05000354/LER-1996-028-01, :on 961220,potential Loss of Containment Integrity Occurred Due to Water Hammer in Drywell Cooler Piping.Administrative Controls Were Implemented to Prevent Defeating of Drywell Cooler Isolation Interlocks
| ML20133P512 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 01/20/1997 |
| From: | Priest J Public Service Enterprise Group |
| To: | |
| Shared Package | |
| ML20133P508 | List:
|
| References | |
| LER-96-028-01, LER-96-28-1, NUDOCS 9701240304 | |
| Download: ML20133P512 (10) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(viii) 10 CFR 50.73(a)(2)(ii) 10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2) 10 CFR 50.73(a)(2)(iv), System Actuation 10 CFR 50.73(a)(2)(v), Loss of Safety Function |
| 3541996028R01 - NRC Website | |
text
NRC F*ORJ 366 U.S. NUCLEAR REGULATORY COLISSION APPROVED LY L MS NO. 3150-0104 (5 96)
EXPlRES 04130*Z8 EA E O0Y 'N0RMA 0N COLL ON RE 6
HRS LICENSEE EVENT REPORT (LER)
UC;NE8 MSE?U Mo^""!! ^n 78 iJUs % 'fo8wl"!
AND RETSR E IAN E
BR CH (T4 33) U NUCL (See reverse for required number of g Gu g Ry gigN g As g g. g g s gggo g digits / characters for each block)
MANAGEMENT AND BUDGET, WASHINGTON. DC 20503.
F ACILif y NAME (1)
DOCKET NUMBEM (2)
P AGE (3)
Hope Creek Generating Station 05000354 1 OF 4 T4f LE 14)
Pot:ntial Loss of Containment Integrity Due to Water Hammer in Drywell Cooler Piping EVENT DATE (5)
LER NUMBER (6)
REPORT DATE (7)
OTHER F ACILITIES INVOLVED (8) mouth oar YEAR sE,ougAt R,Ev;s,igu MOsTH oAv vCAR vCAR
' ^ ' ' ' " " * " '
00 1
20 97 12 20 96 96 028 05000 OPERATING 1
THIS REPORT IS SUBMITTED PURSUANT TO THE REQUlHEMENTS OF 10 CFR 5: (Check one or more) (11)
MODE (9) 20.2201(b) 20.2203(a)(2)(v) 50.73(a)(2)(i)(B) 50.73(a)(2)(viii)
POWER 100 20.2203(a)(1) 20.2203(a)(3)(i) 50.73(a)(2)(ii) 50.73(a)(2)(x)
LEVEL (10) 20.2203(a)(2)(i) 20.2203(a)(3)(si) 50.73(a)(2)(m) 73.71 20.2203(a)(2)(ii) 20.2203(a)(4) 50.73(a)(2)(iv)
OTHER 20.2203(a)(2)(m) 50.36(c)(1) x 50.73(a)(2)(v) sgegigbetragicw 20.2203(a)(2)(iv) 50.36(c)(2) 50.73(a)(2)(vn)
~
LICENSEE CONTACT FOR THIS LER (12)
UAE TELEPHONE NUMBER (include Area Code)
~
Jim:s Priest, Lead Engineer - Licensing and Regulation (609)339-5434 COMPLETE ONE LINE IOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)
CAUIC sv8 TEM COMPONENT MANUFACTURER REPORTA0 E
CAUSE
SYSTEM COMPONENT MANUFACTURER REPORTAR E SUPPLEMENTAL REPORT EXPECTED (14)
EXPECTED MONTH DAY YEAR YES NO SUBMISSION (if yes, Complete EXPECTED SUBMISSION DATE).
X DATE (15)
A28 TRACT (Lernt to 1400 spaces,i.e., approximately 15 single-spaced typewntten lines) (16)
On 12/20/96, Engineering personnel completed evaluations of water hammer concerns discussed in Generic Letter 96-06, " Assurance of Equipment Operability and Containment."
Based upon these evaluations, Engineering concluded that a potential exists for primary containment integrity to be adversely impacted when drywell cooler isolation interlocks are defeated in accordance with emergency operating procedures.
Specifically, these actions could result in a water hammer when attempting to place the drywell coolers in service under specific post LOCA drywell conditions.
Since the chilled water system containment penetration was not designed for the effects of a water hammer, containment integrity could be lost.
This issue represented a condition alone which could have prevented the ability to control the release of radioactive materials, and in accordance with the requirements of 10 CFR 50.72 (b) (2) (iii), a four hour report was made at 1606 hours0.0186 days <br />0.446 hours <br />0.00266 weeks <br />6.11083e-4 months <br />.
The cause of this event is attributed to an inadequate evaluation of the potential for creating water hammer loads in post LOCA scenarios as described in Generic Letter 96-06.
Corrective actions include implementing administrative controls to prevent the defeating of the drywell cooler isolation interlocks in post LOCA conditions.
9701240304 970120 PDR ADOCK 05000354 PM g
u
NRJ f@RJ 346A U.S. NUCLEAR RE.ULATORY COM 18SION (4 96)
LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION FACILITY NAME (1)
DOCKET NUMBER (2)
LER NUMBER C6) l PAGE (3)
YEAR sE u g
00 l 2 OF 4 Hope Creek Generating Station 05000354 96 - 028 -
TEXT (if more space is required, use additional copies of NRC Form 366A) (17)
PLANT AND SYSTEM IDENTIFICATION
General Electric - Boiling Water Reactor (BWR/4)
Drywell Cooler System - EIIS Identifier {VB}
IDENTIFICATION OF OCCURRENCE Event date:
12/20/96 Discovery date:
12/20/96 Date determined to be reportable:
12/20/96 Problem Report 961220079 CONDITIONS PRIOR TO OCCURRENCE Plant was in OPERATIONAL CONDITION 1 (POWER OPERATION).
Reactor was at 100% of rated thermal power.
There were no systems, structures or components that were inoperable at the time of the event that contributed to the event.
DESCRIPTION OF OCCURRENCE On December 20, 1996, the Hope Creek Generating Station Engineering personnel completed evaluations of water hammer concerns discussed in Generic Letter 96-06, " Assurance of Equipment Operability and Containment."
Based upon these evaluations, Engineering concluded that a potential exists for primary containment integrity to be adversely impacted when drywell cooler isolation interlocks are defeated in accordance with Emergency Operating Procedure HC.OP-EO.ZZ-0102A(B).
Specifically, this Emergency Operating Procedure (EOP) directs the operator to place all drywell coolers in service (by defeating the containment isolation interlocks) when high drywell temperatures exist in post LOCA scenarios.
However, this action could result in a water hammer when attempting to place the drywell coolers in service under specific post LOCA drywell conditions.
Since the drywell cooler system containment penetration was not designed for the effects of a water hammer, implementation of the EOP could have resulted in the loss of containment integrity.
Since this issue represented a condition alone which could have prevented the ability to control the release of radioactive materials, a four hour report was made at 1606 hours0.0186 days <br />0.446 hours <br />0.00266 weeks <br />6.11083e-4 months <br /> in accordance with the requirements of 10 CFR 50.72 (b) (2) (iii).
On 12/20/96, administrative controls were implemented to prevent the defeating of the drywell cooler isolation interlocks during post LOCA conditions.
Since the design and licensing bases of Hope Creek do not--
s
n-NRJ FORJ 366A U.S. NUCLEA;1 REGULATORY COM.lSilON
(&M)
LICENSEE EVENT REI' ORT (LER)
TEXT CONTINUATION FACILITY NAME (1)
DOCKET NUMBER (2)
LER NUMBER 46) l PAGE (3)
DuiBYb EE 00 l 3 OF 4 Hope Creek Generating Station 05000354 96 - 028 -
TEXT (if more space is required, use additional copies of NRC Form 366A) (17)
DESCRIPTION OF OCCURRENCE (Continued) require the use of the drywell coolers to mitigate the consequences of an accident, no Technical Specification actions or other compensatory measures were required to be taken in order to implement the administrative controls.
ANALYSIS OF OCCURRENCE On 9/30/96, the NRC issued Generic Letter 96-06 to provide notification about safety-significant issues that could affect containment integrity and equipment operability during accident conditions.
These issues concerned:
- 1) susceptibility of cooling water systems serving containment air coolers to the effects of water hammer and/or two phase flow; and 2) over-pressurization of isolated water-filled piping sections in containment during LOCA or Main Steam Line Break (MSLB) conditions.
l The system which supports the dry' sell coolers at Hope Creek is a closed-loop system supplying cooling water to various non-safety related air handling units. In the event of a LOCA, power to the drywell coolers will be tripped automatically and the containment isolation valves will close.
As required by the Generic Letter, Engineering personnel have evaluated the above issues j
for impact on this cystem.
From this evaluation, Engineering concluded that the water hammer concern had the potential to adversely impact containment integrity under specific post i
LOCA/MSLB conditions when the drywell coolers are placed into service in accordance with Hope Creek's EOPs.
Engineering personnel determined that a water hammer could occur:
- 1) anytime the water in the drywell coolers is greater than 250 degrees F and the isolation interlocks are defeated; or 2) if water in the drywell coolers, which initially is less than 250 degrees F, is heated then cooled prior to defeating the containment isolation interlocks.
Due to the presence of relief valves on the piping in the drywell, water inventory would be lost as the temperature rises, and when the water cools, pressure may decrease below the saturation pressure.
These conditions would enable the formation of voids in the piping, which would result in a water hammer when the drywell coolers are placed into service.
Since the system containment penetration was not designed for these water hammer loads, containment integrity could be challenged during specific post LOCA scenarios.
APPARENT CAUSE OF OCCURRENCE Hope Creek EOP HC.OP-EO.ZZ-0102 directs the operator to defeat the chilled water system isolation interlocks to place the drywell coolers into service during post LOCA conditions.
This action was formally incorporated into the EOPs through revisions made in the late 1980s.
This EOP revision represented an opportunity to identify this problem, but the capabilityi
NRJFDR.5366A U.S. NUCLEAR RE2ULATORY COMilSSION (4-96)
LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION FACILITY NAME (1)
DOCMET NUMBER (2)
LER NUMBER 46)
PAGE (3)
B" WEA
=
Hope Creek Generating Station 05000354 96 - 028 - 00 4
OF 4 TEXY (if more space is required, use additional copies of NHC Form 366A) (17) to utilize this isolation interlock design feature had existed since plant startup.
Justification for this EOP revision utilized information from the approved BWR Owners Group Emergency Procedures Guidelines; however, the potential for creating water hammer loads in post LOCA scenarios was not adequately evaluated when the plant was designed.
The failure to properly perform an evaluation of the effects of water hammer loads is attributed as the cause of this event.
ASSESSMENT OF SAFETY CONSEQUENCES
In accordance with 10 CFR 50, Appendix A, Criterion 16, the containment design requires the containment and associated systems to establish an essentially leak-tight barrier against the uncontrolled release of radioactivity to the environment and to assure that the containment design conditions important to safety are not exceeded for as long as postulated accident conditions require.
Since the postulated water hammer could have resulted in the loss of this required containment integrity, there were adverse potential safety consequences associated with this event.
The postulated water hammer does not impact other systems designed to mitigate the consequences of an accident.
PREVIOUS OCCURRENCES
There have been no similar occurrences previously reported by Hope Creek.
CORRECTIVE ACTIONS
On 12/20/96, administrative controls were implemented to prevent the defeating of the drywell cooler isolation interlocks in post LOCA conditions.
To address the water hammer issue, either:
- 1) revisions to the Hope Creek EOPs will be made to address overriding of the containment isolation interlocks of the chilled water system; or 2) the affected containment penetrations will be reanalyzed or modified to accommodate the water hammer loads.
Final implementation of corrective actions to resolve this issue will be completed by the end of the next refueling outage (RFO7).
I Engineering evaluations of the impact on Hope Creek of the water hammer, two-phase flow and over-pressurization issues discussed in Generic Letter 96-06 have been completed and the results will be transmitted to the NRC (in the response to Generic Letter 96-06) by 1/28/97.-
t.-
O PSEG Public Service Electric and Gas Company P.O. Box 236 Hancocks Bridge, New Jersey 08038-0236 Nuclear Business Unit JAN 201997 LR-N97023 U.S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 2
Dear Sir:
a HOPE CREEK GENERATING STATION DOCKET NO. 50-354 UNIT NO. 1 LICENSEE EVENT REPORT NO. 96-028-00 This Licensee Event Report entitled, " Potential Loss of Containment Integrity Due to Water Hammer in Drywell Cooler j
Piping,_" is being submitted pursuant to the requirements of 10CFR50.73 (a) (2) (v).
3 Si
- erely, Ph j
r Bezilla Ge eral Manager -
Hope Creek Operations i
Attachment-1 JPP SORC Mtg.97-003 C
Distribution LER File 2
4 4
The pmtrisin yourlunds.
951160 REV 694
1 s
- - UAN 'A 01997 Document Control Desk. LR-N97023 i
Attachment'.A 4
- - l The :following items represent comniitments that Public Service Electric'&-Gas: (PSE&G) made to the Nuclear Regulatory Commission
-(NRC) relative to this LER :(354/96-028-00).
The commitments are as follows:
.On112/20/96, administrative controls were-implemented to prevent the-defeating of the drywc11 cooler isolation
'interlocksLin post LOCA conditions.
To address the' water _
hammer issue, either:
- 1) revisions to the Hope Creek EOPs l
will--be made to address overriding of the containment isolation interlocks of the_ chilled water system; or 2) the affected containment penetrations will be reanalyzed or' mcdified'to accommodate the water hammer loads.
Final i
implementation of corrective actions to resolve this issue 3
will lme completed by the end of the next refueling outage (RFO7).
- - Engineering evaluations of the impact.on Hope ~ Creek of the water hammer, two-phase flow and over-pressurization issues discussed in Generic Letter 96-06 have been completed and the results will be transmitted to the NRC (in the response i
to' Generic Letter 96-06) by 1/28/97.
]
i 4
1 1
T.
- ~ -
NRC f ORJ 366 U.S. NUCLEAR REGULATORY CO242.lSSION APPROVED I4Y (Ju.J NO. 3150-0104 EXPlRES 04/30/98 (4-95) 4 hlA ORY I RMATI N COLL ON RE 50 0 HR$
LICENSEE EVENT REPORT (LER)
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^"'
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R T4 35)
U.
NUC4 (See reverse for required number of gUggRv Owuggi g A$gN Og, g
555400 digits / characters for each block)
MANAGEMENT AND BUDGET, WASHINGTON, DC 20503, j
F ACILiiY sdAMs (1)
DOCKET NUMBEH (2)
P ACE (S)
Hope Creek Generating Station 05000354 1 OF 4 Ti f LE (4)
Pct::ntial Loss of Containment Integrity Due to Water Hammer in Drywell Cooler Piping EVENT DATE (5)
LER NUMBER (6)
REPORT DATE (7)
OTHER FACILIT8ES INVOLVED (8) t;ONTH DAY YEAR YEAR SEQU NT AL REV SI N MONTH DAY YEAR
' ^ " " " * " '
028 00 1
20 97 12 20 96 96 05000 taPERATIN G j
THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR 5: (Check one or more) (11)
MODE (9) 20.2201(b) 20.2203(a)(2)(v) 50.73(a)(2)(i)(B) 50.73(a)(2)(vni)
LEVEL (10) 20.2203(a)(1) 20.2203(a)(3)(i) 50.73(a)(2)(ii) 50.73(a)(2)(x)
POWER 100 20.2203(a)(2)(i) 20.2203(a)(3)(n) 50.73(a)(2)(m) 73.71 20.2203(a)(2)(ii) 20.2203(a)(4) 50.73(a)(2)(iv)
OTHER 20.2203(a)(2)(m) 50.36(c)(1) x 50.73(a)(2)(v) sp egigbegtgelow 20.2203(a)(2)(iv) 50.36(C)(2) 50.73(a)(2)(vn)
LICENSEE CONTACT FOR THIS LER (12)
NAME TELEPHONE NUMBER (include Area Code)
Jam:s Priest, Lead Lngineer-Licensing and Regulation (609) 339-5434 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)
CAUSE
SYSTEM COMPONENT M ANUP ACT URE R REPORTA0 E
CAUSE
SYSTEM COMPONENT MANUFACTURER REPORTAB E SUPPLEMENTAL REPORT EXPECTED (14)
EXPECTED MONTH DAY YEAR YES NO SUBMISSION (if yes, Complete EXPECTED SUBMISSION DATE).
X DATE (15)
AUSTRACT (Limit to 1400 spaces.a.e., approximately 15 single-spaced typewntten knes) (16)
On 12/20/96, Engineering personnel conipleted evaluations of water hammer concerns discussed in Generic Letter 96-06, " Assurance of Equipment operability and Containment."
Based upon these evaluations, Engineering concluded that a potential exists for primary containment integrity to be adversely impacted when drywell cooler isolation interlocks are defeated in accordance with emergency operating procedures.
Specifically, these actions could result in a water hammer when attempting to place the drywell coolers in service under specific post LOCA drywell conditions.
Since the chilled water system containment penetration was not designed for the effects of a water hammer, containment integrity could be lost.
This issue represented a condition alone which could have prevented the ability to control the release of radioactive materials, and in accordance with the requirements of 10 CFR 50.72 (b) (2) (iii), a four hour report was made at 1606 hours0.0186 days <br />0.446 hours <br />0.00266 weeks <br />6.11083e-4 months <br />.
The cause of this event is httributed to an inadequate evaluation of the potential for creating water hammer loads in post LOCA scenarios as described in Generic Letter 96-06.
Corrective actions include implementing administrative controls to prevent the defeating of the drywell cooler isolation interlocks in post LOCA conditions.
NRC FORM 366 (4-95)
NRC F OR3 366A U.S. NUCLEAR REGULATORY COCISS10N
(&95)
LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION FACILITY NAME (1)
DOCKET NUMBER (2)
LER NUMBER [6)
PAGE (3) 6(QugQAL YEAR u
H::ps Creek Generating Station 05000354 96 - 028 - 00 2
OF 4 TEXT (if more space is required, use additional copies of NRC Form 366A) (17)
PLANT AND SYSTEM IDENTIFICATION
General Electric - Boiling Water Reactor (BWR/4)
Drywell Cooler System - EIIS Identifier {VB}
IDENTIFICATION OF OCCURRENCE Event date:
12/20/96 Discovery date:
12/20/96 Date determined to be reportable:
12/20/96 Problem Report 961220079 CONDITIONS PRIOR TO OCCURRENCE Plant was in OPERATIONAL CONDITION 1 (POWER OPERATION).
Reactor was at 100% of rated thermal power.
There were no systems, structures or components that were inoperable at the time of the event that contributed to the event.
DESCRIPTION OF OCCURRENCE On December 20, 1996, the Hope Creek Generating Station Engineering personnel completed evaluations of water hammer concerns discussed in Generic Letter 96-06, " Assurance of Equipment Operability and Containment."
Based upon these evaluations, Engineering concluded that a potential exists for primary containment integrity to be adversely impacted when drywell cooler isolation interlocks are defeated in accordance with Emergency Operating Procedure HC.OP-EO.ZZ-0102A(B).
Specifically, this Emergency Operating Procedure (EOP) directs the operator to place all drywell coolers in service (by defeating the containment isolation interlocks) when high drywell t.emperatures exist in post LOCA scenarios.
However, this action i
could result in a water hammer when attempting to place the drywell coolers in service under specific post LOCA drywell conditions.
Since the drywell cooler system containment penetration was not designed for the effects of a water hammer, implementation of the EOP could have resulted in the loss of cor,tainment integrity.
Since this issue represented a condition alone which could have prevented the ability to control the release of radioactive materials, a four hour report was made at 1606 hours0.0186 days <br />0.446 hours <br />0.00266 weeks <br />6.11083e-4 months <br /> in accordance with the requirements of 10 CFR 50.72 (b) (2) (iii).
On 12/20/96, administrative controls were implemented to prevent the defeating of the drvwell cooler isolation interlocks during post LOCA conditions.
Since the design and licensing bases of Hope Creek do not4
Nmc FORJ 366A U.S. NUCLEAR REGULATORY COZlSSION I4-95)
LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION FACILITY NAME (1)
DOCKET NUMBER (2)
LER NUMBER (6)
PAGE (3)
YEAR 8ECU AL NU Hope Creek Generating Station 05000354 96 - 028 - 00 3
OF 4 TEXT (if rnore space is required, use additional copies of NRC Forrn 366A) (17)
DESCRIPTION OF OCCURRENCE (Continued) require the use of the drywell coolers to mitigate the consequences of an accident, no Technical Specification actions or other compensatory measures were required to be taken in order to implement the administrative controls.
ANALYSIS OF OCCURRENCE On 9/30/96, the NRC issued Generic Letter 96-06 to provide notification about safety-significant issues that could affect containment integrity and equipment operability during accident conditions.
These issues concerned:
- 1) susceptibility of cooling water systems serving containment air coolers to the effects of water hammer and/or tw phase flow; and 2) over-pressurization of isolated water-filled piping sections in containment during LOCA or Main Steam Line Break (MSLB) conditions.
The system which supports the drywell coolers at Hope Creek is a closed-loop system supplying cooling water to various non-safety related air handling units. In the event of a LOCA, power to the drywell coolers will be tripped automatically and the containment isolation valves will close.
As required by the Generic Letter, Engineering personnel have evaluated the above issues for impact on this system.
From this evaluation, Engineering concluded that the water hammer concern had i
the potential to adversely impact containment integrity under specific post j
LOCA/MSLB conditions when the drywell coolers are placed into service in accordance with Hope Creek's EOPs.
Engineering personnel determined that a water hammer could occur:
- 1) anytime the water in the drywell coolers is greater than 250 degrees F and the isolation interlocks are defeated; or 2) if water in the drywell coolers, which initially is less than 250 degrees F, is heated then cooled prior to defeating the containment isolation interlocks.
Due to the presence of relief valves on the piping in the drywell, water inventory would be lost as the temperature rises, and when the water cools, pressure may decrease below the saturation pressure.
These conditions would enable the formation of voids in the piping, which would i
result in a water hammer when the drywell coolers are placed into service.
I Since the system containment penetration was not designed for these water hammer loads, containment integrity could be challenged during specific post j
LOCA scenarios.
APPARENT CAUSE OF OCCURRENCE j
Hope Creek EOP HC.OP-EO.ZZ-0102 directs the operator to defeat the chilled water system isolation interlocks to place the drywell coolers into service during post LOCA conditions.
This action was formally incorporated into the EOPs through revisions made in the late 1980s.
This EOP revision represented an opportunity to identify this problem, but the capability I
'\\
NRC FOR3 366A U.S. NUCLEAR REGULATORY COMntSSION (4-95)
LICENSEE EVENT REPORT (LER)
TEXT CONTINUATION FACILITY NAME (1)
DOCKET NUMBER (2)
LER NUMBER C6)
PAGE (3)
SEQ YEAR U
N Heps Creek Generating Station 05000354 96 - 028 - 00 4 OF 4 TEXT (if more space is required. use additional copies of NRC Form 366A) (17) to utilize this isolation interlock design feature had existed since plant startup.
Justification for this EOP revision utilized information from the approved BWR Owners Group Emergency Procedures Guidelines; however, the potential for creating water hammer loads in post LOCA scenarios was not adequately evaluated when the plant was designed.
The failure to properly perform an evaluation of the effects of water hammer loads is attributed as the cause of this event.
ASSESSMENT OF SAFETY CONSEQUENCES
i In accordance with 10 CFR 50, Appendix A, Criterion 16, the containment design requires the containment and associated systems to establish an essentially leak-tight barrier against the uncontrolled release of radioactivity to the environment and to assure that the containment design conditions important to safety are not exceeded for as long as postulated accident conditions require.
Since the postulated water hammer could have resulted in the loss of this required containment integrity, there were adverse potential safety consequences associated with this event.
The postulated water hammer does not impact other systems designed to mitigate 4
the consequences of an accident.
PREVIOUS OCCURRENCES
There have been no similar occurrences previously reported by Hope Creek.
CORRECTIVE ACTIONS
On 12/20/96, administrative controls were implemented to prevent the defeating of the drywell cooler isolation interlocks in post LOCA conditions.
To address the water hammer issue, either:
- 1) revisions to the Hope Creek EOPs will be made to address overriding of the containment isolation interlocks of the chilled water system; or 2) the affected containment penetrations will be reanalyzed or modified to accommodate the water hammer loads.
Final implementation of corrective actions to resolve this issue will be completed by the end of the next refueling outage (RF07).
Engineering evaluations of the impact on Hope Creek of the water hammer, two-phase flow and over-pressurization issues discussed in Generic Letter 96-06 have been completed and the results will be transmitted to the NRC (in the response to Generic Letter 96-06) by 1/28/97.
1- -