ML20247F114

Petition for Rulemaking 50-53 Reopening ATWS Rulemaking Proceeding in Light of Power Oscillations Occurring at Facility Following Dual Recirculation Pump Trip on 880309. Listed Encl Repts Should Be Made Part of Record

ML20247F114
Person / Time
Site:	LaSalle
Issue date:	09/06/1989
From:	Hiatt S OHIO CITIZENS FOR RESPONSIBLE ENERGY
To:	NRC
Shared Package
ML20247F119	List: ML20247F114
References
FRN-54FR30905, RULE-PRM-50-53 PRM-50-53, NUDOCS 8909180083
Download: ML20247F114 (8)

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$ Septemoer 6, 1989 (SWR 3MDS) gg ggp j j gjj;jg COMMENTS OF OHIO CITIZENS FOR' RESPONSIBLE ENERGYF'INQ.

1989)U M {("OCRE")

CN PRM-50-53 (54 FED. REG. 30905, JULY 25, g 4

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The petitioner OCRE herein submits comments and reports to be made part of the record in this petition to reopen the ATWS rulemaking proceeding in light of the power oscillations occurring at the LaSalle-2 BWR following.a dual recirculation pump trip on March 9, 1988. j The following enclosed reports should be made part of the record: i Exhibit 1: AEOD Special Report on the LaSalle-2 event, AEOD j Special Report No. AEOD/SB03, dated June 8, 1988. ]

1 Exhibit 2: excerpt from the NRC Augmented Inspection Team l report on the LaSalle-2 event. j Exhibit 3: letter report by the Advisory Committee on Reactor

' Safeguards, dated June 14, 1989, re Boiling Water Reactor Core Power Stability.

Exhibit 4: "A Report on Reactor Study Issue Number 25" prepared for the Ohio State University Expert Panel by Dr. William P.

Stephany of Nuclear Education & Training Services, Inc.  !

(" NETS"). (This report is part of the comprehensive review of the 1975 General Electric Nuclear Reactor Study (commonly known as the Reed Report) commissioned by the Public LTtilities Commission of Ohio. The attachments to the report are not included herein.)

These reports support OCRE's position regarding the ATWS rulemaking in light of the LaSalle-2 power oscillation event.

The AEOD report, which was largely the basis for OCRE's 1988 )

petition in this matter filed under 10 CFR 2.206, states that the LaSalle event " necessitates that ATWS mitigation be j reviewed in light of this event." Exhibit 1, p. 7, emphasis 1 added. The NRC's Augmented Inspection Team also expressed concern that, "in view of the large magnitude of the APRM ,.

oscillations in LaSalle, the AIT believes that the ultimate power level without scram is unknown, and that the 500%

bounding level assumed in the ATWS investigation may not be bounding. LPRM oscillation magnitudes more than seven times those of the APRMs have been observed in the case of regional ,

oscillations." Exhibit 2, p. 24. These reports illustrate the {

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NRC Staff's concerns.and advice on this matter, which OCRE is

-endorsing.

Significantly, the reason ATWS is still a safety issue today, despite.the ATWS' rule, 10 CFR 50.62, is that the Commission failed to follow the advice of its own Staff in the ATWS rulemaking process.- In NUREG-0460, Volume 4, the Staff made the following comments regarding ATWS in BWRs:

Several events are shown to have significant, periodic oscillations in neutron flux following an initial, large neutron flux spike. The staff has never before encountered

, this' type of accident behavior prediction, and so it has never been specifically considered in previous PCI evaluations. . .

. ?!' combined effects of high neutron flux spikes, resulting in high cladding and boil temperatures, followed by oscillation in flux,. fluid flow, etc., raise questions not only about fuel future, but also about the potential for loss of coolable (rod-like) geometry. The 2200 F, 17 percent oxidation LOCA limits that GE proposes as evidence of coolable geometry are not applicable here because those limits address cladding oxidation and embrittlement effects only. They do not address the potential effects of oscillating mechanical loads on wasted and collapsed cladding that might be " locked onto" the fuel pellets as a result of a BWR ATWS involving a high flux spike, nor do they consider center-melted oxide. NUREG-0460, Volume 4, pp. A-87 to A-89.

The recommended mitigative measure, an automatic, high-capacity standby liquid control system (300-400 gpm), would eliminate or greatly reduce the oscillations. NUREG-0460, Volume 4, pp.

A-64, A-48 to -52, A-43, and 29. Unfortunately, this* measure was not incorporated into the final ATWS rule. OCRE believes that incorporation of the automatic, high-capacity SLCS, along with the other provisions of the ATWS rule, 10 CFR 50.62, pertaining to BWRs, would allay any concerns about power oscillations resulting from the recirculation pump trip, which is a necessary feature to quickly reduce power and reactor pressure to avoid failure of the reactor coolant pressure boundary.

It is significant that independent reviews of ATWS and the LaSalle event also point out the need to reconsider the ATWS rulemaking in light of the LaSalle oscillations. Exhibit 4 ,-

prepared by NETS, a firm which provides consultants and services to the nuclear industry, concludes that "there is a large gap between the ATWS prevention and mitigation recommendations stated in NUREG-0460 and the ATWS Rule stated in 10 CFR 50.62. In light of the recent LaSalle event, the l

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consequences of the recirculation pump trip ATWS mitigation j feature do need to be reviewed, and the concerns expressed by i the NRC Staff in Vol. 4 of NUREG-0460 also need to be looked at again." Exhibit 4, p. 16. The ACRS also recommends that

" considerable attention be given in the longer term to the development of an improved understanding of the conditions that can lead to an ATWS compounded by core power oscillations."

Exhibit 3.

OCRE would urge the Commission to follow the advice of its Staff as given in NUREG-0460, the AEOD report, and the LaSalle Augmented Inspection Report, as well as that of independent reviewers such as the ACRS and NETS.

Respectfully submitted, Susan L. Hiatt OCRE Representative 8275 Munson Road Mentor, OH 44060 (216) 255-3158 3

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NUCLEAF REGULATORY COMMISSION W ASHING TO N, D. C. 20555

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i ME*0RANDUM FOR: Thomas E. Murley, Director

- Office of Nuclear Reactor Regulation Eric S. Beckford, Director Office of Nuclear Regulatory Research FRCM: Edward L. Jordan, Director Office for Analysis and Evaluation '

of Operational Data '

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SL5 JECT: AE00 CONCERNS REGARDING THE MARCH 9, 1988 POWER OSCILLATION EVENT AT LASALLE 2 Erciosed is an AEOD Special Report detailing our concerns about the LaSalle 2 pcwer oscillation event of' March 9,1988. We have reviewed calculations

- , .perfonned by Brookhaven on the BWR Nuclear Plant Analyzer, as well .as the <

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licensee's LER and other foreign and U.S. information. . Although this is the first event of this type at a domestic reactor, similar events have occurred .

in foreign reactors. Based on this review, we classify this event as an . . . . - ;_ . ,

important precursor event with significant safety concerns. . Our.mos; m ; _._

significant concerns and associated recommendations are described below. & qiw 4

1. The LaSalle event raises questions about the adequacy of the analysis used to meet the core stability requirements of GDC-12 when both ,

recirculation pumps are tripped. The event also points out the -

difficulties the operators face in rapid diagnosis of and responte to an event which readily promotes significant complicating factors such as subsequent loss of feedwater heating and reactor water level fluctuations.

Simple and unambiguous procedures are needed to assure prompt proper operator response which ensures jompliance with GDC-12. GE SIL 380 does not provide adequate guidance.

2. During startup and shutdown, BWRs routinely enter regions of potential thennal-hydraulic-neutron kinetics instability. This operation can be avoided without large impact on plant operations by modifying plant operating procedures to increase recirculation flow slightly early in the startup and by inserting control rods sooner during shutdown.

Several foreign reactors operate with power / flow operating restrictions that avoid the unstable regior.. Additionally, reduction or loss of forced recirculation flow during plant transients can result in the plant entering regions of potential instability. Prudent operator action is needed to restore stable plant operation and to avoid actions which could initiate events with more significant consequences. For example, restart of recirculation pumps following loss of feedwater heating or MSIV closure could result in additional reactivity insertion while the reactor was exhibiting power oscillations.

3. This event has implications regarding the reactor transient response to a recirculation pump trip during an ATWS. In particular, the power oscilla-tions may substantially exceed previously predicted values and thus raise questions regarding previous fuel integrity evaluations.

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Thomas E. Murley f-Conclusion

-The March 9 LaSalle event indicates serious deficiencies in the core stability

. analysis for LaSalle and perhaps other BWPs. Further, such undamped power oscillations call for prompt operator recognition and action, yet at LaSalle, L operators were not trained to recognize or respond to such oscillations.

Adequate plant procedures did not exist at LaSalle, and few, if any, plant simulators in the U.S. are capable of modeling these types of oscillations.

'It is not at all clear at this time that we understand the nature and potential consequence of such power oscillations considering such factors as improper or no operator action, alternative core configurations and equipment failures, or -

divergent localized power oscillations. Since it will take time to thoroughly analyze and understand the laSalle event and its implications on other BWRs, we

. conclude that, at least in this interim period, action is warranted to minimize the potential for core instability. Our recommendations in this regard are presented below. -- - -

We anticipate a written response to these recommendations within 45 days as discussed in NRC Manual Chapter 0515. - ^

Recommendation to NRR af-  :...

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Pending a full understanding of the LaSalle event ano its implications, we -

3 believe that all BWRs should be required to- rc . _ ;

(a) Immediately insert control rods to below the 80% rod line following 'i < ;

reduction or loss of recirculation flow or other transients which result '

in entry into potentially unstable regions of the power / flow map.

(b) Increase recirculation flow during routine reactor startups ano insert ~ -

some control rods prior to reducing recirculation flow below 50t during shutdowns to avoid operation 'in potentially unstable areas of the power /

flow map. 4 (c) Immediately scram the reactor if (a) or (b) above are not successful.

Recommendation to RES Review resolution of GIs B-19 and B-59 anc m'S rritigation in light of the LaSalle operating experience.

Please let me know if we can provide any clarification or additional assis-tance. If you have Questions regarding the enclosed Special Report, please call Jack Rosenthal on x24440.

Odn'slsiredBr E O b lan Edward L. Jordan, Director Office for Analysis and Evaluation of Operational Data

Enclosure:

As stated Distribution: See next page

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1 AE00 SPECIAL REPORT /

IINIT : La5611e'2 .' " SPECI AL PEPORT NO. : AE00/5E03 00CKET N0.: 50-374 OATE: June 7, 1988 LICENSEE: Comenwehlth Edison EVALUATOR / CONTACT: J.Kauffman/G.Lanik SU8 JECT: AEOD CONCERNS REGARDING THE POWER OSCILLATION EVENT AT LASALLE 2 (BWR-5) i EVENT DATE: March 9, 1988 - -

SUMMARY

The LaSalle event involved power oscillations caused by neutron flux / thermal .

hydraulic instabilities of a magnitude that were not predicted by design '

6nalysis, unanticipated by the operators, and potentially in conflict with . "

General Design Criterion (GDC) 12. Based on vendor analyses, two NRC Generic -

Issues (Gis) had previously been resolved concerning stability of BWRs; and r this event raises questions regarding the adecuacy of those resolutions.

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Since analyses predicted that these oscillations would not occur.. little ~ " ' "

' guidance and training were provided for operator detection and response. V -'

7 Further, operation in. unstable areas of the.BWR power / flow map has potential .-

. adverse safety consequences. Because LaSalle 2's core was calculated to'be ._; + , " .

more stable than the typical BWR core, other BWRs may be more susceptible to this problem.

Ir light of the present uncertainties, we recomEnd that BWP 'l'icensees should be required to implement proceoures to:

• a) Immediately insert control rods to below the 80s rod line fc,llowing 1 reouction or loss of recirculation flow or other transients which result l in entry into potentially upstable regions of the power / flow map.

b ', increhse recirculation flow ouring routine reactor startups ar.d insert scme control roos prior to reducing recirculation flow below 50t during shutdowns to avoid operation in potentially unstable areas of the cover / flow map.

c) Immediately scram the reactor if a) or b) above are not successful in preventing and suppressing oscillations.

We also recomend that NRR revisit Gis B-19 and B-59 anc ATW5 mitigation in light of the LaSalle operating experience. '

Description of the Event (Compiled from licensee's 50.72 report, Parch 9 T@, and references 1 through 5).

While perfortning the functional test on a differential pressure switch, an instrument maintenance technician inadvertently valved in the variable and reference legs with the equalizing valve open, thereby connecting the variable i and reference legs. This initiated a " pressure equalization" between the variable and reference legs, and resulted in a high " indicated" level to the

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l feedwater level control system, causing the feedwater purnos to begin reducing i finw. Realizing a valving error was made, the reference leg was iranediately )

I isolated from the variable leg. This resulted in a low " indicated" level  !

l . , . spike. . The level spike caused uthen. leve.1 switches, utilizino the same {

reference leg, to also actuate, including the trip of the reactor recirculation pumps from an Anticipated Transient Withnut Scram (ATWS) signel.

Due to the rapid power reduction from 845 to approximately 40: ca': sed by the trip of both recirculation pumps, feedwater heater high level alarms were received and heaters began automatically isolating. This resulted in reduced feedwater temperature and the insertion of positive reactivity due to the negative moderator temperature coefficient. With feedwater level control acecuately handling the level transient, the licensee tried 'to re-establish feedwater heating and to restart the recirculation pumps. Attempts to restart the recirculation pumps were unsuccessful.

With the unit in a high control rod line condition (power was 85* prior to the event) ano low flow condition (natural circulation), the unit started experiencing neutron flux oscillations from rapid creation and collapse of voids in the core region. Approximately 5 rninutes into the event, multiple high and low alarms were recorded by the local power range monitors (LPRMs). _

The average power range tnonitors (APRM) recorders were oscillating between 25%

and 50t of full power with an approximate 2 to 3-second period. Because of ~~'

a limitations of the APRM recorders, the actual neutron flux oscillations (approximately 757 power) were larger than the indications of the APPM, -

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recorders. The control room operators were in the process of menually < : . ~4 "

scramming the unit, when an autnmatic scram occurred on upscale neutron trip ^<- -

(118T on APRPs), imediately prior to the scram, the operators noticed that a rnajority of the LPRM Hi alarms were lit. The setpoint for the LPRM Hi alarms _

is 105; of full scale.

Foreign Operatino Experience 1 __ .

A number of power oscillation events have been reported by the net IPS system.

Power oscillations were reported in 1985 ano 1906 at a foreign EWR-3 in if 5-677 end 681. The oscillationsiwere I4i peak to peak curine natural circula-tion testing. In June 19E? in IR5'-220, e foreign BVR c reportec oscillations of 75 of the "mean" flux durina forced circulation af ter moving one control rod. The reactor trip;ed on APPV Figh flux citer five 4W holi scrent had been reset. These power oscillations had a 2.5 secono period. :n response, operatinn limits were established at that f acility tc prevent operation in the area of instabilities. Another event (IRS-220.2) at this reactor in January, 1983, demonstrated that it is possible to start these power oscillations from nc.rmal operating conditions. IRS-363 reported that in October,1983, during testing at the sarne reactor, divergent, out-of-phase oscillations were experienced. The report describing this event stated that this was "a potential GDC-12 violation." Again, opera ting res trictions' were implemented that reouire rapidly maneuvering the reactor to a stable region following a single recirculation pump trip. Information received as followup to these events indicates that operating instructions were also developea for loss of feedwater heating events, loss of all recirculation flow, and low rec i rcula tinn flow ccnditions. We have also received information that folinwino startup testing at yet another foreign tWP, ope

• ating instructions were implemented

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p to prevent routine entry into, potentially unstable areas. In particular, guidance was developed to prevent routine entry into these dreas during reactor startups ano shutdowns, to require increased monitoring of APPNs and LPRMs in

_-- potentially unstable areas, an? to provide guidance for operator response to certain transients such as loss of feedwater heaters and recirculation pump trips and restarts. In summary, these foreign plants have taken action to

. restrict or prohibit operation in areas of instability. Figure 1 is an example of operating restrictions during startup and shutdown in place at one foreign BVR.

i U.S. Operatino Experience Other than LaSalle, no events involving diverging power oscillations at BWRs I were identified in the SCSS operating experienc6 data base. However,sstartup '

testing and other testing have included inducing power oscillations, observing -

the reactor response, and testing the effectiveness of oscillation suppression

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~ Review of the data base since 1980 did capture 167 events involving a trip of

• cr.e or two recirculation pumps while the reactor was critical. Thus; when combined with routine startups ano shutdowns, it is clear that BWRs are -

.A frequently operated in potentially unstable regions. The nuinber of reported ' .

events is low since there are no reporting requirernents for recirculation pump '

trips, unless it is in conjunction with some other reportable condition. -

Sr.all power oscillations are similarly not reportable. ' - '

Related GDCs and GIs _

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The LaSalle event relates to two GCCs and two Gis: -

" GENERAL DESIGN CRITER10N 10 - Reactor Design. The reactor core and, associated coolant, control, and protection systems shall be designed with ~

appropriate margin to assure that 'specified acceptable fuel design limits are not exceeoed during any condition of'norinal operation, including the ef fects of anticipated operational occurrences."

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" GENET.AL DESIGN CFITER10N 12 - Suppression of Feacter Power Osciliatiuns. The reactor core 'and assuciated coolant, control, and protection systenis shall be cesigned tu assure that ocwer oscillations whict can resu't in conditions exceeoing specified acceptet.le fuel design limits are. rot possibli or can be reliably and readily detected and suppressea."

G! B-19: " Thermal - Hydraulic Stability" and GI F-f 5: "(N-1) Loop Operation in BWRs and FWRs".

These GIs were closed out by the issuance of Generic Letters 86-02 and 86-~09.

Generic Letter 86-02 stated that the approved GE and Exxon rnethods for calculation of core stability decay ratio are ur.certain by 20% and 25t, respectively, in predicting the onset of limit cycle oscillations (decay ratio

= 1.0). The Generic letter noted, "...BWR 4, 5. and Es may not be able to show compliance with GDCs 10 and 12 solely using analysis procedures to prnve that thermal hydraulic instabilities are prevented by desi However, the Generic Letter concluded that BWR 1, 2, and 3s should have'gn." sufficient margin.

It also stated that for cores which do not meet the analytical criteria (decay ratio less than 0.8), the operating limits of GE SIL 380 would be sufficient to provide for detection and suppression of flux oscillations in operating

4 regions of potential instability adequate to demonstrate corrpliance with GOC 10 and GDC 12 for cores loaded with approved fuel designs.

Generic letter 86-00 noted that the review of BWF (N-1) loop operation was ccerlicateo by potential thermal-hyoraulic instability and f et curro vibra tion problems during single loop oper otion. In low flow operating regions , it was necessary to develop special operating procedures to assure that GDCs 10 and 12 were satisfied in regaro to thermal-hydraulic instabilities. Flant Technical Specifications consistent with these procedures were accepted by the staff for reactors which were not demonstrably stable based on analyses using the then approved analytical snethods; details of the operating limitations were developed for GE SIL 380 and contributed to the resolution of GI B-19.

In addition, tests at Brown's Ferry demonstrated that single loop cperation had similar stability characteristics as two-loop operation under the same pcwer/ flow operating conditions. The tests confirmed the staf f's finding that Technical Specifications based on GE SIL 380 which were proposed fur some BWPs were appropriate for the detection and suppression of thennal hydraulic instabilities. The staff expected to approve single loop operation for licensees who submitteo the appropriate ECCS analysis.

Relevant Licensine Actions . , _

The foreign event involving out-of-phase, dit ergent oscillations.. resulted in

~ issuance of a board notification (No.84-062) in March,1984 Stability tests -

dernonstrated that " limit cycle oscillations" could occur within cennissable

. operating space below the rated rod line at natural circulation flow. The -

high power level (1207) scram protection which is baseo on APRM signals would u.

not necessarily prevent violation of critical heat flui limits if such . local instabilities were to occur. The test demonstrated that local thermal _*

hydraulic c5cillations which are out of phase with the APRFs could occur. it was unclear at that time (19f4) how high a local oscillation could reach before detection by an operating crew using then current monitorirg-procedures . +

This board notification was made af ter the issuance of GE 5!L 3rr, which is currently usec as guidance to operators for the ,e type of events. Plant Technical Specification changes were trade for piants undercoine licensing hearincs to oddress the concerrs o'f this board notification.

Previous Vendor Pecorrnenco tions General Electric Co., had r.reviously identified in GE Sit 380 anc ether dccuments thi. the concition of hich rod line and low flow was susceptible tn neutron flur/ thermal-hydraulic oscillations. However, based upon analysis, Co:nmonwealth Edison did not believe such oscillaticr.s woulo occur at LaSalle, and as a result, the SIL was not implernented.

Because this event at LaSalle involved large power oscillations, General Electric Co. has issued Rapid Information Communication Services infcmation Letter (RIC51L) No. 006 Fevision 1 pertaining to BWP core thermal hydraulic stability. The RIC51L supplements GE SIL No. 380 Revision 1 on t>e same ,

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Concerns Recardino This Event e ,. l

1. Stability analysis methods are highly uncertain. LaSalle 2's calculated

,, decay ratiu was approximately 0.6 for this fuel cycle. This means that that the transient reactor behavior that was observec during this event was predicted not to occur. The licensee's review of this event stated that the conditions prest-nt at the start of the oscillations appear to be only slightly more severe than the assumptions used to analyze the LaSalle decay ratio. There is also infomation that indicates that the

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stability analysis for Vermont Yankee was shown by stability tests as

non-conservative (Ref. 6).

2. LaSalle operators were not trained for this type cf event. Because GE antlyses predicted that this event would not occur at LaSalle, GE SIL 380 7' was knowingly not in place and operators not traineo on GE SIL 380 at laSolle, as allowed by Generic Letter 86-02.

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3. GDC 12 may have been violated. Although chemistry sarnples folluwing the LaSalle event did not disclose any fuel damace, the event was potentially a violation of GDC 12 in that undampened power oscillations occurred and no procedures or methods were implernented to reliably and readily detect  ;

. and suppress these power oscillations. - -

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4 Other BWRs may have a susceptibility to unstable power oscillations. -

Because analyses similar to the ones used at LaSalle are used at other ,

plants to meet GDCs 10 and 12, this transient response could occur at -

Other BWRs with decay ratios less than 0.8. Like laSalle, these other' '

BWRs may not have implemented procedures to reliably detect and suppress

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power oscillations. At LaSalle, the operators allowed nearly two minutes of unstable operation before deciding to take action to shut down the_

unit.

5. GE SIL 380 Revision 1, even'if implemented, is inadequate tu ensure compliance with GDC-12. This ra.ises the issue of the adequacy of GL 86-02 in assuring that GDC-12 is met for plants with predicteo decay ratios greater than 0.P. The$1Lhasar.umbefofinadequacies:

--APRM " noise" and not actual rapid power changes is discussed as a result of flow instabilities.

--This noise is said to normally rance between 4-121 (peak-to-peak) of rated power, whereas LaSalle reported power oscillations of nearly full scale (7U power).

--Some of the terms are not defined or cnmmonly uncerstood by utility operations personnel, e.g. " limit cycle oscillation." This makes it difficult to use as the basis for operator guidance and procedures.

--Puwer oscillations rray nut be readily identified and suppressed.

During an event with numerous failures ano alarms, it is not certain that operator attention will be promptly called to power oscillations, especially since the APRM instruments typically have large oscillations (noise up to 10% under nomal 100% power steady state operation) and the APRM recorders do not show the full magnitude of power oscillations ,

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.- due to time delays. Operators might consider any indicated oscillations a s, no ma l . ,

--The basis for the proposed actions is apparently non-conservative or

., sensitive to small parameter changes.

--Guidance is provided without explaining in detail why the actions are taken or the bases for the actions. Even in the case where out-ot-phase oscillations were experienced, GE SIL 380 states that "very large margin to safety limits were maintained." This downplaying of the potential severity of thermal-hydraulic instabilities may mislead operators into thinking that the stability concerns are not important.

6. Operator training on recognizing and responding to power oscillations is-poor. Few, if any, simulators used by utilities are capable of modeling -

the type of oscill6tions that occurred at LaSalle. Since the existing guidance in GE SIL 380 does not state that power oscillations from . ,

O to 120% power are possible and have been experienced, it is likely that very few licensed operators or training instructors were even aware that oscillation of this magnitude could occur. If operator action is necessary to ensure compliance with the GDCs, it is essential that licensed operators be trained regarding the assumptions, conditions, limitations, etc. of - ' '

the operating concerns. However, simple guicance - such as: " reduction or loss of recirculation flow resulting in entry into a potentially unstable area, insert control rods to below the 80% rod line" - that ' J ensures avoidance of the unstable or unanalyzed regions is preferable ~to '

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} .3 reliance on operator memory to ensure operation within analyzed regioris.'

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Improper operator actions could worsen the eveht. The operators at f

LaSalle tried to restart recirculation pumps because their training and procedures allowed therr. to do so. In this event, with a downcomer f.illed -

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. with cold feedwater and an unstable reactor, a successful restart of recirculation pumps would lead to further rapid reactivity' inser' tion with potential adverse consequences: We are also concerned abo'ut the effects M[- that would have occurred if additional reactivity insertion due to void l collapse in response to a turbine trip or an MSiv closure had occurred during the power oscillations. 40ther operater actions, plant conditions, such as end of cycle or cifferent pcwer distribution, or plant transients

. may have resulted in fuel damage.

Several calculations using the BWR Huclear Plant Analyzer were performed

. by Brookhaven at AE00 request. The simulaticn of the LaS611e event is shown in Figures 2 through 5. By pararnetrically ir. creasing loop flow resistar.ces, it was possible to generate power oscillations similar to those experienced at LaSalle. Prelimir.ary results from these runs indicate that large reactivity changes occur during these events. The power oscillations experienced at LaSalle are cyclic interactions of core void famation, flow, and neutron power. The period of the oscilla-tions is about 2.5 seconds while the thermal time constant of the fuel is 5 to 7 seconds; anc consequer.tly, direct gamma heating of the coolant is the likely energy feedback mechanism. This phenomena apparently begins with themal-hydraulic instabilities arising due to relatively large two-phase resistance in the core, while the driving head end flow rate are f

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i low due to loss of furced circulation. Fomation of voids then drives 2 neutron power down which slows further void fortnation, resulting in lower two-phase . flow resistance, and increased natural circulation flow into i

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the bottom of the core. This col'd water increases core reactivity and results in a power increase. The resultant voio formation continues t the cycle of oscillation. Large neutron power oscillations are the result of large reactivity changes.

Preliminary results from the Brookhaven analyzer indicate that large reactivity changes occur during these events. Figure 4, for example, represents the LaSalle base case, where the analyzer calculated 0.5 dollars total reactivity inserted just prior to the reactor trip. -

i l The LaSalle event is an important precursor event. Although-the' l S. - - .

consequences of thi_s particular event were not serious, they could have l been worse in other circumstances. First of all, the potential exists for l localized power oscillations where one half of the core oscillates 180 l degrees cut of phase with the other half; and in that case the APRM trip would not trip the reactor until the amplitude of the local power oscillations was much greater. An actual event of this type is noted in ,

the foreign operating experience. . Secondly, the potential exists for f operator action or plant equipment failure to worsen the esent, for -

example, restart of a recirculation pump or M5!V closure could result in ' ~

additional reactivity insertion. . ..

9. Previous efforts taken in regard to ATWS mitigation may be inadequate. .. .:

The action of tripping recirculation pumps automatically and inducing an " '

event similar ic the laSalle event when it is not clear where the power -

oscillations would stop anc what the effects of these oscillations would be in the absence of an automatic scram, necessitates that:ATW5 - -

mitigation be reviewed in light of this event. -

10. The resolution of G!s B-19 and'B-59 may be inadequate. The analyses which fortn the technical bases for the resolution of these issues have been challenged. The LaSalle event was predicted.by analyses to be prevented by design, but it ucciarred.

potential Actions to Address the Problem

1. We recommend that EWR licensees should be recuired to develop and implement procedures to:

, al Immediately insert control rods to below the BN rod line following reduction or loss of recirculation flow or other transients which result in entry into potentially unstable regions of the power /finw map.

b) Increase recirculation flow during routine reactor startups and insert some contrul rods prior to reducing recirculation flow below 50% during shutdowns tn avnid operation in poter.tially unstable areas of the power /

flow map.

s c) Imediately scram the reactor if a) or b) above are not successful in preventing and suppressi69 oscillations. '

)

2. We also recomend that hRR revisit GIs 8-19 and fi-59 and ATW5 ritigation-1

-' l in light of the LaSalle i,peratin'gexperience.

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' Report No. 50-373/88008; 50-374/88008 l i

Docket No. 50-373; 50-374 License No. NPF-11; NPF-18 I Licensee: Commonwealth Edison Ccmpany P. O. Box 767 Chicago, IL 60690 l

Facility Name: LaSalle County Staticn, Units I and 2 ]

Inspection At: LaSalle Site, Marseilles, IL Inspection Conducted: March 16 through 24, 1988 Inspectors: NRC Augmented Inspection Team Team Leader: M. A. Ring ] } /!26 Dat'e Team Members: R. A. KoprivDNIh /, [/:'ki Date L.E.Philiih h,' ,

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Approved By: W. L. Forney, Chief ~ :i ,

Reactor Projects Branch 1 Date Insoection Summary Insoection on March 16 throuah 2t., 1988 (Recort No. 50-373/88008(DRP);

50-374/88008( DRP))

Areas insoected: Special Augmented Inspection Team (AIT) inspection conducted in response to the dual recirculation pump trip and subsequent core power I

l I . 27

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, a[ oscillations resulting-in a-reactor trip on March 9, 1988, at LaSa11e,' Unit-2.

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. The'. reviewlincluded root; cause determination,- safety significance, performance f; .

'of: operators and equipment, adequacy ofiprocedures, effects on the reactor,

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. reporting' actions and potential. generic implict.tions. l-Results:_ No violations or' deviations were jdentified; however, the licensee,

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e has committed to. procedure:and Technical Specification changes as well as.

further study in the areas;of inherent shutdown r.sechanisms, instrumentation .:

" . capability and uncertainties-in the decay ratio calculations. The licensee's ,

-. interim report, as required by the CAL, is included as attachment 5 to this ,

report.

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4 . i is needed to assess.the' nature and magnitud'e-of neutron-t

flux oscillations and the' safety of restart after an- l

.in' stability event. -

'~ LLaSalle and:some other BWRs do not have high speed data recording instrumentation which can be committed for  !

availability during plant operation. -

4; Oscillation Characteristics Some characteristics of the LaSalle neutron flux oscillations.

were atypical of previous events and have led to concerns about '

-the. applicability of previous' safety analyses. .The magnitude of in phase' limit cycle oscillations previously observed on the

1' APRMs during special stability tests and operating reactor events were. typically in the range of 5% to 15% (peak-to peak) of rated power, and as high'as 25%. This compares to peak-to peak values of'about 100% at the time of the 118% neutron flux trip for LaSalle.

The-estimated value of' local power at the time of trip was greater than 310% and LPRM readings indicate that the core power peak shifted and' increased by 25%. Even though the fuel LHGR limit of 13,4 kw/ft was not exceeded because of the thermal time constant of the fuel, the increased power peaking was unexpected based on Vermont Yankee stability tests, and was.not factored into the generic' safety evaluation performed by GE during review of tne thermal hydraulic stability Generic' 1ssue B-29.

The previous GE safety a'nalyses considered several limiting moderate frequency. transients which were initiated while the neutron flux was oscillating.below the 120% scram setpoint, and included a rod withdrawal error with the flux oscillating up to the 1200 scram level. Additional analyses were performed to evaluate the impact of oscillations that apprcached 300% of rated neutron flux (e.g., regional oscillations) without scram j

. prior to rod insertion and termination of the event. All of these analyses showed that significant fuel thermal margin existed to safety limits. While there are several aspects of these analyses which differ from LaSalle (initial power level ,

i and amplitude of the oscillations; no change in bundle peaking I

factors due to the event, etc.), the AIT agrees that they are j ,

sufficiently representative and conservative to demonstrate that j. I no fuel thermal or mechanical limits were exceeded during the j i

event. However, reliable detection and suppression provisions are necessary to assure protection against future events which could involve regional oscillations to higher power levels.

The licensee was also asked to review the impact of the event on stability considerations addressed in the 1979 GE Generic ATWS report, " Assessment of BWR Mitigation of ATWS" (NEDE-24222).  !

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, potential impact of limit cycle neutron f dux ould not l j

I 500% of rated bundle power following recircul j result in sufficient fuel clad temperature i varIt wasanfurther concluded ycles.was l

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affect fuel integrity. f theto lim t c clad integrity due to prolonged exposure deposited l

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acceptable consequence in view of the importance o l recirculation pump trip ((RPT) to minimize the energythereby ll! l in the suppression poolpressure withinin limits) during an A i

In view of the large magnitude of the '1evel withoutoscillat ons APRM LaSalle, the AIT believes that the ultimate powerd in the ATWS scram is unknown, and that the 500% level assumeLPRM oscillati been observed in investigation may not be bounding.

The licensee reports that more than seven times the case of regional oscillations.

those oferthe APRMs limits) and have i the BWROG is discussing this issud'(inherent powJuly 1, 1988.  !

the licensee will provide a status report on

5. Additional Concerns licensee in Several additional concerns were his presented report. to theThese que the form of questions. response are contained in Attachment 5 t B. Recommendations in items IV.A.I The AIT recommends that the concerns identifiedNRR for generi through IV.A.5 of this report In thebeinterim, examined thebyAIT recommendsin IV.A LaSalle specific resolution. i i d via letter to that revised stability TS as discussed LaSalle Units 1 and 2 and the licensee be author z ed they rem modify interim operating procedures provideThe revised techni ion of high worth with the new T.S. procedures should pumps incorporate the changes sum are operating (Appendix A, Item 3), which include immediate insert rods and observation of ApRM/LpRM noise Itwhen is noThe reactor is and power is above the 80% Rod Control 5 Line.lity minutes) to is suspected.

tripped immediately whenever instabi h is sufficient expected that the time available (greater t an to instability following a two pump trip transientd for reactor trip Proposed procedures permit manual power reduction, avoiding the neeprior to s i

unless the core is unstable by a large marg (n. result in s permit manual action for up to two minutes reverse operating actions which mayinstability when on V. AIT CONCLUSIONS b ed on LaSalle Unit 2 The AIT finds that the core power oscillations o servl error resultin on March 9, 1988, were initiated by a personne 24

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