A Review of Criticality Accidents A Review of Criticality Accidents

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KEWB–B core was designed specifically to extend these measurements to a period of 1 millisecond. It was cylindrical and, during the transient experiments (up to about 5.2 $ above prompt criticality), contained 18 l of UO 2 SO 4 solution. In the KEWB systems, two quenching mechanisms seem to be dominant over a wide range of excursions. The first of these is the rise in neutron temperature and thermal expansion as the core temperature rises, resulting in a prompt temperature coefficient equal to –2 ¢/°C at 30°C. This effect is sufficient to account for the observed yield of excursions starting near prompt criticality, but is inadequate for more violent transient experiments. The second quenching mechanism is bubble formation. 100 The available evidence supports the contention that during the spike, void space, consisting of many very small bubbles (microbubbles) with internal pressures of from 10 to 1000 atmospheres, is created by the fission process. The bubbles later coalesce and leave the system, giving the observed gas production coefficient of about 4.4 l/ MJ. Growth of these microbubbles seems to involve the repeated interaction between fission fragments and existing microbubbles from earlier fissions. Thus, a quenching mechanism proportional to the square of the energy release can be invoked. This model is successful in describing the solution transients, notwithstanding imprecise knowledge of the manner in which the bubbles form and grow. While the KEWB, SPERT, and TRIGA programs were largely oriented toward reactor safety, the CRAC studies were conceived and conducted to further an understanding of process accidents. Transients were stimulated in cylindrical vessels of 300 and 800 mm diameter with highly enriched uranium solutions having concentrations from 48.2 g/ l to 298 g/ l. In most experiments, solution was added to a cylinder at a constant rate until the height required for criticality was substantially exceeded. Some experiments used a neutron source of sufficient size to initiate the excursion as soon as the system achieved criticality, while the absence of such a source in others permitted the system to become superprompt critical prior to initiation and resulted in higher spike yields. The magnitude of the spike yield correlated well with the rate of reactivity addition when a source was present. For periods shorter than 10 milliseconds, the specific peak power was found to vary as the reciprocal period to the 3/2 power, which is in agreement with predictions based on the KEWB results. 110 The CRAC program also provides helpful guidance regarding dose rates to be expected near unshielded solution excursions. For the 300 mm cylinder the exposure at 4 m from the surface of the vessel was about 2 × 10 -15 R/fission and for the 800 mm cylinder about 5 × 10 -16 R/fission. SPERT-1 reactor cores (heterogeneous, moderated, and reflected by water) 99 were of two general types. The first had fuel in the form of MTR type aluminum– uranium plates and cores designed to include the range from under moderation to the more hazardous region of over moderation. The second was composed of canned UO 2 rods about 10 mm in diameter. The uranium enrichment in these rods was 4%. Transients of the plate type reactors have been extensively studied since 1957 in an effort to solve core design problems and to find the limitations of such reactors. In particular, the period and energy release that can cause damage have been carefully determined. The shutdown of a power transient in the SPERT systems is more complicated than in simpler reactors. The model developed includes heating and density change of the water; heating of the core structure, including its own geometry changes and moderator expulsion from such changes; and finally, the boiling of water next to the plates and loss of moderator when water is expelled from the core. When the plate type core was destroyed, the reactivity, period, peak power, and fission energy release were essentially as predicted. The destructive steam pressure pulse starting some 15 milliseconds after completion of the nuclear phase was not foreseen and is thought to have been caused by very rapid transfer of energy from the near molten aluminum plates to the thin layer of water between the plates. The transfer, occurring before any significant volume change took place, and the resulting high pressure destroyed the core. This effect seems to have been involved in the destruction of BORAX, SPERT, and SL-1. The second type of SPERT-1 core 89 (4% enriched UO 2 rods in water) was tested during 1963 and 1964. Transient experiments with this core demonstrated the effectiveness of the Doppler mode of self shutdown and provide a basis for analysis of accidents in similar power reactor systems. Two attempts to destroy the core by placing the reactor on very short periods (2.2 and 1.55 milliseconds) failed. In each case, the Doppler effect was operative and additional quenching developed because one or two fuel pins (out of several hundred) cracked and caused local boiling. The pins were thought to have been saturated with water before the test.
References 1. Stratton, W. R. A Review of Criticality Accidents. Los Alamos Scientific Laboratory, Los Alamos, NM., LA-3611, (1967). 2. Stratton W. R., revised by D. R. Smith. A Review of Criticality Accidents.Lawrence Livermore National Laboratory, Livermore, CA, DOE/NCT-04, (March 1989) 3. Paxton, Hugh C. Glossary of Nuclear Criticality Terms. Los Alamos National Laboratory report LA-11627-MS, Los Alamos, NM, (October 1989). 4. Barbry, F. SILENE Reactor: Results of Selected Typical Experiments. Report SRSC n223, CEA Institute de Protectionet de Surete Nucleaire, Department de Recherches en Securite, Service de Recherches en Surete et Criticite, Centre d’Etudes de VALDUC - SRSC 21120 Is-sur-Tille, France, (September 1994). 5. Lecorche, P., and R. L. Seale. Review of the Experiments Performed to Determine the Radiological Consequences of a Criticality Accident. Y-12 Plant, Oak Ridge, TN., Y-CDC-12, (1973). 6. Dunenfeld, M.S., and R. K. Stitt. Summary Review of the Kinetics Experiments on Water Boilers. NAA-SR-7087, (1963). This report is proprietary information of North American Aviation Co. 7. Accidental Radiation Excursion at the Y-12 Plant June 16, 1958. Union Carbide Nuclear Company, Y-12 Plant, Oak Ridge, TN., Y-1234, (1958). 8. Callihan, D., and J. T. Thomas. “Accidental Radiation Excursion at the Oak Ridge Y-12 Plant—I, Description and Physics of the Accident.” Health Phys., 1, 363–372, (1959). 9. “Oak Ridge Y-12 Accidental Excursion, June 16, 1958.” Nucleonics, 16, Nov., pp. 138–140, 200–203, (1958). 10. Hurst, G. S., R. S. Ritchie, and R. L. Emerson. “Accidental Radiation Excursion at the Oak Ridge Y-12 Plant, Part III, Determination of Doses.” Health Physics, 2, p. 121 (1959). 11. Paxton, H. C., R. D. Baker, W. T. Maraman, and R. Reider. Nuclear-Critical Accident at the Los Alamos Scientific Laboratory on December 30, 1958. Los Alamos Scientific Laboratory, Los Alamos, NM, LAMS-2293 (1959). 12. Paxton, H. C., R. D. Baker, W. J. Maraman, and R. Reider. “Los Alamos Criticality Accident, December 30, 1958.” Nucleonics, 17(4), pp. 107–108, 151 (1959). 13. Ginkel, W. L., C. W. Bills, A. O. Dodd, K. K. Kennedy, and F. H. Tingey. Nuclear Incident at the Idaho Chemical Processing Plant on October 16, 1959. Phillips Petroleum Co., Atomic Energy Div., Idaho Falls, ID, IDO-10035 (1960). 14. Paulus, P. C., A. O. Dodd, K. K. Kennedy, F. H. Tingey, and F. M. Warzel. Nuclear Incident at the Idaho Chemical Processing Plant on January 25, 1961. Phillips Petroleum Company, Atomic Energy Div., Idaho Falls, ID, IDO-10036 (1961). 15. Latchum J. W., F. C. Haas, W. M. Hawkins, and F. M. Warzel. Nuclear Incident at the Idaho Chemical Processing Plant of January 25, 1961. La-54-61A, Phillips Petroleum Co., (4 April 1961). 16. Olsen A. R., R. L. Hooper, V. O. Uotinen, and C. L. Brown. “Empirical Model to Estimate Energy Release from Accidental Criticality.” Trans. Am. Nuc. Soc., 19, pp. 189–191, (October 1974). 17. Hetrick, D. L. Letter to Thomas P. McLaughlin, (14 July 1999). 18. Callihan, D. “Accidental Nuclear Excursion in Recuplex Operation at Hanford in April 1962.” Nucl. Safety, 4(4), pp. 136-144, (1963). 19. Clayton, E. D. Further Considerations of Criticality in Recuplex and Possible Shutdown Mechanism. Hanford Atomic Products Operation, Hanford, Wash. HW-77780 (1963). 20. Zangar, C. N. Summary Report of Accidental Nuclear Excursion Recuplex Operation 234-5 Facility. HW74723, Richland Operations Office, AEC, TID–18431 (1962). 21. Investigation Committee. Final Report of Accidental Nuclear Excursion Recuplex Operation 234-5 Facility. Hanford Operations Office, HW-74723, (August, 1962). 22. Clayton, E. D. “The Hanford Pulser Accident.” Transactions of the American Nuclear Society, 46, pp. 463–464, (June 1984). 23. Nakache, F. R., and M. M. Shapiro. The Nuclear Aspects of the Accidental Criticality at Wood River Junction, Rhode Island, July 24, 1964. Supplemental Report, United Nuclear Corp., New Haven, Conn., Fuels Div., TID-21995 (1964). 24. Kouts, H., et al. Report of the AEC Technical Review Committee, (Nov. 6, 1964). 25. Daniels, J. T., H. Howells, and T. G. Hughes. “Criticality Incident-Aug 24, 1970, Windscale Works.” Trans. Am. Nuc. Sec., 14, pp. 35–36 (1971). 111
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LA-13638 Approved for public releas
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A Review of Criticality Accidents 2
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PREFACE This document is the second
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Tom Jones was invaluable in generat
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C. OBSERVATIONS AND LESSONS LEARNED
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FIGURES 1. Chronology of process cr
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51. MSKS and assembly involved in t
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A REVIEW OF CRITICALITY ACCIDENTS A
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3 Black Sea Baltic Sea Obninsk Norw
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North Atlantic Ocean Irish Sea Wind
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1. Mayak Production Association, 15
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time of the accident. However, the
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determined by a radiation control p
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consequences of this accident, the
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Figure 10. Drum in which the 1958 Y
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The entire plutonium process area h
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During the next shift, radiation sa
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9. Siberian Chemical Combine, 14 Ju
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The accident investigation determin
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10. Hanford Works, 7 April 1962 18,
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heater and stirrer were turned off
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vessel 64-A was added. The dissolut
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additional reflection afforded by t
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15. Electrostal Machine Building Pl
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ased on a radiation survey, that th
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0.5 m 0.5 m 0.5 m 0.5 m 3.0 m 1.4 m
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The investigation identified severa
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19. Idaho Chemical Processing Plant
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20. Siberian Chemical Combine, 13 D
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To Glovebox 6 7 8 6 From Glovebox 6
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on a 0.4 m square pitch grid. The b
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competing reactivity effects proved
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Figure 35. The precipitation vessel
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B. PHYSICAL AND NEUTRONIC CHARACTER
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