ML20245L290

Safety Evaluation Supporting Amend 121 to License DPR-40

ML20245L290
Person / Time
Site:	Fort Calhoun
Issue date:	04/26/1989
From:	Office of Nuclear Reactor Regulation
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ML20245L270	List: ML20245L267 ML20245L283 ML20245L290
References
NUDOCS 8905050368
Download: ML20245L290 (4)

Text

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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO.121 TO FACILITY OPERATING LICENSE N0. DFR-40 OMAHA PUBLIC POWER DISTRICT FORT CALHOUN STATION, UNIT NO. I DOCKET NO. 50-285 .

1.0 INTRODUCTION

By letter dcted January 6,1989, Omaha Public Power District (OPPD)

, submitted an application for an arer.dnent to Facility Operating Licer.se ho. DPR-40 tbt would modify the Fort Calhoun Station, Unit No.1, Techr.ical Specifications (TS) to (1) change the certainrent spray system surveillance .

testing requirements to provide a quantitative value tc define the minimum acceptance criteria, (2) change the Basis of the containment spray system surveillance requiremer.ts b determined from analysis, 3(y)providing reduce the the minimum maxirum spray power levelflow requirements permitted on Ccndition for Operation for Departure from Nucleate Figure 2-7, Limiting (4) revise Figure 2-3, Predicted Radiation Induced Boiling Monitering, NDTT Shift, based on calculaticos using US NRC Regulatory Cuide 1.99, Revision 2, (5) correct the neutron fluence value stated as cccurring at 14 Effective Full Power Years (EFPY) at the inner surfag cf the reactor vesselwayatthecriticalweldlocationfrom1.4x10 n/ car to 1.21 x 10 n/cm2, and (6) change the references ir TS 3.6 from "FSAR" to "USAR" and adding an aoditional reference to USAR Section 14.16.

. 2.0 DISCUSSION 2.1 Containment Spray System Surveillance Requirencnts (Items I and 2)

The Fort Calhoun Station Containment Spray (CS) System censists of three cor.tainment spray pumps which supply flow, via a corrnon header, to two independently isclable spray headers. Each spray header centains 274 nozzles. The present Techr.ical Specification surveillance requirement fnr the spray system states, in part, that the system test will be considered satisfactory if visual observations indicate all components have operated sa tisfa ctory. However,10 rczzles on each spray header are blocked by ventilation ductwork and piping, and one nczzle on one of the headers is missing, thus affecting the operability of these nozzles.

The licensee provided an analysis, " Reexamination of Containment Pressure  ;

Response for the DBA LOCA and DBA MSLB Events at Fort Calhoun Station",

dated December 1988, which was conducted to ceternine the response during the postulated Loss of Ccolant Accident (LOCA) and Main Steam Line Break (MSLB) events, using the revised information en the spray nozzles.

Additionally, the analysis included the revised containment spray pumps start logic which delays thc tire that water is discharged from the spray 8905050368 DR s90426 p ADOCK 05000ggy PNL)

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r headers. The pumps now start on the same signal which opens the header isolation valves (containment spray actuation signal) rather than on the safety injection actuation signal (SAIS).

The pressure response analysis was conducted using the CONTRANS 1 containment code, which was the same as used during the previous analysis in USAR Section 14.16.5, and with the current plant configuration. The single header atomized spray flow was reduced from 3400 gpm to 3155 gpm to account for the blocked and missing nozzles. The computer run conducted with a containment spray delay ' time and fan cooler actuation time of 1200 and 60 seconds, respectively, produced a peak containment pressure of 56.3 psig at 60.3 seconds into the event. The results of this run provided the basis for the LOCA case ualysis.  !

For the MSLB case analysis, the present analysis used the same computer code, the combined SGNIII/CONTRANS code, as utilized in previous analysis.

The analysis produced a peak containment pressure of 58.7 psig at 71.7 seconds for the benchmark MSLB response. As noted, the peak occurs prior to any actuation of the sprays. ,

The summary of the LOCA analysis results show that peak containment pressure is reached at the time of activation of the containment fan coolers. For purposes of single failure, only one containment fan cooler and filtering unit and two cooling units were assumed to operate. The analysis also showed that, if no active heat removal was available, the containment design pressure of 60 psig would be reached at 176 seconds.

With the spray pumps now starting on the CSAS, spray flow from the nozzles would occur after approximately 90 seconds, due to the delays in time for the pumps to come up to speed and to fill the lines and headers after the CSAS. The analysis also shows that peak containment pressure occurs nearly concurrent with the initiation of active heat removal from the sprays; therefore, the containment design pressure would not be exceeded. Thus, the change to the containment spray pump start logic is adequate since spray initiation remains at less than 176 seconds and containment spray is redundant to the fan coolers.

For the HSLB cases, the steam generator blowdown energy rates to containment are generally greater than the removal rates from the combined fan capacity end containment wall. Therefore, peak containment pressure occurs after fan cooler actuation and when the generator reaches dryout conditions. The previous MSLB analysis assumed spray delivery at 55 seconds which strongly influenced peak pressure due to the effect of the sprays on the superheated containment atmosphere. However, even without fans and a delay of spray actuation, the peak pressure is 59.3 psig and thus, neither are needed to control peak pressure.

The staff finds the analysis assumptions, input conditions, and computer code utilized to be satisfactory for both the LOCA and MSLB containment response analyses. Also, the analysis results appear to be in agreement with the conditions. Thus, the staff finds the proposed changes to the number of fully operable spray nozzles and the reduced single spray '2ader atomized flow rete are acceptable.

t 2.2 Faximum Power Level Permitted by LCO for DNB (Item 3)

The Limiting)

Boiling Condition for (DNB Monitoring, Operation Technical (LCO) forFigure Specification Departure from Nucleate 2-7, provides the core power level limitation versus the Axial Shape Index, Y This is one j ofseveralparameterswhicharemaintainedtoensurethattbe. fuel design ,

limits will not be exceeded during a design basis anticipated operational 4 occurrence and the consequences of a DBA will be no more severe than j predicted. The present figure 2-7 defines a core power limit of 100.5% of I rated thermal power for Y between a value of -0.057 and 0.098. However, the USAR safety analysis dere performed with an input assumption of reactor pcwer at 102% (e.g.100% plus 2% uncertainty). Thus, plant 4 operation up to a limit of 100.5%, assuming the uncertainty, may cause these analysis to not be valid. Additionally, the license conditions do not allow steady state power levels above 1500 MWt, which is 100% of rated thermal power. Since this proposed change is a further restriction to ersure that the authorized power level is not exceeded and no safety analysis are affected, the staff finds the change to be acceptable.

2.3 Revice Predicted Radiation Induced NDTT Shift (Item 4)

Regulatory Guide 1.99, Revision 2, provided equations for predicting the shift in nil ductility transition temperature, RT uc to neutron irradiation at the reactor vessel inner surface akT,he t 1/4t depth from the inner surface. The present Technical Specification Figure 2-3 provides curves of this teraperature shift versus the irradiation level based on draft Revision 2 cf the regulatory guide. However, Revision 2 issued a different through wall attenuation equation than that in the draft. This has required a change in the Figure 2-3 curves to correspond with the new equation. The value of the RT shift is used in the adjustmentoftheheatupandcccidowncurvehIothatsufficientmarginis maintained. Since the value of the predicted RT shift used in the generation of the heatup and cooldown curves wasNre conservative than '

that in the proposed change to Figure 2-3, no further correction of these curves was necessary. Therefore, the staff finds this change to the Figure 2-3 to be acceptable.

2.4 Administrative Changes

a. Neutronfluence(Item 5) l The Basis for Technical Specification 2.1.2, Heatup and Cooldown Rate, states that the predicted neutron fluence at the reactor vesselinnersurfaceforthecriticgbeltlineweldat14 Effective Full Power Years (EFPY) is 2.4 x 10 n/cm2 During a previous amendment, No. 114, which changed the heatup and cooldown curves to

correspond to operation through 14 EFPY rather than 15 EFPY, the j statedvaluecfneutronflutrcewasnotchanged,yjheproposedchange i to the TS provides the correct value of 1.21 x 10 n/cmr . This l change is administrative in nature since it causes no impact on any l

analysis and the heatup and cooldown curves had been previously changed for operation through 14 EFPY. Therefore, the staff finds the proposed change to be acceptable.

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b. ReferencestoUSAR(Item 6)  ;

The licensee has proposed changes which provide correct references to the Updated Safety Analysis Report (USAR) rather than the erroneous Final Safety Analysis Report (FSAR) references set forth in Technical Specifications (TS)3.6. The change also adds an additional reference for USAR Section 14.16 to this TS as a clarification. The staff finds the proposed changes and clarifications to be administrative in nature and to correct the, reference information errors. Thus, the proposed changes are acceptable.

3.0 ENVIRONMENTAL CONSIDERATION

S This amendment involves'a change in the installation or use of a facility,.

component. located within the restricted area defined in.10 CFR Part 20 and-changes in surveillance requirements. The staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Comission has previously' >ublished a proposed finding that the amendment involves no significant lazards consideration and there has been no public coment on such finding. Accordingly, the amendment meets the elig(ibility forthin10CFR51.22(c)9). Pursuantcriteria for categorical to 10 CFR exclusion 51.22(b), no environmentalset impact statement or environmental assessment need be prepared in connection with the issuance of the amendrent.

4.0 CONCLUSION

The NRC staff has. concluded, based on the consideration 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 compliance with the Comission's regulations, and the issuance of the amendment will not be inimical to the comon defense and security or to the health and safety of the public.

Date: April 26, 1989 Principal contributor: P. Milano l

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