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| {{Adams
| | #REDIRECT [[05000423/LER-1996-036]] |
| | number = ML20210J798
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| | issue date = 07/25/1997
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| | title = :on 960926,discovered That SR Valves Were Controlled by non-safety Equipment.Caused by Design Error. Upgraded Qualification for 38 Control Circuits
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| | author name = Smith D
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| | author affiliation = NORTHEAST NUCLEAR ENERGY CO.
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| | addressee name =
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| | addressee affiliation =
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| | docket = 05000423
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| | license number =
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| | contact person =
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| | document report number = LER-96-036, LER-96-36, NUDOCS 9708190060
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| | package number = ML20210J793
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| | document type = LICENSEE EVENT REPORT (SEE ALSO AO RO), TEXT-SAFETY REPORT
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| | page count = 5
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| }}
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| {{LER
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| | Title = :on 960926,discovered That SR Valves Were Controlled by non-safety Equipment.Caused by Design Error. Upgraded Qualification for 38 Control Circuits
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| | Plant =
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| | Reporting criterion = 10 CFR 50.73(a)(2)(vii)
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| | Power level =
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| | Mode =
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| | Docket = 05000423
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| | LER year = 1996
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| | LER number = 36
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| | LER revision = 0
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| | Event date =
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| | Report date =
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| | ENS =
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| | abstract =
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| }}
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| =text=
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| {{#Wiki_filter:_
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| NHC f OHM 3b6 U.S. NUCLE All HE GUL A10HY COMMISSION APenovtD BY oms No. 31bo 0104 (4-9b)
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| EKPiMEs 04!30/9a 2'#@&Ma*#4JMinsP!fq'aN#21''ntM
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| *""J"?,r i'AM 'L'NtP**'M',e '"te
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| 'M LICENS!E EVENT REPORT (LER) m,d't 'gfrt,'t,q'mlt*~,,'%~,'ag f#u % ' E UP4'l % d'fs004"#i! M P"'*
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| (See reverse f or required r' umber of i
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| l digits /charat.tersfor each block) f ActLi1Y hAMG tti pocElf0#UM M a m PAut (31 Millstone Nuclear Power Station Unit 3 05000423 1 of 5 TITLt 141 Safety Related Valves Controlled by Non Safety Equiprnent EVENT DATE (6)
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| LER NUMBER (6) liEPORT DATE t7)
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| OTHE R F ACILITIES INVOLVED (B)
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| MONTH DAY YLAR YEAR SEQUENilAL REVISION MONTH DAY YEAR
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| # Aciuty NaMt occatt e uwMn NUMBER NUMBER 09 20 90 90 030 02 07 25 97 OPl.R4 TING 5
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| THIS REPORT IS SUBMITIED PURSUANT TO THE REQUIREMENTS OF to CFR 61 (Check one or more) (11)
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| MODE 19) 20.22011tn 20,2203(aH2Hv) 60.73(aH2)h) 60.73(aH2Hym)
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| POWER 000 20.2203(aH1) 20.2203(aH3)D) 60.73taH2Hiu 60.73taH2Hal LEVEL (10) 20.2203(aH2)h) 20.2203(aH3)Ds) 60.73taH2Hm) 73,71 20.2203(aH2Heu 20 2203teH4) 60,73(aH2)Dv)
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| OTHER
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| ~
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| 20.2203(aH2Hm!
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| 60.36(cH1) 60,73(aH2Hv)
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| Specif y in Abste et below
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| ~
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| 20.2203(aH2)bv) 60.361cH2)
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| X 60.73(a)(2Hvii)
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| LICENSEE CONT ACT FOR THIS LER (12)
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| NAMI flLLPMONE NVMMR lt'*clude A*es Coost David A, Smith, MP3 Nuclear Licensing Manager (080)437 5840 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)
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| CAU$t
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| &YSitM COMPONEN1 MANUF ACTUHER Nt POHT ABLE CAust sYsT(M COMPONE NT M ANUF AC TURE R hie U.NesLL to NPHDs To NPRDs SUPPLEMENT AL REPORT EXPECTED (14)
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| EXPECTED MONTH oAY YEAR YES X NO SUBMISSION Of ves, complete EXPECTED SUBMISSION DATE),
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| DATE (15)
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| ABSTRACT (Limit to 1400 spaces. l.a., approximately 16 single spaced typewrittenlines) (16)
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| On September 20,1990, with the unit in Mode 5 of an extended outage,whde performing an engineering evaluation,as part of the ConfigurationManagement Program, an initial determinationwas made that the High Pressure Safety injection (SlH) and Low l
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| Pressure Safety injection (SIL) systems were subject to degraded performance due to possible mis position;ngof normally closed s fety related air operated valves. Mis-positioningof these 21 valves can be postulated to occur under post accidentharsh l
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| $nvironmentalconditionsdue to failure of non quahfiedpower and controlcircuits. As a result, the potential diversion of SlH and/or SIL flow under accident conditions may be more than the margin allowed within the Loss of Coolant Accident analysis. This condition was reported at 1434 hours on September 26,1990, pursuant to 10CFR50.72(b)(1)(iii)(D) as a condition that could hiive prevented the fulfillment of the safety function of a system needed to mitigate the consequences of an accident. Subsequently,17 Cdditional air or solenoid operated valves have been identified within the following systems: Reactor Plant Component Coohng Water, Containment Vacuum, Reactor Plant Samphng, Post Accident Samphng, and Main Steam to the auxiliary feedwater steam turbine.
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| The cause of the reported condition is a design erro.. The initial plant design did not adequately consider the potential mis.
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| positioning of these valves under harsh environmental conditions or active failure. This design error appears to have been a risult of inadequate interface between the Nuclear Steam Supply System provider and the Architect / Engineer (A/E) or internally within the A/E organization.
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| The potential for valve mis positioning or active failure and the resulting effects has been evaluated. The quahfication for 38
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| :- vilve control circuits will be upgraded, or other corrective actions taken to prevent mis-positioning.
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| 9708190060 970725"
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| ~
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| PDR ADOCK 05000423 S
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| PDR.
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| LRC 70;M 366A U.s. NuCLI AR Rt ouLAloRY Commission LICENSEE EVENT REPORT (LER)
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| TEXT CONTINUATION r ACilliV NAMi t1[
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| DOCKET NUMBLR(2)
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| L ER NUMBER (6)
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| PAoL (3)
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| YEAR sEQutN'IAL HIVibiON Millstoiio Nuclear Power Station Unit 3 05000423 NuvetR NuwmL 2 of 5 90 030 02 TLKT l11more weceis requued. use additions /cornes of NRC form 366Al l17) 1.
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| Qascription of Event On September 20,1990, with the unit in Mode 5 of an extended outage, while performing an engineering evaluation as part of the Configuration Management Prograrn, an initialdeterminationwas made that the High Pressure Safety injaction (SlH) [
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| I and Low Pressure Safety injection (SIL) systems were subject to degraded performance due to possible mis-positioning of normally closed, safety related, air operated valves (AOVs). Mis positicmngof these valves can be postulated to occur as i
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| a r:sult of possible failures related to non-qualified power and control circuits. The 21 valves originally identified to be l
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| cff:cted were the accumulator fill valves (3SIL*AV8878A/B/C/D), the SlH to Sll isolation test valves (OSIL*AV 8872A/B/C/D), accumulator test line isolation valves (3SIL*AV 8877A/B/C/D), accumulator injection fill line isolation valves (3SIL'AV8879A/B/C/D), charging pump test supply isolation valve (3SlH*AV8882), SlH pump hot leg test line isolation valves (3SIL'AV8889A/C), and the Residual Heat Removal syrtem pump hot leg test line isolation valves (3SlH*AV8889B/D) Spurious operation of these valves, coupled with assumed failure of the non-safety related downstream piping, could result in diversion of Emergency Core Cooling Systam (ECCS) flow.
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| Tha 21 SlH/SIL valves are controlled by momentary contact push-button switches in the control room which include a s::1-in circM utilizing contacts from the valve limit switch located inside containment. The associated control circuit penetrations utilize terminal blocks inside containment. These configurations are not currently qualified for post cccident harsh environmental conditions. This circuit design could result in spurious valve operation due to the effects of harsh environmental conditions following a design basis accident (DBA).
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| Following initialidentification of this condition, it was reported at 1434 hours on September 26,1996, pursuant to 10CFR50.72(b)(1)(lii)(D). It was also reported, pursuant to 10CFR50.73(a)(2)(vii)(D) as a condition that could have pr; vented the fulfillment of the safety function of a system needed to mitigate the consequences of an accident.
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| S2venteen ad(tional safety related air operated and solenoid operded valves have been identified where failures of non-qualified control circuits could degrade performance of a safety system function These additional valves are located in the following systems: Reactor Plant Component Cooling Water (CCP), Containment Vacuum, Reactor Plant Simpling (SSR), Post Accident Sampling (SSP), and Main Steam to the auxiliary feedwater steam turbine. The s:venteen additional valves are the turbine dnven auxiliary feedwater pump-turbine exhaust condensate drain line isolation valve (3 MSS *AOV65), three temperature control valves in the CCP Sp tem (3CCP*1V32A/B/C), the Containment Vacuum System inboard containment boundary valve (3CVS*AOV23), six Post Accident Sample System simple isolation valves, and six Reactor Plant Sampling System sample isolation valves.
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| Mis-operation of the Reactor Plant Component Cooling Water valves could degrade performance of the CCP system when required for Safety Grade Cold Shutdown. The affected Containment Vacuum System valve is a containment boundary valve and mis positioning would degrade the redundant containment boundary. Mis-positioning of the R: actor Plard Sampling System and Post Accident Sampling System valves could degrade ECCS performance by cr:ating a flow diversion path via the 3/8 inch safety-related sampling lines between redundant system trains, 11,
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| ==Cause of Event==
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| The cause of the reported conditions is a design error. The initial plant design did not adequately consider the potential mis-positioning of these valves under harsh environmental conditions or active failure. This design error appears to h:ve been a result of inadequate interface between t"e Nuclear Steam Supply System provider and the Architect / Engineer (AT.; or inter' ally within the A/E org snization.~
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| N!.c f oRV 366A U.S. NuCL E AR I.E GuL AT oRY CoM Mis sioN LICENSEE EVENT REPORT (LER)
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| TEXT CONTINUATION F ACILliY N AME (1) ooCKET NUMBER (2)
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| LER NuMBE R (G)
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| PAGE (3)
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| YEAR sEOutN1 AL HEvisioN Millstone Nuclear Power Station Unit 3 05000423 NUMBE R NUMBE R 3 of 5 l 96 036 02 T EX1 (11more space is requaed, vse additronalcopics of NRC form 366A) t11)
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| These are historical conditions identified as part of the Configuration Management Program review process l
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| 111. Analysis of Event 3SIL*AV8878A to D are normally closed valves and are located in each accumulator fill line inside the containment.
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| Th3 valves can be opened by an operator from the main control board when borated makeup water is required to be cdded to the accumulators. These valves are Category I components for the purpose of maintaining pressure boundary integnty. However, they are powered and controlled via circuitry which is not class 1E.
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| 3SIL*AV8889B/D,3SIL*AV8877A/B/C/D 3SIL*AV8879A/B/C/D. 3SIL*AV8872A/B/C/D, and 3SlH*AV8882, 3SlH*AV8889A/C, are located in check valve test lines and are normally closed except for check valve leakage monitonng. The valves are opened by an operator from the main control board for the purpose of performing these procedures to identify check valve leakage. These valves are designed with a restricted port to limit the maxirnum flow l rcte when fully open with full system pressure across the valve. These valves are Category I corponents for the purpose of maintaining pressure boundary integnty. However, they are powered and controlled via circuitry which is not class 1E.
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| The turbine driven auxiliary feedwater pump-turbine exhaust condensate drain line isolation valve (3 MSS *AOVGS) is normally open to drain any condensate collected in the exhaust line while the turbine is in standby. The valve fails closed on loss of power or loss of instrument air, and closes on a turbine start. The CCP System temperature control valves (3CCP*TV32A/B/C) modulate to maintain cooling water temperature within predetermined limits. These valves f;il to the position of maximum coolir.g on loss of instrument air. The Containment Vacuum System inboard containment boundary valve (3CVS*AOV23) is a normally closed valve which fails closed on loss of power or loss of instrument air The associated outboard containment boundary valve is a manuallocked closed valve. The Reactor Plant Sampling System and PASS System valves are normally closed solenoid valves located inside containment which f il closed on loss of power, An interpretation of the Institute of Electncal and Electronics Engineers (IEEE) standards conducted by the Nuclear St:am Supply System (NSSS) provider had determined that there were no credible failure mechanisms for the non-l safety related control system which operates these valves. However, the onginal NSSS design did not specifically cddress failures of these control circuits due to postulated harsh environmental effects. Therefore, it can not be cssured that these valves will remain in their proper fail safe position in the event of a design basis accident. Dunng a DBA, a failure could potentially lead to the diversion of safety injection flow and possibly result in pump runout.
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| The Millstone Unit 3 Safety Evaluation Report (SER) states, for protection systems, that the unit conforms to the design basis requirements of IEEE Standard 3791972, 'IEEE Trial Use Guide for the Application of the Single Failure Cnterion to Nuclear Power Generating Station Protection Systems." IEEE Standard 379 as supplemented by Regulatory Guide 153,' Application of the Single Failure Criterion to Nuclear Power Plant Protection Systems, Revision 0, June 1973,*
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| r:: quires that protective system components which are not qualified for seismic events or accident environments be assumed to failif such failure adversely affects safety system performance. Lack of equipment qualification may serve cs a basis to assume failure With the possible f ailures of non-qualified equipment assumed, the protection system must then be capable of performing those functions required to mitigate the consequences of the specific event.
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| 10CFR50.49, " Environmental qualification of electric equipment important to safety for nuclear power plants? requires th"t nonsafety-related electric equipment whose failure under postulated environmental conditions could prevent satisfactory accomplishment of safety functions are included in the plant environmental qualification program.l l
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| x
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| ,U.s. NUCLEAR F.EoulAT oRY commission LICENSEE EVENT REPORT (LER)
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| TEXT CONTINUATION l
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| F AclLITY NAME (1) oocKET NuMsER (2)
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| LER NuMsER16)
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| PAoE (3)
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| I YEAR SEQulNTIAL REv!SloN Millstone Nuclear Power Station Unit 3 05000423 NuMBti NUMBER 4 og $
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| l 90 030 02 1EKT fit more space is required, vse additionalcopies of NRC form 366A) 111)
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| Th3 potential safety consequences of this condition are significant in that credib multiple failures of these non-qualified power and control circuits causing a potentialloss of the ability to mitigate the consequences of an accident. This could result in off site dose consequences potentially in excess of 10CFR100 limits.
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| However, the likelihood of such a common mode failure is considered to be remote, because most of the affected non-qualified equipment is similar in design and construction to other fully qualified components.
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| :- For the SlH/SIL valves, no significant diversion of ECCS flow can occur unless concurrent failure of the connected 3/4 inch and 1 inch test and drain piping is assumed. Although this piping is not considered safety related, the piping design pressures are significantly greater than would be experienced and the water volume would be contained by normally closed isolation valves downstream. For the Containment Vacuum System (3CVS*AOV23) valve, the line
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| :- r maine isolated by a locked closed manual valve. The normally open turbine steam exhaust line drain (3 MSS'AOV65) valve is on a 3/4 inch diameter line, and is located on a relatively low energy steam line. The Reactor Plant Component Cooling Water System (3CCP'TV32A/B/C) valves, and Main Steam System valve are located in accessible areas outside containment. Although not credited, these valves can readily be vented to their safe position by manual action.
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| For the Reactor Plant Sampling System and Post Accident Sampling System valves, significant diversion of flow would not be expected due to the small diameter (3/8 inch) of the sample lines.
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| ==IV. Corrective Action==
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| S!fety related valves (air operated and solenoid operated) controlled by non safety-grade components whose failure could adversely affect safety system performance have been identified. Initially 21 valves located in the SlH and SIL Systems were identified. Further review has identified an additional 17 valves in the following systems: Reactor Plant Component Cooling Water, Containment Vacuum, Reactor Plant Sampling, Post Accident Sampling and Main Steam to the auxiliary feedwater steam turbine. Recurrence of this type of condition is not expected due to enhancements in th3 design control process made since this occurrence. This completes reviews of plant systems for this condition.
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| Th3 following corrective action will be token:
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| 1.
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| For systems whose safety performance may be unacceptably reduced as a result of air operated oi solenoid operated valves (controlled by non-safety grade, non-qualified components) mis-positioning due to exposure to harsh environmental conditions or active failure, corrective actions will be taken to restore performance to within the l design basis prior to entry into Mode 4.
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| V.
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| ==Additional Information==
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| None fdC FORM 366AN951
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| U.S. Nucl::r Regul: tory Commission B16590/Page 2 Atta6hment: 1)
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| NNECO's commitments in response to LER 96 036 02 2)
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| LER 96 036 02 cc:
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| H. J.
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| Miller, Region l Administrator A. C. Corne, Senior Resident inspector, Millstone Unit No. 3 J, W. Andersen, NRC Project Manager, Millstone Unit No. 3 W. D. Travers, Dr., Director, Special Projects
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| ?
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| l l
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| Docket No 50 g3 B16590 Millstone Nuclear Power Station, Unit No. 3 NNECO's Commitments in Response To (LER 96-036 02)
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| July 25,1997
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| _a
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| 4 List of Regulatory Commitments The following table identifies those actions committed to by NNECO in this document. Please notify the Manager - Nuclear Licensing at the Millstone Nuclear Power Station Unit No. 3 of any questions regarding this document or any associated regulatory commitments.
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| Number
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| ==Commitment==
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| Due None None N/A o
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| 6 4
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| Docket No. 50-423 B16590 I
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| Millstone Nuclear Power Station, Unit No. 3 NNECO's Submittal of
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| &ER 96-036-02)
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| July 25,1997
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| %Ac toRM 366A U.s. NUCLEAH REoVLAToAY Commission
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| {4 9f4 UCENSEE EVENT REPORT (LER)
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| TEXT CONTINUATION F ACILITY NAME (1)
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| DOCKET NUMBER (2)
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| LER NUMBER (6)
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| PAoE 131 YEAR sE QUE N11AL REV1SloN Millstone Nuclear Power Station Unit 3 05000423 NUMBE H NUMBE R S of 5 i
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| 96 030 02 l
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| TlKT rit more spaceis required. use additoonetcol es of NRC form.166A) l11) d
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| ==Similar Events==
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| LERs discussing inadequate design control related conditions are identified below. Various elements of the Configuration Management Program are being conducted to detect design and licensing basis problems. The LERs att LER 96-007-02
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| * Containment Recirculation Spray, Quench Spray, and Safety injection Systems Outside Design Basis
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| {
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| Due to Design Errors.'
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| r LER 96-013-00 ' Residual Heat Removal (RHR) System Design Deficiency Due to NonconservativeOriginal Design i
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| Assumption? (Loss of control air causes RHR heat exchanger (HX) throttle control valves to fail open overstressing the Reactor Plant Component Cooling Water System piping connected to the HX.)
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| LER 96-045-00
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| * Electrical Separation Design Conflict with FSAR?
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| LER 97-010-00
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| * Electrical Calculation Discrepancies in Minimum Voltage Analysis for Class 1E Electrical Systems?
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| LER 97-01100
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| * Hydrogen Recombiner Heaters Potentially Outside of Design Basis Under Degraded Voltage Conoitions?
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| LER 97-015-00 'PotentialVortexing of Recirculation Spray System Pumps?
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| Manuf acturer Data Ells System Code High Pressure Safety injection System........80 Residual Heat Removal / Low Pressure....
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| ..BP Safety injection System Containment Vacuum System...
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| .... B F -
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| Component Cooling Water System...
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| ...CC l
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| Main Steam System....
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| ....SB Post Accident (Monitoring) Sampling..
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| ......lP System Reactor Sampling System (Sampling..
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| ..KN
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| :- and Water Quality System)
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| Ells Component Code V alv e, S olenoid, Flow......................................F SV
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| ' V al y e, S a m ple................................................ S M V V al v e, Te s t....................................................TV Val v e, I s ola t io n............................................... l S V-
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| }}
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| {{LER-Nav}}
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