05000263/LER-1990-001, :on 900313,design Deficiencies Noted in Emergency Filter Train Sys.Ventilation Units & Ductwork Isolated.On 930315,determined That Train B of CR Emergency Filtration Sys Could Not Supply Adequate Flow
| ML20035F082 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 04/14/1993 |
| From: | Andrew Ward NORTHERN STATES POWER CO. |
| To: | |
| Shared Package | |
| ML20035F081 | List:
|
| References | |
| LER-90-001, LER-90-1, NUDOCS 9304200235 | |
| Download: ML20035F082 (11) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i) |
| 2631990001R00 - NRC Website | |
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FAG ( (3) 05000 263 1 OF MONTICELLO NUCLEAR GENERATING PLANT Potential Emergency Filter Train System Inoperability Due to Interaction with Non-Saf ety Related Equipment f dVENT DATE (5) LER NUMBER (6l " REPORT NUMBER (7) OTHER F ACluTIES INVOLVED f8) i " " N^*k ^*C N A*L" ni a A s.At m nos l Mn n.o ns rw un u. sr^" ,y,, 05000 1 wm r.m uw umm 11 18 91 93 001 05 04 14 93 05000 OPERATING THIS REPOR1 IS SUBMITTED PURSUANT TO THE REQUtREMENTS OF to CFR O iCheck one or morel (11) MODE (9) h, 20 402{b) 2D 4051c) . S3.731a)(2)(iv) 73 71 f t>) POWER ] 20 40 maw ' 50 h n) 50 73ta)t2)w) X 73 71(c) j LEVEL (10) y 20 400aH1WI 5032icu21 60 73 aH2)tM OTtiER 20 4D5 tau 1 W) 50 73 aH2t H X 50 73(aH2)(vmHA) "
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(([ 20 405:aH1!M 50 73 aM2Hm) 50 73(aM2)W) UCENSE E CONTACT FOR THIS L ER (12) r.wt Lt%r t navnts pmm. Am. conn Anne k'ard. Superint endent Reactor Systems Engineering (612) 295-1256 COMPLETE ONE LINE FOR EACH COMPONENT f AtLURE DESCRIBED IN THIS REPORT (13) I cA,. a .v m e om mr, & ac W CA X nmM coummrNT ut.NarACTuits P SUPPI E ME N'TAUREPORT EXPECTED (14) EXPECTED M" ~ ns SUBMISS!ON m m er m oescu t < euws un X DATE (15) Aegggygeggg0gafdf f5T$EEN EEt9[e9pIIid[r"iDSak submitted on April 12, 1990. l t The supplemental report concerns the discovery of an additional system design deficiency. On March 13, 1990, design deficiencies in the Emergency Filter Train (EFT) system, and systems which interact with the Emergency Filter Train system were discovered during a special test. Immediate corrective actions were taken to isolate and secure various ventilation units and ductvork to prevent Safety Related/Non-Safety Related systems interaction and ensire operability of the Emergency Filter Train system. An evaluation was completed for the 250 VDC battery for elevated air temperatures and a modification was completed to provide a redundant method to block air flow to the battery room from the EFT. On March 15,1993, it was determined that the "B" train of the Control Room Emergency Filtration system could not supply adequate pressurizing flow when operated just above the low flow trip set point. This was caused by a design deficiency. Administrative contro1r have been initiated to maintain the "B" Filtration train flow controller in manual and the "B" train of the Control Room Emergency Filtration system in lag with the "A" Filtration train in lead An alternate method of measuring system flow though the "B" Filtration train of t.he Control Room Ventilation EFT system has been implemented and surveillance procedures have been revised. 2 9304200235 930414 PDR ADOCK 0500 3 S l
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Description
I On March 13, 1990, with the plant operating at 100% power, Special Procedure j
- 8878 " Emergency Filter Train Filter Fan Low Flow Logic Test" was performed.
l The purpose of the test was to assess whether operability of the Emergency l Filter Train system (Ells System Code : VI) had been compromised by a former t design deficiency (currently eliminated) in the system logic. The test showed that the previous logic design deficiency had no adverse impact on Emergency Filter Train system operability. However, the test disclosed a previously unidentified interaction between one j of the Administration Building ventilation units (V-AC-14) (Ells System Code. t UD) (a Non-Safety Related system) and the Emergency Filter Train. With the i outside air temperature between 40 and 70*F, the ventilation unit supplies a I significant amount of outside air, resulting in pressurization of portions of } the administration building. During the test, outdoor air temperature was 49-50*F (The ' worst case' temperature for maximum building pressurization, as subsequently identified by the ventilation unit's manufacturer). The test { showed that the "B" train of the Emergency Filter Train, when operating alone, I was unable to maintain a positive differential pressure between the Main l Control Room (Ells System Code : NA) and the Administration building (EIIS System Code : MA) as required by Technical Specification 4.17.B.2.b(3). The i "A" train of the Emergency Filter Train was able to maintain the required l positive differential pressure. Currently, the Emergency Filter Train a actuation logic does not automatically trip ventilation unit V-AC-14. V-AC-14 was immediately tripped and secured to ensure Emergency Filter Train j operability. On March 30, 1990 during subsequent investigation of the Emergency Filter Train system design, engineers determined that administration building ventilation supply units V-AC-ll and S-1 (see Figure 1 Simplified l Administration Building Ventilation system drawing), may not trip in the event l of an Emergency Filter Train High Radiation Mode automatic initiation. The l si nal for these ventilation units is initiated by a single Non-Safety Related 5 relay and associated Non-Safety Related switchgear. This is contrary to the design basis for the Emergency Filter Train system which requires all equipment related to Control Room habitability to be single failure proof and l Safety Grade. Failure of these ventilation units to trip during a High Radiation event could potentially pressurize the Administration Building and i degrade the Emergency Filter Train's ability to maintain a positive j differential pressure between the Control Room and the AdministrStion j Building. The ventilation units were immediately tripped and secured to ensure Emergency Filter Train operability. woeu mz m
I NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104 me EXPlRES $/31/95 i I sT:MCE D BJRDEN 39 ra[peggi v0 COMMY WW TW5 INf DRt/A itON COdfCTON RfDUEET (W O HRE. FDAWARD LICENSEE EVENT REPORT (LER) cover.n,ac.m uco mwE 20 Tm wo.nunaN 6 i wn nrconas uwauna wwc ruw m% u s. acau, MGJJTOHV C,OV%nON. WENNGTON, DC 22% Orn, AND TO T4 F ASTREAK REDCON PFCut CT p%C1%, Of F M OF (See teverse ip? reqwred fiumber Of dp$ / char $0te"$ 1?r Fath block) uwW Mi *a AND itJDGE7, W ASHN3 TON. DC 23503 F ACiUTY tiiAML (11 $ Chit NUMHiH p1 FAGE p) MONTICELLO NUCLEAR GENERATING PLANT l 05000 263 1 OF 11
- i*
- Potential Emergency Filter Train System Inoperability Due to
[ Interaction with Non-Safety Related Equipment EVENT DATE (5) LER NUMBER (6i REPORT NUMBER (7) OTHE R FACILITIES INVOLVED (8) '# * *L L'N "* H sEwrN A mas uasm w am e un on v"" y, 05000 [ um wm Dwmuuss l1 18 91 93 -- 90-001 05 04 14 93 05000 OPERATING ITHIS REPC IS SUBMITTED P(IRSUANT TO THE REQUIREMENTS OF to CFR O # Check one or more) (11) MODE (9) N 2040 2 20 405tc) 50 73;aH2J0v) 73 71(b) POWER 20 435,a H1 HU $0 36:ch1) 50 73:aH2Hv) X 73.71(c) LEVEL (10) j007 23 435 ant 6 50.3NcH2) 50 73(a)(2HM OTHER 20 AD5iaMi)im! 50 73;ah?;W X 50 73!aH2Hvi.0(A) W*M * 'h*"*d wm em m %.t unc 20 43NaH1l pv) 50 73iaH2Hm 50 73(aH2HeHB) n, m 23 435;aH1 HW 53 73ta62)W) 50.73(aH2Ht) { I LICENSEE CONTACT FOR THIS LER (17) wn m t m ost Nsuu s pnu e. o n u m Anne Ward, Superintendent Reactor Systems Engineering (612) 295-1256 [ COMPLETE ONE LINE FOR EACH COMPONENT FAtLURE DESCRIBED IN THIS REPORT (13) I c/ a u tv e m.v w u Ac? m c.Aou sys EM counoNoa uANJrACTU4i R l SUPPLEMEfiTAL REPORT EXPECTED (14) EXPECTED vts SUBMISSION r m m oscr o nia/ 3 ns ten X DATE (15) " N @ i N eS5 do W of'fT.iE 5 UM'E En N p'OEf Tr'i$$Na h submitted on April 12, 1990. l The supplemental report concerns the discovery of an additional system design deficiency. On March 13, 1990, design deficiencies in the Emergency Filter Train (EFT) system, and systems which interact with the Emergency Filter Train system were discovered during a special test. Immediate corrective actions were taken to isolate and secure various i ventilation units and ductwork to prevent Safety Related/Non-Safety Related systems interaction and ensure operability of the Emergency Filter Train system. An evaluation was completed for the 250 VDC battery for elevated air temperatures and a modification was completed to provide a redundant method to block air flow to the battery room from the EFT. On March 15,1993, it was determined that the "B" train of the Control Room Emergency Filtration system could not supply adequate pressurizing flow when operated just above the low flow trip set point. This was caused by a design deficiency. Administrative controls have been initiated to maintain the "B" Filtration train flow controller in manual and the "B" train of the Control Room Emergency Filtration system in lag with the "A" Filtration train in lead. An alternate method of measuring system flow though the "B" Filtration train of the Control Room Ventilation EFT system has been implemented and surveillance procedures have been revised. l $s Die / :* S R l 9304200235 930414 ADOCK0500g3 ppR S
.. ~ .. - ~. - _.. -.. ... -.. - ~.. ~ ~. ... ~ ~.. ~. - E l REQUIRED NUMBER OF DIGITS / CHARACTERS FOR EACH BLOCK f BLOCK NUMBER OF NUMBER DIGITS / CHARACTERS ( I LE i 1 UP TO 46 FACILITY NAME i c 8 TOTAL i DOCKET NUMBER 3 IN ADDITION 10 05000 3 VARIES PAGE NUMBER 1 4 UP TO 76 TITLE 5 ^ 2 PER BLOCK 7 TOT AL l ^ 6 LER nut'9BER i 3 FOR SEQUENTIAL NUMBER 2 FOR REVISION NUMBER - f 7 REPORT DATE p pE B CK UP TO 18 - FACILITY NAME t t OT HER FACILfTIES INVOLVED l 8 TOTAL - DOCKET NUMBER 3 IN ADDITION TO 05000 4 i 9 1 OPERATING MODE I + 10 3 POWER LEVEL 'I R 10 W CHECK BOX THAT APPLIES UP TO 50 FOR NAME h a 14 FOR TELEPHONE CAUSE VAR:ES ? 2 FOR SYSTEM 13 4 FOR COMPONENT EACH COMPONENT FAILURE .i 4 FOR MANUFACTURER NPRDS VARIES j S A ED CHECK BOX THAT APPLIES SJoJgcx EXPECTED SuBM:SSiON Due i ,s i ? .'l t i f A t I I i-
NRC FOW 366A U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-01N en EXPIRES 5/31/95 l ESTiuA'ED BJ80EN PER RESPONSE TO COMP Y W7H TH'S k L LICENSEE EVENT REPORT (LER) 5.TE%EN',%RETSE MO L O M l TEXT CONTINUATION AND MO RDS MAMGEUENT BWM (UNBS WW $ NN 3 REGutATORY COMM!SSON. AAS*N3 TON. D0 20SSS0301. AND TO %E FAITRWORK REDJCTON PFCJECT Q152Dd;, oFFCE or j MANAGEMENT AND B.OGET. WASM N3?ON. DC 20503 l nciuri =Aus m ooc.n wuuera m u n wuueta m eAct m j sms m uso. y, NJV66R NUMBER 05000 263 I Monticello Nuclear Generating Plant 93 001 05 2 11. rEn w ~, u.:.. uw .a..nw.- coo s e wn,. m m ) -===
Description
On March 13, 1990, with the plant operating at 100% power, Special Procedure I
- 8878 " Emergency Filter Train Filter Fan Low Flow Logic Test" was performed.
I The purpose of the test was to assess whether operability of the Emergency Filter Train system (EIIS System Code : VI) had been compromised by a former i design deficiency (currently eliminated) in the system logic. The test showed [ that the previous logic design deficiency had no adverse impact on Emergency [ Filter Train system operability. l However, the test disclosed a previously unidentified interaction between one of the Administration Building ventilation units (V-AC-14) (EIIS System Code. UD) (a Non-Safety Related system) and the Emergency Filter Train. With the outside air temperature between 40 and 70*F, the ventilation unit supplies a significant amount of outside air, resulting in pressurization of portions of the administration building. During the test, outdoor air temperature was 49-50*F (The ' worst case' temperature for maximum building pressurization, as subsequently identified by the ventilation unit's manufacturer). The test showed that the "B" train of the Emergency Filter Train, when operating alone, ( was unable to maintain a positive differential pressure between the Main Control Room (EIls System Code : NA) and the Administration building (EIIS System Code : MA) as required 'ay Technical Specification 4.17.B.2.b(3). The "A" train of the Emergency M iter Train was able to maintain the required positive differential pressure. Currently, the Emergency Filter Train actuation logic does not automatically trip ventilation unit V-AC-14. V-AC-14 { was immediately tripped and secured to ensure Emergency Filter Train j operability. On March 30, 1990 during subsequent investigation of the Emergency Filter Train system design, engineers determined that administration building ventilation supply units V-AC-11 and S-1 (see Figure 1, Simplified l l Administration Building Ventilation system drawing), may not trip in the event of an Emergency Filter Train High Radiation Mode automatic initiation. The signal for these ventilatlor units is initiated by a single Non-Safety Related relay and associated Non-Safety Related switchgear. This is contrary to the design basis for the Emergency Filter Train system which requires all i 4 equipment related to Control Room habitability to be single failure proof and Safety Grade. Failure of these ventilation units to trip during a High l Radiation event could potentially pressurize thc Administration Building and degrade the Emergency Filter Train's ability to maintain a positive differential pressure between the Control Room and the Administration Building. The ventilation units were immediately tripped and secured to ( ensure Emergency Filter Train operability, j i NRO F08W 366A (5 92) I
l r NRC FORM 36sA U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104 tsu EXPIRES 5/31/95 [ i ES'1MA*ED BJRDEN PER RESPONSE To couPJ wiH TH:5 i LICENSEC EVENT REPORT (LER) 5"?ZER$%RE.*s"vJ!% O's ^"m"o"a<5 """ss' ion was+N3m De nuam. "wo Te" i TEXT CONTINUATION aE3 couu THE PAPERWORK REDrCN PSCJECT (31S>c104. oFFCE oF i MMA3EMCC AND BASET m A5+NSTON, DC m33 FACIUTv hAME Pl DOCEET NUMBER (2) U R NUMBk N (t) PAGE (3) uum w. NJUBE R NUMBE R 05000 263 OF Monticello Nuclear Generating Plant 93 001 - 05 3 11 [ rta w m ur, a wea -.mam, ac m w.s o n f This same design deficiency exists with the Control Room kitchen and lavatory [ erhaust fan (V-EF ~l6). This fan exhausts air out of the Control Room. It is possible that if the fan does not trip as designed in the event of an Emergency Filter Train High Radiation emergency, it could exhaust enough air from the Control Room keeping it from being pressurized. The fan was l immediately tripped and secured pending further investigation and testing. i Further investigation has revealed a potential concern involving interaction e between the Turbine Building Ventilation Units V-AH-1, V-AH-2, V-MZ-1, V-MZ-4, V-MZ-5, and V-MZ-6 (Ells System Code : UD) and the Emergency Filter Train System. The Control Room is adjacent to the Turbine Building at the Turbine i Operating Floor level. The Turbine Building Ventilation Units (supply fans) are not automatically tripped upon Emergency Filter Train High Radiation mode initiation and failure of non-safety-related Reactor Building Exhaust fans, which also exhaust from the Turbine Building, could result in pressurization of the turbine building relative to the control room. Procedures are in place I to instruct operators to trip the turbine building ventilation units as needed l in a High Radiation event to assure the Control Room remains at a positive pressure with respect to the Turbine Building. i i On April 6, 1990 further review of the Emergency Filter Train design [ determined that a passive break in the Non-Safety Grade portions of the i Emergency Filter Train system ducting (E1IS Component Code : DUCT) serving the Emergency Response Facilities (EIIS System Code : Un may divert pressurizing air from the Control Room to the duct break. Detailed review of the postulated ductwork failure has revealed that a potential problem does exist if one Emergency Filter Train Ventilatiot. unit fails. The dampers supplying l pressurizing air to the Emergency Response Facilities have been secured l closed. Other ductwork and non-Safety-related equipment failures in the i Administration Building have been postulated which may allow contamination to enter the Control Room. For this reason, the non-Safety Related ductwork from the Emergency Filter Train to the Emergency Response Facilities and a return register which is in the B Emergency Filter Train room were blocked. As part of the corr'ectivo actions initiated 'oy the event, a Design Basis / Configuration Management review was initiated. On November 18, 1991 at 0845 it was discovered, through this review, that a single failure of damper VD-9212B,-Battery Room Supply Damper, could prevent the Emergency Filter Train from performing its design function. If the above damper failed to close it J could prevent the Control Room from pressurizing during a radiation release i event which is inconsistent with the Emergency Filter Train design basis. NRc Ffh/ 36EA $Sh
NRC f ORM 366A U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-C.04 I esa. EXPIRES 5/31/95 EswA't0 BJRDEN PER RESPONSE TO COMM W"M TH:$ LICENSEE EVENT REPORT (LER) 5 % 7REf $ $ ".E.,sF J ?o""' M *2 J TEXT CONTINUATION Ac rec Ras vasa wm aws ivws vs.w s mA= RE3MORY COpV S$rCN &&SNWG*ON DO mn(LTi, ANO TO THE PAPEnWORK REDJCTON PRMOT (39M4%, oMCE or uANAst utNT AND pJDt,ET. WASMNGTo% DC MS3 f ACILTTY hAML C) { DOce ET NUMBER (2) LIR NUMS[R tt) PAG [ (3) s.t ww.A. Rr m> DCJUBER NJWBER 05000 263 OF 1 Monticello Nuclear Generating Plant 93 001 - 05 4' 11 l FTuim n.m.. m e...= w. cop. w % % on On March 15,1993, with the plant in cold shutdown, while investigating concerns identified as part of the Design Basis Document saview Program, plant l engineering identified that the "B" train of the Control Room Emergency Filtration system could not supply adequate pressurizing flow to the Control { Room when operated just above the low flow trip set point. Contrary to this, the "A" train of the Control Room Emergency Filtration system was able to pressurize the Control Room when operated just above the low flow set point. l An investigation was initiated to determine the reason for the difference. It was determined that the flow element which provides the input to the flow [ controller (EIIS Component: FCO) was not providing accurate flow indication causing the flow control damper, VD-9111B, (EIIS Component: DMP) to operate at almost the closed position. This resulted in low pressurizing air flow to the [ Control Room when in the High Radiation mode. The flow controller was placed in manual and positioned to maintain the supply damper at the 75% open position when the "B" filtration train is operating. Testing has confirmed that this will supply adequate flow to pressurize the Control Room during all t conditions. Administrative controls have been initiated to maintain the "B" Fi1 tration train flow controller in manual and at 75% and the "B" train of the Control Room Emergency Filtration system in " Lag" with the "A" Filtration train in " Lead". l [ t The March 15, 1993 event was a condition prohibited by Technical i Specifications and is reportable per 10 CFR 50.73(a)(2)(i). i
Cause
i The root cause of the potential Emergency Filter Train inoperability I identified prior to the March 15, 1993 event was design inadequacy. l f 1 1 The Emergency Filter Train was installed in response to the Three Mile Island l Action Plan. The system was designed to enhance Control Room habitability i following a loss of Coolant event. During the final stages of the Emergency i Filter Train construction, a second addition to the plant Administration Building was constructed (see Figure 1). The potential for the.second [ Administration Building addition's Non-Safety Related ventilation system .(V-AC-14) to interact with the Emergency Filter Train system was never considered in the design of V-AC-14. sio direct Safety Related trips from the Emergency Filter Train were included in the design of V-AC-14. j f i i i w PdRC FORW 3%A {$-SE 'h
.U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104 m EXPlRES 5/31/95 ESTIVA'ED BJ8 DEN PE R RESPONLE To COVD Yi W"H TH.S LICENSEE EVENT REPORT (LER) E MEG % 57RE,*$fufTa L C C TEXT CONTINUATION ae rec es uAmum saw mee m swam REGJ#0Rv COMMSS8DN. WASHNGTON. DC P:%5m AC TO i THE FAPERWORA REDUCTCN PROJECT p+5mosi, OFFCE oF l WA%AGEMENT AC BJD'iET. WASMNGTON DC 2 T3 L F Actuiv hAME 0) DOCKET NUMBER 121 E.ER NUMBER (S) PAGE pp LE WEhhA. RE v.iro*e yg, NJMBER A'UMBE R 05000 263 OF Monticello Nuclear Generating Plant 93 001 - 05 5 11 l ru,wmo,uraanuna m. m eoo.. w o-xs on The design of ventilation systems for the original Administration Building and its first addition took into account the need to provide a level of protection for the facilities during a High Radiation Event. With this in mind, l Administration Building ventilation units were provided with an automatic trip
- - I upon initiation of the Emergency Filter Train High Radiation Mode.
Since the ventilation systems serving these areas are not required to be single failure proof, the automatic trips and associated ductwork were designed and installed utilizing Non-Safety Related components. Also, redundant isolation between safety and non-Safety Related portions of the ductwork was never installed. l The design did not take into account the potential interactions with the l Safety Related Emergency Filter Train system, and the Non-Safety Related l ventilation systems and non-Safety-Related ductwork. { i The cause of the March 15, 1993 event was a design deficiency. The flow
- - f element was located next to a bend in the duct.
The proper location of the element should have been in a straight run of duct work. This resulted in a-t high turbulent air flow and improper flow indications. Also, the vendor of the flow element has indicated that an improper flow straight was used for l this application. i i Analysis The original design of the Emergency Filter Train system resulted in conditions where a failure of the Non-Safety Related ventilation units to trip could have potentially resulted in pressurization of portions of the i Admini=tration Building or Turbine Building and degradation of the Emergency Filter train system's ability to maintain a positive differential pressure between the Control Room and the Administration Building or Turbine Building. A failure of the ductwork could have degraded the Emergency Filter Train -} System's ability to pressurize the Control Room, or have allowed unfiltered l airborne ac':wity to be brought into the Control Room. These deficiencies had i the potentid co adversely affect the habitability of the Main Control Room. l The dose received by operators in the Main Control Room has been shown by [ previous design reviews to be the most limiting plant condition during events which release gaseous radiation to the environment (reference Licensee Event Reports 89-29 and 89-40). Analyses have shown thyroid dose to be the limiting Control Room dose condition. Self Contained Breathing Apparatus are available { to protect Control Room operators during a release of gaseous radioactivity. I l I i r NRC FQFn# 366A (542) l-a
INRC FORM 366A U.S. NUCLEAR REGULATORY COMMtSSION APPROVED BY OMB NO. 31504104
- in2, EXPlRES 5/31/95 ES9Ma*ED BJR3EN PER RESPQN$[ TO COMPLY W*H TH1$
LICENSEE EVENT REPORT (LER) COMME,cs REs OLLECTO.JRatN ESwcE mwORMuoN W50RMA70N C N REOJEET: 50 0 HRS FORnARO Na TEXT CONTINUATION AC RWRDS MAM EMEC BMNCM mBB WW E NMAR REGJLATORf COMM:SSON. W ASHN370N. DC 20SS#D30t ANO 10 T**E PApfRWDRK REDUCTON PRMCT {31SM10dj, oFFCE of j i MANAGEMEC AND BJ33ET, W AS**iN3 TON, DC 20503 I f ACILITV NAME m DDCAET NUMBE R (2) LIR NUMBE R (5) FAGL (3) SE Q JL SLA. MViSON 99 NJMBER NJMBER 05000 263 OF l Monticello Nuclear Generating Plant 93 001 - 05 6 11 i Ten m. w.ma r.wea ..mm ca.w wc w uss v n An analysis has been performer' M 'etermine the effect of the Administration Building or Turbine Building p .arizing and the resulting dose received by a Control Room Operator. This analysis assumed that a Loss of Coolant Accident resulting in core damage has taken place, that the Primary Containment (EIIS i System Code : NH) leaks at its Technical Specification limit of 1.2 percent s per day, on a weight basis, and that the Main Steam Isolation Valves leak at .i their Technical Specification limit. The analysis showed that Control Room operator dose does not exceed the limits of 10 CFR Part 100 if the Reactor Building Plenum, Turbine Building, and Administration Building fans are tripped within 33 minutes. Therefore, sufficient time is available for { operators to take manual action to assure Control Room habitability in the [ event of a release of gaseous radioactivity due to core damage. A probabilistic analysis was performed to determine the probability of the Non-Safety Related breakers associated with the Administration Building or Turbine Building ventilation system not opening. This analysis assumed a Loss 3 i of Coolant Accident leading to core damage had taken place. The analysis showed that the probability of a Non-Safety Related breaker failing to open, together with a Loss of Coolant event was extremely small (less than 7x10-7 per yea 1 j l The ductwork for pressurizing air from the Emergency Filter Train system to t'e Emergency Response Facilities h s been blocked. This assures that the ] Emergency Filter Train will be able to pressurize the Control Room in the High Radiation mode of operation as required if the ductwork in the Emergency Response Facilities fails. This is acceptable because it does not affect the ability of the Emergency Filter Train system to pressurize the Control Room if one or both Emergency Filter Train ventilation units is available. In this t l configuration, pressurizing air is supplied to the Emergency Response Facilities providing both Emergency Filter Train ventilation units are I operable. Upon failure of one Emergency Filter Train ventilation unit, no ventilation or pressurizing air is supplied to the Emergency Response 1 Facilities, however, the Emergency Response Facilities ventilation is not required to be single failure proof per NUREG 0696. l i All ductwork connecting the Emergency Response Facilities to the Emergency I Filter Train system has been blocked because of the possibility of contamination migrating from the Emergency Response Facilities to the Control Room through common ductwork. The ductwork has been blocked in a manner so that it may be restored as needed, and procedures have been issued for l restoration of ventilation to the Emergency Response Facilities. This restoration will occur only if both Emergency Filter Train ventilration units are available, all Administration Building Ventilation units have been NRO FORM 366A 1542;
.U.S. NUCLtAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104 5 BL EXPIRES 5/31/95 ESTMA'ED BJRDEN PER RESPONE TO COWPiY W'TH TH'S LICENSEE EVENT REPORT (LER) C7%E$$'@R:"'.*d4L J!?o O, S T E TE CO AND RE:'/ORDS MANA3EuENT B;ANOH iuNBB met U S NJCLEAR REGJLATORY COVM SSON n ASH'NGTON, DC 235W0001. AND TO THE FAPERWORK REDJCTON PRa]ECT p1motod), oFFCE OF MANAGE ME NT AND BJDGET, WASHN3 TON DC 20503 l f ACILITY hAME (t) DOCAET NUMBf R (2) Lt R NUMBLH (6) PAGE (3) uust a a%>. NUVBER NUMBER 05000 263 OF Monticello Nuclear Generating Plant 93 001 05 7 11 ) TEXT (t? more wete n wred use esamord comes or +C form Je#; o h verified tripped, and the ductwork has been verified to be intact, thus assuring that neither Emergency Filter Trait.tnit operability is affected. Manual action for Emergency Response Facilities Emergency ventilation was verified to be acceptable per NUREG 0696 (the boundary for the Emergency Response Facilities is already manually initiated). Blocking of the return register from the B Emergency Filter Train room to the suction of the A Emergency Filter Train ventilation unit is acceptable because i return air from the B train room is not required in the normal or emergency i modes of Emergency Filter Train operation. %e flow through the register is l minimal (400 CE4 <10% of unit ventilation flow), so it will have a negligible effect on either the A or B Emergency Filter Train ventilation unit. ( l During normal operation the Battery Room receives ventilation from the Emergency Filter Train system, but is isolated from the Emergency Filter Train i system upon a high radiation signal. If the damper failed to close during l emergency isolation condition, air from the Emergency Filter Train system used l for pressurizing the Control Room could be diverted through the Battery Room i thus reducing the Control Room pressurization and possibly resulting in l increased operator dose. This postulated failure is unlikely as the damper-is j safety related and is designed to fail closed. However, the damper was l secured in the closed position to eliminate any potential for diverting pressurizing air. This was an acceptable interim corrective action since an additional air supply to the Battery Room with an in line duct heater exists to maintain room temperature during the winter months. A remaining concern l was the operation of the battery during warmer months with no cooling except j outside air. An evaluation of the 250 VDC battery was completed for battery i operation at a room temperature of 107F, the extreme maximum outside air temperature listed in the Updated Safety Analysis Report. The evaluation concluded the battery would meet its load profile objectives at this elevated j temperature. 4 i Since the effect of diverting pressurizing air from the Control Room is to i lower the differential pressure between the Control Room and the Administration and Turbine Buildings, the analysis of operator dose resulting J from a failure of Non-Safety related administration building ventilation units to trip off during a radiation release event (previously discussed in this report) is bounding. As discussed previously, this analysis indicates time exists for operators to take manual action to secure VD-9212B closed. Procedures are in place that require operators to verify VD-9212 is closed following detection of radiation in the outside air. Since the capacity of the battery room exhaust fan is less than the pressurizing air fan, some pressurization capability may still exist even following a failure of VD-9212 to close. NRO FORM MA FS2) j j
I j = .U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 31504104 i N EXPIRES 5/31/95 i ESTsMATED BaRDEN r[R REf7DNEE To couMY w TM s LICENSEE EVENT REPORT (LER) $$7ZE25s',,3DJ, ;LUD ;"S,,g'TH,2 [ TEXT CONTINUATION AND REC RDS MD.AGEMEC BRANOH (UNBB 77%, U S. NUCLEAR REGJLATOW COMU$SIDN. M ASM'N310N, DC 20St$Mt. AND TO THE PAPERWORK REDJCTON PMCT ptMM oMCE OF I M ANAGEMEV AND BJDGET, WAS*ttN370N. DC 70533 F ACILITY 8vaME tt) DOCAE1 NUMBER (2) L1R NUMBER 46) PAGE. (3) [ YEAR i i NUMBE R NUMBER 05000 263 OF ) Monticello Nuclear Generating Plant 93 001 - 05 8 11 h rw w ~. un., wm..n aes,< N8c +o<- :,m <,7, There were no consequences to the health and safety of the public because the l postulated event did not occur. Even in the unlikely event that the damper vere to fail concurrent with a radiation release, Control Room habitability could be maintained by operator action. ti During performance of surveillance procedures required by Technical Specifications, the "B" Filtration train has demonstrated the ability to pressurize the Mai:. Control Room with an indicated flow of 1000 CFM as k controlled 1.y the inaccurate flow element (this corresponds to an actual flow a of about 600 CFM). Forthermore, it has been demonstrated that an actual minimum flow of 600 CFM is needed to maintain a positive pressure in the Control Room. This corresponds to the low flow trip set point for the "B" Filtration train. During an accident condition, the ability of the "B" Filtration system to pressurize the Control Room would degrade over the long term due to filter clogging. Since the inaccurate flow element causes the flow controller to sense higher than actual flow, the Control Room Ventilation EFT system would not have auto-transferred to the "A" F!itration train prior to the "B" Filtration train flow decreasing below that required to maintain pressurization of the control room. Therefore, pressurization of the Control Room would not have been assured. Since Control Room pressure and filter d/p i is monitored during a long term accident by operations personnel, this condition would be detected and corrective action would have been taken to place the "A" Filtration system in operation. Placing the damper at 75 % ensures nearly maximum attainable flow is achieved l (to account for filter plugging) while providing margin to the upper flow l limit to ensure compliance with Technical Specifications and design bases. ) Therefore, the March 15, 1993 event had no consequences to the health and i safety of the public.
Corrective Actions
o 1. The breakers for all ventilation units which could potentially degrade the Emergency Filter Train due to a Safety Related/Non-Safety Related ) ' interaction were immediately secured opened. An analysis (using the Commercial Grade Dedication process) was completed to show that the Administration Building ventilation unit (V-AC-11, V-AC-14, and S-1) breakers, motor contractors, and relay trip logic would be able to perform their intended functions in the event of i an accident. Analysis of the Turbine Building ventilation unit (V-AH-1, V-AH-2, V-MZ-1, V-MZ-4, V-MZ-5, V-MZ-6) breakers was also completed, I sc FoRv mA p42;
.U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104 (5 Sh EXP!RES 5/31/95 EStuA*ED BJOEN PER RESPONSE To COMP y W'TH TH:S t LICENSEE EVENT REPORT (LER) MORuaTION COufCTON REQUEST: 60 0 HRE FORAARD eOveEcs nEaAme see>EN me,TE TO T,E yORuroN TEXT CONTINUATION ^'""E0**""**"'"***""""""*""*" REGJLATORY COuv:55cN, EASMtNGTON. 00 P05b5 0001 AND TO int PAPERWORK REDJCTON PRWECT ptSG01D41. OF FCE of M ANAGEME M AND BJDGET, WASMMTON, DC 2:S::3 FACRITY hAME (1) DOCA [T WUMBER (2) LER NUMBER gg) PAGE g33 LlQukM A NEVib ON g wouarR wuvetR 05000 263 OF i Monticello Nuclear Generating Plant 93 90-001 - 05 9 11 nxt w m sc.n a e,c a enw. coo.n e m= m nw or> showing that the breakers are able to perform their intended functions in the event of an accident. In the most limiting case, the ventilation units are required to trip in a Design Basis Loss of Coolant Accident I (Analysis does not have to be made for a simultaneous seismic event per Generic Letter 87-02). 4 t Two independent methods to trip each unit were identified and evaluated. f Procedures that specify the required manual actions were issued. It was j physically verified that the units could be tripped in time to assure that 10CFR50 guidelines are not exceeded. i The Administration Building ventilation breakers were returned to j service following completion of a 10 CFR 50.59 safety evaluation. j 2. The Control Room kitchen and lavatory exhaust fan remains secured until l 1 modifications can be made to allow fan to operate without affecting { l Emergency Filter Train operability. i 3. Procedures have been issued to trip Turbine Building Ventilation Units i as needed during a radioactive release to keep the Control Room at a l positive pressure with respect to the Turbine Building. A 10 CFR 50.59 i evaluation was completed to justify these actions, j 4 4. Ductwork has been blocked to assure separation of the Safety Related and j Non-Safety Related portions of the Emergency Filter Train ducting. 5. Procedures have been issued to restore the ventilation to the Emergency. Response Facilities, if needed, in a manner which does not affect Emergency Filter Train system operability. A 10 CFR 50.59 review was completed and documented for these procedures. l 1 6. The ductwork connecting the A Emergency Filter Train ventilation unit I with the B Emergency Filter Train room has been block.ed and sealed closed to prevent system interactions. A 10 CPR 50.59 review was completed and documented for this change. 7. A Design Basis / Configuration Management review of the Emergency Filter Train system has been completed. 8. A modification was completed which installed a redundant method of terminating air flow to the 250 VDC battery room in the event Damper VD-9212B failed to close. Is*.; FORM 366A 15-971 - T '-+-.m v-v s= e me wh
NRC F04.9 366A U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104
- 82; EXPIRES 5/31/95 ESTWATED BJRDEN PER RESPONSE 70 COMP Y Ca TH:S d
6 j LICENSEE EVENT REPORT (LER) 8NE DRMATON CDiECToN REOJEST SD 0 HRS FORAARD Covvees nos=Na suRDEN cs?.us To wc ~coR AToN 1 TEXT CONTINUATION AtiD RECORDS MANAGEMENT BRANON,os 3,m3,U.S NJ^ fMN99 7714L RE3wm covo.33.oy, y3,.3 m. ty ra ma rArrawo:w Rtacow PRxEcv iremon,, omte or MANAGEME N'T AND BJDGET. W ASMN3 TON EC P3503 f ACluTV hAME (1) DOChfT NUMBER (2t L1R NUMBE R tt) F AGE (3) f LE QJE ATsA. Ed v4CN g NsuerR NuMeER 05000 263' OF l Monticello Nuclear Generating Plant 93 001 - 05 10 11 ren <m. ua., nw.a.a..mv cw.n e wc ru, xs p r> 9. The 250 VDC battery was evaluated for operation at 107'. The i F evaluation determined the battery could meet the load profile objectives at this elevated temperature. 10. Corrective actions are being developed to eliminate operator actions. 11. The flow controller for damper VD-9111B has been placed in manual and i set at 75%. 12. An alternate method of measuring system flow though the "B" Filtration train to meet Technical Specification requirements of the CRV-EFT system j has been implemented. i 13. Surveillance procedures have been revised. l l l 14. Administrative Controls have been initiated to maintain the "A" Filtration train in " Lead" with the "B" Filtration train in " Lag". l P 15. Technical Specifications require freon and DOP testing be performed at l 1000 CFM (+-10 %). These surveillance procedures have been re-performed using the alternate method of measuring air flow to verify the required 1000 CFM (+-10%) air flow. ]
ADDITIONAL INFORMATION
2 Failed Comoonent Identification None
Previous Similar Events
i None F N3C FORV 366A (5412) ~
l i NGC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104
- s s2; EXPIRES 5/31/95 E STtMATE ) BURDEN PER RESPONSE TO COMPLY WTH THS
} WroRMATON CoufCTCN REQUE ST: 50 0 NAS. FORWARD LICENSEE EVENT REPORT (LER) COMME.cs REsA%3 BURoEN ESwAn To ist uFowico TEXT CONTINUATION AN REOORDS MAWEMEM BWCH WDB ntdh U S NM i REGLUTORY COMM:SSON. m ASHNGTON. DC Posts.coct. AND TO THE PAPERWORK REDUCTON PRCLKCT ptSo-0104;. OFFICE OF MANAGEMENT AND BLCCET WASHINGTON. DC 70503 7 ACtWTY KAME (1) DOCREY NUMBf R (2) EER NUMkf R (S) FAGE (3) bEQUE.NTiA., HEvi$O4 NUMBER NUMBER 05000 263 OF Monticello Nuclear Generating Plant 93 001 - 05 11 11 t TExi pr mee spare a rowe2 nae eamm capes o NRC form x>tA) 517) r I t OA r C.A. - ( t f e r V-FAC-14A V-CAO-143 lI 1E A cA. : CA. ..4 I T v-AC-u
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