ML20206F164
| ML20206F164 | |
| Person / Time | |
|---|---|
| Site: | Monticello, Dresden, Perry, Fermi, Duane Arnold, Clinton, Quad Cities, Big Rock Point, LaSalle, 05000000 |
| Issue date: | 06/21/1986 |
| From: | Norelius C NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | Guthrie S, Gwynn P, Mcgregor L NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| References | |
| IEB-86-001, IEB-86-1, NUDOCS 8606240292 | |
| Download: ML20206F164 (1) | |
Text
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4-I MEMORANDUM FOR:
S. Guthrie,' SRI Big Rock Point P. Gwynn, SRI Clinton L. McGregor, SRI Dresden J. Wiebe, SRI-Duane Arnold W. Rogers, SRI Fermi M. Jordan, SRI LaSalle P. Hartmann, SRI Monticello J. Grofbe, SRI Perry
- A. Madison, SRI Quad-Cities FROM:
C. E. Norelius, Director, Division of Reactor Projects i
SUBJECT:
IE BULLETIN 86-01, MINIMUM FLOW LOGIC PROBLEMS THAT COULD e
DISABLE RHR PUMPS, INSPECTION RESPONSIBILITIES
[
'The subject Bulletin, issued May 23, 1986, requires a licensee response in i
seven days identifying whether or not the problem exists at their facility and if a problem exists they are to submit a written report in 30 days informing l
the NRC of the schedule for long-term resolution of the problem.
l The resident inspector staffs are assigned the responsibility of reviewing these
}
._ responses in accordance with Inspection Procedure 92703 including the adequacy of the response, which in the procedure is assigned to regional personnel.
i The review for adequacy shall include all the requirements of Section 92703-02 l
of the procedure. The review will be documented in an inspection report and l
shall include the adequacy of the response as well as the implementation of the bulletin requirements.
f We understand that a Temporary Instruction (TI) will soon be issued that will give guidance for your evaluations. This Bulletin should not be closed until after 4
the TI is issued.
Any questions on this subject should be.2ddressed to E. R. Schweibinz on j -
FTS 388-5542.
9506240292 860621
_F PDR ADOCK 05000010 e
M Charles E. Norelius, Director Division of Reactor Projects cc:
C. Paperiello, DRS J. Hind, DRSS p gs Tb3 d M/O 8b D. Boyd,-DRP R. Knop, DRP G. Wright, DRP Q
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NUCLEAR REGULATORY COMISSION "Y
OFFICE OF INSPECTION AND ENFORCEMENT WASHINGTON, DC 20555 May 23, 1986 IE COMPLIANCE BULLETIN NO. 86-01:
MINIMUM FLOW LOGIC PROBLEMS THAT COULD DISABLE RHR PUMPS Addressees:
All GE boiling water reactor (BWR) facilities holding an operating license (OL) or a construction permit (CP).
purpose:
The purposes of this bulletin are:
(1) to inform BWR licensees and applicants of a recently identified problem with the minimum flow logic for which a single failure could disable all RHR pumps, (2) to request that licensees affected by the problem immediately provide appropriate instructions and training to plant operators on how to recognize the problem if it occurs and take appropriate mitigating actions, (3) to request that licensees notify the NRC of the existence of the problem at their facility within 7 days of receipt of this bulletin, and (4) to request that licensees inform the NRC of measures taken to correct design or installation problems that are identified as a result of this bulletin.
Description of Circumstances:
During a recent review of IE Information Notice 85-94, " Potential For Loss Of Mininum Flow Paths Leading To ECCS Pump Damage During A LOCA," the Pilgrim Nuclear Power Plant discovered that a single failure under certain accident sequences could result in all RHR minimum flow bypass valves being signaled to This condition close while all other pump discharge valves are also closed.
could result in no flow through the RHR pumps and could lead to the pumps running dead headed with potential for pump damage in a few minutes.
If this single failure occurred in conjunction with an automatic start of the RHR system, RHR This event could disable pump damage may occur if unrecognized by the operator.
RHR functsons including Low Pressure Coolant Injection (LPCI), head spray, As drywell spray, shutdown cooling, torus spray and suppression pool cooling.
a result of the loss of suppression pool cooling over a long period of t'me, core spray pumps could ultimately lose net positive suction head and also be unavailable.
.The NRC staff has judged that the overall probability of a serious core damage accident due to this problem is low.
Nevertheless, such a single failure vulner-ability is not in compliance with the regulatory requirement for independence in emergency core cooling systems and could compromise several important systems.
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IES 86-01 NRY 23. 1986 i
Page 2 of 3 i
i, GE's initial review identified the following plants with some potential for having the single failure problem: Pilgrim; Browns Ferry 1, 2 and 3; Peach Sotton 2 and 3; Duane Arnold; Millstone 1; Nonticello; Quad Cities 1 and 2; Dresden 2 and 3; Cooper; and Fermi 2.
The staff understands that GE has notified these potentially affected facilities and informal communications from GE indicate that the list was narrowed to Pilgrim, Quad Cities avi possibly Dresden during preparation of the bulletin.
The above listed plants are being provided the bulletin by telecopy.
Since there is not complete certainty that i
other BWR's do not have the problem, they are also addressed by this bulletin.
Their copies are being sent by mail.
1 At Pilgrim, to prevent the pump from running dead headed, each pair of RHR pumps is provided with a minimum flow bypass capability.
The minimum flow bypass consists of an orificed flow bypass which allows a flow of approximately I
10 percent of rated flow.
The minimum flow bypass lines for each pair of RHR S"
pumps are connected to a single line and controlled by a single minimum flow' bypass valve.
The minimum flow bypass valve is normally open.
The valve will close upon sensing flow in either of the RHR loops.
'i The current logic configuration for Pilgrim minimum flow bypass valves is that a high flow signal from either the A or 8 RHR loops will close both A and 8 bypass valves.
Thus, a postulated single failure of a flow sensing instrument.
may result in all RHR pumps running without bypass flow.
If this failure occurs during an event with a high drywell signal or low water level signal, such as during small or intermediate size loss of cooling accidents (LOCA) or spurious actuation, the RHR pumps may start and run dead headed.
j One of the potential fixes being proposed by GE is to remove the automatic closing signal from the RHR minimum flow bypass valves.
This fix will result in some of the LPCI flow being diverted through the minimum flow line.
For j
other RHR modes of operation, the valves may be manually closed.
Although safety analyses may justify this interim fix, there are a number of problems that need to be considered.
For example, on many plants the minimum flow bypass valves must be closed during shutdown cooling in order to prevent draining the reactor vessel inventory to the torus.
The minimum flow bypass valves are considered containment isolation valves on some plants.
i t
j REQUIRED ACTION FOR GE BWR FACILITIES Promptly determine whether or not your facility has this single failure '
1.
vulnerability.
2.
If the problem exists, immediately instruct all operating shifts of the j
problem and measures to recognize and mitigate the problem.
3.
Within 7 days of receipt of this bulletin, provide (a) a written report to the NRC which identifies whether or not this problem exists at your facility, (b) if the problem exists, identify the short-term modifications to plant operating procedures or hardware that have been or are being implemented to assure safe plant operations.
IE8 86-01' May 23, 1986 Page'3 of 3 4
4.
If the problem exists, provide a written report within 30 days of receipt of thi~s bulletin informing the NRC of the schedule for long-term resolution of problems that are identified as a result of this bulletin.
Should a licensee determine that any action requested by this bulletin jeopardizes overall plant safety, the NRC should be notified of that fact and provided with appropriate justification for not implementing the requested action.
Such notification shall be made within 7 days of receipt of this bulletin.
The written reports shall be submitted to the appropriate Regional Administrator under oath or affirmation under provisions of Section 182a, Atomic Energy Act of 1954, as amended.
Also, the original copy of the cover letters and a copy of the reports shall be transmitted to the U.S. Nuclear Regulatory Commission, i
06cument Control Desk, Washington, D.C. 20555 for reproduction and distribution.
This request for information was approved by the Office of Management and Budget under a blanket clearance number 3150-0011.
Comments on burden and duplication may be directed to the Office of Management and Budget, Reports Management, Ro A 3208, New Executive Office Building, Washington, D. C. 20503.
If you have questions regarding this matter, please contact the Regional i
Administrator of the appropriate NRC Regional Office or one of the technical contacts listed below.
(
- f e_-
I ames Taylor, Director Offic
.,f Inspection and Enforcement Technical Contacts:
Eric Weiss, IE (301) 492-9005 M. Wayne Hodges, NRR (301) 492-7483
Attachment:
List of Recently Issued IE Bulletins l
IEB 86-01 May 23, 1986 LIST OF RECENTLY ISSUED IE BULLETINS Bulletin Date of No.
Subject Issue Issued to 85-03 Motor-Dperated Valva Common 11/15/85 All power reactor Mode Failures During Plant facilities holding Transients Due To Improper an OL or CP for Switch Settings action 85-02 Undervoltage Trip 11/5/85 All power reactor Attachments Of Westinghouse licensees and DB-50 Type Reactor Trip applicants Breakers i
85-01 Steam Binding Of Auxiliary 10/29/85 Nuclear power facil-Feedwater Pumps ities and cps listed in attachment 1 for action; all other nuclear power facil-ities for information 84-03 Refueling Cavity Water Seal 9/24/84 All power reactor facilities holding an OL or CP except Ft. St. Vrain 84-02 Failures Of General Electirc 3/12/84 All power reactor Type HFA Relays In Use In facilities holding Class IE Safety Systems an OL or CP 84-01 Cracks In Boiling Water 2/3/84 All BWR facilities Reactor Mark I Containment with Mark I contain-Vent Headers ment and currently in cold shutdown with an OL for action and all other BWRs with an OL or CP for information OL = Operating License CP = Construction Permit
res,s et eco entset o.as.soes ionis
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i,l 4 '
e W9 CLEAR PRODUCTS & RNCIDEERING BERTICES BEPARa mrt SAk Jost, CALIFORNIA May 21, 1946 LLC-33-86 i
.w Tet From:
C. 3. Strambeek -
$UBJECT!
PRELIMINARY SAFETY __3TALO& TION FOR RMR _FUNP BTP&SS TALTES During a review of the information.in Inspection & Rnforeement taformation Notice 85-944, a BWR plant. discovered that a single f ailure seuld eseelt to all RMR minimum flow bypeos valves being signaled to -cleme while all other pump discharge valves are also closed. This eendition could result in no flow through the Rkk pumps and could lead to the pump running dead headed with potential for pump damage in a few minutes.
If this single failure occurred at a BUR 3 or 4 coincident with an event that resulted la en automatic etart of the 3RR system..RNR pump damage may occur before operator' recognition and interventarm. This event could prevent containment cooling but. adequate core: cooling would be available.
l The following is a preliminary safety evaluation prepared by SE to respond to the condteson se presently reviewed at the discovering plant. Plasse review this information immediately with'your custasier.
The receamended metions are:
1)
Notify the' plant operators of the potential problem..
2)
Review the amioting NHK min 4== flow bypass legio and piping errainsommet for the described problem or og other exiettag problem. -
The NRC beceau aware of the problem at the discovering plant and requested to know what actions CE was taking and what other planto OE was reviewing.
for,pos'etble similar prob eme. GE identified the potential generic nature of the problem to the NkC TTT~Intomation motice.85-94: Peteattal For Loos of Minim m Flow Pathe.
~
Lead to ECCS Pump Damage During et LOCA.
e.e G. L. Sessi
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F. T. Yran R. W. Neward C. J. Romanek I
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1 SEltERAL $!!LECTRIC esumanauwsusnes ceanAtoes omew uanc com o m curma mus e sm xac, cwasmAum PRE 1,tMIMARY SAFEIT EVALUATION FOR _BWR 3/4 DER PtIMP BYFAtt VALTEB_
INTRCOUCTION_
h Residual Heat Removal %).aystem for SWR 3?s and 4's typically la semprised af evo loope (A and 5) with'two pumps on each loop. N ME '
system performa several. safety functione.
3e providee seen maaltag through its low pressure coolant injection.(LPCT) mode and provides centstassat-cooling through its containment spray and suppression pool eco1ing upde.
It also provides shutdown cooling. t
'Ihe LPCI hode operation. in automarie. - All other modes of RER operatiana are manually initiated.: In its LPCI mode, the RPR pump will etert with as initiation nigual of hish drywell pressure signet er a coincident low reactor water level and low reactor pressure signal. Since the RER eystem is a low preneure system. any event which may result in a high drywell pressure condition may cause the four.RRR pumps to start and not inject until the reactor pressure is low enough for injection or operator actions.
To prevent the pump from running dead.headad, each pair of RER pumps is provided with a minimum flow bypesa capability. The minimum flew bypass j
cons 3sts of an orifice flow pass which allows a flow of approximately 10%
of rated flow. The miniseum flow bypass lines for auch pair of the RER
- pmepa are connected into a single lina and controlled by a single minimum flow bypass valve.
N' minimum flow bypass valve is normally open. The valve will close upon sensing established flov in the PRR loops.
The current logic configuration for the discovering RWR RNR minimum flew bypass valve in that a high flow signal from either the A or 5 side of the 1
RRR loops will close both A and 3 bypass valves. Thus, a postuisted single failure of a flow sensing instrument may potentially result in the EER pumps running without the minimum flow bypass valves.
If thia failure l
occurs during an event with h15h drywell pressure signal, the 3RR pumps say start and run dead headed for some significantly long duration.
l 1
Field experience indicates that RNR pumps have operated dead headed for durationsion the order af half an hour and subsequently were capable of deliverint flow. Nevertheless. pump manuf acturere clasm that lackins well instrumented test de$a en these pompe they cannot substantista phe pump's performance when running dead headed for more than a few minutes. Lacking any concrete dats. it is conservative to savume that the aforementioned einste ratture et an RNR flow, sensing instrument during an event With high drywell pressure conditiona may result in the failure of all SMR pumps.
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This partiew)ar failute might affect the plant's capability for lens ters sentoimment eucting. on the other hand. the more aeoling sepability is not reduced for the senar Rasmitina neefdent event which to the restreu3etion euetion line break with the LFCI iMaction value.feiture.
For this event.
core cooltag is provided by the two more spray peups.
POfENTIAL FM g
Coe of the potentia 3 ' fixma for the problem is to remove the closing signal from the RNR minimum flow bypass valves. This fix vill result is some of the LPCL flow being diverted through the mininasm flow line. per other modes of Rim operation, the valves may be manually closed. The effects of leaving the bypass line opea during LPCI operation are being addressed below. For conservatism, the potential of f act of not closing the miniumin finw bypeus valven for other modes of RKK operatian are addressed also.
IMPACT.0F LPCT_ OPERAT10W Reinoving the signal to close the RNR minimum flow im asa valves will result in the valves staying open during the LFCI operation.
This in effect will reduce the atmount of LFC1 flow to the.reacter vessel by the amount of flow threvnh the minisnam bypass lines. The bypass flow at pus shutoff Icw flow conditions to approminately 101 of the rated flow for each EMR pump.
Bownver, this potential reduction of the LPCI flow hoo imeisnificent effect un the plant'e core cooling espebility. Thie 1e demonstrated by she following evaluation for postuinted large, intermediate and small breair events.
The most limiting accident event for SWR' S's and 4's typically is the r.eximum recirculation suction line break with an assumed single failure cf thu LFCI injectium valve.
For this event,.so LPCI idection flow is assumed and the reduction in LPCI fisw is of no import =nea.
For the next limiting.1arge break accident, the asaamed single failure is a diesel generator failure.
The operable systems are NFCI. ADS. one core spray pump and 2 LPCI pumps.
The reduction in the LFCI flow may increase the core unecvary time. However, plant specific calculations for the distevering the small increase Sg core uncovery produc4e only a plant de.mnstrate that Since the PCT for the 30"F increase in peak cladding temperatura (PCT).
LPCI failure is higher than the FCT for the diesel generator failure by over 100*F, the reduction in the LPCI tiew will not cause the large break event with diesel generator failure to become the limiting event.
For the enae of intensediate breake with diesel generator failers. the reduced LPC1 flow will result in a PCT increase of only 50*F.
Bowever, the difference between the PCT for the intermediate break eventa and the i
l liwiting large break events are larSer than $60'F.
Therefore, the reductivn in the LFCI flow will met cause the intermedtete break case to become "the limities event for the disoevering piant.
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seen as een tas7:4i s.33. tees isins P. e r
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p For the cc.ae. of small break events, the Teduced LFCI flow will have insignificant effect on the FCT. This is because the limiting small break event is with the asenmed ItPCI failvre and the avaitable systems are ADS, two core' spray pumps and four RMR pumps.. For "astell breaks, the two core spray pumps are more than adequeta for core cooling. Therefore, eke reduced LPCI flow will not degrade.the plant's core cooling capability for small break events.
i Ye should be noted that a realistic analysis would indicate that any low a -
preneure pump (LPCI or core spray) would provide adequate core cooling for any accident events.
e
/
IMPACT ON OTHER MODES _ 0F RNR OPERAT105 With the exception of the LPCI usode, all other MUL modes of opatation are agnually controlled and actuated.
Therefore,.the RHR minimum flow bypass -
valveu can he closed by proceduren for the non-LPCI modes of RER operation.
Failure tu close the bypass velve for t.he non-LPCI modas of ENR operation is less limiting than other pntential single failure.
CONCLtfSIONS A safety. evaluation was performed to access the impact of removing the l
automatic closing signal from the RRR minimum flow bypnas-valves. The l
proposed modification is to prevent a single failure of the RER flow sensins instrument causing the loss of all pinimeum flow nyrpass sepability for the four RilR pumps.
The evaluation found that removing the automatic closina signal will result in the plant complying with design basis requirements sua provide adequate cero cooling er containment cooling capetbility. Therefore, the safety evaluation concludes that the proposed.
modification will not result in an unreviewed safety guestion.
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