ML20086F562

Application for Amend to License DPR-40,revising TS 2.1.6 in Order to Change LCO of Main Steam Safety Valves

ML20086F562
Person / Time
Site:	Fort Calhoun
Issue date:	11/27/1991
From:	OMAHA PUBLIC POWER DISTRICT
To:
Shared Package
ML20086F552	List: ML20086F551 ML20086F562 ML20086F568
References
NUDOCS 9112030285
Download: ML20086F562 (14)

Text

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BEFORE THE UNITED STATES NUCLEAR REGULATORY C0KMISS10N in the Matter of )

)

Omaha Public Power District ) Docket No. 50-285 (Fort Calhoun Station )

Unit No. 1) )

APPLICATION FOR AMENDMENT OF .

OPERATING LICENSE Pursuant to Section 50.90 of the regulations of the U. S. Nuclear Regulatory Commission ("the Commission"), Omaha Public Power District, holder of Facility Operating License No. DPR-40, herewith requests that Section 2.1.6 of the Technical Specifications set forth in Anpendix A to that License be amended to change the Limiting Conditions for Operation of the Main Steam Safety Valves.

The proposed changes in Technical Specifications are discussed in the Application for Amendment dated June 28, 1991. This application only provides numerical values for the safety valve setpoints and therefore does not affect the discussion, justification and no significant hazards consideration transmitted in the Application for Amendment dated June 28, 1991. The proposed changes in specifications would not authorize any change in the types or any increase in the amounts of effluents or a change in the authorized power level of the facility.

WHEREFORE, Applicant respectfully requests that Sectinn 2.1.6 of Appendix A to facility Operating License No. DPR-40 be amended in the form attached hereto as Attachment 3.

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t-A copy of this Application, including its attachments, has been sub-mitted to the Director - Nebraska State Division of Radiological Health, as required by 10 CFR 50,91.

OMAHA PUBLIC POWER DISTRICT By t$'--

Acting Division Manager Nuclear Operations Subscribed and sworn to before me this M day of November, 1991.

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, MELYA L EVANS c,s )

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Notary Public

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMHioS10N in the Matter of )

)

Omaha Public Power District ) Docket No. 50-285 (Fort Calhoun Station )

Unit No. 1) )

AFFIDAVIT T. L. Patterson, being duly sworn, hereby deposes and says that he is the Acting Division Manager - PJclear Operations of the Omaha Public Power District; that as such he is duly authorized to sign and file with the Nuclear Regulatory Commission the attached itiformation concerning the. Application for Amendment dated 11/27/91, concerning revision of Main Steam Safety Valve Setpoints'; that he is familiar with the content thereof; and that the matters set forth therein are true and correct to the best of his knowledge, information, and belief.

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Acting Division Manager Nuclear Operations STATE OF NEBRASKA)

, ) ss COUNTY OF DOUGLAS) l Subscribed and sworn to before me, a Notary Public in and for the State of Nebraska on this 21 day of November, 1991 '

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ATTACHMENT 1 OVEST10NS ON FORT CALHOUN MSSV AMENDMENT RE0 VEST Question No, i: Technical Specifications 2.1.6(3) requires only that eight of the 10 main steam safety valves (MSSVs) be operable with their lift settings maintained between 1000 psia and 1050 psia +3/-2% of the nominal nameplate setpoint values, lhe MSSV open setpoint assumed in the loss of Feedwater Flow analysis is 1066 psia, which could be nonconservative compared to the actual settings since safety valves are allowed to operate at higher settings values than the analysis values. Provide the li'. settings for all the MSSVs on Technical Specification 2.1.6.

Answer No. 1: Attachment 3 contains replacement pages for OPPD's

" Application for Amendment of Operating License" submitted by Reference 2, which seeks to amend Section 2.1.6 of the Technical Specificatic~.s. The changes requested pertain to limiting Conditions for Operation of the Main Steam Safety Valves.

The replacement pages only provide numerical values for the safety valve setpoints and therefore do not affect the discussion, justification and no significant hazards considerations contained in Reference 2.

• Question No. 2: See Attachment 2 for reply.

Question No. 3: See Attachment 2 for reply.

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ATTACHMENT 2 Question 2: Explain and justify in de' ail how steam generator and pressurizer safety valves were modelled in the analyses for overpressurization events, especially the model-ling assumptions for the valve accumulation, blowdown, and setpoint tolerance. Identify the locations of the max l mum RCS and secondary pressures for overpressurization events, Response 2: The main steam safety valve blowdown and accumulation model contained in the CESEC code is shown in the attached Figure A. The pressurizer (primary) safety valve blowdown and accumulation model is shown in Figure B.

f The setpoint tolerances were varied and a number of computer runs ovete made to deter-mine thu system response sensitivities with various corribinations of setpoints for the Main Steam Satety Valves (MSSVs) and the Primary Safety Valves (PSVs) for the limiting tran-sients. For Fort Calhoun the limiting design basis transients have been determined to be the Loss of Load event for the primary system and the Loss o' Feedwater event for the sec-ondary system. These limiting transients were established by OPPD and Combustion En-gineering during the benchmarking of the CESEC code and the reload analyses per-

, formed for the 2700 MWt analog plants and Fort Calhoun. The transient resuits indicated that a setpoint drift of up to + 5% for both the MSSVs and the PSVs would satisfy the accep-tance requirements 'or the events. For conservatism, however, a value of +3% was se-lected for incorporation into the facility license change request. Only 8 of 10 main steam safety valves were assumed to be operable, in accordance with the Technical Specification requirements (1015,1025,1040 and 1050 psia)

The maximum pressure location for the primary system for the Loss of Load and Loss of Feedwater events is in the Reactor Coolant System while the secondary peak pressure is -

located at the Steam Generator. CESEC does not have the modelling fidelity to provide more concise locations. If the NRC approves the CENTS code for use by OPPD then a more accurate location can be identified due to the larger number of nodo locations calcu-lated.

Question 3: Discuss the peak RCS and secondary pressures for the relevant overpressur-ization transients such as Turbine Trip, Feedlir e Break, and Locked Rotor analyses with the increased setpoint tolerance.

Response 3: The Turbine Trip without simultaneous reactor trip is called the Loss of Load (LOL) analysis in the Fort Calhoun USAR. The LOL used a conservative, bounding analysis to calculate the peak pressures. Enclosed is Table 1 for the input parameters, sequence of events summary (Table 2) and Figures 1-1 through 1-4 indicating the system re-sponses from the +3% drift case. The +3% drift case was one of thc cases that were run ,

as part of the sensitivity analysis. The safety valve opening point in iicated in the summary (Table 2) is for the first valve accounting for setpoint drift and accumulation.

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_ . _ _ _ . __ _ _ _____._ _ _ _ _ _.~ _._.___.__.. _ _ __. . _ _ _ _.

4 The Feedwater Une Break Accident, analyzed as an overpressure / partialloss of RCS heat removal capacity event, is not part of the Fort Calhoun design basis and was not specifical-ly considered as such in the initiallicense application. In Appendix M of the Fort Calhoun i

Station USAR the feedwater line break is examined and shown to be bounded by the steamline break, as a RCS overcooling / excess heat removal event, in response to an NRC question during closure of NUREG-0737, item II. " 2, OPPD responded to the specific NRC question at the time rather than identifying Uh. R Appendix M as the reason for not providing a steam space volume. It is OPPD's posi, un that in providing this information no commitment was made to revise the Fort Calhoun Station design basis by including the feedwater line break (as an overpressurization event) as a USAR Chapter 14 postulated accident.

The Locked Rotor event is described as the Seized Rotor event for Fort Calhoun. The Seized Rotor event was not specifically analyzed for the increased setpoint drift since the transient does not have sufficient time to develop due to the reactor trip on low flow within the first few seconds of the initiation of the event. This willlimit the temperature and pres- -

sure increases for the event to a level much less than the limiting event, ,

i 2

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FIGURE A -

SECONDARY SAFETY VALVE CHARACTERISTICS .

FRACTIONAL VALVE FLOW AREA FRACTIONAL VALVE FLOW AREA

- WHEN INLET PRESSURE REACHES WHEN INLET PRESSUREDOES NOT -

GREATER THAN OR EQUAL TO 3% REACH 3% ABOVE THE VALVE SET

- ABOVE THE SET PRESSURE DURING PRESSURE DURING ATRANSIENT-A TRANSIENT. .

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1 Table 1 FORT CALHOUN CYCLE 11 KEY PARAMETERS AS$UMED IN THE LOSS O!

LOAD TO BOTH STEAM CENERATORS ANALYSIS farmetrt un,tig

' value initial Ce,re Power Level MWt 1530 (102n)

Initial Core Inlet Temperature 'T 547 Initial Pressurizer Pressure psia 2,053 Initial Steam Generator Pressure psia 815 tuittal hCS Flow' Rate gpm 196.,000 Moderator Temperature Coefficient 10"' ap/'F + 0.5 Fuel Temperature Coefficient 10 ap/*F Least negative-

. predicted during core life.

Fuel Temperature Coefficient Multiplier 0.85 CEA Time to 1004 Insertion (Including Holding coil Delay) sec. 3.1 Scram Reactivity Worth gap

-6.65 Kinetics Parameters A

.004696 Allowable Primary and Secondary Safety Valve Setpoint Drift'  %

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Table' 2-FORT CALHOUN CYCLE 11 SEQUENCE OF EVENTS FOR THE LOSS OF: LOAD EVENT-I_0 MAXIMlZE CALCULATED RCS PEAX PRESSURE '

1 Time (sec)_

Event E Setooint or value 0.0 Loss of Secondary Load ----

i 6.8 Steam Generator Safety Valves Open 1066 psia 8.9 High Pressurizer Pressure Analysis 2422 psia Trip Setpoint is Reached-10.2 CEAs Begin to Drop Into Core ,

10.6 Pressurizer Safety Valves Open 2575 psia 11.3.

Maximum RCS Pressure 2649 psia -

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Maximun. Steam Generator Pressure 1078 psia-4 I

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