ML20085J778
| ML20085J778 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 06/06/1995 |
| From: | Crandall R, Doyle T, Schmidt R NORTHEAST NUCLEAR ENERGY CO. |
| To: | Necci R NORTHEAST NUCLEAR ENERGY CO. |
| Shared Package | |
| ML20085J781 | List: |
| References | |
| DE2-95-472, NUDOCS 9506230017 | |
| Download: ML20085J778 (15) | |
Text
y w
p .:,
- Northeast '
- Nuclear Energy M ein o t
June' 6,1995 ,
DE2-95-472 j To: R. P. Necci 2 C)~3 [ [ f N O E 8 #
From: T. A. Doyle -
Subject:
AC-1 and 11. Enclosure Building Secondary Containment Integrity i Single Failure Deficiencies - Engineering Analysis. ,
i
Reference:
(1) LER 2 94-040 01 " Ventilation Enclosure Building Effecting Integrity" (2) memo DE2-95 254 TAD to distribution, mtg minutes 4/6/95, AC-01 and 11 Design Problems for Secondary Containment ;
Integrity", dtd 4/28/95 i ;
(3) MP-2 Technical Specification, Section 3/4.6.5 ,
(4) FSAR, Chapter 5, 6, & 14.
(5) memo, SD Weerakkody/JK Rothert (PRA) to TAD subj " Single ;
Failures of EBFS and Their Importance to Public Safety", dtd , '
5/25/95.
(6) SWEC evaluation " Enclosure Building Purge & Exhaust", dated ,
4/95. +
(7) Enclosure Building Containment Ventilation - schematic of-2AC-l&ll problems.
(8) W G Council to USNRC, subject "MP-2 Enclosure Building Design", dated 3/1/79.
(9) R. M. Kacich to J. J. Kelley, subject "MP-2 Enclosure Building !
Design", dated 7/8/81
~
Potential single failure deficiencies with respect to damper AC-1 and AC-Il were identified in reference (1). MP2 Design Engineering has undertaken l what has turned out to be a fairly significant task with a very challenging schedule to correct the single failure issue by purchasing and installing new damper equipment and designing some CIAS actuation logic changes.
Reviewing the deficiencies involved clarifying the individual and integrated design bases of several systems associated with Secondary Containment Integrity. These are 1) Main Exhaust ventilation, 2) Containment and ,
Enclosure Building Purge (CEBPS), 3) Enclosure Building Filtration System )
(EBFS), and 4) The Enclosure Building itself. The design basis was difficult to 1 ascertain as there were apparent contradictions between design bases documents, regulatory requirements, and as-built design.
Initial design basis investigations suggested that the LER's conclusions were correct and that the single failure deficiencies were indeed an original oversight. However, as noted in the initial evaluation report ref. (6), there were unexplained inconsistencies with the design bases.
- l 9506230017 950606 0 0f
- 1[
U '
June 6,1995
- DE2-95-472 Page 2 of 8 This memo documents the results of our design basis investigations, which are now complete. When approved, this will provide the basis for a major change in scope for the proposed design, currently being worked.
Design Basis and Engineering Evaluation Conclusions Dampers AC-1 and AC-11 were purchased per Bechtel specification 7604-M526 as non-QA non seismic equipment. The installation of these dampers, and their respective ductwork up to the containment isolation dampers was per the same Bechtel specification;'and installed as non-QA. The Bechtel Plant Data Book, revised in 1977 after Start up, continued to classify these ' dampers non-QA.
Post 1977 NNECO upgraded these dampers to QA status since they received a CIAS signal to close post LOCA. Currently, AC-1 & 11 are QA, non seismic, non EEQ.
The Enclosure Building structure is designed to seismic class I criteria.
However, this criteria wasn t applied to many Enclosure Building penetrations which are documented and confirmed to be non seismic. The licensing basis for the Enclosure Building shows that it was seismically designed, however, it is not expected to support EBFS post an SSE seismic event. Correspondingly, EBFS itself will not be operable (with respect to LOCA mitigation) post SSE. This has been clearly stated to the NRC, evaluated, and accepted, as recently as 1979 ref(9).
The Engineering Evaluation contained in the last section of this document, recommends by the assessment - of " Negligible Risk Significance to Public Safety ", that the A C-1 and 11 proposed modification to ensure single failure _ pro visions, not be installed.
The plant is adequately and safely designed to mitigate the consequences of a LOCA and is at no further risk now than previously expressed at plant original licensing.
Discussion Single Failure Problems The first single failure problem scenario deals with 2AC-1. Looking at ref. (7),
if the Enclosure Building is being purged, and you postulate a single failure to the facility 1 control power, then AC-01 will remain open and EBFS fan 'A' will not get a start signal. EBFS fan 'B' will attempt to draw down the Enclosure Building to the required negative 0.25 in w.g. and most likely not achieve the requirement.
The second problem deals with 2AC-11. As shown on ref. (7), if the Enclosure Building is being purged and you postulate 2AC-11 damper to not close (either facility 2 or mechanical failure), then fans 34A, B, and C (main exhaust fans) will have a direct suction to the Enclosure Building atmosphere (2AC-08 opens on CIAS) and will result in an unfiltered release which may exceed 10CFR100 limits for offsite dose - post LOCA.
L
i - +
June 6,1995 DE2-95-472 !
Page 3 of 8 ;
l Design Basis Investigation Since plant start up in 1975, there have been inconsistent interpretations of design basis for the CEBPS and the Enclosure Building' and their components.
This is evidenced by different NRC submittals and design changes. . I will only discuss ' the final determinations.
The CEBPS is designed to ventilate the Enclosure Building (all modes) and the Containment (modes 5 and 6 only). It was purchased and installed per Bechtel spec. 7604-M526 as non-QA, and non seismic. The portion of the CEBPS which penetrates the containment and ties to EBFS was purchased and installed per spec. 7604-M506 as QA and seismic class I. The purge fan F23, was purchased ,
e per Bechtel spec. 7604-M504 as non-QA, non seismic. The Bechtel Plant Data Book last revised in 1977 supports the same determinations.
Post 1977, dampers which isolate the CEBPS from the Enclosure Building (2AC-01 & 11) were upgraded to QA status. This was to accomplish the isolation .
function from CIAS signals post LOCA. Fan F23 gets a CIAS shutdown signal and thus it's breaker and controls are QA. t The EBFS system was purchased and installed per Bechtel spec. 7604-M534 ar.d 7604-M506 respectively, as QA and seismic class I. The EBFS tie to CEBPS is QA and seismic class I also. .
The Enclosure Building was not in the 1973 MP-2 plans. It was added at the request of the Atomic Energy Commission as a measure to reduce offsite.' doses l post LOCA. The building was designed to be seismic class I. During the latter l part of MP-2 construction, many Enclosure Building penetrations were -
designed and installed non seismic. I In September 1977 NNECO informed the NRC of a fan penetration in the Enclosure Building that was not seismic. They considered it 'a reportable situation. The next year they realized more penetrations were not seismic, and made seismic design improvements. Finally in 1979, ref. (8) clearly defined the Enclosure Building design basis. Although the building was seismically designed, many of it's penetrations are not seismic. After a seismic event (SSE), it will not maintain negative pressure in the Enclosure Building Filtration Region. The sheet metal siding may be damaged and some penetrations may fail. Furthermore, this is within our original ;
licensing basis. Therefore EBFS may not be operable for LOCA mitigation post SSE. Justification for clearly separating the SSE event and LOCA mitigation is the NRC staff's Safety Evaluation Report, section 3.9 for the MP-2 operating license and the NRC NUREG CR-1889 which determines the !
coincident occurrence to have a probability of 1.8x10-12 Reference (9)
I solidified this interpretation to our Licensing basis in 1981.
The FSAR section 5.3.4 assessment of through-line leakage, stated that single-failure criterion is only applied to seismic class I components.
i i
d
/ o.
June 6,1995 DE2-95-472 Page 4 of 8 A review of NRC Questions and Answers applicable to the CEBPS and Enclosure Building during the plant operating license process in 1973-1974 clearly state our position:
Ostn 5.39
"... assess through line leakage from the , containment which may bypass the Enclosure Building."
Answer NNECO stated the following assumptions to postulate the scenarios;
- There is either a seismic occurrence and all non seismic lines are broken, or there is not a seismic occurrence and all non seismic lines are intact.
. The single failure criterion applies only to seismic class I components. ;
i Purge dampers 2AC-03 and 08 were credited for release CEBPS line leakage l (from the containment penetrations) to the Enclosure Building. ]
Ostn 6.17
"...specify containment isolation valves 2AC-06 and 07 Technical Specification leakage limits assuming either 2AC-08 or 03 fail to open (to vent leakage gasses to the Enclosure Building)." .
Answer NNECO discussed expected leakage past the containment isolation valves of 4.8 scfh, and expressed that this wasn't a concern since it was less than 0.1% of that assumed for off site boundary dose analysis. Then they stated,
However, damper leakage (2AC-06 or 07) is considered to be released to the Enclosure Building Filtration Region even with the failure of dampers 2AC-03 or 08 to close.". This is apparently a typo at the end, since the assessment is 2AC-03 or 08 "in the closed position".
Ostn 6.15.4
.. demonstrate flow in purge lines will be inward following a LOCA including failure of AC-01 or 11."
Answer NNECO calculated for AC-01, that with 2 EBFS fans running, that flow would still be into the Enclosure Building through the 48" open damper. Then they stated that this was more conservative than the 2AC-1I scenario.
Review of the NRCs Safety Evaluation for MP-2, dated May 10,1974 came up with the following sections which contribute to our licensing basis:
Section 6-20 " Based on our review of the proposed design and predicted performance of the EBFS, we have concluded that the system meets the intent of the GDC 41, 42, 43, and 64."
Section 7.3 " Engineered Safety Feature Actuation System
i +
i June 6,1995 ;
DE2-95-472 Page 5 of 8 The Unit 2 engineered safety feature actuation system (EBFAS) is functionally identical to the Calvert cliffs system, except for two additional j actuation channels: (1) an enclosure building filtration actuation channel, '
which is actuated automatically by a safety injection actuation signal or by manual actuation from the main control board, and (2) an auxiliary exhaust actuation channel, which is actuated by Jiigh radiaticn in the fuel handling area or by manual actuation from the main control board. The applicants have documented that this system is designed and is being constructed in accordance with IEEE-279. We have evaluated the documentation of the electrical diagrams and conclude that the designs are acceptable."
Section 7.4 " Bypass Status of ESF systems Unit 2 has included a bypass safety status panel to satisfy the intent of Regulatory guide 1.47. In addition to the position indicating lights for valves, pumps, fans, and dampers, each safety related equipment item, which is automatically initiated to satisfy safety functions, is provided with '
a white and blue status light. These lights are located on the safety status panel and are grouped according to their safety function. Normally all the panel lights are off. :
The white light indicates the availability of the control circuit '
and is arranged to energize whenever power to control circuit is lost for any reason . including a blown fuse, tripped or racked out circuit breaker, loss of power, or an equipment item that is administratively bypassed for maintenance.
The blue light indicates that the equipment item is in the safe position or safe operating mode, and therefore, all blue lights in safety function group should be lit when the safety actuation signal exists. Thus, it will be readily apparent to the operator if any of the equipment is not in the safe mode for the safety function required. This design is acceptable."
Probabilistic Risk Assessment The PRA group was asked to evaluate the safety significance of the single failure deficiencies. They produced reference (5), " The Single Failures of EBFS and Their Impact on Public Safety".
PRA concludes the following:
- The public safety impact associated with these single failures is ;
negligible. The benefit determination, when averted person-REM is l used, shows a benefit of $60 over the remaining plant life. (Projected !
plant cost for the corrective design changes is about $400,000).
- Due to the significance of maintaining the functionality for EBFS for design basis events, we recommend " Negligible Risk Significance" as a basis to not perform expensive modifications here. They recommend i l
compensatory actions in light of the single failures-
- AC-Il---Trip main exhaust fans or shut 2AC-08 both from the control room
E l s' e-1 l
June 6,1995 DE2-95-472 Page 6 of 8
- AC-01---Manually start EBFS fan 25A from control room. l Additionally, the single failure scenarios discussed earlier can only occur when the plant is at power and is ventilating the Enclosure Building. This is an infrequent plant operation and is only performed at power, when the ,
Enclose Building gets too hot for comfort. 1,994 the Enclosure Building was only ventilated for 600 hours0.00694 days <br />0.167 hours <br />9.920635e-4 weeks <br />2.283e-4 months <br />. This is 6.8% of the year. Therefore if the ,
calculation of ref(5) has the Core Damage Frequency reduced to 6.8% of the -
assumed 6.0E-6/yr; than the resulting $60 for the plants remaining life can be reduced to $4. t To reduce the risk of the single failures resulting in any significant complication, there are other actions that can be expected without procedure -
changes;
- If main exhaust is still running enough time after the LOCA when containment leakage is highly radioactive; the discharge will go to the ' i MP-2 stack. There rad monitor elements and control room alarms from instrument loop 8132 will tell the operators of the unfiltered release '
condition and they will secure main exhaust fans. '
- If AC-Il sticks open, post LOCA, and main exhaust fans continue to pull air from the Enclosure Building, the supply will quickly dwindle to ,
negligible amounts as the EBFS fans will start to pull 13,900 cfm until vacuum results in the Enclosure Building. At this point, the design in-leakage into the Enclosure Building Filtration Region will allow only about 2500 cfm. The greater suction capabilities of the EBFS fans will remove most of this leakage. Main exhaust fans have suction demands satisfied by other sources (Auxiliary Building, condenser air removal, fuel hall). !
Also, indication of dampers AC-1 & 11 position and EBFS fans A & B status is shown on control board C0lX " Safety Status Panel". The operators will have the indication of the postulated ' wrong' accident positions, although we're not taking credit here for any immediate actions on them.
Engineering Evaluation ;
in assessing how original design could overlook so large an oversight as the '
single failures of AC-1 and 11; it becomes apparent that it wasn't so large an oversight but more a position taken as the result of evaluation of integrated plant systems response and risk significance. ,
3 1
June 6,1995 DE2-95-472 Page 7 of 8 l
i Reasons that come up to address why AC-1 an 11 weren't single failure proof )
designs are:
- The single failures postulated for AC-1 and 11 are only possible when the plant is ventilating the Enclosure Building. The original design may have taken credit for this operation being an infrequently performed evolution and thus not necessary for single failure design philosophy.
The Enclosure Building purging while at power is still an infrequently operation as seen by last years 6.8% occurrence of the operation.
This alone reduces the vulnerability to these single failure occurrences by a f. actor of ten.
. NNECO evaluated that minor leakage past CEBPS containment isolation valves AC-06 and 07 would vent to the Enclosure Building upon a AC-08 I
failed close. This is reasonable as the ducting is about 100' in length t between the containment and the Enclosure Building exit and is low pressure, SMACNA, non seismic ducting. This type ducting normally leaks much higher flow rates than the few cfm from the containment isolation valves.
- Containment isolation valve leakage was estimated to be a very low % of overall offsite dose leakage.
- The original electrical single failure of AC-1 was only a damper failure.
We are assuming a much more conservative failure of one entire ESAS cabinet, resulting in AC-1 remaining open, EBFS fan A not starting and A diesel generator not starting. i The PRA report suggests minimal benefit to performing modifications which ;
will cost about $400,000 to ensure 2AC-01 and 11 are single failure proof.
There is reasonable evidence that the worst case releases postulated will not ,
happen. This is further suggested by the NRC's recent relaxation of LOCA ,
source term's initiation from t=0 to t=30 minutes post LOCA.
i Looking at the scenarios of events, post one of the two single failures described above, it can be entirely expected that the operators will accomplish the reasonable steps required from their indications and existing procedures.
l
., v, [
t- .. ;
June 6,1995 !
DE2-95-47 Pagegof, t
Therefore it is recommended by the assessment of " Negligible Risk Significance to Public Safety" that the AC-1 and 11 proposed modification to !
ensure single failure provisions, not be installed. The plant is adequately and '
safety designed to mitigate the consequences of a LOCA and is at no further
- risk now than previously expressed at plant original licensing. '
l Prepared By Independent Review of Design Basis Information m f $ $ huh h*b'$5 f- j i/- r - -'
Signatures belo w indicate concurrence with the evaluation I c o n clu sio n s.
s PRA Supervisor Safety Analysis Supervisor
$ st / / NW Asbe $<r l p Ann Radiological Assessment Supervisor s/6 sw Licensing Supervisor
-nmywg ;
/ h $N0l0 M bl W WA Ol$
cc:
[
Pli Baumann M O'Meara D Leach J Creamer G Bouchard M Gentry M Ahearn R IIaley J Riley EA DeBarba 1
y~
,W ' . .
I CAcc l ) .
N:rtbcest
% }g NuclearEnergy Ferry Rd. (Route 156), Waterford, CT 06385 Mastone Nuclear Power station Northeast Nuclear Energy Company P.O. Bor 128 '
Waterford, CT 06385-0128 .
l (203)444 -4300 :
Fax (203)444-4277 ,
The Northeast Utsties System Donald B. Mmer Jr.,
senior v r,s.ideni- um. tone Re: 10CFR50.73(a)(2)(v) ;
' March 10, 1995 1 MP-95-080 i U.S. Nuclear Regulatory Commission Document Control Desk Washington, D.C. 20555
Reference:
Facility Operating Ucense No. DPR-65 Docket No. 50-336 :
Licensee Event Report 94-040-01 ,
This letter forwards Licensee Event Report 94-040-01 required to be submitted within thirty (30) days pursuant to 10CFR50.73(a)(2)(v).
Very truly yours, NORTHEAST NUCLEAR ENERGY COMPANY FOR: Donald B. Miller, Jr. !
Senior Vice President - Millstone Station :
BY:
Michael H'. rothers !
Millstone Unit 3 Director i i
DBM/PHB:ljs i
Attachment:
LER 94-040-01 '
cc: T. T. Martin, Region.1 Administrator ,
P. D. Swetland, Senior Resident inspector, Millstone Unit Nos.1,2, and 3 G. S. Vissing, NRC Project Manager, Millstone Unit No. 2 f
L e
I i
b00l ^
~
NRc Form 366 U.S * CLEAA REGULATORY COMMISSION Ar,OVED BY OMB NO. 3150- 0104 ;
, ' (5- 92). EX~1RES: 5/31/e5 ESTlWA'ED SJ8L ...isE3 RESPONSE TO COW *LY wf'M Tras pdFORuAtos OOLLECTION MOJEST: to.o MRS Fon*ARD COMMENTE REGACPdG
., UCENSEE EVENT REPORT (LER) Wg 55A' 't,,3 eWASHINGTON . DCw,l*=3ll' 3386-00o1. ^"a,*g5ANDagagMg TO THE pnpEmwC
(&se reem for recuned nwrieur et egeushmeseus tw each back) DC
~ '
FACUTV esAnst (1) 00CetET seutstEn (8) PaGE (3)
Millstone Nuclear Power Station Unit 2 05000336 1 OF 5 Ventitation Design Deficiency Affecting Enclosure Building integrity EVENT DATE (5) LER NUMBER (6) REPORT DATE (T) OTHER FACILITIES INVOLVED (e)
F M hAME esocM DAY YEAR VEAR SI0d@
gg $$ es0s(TH Osv VEAR DOCKE1 hswaER 05000 -
FACeV 4AnsE DOCEET h.asBER 12 6 94 94 -
040 - 01 03 10 95 05000 opeRAtmo , THIS REPORT 18 SEING SUBMfTTED PURSUANT TO THE REQUIREMENTS OF 10 CFR 4: (Check one or more) (11)
"'l to ermi e ovei ao um>mte is timi POWER Po4D6isH1H4 W mie)r1) X to 73teimM 73 71se)
WL 08) 0 ,,,,,,,,, ,, ,,, ,,,,,,y, o,,,,
re enmm no me>mm so nummA) <s segge,
- ,o ..m , io meme .e n .3m m ., W=w
,0...,,~, . . . . . . .o ne..,
LICENSEE CONTACT FOR THIS LER (12) sannsE TELE **0NE NUWsEn twm aw Cow Philip J. Lutzi, Nuclear Ucensing (203) 440- 2072 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBEo IN THl$ REPORT (13)
REPORTABLE REPORTABLE CA2E SYSTEM COW 80NEN* WANurACTURER TO NPCS CAJEE SYSTFM COW 80NENT RAANUFAC'URER TO N#CS SUPPLEMENTAL REPORT EXPECTEo (14) EXPECTED
"# D" **
YEs no SUBM!sslON ryn como = DeicitD ssswissoN DatE) X DATE (15) j AssTRACT (Lens .1 ao enee 4. aoronmeev is es,m-esecan erse. ari nissi os)
On December 6,1994, at 2223 hours0.0257 days <br />0.618 hours <br />0.00368 weeks <br />8.458515e-4 months <br />, with the plant defueled, it was determined that a release path existed from the Enclosure Building that would allow for a direct discharge to atmosphere following a Loss of Coolant Accident 1 (LOCA) that would not receive charcoal filtration.
Further investigation revealed that there were other postulated single fallars scenarios that resulted in a release path from the Enclosure Building that would allow a direct discharge to the atmosphere without charcoalfiltration following a LOCA if Enclosure Building Purging operations were being performed.
The root cause is a deficiency in the original design.
h4C Form M (5-SO
l
- 4(RC F 366A U.S ; LEAR REGULATORY COMMISSION ysD BY OMS NO. 31s0-e104 A ;
P- -
EXPIRES: s/31/ss .
UCENSEE EVENT REPORT (LER) c "oTJS."=5W."f3" 7allllllFe% FEY.*."?. !
TEXT CONTINUATION **"cll WnRTET "m*al"o"4' M,Ts' en i ll= LiFsWh"* ' Jalr.J' L"*=a mas.eRoran.ac mens..
l recam hWE M DOcE7.madsER e) L5R 84AdsER S - RAGE GR i
j vs R "?.fs:F 'EMT !
Millstone Nuclear Power Station '
Unit 2 05000336 94 - 040 -
01 02 OF 05 rErr . - . .- - 4c,-- = nn
- 1. Description of Event i On December 6,1994, at 2223 hours0.0257 days <br />0.618 hours <br />0.00368 weeks <br />8.458515e-4 months <br />, with the plant defueled,it was determined that a release path !
existed from the Enclosure Building that would allow a direct discharge to the atmosphere during a Loss of !
Coolant Accident (LOCA) that would not receive charcoal filtration. The cause of this event has been i determined to be an oversight in the original design of the discharge flow path for the Hydrogen analyzers.
With the establishment of the system engineering program, the engineer reviewing a work package immediately identified the discrepancy in this non-safety related system and initiated an investigation. l The design basis of the Enclosure Building Filtration System is to collect any leakage from the Containment Structure during a LOCA and process the leakage through a High Efficiency Particulate (HEPA) and Charcoal Filtration system. This method of discharge minimizes the publics exposure to lodine and maintains off site dose less than 10CFR100 limits.
A hydrogen analyzer cabinet and sample hood exhaust fan was found to take a suction on the enclosure I building and discharge approximately 1000 cfm out the Unit 2 Main Exhaust stack. This flow path has .
HEPA ft!!ers but does not have any Charcoal Adsorber filtration. This non-safety related exhaust fan i normally runs to maintain a negative pressure on the sample hood to prevent technicians from being exposed to gas while obtaining routine chemistry samples. The fan has no automatic shut off feature and 4 there are no isolation dampers in the line to prevent a release during an event that would actuate the !
Enclosure Building Filtration System.
- i The Radiological Assessment branch performed an evaluation to determine the effects of this condition. '
Their analysis was based upon a major accident assuming a substantial meltdown of the core with r subsequent release of appreciable quantities of fission products as identified in 10CFR100.11 and <
concluded that the calculated site boundary thyroid dose would exceed 10CFR100.11 limits. l
. \
Following the discovery of this condition on December 6,1994, immediate corrective action was to declare the enclosure building integrity inoperable. The plant was in an undefined mode due to the core being off loaded when the discrepancy was found and declared inoperable. Enclosure Building integrity is not required in Mode 5 or 6, therefore, no additional operator actions were required. '
Further investigation of ventilation systems with penetrations into the Enclosure Building resulted in additional findings. On February 9,1995, at 1300 hours0.015 days <br />0.361 hours <br />0.00215 weeks <br />4.9465e-4 months <br />, with the plant defueled, a potential design deficiency in the enclosure building purge system was identified. In the event of a single facility or component failure, a release path from the Enclosure Building would allow for a direct discharge to the atmosphere without charcoal filtration following a LOCA if Enclosure Building Purging operations were being performed.
Completion of the investigation revealed that there were three system configuration discrepancies. It is important to note that in order for any of these unsatisfactory conditions to exist, Enclosure Building Purge operations must be in progress coincident with the Design sasis Accident and a single facility or component failure must occur.
1 es4C% 336A (6 32)
_- a
t 0 U.S LEAR REGULATORV COMM'SSION A DVED BY OMB NO. 31s0-0104 NAC (6- 9 g)onn 366A EXPREs: s!31/95
' LICENSEE EVENT REPORT (LER) M S?o? % "# E"&"s" b ,"X E s l % "#Js TEXT CONTINUATION E" .**.h'?/h*".,x2 ToIf o"d' MU'o!
- encu.r E'eieS*m man, cow.ocmean.7 oTc/*!w' E *ut O E'8x2 FAC4.rfy seams (il DOCMET De@eBER (t) 4.ERN M R(S) peGe (3 veam E." "Jl..3 Millstone Nuclear Power Station Unit 2 N 94 -
040 -
01 03 OF 05 TEXT tv mere sence s eeocea. ime esecoes' enose er NRc rare assA) rirl Scenario AC-11 AC-8 F23 AC-1 AC-3 EBFS Results Dual CIAS Closed Open Off Closed Open F25A On Enclosure Building integrity Satisfied.
Actuation F25B On Exhaust path is filtered. Offsite dose is less than 10CFR100 limits.
Facility 1 Open Open On Closed Open F25A On Main Exhaust is taking a suction on the CIAS F25B Off enclosure building through AC-11 and initiation. AC-8. No Filtration, Off Site Dose Facility 2 greater than 10CFR100 Limits.
CIAS Fadure Facility 2 Closed Open Stopped Open Open F25A Off The EBFS will be attempting to draw a CIAS For For F25B On negative in the enclosure building with initiation. Purge Purge the outside air damper (AC-1) open Fecility 1 and no outside air isolation. Per FSAR CIAS Failure 6.7.4.1, two EBFS fans are required to meet the design assumptions with AC-1 open. Not Evaluated, reduced :
effectiveness, release path is filtered. !
With dual Open Open Off Closed Open F25A On Main Exhaust is taking a suction on the actuation and F25B On enclosure building through AC-11 and AC-11 sticks AC-B. No Fittration, Off Site Dose !
open. greater than 10CFR100 Limits.
With dual Closed Open Stopped Open Open F25A On Exhaust path is fittered. Offsite dose is actuation and F25B On less than 10CFR100 limits. Reference AC-1 sticks FSAR 6.7.4.1 open.
(See attached sketch) l l
Data was co!!ected for the period in 1994 when Millstone Unit 2 was in modes 1-4 and it was identified that enclosure building purge operated for approximately 10 percent of that time.
Immediate operator response was not required since the plant was defueled and Enclosure Building integrity is not required in this mode.
There were no automatic or manually initiated safety systems actuated as a result of these events.
- 11. faute of Event The root cause of the hydrogen analyzer event is the design and installation of the hydrogen analyzer cabinet ventilation system.
The root cause of the Enclosure Building Purge deficiencies is the original design of the system. The Enclosure Building purge system was not property designed for single failure, coincident with purgin operations. The system does have isolation signals to individual components in the flow path. However, the system is not protected against a single failure of a component of train of equipment.
==c % ae6a n.m
Form 366A U.S. .; LEAR REGULATORY CoMMtSSION A. JVED BY OMS NO. 3190-0104 NR'N
- - EMMREs: s/31/Ds
,, UCENSEE EVENT REPORT (LER) cE 8siE"8 "oIE*s"foEE ** C ." M O TEXT CONTINUATION E' m D f'w"e'taAn ToIfEv' U E"5 l
01 04 0F 05 TEXT te nione essa e ren.nes. ame sesso.us amose er Nec 7erm assn pra 111. Analvsis of Event ,
Based on event investigation, this condition is reportable under the criteria of 10CFR50.73(a)(2)(v),"Any event or condition that alone could have prevented the fulfillment of the safety function of structures or systems that are needed to Mitigate the consequences of an accident."
The Radiological Assessment branch performed an evaluation to determine the effects of the hydrogen analyzer condition. Their analysis was based upon a major accident assuming a substantial meltdown of the core with subsequent release of appreciable quantities of fission products as identified in 10CFR100 i and concluded that the calculated site boundary thyroid dose would exceed 10CFR100 limits. This configuration has existed since initial plant construction and startup.
The Radiological Assessment branch performed an additional evaluation to determine the effects of the enclosure building purge condition. Their analysis was based upon a major accident assuming a substantial meltdown of the core with subsequent release of appreciable quantities of fosion products as identified in 10CFR100 coincident with Enclosure Building purge operations, a single failure of a facility or component and significant leakage from containment into the enclosure building. The Radiological Assessment branch concluded that the calculated site boundary thyroid dose would exceed 10CFR100 limits if the release went undetected.
f IV. Corrective Action i
Following the discovery of this condition on December 6,1994,immediate corrective action was to declare l l
the Enclosure Building integrity inoperable. Since the plant was defueled when the discrepancy was found and Enclosure Building integrity is not required in Mode 5 or 6, no additional immediate actions were required.
Work is in progress to redesign the hydrogen analyzer and sample hood ventilation system to correct this i deficiency prior to Mode 4 when enclosure building integrity is required. Work is also in progress to correct the purge system so that the enclosure building integrity is maintained following all credible system or component failures.
V. AdditionalInformation f Similar LERs: None ElIS Codes Hydrogen Analyzer Cabinet IK-CAB IK-FAN Hydrogen Analyzer Cabinet Fan Containment Leakage Control System BD Reactor Containment Building NG Plant Exhaust System VL i
N40 89Pm MA (S-92)
r-
~
i . .
~
{ E a:
?
= a .g Main Exhaust to na Unit 2 Stack *I cr i
~;
AC-11
' n !"R5F 5 8
C 7 i L25
! a 4
ckna m ._
L Facility 2 i l 5 CIAS 2
i ! lN ogg s
l $
E g $t
-H W
% o- -e m iCmS f5 h' N i g sa AC-7 / EBFS to 2g urmisme m 5
= Chertoef o l 3 a Filtn W e h!g !M e
.s g 5A OAM G1 ams, e., o- - )
hj y.
recar i CLAS Ci4S
-r a r4 VSepse ,
i e, . tiEispi i-Fadity2 CIAS o
!), l: 3E lga ENCLOSURE BUILDNG j l Jg 5 o, fb9 AC4, AC-5, AC-6& AC-7 ,
i9 arenomasdesshnases14 g Nf 4 5 g ijg,lh g O E I Ah 8 j!$
1 h99 #
._1 ll
INTER OFFICE MEMO oxTE: 9-2 M S Millsbe LM4- Wo. 2.
TO! G .1 m r;- d DEPT. - IDCATION: h elaar Unit Director FROM: b .7. VGa O dEnnA DEPT. - IDcATIONt EOb N MI
SUBJECT:
A5 -i (H02) "bnwJ VEsr tmtk ra Cmr/ Euct. "b.D 7urLR MESSAGE The technical evaluations on the subject potential reportable item have been completed.
g, ,4 NL recom= ends that this item D reportable. %+W'#^
2hKs.
Please review the REF, complete Section 5.0 of the attached form NEO 2.25, and route the completed RET to each of the REF participants. ,
"R . Kr. c ich cc: REFFILE(51'>) g g m ,3 . v/.is.o n 2 36mann x d W. N4Chim h.Lt.d:L '. y E ,7 g M Kc. , ,n q lc..s T
G Neeq3 oRICINATOR - DO NOT WRITE BELoW THIS I,2NE SIGNTDs g i
REPLY DEPT.-LOCATION: SIGNED: DATE:
cxn PLEASE RETURN COPY MTTH REPLY
i i
5.0 : DIRECTOR NUCLEAR UNIT _ Date Received '! 3 'l 5 5.1 The item has been determined to be, i Not Reportable Section Nofs).
t[ Reportable per 10CFR50.72 10CFR50.73 Min ',O ' !2IN"d I.U'di / int / 6 b'i 3 i
~ ~
' ~~
10CFR50.9(b) l Operable i Not Operable 5.2 If the item is determined to be reportable, follow applicable station procedures l for NRC reporting requirements.10CFR50.72 items are verbally reported within i 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />; 10CFR50.73 items are submitted under Director Nuclear Unit signature ;
within 30 days; and 10CFR50.9(b) items are verbally reported within 2 days .l then submitted under corporate officer signature within 2 weeks. -;
, : ~
Applicable. LER-Number -. '5- 3L !
5.3 Basis for Director-Nuclear Unit decision (if any differing views exist): j Attach or Note Basis:
i 5.4 File the original REF/OE with the PIR, and provide a copy to the Manager.
Nuclear Licensing, the NSAB Chairman, the individual who identified the issue ,
resulting in the initiation of the PIR, and those individuals who provided an Operability or Reportability Evaluation,
')
j f -
h l~ $
Djp6cWNubear Unit Date -)
i l
l i
NGP 2.25 Rev.8 m e.NEo Page 7.I-3of 3
~
w f -
Northeast - ,
g ,".9gj qge, 1 Nuclear Energy Memo
+
b _
April 28,1995 i DE2-95-254 To: Distubsf I o %
From: T. A. yle
Subject:
Meeting Minutes, April 6,1995, on AC-01 a'nd 11 design problems ,
for Secondary Containment Integrity.
Attachment:
(1) S. & W. report " Enclosure Building Purge System and Enclosure Building Filtration System Design Evaluation."
(2) Action Plan: AC-01 & 11 Design Solution.
(3) Secondary Containment Integrity Problems Sketch, i (4) Evaluations of proposed solutions. l Meeting Obiectives: ,
A meeting was held after the receipt of SWEC's final draft of attachment (1). ;
The meeting objectives were to:
(1) Briefly reiterate the Secondary Containment Integrity problem ,
(2) Review and reconcile comments from the attendees review of attachment (1).
(3) Evaluate the different solutions which were proposed !
and entertain new ones. '
(4) Decide as a team on the favored solutions (options).
(5) Develop a plan of action to best correct the problem with minimal impact on the RFO schedule and present to MP2 management, f Background and Problem Descriotion The problem was first reported with PIR 2 95-126 as a potential design flaw condition from original construction. It appeared that there was an unanalyzed breech of secondary containment integrity through dampers 2-AC 01 AND 2 ACll. Attachment (3) depicts a schematic showing how three plant ventilation system; Containment and Enclosure Building Purge System, ,
Main Exhaust System, and Enclosure Building Filtration System, all ventilate the enclosure building.
4
e Page 2 of 4 DE2 95-254 April 28,1995 After a LOCA the EBFS system must take suction on the enclosure building and draw it down to a vacuum. The other two systems must isolate as necessary to support the building integrity and vacuum.
Problem ( 1 ). 2 AC-01 : If the single failure after a LOCA is postulated to be a loss of facility 1 CIAS, then AC-1 will not close and F-25A will not start.
EEFS (F-25B) will not be able to draw a required vacuum in the enclosure building with AC-01 stuck open. This is a FSAR and Technical Specification requirement to limit off-site doses after a LOCA.
Problem (2). 2 AC- 11 : If the single failure after a LOCA is postulated to be a mechanical failure of 2AC-11, stuck open, then F-34 fans (1,2,or 3) will draw on enclosure building air and discharge through the MP-2 stack.
This will constitute on unfiltered exhaust from the site and analysis shows that our 10CFR part 100 offsite dose limits will be exceeded. .
SWEC was asked to do detailed design basis investigation to independently assess the situation. Attachment (1) reports on the investigation, evaluates the findings, and proposes solution options for the two problems.
Renort Com ment s The group provided good comments on the report contents specifically:
MP-2 single failure licensing criteria was added (IEEE 279/271),
Technical Specifications requirement for 60 second draw down was added to the report.
Ray Crandalls evaluation of LLRT leakage was moved to the evaluation section.
Clarification's were made with respect to non-QA and QA equipment assumptions for operations after t=o for the accident and single failure analysis.
Added applicable specifications for EBFS and the Enclosure Building.
Seismic design and installation specifications at F-23 area should be verified.
Evaluation of the Options Attachment (1) details the different proposed solutions. In addition to the options in the report, an alternative was suggested involving operator action.
The NRC will now allow a 30 minute delay from time t=0 of the accident, to the :
time where the radiological source term must be assumed.
l l
- j. ,
Page 3 of 4 DE2-95-254 April 28,1995 i
Using this allowance, the EOP's could be revised to require an operator to manually shut and verify 2AC-01 and 2AC-II. The operator action would thus establish secondary containment integrity before the source term was realized. ,
The group assumed that MP-2 management would prefer a design modification solution versus the operator action solution.
The five design solutions were evaluated as shown on attachment (4). the major considerations to the evaluation were:
The solution must be technically sound and reliably re established secondary containment integrity upon a single failure,
+
The solution should have minimal impact on the MP-2 RFO and Start. a up schedule. Enclosure building integrity is required in modes 1,2,3, and 4.
Assuming that the majority of the modifications will be implemented while the plant is operating, the retest for the specific solution should be one the can be done while the plant is on line.
Decision and Action Plan The team made a decision to elect option three from the report. This involves:
(1) A new safety related counter balanced gravity damper will be installed in the suction duct to F-23 (Purge Fan). This will open when the purge fan operates and when the purge fan shuts down, the damper will close.
It will be weighted such that EBFS will be able to establish it's design vacuum without the damper opening. F 23 calculations and operating curve were analyzed and the fan should pose no problem to the fix.
(2) 2AC08, which presently opens on a facility one CIAS signal, will be redesigned to shut on the CIAS signal. This will require a new operator for 2AC08, and some minor switch rewiring at the main board in the control room. The damper for 2AC08 will also be replaced as it presently is a field fabricated item and would pose a problem if we tried to reuse it.
This solution requires that the LLRT leakage through 2AC07 and 2AC08 be evaluated as an additional release during the accident. This has been evaluated conceptually to be insignificant.
Implementation of these two solutions was discussed. It is believed presently, that both modifications can be performed while the plant is on line. The wiring at the main board may be elected to be done prior to start up.
The new qualified equipment has lead times as long as 15 weeks (dampers).
However, we believe we can cut that time in half with negotiations with the vendors.
i- .
- e Page 4 of 4 '
DE2 95-254 April 28,1995 ;
i To support doing the construction on line and ensuring secondary j containment integrity two jumpers / bypasses must be installed prior to mode 4. >
They will cause 2AC-01 and 2AC-Il to fail in the shut position. This ensures .
integrity and will allow continued made changes. The result of the B/J's is that '
the enclosure building will not be able to be purged. P. Baumann explains that this won't be a problem until July based on last years experience. The dampers y could be opened and the building purged if dedicated operators stood by to l isolate 2AC-01 and 2AC-Il in the case of an accident. '
The testing requirements after these modifications will involve purge (F-23) fan and EBFS (F-25 A&B) fan operations and per Phil should not present a problem while at power. The retest for the logic change at ZAC-08 was discussed with Rick Halleck and Gus Filippides after the meeting. Presently, they believe the electrical retest can be done at power as well.
An action plan to confirm the results shown above and enable all activities associated with the project to be performed with minimal impact to the RFO schedule is shown on attachment (2). As conceptuali:ed now, this action ,
plan will not effect the RF0 critical path.
The plan is conceptual at this point and if management concurs with the decision made and action plan developed; than the plan will be implemented. !
Please call me if there are any questions.
TAD /Imj Distribution:
S. Sudigala R. Necci !
M. Gentry R. Grebasch .
W . Stairs J. Riley i M. Bain M. Ahern !
D. Dube M. Kai M. Fortini M. O'Meara G. Filippides J. Regan '
S. Wainio B. Price ;
J. Padden G. Bouchard Attendes:
L. Pazura J. Creamer B. Ekvall D. Leach R. Halleck R. Wells G. VanNoordennen R. Crandall A. Lassonde C. Dempsey P. Baumann
1
- ff'/7 W d d f /
. STONE 8 WEmSTER ENGINEERING CORPORATION 245 SUMMER STREET, BOSTON, MASSACHUSETTS 02210
,.a............, ........m.....
- 'ir"'..,. "!.*J!%"'.
- 'TI.' n" "' :::R:::"
i'"C.'s *h a as :;:2'."&;
- !..:.t"
'l,. " O . ..
Northeast Nuclear Energy Company April 11,1995 P. O. Box 128 Waterford,CT 06385
Subject:
Enclosure Building Ventilation Systems Design Evaluation
Dear Mr. Doyle,
Enclosed you will fmd Stone & Webster's evaluation of the enclosure building ventilation systems in cccordance with your work release dated March 8,1995. The report provides an independent evaluation of the design, areas of the design that warrant further review / investigation, and a recommended course of action.
The attachment is a facsimile. I will forward the original document to you in a few days. We look forward to our continued support of Millstone Unit 2 in this matter.
Very Truly Y.ours.
/
' ) WP' k Don' Leach Project Manager Enclosure
]
e l
is e e
- s T o N E & W E B s T E R 198 6
..