B11573, Forwards Response to SER Confirmatory Item 23 Re Max Calculated External Differential Pressure on Containment, Identified in NUREG-1031.Revised Analysis Performed.Results Acceptable.Revs Will Be Included in Future FSAR Amend

Forwards Response to SER Confirmatory Item 23 Re Max Calculated External Differential Pressure on Containment, Identified in NUREG-1031.Revised Analysis Performed.Results Acceptable.Revs Will Be Included in Future FSAR Amend

ML20127E706
Person / Time
Site:	Millstone
Issue date:	06/17/1985
From:	Opeka J NORTHEAST NUCLEAR ENERGY CO.
To:	Youngblood B Office of Nuclear Reactor Regulation
References
RTR-NUREG-1031 B11573, NUDOCS 8506240626
Download: ML20127E706 (9)

Text

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NORTHEAST UTILITIES con.,.i Omc.. . s ioen sire.i. Boriin. conn.ciicui 1 . s- EsEIcN=. P.O. BOX 270 a w ana " "**""

HARTFORD CONNECTICUT 06141-0270 L L J ((', [1**d' '5"", (203) 66s-s000 June 17,1985 Docket No. 50-423 B11573 Director of Nuclear Reactor Regulation Mr. B. 3. Youngblood, Chief Licensing Branch No.1 Division of Licensing U.S. Nuclear Regulatory Commission Washington, D.C. 20555

Reference:

(1) B. 3. Youngblood letter to W. G. Counsil, Issuance of Safety Evaluation Report - NUREG 1031 - Millstone Nuclear Power Station, Unit No. 3, dated August 2,1984.

Dear Mr. Youngblood:

Millstone Nuclear Power Station, Unit No. 3 Response to SER Confirmatory Item #23 Attached is Northeast Nuclear Energy Company's (NNECO) response to SER Confirmatory Item #23 concerning the maximum calculated external differential pressure on containment which was identified to NNECO in Reference (1).

As indicated in the response, a revised minimum containment pressure analysis has been performed using the most severe allowable initial conditions. The results of this analysis show that the minimum calculated containment pressure is above the minimum design pressure and is, therefore, acceptable.

The appropriate portions of Section 6.2 of the FSAR have been revised to reflect this most current analysis and will be included in a subsequent amendment to the FSAR.

This information should be sufficient to resolve Confirmatory Item #23.

However, if you have any further questions, please contact our licensing

! representative directly.

Very truly yours, NORTHEAST NUCLEAR ENERGY COMPANY et. al.

BY NORTHEAST NUCLEAR ENTRGY COMPANY Their Agent g 3. F. Opeka F. QA O E

Senior Vice President O

t

'bO \

STATE OF CONNECTICUT )

) ss. Berlin COUNTY OF HARTFORD )

Then personally appeared before me 3. F. Opeka, who being duly sworn, did state that he is Senior Vice President of Northeast Nuclear Energy Company, an Applicant herein, that he is authorized to execute and file the foregoing information in the name and on behalf'of the Applicants herein and that the statements contained in said information are true and correct to the best of his knowledge and belief.

AMjid -l @M$f Notary Pyblic My Commission Expires March 31,1988 I

i

SER Confirmatory Item #23 - Maximum External Differential Pressure on Containment (SER Section 6.2.1.5)

The staff has reviewed the applicant's input parameters used in the minimum containment pressure analysis including initial containment conditions, containment net free volume, passive heat sinks, heat transfer to passive heat sinks, and containment active heat removal, and it has found them acceptably conservative and in conformance with BTP CSB 6-1, " Minimum Containment Pressure Mode. for PWR ECCS Performance Evaluation," with one exception.

This exception is the nonconservative assumption of an initial containment pressure of 9.5 psia, which is greater than that which may be encountered under limiting normal operating conditions. The staff is unable to form a conclusion on the acceptability of the analysis, however, because the applicant has not provided the mass and energy release data and the calculated containment pressure response. Once this information is received, the staff will report the results of its review in a supplement to the SER. This is considered a confirmatory item.

Response: (June 1985)

Refer to revised FSAR Section 6.2.1.1.3.5.

A revised minimum containment pressure analysis shows the calculated minimum pressure of 8.03 psia to be above the design pressure of 8.00 psia using the most severe allowable initial conditions.

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5. The rate of mass and energy release to the containment from 1.9 the break during the core reflooding period

6. The mass of nitrogen added to the containment from the gas 1.10 accumulators Following the core reflooding period, the containment 1.12 depressurization systems and containment passive heat sinks remove 1.13 energy from the containment atmosphere at a rate sufficient to reduce 1.14 the pressure to below atmospheric pressure in less than 60 minutes. 1.15

! The depressurization time is a function of the following parameters: 1.16

1. The containment free volume 1.18

2. The mass of air inside the containment structure 1.19 I

3. The rate of heat removal (or addition) from (or to) the 1.20 containment atmosphere by the passive heat sinks within the containment structure 1.21

4. The rate of heat removal from the containment atmosphere by 1.22 the containment heat removal systems (this is significantly 1.23 i dependent on the ultimate heat sink temperature).

5. The rate of mass and energy release to the containment from 1.24 j the break following the end of core reflooding

6. The mass of nitrogen added to the containment from the gas 1.25 i accumulators l After the containment is depressurized, the depressurization systems 1.27 continue to remove energy from the containment at a rate sufficient 1.28 to maintain the containment subatmospheric pressure. The passive 1.29 1 heat sinks add net energy back to the containment atmosphere following depressurization. The containment experiences a 1.30 subatmospheric pressure peak when the RWST empties, terminating

, quench spray.

i 6.2.1.1.3.2 containment Analysis Analytical Model 1.32 The LOCTIC computer program which is used to model the containment 1.33 system, the passive heat sinks, and the containment heat removal 1.34 systems was developed by Stone & Webster. A topical report (LOCTIC 1.36 -

1971) described in detail the assumptions used and the mathematical

. formulations employed. Comparisons made with the standard CONTEMPT-LT (Aerojet 1975) code show excellent agreement. LOCTIC has been 1.37l84-th 1.38 under development for many years and was used in the design of six operating stations

Surry Power Station Units 1 and 2 (USAEC 1972b), 1.39 Maine Yankee Atomic Power Station (USAEC 1972c), North Anna Station Units 1 and 2 (USAEC 1970), and Beaver Valley Power Station Unit 1.41 No. 1 (USAEC 1974c). Other station designs that have utilized the 1.42

, LOCTIC code are the North Anna Power Station Unit 3 (USAEC 1972a) and the Beaver Valley Power Station Unit 2 (USAEC 1974a). 1.43 4

] Amendment 14 6.2-5 July 1985 i -

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u1217912sro14i 05/30/85 241 MNPS-3 FSAR opposite side, and some are_ modeled as being insulated (ending 1.9 boundary).

Resistance to heat transfer at the liner-concrete interface is 1.10 considered in the containment analysis by use of a conservatively low 1.11 value of thermal contact conductance of 100 Btu /hr/ft 2/*F (Gido 1978). Since the steel liner is used as a form for pouring of the 1.12 concrete, and since the concrete mix is very wet, the liner, in 1.13 effect, becomes " glued" to the concrete.

The model considers transient heat conduction to the containment 1.14 structure through the composite thermal resistance made up of the 1.16 paint film on the steel liner, the liner itself, the liner-concrete

- interface, and the concrete. Section 6.2.1.1.3.2 discusses the mesh 1.17 sizing for the passive heat sinks.

'6.2.1.1.3.5 External Pressure 1.19 Inadvertent operation of the containment heat removal systems will 1.20 cause a decrease in the pressure inside the containment, thereby 1.22 increasing the normal external pressure differential on the containment structure.

The analysis of maximum external differential pressure assumes 1.23 inadvertent actuation of the quench spray system caused by a single 1.24 spurious containment depressurization actuation (CDA) signal. Note 1.25 that two active component failures are necessary to generate a y ,, ,

spurious CDA signal.

The minimum internal pressure is determined by modeling an 1.26 inadvertent quench spray pump start using the computer program 1.27 CONTEMPT-LT (Aerojet 1975). The minimum internal pressure is 1.28 calculated to be 8.03 psia at the time the quench spray is assumed to be manually deactivated (10 minutes). The rate of containment 1.30 pressure decrease at 10 minutes is small. If the quench spray is 1.31 allowed to run beyond 10 minutes, it would require several additional 1.32 minutes for the pressure to drop from 8.03 psia to the minimum 1.33 containment design pressure of 8.00 psia. ,q.,3 The parameters used in calculating the minimum containment pressure 1.38 are shown in Table 6.2-78. Each parameter was selected as 1.39 representing the most severe allowable initial condition for minimizing pressure. As a result, the calculated pressure is a lower 1.40 .

bound estimate.

See Tables 6.2-2 and 6.2-3 for additional containment design 1.42 evaluation parameters.

6.2.1.1.3.6 Loss-of-Coolant Accident Results 1.44 The loss-of-coolant accident (LOCA) containment transient analysis 1.45 was performed using the LOCTIC computer code (Section 6.2.1.1.3.2) 1.46 for a spectrum of pipe break locations and sizes. 1.48 Amendment 14 6.2-12 July 1985

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u12179123rc14i 05/30/85 241 MNPS-3 FSAR The results of the peak containment pressure analysis are tabulated 1.49 in Table 6.2-4. The mass and energy release data used for this 1.50 analysis are given in Section 6.2.1.3. The initial containment 1.51 conditions which yield the highest peak calculated containment pressure are the maximum pressure, temperature, and relative 1.52 f

Amenchnent 14 6.2-12a July 1985

i u1217912sra141 05/23/85 240 MNPS-3 FSAR i

l rate test described in Section 6.2.6.1 adequately demonstrates the 1.9 leak tightness of the containment. 1.10 i An evaluation of in-leakage following a LOCA shows the containment 1.11 pressure to be effectively subatmospheric at -0.5 psig 30 days 1.12

- following the accident. The inleakage analysis is based on the 1.13 maximum specified out-leakage rate of 0.9 percent per day at j approximately 45 psig adjusted to the pressure differences determined 1.14 g ,, , , ,

i to be present following a LOCA.

! The maximum in-leakage rate to the subatmospheric containment during 1.15 j normal operation is approximately 14 scfm at 9.5 psia, the lowest 1.16

normal operating containment pressure. This corresponds to the 1.17

. out-leakage rate of 0.9 percent per day at 45 psig adjusted for the pressure differential and other important flow paramaters. 1.18 I The containment structure enclosure will be evacuated by the 1.19 i supplementary leak collection and release system (SLCRS) to slightly 1.20 negative pressure immediately following the design bases accident ,

1 initiation of the engineered safety features actuation system 1.21 '

1 (ESFAS). This will ensure all leakage from the primary containment 1.22 (0.9 percent per day) is passed through the high-efficiency j

particulate air (99 percent efficient) filters of the SLCRS prior to 1.24 release from the containment structure enclosure, engineered safety 1.25 l feature building, main steam valve building, hydrogen recombiner j . building or auxiliary building which are all connected to the SLCRS. 1.27

! This filtration will ensure the reduction of primarf leakage from 1.28 0.9 percent per day to less than 0.1 percent per day released to the 1.29 environment. The SLCRS will be tested prior to loading fuel to 1.30 i verify that a slightly negative pressure can be obtained and I maintained following an ESFAS actuation in the areas mentioned above. 1.31 This test will'be conducted again at each refueling or at intervals 1.32-not to exceed 18 months. Some leakage through piping systems may 1.33 i bypass the secondary containment. This leakage is limited to the 1.34 4 design leak rates through these piping systems. The bypass leakage 1.35 penetrations, identified in Table 6.2-65, are tested in accordance i with Section 6.2.6.3, and the combination of their leakage rates is 1.37 i compared with the maximum allowable rate (9 scfh). When the actual 1.38 l

leakage rate approaches this limit, corrective action will be taken. 3

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! 6.2.7 References for Section 6.2 1.40

) Aerojet Nuclear Company. CONTEMPT-LT -

A Computer Program for 1.41 j Predicting Containment Pressure-Temperature Response to a Loss-of- 1.43 //-/4 i Coolant Accident. ANCR-1219, June 1975. 1.44 ,

{ Aerojet Nuclear Company, 1976. RELAP4/ MOD 5: A Computer Program for 1.47 i Transient Thermal Hydraulic Analysis of Nuclear Reactors and Related j Systems. User's Manual Vol I-III, Report ANCR-NUREG-1335. 1.48 j American Nuclear Society (ANS) 1978. Decay Heat Power in Light Water 1.51 4 Reactors. ANS Standard, June 1, 1978, Revised September 1978. 1.52 t

i i Amendment 14 6.2-85 -

July 1985 f

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05/23/85 240 ul217912src141 MNPS-3 FSAR p

Atomics International Division Rockwell International. Test 1.54 Procedure - Hydrogen Analyzer Systems, No. N019DTP120003.

Baer, Robert L. (Office of Reactor Regulation Division of Project 1.55 Management, (USNRC) 1978. Letter to Mr. Gordan Pinsky (Owens-Corning 1.56 Fiberglass Corporation).

Bloom, G.R., et al. Hydrogen Distribution in a Containment with a 1.57 High Velocity Hydrogen-Steam Source. Presented at the Second 1.58 International Workshop on the Impact of Hydrogen on Water Reactor Safety, Albuquerque, New Mexico, October 3-7, 1982. 1.59 i '0' I f Brocard, D.N. Buoyancy, Transport and Head Loss of Fibrous Reactor 1.60

. Insulation. NUREG/CR-2982, U.S. Nuclear Regulatory Commission. 2.1 Prepared by Alden Research Laboratory, Worcester Polytechnic 2.2 Institute, Holden, Massachusetts. November 1982. 2.4 p %-ex The Mathematics of Diffusion. Oxford University Press, 2.6 Crank, J. '"'"

1956, pp 186-199.

Gido, R.G. Liner-Concrete Heat Transfer Study for Nuclear Power Plant 2.7 Containments, Los Alamos Scientific Laboratory, LA-7089-MS Informal 2.8 Report NRC-4, issued January 1978.

Gido, R.G. Subcompartment Analysis Procedures. Los Alamos Scientific 2.10 Laboratory. NUREG/CR-1199, LA-8169-MS, Informal Report R-4. 2.11 fo,5)

December 1979. 2.12 -

Hanover, Stephen H. (Chairman Advisory Committee of Reactor 2.13 Safeguards) 1969. Letter to Hon. Glenn T. Seaborg (Chairman USAEC) 2.14 Report on Brunswick Steam Electric Plant.

Hanover, Stephen H. (Chairman Advisory Committee of Reactor 2.15 Safeguards) 1969. Letter to Hon. Glenn T. Seaborg (Chairman USAEC) 2.16 Report on Edwin I. Hatch Nuclear Plant.

Hilliard, R.K., et al. 1970. Removal of Iodine and Particles from 2.18 Containment Atmosphere by Sprays. Battelle-Northwest, Richland, 2.19 Wash. BNWL-1244.

Hilliard, R.K. and Coleman, L.F. Natural Transport Effects on 2.21 Fission Product Behavior in the Containment Systems Experiment.

BNWL-1457, Battelle Pacific Northwest Laboratories, Richland, 2.22 ,

Washington. December 1970. 2.23 90'o 17 IDCOR Program Report, Technical Report 12.2, Hydrogen Distribution in 2.24 Reactor Containment Building. September 1983. 2.25 Idel'chik, I.E. 1960. Handbook of Hydraulic Resistance, Published 2.27 persuant to an agreement with the U.S. Atomic Energy Commission and the National Science Foundation, Washington, D.C. 2.28 l

Amendment 14 6.2-86 July 1985

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u1217912sra14k MNPS-3 FSAR l

TABLE 6.2-78 1.8 INPUT DATA FOR MINIMUM CONTAINMENT 1.10 PRESSURE ANALYSIS 1.11 Minimum initial air partial pressure 8.9 psia 1.14 Minimum initial containment temperature 120'F 1.16 Minimum RWST water temperature 40'F 1.18 Number of inadvertently activated 1 1.20 quench spray pumps 1.21 Maximum quench spray system flow 4,600 gpm 1.23 (one pump operating) 1.24 Outside air temperature O'F 1.26 Secondary containment temperature 28 F 1.28 Quench spray effectiveness 100% 1.30 Heat transfer regime modeled at Turbulent natural 1.32 the heat sinks convection 1.33 Amendment 14 1 cf 1 July 1985 l

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