ML20038D126
| ML20038D126 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 12/10/1981 |
| From: | Crutchfield D Office of Nuclear Reactor Regulation |
| To: | Finfrock I JERSEY CENTRAL POWER & LIGHT CO. |
| References | |
| TASK-06-02.D, TASK-06-03, TASK-6-2.D, TASK-6-3, TASK-RR LSO5-81-12-031, LSO5-81-12-31, NUDOCS 8112160082 | |
| Download: ML20038D126 (39) | |
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/2 /6 December 10, 1981 Dockett.No. 50-219 LS05 12-031 A
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$k Mr. I. R. Finfrock, Jr.
Vice President - Jersey Central 6
Y b9 Power & Light Company
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<g Post Office Box 388 5
Forked River, New Jersey 08731 E
A Qigi,'.g W.
Dear Mr. Finfrock:
SUBJECT:
SYSTEMATIC EVALUATION PROGRAM (SEP) FOR THE OYSTER CREEK NUCLEAR POWER PLANT - EVALUATION REPORT ON TOPICS VI-2.D AND VI-3 Enclosed is a copy of our draft evaluation of SEP Topics VI-2.D, "Hass and Energy Release for Possible Pipe Break Inside Containnent," and VI-3, " Containment Pressure and Heat Removal Capability." This evalua-tion corpares your facility, as described in Docket No. 50-219, with the criteria currently used by the regulatory staff for licensing new facilities. Appendix A to our draft evaluation is a draft Technical Evaluation Report from our contractor, Lawrence Livermore National Laboratory. Please infom us if your as-built facility differs from the licensing basis assumed in our assessment. Comments are requested within 30 days of the receipt of this letter so thattthey 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 s b.fect are modified before the integrated assessment is conpleted.
u 5,60 +
Sincerely, lll Dennis M. Crutchfield, Chief Operating Reactors Branch No. 5 g o:
Division of Licensing g
Enclosure:
Draft SEP Topics VI-2.0 and VI-3 cc w/ enclosure:
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3 Mr. I. R. Finfrock cc G. F. Trowbridge, Esquire Gene Fisher Shaw, Pittman, Potts and Trowbridge Bureau Chief 1800 M Street, N. W.
Bureau of Radiation Protection Washington, D. C.
20036 380 Scotts Road Trenton, New Jersey 08628 J. B. Lieberman, Esquire Berlack, Israels & Lieberman Commissioner 26 Broadway New Jersey Department of Energy New York, New York 10004 101 Commerce Street Newark, New Jersey 07102 Natural Resources Defense Council 917 15th Street, N. W.
Licensing Supervisor.
Washington, D. C.
20006 Cyster Creek Nuclear Generating Station J. Knubel P. O. Box 388 BWR Licensing Manager Forked River, New Jersey 08731 GPU Nuclear 100 Interplace Parkway Resident Inspector Parsippany, New Jersey 07054 c/o U. S. NRC P. O. Box 445 Deputy Attorney General Forked River, New Jersey 08731 State of New Jersey Department of Law and Public Safety 36 West State Street - CN 112 Trentoq, New Jersey 08625 Ocean County Library Brick Township Branch 401 Chambers Bridge Road -
Brick Town, New Jersey 08723 Mayor
-Lacey Township 818 Lacey Road Forked River, New Jersey 08731 Commissioner Department of Public Utilities State of New Jersey 101 Commerce Street Newark, New Jersey 07102 U. S. Environmental Protection Agency Region II Office ATTN:
Regional Radiation Representative 26 Federal Plaza New York, New York 10007 L
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- SAFETY EVALUATION REPORT ON CONTAINMENT PRESSURE AND HEAT REMOVAL CAPABILITY SEP TOPIC VI-3 AND MASS AND ENERGY RELEASE
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FOR POSSIBLE PIPE BREAK INSIDE CONTAINMENT, SEP TOPIC VI-2.D FOR THE 0YSTER CREEK NUCLEAR POWER PLANT e
DOCKET ^NO. 50-219 i
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TABLE OF CONTENTS Page I.
Introduction 2
I I'.
Review Criteria 2
III.
Related Safety Topics 3.
IV.
Review Guidelines 3
V.
Evaluation 4
j VI.
Conclusions 6
l.
Appendix A:
Containment Analysis hnd Evaluation for the Oyster Creek Nuclear. Power Plant.
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I.
Introduction The Oyster Creek Nuclear Power Plant, Unit 1 began commercial operations in 1969.
Since then the staff's safety review criteria have changed. As part of the Systematic ~ Evaluation Program (SEP), the containment pressure and heat removal capability (Topic.VI-3) and the mass and energy release for possible pipe break inside containment (Topic VI-2.D) have been re-evaluated.
The purpose of this evaluation is to document the deviations from current safety criteria as they relate to the contair. ment pressure and heat removal capability and the mass / energy release for possible pipe breaks inside containment.
Furthermore, independent analyses in accordance with current criteria were performed to determine the adequacy of the containment design basis (e.g., dssign pressure and temperature).
The. significance of the identified deviations, and recommended corrective measures to improve safety, will be the subject of a subsequent, integrated assessment of the Oyster Creek plant.
II. Review Criteria The review criteria used in the current evaluation of SEP Topics'VI-2.0 and VI-3 for the Oyster Creek plant are contained in the following documents:
(1) 10 CFR Part 50, Appendix A, General Design Criteria (GDC) for Nuclear Power Plants:
(a) GDC 16 - Containment design; l
(b)
GDC 38 - Containment heat removal; and (c) GDC 50 - Containment design basis.
(2).10 CFR Section 50.46, " Acceptance Criteria for Emergency Core Cooling, System for Light Water Nuclear Power Reactors."
i (3) 10 CFR Part 50, Appendix K, "ECCS Evaluation Models".
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(4) NUREG 75/087, Standard Review Plan for the Review of Safety Analysis s
Reports for Nuclear Power Plants (SRP 6.2.1, Containm.ent Functional Design).
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 break 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.B, ESF Switchover from Injection to Recirculation Mode (Automatic ECCS Realignment)
(4)
IX-3, Station Service and Cooling Water Systems (5)
X, Auxiliary Feedwater System (6)
USI-A24, Qualification of Class IE Safety Related Equipment IV. Review Guidelines General Design Criterion (GDC) 16 of Appen' dix A to 10 CFR Part 50 requires that a reactor containment and associated systems shall be provided to establish an essentially leak-tight barrier against the uncontrolled release of radioactivity to the environment.
In addition, GDC 16 requires that the containment design conditions important to safety are not exceeded for as long as the' postulated accident conditions require. GDC 38 requires a containment heat removal system be provided whose system safety function shall be to reduce the containment pressure and temperature following any loss-of-coolant accident (LOCA) and maintain them at an acceptably low level; furthermore, the
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system safety function shall be achievable assuming a single failure. GDC 50 requires that the containment structure and the cont'inment heat removal a
system shall be designed.so that the structure can accomodate, with sufficient margin, the calculated pressure and temperature conditions resulting from any LOCA. This margin as obtained from the conservative caiculation of mass and energy release and the containment model is discussed in the Standard Review Plan (SRP) Section 6.2.1, Containment Functional Design.
The containment design basis includes the effects of stored and generated energy in the accident.
Calculations of the energy available for release should be done in accordance with the requirements of.10 CFR Part 50, Section 50.46 and Appendix K, paragraph I.A, and the conservatism as specified in SRP 6. 2.1. 3. The mass and energy release to the containment from a LOCA should be considered in terms of the mass and energy release during blowdown. Break locations should include recirculation line breaks and steam line breaks. The review also includes the analysis of a postulated single active failure.
By reviewing the licensee's analysis, deviations from the current criteria are identified and independent analyser are performed, as required, to evaluate the significance of these deviations.
In the analyses, "the best estimate" method is used; i.e., by using actual" plant design data, a best estimate, but still reasonably conservative, containment analysis is obtained. The evaluation is completed by comparing the results with the containment design basis.
V.
Evaluation In the case of-BWRs it is necessary to evaluata the effect of pipe breaks below the core for maximum containment pressure and pipe breaks above the core, for maximum containment temperature.
Based on our review of the existing 4-
docket of Oyster Creek, the break locations analyzed include both breaks below and above the core.
In the Oyster Creek FSAR a spectrum of recirculation line breaks were analyzed to determine the peak post-accident pressure'in the drywell and wetwell. All of the resultant peak drywell and wetwell calculated pressures
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were determined to be below.the drywell design pressure of 62 psig and wet
- ell design pressure of 35 psig. The maximum calculated peak drywell pressure and temperature was determined to be 35 psig and 281 F resulting from a DBA LOCA. The resultant peak wetwell pressure and te.aperature were 21 psig and 1300 In addition to reviewing the applicant's analysis, a confirmatory F
analysis was performed which is presented in Appendix A of this report. Mass and energy release rates utilized in the analysis were ' calculated using RELAP-4 MOD 7 in accordance with current criteria. Calculation of the post-accident containment pressure and temperature was done using CONTEMPT-LT/028.
In this case the analysis which was run was a DBA LOCA. The calculated transient reflects a post-accident peak drywell pressure of 37 psig and a peak temperature of 285 F.
The drywell design pressure for Oyster Creek is 62 psig. Both the utility analysis and our own show the peak pressure is well below the drywell's design value.
In addition to recirculation line breaks,'the current criteria states that steam line breaks above the core must be considered. The licensee has performed a spectrum of steam line breaks which are doc'mented in the report u
i entitled " Environmental Qualification of Safety-Related Electrical Equipment
-0yster Creek Nuclear Generating Station" submitted by Jersey Central Power
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and Light Company on October 2,1980.
The main steam line break analyses include 0.01, 0.75, and 2.0 sq. f t. break sizes.
The 0.75 sq. ft. break was found to produce the worst temperature condition inside the containment of 3110 All the resultant peak calculated pressures were well below the F
design pressure of 62 psig.,
In addition to reviewing the applicant's analysis, a confirmatory analysis c
was also performed for a steam line break. This is given in Appendix A.
This analysis included break sizes of 0.01 and 0.75 sq. ft.
The containment response was calculated assuming that the only.available heat sink was the pressure suppression pool. The results of these analyses showed that' a 0.75 sq. ft. break resulted in the most' severe temperature conditioris in the 0
containment. The peak drywell temperature was 312 F.
The temperature-profile is more severe than that resulting from'a recirculation line break.
Therefore, the temperature profile resulting from 0.75_ sq. f t. MSLB should be used to support equipment qualification efforts. The peak drywell pressure was 19 psig. Therefore, for a steamline break the calculated peak pressure is less than a DBA LOCA.
Based on our confirmatory analysis results and the applicant's analysis results, we find the applicant's calculations of the pressure and temperatures produced by the DBS LOCA acceptable.
The applicant's calculated peak tempera-
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ture of 313 F for the 0.75 sq. ft. PSLB is also acceptable, along with the calculated containment pressure for use in the post accident environmental qualification of equipment important to safety and located inside cohtainment.
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VI. Conclusions We have reviewed the analyses submitted by the licensee and have found them to be within the design limits of the Oyster Creek plant. However, due to lack of information regarding the analysis assumptions including initial-conditions, a confirmatory analysis was performed and is reported in, Appendix A.
The confirmatory results were found.to be comparable to the licensee's results and, therefore, we conclude that the licensee has satisfactorily
demonstrated the adequacy of the containment, functional design.
The containment atmosphere conditions during a'O.75 sq. ft. steam line break should be provided as input to the equipment qualification of safety i
related equipment' effort.
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. Appendix A SEP Containment Analysis and Evaluatibn for the Oyster Creek Nuclear Power Plant 1
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Contents Page 1.0 Introduction aad Background 4
2.0 Containment Functional Design -
4 2.1 Review of Oyster Creek Containment Design 6
2.2 Review of Oyster Creek Pipe Breaks Inside the Reactor Coolant Pressure Bouildary 7
2.3 Reanalysis of Oyster Creek Containment
- 9 Design 3.0
' Recirculation Line Break 9
3.1 Containment Response to a Recirculation 11 Line Break 3.2 Containment Response Results 12 4.0 Main Steam Line Pipe Break 13 4.1 Containment Response to a Main Steam Line Break.
14 4.2 Containment Response Results 15 5.0 -
Conclusions 15 6.0 References 16
[
List of Tables Table List of Tables P_ age.
a e
1 Recirculation.Line Blowdown Mass and Energy Release Rate Data 17-2 Containment Model Input Data 18 3
Main Steam Line Break Mass and Energy Release Rate 2
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Data (.01 ft break) 19 4
Main Steam Line Break Mass and Energy Release Rate Data (0.75 sq. ft.)
20-5 Containment Heat Sink Data 21
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e S
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List of Figures Figure Page
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1 Drywell Pressure Response to a Double-Ended Recirculation Line Break 22 2
Drswell Atmosphere Temperature Response to a Double-Ended Recirculation Line Break 23 3
Wetwell Pressure Response to a Double-Ended Recirculation Line Bre'ak 24 4
Wetwell Atmosphere Temperature Respons~ to a Double-Ended e
Recirculation Line Break 25 5
Wetwell Pool Temperature Response to a Double-Ended Recirculation Line Break 26
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6 Drywe'll Pressure Response to a.01 ft2 MSLB 27 7
Drywell Atmosphere Temperature Response to a.01 ft2 MSLB 28 8
Wetwell Pressure Response to a.01 ft2 MSLB 29 9
Wetwell Atmosphere Temperature Response to a.01 ft2 MSLB 30
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10 Wetwell Pool Temperature Response to a.01 ft2 MSLB 31
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11 Drywell Pressure Response to a.75 ft2 MSLB 32 12 Drywell Atmosphere Temperature Response to'a.75 ft2 MSLB 33 13 Wetwell-Pressure Response to a.75 ft2 MSLB 34 14 Wetwell Atmosphere Temperature Respons'e to a.75 ft2 MSLB 35 15 Wetwell Pool Temperature Response to a.75 ft2 MSLB 36 4
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1.0 Introduction and-Background As part of the -Systematic Evaluation Program (SEP), the ' containment functional design capability of the Oyster Creek Nuclear Power Plant has.been re-evaluated. The purpose of this report is-to document the resolution of.SEP S'afety Topic VI-2.D, Mass and Energy Release for Possible Pipe Break Inside Containment, and Safety Topic VI-3, Con'lainment Pressure and Heat Removal -
Capability, and deviations from current safety criteria as they relate to the 4
containment functional design. The significance of the identified deviations and recommended corrective measures will be the subject.of a subsequent integrated assessment of the Oyster Creek plant.
The containment structure encloses the reactor 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.
f Furthermore, equipment having a post-accident safety function must be 1
i environmentally qualified for the resulting adverse pressure and temperature L
conditions.
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4 2.0 Containment Functional Design Oyster Creek is a 1930 MWt General Electric Mark 1 BWR which uses a primarycontainmentconsistingofa_drywell,orpressurbabsorptionchamber, and inter. connecting vent pipes. The pressure absorption chamber is a steel pressure vessel in the shape of a torus located below and encircling the J
drywell. The chamber is approximately half filled with water. The vent system from the drywell terminates below the water. level in the pressure absorption chamber, so-that in the event of a pipe failure in the drywell, the 4
4.
4 released steam passes directly to the water whtre it is condensed. This transfer of energy to the water pool rapidly reduces'the pressur.e in the drywell an'd substantially reduces the potential for subsequent leakage from the, primary containment.
In addition to the pressure absorption chamber, independent auxiliary cooling systems are provided for the reactor and containment cooling under
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various normal and abnormal conditions. These are:
1.
A shutdown cooling system (for decay heat) is provided which circulates water from the reactor through heat exchangers and back to the reactor.
2.
An isolation condenser system, consisting of two condensers, removes decay heat from the reactor while it is still under pressure. Heat from this system is rejected to the atmosphere as steam.
3.
Two separate and independent core spray systems circulate water from the pressure suppression pool to the reactor. The water from these
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systems is distributed directly to the reactor by spray headers mounted inside the plenum and above the core.
Either of these systems provides for cooling the reactor in the event of a '
loss-of-coolant accident (LOCA) due to' a large coolant break. An automatic depressurization system is provided to rapidly reduce reactor pressure to allow the core spray system to maintain continuity of core cooling in the event of a LOCA due to a small coolant line break.
4.
A primary containment cooling system provides water from the pressure suppression chamber pool, which is pumped through independent heat exchangers, and discharged through spray nozzles into the drywell.
Water entering the drywell is returned by gravity to the pressure suppr.ssion pool chamber to complete the cycle.
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In the event of loss of offsite power and failure of one dieseT generator, minimum containment cooling is provided by two low pressure ' coolant injection
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(LPCI) pumps. These pumps are manually switched from core injection mode into the residual heat removal (RHR) mode.
In the RHR mode, water from the wetwell
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is passed through two RHR heat exchangers and returned to the wetwell. A
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containment spray system in.the drywell is provided, but is not Safety Clais 1 and, therefore, is not given credit.in the containment analysis.
2.1 Review of the Oyster Creek FSAR Containment Design Analysis There are two separate calculations which make up.the containment design analysis. First is the mass and energy release analysis for postulated LOCA's. This provides the time dependent mass and energy input'into the containment structure.
The second calculation is the containment response to the mass and energy input to the containment structure. This resuli;s in the time-dependent containment temperature and pressure profile. The severity of the containment response depends on the magnitude and nature of the break location.
If the break is below the core the break flow will be initially single phase liquid.
Th'is results in a fast 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 resul'.s in a much longer blowdown of the mass and energy release to the containment at a much hi.gher enthalpy.
Because of these effects, breaks'below the core are found to produce the most severe pressure response in the containment and steam line breaks above the core produce the most severe temperature response.
The acceptance criteria used to evaluate the Oyster Creek Containment Design Analysis was based on the Standard Review Plan (SRP) Section 6.2.1.
For the containment design analysis to be found acceptable both the mass and
energy release a,nd containment response calculations must meet the acceptance criteria specified in the SRP.
2.2 Review of Oyster Creek Pipe Breaks Within the R-actor Containme~nt System _
The SRP specifies several acceptance criteria applied to the mass and energy release analysis for system pipe breaks. Among these are the break location.
In the Oyster Creek FSAR the most severe mass and energy release rates calculated for containment design were done assuming a double-ended recirculation line break (1,2)
The power level was assumed to be at 1860 Mwt. The energy sources considered include the vessel internal energy and decay heat. The MSIV's are assumed to begin closing at 0.5 seconds and require 3 seconds to fully close.
In the containment response calculation no credit is taken for the heat sinks. The calculated peak post-accident containment pressure resulting from a double-ended recirculation line break was 36 psig. The peak wetwell pressure was 25 psig and the peak drywell temperature was 285 F.
In addition to recirculation line breaks, the current criteria states that steam line breaks above the core must be consid'ered. The licensee has performed a spectrum of steam line breaks which are documented in the report entitled " Environmental Qualification of Safety-Related Electrical Equipment -
Oyster Creek Nuclear Generating Station" submitted by Jersey Central Power and 1.ight Company on October 2,1980.(4) The input and boundary conditions used in this analysis are in accordance with current criteria as. presented in the following discussion.
The basic model used in the analysis was derived from the Exxon Nuclear large break model described in Exxon report XN-75-55.(3)
This'model was G.
', used by Exxon Nucle r to calculate the core flow during blowdown for ncn-jet; pump ';iling water reactors. The analytical methods used were, approved by NRC to meet the requirements for ECCS calculations, i.e.,10 CFR 50 Appendix.K.
The break was assumed to be in one of the main steam lines upstream of the main steam isolation valve (MSIV) in that line. The break discharge coefficient was set equal to unity. The critical flow models used are extended Henry-Fauske for subcooled flow and Moody for saturated flow as specified in the SRP.
In order to be conservative with respect to the containment rcsponse, a very large phase separation velocity was used in the. upper downcomer bubble model. The exact value of phase separation velocity used was not reported.
The feedwater flow was maintained in the automatic control mode.~ Reactor scram occurs on high drywell pressure of 2 psig.
The main steam line breaks reported incude 0.01, 0.75, and 2.0 sq. f t. in size. The 2.0 sq.ft. break represents the largest break size that can blowdown without initiating the automatic containment spray system.
The 0.75 sq. f t. break represents the largest break size in which the blowdown remains single phase throughout the blowdown. The 0.01 sq. ft, break was chosen to be the lower bound of the spectrum of breaks analyzed.
The containment response to each of the st'eam line breaks was also pr'sented in the report. For each of the breaks analyzed, it was assumed that e
the break flow would eventually degrade to become decay' heat steam flow. The containment model used to calculate the containment response for each steam line break was run using different assumptions. These were a containment model 'ith sprays and heat sinks, no spray and with heat sinks, and no sprays w
and no heat sinks.
Since the containment spray system is not built to Safety Class 1, the spray system cannot be accounted for in this analysis. Therefore 6,
the appropriate model to be used for analyzing the containment response calculation is the case with no sprays and heat sinks available., In addition,
- the residual heat removal system was not accounted for. The containment-
- response to the 0.75 sq. ft. break produced the most seve.re temperature response under these assumptions. The. calculated peak post-accident drywell pressure resulting from a 0.75 sq.'ft. steam line break was 19'psig. The peak post-accident drywell temperature was 3110F I
2.3 Reanaly' sis of Oyster Creek 2 Containment Design As mentioned earlier in Section 2.1, Review of Oyster Creek Containment
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Analysis, there are two separate calculations which make up the containment design analysis, the mass and energy re. lease rates and the containment response. The mass and energy release rates calculation can be the result t.'.
4 either a recirculation line break or a steam line break. The recirculation line breaks result in the limiting condition for calculating the peak. pressure i
T inside the-containment. The steam line pipe break analysis is the most limiting case'for temperature conditions inside the' containment.
Both of these analyses were performed and are discussed below.
i 3.0 Recirculation Line Pipe. Breaks For a recirculation line break a DBA LOCA generates the highest containment temperatures and pressures for breaks which occur below the core-mixture level. The LOCA analysis was performed using the RELAP4-MOD 7 computer code. The RELAP4 input deck was obtained from Jersey Central Power and Light t
l Company at the request of NRC. 'The deck was used for analyzing oper,ational transients using the computer code RETRAN. RETRAN is a EPRI developed.
hydraulic transient code based cn RELAP4-MOD 3. - It was carefully reviewed for
code options and in1tial and boundary.. conditions. The plant physical description was assumed to.be as-built. Additional information required was taken from the Oyster Creet. AR.
The initial and boundary cc..ditions for this analysis were defined to j
satisfy the requirem '
,f the Standard Review Plan as discussed below. The following is a listing of the initial conditions and a summary of the assumptions used in this analysis.
1.'
The reactor is operating at 102% of design power at the, time the recirculation pipe breaks. This maximizes the, core heat generation
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rate.
2.
A complete loss of normal a-c power occurs simultaneously with the pipe break.
In addition, the single failure assumptiori was a loss'of one diesel generator.
3.
The recirculation loop pipeline is considered to be instantly severed. This results in the most rapid coolant loss and depressurization with coolant being discharged from both ends of the break. The break area is assumed to be 6.29 sq. f t. and represents.a double-ended br~eak of one of the 24 inch diameter recirculation lines.
4.
Tne reactor is assumed to go subtritical at the time of accident i
initiation due to void formation in the core region. Scram would also occur in less than one second due to a.high drywell pressure signal. The difference between shutdown at time zero and one second is negligible.
5.
The sensible heat released in cooling the fuel and the core decay.
heat-are included in the reactor vessel depressurization calculation. The rate of energy release is calculated using a conservatively high heat transfer
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coefficienttbroughoutthedepressurization. This~ maximizes the' heat-e
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removal rate into the containment. ' Calculations of heat transfer from surfaces exposed to liquid were based on nucleate boiling heat-transfer. For surfaces exposed to steam, the heat transfer calculatidn was based on forced convection.
6.
The main steam line isolation valves are assumed to start closing at
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0.5 seconds after the accident, and are assumed to be fully closed I
within 3 seconds.
By assuming rapid cloture of these valves, the reactor vessel. is maintained at a high pressure, which maximizes the discharge of high energy steam and water into the primary containment.
7.
The feedwater flow is' assumed to stop instantaneously at time zero.
This conservatism is used because the relatively cold feedwater flow,,
if considered to continue, tends to depressurize the reactor.. vessel, thereby reducing the discharge of steam and water into the primary' containment.
8.
The vessel depressurization flow rates are calculated using a discharge coefficient of 1.0, with the Henry Fauske correlation for subcooled and Moody correlation for saturated fluid. A 14.7 psia back pressure was assumed to maximize 51 ass and energy release throughout the blowdown. The blowdown calculation was run until.the primary system pressure dropped below the drywell design pressure of 62 psig. At this time the 1.2 ANS decay heat curve was used.
i The results of this analysis are the time dependent mass and energy release rates presented in Table 1.
3.1 Containment Response Calculation to a Recirculation Line Break The input data for the containment response. calculation consists.of the i
mass and energy rel'eise to the contaire. nt,'a descriptive of the co'ntainment heat removal systems and containment heat sink data. The mass and energy release rates data used were taken from the blowdown of 'the recirculation line presented in the previous section.
The containment heat removal system consists of a pressure suppression pool, residual heat removal. system, and containment sprays. However, since the containment sprays are not Class 1 Safety Grade, they were not taken into account. The containment heat sink data was also conservatively omitted since this analysis is primarily for determining the maximum post-accident pressure and the effact of heat sinks would be negligitle in the short-term.
The Residual Heat Removal (RHR) system consists of 4 low pressure pumps which take water from the suppression pool and pass it'through two heat exchanges before it returns to the suppression pool. With loss of offsite power this system is reduced to 2 low pressure pumps. The RHR system is manually operated and assumed to start at 600 sec. after the break.
Prior to activation of the RHR system, the low pressure pumps add liquid to the reactor vessel. At the end of blowdown this flow discharges through the recirculation line break into the drywell pool.
The containment response calculation was done using the CONTEMPT-LT/28 computer code.
The program model represents the containment as consisting of three-regions; the reactor vessel, the drywell and the wetwell. The physical J
model was obtained from Jersey Central Power and Light Company at the request of NRC. A summary of the containment heat removal systems is given in Table 2.
3.2 Co'ntainment Response Results The containment pressure and temperature response to a recirculation-line break are shown in Figures 1 through 5.
The calculated transient reflects a,
peak post-accident containment drywall pressure of 37 psig and a temperature
.of 285 F. 'The post-accident containment wetwell pressure.and temperature are 16 psig and 142 F.
The containment design pressure for the drywell is 62 psig and the we'twell is 35 psig. There is,.therefore, a substantial margin between the peak calculated pressure and the containment design pressure.
4.0 Main Steam Line Pipe Breaks Analyses of the containment response to a steam line break were also made. This analysis is performed to determine the most severe long term pressure and temperature condition in the containment following a pipe break.
The containment long term resp'onse is c.alculated for a' 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> period. To determine the cost severe long term conditions in the containment the most limiting steam'line break size must be found. For BWRs the most limiting 4
break size was found by running a spectrum of break sizes; and taking into consideration vent clearing in the suppression pool and the rate of blowdown.
The smallest break size of interest is the one which just clears the vents during the six hour period. The largest break size of interest is the one which requires 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> to depressurize the reactor. The single active failure assumption was loss of offsite power with the f'ailure of one diesel generator. The initial power was specified as 102% of the design rating. The input and boundary conditions for this analysis were defined to satisfy the requirements of Section 6.2.1 of the Standard Review Plan and are discussed 1
below.
The break was assumed to be in one of the main steam lines. The break discharge coefficient was set equal to unity. The critical flow models used were extended Henry-Fauske for subcooled flow and Moody for saturated flow at the break junction. The RELAP 4 input deck used was based on the same one e,
used in the recircQ1ation break analysisi IThe break location was iaoved to the main steam line upstream of the MSIV in that line.
Main steam line breaks ranging in sizes from a double-ended guillotine to 2
2 a.01 ft were analyzed. A break size of 0.75 ft was included to verify the containtaent analysis submitted by Jersey C'entral Power and Light ' Company.
In order to be conservative with respect to the containment response,"it is necessary to maximize the steam that would exit from the break. This was done by assuming an infinite phase separation velocity in the upper plenum which prevented any entrainment. The feedwater flow was assumed constant at 2064 lbm/sec at 315.5 'F' until the reactor scramed on high drywell
~
pressure of 2 psig.
The steam flow was also assumed constant at 2064 lbm/sec at 548.8 *F
^
until the reactor scramed.
For the O'.01 sq. ft. break, th'e-reactor scramed at 85 seconds, for the 0.75 sq. ft. break this occured at 5 seconds.
Following blowdown, the mass and energy release rates reduce to that of a steam decay heat curve. The mass and energy release rates for the 0.01 ft2 2
and 0.75 ft steam line break are presented in Tables 3 and 4, respectively.
Wa 4.1 Containment Response to a Main Steam Line Break The inputi data for the containment response calculation consist of the ma'ss anc' energy release to the containment, a description of the containment heat removal systems and the available containment heat' sink data. The mass and energy release rates were taken from the blowdown rates presented in the previous section.
Following the blowdown, the mass and energy release rates are re'duced to decay heat steam. As before, the only containment heat removal system modeled was the residual heat removal system mentioned in Section~3.1, Containment Response to a Recirculation Line Break. This system was assumed De 14 -
~. "
to start at 600 seconds. Since this analysis is intended'for long ttrm
~
post-accidint containment conditions, the effect of containment heat sinks was
~
also included. The containment heat-sink data use'd is presented in Table 5.
m.
4.2 Contadnment Res~ponse to Main Steam Line Break Results The (e:ntainment response results were calculated using CONTEMPT-LT/28 over a six hour period. A two volume model representing the drywell and wetwell was used over the entire six hour period. The containment pressure and 2
temperature response to the.01 ft steam line break are shown in Figures 6 through 10. The calculated-transient reflects a peak port accident drywell pressure of 19'psig and a temp'erature o,f 310 F.
The resulting wetwell 0
pressure and temperature are 17 psig and 1.27.0F.
The response to a 0.75 ft2 steam line break is shown in Figures 11 through 15. The calculated e
transient in this case reflects a peak post accident drywell pressure of 19 psig and a temperature of 312 F.
The resulting wetwell pressure and 0
temperature are 18 psig and 139 F.
Both the peak drywell and wetwell pressures are substantially below design 2
for both cases. The post accident temperature for the.75 ft steam'line break results in the most severe temperature co'nditions and should be used for equipment qualification of s.afety related equipment.
i
'5.0 Conclusions Based on our review of the Oyster Creek docket and our subsequent analysis, we conclude that the Oyster Creek containment design pressure meets current NRC criteria. The containment atmosphere conditions as a result of a 2
0.75 ft steam line break provides the most severe temperature conditions for equipment qualification of safety related equipment.
'. 6.0 References 1.
Oyster Creek-Unit No.1 Facility Description and Safe,ty Analysis Report 2.
Oyster Creek Unit No. 1, Amendment ll' Answers to 109 AEC Questions Regarding Additional Plant Information, June. 21,1967.
3.-
Safety Evaluation Report by the Office of Nuclear Reactor Regulation of the Exxon Nuclear Co. WREM based NON-JET Pump-Boiling Water Reactor ECCS Evaluation Model and Application to the'0yster Creek Pl~ ant, Dec. 10, 1976.
4.
Letter from Jersey Central Power and Light Co. to Dennis Crutchfield on October 26, 1980, " Environmental Qualification of Safety Related Electrical Equipment-Oyster Creek Nuclear Generating Station, Oct. 31, 1980.
+
e t
b 16 -
-~
TABLE 1 Double-Ended Recirculation Line Break 2
~
Release Rate Data (6.29.ft Break)
TIME.
' FLOW ENERGY' (seconds)
(1bm/sec)
(Btu /lbm) 0.0 3.69E4 546.
1.0 3.69E4 546.
2.0 3.69E4 544.
'3.0 3.69E4 543.
4.0 3.69E4 541.
5.0 3.68E4 538.
10.0 1.98E4 609.
15.0 9.6SE3 676.
20.0 7.32E3 691.
25.0 5.18E3 700.
30.0 2.18E3 711.
Decay Heat at 30 seconds (1.2 ANS) 30.0 6.75El 1200.
100.0 6.17El 1200.
400.0 4.39El 1200.
1000.0 3.45El 1200.
4000.0 2.38E1 1200..
10000.0 1.80E1 1200.
i 40000.0 6.33E0 1200.
TABLE 2 Containment Model Input Data (Taken from Oyster Cresk FDSAR),
Drywell/Wetwell Data Drywell Wetwell Free Air Volume (cu. ft.)
180,000.
127,000.
Initial Pool Water Volume (cu. ft.)
0.0 83,400.
Initial Temperature of Atmosphere (oF) 135.
120.
Initial Temperature of Pool (oF)
~
135 7 12d.7 Initial Pressure (psia) 14.
I4.
Relative Humidity (percent) 100 100 Pool Surface Area (sq. ft.)
1,373.9 9,219.-
HTC Moltiplier 1.0 1.E6 Mass Transfer Multiplier 1.0 1.0 Vent System Vent Pipes Number 10 Internal Diameter 6 ft. 6 in.
Break area / vent pipe area 0.0194 Downcomer Pipes Number 120 Internal Diameter.
I ft. 11.5 inch.
Submergence Below Absorption Pool Water Level 4 ft.
~
Vacuum Breakers Number valves 14 Vent area 1.75 ft2 per valve Actuation set-point-0.5 psi for full open Residual Heat Removal System iteat exchanger surface area 6200. ft2 Overall heat transfer coefficient 398.6 Btu /hr ft2 0F Coolant inlet temperature 859F Secondary coolant flow rate 2.0696E6 lbm/hr Pr'imary coolant flow rate 2.122E6 lbm/hr
- (
--w
,y y
9 9
9--
TABLE 3 MSLB Mass and Energy 2
2 Release Rate Data (0.01 f t Break)
TIME FLOW ENERGY (seconds)
(1bm/sec)
(Btu /lbm) 0.0 0.0 0.
1.0 22.18 1197.
10.0 22.18 1197.
4 400.0 22.18 1197.
Decay Heat at 400 Seconds (1.2 ANS) 400.0 4.39El 1200.
1000.0 3.45L1 1200.
4000.0 2.38E1 1200.
10000.0 1.80E1 1200.
40000.0 6.33E0' 1200.
9 0
I G,
i
TABLE 4'
~
MSLB Mass and Energy 2
Release Rate Data (0.75 ft Break')
TIME FLOW ENERGY (secor. ids)
(lbm/sec)
(Btu /lbm) 0.0 1665.
1198.
1.0 1665.
1198.
5.0 1665.
1198.
10.0 1369.
1201.
20.0 980.
1201.
40.0 875.
1198.
60.0 729.
1197.
80.0 667.
1196..
120.0 485.
1197.
160.0 394.
1198.
'200.0 301.
1197.
Decay Heat at 400 Seconds (1.2 ANS) 400.0 4.39El 1200.
1000.0 3.45El 1200.
4000.0 2.38E1 1200.
10000.0 1.80E1 1200.
40000.0 6.33E0 1200.
TABLE 5
~
~~
Containment Heat Sink Data (Taken'from Reference 4)
Material Thickness Area 2
(Inch)
(ft) f Drp ell Sphere--Lower Steel 1.154 2542.
Insulation 2.5 Concrete 78.
Drpell Sphere--Middle Steel 0.770 4220.
Insulation 2.75 Concrete 78.
Dr pell Sphere--Upper Steel O.772 3085.
Insulation 2.75 Concrete 78.
Drpell Transition Steel 2.566 1433.
Insulation 2.5 Concrete 78.
Drpell Cylinder Steel 0.640 2574.
~
Insulation 2.5 Concrete 78.
Drpell Head Steel.
1.188 428.
Torous Steel 0.385 14952.
Material Properties Material
' Thermal Conductivity Feat Capacity (Stu/hr. ft F)
(Btu /hr. ft30 )
2 p
Concrete
'0.92 22.6 Steel
,27.0 58.8 4
3.74 Insulation 0.g2
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Wetwell Pool' Temperature Response to a.75 ft MSLB e
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