ML19242C976

Summary of 790327 Meeting W/Util in Bethesda,Md Re Stretch Power Primary Pressure Increase & Conversion to Exxon Fuel

ML19242C976
Person / Time
Site:	Fort Calhoun
Issue date:	06/28/1979
From:	Conner E Office of Nuclear Reactor Regulation
To:	Office of Nuclear Reactor Regulation
References
NUDOCS 7908140294
Download: ML19242C976 (14)

Text

9 MEETING

SUMMARY

DISTRIBUTION ORB #4 Mr. Theodore E. Short Assistant General Manager Omaha Public Power District 1623 Harney Street Omaha, Nebraska 68102 Do:ket File R. Reid A.u em V. Noonan L PDR P. Check ORB #4 Rdg G. Lainas NRR Rdg G. Knighton H. Denton Project Manager E. G. Case OELD V. Stello 0I&E (3)

D. Eisenhut R. Ingram R. Vollmer R. Fraley, ACRS (16)

W. Russell Program Support Branch B. Grimes TERA T. J. Carter J. R. Buchanan A. Schwencer Meeting Summary File D. Ziemann NRC Participants T. Ipcolito W. Gammill K. Mcge P. Kapo J. Burdain M. Chatterton E. Adensam S. Weiss 65S 23

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UNITED STATES i

NUCLEAR REGUL ATORY COMMISSION 3.S" d "m C

WASHINGTON, D. C. 20555

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June 28, 1979 Docaet No. 50-285 LICENSEE: GMAHA PUBLIC POWER DISTRICT (OPPD)

FACILITY:

FORT CALHOUN STATION

SUMMARY

OF MEETING HELD ON MARCH 27, 1979 TO DISCUSS THE STRETCH POWER PRIMARY PRESSJRE INCREASE AND CONVERSION TO E,YYON FUEL FOR FORT CALHOUN The meeting was held at the Commission's offices in Bethe?ca, Maryland. A list of attendees is given in Enclosure 1.

Introfuction In cas communications with OPPD, the NRC was notified of their intension to request an amendment for Cycle 6 operation of Ft. Calhoun at the stretch power level of 1500 MWt and an increased primary cressure of 2250 psi.

OPPD hac reouested this meeting to discuss the staff review and present the anticioated schedule. The OPPD meetina agenda is shown in Enclosure 2.

Summarv After ocer.ing remarks by the NRC (Conner), OPPD (Morris) covered agenda Item I (En:losure 2). OPPD will submit an application for power level increase from 1420 to 1500 MWt and RCS pressure increase from 2100 to 2250 psi.

Part of the analyses will be for 1560 MWt, the expected mar.imum achievable ocwer level.

The FSAR oresents analyses of equipment at 1500 MWt and core parameters at 1420 MWt. OPPD plans to return to operation after the Cycle 6 refueling outage. in wnicn, Exxon Nuclear Company (ENC) fuel is leaded for the first time, at a ocwer level of 1500 MWt and RCS pressure of 2250 osi. The startup for Cycle 6 is planned for March 15. 1980.

In the presentation of agenda Item II, OPPO (Gasper) stated that Ft. Calhoun operatina at 1560 MWt, 2250 osi and 545 F would have equal core parameters to a Calvert Cliffs tyoe reactor operating at 27C0 MWt, (stretch power rating).

Ccmoustion Engineering (CE) has perfor ned a feasibility study and OPPD has evaluated the olant eouipment for stretch cower coeration.

Important findings were:

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turbine generator plus support equipment (heaters) can handle 1560 MWt, 2.

main condenser flow and delta temperature limit (ETS) are adequate for 1500 MWt, but must increase either value for 1560 MWt operation, 3.

1500 MWt may be obtained without a RCS pressure increase, but 2250 psi would be necessary for 1560 MWt operation, 4.

no changes in plant instrumentation or equipment appears necessary to accomplish 1560 MWt operation.

Much of the analyses performed by CE will support operation at the 1560 MWt power level.

OPPD plans to call for a two step approach; first, power increase to 1500 MWt with RCS pressure of 2250 psi, and then a later application to increase power level to 1560 MWt.

Item III of t ie aganda was presented by CE (Schaebrean).

Important points made on the planned review are as fallows:

A. - Ft. Calhoun steam generator design pressure is 2500 psi, operational limit at 2100 psi was a result of special requirements at the time of licensing. The RCS fatigue analysis will be recqalyzed using 14,000 instead of 15,000 stress cycles for 2250 psi operation.

B. - Containment has been analyzed for 1500 MWt operation.

C. - The range of the RCS temperature instrumentation will be reviewed. No changes are expected to be needed in the safety instrumentation.

D. - CE plans to revise the analysis for overpressure protection at the higher power level and RCS pressure.

E. - The f si clad collapse analysis will be redone.

F. - The technical specification limits f;r the primary components will be reviewed.

G. - A performance evaluation of the auxiliary systems will be completed.

H. - The stresses for the core shroud bolts and other reactor vessel internals will be analyzed.

In the opening remarks for agenda Item I'/, ENC (Owsley) stated that many of the ENC references presently docketed at the NRC will be referenced at the Ft. Calhoun fuel design. Enclocure 3 shows the slides used in the presentation.

Highlignts of the planned review are as follows.

A. - The ENC fuel to be used at Ft. Calhoun will be acceptable for an exoosure of 36,000 MWD /MT, a value 10 percent higher tnan other ENC reload fuel, hRVcv) ms;

??G B. - The themal-hydraulic analysis will be perfomed using a staff approved generic methodology.

C. - The neutronics analysis will utilize a staff approved methodology.

ENC will perform setpoint analysis. CE will continue to perfom INCA. OPPD has obtained adequate information from CE to perform PDQ and related core analyses.

D. - ENC will use a staff approved plant transient analysis looking at the set of most limiting transients. The NRC (Weiss) pointed out that a new standard listing of A00 and accidents should be con-sidered in determining which plant transients should be analyzed.

E. - ENC (0wsley) pointed out that a staff approved methodology will be used for the large break LOCA analysis. The staff (Conner) pointed out that no small break LOCA analysis was listed on this slide.

The staff made it clear that the small break LOCA must be analyzed for Ft. Calhoun.

Agenda Item V was presented by OPPD (Jaworski).

Infomation transfer between CE and ENC has been adequate. Review of the Ft. Calhoun technical specifications, plant procedures and plant system descriptions will be perfomed by OPPD., OPPD;s listing of transient analysis for stretch power was distributed. The staff (Adensam) questioned whether the radio-logical consequences of accidents would be reanalyzed. Also dis-cussed were the need to review changes in the site (population, etc.),

centrol room dose to operators after LOCA, new meteorological data, and leakage limit for ESF systems located outside the containment.

OPPD indicated that population and meteorological data had been docketed for the proposed Unit No. 2.

The staff (Conner) mentioned the possibility that infomation necessary to supplement the Final Environ-mental Statemenc by issuance of an Environmental Impact Appraisal (as was done for Calvert Cliffs) may be required if the stretch power level had not previously been addressed.

OPPD said that one CE test fuel assembly would reach an exposure of 46,000 MWD /MT during Cycle 6 if a contractural agreement was reached with DOE. This fuel assembly would be discharged at the end of Cycle 6 and some of the fuel pins shipped to a hot cell for examination.

The staff (Chatterton) requested that physics testing be croposed to confim the design parameters, expecially at the increased power level.

A copy of the brancn position on startuo testing was provided to OPPD.

The staff (Weiss) next questioned ENC handling of peaking factor uncertainties. ENC (0wsley) responded that they would use an apcroach similar to the Westinghcuse met. hod that is currently being reviewed by t

the staff. ENC was requested to provide a submittal covering the handling of these uncertainties.

The staff (Weiss) also requested documentation of a comparison of OPPD calculated PDQ values with those calculated by CE. OPPD (Gasper) responded that the calculations are the same. This should be addressed in the Cycle 6 reload submittal.

In summary, OPPD (Morris) provided the following schedule for submittals:

June 1, 1979 Setpoint Methodology August 1, 1979 ECCS Analysis October 1, 1979 Results of CE Review November 1,1979 ENC Analysis January 1980 Plant Shutdown March 15, 1980 Initial Criticality for Cycle 6 The staff questioned the effect of not granting the RCS pressure increase on return to operation at 1420 MWt for Cycle 6.

OPPD responded that the 2250 psi analysis would not bound operation at 2l00 psi.

The staff (Conner) discussed the ACRS subccmittee review and possible appearance before the full comittee. For power levels initially rev sed by i

the staff and the ACRS, the subcommittee reccanends whether a full ccmittee review is required. For the 1560 MWt power level, above the FSAR value, a full comittee review is required.

It was pointed out by Mr. Reid, that the ACRS generic action list and any open item from the ACRS letter on Ft. Calhoun operation will need to be addressed at the ACRS meetings.

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E. L. Conner, Project Manager Operating Reactors Branch ?4 Division of Operating Reactors

Enclosures:

1.

List of Attendees 2.

Meeting Agenda 3.

Slides Used 4

Transient Analysis cc w/ enclosures:

See attached pg.

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. s,i ATTENDANCE LIST March 27, 1979 MEETING WITH OPPD NRC CE Monte Conner D. M. Fayer R. W. Reid R. R. Mills K. G. Hoge Alfred Schornbrau Pete Kapo John F. Burdoin Margaret Chattartoi' Elinor Adensam Sy Weiss ENC OPPD Gerald Owsley G. Short Jim Morgan T. R. Robbins Larry Nielsen Richard L. Jaworski K. P. Galbraith Josepn K. Gasper Oli S. Wang Ken Morris

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NUCLEAR REGULATORT CCHMISSION OMAHA PUBLIC POWER DISTRICT MKETING FORT CALHofE STATION STRETCH POWER MARCH 27, 1979 ROOH P-130. PlirT.r.TPS BUIT.nTwn BEnfESDA, 'HARYIAND AGENDA I.

Introductico and Opening Remarts - OPPD A)

Review of Fort Calhoun Station's Licensing and Operatin6 R.istory B)

Present Licensed Power Level and Pressure i[

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Proposed Licensed Power Level and Pressure tr-M D)

Purpose of Today's Meeting N1(

NTf53 II.

Scope of Stretch Power Analyses and Philosophy - CPPD Whsa Eif M A)

Results of CE Feasibility Study

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Scope of Analyses D'rM C)

Philosophy Used in Performing Analyses III.

Review of Combustien Etsgineering's Role in Stretch Power Pr.}ect - C3 A)

Steam Generator Loading and Ibl^aAing Cycles 1)

Rasults of Feasibility Report 2)

Addendum to Strees Report and Specification Revisial O'

3)

Reduction in Allowable Number of Stress Qycles 1)

Cycles to Date 4

B)

Con tainment Pressure Evaluaticn 6JJ J J '/

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FeeAlbility Report Resulta 2)

Basia for Repeat Ana. lysis 3)

PSAR Reference to Ccatainment Pres = 1tre Evaluatica C)

Instrument Acceptability 1)

Temperature and Pressure Ccaditions to be Reviewed 2)

Linitaticna of Review (Nor a1 Operstica at Stretch Power)

2 D)

Overpressure Protectica Analysis at Operating Conditicos 1)

Secpe Descriptico 2)

Events to bra Baviewed E)

Fue1 C1ad Collansa 1)

Description of Zhvelope Approach 2)

Limits of Anm.lysia F)

Primary Component Technical Specificatica Review 1)

Approach 2)

Limita of Review G)

Auxiliary Systens Capability 1)

Listing of Systems 2)

Scope of Investiga*1co H)

Reactor Vessel Internals Stressen 1)

Peasibility Report Results 2)

Heat Generation Investigation 3)

Temerature Distributico Investigatica 14 )

Stress EvaluaH en e-

.IV.

heview of Exxon Nuclear's Role in St,, retch Pcarer Project - Descriptica I

of - ENC - Technical Informatica Relating to Fuel raa Analyses A)

Mechanical l

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Thermal Hydraulics j

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ECCS Analysis E)

Plant Transient An V.

Other St wtch Pcuer ceasiderations - CPPD A)

Administrstico B)

Technical Specification Review 6 L; C m

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Revision of Plant Procedures D)

Revision of Plant System Descriptiens E)

Append 14 I Opdating F)

Review of Radiological Consequences VI.

Su:= mary - OPPD A)

Review of Schedulo B)

Review Ibens Identifissi During 14eeting

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MECHANICAL DESIGN METHODOLOGY - PREV 100 SLY APPROVED FOR PWRS FORT CALHOUN APPLICATION DESIGN CRITERIA DESIGN FEATURES PERFOP.MANCE ANALYSIS e g:

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THERMAL-HYDRAULICS METHODOLOGY - PREVIOUSLY APPROVED FOR PWRS FORT CALHOUN APPLICATION STEADY-STATE FUEL ANALYSIS HYDRAULIC COMPATABILITY STEADY-STATE DflB L*

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IlEUTR0flICS iiETHODOLOGY - PREVIOUSLY APPROVED FOR OTHER PWRS FORT CALHOUN APPLICATION REACTIVITY COEFFICIENTS CYCLE REACTIVITY POWER DISTRIBUTION ANALYSIS CONTROL REQUIREMENTS R00 WORTH 3 655

2. ' 6

o. PLANT TRANSIENT ANALYSIS METHODOLOGY PTS MODEL - PREVIOUSLY APPROVED FOR PWRS FORT CALHOUN APPLICATION CALCULATE AND EVALUATE CONSEQUENCES OF MOST LIMITING TRANSIENTS 0

Flow REDUCTION 0

REACTIVITY INSERTIONS 0

OVERPRESSURIZATION REVIEW ALL TRANSIENTS ANALYZED IN SAR

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E-ECCS ANALYSIS METHODOLOGY - MODEL PREVIOUSLY APPROVED FOR PWRs FORT CALHOUN APPLICATION LARGE BREAK SPECTRUM ALLOWABLE LHGR Exposure SENSITIVITY STUDY ENC FUEL CE FUEL

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TRANSIENT ANALYSIS FOR STRETCH POWER Item Incident Status for Stretch 1.

CEA Withdrawal To be reanalyzed 2.

Boron Dilution To be reanalyzed 3.

CEA Drop To bc reanalysed 4.

Malpositioning of PLR prohibited during operation 5.

Loss of Flew To be reanalyzed 6.

Seized Rotor To be reanalyzed 7.

Idle Loop Startup Prohibited during operation 8.

Turbine / Generator Overspeed Not affected by stretch pwr 9.

Loss of Lead To be reanalyzed 10.

Loss of Feedwater To be reanalyzed 11.

Excess Load To be reanalyzed 12.

Steam Line Rupture To be reanalyzed 13.

Radiological Consequences of Steam Line Rupture FSAR is adequate 14.

CEA Ejection To be reanalyzed 15.

Radiological Consequences of CEA To be reanalyzed is fuel Ejection failure is predicted by ejection analysis 16.

Steam Generator Tube Rupture To be reanalyzed 17.

LOCA To be reanalyzed 18.

Containment Pressure Evaluation Assessment by CE for 2250 psia 19.

Generation of Hydrogen in the Containment FSAR is adequate 20.

Waste Gas Incident FSAR is adequate 21.

Fuel Handling and Fuel Loading Incidents FSAR is adequate 22.

Waste Liquid Incident FSAR is adequate 23.

Maximum Hypothetical Accident FSAR is adequate 24.

RCS Depressurization To be reanalyzed 25.

Malfunction of One Steam Generator To be reanalyzed 26.

Radiological Consequence of Feedwater Line Rupture Outside Containment FSAR is adequate 27.

Control Roca Dose to Operators Post LCCA FSAR is adequate 23.

Appendix I Specification Completed for 1500 MW

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