05000000/LER-1986-008-01, :on 860414,not All Required Valves Listed in Valve Position Verification Surveillances.Caused by Difference of Opinion Re Valves.Discrepancies Corrected
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(1) |
| 0001986008R01 - NRC Website | |
text
_ _ - - -.
pge,. aos U s. NUCtaAa s. wLAToxv ComuniesioN Arezovso ceo mo sino.cios LICENSEE EVENT REPORT (LER)
- ACILITY StAaet nl DOCK 8T NUMWE A (2)
PAGE135 Salen Generating Station - Unit 1 o i s l o 1 o l o l 2 l l2 i lorgg Not All Required Valves Listed In Valve Position Verification Surveillance EVENT DATE (5)
LER NuesetRits RSPORT DATE (71 07MER,ACILITit8 tNVOLVED IB)
SEQv f ak j Mo TM OAV vtAR 8 Acik'T v h AWee DOCKET 810MetF.isi MoNTM QAv v8AR YEAR
_Salen - Unit 2 o 15 l D ! o l o l3 I li 1 0l4 1l4 8 6 8l6 0 l0 l 8 0l0 0 l5 1l 4 8h o isto to ru, l l
~
'~
.7,,,,
Twas maconT is sueMetveo evneUANT To Tna neouintusnets or io Cen t rCnne
., m we.-a,> iin 6
=
=w sonwtrw rs
n.,
o r oj o
= =winiun m.wwm 3
mnwaH.i n
.Huni rs.riw M 4054.H1HW) m.mwai sonwH H.<n guegsyg;A,p;,,
}
to 734ala)n) 00.734sH2HetidHA)
J86A/
30 assesin HM f10.7SteH2itM 90 73(sH2HveNHel N sosteH1Het go.731sH2Hal go.736eH2Hs)
LICENBEE CONTACT POA YMit Lf m (12)
TE LEPHONE NUM8tR ARE A COOL J. L. Rupp - Operations Licensing Engineer 6 3 0, 9 3 39 i 4;3 i O9 i
3 i i
COMPLETE ONE Laht POR SACM COMPONENT F A8 LURE D4HCRISED IN Twas napont (131
" A Ny C.
R{o Ajs
CAUSE
SYSTtw COMPONENT A C.
m TA yg
CAUSE
Svt'EM COMPO4 TNT g
pp k_
I 1 i f I # 1 i
i i !
I i i I
I I I I I I i
! I t i I f I
su.eLaueNtAL napoaf sme:CTeo ne, MONTH oav vpm l
] ves tu.,
arr,er,o suserimoh cA rv
] No
- E*,n*l'N,"
I r
A=T. AC T m,,.. -..
,.., n.,
g g
Following a routine review of Unit 2 S$rvice Water System operating procedures, it was determined that several valves should be added to Surveillance Procedure SP(0) 4.7.4.a.
This thirty-one day surveillance requires verification that each valve servicing safety related equipment (that is not locked, sealed, or otherwise secured in position) is in its correct position.
As a result of these findings, the decision was made to immediately review all valve surveillance, including containment integrity, the Auxiliary Feedwater System, the Emergency Core Cooling system and the boron injection flow path.
This review, which was completed on April 14, 1986, revealed no problems with any surveillance except SP(0) 4.6.1.1.a (containment inte rg rity) ; two valves were required to be added to this surveillance.
The valve surveillance were updated to reflect all identified discrepancies.
The revised Unit 2 surveillance were then compared with the Unit 1 surveillance, which were also revised as necessary to ensure compliance with the surveillance requirements.
The majority of the discrepancies involved the service water surveillance.
These valves were originally omitted from the surveillance because of a belief that the normal operation of the serviced equipment provided adequate l
assurance of the correct valve positions.
The other two valves were isolated cases resulting from apparent oversights.
8708210529 870819 PDR FOIA "4,'"" '"
GORDONB7-512 PDR
j
\\
LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Salem Generating Station DOCKET NUMBER LER NUMBER PAGE Unit 1 85988272 86-588-88 2 OF 5 PLANT AND SYSTEM IDENTIFYCATINr Westingbouce - Pressurized Water Reactor Energy Industry Identification System (EIIS) codes are identified in the text as [XX].
q l
IDENTIFICATION OF OCCURRENCES I
Not.All Required Valves Listed in Valve Position Verification Surveillance
)
Event Date:
04/14/86 Report Date: 05/14/86 This report was initiated by Incident Report No.86-111 CONDITIONE PRIOR TO OCCURRENCE:
Unit 1 - Mode 6 - Refueling Outage Unit 2 - Mode 1 - Rx Power 100 % - Unit Load 1130 MNe DESCRIPTION OF OCCDRRFMCE:
During a routine review of Unit 2 Service Water System [BIl operating procedures, it was questioned whether the computerized Tagging Request and Inquiry System (TRIS) valve lineup f or Surveillance Procedure SP(0) 4.7.4.a contained all the required v alv es.
i Surveillance Procedure 4.7.4.a requires at least two (2) service water loops to be demonstrated operable at least every thirty-one (31) days by verifying that each valve (manual, power operated or automatic) servicing safety related equipment that is not locke.d, sealed, or otherwise secured in position, is in its correct position.
Ninety-four (94) questionable valves were identified to the Systems Analysis Group of the Engineering and Plant Betterment Department.
Each valve was reviewed to determine the applicability of SP(O) 4.7.4.a; the results are as follows:
1.
Fif ty-three (53) of the valves were associated with the service water pumps, various room coolers and the chillers.
Although the proper position of these valves is verified by the normal operation of the serviced equipment, it was recommended that they be added to SP(0) 4.7.4.a.
2.
SP (0) 4.7.4.a was not applicable to nine (9) of the valves.
1 i
I i
LICENSEE EVENT REPORT (LER) TEXT CONTINUATION l
1 Salem Generating Station DOCKET NUMBER LER NUMBER PAGE Unit 1 85809272 86-888-99 3 OP 5 DESCRIPTION OF OCCURRENCES (cont'd) 1 3.
Twenty-ei ght (28) of the valves were associated with the Emergency Diesel Generators [LB] and the Containment Fan Coil Units [BK].
Because their position is verified by the satisf actory perf ormance of a surveillance (on at least a monthly basis) for the serviced equipment, it was determined that they need not be included in SP(0) 4.7.4.a.
4.
The f unction of two (2) of the valves is to maintain the i
pressures and flows equalized between the Nuclear Service Water Headers [BI].
It was determined that the position of these valves (normally open) should be periodically verified to ensure service water flow to all saf ety related equipment in the event of a loss of offsite power, concurrent with the loss of one (1) emergency diesel generator.
Therefore, it was recommended that these valves be added to SP(0) 4.7.4.a.
l 5.
The last two (2) valves cross connect the chillers, which provide the air conditioning required to cool the control room, computer room and equipnent room.
It was determined that the position of these valves (no rmally closed) should be periodically verified to ensure that service water flow is maintained to at least one chiller in the event of a loss of i
either service water header.
As a result of these findings, the decision was made to immediately review all valve surveillance f or containment integrity, the Auxiliary Feedwater System [BA], the Emergency Core Cooling System (BO) anc the boron injection fl ow pa t h.
Note: The valve surveillance for the Component Cooling Water System [CC] were not included because a similar review for this system was performed last year.
This review, which was completed on April 14, 1986, revealed no problems with any of the surveillance except SP(O) 4.6.1.1.a (containment integrity).
SP(O) 4. 6.1.1. a r eq ui r e s containment integrity to be
(
demonstrated at least every thirty-one (31) days by verifying l
that all penetrations not capable of being closed by operable containment automatic isolation valves and required to be closed during accident conditions are closed by valves, blind flanges, or deactivated automatic valves secured in their i
positions.
I At this time (April 14, 1986), a list of six (6) questionable valves were identified to the Systems Analysis Group for their review.
In addition, an incident report was generated for the potential deportability of the findings.
Each valve was reviewed to determine i
the applicability of SP(0) 4.6.1.1.a.
The results, which were l
received on April 25, 1986, are as f ollows:
l a-
)
LICENSEE EVENT REPORT (LER) TEXT CONTINUATION l
Salem Generating Station
. DOCKET NUMBER LER NUMBER PAGE l
_ Unit 1 85358272 86-988-SS 4 OP 5_
DESCRTPTTON OF OCCURRENCE: (cont'd) i 1.
Four (4) of the valves, which are located on the Reactor Coolant Pump [AB] seal injection lines, did not meet the applicability of the surveillance since these penetrations do not require isolation during an accident.
2.
One (1) valve, a manual drain valve which is located outside containment on the coolant charging line [CB] was determined to represent a potentially unisolated containment penetration if the f ailure of a check valve (located inside containment) is assumed as the single failure.
Theref ore, it was recommended that this valve be added to SP(0) 4.6.1.1.a to verify its closed position on a periodic basis.
3.
The last item, is a manual drain valve which is located inside containment on the cooling water inlet line to the Excess Letdown Heat Exchanger [CB].
This was determined to be a potentially unisolated containment penetration if the f ailure of the remote operated heat exchanger cooling water inlet isolation valve (located outside containment) is assumed, It was recommended that thiu valve also be added to SP(0) 4.6.1.1.a.
1 The TRIS valve lineups for the valve surveillance were updated tu reflect all identified discrepancies.
The revised Unit 2 surveillance were then compared with the Unit 1 surveillance, which were also revised as necessary to ensure compliance with the surveill ance requirements.
l APPADRNT CAUSE OF OCCURRENCES The procedural inadequacies resulted f rom an apparent difference of opinion as to which valves should be included in the thirty-one (31) day surveillance.
The majority of the discrepancies involved the service water surveillance, and it appears that the originator of the surveillance believed that the normal operation of the serviced i
eq uipme nt provided adequate assurance of the correct valve positions.
Therefore, it was not originally deemed necessary to include these valves in the routine surveillance.
In one isolated case involving the manual drain valve on the coolant charging line (1CV288), the event was caused by the implementation of a system design change.
The system originally contained an additional valve (ICV 287) in series with ICV 288.
ICV 287 was a
" locked closed" valve, thus satisfying the containment integrity r eq uir ement.
This valve was subsequently removed from the system by Design Change Request 1EC-1304; however, due to oversight, ICV 288 was not added to SP(O) 4.6.1.1.a.
~
LICENSEE' EVENT REPORT (LER) TEXT CONTINUATION Salen Generating Station DOCKET NUMBER LER NUMBER PAGE Unit 1 85888272 86-898-88 5 OP 5 APPADRMT CAnEE OF OCCUDDRNCEt (cont'd)
- Concerning the drain valve (1CC179) on the cooling water inlet line to the Excess Letdown Heat Exchanger, the FSAR shows the remote operated cooling water inlet valve (1CC215) as normally closed; therefore, isolating the penetration.
The originator apparently assumed the penetration-did not fit the applicability of the s u rv eill an ce.-
H owev er, the penetration does require isolation during an accident, and the surveillance is theref ore applicable.
ANALYSIS OF OCCUDDRNCEr The surveillance discrepancies were self identified and corrected in a timely manner.
Each item was reviewed for its potential safety significance, and it was determined that none posec a threat to the health or safety of the public.
However, because the surveillance requirements were not fully complied with and, because of the two (2) potentially unisolated containment penetrations, this event is reportable in accordance with the Code of Federal Regulations,10CFR 50.7 3 (a) (2) (i) (B) and 10CFR 50.73 (a) (2) (v) (C).
Because the findings involved both Unit 1 and Unit 2 surveillance, this Licensee Event Report is being submitted as a Unit 1 LER.
CODDRCTIVE ACTIONr i
.As previously stated, all identified discrepancies f or both Unit 1 and Unit 2 have been corrected.
No f urther action is deemed necessary.
3MkL op General Manager-Salem Operations JLR:tns SORC Mtg 86-035 l
l
i PHILADELPHIA ELECTRIC COMPANY 2301 MARKET STREET P O BOX 8699 PHILADELPHIA A, PA.19101 r
mm e4b4ooo January 16, 1987 I
Docket Nos. 50-277
(
50-278 i
j U.S. Nuclear Regulatory Commission l
ATTN:
Document Control Desk l
Washington, DC 20555
SUBJECT:
Licensee Event Report Peach Bottom Atomic Power Station - Ulits 2 /3nd 3 l
This LER concerns failure to establish a firewatch required by the Technical Specifications.
Reference:
Docket Nos. 50-277 and 50-278 Report Number:
2-86-25 Revision Number:
00 Event Dates:
December 10 and December 16, 1986 Report Date:
January 16, 1987 Facility:
Peach Bottom Atomic Power Station RD 1, Box 208, Delta, PA 17314 This LER is being submitted pursuant to the requirements of 10 CFR 50.73(a)(2)(1)(B).
We regret the delayed submittal of this LER.
On January 14, 1987 it was identified that a similar reportable event occurred on December 10, 1986, as well as on Decertber 16, 1986.
Very truly yours, R.
H. Logue 1
Assistant to the Manager Nuclear Support Department i
cc:
Dr. Thomas E. Murley, Administrator, Region I, USNRC T.
P. Johnson, NRC Resident Inspector h*[L - S9 *Sl0' gw;e70sDS- \\ 4
\\ v.. s. n. ses us u s e s e. s o. u a e, 1 .n.o.e e e.e a vu o.e. l UCENSEE EVENT REPORT (LER) 8"*""*88 m.s u, -... Peach Bottom _ Atomic Power Station - Unit 2 eIsistelol21717 1lod OI 3 i....... Diesel Generator Room Cardox Defeated Without a Firewatch .... e.. i.. u a.v.n a.. mesi uu ivi ev.ia eu.st n mse n. .s... u, use nsa "tmlr .. ;,i: u, nn o.s.. n .se essn ss a.a. PBAPS - Unit 3 e;s10te;ei2t718 i l2 1lC 8 6 816 0 l2 l 5 0 l0 0l1 1l6 8 l7 e is so s ~ oo, i i r ga meat a svn ma e swas...t to tas sa was.a.vi ee w va i _sca.. .w , viu ,,n,,,,, N m.sa me so se msn ww.e nsi.e ss..s i.es esmwm ess awe.m tuiw g gagsp4.g I 1 0 _10 m.se.winsi m a. ie ms3 ai ne. .se n .. m es me.nume mJ s.o.m meews nes n me.in me a ee.mimu w> a ) pt s.a. 4.D.f.81 #ea v.is 6i s insi j ..........n. ..s. 1 B 4,1,,5 0, t. i s W. C. Birely. Senior Engineer - 1.icensing Section 2, 45 i . a u e.. o, e.. u.....,. n,.....n. .... n.., j u.,n .,,n.i' .e 53;p 3 ppt;p.~,.Q4 ' 13;;5 ayStir ;'gp; un nin. Sp y' ~, 2 I t 1 i I 1 I 'f' e i e # 1 l t p' I I i ! 1 1 1 I I I I I I l s., i n.. w .i.i.v 6 m es, m icn e... we.,u.o. f ]. n..,...,., m c er,.. w... u e e, ~;q.. l e..ci.............,.,.,..... Abstract: 2-86-25 On December 10, 1986 and December 16, 1986 between approximately 0100 and 0300 hours, dedicated fire watches as required by the Technical Specifications were not established in emergency diesel I generator rooms due to failure to closely adhere to procedural l guidance. During a surveillance test of the E-3 and E-4 diesels on December 10 and of the E-1 and E-2 diesels on December 16, the automatic injection mcde of the carbon dioxide fire suppression (Cardox) system for the diesel being tested was defeated for more than one hour without establishment of a dedicated fire watch in accordance with the Technical Specifications and Administrative Procedures. The manual injection mode of the cardox system and i diesel room cardox injection alarms were operable. Operations l personnel failed to closely' adhere.to guidance regarding the defeat of the automatic injection mode of the Cardox system in Procedure S.8.4.A (manual start of diesels) and Administrative Procedure A-12.1 for controlling Technical Specification fire watches. + To prevent recurrence, these events have been reviewed with all operations personnel and the importance of automatic Cardox injection capability has been re-emph.asized. . n.... A.) i
1 1 .a... wa na ve6saa su esa, so .u LICENSEE EVENT REPORT (LER) TEXT CONTINUATION anno.seow an uw. en,. ass ense secnst s **=6 na Dec a s t awn.a s e his 6ss sweessa oss ,ast saa Peach Bottom Atomic Power u g-igp p g.,. - StatCon - Unit 2 8 1510lo10l 21 717 8;6 010 0 l2 or o l3 l 012(5 Unit Conditions Prior to the Event i December 10, 1986: Unit 2 at 100% and Unit 3 at 98% Reactor Power December 16, 1986: Units 2 and 3 at 100% Reactor Power Description of the Event: On December 10, 1986 and December 16, 1986 between approximately ] 0100 and 0300 hours, dedicated fire watches required by the Technical Specifications were not established due to failure to l closely adhere t'o procedural guidance. During performance of Surveillance Test ST 8.1, " Diesel Generator Full Load Test" on the E-3 diesel on December 10, 1986 and the E-1 diesel on December 16, 1986, an NRC inspector observed that the automatic injection mode of the carbon dioxide fire suppression j (Cardox) system in the E-3 and E-1 diesel rooms was defeated j without a dedicated fire watch. The inspector brought this to the attention of Shift Supervision and tne technical staff, and I it was concluded that a Technical Specification'non-conformance had existed. Technical Specification 3.14.B.4.a requires a dedicated fire watch to be established for the unprotected area within one hour if the diesel Cardox system is inoperable. The plant operator defeated the automatic injection mode of the Cardox system in the E-3 and E-4 diesel rooms on December 10 and t.::e E-1 and E-2 diesel rooms on December 16 upon entering and left each defeated for approximately 1 hour and 10 minutes. The prerequisites of System Operating Procedure S.8.4.A, " Manual Start of Diesels", state: "When leaving any diesel room, always rearm the injection (Cardox) system by operating the disable switch..." and "Cardox shall not be disarmed for more than 15 l minutes without shift supervision approval (will require a Technical Specification fire watch)." Administrative. Procedure A-12.1, " Procedure for Controlling Tech. Spec. Firewatch and Firewatch Patrols", states: "If the activity for which the Cardox System was defeated is still in progress 15 minutes after the defeat switch was placed in the ' Defeat' position, immediate arrangements should be made to provide a Tech. Spec. Firewatch in the effected room within one hour after the defeat switch was criginally placed in the ' Defeat' position." These events cccurred due to failure to adhere to the above guidance. The EIIS code for the af fected system is KP, fire protection. .u.... s A~ L
est, e 3mee t#J awCb842182wu,pa, gen.eme. LICENSEE EVENT REPORT (LERI TEXT CONTINUATION anao.seo e a nee.ein. s **es emas .... u i,.... n.
- aca a-a in o.,,...
.... im Peach Bottom Atomic Power . ggp p =;.g; Station - Unit 2 O3 o l5 lo lc lo l2 l7 l 7 816 0 l2l5 010 g or 0l3 ,,,,,s - .. ~ a. - anco-- m.ron Consequences of the Event: f There were no adverse consequences and the safety significance of this event is minimal. The automatic injection mode of the diesel Cardox system was defeated; however, the manual actuation mode remained operable. Additionally, the diesel room Cardox injection system alarm (red revolving beacon and horn), which actuates on a valid automatic Cardox injection signal, remained operable. Upon actuation of this alarm, the operator would rearm the automatic Cardox injection system and evacuate the room.
Cause of the Event
The cause of this event was failure of licensed and non-licensed operations personnel to closely adhere to the precedural guidance of Procedures S.8.4.A and A-12.1.
Corrective Actions
This event has been reviewed witn all operations personnel. The importance of maintaining automatic Cardox protection has been re-emphasized. For additional assurance that this event does not recur, Surveillance Test ST 8.1 has been revised to include a caution statement regarding the requirements associated with defeating the automatic injection mode of the Cardox system. This procedure revision will be reviewed with operations personnel. Previous similar occurrences: LERs 2-82-35, 2-84-14, 2-84-15, and 2-86-02 involve failure of operations personnel to establish fire watches in accordance with the Technical Specifications. LER 3-86-17 involves failure of security personnel to implement Technical Specification fire watch patrols as~ directed. h O O }}