ML19322C477
| ML19322C477 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse, Arkansas Nuclear, Rancho Seco, Crane |
| Issue date: | 03/19/1979 |
| From: | Foster J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | Creswell J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| References | |
| TASK-TF, TASK-TMR NUDOCS 8001170807 | |
| Download: ML19322C477 (14) | |
Text
CRESO4LL I*
FE 0 TO PERSONAL FILE On Friday l M'rch 16,;1979 (from 9:00 to 11:30 a.m.) C. Morelius, L. Speccard, J. Streeter, J. Foctor, T. Tambling, J. Kohler and J. Creccwell met to discuss the findings of a recent investigation at Davic-Bessie, including possible itema of noncompliance.
C. Norelius opened with a general comment on the investigation, and commente on the purpoco of the meeting: to resolve differences of interpretation on the findings of the invcttigation and clarify remaining iccues.
A pa$lage reprocenting major pointo found was followed throughout the presentation by Kohler and Foster. This packa6e is attached.
Discuccions were held on ceveral pointc, includinC infomation gathered, iccuca not covered, technical significance of the two areas investigated. J.:Cresswell indicated that he ind passed his concerns regarding Loss of Pressurizer Level Indicatior.(LOPLI) on to the ASLB.
i Mr. Cresswell posed several questions during the discuccions, including:
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- 1. Were the invectigators certain that the previous instances of=LOPLI had been-
" reviewed by the NRC?'
Ancwor: Yes, although no report had been made of the Arkancas events, each of the tranciento had been reviewed, and LOPLI was precent during the evento. In como cases, it may be that no mention of LOPLI was specifically noted.
- 2. Had the investigators reviewed the performance of the Once Throu6h Steam Generators during the November tranc ent? This was fmm a remark in paragraph 3 of a memo demonctrating tinlinecc o CO LOPLI evaluation.
Answer. No. Not within the invecti tion's coope.
- 3. Had anyone questioned the " sanctity" of the thirty-second ECCS criteria during the investigation?
Ancwer:iNo. Not within scope of the investigation or expertice of the invecticators.
- 4. Had any part of the TECO procedure for initiating an FCR been violated during incue and implementation of FCR'c related to the undervoltage relay cetpoint?
Answer: Not investigated.!Nct within scope of the investigation.
- 5. Had the' investigators considered memos from Mr. Buck, which advised of
' outstanding W R's?
Answer: Yes, and the memoc will be documented in the investigation report.
- 6. Would it be useful for J. Smith to review the July 18, 1978 sumbittal to NRR;
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which resulted in the approval of Ammendment #_7?
Answer:' No. NRR han. reviewed and approved the submittal.
Inc1gsion of an item of noncompliance related to a lift wire procedure which violated a Technical Specification was discucced. This iccue was not included. in the investigation ac no new infomation was required to establich non-compliance.
Following long discussion, it was agreed that no items of noncompliahce be cited with the investigation report, but concerns as to TECO'c management would be contained in the report transmittal letter.
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- . Mr. Creswell
- attempted to e:rcuse himself approximately 3/4 of the way through the
" meeting. He stated that "this has become a manaCement discussion now....I have some memos to write." Mr. Streeter indicated that he should stay until the conclusion of the meeting.
Throughout the meeting,.Mr..Cresswell's attitude suggested that he thought that the p sues he had raised had not been properly investigated in that the investigation had not been widened to other issues. Questions he raised on OTSGs, K R procedures, ect., appeared to be grasping at straws to find some significance to the issues he had raised.
On a more personal level, and on the basis of statements Mr. Cresswell has made to me over several weeks, the following opinions are made:
- 1. He will not give up his" concerns" even in the face of overwhelming evidence that they are not sicnificant.
- 2. He vill pursue any aspect of the issues raised which he can, within the regular inspection procr I refer specifically to perfomance of OTSG's during the November transien pA6c40(/Agr.'
- 3. He has several other concerns related to Davis-Bessie which were not given to the investigators. He will continue with these, and object to any resolution of these issues.
4.'Iffeel that his attitude borders on paranoia, and that he will become increasingly-distrustful of regional management. I believe that he now feels that I was " soft"
.on TECO, and so was Joel Kohler.
- 5. If the progressive distrust of regional personnel proceeds unabated, he will cc@ lain to higher levels of NRC and finally "go public" perhaps with the support of some anti-nuclear group. I hope that this does not occur, as I feel that he is technically competent, and sincere. However, his outlook on Davis-Bessie clouds his technical judgement.
- 6. Sono concerns as to TECO management are well-founded. However, there is no
' evidence that they knowingly violated NRC regulations.
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I SCOPE I. IASS OF PRESSURIZER LEVEL INDICATION (LOPLI)
E A.iTQINESS OF EVALUATION OF IDPLI BY TECO POLI 4 WING.11/29/77 TRANSIENT
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B. POSSIBLE GENERIC OCCURRENCES OF 1hPLI AT OTHER B&W PLANTS INFORMATION TO BE GATHERED:
- 1. OTIER OCCURRENCES AT B6M PLANTS
- 2. CHRONOLOGY OF TECO EVALUATION II. UNDERVOLTAGE RELAY SETPOINT A. TIMELINESS OF IMPIEMENTATION OF TECH. SPEC. CHANGE B. REASONS FOR ABOVE.
INFORMATION TO BE GATHERED: CHRON014GY OF EVENTS AND REASONS FOR SAME.
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2 LOSS OF PRESSURIZER IIVEL INDICATION (LOPLI) 1.14PLI HAS OCCURED AT DIER BW PIANTS AS NUIED:
ARKANSAS 10/74, 12/74, 5/75 THEE MILE ISIAND 4/78, 11/78 RANCHO SECO 74, 75, 78 (several)
b TE BW SIMULATOR IS PROGRA)MED TO IAOSE PESSURIZER IZVEL UNDER CERTAIN CONDITIONS.
- 4. BW STATED THAT IDPLI WAS AN OPERATIONAL "EADACE" BUT NUI A SAFETY PROBLEM, AS PRESSURIZER IEVEL CANNOT FALL BELOW A CERTAIN LEVEL WITHOUT A DROP IN T5fE RCS PRESSURE WHICH WILL INITIATE HPSI.
- 5. THE BW ANALYSIS OF TE PENOMENOM INDICATED THAT CORE COVERAGE WILL BE MAINTAINED AND ECCS WILL OCCUR AS ANALYZED.
- 6. BW ADVISED TECO OF TE POSSIBILITY OF LOPLI DUE TO MAIN STEAM SAFETY RELIEF B14WDOWN SETTINGS PRIOR TO POWER OPERATION.Sdf/N4f Nef /AP/MAPANflD.
- 7. ON NOVEMBER 29, 1977,,A REACTOR TRANSIENI RESULTED IN LOPLI AT DAVIS-BESSIE.
- 8. IN ADDITION TO STEAM SAFETY RELIEF VALVE"SEITINGS BEING OUTSIDE OF RECmMENDED SETTINGS, SIZE AND IESIGN OF TE AUXILIARY FEEDWATER PUMPS AT DAVIS-BESSIE 1
CONIRIBUTED SIGNIFICANILY TO IMPLI. THIS IS A DESIGN DIFFERENE BEIVEEN DAVIS-BESSIE AND OLDER BW PLANTS. TE RESULTING I4PLI IS COMPOUNDED BY THIS ESIGN
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BUT THE CONCLUSION OF THE VENDOR IS THAT SYSTEM RESPONSE IS IDENTICAL.
- 9. DOCUMENTATION EXISTS THAT TECO DISCUSSED I4PLI WITH BW SOON AFTER TE TRANSIENT AND CONCLUDED THAT NO SAFETY PROBIIM EXISTED. TECO WAS EXPLORING CORRECIIVE ACTIm REGARDING AUXILIARY PEEDWATER PUMP FLOW CONTROL AS EARLY AS FEBUARY 10, 197$. NO NEW I4PLI ANALYSIS WAS PLANNED UNTIL INFORMATION WAS REQUESTED DURING AN RIII INSPECTION.
- 10. THE BW ANALYSIS OF TE DAVIS-BESSIE TRANSIENI SHJ4 THAT TE PRESSURIZER DID NOT VOID DURING TE EVENT. TE ANALYSIS IDENTIFIED ONE SPECIALIZED CASE WHERE
' THERE WOULD BE VOIDING OF TE PRESSURIZER IF SECONDARY STEAM PRESSURE FALLS
' BELOW ANTICIPATED VALUES,
- 11. NO ATIEMPT WAS MADE TO JUDGE TE TECHNICAL. AEQUACY OF TE ANALYSIS ESCRIBED BY BW, QUESTIONS BEGARDING THIS ANALYSIS 3HOULD BE DIRECIED TO NRR.
- 12. NRR HAS REVIEWED THIS ISSUE AND IETERMINED THAT THERE IS NO SAFETY QUESTION.
" DISCUSSION WITH NRR ON MARCH 13, 1979, INDICATED THAT NRR HAS NO OPEN ITEMS
, ON THIS ISSUE.
13l TE INVESTIGAT018i JUDGEMENT IS THAT TECO PERFORMED A PROPER REVIEW NO ITEMS OF NONCOMPLIANCE WERE INDICATED.
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-DISCUSSION-h
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3 UNERVOLTAGE RELAY SETPOIh7 (UVRS)
' CHRON01DGY:
7/27/76 MILISTONE BULLETIN 10/1/76 UNERVOLTAGE REIAYS INSTALLED IN PIANT (N(7I ORIGIONAL ESIGN), SET TO 10 SECOND DELAY ON 907. UNDERVOLTAGE.
12/8/76 SER ISSUED UNDERVOLTAGE STILL UNDER REVIEW 7
g 4/22/77 PLARI LICENSED. FOUR MONTH ALIDWANCE FOR FINAL REVIEW OF UNERVOLTAGE
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IS LICENSE. CONDITION. ADMINISTRATIVE PROCEDURES IN EFFECT FOR VOLTAGE.
10 SECOND DELAY CALIBRATED 4/11/77.
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t 7/18/77 TECO SUBMITIAL TO NRR, AIELYSIS OF UNDERVOLTAGE. EIAY TO BE SET TO j
f 9 SECOND DELAY TO COPE WITH CASE WHERE THERE IS A SA AND VOLTAGE IS 1
IISS T!!AN 90%, MORE THAN 597. FOR MORE THAN NINE SECONDS. THIS WAS 'IO ELIMINATE THE POSSIBILITY OF REDUCED MARGINS FOR ECCS.
16/77 FCR 217 IMPLEMENTED REIAY SETPOINT AT NINE (9) SECONDS + or - 1.5.
(ALLOWABLE RANGE OF 7.5 to 10.5 SECONDS),
10/5/77 FCR 217 IMPLEMENTED. REIAYS SET TO 9 SECONDS.
10/27/77 SUBMI' ITAL OF AMMENINENT f 7. SETPOINIS CHANGED TO 7 + or-1.5 SECONDS.
THIS CHANGE WAS TO ASSURE THAT THE PELAYS WOULD ACTUATE WITHIN NINE SECONDS, INCLUDING A MAXIMUS.5 SECOND DRIFI ( IN ACCORDANCE WITH REG GUIDE 1.105).
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10/28/77 FCR 430 WRITTEN TO SET RELAYS AT 7 + or - 1.5 SECONDS, INSTALL PUSHBUTTON
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TO DEFEAT REIAY IF OPERATION PROBLEMS RESULT FRCH RELAY SETPOINT CHANGE.
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10/29/77 AMMENDMENI' f 7 APPROVED. MODIEICATION: UNDERVOLTAGE RELAYS REVISION: CHANGE RELAY SETPOINT
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TECH. SPEC. CHANGE: 7 + or - 1.5 SECONDS 4/78 (approx) PLANI GOES TO COLD SIPJTDOWN TO DEAL WITH FUEL PROBIIM.
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6/12/78 SRB REVIEW OF FAILED SFAS TEST FINDS THAT RE1A Y SETPOINT CHANGE HAS NOT BEEN ACCOMPLISHED.
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6/15/78 FCR 430 IS IMPIIMENIED, REIAYS SET TO 7 SECONDS.
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6 /23/78 IZR 061 EPORTS DISCOVERY OF IACK OF RELAY SETPOINT CHANGE.
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j J TE[ LICENSEE DID NOI._IMPLEMENI AMMENIHENT i 7 FOR A PERIOD OF SEVEN NONTHS.
i THE RESULT OF THIS FAILURE WAS AS FOLwWS:
- 1. REGUIATORY GUIDE 1.105, CONCERNING SETPOINT DRIFI WAS NOT FOLwWED, WHICH WOULD ADD AN ADDITIONAL.5 SECOND MARGIN TO RESPONSE TIMES.
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- 2. TE NINE SECOND TIME IELAY WAS IN EFFECT, BUT THE.5 SECOND DRIET WAS NOT INCLUDED IN TIE SETTING, AND COULD THEORETICALLY HAVE DRIFTED THE SETPOINI TO 9.5 SECONDS, THEREBY DELAYING ECCS BY.5 SECONDS UNDER CONDITIONS OF SA AND LOW VOLTAGE UNDER 907., MORE THAN 59% FOR OVER 9 SECONDS. HOWEVER, THE LICENSEE REPORTED THAT THE AVERAGE RELAY SETPOINI WAS FOUND TO BE 8.99 SECONDS (IER-78-061).
- 3. THE TECHNICAL SPECIFICATIONS ALLOW + or - 1.5 SECOND RANGE FOR RELAY SETPOINT.
' BY NOT IMPLEMENTING AMMENDMENI f 7, UTILITY PERSONNEL COULD THEORETICALLY #ME SELECTED A REIAY SETPOINI ABOVE NINE SECONDS. AS A WORST CASE, THIS WOULD HAVE PLACED THE RELAY SETPOINT 2 SECONDS FROM THAT REQUIRED. THIS 2 SECOND CASE INCLUDES A THEORETICAL.5 SECOND DRIET IN A NONCONSERVATIVE DIRECTION. HOWEVER, THIS WAS NOT THE CASE, AS SETPOINI DOCUMENIATION INDICATED THAT THE RELAYS WERE SET FOR NINE SECONDS.
IN ORDER FOR THE LICENSEE TO FAIL TO MEET THE 30-SECOND INJECTION INIO THE VESSEL ASSUMING THE FSAR MSS OF OFFSITE POWER TRANSIENI, TE FOLwWING THINGS WOULD HAVE TO '.4 CUR SIMULTANEOUSLY:
- 1. LOCA
- 2. DEGRADED GRID V0".TAGE GREATER 'IHAN 597. BUT LESS THAN OR EQUAL TO 907. ON THE INCOMING 4.16 K%USES FOR A TIME PERIOD OF GRRATER THAN 9 SECONDS.
- 3. THE 907. RELAY SETPOINI WHICH WAS VERIFIED TO HAVE BEEN SET TO 9 SECONDS WOULD HAVE TO DRIPI NONCONSERVATIVELY TO 9.5 SECONDS.
THE NET RESULT OF THIS EVENT WOULD BE THAT ECCS WOULD BE IELAYED A MAXIMUM OF
.5 SECONDS.
THE INVESTIGATORS NOTED THAT FOR ITEM 3, THERE IS AN EQUAL PROBABILITY THAT THE RELAYS WOULD DRIPI IN A CONSERVATIVE DIRECTION.
IT IS OUR OPINION THAT THE PROBABILITY OF ITEMS 1, 2, and 3 OCURRING SIMULTANEOUSLY IS OF A SUFFICIENTLY SMALL MAGNITUDE NOT TO REPRESENT A SAFETY CONCERN ( ALL TIMES ARE FOR ANALYZED EVENTS WHOSE ACCURACY WE CANNGI VERIFY).
AS T JUNE 15, 1978, THE SETPOINT ASSOCIATED WITH THE 90hMIERVOLTAGE RELAY IS SET AT 7.0 + or - 1.5 SECONDS IN COMPLIANCE WITH TE TECH. SPECS.. (gf/J[
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?TECO MANAGEMElfr MFICIENCIES
~.' C(39tUNICATION PROB 2M. FAILURE TO PROPERLY CbMMUNICATE WITH LICENSING WHEN A TECHNICAL SPECIFICATION CHANGE WILL BE ACCOMPLISHED, ALSO, FAILUPI TO C0!frACT LICENSING TO ADVISE THEM THAT A TECHNICAL SPECIFICATION
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CHANGE WAS ISSUED WHICH THE FACILITY WAS NOT READY TO CCDIPLY '.*ITH ( RELAY SETPOINT).
2.i FAILURE TO PROPERLY REVIEW FCR'S.TO DETERMINE IF A CHANGE TO THE STATION TECHNICAL SPECIFICATIONS IS NEE E D DUE TO A FCR.
- 3. ENGINEERING REQUIRED TO IMPEMENT FCR'S RELATED TO TECHNICAL SPECIFICATION CHANGES NOT PERFORMED PRIOR TO THE APPROVAL OF THE TECHNICAL SPECIFICATION CHANGE. THIS RESULTED IN A DELAY IN IMP 12MENTING TECHNICAL SPECIFICATION RELATED CHANGES.
- 4. IACK OF TRACKING OF FCR'S TO INSURE PROPER AND TIMELY IMPLEMENTATION, IAP.GE BACKLOG OF FCR'S,
- 5. INACCURATE CCMMENTS INCLUDED ON FCR FORMS (SUCH AS REASON FOR UNDERVOLTAGE SETPOINT ).
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__ ENFORCEMENT
- 1. TECHNICAL SPECIFICATION VIOLATION IN THAT TE PLANT OPERATED FOR SEVERAL MONIHS WITHOUT IMPLEMENTING THE 7 + or - 1.5 SECOND SETPOINI REQUIRED BY TECH. SPEC.(AMMENIIIENI #7). THIS WAS REPORTED BY THE LICENSEE ON JUNE 23,
. 1978, AND IS RELATED TO FCR IEFICIENCIES WHICH HAVE BEEN DISCUSSED WITH
, TECO MANAGEMENI AND COVERED IN DIER INSPECTIONS.
- 2. NO ITEMS OF NONCOMPLIANCE WERE OBSERVED IN RELATION TO IDSS OF PRESSURIZER
^. LEVEL INDICATION.
.3. AS NOTED, FCR IEFICIENCIES ARE COVERED IN 0THER REPORTS.
4.THE "LIFI WIRE PROCEDURE" WRITTEN TO COPE WITH DIFFICULTIES EXPERIENCED FOLLOWING ADJUSTMENI OF THE UNDERVOLTAGE RELAYS IS NUI TREATED IN THE INVESTIGATION REPORT, BUT APPEARS TO BE CIEAR.-CUT NONCCMPLIANCE.
- 5. DEFICIENCIES NOTED UNDER"TECO MANAGEMENI DEFICIENCIES" APPEAR TO BE MAJOR FINDINGS OF THE INVESTIGATION AND WERE COVERED IN THE EXIT INTERVIEW WITH TECO PERSONNEL. TESE POINTS SHOULD BE FURTER HIGHLIGHIED IN THE REPORT TRANSMITTAL LETTER.
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ARKANSAS PbWER & LIGHT COMPANY
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FROM:
William Cavanaugh Y'
SUBJECT:
Arkansas Nuclear One-Unit 1 Pressurizer Level Setpoint
' (Fil e : 3740)
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Reference:
1. JWA-848 6
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NDC-2360 l
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Letter, Govers to Cavanaugh 3/3/75 h
y-Attached is reference 3 from B5W which provides their answers to
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PSC comments on loss of level indication in the pressurizer.
following a reactor trip.
From that letter, it can be seen that
.as long as water remains in the pressurize 2* the core will remain
~ covered.and the HPSI setpoint will not be reached.
If the pressuri:er
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pressure drop mentioned in their letter.
3 If you have further questions, please contact us.
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- cc: Mr. D. A. Rueter 7
Mr. M. L. Pendergrass
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April 3, 1975 Telephone: (804) 3G5111 j
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.Mr. W. Cavanaugh, Ill Manager, Nuclear Services
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Arkansas Power & Light Company o g. t(,,1 \\j f P.O. Box 551 Little Rock, Arkansas 72203 h
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Subject:
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Pressurizer Leve1 Setpoint I
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Jteference:
NDC 2360, 3/3/75
Dear Mr. Cavanaugh:
4 NDC 2360 expressed concern over the momentary loss of pressurizer level indication
,A following a reactor trip and requested additional Information to clarify that maintaining RC pressure above 1500 psig (HPSI automatic actuation setpoint) would Y
ensure that the reactor core remains covered with water.
4 This protection can be demonstrated by using a very simple principle:
reactor coolant system pressure is determined by the saturation pressure for the hottest
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water in the reactor coolant system.
In all operating situations except extreme cccident conditions, this water is, of course, pressurizer water at about 650 F, 0
'I corresponding to a saturation pressure of 2155 psig while the average water h
temperature in the reactor core of 579 F has a saturation pri.ssure of about 1300 psig.
Within about 20-30 seconds af ter a reactor trip, all water in the reactor k.:
coolant system (except pressurizer water) will be below 5790F as the reactor Q
power-sustained differential temperature across the core collapses and as the I 'i reactor coolant system is cooled to about 550 F (due to turbine bypass valves being set to control OTSG pressure at 10'10 psig)
Even though the pressurizer water out-surge during system cooldown will allow system pressure to fall below 2155 psig,, data from reactor trips at B&W's operating plants.shows that RC pressure g.
remains well above 1500 psig. With the RC cooldown established by means of the turbine bypass valves' pressure setpoint, RC pressure will not drop to 1500 psig wn'le'ss the pressurizer is completely drained. ilf the pressurizer were to drain
'somplitely, RC pressure.would drop rapidly to-the' saturation pressu're for the
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,[ hottest watei' remaining in the RC' system.' The te'nperature of this water would be between 5500F and 579 F with a resulting RC pressure of 1010_psig to 1300 psig.
This resulting RC pressure band if the pressurizer were to apty following a reactor trip is we:11 bel.ow thezl500 psig HPSI automatic initiation setpoint.
Thus 1500 psig is an adequate low pressure setpoint for ensuring that the reae. tor core remains covered wi th water.
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you have any further questions in this matter, please advise.
.Very truly yours, l
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J. D. Phinney', Manager l'
Operating Plant Services & Maint.
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J. V. Anderson l+
J. A. Ba i ley pt R. P. Locke tt, J r.
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THE BABC0CK & WILCOX COMPAllY POWER GEllERATI0li GROUP T3 l
R.P. WILLIAMSON - NUCLEAR SERVICE From C.W. TALLY - CONTROL ANALYSIS (EXT. 2G03) m us.s Cust.
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Date SPR 396 FEBRUARY 10, 1978
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Reference:
1.
14tter Bh*T-1609, J.A. Lauer to C.R. Domeck, T1.2/12B, dated December 5, 1977.
l Engineering has evaluated the transient described in SPR 396 resulting in the 1
following comments-1.
ne classification of the transient in Reference 1 was correct and no further com:nent on this aspect is required.
2.
He decrease in pressurizer level (off-scale low) is indicative of rapid steam generator level. increases following the initiation of AFW. This undesirable effect is symptomatic of high level setpoints. 3 Conversations with Fred Miller _ of TECO Engineering "have7c'onfirmed 'IECO's awarenessof this problem and their desire to have-it rectified.
In view of the fact that Davis-Besse I has elevated loops, there should be little difficulty in decreasing the level setpoint with appropriate analysis. The funding for this work will be pursued through Project Management.
3.
Engineering has been unable to satisfactorily resolve the dissimilar behavior of the two OTSG's during the transient. During the 5 to 15 minute period of the transient, the two steam pressures moved in opposite directions and were considerably apart. ne plant computer printout says a main steam line' warm up isolation valve was open during this time ("22:55:56 2688 FN STN Line 2 WU ISO VLV CIDS"), but TECO Engineering says the valve indicator is wired backwards, indicating that it actually was closed until 22:55:56, when an operator openedit.If indeed it was closed until
- this time, there appears to be no logical explanation for the steam pressure differences. His should be passed on to 'IECO Engineering, since Plant Design has no further information with which to I
investigate this anomaly.
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TOLEDO EDISON COMPANY.
DAVIS-BESSE UNIT ONE NUCLEAR POWER STATION i
SUPPLEMENTAL INFORMATION FOR LER NP-32-78-07 i
DATE OF EVENT: June 12, 1978 FACILITY: Davis-Basse Unit 1 IDENTIFICATION OF OCCURRENCE:
Incorrect setpoints on essential bus undervoltage
- relays, i
i Conditions Prior to Occurrence: The unit was in Mode 6 with Power (MWT) = 0, and Load (HWE) = 0 i
Description of Occurrence: On June 12, 1978, during the St stion Review Board review q
cf the " Safety Features Actuation System (SFAS) 18 Month Test", ST 5031.07, it was found that the time delay setpoints of the essential bus undervoltage relays were incorrect and that the monthly channel functional test was not being performed.
The initial investigation showed the Facility Change Request (FCR)77-217 which tas implemented on October 4,1977, called for the time delay to be set at 9 =
ceconds. FCR 77-430 was prepared on October 28, 1977, to correct the setpoints'to 711.5 seconds, but had not yet been issued for implementation on June 12, 1978.
This occurrence is being reported in accordance with the provisions of Technical Specification 6.9.1.8f.
Designation of Apparent Cause of Occurrence: The cause of this occurrence is d
procedure inadequacy.
Analysis of Occurrence:
There was no danger to the health and safety of the public cr to unit personnel. The intent of the 7 1 1.5 second time delay setpoint is to ensure,that a bus trip will occur in 9 seconds af ter the bus voltage degrades te less than 90% of the normal voltage. The average time delay setting of the relays cas found to be 8.99 seconds, i'
Corrective Action: FCR 77-430 was immediately implemented and at.that time it was
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also found that the voltage setp'oints were incorrectly set to a maximum of 2.5%
less than the tuhnical specification minimum. One relay was found to be defective and was replaced. The time delay and voltage setpoints were adjusted, to values in compligoce with Table 3.3-4 of Technical Specification 3.3.2.1. -A modification -.. --
(T-2870) was prepared for a test to be performed in conjunction with 'ST 5031.07 to satisfy the monthly functional check.
new ' surveillance test procedure will
'ba written to assure the monthly functi test is coupleted when the unit is in
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the applicable modes. This work was cospleted on June 15, 1978 under Maintenance f-Work' Order 78-1397.
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. ' y-Psilure Deta: This is not a repetitive occurrence.
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