Notice of. Violation. - 2'- various conditions. After the 3-20-90 event, t both engines have been subjected to a comprehensive test pro 9 r Subsequent to this test program, DGIA and 0G18.have been starte J 1 cast 18 times each and no fa11eres or problems have occ of these starts." These statements are inaccurate in that they represent that at leas consecutive successful starts without problems or failures had occurred-on the DGs for Vogtle Unit I (1A OG and it OG) following the comp i of the comprehensive test program of the control systems for these i when, in fact following completion of the comprehensive test p the control systems, there were no more than 10 and 12 consecuti; successful starts without problems or failures for IA OG and 18 OG respectively. number of consecutive successful starts en LA D problems or failures could have had a natural tendency or capabilit cause the NRC to inquire further as to the reliability of the DGs. y to C. cover letter dated JuneContrary to the above, information provide 29, 1990 was inaccurate and incomplete in material respects as evidenced by the following three examples:. The letter states that: ) Comoany (GpC) hereby submits the enclosed revised re event which occurred'on March 20, 1990. clarify the information related to the number of successful dieselTh generator starts as discussed in the GPC letter dated April 9,1990...." 1 1. The LER cover letter is incomplete because-the submittal did not provide information regarding clarification of the April 9,' 1990 1etter. The incompleteness was material in that the NRC subsequently letter. requested GPC to enke a submittal clartfying the April 9,1990 i r 4 t The letter states that: "If the criteria for the completion of the. test 4 Vogtle Electric Generating Plantprogram is understood to be the f VEGP) procedure 14960-1 " Diesel Generator Operability Test," then(there were 10 succes'sful sta Diesel Generator IA and 12 successful starts.of Diesel Gene between the completion of the test program and the end of April 19, 1990, the date the LER-424/1990-06 was submitted to the NR Tie number of successful starts included in the original LER (at least

18) included some of the starts that were part of the test program.

difference is attributed to diesel start record keeping practices and The the definition of the end of the test program." 1 4 4 es,4 sost tt ce houd

' Notice of Violation - 3.- '2. The last sentence in the above parafraph'is inaccurate because diesel record keeping practices were not-a cause of the difforence in number of diesel starts reported in the April 19. 1990 LER and the June 29 1990 letter. errors unrelated to any problems with the diesel gener keeping practices. The inaccuracy was material in that' tt could have led the NRC to erraneously conclude that the correct root causes for the difference in the number of diesel starts reported'in the ' april-19,1990 LER and the June 29,1990 letter had been identified by GPC. 3. The last sentence in the above paragraph is also incomplete because it failed to include the fact that the root causes fo difference in the number of diesel starts reported-in the april 19, 1990 LER and the June'tt, 1990 letter were personnel-First, the Vogtle plant General Manager who directed the errors. Unit Superintendent to perfore the start count-(which formed the basis for the April 19, 1990 LER) failed to issue adequate instructions as to how to perform the count and did not adequately assess the data developed by the Unit Superintendent. _ in addition, the Unit Superintendent made an error in reporting his count. Second, the Vogtle Plant General Manager, the General Manager for Plant support and the Technical Support Manager failed to e ariff ar.d verif/ the starting point for the count of successful consecutive DG starts reported in the April 19, 1990 LER. The incompleteness was material in that,-had correct root causes for the difference in the number of diesel starts reported in the April 19, 1990 LER and the June 79,.1990 letter.been presented, this information could have led the NRC tePseek'further infomation. D. Contrary to the above, information by GPC in a letter dated August 3/e,provided to the NRC Region II Office in material respects as evidence ( by the following two examples The letter states that: original LER appear to be the result of two factors.*The c9nfusion in t First, there was confusion in the distinction Ietween a successful start and a valid test... Second, an error was made by the individual who performed the count of DG starts for the NRC April 9th letter.' 1. These statements are inaccurate in that confusion between a successful start and a valid test was not a cause of the error regarding 04 start counts which GPC made in its April 9, 1990 letter to the NRC. e4

s ui sost si ce uses

c n f q. Notice of Violation The inaccuracy was material. in that it could have led the ifRC to' erroneously conclude that the correct root causes for the error in i the April 9, 1990 letter had been identified by GPC. 2. The statements are also incesplete. While an for'the' April 9,1990 letter, the root causes of the error in that i letter were not completely identified by GPC. ' Specifically,'the Vogtle Plant General Manager who directed the Unit Superintende to perform the start count failed to issue adequate instructions as to how to perform the count and did not adequately assess the data developed by the Unit Superintendent. In addition, the Unit J Plant General Manager. Superintendent did not adequately l } The-incompleteness was material in that, had the correct root causes for the error in the April 9,1990 letter regarding DG start counts been reported NRC to seek further informa, tion.this information could have led the These violations in the aggregate represent a Severity Level !! problem (Supplement VII. Civil Penalty -)5200,000 Secause Georgia Power Company responses oursuant to the prov(Licensee) has already provided written. Modified tiotice of Violation, the Licensee is not required to response to this Nodified Notice of Violation. Within 30 days, the Licensee may pay the civil penalties by. letter' addres to James Lieberman, 01 rector, Office of Enforcement. U.S. -Nuclea Commission One White flint North 11555 Rockville hiket R 2738 with a check, draft, money o,rder er electronic transfer pay,able to the Treas;urer of the United States in tho'a, mount of $200,000 or may, protest imposition of the civil penalties in whole er in part, by a written answer addressed to the Director Office,of Enforcement, U.S. Nuclear Re9ula Commission. Should the Licenses fail to answer within the time specified, an order imposing the civil penalties will be issued. to file an answer in accordance with 10 CFR 2.205 protestinj1 the civi penalties. in whole er.in part, such answer should be' clear y marked as an " Answer to a Notice of Violation." Upon failure to pay any civil penalties due which subsequently has been determined in accordance with the applicable provisions-of 10 CFR 2.205, t matter may be referred to the Attorney General, and the penalties, unless compromised, remitted, or mit194ted may be collected by civil action pursuant to Section 234(c) of the Act. 42 U.$.C. 2282c. e h,d h

Notice of Violation The response noted above (letter with payment of civil penalties or Answ a Notice of Violation) should be addressed to: James Lieberman, Otractor f North, 11555 Rockville Pike, Rockville, mOffice of Enforceme 20852-2738, with a copy to the Regional Administrator, U.S. Nuclear Regulatory Commission, Region !! a copy' to the NRC Resident Inspector at the Vogtle facility. Dated at Rockville, Maryland this day of February,1995 a t 4 J i .,g., esiti sess st as = a

q L APPEN0!X 3 ~ EVALUATION AND CONCLUSION On May 9.1994, a Notice of Violation and Proposed leposition of Civil Penalties (Notice) _was issued for violations identified during an NRC inspection and investi the Notice on July 31,gation. Georgin Power Company 1994, includ ng a Aeply pursuan(t to 10 CFR t.201 and Answer pursuant to 10 CFR 2.205 and supplemented response on February 1,19 Based on its supplemental response, the Licensee admitted Violations A, C, and E, and denied Violation B. Violation B Is as follows: The NRC's evaluation and conclusion regarding-- Rastat - nt af Vialattan R by GPC in an April 9,1990 letter was incomplete i Specifically, the letter states, when discussing the air quality of the DG starting air system at the Vogtle facility, that: air quality of the O 'GpC has reviewed concluded that air qu/4 air system including dewpoint control and has ality is satisfactory. H than expected dowpoints were later attributed to faultyInitial reports of h instrumentation." This statement is incomplete in that it fails to state that actual high dew points had occurred at the Vogtle facility. that the causes of those high dew points included failure to use airIt ais dryers for extended periods of time and repressurtration of the DG air j start system receivers following maintenance. j The incompleteness was material.. In considering a restart decision, the NRC was especially interested in the reliability of the DGs and i specifically asked that GPC address the matter in its presentation on restart. The NRC rolled, in part GPC in its letter of April 9,1996 in reaching the decision to allowupo Vogtle Unf t I to return to power operation. fr arv of Licensas's Raar te ta Violattaa R GPC argues that its April 9,1990 letter addressed, accurately and completely, the on-going events related to concerns about dowpoint data, i The statement about initial reports referred to a high dowpoint reading i measured on March 29, that was first reported to NRC representatives in the April 5-9, 1990 period dowpoint measurements taken(during the recovery from the S To suggest that the letter either sought to identify or explain an higher than expected dowpoints is to take GPC's statement out of context. This' would give it a meaning which is inconsistent with the actual t understanding of GPC and NRC representatives at the time. Prior to the NRC's decision to allow Unit 1 to return to power operation, SpC kept the NRC informed of actual high dowpoints on the 1A DG control air and n., ts u s ssst st*te I

7 6 1 1 Appendix provided oral informatter. on other engines. of the NRC substantiate the context and meantng of-the state i understanding of the statement's meaning, by NRC representatives and information conveyed to the NRC prior to restart. GPC argues that' the April 9,1990 letter identified certain short-term. corrective actions. GPC contends that there can be little doubt th the letter was discussing the current situation and it is. unduly j strained to say the statement was intended to describe all past.. saintenance issues. expected dowpoints in the distant past attributed to sy being out of service and system repressurization followiry maintenance was not reasonably necessary to consistely describe the short-term Moreover, changes in preventive maintenance pra 'tigh dowpoint readings after the SAE.. wre distant dowpoint measurements such less informative about air - quality than recent data Applying a rule of reason, the information in + the April 9 letter was a complete explanation of the basis.for SpC's closttre of dew point concerns which arose subsequent to the SAE. Based on the above arguments, GPC requests ~ that'Violatton 8 be withdrawn. NRC Evaluation of Licennas's Rennanna to Violatlan 8 \\ Upon further review, the NRC concludes that GPC's statements rega air quality presented in the April 9,1990 letter were sufficient in ' i air quality was acceptable. scope and GPC had an adequate technical b in response to the event, in order to deterniine if air quality was a root cause of the DG performance on March 20 GPC inspected air filters on the contrul air system that had been pulle,d in early Merch 1990. They also conducted an internal inspection of the DG air receivers after the March 20 event. Dewpoint measurements on March 29 for DG 1A air receivers that were ostside specified acceptance criteria were determined to be due to a faulty instrement. GPC replaced the instrument and the resulting readings were satisfactory. This violation was premised on the NRC's conclusion that the reference to " initial reports of higher than expected dowpoints" was part of GPC's effort to present a comprehensive review of past air quality problems,- including problems occurring prior to the SAE. The NRC relied on information contained in Inspection Report 50-424,425/90-19 Supplement 1, that indicated that there had been high dewpoint readings related to air dryers being out of service and system repressuriaation in addition to those attributable to faulty instrumentation. The NRC belleved that the high dowpoint readings referenced in the report preceded the SAE. This information led the NRC to conclude that the information on air quality contained in the April 9 letter was incomplete. The NRC did not view the April 9 letter as focusing the discussion on air quality to only activities contemporaneous with the i ,s m sui ri ce --w oe-m c- -, - ~.eyv we-

1 Appendix ,3, event and subsequent recovery. The MC agrees with_ GPC that historical information was not'necessary for a restart decisio therefore, the April 9 letter was not incomplete. Based on the above evaluation, the MC concludes that V be withdrawn. o l' 4 I et 4 es,4 ssst et ce woes

jaun ^ es..a. s na so.e Aseg k ? UfMTED STAtt$ - NUCt.5AR REGULATORY 00 MMS $10K WAteessefest, p.o. speas.eset FebrJary 13, 1995 Thomas V. Greene Southern Nuclear Operating Compaty 40 Inverness Center perhuny 81rsingham, Alabama 38201

SUBJECT:

GEORGIE R. FREDERICK, HARRY MAJOR

Dear Mr. Greene:

Demand for InformationThis refers to your letter dated August 9,19M in Licensee) and sent to yo(OFI)~ issued to the u by our. letter dated la Power Ceepony (GPC or 9 1994.' The DFI addressed your contributions to GPC's failure to provide t ERCwithinformation= material respects.regarding the Vogtle diesel generators that was cesplete an i The NRC has reviewed your response to the DFI in conjunction with SPC's response to the DFI and GPC's initial and supplemental responses to the Notice of Violation and Proposed Imposition of Civil penalties (Notice) that was aise. issued on May 9, 1994. After evaluating the responses, including your dental of;de failures as described in the DFI, the NRC maintains that our perfonaance 3 ur of the five violations and associated performance failures occurred as stated in the i Notice and DFI. actions in an effort to ensure the accuracy and completeness i i .provided to the NRC in the fkture, the DFI response fails 1 specific performance failures. The NRC has given careful consideration to the i performance failures to ensure future temp 11ance. question as i i The MC bas considered that ' your performance failures were limited to the submittal of one letter (June 29, 1990, and has considered the effect that SPC's general corrective actions will l)kely have on you and the remedial effect that the i and DFI process itself has Itkeiy had on you. After considerins all of the circumstances in this matter, the NRC has concluded that ne further action should be taken regarding your actions. I have included a copy of the Modified Notice of Violation and proposed i-1.nposition of Civil Penalties that is being issued to GPC on this date to i emphasize the seriousness with which the NRC views the violations and [ associated performance failures en the part of the individuals involved in the Y i l I m

x P Thomas V. Greene 2-circumstances of the violations. P NRC-licensee, you have.an individual responsibility and O ensure that all information provided to the NRC, whether o is complete and accurate in all material respects. y to w ng, S1ncere1y, ( N.t.7Ed h -Mllboan

1. Mputy Executive Director V for Nuclear Reactor Regulation, Regional Operations and Research i

Enclosure:

As Stated cc w/o

Enclosure:

i Georgia Porer Company [ \\ .si. I 'I i I f I i

[ 7 e.. = uestas starse s MLEAR REStJLATORY COMMISSION wAsemeeron o.o.mement February 13, 1995 George Beckhold Jr. Southern Nuclear, Operating Company i 40 Inverness Center parkway ~ tirmingham,. Alabama 38201 $UBJECT: RE$p0NSC TO DDIAND FOR INF0NIATION RESARDINS S

Dear Mr. Beckhold:

{ response to the subject Demand fThis refers to your letters date Pouer Company (SpC or Licensee) or Infomatten (by o)ur letter d! DFI and sent to you May g lyet. i to provide the NRC with information regarding the Vo that was complete and accurate in all esterial rupsets. The NRC has reviewed your responses to the DFl in con.iunction with of Violation and Proposed Imposition of Civil penalt issued on May 9, 1994. After evaluating these responses the NRC mai occurred as stated in the Notice and DFI.that four of the five v i In your supplemental response to the OF1, you acknowle responsibilities in the events vederlying the enforce:dpod your role and contrary views and cautious in formulating conclusions.t ent action and expressed would like to continue to be employed in your current position until such You stated that you as you have reacquired the confidence of the licensee and your err:yer. i also stated that you understead that additional assurance was requir You HRC. In an effort to provide this addittenal asserence you stated that yo have requested that your current esplayer Southern Nrcieer Opera provide manager'you with an opportunity for intens,ive training in the areas of a communicetor with co-workers and regulatory as obilgations an ear ladestry and as a 1 i the NRC after you have coupleted this effort.gencies. You consitted to notify your frfomance in the areas of communication effectiven i detti attention to accountability for actions, and any other standard that your esplayer l-identifies. You also consitted not to seek a line management position over l licensed activities at any nuclear power satisfactory completion of this training.pitat licensed by the NRC untti after Thereafter if you are nominated for a position in line management within 3 years of Fe,bruary 1 committed to infom the NRC of that nominatten at least 50 days, prior to 1998, you assuming the positten. You expressed your view that with these additional steps on improving your management style, and reinforcement of your curren sensitivity to the very high standard legitimately required in connunications with the NRC, the NRC can have confidence in your involvement in licensed f -.~ i

George lockhold, Jr. 2-sensitivity to the very high standard legitimately required in with the NRC, the NRC can have confidence in your inv activities in the future. mmunications censed additional actions shouldThe NRC has given careful cons ensure future compliance. be taken with regard to your performance f e er The NRC has consid6 red your acknowledgeme role and responsiH11ttes in the events underlying the enforc effect that GPC's general corrective actions will likely have remedial effect that the enforcement and DFI process itself h n action, the you, and your commitments to take the actions noted above the to your commitments, no further action s,hou e y had on , subject imposition of Civil Penalties that is being issu ng your actions. roposed emphasize the seriousness with which the N s date to circumstance NRC-licensee,s of the violations. you have an individual responsibility and account n the ensure that all information provided to the NRC, whether orally or i is complete and accurate in all material respects.

writing, Sincerely, p [.

hf: D mes L. Nilhoan eputy Executive Director for Nuclear Reactor Regulation Regional Operations and Researc,h

Enclosure:

As stated cc w/o

Enclosure:

Georgia Power Company l l l f L

1 l( U$NTED eTATs3 i NUCLEAR RESULATORY COMMISSION WASHINetosi,e.c.IWeOHept i Fetruary 13, 1995 Harry Majors. Southern lluclear Operating Company 40 Inverness Center Parkway Bimingham, Alabama 35201

SUBJECT:

RESPONSE TO DDIAND FOR INFORMATION REGARDill6 THOMAS V. GREENE, GEORGIE R. FitEDERICK, HARRY MAJORS, AND NICHAEL W. HDRTON

Dear Mr. Majors:

.This refers to your letter dated August 5,1994 in response to the subject ~ Demand for Infomation (DFI) issued to the Geo ia Power Company (GM or i Licensee) and sent to you by our letter dated 9, 1994. The DFI addressed your contributions to GPC's failure to provide t flitC with information. regarding the Vogtle diesel generators that was complete and accurate in all i material respects. ,1 The.futC has reviewed your response to the DFI in conjunction with SM's-response to the DFI and GM's initial and supplomaatal responses to the llotice of Violetten and Proposed Imposition of Civil Penalties (Notice).that was also issued on May 9, 1994. After evaluating the responses, including.your denial of your perfemance failures as ; described in the DFI, the NRC maintains that four of the five violations ard asso: lated performance failures occurred as stated in the-Notic~e and DFI. Although GK has identified a variety of general corrective actions in an effort to ensure the accuracy and completeness of information i provided to the NitC in the future, the DFI rosponse fails to identify individualized corrective actions taken er planned by Spc to address your specific perfomance failures. The hitC has tiven careful. consideration to the question as to whether additional actions should be taken with mgard to your performance failures to ensure future compliance. The INIC has considered that your performance failures were limited to the submittal of one letter-(June 19 19901 and has considered the effect that EPC's <peneral corrective actions w,ill Idely have on you, and the remedial effect tist the enforcement and DFI process itself has likely had on you. After considering all of the l circumstances in this matter, the NRC has concluded that no further action should be taken regarding your actions. I I have included a copy of the Modified Notice of Violation and Proposed i Imposition of Civil Penalties that is being issued to GPC on this date to emphasise the seriousness with which the NRC views the violettens and + associated performance failures on the part of the individuals involved in the i

q Harry Majors-2- l circumstances of the violations. You are reminded that, as an employee of an I NRC-Itcensee, you have an individual responsibility and accountability to. 1 ensure that all information provided.to the NRC, whether orally or in writing, is complete and accurate in all material respects. Sincerely, [ L YA. L-a L. Milhoan ~ puty Executive Director for Nuclear Reactor Regulation, Regional Operations and Research

Enclosure:

as Stated cc w/o

Enclosure:

Georgia Power Company i i 1 1 l l \\

\\l unna owne j NUCLEAR REGULATORY COMMISSION waewswovow, o.o. seassenes . \\**=** i Mbruary 13, 1995 Kenneth licCoy Southern lluclear Operating Company 40 Inverness Center Parkway Birmingham, Alabama 35201

SUBJECT:

RESPONSE T0 DEMAle FOR INF0lplAT!0N REGARDING KE101ETH McC0Y i

Dear Mr. McCoy:

i This refers to your letter dated August 5 1994 la response to the subject Demand for Infomation (DFI) issued to the, Georgia Power Company (GPC or Licensee) and sent to you by our letter dated May 9,1994. The DFI addressed your contributions to spC's repeated failures to provide the llRC with information reparding the Vogtle diesel generators that was complete and accurate in al material respects. j The NRC has reviewed your response to the DFl in conjunction with GPC's response to the DFI and GPC's initial and supplemental responses to the flotice 1 of Violation and Proposed Imposition of Civil Penalties (Notice) that was also issued on May 9, 1994. After evaluating the responses, including your denial of your performance failures as described in the DFI the NRC maintains that four of the five violations and associated p rform,ance failures occurred as stated in the Notice and DFI. Although 'iPC has identified a variety of general corrective actions in an effort to ensure the accuracy and completeness of information provided to the IIRC in the future, the DFI response fails to identify individua11 sed corrective actions taken er planned by GPC to address your specific performance failures. The NRC has given careful consideration to the-question as to whether additional actions should be taken with regard to your performance failures to ensure future compliance. The NRC has considered that your performance failures were limited to the submittal of tuo letters (June 29. and August 30, 1990)l likely have en you and the remedial effect and has considered the effect that GPC's eneral corrective actions wil ! hat the enforcement and DFI process itself has likely had en you. After considering all of the circumstances in this matter, the NRC has concluded that no further action should be taken regarding your actions. I have included a copy of the Nodified Notice of Violation and Proposed Imposition of Civil Penalties that is beins issued to GPC on this date to emphasise the seriousness with which the NAC views the violations and associated performance failures on the part of the individuals involved in the 4 ~ * ~ r ,,,e,m.-,en,.e.., .-.,e-..n, I

1 Kenneth McCoy l circumster.ces of the violations. You are reminded that, as an employee of an NRC-licensee, you have an individual responsibility and accountability to ensure that all information provided to the NRC,-whether orally or in writing, is complete and accurate in all material respects. Sincerely, f /. M mes L. Milhoan l puty Executive Director for Nuclear Reactor Regulation, Regional Operations and Research

Enclosure:

As Stated cc w/o

Enclosure:

Georgia Power Company i I I

n.4 i h p_ UNITED eTATee y\\e...e NUCLEAR REGULATORY COA 4 MISSION wasumeton, s.o. seassa I February 13, 1995 t Georgie R. Frederick GeorgiapowerCompany Vogt e Electric Generating Plant River Road Waynesboro, Georgia 30830

SUBJECT:

Resp 0NSE TD DDIAle FOR INF0ftl4T!0N REGARDING THOMAS V. GREENE! GEORGIE R. FREDERICK, HARRY M4JDRS, AlO NICHAEL N. HORTON i

Dear Mr. Frederick:

This refers to your letter dated July 28, 1994 in response to the sub, ject. l Demand for Information DFI ' Licensee 1 and sent to yo(u by) our letter datedissued to the Geo la Power Co g t 9, 1994. The Ol! addressed your conhributions to SpC's failure to provide NRC wi n. taformation regarding the Vogtle diesel generators that was cosplete and accurate in all material respects. { The NRC has reviewed your response to the DFI in con.1 unction with GPC's L t response to the DFI and GPC's initial a::d suople v al responses to the Notice i i of Violation and proposed leposition of Civil pensities (Notice) that was also issued on May 9, 1994. l After evaluating the responses including your dental of ;eur performance failures as described in the Off, the NRC cahtains that Jour of the five i violations and associated performance failures occurred as stated in the Notice and DFl. Although GPC has identified a variety of general corrective s actions in an effort to ensure the accuracy and coop 1stoness of infomation i provided to the F.C in the future, the DFI response fails to iden+1fy individualized corrective actions taken or planned by SPC to address your i specific performance failures. The NAC has given careful consideration to the question as to whether addittenal actjens should be taken with regard to year perfomance failures te ensure future compliance. The NRC has considered that our performance failures were limited to the submittal of one letter y(June 29,ill lif and has considered the effect that SpC's peneral corrective 1990) sctions w tely have on you and the remedial effect t ut the enforcement-and DFI process itself has likely had on you. After considerins all of the circumstances in this matter, the NRC has concluded that no forther action i should be taken rstarding your actions. j i I have included a co>y of the Meditted Notice of Violation and proposed !aposition of Civil 4nalties that is being issued to GPC on this date to i emphasize the seriousness with which the NRC views the violations and l associated performance failures on the part of the individuals involved in the i e-

n- } .........a a .1 'l l Georgie R. Frederick i - circumstances of the violations. NRC-licensee, you have an individual responsibility and accountab ensure that all information provided to the NRC, whether ora 11y or 11ty to is complete and accurate in all material respects. Sincerely, i n s L. M11hoan f))eputy Executive Director [/ for Nuclear Reactor Regulation. s Regional Operations and Research

Enclosure:

As Stated cc W/o Er.;1osure: Georgia Power Company i l s t r f e

[\\ v.,ms,s w ee s U \\*...* wasseneten, o.o. asasseem rebruary 13, 1995 Michael W. Horten Southern Nuclear Operating Company 40 Inverness Center Parkway Bimingham, Alabama 35201

SUBJECT:

RESPONSE TO DEN 4W FOR INFORMATION REGARDING T60145 V. GRE GEORGIE R. FREDERICK, HARRY MAJOR $, A m MIC6 EEL W. NORTON-

Dear Mr. Herten:

This refers to your letter dated July 29 Demand for Informatten (DFI) issued to tb Geo1994 in response to the subject Licensee) and sent to you by our letter dated 14 Power. Company (SpC er 9, 1994. The DFI addressed your contributions to spC's failure to provide e M C with information regarding the Vogtle diesel generators that was complete and accurate in all material respects. The NRC has reviewed your response to the DFI in conjunction with SpC's response to the DFI and GPC's initial and supplemental responses to the Notice of violation and Proposed Impositten of Civi1~ penalties (Notice) that was also - issued on May 9,.1994. After evaluating the responses including your dental of year perfernance failures as described in the DEI the NRC maintains that four of the five violatiens and associated perform,ance failures occurred as stated in the Notice and DF1. Although spC has identified a variety of general corrective actions in an effort to ensure the accuracy and coupleteness of information provided to the MC in the future, the DFI response falls to identify individualized corrective actions taken or planned by Spc to address your i specific performance failures. The NRC has given carsful consideretten to the guestion as to whether additional actions should be taken with regard to your perforisance failures to ensure futere compliance. The NRC has considered that your performance failures were limited to the submittel of one letter (June 19 1990 actionsw,illI]Eelyhaveonyouand has considered the effect that 8PC's general c and the remedial effect tant the enforcement and DFI process itself has likely had on you. After cohsidering all of the i circumstances in this matter, the NRC has concluded that no farther action should be taken regarding your actions. I have included a copy of the Modified Notice of Violation and proposed Impositten of Civil Penalties that is being issued to GpC on this date to emphasias the seriousness with which the NRC views the violations and associated performance failures on the part of the individuals involved in the a

1 Michael W. Norton circumstances of the violations. You are reminded that, as an employee of r... NRC-licensee, you have an individual responsibility and accountability to ensure that.all information provided to the NRC. whether orally or in writing, is complete and accurate in all materfal respects. Sincerely, I b k. F s L. Nilboan puty Executive Director for Nuclear Reactor Regulation Regional Operations and Researc,h

Enclosure:

As Stated cc w/o Er..losure: Georgia Power Company a i r k

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