ML20053E908
| ML20053E908 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 06/03/1982 |
| From: | James Shea Office of Nuclear Reactor Regulation |
| To: | Counsil W NORTHEAST NUCLEAR ENERGY CO. |
| References | |
| TASK-06-02.D, TASK-06-03, TASK-6-2.D, TASK-6-3, TASK-RR LSO5-82-06-002, LSO5-82-6-2, NUDOCS 8206100198 | |
| Download: ML20053E908 (61) | |
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i Docket No.80-245 LS05-82 002 Mr. W. G. Counsil, Vice President Nuclear Engineering and Operations Northeast Nuclear Energy Company Post Office Box 270 Hartford, Connecticut 06101
Dear Mr. Counsil:
SUBJECT:
SYSTEMATIC EVALUATION PROGRAM (SEP) FOR THE MILLSTONE 1 NUCLEAR POWER STATION UNIT NO. 1 - EVALUATION REPORT ON TOPICS VI-2.D AND VI-3 Enclosed is a copy of our draft 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." This evaluation compares your facility, as described in Docket No. 50-245, with the criteria currently used by the regulatory staff for licensing j
new facilities. Appendix A to our draft evaluation is a draft Tect.nical Evaluation Report from our contractor, Lawrence Livermort National Laboratory.
Please infom us if your as-built facility differs from the licensing basis assumed in our assessment. Coments are requeJted within 30 days of the receipt of this letter so that they may be considered in our final evaluation.
This evaluation will be a basic input to the integrated safety assessment for your facility unless you identify changes needed to reflect the as-built conditions at 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.
$$0Y Sincerely,
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James Shea, Project Manager
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Operating Reactors Branch No. 5 p. M s'a#O Division of Licensing oo Or 4
Enclosure:
SE Draft SEP Topics VI-2.D and VI-3 S0.4 cc w/ enclosure:
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t Mr. W. G. Counsil CC William H. Cuddy, Esquire State of Connecticut Day, Berry & Howard Offict of Policy & Management Counselors at Law ATTN:
Under Secretary Energy One Constitution Plaza Division Hartford, Connecticut 06103 80 Washington Street Hartford, Connecticut 06115 Ronald C. Haynes, Regional Administrator Nuclear Regulatory Commission Region I Office 631 Park Avenue King of Prussia, Pennsylvania.19406 Northeast Nuclear Energy Company ATTN:
Superintendent Millstone Plant P. O. Box 128 Waterford, Connecticut 06385 Mr. Richard T. Laudenat Manager, Generation Facilities Licensing Northeast Utilities Service Company P. O. Box 270 Hartford, Connecticut 06101 Resident Inspector c/o U. S. NRC P. O. Box Drawer KK Niantic, Connecticut 06357 Fftst-Selectman of the Town of Waterford Hall of Records 200 Boston Post Road Waterford, Connecticut 06385 John F. Opeka Systems Superintendent Northeast Utilities Service Company P. O. Box 270 Hartford, Connecticut 06101 U. S. Environmental Protection Agency Region I Office ATTN:
Regional Radiation Representative JFK Federal Building Boston, Massachusetts 02203
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1 SAFETY EVALUATION REPORT ON CONTAINMENT PRESSURE AND HEAT Re40 VAL CAPABILITY SEP TOPIC VI Ato MASS AND ENERGY RELEASE FOR POSSIBLE PIPE BREAK INSIDE CONTAINMENT, SEP TOPIC VI-2.0 FOR THE
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TABLE OF CONTENTS Page 1
I.
INTRODUCTION ---------
2 II. REVIEW CRITERI A -----------
3 III.RELATED SAFETY TOPICS ------
4 IV. REVIEW GUIDELINES 5
V.
EVALUATION -
7 VI. CONCLUSIONS --
APPENDIX A - SEP CONTAINPENT AN/ LYSIS AND EVALUATION OF A-1 MILLSTONE 1 POCLEAF POWER STATION A-2
1.0 INTRODUCTION
AND BACKGROLND --
2.0 CONTAItNENT FUNCTIONAL DESIGN A-3 2.1 Review of Millstone 1 Containment Design Analysis -
A-5 A-5 2.2 Review of Millstone 1 Primary System Pipe Breaks ---
A-7 3.0 ANALYSIS OF MILLSTONE 1 CDNTAINMENT DESIGN --
A-7 3.1 Recirculation Line Break -
3.1.1 Containment Response to a Recirculation A-8 Line Break -----
A-9 3.1.2 Containment Response Results A-9 3.2 Main Steam Line Pipe Breaks -
3.2.1 Containment Response to a MainsSteam Line A-12 Break --
3.2.2 Containment Response to Main Steam Line Break A-12 Results A-14
4.0 CONCLUSION
S -
APPENDIX B - CS&A - 125-81, J. D. Atchison letter to D. G. Vreeland, B-1 12/17/81 ----
LIST OF TABLES TABLE TITLE PAGE A-15 1
RECIRCULATION LINE BREAK ASSUMPTIONS -------------
2 DOUBLE-ENDED-GUILLOTINE RECIRCULATION LINE BREAK' 2
RELEASE RATE DATA (5.62 FT -8REAK DRESDEN 2 ----------
A-16 A-17 3
CDNTAINMENT MODEL INPUT DATA ----
A-18 4
MAIN STEAM LINE BREAK ASSUMPTIONS -
2 5
0.75-ft MAIN STEAM LIE BREAK BLOM)0WN - MILLSTONE 1 A-19 6
0.1-ft MAIN STEAM LINE BREAK-BLOWDOWN - MILLSTONE 1 ---
A-20 7
0.01-ft MAIN STEAM LINE BREAK - MILLSTONE 1 -
A-21 3
MILLSTONE 1 CONTAINMENT DESIGN CONDITIONS VERSUS A-22 CALCULATED ACCIDENT CONDITIONS -
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.a LIST OF FIGURES FIGURE TITLE PAGE I
Containment Pressure Response to a Double-Ended Guillotine Recirculation Line Break --
A-23 2
Containment Temperature Response to a Double-Ended-Guillotine Recirculation Line Break --------
A-24 2
3 Mass Flux vs. Total Depleted Energy for the 0.75 ft and 2
0.10 ft MSLB A-25 2
4 Revised Blowdown Data for the 0.10 ft MSLB ----
A-26 2
5 Drywell Pressure Respanse to a 0.75 ft MSLB ----
A-27 2
6 Drywell Atmosphere Temperature Response to a 0.75 ft MSLB -
A-28 2
7 Wetwell Pressure Response to a 0.75 ft MSLB A-29 2
8 Wetwell Atmosphere Temperature Response to a 0.75 ft MSLB -
A-30 2
9 Wetwell Pool Temperature Response to a 0.75 ft MSLB -----
A-31 2
10 Orywell Pressure Response to a 0.10 ft MSLB -----
A-32 2
11 Wetwell Atmosphere Temperature to a 0.10 ft MSLB --
A-33 2
12 Wetwell Pressure Response to a 0.10 ft MS W -----------
A-34 2
13 Wetwell Atmosphere Tenperature Response to a 0.10 ft MSLB -
A-35 2
14 Wetwell Pool Temperature Response to a 0.1J ft MSLB ---
A-36 2
15 Drywell Pressure Response to a 0.01 ft MSLB ---
A-37 2
16 Drywell Atmosphere Temperature Response to a 0.01 ft MSLB -
A-38 2
17 Wetwell Pressure Response to a 0.01 ft MSLB --
A-39 2
18 Wetwell Atmosphere Temperature Response to a 0.01 ft MSLB -
A-40 2
19 Wetwell Pool Temperature Response to a 0.01 ft MSLB A-41 s
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I.
INTRODUCTION The Wilstone 1 Nuclear Power Station began commercial operations in 1971.
Since then the staff's safety review criteria have changed.
As part of the Systematic Evaluation Program (SEP), the mass and energy release for possible pipe break inside containment (Topic VI-2.0) and the containment pressure and heat removal capability (Topic VI-3) have been re-evaluated.
The purpose of this re-evaluation is to document all deviations from current safety criteria as they relate to the containment pressure and heat removal
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capability and the mass and energy releas'5 for pos51ble pipe breaks inside containment.
Furthermore, independent analyses in accordance with current criteria were performed to determine the adequacy of the containment design bases (e.g., design pressure and temperature) and to provide input for SEP Topic III-12, Environmental Qualification of Safety-Related Equipment. The significance of the identified deviations, and recomended corrective measures to improve safety, will be the subject of a subsequent, integrated assessment of the Millstone 1 plant.
The SEP Analysis and Evaluation and plotted results are given in Appendix A.
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II. FEVIEW CRITERIA The review criteiia used in the cdrrent evaluation of SEP Topics VI-2.D and VI-3 for the, Millstone 1 plant are contained in the following documents:
(1) 10 7R Part 50, Appendix A, General Design Criteria (CDC) for Nuclear Power Plants:
(a)
GDC 16 - Containment design; (b) GDC 38 - Containment heat removal; and (c)
GDC 50 - Containment design basis.
(2) 10 ER Section 50.46, " Acceptance Criteria for Emergency Core Cooling System for Light Water Nuclear Power Reactors".
(3) 10 CFR Part 50, Appendix K, "ECCS Evaluation Models".
(4)
NUREG-0800, Standard Review Plan for the Rev'iei of Safety Analysis Reports for Nuclear Power Plants (SRP 6.2.1, Containment Functional Design).
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III.
RELATED SAFETY TOPICS The review areas identified below are not addressed in this report but are related to the SEP topics of mass and energy release for possible pipe breaks inside containment and/or containment pressure and heat removal capability.
(1)
III-1, Classification of Structures, Components, and Systems (Seismic and Quality).
(2)
III-12, Environmental Qualification of Safety-Related Equipment.
(3)
VI-7.8, ESF Switch-over from knjection to Recirculation Mode (Automatic ECCS Realignment).
(4).
IX-3, Station Service and Cooling Water Systems.
3 (5)
X, Auxiliary Feedwater System.
(6)
USI-A24, Qualification of Class 1E Safety-Related Equipment.
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IV.
REVIEW GUIDELINES General Design Criterion (GDC) 16 of Appendix A to 10 CFR Part 50 requires that a reactor containment and associated systems shall be provided to establish a leak-tight barrier against the uncontrolled release of radioactivity to the environment.
In addition, CDC 16 requires that the containment design conditions important to safety not be exceeded for as long as the postulated accident conditions require.
GCC 38 requires a containment heat removal system to be providedwhose safety. function shall be to reduce the containment pressure and temperature following any loss-of-coolant accident (LOCA) and to maintain them at acceptably low levels; furthermore, this safety system shall function with a single failure. GOC 50 requires that the containment structure and the containment heat removal system shall be designed so the structure can accommodate, with sufficient margin, the This calculated pressure and temperature conditions resulting from any LOCA.
margin was cbtained from the conservative calculation of mass and energy release, and the containment model is d'.scussed in the Standard Review Plan The containment design (SRP) Section 6.2.1, Containment Functional Design.-
basis includes the effects of stored and generated energy in the accident.
Calculations of the energy available for release should be performed in accordance with the requirements of 10 CFR Part 50, Section 50.46, and The Appendix K, paragraph I.A, and the conservatism specified in SRP 6.2.1.3.
mass and energy release to the containment from a LOCA should be considered in Break locations should terms of the mass and energy release during blowdown.
The review also include recirculation line breaks and steam line breaks.
includes the analysis' of postulated single active failures of components in the secondary system.
By review of the licensee's analysis, deviations from the current criteria are identified and independent analyses are performed,s as required, to evaluate the significance of these deviations.
The evaluation is completed by.
comparing the'results with the containment design bases.
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V.
EVALUATION In the case of BWRs with Mark I containments, it is necessary to evaluate the effect of pipe breaks below the core; for maximum containment pressure and pipe breaks above the core for maximun containment temperature.
In the Millstone 1 FSAR, a full double-ended-guillotine (DEG) recirculation line break was analyzed to determine the containment design pressure. The initial and boundary cond!tions used by the applicant were reviewed.
In exception to current design criteria, the FSAR analyses were performed at 100% reactor full power condition, not the 102% required in current criteria.
- However, later docket material in support of reload licensing shows reanalysis of the containment pressure based on 102% reactor full power conditions.
This is acceptable by current NRC criteria.
The FSAR maximum calculated contairinent pressure is 43 psig in response to the design basis accident (DBA) LOCA break. This is well below the containment design pressure of 62 psig. A confirmatory analysis was performed and is presented in Appendix A of this report. The confirmator'y analysis response was calculated using CDNTEMPT-LT/028. The calculated maximum containment pressure was 44 psig.
Hence, the confirmatory analysis agrees with the FSAR calculated maximun containment pressure due to a DBA LOCA, and both values are below the containment design pressure of 62 psig.
U The Millstone 1 FSAR gives a containmer.t design temperature of 281 F.
The confirmatory containnent analysis in response to a DSA LOCA gives a calculated maximun drywell temperature of 291 F.
The design temperature is exceeded for approximately 20 seconds during the calculated transient.
However, it must be kept in mind that drywell structure will lag the temperature response of the atmosphere.
As a result, the structural response will not exceed the design temperature of 281 F.
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VI.
CONQ.USIONS The deviations of'the Millstone 1 plant' from current criteria have been identified in Section V, above. From the independent contairment analyses reported in Appendix A, it is concluded tnat the, Millstone 1 contalment design pressure meets current ' criteria. It is evident from the three MSLS cases analyzed that the containment design temperature is exceeded by postulated medlun size main steam line break accidents.
In the related matter of containment e@lpment qualification (SEP Tcpic III-12), the Millstone 1 FSAR specifies a maximum emergency environment of two 0
hours at 320 F and a long term emergency environment of ten days at 2
281 F.
The calculated temperature re:ponse for a 0.1-ft MSLB exceeds the 320 F value (by 8 F at the peak) for approximately eight minutes.
This is not significant when the overall qualification envelopes of 320 F for two i
hours is considered.
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me APPENDIX A SEP CONTAINMENT ANALYSIS AND EVALUATION OF 1
MILLSTONE 1 NUCLEAR POWER STATION' 4
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l.0 INTRODUCTION AND BACKGROUND As'part of the Systematic Evaluation Program (SEP), the containment functional design capability of the Millstone 1 Nuclear Power Station has been reevaluated.
The purpose of this report is to document the resolution of SEP Safety Topic VI-2.0, Mass and Energy Release for Possible Pipe Break Inside Containment, and Safety Topic VI-3, Containment Pressure and Heat Removal Capability, and deviations from current safety criteria as they relate to the containment functional design.
The significance of the identified deviations and recommended corrective measures will be the subject of a subsequent integrated assessment of the Millstone 1 plant..
The containment structure encloses the reactor system and is the final barrier against the release of radioactive fission products in the event of an accident.
The containment structure must, therefore, be capable of withstanding, without loss of function, the pressure and temperature conditions resulting from postulated LOCA' and steam line break accidents.
Furthermore, equipment having post-accident safety functions must be environmentally qualified for the resulting adverse pressure and temperature conditions.
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2.0 CONTAINMENT FUNCTIONAL-DESIGN > !.
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Millstone 1 is a 2011 MWt Genera'l Electric Mark' I BWij which hasDa primary y
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The pressure suppression chamber Jis a steel
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The chamber is approximately half filled with water.
The veat k[
system from the drywell terminates belo* the water levein the pressure q,7 suppression chamber, so that in the event of a pipe failure in tb drywell, the released steam passes directly to the water, where it ds coni?nsed.
This transfer of energy to the water poo.L rapidly reduces th6 port-accident pressure in the drywell and substantially reduces'the potsntial for subsequent leakage from the primary containment.
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In addition to the pressure absorption chamber, independe,at auxillary cooling systems are provided for the reactor and containment cooling under various normal and abnormal conditions.
These are:
1 (1)
A low pressure coolant injection (LPCI) containment cooling system which serves three functions:
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(a) To inject water into the reactor vessel 1 subsequent to a
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(b) To remove heat from the water in the suppression chambe't;,
(c) To spray water into the drywell and/or the suppression chamber as an augmented means of removing energy sfrom the cogtainment as t
required.
(2)
A shutdown cooling system to renove reactor decay heati during shutdown.
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An isolation condenser to remove decay heat from the core when the reactor is isolated.
(4)
A feedwater coolant injection (FWCI) system to remove decay heat and to provide coolant inventory control and heat dissipation from the core to the suppression chamber under postulated small break accidents.
If the FWCI system should fail to operate, an autcmatic depressurization by blowdown will be employed through ' automatic opening of relief valves which vent steam to the suppression pool.
This blowdown will depressurize the vessel in sufficient time to allow the core spray or the,1,0w pressure coolant injection (LPCI) function of the ECCS to adequately cool the core and prevent any clad melting.
(5)
Two core spray systems designed to pump water under accident conditions from the pressure suppression chamber pool directly to the reactor core by indq]endent spray headers or spargers mounted in the reactor vessel above the core.
(6)
An auxiliary coolant supply system via a cross-tie between the service water system and the condensate storage system which makes available an inexhaustable supply of. cooling water from the Long Island Sound to the reactor core and containment independent of all other cooling water sources.
In the event of loss of offsite power and failure of one diesel generator,.
minimum containment cooling is provided by three low pressure coolant injection pumps.
After ths core is flooded, these pumps are manually switched from a core injection mode to a containment cooling mode.
Water from tne wetwell is passed through a heat exchanger and retprned to.the wetwell.
A containment spray system is provided to spray cooled water to either the drywell, the wetwell, or to both.
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2.1 Review of the Mill <, tone 1 Containment Desion Analysis n
First is Two separate calculations make up the containment design analysis.
This the mass and energy release rate calculation for postulated LOCAs.
provides the time-dependent mass and energy input into the containment Second is the calculation of containment response to the mass and structure.
energy input to the containment structure.
This results in the time-dependent containment temperature and pressure profile. The severity of the contalment response depends on the magnitude and nature of the break location.
If the break is below the core, the break flow will initially be single-phase liquid.
This results in a rapid blowdown of the mass and energy release to the containment at a relatively low enthalpy.
If the break is above the core, the break flow will be mostly single-phase steam.
This results in a much longer blowdown of the mass and ' energy release to the containment at a much higher enthalpy. Because of these effects, breaks below the core produce the most severe pressure responses in the contalment, and steam line breaks above tne core produce the most severe temperature responses.
The acceptance criteria used to evaluate the Millstone 1 Containment Design Analysis were based on the Standard Review Plan (SRP), Sections 6.2.1.1.C, 6.2.1.3, and 6.2.1.4.
For the containment design analysis to be found acceptable, both the mass and energy release and the containment response calculations must meet the acceptance criteria specified in the SRP.
2.2 Review of Millstone 1 Primary System Ploe Breaks The SRP specifies several acceptance criteria to be applied to the mass and energy re' lease analysis for primary system pipe breaks.
Among these are the break location.
In the Millstone 1 FSAR, the most severe mass and energy release ute :alculated for containment design was'done assuming a double-ended recirculation line break.
The input and boundary conditions used in this analysis are in accordance with current criteria except for initial reactor power.
The FSAR 03A LOCA mass and energy release eclculation assumes the initial reactor power at 10CM full power, not 102% as required by current criteria.
However, later docket material documents an analysis of 102% of fu?1 power initial conditions, which is acceptable.
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The FSAR calculated peak post-accident containment pressure resulting from a double-ended recirculation line break is 43 psig.
The peak wetwell pressure 0
was 25 psig.
The peak drywell temperature was 291 F.
In addition to recirculation line breaks, the current criteria state that steam line breaks above the core must be considered.
The' licensee has not performed a steam line break analysis specifically for containment design evaluation.
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3.0 ANALYSIS OF MILLSTONE 1 CONTAINMENT DESIO4 The lacirculation line break resul'ts in' the limiting condition for calculating the peak pressure inside the containment. The steam line pipe break analysis is the most limiting case for temperature conditions inside the containment.
I Both of these analyses were performed.
3.1 Recirculation Line Break There are no reactor coolant system (RCS) blowdown decks available for the l
Millstone 1 plant. The specific irput. decks for reactor blowdown analysis are
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proprietary to the General Electric Company and are not available in the open I
literature. A search of the Millstone 1 docket material did not provide any documented LOCA mass and energy release data.
In this situation, it was decided to cpply ths LDCA blowdown data from the most similar BWR plant as the best engineering estimate.
A literature search concluded that the best available LOCA blowdown data would be from the Dresden 2 plant, which was also under Inview for SEP.
Refer to Appendix B for more details pertaining to the 4
selection of the Dresden 2 LOCA blowdown.
The Dresden 2 RCS design is nearly identical to that of Millstone 1.
The full 2
DEG recirculation line break area is 5.62.ft for Dresden 2 versus 5.82 2
ft for Millstone 1, assuming the equalizer line valve is open for both plants.
The initial reactor pressure is 1000 psig at Dresden 2 versus 1035.
i psig at Millstone 1.
Hence, the saturated liquid enthalpy. during the initial recirculation line blowdown will differ by 5 Btu /lbm, which is less than a 1%
' difference. Due to its larger physical size and higher power rating, the Dresden 2 reactor coolant. system mass and energy inventory at full power is approximately 25% larger than the Millstone 1 RCS mass and energy inventory.
Therefwe, the application of the Dresden 2 LOCA blowdown to the Millstone 1 containment model will provide a conservatively hight. calculated containment -
response. Table 1 lists the recirculation line break assumptions that were used in the Dresden 2 SEP analysis (Re: Menorandum T. Spets to l
G. Lainas, December 2,1981, Dresden 2 TER, SEP Topics VI,2-D and VI-3).
These assumptions remain valid for Millstone 1.
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The resulting recirculation line break mass and energy release rates are shown in Table 2.
3.1.1 Containment Response Calculation to a Recirculation Line Break The irput data for the containment response calculation consists of the mass and energy release to the contairment, a description of the containment geometry, heat removal systems, and containment heat sink data.
The mass and energy release rate data used were taken from the blowdown of the recirculation line described in the pr.evious section.
The contairment heat removal system consists of a pressure suppression pool, LPCI containment cooling subsystem, containment sprays, and contairnent fan coolers. For this analysis, the containment fan coolers will not operate due to the assumed loss of auxiliary ac power. The contairment sprays must be manually activated by the operator and therefore, conservatively will not be accounted for in this analysis.
The LPCI contairment cooling subsystem will be assumed to be switched from core flooding made to wetwell cooling mode at 600 seconds.
This switchover requires operator action.
The pressure stppression pool and vent and downcomer codel information was taken from the Millstone 1 FSAR and subsequent docket material.
No Millstone 1 containment heat sink data were available in the coen literature or in docket material.
Therefore, Oyster Creek containment heat sink data were used as the best available substitutes.
The containment response calculation was performed with the CONTEWT-LT/028 computer code.
The nede model is composed of three regions: reactor vessel, 3
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drywell, and wetwell.
The geometric descriptions and initial boundary conditions were obtained from the 14111 stone 1 FSAR.
A summary of the contalment input model characteristics are given in Table 3.
3.1.2 Containment Resconse Results The containment pressure and temperature responses to a recirculation line break are shown in Figures 1 and 2.
The calculated transient reflects a peak post-accident containment drywell pressure of 44 psig and a temperature of 291 F.
The peak containment wetwell pressure and temperature are 14 psig 0
and 144 F.
The containment desigrtpressure for the drywell and the wetwell is 62 psig.
There is, therefore, a substantial margin between the peak calculated pressure and the containment design pressure.
3.2 Main Steam Line Pipe Breaks, Analyses of the containment response to various steam line breaks were Break sizes of performed to reveal the most severe temperature co'ndition.
2 2
2 0.01 ft, 0.10 ft and 0.75 ft were examined to identify the most limiting steam line break. Blowdown data consisting of mass and energy These release rates were provided by Northeast Utilities, for these breaks.
data were developed by General Electric based on their licensirg code as part of the equipment qualification effort.
(Ref.: Letter, W. B. Counsil, Northeast Utilities to H. R. Denton, NRR, June 9,1981). The blowdown The staff calculation was performed using the assumptions given in Table 4.
has reviewed the assumptions listed in Table 4 and agrees with them with the exception of the manual actuation of the ADS valves at 600 seconds. Also, the utility has indicated to the NRC that the Low Pressure Core Spray (LPCS)
The injection was assumed to be constant at full flow once it was initiated.
result of these two actions was that the resulting blowdown exhibited a switchover from the steam to the liquid phase.
The time that this switchover 2
occurred ranged from about 2007 seconds for the 0.01 ft break to' 900 2
seconds for the 0.75 ft break.
The switchover is accompanied by a corresponding drop in enthalpy of water as it goes from steam to liquid.
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Although this switchover is possible in the accident sequence it appeared more plausible to the NRC that operator action would be taken beyond the 10 minute time limit into the transient to avoid the condition of having solid water above the level of the main stcam line.
It seemed appropriate to take into account the operator action which would likely be to. cycle the LPCS pump to control the water level in the vessel.
The resulting blowdowri, then, would consist of pure steam or a two phase mixture.
Rather than try to determine the relative percentages of steam and liquid in the blowdown it was decided to conservatively specify all the blowdown to be steam with its accompanying enthalpy in the vicinity of 1200 8tu/lbm.
Based on the considerations centioned and after examining the utility transmittals it was felt that enough information nad been submitted to perform the containment response calculations without deriving new blowdown output from the RsLAP code.
The blowdown data was moffled to eliminate the crossover from steam to liquid.
The initial valu;s of flow rate and enthalpy 2
2 were taken, as is, up until ADS actuation.
For the 0.75 ft and 0.01 ft break the blowdown values were held constant from this point on through the 2
remainder of the transient.
For the 0.10 ft break, the blowdown was 2
correlated to the 0.75 ft break after that break flow was adjusted.
2 0.75 ft Main Steam Line Break The blowdown data submitted by Northeast Utilities was revised to eliminate the switchover from pure steam to liquid at 900 seconds.
The revision consisted of simply maintaining the blowdown values of vapor mass flow rate, 60.19 lb/sec, and enthalpy, 1169 Btu /lb constant from 816.3 seconds to 3000 seconds.
Table 5 contains the revised data. The effect of this modification compared with the original utility data is marginal since the peak temperature is the same for both sets of data.
2 0.10 ft Main' Steam Line Break 2
The blowdown curve for the 0.10 ft break exhibits a sharp change in slope in the time frame immediately following 600 seconds, the point at which the A-10 i
automatic depressurization system valves actuate.
The ADS actuation causes an increase in the rate of reactor vessel depressurization which manifests itself by a decrease in vapor flow rates.'
To arrive at a modified version of this blowdown it was decided to correlate 2
with the 0.75 ft break the relationship between the integral of the product ~
of steam flow rate and enthalpy over time versus the mass flux, i.e.,
P m H dt vs. m/ Abreak i.e.,
TOTAL DEPLETED ENERGY vs. MASS FLUX steam flow rate, lbm/second where m
=
enthalpy, Btu /lbm H
=
2 b
mass flux, lbm/second - ft
=
A 2
break (gbreak = cross sectional area of break, ft )
First the curve was plotted, in Figure 3, using the revised data for the 0.75 2
2 ft MSLB given in Table 5.
Then the data was plotted for the 0.10 ft MSLB through 611.4 seconds and extended graphically so that it asymptotically approached the curve generated from the 0.75 ft break.
This was done as an assumption that both breaks would converge.
From this curve, representative points were selected and from them the values of mass flow rate and tLne were determined.
The revised blowdown is shown in Figure 4 and compared with the
' original.
For example, point 1 on Figure 3, indicates that the mass flux 2
equals 1031.lb/ft seconds.
Thus m = 103 lb/sec.
(change in total energy depleted)/(m)(h) = 185.1 sec.
AT
=
~
where H = enthalpy, 1205 Stu/lbm T total = 611.4 + 185.1 = 796.5 sec.
Table 6 gives the revised blowdown data and Figure 4 shows the revised blowdown curve.
A-11 4=.*ws m e em a uw
2 As with the 0.75 ft break, the effects on containment temperature of this modifications compared to the utility deck is marginal.
The peak temperature is the same for both sets of data.
2 0.01 ft Main Steam Line Break The data submitted by Northeast Utilities at time t = 0 consists of pure steam with a flow rate of 21.16 lb/sec and an energy content (enthalpy) of 1191 Btu /lbm.
The revised data contained in Table 7 simply consists of maintaining the blowdown at these levels for the duration of the transient.
The effect of this change was to raise the peak temperature up to about 300 F and shift the peak to 3000 seconds.
The utility data yielded a peak temperature of 261 F at 2000 seconds. By comparison with the Oyster Creek nuclear power 2
plant, a 0.01 ft MSLB break was examined by LLNL and the peak temperature 0
was determined to be 310 F.
This break was also examined by LLNL for 0
Dresden 2 and the peak temperature was determined to be 328 F.
3.2.1 Containment Response to a Main Steam Line Break The containment response to a Main Steam Line Braak was calculated using the CONTEMPT-LT/028 computer code.
The input data comprised the same geometric containment model as that used to analyze the prev'ious LOCA break described in 2
Section 3.1.1.
The three MSLB cases of 0.75, 0.1, 0.01 ft were analyzed using the blowdown data shown in Tables 5, 6, and 7.
In each analysis, when blowdown input ends, the mass and energy release rate to the drywell is calculated by CONTEMPT based on ECCS injection and 120 percent of ANS standard for core decay heat. The containment response analysis assumes a loss of auxiliary ac power and both FWCI and LPCI failure. One core spray is assumed to operate.
No credit is taken for the manually initiated containment sprays.
v 3.2.2 Containment Response to Main Steam Line Break Results The calculated pressure and temperature responses to the postulated main steam line breaks are shown in Figures 5 through 19.
The calculated containment A-12
4 9
2 response to the.75 ft MSLB shows a maximum drywell pressure of 19.2 psig 0
and a maximum temperature of 325 F.
The wetwell maximum pressure and temperature are 17.1 psig and 157 F.
2 The calculated containment response to the 0.1 ft MSLS shows a maximum 0
drywell pressure of.17.6 psig and maximum temperature of 328 F.
The wetwell maximum pressure and' temperature are 15.5 psig and 125 F.
The calculated containment response to the 0.01 ft MSLB shows a maximum 0
drywell pressure of 16.7 psig and maximum temperature of 300 F.
The wetwell maximum pressure and temperature are 14.7 psig and 111 F.
As the drywell and wetwell design pressures for Millstone 1 are 62 psig, the results above confirm that the containment pressures due to steam line breaks are substantially below the design pressures.
The maximum post-accident 2
drywell temperature was approximately the same for the 0.1 ft and the 0.75 2
ft MSL8s.
The design temperature of the Millstone 1 containment is U
2 2
281 F.
Both the 0.1 ft and 0.75 ft MSLB cases exceed the design tenperature.
O e
A-13 e
~-
4.0 CDNCLUSIONS Based on the review of the Millstone 1 docket material and the above discussed zeanalysis, it is concluded that the Millstone 1 contairunent design pressure meets all current NRC criteria. The containment, atmosphere temperature profile as the result of a 0.1-ft MSLB exceeds the containment design tempera ture. The drywell atmosphere reaches a maximum temperature of 328*F which exceeds the containment structure design temperature of 281 F.
S O
G 4
e e
A-14 e
~
TME l RECIRCULATIOR LINE BREAK ASSUMPTIONS BASED ON DRESDEN 2 DEG BREAK i
- 1. Reactor initial condition is at 102% of full power.
- 2. Recirculation pump suction line instantly separates to a full double-ended 2
guillotine break. Break area = 5.62 ft,
- 3. Equalizer line is open.
- 4. Loss of offsite ac power and diesel generator.
- 5. Main steam isolation valves start closing at 0.5 second and are fully closed within 3 seconds.
- 6. Feedwater flow stopped at time of accident.
- 7. Mass discharge through the broken pipe calculated using Moody critical flow with a multiplier of 1.0.
- 8. All reactor vessel mass is discharged through the break.
9.
Reactor scrams at time zero.
4 O
4 4
A-15 4
-~
~
a
TABLE 2 4
00UBLE-ENDED-GUILLOTINE RECIRCULATION LINE BREAK RELEASE DATA (5.62 ft2 BREM)
Time Flow Energy (seconds)
(lbm/sec)
(8tu/lbm) 0.0 27211.
552.
1.0 27211.
552.
i 2.0 27211.
553.
3.0 27211.
553.
4.0 27211.
554.
5.0 27211.
556.
10.0 27211.
562.
15.0 27218.
572.
20.0 19412.
591.
25.0 3012.
570.
30.0 1659.
711.
35.0 1037.
1198.
45.0 516.
1189.
55.0 291.
1180.
OECAY HEAT at 55 seconds (1.2 ANS)
D e
9 S
A-16 e
5-c e
~
w
+
TABLE 3 CONTAIN!ENT MODEL INPUT DATA (TAKEN FROM MILLSTONE 1.FSAR)
Drywell/Wetwell Data Drywell Wetwell 3
Free Air Volume (ft )
146,900.0 114,600.0 3
Initial Pool Water Volume (ft )
0.0 94,000.0 Initial Temperature of Atmosphere (OF) 150.0 95.0 Initial Temperature of Pool (DF) 150.0 95.0 Initial Pressure (psia) 15.7 14.7 Relative Humidity 1.0 1.0 2
Pool Surface Area (ft )
1271.8 9151.4 Vent System Vent Pipes Number 8
Internal Diameter
- 6 ft. 9 in.
Vent Tubes Flow Area, Total 286.3 ft2 Downcomer Pipes Number 96 Internal Diameter 2 ft.
Submergence Below Absorption Pool Water Level 4 ft. 9 in.
l l
A-17
TABLE 4 MAIN STEAM LITE BREAK ASSUMPTIONS 1.
Scram at time zero.
2.
FWCI failure.
3.
No credit for isolation condenser.
4.
No LPCI operation.
5.
One of two core sprays operate.
6.
Feedwater controller maintains normal reactor vessel water level until manually shut off at 600 seconds.
7.
Reactor initially at 102". of full power.
8.
Steam line isolation valves close within 3.5 seconds of break.
9 4
e e
e e
G A-18 4
TABLE 5 4
0.75 ft2 Main Steam Line Break Blowdown TIME (s)
MASS FLOW RATE (1bm/s)
ENTHALPY (BTU /lbm) 0.0 1587.0 1191.0 32.7 1247.0 1199.0 83.3 662.4 1205.0 163.1 385.3 1201.0 248.4 238.3 1195.0 328.0 172.3 1189.0 405.1 135.1 1185.0 488.4 109.6 1181.0 576.7 91.9 1177.0 660.6 77.67 1174.0 738.5 67.53 1171.0 816.3 60.19 1169.0 3000.0 60.19 1169.0 c
6 i
4 e
A-19
TABLE 6 2
0.10 ft MAIN STEAM LINE BREAK BLOWOOWN TIME (s)
MASS FLOW RATE (1bm/s)
ENTHALPY (BTU /lbm)
.- 0.0 211.6 1191 73.6 200.5 1193 144.4 177.2 1197 216.2 169.0 1198
'291.7 148.3 1201 376.2 141.5 1202 454.7 124.7 1203
^
535.7 120.3 1204 611.4 103.1 1205 796.5 90.0 1200 981.7 77.0
~
1200 1198.1 70,0 1200 1317.1 50.0 1200 1900.0 40.0 1200 2316.7 30.0 1200 2816.7 24.0 1200 3233.0 19.0 1200 s
A-20
^
4 4
TABLE 7 1
2 0.01 - ft MAIN STEAM LINE BREAK BLOWOOWN TIME (sec)
MASS FLOW RATE (lbm/s)
ENTHALPY (BTU /lbm) 0.0 21.16 1191 i
5000.0 21.16 1191 q
e
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TABLE 8 J
MILLSTONE 1 CDNTAINMENT DESIGN CONDITIONS VERSUS CALCULATED ACCIDENT CONDITIONS l
I Event Contairrnent Design Calculated l
DBA LOCA 62 psig and 2810F 44 psig and 2910F 0.01 ft2 MStB 62psigand281k 16.7 psig and 3000F 0.1 ft2 MSLB 62 psig and 281aF 17.6 psig and 3280F 0.75 ft2 MSLB 62 psig and 281oF 19.2 psig and 3250F s
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APPO4 DIX B CS&A-125-81, J. D. Atchison Letter to D. G. Vreeland, 2/17/81 5
b B-1
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ENERGY HNCORPORATED REF: CS&A-125-81 February 17, 1981 Mr. David Vreeland, L-90 University of California Lawrence Livermore Laboratory P.O. Box 808 Livermore, CA 94550
SUBJECT:
SEP Containment Analysis, Recommended Action on Millstone I
Dear Dave:
The purpose of this letter is to discuss the possible methods of obtain-ing the necessary pipe rupture blowdown data for Millstone I to be used for contaiment analysis.
The following recommendations are different than those proposed in the telephone discussion of 2/10/81 involving G. R. Sawtelle, J. D. Atchison, and D. G. Vreeland.
Since that time, further reference research and discussion concerning the Millstone I pl ant have led to a new conclusion.
The associated background and reasons for this decision are given below.
Of all the plants to be analyzed in the SEP Contaiment Analysis, only Millstone I does not have an available RELAP deck for blowdown calcula-tions.
The two alternative solutions to this problem are to either create a Millstone I RELAP deck from scratch or use a RELAP deck from a nearly identical BWR system plant that would be available to us.
The first alternative is not desirable given the scope of this project and budget constraints.
Therefore, an attempt was made to find a nearly identical BWR sister plant to Millstone I that would have an available RELAP deck.
The search narrowed down the choices to two plants, Oyster Creek and l
Dresden II.
Both plants are included in the SEP Containment Analysis uork, so the RELAP blowdown decks are immediately available.
Plant l
parameters comparing Millstone I, Oyster Creek, sand Dresden II are shown in Table I.
I Based on available data for BWRs and available RELAP decks for blowdown calculations, Oyster Creek is the best match for rated thermal power and reactor coolant inventory.
The main problem is that Oyster Creek has a.
lightly smaller reactor vessel than Millstone I and it is a nonjet pump HEADOUARTERS. ONE ENERGY OR, P. O. BOX 736. IDAHO FALLS. IDAHO 63401 (208) 5291000. TWX: 910-978-5979 ENERGYINC IDAH OFFICES IN WASHINGTON D C.
. RICHMOND. VIRGINIA. SEATTLE, % ASHINGTON AND ALOUQUEROUE NEW MEXICO 57-8 l
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Mr. David Vreeland February 17, 1981 plant. The major change that would have to be implemented to the Oyster Creek deck is the renoding of the recirculation loops and inclusion of
- the jet pumps.
This would be necessary in order to obtain the correct recirculation line break flow areas and blowdown rates.
The same would have to be done to the main steam lines for the steam line break case as the two plants have different main steam configurations.
However, at the present time we have on hand none of the information required to compare the Oyster Creek and Millstone water inventories at normal operating conditions.
This information along with the downcomer flow areas is critical ta the correct blowdown response.
The RELAP deck would then have to be debugged to ensure proper results.
There is extremely limited information available in the Millstone I FSAR as is needed to ensure the validity of the Oyster Creek deck modifications.
After contacting NUSCO, it is not known at this time if the updated information required can be obtained in a timely manner.
Therefore, this alternative will not be pursued any further.
The other choice is to use the blowdown data already being generated for the Dresden II plant as part of this SEP analysis.
Millstone I and Dresden II are the same generation jet pump BWR plants and have nearly identical configurations and safety features.
Dresden II is a larger plant at 800 MWe versus Millstone I at 560 MWe.
It is proposed that the best course of action is to apply the Dresden II blowdown to the Mill-stone I containment analysis.
The following reasons substantiate this action along with the comparison data of Table I:
(1)
The resulting Millstone I containment analysis will definitely be conservative.
Whereas with the other option of using the Oyster Creek deck, the uncertainties involved were not on the conservative side.
(2) Dresden's drywell free air volume is only 8% larger-than Millstone.
Both plants have the same vent pipe flow area and configuration.
l (3)
Fran FSAR infomation, the design basis accidents (double-ended recirculation line rupture) for both plants have the l
l same sequence of events following the rupture.
Both plants use the same ECCS equipment to control the accident, namely LPCI and core soray.
The two important setpoints in this analysis are high drywell pressure and low-low water level.
These setpoints are compared in Table *I.
The ECCS systems for both plants are shown in Table II.
There are slight differ-i ences in the two plants' systens.
(4)
For the double-ended steam line break inside the drywell, both plants have the same sequence of events following +he pipe rupture.
The important setpoints in this analysis are high a
steam fl ow, MSIV closure scram setpoint, and high drywell pressure.
These are shown in Table I.
Both plants use the i
57-8 3
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Mr. David Vreeland February 17, 1981 assumption of 10.5 seconds for complete MSIV closure.
Both plants use core spray and LPCI to control the accident.
(5) Dresden's FSAR peak pressure response to the design basis accident is approximately 10% higher than Millstone.
Mill-stone's FSAR peak pressure response is approximately 30% under its design value.
Taki ng the above factors into consid-eration, using the Dresden II SEP blowdown applied to the Millstone I containment should give a Millstone peak pressure response that is within the design values and is also very conservatively calculated.
(6) Using the Dresden II blowdown will save several weeks of time and expense that would be needed for the other alternative to modify the Oyster Creek RELAP deck, debug it, and run the blowdown cases.
The only drawback at this time concerning the recommended method is the lack of FSAR design basis accident mass and energy release data for both Millstone I and Dresden II.
If this were available, it would confirm the validity of the recommended approach.
However, based on engineering judgment and the deficiencies of the other alternatives, the recommended course of action will give the most valid and conservative results within the framework and objectives of this project.
Based on this letter and our recent telephone conversations on this subject, work will proceed by applying the Dresden II blowdown data to the Millstone I containment.
The contents of this letter show this to be an acceptable methodology.
Very truly yours, A
n.
w h-
//(eN =,s
~
John D. Atchison Engineer
.JDA:db Enclosures l
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e 57-8
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TABLE I BWR PLANT PARA!ETERS*
Parameter Millstone I Oyster Creek Dresden II Thermal Power (MW) 2011 1600 (now 2527 uprated to 1930MW)
Operating Pressure (psig) 1000 1000 1000 6
6 6
Recirculation Flow (lb/hr) 69 x 10 61 x 10 98 x 10 6
6 6
Steam F1ow (1b/hr) 7.94 x 10 5.85 x 10 9.945 x 10 Circumscribed Core Dia.
177.1 in.
170.55 in.
189.7 in.
2 2
2 Heat Transfer Surface Area 50,796 ft 49,137 ft 62,640 ft 2
Average Heat Flux (Stu/hr-ft )
129,640 107,470 131,860 2
Max. Heat Flux (Btu /hr-ft )
310,000 295,600 312,800 Core Subcooling (Btu /lb)
'23.5 25.7 22.4 Core Ave. Void Fraction 38.9%
32%
29.9%
Core Ave. Exit Quality 13.1%
9.8%
10.1%
Volume Ratio (cold) 2.41 2.38 2.41 Water /UO2 Number of Fuel Assemblies 580 560 724 Fuel Rod Array 7x7 7x7 7x7 Fuel Rod Pitch (inch)
.738
.738
.738 Fuel Pellet 0.D. (inch)
.488
.488
.488 Clad Thickness (inch)
.0355
$.h355
.032 Clad 0.D. (1..:h)
.570
.570
.563 Active Fuel Length (inch) 144 144 144 i
Number of Control Rods 145 137 177 Core Equivalent ~ Dia. (inch) 163.1 160.21 182.2 l
57-8 t
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TABLE I (continued)
Pa rameter Millstone I Oyster Creek Dresden II Reactor Vessel ID 18'8" 17'9" 20'11" Reactor Vessel Height 64'8" 63'10" 68'7" Number of Recir. Loops 2
5 2
Recir. Pipe Size 28" 26" 28" Number of Jet Pumps 20 none 20 Type of Primary Containment pressure pressure pressure suppression suppression suppression Drywell Cylindrical Dia.
34'2" 33' 37' Drywell Spherical Dia.
64' 70' 66' 3
3 3
i 146,900 ft 180,000 f t 158,236 ft Drywell Free Air Volume Number Vent Pipes 8
10 8
Vent Pipe 10 6'9" 6'6" 6'9" 2
2 2
Vent Tubes Total Flow Area 286.3 ft -
331.9 ft 285 ft Vent Header ID 4'9" 4'7" 4'10" Number of Downcomer Pipes 96 120 96 Downcomer Pipe ID 2'
l'11.5" 2'
3 Wetwell Water Volume 98,700 ft 83,400 112,203 max.
3 83,500 ft min.
3 Wetwell Free Air Volume 125,100 ft 127,000 117,245 max.
3 109,900 ft mi n.
57-8 t
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TABLE I (continued) 1 Pa rameter Millstone I Oyster Creek Dresden II 4
Torus ID 29'6" 30' 30' Torus Major Diameter 102' 101' 109' Number Main Steam Lines 4
2 4
Main Steam Line Dia.
20" 24" 20" Protection System Setpoints Millstone I Oyster Creek Dresden II Reactor High Pressure 1085 psig 1050 psig 1070 psia Reactor Low Level 2" above l' below 1" above bottom of
- normal bottom of separator separator Reactor Low-Lm/ Level 49" below 5' below 59" below bottom of normal bottom of separator separator High Neutron Flux 120% rated 120% rated 120% rated power power power Drywell High Pressure 2 psig 2 psig 2 psig Scram on MSIV % closure 10%
10%
10%
Main Steam High Flow 120% rated 120% rated flow flow
- All values are taken from the respective plant's (SAR and Tech. Spec.
The information may not be up to date, especially Oyster Creek which has been uprated in power.
l 1-57-8
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TABLE II EMERGENCY COOL COOLING SYSTEMS Design Function - Plant Number of Pumps Coolant Flow
- Pressure Range
- Core Spray Millstone I 2 (100% each) 3600 gpm 0 90 psi 245 to O psi Dresden II 2 (100% each) 4500 gpm 0 90 psi 260 to O psi LPCI Millstone I 4 (33% each) 7500 gpm 0165 psi 235 to 0 psi 15,000 gpm'0 0 psi Dresden II 4 (33% each) 8000 gpm 0 200 psi 275 to O psi 14,500 gpm @ 20 psi FWCI Millstone I 3 (100% each) 8000 gpm 1125 to 100 psig HPCI Dresden II 1 (100%)
5600 gpm 1125 to 150 psig i
- RV internal pressure to drywell pressure differential.
s' 57-8
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