ML20206L530

Safety Evaluation Supporting Amend 31 to License NPF-47

ML20206L530
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Site:	River Bend
Issue date:	11/22/1988
From:	Office of Nuclear Reactor Regulation
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NUDOCS 8811300046
Download: ML20206L530 (9)

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SAFETY EVALUATION BY THE OFFICE 0F NUCLEAR REACTOR REGULATION RELATED TO AMENbrENT NO. 31 TO FACILITY OPERATING LICERSE NO. NPF-47 GULF STATES UTILITIES COMPANY RIVERBENDSTATION,UNITJ DOCKET NO. 50 158 ENCLOSURE 1

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1.0 JERODUCTION By letter dated April 6 1988, as supplemented October 20, 1988, Gulf States Utilities Corrpany (CSU),(the licensee) requested an amendment to Facility Operating License No. IlPF-47 for the River Bend Station, Unit 1.

The proposed arendment would modify the Technical Specifications to allow single recirculation loop operation.

Sirgle recirculation loop operation (SLO) at reduced power is highly desirabic when one loop becomes inoperative during rnaintenance or testing activities.

This evaluation provides the results of the HRC staff's review of the licensee's evaluation of accidents and abnorral uperational transients with only or.e recirculation purp operative. This evaluation is performed for a PSX8R fueled core on an equilibrium cycle basis up to a rayirur pou r of approximately 70%

of rated. The analysis and evaluation are applicable to both the initial fuel cycle and reload cycles.

Reference I also addresses Ttchrical Specification changes related to Thermal-Hydraulic Stability considerations during SLO. The staff has reviewed the proposed changes ard included an evaluatio'n in Section 2 of this Safety Evaluation.

This. evaluation also addresses the proposed recirculation flow and differential temperature limits to avoid thertal stratification that could result in unaccep-table thermal stress levels in the bottom head region during SLO.

2.0 EVALUATION The licensee provided a General Electric (GE) report entitled "Single Loop Operation Analysis for River Bend Staticn, Unit 1" (Ref. 2). The GE report evaluated the SLO safety issues pertaining to the River Bend Station to justify extended operation with one recirculation loop out of service. The staff evaluation of the SLO safety issues and the proposed Technical Specification changes follows, i

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2.1 HCPR FUEL CLADDING INTEGRITY SAFETY LIMIT The net effect of increased uncertainties in the core total flow and TraversingIn-CoreProbe(TIP)readingsforthesingleloopoperationisa 0.01 incremental increase in the minimum critical power ratio (MCPR) fuel cladding intejrity safety limit. Operating with one recirculation loop results in a r.aximum power output which is about 30% below that which is attainable for two-pump operation. Therefore, consequences of abnormal operation transients from one-loop operation will be less severe than those from a full power two-loop operational rode as provided in the River Bend Updated Safety Analysis Report (USAR).

The transient peak value results and Critical Power Ratio (CPR) results for theLoadRejectionwithBypassfailure(LREPF)andFeedwaterControllerFailure (FMCF) with maximum demand are surt.arized in Table 1.

TABLE 1

SUMMARY

OF TRANSIENT PEAK VALUE AND CPR RESULTS LRBPF FWCF Initial Power /Flcw (% Rated) 70.?/53.6 70.2/53.6 Peak Neutron Flux (i t'BR) 70.3 84.4 Peak Heat Flux (% Initial) 100.3 107.4 Peak Dome Pressure (psig) 1169 1153 Peak Vessel Bottom Pressure (psig) 118?

1165 Required Two Lcep Initial PCPR Operating Li:rit at SLO Condition 1.39 1.39 delte-CPR 0.05 0.1?

Transient l'CPR 1.34 1.27 SLl'CPR at SLO 1.07 1.07 This table shows that for the limitirig transient events atelyzed here, the MCPRs are all above the sirgle-loop operation safety limit value of 1.07 so that there will be r.c fuel failure due to boiling transition. The peak vessel pressures are all below the AS!'E code value of 1375 psig. Therefore, the pressure barrier integrity is naintained under single-loop operation l

conditions. The staff finds this acceptable.

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2.2 MCPR Opera _thS,pjn,it 2.2.1 Accidents (Other Than LOCA) and Transients Affected by One

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f.k.C.ilcuj,a,t,ip,n_ L8Mt pJ15'e'rFice One Purp Se,irure Accid,en,t e

A plant specific analysis was not perforned for this event. Previous analyses for tha Grand Gulf plant has shcwn that the event usults in a PCPR value significantly above the SLO safety limit PCPR. This has been confirmed for other BWRs (Refs. 3 and 4).

2.2.2 Abnormal Operating Transients Although the increased uncertainties in core total flow and TIP readings resulted in a 0.01 increase in MCPR fuel cladding inter.ity safety limit during single-loop operation, the limiting transients analyzed in the GE report indicate that there is more than enough MCPR rnargin during single-loop operation to compensate for this increase % safety liinit. For single loop operation at off-rated conditions, the stJiJy state operating MCPR limit is established by the power dependent and flow dependent MCPR curves. For the

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nost limiting transient events analyzed, the GE report also shows that the t

present power dependent MCPR limits are bounding for single-loop operation.

Further, the present fle N '. dent FCPR limits are also bou'iding for single-l loop operation since the iaaxiu m core flow runout during single loop operation is only about 54% of rated, iie transient consequence from one-loop operatico is therefore bounded by prev iMiy submitted full power analyses. This is acceptable.

2.2.3 Pnd Withdrawal Er.or The rod withdrawal e ror at rand pwer is given in the USAR for the initial core and in cycle-d> pendent re%ed supplen. ental submittals. These analyses were perforred tc deroonstrate tr.at, even if the operator had igt.ored all instrunent ir.dications ard alarrts during the course of the transient, the red bicck system would stop rod withdrawal at a ninimum critical pcwer ratio which is higher than tFr fuel cladding integrity safety linit. The GE report also shows that correction cf the rod block equation for sirigle-loop operation assures that the VCPR safety limit is not violated.

One-purp operath.r. ruults in backflow through 10 of the 20 jet purps while flow is being supplied to the lower plenuni from the 10 active jet punps.

Because of this backflow through the ir. active jet purcps, the present rod block ec,vetion and ApKM settings were trodified for use during one-purp operatior. The staff has found thera acceptable.

The staff finds that one-loop transients and accidents other than LCCA, whict is discussed below, are bounded by the tuc-loop operation er alyses and are, therefore, acceptable.

2.3 Stability Analysjs With one recirculation loop not in service, the prinary contributir.g factors to the stability perforcance are the power / flow ratio and the recirculation 1 cup characteristics. At forced circulation with one recirculation loop not in operation, the reactor core stability is influenced by the inactive recirculation loop. Staff evaluations have considered whether increased noise in SLO was being caused by reduced stability margin as SLO core flow was increased. Results of analyses and test indicates that the SLO stability characteristics are not significar.tly different from two-loop operation. At low core flows, SLO ray be slightly it.ss stable than two-loop operatier, but as cort ficw is increased and reverse flow is established, the stability I

4 perforniance is similar. At higher core flows with substantial reverse flow in the inactive recirculation loop, the effect of cross flow on the flow noise results in an increase in systen. noise (jet pump, core flow and neutron flux noise), but core thernal-hydraulic stability inargin is very high, similar to two-loop operation. GE has developed a Service Infornation Letter-380, Revision 1 (Reference 5) informing plant operators how to recognize and suppress unanticipated oscillations when er. countered during plant operation.

i The NRC has approved the recorcendation of SIL-380 for incorporation into BWll Plant Technical Specifications. The licensee has incorporated the surveillance requirements recosteended by SIL-380 into the River Bend Technical Specifications and has proposed irodificatiers applicable to the SLO inode. The staff finds this acceptuble.

In a related matter, the NPC has identified generic safety icplications regarding power oscillations in Eoiling Water Reactors and has recently issued an fRC Bulletin No. 88-07 (Ref. 7) dealing with this subject. The licensee fcr River Bend has responded to the Eulletir. by Reference 8 and has identified a revisien to a statiori Abnormal Operatir.g Precedure (ACP) in accition to confirr.ation of the bulletin actico iters. The NRC herein acknowledges the licensee's response and notes that the AOP revision will be reviewed urder an NRC Regional Office inspection it. accordance with a Tenporary Instructico procedure.

T.4 Loss of Coolant Accident Analy,sjs SAFE /REFLOOD calculations were perforr.ed for a full spectruri of large break sizes for the recirculation suction line breaks for the single-loop operation roede. The small differences in ur'covery time and reflood tine for the lir.11 ting break size, i.e.,183 seconds for the single-loop vs.184 seconds for the two-locp cperatier., wculd result ir a sitall change in the calculated peak The rayiruri average plarar linear heat generatien rate cladding)ter perature. reduction factor for the raost liinitir.g single-loop operatiori fcr (MAPLHGR P8XER fuel is 0.84 which is conservative.

i In the event of a snall break LOCA, the slight increase (50*F) in peak clad terperature (PCT) is offset by the effect of the decreased MAPLHGR (equivalent to 300'F to 500'F PCT) for the single-loop operation. The calculated PCT values for small breaks will therefore be well below the 1547*F PCT value previously analyzed for small breaks. The LOCA analyses applicable to the I

River Bend SLO rode have been perforrted using rethodology apprrved by the staff l

(Ref. 9) and the results are acceptable.

I 2.5,Contai,nn,ep,t, Analysis The GE analysis indicates that under SLO conditions limiting case accidents i

would result in peak containment pressures, containrent terperatures, and suppression pool teraperatures which are less severe than those estiinated for i

design basis accidents under two-loop operation. GE also evaluated the chugging, condensation oscillation and pool swell loads under SLO conditions l

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5-and stateJ that these loads slightly exceed those estimated for accidents during tvo-1cep operation.

In response to a staff request, the licensee providec additional discussion and quantification of the containrrent loads for SLO conditions. The information provided in Reference 10 verified that the evaluatico specified in Appendix 6A of the River Bend Updated Safety Aralysis Report (UPAR) is applicable to SLO at the limiting operating point. The analysis was made using staff-approved inethodology and a model based on the MARK 111 containrent test program. The margins for pool swell loads and condensation loads were a few percent greater than two-loop limiting conditions and well within the design values. Chugging loads do not increase for SLO cotiditions. The staff finds this acceptable.

2.6 Hiscellaneous_I,rnpact Evaluation Antj,cjpated Transient Without Scram Since the SLO initial power / flow condition is less than the rated condition used for the two-loop ATWS analysis, GE fcur.d the transient response it.ss severe and therefore bour.ded by the FSAR analyses. This is acceptabic.

f.u,e,1,f e cha nj,c a l, Pgfortna n ce Due to the substantial reverse flow established during SLO, both the Averagc Power Rarige Tonitor (APRM) noise and core plate differer.tial pressure rioise are increased slightly. GE has stated that the APRP fluctuation should not exceed the fuel rod and asserbly desigr. bases.

This is acceptable.

Vessel Irternal Vibration OE irposed a recirceletien purrp drive flev 14rrit for single-loop operatior7, which is about 33,000 gpri for rated reactor water terperature and pressure.

This is based on reasured prototypical value fror, the Kuo Sheng 1 plarit which has been accepted by the staff as the valid prototype for River F e r.d. Ilith raaxirum flow thus lin.ited, vibration levels of the reactor iriternal cor ponents will be within acceptar.ce limits during SLO at River Berd Statior.. This is acceptable.

[e,tpgp_Opertbility Jet purrp surve113crce is only required for the operating lotp. The licenset has proposed nodifications to the River Eerd TS to accoprodate the SLO mode. These changes are acceptable tc the staff.

2.7 Rgn,al_StressLimits The licensee considered the possibility that thermal stratification nay occur in the botton head of the reactor pressure vessel during single loop operation.

Therrial stratification r:ay occur if a stagriant layer of cold water forrs near the bottorr tead.

If the water suddtnly r:ixes with warn water such thet the terperaturt iri the bcttori head suddenly incrtasts, then penetratior.1 ir the botter head r ay opand at a rate differer.t fror the botton. head. This rnay result in the forr.stiori of crar.Ls at the perietrations.

6-To avoid single loop operation at low power or.ow flow conditions that may allow thermal stratification to occur, the licensee has proposed limits of operation at greater than 30% rat.ed thermal power and greater than 50% rated recirculation loop flow in the operating loop in order to increcse power or flow. The licensee has stated that operation in the region of the power-flow map above these limits would not lead to thermal stratification. However, operation at or below the 30% rated thermal power or at or below the 50% rated recirculation loop flow would be permitted if the following differential temperature requirements are met within 15 minutes prior to an increase in thermal power or increase in recirculation loop flow:

a.

Less than or equal to 100'F between the reactor vessel steam space coolant and bottom head drain line coolant; and b.

Less than or equal to 50'F between the reactor coolant within the loop not in operation and the coolant in the reactor pressure vessel (not applicable if the loop is isolated); and c.

Less than or equal to 50*F between the reactor coolant within the loop not in operation and the operating loop-(not applicable if the loop is isolated).

The licensee has proposed that these limits be incorporated in Technical Specification 3.4.1.1, action statement f, Surveillance Requirement 4.4.1.1.4, and Bases 3/4.4.1.

Current Technical Specification 3.4.1.4 contains similar temperature differential restrictions to prevent undue stress on the reactor vessel with regard to idle recirculation loop startup. This technical specification states that an idle recirculation loop shall r.ot be started unless the temperature differential between the reactor pressure vessel steam space and the botton head drain line coolant is less than or equal to 100'F, and:

a.

When both loops have been idle, unless the temp *rature differential between the reactor coolant within the idle loop to be started up and the coolant in the reactor pressure vessel is less than or equal to 50'F, or b.

When only one loop has been idle, unless the temperature differential between the reactor coolant within the idle and operating recirculation loops is less than or equal to 50*F and the operating loop flow rate is less than or equal to 50% rated loop flow.

l Based on its review, it is the staff's judgement that for greater than 50%

rated recirculation loop flow, and greater than 301 rated thermal power, there 4

l will be adequate circulation of reactor coolant during singleloop operation to 1

assure that there will not be thermal stratification that could lead to unaccep-l table stresses in the bottom head of the pressure vessel. The staff also finds that the proposed differential terrperature limits for operation at or below E0%

rated recirculation loop flow, or at or below 307 rated thernal power are l

consistent with the current Technical Specificaticn 3.4.1.4 i

. The staff concludes that proposed Technical Specifications 3.4.1.1, action f, Surveillance Requirement 4.4.1.4, and Bases 3/4.4.1 are acceptable.

3.0 TECHNICAL SPECIFICATION CHANGES The licensee has proposed to change TS Limiting conditions of Operation, Bases Sections and the corresponding descriptive sections thereto so as to be in conformance with the GE analyses to implement SLO. The staff has reviewed the changes and finds them consistent with results of the GE analysis and also with TS changes approved for other BWR/6 facilities for single-loop operation. The staff concludes that these TS changes are acceptable.

3.1 Specification 2.1.2,page,2-1 a

3 The safety limit Minimum Critical Power Ratio will be increased by 0.01 to 1.08 for single-loop operation. This number is to account for core flow and T!P reading uncertainties, which are used in the statistical analysis of the safety limit.

3.2 Tabie 2.2.2-), page,2,4 The APRM Reactor Protection Systen Instrumentation Trip Setpoints are modified a

to account for backflow through half the jet pumps. The setpoint equations will be changed in the RBS Technical Specifications. The changes are siciilar to other plant TS and are acceptable to the staff.

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3.3 Specification 3/4.2.1,_page,3/4,2:1 The '.imiting Condition for Operation was changed to reflect the 84 percent i

reduction in APLHGR values for single-loop operation. The number is derived from LOCA analyses initiated from single-loop operation as discussed in Section 2.4 of this Safety Evaluation.

3.4 Spe,c,i,fj,ca,tjen,3/,4.2.2,,page 3/4 2-7 The setpoint equations for the APRM setpoint changes in Table 2.2.1-1 are identified.

3.5 ' Table 3.3.6_-j g age,3/4,3-62 j

Control Rod Block Instrumentation Setpoints will be modified to account for back flow through the inactive jet purps. These changes are similar to previously approved SLO Technical Specification changes on other plants and are acceptable to the staff.

j 3.6 Specification,3/4.4J,, p,ag e s,3/4,4 -1, t h,roug h,3/4,4 -5 The Technical Specifications related to the Recirculation Loops are snudified to reflect single-loop operation censiderations discussed in this Safety Evaluation. This includes replacerent of Figure 3.4.1.1-1 to identify the detect and suppress regions of the power-flow map associated with l

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thertal-hydraulic stebility.

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3.7 Bases Section Changes The Bases Section changes related to the proposed SLO mode and identified in the licensee's subr.ittal were reviewed by the staff for consistency with the changes discussed above. The staff finds the bases discussions accurately reflect the bases for the changes and are acceptable as proposed.

4.0 ENVIRONMENTAL CONSIDERATION

Pursuant to 10 CFR 51.21, 51.32 and 51.35, an environmental assessment'and finding of no significant impact was published in the Federal Register on November 17, 1988 (53 FR 46516).

Accordingly, based upon the environmental assessment, the Commission has determined that issuance of this amendaent will not have a significant effect on the quality of the hunian environe.ent.

5.0 CONCLUSION

The staff has concluded, based on the considerations discussed above, that:

(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (2) such activities will be conducted in cotpliance with the Corrission's regulations, and the issuance of the amerdrent will not be inimical to t.o cotton defense and security or to the health and safety of the public. The staff therefere concludes that the proposed changes are acceptable and they are hereby incorporated into the River Bend L' nit 1 Technical Specifications.

Dated: November 22, 1988 Principal Contributors:

M. McCoy, W. Paulson

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i 5.0 MFERENCES 1.

Letter,J.C.Deddens(GSU)toDocumentControlBranch(NRC),

"Application for Amendment to Operating License No. NPF-47," April 6, 1988.

2.

"Single Loop Operation Analysis for River Bend Station, Unit 1,"

NEDO-31441(DRFNo.A00-02463), General Electric Company, May 1987.

3.

Letter, D. M. Musolf (tiSP) to Director of ONRR (NRC), "Request for i

Amendment to Operating License Nc. DPR-22 " March 24,1986.

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4 T. L. Riley (Clinton Power Station), "Pump Seizure During Single Loop Operation,"

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L3086(12-15)-6, December 15, 1986.

5.

"BWR Core Thermal Hydraulic Stability," General Electric Corpany,

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February 10, 1984 (Service Information Letter-380, Revision 1).

l 6.

Letter, C. O. Thoras (faC) to H. C. Pfefferlen (GE), "Acceptance for Referencing of Licensing Topical Report NEDE-24011, Rev. 6, Amendment 8 Thermal Hydraulic Stability Amendment to GESTAR 11," April 24,1985.

7.

NRC Eu11etin Nc. 88-07: Power Oscillations in Soiling K'ater Reactors (BWRs), June 15, 1988.

i 8.

Letter, J. E. Bocker (GSU) to Document Control Desk (NRC), Response to IIRC Bulletin 88-07, September 8, 1988.

9.

Letter, H.N. Berkow (NRC) to J.F. Quirk (GE), Safety Evaluation of General Electric ECCS Evaluation ilethodology for Single Loop Operation, dated Parch 5, 1986.

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10.

Letter (REG-29C71), J. E. Booker (05U) to Docurent Control Dest (NRC)

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dated October 20, 1988.

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