ML20207A230
| ML20207A230 | |
| Person / Time | |
|---|---|
| Issue date: | 10/09/1985 |
| From: | Bassett O NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES) |
| To: | Ernst M NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES) |
| Shared Package | |
| ML20207A169 | List:
|
| References | |
| FOIA-86-678 NUDOCS 8611100106 | |
| Download: ML20207A230 (12) | |
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Ross Bassett Morrison Beckn MEMORANDUM FOR: Malcolm L. Ernst. Deputy Director ld Division of Risk Analysis and Operationsfo feld r/f Office of Nuclear Regulatory Research Silberberg FROM: 0. E. Bassett, Director Mitchell Division of Accident Evaluation u Office of Nuclear Regulatory Research g
SUBJECT:
STEAM EXPL0SIONS Tel foN Enclosed is~ a copy of a comment on the frequency of massive in-vessel steam explosions ({ containment failure mode) prepared by Jorah Hopenfeld. He prepared it as a contribution to the discussion of this mode of early contain-ment failure in the Report of The American Physical Society's Study Group on "Radionuclide Release from Severe Accidents at Nuclear Power Plants." I am sending it to you so that it may be considered in the evaluation of containment behavior being prepared for NUREG-1150.
The preliminary evaluation of containment behavior for the Surry Plant, as given in Appendix B to NUREG-0956, does not include any probability of the oc failure mode in the central estimate and adopts the WASH-1400 value of 0.01 in the pessimistic case. For the reasons given in the enclosure, the probability of the A failure mode may have been underestimated and that for some plants it may represent a significant cause of early containment failure.
If you would like to discuss this matter further, please feel free to contact Mr. Hopenfeld.
Oricir.el C ' d U2 I)av$ d I. mn Ate O. E. Bassett, Director Division of Accident Evaluation Office of Nuclear Regulatory Research
Enclosure:
Coment on The American Physical Society Conclusions Concerning Steam Explosions cc w/ encl:
F. Gillespie, DRA0 J. Murphy, DRA0
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COMMENT ON THE AMERICAN PHYSICAL SOCIETY CONCLUSIONS CONCERNING STEAM EXPLOSIONS l-In-its recently completed study on radionuclide release from severe accidents at nuclear power plants (Reference 1) the American' Physical S6 fety (APS) concluded that steam explosions are not likely to cause contain.. ant failure.
This assessment comes in the wake of a similar assessment conducted ten years earlier by the Rgactor Safety Study, WASH-1400 (Reference 2), where< a much larger probability, 10~ , was assigned to containment failure from steam explosions.
The contrasting results of these two studies raises the question, what is the new information which has been generated in the intervening period which would result in such differences. The purpose of this note is to. address this question.
- Both the APS and the WASH-1dOO studies relied on' subjective judgment to arrive at their respective predictions which, in the absence of sufficient experi-mental data on the behavior of.large steam explosions, is critical to the probability estimates. However, when such judgment'is based on analytical models these models must be scrutinized relative to their underlying assump-tions and agreement with experiments. Even though some of the experts from which the APS sought advice heavily relied on analytical models, the APS study, as discussed below, presents only a cursory examination of these models.
Appendix AIII cites the conclusions of experts from Sandia and experts from the IDCOR program, all agree that steam explosions are unlikely to cause contain-ment failure in a large dry PWR. Sandia disagreemer,ts with various aspects of the IDCOR model were also cited. Although the information was available to the APS, the study makes no mention of the fact that when compared to actual data the IDCOR rrodel fails the basic requirement of giving reasonable agreement with experiments. The model employed by IDCOR was originally proposed by Henry and .
Fauske (Reference 3).
IDCOR tested model predictions with pressure measurements from large explosions which were conducted in China Lake. In these tests liquified natural gas (LNG) and water were used as the' working fluids. IDCOR concluded that since the theoretical predictions were commensurate with the experimertal data the mathematical model could be applied to predict steam explosions in the reactor system. The application of this model then led IDCOR to the conclusion that it would be physically impossible for steam explosions to fail reactor containment.
, Lawrence Livennore National Laboratory, independently of IDCOR, also compared the experimental data from the same LNG tests with the Henry and Fauske (Reference 3), model but reached an entirely different conclusion. Following a
. detailed study McRae (Reference 4) noted that the pressure rise as measured in I
China Lake represented only a fraction of the total energy which was released from the explosions-because they occurred below the water pool. McRae has
, calculated that without applying a correction for the pressure attenuation of l
the shock wave under water, and using airblast data alone, one can easily Enclosure
E underestimate the yield of an explosion by a factor of 1000 even if the explosion had occurred only 1 foot below the surface. Using the Henry and Fauske concept of premixing requirements and mar.imum thermodynamic efficiency McRae shows with a specific numerical example, how the model may underestimate the actual energy release. Since IDCOR did not correct for energy attenuation ,
in the water, their upper bound calculations of the pressure pulse could be off by a factor of 1000 or even more because some of the explosions occurred three feet below the surface as noted in the data report (Reference 5). McRae also noted that the energy release depends on the spill rate while the fluidization model shows no such dependence.
In addition to relying on the judgment of the experts from Sandia and the IDCOR program the APS conducted its own analysis of the China Lake experiments pointing out that the scale of these experiments is similar to that of reactor accidents and that the measured pressure at distance equivalent to the contain-ment vessel decays to less than one atmosphere. Although the APS notes that the tests were carefully examined no mention was made of the fact, as already discussed above, that this low pressure is sensitive to the water depth at which the reaction took place and therefore it may bear no relation to the pressure at the top of the vessel.
The APS points out, and rightly so, that unfortunately the probability of containment failure must remain judgmental. Since, as discussed above, the APS conclusion that steam explosions are unlikely cannot be substantiated by data, it is recommended that steam explosions be not excluded from risk assessment studies. It should be added that when such judgment is based on models of obvious deficiency the results must be seriously qualified, neither IDCOR or the APS met this requirement.
REFERENCES
- 1. Report of the American Physical Society of the study group on radionuclide release fron severe accidents at nuclear power plants. Review of Modern Physics, Volume 57, Number 3, Part II, July 1985.
- 2. NRC, 1975, Reactor Safety study, Study Director N. C. Pasmussen, WASH-1400, Nuclear Regulatory Commission.
- 3. R. E. Henry and H. K. Fauske, " Required Initial Conditions for Energetic Steam Explosions," Fuel Coolant Interactions, HTD, ASME (1981).
- 4. -T. G. McRae "Large-Scale Rapid Phase Transition Explosions " UCRL-88688, May 1983
- 5. " Burro Series 40-m LNG Spill Experiments," UCRL-86704, August 15, 1901.
s T l October 1, 1985 FY 1986 Program Brief Program: DAE
Title:
Corium-Coolant Interactions Fin No.: A1030 Contractor: Sandia
. Site: Albuquerque S*.a te : New Mexico NRC Technical Monitor: J. L. Telford Principal Investigators: M. Berman, B. Marshall, M. Corradini (Un. Wisc.) ,
(Consultant: R. Anderson, ANL)
Objective: To perf m designed experiments, conduct analyses, and develop or improve models for the phenomena governing coriuc-coolant interactions.
Budget Activity: 601930 Oblig: $ 585K FY 1986 Scope:
(10/1/85 - 9/30/86)
- 1. Conduct a set of designed experiments in the FITS vessel to investigate the effects of three independent variables on energy conversion ratio, hydrogen generation, steam generation, debris characteristics, quench rate, and the fraction of corium participating in the interaction.
A. The independent variables are: water depth, number of holes in a submerged stainless steel plate, and delayed trigger at time of corium stratification on the bottom of the water chamber.
(1) The two levels for water depth are: " low" is 29.7 inches; "high" is 52 inches. The depth of water is measured from the bottom of the water chamber to the top of the stainless steel plate with about 0.25 inches of water above the steel plate.
(2) The two levels for " number of holes" are: " low" is four (4) holes arranged in a " triangular" array, each 1.6 inches (4 cm) in diameter, with the hole spacing center to center of 2.1 inches (5.3 cm); "high" is seven (7) holes arranged in a
" triangular" array, each 1.6 inches (4 cm) in diameter, with the hole spacing center to center of 2.1 inches (5.3 cm). These holes are in two different stainless steel plates each of which will represent a submerged lower grid forging in their respective tests. '
(3) The two levels for delayed trigger are: " low" is no trigger; I "high" is a delayed trigger. The trigger is to be fired after the corium is stratified in a pool on the bottom of the water chamber.
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fj Fin A1030 2 October 1, 1985 B. The fixed experimental conditions are:
(1) Corium mass is about 50 kg to be verified in Item 2 below.
(2) Water temperature is saturation.
(3) System pressure is one atmosphere (1 bar).
(4) Rigid (wall and bottom) steel water chamber of a selected (fixed) diameter. This allows any explosions to expand in only one direction.
(5) Melt entry velocity is about 4 m/s.
C. The experimental matrix is given below.
Corium-Coolant Conversion Ratio Series Test Water Depth Number of Holes Delayed Trigger 1 52 in. 4 no 2 29.7 in. 7 no 3 29.7 in. 4 yes 4 52 in. 7 yes
- 2. Conduct one or two tests, not in the FITS vessel, to accurately and precisely measure energy conversion ratio under the same experimental conditions given in item 1 above, as preparation for the four tests described in item 1 above except that iron thermite melt may be used for these tests to reduce cost.
- 3. Conduct test 3D to comp ete the FITS D Series for 20 kg melts. Analyze these four test results for 20 kg melts to determine the effects of the other five independent variables. . .
- 4. During the second half of the fiscal year, conduct repeat tests from the FITS D Series.
- 5. As proposed in your 189a. the University of Wisconsin subcontract is acceptable. We estimate that the cost of this effort will not exceed $65K and that the scope will include:
- a. Analyzing test results and comparisons to model or code predictions.
- b. Corium-coolant interaction model or code predictions for experimental or full scale conditions.
- c. Corium-coolant interaction model development or improvements.
- d. Technical assistance on any aspect of corium-coolant interactions, as required.
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Fin AIO30 3 October 1,1985
- 6. Analyze the experimental conditions or approach of item 1, given above, for possible improvements; provide the written analysis to the NRC Technical Monitor before start of tests. All work, tests, and experiments not specified in this work brief, or changes in experimental conditions (e.g., as given in Items 1, 2, and 3) previously specified will be reviewed and approved in writing by the NRC Technical Monitor before start of work or testing.
- 7. Develop or adapt improved measurement systems for mixing, expansion, quench rate, energy conversion ratio, and related pressures for Items 1, 2, and 3 given above. Post test examination of debris will include characterization using scaning electron microscope pictures comparing edge vs. center for samples of debris particles. Candidate measuremer.t devices include: impact plates, accelerometers, strain gages, and conductivity or temperature probes. Measurement system effectiveness will be demonstrated before start of experiments.
- 8. Provide technical consultation on specific topics related to corium-coolant interactions and make preser.tations, as requested by the NRC Technical Monitor.
- 9. Approximately six progress review meetings will be held during FY 1986.
- 10. Provide preliminary data reports on a timely basis (e.g., within 60 days) to the NRC Technical Monitor after completion of each test for work specified in Items 1, 2, 3, and 4 given above.
- 11. Provide monthly reports itemizing project costs and progress by subtask activity and travel. Summarize project activities for the past month and identify any problems that might affect project milestones. This report will be submitted to the NRC Technical Monitor by the 10th of each month.
- 12. Provide a final draft report by September 30, 1986 for work Items 1 to 5 given above. The fir.al draft report will contain a complete analysis of all test results for Item i given above (4 tests) and item 3 given above (FITS D Series for 20 kg). For Items 1, 2, and 3, given above, a schedule is: complete the tests by April 1, 1986; complete post-test chemical, physical, and data analysis by June 1, 1986; complete draft report and issue for connents by August 31, 1986.
- 13. Provide copies of any trip reports to the NRC Technical Monitor for trips which used project funds.
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October 1, 1985 FY 1986 Program Brief Program: DAE l
1 1
Title:
Corium-Coolant Mixing Fin No.: A2260 Contractor: Argonne Site: Chicago State: Illinois NRC Technical Monitor: J. L. Telford Principal Investigator: B. W. Spencer J
(Consultant: R. P. Anderson)
Objective: To perform designed experiments and analyses for the phenomer.a governing corium-coolant mixing.
Budget Activity: 601930 FY 1986 PROGRAM: S 205K FY 1986 Scope:
(10/1/85 - 11/30/86)
- 1. Conduct a set of designed experiments in the COREXIT Facility to investigate the effects of two indeper. dent variables on corium breakup, mixing, quench rate, hydrogen generation, ard debris characteristics.
A. The independent variables are: water depth and number of holes in a submergered stainicss steel plate.
(1) The two levcis of water depth are: " low" is 23.8 inches (60.5 cm); "high" is 42 inches (106 cm). The water depth is measured from the bottom of the water chamber to the top of the steel platt with about 0.25 inches of water above the steel plate.
(2) The two levels for number of holes in a submerged stainless steel plate are: " low" is four (4) holes arranged in a
" triangular" array,0.8 inches (2cm)indiameter, spaced 3cm center to center; "high" is seven (7) holes arranged in a
" triangular" array, 0.8 inches (2 cm) in diameter spaced 3 ce center to center. These holes are in two different stainless steel plates each of which will represent a submerged lower grid forging (13.5 inches thick) in their respective tests.
l B. The fixed experimental conditions are:
(1) Delivered melt temperature is about 2800*C.
(2) The cover gas is argon,
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Fin No.: A2260 2 October 1, 1985 (3) A nonzero subcooled water temperature, to be detennined during pretest analysis.
(4) System pressure is one atmosphere.
(5) Melt entry velocity is about 4 m/s.
(6) Delivered corium melt mass is 10 kg.
C. The experimental matrix is given below.
Corium-Coolant Mixing Series Water Number Test Depth (cm) of Holes 1 60.5 4 4
2 106 4 3 60.5 7 4 106 7 D. The measurerents made before, during, or after each test include:
water and cover gas pressure and temperature, conversion ratio (in the event of an explosion), physical debris characteristics, chemical debris characteristics (including e.g., XRD, SEM, oxidation states),
hydrogen generation, quench rates, X-ray motion pictures, optical motion pictures in the pipeway exiting into the expansion vessel, and amount of breakup and mixing.
- 2. Analyze the experimental conditions or approach of item 1, given above, for pcssible irprovements; provide the written analysis to the NRC Technical Moniter before start of tests. All changes to the specified experimental conditions must be approved by the NRC Technical Monitor.
- 3. Provide technical consultation on specific topics related to corium-coolant mixing and make presentations, as requested by the hRC Technical Monitor.
- 4. Approximately four progress review meetings will be held during FY 1386.
- 5. Provide quick-look data reports to the NRC Technical Monitor on a timely basis following completion of each test.
- 6. Provide monthly progress and cost letter reports itemizing by subtask activity and any travel. Summarize project activities for the past month and identify any problems that might affect project milestones. This report will be submitted to the NRC Technical Monitor by the 10th of each month.
- 7. Provide a final report by November 30, 1986 including results from all four tests, interpretation, and model comparisens.
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O October 1, 1985 FY 1986 Program Brief Program: DAE
Title:
Corium-Coolant Mixing Fin No.: A2260
- Contractor: Argonne Site: Chicago State: Illinois NRC Technical Monitor: Task A - J. L. Telford Task B - J. L. Telford and J. Hopenfeld Principal Investigator: Task A - B. W. Spencer Task B - R. P. Anderson Objective: To perform designed experiments and analyses for the phenomena governing corium-coolant mixing.
Budget Activity: 601930 FY 1986 PROGRAM: TASK A - $ 205K TASK B - S 160K TOTAL T71s5K FY 1986 Scope:
(10/1/85 - 11/30/86)
Task A - B. W. Spencer
- 1. Conduct a set of designed experiments in the COREXIT Facility to investigate the effects of two independent variables on corium breakup, mixing, quench rate, hydrogen generation, and debris characteristics.
A. The independent variables are: water depth and number of holes in a submergered stainless steel plate.
(1) The two levels of water depth are: " low" is 23.8 inches (60.5 cm); "high" is 42 inches (105 cm). The water depth is measured from the botte:a of the water chamber to the top of the steel plate with about 0.25 inches of water above the steel plate.
(2) The two levels for number of holes in a submerged stainless steel plate are: " low" is four (4) holes arranged in a
" triangular" array, 0.8 inches (2 cm) in diameter, spaced 3 cm centertocenter;"high"isseven(7)holesarrangedina
" triangular" array, 0.8 inches (2 cm) in diameter spaced 3 cm center to center. These holes are in two different stainless steel plates each of which will represent a submerged lower grid forging (13.5 inches thick) in their respective tests.
B. The fixed experimental conditions are:
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9 Y Fin No.: A2260 2 October 1,1985 s
(1) Delivered melt temperature is about 2800*C.
(2) The cover gas is argon.
(3). A nonzero subcooled water temperature, t'o be determined during pretest analysis. )
(4) System pressure is one atmosphere.
(5) Melt entry velocity is about 4 m/s. '
(6) Delivered corium melt mass is 10 kg.
C. The experimental matrix is given below.
Corium-Coolant Mixing Series Water Number .
Test Depth (cm) of Holes 1 60.5 4 2 106 4 3 60.5 7 4 106 7 D. The measurements made before, during, or after each t$rt include:
water and cover gas pressure and temperature, conversion ratio (in the event of an explosion), physical debris characteristics, chemical debris characteristics (including e.g., XRD, SEM, oxidation states),
, hydrogen generation, quench rates, X-ray motion pictures, optical motion pictures in the pipeway exiting into the expansion vessel, and amount of breakup and mixing.
- 2. Analyze the experimental conditions or approach of item 1, given above, for possible improvements; provide the written analysis to the NRC Technical Monitor before start of tests. All changes to the specified experimental conditions must be approved by the NRC Technical Monitor.
- 3. Provide technical consultation on specific topics related to corium-coolant mixing and make presentations, as-requested by the NRC Technical Monitor.
l 4. Approximately four progress review meetings will be held during FY 1986.
I 5. Provide quick-look data reports to the NRC Technical Monitor on a timely .
basis following completion of each test.
L 6. Provide monthly progress and cost letter reports itemizing by subtask j activity 'and any travel. Summarize project activities for the past month and identify any problems that might affect project milestones. This report will be submitted to the NRC Technical Monitor by the 10th of each month.
- October 1, 1985 Fin No.: .A2260 3
- 7. Provide a final report by November 30, 1986 including results from all four tests, interpretation, and model comparisons.
' Task B - R. P. Anderson
- 1. Conduct a set of designed experiments for the Stratified Contact Mode to investigate the effects of three independent variables on gas phase pressure, liquid phase pressure, depth of mixing, and energy conversion ratio. The independent variables are: surface area of interaction, separation space, and temperature difference. Photography of the experiments will be used extensively. The materials used will be freon and water or other materials (e.g., salt or tin) as approved by the NRC Technical Monitors.
- a. The text matrix, experimental conditions, and schedule will be jointly determined by the Principal Investigator and the NRC Technical Monitors (with advice from consultants).
- b. This series of experiments will be conducted first.
- 2. Conduct a set of designed experiments for the Pour Contact Mode to investigate the effects of four independent variables on gas phase pressure, liquid phase pressure, and energy conversion ratio. The independent variables are: liquid slug height, mass ratio, compressible volume, and temperature. Photography of the experiments will be used y
extensively. The materials used will be freon and water or other materials (e.g., smelt and water) as approved by the NRC Technical Monitors.
- a. The test matrix, experimental conditions, and schedules will be jointly determined by the Principal Investigator and the NRC Technical Monitors,
- b. This series of experiments will be conducted second.
- 3. Provide technical consultation on specific topics related to corium-coolant mixing and make presentations, as requested by the NRC Technical Monitors.
- 4. Approximately four progress review meetings will be held during FY 1986.
- 5. Provide quick-look data reports to the NRC Technical Monitors on a timely basis following completion of each experimental series.
- 6. Provide monthly progress and cost letter reports itemizing by subtask activity and any travel. Summarize project activities for the past month and identify any problems that might affect project milestones. This repor will be submitted to the NRC Technical Monitors by the 10th of each month.
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- Fin No.: 'A2260- .4 October 1, 1985 s.
7.. Provide a' final report by September 30, 1986 including results from all tests, interpretation, and any model-comparisons.
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