ML20044C681
| ML20044C681 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 09/17/1991 |
| From: | Beall J NRC |
| To: | Rosenthal J NRC |
| Shared Package | |
| ML17056C371 | List:
|
| References | |
| CON-IIT07-772-91, CON-IIT07-773-91, CON-IIT07-776-91, CON-IIT7-772-91, CON-IIT7-773-91, CON-IIT7-776-91 NUREG-1455, NUDOCS 9305070179 | |
| Download: ML20044C681 (15) | |
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,1 N'7' UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF THE RESIDENT INSPECTOR NINE MILE POINT FAXSIMILE TRANSMITTAL FORM 9 /? 'f/
DATE To:
AC_W l O_.S&A/THA L FROM:_.
Tim bra LL PAGE 1 0F Y
(INCLUDING THIS FORM)
CONFIRMATION NUMBER (315) 342-4041
SUBJECT:
bf5E Af,E THE DQF1 [ftT $o7s 5 ks & op,_tytTH Tfff fMDwkTGR Uw6 AMD TMS
$CC C//fCY VANE [M)/gQ]gg COMMENTS:
Yott SEQutsTfb
'TM$6 W N CA/ TM Y VfAG AUAnARtf.
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9305'70179 911031 0
PDR ADOCK 05000410; S.
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iteview the ro e c ruo w a m.i carrect) m - ' om e co on <
j AMdl* PEN G fit ' ; 'wt n t prnh1WM.,i t. itf f i ( j i 1 ed i r'1 On 1..
/tiii
!l in the following componente/ systems:
'l Feed water During August 13 event, control room operators at*empted uncuccensfully to open one valve (64B) in the non-eafety feeawater system.
Normal operating proceaure was to equalize preeeure acrocs the valve prior to attempting to open.
The operatore elected not to open the bypass valves (local).due to alarming ARM's and other access problems (associated with the SAE).
The operators were cognizant of the capabilities of tb componente and determined that opening under the differential pressure was within th-design of tne hardware.
elMPC's root cause analysis indicated the
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potential causes to be loose bolte on actuator to valve area, unbalanced torqua r
l switch, or the torque switch bypass set too low.
The bolts were tightened and torqued.
Bo11ts on those non -satety valves had no vendor prescribed torque values.
Ot.her valvea ware inspected for similar problems and none were j
found.
Safety-related valves have torqu-values clearly erecified.
The torque switch was balanced ani do not require any adjustment.
i Terque switch bypass was increased from 6% to 12%.
Safety-related valves are-bypanned for 1
above 90%.
So no similar problems exint-on safety-related valves.
The non-cafety vendor manual was updated to specity torque values for actuator bolts.
Thin was not a problem with other manuals revieued.
Also as part of the corrective action plan, l
the licensee stated that either the valves would bd demonstrated capable of opening under the event-caused differential pressure, or that the system procedures would be I
revised to eneure that the valves would not be called upon to open (a3 ready be open).
-]
The licensen found duplicating the differential pressure conditions experienced l
during the event not to be feasible and stated that the applicable procedures would be revised.
The inspector found no evidence of a generie maintenance problem and no implications of other systems problems.
The previous post-maintenance (modification) testing was reviewed by the inspector and found acceptable.
The inspector found that the operators called upon the valve to open under conditions within ite design capabilities, but not covered by procedure and hence never tested.
l The inspector had no concerns with respect to l
the feedwater valves or the actions of the operators.
RCIC (Check Valve Indicat$on Problems)
Position indication problems were experienced during the August 13 event.
Similar l
indication problems had been experienced about twice a year-previously.
The inspector noted that several facilities had a repetitive failure history with similar components.
The inspector reviewed the work packages for l
the valves including a post-modification l
hydrostatic teet procedure and identified no deficiencies.
The hydrostatic test was completed satisfactorily;with no leakage observed at 1045 psig, l
IST Surveillance Procedure N2-OSP-ICS-CS001,
" Cold Ghutdown IST Surveillance," was modified such that the leg between AOV 156 and 157 is filled with water prior to valve stroking.
This is intended to prevent water elugging which had been damaging the limit switch and causing the indication problems.
l The licensee also plane to revise OP-35,
'RCIC Operating Procedure,' to ensure water is in the leg.between ADV 156 and 157 prior to system operation.
Both valves were tested successfully on September 8, 1991, with no indication problems experienced, i
(
e*
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r l-j In summary, the inspector found the problema j
to have been valve position indication only and did not impact system performance.
Licensee efforts to address the problems were found to be acceptable.
Modifications completed during the current outage may have solved the indication problems.
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50610 Federal Register / Vol. 54. No. 235 / Friday. December 8,1989 / Rules and Regulations Act, to make these regulatory provisions North.11555 Rockville Pike Rockville, it has been concluded that the violation effective as specified, and handlers have Maryland 20S52. between 7:30 a.m. and involves a significant regulatory-.. -
been apprised of such provisions and 4:15 p.m weekdays. Copies of concern.The Commission acknowledges the effective time.
- t-comments received may be examined at that inclusion of the root cause of a the NRC Public Document Room. 2120 L violation as an escalation factor when List of Subjects in 7 CFR Part 910 Street NW Lower Level. Washington.-
considering a civil penalty is a change Arizona. California. Lemons.
DC.
from past practice. Further, the Marketing agreements and orders, ron runmta enronuAnoN CONTACT:
Commission recognites that For the reasons set forth in the -
James Lieberman. Office of consideration of only one root causee preamble,7 CFR part 910 is amended as Enforcement. U.S. Nuclear Regulatory (maintenance) as a specific escalating 1 i
i follows:
Commission. Washington, DC 20555.
factor focuses on only a fraction of the Telephone (301) 492-0741.
possible casual factors that may be DART 910-. LEMONS GROWN IN SUPPtiMENTARY INFoRMATioec On involved in a particular violation..
CALIFORNIA AND ARIZONA March 23.1988, the Commission issued a By this change.- the Commission is not 1.We authority citation for 7 CFR Policy Statement on Maintenance of establishing a new group of civil penalty part 910 continues to read as follows:
Nuclear Power Plants (53 FR 9430) actions. Consistent with current which stated the Commission's practice, a violation will be considered Authority: Secs.1-19. 48 Stat. 31. as expectations in the area of maintenance for escalated action (Severity Level L IL 4
cmededd M. 801-674.
and its intention to proceed with a or III violations) based on the violation.
l 2.Section 910.995 is added to read as rulemaking on maintenance.
lacluding its impact circumstances, and follows:
Subsequently, on November 28.1988. the root causes. Special escalation will only j
Note: This section will not appear in the Commission published a Notice of -
apply if the violation or problem area Code siFederalRegulations.-
Proposed Rulemaking(53 FR 47822)-
(aggregated violations) has a directed toward improving the maintenance root cause.
l 1-j l 99 0.995 Lamon Regulation 695.
effectiveness'of maintenance programs.
De Commission concludes that f
, l The quantity oflemons grown in
%e Commission recognizes that the modifying the Enforcement Policy to California and Arizona which may be industry and individuallicensees have permit increased civil penalties for i
l handled during the period from made improvements in their Severity Level L IL orIII violations t-i 'l December 20.1989. through December maintenance programs.'Indeed, the which occur 90 days orlater after the -
16.1989,is established at 300.000 Commist.lon has seen noticeable date of this notice and which result from cartons.
progress by the industry over the past.
maintenance deficiencies may provide a Dated: December 6.1989.
four years in the area of nuclear power furtherincentive to ensure that all Charles R. Brader, plant maintenance.The Commission licensees place appropdate attention on also recognizes that the industry is ~
maintenance of equipment whose failure DimetorFruit and Vegetable Division.
committed to continue to improvee could significantly impact safety. Use of
[FR Doc. 89-28882 Filed 13-7-89; 845 am]
maintenance.Nevertheless. NRC the Commission's enforcement program l
maintenance teaminspections have in this manner to emphasize the confirmed that furtherimprovements are importance of meeting existing necessary, especially with regard to requirements related to maintenance is NUCLEAR REGULATORY effective implementation of warranted because of the varymg COMMISSION maintenance programs. In view of the quality oflicensee maintenance 10 CFR Part 2 progress made to date, as well as the programs includingimplementation, industry's expressed commitment to and the decision to hold in abeyance the Policy and Procedures for continue to improve maintenance. the rulemaking on maintenance. By this -
Enforcement Actions; Policy Commission has decided to hold revision to the Enforcement Policy. the Statement rulemaking in abeyance for a period of Commission is putting IIcensees on 18 months from the effective date of the notice that the decision to tiefer a AGENCY: Nuclear Regulatory Revised Policy Statement on maintenance rule does not mean the i
Conumssion.
Maintenance of Nuclear Power Plants Commission does not expect a senous Acnosc Policy statement: modification.
which was published elsewhere in this licensee effort in the maintenance area.
issue.ne Commission will assess the It is expected that the revision to the I
SUMMARY
- 'Ite NRC is publishing a need for rulemaking at the conclusion of Enforcement Pohey wdl remain effective modification to its Enforcement Policy to tMs la month period, based upon atleast until the Commission 1
add an additional civil penalty industry initiatives and progress in reconsiders the need for rulemaking in adjustment factor for violations improving maintenance.
the maintenance area.
involving maintenance deficiencies.nis The Commission bebeves that a Since this action concems a general policy is codified as Appendix C to 10 str ng maintenance program can make a state nent of policy no prior notice is CFR part 2.
significant contribution to safety. In the required and, hence, this modification to ErrEcTivs DATc December 8,1989.
Revised Policy Statement on the the Enforcement Policy is effective upon However.It will only be applied for Maintenance of Nuclear Power Plants, issuance. However, the modification for l
violations which occur after March 8.
the Commission stated its intention to maintenance will only be applied for 1990. Comments submitted on or before emphasize maintenance in enforcing violations which occur 90 days or later February 6.1990. will be considered.
existing requirements for power after the date of publication.
ADDRESSES: Send Comments to:
reactors. Consistent with that position.
List of Subjects in lo CFR Part 2 Secretary U.S. Nuclear Regulatory the Enforcement Policy is being reviced Commission. Washington. DC 20555.
to provide such emphasis by addmg Administrative practice and l
ATTN: Docketing and Service Branch.
maintenance failures as an escalating procedure. Antitrust. Byproduct Deliver comments to One White Flint factor in assessing civil penalties where material. Classified information. Civii l
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Federn! Register / Vol. 54. No. 235 / Friday December 8,1989 / Rules and Regulations l
50611 s
p:nsity. Enforcement. Environmental maintenance on March 23,1988 (53 FR and endorses industry maintenance
{
protection. Nuc! car materials. Nuclear 9430), and a proposed rule on November initi-tives; however, recent NRC J
power plants and reactors. Penalty. Sex 28.1988 (53 FR 47822). The Commission inspecticas oflicensee maintenance discrimination. Scurce material. Special recognizes that the industry and programs and their implementation. and nuclear material. Violations and Waste individuallicensees have made evaluations of plant operational data J
tre:tment and disposal improvements in their maintenance indicate that many licensee programs. Indeed, the Commission has. maintenance programs need further PART 2-RULES OF PRACTICE FOR seen noticeable progress by the industry improvement. For example, there l
DOMESTIC UCENSING PROCEEDINGS over the past four years in the area of remains a wide variation across the nuclear power plant maintenance.ne industry in the effectiveness of the 1.ne authority citation for part 2 Commission also recognizes that the implementation ofmaintenance continues to readin part as follows:
industryis committed to continue to programs. Areas of weakness include s
Au% Sec.161. ca Stat. 94a. as improve maintenance. Nevertheless, engineering support, root cause analysis.
NRC maintenance teaminspections trending, and recordkeepin6-n 1242 a e
C have confirmed that further,
ne Commission believes that good 4
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- 2. Appendix C. section V,B is improvements are necessary, especially ' maintenance is a key factor in achieving cmended by adding s'ection V.BJ with regard to effective implementation - and maintmining a high level of safety in directly after parsgraph 3 of section of maintenance programs. In view of the plant operations throughout thelife of a ll l
V.B.6 to read as follows:
progress made to date, as well as the nuclear power plant by helping to -
i l
industry's expressed commitment to ensure that equipment will perform its Appendix C-GeneralStatement of continue to improve maintenance, the intended function when required.In Policy and Procedure for NRC Commission has decided to hold addition. a well-documented and Il Enforcement Actions rulemaking in abeyance for an 18 month executed maintenance program is period to monitor industry initiatives expected to be significant in plant life i.
i V. Enforcement Actions.* *
- and progress and, at the end of this 18 extension decisionu Because month period, to assess the need for maintenance plays such an important rulemaking in this area.Ris revised and integral role with plant operations B. CidPenchy * * *
- 7. Maintenana-Related Ceuse policy statement is being issued to-in assuring public safety, the The base civil penalty may be increased e describe the Commiazion's expectations Commission is convinced : hat continued industry attention and improvement in
- much as 60% for cases where a cause of a during this 18 month period, as well as -
- the maintenance area is needed not only
'p%
the Commission's planned actions,.,,
during and at the conclusionof this.
to improve maintenanea at some nuclear
- ega, purpoons of application of this factor, a canse
' of the violation sha!!be e Wed to be period.This policy statement contains a ' power plants today, but to ensure voluntary solicitation of reporting and performance of effective maintenance at maintenanrwelated if the violation could -
' record keepmg that is subject to the * ;[. all nuclear power pknts in have been prevented by haplementing a.
. Paperwork Reduction Act of1980 (44 ~.
herefore, the Commission has decided ~
mainten== program consistent with the U.S.C. 3501 et seq.),it will be submitted' [ to hold rulemaking in a scope and activities defined by the Revised to the Office of Management and Budget j
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- h Nta.In M
1
- forreviewandapprovalof the.
. date of this revised policy statement to.
Pow
'. factor. consideration will be given to, among. Information collections.
' permit the Commission to monitor
.cther things, whether a failure to perform.
miintenance or improperfy performed EFFECTIVE DATE:This revised policy:
. industry initiatives and progress in maintenance was avemstic failure.
statement is effective December 8.1989.
improving maintenance and to evaluata the need for additional rulemaking.
The degree of the programmatic failure will Fon FURTHER IPtFORRAADO98 COffTACT.
ne Commission is issuing this c ns ered in applying this factor.
Moni Day. OfBee of Nuclear Regulatory revised policy statement to describe the Research. U.S. Nuclear Regulatory Commission's expectations and future i
e.MaSand, tMs 5th day Commission. Washington. DC 20555,
[
"h actions plannedin the maintenance telephone:(301) 492-3730.
area, and to restate the Commission's For the Nuclear Regulatory Commission.
Background
views with respect to what constitutes Samuel J. Chilk.
SecretaryoftheCommission. -
On March 23.1988 (53 FR 9430), the an effective maintenance program.
[FR Doe, e9-28742 Filed 12-7-8m 8:45 am]
Commission published a Policy Revised PoIIcy Statement Statement on Maintenance of Nuclear power Plants which stated the ne Commission desires to have in owwa cooe neSes.=
Commission's expectations in the area place an industry-wide program that to CFR Part 50 of maintenance and the intention to -
will ensure effective maintenance is achieved and maintained over the life of proceed with a rulemaking on Maintenance of Nuclear Nwer Plants;.. maintenance. Subsequently on1 each plant.ne Commission expects Revised Policy Statement November 28,1988, the Commission +
. each licensee to assume responsibility for assuring that an effective -
Accxcy: Nuclear Regulatory published a notice of proposed rulemaking (53 FR 47822) directed maintenance program is or has been Commission, toward improving the effectiveness of
. developed. Implemented and maintained 4
maintenance programs. -
. at his facility.ne Commission Acnope Revised policy statement.
sunenAARY:The Commission believes NRC's rulemaking initiative served to recognizes that the Nuclear safety can be enhanced by improving
- increase industry attention on this -
Management and Resources Council nuclear power plant maintenance across important aspect of nuclear power plant (NUMARC) rnd the Institute for Nuclear the nuclear industry. Consistent with safety.ne Comrnission acknowkdges power Operations (INpO) can this belief, the Commission previously industry's effort and progress directed contribute. through their leadership, to published a final policy statement on t ward bpmvements in maintenance an industry-wide program for improving
Federal Register / Vol. 54. No. 235 / Friday. December 8,1989 / Rules and Regulations 50612 and maintaining effective maintenance plants. In this regard, the Commission solicited industry participation in a joint and encourages such leadership.
has issued for comment a standard for NRC/ licensee project with the objective During the next 18 months, the maintenance in the form of a draft of sharing and comparing development Commission intends to closely monitor regulatory guide and announced its work on maintenance effectiveness individual licensees and the industry a:
availability fn the Federal Register (54 Indicators.
j a whole and assess the need for FR 33988: August 17.1989).The Finally, the Commission reemphasizes r
additional rulemaking in the area of Commission also intends to hold a its previous views with respect to maintenance.%is monitoring will workshop early in 1990 to promote elements of an effective maintenance further dialogue on the standard.%e program. Specifically, the Commission -
include completion of the ongoing industry and the public are encouraged expects the scope of each licensee's Maintenance TeamInspections to assist in the refinement of this maintenance program to include all (including some selected reinspections) standard or propose a suitable systems, structures and components end review of other inspection results, attemative standard for NRC addressed by existing regulations and end performance indicators: and endorsement (to be considered, any licensee commitments and described in industry's and individual licensee's alternative standard would need to be the documents (e.g., Final Safety
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I performance. commitments, and progress toward improvement. Industry proposed to the Commission by March Analysis Report) required by to CFR groups andindividuallicensees are 1.1990).%e Commission intends to 50.34, whose falhue muld significantly have a standard available for use in impact the safety or security of the encouraged to provide information to appmximately1 year and will facility.%is includes systems, document their commitments and to encourage voh.ntary industry use and structures,and components in the demonstrate their performance and improvement in maintenance. In adoption of this standard. Adoption and balance of plant, since experience has addition, the Commission intends to use of an acceptable standard willbe a shown that failures in many balance of consideration in evaluating industry's plant systems, structures, and continue development of a rule on and individual licensee's commitment to components can and do have an impact maintenance so that at the end of the is month period,if rr.lemaking is achieving and sustaining effective on plant safety or security.
determined to be necessary, the maintenance.
In addition, the Commission defines Commission willbe in a position to' Anintegralpartof aneffective maintenance as the aggregate of those promulgate such a rule.
maintenance program is the monitorin8 actions which prevent the degradation and feedback of results.ne -
or failure of. and which promptly restore In enforcing existing reqmrements Commission believes that such the intended function of, structures.
over this time period, the Commission intends to emphasize maintenana by programs should utilize quantitative systema, and components. As such.
i assessing whether a safety significant information regarding operational, maintenance includes not only the violation (i.e SeverityIevelIII or history, especially component failures activities traditionally associated with higher} oflicense conditions or and system reliability / availability to identifying andmm:ing actual or regulations could have been prevented if monitor and adjust the maintenance potential degraded conditions (i.e an effective maintenance pmgram had program. Performance indicators that repair, surveillance, diagnostic been implemented. Accordingly, the are based upon actualcomponent -
examinations, and preventive measures)
Commission, by separate action is reliability, system reliability / availability but extends to include all supporting modifying its enforament policy to and failure bistory pmvide a useful functions for the conduct of these provide that. for safety-significant indication of maintenance effectiveness. activities. Accordmgly, each commercial violations where a civil penaltyis Such measures are most effective when - nuclear power plant should either have appropriate, the amount of the penalty they are based on a well-structured and in place or develop and implement a for such a violation may be escalated component. oriented system. e.g., the well. defined maintenance program to where a programmatic inadequacyin Nuclear Plant Reliability Data System assure that the above is accomplished.
maintenance is a root cause.
(NpRDS).to capture and track Activities and supportmg functions that Furthennore, plant specific orders or equipment history data.ne should be considered in a maintenance letters requesting information pursuant Commission encourages the use dthe program. as defined in this policy to to CFR 5054[f)may beissued where industry-wide NPRDS data for this statement, are listed below:
purpose. Including improved industry
(# Afdnfenonce Afonogement and poor or dechning maintenance performance raises safety issues. -
use of and participation in the NPRDS to Techn&gy Additional Commission actions and gauge the effectiveness of maintenance.
expectations are discussed below.
Licensee reportmg of such data to the Corrective and preventive
%e Commission believes that the system in a timely and complete manner maintenance programs (ths latter may and licensee use of such data to monitor include reliability. centered and development and use of a component failures and system
. predictive maintenance activities) to comprehensive performance-based-reliability / availability for comparison integrete and focus these activities on standard for maintenance, which w+th overall plant goals or standards, structures, systems, and components provides guidance and requirements on ' represents one acceptable element of. whose failure could significantly impact the scope. goals, performance and activities associated with an effective maintenance monitoring,
. safety and to pdoritize preventive...
maintenance program, is important in
%e Commission intends to develop..
maintenance tasks. Maintenance
' assuring that maintenance is improved.-
validate, and usemaintenance~ -
management and technology should,
where necessary, and remains effective effectivenessindicators.He include consideration of:
over the life of each plant.nerefore.
Commission also encourages (i) planning.
during the next 18 months, the development and use of.such indicators (i) Scheduling-Commission intends to continue to by licensees and the industry such that EI bt"I*3-develop, on a cooprative basis with the the progress of improvementin (iv) Shift Coverage..
maintenance can be closely monitored.
industry and public, a maintenance +
' standard for commercial nuclear power To that effect. the Commission has (v) Resource Allocation...
)
1 1
I Federal Register / Vol. 54. No.-235 / Friday. December 8.1989 / Rules and Regulations 50GU (vi) Controlof Contracted DEPARTMENT OF THE TREASURY savings associations, one by a savings i
association and its nondiversified
?
Maintenance Sarvices.
Office of Thrift Supervision savings and loan holding company, two j
(vii) Availability of Parts. Tools, and Facilities.
12 CFR Part 545 cMporation of a san.gs asmaadon (viii) Measures of Maintenance involved in out of-state mortgage i
[No.89-358]
Pmgrazn EHeckeness.
banking activities, and two by law Erms (ix)Internalcommunications between RIN tS50-AAos on behalf of clients.
l the maintenance organization and plant All of the commenters supported the j,
operations and support groups, as well Agency Offices proposal and strongly favored the i'
as communications between plant and Date: October 31.1989, adoption of a finalrule.None of the i
AGENCv: Office of Thrift Superivision.
commenters expressed any concern l
and the plant c]o einan Tressury.
about the proposal to expand the (x) Extemal communications between ACTION: Final rule.
geograpical area in which lending I,
the plant maintenance organization and offices can be established.Several individual vendors to consider their suuuARY:%e Office of Thrift commenters raised questions about the recommendations or requirements.
Supervision ("Omce") is amending its activities and operations of these oSices regulations at 12 CFR part 545 to remove includi tg state authority over such g
the current restriction banning Federal offices, operation of such offices under a Ensure engineering support to savmgs associations from estabWhtne trade name, exportation of interest -
maintenance. inc!=iino root cause agency offices to originate and service rates, and the extent of activities to loans outside the same state as the which they can engage.'Ite proposed analysis and updating the maintenance home office of the savings association or rule was limited to the question of program as a result of plant the same state of any association's location of such offices and not to the modifications.
branch office.This change affords expansion or modiScation of the range (J)Rodiotion Exposure Control Federal savings associations the of activities that such offices could Ensure radiological exposure control flexibility required to effectiveIy and gg gg, in D8 m % mhnam
[
,n basis. Consequently, the Office does not t e pe ti
- s. A on
.the believe it is appropriate to respond to ac M es.
Office is amanding 12 CFR 545.96(d) to s
uin e.
(4) Maintenance Personnel require notification in writing of agency - about the activities of su d
Qualification and Tinuung opdnp and closings m DATEdanuary 8.1990.
opinions may be requested from the Develop and apply snaintenance FOR FURMERINFORMATION CONTACT:
Chief Counsel's office.De commenters personnel quahfications and training Cindy 1. liansch. Mnancial Analyst. '
noted that aBowing agency omces (also
~
requirements.
(202) 90%7488; or Cheryl Mar'tm.
referred to as loan production offices) to (5) Quality Assurunce-Regional Director. (202) 90%78G9: or '
operate outside an association's
. Meen Willard. Deputy Director. (202) branching territory would have Ensure use of quality assurance and 906-6789; or Patrick G.Berkakos, numerous benefits, such as:
quality control to maintenance-related Erectw Corporate Activities.
(1)1ncreasing the efficiency of Federal activities.
Supervision (Operations). Omce of savings associations by eliminating the (6] Documentation Thrift Supervision (Operations). Office need to create a subsidiary mortgage i
p 1
- Stred, organization and thus reducing loan Develop equipment history and origination costs; trending, maintenance record-keepm, g, SMEMENTARY WORMATION:In order (2) senefiting consumers as a result of and maintenance procedures.
to allow Federal savings associations to increased competition among lending (7) Testing andReturn to Service operate more efficiently and to complete institutions and reduced origination i
more effectively in existing markets, the I
Develop and use post-maintenance FederalHome LoanBankBoard.
(3)r!nninating separate testing and return to service piecwuis.
g.. Board"). predecessor to the Office.
recordkeeping. issuance of stock, and in accordance with the above,the proposed to amend 12 CFR 545.90 to I""*
"""I"8****"8'"*"
~
I i
Commission intends to monitor remove the restriction that Federal between parent and service corporation.
individuallicensee and industry savings associations may not establish thereby reducing costs and increasing commitments, peformance, and agency ofHees to originate and service competitiveness, improvement in ma'ntenance over the loans outside the same state as the (4) Permitting Federal savings next la months, and to evaluate the -
home omce of the savings association or associations to oIIer the same lending need for additional ulemaking to ensure the same state of any association's services as nationalbanks; that effective maintenance is achieved branch office.The Board proposed to
- 15) Assisting Federal savings and maintained over the life of each.
remove the restriction thereby allowing associations to originate high quality nuclear power plant. -
Federal savings associations to residentialmortgageloans; estabihh agency offices on a nationwide
" Dated at Rockville.MD. this 5th day of basis without regard to the location of -
- (B) Permitting Federal savings yecember 19e9.
~
the home or branch office (s). Board Res.. associations to open an office in strong For the Nuclear Regulatory Commission.
No. 89-1708,54 FR 30555 []uly 21.1989).
lending markets more quickly: and Samuel 1. ChA The comment period expired on (7) Assisting Federal savings Socnetaryof the Commission.
September 19.1989. The Office received associations to increase profits. in light (FR Doc. eMa741 Filed 12-7-87. 8 45 aml sixteen comments in response to this of changes due to the Financial proposal. Ten were submitted by Institutions Reform. Recovery and l
on.uwa coor nem I
4 4'/322 Fodseal Register / Wol. 43. No.as 7 Island:y, NovImber 28,1968 /.. Proposed Rules BluCLEAR REOULATORY CORIlll8SION 10 CFR Part 60 Ensuring the Effectivenees of teeintenance Programs for Nuclear Power reente Assucy NuclearRegulatory Commi.nion.
AcTsosi. Proposed rule.
suenansry:he mamission is proposing e
to amend its regulations to require connnercial nuclear power plant licensees to strengthen their maintenance activitics in order to reduce the likehbood of failures and events consed by the lack of effective maintenance.no Commission believes safety can and must be enhanced by deBains an adequate maintenance program to e mre the effectiveness of such programs throsphout the nuclear endantry.Do proposed rule requires plant maintemenos programs to include speciBc activities, including the monitortog of the effectiveness of plant maintenance psoprams.
oAft:Comument period expues January 3r,1ses. r-*= sensived after this date will be cameldened ifit is practical to do so. but esserence of amesideration osanot be given except as to an===its received en or before this date.
meenenses: Mail written a===aats to:
Sometary UANuclearRegulatory
, c-w Washington, DC 30655.
. Attention:Dedeting andService Branch.
-~-
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Federal Regineer / Vol. 53, No. 728 / Monday. November 28,1988/ Propueed Rules d7323
/l i
Deuver-====ts to:11156 Rocim!!e activities asy condested to preserve or faitiatives pmmote industry Piha, Recimus.MD between y:30 am restore, with prompt repair, the reopencibihty for problem identification I
and tis pm weekdays..
availability, M
- andrehability and reesistion; of plant structures, system, and
- 2. Prescriptive rulemaking options I
Copies of the paper en d- "#ef components. N program absold clearly may hopedeladestry initiatives and 6=
i options, transcript and r.1 I
the Public 2 M. draft NUREIG denne the ra==*= and activities bility toimprove maintenance; i
report, draft regulat'ory analyis.
included, as well as the management i
environmental========s and finding syswas md to controltboes acevites.
3.Rulemaking should be directed
]
of no significant impact, the empporting Further,the program abould imelade toward spealfytag the NRC's j
stateenest submited to DMB, and feedback of specific results to enmuo expoetations in intenance and j
]
comunents received may be exaedaad at: corrective acnons,,..
foroverall _requinaslicenseemonitoringof the
)
the NRC PublicDocument Room.2120 L program evaluation, and the ' '
effectivenrss of maintenance programs.
I Street, Lower Level, NW., Washington, identification of possible compenset or. ' %ersiose, the ts-n.sa.s p,,p3.,,.
DC.
system doeign problems,e.' -
. meintensass rule which gives incentive with the rule woeld be vetfleilby NRC forindusty to develop a standard for a ronpusman esequus moss costrace.
Mom Dey Olhos of Nuclear Regulatory audit and inspection.
mainteennes program,which NRC may h propond rule deeJ ad rupdm endorse in a Regulatory Calde.
j j
Research, U.S. Nuclear Regulatory 1r=r=*=d=== for $8=8-*=-- Standard i
j WashingtonJC sosas, Indice re 4
I However, each Beensee would be The t'a==i-i== encouragw industry j
oummameram meonmanne required to have his own eyeesa fer to develop a Maintenance Standard monitoring main==== effaciveness which willprovide guidanas for 1
e
1 Background
. which would be subject to NRCreview.. complying with regnarements of the i
On March 211988,&e Camiselon The Commissica solicits comuments on y,_,
f rule.%s Co=m3==3an believes f
j published a Anal policy Statement on se appucadon andmfuloses of MPIs eat h dev& pant of a shnded dl i
3j Mainwnence of Nuclear Power Plants.
as part of the rule, and whether a set of guide currentindustry initiatives in the policy Statement, the Comunission MPIs exists which couldindicate the towards developing andimplementing 1
stamd eat it expeckd to pubEsta a effectiveness of last maintananas acceptable malatenanos pograms and programs. In adtion, the Comunissionthat stiHty participation la preparation -
Notice of Proposed R in i
nearfate and pmvided,'gara solicits feedoack on whether to require of a Maine===a= Standard will provide l
I'*g*
r y -
"I
reporting a speciBc set of Mple to the additionalineentive and responsibility NRC as part of the rule.
forimproving plast maintenance
}
contiminay evaluaw the opmtional '
i performance of nuclear power plants.
Public Weeksidy 7ComisshM W% a Analysis of opmtional events has De Commission held a Public Regulatory Guide to provide guidance y
shown that,in see cases, nnclear Workshop ca.luly 11-13.1tgesin for complying with the rule if industry g
l Power plant equipment is not being Washmston, DC to solicit early input for does not develop an adequate standard maintained at a level to ensure that the the formulation of the rule from the However, the Commission prefers to equipment will perform,with a high public and regulated industry. Prior to endorse enindustry-developed degree of relisbuity,its intended the Workshop, a paper on rulemaking standard.To meet the Commission's function with vegoired. A heited E options was distributed to interested plans forimpleasentation of the rule the examination of nuclear power pisnt parties to facilitate Workshop industry commitment to develop a maintenance programs bas found a wide discussions.De paper on rulemaking Maintemence Standard should be made g
variation la the effec'iveness of these options 8 and the transcript and
- now and a Anal standard should be At some plants, assintenance proceedings (NUREG/C8%ogg):of the proped nolate than Sepumk 1' e been a significent contributor to Workshops are available for inspection gges.He Commission expects to l
plant reliability problems and hence,is in the NRC Public Docenent Room.2120 publish a Regulatory Guide endorsing a l
of safey concern.De Canaission L Street. imwer laval, NW., Washington, Maine =nanne Standard or providing believes safety can and most be DC.
NRCguidancein Novemberless.He enhanced by stmngthening the As a result of Workshop discussions.
comprehensive m requirements of
- effectiveness of maintenance programs the Commission has come to the the proposed uld be required to 4
throughoet the nuclear industry and this fonowing conclusions' b% WW 6 Wm is the objective of this proposed rule.
1.Rulemaking should enomage following puhucation of the Baal rule.
i Industry initiatives directed toward
%e peoposed rule daansa those i
non-ipri metas es &==&amiaa considers l
It is the objective of the f5===1*=laa ascessary for an acceptable i
that all campaaaats, systemas and
- "*"*"" bem viciar seen ir.amensw Maintenance Program.To be i
Wrectmes of nuclem power plants be g,", e, acceptable, any standard developed to effectively unanntained so that plant a
g intpleinent the rule should have the alpdp===t will perform its intended Nuclear Power Plants " dated jume 37.1ses, function when required.De scope of a cap.ee of NUllsC enrise sesene may be following characteristics:
Punewed emush er US Camuman hanks
- Sbeeld defdue the plant systems.
i the gesposed rule is intended to cover pcMM,".,s,eer.rnse7,17g' structures and aa== pan==ts inchaded in I
all systema, structures and componenta o
the -la*=e program (the scope of including those in the Balance of plant w+ ocassis c.,tes mer ets. h, (BOP).To n " this objontive, the punheemd hem to NaammalTutmiset Animummahm the rule sovers all systeses, structures semce u.s.Depwessai tosummessompaa and aa-panants '-eh those in the
,i propo*ed rule would repare each noyal mood. senasBeld. VA statt. A copy is commercial nuclear power plant to ovesable for inspectaan er empruir for a fee in the
- S'hould require a systematic l
develop and implement a welialeAned NRC Pubbc N-==t asemL rise L semM. Imwar e
inval. Nw.Washimsest DC.
evaluation (" systems approach") of the l'
program to assure that main enmara
i l
j
)
l l
Federal Register / Vol, 53, No, F.28 / Monday, November 28, 1988 / Proposed Rules 47825 be ment co,noom,w,,t,he n,eed for a accomplish with side role t.ha.t osmoot more better existme programa " I share their com me
, e,a,,,,
, e,. emaida,.
e5e
,. mil o t,ve m - &.t &e - of rule -d i.
7.Se Comenisolon believes that the withat a ngelament Ibow not sesmal a e einse won and ham rether than eenefactory enower.1 deset believe the sees enhance the puenction af the pubi c.
inclueim of balance of plant (BOP) he base mee est Wasmesa de not how Regardag mimpmewecen."th i
equipment in the._, _ _ _ _ maintenance estabbehad maintomasse Meet t====== appasse to be saying that emce r
avie le -'y and mper. Howem, importantly to me, them has ao e5ecew =naa==a-is messesary to P
the a===i==== also recognises that demoastration that this runs weeld taprove maintata adequate protection, e rule r
some lioonese mainte==- programs, as implementanen of exteting Inspame. Neither abould be excepted under i so.sog(e)(4). nie presently osaagered, apply to how I been ymvided with -y mar esempson woeld preldbit stat from taking structwes, and consponents documentation on what to prehlsm to and implemsstation easte into sonsideration.
that are,wt t geestion,irrelevemt to how, spednaally, this rule win an It. On the However,it would regare that a documented I
contrary, the trende ste5 has provided show evaluanon be prepared for public comunent.
P"Mcum.of pubbc health and safe cootmoed kaprovement la es maatamenos herefore, my opposties to the exception te l
fun.m
-"' hasards associal ares.
not to the excepuce itself but to the with the operettom of the nuclear power ne proposed rule the=='aa-is now precedentialasture of the use of the adequate r
Plant.%e Commission requeste public publishing faile to provide a heels for protecnos ergimm,t. lat see state that I too j
conument commerning what limitation, if determintap when a meistenesse program ie seengly behow that assotrue semintenance is j
any, should be placed on the Baal eNective or when improvemente are
-ry to mesmo that nuclear power maintemence rule to powride sonne "a
."We are even delayas plante am esfe and to provute adequete licenese Bezibihty in tide regard.
tion of the aaa==paartog Reguletary protection to the public. I also bolsve. lut se l
3.ne amuniandan bebeves that G"kI' "'ul the Saal relo. Without being strongly, ebet this rule is not necessary to r
1 individual wwker accontability plays afforded the opportunity to soview this provide thatproseccan, and that as the ACRS Asqpiscoseanden decament, to Commission le noted it may well have the opposite effect. I j
animportant role in an e5ecti nuit in the possuem of approvtse a spensee beltsve that we eennot a5ard to be carelese maintenann= program.De rule. It is no wonder that this ed==ahat about the use of the "adegente protection" is, therefore, soliciting connesets on the would elicit auch widespread appee6 tion.no argument for exception to the backfit rule.
4 means forincorporating this pubh'c is being enhed to esammet en a sale of he Comedesion to in Mtigation about h considerationlato a hcomese's form but no substanca.1 believe it would be very issue.no --a==-
addr seed e e
7 maintenanno progress.
more productive to delaylessense of this poest to detail under the hombag " Adequate 9.De rammesias 9 desires to proposed mis until the draft ugulatory guide protechen" la the Aespamme to Comments on estabush cr*teria within the le eveilable for m===t. Only then can we the Seal 10 CFR Part so Aerision ofBackfit maimmance rule which would fores the receive seenmsful onesesnes en the process forpower Asoctore. IAt us re= ember basis fe determining when a rul*==hmt package.
that there had been concerne that in dealing -
a I am concerned that this rule goes beyond with the backfit rule, the Commission would maintenance program is fuDy e5ective our authority. I onesot ayee with a rule that use the phrew "edequate protection" and additionalimprovement is not would heve the NRC regulating==*=*===='*
arbitrerGy.no Camusieston could i
warranted front a safety m**=lpaint.
on all systems, structures and aampaa= ate unwittingly be yvtes wedence to that view.
~
i Such crtiaria udght be either
- regardless of whether they have e sexes to piansa==ny,it seems to me that the.
quantitative or guahtative and could be radiological safety or met. I aan treebied by N===='on position on adequate protection based on epom6c menearable attribetes, the statude demonstmed when we ausseet is internally inconsistent. ne Comeuseum on overallplant performa== on.
public -===to on what heitettone,if any, semie to recogmaa that when it states that program results, or on other attribetes.
ebould be placed as &e Saal mio to address this role le asedal to maintain adequate j
%e Cc==ia=6aa requests public structune, eyewee and an-ya'= ate that are protecnon. it is saying that the arrent "without guest /ast safejremst fuey EImphasie) operating plante now pose undue risk to the J
"'=*= mat concerning the need for such tc ee
-- - of pubtle hesie and enfery."
hubiac wMeh we are pedy m;erahng. If I criteria, the form of smch criteria, and nh cleedy ahdientes ser===r====May to iieve that.1 would suggut (se rm sun the criteria themesives.
ebow that a segulation le aseded. We asset would the seat of the Canumoeion) thet this
- 10. Are performance indicators that ask oeruelven: ese we pseemedag with this rule beoems tuumedsetely eNeceve. hk is are being used by industry, may.be used.eulemaking tx the enke of the rule sneelR As sleesty set es seas. As the====*== in r
in the future, or have been seed in the ettested to by the asses whee he the very same samment shows," * *
- the past, appropriate candidates as Comuniesian cited Hama-the NRC already peepened sels==diaan and standardiase quanutadve mensores of mala ananna has the methartty to emieres enspilense la the provassely adotaqr(my emphesse) n i
maintesesos ama.
Comedesien segoiremente, both explicit and eHecuveneest he Ome== las'an is ne assumente edemanad by both the esaN taipacit.to technical specifiutions, pardQ Min W w and the Commissies in to aamply w6th ha=====
analysis reporte, and 10 CFR e
analysis concerning istdicators er the therequirusessteof the rule have part 80. Appendia EL"It seems to me that the use ofindicators of component played a sigstBeast sole h dedulet not to e--a.=na= can't beve it both ways.
j relaability as maintenemos performance support this proposed De staN
. Iraquest osamente en my views.
Indicatore.
argument for to seis's compusess with 11.Should an industry-wide I so. sos bee been made se the beste of east.
impost na-pana=e failure reporting eyeton, e.3, ne star states,that es beddit analysis
. AlW Mamaission believes NPRDS. be used by all pie.nte in order to
,,,,,,,8 g W,g ge 6
e that this M rule le by virtue of to m
support the sharing of generic puhuc heale and without ear GR BO.10g(s)(4) met subject to the anaintenance experience and facilitate addittemelemet."I as ofes
^w _ - t for a beckfit finding and 4
monitoring of maintenance effectiveneef assumptione madein es and analysis, it has nevertheless performed 11 Pa'"a'aa'aan' Roberts had the regulatory emelysis and esquest osamenes en an analysis of oost and other backfit following views:
both shese desumente.Ialas sequest factors as an ehernative ground for I cannot Isin the ensiedty to empporting the coenmente en thevisue of to AERE.Her praamadine with the proposed rule and
. _. - f r=d==airemy se sneinteammae.h state that " * *
- em m deemstatseine of to facilitate public comment on the order to have the beneSt of the public's
.-- n and espealaur to way in which baciftleave'
-==ata it has been say casteen to agree to they am '.ypiemliy endoned, hat lead as to publiontion of proposed. ' ^7 Icannot believe that, onder a rule, a move toward he ele impact of the proposed do so in this botanos.1 beve asked one uniforndty weeld eccw, and this le Ekely to retinirement on Beensees should be r =A===ntal question. What are we trying to decrease the eSectivenese of some of the negligible.The accompanying draft 4
.-,,y
Federal Register / Vol. 53 No. 228 / mnday, Nove'nber 28. 1988 / Proposed Rul:s 47827 costs, due to cost savings from reduced modifications into the maintenance small compared to the other costs and plant downtime. Further, a proactive program:
benehts ofimproved maintenace, and well-planned maintenance program (5) Equipment history and trending:
- 5. Installation and contmuing costs decreases costs of corrective and repeat (6) Maintenance record keepmg:
associated with backfit. includmg the (7) Management of parts, tools, and cost of facihty downtime or the cost of maintenance.
The following discussion presents the facilities:
construction delay.
summary of the backfit analysis. Further (8) Maintenance procedures:
For 100 operating reactors, the total details may be found in the Regulatory (9) Post-Maintenance testmg and estimated cost associated with the return-to-service activities:
proposed maintenance rule is -519B Analysis for the proposed rule.
( )Measms f ve all at tenance million.The minus sign denotes a cost Anal sis of l 50.109(c) Factors pmgram effecMeness:
savings.This estimate breaks down as Y
- 1. Statement of the specific objectives (11) Maintenance management and gotjgw,;
that the backfit is designed to achieve, organization in the areas of:
The purpose of the maintenance rule (i)pla -
is to improve maintenance effectiveness, (ii) Sched -
inssey cost ement o,,ese 80**
and thereby enhance overall safety. by (iii) Staffing.
establishing basic requirements for plant (iv) Shift coverage, and maintenance programs. In establishing (v) Resource allocation:
"*""*"*",* P,'s",
j
,n (12) Control of contracted u,,,,,n no, w 23 these requirements, the Commission maintenance services:
neg# a.warunops a4 intends to consider the industry-wide efforts that have already been initiated.
(13) Radiological exposure control wnpeemeni epomments sat son.a (in uding ALARA)during mamtenance P'en's The objectives of the maintenance a
es, u,,,t.o, rule are as follows:
(1) To define NRC's expectations for (14) Ma,intenance personnel u.nt,eo. ymem un plant maintenance: and qualification and training:
Processes is (15) Internal communications between inmuseed stamng to mace ove..
(2) To improve licensee maintenance the maintenance organization and plant p,u["*,,,
50 programs by requiring the effective conduct of a set of functions and operations and support groups:
e m,9 m
(16) Communications between plant Cost seveige due to eneroved evassomty activities.
With implementation of the rule, it is and corporate management and the and reducea correctwo maaenence.
- uoo expected that the current wide variation maintenance organization: and Tow not sinusey costs a
in maintenance' performance will be (17) Consideration of maintenance reduced so that the performance of recommendations or requirements of Notes:1. Negative signs denote cost mdividual vendors.
savings: 2. Values in table are rounded.
plants that lag behind the industry as a Criteria for acceptability of the conduct of the above activities will be
- 6. The potential safety impact of whole will be brought up to the level of performance of the majority of the industry. Second, the overall average provided in the proposed Maintenance changes in plant or operational l
level of industry's performance should Standard or Regulatory Guide.
complexity,inclutling the relationship to
- 3. Potential change in the risk to the proposed and existing regulatory also improve.
Public from the accidental offsite release requirements, An important part of the structure of of radioactive material.
De proposed rule would require the rule is to achieve improved maintenance performance in a way that Implementation of the proposed certam elements in a plant maintenance maintenance rule will result in an program and should not add to plant or allows licensees the flexibility to estunated total risk reduction to the operational complexity. Improved determine the details of their individual maintenance program so that plant, public rangmg from 50.000 to 500.000 maintenance should result in a decrease person-rem with a point estimate of of challenges to safety systems and specific factors can be taken into about 250.000 person-rem.
forced outages, and therefore, should account. This flexible approach will
- 4. potentialimpact on radiological decrease the complexity of operations.
enhance both aafety and cost-exposure of facility employees. A large Re proposed maintenance rule is effectiveness, compared to a rigid and fraction (two thirds to three fourths) of related to the following existing prescriptive rulern ski"F approach.
the occupational radiation exposure-regulatory requirements:
- 2. General description of the activity incurred at nuclear power plants is (1) Surveillance requirements for required by the licensee or appbcant in associated with maintenance, on the '
safety system is required in the order to complete the backfit.
order of 303 person-rem per reactor-year technical specifications.nese The licensees will be required to have a documented and effective in 1987. Improvements in maintenance requirements are not duplicated for the maintenance program which shall programs can affect collective proposed maintenance rule.
occupational exposures both poshively (2) In-service inspection requirements include the following activities:
and negatively. Increases in are covered under 10 CFR 50.55a. Codes (1) Technology in the areas of:
maintenance activity due to expanded and Standarda.These requirements of a (1) Corrective maintenance, (ii) Preventive maintenance, preventive maintenance or more prevertive maintenance program are not (iii) Predictive maintenance, and aggressive corrective maintenance (to dup!Icated under the proposed (iv) Maintenance Surveillance:
reduce backlogs, for e. ample) will tend mainten:ince rule.
(2) Engineering in support of to increase exposure, and productivity (3) Criterion 1. Appendix A.10 CFR and reductions in the amount of rework Part 50, concerning General Design maintenance:
(3) Quality assurance and quality will tend to reduce exposures.He net Criteria (GDC). requires that a quality control of maintenance activities; effect of these positive and negative assurance (QA) program be established (4) Incorporation of plant trends is believed to be beneScial, but and implemented in order to provide
I Federal Regist:r / Vol. 53. No. 228 / Mondsy, November 28, 1988 / Proposed Rul:s 47829 Nuclear Reactor Regulation, that a comprehensive documented maintenance program is being maintained and implemented, which addresses all elements and activities in l
paragraph (b) of this section including measures to monitor the effectiveness of the mamtenance program and to improve the program where appropriate, in addition, each licensee shall develop (insert a date 3 months after the effective date of the amendment) a timely and expeditious plan and schedule (including Key Milestones) for meeting the requirements of this section.
Dated at Rockville, Maryland, this 21st day of November.198&,
For the Nuclear Regalatory Commission.
Samuel I. Chilk.
Secretary of the Commission.
[FR Doc. 88-27331 Filed 11-25-88: 8 45 ami a u mo cooe n 4 w l
I 1
gr.
E OR NN,
enc p amain u s muctaan atoutatomy couwss,o=,
LICENSEE EVENT REPORT (LER) TEXT CONTINUATION
.movio ous ~o no-m unmis was o _., _....
l Nine Mile Polut l'ni t 2 0 l5 l0 l0 l0 l 4l1 l0 Sl4 0l1l4 Ol0 0l1 0F Ol 6 mvn a
e, mm4wnn i
1.
DESCRIl' TION _0f_THfm*VI.NI on April 13, 1989 at approximately 1101 hours0.0127 days <br />0.306 hours <br />0.00182 weeks <br />4.189305e-4 months <br />, Nine Mile Point Unit 2 (NMP2) experienced a reactor scram due to a turbine trip. At the time of the event the reactor mode switch was in "RUN". (0perational Condition 1) with the i
reactor at 1007. rated thermal power, and the reactor pressure and temperature 3
at 1003.5 pounds per square inch gauge and 546 degrees Fahrenheit respectively.
l t
The turbine trip resulted from a generator protective circuit relay actuation. The turbine trip initiated a fast transfer of house service loads from the station normal service transformer to the station reserve l
transformers. S h hgear 2NPS-SWG003 failed to transfer.
This caused a loss l
of f eedwater since the operating feedwater pumps (WS-PIB and WS-PIC) were being powered from 2NPS-SWG003. The complete loss of feedwater coupled with the normal operation of the turbine bypass valves (TBV's) to control reactor pressure caused reactor water level to decrease to the Level 2 (108.8 inches) setpoint. The lowest the reactor water level reached during the transient was approximately 98 inches. The level 2 setpoint caused the automatic initiation of the High Pressure Core Spray (CSH) and Reactor Core Isolation Cooling (ICS) systems, which injected water from the Condensate Storage Tanks (CST) to the reactor vessel to restore water level. NMP2 entered into an Unusual Event and its Emergency Plan based on Emergency Core Cooling System (ECCS) injection on a valid initiation signal.
When reactor water level was increased to normal, CSH injection was secured.
ICS injection was automatically secured at Level 8 (202.3").
Operators continued to monitor reactor water level and believed that all vessel injection was secured.
However, feedwater was continuing to be injected. Power was lost to motor orerated f eedwater regulation valve 2WS-LV10B, causing the valve to f ail as is.
This condition was not recognized by the operator since the indication of i
the valve position and the position demand failed downscale. The operator 1
I believed 2FWS-LV10B had closed as had 2WS-LV10C, the complementary hydraulically operated regulation valve. Therefore, with the combination of cold water injection, steam line drains, and steam consumption by ICS, reactor pressure was lowered to the point where condensate booster pump discharge pressure exceeded reactor pressure. These conditions permitted feedwater flow l
to the vessel through the open LV10B valve. The operator recognized that feedwater flow was increasing causing reactor water level to inerense and informed the Station Shift Supervisor of these conditions.
The station shitt Supervisor then ordered the remaining condensate booster pump secured.
The maximum reactor vessel water level recorded during the transient was approximately 258 inches. Water level then decreased due to boil off.
When Level 8 had cleared its setpoint the ICS pump was used to maintain watei level. The CSH pump was tripped af ter it was verified to be no longen needed to maintain reactor vessel level.
The f ailure of the 13.8kv switchgear (2NPS-SWG003) to transf e r was caused by a
" positive interlocking roller" not being fully engaged. This positions a limit switch, which provides a breaker position interlock to the closing circuit.
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LICENSEE EVENT REPORT (LER) TEXT CONTINUATION
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4 ll a 0 l0 olj OF uln nro,--
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DESCRIEIl0N OF THE EVENT (Cont'd)
Complications occurred during the event when a licensed operator inadvertently de-energized the remaining 13.8kv power board (N2-NPS-SWG001). The licensed operator observed a " Red Flagged" condition with no position indicating light (due to a loose light bulb). The operator attempted to check voltage on the af f ected board but observed the wrong meter (which indicated zero). When the control switch for breaker 1-1 was placed in the reset (or lockout) position, power board N2-NPS-SWG001 was de-energized. The operator immediately realized his mistake and re-energized N2-NPS-SWG001 in accordance with station procedures.
During the momentary loss of 13.8kv power to the house service loads, the remaining Circulating Water (CWS) pumps were de-energized. The decrease in the Main Condenser water box level af ter the loss of power to the CWS pumps, prevented the immediate restart of the CWS pumps and the subsequent loss of the Main Condenser as the primary heat sink. The SSS directed the use of the Steam Condensing mode of the Residual Heat Removal (RHS) system loop A.
The Main Steam Isolation Valves (MSIV's) were directed to be closed as the Main Condenser vacuum decreased to 10 inches of mercury. As the inboard MSIV's were being closed, a Group I Isolation was received due to low condenser vacuum and a fast closure signal for the MSIV's was received (closing all the MSIV's).
The ICS system was used to maintain reactor vessel water level and the Steam Condensing mode of RHS was used to maintain reactor vessel pressure control.
A reactor cooldown to ambient conditions was then performed.
i Uninterruptible Power Supply 1D (UPS-lD), tripped due to an overload condition. This resulted in a loss of approximately one half of the Gai-tronics system in the plant, a total loss of Gai-tronics in the Control Room (affecting communications with plant operators outside the Control Room) and a partial loss of emergency lighting.
1I.
CAUSE_0E..IHE_IVENT I
A root cause analysis was performed using Site Supervisory Procedure S-SUP-1 l
" Root Cause Evaluation Program". The root cause for this event was determined to be loose wire connections in the Main Generator Potential Transformer cubical, 2 CMS-CUB 01, for circuit 2SPGZ03. This is attributed to poor installation compounded by vibration in the area of the connections.
I l
aseC Pero seeA u 8 huCLE&a R$3utef 0av coumissiosu LICENSEE EVENT REPORT (LER) TEXT CONTINUATION
- aovio ows =o vio-cio.
LArimf 8 4 31 ss F AC3UTY kAmst HI M e-een g; LS8 hund888ISI P A 04 138 vs a j
u g,* 6 t*4*,p 4
Nine Mile Point Unit 2 0 l5 l 0 l 0 l 0 l 4 l 1 K) 8l4 1 1l 4 0 l0 d l1 oF lllS rurrn w
s=,wer assaon III.
ANAIXSIS OF THE EVENT This event is reportable under 10CFR50.73(a)(2)(iv):
"Any event or condition that results in manual or automatic actuation of any j
Engineered Safety Feature (ESF), including the Reactor Protection System (RPS). However, actuation of an ESF, including RPS, that resulted from and was part of the preplanned sequence during testing or reactor operation need not be reported."
The reactor scram was due to a turbine trip, which was in response to the actuation of a generator protection relay. The reactor scram is a protective j
function and therefore poses no adverse safety consequences.
i The f ailure of the 13.8kv power board (N2-NPS-SWG003) to transf er and subsequent loss of the remaining 13.8kv power board (N2-NPS-SWG001) did not pose a threat to the health and safety of the General Public as the three (3) divisions of Emergency Core Cooling Systems (ECCS) were operable with power sources frem off-site and Diesel Generators. Only two of the three ECCS divisions are utilized to achieve safe shutdown.
CSH and ICS automatically initiated at Level 2 (108.8 inches), to restore level as designed.
The problems that were encountered with tle overload condition of UPS-1D (partial loss of communications outside the Control Room and partial loss at emergency lighting) did not compromise the safety of the general public as the safe shutdown of the plant can be achieved from the Control Room.
i An evaluation of 10CFR50 Appendix R, Section III.J. Emer gency Lighting" requirements was conducted. The evaluation concluded that there is no impact on the Appendix R safe shutdown analysis.
I Engineering is, however, conducting a review of associated essential light circuits to reconfirm that cable routings do not traverse openly in postulated fire areas (i.e., UPS-1D circuit cables through the Control Room and Relay Room areas).
If unanalyzed conditions are found, a supplement to this repott
]
will be issued.
Transient recording indicated that water level was slightly at ove the lowest elevation of the Main Steam Line (MSL) nozzle. However, the level trends indicate that water level did not reach the Main Steam Line (f1SL). A firm conclusion that water did not flow down the Main Steam Lines :ould not be made.
However, if water flowed down the Main Steam Lines, it was f or a very short duration.
Based on prior Engineering analyses performed on previons vessel overfill, the effects of potential transients created, if water had entered Main Steam Lines, were within plant design margins.
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ANALYSIS.OE.lHE EVENI (Cont'd)
In accordance with the requirements of Technical Specification Sections j
t 3.5.l(f) and 6.9.2, Emergency Core Cooling System (ECCS) Injections, the following data is provided:
For the HPCS nozzle s
- Total accumulated initiation cycles to date = 4
- Current usage factor value remains below 0.70 l
The duration of the event was approximately 8.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> from the time the event I
was initiated (1101 hours0.0127 days <br />0.306 hours <br />0.00182 weeks <br />4.189305e-4 months <br />) until the Shutdown Cooling Mode of RHS was controlling reactor temperature and pressure (approximately 1930 hours0.0223 days <br />0.536 hours <br />0.00319 weeks <br />7.34365e-4 months <br /> ).
i l
IV.
COBRECIIVE_ ACTIONS 1.
Relanded the disconnected wire and tightened the other potential 1
transformer connections.
2.
The Electrical Preventative Maintenance Procedure N2-EPM-GMS-R693 will i
be revised to ensure the integrity of the wire connections in the areas of high vibration is checked.
3.
The Electrical Preventative Maintenance Procedures will be revised to check interlock mechanism for proper operation.
i 4
A Lessons Learned Transmittal was generated by the Operations Department to discuss the problems associated with this event.
l 5
The licensed operator involved in de-energizing N2-NPS-SWG001 was counseled on self-verification and communicatinn.
o.
Operation Procedures have been revised to ensure visual veritirntion of the interlock roller positions on 13.8kv breakers.
7.
Operation Procedures have been revised to ensure that feedwatet.
Condensate Booster and Condensate pump powne suppiv linenpa
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s ep.ii n t e for running pumps.
I R.
Operation Procedure was revised to provide improved disertion t ot watet level control following a reactor scram.
9 The electrical loads on UPS-ID have been reduced to prevent a ti-ip on overIcad. Additional modifications are being considered to !inthet reduce UPS-ID loads.
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ADDlIlCliAL INFORMATION A.
Failed Components:
- 1. Failed Ccmponent Identification: 2VBB-UPSID Component
Description:
Uninterruptible Power Supply for Station Lighting /Comunication Component Vendor:
Exide Power Systems Division j
- 2. Failed Component Identification: 2NPS-SWP-003-1 Component
Description:
3000 Amp Breaker Component Vendor:
General Electric Company II.
Nine Mile Point Unit 2 has not experienced a reactor scram caused by a similar event.
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A.B.
Davis, Region III I
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Alexich P.A.
Barrett.
J.E.
Borggren R.F. Kroeger i
NRC Resident Inspector"1 J.G. Glitter, NRC R.C. Callen G. Charnoff, Esq.
Dottie Sherman, ANI Library.
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'""'* REACTOR PROTECTION SYSTDi ACTUATION!DUEiTO MALFUNCTION OF CONTROL ROOM INSTRUMENIATION DISTRIBUTION INVERTER d'2d4RtSMMIsfdM' S VENT DAf t ISI t e m hUtes E R 443 -
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MWQ x n yn%,r0 On August 14, 1989 at 1601 hours0.0185 days <br />0.445 hours <br />0.00265 weeks <br />6.091805e-4 months <br />, a,m ' Reactor Protection Syste= (RPS) sn actuation (reactor trip) occurred when operators transferred the Centrol Roe = Instrumentation Distribution (CRID) IV (vital bus) inverter to its
~
nor=al Class IE power supply'and the inverter failed. When the CRID IV inverter failed, a reactor trip signal was initiated due to the Reactor Coolant Pu=p (RCP) circuit breaker position indication open (f ed f rom CRID IV).
Prior to the trip (at approxie.ately 1540 hours0.0178 days <br />0.428 hours <br />0.00255 weeks <br />5.8597e-4 months <br />), the CRID inverter had transferred to its alternate non-class IE power supply at the same time that a control power fuse had blown on Power Range Nuclear Instrumentation l
System (NIS) Channel IV (N-44). Subsequent investigation determined that the CRID inverter failure was due to a failed silicon controlled rectifier (SCR) in the static transfer switch. This also resulted in the failure of fuses and power supplies in various components fed f rom the CRID.
The faulted SCR's were replaced and the CRID inverter declared operaH e.
All cceponents fed from the CRID vere inspected and, where necessary, fuses and/or power supplies were replaced.
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Description of Event M
On August 14, 1989 atapproximately21540. hours.~operatorsreportedthat the control power fuse (EIIS/EF-TU);for; Power Range Nuclear Instrumentation Channel (NIS) (EIIS/IG-CH) N-44 had blevnM At.the same time, the Control Room Instrumentation Distribution ((CRID)iIV (vital bus) inverter (EIIS/EF-INV) had transferred to its alternate noti { class;1E power supply. At the time,
~
it was believed that the blown fuse may have' caused CRID IV inverter to transfer. Operationa pulled lthe[ control ~and instrument fuses f or N-44 per procedure and contacted -Instrumenta' tion?and Control (I&C) personnel to trip the remaining bistables associated'vith N-44, and to determine the cause of the blown fuse.
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While 16C vas preparing to troubleshoot N 44, Operations prepared to transfer the CRID IV inverter back tolits normal Class IE supply (EIIS/EF).
This power supply transfer was~1nitiated to provide the availability of an alternate power source (in thofevantlof[the failure of the prieary power source) during the troubleshooting process. k'han the non-licensed operator (NLO) in the CRID inverter room pushed the jInverter to Load" push button on the inverter, the alternate source" pilot light vent out and the inverter f ailure and f an failure lights;cac.a on.bThe " Inverter Supplying Load" pilot light came in dimly and then vent out. He then pushed the push button a second time and believed,that_the inverter transferred properly, as there were no indications to the.. contrary. However, inverter output voltage was indicating approximately"84;VAC.in the control room (normal l
voltage is 120 VAC). The reactorftriploccurred when the inverter lov l
output voltage caused the Reactor Protection System (EIIS/JE) to sense the
- 24 Reactor Coolant Pump (RCP) (EIIS/AB-P) breaker position (f ed f rote CRID IV) indicating open above Permissive P8.
Following the trip sequence [ opening'of the reactor trip breakers (E1IS/JE-BKR),
turbine (EIIS/TA-TRB) trip, insertion of reactor control rods (EIIS/AA-ROD),
feedvater isolation (EIIS/JB) and autor.atic starting of the motor driven auxiliary feedvater pumps (EIIS/BA-P)], Operations personnel irrediately implemented Emergency Operating Procedure 2 ORP 4023.E-0 to verify proper
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response of the automatic protection system and to assess plant conditions for indicated appropriate recovery actions. Due to the f ailure of CRID IV, various control room instrumentation indications and components were unavailable. These included the protection syster status lights, #24 RCP operating parameter indication (EIIS/SB-PI), Steam Generator vide ranFe level indication (EIIS/AB-LI) Loop 4 indication for auxiliary water flov (EIIS/BA-FI), main steam pressure indicators MPP-212 and 242 (EIIS/SB-PI),
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one channel of steam generator narrov range level (EIIS/JB-LI), and loss of the use of stear dump system (EIIS/SB-V).
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Operators tripped the #24 RCP Qasktheydad:nofindicationavailable.
They manually activated the mainfurbine75olitioidl(SOL) trip (EIIS/TA-SSV) and the AMSAC (EIIS/JG), becauseltheyjhid[tiFatatusL11ght for main turbine trip, although the turbinefstop}M M @l W Fyali M(Ells /TA-V) position indication indicated full closed.
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Two rod bottom lights did not lighti(rod position indicators [RI} showed i
the rods at less than 25 steps).~g Par' procedure, operators emergency borated for five minutes. It vasilaterl determined that a leaking capacitor ononerodbottombistableand%burnedfoutresistorontheotherhad caused the problem and was notirelateditolthe CRID IV failure. These items weresubsequentlyrepairedRBecauseTofithellossofsteamdumps, operators usedthemainsteampoweroperatedrelief[ valves (EIIS/SV-RV)tostabilize the plant until steam dump control was're-established.
~ #$$$&Y W Operaters also noted that ti$isispeedtindicationlL(EIIS/BA-SI) for the Turbine Driven Auxiliary Teedpu=p had failed,T it[vas'later determined that this was due to a broken lead on the' magnetic' pickup for the tachometer and was notrelatedtotheCRIDIVfailure[.%Thiswasrepairedandthepumptested per Operations procedure.,
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Tollowing the trip. I&C personnel began investig,ation into the cause of the CRID IV inverter failure, and the possible affects of the low voltage condition (84 VAC) on components fed from the CRID IV inverter. This included an examination of Train B of the Solid State Protection System (SSPS), and N-44 and other miscellaneous ' drawers in the Nuclear Instrumen-tation system.
T;y QL fi j3 f Fh It was found that the lov voltage condition had caused high current flow to leads fed from CRID IV. resulting in_a number of individual breakers tripping open, fuses blowing and power supplies burning up.
s
.y Attachment No. 1 is a listing of those components that were found to have been adversely affected by the inverter failure.
The cause of the low inverter output which led to the reactor trip was determined to be due to a failed silicon controlled rectifier (SCR) in the static transfer switch. This SCR had experienced a rare failure mode vith the SCR gating close which blocks the output. The remaining SCR conti.wed to function, producing sufficient voltage to allow the transfer to occur, but unable to supply sufficient power to the equipment.
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' tested [to ensure operability.
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Cause of the Event C%64k mapO ulted SCR in the CRID IV inverter The cause of the event is due..to;a,lted[in;very lov output voltaFe static switch circuitry.$'Ihisjresu (84 VAC) resulting in a reactor tripifros j24' Reactor Coolant Pump breaker position indicating open above PermissiveiP8@7heilov voltages also resulted in P
high current demand by equipment supplied by,CRID IV, causing damaged cocponents and blown fuses y t ] f 3 y.T*M.
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This event is being reported [in.accordance,with 10 CFR 50.73(a)(2)(iv) as an event that resulted in an unplanned automatic actuation of the Engineered Safety Features,;includingithe'l Reactor Protection Systeci.
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The automatic protection responsesh including' reactor trip and its associated actuations were~ verified 7to.have functioned properly as a result of the reactor trip signal. ;Besad[onfthe.'above, it is concluded that the event did not constitute an unreviewed.' safety question as defined in 10 CFR50.59(a)(2) nor did it adversely impact the health and saf ety of
' Yh fj%l$ \\ f hNhk Corrective Action
) $.GjbJ.g:,j The CRID IV inverter was repaired,f+d declared operable on August 16, vm
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v c an 1989. All components povered from'the CRID'1V vere evaluated for damage
~
and repaired / tested as necessary to verify operability. I6C has developed a guideline for checkout of a CRID inverter to verify that the inverter and static switch are operating correctly if a CRID should auto transfer.
This guideline vill be performed before operators are instructed to transfer back to the Class IE power supply and thus eliminate the chance of a reactor trip and potential equipment damage due to the causes of this event.
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Manufacturer:
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EIIS Code:
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wyT,3 hl Rack 12:
MPP-212 SGl1 Steam; Pressure sed MPP-242 SG;4; Steam"Pressureet.i.
BLP-110 SGLoop;1fleve differential pressure transmitter BLP-120 SG LeopI2flevel ifferential pressure transmitter BLP-130 SG Loop?-311evel. ifferential pressure transmitter j
BLP-140
'SG Loop [4}1eia1{ differential pressure transmitter KPS-153 Pressurizer [piessUseftransmitter PPP-300 LoverlContain$enti?prissurizer transmitter RWSTile' al(Erahamitters ELS-951 v
FFI-240 Aux;Feiedvsteito*SGli4N
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NTP-240 Loop.;4'.RTDiT e
FFI-241 Feedvatei2to flov' indicator transmitter n
Cabinet 22
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ILA-131 AccumulatorjTank Level Transmitter i
ILA-141 Accumulator /TanF[,4[LevelTransmitter IPA-131 AccumulatorqTank!3 Level Transmitter i
i IPA-141 AccumulaterlTank]4flevelTransmitter j
QTI-240 RCPLoopj4llov' bearing" J
RCPLoopJgfl[i.ealS QTI-40 s
NTA-252 Pressurizer; vapor temperature IFI-54 Loop 4~ cold" injection QFA-240 Seal water, injection transeitter QDA-40 RCP Loop.4 seal water
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ITR-311 Residual heater'fleRTD QTC-302 Letdown heat exchanger RTD NTA-152 Pressurizer. relief. discharge temperature IPA-310 RHR pump #1'_ discharge QPC-301 Letdown heater low pressure IPA-250 Boron injection. tank ITI-310 Residual heater #2 outlet QLC-452 Volume contro1{ tank ITI-311 Residual heater #2 outlet QRV-303 Letdown to volume control tank diversion valve FRV-240 Loop 4 feedvater control valve CRV-470 Letdown Hx CCW valve QRV-301 Letdown heater' control valve I e '. -
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- s[I control Bank B. position' control Bank C^ position p&
w Control Bank D position 2,?
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.i Bank B limit b
Bank B positi
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Bank C position 1 D Bank D limit W M.h T ~
Bank D positioti D"E.
Average power 01, 02, 03,& 04
. y c:e PPA-313 Upper. containment pressurizer transmitter Rod control auto rodslinlC}<:
Rodcontrolautorodsouth.
Rod control rod speed demand _
f h'qaq;1 Incore E @N5h. #
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UCENSEE EVEN R CONTINUATION 1
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SSPS Train B #2~15V Pows' p
t Demux48VPower'Suppif h'
Demux 15V Power Supply
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NR-44 Power' range) jrecorde h.
P SG-14 Overpower; records
-q M FRV-210 SGwaterlieve1[controivalvejLoop1 auto /manualstation
{
NRV-164 Pressuriser!waterTs~pysyjvalve auto / manual station Feedvater differentialipressure[ con}TK";f 2,"S" recire manual station QRV-450 Boric ai:id;transferspump troller Pressurizer' safety andjrelia ' valve flow monitor SG-31 Incoreithermociup iTrai B' recorder CRV-341 Witrogen)suppipitd@cumLlai;or tanks vent valve controller
~
N-44 Power; range;contr over fuses Comparator and rate' draw's Instr. power and',c'ontrolf e uses$ N Audio count rate'drawarf j
1.
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2.
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Miscellaneous contro1#and indication drawer instr. fuses Wigg@9F hN g,
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Damaged Components
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@l 2-QDA-40 RCP #4 No. I sealfdifferential pressure transmitter
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Miscellaneous control and indication._drsvar low voltage power supplies (2)
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P Decemoer 28, 1981 h,,
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l lNl Mr. James P. O'Reilly,;01 ectbt; jeg o II, qp:fghh Office of Inspection and Enforcement y
] r[W@'W U.S. Nuclear Regulatory, Commission 4t is p
101MariettaStreet, Suite ($g 3100' Atlanta, Georgia 30303Ld gy/$h
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Dear Mr. O'Reilly:
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.REPORQBLE OCCURRENCEj'y 53 335
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am ree LUCIEUNITilf
'N%d.S Usm:
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rmaw DATE OF-OCCURRENCE NOVEMBER 25, 1981 l
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,.. ELECTRICAL _ SYSTEM' k.2%Nf
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The attached Licensee Event Rep' ort spbe n Technical Specification 6.9 to proiiddl30"gjsubmitted in accordance with
?Mu g j ',g r,' day l#";
notification of the subject
.8 occurrence.
.Wggg..ty g, 4 ujm Very truly yours, M agg g;Nhfp[
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Director of Nuclear Energy Jo./
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.ry;* : mp Attachment kN$N b n_',
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Director, Office of Inspection.-and Enforcement (30) cc:
Harold F. Reis Esquire-gigg File 93341 SL g;) pq '
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ICENSEE EVENT. REPORT a.n
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k EASE PRINT CA TYPE ALL At0Vlato iNPORNIATioN)
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IVINT Of SCRIPTION ANo #RCS ASLE CON $tCUENCES D r*
i o,1 l While preparing to start tipYaf ter? refueling, B2 Station Service Transformer ]
i 1
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Ahnne _1 minnten inter tho ? j 1
F5'TT'1 i f ailed, deenernizine the B2',,480V' Load Conter.
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tie breakers feeding the'AB-DC bus from'thi B bus tripped, deenergizing the 1 i
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nnent nnd 1?nV vien1 I
I.>islIAB DC bus.
Loss of this ' caused 'a" lone of AB ennern1
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AC which feeds all' control, roots alarms but NOT instruments. The DC bus, 1
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,,,,,,120V AC and alarms were@stiore,..,.,,,ithin 15 minutes.
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io,,ii See LER 335-79-28 for related events.
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.a O AUSE CISCRieTION AND CORRIC',*lV{ ACTlQN3_
!The B to AB tie breakers were thoroughly tested and inspected. A loose
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t i i I L nent enrminntinn tw-iover current trip was set 'onE"Lhcm both breakers.
,y The trip was reset to ;
- Hi per design and all connectidnssvere' {orqued. All similar breakers werc!
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See attached' contin'uation* sheet.
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I lA.j@lOPERATOROBSERVATION l'IsIl.nj @ l n!n Inl@l m
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1.ER 335-81-053 CAUSE DESCRIPTION AND' CORRECTIVE ACTION' CONT.
is felt that a combinatio%@frthe;1oose power connection in
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' nio It conj unction with.thdTEddit'id6Elkliia'd,placed on the "AB" DC Bus
- 480V!I3adTCchter*.. caused the "B" side breaker would natura11yJtr._ip..to.n.f.u.nd.e)rv,.ol breaker to trip. On'c'E It must be noted that the vital 120V'AC*does mot 7 supply;. power to instrumentation in Ro om ins t rumen t a t ion. va s..I.ns t.los t.7,,.rument-AC; system does tha the Control Room and: thej g,
- nock vep%
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A The erroneous "Lo" settingfon)the overcurrent trip, at first suspected, is now not consider'd,'arcause of this occurrence.
t e
n/Ad$7598?E/1, The Vital AC invert'eh is, powered from'the "B2" 480 V LC Via
)
"AB" 480 V MCC. The."AB"/DC'BEsiis ithe backup power source.
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Comm nwsalth Edis:n i
Byron Nuclear Station l
4450 North German Church Road l
Byron, filinois 61010 1
( >c t obo 3,
1990
- 1. t :
ilYROri 90.0963 ti. !i. tiuclear Regulatory Commission f io ciuna n t Cont.: ol Desk Hanhington, D.C.
20555 c-l >o a fil : ;
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The enc l os*tt Licensen Event Repos t f r om Byron Generating St.at.lon is being i e annmit t ett to you in accordance witti t.he :egulrements of 10CFR50. 7 'i( a > ( / J t r v i.
Thir. report in number 90-006: Docket tio. 50-455.
S i ne..r e l y,
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R.
Pleniewicz
/
lit at ion Manages ity s on !Jur lean Power Station Irl' mlm I; o.
l. o; o i.: :
1.irensee Event Itepor t. tio.90-000
- t.. Itai t Itav i s, 13RC Region ill Mmi ni s t e nt ni W -- t i opp, 13 RF Senior Hesilent inspectai i til".
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1.lCENSEE EVENT REPORT (LER) g racil.ty Name tI)
Docket Number (2)
Pe gg_L}L 0151 O! 010141515._ Lbg!fJb
____ __Bnrn _ Unit 2 Titio (4)
Inag.grtenL i.tgin A_$gfgly iniection Sional Due to Miscomunication and ProLiiLral Deficiencv____
Lvont Date ($).
_._ ___,1[R_!b dir l.61 RrporLDale (7)
Qtherlacultica_ involved 10)
/, Sequential
/// Revision Month Day Year
,_FAgiltlyyames_ _22Lt.tt. PNmhg.r ( 5 )
Year 1/,/,f Month Day fear fff H
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///
NumbtL NONE
_9LILOLOLOLLI 0h 0 la
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...D.Lo_16 oIo 1.Lt_ 0..lL S.It.
...0151 DL01.01 1 1
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THIS REPORT 15 $UDMITTED PUR$UANT TO THE REQUIREMENTS OF 10CFR fil't li A I I Nf.
LChr.ct gne or more of the followinal fill
,,,y h.
20.402(b) 20.405(c)
JL 50.73(a)(2)(iv) 71.71(b) rowtp 20.405(aHIHil
_ 50.36(c)(1) 50.73(a)(2)(v) 71.71tc) l i vi t 20.405t a H I H il) 50.36( c H 2) 50.73( a H 2 H vi i )
Other ispecify (10)
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LICENSEE CONTACT FOR THIS LER f12)
Name f[(iP_HQN.( HVM0(R _
AREA CODE r 9ierA.OmatinLiO2 ngir Ext. 2 HB LLLLLLlLlLLd 5L4L41 '
i CQ& LETE ONE LINE FOR EACH COM ONENT FAILURE DESCRIBED IN THIS REPORT (131
( AU5f.
SY5f f M COMPONINT MANUFAC-REPORTABLE
/
CAUSE l SYSTEM COMPONENT MANUFAC-REPORTABLE //////
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__3GPf1@@LAL REPORT EXPECTED f 141 Espected M_gn_th_.{ pat i fese Submission j
Date TIS)
L l'es il' yes... (Q!"Plett. LKPLLIED_5VEHl531QN.DATEl I H0 E_lL
- 1. [3 ? 11 Alt'.19 A: I itimit to 1400 spaces, i.e, approximately fifteen single-space typewritten lines) (16) on toi 10 at erpro imately 0805. with Unit 2 in Mode 5 (cold shutdown) 2B05 3.2.1.1.a-1, " Unit i=o train A Manual satet, Intettion Initiat6on and Manual Phase A Initiation Surveillance," was being performed per the refueling nut age u hedule.
Af ter perf orming the normal Saf ety injeClion (51) DE), it was noted that the 2C Reactor Containment Fan Cooler (RCFC) lo= speed fan breaker did not close. Attempts to close the breaker were unsuerpssful. At 0820 the 480 volt Bus that feeds the breaker was de-energized to allow removal of the breaker.
At 0A50 =hile strippemg the Bus of its Alternating Eurrent load, Instrument Inverters 211 and 213 were de-eneeuired due to a cowinunications breakdown. When the Instrument Buses were de-energized, the Pressortier P r e a. s n e u in= Si and Steamline Pressure low SI blocks were lost on Train A.
The Unit Reactor Operator =as.,na-are j
o.it sh, binr ei s had been lost and the surveillance did not contain an emergency emit section to provide e-stneation uu. dance.
At 0902, the Train A reatter trip breaker was closed per the surveillance and a 5atetv
)
Inie.e.no siunal resulted due to a loss of the Reactor Trip interlock (P 4) =hile c yc lino the reac t or trip breaber.
j incl de a procedure revision to the manual SI surveillance to it.rlude an emercenry e.it
.or.
.e..e a.tions i
5,,..no th s event will also be included in Operator Required Listening. In addition. placards have been plated on th, ontent Board to verify the necessary blocks prior to cycling the reactor trip breabers.
repnrtable pursuant to 10r.rR30.71(a)(2)(lv) as a result of the Automatir (ngineered Saf et y featnre This e.pnt i*
Aseqatenn i
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LICENSEE E EN T
T kT50N rnrm Der, R FACitlIY NAME,(1)
DOCKET NUteERj(2)
- LER NVM ER (6)
PA2t_ ( 3L
{b5 Mf Year 7
///
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///
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_Syron jhilt. 2 0I510I h
0l0l6 Ol 0 0 LJ!_ LM IEXT EnergyIndustryIdentificationSystem(E!!$[ Code's?arelldentifledinthetextas(XX) m,4 ya pqW jb dy!q!dfWM wh g '
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M(LCQNQWQNS PRIOR 10 EVENT:
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Fvent Date/ Time 9-03-90 /
0903 MM
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Unit 2 MODE _L_ - Cold Shutdown
. Rx. Power,,7.[_QQF RCS (AB) Temperature / Pressure _12'f/_JM_nig_
Nh[ldhf[I H.
PM W P110N Of EVENT:
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Priortothisevent,on9-3-90atapproximately"08057jithUnit2inMode5at152*Fand365psig.
Operating surveillance 2005 3.2.1.1.a-1
," Unit (T k irain,A, Manual Safety Injection Initiation and Manual Phase A Initiation Surveillance." was belng performed'perjthe refueling outage schedule. Train A Solid State Protection system (55PS) was in NORMAL $frainf B_SSPS [was In TEST. Train A Emergency Core Cooling 1
System (ECCS) pumps were RACKED TO TEST.% Train B ECC5 pumps,were operating as required. Pressurizer level was 70L Instrument Air (IA) (LD) containment $iblatib(Yalve 21A065 had been reopened to restart charging and maintain pressurizer level.
, j$hshkh5 7 W Mp.,M & ni,.
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,m After performing the manual Safety Injection,($1);(JE) h [ f froe' main control room panel 2PM05J. It was noted the 2C Reactor Containment Fan Coole{(RCFC){high that ee an breaker (VQ) [VA) had tripped but the low speedfanbreakerdidnotcloseasexpected{jA1ccalynspection.'ofthebreakercubicalidentifiedno 4
enternal problems. Anattemptwasmadetostart!theifanjrompanel2PM06Jasamethodoftroubleshooting the problem with the breaker.Qh[1f0ll speed breaker closed, but running amperes (amps)
^
the location of on the control board indicated excessive Currenti(appronlmately,335 amps). The control switch was taken to i
pull-to-lock (PTL) but the breaker did not open'."dANttempt was made to open the breaker locally but the breaker did not open. At 0820, 480 volt safetyYelatid. Bus 231X which feeds the 2C RCrt was de-energized, and the suspect breaker was removed from it(CubiClefjf[A't)C850/ while Bus 231X =as de-energized. Equipment Operatorsstationedatthe120voltInstrument: Inverters l2jjand213weredirectedtoshedtheAlternating Current (AC) from Bus 231X. However, they opentI (icdr breakers (AC and direct current (DC)) at the
~
inverter panels due to a comunlcations breakdo'wn7 Th[following components were lost:
.?k.
.h Q
jy/ R Instrument Bus 211 yh.
OkJYESN ((O ygfi j
l N31 Source Range NTS Intermediate Range N41 Power Dange channel Instrumentation and Control Cabinets (2PA01J. ZPA05J. 2PA13J. 2PA09J. 2PA45J. 2rA27J)
Main Control Board Power (Channel 1) 55P5 Channel ! Input Cabinet for 2PA09J and 2PA10J
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FACILITY NAME (1)
DOCKET l M BE 8?tER HUNER M)
Pa c e 13L._
Th!-
fYe' $
///
Sequential
///
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' odes"areildentified in the text as (XX] ' QYFN $@fQ 3% $g$ '.&gle B. QLKRifl10E OF EVENT: (Cont) 9y W Instrument Bus 213 'QW M pd-N43 Power Range channel ,74%~ Mh 2PA07J,p<2PA15J) InstrumentationandControlCabinets(2 PAD HainControlBoardPower(Channslj!!Q hh $$PS Channel !!! Input Cabinet for 2FA09J!snd 2PA10JW , $$$h$hhh l At 0854, Dus 231x was restored. i At 0855,i!nstrumentflnverter. 211 was re-energized. Instrument Inverter ~ i 213 could not be started. InstrumentBu's213wasre-energized lfromtheConstantVoltageTransformerat 7tT@k $ Qh 0910. $kbhh, NN whentheInstrumentBuses211and213~werede-energizedh'nce,PressurizerPressureLowSIandSteamline the PressureLow$1blockswerelost' entrain [Nnofutomat'kH$wher7si P-4 (Engineered Safety feat.res Actuation System Reactor Trip Interlock) was present [ Safety.Injectionoccurred. Th, N50 was una-are that the blocks had been lost and the surveillance.was; resumed.' f s Mk[7hk$hhhhhE At0902,theATrainreactortripbreakepRg[JG)[wasposedtoresettheFeedwaterIsolationsignalrer the surveillance. ASafetyInjectionstgnaltresulted[dueto:alossoftheautomatic$1blockwhenP4was clearedbyclosingthetrainAreact$(trjp eakekkihej51[onlyaffectedonevalvesinceallother aquipment was in the 51 actuated state as a.resulfoflthe manual 5! initiated for the test. valve 2!A065 closed which resulted in a loss of chaiglSg8 pee's'sidfdhe~v'el' began drorping about 1% per minute. lW h $ $ $kNh k h ' l An investigation was initiated by the Jhif t}ControldRoom Engi,neer (SCRE, Senior Reactor Operator) and the Shif t Engineer (SE. Senior Reactor Operator)h0ncQhegoot"cause for the SI was determined, immediate act ions were taken to re-establish the tow Steamldn,e and Pressuriser Pressure blocks and the 51 signals were blocked. The Feedwater Isolation signal 1was; reset' -[j@ , o&s.
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Comennent Cooling (CC) (CCI to the 28.Residualysat (RH){[BP) Heat Exchanger was lost for appronimately 10 minutes during each of the two manual,5!jinttlatlon tests'as' expected due to procedural instructions. Due to the planned loss of CC to the RH Heat Exchangeffkluring th's surveillance, the temperature in the Reactor Coolant System (RCS) (AB]increasedapproximatelbl3'ThhDuringthesurvelliance,thepressurizerlevel decreased to 20%. with the exception of these'evolutlons', plant conditions remained stable during this event. ~
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l the NRC Operations Center was notified of the event via the Emergency Notification System pursuant to 10CTR50.721b)(2)lii) as a result of the automatic Engineered Safety Feature actuation. This report is I submitted pursuant to 10CFR50.73(a)(2)(lv).; W_ S-i 7 - a ., [w c. l.AV5L OLLYLMI: l fL~ },% ihe root cause of the initiating failure of the 2C RCFC' low speed f an breaker is unSown at this time. The breaker has been sent to Westinghouse for a failure analysis. A supplemental repor*.ill be issued hen a ,\\ ennt rause is determined. .j', ~ -
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\\ the inadvertent de-energigation of the Instrument Inverters was the result of a cognitive personnel error the part of the Unit 2 N50 for not clarifying the directions he had given the Equipment Operators in the ] on 4 h N P$:. ' i t 0611R /00 7 FR-S i s 2,, .h4 Ah% o -
m l , m..y. ' -., 3 ... 4 % ~ e tfCtN,.EF r9tNr'R s0RT77ttRwT rxtrc0ETW;Ar ten rer-c t.z c_ a FACilliv HAMC,(1) DOCKET. NUte E LER"HUMBER (6) Past_13) SQ .jYe @Mg Sequential y Revision bb N' fil Number LLL lidtL gyrgn. Unit 2 l0i 0 01016 0l 0 Q_14 Of Ll5. IExi EnergyIndustryIdentification1$fstem# E'$: IdWafelj& dentified in the test as (XX] ' &-., t y_% ^ fni[ ' ' i ..%yG ;' l. ' ' ' - ~$; M"bg d, j ~. ' %. 9 2 C. (M5LQf EVENT: (Cont) 5 yg
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.7 m.%,: the intermediate cause of the $aretylInjectic psj na,1,rghejurveillance was re-entered, was the loss of eralssfve;P-ll(lossofPressurizerChannelsP455 thetwoInstrumentBuseswhichresulted,jn oss and P-457) which caused a loss of the"bloi 'E og$ieaml. lye [and. Pressurizer Pressure Safety injection. ,ro t al; eficiency since there was not an emergency Therootcauseoftheinadvertent;$1 s[du o I emitsectiontoinstructtheOperators'how3o, restore /verJJyplintequipmenttonormalconfiguration. ' TDM@M$.hPM$s of the Instrument Inverter;213(tojestart[wasj;jf vestigated under Nuclear Work Request (tNR The failure this resistor open, the capacitors injthej!nverter Could,'not,en and caused the IFU fuse a.elle.d cp With It was determined that the.I,m res,istoL ad h - R. 16632. tidf!nMh,be charged prior to starting the Inverter.[wouldh The current drawn by closing the.lthbt)Mbrekl[r fuse causing it to blow. M OY,& k A khL o. iAn fy Analysis:
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,M.4 mhay wwc Lym riant and pubilc safety was not affected'b this event.gNo. water was injected into the RCS during this Only one valve (2!A065) waslaf feAtedEb'jit@$afet[I'rijsetion due to the testing that was in event. progress at the time of the event (Trafn A7 E.N.n n lfi'n'dlT M in 5 in TEST). With Unit 2 in Mode 5, no ORMA ~ n p e-n Safety Injection signals were required to e g erabl g H the Unit in Mode 5 and both trains of ECC5Jngorma{g aditwo Instrument Inverters e operatl_on{bothtrainsofECCSwouldhaveactuatedon l a safety Injection signal. Undermoreseveregtlal; conditions,(i.e.lossoftwoIn..rumentInvertersin both trains of ECC5 would have"achted,ch'j$4fe[ty(1,njectica signal and resulted in a reactor Mode 1) Thetemperatureincreasein,the\\RC$gould[noty[sve'oCCurredduringnormalplantope trip. in any ande. bystartingtheOCCpumpifnecessary[PrEsugieev'eUco'uldhavebeenrestoredbyresettingthe51 }Nih Y (( relays and restoring Charging. )Th1 The most significant equipment that lost [powe(Srgng}his_eventwasNuclearInstrumentation(NR) (IG) Source Range Channel N31. Because Tralti 81$$P waslJnjiESTE$ourceRangeChannelN32BoronDilution PreventionSystem(BDPS)wasnotavailable%[wer(eadtheho'andthereto,th[shor5 duratio Du Technical $pecification Action Requirement were no safety consequences. In addition count rate was always available d N32}}@ % Y N i f. C W CC.11YE. E l10!!5: p .= x. The 211 Instrument Inverter 1R resistor and 1fU fuse,were replaced. The Inverter =as re-energized. No further corrective actions to the Inverter are planned.=
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.. ~j, b,,:. ~. - y. As interim corrective action, the 2C RCFC low speed f an breaner was replaced under tNR B79257. The failed breaker was sent to Westinghouse to determine the; failure mechanism. Corrective action will be determined after completion of the inspection by Westinghouse'and' doc'umented in a supplemental report. Action Item knord ( AIR) 90-228 is tracking this itee.JjfGjj $NiNlL + An Event (valuation Board was held with Senior $tation Management and the personnel involved with this
- event, the following preventive actions were agreed upon at the board.
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U Gli a s;. LIEENSEE' EVENT REPORT ER TEX ONTINUATION rerm Rev 2. 0_. e FACIL!!Y,NAME,(1) . DOCKET hE tNPEER (6) Page (3) (({ Sequential Revision g// .p% /// Numbgr_ _ Number / hY 0 1 Q j. 6 0J.__0__-Q_l5_._Ql._0_l5. Dyron,_NnlL1 0 ^ Text Energy Industry Identification? it ~6de le. identified in the text as [xx) .g; nW 7p
- ($ngy 1/2005 3.2.1.1.a-1/2 will be ate,an Emergency Enit section. AIR 90-222 is trackingthecompletion[dl[l~rev 1.
[ pgegs d e stening to stress the importance of proper This event will be included} ' ra 'ratordtequ te 2. communication. AIR 90-226"I ct kl' Break (eQ Scar stating to verify the necessary blocks are in 1. Placards have been placed ,eg P ace prior to cycIlng the,Ref eto l p, TechnicalStaffwillInyestigat$the ydk9htroke Instrument Air isolation valve 2!A065 during the 4. Manual 51 survelliance.Y AIR 90-223' ek his item.- kk sh f. ERLYJ M _Q((URRENCES: The following LER "ar ' In a tha.ticomunication breakdowns contributed to the events.',Nh h.M hh Numttr Millis Wi ,M3k, he, pMh 90-009 (Docket 454) (Hissedi5ampi,.e Reju,jred byjagI.noperable Radiation Monitor Due to ~ Niscomunicatiog;and Personnel Error. j $ 5h Con,ta,lnmenQurgeJotgegfp?5
- d. Operable 100 hours Prior to Core Alterations 90-001 (Docket 454)
Due,'to iscosuunicatlon-andilnappropriate Tracking Mechanism. mg 9 v , ?}Q.y fl:- .m- ,,DB F G. Cont 0NLNT rAI M E DATA: y% / Lb> xy; RH00EL pfGfARTf.fD inverter Instrument Bus 213 NOMENCLATURE $; 1:(MRSER NUMBER @i', MANUFACTURER e4 @k'4 ~ Westinghouse Resister ' 9jm-443A322ill7- 'M 50P150' i N/? Gould fuse ., Ni@ $N The RCFC breaker has been returned to Westinghouse or }fal urel analysis. The component failure information will be provided in a supplemental report [ fh n e.,. ~ ~ [ , [ -m. m l
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hh "44 I U l e o una amm so,o ne ... i o m., y ) P L OfilD A f*OWf it 4 6 16.617 ( t i'N A N S llAY l3:HI L-87-207 10 Cril 50.73 i l ? ? t). S. Nuclear Regulatory Commission Atin: Document Control Desk l Washington, D. C. 20555 Gentlernent i r Re St. Lucle Unit I j lbcket tJo. 50-335 i llepor table Event: 87-10 Date of Event: April 14,1987 l fleoctor Trip Due to Loss of l Instrument Buses Caused by Personnel Error i The ot toched Licensee Event Report is being submit ted pursuont to the requirements of 10 Cl R 50.73.o. to ;,tovide notificotton on the sub}cet event. l Very truly yours, (Yth C. G. Woc Group Vi ' resident fluclear i nergy COW /MSD/gp A t iochmc., t I h. I. ! Icison Croce, Regional Administrator, Reglon 11, U5tJRi: cc: Senior Hesident inspector, USNRC, St. Lucie Plant i E JW4 /030/1 p
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LICENSEE EVENT REPORT (LER) aoc ' ' ' *v** ' ' "' st. Lucie Unit 1 015lo10181313I5 ' ! 'I Ul 3 ..c.,,,,...... REACTOR TRIP DUE TO LOSS OF INSTRUMENT BUSES CAUSED BY PERSONNEL ERROR 0 v8410 10 la tga =WesG8 A sid ASPot, DaTe 179 of est A 8 ACitettI8 emvotvt D see
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e,ge ,. g.o.,,j o g i i i I ! l I ! I 1 i 1 i i t I i l 1 ! I i 1 1 I l l I ! I l i un%'= D.. 88 SupFLIuthf ah 80 0e9 e afdCTBD ege ........o% ]......-.. e..e r.r o s, e w :.o es.s. I"] so 1 i 1 ......c....... .................ne. ADSTRA_CT nn April 14 1987, at 1204 hours, St. Lucie Unit I was tripped from Mode 2 (power 1 %. average coolant temperature y,325 degrees F.) due to the loss of 5 two instrument inverters and the subsequent loss of the IMD and IMH 120 Volt AC Dusnes (EIIS:ET). The loss of the AC busses resulted in the actuation of the caenergize to actuate functions of the Reactor Protective System (HPS) (Ells:JC) and Engineered Safety Features Actuation System (ESPAS) ( EIIS:JE). The root cause of the event w&s a cognitive perso nel error on the part of a utility non-licensed operator in following a pl.et approved procedure for the normal operation of the 120 Volt AC Class lE System. Immediate corrective actions included the resetting of all trip signals and the restoration of all affected equipment to the normal operating status, and I the counseling of the non-licensed operator by his supervisor on the need for greater attention to detail while performing his job functions. Longterm corree-tive actions include the revision of the procedure to include a caution statement in the section discussing the removal of Instrument Inverters from service. and a review of this event by the plant Training Department to determine the appropriate training requirements and methods. l l .8705210283 8705148 PDR ADOCK 05000335 %. c..... 5 PDR 4.a
l 21C pensi asas y 6 eiUCtl&2 AIsut&f 03T COMWitS80's LICENSEE EVENT REPORT (i.ER) TEXT CONTINUATION
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' A00 lb 80 8 "4 vta. ',*6 St. Lucie Unit 1 0l0 0l 2 0F 0l3 0 l 1l0 0 l5 l0 l0 l0 p l3 l 5 8 l7 tan.n w =ac w assa smn DESCRIPTION OF RVENT At 1204 hours on April 14, 1987, St. Lucie Uni; 1 was in Mode 2 (reactor power 151, average coolant temperature 2,325 deg. F). Low power physics testing was in progress after a scheduled refueling outage. A Huclear Plant operator (NO) was in the process of tagging the ID Instrument Inverter (E!!S EF) out of service for preventive maintenance. He was transferring the inserter to the IB Maintenance Bypass Bus when he made an error in following the procedure and opened the inverter output breaker for the ID Instrument Inverter, rather than the 18. This resulted in the loss of power to the IMD 120 volt AC Bus. Aware that he had made an error, the operator attempted to rostoro power to the AC bus by transferring the 1D Inverter to the 1B Haintenance uypass aus. The attempted transfer was unsuccessful, and resulted in the loss of the ID Instrument Inverter and the 1MB 120 Volt AC Bus. As these busses supply the Reactor Protective System (RPS) (EIIS JC) and the Engineered Safeguards Actuation System (ESFAS) (EIIS JE), the loss of all power resulted in actuation of all do-energize to actuate RPS reactor trip channels and ESFAS signals. The plant response was as expected for such an event. The ESFAS signals actuated were Saf ety Injection Actuation Signal (SIAS), Main Steam Isolation System (MSIS), Auxiliary eedwater Actuation Signal (AFAS), and Containment Isolation Signal (CIS). Hy 1206 hours, all verified spurious ESTAS and RPS signals had been resets the plant was stabilized in Mode 3 by 1220 hours. CAUSE OF EVENT The root cause of this event was a cognitive personnel error on the part of a ut ility non-licensed operator in following a plant approved proceduro for the normal operation of the 120 Volt AC Class 1E System. His initial mistake which resulted in the loss of the 1D Instrument Inverter introduced additional stress on the NO, and contributed to the making of the second error and the j loss of the in Inverter. There were no unusual characteristics of the work location which contributed to the personnel error. t l i
4 aslCfe,= Mea y 3 eeuCat AA E40VL A102v C0edheite*0ee 1.lCENSEE EVENT REPORT (LER) TEXT CONTINUATION ....ovio ous ~o n eo-e+ tspists s>3 m 'tClut ? *satse on DOCEST8svassan Mi gga asuuesa ses past sp !'W'"
- ' '?'T St. Me e i 0 l 3, 0l1l0 0l0 0l3 0F 0 l5 l0 l0 l0 l3 l 3l5 8l7 stx, m - <. m. m one se smas nn ANALYSIS OF EVENT This event was deemed reportable as per the requirements of 10 CFR 50.73.a.2.1v, any event or condition that results in the manual or automatic actuation of any Engineered Safety Feature (ESP), including the Reactor Protection System (RPS).
Plant response to this event was as expected for the existing plant conditions. The RPS actuation resulted in an uncomplicated reactor trip. The SIAS actuation did not result in ECCS discharge into the RCS because the RCs was at normal operating pressure. The MSIS had no effect because the main steam isolation valves and main feed isolation Valves were already closed. The AFAS had no significant effect because the system was already in use main-taining S/G 1evel. The CIS valve closures did not adversely impact plant operation. All safety functions were maintained throughout the event. There was no adverse affect on public health and safety. CORRKCTIVR ACTIONS Immediato corrective actions included the following 1. All 1:SFAS and RPS signals were verified to be spurious and remot. All oquipment was restored to normal operational status. ?. The NO was counseled by his supervisor on the need for greater attention to detail while performing his job functions. Additional long-term corrective actions include: 1. The mystem operating procedure will be revised to include a Caution Note in the section concerning taking Instrument Inverters out of service. 2. The plant training department Will review this event to determine the appropriate j training requirements and methods. ADDITIONAL INPORMATION All components functioned normally during this event. ? For a previous reactor trip caused by a loss of an instrument inverter, see 1,1censee Event Report 335-82-71. I
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>gg y; w, if's;- Wisconsin cinc; eat gm (414)221 2345 231 W. MICHIG AN.P,0. DOX 2046,MILWAt)KE j' "J ?d ' ' ;:t g Y)l VPNPD-87-511 W ff 10 CFR 50.73 NRC-87-118 'l ge{ 20, 1987 %~i6 Ch November xpy gy Wil 3 ,,w, U. S. NUCLEAR REGULATOR COMMISSI Document Control-Des)d 2 Washington, D. C L 20 ? e
- q; ry' Gentlemen
9 h f 9, DOCKET 50-266 /2 si j LICENSEE EVENT REPORT 87 004 01 h b. LOSS OF THE RED INSTRUMENT + BUS 1 POINT BEACH NUCLEARrPLANT;nUNIT-8 Attached is Licensc% DxPMR ~l@ o? Event Report 004-01 for Point Beach Nuclear Plant, Unitil',ldet' ailing lbs's of the red instrument bus during a replacementibfl batticryjecil. The battery and associated DC bustare common both units. Mb?
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LER 87-004-01 is filed pursuan togi CFR'50.73(a)(2)(iv), "Any l operation or condition' that!resulto'd[iRmanual or automatic actuation of any engineered! safe feature-(ESF), including the RPS' reactor protection;sy@steU ..k We had noted in th@e:origina LERL hathwe, expected this supplement to be filedibi[loctobedl~,%1987. The depth of the investigation has' caused thc7rcpdrtitoRbe delayed. We discussed j the timing with the senior >Rc'sidentBInspector. . &;in. wn m : If any further informatio:ay ngp k +1,.p v. nJisbrcquired; please contact us. 4 Q$d%r@ * ]:kUMg-;;E Very truly yours, .gy; . :. 9: h ) ' dih(h.d.b ' (/ ~ :. w;2; % :/~m m,W". i L c *,rt 'MjyJ;j@Qi[m.*c - 1 4 g M'L i@pjacgisf - C. W. Fay Vice President 4 734 ^ggp.. 5 Nuclear Power % - igfb$s 1 ff Q;h g;(a m4 Enclosure
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.uwge. n wa ..c.uS'o. h. Beach Nuclear Plant j ] ' N., ~ oisIo1o1ol216l6 1 loFl 017 ln occ.. .. u i f 6 f L G 644 M Loss of the Red Instrument Bus %,,w.a ', twtNT Daf t tad Lga tsuMBea tel Strosif DAf 8173 - Of at A # AC6 kite 8 8 earvotWSD WI '8j ,8, heOeifn DAV' ttAn ' 4d ' '"s e Docaat seuwstaisi 6 WOtim QAy Y8AA TSAA n mn 4' 018101010 1 1 I 53 3 0l 5 1l E 8 7 8 l7 0l0l4 0l 1 Ifl pl6 h7 d N/A eisieioio, t, ~ ~ 1 ,,,,R.,,,,, v== ainoai = we-iruo eva uini to v.ea asoua mauri o, is Can s no .-.e , nii N aRJit! M.aStesi { 90.f &ent316ml 73 7t hi a M aatenHlH8 SSMlailli,l D$.f DesHIllel IIIll*l n., i m ..n H., Masai.,- g,,..f p,,,= M.n iaH. 38 asseninitsl gg.33sjggg, pg,f)tsigneesiHal J.6A s M aestsH1Herl $4J 4es)GH44 08.f SelLB16eest tal $$ a05m)f189 4 M.? &asigilmil te fleslGHal Locate.88 Coer1&CT Pos Teens Lea fits 'nd set 1 g gg e owg sovn g g si 1 4 amia COGG l 4 221,,2i 8, li 1 C. W. Fay, Vice President-Nuclear Power 4 i i i i COMPLEf t Oss8 Lue4 Poe SACM COMPoester? Faltuat DescanBSD em Twas Stroef 113# [ 88 ^ l8 "O CauS4 SygT8u CO6aPote n t " '[ "g','
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N' ye,MS s 33C I IElI W111210 r i e i 1 i i a2%V:u t t i ; 9 %ft:4g i .i REPLBaseart AL aepoaf 8 EPSCT 0 011er 4041M CAv v{AA D see use ~ umeno ue as on aAno D 's.o' ,sC,n..-u= _ a, ii., On May 15, 1987, with Unit 1 and Unit 2 shut down, the station battery i (DOS) breaker was opened to isolate the station battery for cell replacement. When the battery (DOS) breaker was opened, the battery charger (D07) caused a voltage spike on the DC bus (D01). The voltage spike affected the associated instrument bus power supply inverters (IDY01 and 2DY01) for both units and the standby (swing) inverter DYOA. This initiated a reactor protection system actuation in both units. The voltage oscillation was cleared within 10 seconds by manual reclosure of the battery breaker. The loss of voltage on the Unit I red instrument bus resulted in a 2/4 power range reactor trip signal from Channels N41 and N43. N41 tripped when its power supply was deenergized. N43 was in trip due to modification work. It also resulted in a 1/2 intermediate range reactor trip signal. The voltage perturbation on the Unit 2 red instrument bus resulted in a 1/2 IR N35 reactor trip signal as well as a 1/2 source range N31 reactor trip signal. All reactor protection system circuitry functioned as designed during this event. An investigation into the circumstances of this event has revealed that without the filtering effect of the battery, the battery chargers will produce voltage perturbations on their respective loads ( the instrument buses ). Plant modifications are being proposed to, p eliminate this problem. 8711300155 8711 8 h h PDR ADOCK 0500 266 S PDR u... +,~_ [' .,\\. = '...., ^ _; _l.. ' '. - f a,,C ; e
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l LICENSEE EVENT REPORT (LER) TEXT CONTINUATION amovio owe no mo-e+ ...... m n l ,ac.uv... i m ooc.. u. u, j " t!.W." t'rJ: I 'saa Point Beach Nuclear plant 015 lo j o l0121616 8!7 01014 01 1 01 2 0F 0 l7 ana-- ..-a .me n. nw mn l i BACKGROUND AND EVENT DESCRIPTION on May 15, 1987, with both Unit 1 and Unit 2 shut down, a cell replacement in the DOS battery was scheduled. The battery was being isolated to replace a faulty cell (#45) which had indicated a downward trend in voltage. Because both PBNP units were off line with their respective reactor trip breakers open, the battery cell replacement was planned to be performed with the battery isolated from the DC bus. The cell replacement was planned with the battery isolated because an additional battery cell had to be removed to gain access to the faulty cell (#45) located behind a vertical support and the other cell. The battery (DOS) consists of 59 individual cells. The cells are monitored individually for degradation. When cell performance is degrading, the cell normally can be replaced without disconnecting the battery from its DC bus. Because the location of defective battery cell #45 requires the removal of another cell to facilitate access, it was considered easier and safer for the individuals performing the cell replacement to disconnect the battery (DOS) from the associated DC bus (D01). Since both units were shut down with their reactor trip breakers open, the opportunity was present to replace battery cell N45 with a new cell without even remotely jeopardizing the power operation of either unit. During a meeting to approve the application of Routine Maintenance Procedure 22, a subcommittee of the Manager's Supervisory Staff discussed the potential effect on the DC bus (D01) and instrument bus (YO1). It was believed that the battery charger would sustain the DC bus without difficulty. Therefore, an explicit statement was not included in the procedure to alert operating personnel that an RPS actuation might occur. Using Routine Maintenance Procedure 22, replacement of battery cell
- 45 was started at 11:13 AM on May 15, 1987.
When the breaker between the battery (DOS) and DC bus (D01) was opened manually, both Unit 1 and Unit 2 reactor protection systems actuated. The battery charger voltage spike blew the fuses in the power supply to both the " master" and " slave" control circuits associated with the Unit 1 inverter (lDY01). This caused a loss of inverter output voltage and subsequently a loss of voltage in the Unit I red instrument bus (1Y01). I l =.....
l '~ ..e.. m ...uci... v6.,o,co..wo LICENSEE EVENT REPORT (LER) TEXT CONTINUATION unovio o s no mon ispiagg 3/3:46 C 'L 'I ' "'#8 (" DQctti asutIB4 A 626 40e suunasta les
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. = - . ace ma4unn The battery charger voltage spike blew the fuses on the power supply line to the " master" control circuit associated with the Unit 2 inverter (2DY01); however, the " slave" control circuit remained operable. Although the Unit 2 red inverter (2DYO1) remained in service, it experienced a voltage fluctuation while the battery breaker was open. The standby inverter (DYOA) operating in a standby mode powering a dummy load was also affected by the battery charger voltage spike. Fuses in the power supply for both the " master" and " slave" control circuits also failed, resulting in loss of inverter output voltage. The standby inverter does not have an automatic transfer capability to either the Unit 1 or Unit 2 instrument bus. PLANT SYSTEM RESPONSES The Unit 1 and Unit 2 reactor protection systems (RPS) responded appropriately to the red instrument bus (Y01) power problems. The nuclear instrumentation system actuations were in accordance with their design. Since the reactor trip breakers were open, no additional systems were challenged. Loss of voltage on the Unit 1 red' instrument bus deenergized power range (PR) channel N41. This resulted in a trip signal from that channel. PR channel N43 had a trip signal inserted due to modification work being performed to upgrade an Appendix "R" source range instrument to meet the requirements of Reg Guide 1.97. This resulted in a 2/4 PR trip logic. Intermediate range (IR) channel N35 was also deenergized, resulting in a 1/2 intermediate range reactor trip signal. A 1/2 source range (SR) reactor trip signal did not occur when N31 was deenergized because the 2/4 PR trip logic cleared the P10 permissive which automatically blocked the SR reactor trip logic and deenergized the SR high voltage. (The IR and PR low setpoint require a manual action to block.) Both channels of SR indication were lost for approximately 35 minutes while the inverter (1DY01) and the red instrument bus (1Y01) were being returned to service. No refueling operations were in progress, and no reactivity changes were made while both SR detectors were out of service. Voltage fluctuations on the Unit 2 red instrument bus deenergized PR channel N41. This resulted in a trip signal from channel N41. Since no other PR instrument channel was in a trip condition, neither the 2/4 PR logic required for RPS actuation nor the P10 permissive logic made up. Thus, no power range trip occurred and
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macp.,. assa u 8 asuCL E AA St ouk A10av COMaossion LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Amo so o~s ao ae-ei$a t aPent $ B r3t M PLC8Leiv names sie DOCEtt 88UMDB A L24 tga seuusik sei P AGE (3' "=; "'JJ,Y: Point Beach Nuclear Plant ol5{ololo12l6l6 81 7 0l0l4 0l1 0 l4 of 0p ,sa -. n < ac o.- xu nn the source range instruments were not affected. IR channel N35 was deenergized resulting in a 1/2 IR reactor trip signal. SR channel N31 was also deenergized resulting in a 1/2 SR reactor trip
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SYSTEM DESCRIPTIONS PBNP has four 125 volt DC power buses. Each bus receives DC power from either a battery, normal battery charger or swing battery charger. The swing battery charger is shared with one other DC bus and is used only when a normal battery charger is out of service for maintenance. The swing battery charger is connected to only one DC bus at a time. Each DC bus supplies DC power to two instrument bus AC to DC power supply inverters. One inverter is associated with an instrument bus in Unit 1 and the other inverter ir associated with an instrument bus in Unit 2. Each DC bus can also supply DC power to a swing inverter which can supply power to either the Unit 1 instrument bus or the Unit 2 instrument bus. The swing inverter is connected to only one instrument bus at a time and is used only when a normal inverter is out of service for maintenance. No automatic transfer (,apability is installed for either the swing charger or the swing inverter. Two of the DC bue.cs utilize Westinghouse battery chargers (130 volt DC, style number 130RF-400), Westinghouse inverters (model number 125CTT 10 kVA) and 1800 ampere-hour batteries (C&D, type LC-25). The other two DC buses utilize Power Conversion Products battery chargers (model number 35-130-500), Elgar inverters (model number 253-1-103), and 1500 ampere-hour batteries (CtD, type LC-21). The DC bus involved in this incident utilize 1 the Westinghouse chargers, inverters, and a C&D 1800 ampere-hour battery. GENERIC IMPLICATIONS None determined at this time. CAUSE Based on subsequent te<:'_ing, the failure mechanism for the westinghouse inverters may have been due to excessive current demand by the ferro-resonant circuit when the bus voltage dipped to less than 105 volts (charger ripple voltage). ..c.o. = g ~ - k_. .? . l ) K ihk ll
anac p.'* 3bea U S 8.uCLl&R 840Wh41DAT COe4adi&&eose LICENSEE EVENT REPORT (LER) TEXT CONTINUATION areaovto o as ao mo+o. gapint s 4t3in eaCskif f asaane na DuGEtt IgualBS A ul 63 a munast a its P A G A 18 8 " 'd ((*h' 88 vlaa nl nl 4 nl 1 0 15 M 0D Point Beach Nuclear Plant 0l5l0l0l0l9lglg g] 9 nn,, a=s.-mamon The voltage fluctuations on the DC bus were sufficient to cause the inverter silicone controlled rectifiers (SCRs) to misfire blowing l the power supply fuses described above and resulted in deenergizing the Unit 1 red instrument bus and caused a voltage fluctuation on the Unit 2 red instrument bus. The cause of the RPS actuation was a direct result of these red instrument bus power supply problems. Operational experience with the Westinghouse inverters has demonstrated that they are sensitive to voltage perturbations on the DC bus. The Westinghouse inverters can operate satisfactorily with voltage fluctuations where the voltage change is such that voltage does not dip below 105 volts and the rate of voltage change is not greater than 11 hertz. The NIS is designed to actuate the RPS when power is lost or momentarily interrupted. The DC bus voltage fluctuations were caused by the battery charger operating without the battery's filtering characteristics. The reason for this was investigated and is believed to have occurred because of either the design characteristics or the adjustments of l' the battery chargers. The battery chargers may not have sufficient DC filtering or feedback response to maintain a steady DC bus i voltage with the battery disconnected from the bus. Ripple on the output voltage of the Westinghouse battery chargers is not specified in the technical manual; however, specifications for the Elgar chargers which have equivalent output filtering capacitors have ripple specifications of approximately 2.6 volts RMS. A check with Elgar revealed that their inverters could probably handle 2.6 volts of ripple with no detrimental consequences. Specifications for a battery charger / eliminator list ripple in the output of approxi-mately 30 millivolts and utilize output capacitors which are i approximately double the size of those in the Westinghouse chargers. I Additionally, the input filtering of the Westinghouse instrument bus inverter is not sufficient to prevent DC voltage perturbations from affecting the instrument buses. REPORTABILITY l This LER is provided pursuant to 10 CFR 50.73(a)(1)(iv), "Any event or condition that resulted in manual or automatic actuation of any engineered safety feature (ESF), including the reactor protection system (RPS)." The Energy Industry Identification System component function identifier is BYC and system designation for the cl.arger is El.
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E sPents 3r3:46 'aca6i, y masse na DQcati enume64 430 kga asuusta 164 F A GE 13' "tW:' T6*3: =n" Point Beach Nuclear Plant 0l5l010l0l2l616 81 7 -- O l0 l4 0 l1 0 l6 0F 0p i rm - -. - -c w . m. SAFETY ASSESSMENT It has been determined that if either battery associated with the DC buses supplied by Westinghouse or blue battery chargers is disconnected from its associated DC bus, voltage perturbations on the associated instrument buses in both Unit 1 and Unit 2 may result. A dual unit turbine runback and possibly a dual unit reactor trip might occur as a result of the voltage perturbations. PBNP is designed and staffed to handle such an occurrence. While loss of one instrument bus in both units might occur until the battery is restored, redundant instrumentation from the remaining three instrument buses 3s available for each unit to safely operate. The loss of SR indication for 36 minutes in Unit 1 did not present a safety hazard. The unit was in refueling shutd: , status with boron concentration of 2181 ppm and all control rods fully inserted in the core. No reactivity changes were taking place, and one IR detector was available for monitoring neutron flux. The safety and health of the public and plant employees were not affected during this event. The ideal time to remove a battery from service presented itself with both units shut down and reactor trip breakers open. In this condition, actuation of the RPS presents no significant challenge to the safe condition or control of the reactor. We know of no occurrence where similar DC bus voltage perturbations have occurred at Point Beach Nuclear Plant. CORRECTIVE ACTION The Westinghouse battery chargers were tested to confirm normal operation. No abnormalities were identified when each charger was supplying power to both the battery and its associated DC bus. When the battery breaker was opened, voltage fluctuations occurred when either normal battery charger or the standby battery charger was supplying power to the DC bus. Additional testing with a resistive load bank has also been done. The load bank was connected to one of the Westinghouse battery chargers in parallel with an inverter. Load was then varied and the DC bus ripple and inverter output ripple were measured. At a load of approximately 70 amps, the DC bus ripple was 14.5 V at 10.8 Hz. At a load of approximately 120 amps, the DC bus ripple was 34.1 V at 10.8 Hz. It was not possible to increase load any further without exceeding the voltage input limit of the inverter and possibly tripping it. . 3. o.. m. .a
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UCENSEE EVENT REPORT (LER) TEXT CONTINUATION w aovo oiseac s'u + n.* n sa,es .u.ui. - i m eos.n aw.esa a
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t 2' :' 7:,*,t: n** Point Beach Nuclear Plant e is ls le le 1216 l6 8l7 0l0l4 0l1 0l7 0' 017 I Normal load on the DC bus is greater than 130 amps. The inverter i output had no noise as a result of the load bank loading. These tests indicated that the cause of the problem is in the design or adjustments of the battery charger and not in the DC system or instrument bus inverters. It was concluded that to improve system reliability, the battery chargers must be modified or possibly adjusted to allow operation when the battery is disconnected. It is currently believed that the voltage perturbation may be eliminated by adjustment of the electronic control circuitry associated with the battery chargers. If adjustmonts will not eliminate the voltage perturbations, modification or replacement of the battery chargers may be considered. 1 G l l 1 l l g,....
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&y.$ jy M ^?g STONEDISON N 6Qy(i!!floadC7d PY75iition 3,$ assachusetts 0236o N k 8d+r g.7. ; .f ,.ga W aa gfQ George W. Davis ,j 7 sen'or vice Pres 4ent-Nuclear, UM April 24 1991 ' s. ' kL%I Q' BECo Ltr. 91- 060 a m,,, /. S$i. .g@, M U.S. Nuclear Regulatory Commiss o 7) Attn: Document ControliDesk 04 Hashington 0.C. 2055
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Dear Sir:
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The enclosed Licensee Event Repor (LER),91$006-00,"HighPressureCoolant Injection and Reactor Core? Isolation TrippedInverters",issubmittedlin'lCoo!!ng!SystemsBecameInoperableDueto accordance with 10 CFR Part 50.73. yMR Please do not hesitate'ts @* contact me , ife7 there;are any questions regarding this report. af@/*% f Mp , W s y @%; ; ~./)w 4% / .. ttQ -( " chy pym. My J g' ,G. . Davis (m.b ,." Myh
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Enclosure:
LER 91-006 IR u y a-Mr. Thomas T. Hartin M d ' cc: RegionalAdministrator,'% Region ~-IT i U.S. Nuclear RegulatoryjCommission-475 Allendale Rd. 4 /; % 2 King of Prussia, PA:'19406 0 m.
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i i i f' i i i i i i i o tu'*Llut NT AL atPon t t arg Cit 0 04s t SW * * + vos'- 04. itsu )e&b&hh,- 9m o sv..uw } v e s no.., ,koj '~ ca.-o. I **r Crto suswssooN onto l l l a.,acCrm.,,......... . ng g g y g On March 26, 1991 whenstartingtheSB'dReactorJRecirculationPump,theHigh Pressure Coolant Injection (HPCI) and' Reactor), Core Isolation Cooling (RCIC) System inverters tripped on high voltage. ;The:HPCI!and RCIC Systems were inoperable for nine minutes. The cause of the inverter 4 trips;wasTa voltage fluctuation that ~ occurred during pump start. The load required'byithe' pump start caused the battery charger that supplies DC voltage to the inverters to overcompensate resulting in a voltage surge. The trip setpoint of the inverters was exceeded during this surge. Corrective action was taken to reset the inverters. An engineering evaluation has been initiated to investigate enhancements to preclude the inverters from tripping as a result of large pump starts. Interim measures will include an administrative change to caution operations personnel that the potential for inverter trips exists. The inverters were manufactured by Topaz Electronics, Model No. 125-GW-125 (60). 8 The event occurred at power operation with the reactor mode selector switch in the I RUN position. The Reactor Vessel (RV) pressure was approximately 956 psig and the RV water temperature was 542 degrees Fahrenheit. The reactor power level was 30 percent. This report is submitted in accordance with 10 CFR 50.73(a)(2)(v)(0) and the event posed no threat to the public health and safety. haC 8 m Jte 4 et 1N + ~u o. a saa.
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OFROvtD oms No 3160 0 04 639 mum 3a mai. v== UCENSEE EVENT. REPORT, LE-fy ',',lj",*,t,',a ',,u"SR,g"o7,jg51,'igO,' a'n','O o TEXT CONTINUATl_0N K T,"",,',*,'o',l;io'."d =',7,"M"^,'M 'u"%'j?"f,i M WW. R iv,1,',,"M t*,",aC M 're'2 6", 5 C 2 lk? dEcr # .........., 2 n d ce uawaosusur ano suocomamosos oc moi emuiv waus m vustR tin uvusin is, race ni " $13 M*,0 n na" Pilgrim Nuclear Power Station 6 i' ,4 2 l9 [!3 9 11 0 l 0l6 ol 0 ol7% ol 4 tune ~. . =m mac u,,, w w nn y e w e EVENT DESCRIPTION ) On March 26,1991at0043houMit BMRiactorRecirculationPump, the High Pressure Coolant Injectib5T(HPCI) dlReactor3 Core Isolation Cooling (RCIC)SysteminverterstrippedfonihighivoltageMThis7madetheHPCIandRCIC Systems inoperable. The Recirculation.umphasibeing; started because of a prior i event that resulted in a lockout of1th emergencyi4160. volt bus A-6 and subsequent trip of the 'B' Recirculation Pump!(see LER The pump was startedinaccordancewithProcedure!2:2!8.4]j91iOO5100jfordetails). (Reyq33)h" Reactor Recirculation System". Theinverterstrippedwhenjthelpupiwasistartedandtheinverterfailure alarms were received at controliroom" anels 03Cfandc904L for HPCI and RCIC, respectively. [ y Mro, Correctiveactionwastakent['rese, over,terslatiOO52hoursattheassociated panels in the Control Room. Failure and HalfunctioniReport 91-103 was written to document the event. The NRC Operations Cent'erJwas! notified as required by 10CFR 10,1991 at 0927shours3]ulh'11stca14was made because at the ti 50.72 on April theinverterstripped,theHPCIdandlRCIC Limiting Condition of Operation was not"ASystems ere<not considered inoperable. A ntere ithe inverters were promptly reset, nine minutes after tripping? !@ M ) M V2 l V The event occurred at power operation $with h ~ eactor. mode selector switch in the RUN position. The Reactor Vesseli(RV)fpressuretwasiapproximately 956 psig and the RV water temperature was 542 degrees (FahrenhelthThe reactor power level was approximately 30 percent, z g% Jg w[.sh. EM CAUSE l w The cause of the inverter trips wastatfluctuation'of4the input DC voltage that resultedwhenthe'B' Recirculation:PumpTwasistartedMAreductioninvoltage occurredonthe4160VACbutt-6due[tojthelload!demandcausedbythepumpstart. This also caused a voltage reduction on the 480V AC buses that feed the battery chargers. The battery chargers supply DC powerfto.the HPCI and RCIC inverters. The battery charger maintains a constant (DCl output provided the AC input does not vary by more than 10 percent. Whenithe; Recirculation Pump was started, the input voltage to the chargers went below its ;10jpercent input voltage margin. With the input voltage reduced the battery chargerioutputivoltage was also reduced. The battery charger responded by overcompensatingifor:the-low output voltage. This resulted in a voltage surge thereby causing;thelinverters to trip. , ! 2..:W The trip range of the inverters was not' sufficient to endure the transient. The inverters are calibrated to trip at approximately:140V DC. Values obtained from plant recorders at the time of the trips.were]l45V and 149V DC for the HPCI and { RCIC inverters, respettively. The.inverterstconvert 125V DC power to AC power for the HPCI and RCIC flow controllers and squarejroot converters. With the inverters tripped, the systems would not automatically' reach' rated speed nor full flow conditions. The inverters were manufactured by: Topaz Electronics Model No. 125-GW-125 (60). o PMN !. N Nl g mm
m s@4 2 5: C E h pg cJow >= wgmusumo miouw couuyo m. UCENSEE EVENT. REP.OR (LER p - n.= q 'St'!,oavgfJU "'nt,'vit,'og"l,y,T," '.= TEXT CONTINUATION, 4 l' c?"37,73,",1,LoMMf',M"s"*/'d? LT "',g,"f5 Mh wA ., j$ o,??a'/A".'Jt",7a'&*/a*J.4mcw~ oc msWWa 0 myg uurotua%r amo eaco.was PLCaLIT Y haua th Q' Jy DCM%I AVMOIR (24 *'WQy? g g g g,9yg g g ig, ,,g g g); (di .([Qhg'[ $ g,, t g,n
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- p-immediatecorrectiveactioRwas folle the Alarm Jesponse Procedures (ARPs) 903C and 904L for the trip of the HPCIf and!RCICjinverters2'The applicable breakers in distribution panels 04 and 05.wereichickeditoiverify 125V DC power was available to theinverters,andtheinverters$MTwerei i
An Engineering Service Requests (9]pi@ reset at.-0052! hou NG 249)g$@Igenera$6 Mgr
- was ted to investigate adjusting the trip setpoints on the invertersforjinstal,linglanlinverter that can accommodate such voltage fluctuations. Inadditiondthelbattery!chargerresponsewithrespecttoAC supplyvoltagefluctuations.islbeing[ evaluated;$An
- Update to this report will be submitted if significant new informatio 'becomes7available.
~ ll$&Q" Qyj%f5R8 Interim measures to be taken_Lincihdera angeitoIProcedure 2.2.84 to caution operations personnel of the.potentialiforlinverteritrips when placing the j Recirculation Pumps in serviceypJheIprocedure[will require the operators to promptlyresettheinverters1asirequiredsbythe!ARP;(kHithregardstothelate notificationtotheNRCOperationstCenterR{lnightforderwasissuedtoinstructthe a operators that when the HPCI and;RCIClinverters; trip'the systems are to be considered inoperable until thefcircuit " TresetMTheappropriatenotifications will be made. mig jhMh $N%e%Q d7 SAFETY CONSE0VENCES 4 i dQNk n[q s M$ Theeventposednothreattothe"public[hda),tgandfsafety. WMMM ' ' Sfp The trip of the HPCI and RCIC inverterstwas heldesigned response. The inverters were reset nine minutes after; tripping %If; theisystems were required to function during the nine minutes, the circuitry 3couldihave.been reset immediately and the systems would have been available Q @ p g; Q P <.x: During the time HPCI and RCIC were inoperable'due to the tripped inverters, the automatic actuation of the Automatic.Depressurization System was capable of reducing the Reactor Vessel pressure foralow, pressure cooling provided independently by the Core Spray System and; Residual' Heat Removal System / Low Pressure Coolant Injection mode. >;y y _ a. The report is submitted in accordance withl50[CFR 50.73(a)(2)(v)(D) because the HPCI and RCIC Systems became inoperable g% 1{ Thb-4 l) ? f... + e g ~ ,4 dd$ T~ n 'N i 2 s lV i i -
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- 0l0 0l4 or 0 l4 rt = ~. m. m,, m.m,,. 2 6 SIMILARITY TO PREVIOUS EVENTS ? A s.:. A review was conducted of Pilgr Stat on censee vent Reports (LERs) issued since January of 1984. The!~reviewfo inoperableduetotripped. inverter (siW.cusedfonlLERs'hereHPCIand/orRCICbecame Jhehevjewirepealed one similar event. LER 85-029-00 involved receipt ofathe*HPCIlinverterJcircuitry failure alarms and an ATHS trouble alarm in the Control 3 Room?WItTwaifdetermined that the HPCI inverter ^ and the breaker feeding the ATHS! inverter /hiditrippedu Immediate corrective action was to reset the HPCI inverter and?ATHS!breakegRtTheicause of the trips was s determined to be a fluctuatjonlofftheppu.tIDCivoltage;; WsWXE H ENERGYINDUSTRYIDENTIFICATIONSYSTEHkEIIS)? CODES , n._ ~ MCM The EIIS codes for this report >:arefas lows as g m COMPONENTS .M> ' u, CODES se 5 D M pg.b.Mp yINVT Inverter Charger. battery yg jBYC SYSTEMS ~ h .) !10 p High Pressure Coolant Injection (HPCI ystem p BJ Reactor Core Isolation Cooling (RCIC ste ~ BN ~ Low Voltage Power System (480V AC)M*- EC DC Power System %ig EI 6 ,h 16 L 4$4d
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Reference:
N DocketNo.50-336Ssg!RO-50-336/82-002 M hp Reportable Occurrence ,.e y&y#*+gy&p.;,, a 9> syn w ~ ' $n$ g;n$ g,4 p+M,.,J x,.. ,8 d f
Dear Mr. Haynes:
4p .w ,,.,;K -- Mb; ~ .,M.e y. s n 'n% This letter forwards Licensee, Event,Repopt:82-002irequired to be submitted within thirty days-pursuantitojMillstone Unit 2 Appendix A Technical Specifications (Section.6!9.1179.b,3 conditions leading to operation in a degraded mode permitted;byla.il,imiting condition for An additional threelcopieslofi@the; report are enclosed. operation. 4 e; g ! 9 5 M %M y & @ @' @ g ' Yours;truly,g w@gWg '40RTHEAST NUCLEAR ENERGYiCOMPANY ',, yy' m;g# c Q5 ,fff.g? <?g/ roczka# #Nis Static Superinten ents c Millstone Nuclear Power Station l1% R"C %~*'Q- }f & EJM/SS:mo ^
Attachment:
LER R0-50-336/82-002~ 4 cc: Director, Office of Inspection'and Enforcement, Washington, D.C. (30) Director, Office of Management'Infortration and Program Control, Washington, D.C. (3) .2 .Y U. S. Nuclear Regulatory Comission,1c/o. Document Management Branch, Washington, D.C. 20555 'i f ", i ,J Ll;, ], ( m;(- \\ y.3 y :;, y 4y'.,' W $.p'h 4 W n s.w m . vj ) .;y:%g 7 8202090000 B20201 ~ W: E %,. PDR ADDCK30000336 - 11} r SM/6 $ (,pO S My o pd 4.jilp;43 i/
,u w , A shn M WhuRM L A W N n QSEE EgNREgTgf; N werwrbw3M4' i.ON inot ; i rf g ( ~j {~~j ~ j l".Yl - o, ?N EASE P INT On TYPE ALL ntouinID INI ont.1ATiofd J.h 9 7 $@ U) l. CJ J1_IQtij S j 2 j@[0_[01 -j,110 l%@ i -al 10))@l 4l 1l 111l 1JOLLJ O a 9 i u r v.. i not i4 is ,uctm,eNUuutsw w e m to uuN;c nPc a si c^r w l_t )@[ Oj 5j 0 l 0 l 0 l 3 l 3'i'.6fl@u:sbl/lhh:0i]6 tl8~ l2 l@[1 [ k P4' T lO E ['$2r a is ntn,HvoAit na Doce.u Nuuuta g.y ugp cvtNtoAis; m co u rvrNT ra crun noN ANo rnocAnLE CONSEOutNCES hhh{ J3] l With the plant shutdown for refueling,$uses . inverters 2 and 6 blew resulting ..J kh of the overpressure circuit I jj] dn loss of vital instrument pa e guqM* WWce qi] l for shut down cooling which c%dWs@@72-Sl'-652)to7close. ausedivalve Shutdown coolinnas ._j .ve% W M@eMppt 3 j s] [ restored in 7 minutes. The plant vastoperatedlin,accordance with T.S. A.S. 3.9.8.2, _ j tio similar LER's Mg y gd g bm m _ _ _._ _.J .~w-l tnS M , J rq l during this event, jaggg 9% N U k MIIQd M b d q J} [. -.. _ _j is tNfW6W.h W j 3j3] l C AUSE CAUSE _ I G,,'. t COMP. VAtyt I,UuCODE SUSCODE .,( GDE CODE $UuCODE ** s OMJ COMPOf4EN1 GoDE L
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- transferred from inverter 2 to inverter 6MA;1args current flow resulted, which I
gji my ;w+sw blew the fuses in both invertersM A' departmental memo was issued to all tech-I ,,,)g wpmw:. cam y nicians to check all leada priort to connection'.to' test equipment. j , j.,, l ~ %pygg g 1 'rra l w B 9 % Poe.t H 01HERSTATUS ILC HV DISCOVluY I.;t scruP 1:0N Yi AS 5 ijyj (3J @ In lo lo l @ l f!A l ljLj@l Operator observation l e3 H 9 10 12 11 44 g 45, 46 Ac tivi t y (ON1LNT } L OC AliUN Ur HE L E ASE
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- ATTACHMEigq0]LER182-002i h
Event Description and probablhConsequences FC c$$ l . MMHMS ' During refueling operations a"preventat ma ntenance test was being performed A short citiUit tistVeadwasconnectedtotheprinted / on static switch VS2. circuit card of the switch.idThisfcaifsedt icomponent on the card to fail and a silicon controlled rectifier!(SCRjf2){toi ecomeTcontinuously gated. The load on Inverter #2 was transferred:todg(ver.terJf6bbut because SCR #2 was on con-tinuously the inverter outletshere connected;iniparallel. Due to frequency differences a large current 1 flow;resultedifromicach inverter causing the input which in turn caused contact closurelin}ditallinstrument AC in panel # VA fuses to blow. Thisresulted;inDossiof This caused valveL 2-Sl-652f to,@. clos l overprotec the eR$' cooling. .9 %n % i M Q fl Health physics was notified to~ evacuate;the%$ & M~k
- steam generators. Maintenance was notified to assist in opening 2-Sl-652yf%fMW M:f@ijdjOW% +
The loss of shutdown cooling prevents 1 decay. heat removal from the reactor vessel. Since shutdown cooling was? restored 51nialshort period of time, i.e., seven (7) minutes and reactor'coolantitemperaturelincreased from 84 to 114 F, there was no significant effect;on1the p%%4lantR$Nik . 9nyg%:p Cause Description and Corrective Actions 3$fd@6 The cause of this event was a short.circuitedI e&t stilead used in performing a preventative maintenance test. SA' departmental; memo has been issued to test technicians to check all test. leads for'short' circuits prior to connecting to N ~,g;g i [Mfgggn;;gd N$MMfbh!R equipment. pWm ',' & bAf &w ; m w?]&l W s am hkh y.v;tyx pp,. a. e4Qp;w w" W l?W,Hi %gygg gbg.i:- .q.w y, [ 4 s 3 + it ' s a 1 -q. ?.$ *% ,,J.p I,4 f.&j?f ff g Nfh.x.xkhhi'I Yf 4 5k pfg$hf$wi$k M 4
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' ' i. W - 3 jef - At 0230 and agaik fid65e @ 4 o 9:.tkbormalandRemoteShutdown Panelcontrolandjin$j@en.tttion2D fes/fortheReactorCore Isolation Cooling:S t A m eu 0 -.o i .No II. CAUSE wm -Q ,dg% Power is supplied = to;the; dis}fyomJl25J 0jBusT211Y, through a Topaz Reactqrj ore] lation Cooling System (BN) instrumentation andjon't'r ,{highvoltagetripwhichturns the inverter off whenithi}has W rotect Inverter. This inverter e 15p6t.$6 ~i$' Ghitp l34.8 VDC and 130.5 VDC s reaches 147 VDC. The setpoint ofthehighvoltageltrihihid[di dido for the normal and Remote Shu3do h{aneginyertersrespectively. 3 At thetimeofthiseventg(@a,tt argpvasjinprogressonthebattery which supplies thesejinverger,s, heg harge progressed, the battery voltage drifted up toitho[poig at theshighivoltage trip point was exceeded and the inverteisduged[oj[$'hin[the charging voltage was adjusted down to normal ~. hejinve es'at1 automatically and operatednormally..jg g- 'Mg 97 7 III. PROBABLE CONSEQUENCES'0FITHE OCCURRENCES mem n '24% power. At the time of this eventlytaSalle t The plant 'was wasmaintainedina" safe {oy}srh't'iEgthe other ECCS sys hiidifi6nf because the HPCS,(BG) and Eia p Q "S e ~cgyC E s g. 4.; a. 7 4.4 % e . gyg .c .c 4 IV., CORRECTIVE ACTION i ffi V y.: $ ~.. D..MJ (' 3Q S Q G'.T..
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~ The battery charging voltage:vas?a'djusted down and the inverters reset. WorkRequestL37308waswritted[E0herify;thehighvoltagetrippoint setting. The settings weretfoundfitI134,8 VDC and 130.5 VDC and were reset to 147.0 VDC and 147.2 VDC'for ttie' normal and Remote Shutdown PanelinvertersrespectivelyMA} survey lofallsafety-relatedinverters l will be performed to ensure thatithe high> voltage trip setpoints are periodically calibrated (Action 11temjRecord 1-84-67090). ' y&c % w %b r .",gh?Mp W. e V. PREVIOUS EVENTS .., ygjgK ' hkyk No previous events have been; reported s,, ' ypy&~hyRMQf,;f a HMP k- -A @i4fQNM..N L ^ Y;, +dk% '* > M \\
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- i W4'W*W f - Q;y, K& a 364 "U p3, j _- U s NUCLt AR RIGut ATomv CowMlE8foN }} $ arenovgo ous No Jit04n04 LICENSEE EVENT REPORT (LER) ' ' ^ " " " " " ..C,u r,.... n i Turkey Point Unit 4 ooCu 1.u.. in, t ..a a o [s ; o j o j o l 2 5[! l1 lorl0 l 4 1 n r a... Engineered Safety Feature Actuation - Reactor Trlp ive=1 o*t.
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. ACstef v hawes Dmit%vhtW' Turkey Point Unit 3 o i s ; o ; o ; o ; 2, 5; O 0 j6 2 l0 85 3l5 0l1 l7 - 0l0 0l7 2l2 Sl5 N/A 1 o,s,o,o,o, I tais mercar is sutwirve o,umsuant to vi.e as ovianuswrs on to Ca n 3,ca.ra .,--. < ea. *.<,.--,, nis o,,, u,,a n aorm .om a n amnn. amw A u.mwm so n wnin e n vnw u.rm nen. i n a inne so.m m m u.ru nan.c ) "?.P 1, 0,0 n m., o, r a, ..u.~, ~ n aumam u.tw nxua u n winn,wnsi - L':o '" " ' "' ' ' n m.- n., . nma., . n.n,u n., n mamn.i sonanrn.m so.rmn n. Lactesgas CONTACT Pos TMia Lin H3) Navt T E Lle-ONE NvMet R aat A CDOt R. D. Hart, Licensing Engineer 30;5 2;4i 5 ,2;9lI 0 j i t CouPt ITE oNG tiNI fon (ACM conseroa gNT PAstumE of temisto its TMtf Af roaf itz, Ca 55 systew coupogr%, w ,ac P f *Om r 8 [ v g y C ore f hhh X E,F F,U, S,1,5, 6 Y X E,F F, U, 5,1 i i5 6 Y N X E;F F;U; ; 5 ;1;5 ; 6 Y j g, i l o............ o o. .o,. o.. ~} n n ura. a.,. r.orcruo suomou carri T} e,o l aEi.C,u,. .=3u. ..'$~,*. 5,. .. ~.,ne vent: On ne 2 195 Unit 4 experienced a reactor trip from 100% full power. The 4C inverter that was in service supplying power to the '20 volt vitalInstrument panel 4P06, tripped de-energized level controller LC-460C and the pressurizer spray valve controllers (causing the Loss of 4P06 valves to.e nain at their list demand position). De-energizing of LC-460C generated a false indication of lov pressurizer level (less than 14'16) which de-energized the pressurizer heaters (contro and backup) and initiated letdown isolation. operation of power operated relief valve (PORV), PCV-4-455C. Loss of 4P06 also resulted in the lo PORV PCV-4 A56 was avail 6te with its associated block valve, MOV-4-535 closed due to slight leakage through PCVA-4 56. conditions, along with a turbine runback due to loss of power to nuclear instru nentation system These channel N-41, resulted in the reactor coolant system pressure increasing until it pressurizer high pressure reactor trip setpoint of 2370 psig resulting in a reactor trip. reached the Cause of Event: The loss of the 4C inverter occurred while attempting to energize the 3C inverter onto the 3B 120 volt UC bus. inverter's charging capacitors prior to energizing the Inverter onto the bus.The proce This step was not executed causing the loss of the 3C inverter which resulted in a DC bus transient which in turn caused a loss of the 4C inverter. Corrective Actions: The following corrective actions were taken following the event:
- 1) Power to the vital instro nent bus for panel 4P06 was re-established and the af fected equipment was returned to normal lineup.
- 2) The 3C and 4C inverters were inspected and checked as per maintenance instructions and no i
significant problems were found.
- 3) A post-trip revie.v was comp!cted and no abnormal operating conditions were identified. Follo completion of necessary reoairs, inspections, and testing, the unit was returned to full power at 9:00 p. n., on June 23,1985.
- 4) The long ter n corrective action will be to replace the inverters with a model of a different'.
enanu f acturer. Replacement of the inverters is currently scheduled to begin in July 1955 The health and safety of the pub!!c were not affected. c -etmemom nm72P
M @_ @ mMM" o 4RCf e 3644 -Q f %;;7% t, U S %UCL L A85.iGUL A f gat y CowMd$ son LICENSEE EVENT REPORT (LER) TEXT CONTINUATION .m ma v i s.~. 8 ACILit y 4Awa na ,.%, e n DOCht! %./M&ts4 ut ten w watn 6 l PAGEir %g;r ! M Turkey Point Unit 4 o [s [o l0 lo l2 l5 l1 8 ; 5 0 l1 l 7 O ; O l0 l 2 0 l4 or m, - u-.. n, _ sac n wmm Event: a-On June 20,1935, at 3:13 p.m., Unit 4 experienhed a reactor trip from 100% full power. the 4C inverter that was in service supplying power lto 120 volt vital instrument panel (4P06)At 3:17 p.m., Loss of power to panel 4P06 resulted in a~1osstof power to the nuclear instrumentation system (NIS) , tripped. power range channel N-41. This channel generated a "NIS Rod Drop" signal which generated a turbine runback to 70% power. Loss of 4P06 de-energized level controller LC-460C and the pressurizer spray valve controllers the spray valves to remain at their last demand position, approximately 10% open). LC-460C generated a false indication of low pressurizer level (less than 14%) which de-energize pressurizer heaters (control and back-up) and initiated letdown isolation. the loss of automatic operation of power operated relief valve (PORV) PCV-4 455C. Loss of 4P06 also resulted PORV PCV 4.456 was available with its associated block valve, MOV-4-535, closed due to slight PCV-4 -4 56. leakage These conditions resulted in the reactor coolant system (RCS) pressure increasing until itthrou reached the pressurizer high pressure reactor trip setpoint of 237 pressurizer spray valve contro!!ers. At this time, a RCS cooldown was in prog,ress due to the feedwater transient experienced as a result of the loss of the 4C inverter. The main steam isolation valves (MSIVs were closed at 3:20 p.m., to help reduce the cooldown. ~ The "B" and "C" reactor coolant pumps (RC were stopped at 3:20 p.m., to stop the pressurizer sprays effect on RCS depressurization. "C" RCS loops supply flow for the pressurizer spray valves. Pressurizer pressure recovery was initiat The "B" and pressurizer heaters were energized by holding in relay LC-460CX and the "B" and "C" RCPs were subsequently started. Loss of 4P06 also caused the "A" steam generator (SG) feedwater level control to transfer from automatic to manual remaining at a demand setting of 100% feedwater flow. Loss of automatic level control along with continuous feedwater flow resulted in the "A" SG level increasing until it reached Hi-Hi level setpoint (80%) which tripped both SG feedwater pumps at feedwater isolation signal and an automatic start of the auxiliary feedwater (AFT) pumps w 3:19 p.m. This resulted in a feedwater to the SGs until the "B" SG feedwater pump was started at 3:32 p.m., establishin feedwater train to the SGs. 4-ONOP-003.6, " Loss of 120 V Vital lastrument Panel 4P06", a regained. s At the time of this event, the "A" emergency diesel generator (EDG) was out of service (005) fo preventive maintenance. Therefore, a cooldown was commenced for Unit 4 at 5:55 p r power operation. p.m., when the "A" EDG was placed back in service. The unit was then returned to hot standby. Af ter the trip, the containment airborne activity levels recorded on the process radia instrument R-il (particulate) reached the high activity alarm setpoint. nis caused containment ventihtion to isolate and control room ventilation to switch over to the recirculation node as de vital instrument panel (4P06), tripped.At 6:03 p.m., while cooling down Unit 4, the 4C i The loss of power to 4P06 resulted in a loss of power to the nuclear instrumentation system (NIS) power range channel N-41, intermediate range channel N-3 source range channel N-31. Loss of power to NIS channels N-31 and N-35 generated a source ra flux (N-31) and an intermediate range high flux (N-35) reactor trip signal. The reactor trip breakers were closed prior to the event, due to the performance of Operating Procedure 1004.2, " Reactor signal, thus completing the reactor protection system logic. ONO panel 4P06 with the 4C inverter at 6:35 p.m., and the lost instrumentation on Unit 4 was regained. ..c,.w..
.,. ~..,;.. e EbhSN[ ),afg- =- wwww tu -o . ~ i -~ UCENSEE EVENT REPORT (LER) TEXT CONTINUATION accenuw w "+ - MM&&T. o~n n - P ActLI,v N Anat att DQCEtt hWM88 R (2) 3 (In huusinis, Pabl Ib k " fC. "'M l a y Turkey Point Unit 4 o ] S T O M fo @ }5 @ ;8 l5 -- O f l7 __. 0 [0 0 l3 0 l4 or n u %. u. .,.~ oc %.m.n m ,,.ww Cause of Event: , M WQ$ W -. a%W J/ The loss of the 4C inverter occurred while attempting to energize the 3C inverter onto the 3B 120 volt DC bus for post maintenance no load check 00thThe 3C Inverter had been out of service as part of a preventative maintenance task action planitdnhance the overall reliability of the inverters. The procedural requirements for this evolution require charging the 3C inverter's charging capacitors prior to energizing the inverter onto the bus. This step was not executed, thus when the 3C inverter DC input breaker was closed onto the bus, an overcurrent condition occurred which tripped off the 3C and 4C inverters. A Chemistry Department evaluation for a similar event revealed that the two most probable reasons for the high R-ll readings are: 1) Particulate matter inside containment was shaken loose due to the reactor trip, 2) A possible system gas leak in containment is leaking gas into the containment which is decaying to a particulate. M = -. The cause of the second loss of the 4C inverterhas'due' to a loss of the 3C inverter. i The cause of loss of discovered the 3C inverter could not be positivelyLidentified. The most probable cause is attributed to a ground discovered on the 3C normal containment cooler. '
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Analysis of Event: g A post-trip review was performed to assess the proper operation of safety-related equipment. inverters were inspected and checked, they were placed back in service and run for a period of 24 hours Af ter the without any mode change on either u11t to ensure reliability. During this period, Unit 3 was in cold shutdown and Unit 4 was in hot standby. No problems were encountered during this period. related equipment were verified to have functioned as designed upon actuation of reactor protectionOthe system features. Similarly, the post-trip review established that the transient behavior of pertinent plant parameters for the reactor coolant system and SGs responded as expected of a reactor tri kind. Specifically, the RCS pressures and temperatures were deterrnined to be following an expect pattern based on the conditions leading up to this transient. Based on the above, the health and s the public were not affected. w, Corrective Actions: ~ The following corrective actions were taken: 1) Power to the vital instru nent bus for panel 4P06 was re-established at 3:35 p. n., from the 4C inverter and affected equipment returned to normal lineup. 2) The groond on the 3C normal containment cooler was cleared. The 3C and 4C inverters were inspected and checked as per maintenance instructions. For the 4C inverter, the DC input breaker trip settings were changed and fuse F6 was replaced. replaced and the 9C input breaker trip settings were checked.For the 3C inverter, fuses FL and F6 w the original sequence of events related in the "Cause of Event" above were repeated.Upon co inverter did not trip, thus verifying that the changed DC input breaker trip settings eliminated the The 4C interaction between the 3C and 4C inverters. 3) Af ter suf ficient time had elapsed to allow the activity to decay off, containment and control roo n ventilation were returned to its normal configuration at 7:50 p.m., on June 20,1985. 4) A post-trip review was completed and no abnormal operating conditions were identified. Because of the second loss of the 3C and 4C inverters, Plant Management decided to inspect the twelve inverters and check the DC input breaker trip setpoints and then run the twelve inverters for 24 hours with no change in operating mode for either unit. 24 hour period. No problems were encountered during the
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$$DN " t W." 1 NG "aa Turkey Point Unit 4 ENMN# o is lo lo io l 215 l 1 si 5 01 11 7 o lo o 14 o' oF n e r -. u-., m + xu nm g y m, ; ((Jd, % . w.;. - _j '/'* Corrective Actions: (continued) A[]UME.,[f[, Change /'dodification (PC/M) 85-103@wasNcompleted for Unit 4
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to modily the logic for initiating a turbine runback caused by"a#ni$3tivs'ilux rate input (NIS signal) from I out of 4 channels to 2 out of 4 channels. This logic. willi prevent challenges to the reactor protection system by not initiating a turbine runback due'to'a' single NIS channel " Rod Drop" signal. my@s$,%. ~ -
- 6) Upon sucessful completion of the 24 hour run onithe Inverters, and impic.nentation of PC/M S5-103, the umt was returned to full power operation 'at 9:00 p.m., on June 23,1935.
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- 7) Of f-Normal Operating Procedures (ONOP) 003.6;-: 003.9 were reviewed and revised to clarif y immediate and subsequent operator actions to be;taken upon loss of a 120 volt vital instrument panel.
As part of this revision, a method of closing, the pressurizer spray valves was included. The.,e procedures also include a list of control [ room indications lost l j on f ailure of a vital Instrument panel.
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.2 S) Training was given to the operating shif ts'on the' revlhions to ONOP 003.6 - 003.9 by havmg the Plant Supervisor - Nuclear review the revised procedures with his operating crew. i . NM'
- 9) PC/M S4-210 was previously implemented on Unitl3 to provide redundant rod position indication (itPI) signals into the turbine runback initiating logic. : This allows disabling of the NIS signal to the turbine runback logic.
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- 10) An evaluation will be performed on the existing' breaker coordination of the 120 volt instru.nent l
AC system. g 9pg: ~ l l) The induidual involved in the event was counseled "to exercise greater care in using plant procedures. 12)The long term corrective action wi!! be to replace the inverters with a model of a dif ferent manuf acturer. As part of this replacement, a.. regulated 120 volt AC alternate power supply (constant soltage transformer) for each of the eight(S) normal vital inverters will be installed Each replacement inverter has a static transfer switch that will automatically transfer the load to the alternate power supply upon loss of a normal inverter, to allow transition time in manually switchind over to the spare inverters without inducing transients in the sital AC power system. implementation of this replacement enhances plant safety as the a<ailability of vital AC power is unproved. lleplacement of the in<erters for both Units 3 and 4 is currently scheduled to begm in July 198 5. Similar Previous Occurrences: LER s 250-S 4-003, 250-34-014, 250-34-026, 251-34-011, 251-34-021, 251-34-022, 251 -8 5-012, and 2 51 35-013. l l l L I i )...... j ~ Y l W&_ j _}}