ML17309A250

Forwards App B to Evaluation of SEP Topics VI-2.D, Mass & Energy Release for Possible Pipe Break Inside Containment, & VI-3, Containment Pressure & Heat Removal Capability. Util 820201 Main Steam Line Break Analysis Acceptable

ML17309A250
Person / Time
Site:	Ginna
Issue date:	04/12/1982
From:	Crutchfield D, Crutcifield D Office of Nuclear Reactor Regulation
To:	Maier J ROCHESTER GAS & ELECTRIC CORP.
References
TASK-06-02.D, TASK-06-03, TASK-6-2.D, TASK-6-3, TASK-RR LSO5-82-04-029, LSO5-82-4-29, NUDOCS 8204140143
Download: ML17309A250 (12)

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April 12, 1982 Docket No. 50'-244 LS05-82-04-029 Hr. John E. Haier, Vice President Electric and Steam Production Rochester Gas 81 Electric Corporation 89 East Avenue Rochester, New York 14649 Ol rP

SUBJECT:

SYSTEMATIC EVALUATION PROGRAM (SEP)

FOK THE R. E.

GINNA NUCLEAR POHER PLANT - EVALUATION REPORT.-ON TOPICS VI-2.D AND VI-3 (DOCKET NO. 50-244)

References:

1.

Letter from D. Crutchfield of NRC to J. Haier of Rochester Gas 8 Electric Corporation, November 3,

1981, on the same subject.

2.

Letter from J. Haier of Rochester Gas 5 Electric Corporation to D. Crutchfield of NRC. February 1,

1982, on the same sug.ect.

Enclosed is a supplement (Appendix B) to our evaluation of SEP Topics VI-2.D, "Mass and Energy Release for Possible Pipe Break Inside Contain-ment," and V1-3, "Containment Pressure and Heat Removal Capability" (Reference;1,).

This Appendix evaluates the recent main steam line break analysis you supplied (Reference

2) which we found to be acceptable.

Our conclusions on the loss of-coolant accident, provided in Appendix A, I /

remain unchanged.

However, our main steam line break accident conclu-pz~)

sions have been modified.

Based on this change the main steam line break accident is no longer the pressure limiting event.

Therefore, the loss of coolant accident conditions will be used to define containment accident conditions.

P. y/o~g This completes our evaluation of these topics and will be a basic input to the integrated safety assessment for your facility.

This assessthent may be revised in the future if your facility design is changed or if NRC criteria relating to this subject are modified before the integrated assessment is completed.

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Dohket No. 50-244 LS05-82 Mr. John E. Maier, Vice President Electric and Steam Production Rochester Gas

& Electric Corporation 89 East Avenue Rochester, New York 14649

SUBJECT:

SYSTE<'IATIC EVALUATION PROGRAM (SEP)

FOR THE R.

E.

GINNA NUCLEAR POWER PLANT - EVALUATION REPORT ON TOPICS VI-2.D AND VI-3 (DOCKET NO. 50-244)

References:

1.

Letter from D. Crutchfield of NRC to J. Maier of Rochester Gas

& Electric Corporation, November 3,

1981, on the same subject.

2.

Letter from J. Naier of Rochester Gas

& Electric Corporation I'to D. Crutchfield of NRC, February 1,

1982, on the same subject.

Enclosed is a copy of Appendix B to our evaluation.of SEP Topics VI-2.D, "Mass and Energy Release for Possible Pipe Break Inside Containment,"

and VI-3, "Containment Pressure and Heat Removal Capability" (Reference 1).

This Appendix to our evaluation addresses the comment you supplied on the subject (Reference 2).

This evaluation will be a basic input to the integrated safety assess-ment for your facility.

This assessment may be revised in the future if your facility design is changed or if NRC criteria relating,to this subject are modified before the integrated assessment is completed.

Sincerely,

Enclosure:

As stated Dennis M. Crutchfield, Chief Operating Reactors Branch No.

5 Division of Licensing AD;SA;DL GLainas 4/

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. Mr. John E. Maier CC Harry H. Voigt; Esquire

LeBoeuf, Lamb, Leiby and MacRae 1333 New Hampshire Avenue, N.

M.

Suite 1100 Washington, D. C.

20036 Mr. Michael Slade 12 Trailwood Circle Rochester, New York 14618 Ezra Bialik Assistant Attorney General Environmental Protection Bureau New Yo'rk State Department of Law 2 Morld Trade Center New York, New York 10047 U. S. Environmental Protection'Agency Region II Office ATTN:

Regional Radiation Representative 26 Federal Plaza New York, New York 10007 Herbert Grossman, Esq.,

Chairman Atomic Safety and Licensing Board U. S. Nuclear Regulatory Commission Washington, D. C.

20555 Ronald C. Haynes, Regional Administrator Nuclear Regulatory Commission, Region I 631 Park Avenue King of Prussia, Pennsylvania 19406 Res ident Inspector R. E. Ginna Plant c/o U. S.

NRC 1503 Lake Road

Ontario, New York 14519 Director, Bureau of Nuclear Operations State of New York Energy Office Agency Building 2 Empire State Plaza

Albany, New York 12223 Supervisor of the Town of Ontario 107 Ridge Road West

Ontario, New York 14519 Dr. Emmeth A. Luebke Atomic Safety and Licensing Board U. S. Nuclear Regulatory Commission Washington, D. C.

20555 Dr. Richard F. Cole Atomic Safety and Licensing Board U. S. Nuclear Regulatory Commission Washington, D. C.

20555

APPENDIX 8:

STAFF RESPONSE TO LICENSEE COMMENTS This appendix contains the staff's comments on the information received from-the Rochester Gas and Electric Corporation (RG8E)

(Reference

81) regarding Reference 82.

Reference 82 transmitted the draft evaluation of SEP Topics VI-2.D and VI-3 for the Ginna plant to the licensee.

Res onse to Comments Corsnents 1, 2, 5, 6, 7 and 10 (see Attachment 1 to this Appendix) discuss data sources and the conservatism of the assumptions used in the LOCA analy-sis presented in Appendix A.

We concur with the licensee's view that the LOCA analysis in Appendix A is conservative.

Since the resulting pressure and temperature profiles are within design limits, the licensee has not

'presented an additional analysis.

Therefore, the results and conclusions of the LOCA analysis in Appendix A remain unchanged.

The Ginna design basis temperature profile for equipment qualification, iden-tified in a telecon with the licensee (Reference

83) as Figure 2 of Appendix 6E to the FSAR, exceeds the above cited LOCA temperature profile except for the time period between 10,000 seconds and 20,000 seconds.

However, RG8E proposed to revise the design basis temperature profile by increasing the temperature to 250'F for the period between 10,000 seconds and 20,000 seconds, and to use the revised temperature profile (shown in Figure 81) for the environmental qualification of electrical equipment.

We have concluded that this approach is acceptable, since the revised design basis temperature profile for equip-ment qualification exceeds the temperature profile based on the LOCA analysis in Appendix A.

It should also be noted that the containment pressure envelope (Figure 1 of Appendix 6E to the FSAR), which is shown in Figure B2, remains valid.

This pressure profile has been used for equipment qualification, as discussed in Reference B4.

Comment 8 notes that the term "Emergency Core Cooling System (ECCS) flow,"

used on page 18 of Appendix A, is not correct and should read "accumulator flow."

We concur with the licensee's comment.

Comment 9 points out an error on page 21 of Appendix A; the containment de-sign pressure is 75 psia.

We agree.

Corwents 3, 4, and 11 pertain to the Main Steam Line Break (MSLB) analysis, which has been reanalyzed by the licensee in Attachment B of Reference Bl.

The following, discussion is an evaluation of the licensee's analysis.

Evaluation of Licensee's MSLB Anal sis In Attachment B of Reference Bl, the licensee submitted..an analysis of the containment temperature and pressure response following a MSLB.

The analysis was performed in the following manner:

The licensee reconstructed the worst case (Case

5) presented in Appendix A; the licensee performed a study of sev-eral identified conservatisms to show the impact of more realistic assumptions and, finally, the licensee reanalyzed the Hot Zero Power (HZP) and Hot Full Power (HFP) cases.

The results show that for the worst case the calculated peak containment pressure is less than the design pressure of 75 psia (60 psig) and the containment temperature exceeds the design temperature for a-short time.

3-The licensee used the CONTEMPT-EI/288 computer code, which is a modified ver-

. sion of the staff's CONTEMPT-LT code.

The comparative study in Reference Bl shows that for the worst case MSLB there is not much difference in the peak calculated containment pressure and temperature from these two computer codes.

We, therefore, concur with the licensee's finding that using the CONTEMPT-EI/

288 code is acceptable for the Ginna containment sensitivity study.

The li-censee's MSLB analysis was based on newly identified heat sinks, the use of the Uchida film heat transfer correlation following blowdown and the use of four fan coolers for post-accident containment heat removal.

The blowdown data were adopted from Appendix A for the HFP case and from an Exxon Nuclear Company (ENC) calculation for the HZP case.

The newly identified heat sinks include the accumulators and ducting, which is acceptable.

Using the Uchida heat transfer correlation for MSLB analyses is acceptable since it is recommended in Appendix 8 to NUREG-0588, Interim Position on Environmental gualification of Safety-Related Electrical Equip-ment.

The Ginna containment is equipped with four fan coolers.

(Contrary to statements in Appendix A about the use of four fan coolers, the input data erroneously accounted for only one fan cooler.)

A comparison of blowdown data from Appendix A and from the ENC calculation (Figure 3 of Reference 81) shows good agreement between the results, and is, therefore, acceptable.

The licensee's calculated containment pressure profi'les for the HZP and HFP cases are shown in Figures 5 and 7 of Reference 81.'he peak calculated pres-sures for the HZP and HFP cases are 72.4 psia and 63.2 psia, respectively, which are less than the containment design pressure of 75 psia.

The calculated

containment temperature profiles for the HZP and HFP cases are presented in Figures 4 and 6 of Reference 1.

Based on our review of the licensee's MSLB

analysis, we conclude that the resulting containment pressure and temperature profiles are acceptable.

The containment temperature response following a MSLB exceeds the Ginna de-sign basis temperature profile used in qualifying electrical equipment (Fig-ure Bl).

However, the time over which this occurs is short; i.e., less than 60 seconds.

Also, the Ginna plant has an automatic spray system designed to accommodate a single active fai lure.

Based on this fact and the guidance in Reference 85, the containment temperature/pressure profiles for the worst case LOCA are acceptable for use in equipment qualification.

Figures 81 and 82 bound the LOCA temperature and pressure profiles, respectively.

Conclusions Based on our review and evaluation of Refer ence 81, the following conclusions can be made:

1)

The conclusions drawn in Appendix A for the LOCA analysis remain unchanged; 2)

The calculated peak containment pressure for a MSLB accident is less than the containment design pressure; and 3)

The design basis temperature and pressure profiles, as shown in Figures Bl and 82, are acceptable for use in equipment qualification.

Reference:

81. Letter from J. Maier of Rochester Gas and Electric Corporation to 0. Crutchfield (NRC), February 1, 1982,

Subject:

SEP Topics VI-2.D and VI-3, R.

E. Ginna Nuclear Power Plant, Docket No. 50-.244 B2. Letter from D. Crutchfield (NRC) to J. Maier of Rochester Gas and Electric Corporation, November 3,'981,

Subject:

SEP for the R. E. Ginna Nuclear Power Plant - Evaluation Report on Topics VI-2.0 and IV-3.

B3: Telecon between G. nobel (RG8E) and C. Li (NRC) on March 12, 1982,

Subject:

Clarification of Ginna design basis temperature profile for equipment qualification.

84: Technical'valuation Report of Equipment Environmental gualification for the R. E. Ginna Station, March 18, 1981.

85: Guidelines for Evaluation of Environmental gualification of Class IE Electrical Equipment in Operating Reactors,

November, 1979.

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