ML20055A749
| ML20055A749 | |
| Person / Time | |
|---|---|
| Site: | La Crosse File:Dairyland Power Cooperative icon.png |
| Issue date: | 07/13/1982 |
| From: | Crutchfield D Office of Nuclear Reactor Regulation |
| To: | Linder F DAIRYLAND POWER COOPERATIVE |
| References | |
| TASK-06-02.D, TASK-06-03, TASK-6-2.D, TASK-6-3, TASK-RR 5-82-7-20, LSO5-82-07-020, LSO5-82-7-20, NUDOCS 8207190403 | |
| Download: ML20055A749 (40) | |
Text
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/d July 13,1982 Docket No. 50-409 LSOS-82-07-020
,e Mr. Frank Linder General Manager Dairyland Power Cooperative 2615 East Avenue South Lacrosse, Wisconsin 54601 j
Dear Mr. Linder:
SUBJECT:
SYSTEMATIC EVALUATION PROGRN1 (SEP) FOR THE LACROSSE BOILING WATER REACTOR - EVALUATION REPORT ON TOPICS VI-2.0 AND VI-3 Enclosed is a copy of our draft evaluation of SEP Topics VI-2.0, " Mass and Energy Release for Possible Pipe Break Inside Containment," and VI-3, " Containment Pressure and lleat Removal Capability." This evaluation corapares your facility, as described in Docket No. 50-409, 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 froa the licensing basis assumed in our assessment.
Comments are requested 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 HRC criteria relating to this subject are modified before the integrated assessment is completed.
Sincerely,
Nf fra ast Et (sQ Dannis M. Crutchfield, Chief Operating Reactors Branch No. 5 g.
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4 Mr. Frank Linder CC Fritz Schubert, Esquire U. S. Environmental Protection Staff Attorney Agency Dairyland Power Cooperative Federal Activities Branch 2615 East Avenue South Region V Office La Crosse, Wisconsin 54601 ATTN:
Regional Radiation Repr'esentative 230 South Dearborn Street O. S. Heistand, Jr., Esquire.
Chicago, Illinois 60604 L
Morgan, Lewis & Bockius 1800 M Street, N. W.
Mr. John H. Buck Washington, D. C.
20036 Atomic Safety and Licensing Appeal Board U. S. Nuclear Regu.latory Commission Mr. R. E. Shimshak Washington, D. C.
20555 La Crosse Boiling Water Reactor Dairyland Power Cooperative Mr. Ralph S. Decker P. O. Box 275 Route 4, Box 1900 Genoa, Wisconsin 54632 Cambridge, Maryland 21613 Mr. George R. Nygaard Charles Bechhoefer, Esq., Chairman Coulee Region Energy Coalition Atomic Safety and Licensin'g Board 2307 East Avenue U. S. Nuclear Regulatory Commission La Crosse, Wisconsin 54601 Washington, D. C.
20555 Dr. Lawrence R. Quarles Dr. George C. Anderson Kendal at Longwood, Apt. 51 Department of Oceanography Kenneth Square, Pennsylvania 19348 Univer'sity of Washington
~
Seattle, Washington 98195 U. S. Nuclear Regulatory Commission Resident Inspectors Office James G. Keppler, Regional Administrator
. Rural Route #1, Box 276 Nuclear Regulatory Commission, Region III Genoa, Wisconsin 54632 799 Roosevelt Road Glen Ellyn, Illinois 60137 Town Chairman Thomas S. Moore j
Town of Genoa Atomic Safety and Licensing Appeal Board l
Route 1 U. S. Nuclear Regulatory Commission l
Genoa, Wisconsin 54632 Washington, D. C.
20555 Chairman, Public Service Commission of Wisconsin Hill Farms State Office Building Madison, Wisconsin 53702 i
' Alan S. Rosenthal, Esq., Chairman Atomic Safety and Licensing Appeal Board U. S. Nuclear Regulatory Commission i
Washington, D. C.
20555 l
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... _.. _... _ _.. ~.. _... _.
n SAFETY EVALUATION REPORT ON CONTAINMENT PPESSURE AND HEAT REMOVAL CAPABILITY SEP TOPIC VI-3 AND MASS AND ENERGY RELEASE FOR POSSIBLE PIPE BREA.K INSIDE CONTAINMENT, SEP TOPIC VI-2.0 FOR THE LACROSSE NUCLEAR POWER PLANT DOCKET NO. 50-409
'r TABLE OF CONTENTS I.
Introduction 1
II.
Review Criteria 1
III.
Related Safety Topics 2
IV.
General Review Guidelines 2
V.
Evaluation 3
VI.
Conclusion 4
j VII.
References 5
Appendix A: SEP Containment Analysis and Evaluation for the Lacrosse Nuclear Power Plant, i
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I.
Introduction The la Crosse Nuclear Power Plant began comercial operations in 1969.
Since then the United States Nuclear Regulatory Commission staff's safety review criteria have changed. As part of the Systematic Evaluation Program (SEP), the containment pressure and heat removal capability (SEP Topic VI-3)
{t[
and the mass and energy release for possible pipe break inside containment
,;f (SEP Topic VI-2.6) have been re-evaluated. The purpose of this evaluation is to document any existing deviations from current safety criteria that pertain l
to the containment pressure and heat removal capability and the mass / energy release for possible pipe break inside containment.
Independent analysis in t.
accordance with current criteria were performed by LLNL to determine the
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adequacy of the containment design and to provide input for unresolved safety fd k
issue (USI) A-24, Qualification of Class lE Safety Related Equipment.
The w
M significance of any identified deviations, and recommended corrective measures to improve safety, will be the subject of a subsequent, integrated p
piu assessment of the Lacrosse plant.
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II. Review Criteria The review criteria used in the current evaluation of SEP Topics VI-2.0
[
and VI-3 for the La Crosse plant are contained in the following ' documents-o (1) 10 CFR Part 50, Appendix A, General Design Criteria (GDC) for f
Nuclear Power Plants:
(a) GDC 16 - Containment design; p
(b) GDC 38 - Containment heat removal; and j
1 (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."
l (3) 10 CFR Part 50, Appendix K, "ECCS Evaluation Models".
(4) NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants (SRP 6.2.1, Containment Functional Design).
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kII. 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 fer possible pipe break inside containment, and/or containment pressure and heat removal
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capability.
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(1) III-1, Classification of Structures, Components and Systems (Seismic and Quality) f[
(2)
VI-7.B. ESF Switchover from Injection to Recirculation Mode (Automatic ECCS Realignment)
(3)
IX-3, Station Service and Cooling Water Systems 1
(4)
X, Auxiliary Feedwater System I
(5) USI-A24, Qualification of Class lE Safety Related Equipment b
E IV. General Review Guidelines General Design Criterion (GDC) 16 of Appendix A to 10 CFR Part 50 l
requires that a reactor containment and associated systems shall be provided
[
to establish a leak-tight barrier against the uncontrolled release of l
radioactivity to the environment. In addition, GDC 16 requires that the containment pressure and temperature design conditions important to safety ar*
f GDC not exceeded for as long as the postulated accident conditions require.
38 requires that a containment heat removal sy' stem be provided to reduce the containment pressure and temperature following any loss-of-coolant accident (LOCA) and maintain them at an acceptably low level. This safety system is to function assuming a sing 1'e f ailure. GDC 50 requires that the containment structure and the containment heat removal 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 and the containment model are discussed in the Standard Review Plan (SRP) NUREG-0800 l
Section 6.2.1, Containment Functional Design; t.he, margin is obtained from the conservative calculation of mass / energy release. The containment design basis includes the effects of stored and generated energy from the accident.
Calculations of the energy available for release should be made in accordance. - - -
- = = = - =
with the requirements of 10 CFR Part 50, Section 50.46 and Appendix K, In general, paragraph I.A, and the conservatism as specified in SRP 6.2.1.3..
calculations of the mass and energy release rates for a loss-of-coolant accident should be performed in a manner that conservatively establishes the containment internal design pressure and temperature (i.e. maximizes the post-accident containment pressure and temperature).
By reviewing the licensee's analysis, deviations from current licensing criteria can be identified and independent analy,ses performed, to evaluate the In the analysis, "the best estimate" nethod significance of these deviations.
is used; i.e., by using actual plant design data, this best estimate analysis The evalu-remains a reasonably conservative analysis of containment response.
g ation is completed by comparing the results with the containment design basis.
V.
Evaluation _
.I 4
In the case of BWRs it is necessary to evaluate the effect of pipe breaks i
j below the level of the core for maximum containment pressure and of pipe breaks Based on our re-above the level of the core for maximum containment temperature.
view of the existing docket for Lacrosse, the break locations analyzed by Dairy-
-f land Power Cooperative are for breaks only occurring below the level of the core.
f k
i In the Lacrosse BWR Hazards Summary Report a spectrum of recirculation k
line breaks was analyzed to determine the peak post-accident pressure.U All of
.l the resultant peak calculated pressures were detemined to be below the contain-
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The maximum calculated peak pressure was deter-ment design pressure of 52 psig.
The post-accident contain-mined to be 48.2 psig for a recirculation line break.
The containment design temperature is p
ment temperature conditions reached 272 F.
The initial conditions and assumptions prescated in the report were not 280*F; adequate to determine whether or not the analysis was consistent with current cri Therefore, in addition to reviewing the applicant's analysis, a confima-j teria.
tory and independent analysis was performed by LLNL for the USNRC, which is pre-I Mass and energy release rates utilized in sented in Appendix A of this report.
f the analysis were calculated using RELAP-4/ MOD 7 in accordance with current -. -. -
.3
criteria. Calculations of the post-accident containment pressure and temperature responses were made using CONTEMPT-LT/028. One of the analyses made was the double-ended recirculation line break. The calculated transient reflects a post-accident peak drywell pressure of 43 psig and a peak temperature of 265*F. These results are plotted in Figures 1 and 2.
Both the utility analysis and our analy-sis show that the peak pressure and temperature are below the containment's design values.
In addition to the recirculation line break case, the current criteria state that steam line breaks above the level of the core must be considered. The licen-see did not perform main steam line break analyses. Therefore, independent analy-ses were performed. These are discussed in Appendix A.
The analyses were per-2 2
2 fomed for three main steam line break sizes, 0.01 ft, 0.10 ft and 0.634 ft,
2 The 0.01 f t break analysis was used as a bounding case to determine the amount of time the reactor operator would have to initiate containment sprays and/or the Automatic Depressurization System. The analysis indicated that the containment
-would reach the design pressure 56 minutes after onset of the break. The design temperature would be reached in 53 minutes. Operator action would then be needed and should become effective within 53 minutes.
If accomplished on time, the con-tainment design limits would not be exceeded.
Due to their more rapid depressurization rates, operator, action would not j
be required for either the 0.1 ft or the 0.634 ft break; these blowdowns could i
be accommodated solely by the containment's passive heat sinks. Of the two, the 0.634 ft break, corresponding to a double-ended rupture of an eight-inch main l
steam line, caused the more severe containment response of 33 psig for pressure
(
and 250*F for temperature.
VI.
Concl usions The analyses submitted by the licensee have been reviewed and found to be j
within the design limits of the Lacrosse plant. A confirmatory analysis was I
performed for the recirculation line break accident with the resulting containment responses of 43 psig maximum for pressure and 265*F maximum for temperature. These l
results yield more adverse temperature and pressure responses than the main steam line break analyses and should, therefore, be provided as input to the equipment qualification of safety-related equipment, USI A.24.
l l
I 2
Independent analyses were performed for the 0.01 ft, 0.10 ft2 and 0.634 ft main steam line break. The latter two break sizes produce contain-ment responses less severe than the recirculation line break accident. The former, the 0.01 ft break, was found to require operator action which should become effective 53 minutes into the transient. This time frame is adequate for the reactor operator to activate manually the containment sprays or MDS.
Having done so the upper-bound analysis indicates that the containment response would not exceed design values.
VII. References 1.
Lacrosse BWR Hazards Summary Report, Dairyland Power Cooperative,1967.
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LA CROSSE LOCR CONTEMPT-LT/028 02/05/82 05:21:29 DHTYK0U:
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Figure 1. Containment Pressure Response to a 3 59 sq. ft. cold Leg Discharge Break e
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f DIVISION OF LICENSING CORRIGENDA FOR APPENDIX A LAWRENCE LIVERMORE NATIONAL LAB 0PATORY TECHNICAL EVALUATION REPORT FOR SEP TOPICS VI-2.D AND VI-3 Page 1 Lacrosse is a 165 MWt Allis-Chalmers Boiling Water Paragraph 3 Reactor (BWR).
Sentence 1 Page 9 This is the amount of time that exists for the reactor Paragraph 2 operator to take appropriate action such as to manually Sentence 4 actuate the containment sprays or the manual depressuriza-tion systen (M_DS).
Page 11
- Assume steam decay heat curve at 30.0 sec.
Foctnote m
~ -. _ _ _.
APPENDIX A:
TECHNICAL EVALUATION REPORT
+i 2
SEP Containment Analysis and Evaluation y
I1 for the La Crosse Nuclear Power Plant Contents Page II!
i:hi*
18 1
V 1.0 Introduction and Background
- j 1
2.0 Containment Functional Design Description 2.1 Review of La Crosse Containment Design 3
j y
2.2 Review of Pipe Breaks Inside the Reactor Coolant Boundary 3
4 kj 2.3 Reanalysis of La Crosse Containment Design kI 4
3.0 Recirculation Line Break Analysis-3.1 Containment Response to a Recirculation Line Break 6
6
]
3.2 Containment Response Results 7
4.0 Main Steam Line Pipe Break Analysis t
4.1 Containment Response to a Main Steam Line Break 8
[
0 8
4.2 Containment Response Results i
10 5.0 Conclusions 10 6.0 References
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List of Figures P, a_ge Figure Containment Pressure Response to a 3.59 sq. ft. Cold Leg 1
18 Discharge Break 19 Containment Temperature Response to.a 3.59 sq. ft. Cold Leg 1
2 Discharge Break i
20 Containment Pressure Response to a 0.01 sq. ft. Main Steam 3
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Line Break
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21 Containment Temperature Response to a 0.01 sq. ft. Main 5 team j
4 hi il*
Line Break
[M 22 Containment Pressure Response to a 0.1 sq. ft. Main Steam
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Line Break
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i 23 Containment Temperature Response to a 0.1 sq. ft. Main Steam 1
Line Break
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24 Containment Pressure Response to a 0.634 sq. ft. Main Steam 7
Line Break h
25 Containment Temperature Response to a 0.634 sq. ft. Main Steam
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8 Line Break 3
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List of Tables
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.g List of Tables Page Table 11 J-f Recirculation Line Blowdown Mass and Energy Release Rate Data I
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- y Initial Conditions for Containment Response Calculations 2
W;f 13 Main Steam Line Break Mass and Energy Release Rate Data 3
(.01 sq. ft. break) s 14 Main Steam Line Break Mass and Energy Release Rate Data 4
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(.1 sq. ft. break) q 15
'gj Main Steam Line Break Mass and Energy Release Rate Data 5
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(.634 sq. ft. break) 43 Si 16
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6 Containment Heat Sink Data b
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1.0 ' Introduction and Background j
?'.,
As part of the Systematic Evaluation Program (SEP), the containment j
functional design capability of the la Crosse Nuclear Power Plant has been E
re-evaluated. The purpose of this report is to document the resolution of SEP Safety Topic VI-2.D, Mass and Energy Release for Possible Pipe Break Inside h
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.(1) The significance of the identified b
deviations and recomended corrective measures will be the subject of a subsequent integrated assessment of the La Crosse plant.
k N
The containment structure encloses the reactor and is the final barrier against the release of radioactive fission products to the atmosphere in the event of an accident. The containment structure must, therefore, be capable A
of withstanding, without loss of function, the pressure and temperature conditions resulting from postulated LOCA and steam line break accidents.
Furthermore, equipment having.a post-accident safety function must be capable g
of withstanding the resulting adverse pressure and temperature conditions.
I l
5 2.0 Containment Functional Design Description i
~
La Crosse is a 165 MWt General Electric Boiling Water Reactor (BWR).
In La Crosse water enters the bottone of the reactor vessel through four 16 inch pipes and passes upward through the core passing along the fuel rods. Boiling produces steam and is separated in the steam dome. From there, the steam leaves the vessel through two 8 inch pipes. These lead to a single 10 inch line that passes from the containment building to the turbine building. The hot water which is removed from the steam in the steam dome exits the steam dome through four 16 inch pipes. These combine to make up the two 20 inch recirculation lines. The water is returned to the reactor vessel by two recirculation pumps.
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The steam line can be isolated by a hydraulically operated isolation This valve can be controlled remote.
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valve and can be closed in 10 seconds.
1 l for canually from the control room and is closed automatically upon signa s low reactor water level, low steam pressure at the turbine stop valve, or low I
g The steam line can also be isolated by the turbine i
lj main condenser vacuum.
However, this valve is not automatic and is building steam isolation valve.
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H controlled from the control room.
The main feed return line has a check valve inside the conta building and a remotely. operated shutoff valve in the turbine building.
The containment structure consists of a cylindrical steel vessel, 60 fe 264,160 cubic feet and The vessel has an internal free volume of The containment in diameter.
is designed to withstand an internal pressure of 52 psig.
d structure encloses the reactor vessel, primary recirculation pipes, an 5
equipment needed to operate the emergency core cooling system (E containment heat removal system.
The containment heat removal systems consist of a containment spray The containment spray system is manually system and passive heat sinks. Water is supplied to the building spray operated.
The piping storage tank located at the top of the containment vessel.
k connection to the emergency core spray system is on the bottom of the ta
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The connection to the spray headers of the building spray system is a The bottom of the " standpipe is at a sufficient standpipe within the tank.
elevation above the bottom of the tank so that 15,000 gallons of water i
available for the emergency core spray system at all times except dur ng The minimum amount of water available for containment spray at Building spray is delivered by gravity feed at refueling.
full power is 11,300 gallons.
The containment spray system is not built to 1000 gpm to the spray headers.
safety class 1.
In addition to the containment spray system, containment heat remova The containment heat brought about by the presence of passive heat sinks.
sink data is present in Table 6.
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Review of the La Crosse Containment Design Analysis _
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2.1 The design analysis.
Two separate calculations make up the containmentfor postulated LOC]
7 This pro-first is the mass and energy release analysisinput from the primary system in g
jj vides the time dependent mass and energyThe second calculation is the con i
.fi r'ce containment structure.
The containment response results in the l
lf The severity of this mass and energy input.
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[i ofiles.
time-dependent containment temperature and pressure prd nature of the mass a
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the containment response depends on the magnitude anIn turn, the magni f
cnergy release from the postulated LOCA.
dependent on the break j
the mass and energy release to the containment isIf i lly This results in a f ast blowdown of the mass and energy y location.
r thalphy. If the break is
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single phase liquid.
ralease to the containment at a relatively low en This results (j
h steam.
above the core the break flow vill be mostly single p ase tainment at
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in a much longer blowdown of the mass and energy re Q
a much higher enthalpy.
ti ent and found to produce the most severe pressure response in the con
_y temperature response.
B steam line breaks above the core produce the most severe Nj Containment Design The acceptance criteria used to evaluate the La Crosse l
i 6.2.1.
For the Analysis was based on the Standard Review Plan (SRP) Sect onsu i
containment design analysis to be found acceptable the ret response cal j
mass and energy release calculation and the containmen Qj h SRP.
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must meet the acceptance criteria specified in t e j,-1 Q
Boundary Review of Pipe Breaks Inside the Reactor Coolant Pressure bJj 2.2 h
ass and The SRP specifies several acceptance criteria applied to t e mA
{fj energy release analysis for primary system pipe bre
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p d is (2) In this analysis N
break location.
described in the La Crosse BWR Hazards Sumary Report. tion for conta r
the most severe mass and energy release rate calcula 9
l tion line. The 1
design was done assuming a double-ended break in the recircu a
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break location was on the pump discharge side at a po n I -
4 m
the reactor vessel. The maximum calculated peak pressure was determined to 48.2 psig. The peak post accident containment, temperature conditions was A substantial amount of information needed to evaluate the analysis 0
272 F.
is not contained in this report (e.g., infomation pertaining to the choke flow correlation and the heat transfer assumptions used in the analysis).
Without this information it is' not possible to conclude whether or not the 5}lj containment design analysis presented in the La Crosse BWR Hazards Sumary
$j
- h Report is adequate.
2.3 Reanalysis of La Crosse Containment Design As mentioned earlier in Section 2.1, Review of La Crosse Containment Analysis, there are two separate calculations which make up the containment
- 7 design analysis, the mass and energy release rate and the containment j
The mass and energy release can be the result of either a j
response.
recirculation line break or a steam line break. The recirculation line break
[
results in the worst condition for calculating the peak pressure inside the j
A containment. The steam line pipe break analysis is the worst case for temperature conditions inside.the containment.
p B
As pointed out in the previous section, the analyses submitted by Dairyland Power Cooperative lacked necessary information regarding initial j) conditions, assumptions or complete results to determine whether or not the current criteria were met. Both a recirculati,on line break and a steam line b
break analysis was performed again by LLNL and are discussed below.
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'A J
3.0 Recirculation Line Pipe Breaks i
For a recirculation line break a design basis accident (DBA) LOCA 5
i generates the highest containment temperatures and pressures for breaks which I
occur below the core mixture level. The LOCA analysis was performed using the RELAP4-M007 computer code. The RELAP4 input deck was obtained from Dairyland f
Power Cooperative at the request of the NRC. The deck was reviewed by LLNL to evaluate the selected code options, initial conditions and boundary conditions.
The plant physical description was assumed to be as-built.
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Additional information required was taken from the La Crosse BWR Hazards
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Sumary Report.(2)
Y jl The initial conditions and boundary conditions for this analysis were C'
selected to satisfy the requirements of the Standard Review Plan section-W
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6.2.1.
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1 The following is a listing of the initial conditions and a summary of the~:g assumptions used in this analysis.
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The reactor is operating at 102% of design power at the time the 1
This will produce the maximum core heat
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recirculation pipe breaks.
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generation rate.
h A complete loss of normal offsite AC power occurs simultaneously with t
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The recirculation pump discharge pipeline is considered to be instantly h)
Coolant being discharged from both ends of the f
3.
severed at both ends.
The break results in the most rapid coolant loss and depressurization.
break area is assumed to be 3.59 sq. ft. and represents a double-ended dF break of one of the 20 inch diameter recirculation lines.
j The reactor is assumed to go subcritical at the time of accident Scram would normally occur in less than one second due to a 4.
initiation.
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high drywell pressure signal.
The censible heat released in cooling the fuel rods and the core decay d heat are included in the reactor vessel depressurization calculation.,j 5.
The rate of energy release is calculated using a conservatively high he This maximizes the s
transfer coefficient throughout the depressurization.
e Calculations of heat transfer from
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heat removal rate from the core.
surf aces exposed to liquid were based on nucleate boiling heat transfer
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For surf aces exposed to steam, the heat transfer calculation was base forced convection.
The main steam line isolation valves are assumed to be clo I
By assuming closure of these valves, the reactor 6.
start of the accident.
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vessel is maintained at a high pressure, which maximizes the disc arge o e
high energy steam and water into the primary containment.
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' The feedwater flow is assumed to be closed at time zero. This conserva-3, 7
tive assumption is made because the relatively cold feedwater flow, if j1 considered to continue, tends to depressurize the reactor vessel, and j'
3 causes a reduction in the discharge of steam and water into the primary.
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-g containment.
'kl The vessel depressurization flow rates are calculated using a discharge Y;
8.
coefficient of 1.0, with the Henry Fauske correlation for subcooled and h
Moody correlation for saturated fluid. A 14.7 psia back pressure was assumed to maximize the mass and energy release throughout the blowdown.
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The blowdown calculation using RELAP 4 was run until the primary system At 5$
pressure dropped below the containment design pressure of 52 psig.
this time 1.2 times the ANS decay heat curve was used.
_.$j Emergency core cooling system was not modeled since sufficient h
s 9.
information was not provided in the RELAP model obtained from Dairyland d
Power Cooperative.
m.+
~
The results of this analysis are the time dependent mass and energy
]
t}?
release rates presented in Table 1.
5
..h a
3.1 Containment Response Calculation to a Recirculation Line Break 3
The input data for the containment response calculation consists of the y
y mass and energy release to the containment and a description of the containment heat removal systems. Passive con'tainment heat sinks were the
. *~
only heat removal systems accounted for since the containment spray system is manually operated and not safety class 1.
The containment heat sinks modeled N
aro described in Table 6.
The mass and energy release rate data used were taken from the blowdown calculation of the recirculation line break presented in the previous section.
The containment response calculation was made using the CONTEMPT-LT/28 computer code. The program models the containment as a one volume dry The initial conditions used in the analysis are stenarized in i
containment.
o Table 2.
5YU.
l
.,-w.
pee-PP.*"'"-'
S
[.2 Containment Response Results
$a The containment pressure and temperature response to a recirculation line y
break are shown in Figures 1 and ;2. The calculated transient reflects a peak
[
post-accident containment pressure of 43 psig and a temperature of 270 F.
_ jej 0
gA.;
This compares with a containment pressure of 48.2 psig calculated by Dairyland y/d The containment design Power Cooperative presented in the FHSR for la Crosse.
J,$
pressure is 52 psig. There is, therefore, an 8% marnin % tween the peak co calculated pressure and the containment design pressure.
A,y
- (..
T-_-~
u'.-
4.0 Main Steam Line Pipe Breaks
'??
~
Analyses of the containment response to a steam line break were also This analysis is performed to determine the most severe long term made.
,. A pressure and temperature condition in the containment following a pipe break.
The blowdown calculation was done using RELAP4-M007. The input deck used was
.Ti the same one as that used in the recirculation line break with the break
,e location moved to the main steam line. Three break sizes were run, a 0.01, W
The 0.634 sq. ft. break represents the area of a l[jf 0.1 and 0.634 sq. f t.
double-ended break of the 8 inch steam line.
44 The initial conditions and boundary conditions for this analysis were d
g selected to satisfy the requirements of the Standard Review Plan section
]'l
.4 The following is a listing of the initial conditions and a sumary of the assumptions used in the analysis.
.)
6.2.1.
. n'
- gg The reactor is operating at 102% of design power at the time the steam g
1.
line breaks.
A complete loss of normal offsite AC power occur simultaneously with the 2.
pipe break.
The reactor is assumed to go subcritical at the time of accident 3.
initiation.
?'
The sensible heat released in cooling the fuel rods and the core decay
[
4.
The rate heat are included in the reactor depressurization calculation.
~ '
of energy release is calculated using a conservatively high heat transfer
.q
- 't
-4 4
,.l 43 f
____.________y_
i 4
This maximizes the heat coefficient throughout the depressurization.
)
Calculations of heat transfer from surfaces removal rate from the core.
For sur-exposed to liquid were based on nucleate boiling heat transfer.
I faces exposed to steam, the heat transfer calculation was based on forced if
,IQ convection.
fd The main steam isolation valves are assumed to be closed at the J'5 l5.
initiation of the accident.
(q Th2 feedwater flow is assumed to be closed at time zero.
yl The vessel depressurization flow rates are calculated using a discharge ~}
6.
d coefficient of 1.0, with Henry Fauske correlation for subcooled and Moo y]
7.
A 14.7 psia back pressure was assumed l!
correlation for saturated fluid.
The
.a to maximize the mass and energy release throughout the blowdown.
- as blowdown calculation using RELAP4 was run until the mass and energy
,h At this time the blowdown rate was assumed J/f release rate stabilized.
l constant for a conservative length of time to ensure the reactor vesse Then 1.2 times the ANS decay heat curve was used.
]
was depressurized.
Emergency core cooling system was not modeled since sufficient
.f information was not provided in the RELAP model obtained from Dairylan 8.
9
- y Power Cooperative.
34 The results of this analysis are the time dependent mass and energy 1y l
e release rates presented in Table 3, 4, and 5. The RELAP 4 code ana yses wer 2
f 0. 01 ft caconds and An cacnnds for tha
$5 enrei A not ta 2nn creands, 1Fn 2
From these points the blowdown
';I 0.10 ft and 0.634 ft breaks, respect 1veTy.
2 i
~.J l break and was conservatively held constant until 4000 seconds for the 0.01 ft nf 2
2 and 0.634 ft breaks.
p*
400 seconds for both the 0.10 ft 1
Containment Response to a Main Steam Line Break __
4 q
4.1 j
The input data for the containment response calculation consist of the ]
ti t
mass and energy release rates to the containment and the available con a nmen
'd The containment heat sinks modeled are described in Tab p
d l
y heat sink data.
The mass and energy release rates were taken from the blowdown rates pre e,
e in the section 4.0.
.A9 l
9..!
4
[
~
i 1
The containment response calculation was made using the same CONTEMPT model used in the recirculation line break analysis, section 3.1.
The initial conditions used in the analysis are sununarized in Table 2.
4.2 Containment Response to Main Steam t.ine Break Results
)
Figures 3 and 4 show the containment pressure and temperature responses, respectively, for the 0.01 ft break. These curves serve as an upper bound since the blowdown was held constant from 200 seconds to 4000 seconds. Inspec-tion of these curves indicates that the design pressure of 52 psig and temperature of 280*F would be reached in 56 minutes and 53 minutes, respectively. This is the amount of time that exists for the reactor operator to take appropriate action such as to manually actuate the containment sprays or the Automatic Depressuriza-tion System (ADS). These actions having been taken and become effective by 53 minutes will insure that the containment will not exceed its design pressure and temperature limits. The staff believes that the Lacrosse plant reactor operators would be able to respond within this time frame to take action which would termi-nate the transient and limit the containment response to within allowable design values.
2 The response to a 0.1 ft steam line brehki are shown in Figures 5 and 6.
The calculated transient in this case reflects a peak post accident containment pressure of 21 psig and a temperature of 220*F. The response to the 0.634 ft break is shown in Figures 7 and 8.
This represents a double ended break of the eight inch steam line. The calculated transient reflects a peak post accident pressure of 33 psig and a temperature of 250"F. Neither the 0.10 ft nor the 0.634 ft break require operator action due to the fact that these larger break sizes cause the primary system to depressurize in a much faster time frame than the smaller 0.01 ft break.
}
2 Assuming operator action as discussed above for the 0.01 ft break, the peak containment pressures are substantially below the design value of 52 psig for all three cases. The post accident temperature for the 0.634 ft steam line break re-c!
suits in the most severe temperature conditions for a main steam line break but l
is still less than 265*F resulting from a recirculation line break.
! l I
t A
..v.n _
- = - -. - ~ - -
-~
s l
1 a
..u-l
- )
e s
5.0 Conclusions Based on this review of the Lacrosse docket and the subsequent analysis j
performed by LLNL, it is conclud6that the Lacrosse containment design pres-sure meets current NRC criteria. The containment atmosphere conditions as a; result of re' irculation line break provides the most severe temperature co,- <
1 l
~
ditions for equipment qualification of sa[fety,related equipment.
e
,4 6.0 References s
i b
1.
NUREG 0485, Vol. 3, No. 4, March 1, ~1981, Systematic Evaluation Program.
f l
La Crosse BWR Hazards Sumary Report.Dairyland Power Coope.ative,19 f
2.
k f
I
'[
i
',/
p f,',
4 4
i
,e 4
I a
e i
t a /
=
- s l
,r '
ik e
1 s
s y
y.-
E e
t f
t
'4
(
i Is j._ ?
f
, l 0 <* '\\
/-
I#'
y i
-e
~" - - - -- --- -
~m_x_______
m
1
~
a:.
F Table 1.
Recirculation Line Blowdown Mass and Energy Release 2
i Data (3.59 ft Break)
Time Flow Energy (seconds)
(1bm/sec)
(Btu /1bm) j
, 0.0 21180.
593 l
, 0.1 21180.
593.
j 0.2 11473.
555.
i 0.4 13203.
578.
l 0.6 15308.
576.
0.8 16255.
571.
I' i
1.0 15618.
566.
1 l
1.2 14320.
564.
1.4 11985.
561.
2.0 11504.
593.
2.2 13413.
593.
2.8 5440.
634.
I~
3.0 10412.
608.
3.4 5270.
599.
4.0 8104.
615.
4.5, 4178.
693.
5.0.
2856.
665.
6.0 3330.
635.
7.0 1158.
687.
8.0 1374.
615.
9.0 928.
608.
10.0 428.
695.
10.5' 330.
684.
30.0*
5.77 1200.
s 100.0 5.27 1200 5"
400.0 3.75 1200.
1000.0 2.95 1200.
i*
4000.0 2.03 1200.
g#
10000.0 1.54 1200.
e
- Assume steam decay heat cure at 30.0 sec. m,-_
r t
Table 2.
Initial Conditions for Containment Response 2'
Calculations (taken fronIReference 2) i c,
[.(([
Containment 3
264,160 ft Net free volume Ni 14.7 psia aey Pressure 80 F r,
. t.4 0
Temperature
- 3) q 100 percent
,j Relative humidity 2
2827. ft Liquid pool surf ace area 0
100 F Outside air temperature
- 'kE 1.0 Heat transfer multiplier YM 1.0
. wl Mass transfer multiplier li ',Y, fi
- k. ;-i.,' r;.
g o:q 5: "Q T4 k
~'
L.,
}.
ts-4
_'[
ii
-v f
- ,): s s*
r
"'*r
.A b'.
th.$d
- - :: pi
, :S,U ',,
s.
$r g
he P,.~,..'<
g 4 p.
,p-
< ~
1 '4
- 7..
ij. '..
pp.3
- f. '
i
O.
i:::
Ea r.1.
- 3 i
.l
}
~ - - -
2 Table 3.
Main Steam Line Break Mass and Energy Release j
A Rate Data (0.01 sq. ft. break) z 4
o 4
/
Time Flow Energy t
y (seconds)-
(1bm/sec)
(Btu /lbm) l '. - '
cs V-
.c 0.
28.0 1187.
T 4
1.
28.0 1187.
2.
27.8 1186.
1 3.
27.7 1184.
, f a; 7 4.
26.9 1183.
M.
4
, ;g ';jj 5.
26.6 1181.
., f, }., '
!,' Cd.i.
~1 10.
25.9 1183.
15.
25.7 1183.
4 p[,"m{jQQ 25.6 1183.
[
_ 20.
30.
25.4 1184.
]
1185.
G 50.
25.0,
24.7 '
1185.
60.
4 %
80.
24.3 1185.
~
b. ::..;.j+y 90.
24.1 1186.
95.
24.0 1186..
pc, -Q E.. e UM 100.
23.9 1187.
. ww 150.
22.9 1189.
scj N [ p t es..m
)(aA:.:.
21.9 1190.
4000.
21.9 1190.
it; s..yy, e.
[: 's,'Ijd : $l
. q r'p;
- s... i !,'k,
...e[,.t
.5
/Fl
. t? -llh
- i i-
_y j
,;I D
4
.^
[' %,q
. - ;4 a n.:w
' 2 Il N.
r-
- .4 hi, " Mp t
m.y h
,)
~
Table 4.
Main Steamline 8reak Mass and Energy Release Rate Data (0.1 sq. ft. break)
Time Flow Energy (Seconds)
(1bm/sec)
(Btu /lbm) 0.0 257.9 1177.
1.0 250.4 1175.
2.0 250.4 1175.
3.0 241.5 1175.
4.0 234.7 1178.
5.0 229.4 1183.
6.0 224.4 1185.
7.0 221.7 1185.
8.0 218.0 1185.
9.0 215.8 1186.
10.0 212.7 1186.
15.0 214.8 1112.
20.0 253.9 946.
25.0 246.1 925.
30.0 218.4 959.
35.0 178.6 1068.
40.0 156.0 1131.
45.0 149.4 1149.
50.0 137.9 1131.
60.0 113.5 1201.
70.0 101.2 1202.
80.0 90.1 1202.
90.0 81.1 1202.
100.0 73.5 1202.
110.0 66.8 1201.
130.0 55.8 1200.
150.0 47.8 1199.
400.0 47.8 1199.
401.0*
3.75 1200.
1000.0 2.94 1200.
4000.0 2 80 1200.
10,000.0 1.54 1200.
Assume constant out to'400 seconds the,n steam decay heat curve.
l'.
~
~-
Table 5.
Main Steam Line 8reak' Mass and Energy Release Data
(.634 sq. ft. Break)
Time Flow Energy (Seconds)
(1bm/sec)
(Btu /lbm) t O.
1540.
1153.
0.1 993.8 1091.
0.2 682.2 1064.
i 0.3 645.6 1108.
0.4 626.6 1129.
0.5 623.6 1137.
1.0 599.6 1143.
2.0 549.6 1147.
3.0 515.2 1150.
4.0 495.2 1152.
5.0 460.2 737.
6.0 770.6 710.
7.0.
724.8 728.
8.0 707.6 755.
~
9.0 618.8 791.
10.0 522.4 827.
15.0 508.0 753.
20.0 737.0 698.
25.0 327.4 814.
30.0 498.6 858.
35.0 281.0 907.
40.0 209.8 987.
45.0 223.8 939.
50.0 183.0 1003.
60.0 104.0 943.
400.0 104.0 943.
401.0*
3.75 1200.
1000.0 2.94 1200.
r 4000.0 2.80 1200.
10000.0 1.54 1200.
Assume constant flow out to 400 seconds then steam decay heat curve.
_ = _ _ _ _ _ _ _ -
i 4
Table 6.
Containment Structural Heat Sinks (Taken fro:n Reference 2)
A. Heat Sink Descriptions i
- 1. Containment dome I :l '
.a.
Surf ace Area, f t2 5670.
h Composition, thickness ft i
0.05 Steel 0.159 Insulation 4
2.
Misc. steel equipment surf ace Area, ft2 39300.
Composition, thickness ft 0.0417 Steel fjl 3.
Shadow shield h
2 14620.
Surf ace Area, ft
!d Composition, thickness ft 0.75 Q-Concrete g.
.N 4.
Outside biological shield h,
Surf ace Area, ft2
- 4710, M.
Composition, thickness, ft 5.5 Concrete r:
E 5.
Ceilings Surface Area, ft2 3600.
A Composition, thickness, ft 1.0
};
Concrete M
!\\
6.
Floors I5 Surface Area, ft2 3600.
r i
Composition, thickness, ft 1.0 j
Concrete I
t 7.
Walls t
Surf ace Area, ft2 14380.
Composition, thickness, ft 2.4 Concrete r -
y Containment Structural Heat Sinks (continued) 5, Tdble 6.
r 8.
Pump room wall Surf ace Area, f t2 1090.0
/
Composition, thickness, ft 2.4 cf Concrete J
.;:h B.
Material Properties Thermal Volumetric Heat
.2 Material Conductiyity Capacity
.(Btu /hr ft2 0F)
(Btu /hr ft3 op) 4 j.;
27.00 58.80
,j oteel 0.9202 22.62
'd Concrete 0.020 2.0
.:q Insulation 53 i.-.hk j., ~ '
' '5 s
NOTE:
j All heat sinks modeled as rectangular slabs with one side exposure to the 1.
containment atmosphere and the other insulated.
4 Heat transfer based on Uchida correlation throughout the transient.
%e 2.
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Figure 5.
Con'tainment Pressure Response to a O.10 sq. ft. Main Steam Line Break I
Ndl+a' Id kh,%yt.$ThK'h.
M, _
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