ML20137C016

Forwards Responses to 850916 Request for Addl Info Re Main Steam Line Break in Doghouse.Heat Transfer Rate Maintained in Excess of Value Calculated W/Combination of Tube Bundle Uncovery & Superheated Steam Heat Transfer

ML20137C016
Person / Time
Site:	Mcguire, Catawba, McGuire, 05000000
Issue date:	11/15/1985
From:	Tucker H DUKE POWER CO.
To:	Adensam E, Harold Denton Office of Nuclear Reactor Regulation
References
TAC-59005, TAC-59006, NUDOCS 8511260408
Download: ML20137C016 (4)

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1 DUKIs Powicn Coxi>m l'.O. ISOX W11150 CitAHLOTTit, N.C. 215242 stAL 15. Tt:0KEH TELEPHONE vuareessue=,

(704) 373-4fklt mies.aan reessesmes November 15, 1985 L

Mr. liarold R. Denton, Director Office of Nuc1 car Reactor Regulation U. S. Nuclear Regulatory Commission l Washington, D. C. 20555 Attention: Ms. E. G. Adensam, Chief Licensing Branch No. 4

Subject:

McGuire Nucicar Station Catawba Nuc1 car Station Docket Nos. 50-369, -370 -413. -414 Dear Mr. Dentont in response to your letter dated September 16, 1985 which requested additional information regarding main steam line break in the Doghouse, please find attached our responses. i questions 1 through 3 have been answered on behalf of Duke by Westinghouse.

Question 4, which was asked on the McGuiro dockets has been answered by Duke for both McGuire and Catawba.

If there are any questions regarding these responses, please advise.

Very truly yours,

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conservative and that the results submitted in WCAP-9226, " Reactor Core Response to Excessive Secondary Steam Releases," (as referenced in the FEAR) may not be valid for all conditions. Address, in detail the consequenses of modeling superheat and its effect on core response. , Provide detailed comparative analyses of the results calculated with the two models (saturated versus superheated steam models). Response to Question 2 The analysis results in WCAP-9226 address two general categories of steamline break analyses.

• The first category addresses steamline breaks which occur while the reactor is' critical and operating at power prior to reactor trip. Tube bundle mcovery is not predicted prior to reactor trip and the point of minimun DNBR during these transients and therefore there is no impact on the analysis resulta.

The second category addresses stearr.line breaks which occur while the reactor is in hot zero power condition with the most reactive rod stuck out of the core. These transients are also analyzed on a plant specific basis in the McGuire plant FSAR. The analysis methodology used in these analyses conservatively force a delay in the tube buncr'e mcovery well past the predicted tube bundle mcovery point. Therefore, a heat transfer rate is maintained in excess of the value which would be calculated with the combination of the predicted tube bundle mcovery and the superheated steam heat transfer. This is conservative since it maximizes the cooldown by the faulted steam generator and consequently maximizes the return to power. Therefore, the results presented in the McGuire FSAR and WCAF-9226 recain valid. RESPONSES 1D NRC QU5TIONS m STEAM LINE BREAK DOGHOUSE ANALYSES QUETION 3 As a consequence of equipment failures from adverse environmental conditions, the main steam isolation valves in the affected Doghouse were assGned to reopen. This led to blowdown of the two steam generators: e (a) Provide the details and justification of the reactivity methodology used in the analyses as well as the nodalization and the primary coolant mixing coefficients applied in the reactor vessel. (b) Provide the details and justification to support the mixing coefficients applied to the reactor coolant as well as the reactivity feedback models. (This information was requested by the NRC in 1983 on the WCAP-9226 submittal but Westinghouse has not responded to the request. ) (c) Describe how the analyses were performed when assuming the stuck rod cluster control assembly to be positioned in loop 1 core sector versus loop 2 core sector. Response to Question 3 (a) The system nodalization methodology is described in Reference 2. The information on primary coolant mixing coefficients and reactivity methodology is being handled generically as discussed in Reference 3. (b) This information request is being addressed generically as described in Reference 3. (Copy enclosed) (c) As described in Reference 2, the average density, water temperature, and boron concentration used in LOFTRAN can be weighted to a particular core sector (loop). This method is used to conservatively, simulate the effect of a stuck rod in the indicated loops on the core average reactivity calculation. For both cases, " Loop 1" was asstaned to be the loop which continued to blow down after the initial steamline isolation signal. " Loop 2" was assumed to be the loop which had a steamline isolation valve reopen as a result of the superheated steam generated after tube bundle tacovery in the Loop 1 steam generator. One set of analyses assumed that the core sector associated with Loop 1 contained the stuck rod and therefore weighted the Loop 1 core sector density temperature, and boron concentration heavily in the aver. e core rea,ctivity calculation. The second set of analyses assumed tha*L the core sector associated with Loop 2 contained the stuck rod and therefore weighted the Loop 2 core sector density, temperature, and boron concentration heavily in the average core reactivity calculation. m

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RESPONSES 'IO NRC QUESTIONS CN STEAM LINE BREAK DOGHOUSE REFERDIGS

1. Letter from E. P. Rahe, Jr. to C. O. Thomas, ENac-E-3009 1 dated February 27, 195 ,

Subject:

Topical Reports WCAP-8822-P-S1 and WCAP-c860-S1, " Mass and Energy Releases Following A Steam Line Rupture".  ;

2. Burnett, T. W. T., et al, "LOFTRAN Code Description," WCAP-7907-P-A (Proprietary), WCAP-7907-A (Non-Proprietary), April 1984.

3 Letter October dated from E. P. Rahe, Jr. (M) to H. L. Thompson (NRC), ENRC-85-3077, 29,195 G

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% j Westinghouse Water Reactor Electric Corporation Divisions Qyjgnny,,n,,mm3 3 October 29,1985 NS-idic-i.5-3077 Mr. Hugh L. Thompson, Director Division of Licencing U. S. Nuclear Regulatory Comission Washington, D.C. 20555 H.S. 528 (Old Phillips Bldg.)

Subject:

Schedule for Response to Request for Additional Information on WCAP-9226 (P) and WCAP-9227 (NP)

Refs.: NS-TMA-2080, Anderson (W) to Stolz (NRC), 5-17/79 NS-INA-2285, Anderson (W) to Hiller (NRC), 7-30-80 NS-EPR-3007, Rahe (W) to Thomas (NRC), 2-19-C5

Dear Mr. Thompson:

During the course of the NRC review on the Wertint.cuse c topical reports on the steamline break methodology (WCAP-9226 and WCAF-9227), Westin@ouse has provided responses to the staff's requests for additional information on several occasions (referenced letters). There are, however, a number of questions for which responses have not yet been formally supplied, although they have been discussed with members of your staff. Information was also supplied to Argonne National Laboratories in the stumer of 1983 in lieu of fermal responses to some of the requests for additional information in order to facilitate the staff's review of the topical reports.

We'stin@ouse has been preparing responses to the remainder of the questions during this year. We expect to have completed our work in the first quarter of 1986. Accordingly, Westinghouse will submit responses to the retraining questions by April 1,1986. -

If you have any questions, please contact Mr. J. L. Little of my staff (412/374-5054).

Very truly yours,

\

E.P. Rahe,Jf ~

Manager Nuclear Safety Department HIO:pj L 'l(,f

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QUESTION 4: Provide your evaluations of offsite dose, including in particular the case with two steam generator blowdown.

Response to Quention 4: McGuire Main Steam Line Break Dose Evaluation The following two accident scenarios were considered for the MSLB offsite dose evaluations (1) single steam generator blowdown, (2) two steam generator blowdown. Note that the generation of superheated steam does not affect the dose analysis since the entire mass of the defective steam generator (s) is released.

Accident Assumptions

1. The primary and secondary side concentrations are at the following Technical Specification limits when the accident occurs: i Primary - 1 pCi/gm dose equivalent I-131 100/Eypci/gm gross activity (assumed to be all noble gases) l Secondary - 0.1 pCi/gm dose equivalent I-131 ,

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2. The spectrum of iodine isotopes in the primary and secondary coolant is the same as the Design Basis primary coolant spectrum.

3. The primary-to-secondary leak rate of 1 gpm is distributed between the  !

steam generators for the duration of the accident as One Steam Generator Blowdown 0.347 gpm for defective steam generators 0.653 gpm for nondefective steam generators Two Steam Generators Blowdown 0.694 gpm for defective steam generators 0.306 gpm for nondefective steam generators

4. All noble gases which leak to the secondary side are released via tne steam reicase.

5. Initial steam release for the defective steam generator (s) terminates in 30 minutes. Steam release for the nondefective steam generators terminates in 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

6. The iodine partition factor in the defective steam generator (s) is 1.

7. The iodine partition factor in the nondefective steam generators is 0.01.

8. Mass Releases From MSLB Accident 1 Steam Generator Blowdown Case Time Mass Release From Faulted S.G. Mass Release From Intact S.Gs.

(hrs) (1bm) (Ibm) 0-2 108.500 ,

378.248 2-8 ------- 963,012 2 Steam Generators Blowdown Case Time Mass Release From Faulted S.Cs. Mass Release From Intact S.Cs.

(hra) (1bm) (ibm) 0-2 217.000 378.248 2-8 -------

963.012

9. Accident Dispersion Factors (X/Q)

Exclusion Area Boundary - 9.00E-04 sec/m3 Low Population Zone - 8.00E-05 sec/m3 Two reactor coolant activity cases were evaluated for the MSLB accident.

Case 1 A preaccident iodine spike exists with iodine concentrations that are assumed to be the maximum permitted by Technical Specifications for full power operation, i.e., 60 pCi/gm dose equivalent I-131.

Case 2 In this case the reactor trip and/or primary system depressurization associated with the MSLB creates an iodine spike in the primary system.

The increase in primary coolant iodine concentration is estimated using a spiking model which assumes that the iodine release rate from the fuel rods to the primary coolant increases to a value 500 times greater than the release rate at the Technical Specification equilibrium value. The iodine spike continues for the duration of the accident.

Results One Steam Generator Blowdown - Offsite Doses (Rem)

Exclusion Area Boundary Low Population Zone Whole Body Thyroid, , Whole Body Thyroid

Case 1 1.48E-02 5.49E+00 4.00E-03 1.32E+0d case 2 1.85E-02 4.02E+00 1.00E-02 2.46E+00 Two Steam Generator Blowdown - Offsite Doses (Rem)

Exclusion Area Boundary Low Population Zone Whole Body Thyroid Whole Body Thyroid Case 1 2.21E-02 1.09E+01 5.33E-03 2.57E+00 l

l Case 2 2.96E-02 7.94E+00 1.73E-02 4.82E+00 The conservatively calculated doses are well within the exposure guideline values set forth in 10 CFR Patt 100, Section 11.

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r QUESTION 4: Provide your evaluations of offsite dose including in particular the case with two steam generator blowdown.

Response to Question 4: Catawba Main Steam Line Break Dose Evaluation The following two accident scenarios were considered for the MSLB offsite dose evaluations (1) single steam generator blowdown, (2) two steam generator blowdown. Note that the generation of superheated steam does not af fect the dose analysis since the entire mass o.f the defective steam generator (s) is released.

Accident Assumptions

1. The primary and secondary side concentrations are at the following Technical Specification limits when the accident occurs:

Primary - 1 pCi/gm dose equivalent I-131 100/EypC1/gm gross activity (assumed to be all noble gases)

Secondary - 0.1 uCi/gm dose equivalent I-131

2. The spectrum of iodine isotopes in the primary and secondary coolant is the same as the Design Basis primary coolant spectrum.

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3. The primary-to-secondary leak rate of 1 gpm is distributed between the '

steam generators for the duration of the accident as One Steam Generator Blowdown 0.347 gpm for defective steam generators 0.653 gpm for nondefective steam generators Two Steam Centrators Blowdown 0.694 gpm for defective steam generators 0.306 gpm for nondefective steam generators

4. All noble gases which leak to the secondary side are released via the steam release.

5. Initial steam release for the defective steam generator (s) terninates in 30 minutes. Steam release for the nondefective steam generators terminates in 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

6. The iodine partition factor in the defective steam generator (s) is 1.

7. The iodine partition factor in the nondefective steam generators is 0.01.

8. Mass Releases From MSLB Accident 1 Steam Generator Blowdown Case Time Mass Release From Faulted S.C. Mass Release From Intact S.cs.

(hre) (1bm) (1bm) 0-2 175,700 ., 403,650 2-B -------

968,638 2 Steam Generators Blowdown Case Time Mass Release From Faulted S.Cs. Mass Release From Intact S.Gs.

(hrs) (ibm) (1bm) 0-2 351,400 429.120 2-8 -------

924,777

9. Accident Dispersion Factors (X/Q)

Exclusion Area Boundary - 5.50E-04 sec/m3 Low Fopulation Zone - 1.80E-05 sec/m3 Two reactor coolant activity cases were evaluated for the MSLB accident.

Case 1 A preaccident iodine spike exists with iodine concentrations that are assumed to be the maximum permitted by Technical Specifications for full power operation, i.e., 60 pC1/gm dose equivalent I-131.

Case 2 In this case, the reactor trip and/or primary system depressurization associated with the MSLB creates an iodine spike in the primary system.

The increase in primary coolant iodine concentration is estimated using a spiking model which assumes that the iodine release rate from the fuel rods to the primary coolant increases to a value 500 times greater than the release rate at the Technical Specification equilibrium value. The lodine spike continues for the duration of the accident.

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Results .

One Steam Generator Blowdown - Offsite Doses (Rem)

Exclusion Area Boundary Low Population Zone Whole Body Thyroid . Whole Body Thyroid Case 1 1.03E-02 4.22E+00 9.38E-04 3.23E-01 Case 2 1.26E-02 3.32E+00 2.29E-03 5.77E-01 Two Steam Generator Blowdown - Offsite Doses (Rem)

Exclusion Area Boundary Low Population Zone Whole Body Thyroid Whole Body Thyroid Case 1 1.61E-02 8.37EH)0 1.28E-03 6.34E-01 Case 2 2.07E-02 6.59E+00 3.97E-03 1.14E+00 The conservatively calculated doses are well within the exposure guideline values set forth in 10 CFR Part 100 Section 11.

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