ML19029A881
| ML19029A881 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 04/09/1980 |
| From: | Valore C Lower Alloways Creek Township, NJ |
| To: | Atomic Safety and Licensing Board Panel |
| References | |
| Download: ML19029A881 (132) | |
Text
{{#Wiki_filter:~--* *-:-:--* UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION Before the Atomic Safety & Licensing Board IN THE MATTER OF PUBLIC SERVICE ELECTRIC
& GAS COMPANY (Salem Nuclear Generating Station, Unit No. 1)
RESPONSE TO THE ATOMIC SAFETY AND LICENSING BOARD ORDER DATED FEBRUARY 22, 1980 The Intervenor, Township of Lower Alloways Creek, hereby submits the technical report of Dr. Richard E. Webb in response to the Atomic Safety & Licensing Board Order dated February 22, 1980 to wit:
"In the event of a gross loss of water from the spent fuel storage pool at Salem 1, what would be the difference in consequences between those occasioned by the pool with the expanded storage proposed by the Licensee and those occasioned by the present pool.
Dr. Webb's report consists of two parts. Part 1 is attached. Part 2 will be hand delivered by Dr. Webb to the Nuclear Regulatory Commission Offices in Washington, D.C. on April 10, 1980. Dr. Webb's qualifications have been previously submitted. TOWNSHIP April 9, 1980
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION Before the Atomic Safety and Licensing Boa~d In the matter of PUBLIC SERVICE ELECI'RIC
& GAS COMPANY Docket No~ SQ-272 (Salem Generating Station .
Unit #1) CERTIFICATE O~ SE~VICE I hereby certify that copies of Richard E. Webb 1 s ~epo~t, PArt l in response to Board order dated February 22, 1980 in the above captioned matter have been served upon the attached list by deposit in the United States mail, at the Post Office in Northfield New Jersey with proper postage thereon, this9th day of April I 1980. Dated: April 9, 1980
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Gary L. Milhollin, Esq. Richard Fryling, Jr., Esquire Chairman, Atomic Safety Assistant General Solicitor
& Licensing Board Public Service Electric &
1815 Jefferson Street Gas Company Madison, Wisconsin 53711 80 Park Place Newark, N. J. 07101 Glen O. Bright Member, Atomic Safety Keith Ansdorff, Esquire
& Licensing Board Assistant Deputy Public Advocate U. S. Nuclear Regulatory Qxrrnission Departrrent of the Public Advocate Washington, D. C. 20555 Division of Public Interest Advocacy P. o. Box 141 Dr. James C. Lamb, III Trenton, New Jersey 08601 Member, Atanic Safety &
Licensing Board Panel Samra T. Ayres, Esquire 313 Woodhaven Road Department of the Public Advocate Chapel Hill, N. C. 27514 520 East State Street Trenton, N. J. 08625 Chairman, Atomic Safety arrl Licensing Appeal Board Panel Mr. Alfred C. Coleman, Jr. U. S. Nuclear Regulatory Comnission Mrs. Eleanor G. Coleman Washington, D. C. 20555 35 "K" Drive Permsville, N. J. 08070 Chairman, Atomic Safety & Licensing Board Panel Office of the Secretary U. S. Nuclear Regulatory Corrrnission Docketing and Service Section Washington, D. C. 20555 U. S *. Nuclear Regulatory Conmission Washington, D. C. 20555 Barry Smith, Esquire Office of the Executive Legal Director June D. MacArtor, Esquire U. s. Nuclear Regulatory Carmission Deputy Attorney General Washington, D. C. 20555 Tatnall Building, P. O. Box 1401 Mark L. First, Esquire *. .. Delaware 19901 Dover, Deputy Attorney General Mr. Frederick J. Shon Departrrent of Law & Public Safety Atomic Safety ard Licensing Board Environmental Protection Section U. S. Nuclear Regulatory Carmission 36 West State Street Washington, D. C. 20555 Trenton, N. J. 08625 Mary O. Herderson, Clerk Mark J, Wetterhahn, Esquire Township of Lower Alloways Creek for Troy B. Conner, Jr., Esq. Municipal Building 1747 Pennsylvania Avenue, N. W. Hancock's Bridge, N. J. 08038 Suite 1050 Washington, D. C. 20006
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THE ACCIDENT HAZARDS OF SPENT FUEL STORAGE AT THE S,ALEM NUCLEAR POWER PLANT, SALEM COUNTY, NEW JERSEY BY Richard E. Webb March 19 :1 1980 Supplement to Author's February 27, 1979 Report
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Introduction In its Order of February 22, 1980, the Atomic Safety and Licen-sing Board of the U. s. Nuclear Regulatory Commission posed the* following question:
"In the event of a gross loss of water from the storage pool, what would be the difference in consequences between those occasioned by the pool with expanded storage and those occasioned by the present pool?"
The Board stated:
"We will accept in answer whatever measure of consequences each party sees fit to present; rowever, we encourage all to use some common measure,, perhaps the potential dose to an individual who remains at the exclusion area boundary for a given period. We expect, of course, that each party will postulate and make appropriate calculations for some specific sequence of events, including heating, possible melting, and possible dispersion mechanisms."
This author's February 27, 1979 testimony report, "The Accident Hazards of Spent Fuel Storage at the Salem Nuclear Power Plant", supplies the basic answer to the Board's question. The purpose of this supplement is to substantiate his earlier report by means of more detailed analysis and calculations. This supplement treats: (1) The quantities of radioactivity that would ultimately be stored if the high-density storage racks are installed and utilized. (2) The site contamination levels that could result following a reactor accident, which would force evaluation of the spent fuel pool and thereby cause the pool to boil d~--a "gross loss of water". (3) The time it takes for the pool to boil dry.
* (4) The spent fuel heat up potential following loss of water, which will show that the zirconium fuel rod cladding would
catch on fire. (5) The zirconium fire and explosion hazard. (6) The potential for near full release of radioactivity-- mainly, strontium-90, cesium-137, and plutonium due to overheating of the spent fuel material. (7) The potential for escape of the radioactivity into the atmosphere, by means of rupture of the spent fuel storage building, due to hydrogen explosion, zirconium explosion, molten fuel/Water interaction (steam explosion), over-pressure~ or simple venting. (8) The potential harmful health consequences of the release of radioactivity--strontium-90 and cesium-137; particular with respect to cancer risks. (9) The conceptual possibility of a formation of a fast neutron reactor and a nuclear explosion hazard with the possible release of plutonium vapor. (10) The potential harmful consequences of plutonium release. (11) The question of the genetic hazards of the releases of radioactivity. (12) The possible consequence3 of a loss of water in the pre-sent spent fuel pool, involving open, low density, storage racks; and, (13) Finally, the potential consequences of a gross loss of I water in a storage pool containing only 67 spent fuel assemblies at any one time--the original license--will be examined for comparison with the hazards of a fully loaded "high-density" storage racks containing 1170 spent fuel assemblies (the new rack capacity being applied for).
Appropriate conclusions and recommendations will be offered to con-clude this Supplement: Radioactivity Inventory The spent fuel storage capacity of the proposed high density storage racks is 1170 spent fuel assemblies for each spent fuel storage pool at the Salem Plant '(2 pools total,, one for each reactor). (~).The Salem Utility--Public Service Electric and Gas Co. (PSE&G )-- states that this capacity will hold 18 annual fuel discharges from
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the core,, and that each discharge will consist of one third of the reactor core of spent fuel. .(2} This means that 65 spent fuel assemblies would be removed annually and stored in the pool. The average fuel burnup for spent fuel assemblies is about 35,,000 megawatt*-days per metric ton of uranium (MWD/MTU). '(3) Each spent fuel assembly contains o.46 metric ton of uranium (MTU). (4) The amount 'or radioactivity per metric ton of fuel at a burnup of 35,000
': I MWD/MTU,1s as follows:
Strontium-90 9.0 x 104 curies/MTU besium-137 1.16 x io5 curies/MTU Plutonium 79.5 Kg of Pu-239 equivalent/MTU. These :rigures are proved in Calculations No. 1 and No. 2 included
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at the end of this Supplement. For fuel burnup at 25,000 ~/MTU, NUREG-0404 data results in 56.8 Kg of Pu-239 equivalent radio-activity per MTU (see calculation No. 2). (5) It is assumed that .the value is proportional to fuel burnup. Hence, the 79.5 Kg figure is an extrapolated value to apply to 35,000 MWD/MTU. To arrive at the radioactivity in storage at full capacity for each storage pool (after 18 refueling discharges to the pool), we I 7 . _ l l::.!!3.. ,,. ) use the formula: ~ L Ti e ri C X
- 46 X 65 L,. Y.a. ,
_ 1...11i. t1 n .:1. t> Where the term e T~ is the decay factor for a batch Of spent fuel in storage lf1 years. For plutoniUI]l, we simply multiply the Kg of Pu-239 equivalent by the mass of spent fuel in MTU, or .46 MTU/ass'y x 1170 ass'ys. The results are as follows: Radioactivity Inventories at Full Capacity of High Density Racks (1170 spent fuel assemblies). Strontium-90 40.8 x 106 curies 6 Ceaium-137 51. 7 x 10 ctiries Pu-239 Equivalent 42.8 x 103 K~ (42.8 tonnes
- Since the proposed storage increase of ~pent fuel is 1105 spent fuel assemblies, the net increase in Sr-90 storage would be 37.8 x 106 curies per pool, or 75.6 x 106 curies total for two pools. The NRC's Reactor Safety Study (WASH-1400) assumes that the core of a large power reactor will contain 3.7 x io6 curies of, Sr-90. (Appendix 6, p.3-3). Therefore, the proposed increase is equivalent to 20.4 reactor cores. In my original report of February 27, 1979 report (p.22) I gave a figure of 24, which the Board cited in its Order, but which is slightly in error. The true figure is 20o4.
Expressing the plutonium inventory in kilograms of Pu-239 equivalent is convenient, since contamination limits on ground of /'r.J- 2.3 fl fallout are expressed in micro-grams per square meter. Site Radiation Levels following .§_ Reactor Accident The most probable possibility for losing wa*ter in the storage pool is a severe reactor accident, which would likely result in extremely high radiation levels on the reactor site for extended periods (a year or so), and which consequently would make the plant inaccessible for maintaining the pool water cooling system. The cooling systems could then break down; and then the water would boil dry in a period of about one month. The potential radiation levels in the area of the spent fuel pool and the auxiliary building are calculated to be in the tens of thousands of rems per hour. after thirty days of decay following the postulated reactor accident. See Calculation No. 3. Since a dose of 500 rems to a person is lethal by acute radiation sickness, and since lesser doses would substantially increase a person's chance of contracting c~ncer, one can appreciate the potential inaccessibility of the reactor plant following a* reac.tor accident. To calculate the potential dose rates, it was assumed that ~00%
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of the reactor core inventory o.f .fission products were released in a reactor accident and available for .fallout. Specifically, I. assumed those radio~nuclides listed in Table l of Calculation No. 3, which were taKen from the NRC'a Reactor Safety Study. To accurately predict the fallout concentrations is not possible. Therefore, plausible as~umptions are required. Accordingly, one may assume that a fraction, f, o.f the core fission products falls out next to the 'reactor building and uniformly over a circular area o.f rad~us, R
- Allowing for decay of the .fission products (the noble gases were excluded, Of course, since they do not fall out), I calculated the .following dose rates for various .fractions, radii, and decay times .following the reactor accident:
f R Radiation Levels, r/h~ days si'ur~ Y'e/t>qu frou.t 'f'f11'ef-ol"'. 10% 200 feet 27,600 r/hr 30 days 10% 200 feet 17,000 r/hr 60 days 10% 200 feet 4,920 r/hr 180 days 100% 2000 .feet 2,760 r/hr 30 days 5% 300 feet* 6,130 r/hr 30 days 5% 300 feet* 3,790 r/hr 60 days 5% 300 feet 1,090 r/hr 180 days 1% 150 feet* 4,910 r/hr 30 days 1% 150 feet 3,030 r/hr 60 days 1% 150 feet 875 r/hr 180 days I More realistically, perhaps, one may assume the atmospheric diff'usion/ fallout equation used in the WASH-740 report. (6 For a heavy rain
fall (to wash out the fission products from the air) and large particle aizes--a heavy fallout rate--and a low wind speed (5 miles/ hr), the equations predict the following dose rates at a distance of 60 meters (200 feet) from the point of release: Heavy Fallout Rate Decay Time Dose Rate 30 days 61,300 rem/hr 60 days 37,800 :t>em/hr 180 days 10,9QO rem/hr For lighter fallout rates (30%) and a stronger wind (12.5 miles per hour), the dose rates were calculated aa follows: Lighter Fallout Rate Decay Time Dose Rate
>. 30 days 7, 560 rems /hr 60 days 4, 700 rems/hr 180 days 1,350 rems/hr The foregoing figures are the radiation levels due only to gamma rays (external body radiation). There would be the addition al hazard of inhaling the radioactivity and skin absorption, and also of local areas of more intense gamma :t>adiation. These risks and dos*e rate* potentials are incalculable.
One may question the assumption of 100% fission produce release from the core. However, an upper limit of the release substantially less than 100% cannot be established other than by full-scale reactor
destructive tests. (7 The Reactor Safetz Studz. asserts that the elements 'j.r, Nb, Ce, and Nd are not volatile. If we exclude these altogether, the dose rates given above would only be re~uced to 33% of the stated values. So, the overall conclusions are not changed. We can conclude, therefore, that a severe reactor accident wo~ld render the spent fuel pool unmaintainable for long periods of t1me--t1me enough for the pool to boil dry, as we shall see next. Time for the Storage Pool to Lose ~Water ~ Bollin~ In my February, 1979 report (tea timony) I as s~erted that the pool could lose its water by boil off following a breakdown in the cooling system in about 4 days to 2 weeks. These figures are sub-stantiated by detailed calculations. in Calculation No. 4. These
#'fl calculations show that-1 a reactor accident occurring any time during a yearly reactor operating/refueling cycle, the water would boil down to the tops of the fuel in 13 days or less, or less than six days if the reactor accident occurs within 30 days following a refueling. The time to boil down the pool ~o where 90-~ of -the fuel height is uncovered from the start of the loss of pool cooling is 21 days or less--10.5 days, if the reactor accident occurs within
. :fr11H1 tler'et1C't-0 Y' t:1rcr'tlt>V1 £ 30 days following refueling. Since the radiation levels.,.( could be
enormously high du1*1ng such a period, 1 t would have to be a::rnumed s.--te that the spent fuel would be uncovered well before the radiation A levels decayed enough to permit maintanence of the pool(s). 'rhe above figures assume a normal refueling discha1"ges _of 65 spent fuel assemblies and a pool full of spent fuel (1170 assemblies). If a full :reactor core were discharged into the spent fuel pool, the heat load could be about double, decreasing the time-to-boll-dry by 50%, making matters worse. Removing the core from a reactor eliminates the possibility of a reactor accident in that reactor; but the adjacent reacto~ could still suffer an accident. Though, it may seem that the probability of a reactor accident occurring during storage of a full co1"e dinchai-*ge with a pool nearly fully load is very low, it must be recognized that econom..tc pressur;;s would be strong to operate the other reactor despite safety problem, if one :reactor were shut down and its core removed to its pool. twe So, it could be a3sumed that probability of a reactor accident A. occurring following a full core discharge from the other re=i.ctor ls,. high. As will be elaborated on in the next sections, the danger poj_nt of spent fuel heatup following loss of water cooling is 900°c tem-perature--the point at which the zirconium cladding would ignite in air to initiate a self-sustaining fire in the storage pool~ (8 An upper bound of the fuel heatup rate is calculated by assuming adiabotic heat up. See Calculation No. 5. The average heat gener-ation rate per spent fuel assembly would be 1.0 KW for one year decay from the last refueling in a fully loaded pool At this rate, assuming the newer spent fuel dissipates its heat to the older spent fuel (a doubtful assumption), the adiabatic heat up to 900°c would
require 2.6 days or less. If the heat of new spent fuel does not diffuse well to old spent fuel, the heat up to 900°c could take only about 12 hours following fuel uncovery for one year decay since the last refueling. For less than one year decay of the most recent spent fuel discharge, the heat up time would be less than 12 hours. Clearly, the adiabatic heat up times are short. For non-adiabatic heatup--accounting for heat removal by air convection and thermal radiation--the heatup would be slowed but evidently not significantly, according to analysis by Sandia Laboratories (see page 54 of the Sandia report discussed in the next section). Sandia finds that the spent fuel heat-up to the danger point of 900°c would be of the order of 14 hours for one year minimum decay time for spent fuel stored in the pool. It remains, therefore, to calculate the heatup potentials following a loss of pool water. Spent Fuel Heatup Following Loss of Water/Exposure to Air As mentioned before, the fuel rod tempeI*ature of 900°c is the point at which the spent fuel cladding is expected, and should be assumedJto ignite in air (zirconium fire). (8 The.question addressed in this chapter is whether the spent fuel would heat up to the zir-conitim ignition temperature following loss of pool cooling or water.
/Qpse It will be shown that regardless of the time,.1 at which the accident occurs since the last placement of' "new" spent f'uel f'rom the reactor to the pool--the minimum decay time for heat decay of the spent f'uel in storage--zirconium ignition would occur for the applied-for re-racking design, and for any case of high density storage (10.5 inch center-to-center spacing) in which conceivable mitigating design changes in the high density rack and spent fuel storage might be made.
To calculate the fuel temperature excursion following a loss of pool water (boil off), however, is a major mathematical task which has yet to be adequately performed by the industry and Government. The work of Ao Benjamin, et al., of Sandia Laboratories, as reported in NUREG/CR-0649, Spent Fuel Heatup Following Loss of Water During Storage, March 1979, is the only available official analysis of the spent fuel heatup potential following a gross loss of water from the storage pool. (This report is referred to herein as the Sandia Report.) The report analyzes, by calculations, the spent~fuel heatup for the case of high density storage of pressurized water reactor fuel (Zion reactor/Westinghouse design) in a reactor site pool, assuming primarily perfect ventilation of the storage pool building to discharge the heat of the spent fuel (heated air) to the outside atmosphere. (The report's analysis applies to the applied-for license for ~pent fuel storage compaction for Salem, since the Salem reactors are Westinghouse designs, and since the high density rack design is practically the same as the proposed rack# design for Salem: 10.25 inch cell-to-cell pitch (center-to-center spacing versus 10.5 inch for Salem.) The data in the Sandia
report indicates that for certain worst-case conditions, the spent fuel would heat up to the zirconium fire temperature, and suggests that design improvements for spent fuel storage could be made which may preclude a zirconium fire. It addresses the important consider-ation of imperfect building ventilation and reports that this factor would greatly worsen the heatup; but the report does not definitely conclude that the storage building ventilation is not adequate to prevent severe spent fuel overheating. Therefore, Sandia's calcu-lations need to be supplemented to investigate the whole range of possibilities and conditions for spent fuel pool heatup following loss of water in the pool at Salem. This chapter is intended to supply such a supplement. Firstly, the Sandia analysis makes the conservative assumption that the spent fuel pool will contain a whole reactor core load of fuel recently* discharged from the reactor after extended power operation in addition to the spent fuel stored in the pool from regular, past refueling discharges from the reactor. A whole core discharge maximizes the heat up pot.ential, by increasing the rate of heat generation in the pool by 40% to 110% compared to regular storage, depending on decay times. (By regular storage, I mean that which would be normal storage with the high density racks.) This difference is substantial, so that one cannot merely assume that the Sandia analyses based on whole core discharges applies to the case of regular storage. On the other hand, the whole-core discharge assumption can provide an approximate, upper bound pre-diction for the regular storage case. However, the Sandia Report (p.54) compares their predictions of the spent fuel temperatures
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- during heatup for the cases of whole core discharge and regular
("normal") storage and finds that the difference between the two cases is insignificant after one year decay since the last placement of fuel into the pool. But whether the difference remains insig-nificant for lesser decay times(of the most recent batch of spent fuel placed in the pooJ)cannot be determined from the report. It is of primary interest to determine whether the spent fuel zirconium cladding would ignite with only one-third core refueling batches of spent fuel stored in the pool--the regular situation of spent fuel storage under the present plans for spent fuel storage at Salem (and at all other reactors). It may be that the spent fuel heatup under the condition of' regular storage may be substantially less for lesser decay times (when the loss of water event occurs); but* calculations presented in this present report indicate that probably there should not be any substantial difference between the case of whole core discharge and that of regular storage.* Secondly, the Sandia Report does not analyze the particular accident of most concern: spent fuel heatup following the break-down of the pool cooling systems. Rather the Sandia Report merely assumes rapid drainage of the pool water (Presumably by sabotage, ear.thquake, or dropped spent fuel shipping cask.) It treats two cases: (1) complete drainage, so that natural_ (free) convection air flows may enter the bottom of the spent fuel and pass up through the fuel; and (2) incomplete drainage, where the sudden new water level exposes most of the fuel, but plugs the inlet flow holes at the bottom of the storage racks, thus depriving the fuel of free air circulation through the fuel. The case of complete .drainage is
amenable to conventional fluid flow and thermal calculation, though the rigorous computational effort is formidable. However, t.;he caue of incomplete drainage (bottom inlet air flow blockage) is not a straight forward physical problem, for the heated steam rising up through the spent fuel assembly tends to inhibit cold ai.r flow entering the fuel through the top opening! and, hence, any model constructed to make a prediction would be purely hypothetical,* requiring experimental verification. Sandia analyzed the complete drainage case by conventional natural air circulation/convection theory--cold air entering the bottom of the spent fuel, and rising up the spent fuel assemblies upon heating, and exiting at the top~ See Figure l
- For the case of incomplete drainage, the Sandia Report provides only crude bounding calculations to estimate whether spent fuel would overheat. I shall now review and comment on the Sandia analysis and results as presented in their report.
As previously mentioned, Sandia did not attempt to calculate_ the case of boil-off of the pool water~-that is, a transient analysis of the boiling off of the pool water and the consequent and con-current spent fuel heatup. It is important that this case--the most likely process leading to a gross loss of water, since it is credible for the case of a severe reactor accident and resultant site con-tamination)--be analyzed. The analysis, however, would be a formid-able task. A completely rigorous model would have to include the following factors: (1) heat removal by steam exiting the tops of the fuel; (2) free air convection heat removal, entering the tops of the fuel;
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c'E\:1'::.::.?l~~~f!:.1:\.~;;;,;,~:..:c'.0:.1""-~E~*~'"-'l!i!;!w:,~-:::. ,'Ji::. ~-L'.t,c11&'lJ:1'!.::it'.'.°!t'.'.\';1k'i~:!'t~'."'.:~.:.t-L:;~::°"':t:.t't.~~~~"'"~"""'""'::****.** e . (3) thermal radiation heat dissipation at the tops of the fuel;
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(4) heat dissipation from the building exterior surfaces; (5) thermal radiation heat transfer from fuel rod to fuel rod, and from spent fuel holder to holder; (6) vertical heat conduction in the spent fuel, cladding, and holder; and lateral heat conduction across fuel rods and holders; (7) lateral air convection within fuel assemblies; (8) thermal radiation streaming within and out from the voided, flow channels between fuel rods; (9) *postulated building leak rates or ventilation flows; (10) heat transferred to the water; and walls of the pool and into the ground; etc. As there would be about 338,000 fuel rods in the storage pool, there would be at least 100 million nodes for a mathematical finite-difference mesh to attempt such a rigorous calculation, since the fuel rod length would have to be divided into, say 20 segments, and each segment divided into. say 8 rings, four quadrouts each. Though I'm not abreast with the state-of-the-art of compulational science, I strongly doubt that such a calculation is practical. Therefore, a rigorous upper bound of the heatup potential would have to be developed by calculational methods and heatup theory which are at least practical. As previously noted, an upper bound of the adiabatic heatup time to the zirconium ignition temperature (which is only slightly greater for a steam atmosphere, namely, ll00°c) is 4.9 hours for
spent fuel aged one year--the time period between refuelings (see . Calculation No. 5). This is to be compared with the time per:tod of 8.7 days to boil the water down from the tops of the fuel to 10% of the fuel height (neglecting vertical heat conduction within the rods and holders 3 and thermal radiation streaming, which should be neglectable. See Calculation No. 4). Clearly, this shows that there may be the potential for heating up the zirconium to ignition temperature (9oooc in air/1100°c in steam) well before the water can boil away from the fuel to allow air convection cooling by air entering the bottom of the fuel. The Sandia Report presents calcu-lations to predict the heatup for a partially drained pool, which provides some insight into the potential for overheating in a boil-down process. These calculations attempt to put bounds on heat dissipations, and show that the heat generation rate could exceed the heat removal rates, implying a severe.heatup potential. How-ever, the Sandia analysis assumes a pool full of relatively new spent fuel--all one year decayed fuel--instead of the actual case of successively increasing ages of spent fuel batches in the pool (1,2,3,4,5,6,---years aged) The effect of the assumption is that the total heat generation in the pool would be about five times the actual case planned for Salem for a minimum decay time of one year. This assumption may not be- conservative as a more rapid boil off of th~ water might lessen the time when the air inlets are blocked by water. Therefore, until a more refined analysis of a boil down transient is performed, we may assume t?at the heatup in a boil down transient would be worse than Sandia's crude sudden partial drainage model.
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On the other hand, however, the case of incomplete loss of water or a boil down transient could conceivable be rendered no worse than the complete drainage situation. Specifically, Sandia suggests that holes could be made at various points al~ng the side walls of the spent fuel holders, which would promote air circulation as the holes are uncovered during the boil-down. However, whether this expedient would be adequate would have to be investigated by a rigorous analysis outlined above, or at least one which can be shown to bound the heat up. I am not convinced of the cooling potential by the suggested holes in the spent fuel holder side walls. Air would have to flow from the edge of the pool through the gaps between the holders., which would offer flow resistance. The Sandia Report offers no mathemati-cal analysis of the potential for cooling by the side holes. Certainly, assuming complete pool draining--natural air circu-lation through the entire length of the spent fuel assemblies-- would provide a lower bound of the spent fuel heat up potential; or at least we may assume so. If such a lower bound predicts zir-conium ignition, the possibility of a spent fuel catastrophy would be proven for the more probable boil-down accident possibility. Fortunately, natural air circulation is more amenable to analysis and calculation. The Sandia Report gives the results of calculations using its. SFUEL computer model of a loss of water accident in a spent fuel pool, assuming complete drainage/natural air circulation. The model includes most, but not all, of the essential physical processes; namely, thermal radiation between spent fuel holders (holder to holder), air heatup and flow, vertical heat conduction in the solid
materials, thermal radiation from the tops of the fuel. However, the theory does not include the effect of the resistance of heat flow internally across fuel rods and the resistance due to rod-to-rod thermal radiant and air convection heat transfer. We shall return to these shortcomings later. First, Sandia applied its theory to the case of perfect building ventilation; that is, the assumption was made that the heated air exiting from the spent fuel was complet~ly vented to the atmosphere and that cold air was admitted into the building to complete the air circulation cycle. For *the high density rack design applied for Salem, the Sandia Report indicates that for the case of a whole reactor core discharge into the pool, the zirconium would ignite for decay time: of 280 days or less (pp.55, 60).* However,, this figure might not apply to Salem. PSE&G 1 s safety analysis does not specify the size of the gap between the pool wall and the outer-most spent fuel holders. If there is no sizeable gap--Wall-to-wall 11 storage-- the air flow resistance for the natural circulation would be much greater, resulting in greater spent fuel heatup temperatures. The Sandia theory predicts zirconium ignition for decay times 700 days or less for the case of.wall-to-wall storage. However, since these results are based on the assumpt.ion of a full core discharge,. they do not of themselves prove that the zirconium would ignite for i ' regular storage. In addition, the Sandia Report suggests that the heatup potential can be considerably reduced by removing the 20 or so"control rods" (non-fuel-rods) from each spent fuel assemblies and remoying strutural obstructions to air flow in the gap between adjacent spent fuel holders. (See the figure on page .) These changes would reduce air flow resistance and thus promote air cooling. '
The Sandia Report does not calculate the combined effect of these changes; but, it does state that the fuel theory predicts that opening the space between the holders would reduce the decay time to 80 days or less for zirconium ignition (p.60). That is, after 80 days following the whole core discharge from the reactor (reactor shut down), a complete loss of water from the pool would not result in zirconium ignition. (Remember, incomplete drainage or a boil-down would be worse.) Still, zirconium ignition would be possible within the first 80 days. Again, this assumes full core discharge and perfect ventilation. The Salem spent fuel building, however, is not perfectly venti-lated in the sense of Sandia's assumption; for that would require large openings in the building. (Of course, we strive to seal the building, and control the ventilation to direct the vented air through filters in order to minimize leakage of radioactivity into the atmosphere.) The vents presently in the bu1.lding have limited flow capacity, and presumably require electric blowers to force the air flow, as the air undoubtedly is passed through a bank of air filters for normal control of radioactivity emissionse Therefore, in a loss of pool water event, the heat of the spent fuel would go to heating up the air within the building, as well as the fuel and building structure itself o Heating the air decreases its density and increases its ""1'iscosity" or drag (unlike liquids). (The vis-cosity is relatively independent of the air's density.) Thus, as the building air heats up, its heat absorption capacity per unit volume of air would decrease, due to its decreasing density, and the viscosity of the air would increase, due to its increasing temper-ature. The combined effect would be to reduce the flow of air through
..
- the fuel and, hence, the rate of heat removal from the fuel. This effect, plus the increased temperature of the air entering the spent fuel at the bottom, due to the heating up of the air in the building, would make the spent fuel heat up much greater than the case of perfect ventilation (to be shown). Eventually, the temperature of the building would rise to a high enough temperature to dissipate the heat to the outside atmosphere, neglecting a fire. However, the data given in the Sandia Report indicates that zirconium ignition and/or a fuel melt down would occur before that point. However, we can't yet definitely conclude this from the Sandia Report, since the Sandia's SFUEL theory was not applied to the Salem case of high-density storage, imperfect or no ventilation, and normal storage (no full core discharge to the pool). Hence, there is the need to supplement Sandia's analysis.
Based on my calculations, it appears that the spent fuel heatup potential is drastically greater for the real case of imperfect ventilation than the ideal case of perfect ventilation. Using a crude thermal model of the spent fuel building, I calculate that in 12 hours the building air temperature would rise to about 200°c (assuming an initial temperature of 27°c), or a rise of 173°C. (See the table on the next page; and Calculation No. *6). This assumes that the building air is vented to the outside at the rate of two complete air changes in one hour, which is standard, according to the Sandia Report. We might further assume that the ventilation system would fail in the event of a seve~e reactor accident, just as *We would have to assume the pool water cooling system fails, due to the site evacuation. But for the case of no ventilation, I
Building Air Temperature as a Function of the Time (hours) after a Loss of Pool Water Temperature Time Temperature Rise o.o hours ' 300°K/27°C o0 c 0 1.4 hours 50°c 23 c 3.6 hours 86°c 59°c 6.9 hours 132°C l05°c 9.4 hours 163.2°0 136°c 12.8 hours 199.0°c 172°C 15.0 hours 220.7°c 194°c 0 19.5 hours 258.5°c 231.5 c 22.5 hours 280.0°c 253°c 429°c 402°c ~finity Rate of "f uilibrium) Ventilation: Vfl e Vc-ll1JW'°-" Building Air Ventec/ --- ">in ~ hour. Assumptions: One year Decay of most recent spent fuel.Heat Generation in pool is 1240 KW. Building Dimensions : 40 1 x 30 1 x 30' * '
- 5 inch concrete wall. -
calculate a negligible difference: an air temperature of 202°c at 12 hours, starting from 27°c. These results assume one year decay of the most recent batch of spent fuel--one year minimum decay tirne-- .and a fully loaded pool (1170 assemblies). The steady-state,
. 0 equilibrium temperatures are 497 C for no ventilation and 429°c with the standard ventilation rate (see Calculation No. 6). We can conclude, therefore, that the building air temperature would rise 0
by about 200 C in a short period of time--12 hours, with a maximum potential of about 4oo 0 c! Natural air convection calculations which I have made show that the building heatup will have a disastrous effect on spent i'uel heatup, even if the mitigating measures of control rod removal and opening the space between holders were taken. Author's Analysis of Spent Fuel Heatup My theoretical calculations of the spent fuel heatup, assuming complete pool drainage, are presented in Calculation No. 7, along with the associated theoryo Briefly, I assume heat removal by air convection only: that is, I neglected thermal conduction along the fuel rods, thermal radiation heat losses and thermal radiation heat transfer laterally--rod-to-rod/assembly-to-assembly, and thermal
radiation streaming between the rods a These assumptions are justified in the attached calculations No. 8, No. 9, Nos. 10-A and 10-B, and No. 11. In No. 8 it is shown that if the heat of' the f'uel rod were to flow only in the fuel rod and the~ ends were cooled to 27°c, the temperature at the mid-height would be 900°c--the zirconium fire point--for a rod heat generation of only 5% of the decay heat rate OP* fl-re.it) for one year decayed spent fuel. The thermal conductance~of the spent fuel holder is about the same for the fuel rods; so the holder would not change this result by absorbing heat from the fuel and then transmitting it by conduction away. Therefore, one can neglect vertical thermal conduction. Thermal radiation heat emission can not be so easily neglected. Calculation No. 9 shows that emission rate, governed by the Stephan-Boltzman Law,e~fr~, is 7.5 watts/cm2 for a 900°c surface temperature, compared to 1.77 x lo-2 w/cm2 for the surface of a fuel rod after one year decay--5.1 KW per fuel rod assembly (270 rods). However, a fuel rod surface will absorb thermal radiation energy from adja-cent fuel rod and holder surfaces; so the surface-to-surface heat exchange must be accounted for, of course. Calculations No. 10-A and 10-B investigate this. Calculation 10-A derives an effective thermal conductivity of a row of parallel fuel plates representing rows of fuel rods. The effective value is found to be less than the thermal conductivity of solid uo 2 , for the assumed temperature gra~ient equal to a 900°c drop over 12 feet width of fuel assemblies-- roughly the distance from the center plane of the pool to the edge of the pool. Since this effective value is less than the conductivity of the fuel rod in the vertical direction, and since vertical con-ductance can be neglected, we can neglect it for the lateral direction.
This is confirmed in Calculation No. 10-B, which calculates the lateral heat transfer by thermal radiation from rod-to-rod and be conduction across each rod. Assuming no air convection heat trans-fer, and a pool full of spent fuel having an average decay heat rate of 1.0 KW per assembly--roughly the average for 1170 spent fuel assemblies that would be stored at Salem in each pool, (compared to 25 KW and 5 KW per assembly for 30 day and one year decay, respec-tively), I calculate that the maximum temperature at the center of the pool would be 15,84o0 c. As will be seen shortly, air convection alone would limit the fuel temperature to 6oo 0 c if all the fuel assemblies were 5 KW per assembly, or 5 times the 1.0 KW value on which the 15,850°c figure is based. Thus, we may assume that lateral thermal radiation can be neglected without appreciable error. The
.f Sandia Report presents a graph (p. 54) which supports this conclusion.
It compares spent fuel peak temperatures following loss of wate.r for the cases of whole core discharge (195 fuel assemblies* and normal (regular) discharges (65 spent fuel assemblies). (Each case would involve the rest of the pool filled with regular, annual discharges.) Sixty five fuel assemblies would admit a greater potential for dis-sipating their heat laterally to adjacent,cooler aged spent fuel than 165 assemblies, and yet the Sandia analysis found an insignifi-cant difference in peak temperature (about l0°c). Furthermore, the Sandia analysis assumes that the lateral heat transfer is fuel assembly-to-fuel assembly, not rod-to-rod, thus in effect assuming infinite lateral conductance within a fuel assembly--an assumption which is not supported in the report. So, the Sandia mathematical model exaggerates the lateral heat transfer, and yet the difference
PWR SPENT FUEL. 17xl7 PIN ARRAY 1600 HIGH DENSITY STORAGE CONFIGURATION .. 1611 DOWNCOMER AND LARGE BASEPLATE HOLES , 1400 1 YEAR MINIMUM DECAY TIME .* OXIDATION EFFECT p 1200 INCLU OED, ALL CURVES WORST CASE LOADING FULL CORE DISCHARGE NORMAL DISCHARGE 400 FULL CORE DISCHARGE CONTROL RODS REMOVED 200 OL-~--1~~--'-~~--'-~~--i-~~---~~--i 0 4 8 12 16 20 24 .
.TIME AFfER POOL DRAINAGE (hrs)
Figure 16. Effect of Fuel Loading and Control Rod Removal on Heatup of PWR Spent' Fuel in High Density Con-figuration, Well-Ventilated Room
~-,,,,,~~,*~'"'"~':~~~~-~*-"~ ~ . . between whole core discharge and normal discharge was found insignifi-cance. Note: In my original testimony of February 27, 1979, p.11, Qf'f't1"S . I state that this Sandia assumption to be "crucial". My subsequent t1 of vz,,t'He research revealed that this statement.1 is in error and thus should be ignored. It was based on an original prediction that the peak air temperature in the spent fuel could not be less than 1300°c regard-less' of how low the decay heat level is. I interpreted this result as attributed to a stalling effect, and that therefore the thermal radiation must play a crucial role in heat dissipation. However, I subsequently uncovered an extraneous root in the mass flow momentum equation, } d? ::=. o , which my earlier computer calculation con-verged on. Careful re-checking uncovered this error. My corrected theory predicts monotonically decreasing peak air temperatures with decreasing decay heat levels, as 011'1!? ..uo~1c1 t'wtv/t,'vt!lr e('f",,otPc't-. Finally, there is the question of the thermal radiation streaming up and down the gaps between fuel rods and out from the ends of the assembly. However, this effect too may be neglected, because of the close packing of the,fuel rods. Calculation No. 11 shows that the effective thermal conductivity for vertical streaming is less than the conductivity of the solid rod; so it may be neglected, since thermal conductance within the rods can be neglected. The streaming of thermal radiation from the tops of the assemblies will be appreciable; but the effect is limited to the top 2 cm of the rod (0~5% of the assembly length), because the solid angle subtended by the openings at the ends of the fuel coolant (air) channels to points deeper down into the fuel assembly is negligible. Overall, therefore, one can assume air convection alone and
. ~~~-~m-~~<i~~~~i~:t~:::t:~'.:i;;::rm:;'."[f::~;;f".~~.,;*~:'~\:~'I".':r3::-~~r,~':'ffl-:~mf.If~r~;:'.x""t~r:;"'~~A7Yl"~tr".~~~~*:f5f:~~~,,,,.,,,,",~'--.-'*"*"~**""~*
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make reasonable predictions of the spent fuel heat up potential. This greatly simp~ifies the calculation, by allowing analysis of a single fuel rod/air channel to.represent the spent fuel. This I have done in Calculation No. 7--"Spent Fuel Heatup Calculations". My model assumes steady-state, that is, the heated air exiting from the fuel assembly is cooled by a cooler (venting). I assume a resistance-free downcomer for allowing the cold air to circulate .
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to the inlet of the fuel air channels. This would be space between
/ .
the edge of the pool and the outer most fuel holders. However, PSE&G's safety analysis does not indicate the width of this gap. If the gap is small, the resistance could be substantial, according to the Sandia Report,, page 55, figure 17,, which is reprinted on the next page. See the curve labelled "wall-to-wall" storage. The curve indicates a zirconium fire for a two year minimum decay time for spent fuel in the pool,, and thus a fire .hazard at all times during reactor life. Therefore, it is important to determine the gap width. My model also accounts for the flow resistance of the rods spacer grids, which is neglected in the Sandia analysis, and for the six inch inlet hole at the bottom. The results of my calculations are displayed in the following two figures, ~and 3, and in the tables in Calculation No. 7. Figure a_ compares my predictions with those by Sandia. As is seen, the agreement is excellent (There was no fudging of data. The input data are simple basic data on properties and fuel design, and are given in Calculation No. 7.) The results show that for most of the refueling cycle year, a zirconium fire hazard would exist* *out to 270 day minimum decay time. The Sandia Report sug-
1400--~~-.-~-,-~r-~r--r-r-r-r-~~~---~-r--T-1"'~~~........-. I - PWR SPENT FUEL, 17x 17 PIN ARRAY FULL__~o~~-D! S_CHARGE-lo.AfilN~ 1200 (ANNOTATIONS GIVE STORAGE CONFIGURATION AN BASEPLATE HOLE SIZE. REFER TO Fl GS. 2 AND 3 FOR DESIGN D TAILS)
"l (YEARS) 2 1000 - HI DENS.
p
- ---OXIDATION EFFECT I
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MINIMUM DECAY TIME ( DAYS) Figure 17. Summary of Heatup Results for PWR Spent Fuel,- Well-Ventilated Room
- I ) l and wit~ ~ontrol rods inta;ct. As shown in Figure 17, the variation is c~~tic~l deca~itimes over the cases considered is extremely large, ranging: :from well under 10 days for the open frame configuration td nearly 2 years for the high density configuration ~ith wall-'to~wall placement. These results should be.co~sidered in coritext with the fact that according to current practice,/ . decay. ,times *'
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geats that the maximum decay time for a fire could be reduced to 80 days, if the space between the holders were opened to allow for air convection in those spaces. My analysis confirms this--see 0 . Calculation No. 7,. p. 19. I calculate a 957 C peak clad temper-ature for 80 d.ay fuel assuming the .953 inch gap between holde~s would increase the flow area and associated "hydraulic diameter" of the air channel between fuel rods. Th*~ Sandi*a Report further suggests that the maximum decay time for a fire could be also reduced by removing the control rods from' the fuel assemblies, to promote air flow (25% of the 289 rods in the assembly are control rods and the rest are fuel rods, except for an instrument rod); but the report does *not quantify the matter. It notes that for one year minimum decay of spent fuel in the pool,, the peak temperature would be reduced 200°c--from 610°c to about 41o0 c. I calculate a 112°c drop, and a reduction of the maximum decay time for a fire from 270 days to 200 days. The Sandia Report does not consider the combined effect of the two changes. I calculate that the com-bined effect of control rod removal and gap opening is to reduce the maximum decay time for a fire hazard to 70 days (beyond which the temperature would not reach 900°c). Whether such changes are practical is another question, for removing control rods increases the re-criticality hazard. We can conclude, therefore, that even if the mitigating measures were taken, if practical, and if the building were perfectly venti-lated, and if a large enough gap at the edge of the pool were pro-vided, a fire hazard still could not be prevented for 20% of the time of reactor operation (time between refuelings); namely, for a period of 70 days at least since fuel discharge from the reactor.
However, perfect ventilation would require opening the building to the outside; but this is obviously not an acceptable health and sa.fety measure. As previously discussed, the building's ventilation system would not be adequate to vent the heated air from the spent fuel. In 22.5 hours, the building temperature would rise to 280°c (over 500°F). I factored the building air heatup into the spent fuel.air convection theory by assuming successively greater tem-peratures at the air inlets to the spent fuel assembly. Figure 3 plots the results for one year decayed spent fuel. It shows that for' the PSE&G's applied-for storage design (assuming no wall-to-wall storage), the peak fuel temperature would rise from 6oo 0 c to the fire temperature (goo0 c) for only a 50°c increase in the building air temperature--from 27°c (83°F)to 75°c (167°F)--Just 12% of the building air heatup potential. Even if the control rods were removed and the gaps between the spent fuel holders were cleared of air flow obstructions,, the spent fuel would heat up to a fire temperature (900°c compared to a 280°c level for perfect ventilation) within a days time. See curve #5. Again, this is for a minimum decay time of one year for spent fuel *in the pool, high density racks, and free downcomer. These are the important results which the Sandia study does not reveal. It can be concluded, therefore, that the applied-for high density stor~g~ will present a zirconium fire hazard at all times during the, reactor life, due to possible loss of pool cooling/water consequenc'e o~ a reactor accident, or other causes of a loss or water. Indeed, it might be that for the present rack design--open-racks, 21 inch center-to-center spacings--a fire hazard will exist, due to the:imperfect ventilation of the building. This question
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will be addressed later (chapter ). It must be remembered that the above analysis is optimistic: (a) No wall-to-wall storage was assumed; (b) The case of water boil down--water blocking air inlets-- was neglected; (c) No whole core discharge was assumed; (d) No debris exist in the fuel to block air flow; and (e) No thermal distortions occur to block air flow. Also, the analysis is theoretical. There is always the possibility that an actual heatup ~ould be worse than predlcted--a point always to be kept in mind. See my February 1979 report. In conclusion, the applied-for spent fuel storage re-racking will carry a 100% probability of a zirconium fire following a.loss. of water in the storage pool! The consequences of such a fire will be inquired into next!
Consequences of Zirconium Fires Fission Product Rel~ase Zirconium alloy fuel rod cladding is extremely reactive sub-stance. It's fire and explosion hazard is extreme and is well enough documented to prove this point. See the U.S. Atomic Energy Commission report; ffZirconium Fire and Explosion Hazard Evaluation", TID-5365, August 7, 1956, and The Technology of Nuqlear Reactor Safety, Thompson and Bekerly, Vol. 2, Chapter.17, "Chemical Reactions .. , pp. 45)-456 on zirconium air reactions, by L. Baker and R. LilMATAINEN. (zirconium is used as the fuel in photo flash bulbs~ according to Dr. Earl Gulbra,den
.. of the University of Pittsburgh--20 mg per bulb (in a pure oxygen atmosphere). At 5.1 x 10 4 gm of Zirconium per spent fuel assembly, each storage pool would contain the equivalent of J billion:
flash bulbs) The heat. of reaction,. zr"t* .02-7 Zr02 t- Q(heat), is - Q ~ 262 Kcal/mole of Zr. Calculation No. 12 shows that the thermal equilibrium temperature of a given mass of Zirconium, plus a mass of air needed to oxidize the mass of Zr, and the mass of U02 in pro- . portion to the U021,...... Zr ratio in the spent fuel rods, is )447°c, neglecting vaporization of the uo 2 and Zr and neglecting thermal radiation and convective heat losses., Therefore, the potential exists for the fire heat alone to melt the spent fuel, since the melting temperatures of U02, Zr, and Zr02 are 285o 0 c, 186o 0 c, and _2700°c respectively. Should the fire start and stop at the tops of the fuel, the melting and/or crumbled fuel rods and Zr02 particulate could plug up the air channels in the fuel rods, thus* depriving the spent fuel of air convection heat removal. Calculation No. 13 shows that the radioactive decay heat cannot be dissipated by conduction
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"alone, so the rest of the fuel would overheat and the spent fuel would collapse into an uncoolable rubble and a melt. Seventy two percent of the fuel would be molten, assuming the crust does not tend to sink in the melt. The question iss What would be the con-sequences of such a fire in terms of fission-product releases that is, what fractions of the Sr-90 1 Cs-lJ?, and Pu would be released? N-etwu iue
~~Sandia Report,* nor any other Government and industry report,
- addresses and treats this question. In this chapter it is argued that the zirconium fire could burst the spent fuel building,--tAeii:-
opening it up to the outsider'""'and then v*rtually
- al~ Sr and Cs could smolder out of the spent fuel--a possibility which cannot be ruled ca~ he out , nor~ the probability4 shown to be less than 100% of a full release **
Calculation No. 12 shows that if the zirconium reacted only with the oxygen present in the building (no replenishment), the heat of the reaction can potentially heat up the nitrogen (depleted air) to a pressure of 132 pounds per square inch (psig). This assumes that the zro 2 and uo 2 in proportion to the Zr consumed would absorb the heat of reaction as well, according to their heat capacities .* wori"ttt (About 7.4 spent fuel assembliesA of zirconium would be consumed, . . out of 1170 in the pool, or .6% of the pool load of Zr.} If o.64% of the spent fuel pool load of Zr reacted with water_(residual water or water from a ruptured pipe), enough hydrogen would be generated (Zr+ 2 H2 0-'? Zr02 f- 2 H2 ) that upon detonation in the air in the building, a pressure of 659 Psi could potentially result. (See Calcu-lation No. 14.) The spent fuel storage building is not designed for such accidents 1 and I assume that it could not take more than Lo psig pressure without bursting. * (Moreover, the heat of the fire would
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weaken the building. ) We wetrl:d estimate the explosion potential for
/I molten/fuel/zirconium interacting with water (a physical, steam explo-sion), upon sudden water entry after a firej and severe explosions
.J.v s /"'!If can be predicted. The heat of ~{4ee!en of U02 ls 282 joules per gram, and its specific heat is about .4 j/gmoc. ThVs, the thermal energy per gram of molten fuel is about 1400 joules/gmz 282 +-. 4~ 28oo 0 c == 1400.
- If we assume four spent fuel assemblies worth of molten fuel interacted with water, we'd have a potential mechanical energy release of about 1 1400 'A
- 4-6 >( 106 X 4 X, 2 = 5.. 16 ~ 106 joules~
assuming 20% conversion efficiency ** (Each spent fuel assembly con-tains
- 46 ~ 10 6 grams of U02, about. ) Since 2 ><, 106 joules= 1 lb. of TNT, the explosion potential would be 260 lbs of TNT equivalent. This, of course, is a crude calculation; but it does show a potential for rupturing the building. No further refinement is practical and be credible. Therefore, we must conclude that once the zirconium ignites that it will completely rupture the spent fuel building. .The large openings would then sµpply fresh air to feed a sustaining Zr fire and the drafts would be the mechanism for the fission* products escaping into the atmosphere.
No one can reliably predict accurately the force of the explosions or pressurizations of the building as a result of the fire, the course of the fire, and the fission product release)to prove that the release of fission products would not be full. If one merely assumes that the spent fuel collapsed and rested on the floor of the pool without any chemical reactions occurring with the overheated concrete or Zir-
- f us er:/
conium, and that the fuel would be a solid mass of fuel, the Sr and Cs* twrouJJi <:< .sci/cl J.Hq t-Y'i x would have to diffuse great distances,,.! to the surface of the mass: /153 clll1.
. ~*
; w u/c /,,, r V)( 17 /e) fl83 &RH** In a computer model of the processA and using the diffusion parameters for Cs and Sr in U02 given in the Battelle report BMI-1779, p.111, I found that 0.0% of the Sr and 5% of the Cs would be released after 1000 hours (41 days)--a very low release. (See Calculation No.
15.) This is due to the cooling of the medium near the surface by thermal radiation and air convection. The Cs and Sr diffusion co-efficients near the surface are low enough, due to the relatively low temperature near the surface as to impede the diffusion of these substances. Deeper into the interior, the temperatures, and hence, the diffusion 'coefficients would be high, but the concentration gradients would be lows hence, little diffusion. *This calculation however, can only be considered a lower bound. The process will be extremely complex and.unpredictable. The Sandia Report "Core-Melt down Experimental Review", SAND 74-0J82, August 1975, discusses, for e'ft"a1111pf-e,, Gt. <:<ud J reactor core meltdown~ the production 01carbon dioxide gas from 4
* .Ivr tlte sPewt f<>~f >11elt concrete decomposition. A the co 2 could (would) be formed and blow through the spent fuel melt/debris, carrying with it fission products Cs/Sr. The non-molten part of the rubble w~uld be porous,allowing Sr and Cs evolving from high temperature regions to pass up through the debris ~pores. Indeed, the fuel particles (broken pellets) need not melt for high rates of release of Sr and Cs from the particles.
Based on available experimental data of Sr and Cs release fractions from uo 2 rods at high temperatures given in appendix C o~ Appendix VII of the Reactor Safety study, I calculate for fuel rods at 2000°c, well below the melting temperature of 2850°c, that 76% of the Sr-90 would be released in one day,and, of course, all of the Cs, since it is more
-tJ.te volatile,* One could make a .model ofA ~process, which I might *I used the f'ormula F.:: 1 - !-
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attempt. *But the results would still be founded on many hypotheses. There is 'the C02 evolution, the smoldering/quiescent fire of zirconium. Surely, the Cs and Sr and other, shorter lived fission products, would migrate toward the surface. The low temperatures region of the *medium would be confined to a thin layer at the surface (about 5 inches thick), Though the fission products might at first condense on the surface of the pores of the medium, they would carry with them the heat source, which would tend to raise the temperature in the layer. Also, the fission products wo~ld only be coated on surfaces, not mixed in the solid matrix. However, a surface fire, if on-going or recurring, could then vaporize the condensed fission products that migrated to the surface. Furthermore, the solid material above the melt would so tend to sink in the melt, I assumer an~what effects would occur are purely speculative. There would be convection currents in the melt. Furthermore, one cannot say how the spent fuel might end up after the zirconium fire and explosions. A porous mass of fuel containing zirconium could reignite and the trapped nitrogen and/or hydrogen* (from Zr/H20 reaction) could. explode the debris and disperse it. A final fire engulfing the debris could then complete the fission pro-duct release. The zirconium fire and explosion hazard domi~ates the processes by which the fission products would be released. As Baker and LIIMATAINEN i ' of Areonne Laboratory concluded, there is "No satisfactory theory exists which describes the course of such fires"r namely, fires involving "aggregates\) *o f Zirconium pieces. They state ~hat "the behavior of larger fires are extremely complex processes." (p. 4j6).
/vJ / '1.S-S-There bl the large Zr fire.... at the Bettis Atomic Power Laboratory in
West Mifflin, Penn *. to consider, in which 159,000 pounds of zirconium were involved--that's tRe- equivalent to 700 spent fuel assemblies worth of Zr. See the AEC's Serious Accident Bulletin No. 84, August 15, 1955. According to the AEC report, the Zr was stored in a number of bins, and the fire spread rapidly from bin to bin, and initial flames reached 80 feet into the air. Those who observed it said that the fire produced a sound like a jet plane noise, indicating severe, high velocity air/nitrogen drafts. Un-estimated but substantial quantities of the Zr and Zr02 (reaction product) were found throughout the plant property (which is a huge laboratory). Dr. Benjamin Lustman, to/tf ffl~ a zirconium expert who worked at Bet*tis at the time_,/\that the workers who saw the fire characterized it as "spectacular" with "severe drafts", and "enormous winds" and bright light like a magnesium flare.* Any model to predict the course of a spent fuel fire,and its consequences in terms of fission product release fractions)would, therefore, be highly hypothetical. Only a real spent fuel pool loss of water accident or full-scale experiments would establish tJ\e fission product release potential; and many such experiments or accidents would be needed, because of the complexity of the processes and their haphazard nature. Of course, such experiments are not practical *. Therefore, it ~be assumed that 100% of ~e Sr-90 ~d Cs-137 are released into the atmosphere. the plutonium is It ought also to assume that 100% of released as well. Next, I shall estimate the radiation dose effects and potential
- It is noted that a f 1re of a reactor spent fuel element has already occurredt At the NRU reactor (See The Technology of Nuclear Reactor Safety, Vol. I, P. 688). An expe.*secr8ingle fuel *element (metalic ~
uranium and aluminum clad) ignited. The efforts to suffocate the fire were d~fficultJdue to the high radiation levels in the pit (10,000 Rem/hour according to some estimates).
harmful consequences of such 100% releases of Sr-90, Cs-1)7, and Pu.
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p _ ... References for pct '('T C?u e
- 1. "Safety Evaluation by the off ice of Nuclear Reactor Regulation Relating to the Modification of the Spent.Fuel Storage Pool, Facility Operating License No. DPR-?O, Public Service Electric and Gas Co ** Salem Nuclear Generating Station, Unit No. 1, Docket No. 50-272, page 1-1.
- 2. "Description and Safety Analysis--Spent Fuel Storage Rack Re-placement, No. 1 Unit, Salem Nuclear Generating Station, Docket No. 50-2?2, Tl 79/1/;o, Revission 1 (Attachment No. 1), pp.1,4.
). A. s. Benjamin, "Spent Fuel Heatup Following Loss of Water During Storage", Sandia Laboratories, March, 1979, NUR.EG/CR-0649, SAND 77-1371,R-3, PP* 2)-24.
- 4. Id., P. 21.
- 5. Draft Generic Environmental Impact Statement on the Handling and Storage of Spent Light Water Power Reactor Fuel, March 1978, U.S. NRC, NUREG-0404, Vol. 2.
- 6. Theoretical Possibilities and Consequences of Major Accidents in Large Nuclear Power Plants, U.S. Atomic Energy Commission, WASH-740, March 1957, Appendix E.
7 .* The Accident Hazards of Nuclear Power Plan~~'- R. Webb, University of Massachusetts Press, 1976, .PP .. 82-83. Though the discussion pertains to Sr-90, the principles discussed apply to all fission products.
- 8. Sandia Report, above ref. 3, p. 47.
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- 9. Id. , p .* 6J.
- 10. Westinghouse Reference Safety Analysis Report, June 19?2, p.4.2-90 **
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