ML20031D587
| ML20031D587 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 09/25/1981 |
| From: | Prelewicz D CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.), NUS CORP. |
| To: | |
| Shared Package | |
| ML20031D553 | List: |
| References | |
| ISSUANCES-OLA, NUDOCS 8110130497 | |
| Download: ML20031D587 (8) | |
Text
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STATEMENT OF DANIEL A.
PRELEWICZ O
CONCEREING SPENT FUEL POOL TOILING My name is Daniel A.
Prelewicz.
I reside at 6901 Keats Court, Rockville, Maryland.
Since June 15, 1978, I have been employed by NUS Corporation, an engineering ser-l vices firm in Rockville, Maryland.
I am cur rently Manager of the Safety Analysis Department at NUS.
Prior to my current position, I was a Senior Engineer in the Thermal and Hydraulics Development Secticn at the Westinghouse Bettis Atomic Power Laboratory in West Mifflin, Pennsylvania.
A resume which describes my background and qualifications is attached.
INTRODUCTION In his affidavit responding to Christa-Maria Con-tention 8 and O'Neill Contention IIIE-2, Mr. Blanchard described the situation in which the spent fuel pool cooling system is cendered inoperable following a Loss-of-coolant Accident (LOCA).
Should this situation occur, the water in the pool will slowly heat up.
As the temperature of the water increases, heat will be absorbed by the pool concrete walls and floor.
Heat will also be dissipated from the pool through the pool surface primarily by evaporation.
Should these mechanisms be insufficent to remove all of the heat
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generated by the spent fuel, the water in the pool will
.8110130497 811005 PDR ADOCK 05000135 T
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_ O eventually reach a boiling condition, and the pool walls and floor will begin to heat up.
This causes thermal stresses in the walls and floor.
Therefore, it is necessary to analyze the effect of this condition on the pool walls and floor.
In his affidavit, Mr. Sacraro discusses this analysis which he has performed based on the thermal gradients described below.
SPENT FUEL POOL BOILING A conservative calculation of the time to reach boiling has been performed by assuming that no heat is removed from the pool by any means.
I further assumed in a conservative manner that the temperature of the fuel pool water at the time of the accident is the full-core offload 0
steady-state temperature of 101 F.
These assumptions yield a minimum time to reach boiling of 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />.
When boiling begins, steam bubbles will be generated in the fuel racks and rise to the surface of the pool.
This occurs when the water temperature reaches the saturation temperature corresponding to containment building pressure.
As noted by Mr. Blanchard, the containment conditions will f
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return to ambient within a few hours after the accident.
Therefore, the pressura in the containment is taken at 14.7 I
psia.
At this pressure, the water will boil when it reaches the saturation temperature of 212 F.
The temperature will
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not increase beyor.d this value provided that the pressure
. s does not increase.
As the fuel pool heats up, heat will be j
transferred from the spent fuel pool water to the concrete walls and floor.
This heat transfer will result in tem-perature gradients in these structures.
Eventually, a steady-state condition will be reached wherein the tem-perature of the walls and floor no longer changes with time.
The walls and floor will reach steady-state conditions in approximately eight days.
There will, however, be a spatial temperature gradient in the wall with the temperature de-creasing with distance from the wall / pool interface.
As the concrete pool walls / floor heat up and expand, stresses will be generated within the pool structure.
The thermal conditions which lead to these stresses must be determined for use in a stress analysis which is require 6 to evaluate the structural integrity of the pool.
Steeper gradients result in larger thermal stresses.
Therefore, upper bound values for the thermal gradients were determined as input for the stress analysis prepared by.%r. Sacramo and discussed in his affidavit.
CALCULATION OF THE POOL WALL TEMPERATURE HISTORY The thickness of the pool wall varies with loca-tion.
In order to bound the thermal gradient, two locations l
were considered in the analysis:
the location of minimum
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wall thickness and the location of maxicum wall thickness.
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-4 The minimum wall thickness is 2 feet and the maximum thick-ness is 6.75 feet.
These are bounding values for the spent iuel pool.
The pressure in the pool will increase with depth due to the hydrostatic head of the water in the pool.
While the temperature of the water will not rise significantly above the saturation temperature at the surface, I have conservatively assumed that the temperature of the pool water increases with depth.
At any depth, the temperature was assumed to be equal to the saturation temperature at that septh.
The saturation temperature is a physical upper limit.
Energy added to water at this temperature will go to produce steam rather than to raise the temperature of the water further.
At the depth of 21.4 feet (top of the fuel assemblies) where the wall is 6.75 feet thick, the saturation temperature is 237 F, while at a depth of 8 feet where the wall is 2.0 feet thick, the saturation temperature is 222 F.
The water in the pool serves as a heat transfer medium to the pool walls and floor.
Following the assumed LOCA and failure of the spent fuel pool cooling system, the water in the pool was taken as increasing from 101 F (the full-core offload steady-state temperature) to the saturation temperature linearly with time over a 20-hour period.
Thereafter, the temperature remains at saturation.
O
i 1 O To determine a limiting maximum value for the thermal gradient through the wall, tha pool water tempera-ture history is used as a boundary condition for the pool wall and floor temperature analysis, and it is further conservatively assumed that the temperature of the inner surface to the concrete is equal to the pool water tempera-ture.
That is, a high value for temperature and a low value for thermal resistance are used at the inside (i.e., pool side, of the concrete wall.
Similarly, a conservatively low temperature of 80 F and a natural convection cooling thermal resistance are used on the outside of the concrete wall.
Following a LOCA, the temperature inside the containment would be expected to be considerably higher than 80 F.
The temperature history of the wall is calculated with the above boundary conditions based on conduction heat transfer within the concrete wall.
A computer code, HEATING 5, was used to perform the calculation.
The results of this calculation provide the temperature of the concrete as a l
function of position within the wall and time.
Mr. Sacramo l
developed the thermal gradients required for the stress analysis from these results.
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l DANIEL A. PRELEWICZ MANAGER, SAFETY ANALYSIS DEPARTMENT EDUCATION California institute of Technology. Ph.D., Applied Mechanica,1970 State University of New' fork at Buffalo, M.S., Engineering Science,1967 State University of New York at Buffalo, B.S., Engineering Science,1966 l
Dartmouth College, Two Phase Flow and Heat Transfer in Nuclear Power Reactor Safety, Summer Ccurse,1975 1
I REGISTRATION i'
t Professionc! Engineer (Mechanical), Pennsylvania,1975 EXPERIENCE i
NUS CORPORATION,1978-Present j
Westinghouse Electric Corporation,1973-1978 Washington University (St. Louis),197D-1973 IWUS - Provides management supervision of activities in thermal hydraulics and safety analysis, i
int.luding the areas of nuclear plant transient and socident analysis, core thermal analysis, liydrodynamic loading, containment and subcompartment analysis, operational safety, and ther.
1 mal design. Lead thermal analyst on Title I design of upgrade fuel for the TREAT reactor. Developed and conducted a safety analysis training program for utility staff. Served as consultant on FLECHT.
SEASET raffood heat transfer test program. Developed computer code for analysis of natural circulation cooling of spent-fuel pools under off-nominal, two-phase conditions. Recently per.
formed analyses of blocked bundle steam flow, containment fire effects, spent fuel pool natural r
circulation cooling, BWR turbine trip, and HEL8 outside containment. Conducted technology l
surveys on computerized operator aids (SPDS and DASS), BWR core spray effectiveness, and pipe whip effects.
Westinghouse Corporation - Served as technical lead in the development and qualification of computer progre.ms, which included FLASH 6, for accident analysis of nuclear power reactors; provided consultation on the use of computer methods for safety analysis; and analyzed bench-mark problems to qualify safety relcted computer codes. Designed and procured a test loop for
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hydraulic pressure surge testing related to LOCA and chedt velve slam structural loadings, Conducted pressure surge tests, including tests of the effects of elastic and plastic coupling, and i
compared results with computer calculations. Analyzed thermal hydraulic test dets related to the i
l performance of reactor plants during accident conditions. Performed containment analysis of a heavy water test loop.
Washington University (St. Louis) - Served as Assistant Professor in Department of Applied Mathematics and Computer Science with Joint appointment in the Department of Mechanical Engineering. Co-investigator on a research project ** Concepts for a Theoretical and Experimental i
Study of Lifting Rotor Random i.oede and Vibrations," funded by the U.S. Army. In this project, j
methods were developed for identifying parameters of helicopter rotor dynamics in snelytical models based on test data. Techniques were also uoveloped for analyzing time-verying linser systems subject to random excitation. Teoching included both graduate and undergraduate a
8 courses in Engineering Mathematica, Numerical Analysis, Stochastic Processes, and Computer gO s-i i
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DANIEL A. PRELEWICZ l
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i HONORS California Institute of Technology l
NDEA Fellow (1966-1969)
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NSF Treinee(1969-1970)
MEMBERSHIPS I
American Society of Mechanii.al Engineers American Nuclear Society American Institute of Aeronautics and Astronautics i
I Sigma XI l
Tau 8etaPi 1
l REPRESENTATIVE PU8UCATIONS
" Evaluation of Flow Redistribution Due to Flow Blockage in Rod Bundles Using COBRA Code Simulation"(coauthor), EPRI NP 1662 Project 1380 2, Final Report,Jan.1981.
" FLASH 6: A FORTRAN IV Computer Program for Reactor Plant Loss-of-Coolant Acci ntAnalysis (LW8R Development Procrsm)"(enguthork ERDA Research and Development Report WAPD TM-1243, July 1976.
" Hydraulic Pressure Pulses with Structural Flexibility: Test and Analysis (LW8R Development Program)," ERDA Research and Development Report WAPD TM 1227, April 1976. Abstract published in Transactions e/ the American Nuclear Society, Vol. 24 Nov.1976, pp. 294-295.
Presented at the ANS ENS International Meeting, Washington, D.C., Nov.14-19,1976.
" Computer Experiments on Periodic Systems identification Using Rotor Blade Transient Flapping-Torsion Responses at High Advance Ratio"(coauthor), Proceedings of the Specialists Meeting on Actorcra# Dynamics, NASA Ames Research Center, Moffett Field, California, Feb.13-15,1974.
" Response of Linear Periodically Time Varying Systems to Random Excitation,"AJ.A.A. Journal, Vol.10. No. 8, Aug.1972, pp.1124-1125.
" Range of Validity of the Method of Averaging," California Institute of Technology Dynamics Laboratory ReportDYNL 102,1970.
" Mariner IV and V Disturbance Torques and Umit Cycles," Jet Propulsion Laboratory Technical l
Report 32 1305, Oct.1,1968.
" Solar Radiant Heating of a Rotating Solid Cylinder"(coauthor), Quarter & ofApp//edMethematics, Vol. XXV, No. 3, Oct.1967, pp. 324-326.
"Nonsteady Coaxial Arcs in Fully Cc':-;+1 Plasma Flow"(coauthor), A./.A.A. Journal, Vol. 5, No. 7, July 1967, pp.1320-1324. Proprinted as A.I.A.A. Paper 66-480 and presented at the 4th Aerospace Sciences Meeting, Los Angeles, California, June 27-29,1966.
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LNITE:D STATES OF NERICA NUCIEAR RDWIAIORY COMESSION IEPORE TIE A10GC SAFEIT AND LIGNSING N In the Matter of
)
) Docket No. 50-155 OIA CONSUtERS ICER COMPANY
) (Spent Fuel Pool
) lixlifistion)
(Big Rock Point Nuclear Power Plant)
)
j AFFIDAVIT OF RAWND F. SACRMO
'I District of Columbia: SS I, Raymcnl F. Sacramo, Principal Engineer, Engineerirg Mechanics Department at NUS Corporation, of lawful age, being first duly sworn, upon my oath certify that the statments ard information contained in the seventeen-page Statscnt concerning Christa-Maria Contention 8 and O'Neill Ccntention IIIE-3 ard the attached Table 1 and resume are true and v mt to the best of my knowledge and Delief.
Dceciated at Rockville, Maryland.
4 J,. n a,,
e-w~ O Subscribed and sworn to before me this 28th day of Septaber,1981.
AM M p rth a n
'ffblic in and for or Maryland and of Mtx*rmay I
My ommissicn expires
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