ML20199J487
| ML20199J487 | |
| Person / Time | |
|---|---|
| Site: | Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png |
| Issue date: | 01/29/1998 |
| From: | Fredrichs T NRC (Affiliation Not Assigned) |
| To: | Mellor R CONNECTICUT YANKEE ATOMIC POWER CO. |
| References | |
| NUDOCS 9802050327 | |
| Download: ML20199J487 (14) | |
Text
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Jtnuary 29, 1998 Mr. Russell Mellor Vice President, Decommissioning Connecticut Yankee Atomic Power Company 362 injun Hollow Road East Hampton, Connecticut 06424 3099
SUBJECT:
REVIEW OF PRELIMINARY ACCIDENT SEQUENCE PRECURSOR ANALYSIS OF CONDITION AT HADDAM NECK PLANT
Dear Mr. Mellor:
Enclosed for your Information is a copy of the final Accident Sequence Precursor (ASP) analysis of the operational condition at the Heddam Neck Plant reported in Licensee Event Report (LER) No. 213/96 016. This final analysis (Enclosure 1) was prepared by our contractor at the Oak Ridge National Laboratory (ORNL), based on review and evaluation of your comments on the preliminary analysis and comments received from the NRC staff and from our independent contractor, Sandla National Laboratories (SNL). Enclosure 2 contains our responses to your specific comments. Our review of your comments employed the criterla contained in the material which accompanied the preliminary analysis. The results of the final analysis indicate that this event is a precursor for 1996.
Please contact me at (301) 4151112 If you have any questions regarding the enclosure.
We recognize and appreciate the effort expended by you and your staff in reviewing and providing comments on the preliminary analysis.
Sincerely.
ORIGINAL SIGNED BY:
Thomas L. Fredrichs, Project Manager Non Power Reactors and Decommissioning Project Directorate Division of Reactor Program Management Office of Nuclear Reactor Regulation Docket No. 50 213
Enclosures:
As stated N Bn C p g g s s C O M cc w/ent.losures:
See next page QlSTRIBUTION:
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WASHINGTON. D.C. 3066H001 January 29, 1998 Mr. Russell Mellor Vice President, Decommissioning Connecticut Yankee Atomic Power Company 362 injun Hollow Road East Hampton, Connecticut 06424 3099
SUBJECT:
REVIEV/ OF PRELIMINARY ACCIDENT SEQUENCE PRECURSOR ANALYSIS OF CONDITION AT HADDAM NECK PLANT,
Dear Mr. Mellor:
Enclosed for your information is a copy of the final Accident Sequence Precursor (ASP) analysis of the operational condition at the Haddam Neck Plant reported in Licensee Event Report (LER) No. 213/96-016. This final analysis (Fnclosure 1) was prepared by our contractor at the Oak Ridge National Laboratory (ORNL), based on review and evalur, tion of your comments on the preliminary analysis and comments received from the NRC staff and from our independent contractor, Sandia National Laboratories (SNL). Enclosure 2 contains our responses to your spr ic comments. Our review of your comments employed the criteria contained in the rv, stial which accompanied the preliminary analysis. -The results of the final analysis indicate that this event is a precursor for 1996.
Please contact me at (301) 4151112 If you have any questions regarding the enclosure.
We recognize and appreciate the effort expended by you and your staff in reviewing and providing comments on the preliminary analysis.
Sincerely, lW Y
Thomas L. Fredrichs, Project Manager Non Power Reactors and Decommissioning Project Directorate Division of Reactor Program Management Office of Nuclear Reactor Regulation Docket No. 50 213
Enclosures:
As stated I
cc w/ enclosures:
See next page
e Connecticut Yankee Atomic Power Co.
Haddam Neck Plant Docket No. 50 213 cc:
Ullian M. Cuoco, Esq.
Resident inspector Senior Nuclear Counsel Haddam Neck Plant Northeast Utilities Service Company clo U.S. Nuclear Regulatory Commission P. O. Box 270 362 Injun Hollow Road Hertford, CT 06141 0270 East Hampton, CT 06424 3099 Mr. Kevin T. A. McCarthy, Director Mr. James S. Robinson Monitoring and Radiation Division Manager, Nuclear Investments and Department of Environmental Administration Protection New England Power Company 79 Eltn Street 25 Research Drive Hartford, CT 06106 5127 Westborough, MA 01582 Mr. Allan Johanson Mr. G. P. van Noordennen Assistant Director Manager Nuclear Ucensing Office of Policy and Management Northeast Utilities Service Company
' Policy Development and Planning 362 Injun Hollow Road Division East Hampton, CT 06424 3099 450 Capitol Avenue MS#52ENR P. O. Box 341441 Regional Administrator Hartford, CT 06134 1441 Region i U.S. Nuclear Regulatory Commission Mr. F. C. Rothen 475 Allendale Road Vice President Work Services King of Prussia, PA 19406 Northeast Utilities Service Company P. O. Box 128 Board of Selectmen Waterford, CT 06385 Town Office Building Haddam, CT 06438 Mr. D. M. Goebel Vice President Nuclear Oversight Northeast Utilities Service Company P. O. Box 128 Waterford, CT 06385 Mr. J. K. Thayer Recovery Officer, Nuclear Engineering and Support Northeast Utilities Service Company-P. O. Box 128 V.aterford, CT 06385
s1 ITCIDSURE 1 LER No. 213/96-016 LER No. 213/96-016 Esent
Description:
Potentially inadequate RilR pump NPSM following a large or medium-break LOCA Date of Event: August I,1996 Plant: Haddam Neck Event Summary The Haddam Neck licensee determined that the net positive suction head (NPSil) available to a residual heat removal (RilR) pump during the sump recirculation phase of a loss of coolant accident (LOCA) would be inadequate for many emergency core cooling system (ECCS) configurations [Ref. Ij. Inadequate NPSH will cause pump cavitation that, if severe or prolonged, will fail a pump. One RHR pump is used during the sump recirculation phase at Haddam Neck; use of RHR pump B would have almost cenainly resulted in its failure due to low NPSH following a large or medium break LOCA. RilR pump A was also vulnerable tc failure for some potential break locations and ECCS component unavailabilities. The estimated increase in the core d
damage probability (CDP) contribution for a 1 year period from inadequate RilR pump NPSH is 1,1 = 10 4
This is above the nominal CDP for the same period of 3.7 x 10 Uncertainties in the frequencies oflarge-and medium break LOCAs (none have occurred), the reduced NPSil following such a LOCA, and the likelihood of pump failure with decreasing NPSH contribute to a substantial uncertainty in this estimate.
Event Description On August 1,1996, with the plant in cold shutdown, the Haddam Neck licensee determined that the RHR pump NPSH would not always be adequate during the sump recirculation phase of a LOCA. Past calculations that had been used to demonstrate adequate RHR pump NPSH were detennined to have utilized an erroneous assumption that there would be sufficiem overpressure in the containment to supply adequate suction pressure to meet the required NPSH.
The NPSH calculation estimates the pressure at the pump suction based on the temperature (and hence saturation pressure) of the sump water, the elevation difference between the free water surface in the containment sump and the pump suction, and the head loss through the flow path. The calculated pmnp suction pressure is compared to the minimum allowable pressure specified by the pump manufacturer to determine if adequate NPSil will exist following a LOCA. Inconsistent with Safety Guide 1, the original Haddam Neck calculation also considered containment overpressure and assumed that sufficient overpressure would exist in the containment to meet the required pump NPSH.
Analyses performed prior to submittal of the LER indicated that the assumed containment overpressure would not exist at all times during the sump recirculation phase following a large break LOCA. These analyses indicated that if motor operated sump valve RH MOV 22 failed to open and the alternate sump recirculation I
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LER No. 213/96-016 flow path through RH V 808A was used, the containment pressure required for adequate RHR pump NPSil would probably not exist during recirculation.
Later analyses performed by the licensee. and completed aRer the LER was submitted, indicated that the NPSil problem existed independent of the sump recirculation flow path that was used and was therefore more serious than initially thought. During a telephone call on hiny 20,1997, with ASP Program staff, licensee personnel prosided additional infonnation on the results of the later analyses. These analyses indicated that if RHR pump B was used during the recirculation phase, adequate NPSH would not exist for LOCA break siics greater than 0.05 A* (large and medium break LOCAs), independent of the sump recirculation flow path that was used. This condition would occur beginning ~l.5 h aRer the LOCA, if RHR pump A was used, adequate NPSH would not exist if RH hiOV 22 failed to open. If RH hiOV 22 successfully opened, NPSH would be adequate for ccid leg breaks, but would be ~80% of the required amount for hot leg breaks.
Additional Event-Related Information At lladdam Neck, post LOCA ECCS operations are divided into three phases-injection, short term recirculation, and long term recirculation. During the injection phase (which is not a concern in this analysis),
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borated water is drawn from the refueling water storage tank (RWST) and injected into the reactor coolant l
system (RCS). The high pressure safety injection (HPSI) pumps inject into the four cold legs and the low pressure safety injection (LPSI) pumps inject into the reactor vessel head The charging pumps are also used if offsite power is available.
When sufficient water has been ejected from the break, filling the containment sump (130,000 gal), short term recirculation is initiated. In this mode, the LPSI pumps are stopped and one RHR pump is started. This pump takes suction from the sump and supplies water to the suction of one HPSI pump, which delivers water to two of the four cold legs Aner a predetermined time has elapsed, two-path recirculation is initiated to prevent boric acid precipitation. In this alignment, one RHR pump again takes suction from the sump and supplies unter to one charging pwnp, which delisers water at high pressure to the loop 2 cold leg. The RHR pump is also aligned to supply low pressure water directly to the upper reactor vessel head (core deluge) during two-path recirculation. The procedures for initiating sump recirculation (Refs. 2 and 3) are complex and involve the manipulation of many valves.
The Haddam Neck design is unusual in that it utilizes a single motor-operated valve (RH hiOV-22) to isolate the containment sump from the RHR pumps. This vahe is opened when short-term recirculation is initiated.
If RH hiOV-22 fails to open, procedure ES l.3 (Ref. 2) instructs the operators to locally unlock and open a parallel manual valve, RH V 808A, to provide recirculation flow, Haddam Neck procedures require sump switchover to be initiated quickly-in 10 min or less following a large-break LOCA (Ref. 4).
Additional information concerning this event and related esents is included in NRC Infonnation Notice 96 55," Inadequate Net Positive Suction Head of Emergency Core Cooling and Containment Heat Removal Pumps Under Design Basis Accident Conditions"(Ref. 5) and NRC Generic Letter 97-04," Assurance of Sufficient Net Positive Suction Head for Emergency Core Cooling and Containment Heat Removal Pumps" (Ref. 6).
2
Oi LER Na. 213/96 016 Modeling Assumptions This analysis assumes that, if RilR pump B is selected for sump recirculation or motor-operated valve Ril MOV 22 fails to open, adequate NPSil would not exist to support RilR pump operation following a large or medium break LOCA. As described above, break sites above 0.05 R are of cor.cern. This includes 2
all large break and some medium break LOCAs (medium break LOCAs include break areas between 0.02 and 0.2 R'). 3ecause data do not exist that would allow the medium break LOCA initiating event category to be further divided to allow only break sizes greater than 0.05 R: to be addressed, all medium break LOCAs were assumed to result in low RilR pump NPSH (the inability to subdivide the medium-break LOCA category may add some conservatism to the analysis).
If RilR pump A is selected for sump recirculation and Ril MOV 22 successfully opens, the analysis assumes adequate NPS11 would exist to support operation of that pump provided the LOCA occurred in the cold leg.
If the LOCA occurs in thc hot leg, operator throttling of discharge flow is assumed to be required to support long term operation of the pump.' Recovery of recirculation flow through use of pump A is also asstmed to be possible, if rump B is initially used.
The procedures for sump recirculation (Refs. 2 and 3) instruct the operators to check pump current following RiiR pump start to confirm that cavitation is not occurring and to throttle discharge flow if necessary.
I ilowever, initial RHR pump start would occur when containment pressure is still high and when adequate i
NPSH is available. Although Refs. 2 and 3 also instruct the operators to monitor RilR pump current during recirculation, the licensee noted two factors in the May 20, 1997, telephone call that would complicate throttling and increase the likelihood that low RHR pump A NPSH will not be detected and corrected before pump damage occurs:
(1) Two of the foui ilPSI discharge valves are closed during sump recirculation. If the LOCA occuned in an RCS loop associated with one of'he two open valves (resulting in the loss of some 11 PSI flow out the break), there is essentially no range through which RHR flow can be adjusted to reduce cavitation and at the same time provide adequate decay heat removal from the core. The licensee believed that throttling would be scry diGicult in this situation. lThis may be more of a concern following a cold leg break, since HPSI flow is injected into the cold leg. Ilowever, because the licensee estimated a suction pressure for RHR pump A following a hot leg break at ~80% of the vendor specified NPSH, this analysis assumes the throttling concern applies to a hot leg break as well. This may be pessimistic.]
(2) Normally, cavitation is indicated by oscillating pump current. For this event, the licensee concluded that the operators would probably not observe oscillating current, but instead see a
- No information is available conceming the expected lladdam Neck RIIR pumn perfonnance at reduced NPSil.
Ilowever, Ref, 5 provided this information for another low-pressure, high-capacity pump--the containment spray pump at Maine Yankee. The manufacturer of that pump indicated that the pump could operate indefinitely at 95% of required NPSil and for 15 min at 75% of requ. red NPSil The pump manufacturer also stated that sinular pumps are routinely operated for 1 to 3 min at 50% of required NPSil without sustaining damage. The lladdam Neck licensee estimated a suction pressure for RilR pump A following a hot-leg break at ~80% of the vendot specified NPSil.
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e LER No. 213/96-016 gradual reduction in pump cunent. This might lead them to open the discharge valves, exacerbating the cavitation problem.
The Accident Sequence Precursor (ASP) Program typically considers the potential for core damage following four postulated initiating events in pressurized vcater reactors: transient, loss of offsite power, small break LOCA, and steam generator tube rupture. Supercomponent based linked fault tree models are available for each of these postulated initiating events. The two initiating events that are of concern in this analysis (i c.,
large and medium-break LOCAs) are not currently modeled. Ilowever, for both of these initiating events, the unavailability of sump recirculation, which would occur following the loss of the RllR pumps as a result of cavitation, is assumed to result in core damage in all probabilistic risk assessments. Therefore, the signiGcance of the etcnt can be estimated directly from the probability of recirculation failure and the probability of a large or medium break LOCA in the unavailability period. The longest unavailability period used in an ASP analysis is I year.
An event tree depicting the potential sequences to core damage allowing a large or medium break LOCA is shown in Fig.1. The cvent tree kt not chronological; it has been structured to reduce the number of sequences that must be addresed The event tree includes the following branches.
1 Afedwm or Large Break LOCA Initiating Event @f L LOCA). The initiating event is a large or medium.
d break LOCA. The frequency oflarge and medium break LOCAs is estimated to be 2.7 = 10 / year and 5.0 = 10"/ year, respectively. These values are based on a survey of large and medium break frequencies performed in support of the analysis of Turkey Point LER No. 250/94-005 in the 1994 precursor report (see Appendix H to Ref / for additional information). The probability of a large-or medium-break LOCA in a 1 year period is therefore 7.7 = 10' Reactor Tr/p (R7). Failure of the reactor ta trip is assumed to result in core damage for a medium break LOCA (reactor trip success following a large-break LOCA is not required since void formation resulting from the break terminates the nssion process).
Consistent with the Integrated Reliability and Risk Analysis System (IRRAS)-based ASP models, a nonrecoverable failure to trip probability of 2.0 = 104 is utilized This is weighted by the likelihood that the LOCA is a medium-break LOCA (0.65), resulting in an overall 4
branch probability of 1.3 = 10.
IJ'SlHPS/ Injection (INJEC7). Failure ofinjection using the LPSI system results in a loss of short term RCS makeup and core damage following a large break LOCA. Consistent with the IRRAS based ASP models, a failure probability of 9.4 x 10"is estimated for the two-train LPSI system. The Haddam Neck Individual Plant Examination (Ref. 8), states that HPSI provides success following a medium-break LOCA. The HPSI and LPSI systems are both two-train systems, with similar failure probabilities (based on the system design, the llPSI system may be slightly more reliabic). To simplify the analysis, the LPSI failure probability was used for the injection branch following both a large and medium break LOCA. This has no impact on the analysis results.
Operator Opens RH-A/Ol' 22for Rectre (IU/-Afol'-22). Failure of the motor operated sump isolation valve to open for recirculation is assumed to result in core damage. Inadequate RHR pmap NPSH exists in this
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t 1.ER No. 213/96 016 case, independent of the RilR pump that is selected for recirculation. Consistent with the IRRAS based ASP 4
models, the probability of motor operated salve R11 MOV 22 failing to open is estimated to be 3 = 10 tother potential Rlin system railures, such as the common cause failure of the RilR pumps, can also result in failure of recirculation. These failures are untclated to this event and contribute to a lesser extent to the overall failure of recirculation. They have not been arldressed in this analysis.]
RRR rump A Sclected(SELECTA). Success foi this branch implies that the operators initially select pump A for recirculation. The licensee stated that the operators were aware of flow restrictions that existed in the B RilR train and that the operators had stated in informal inter icws that they would prefer to use the A pump whco initiating recircuir... To reflect this, the analysis assumed that the A pump would be initially used with a probability of 0.75 (0.25 probability that RilR pump A would not be used).
RCS Cold-Lcg Break (COLD LEG), Success for this branch implies that the LOCA occurred in one of the RCS cold legs. To recognize the greater likelihood of a break in a cold leg because of the larger number of cold leg pipe segments and welds,' this analysis assumes a probability of 0.6 that a LOCA will occur in a cold leg.
RHR rump A 7hrott/cd(7#RO771E A). Faih-c to throttle RilR pmnp A discharge flow following a hot leg break is assumed to result in pwnp failure due to cavitation. The probability of failing to throttle RilR pump flow was assumed to be 0.5 if the llPSI discharge valve ast.ociated with the faulted loop remained open (see factors I and 2 abose). If the valve associated with the fr.:'ted loop was one of the two that was closed, the probability of failing to throttle RilR pump flow was assumed to be 0.12 (ASP recovery class R3, as described in Appendix A of Ref. 7, see factor 2 above). Combining these values with the probability that the 11 PSI valve associated with the faulted loop would be one of the two that was closed (0.5) results in an estimated branch failure probability of 0.31.
Rectrculation Recovcred(RECIRC). Sump recirculation can be recovered using RilR pump A following the failure of pump B,if pump B is initially selected for the recirculation function. Following the failure ef the B puinp, the operators would have to determine that cavitation was a potential cause of failure, and, if the break was in a hot leg, throttle the A pump discharge Dow at the time the pump was started." Two event tree branches address this possibility. The first branch is associated with the recovery of recirculation following a cold-leg break. A failure probability of 0.12 is assumed (ASP recovery class R3, as described in Appendix A of Ref. 7). The use of this probability recognizes the substantial burden and lack of specific procedural guidance that would exist following the failure of RilR pump B. The second branch is associated with the recovery of recirculation following a hot leg break. The failure probability for this branch,0.31, also recognizes the anticipated difliculties associated with throttling RilR pump discharge flow if the llPSI discharge valve associated with the faulted RCS loop remains open.
- Ref. 9 provides a discussion of the factors that innuence the hielihood of pipe break.
" Procedure ECA.1.1, " Loss of Emergency Coolant Recirculation," (Ref.10), instructs the operators to (1) reinitiate injection flow from the RWST in the event of a failure of recirculation and (2) not restart an IWR pump prior to consulting with the Plant Engineering Staf1' 5
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0 I
LER No. 213/96-016 Analysis Results Sequences shown in Fig. I that can result in core damage are described in Table 1. Applying the branch probabilitics (provided in the Modeling Analysis section) to the model for the event results in en estimated CDf' for large and medium break LOCAs over a 1 year period of 1.1 = 10d. The nominal CDP ovos a 1 year period estimated using the ASP IRRAS models for Haddam Neck is 3.7 = lot The RilR pump NPSil problem increases this probability to 1.5 = 104. This value is the conditional core damage probability (CCDP) for a 1 ycr.r period in whien the potentially inadequate NPSil existed.
The dominant core damage sequence for the event involves a postu!ated large or medium hot leg break (medium-break LOCAs are more likely);
e successfut reactor shutdown, injection, and opening of Ril MOV.22; RilR pump A initially selected by the operators for sump recirculation; and failure to throttle RilR pump A discharge now to provide adequate NF SH.
This sequence is highlighted on the event tree in Fig.1.
A greater than usual uncertainty is associated with this estimate. THs tmcertainty is dominated by the uncertainties in the frequency of a large or medium break LOCA (none have occurred), the available NPSil following such a LOCA, and the likelihood of pump failure with decreasing NPSil.
Acronyms ASP accident sequence precursor CCDP conditional core damage probability CDP core damage probability llPSI high pressure safety injection IRRAS Integrated Reliability and Risk Analysis System LER licensee event report LOCA loss-of coolant accident LPSI low pressure safety injection MOV motor operated vahe NRC Nuclear Regulatory Commission RCS reactor coolant system RWST refueling water storage tank j
References
August 29,1996.
6
LER No. 213/96-016
- 3. Iladdam Neck Procedure ES l.4," Transfer to Two Path Recirculation," Rev.13.
4.
Transcript of December 4,1996, Region I enforcement conference for Haddam Neck.
S. NRC information Notice 96 S$, Inadequate Net Positive Suction HeadofEmergency Core Cooling and Containment Heat Removal Pumps under Ikstgn Basis Accident Condottons,0ctober 22,1996.
6.
NRC Generic Letter 97 04, Assurance ofSufflctent Net Positive Suction Headfor Emergency Core Cooling and Containment Heat RemovalPumps,0ctober 7, l997.
7.
R. J. IIelles, J. W. Cletcher, D. A. Copinger, D. W. Dolan, J. W. Minarick, L N. Vanden lieuvel, Martin Marietta Energy Systems, Inc., Oak Ridge Natl. Lab.; and Science Applications International Corp.,
Precursors to Potential Severe Core IMmage Accidents: 1991, A Status Report, U.S. NRC Report NUREG/CR 4674 (ORN1/NOAC 232),Vols. 21 and 22, December 1995.
K.
Haddam Neck Plant Individual Plant Examinatwnfor Severe Accident l'ulnerabiltttes, June i993.
9.
- 11. M. Thomas," Pipe and Vessel Failure Probability," Re/tability Engancering, Vol. 2,1981, pages83-124.
- 10. Iladdam Neck Procedure ECA l.1," Loss of Emergency Coolant Recirculation," Rev. I1.
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4 12R No. 213/96 016 ll 5
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8 8
8 8
8 8
8 g
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Fig.1. Dominant core damage sequence for LER No 213/96 016.
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- LER No. 213/%016 1
j Table 1. Sequence descriptions for analysis event tree i
Sequence Sequence Description probability 3
Large or meditun hot. leg break, successful reactor shutdown and 7.1 = 10'$
I injection. Ril MOW 22 opens for sump recirculation, RilR pump A 1.clected for recirculation, and failure to throttle RiiR pump discharge t
flow for NPSil.
i 5
Large or medium cold-leg orcak, successful reactor shutdown and 1.4 = 10 $
injection, Ril MO%22 opens for sump recirculation, RilR pump B initially selected for recirculation, and failure to recover recirculation using pump A.
7 Similar to sequence $ except a hot leg break occurs.
2.4=10-5 i
8 Large or medium break LOCA, successful reactor shutdown and 2.3 = 10*
injection, and failure of Ril MOW 22 to open for sump recirculation. -
9-Large. or medium break LOCA, successful reactor shutdown and -
7.2 x 104 failure ofinjection.
10 Large. or medium break LOCA and failure to shut down the reactor 1,0=10*
(failure to trip for a medium break LOCA).
4 3
1 h
9 4
.. - - - _. -,,., -,. _. -. _,, ~-, - - _ _,., _,, _ -. -
ENCIDSURE 2 LER No. 213/96-016 LER No. 213/96-016 Event
Description:
Potentially inadequate RilR pump NPSil following a large or medium break LOCA Date of Event: August 1,1996 Plant: lladdam Neck l
Licensee Comments
Reference:
Teleconference with 0, llolfa, CYAPCO, and M. Fairtile and P. D. O'Reilly, U.S. NRC, October 9,1997.
Comment 1 in the preliminary analysis, the third sentence of the third paragraph under Afodeling Assumptions incorrectly states that RilR pump current is not monitored after sump recirculation has been initiated. Procedures ES 1.3, step 14, and ES 1,4, step 6, both require monitoring of pemp current for the running RilR pump after recirculation transfer has been completed.
Response i The analysis has been revised to state that both procedures require pump current to be monitored. The third sentence of the third paragraph under Afodeling Assumptions was deleted: The paragraph now reads as follows:
The procedures for sump recirculation (Refs. 2 and 3) instruct the operators to check pump current following RilR pump start to confirm that cavitation is not occurring and to throttle discharge flow if necessary. Iloweser. initial RllR pump start would occur when containment pressure is still high and w hen adequate NPSil is available Ahhough Refs. 2 and 3 also instruet the operators to monitor RilR pump current dunng recirculation, the beenwe noted two factors in the May 20 1997, telephone call that would complicate throttling and increase the likehhood that low RitR pump A NPSil will not te detected and conected before pump damage occurs:
(1)
Two of the four llPSI discharge valves are closed dunng sump recirculation if the t.OCA occurred m an RCS loop associated with me of the two open valves (resulting in the loss of some llPSI Ilow out the break), there is essentially no range through which RilR flow can t<
1 adjusted to reduce cavitation and at the same time provide adequate decay heat removal from the core. The beensee beheved that throttling would te very dillicult in this situnuon. (This may be more of a concern following a cold-leg break, since llPSI flow is injected into the cold i
'eg. Iloweser. tecause the bcensee estimated a suction pressure for RilR pump A following a hot leg break at -80's o. the sendor specified NPSit, this analysis assumes the throtthng concern opphen to a hot-leg break as well. His may te pessimistic.]
(2)
Normally, c6*itation is indicated by oscillating pump current. For this event, the licensee -
concluded that the operators would probably not observe oscillating current, but instead see
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5 0
l LElt No. 213/96 016 l
t a produal reductke in pump current. This might lead them to open the diuharge sattes..
exacerbating the cavitahim problem, i
Note that because of the expected difficuhics in throttling RHR pump discharge flow, as i
stated by the licensee in a May 20,1997, telephone call and summuired later in the sanx paragraph, this change did not effect the operator error probabilities assumed for this action.
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