ML20137J540

Forwards Generic Ltr for Review & Issuance Re Resolution of Generic Issue B-59 Concerning (N-1) Loop Operation in BWRs & PWRs

ML20137J540
Person / Time
Issue date:	08/20/1985
From:	Bernero R Office of Nuclear Reactor Regulation
To:	Thompson H Office of Nuclear Reactor Regulation
Shared Package
ML20136A677	List: ML20136A670 ML20136F083 ML20136F944 ML20136F952 ML20136G046 ML20137J540 ML20137X272
References
TASK-B-59, TASK-OR NUDOCS 8508300394
Download: ML20137J540 (5)

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AUG2.ORG5 MEMORANDUMFdR: H. L. Thompson, Director Division of Licensing FROM: Robert M. Bernero, Directer Division of Systems Integration

SUBJECT:

GENERIC LETTER FOR ISSUE B-59 (N-1) LOOP OPERATION IN SWRs AhD FWRs

Reference:

Letter from H. R. Denten to R. N. Bernero, " Schedule for Resolving and Completing Generic Issue No. B (N-1) Loop Operation in BWRS and PWRS."

The above referenced letter directed DSI to work with SL to develop a generic letter to inform licensees and applicants of the resolution cf generic issue No. B-59. We have enclosed a generic letter on this subject for your review and issuance. It shculd be noted that this letter references a generic letter on B Thermal Hydraulic Stability which should be issued prior to the S-59 letter.

OriginalSign93 Zg:

RobertE.Bernero* ,

Robert M. Bernero, Director Division of Systems Integration

Enclosure:

As stated cc: H. Denton D. Eisenhut L. Riani T. Speis W. Russell

Contact:

G. Schwenk, DSI:CPB i X.29421 DISTRIBUTION Central Files r/f / h CFB r/f e . -

R. Derriero L. Rubenstein Y W 1 C. Be-itnger k. b fi~

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ATTACHMENT

_ GENERIC LETTER T0: All Licensees cf Operating BWRs and PWRs and License '

Applicants

SUBJECT:

TECHNICAL RESOLUTION OF SENERIC ISSUE NO. B 59 - (N-1) LOOP

, OPERATION Ili BWRs AND PWRs i

The staff has been studying (N-1) Operation in BWRs and PWRs under Ger,eric issue ilo. 3-59. We have recently completed our review cf this issue and the '

perpose of this letter is to inform you of our findings on the resolution of  :

1 Geo9ric Issue ilo. B-59.

The ma.jor!ty of the presently operating BWRs and PWRs are desigr:ed to operate ,

with less tien full reactor coolant flow. If a PWR reactor coolant pump or a BWR recirculation pump becomes inocerative, the ficw provided by the remaining

-(N-1) loops is sufficient for steady state operation at a power level less than full power. Although the FSARs for the IIcensed BWRs and PWRs present (N-1) loop calculations showing allowable power and protective system trip set-points, the NRC staff has disallowed this madE of operation for most plants I primarily because of insufficient ECCS analyses as well as thermal-hydraulic J stability concerns associated with BWRs. At present. BMR and PWR ficensees have Technical Specifications which require shutdown within several hours if one of the reactor coolant loops becomes inoperable.

j The staff recently completed e safety evalu.ation report (SER) (Ref.1) for the l request by Seaver Valley Unit I for (N-1) loop operation. Based on that SER, it

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is expected that Beaver *! alley Unit I wilt be authcrized to operate with (N-1) j i

loops when the Technical Specific 6tions are revised and updated appropriately i

in the near future. As a result of our reviets of the Beaver Valley Unit 1

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GENEMC LETTER 2 submittal the staff expects that there will not be any unacceptable consequences 1 associatedwitn(N-1)IcopoptrationinPWRs. However, the specific design of each plant will be reviewed using the same criteria for acceptability as was

used for Etaver Valley. Therefore, it. is anticipated that if other PWR owners sutcit their (N-1) Icop ar.alysis for a similar revis as was performed on Ecaver Valley Unit I that their pl.ents will also te authorized to operate with (N-1) loops in service, Tiw review cf BWR (N-1) loop operation has been CO3 plicated by potential thermal-hydractic instability ar.d jet pucip vibration problems during single icop operation (SLO). In low flow operating regions, it has been nec,essary to deveicp speciel operating procedures to assure that General Design Criteria 10 and 12 are satisfied in regard to therraal-hydraulic fr. stabilities. Technical

, !pecifications consistent with these procedures have been accepted by the staff fcr reactors which are not den:onstrably stable based on analyses using approved '

analytical roethods; eetails of the requirment were developed as the technicai .

resolution of Generic Issue B-19 and are the subject of our Generic Letter

*(No. ) (Ref. 2). In addition, in an Effort to re3olve certain plant-specific

concerns about jet purp vibration or thermal hydraulic instability irt the single loop operating acde at Browns Ferry Unit 1, TVA completed tests fcr that Yeactor ,

on February 9,1985, and these concerns have been resolved. These tests demonstrated that SLO has similar stability characteristics as two lecp  ;

operation under the sarce power /ficw operating conditions. ney also confirmeo

, the staff's finding that Technical Specificattens based on GE SIL 380 which have been proposed for scue BWRs;are appropriate for the detection and #

suppression of thernal-hydraulic instabilities. Recently, Technical Specifications based on GE SIL 3.80 were submitted for Duare Arnold. Penr.ar.ent .

Sta has been approved for Ouane Arnold (Ref. 3), and the staff expects to approve permanent SLO for other owners who have snbmitted SLD ECCS analysis as soon as Technical Specification changes sirwilar to those for Cuane Arnold are submitted.

• Note the B-19 letter should predate the issuance af our letter on B-59.

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t SENERIC LETTER 3 i This Generic Letter does not' involve any reporting requirements'so that no OMB clearance is necessary, i

Sincerely, t ,

e H. L. Tftompscn, Director ,

4 Divisfon of Licensing i

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GENERIC LETTER ~4

REFERENCES:

1. Letter from S. Varga (NRC) to J. J. Carey (Duquesne Light Company).

" Beaver Valley Unit 1 - Operation With Two Out of Three Reactor Coolant Loops - Safety Evaluation," dated July 20, 1984, Docket No. 50-334,

2.

• Generic Letter - B-19*

3. LetterfromMohanThadani(NRC)toL.Liu(IowaElectricLightandPower Company),datedMay 28, 1985, Docket No. 50-331.

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NUCLEAR REGULATORY COMMISMGX f wasnieentong. noses i ,

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h i MEMORANDUM FOR: . Robert M. Bernero Director I Division of Systems Integratial '

,y FROM: Harold R. Denton, Director '/

1 Office of Nuclear Reactor Regulation v b[l7k

SUBJECT:

SCHEDULE FOR RESOLVING AND COMPLETING GENERIC ISSUE  :

NO,B-59-(N-1)LOOPOPERATIONINBWRsANDPWRs i.

Generic Issue No. B-59, "(N-1) Loop Operation in BWRs and PWRs" has been prioritized as a " Regulatory Impact" issue that is resolved for BWRs and PWRs as explained in the enclosed evaluation. BeaverValley'sSER(MPAE-05) is the basis for resolution for PWRs. For BWRs, the issue is resolved on the basis that Generic Issue B-I9, " Thermal Hydraulic Stability" is resolved, the plant-specific tests at Browns Ferry and the review of licensee submittals for MPA E-04, "BWR Single-Loop Operations." Since the issue was resolved by several different means, you should work with the w, ~ Division of Licensing to develop a generic letter to inform licensees and applicants of the resolution. -

In accordance with NRR Office Letter No. 40.." Management of Proposed Generic Issues," there is no resolution to this issue to be monitored by the Generic Issue Management Control System (GIMCS). However, the attached prioritiration evaluation will be incorporated into NUREG-0933. "Prioritization of Generic '

Safety Issues," and is being sent to other NRC offices, the ACRS, and PDR for

• coments on the technical' accuracy and completeness of the prioritization

evaluation. Any changes as a result of coments will be coordinated with you.

Should you have any questions pertaining to the contents of this memorandum, pleasecontactLouisRiani(24563),

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Harold R; Denton, Director Office of Nuclear Reactor Regulation Enclosure Prioritization Evaluatico .

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[3 CLOSURE

i. PRICRITIL4TICH EVAWATION Cer.ari: Issue No. B-59

"/H-1) LOOP 6PERATION IN BWRs AND PWRt" 4

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ISSUE NO. B-59: (N-1) LCOP OPERATION IN BWRs AND PWRs

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DESCRIPTION ,

Hfstorical Background 1he majority of the presently operating BWRs and PWRs are designed to opersate with less 'than full reactor coolant flow. If a PWR reactor coolant pump or a '

BWR recirculation pump becomes inoperative, the flow provided by the .

! remhining (N-1) loops is c7fficient for steady state operation at a power level less than full' power. Although the FSARs for the licensed BWRs and PWP.s present (N-1) 1 cop calculations showing allowable power and protective

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system trip set-points, the NRC staff h.as disallowed this mode of operation for most plants primarily due to insufficient analyses. At present BWR and PWR licensees have Technical Specifications which require shutdown within a fahly short time if one of the reactor . coolant loops becomes inoperable.

A? lowing (N-1) loop ope /ation gives utility operators more flexibility in I deciding whether tu shut down a plant or let it, operate at a reduced power level, in this issue, (N-1) logp Operation is restricted to that resulting from a single reactof coolant pump failure. When fixing an out of-service .

pump be omes a major task, it is not expea:teti that the pumps will be repaired

while the pir.nt is on-14e. By contituirig operation in the N-1 mode, the f spair work say be postpc,ned untJ1 a scheduled refuelir.g time.

In c6nnection with multi plani action (MPA) E-05 . safety ev.aluatic; fepcrt (SER) was torpht.ed in Juli 19p4 (reference G) for the tequest by Beaver Valley Unit No< 1 (BV-1) for N"1 loop operation. Based on that SER, it is' i

egected that BV-1 will be authorized to operat'a with N-11 cops when the TechnictiSpscifTeationsarerevisedEndupdctedappropriatelyinthenear future. T e 55E for BV-1 repres6ntt the resolution of this isste for FWRs so .

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that this issue is resolved for PWRs. For PWRs the program manager recently .

requested the Director of DL to c'ese out MPA E-05 on the basis that there are no other active PWR applications for N-1 operation and, further, that none are expected in the foreseeable future (reference H). On the other hand, MPA E-04, "BWR Single Loop Operation," covers ten licensing actions on BWR submittals for N-1 or single loop operation (SOL) for seven licensees. '

The staff has reviewed the requests and submittals from the BWR licensees and has approved them such that N-1 loop operation for BWRs would be authcrized if tb licensees submit the appropriate changes to their Technical Specifications. The question of potential thermal-hydraulic instability problems during sing'le loop operation for BWRs and how restrictive the Technical Specification changes would have to be was raised in References I i andy,butthisissuewasresolvedinGenericIssueB-19. However, in an effort to resolve certain plant-specific concerns about thermal-hydraulic instability in the Browns Ferry plant, TVA completed tests at the Browns Ferry plant on February-9, 1985, and those concerns have been resolved. The tests demonstrated that technical specifications based on GE SIL380, which have

, been proposed for some BWRs and approved by the staff, are unlikely to result in any limitation on the achievable power level in SLO. They also indicated (pending verification by data analyses) that SLO is not significantly less stable than two loop operation under similar power / flow operating conditions.

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,e g e Permanent SLO has been approved for Peacn Bottom Unit 3, Quad Cities Units 1

& 2, and Dresden Units 2 and 3 and will soon be approved for Duane Arnold.

l The staff expects to approve permannt SLO for the SLO applicants when

{ appropriate Technical Specification changes have been submitted.

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The prioritization analysis will be limited here .to N-1 loop operation for BWRs inasmuch as the issue is essentially inactive for PWRs. This

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i prioritization has been completed with technical assistance from the Pacific Northwest La'boratories (reference F).

SAFETY SIGNIFICANCE  ;

In the event that a loop becomes inoperative in an operating plant, it is not always feasible to place it back in service by the repair of the failed pump while the plant is on-line. The plant operation with the N-1 loops, however, will not differ from operation with all loops except for the requirement to operate at a decreas,ed power level for the lower flow condition and with corresponding instrument / control set point limitations. The accident I

sequences would be essentially the same as with all loops in operation and .

,. there will be no change in accident initiator frequencies. Moreover, the loss of a loop because of pump malfunction would not impair the function of the ECCS and the other on demand systems should an accident initiator arise.

There had been some concern that operation with one loop out of service could result in thermal-hydraulic instabilities and possibly core damage at low flow conditions as well as with jet pump vibration at high flow conditions that could lead to damage to the reactor internals. But this matter has been adequately resolved. Therefore, the safety issue resolution would affect public risk or occupational exposure only slightly and might reduce risk because power and fission product levels would ,he smaller than at full power.

The purpose of this change is to reduce the impact on licensees.

j POSSIBLE SOLUTION I

The purpose of this task is,.to, develop a set of acceptance criteria, review guidelines and technical specifications changes for the (N-1) loop l authorization requests. This set of criteria,* guidelines and technical l specification changes will encompass accident scenarios (both LOCAs and non-LOCAs) to be analyzed by the licensees, computer models acceptable to NRC

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l for these analyses and acceptable input parameters in terms of reactor operating conditions (such as allowance for uncertainties in power level and fluid measurement). This has already been accomplished for PWRs by virtue of the completion of the Safety Evaluation Report for Beaver Valley Unit No. 1 (Reference G). In addition, the BWR analyses have been reviewed and accepted by the staff and the appropriate generic Technical Specification changes to allow single loop operation for BWRs have been identified.

PRIORITY DETERMINATION Frequency /Coneequence Estimate When operating a nuclear plant at a power level proportional to a reduced number of loops, the safety margins are somewhat increased from those at full power, but this increased margin is not regarded as contributing to a significant reduction in risk. Therefore, any potential risk reduction associated with this issue is perceived to be negligible. Moreover, no-

. additional occupational exposure is anticipated for this issue inasmuch as major loop repair is likely to be done during scheduled downtimes.

Cost Estimate

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To estimate the cost to industry, it is assumed that the amount of work j

! performed by Duquesne Light for Beaver Valley Power Station'No. 1, (BV-1), to i analyze plant performance will be comparable to that required for BWR plants l 1

(reference B). l

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.. t BV-1 analyzed a (N-1) loop large break LOCA and 12 non-LOCAs. Accidents involving the partial loss of forced reactor cdolant flow, startup of an inactive reactor coolant loop, single

• reactor coolant pump locked rotor, and complete loss of forced reactor coolant flow were analyzed in the original l

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5-FSAR. Therefore, they were not reanalyzed. This leads to 13 transient l scenarios to be analyzed.

Using a resource requirement of 5 man-wk and 15 computer hours for each case leads to a total of 65 man-wk and 195 computer hours. Another 30 man-wk per -

plant are allowed for preparing technical specification changes, modifying and upgrading procedures and/or systems and familiarizing operations staff with upgrades. Using the industry rate of $2270/ man-wk and an estimated computer cost of $1000/hr, the total implementation cost is estimated to be approximately $420,0,00 per plant for BWRs as well as PWRs. In addition, the plant specific tests"run by TVA at the Browns Ferry plant on the weekend of February 9, 1985, required operation at reduced power ranging from 50% to 65% ..

for about 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. On this basis it is estimated that the cost to conduct the test, obtain replacement power, and reduce the data will not exceed

$150,000.

The labor and analysis required for operation and maintenance of the resolution of this issue by the licensee is estimated to be negligible.

With the implementation of (N-1) loop operation, plant downtime can be reduced. The results of reference A indicate that the main contributor to (N-1) loop operation are pump seal failures. For.0conee 1, 8S% of pump failure events are due to pump seal problems and 99% of pump maintenance time is on seal fixes. Since the non-seal failures only contribute 1% of the total maintenance time in the Oconee case, we ignore them for this analysis and use pump seal failure probability as the probability of losing one loop and operating under (N-1) 1.oop conditions. Wh!1e the failure frequency of PWR pump seals that contribute significantly to core melt frequency is only 0.02 per plant year (reference C), seal failures'that result in the loss of one loop are estimated to be at a rate of 0.5/py for both PWRs and BWRs. Some l

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of these are,C0CAs or would become LOCAs, but these can be isolated by the BWR recirculation loop valves.

If a plant is base loaded, it is more economical to shut down and repair a seal that fails more than 20 days before the end of a 540 day cycle than continue with one-out-of-two loop reduced power operation. But all plants are not run at full power. Also one loop operation allows flexibility in shutting down to make repairs. Therefore, it is assumed that out of the 0.50/py events, at best only one-third of the events will be continued as (N-1)loopoperation i.e.,0.17/py.

The savings in tems of the avoided outage is estimated from Reference D to .,

be10 days (averageextraoutagetimeperpumpsealfailure). Therefore, the i savings from avoidance of outage per reactor-year is:

(10 days)(1 loop /2 loops)($300,000/ day)(0.17/py) = $255,000/py It is noted that the licensee implementation phase covering the analysis and evaluations of the potential accident sequences have already been submitted to the NRC in many cases for BWRs. Therefore, it will be assumed that the i

remaining plant analyses and/or re-analysis that may be required by the NRC

> staff will affect one-half of the total number ,of, reactors. Also, it will be assumed that the cost savings resulting from the avoidance of outages with (N-1) loop operation is the same each year for the industry. Further, l assuming an average reactor lifetime of 28 years and 44 BWRs, the total industry costs are estimated to be as follows:

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Implementation: (1/2)($420,000)(44BWRs)=$9,240,000 TVA test at Browns Ferry: $150,000 .

1 Operation and Maintenance: Zero i

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, Out$geAvoidance: -($255,000/py)x(44BWRs)=$11,220,000/yr ,

The present worth, PW, of the annual savings from outage avoidance over the average reactor lifetime of 28 years at a real discount rate of 5% is:

0 PW =($11.22x10)[(0.05)-I(1-(1+0.05)-28]=$168.300,000 BWR The cost to NRC of developing a set of acceptance criteria and review guidelines and technical specification changes for the issue are negligible inasmuch as these haye already been identified. Some additional effort will be required to revisd Chapter 15 of the Standard Review Plan to reflect the criteria needed to review N-1 loop operation. The revision to the SRP may .

require approximately 4 man-weeks. In addition, NRC labor to support SER implementation should be minimal at about 1 man-wk/ plant.

The total costs for the NRC are:

Development of Safety Issue Resolution: Zero

! Revision to the SRP: (4 man-wk)($2,270) = $9,080

Implementation: (1 man-wk/ plant) ($2270/ man-wk) (44 plants)

= $99,800 ,.,.

I Total NRC costs are: $110,000 approximately. i CONCLUSION ,

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It is concluded that this is a regulatory impact issue which has been resolved for BWRs and PWRs. For PWRs,the issui has been resolved on the basis of the Beaver Valley SER (MPA E-05) and for BWRs the issue has been resolvedo}nthebasisofGenericIssue,B-19,theplant-specifictestsat l =

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Browns Ferry $nd the review of licensee submittals under MPA E-04, "BWR Single-Loop 5peration."

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. REFERENCES A. Clark, R. H. and Barrow, W. E.,1979, Limiting Factor Analysis of High Availability Nuclear Plants, EPRI-NP-1138, Electric Power Research Institute, Palo Alto, California.

B. Dunn, C. N. (Duguesne Light) October 27, 1978, Letter to A. Schwencer (NRC), " Beaver Valley Power Station, Unit No. 1, Docket No. 50-334 Request for Amendment to the Operating License - No. 35."

C. Kolb, G. J., et al, 1982, Interim Reliability Evaluation Program:

Analysis of the Arkansas Nuclear One - Unit 1 Nuclear Power Plant Vol.1

, of 2, NUREG/CR-272-87, U. 5. Nuclear Regulatory Comission.

D. Makay,E.,ND, Adams,M.L.,1979,OperationandDesignEvaluationofMain Coolant Pumps for PWR and BWR Service, ERPR-NP-1194, Electric Power Research Institute, Palo Alto, California.

E. Olsen, E. A. J., 1981, Nuclear Unit 0)erating Experience-1978 and 1979 Update, EPRI-NP-2092, Electric Power Research Institute, Palo Alto, California.

F. NUREG-2800, " Guidelines for Nuclear Power Plant Safety Issue Prioritization Infonnation Development," February 1983; Supplement 1 May 1983, Supplement 2, December 1983.

G. Safety Evaluation Report, " Beaver Valley Unit 1 Operator With Two Out of Three Reactor Coolant Loops," July 20, 1984.

H. Memo from D. Wiggington to Eisenhut, "Closecut of MPA E-05; Westinghouse N-1 Loop Operation," January 11, 1985.

I. Note from Clark, R. J., to Lainas, G. C., " Status of Single Loop Operation for BWR," October 2,1984.

J. Memo from D. Eisenhut to R. Bernero, "BWR Thermal-Hydraulic Stability Technical Specifications," November 16, 1984.

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I NUCLEAR FUEL AND SERVICES DIVISION e SAN JOSE, CAUFORNIA 95125 3

February 10, 1984 -

File Tab A SIL No. 380 Revision 1 Category 1 BWR CORE THERMAL HYDRAULIC STABILITY The possibility of thermal hydraulic instability in a BWR has been investigated since the startup of early BWRs. These early tests oscillated response ~of a control the core. rod within one notch position and measured the For modern higher-power density reactors, pressure pertubation techniques were developed to measure the core stability been margins.identified previously Based on these' tests and analytical models, it has high power / low flow corner o(f theService Information Letter 380) that the regionofleaststabilitymargin. power /flowmap(Figure 1)isthe startup/ shutdown, during rod sequence exchanges and as a result of aThis recirculatiori pump (s) trip event.

Service Infonnation Letter 380 dis-cussed the possibility of increased neutron flux noise and recornended appropriate operator action in the event that neutron flux noise of increased magnitude occurs. 'As the result of new stability test data, obtained. information additional on BWR thennal hydraulic stability has been As such, this revision of SIL-380 is made to reflect the new information and to provide additional operating recomendations in the unlikely event that thermal hydraulic instability induced neutron flux oscillations occur. This SIL-380, Revision 1, replaces SIL-380 issued August GE BWR 1982 fuel. in its entirety and applies to General Electric BWRs using i

DISCUSSION BWR cores typically operate with the presence of global neutron flux noise in a stable mode which is due to random boiling and flow noise.

This noise, although exhibiting a dominant frequency of 0.3 to 0.7 Hz (the natural frequency of the BWR), does not result in sustained limit cycle oscillations since the system is in a stable mode. This occur-rence of neutron noise is best characterized by the Average Power Range Monitor (APRM)signalwhichtypicallyshowsneutronfluxnoiselevelsof 4-9% (peak-to-peak) tion pumps in operation. at rated power / flow conditions with two recircula--

(SLO), neutron noise levels.of 4-12% of rated (peak-to-peakDuring have been s:

report.2d for the range of low to high recirculation pump spe)ed. .

A GEllERAL $ ELECTRIC E!"TI"/E"#OO3s'm"&"E "m*ufu# m"oEs *=2ME@SETE"w'A"oOF'** " """"" ""

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SIL No. 380 R;visicn 1 As the meters (power / flow conditions hydraulicpressure

sub tandom per/reacto,r tic feedback kine'mechanismcooling,are cpower chang power tions are most and flow at the doturbations~may result id an be en. e thermal

- power tionsplant demonstrated / flow map o (likelyt to'minant freiuency of 0 3 Figure the 1).ccur.at the high pow. o0 ations s

I Previous tability tester occurrence / low.7 flow corner Hz.

ofThese rod lin(e and observed tion on natural the ngs limit cir of APRM cycle r'eco a

neutron das fluxs t s the a

r ers oscilla-individual Local LPRMs w'ere plant have oscillating culation ange Mo and es.

in phase.nitors also deonstrat were easilyflow.)

Power Rof several su LPRM(In addition,ppressed by the at eadilythe e rated co ed oscillations at natural rod line suppresse.- The oscillations the occurrence cir indicated that ~

)of limit allR limit cy atperce another t d by minimal culation.and control rod several ncycle neutron flux however,cle the oscillationswere again wouldobservable occurabove the onrated the A than thosecharacteristics previously RMs and the detailed oscillated out of test of data phase insertion.

of the observed oscillation

~1t e thatwas pr than the core average observed these most at other stabilit y tests. test recent e;

the core average .

with the A'PRM signal were Although anddifferent at control rod very insertionthe large margin etectable to s oscillations safettheshowed ,were local higher amplitudes that dsom oscillatio Four hundred twenty and easily y limits suppressed was maintained by mi i a demonstrated (including 150 years of hireactor years n mal circulation flow rate gh power density plante opof BWR these instabilities less li arethat instabilitie( \

i '~However,kely to~of power been a function occur / flow ne.in the lower poa '

above the encountered at operating rtheaddition ratioIn abovesinceove tests a unique operating statesrated th limit rod line cycle demonstratoscillatiwer BWR/2-3). dens In eactors at minimum forc ons that have these oscillationssumary, as demonstratedrculation b ur at increase if possiblecan be readilyperience suppressed byare core'averagelocal controon the exh)ibit neutron mo at BWRs.

In regions may o addition.l:fod the insertion or (nitoring system and therefore the operator core flow Because "inf ormation

'of their mit low pto bserve characteristicss cycle oscil different rom those an,dthatmitigate fmost .

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t SIL No. 380 Revision 1 RECOFMENDATIONS N General Electric recommends that BWR operators using GE BWR fuel monitor

,the inherent neutron flux signals and. avoid or control abnormal neutron flux oscillations (with particular attention to the region of sensitivity in Figure I where the probability of sustained neutron flux oscillations increases) as follows:

1. Become familiar and svare of your plants normal average power range monitor (APRM) and local power range monitor (LPRM) peak-to-peak neutron flux for all operating regions of the

. power / flow map and .for all operating modes (e.g., two loop and  ;

single loop operation). In particular establish an expected APRM and LPRM peak-to-peak signal for your plant at various i operating states and also for special operating modes (i.e..  ;

SLO) if these modes will be used. The expected APRM noise amplitude can be easily determined from past steady state l strip chart recordings or can be established based on, current operating conditions. {

2. Whenever making APRM or LPRM readings, verify that the neutron i

flux noise level is normal. If there is any abnormal increase  :

in the naturon flux response follow the recommendations in ,

Section 6d to suppress the abnormal noise signal.

I i 3. The LPRM gains should be properly calibrated'as per current plant procedures. This will permit the LPRM upscale alarm t trip setpoints to be set as high as full scale while providing I appropriate indication against unacceptable reduction in thermal margin because of power cacillations. The LFRM upscale alarm indicators should be regularly monitored and all upscale alarms should be investigated to determine the cause i

and to assure that local limits are not being exceeded.

4. Whenever changes are made or happen that cause reactor power

! to change, monitor the power change on the APRMs and locally on the LPRMs surrounding control rod movement to become j

familist with the expected neutron flux signal l

characteristics.

! 5. If a recirculation pump (s) trip event results in operation in j region 1 of Figure 2:

I i a. Immediately reduce power by inserting control rods to or below the 80% rod line'using the plant's prescribed

• control rod shutdown insertion seguance.

I 4

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' SIL No. 380 Revison 1 -4 ~

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After insgrtig control Tods, frequently monitor the APRMs and monitor. the local regions of the core by using e the control rod select switch to display the various

,LPRM strings vliich surround the selected control rod. A I

• minimum of nine control rods should be selected to adequately. display LPRMs representing each octant of the acore and the core center (Figure 3).

• bnormal increase in the espected signals, insertIf there is_any additional using controlprods to suppress the oscillations the plant's insertion sequence.rescribed control rod shutdown .

c. .

• After inserting control rods, monitor the LPRM upscale alarm indicators and verify (using recommendation 5b) that any LPRM upscale alarms which are received are not the result of neutron flux limit cycle oscillations .

d.

low to high frequency speed for flow control plants),

rod line. the operation should be performed below the 80%

a.

Once pumps have been restarted and recovery to power is to commence, follow the reconsendations in Section 6 '

6. . -

M en withdrawing Figure 2: control roda during startup in region 2 of a.

Monitor the APRMs and the LPRMs surrounding control rod movement is being reduced continually for a as power is bains increased or flow neutron fluz response.ny abnormal increase in the normal '

b.

Monitor (using recomm' thee LPRM ndation 5b upscale alara indicators and verify which limit cycle are'oscillations.

received are no)t the result of neutron fluxt c.

Operate the core 13 as symmetric a mode as possible to avoid asymmetric power distributions. When possible, quadrant mirror (sequenc'e 3) symmetric Control patte rod movement should be restricted to no more tha

~

should.be within 2 feet of each other Por ,

at all tim BWR/6 plants with ganged tod withdrawal control rods l

i '

should be symmetric rod patterns.,

moved tn ganga-as much as poss,ible to maintain -

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SIL No. 380 Revision 1 -

d.

s If there is any abnormal increase in the normally expected neutron flux response, the variations should be suppressed. It is su'gg'ested that the operation which caused the increase in. neutron flux response be reversed.

if practical, to accomplish this suppression; control rod insertion or core flow increase (PCIONR's should be followed during flow increases) will result in moving toward a region of increased stability.

e. An alternative to recommendation 6a-d is to increase core flow such.that operating region 2 of Figure 2 is avoided.

PCIOMR guidelines should still be followed.

7. When performing control rod sequence exchanges:

a. Follow recommendations 6a-d. or

b. Perform control rod sequence exchanges outside of, regions 1 and 2 of Figure 2.

8. When inserting control rods during shutdown, insert control rods.to or below the 80% rod line prior to reducing flow into region 2 of Figure 2 (i.e., avoid region 2 during shutdown).

9. Should any abnormal flux oscillations be encountered, data should be recorded on the highest speed equipment available and all available power, flow, power shape, feedwater, pressure and rod pattern information documented for subsequent evaluation and operational guidance.

1 Prepared by: G.A. Watford Aproved by:

D.L. Allred. Manager r> Issued by:

bd/

R.E. Bates. Specialist Customer Service Information Customer Communications Product

Reference:

A71 - Plant Recommendations 1

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UNITED STATES ATTACMENT I o

NUCLEAR REGULATORY COMMISSION j t t8 wasMNGTON. D. C. 20656 .  !

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Docket Nos.: 50-373 23 g ,

50-374 ,

50-397 MEMORANDUM FOR: Chairman Palladino Comissioner Gilinsky Comissioner Roberts Comissioner Asselstine Comissioner Bernthal FROM: Darrell G. Eisenhut, Director '

Di ision of Licensing

SUBJECT:

BWR CORE THERMAL HYDRAULIC STABILITY (B0ARDNOTIFICATIONNO.84-062)

In accordance with NRC procedures for Board Notifications the enclosed information is provided to the Comission. The enclosure is a description of recent thennal hydraulic stability tests at a foreign boiling water reactor. As stated in the enclosed memo current US BWR Technical Specifications place restrictions on US BWR single loop operation. In addition, the GE calculations show that considerable margin exists even with the large neutron flux oscillations. The staff is presently working with applicants and with BWR owners group in the review of BWR Technical Specifications regarding this issue.

By copy of this memorandum we are also informing all applicable BWR Atomic Safety and Licensing Boards, the appropriate BWR Atomic Safety and Licensing Appeal Boards and the corresponding parties. ' j f\ W l . ie IIector Division of Licensing Enclosure February 27,1984 Memo Mattson to Eisenhut cc: See Next Page L PP -

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cc: SECY(2) -

OPE OGC EDO Clinton ASLB (Clark, Ferguson, Paris)

Perry ASLB (Block, Bright, Kline)

Perry ASLAB (Kohl, Buck, Edles)

ShorehamASLB(Brenner,Ferguson, Morris)

ShorehamASLAB(Rosenthal,Edles,Wilber)

Zinner ASLB (Frye, Hooper. .Livingston)

Parties to the proceedings h

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UNITED STATES ,

g ; 3, g NUCLEAR REGULATORY COMMISSION n.- l WASHINGTON, D. C. 20655

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FEB 27W _

~

MEMORANDUM FOR:

Division of Licensing FROM: Roger J. Mattson, Director Division of Systems Integration

SUBJECT:

BOARD NOTIFICATION - BWR CORE THERMAL HYDRAULIC STABILITY

1. Item for Notification The staff has been informed by General Electric (GE) of recent thermal hydraulic stability tests at a foreign reactor, a boiling water reactor with relatively high power density, which demon-strated the occurrence of limit cycle neutron flux oscillations at natural circulation and several percent above the rated rori line.

It was predicted that limit cycle oscillations would occur at the operating state tested and the oscillations were observable on the APRMs and suppressed through control rod insertion. The charac-teristics of the observed oscillations, however, were different than those previously observed in other stability tests. Exami-nation of the detailed data of this test showed that some LPRMs oscillated out of phase with the APRM signal and at an amptitude as great as six times the core average measured by the APRMs.

These data are important since they confirm the possibility of local thermal-hydraulic instabilities which have been postulated but not observed in similar stability tests performed in the Unitel States.

The staff is presently working with applicants and with the BWR Owners Group to review the standard Technical Specifications to assure that they properly protect against the potential for insta-bilities. If changes are to be required, they would follow staff procedures for generic Technical Specification changes, including CRGR review.

2. Relevancy and Materiality, Stability tests on a BWR/4 reactor demonstrated that limit cycle oscillations can occur within permissabic operating space below the rated rod line at natural circulation flow. The high power level (120 percent) scram protection which is based on APRM signals would not necessarily prevent violation of critical heat flux (CHF) i limits if local instabilities occur. l i

l

Contact:

G. Schwenk, DSI:CPB X-29421 M -

N OMO U%gp j

i, ,* .

D. G. Elsenhut II N Core designs are known to decrease in thermal hydraulic' stability with higher power density and with higher gap conductance which is associated with fuel designs having smaller fuel pin diameters. ,

The ccnditions which led to local instabilities in the foreign test are notthan rather clear. global oscillations wttn all fuel assemblies in phase Until core stability behavior can be better pre-dicted and explained, the staff believes that the new infomation

' showing that local oscillations non-characteristic of the APRR signals may occur is relevant to all BWR designs. but is of greater significance to those having higher power density and lower damping characteristics as typified by the BWR/4. 8WR/5, and BWR/6 de:fgns.

3. Significance of Test Infomation The test demonstra'ted that local themal hydraulic oscillations which are out of phase with the APRMs can occur. This raises the possibility of local oscillations occurring in the plant which might not be detected by the operator without monitoring of LPM signals and/or LPRM upscale alarms.

It is unclear at this time how -

high a local oscillation could grow .before detection by an operator using current monitoring procedures. Since we cannot predict the limiting magnitude for such an oscillation, we can not be certain that CHF limits would not be exceeded. It is for this reason that we will examine BWR Technical Specifications to assure that they provide adequate assurance that operating regfons of potential instability are avoided and that neutron oscillations of unac-captable magnitude are detected and suppressed.

The staff concludes that the new infomation from the foreign stability tests does not pose an immediate safety concern for continued BWR operation prior to orderly examination and possible change of Technical Specifications for the reasons which follow:

(a) Current BWR Technical Specifications place restrictions on operation under conditions of natural circulation or single loop operation such that the frequency of operation in regions i with low stability margin is very low. In addition, the core l designs for most operating reactors are sufficiently stable that in the limit less cyclestable operations pemissible areobelieved to be unlikely even  ;

t such oscillations should occur perating regimes. Even if }

power level APRM scram protection would be adequate in mostit is belie!

cases.  !

(b) The magnitude of themal hydraulic instability induced neutron flux oscillations is considerably higher than the oscillations ,

t in the average cladding heat flux because of delays caused by '

the fuel themal time constant. General Electric Company l

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D. G. Eisenhut g7g calculatfor,s for the case Nhere tiee neutron flux oscillates up to 120 percent of rated (RpS trip point) indicate that the surface heat flux has a peak oscillation amplitude of only .

5 percent of its rated value. Thus, a considerable margin exists to safety limits even if very large neutron flux oscillations occur.

(c) The oscillations are readily detectable with proper reonitoring and can te easily suppressed by inserting control reds.

(d) General Electric Company is in the process of providing to all its operating plants guidance on the prop.er raethods to mo.7ttor for themal hydraulic instabilities and on the actions that should be taken to suppress such oscillations if they sheuld cccur. . BWR owners have been made aware of the problem and of pending actions.

4 Relation to Project 1 The foreign stability test results relate to all BWR reactors. It is recormended that appropriate boards be notified of the new information and of the staff's plans to work with the BWR applf-cants and licensees to assure that the Technical Specifications for all EWRs properly protect against the potential or instabilities.

f gh

• 4:04 Hoger J. P ttson, Director Division of Systems Integration o

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