ML20030A357
| ML20030A357 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 11/14/1961 |
| From: | CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | |
| References | |
| NUDOCS 8101090389 | |
| Download: ML20030A357 (40) | |
Text
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SECTION 13 MAXIMUM CREDIBLE ACCIDENT
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- 13. 1 GENERAL CONSIDERATIONS i
- 13. 1. 1 The safety analyses up to this point have dealt with measures designed to protect against release of dangerous amounts of
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radioactive materials from the system and have shown the very high degree of assurance provided by the plant design against accidents that could create radiation hazards. In the i
extremely unlikely case that, in spite of all the precautions, an accident releasing hazardous amounts of radioactive materials from the reactor should occur, the relbased i
materials would be contained within the containm'ent vessel housing the reactor.
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- 13. 1. 2 The " maximum credible accident" (MCA) for the Big Rock Point reactor is defined as an instantaneous rupture of the primary coolant system, creating an opening of 7. 06 square t
fe e t..
This is equivalent to twice the flow area of the largest coolant pipe, namely, recirculation pump discharge line, connected to the reactor vessel. The rupture of this line is taken as the basis for the MCA because a large rupture'of the reactor vessel is not considered credible on the basis of judgment presented in paragraph 13. 2 below.
13.1. 3 To indicate more fully the scope of the potential effects of such an accident, it is assumed that the enclosure leaks at the maximum permissible rate (0. 5% per day at the design pressure).
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- 13. 1. 4 The postulated accident should result in radiological conse-quences of little significance because the core spray system is designed to prevent signifidant Eore damage in such I
an event, thus limiting fission product release to a small fraction of the fission product inventory in the core. However, in order to illustrate the effe'ctiveness of the various barriers k
against release of fission products, the MCA analysis includes examples of varying degrees of effectiveness of the core spray.
The spectrum of consequences analyzed includes those in which:
(a) core spray is fully effective; (b) core spray prevents damage to 90% of the fuel; andJ(c) core spray is assumed to be unavail-able to cool the core, and the core melts down from decay heat.
13.1. 5 Although a specific event is described to facilitate analysis of the MCA, the assumed accident conditions ~should not be inter-
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preted as a specific weakness of design. Rather, the conditions chosen for analysis are judged to be at least severe enough to constitute an upper boundary for credible operating-accident
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severities and, accordingly, are suitable for demonstrating the effectiveness of containment.
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Secticn 13 Page 2 13.1.' 6 '
With respect to the case within t'ae above indicated spectrum of consequences analyzed, in which the core spray system is fully effective, there would be only a very small fraction of the core fission products released to the containment vessel.
The result of such release would have negligible effect on plant personnel and environs. Thus, this case is discussed no.fu rthe r.
It is a measure of the strength'of the Big Rock I
Point plant design that only a MCA, Ghi~ch is compounded by a second failure, such as reduced effectiveness of core spray, need receive further analysis in order to develop circumstances by which the total hazard to the health and safety of the public can be judged.
- 13. 2 BASIS FOR "MCA" LIMITATIONS
- 13. 2.1 Power Boiler Statistics
- 13. 2.1.1 Statistics on reactor vessels are not available because of f
the relative newness of this application. However, very good statistics are available on power boilers in the United States and these statistics seem generally applicable to reactor vessels. It is conservatively estimated that there are 400-500 boilers in the United States designed to ope rate at a pressure over 600 psia, and that they represent not less than 4000 boiler-years of operating experience. The first such boiler was designed over 30 years ago. An ex-amination of the Hartford Steam Boiler Inspection and Insurance Conipany power boiler statistics shows no failure of steam drums in power boilers designed to operate over 600 psia. In no case have the materials used in these drums been superior to those used in fabrication of reactor j
vessels. Statistics on all types of boilers and unfired l
pressure vessels reveal the following relevant points:
- 13. 2.1. 2 A number of boiler explosions have occurred in small low pressure heating boilers. The lack of such failures in power boilers operating over 600 psia is attributed to the additional care in fabrication and operation of these vessels.
- 13. 2. 1. 3 Those failures that did occur were often attributable to lack of operating training and supervision.
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13.2.1.4 The boiler explo: ion statistics indicate a large reduction in incidence of failure over the years, with the trend con-i tinuing in the direction of improvement.
- 13. 2. 2 Brittle Failure 13, 2. 2. 1 The following considerations regarding design, fabrication, and operating conditions for the reactor vessel provide.
reasonable assurance that the vessel will not fail in a brittle manner.
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Sectihn 13 Page 3
- 13. 2. 2. 2 The vessel steel in the high flux region opposite the core has a nil ductility transition (NDT) temperature less than 10
- F.
This low NDT temperature provides considerable urgin for radiation effects. Over the lifetime of the reactor, parts of the vessel may receive integrated neutron exposure of about 1019 neutrons per sq. cm.,
which at room temperature would not be expected to raise the NDT over about 140 to 200*F depending on the neutron energies considered. Since irradiation takes place at elevated temperatures, radiation effects are expected to be even less.
- 13. 2. 2. 3 The vessel steel is low in notch sensitivity, and vessel notches are kept to a minimum by careful design, mate-rial selection, quality control, and inspections.
- 13. 2. 2. 4 The vessel is built in accordance with requirements of the ASME Boiler and Pressure Vessel Code as modified for nuclear reactor vessels.
- 13. 2. 2. 5 High loadings are avoided while metal temperatures are in the nonductile range. (This is'an inherent characteiistic of boiling water reactors controlled on saturation conditionc. )
- 13. 2. 3 Ductile Failure
- 13. 2. 3. 1 The allowable design stress for an ASME Code vessel is less than one fourth the ultimate strength of the material.
A ductile shear type failure resulting irom loadings which exceed the ultimate strength of the material over a signifi-cant section of the reactor vessel could occur only from extreme overpressure, or shpck waves from a rapid nuclear excursion or chemical reaction.
- 13. 2. 3. 2 The safety valves are sized to prevent reactor pressures in excess of those allowed by code. The reactor safety system and reactor characteristics preclude any large nuclear ex-cursion or chemical reaction. Therefore, a ductile failure is not considered credible.
13, 2. 4 Nuclear Excursion
- 13. 2. 4. 1 The MCA is not assumed to involve a significant coincident energy contribution from a nuclear excursion for the follow-ing reasons:
- 13. 2. 4. 2 The very low probability of an excursion, considering the many safeguards incorporated in the design of the reactor and safety systems to prevent such an accident, including:
Control of reactivity insertion rates; i
Multiple scram circuitry; l
Inherent negative coefficients of reactivity; and Availability of strong procedural control.
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Section 13 Page 4 1
- 13. 2. 4. 3 If, nevertheless, a nuclear excursion did occur, none of the excursions postulated for this reactor are believed capable of initiating a primary system rupture. Thus, simultaneous occurrence of an excureion and a pipe break is not considered to be credible.
13, 2. 5 Metal-Water Reaction The primary system rupture accident is not considered to involve a significant coincident energy contribution from a metal-water reaction for the following reasons:
- 13. 2. 5.1 Metal-water reactions of safeguards interest with the Big Rock Point reactor are those involving zirconium, which is contained in the zircaloy fuel channels, and the stainless steel fuel cladding. There is no experimental evidence which suggests a probability of a reaction involving stainless steel and water. In the event that stainless ' steel fuel fla_dding is melted by dec.ay heat as a result of the postulated MCA, the process would be relatively slow and it is not expected that any appre-J ciable amount of elemental iron would be produced in particles small enough to react with water.
- 13. 2. 5. 2 With respect to zircaloy fuel ~ channels, the probability of a zirconium-water reaction is remote. It has been demonstrated conclusively that a violent reaction of zirconium and water cannot occur when the metal temperature is below the melting point. The zircaloy channels would not reach melting tem-peratures until sometime after the failure of fuel cladding and subsequent temperature rise in UO2 fuel in the event of the postulated MCA. Experimental studies indicate that this reaction cannot proceed rapidly unless the molten metal is dispersed into particles with a diameter of one millimeter or smaller. It is doubtful whether particles of such small size could be produced in view of the sequence involved in heating up to the melting temperature.
- 13. 2. 5. 3 In view of the above considerations, it is concluded that there would be no significant energy contribution from a metal-water reaction in the event of the postulated MCA. Any energy
. contribution that might be made would not affect the peak i
enclosure pressure because of the several minutes after blowdown before core temperatures reach melting levels.
Thereafter, the small amount of energy that might be con-tributed by a metal-water reaction would have negligible effect on the enclosure pressure transient.
- 13. 3 BASIS FOR CONTAINMENT DESIGN
- 13. 3.1 The basis for containment design was taken as that operating condition where the internal energy of the coolant is a maxi-This is the care at " hot standby" where some of the mum.
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Section 13 Page 5 primary system volume, norn1 ally occupied by steam during operation, would be filled with hot water which has a greater energy content per unit of volume.
- 13. 3. 2 The analysis of the " maximum credible accident," on the other hand, is based on reactor conditions where the fission product burden of the core is at a maximum. Since this condition only exists during full power operation, total system energy at the time of the accident will be less than that used as a basis for containment design. Thus, the peak enclosure pressure resulting from a primary system rupture during full power operation will be lower than the peak pressure used as basis for containment design (20 psig, compared to 23 psig).
As indicated in Section 3 of this report, the actual design pressure.was established at 27 psig at an early stage in plant design.
- 13. 4 "MC A" PRESSURE CALCULATIONS
- 13. 4. 1 Peak Pressure Calculations 1
- 13. 4.1.1 The largest process line to the reactor...which would give the most rapid release of total system energy to the con-tainment vessel, is assumed to suffer a complete instan-l taneous circumferential break with all primary coolant l
issuing from both sides of the break. Accordingly, the 20-inch recirculating pump discharge line is assumed to break at a point near the bottom of the reactor vessel.
- 13. 4.1. 2 Pressurized hot water partially choked by steam formation initially flows from the break, followed by steam after exhaustion of the water. The characteristics of the designed system are taken into account in calculating the flow rates associated with blowdown of the system contents.
- 13. 4.1. 3 No energy contribution from a nuclear excursion is included on the basis of conside rations given in paragraph 13. 2. 4.
- 13. 4.1. 4 No energy contribution from a metal-water reaction is included on the basis of considerations given in paragraph i
- 13. 2. 5.
- 13. 4.1. 5 The conditions in the containment vessel free space, prior to the accident, are assumed to be 100*F temperature and 100% relative humidity and atmospheric pressure.
- 13. 4.1. 6 The calculations of the resultant peak pressure are based on an assumption of thermal equilibrium between water, steam, and air in the containment vessel free space.
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- 13. 4.1. 7 The free volume (9. 4 x 10 cu. ft. ) assumed in the contain-ment vessel was that occupied by atmospheric air prior to i
the accident.
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Secticn 13 Page 6
- 13. 4.1. 8 Heat is added to the free space by cooling of the fuel during the time interval between the break and peak pressure.
- 13. 4.1. 9 Heat transfer to the cold masses within the containment vessel is assumed to occur using heat transfer coefficients of. 700 2
and 240 Btu /hr-ft
- F.for' steel and concrete, respectively.
l Heat transfer to these cold masses is assumed to occur from the time of the accident until these surfaces are in thermal equilibrium with the containment vessel atmosphere.
Heat losses to the outside of the contaimunt vessel are assumed not.to occur until after the containment vessel i
pressure has reached its peak and started to decline.
- 13. 4.1. 10 On the basis of the above assumptions the peak pressure resulting from a primary system rupture during reactor operation at 240 Mwt has been calculated to be about 20 psig, and occurs approximately 16 seconds after the break, and is shown in Figure 13.1.
- 13. 4. 2 Post Accident Pressure Reduction
- 13. 4. 2. 1 After the initial pressure peak in the " maximum credible accident," pressure and tempe rature in the enclosure would undergo changes with time as a function of the following competing mechanisms:
(a) There would be heat losses from the enclosure atmos-phere to the enclosure shell, to the solid structures in
.the enclosure (which, on the average, would be initially at a lower tempe rature than the vapor space), and to the environment outside.
(b) There would be heat gain from radioactive decay, and from cooling of the primary system metal masses.
- 13. 4. 2. 2 The calculated enclosure pressure and temperature transient curves are shown in Figure 13.1.
As indicated by these i
curves, operation of the post incident spray system and core spray system results in containment vessel temperature and pressure subsidence sooner than would be the case in the event neither ope rated. Both of these spray systems would i
be brought into service automatically in accordance with the descriptions given in earlier sections of this report.
- 13. 4. 2. 3 As indicated in Figure 13.1, the post incident and core spray systems operate to bring the pressure down to slightly above l
atmospheric pressure after about 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> following the rupture.
It was assumed that after about 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> the core spray system would begin recirculating water frcm the containment vessel through the system's heat exchanger. The water level in the containment vessel would have risen to about elevation 587 feet, which would be sufficient to provide suctibn head for the core spray recirculation pumps. The method of operation, then, is that the post incident and core spray systems would be operated in combination, with the water being supplied from the fire-
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protection system for about the first 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />, followed by water recirculation through the core spray system while the post inci-dent system continues to spray fire protection system water for
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approximately an additional hour. 'This method'of operation would have used approximately 1/3 of the available capacity in the containment vessel for water accumulation under the
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given conditions. Thus, intermittent operation of the post incident spray system after the initial 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> period would allow control of the containment vessel pressure for a longer period of time without recirculating coritainrhent vessel
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water, which may be highly contaminated with fission products,. through the post incident spray system.
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- 13. 4. 2. 4 As described above, both the post incident and core spray systems are assumed to operate, however, the principal purpose of the core spray system is to protect against fuel damage rather than affect tha containment vessel pressure 3
transient. Operation of the core spray system would be auto-matically initiated when the reactor system pressure dropped to 200 psig, which would occur about 15 seconds after the sys-i tem rupture. The additional cooling provided by the core spray system (in comparison to non-operation of the core spray system), woula not reduce the containment vessel pressure 4
until after the post incident sprays operate, because this initial core spray cooling capacity would be effectively used in removing heat stored in the core and primary metal masses.
In the event that core spray system cooling was unavailable to the reactor core, the stored heat in the core and primary
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metal masses would be released to the containment vessel atmosphere at a slower rate; this rate was found to be dependent on the transit time through t; vruptured system and through
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the equipment insulation. This transit time was calculated to be in the range of about 100 seconds.
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- 13. 5 CORE HEATUP TRANSIENT
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Figure 13. I shows the effect on the Big Rock Point reactor core i
from the postulated loss of coolant accident. In order to develop circumst.mces by which a spectrum of consequences could be
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analyzed, assumptions were made as indicated in paragraph
- 13. 1. 4.
This included the assumption that cooling was unavail-able to the reactor core which would result in core meltdown,
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and the assumption that cooling was partially effective so as to limit the extent of fission product release to 10 percent of the core meltdown case. It is nettd that in connection with the assump-(
tion of unavailable cooling for the cure that the core spray system is assumed to be operable in all other respects; these conditions would result only in the event of a highly improbable simultaneous
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break of the core spray system water line inside the containment vessel. As indicated in paragraph 13.1. 6, a fully effective core spray would preclude significant core damage, which would
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result in little or no off-site radiological effect.
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Section 13 page 9
- 13. 5. 1 Cooling Unav.ailable to the Core
- 13. 5. 1. 1 When the reactor cooling water is lost from the core through the primary system rupture, the fuel rods would become blanketed with steam which would act to insulate the rods.
The power generation heat, which had not migrated out of the fuel pellets, and the heat from radioactive decay, would raise the temperature of the fuel and cladding. The analytical model for calculating the resulting core temperature distributions utilized ten core axial nodes in four fuel rod locations with a typical fuel bundle in each of five core radial regions. This model was utilized during the initial seconds of the trantient. After about 20 seconds, all core r. odes had a temperature drop across the fuel pellet of less than 200*F, and the temperature drop across the, cladding was ne gligible. With this redistribution of temperatures apparent in the fuel at this time, the initial model was replaced by a two node model, one node of the fuel and cladding and one node at the surface of the cladding. The two-node model was employed until all of the cladding reached perforation temperature (1600*F), at which time, the surface node was removed from the model, and the single node.o'f fuel and clad temperature was used to the completion of fuel melting.
- 13. 5.1. 2 The core heat was based on reactor shutdown in four seconds after the primary system break. The heat contribution from radioactive decay was based on long-term operation of the core at a steady state power level of 240 thermal mega-watts, with decay heat in accordance with the Stehn and Clancy decay power plot. (It is to be noted that this decay heat is used twice in this evaluation: once, assumed to leave the fuel to increase co ntainment vessel pressure, and here, retained to heat the fuel. )
- 13. 5.1. 3 The above analytical model gave results which are plotted on i.
Figure 13. I showing fraction of fuel rods reaching perforation temperature and volumetric fraction of fuel reaching tempera-tures over 5000*F, both as a function of time.
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- 13. 5. 1. 4 The above analysis is believed to be conservative. It is 2 uel would stop short of f
expected that the melting of UO
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a complete meltdown as a result of heat losses by radiation I
to the vestsel wall and then to the containment vessel atmos-phere, in addition to heat losses by conduction from fuel pellets which would have dropped out of the core to the f
bottom of the reactor vessel. Analytically, no credit was taken for either such radiant heat losses or such fuel i
pellets which would be free to drop once their temperature reached about 2550*F, the clad melting temperature.
- 13. 5. 2 With Core Spray Cooling l
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- 13. 5. 2. 1 In the Big Rock Point reactor, the core spray system starts automatically on simultaneous signals indicating ' low reac-l l
tor pressure and low reactor water level. Water (at an I
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assumed temperature of 50'F) wattld be pumped into the reactor at about 400 gallons per minute and 200 psig, to spray approximately 4 to 5 gpm into each fuel channel.
This amount of water provides about 1. 25 times that required for adequate core cooling on a basis of 5% of
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240 Mwt decay heat rate.
- 13. 5. 2. 2 Although it has been conventional to analyze the " maxi-mum credible accident" (MCA) on the basis of
_100%_ core meltdown, the most probable "MCA" is considered to be represented by the included 10% " melt-down" case. Considerations bearing on this conclusion 4
include: (a) even with a partially effective core spray, the total melting would be significantly less than 100%,
(b) radiant heat losses from the core would preclude some outer rows of fuel rods frorn ever reaching melting tempera-tures even without any core spray cooling, and (c) when the cladding melted, the fuel pellets would fall out of the core, thereby providing means for heat transfer by con-duction, which would preclude reaching melting tempera-i tures for those pellets.
- 13. 6 EMISSION OF FISSION PRODUCTS TO CONTAINMENT VESSEL Fission products in the reactor fuel become available to the containment vessel free volume, at a rate dependent on the achievement of high temperatures in the core. The following paragraphs discuss the bases and assumptions on which the analysis was made.
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- 13. 6. 1 Fission Product Release Froin Fuel t'
- 13. 6' l.1 As indicated earlier, the fission product inventory of the reactor core was based on long-term oper ation at a steady state power level of 240 thermal megawatts up to the time g
of the "MCA" rupture. The quantity of the various fission products in terms of curies per megawatt is consistent with ORNL-2127 information.
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- 13. 6.1. 2 The initial release of fission products would occur when
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the fuel rod perforates and allows the escape of the gases contained in the fuel rod plenum. These gaseous fission products would consist of noble gases and halogens, iodine
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and bromine. The design of the fuel cladding on the basis of experimental work at VAL, provides adequate capacity to withstand the internal pressures generated by a 50%
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release of noble gases and halogens at the end of fuel life (15,000 MWD /T) under transient conditons. The core was considered to be made up of 1/3 fuel at 15,000 MWD /T,1/3 i
fuel at 10,000 MWD /T, and 1/3 fuel at 5,000 MWD /T. Using appropriate power peaking as would occur in the core, it was estimated that an overall average of no more than 20% of the i
total core inventory of noble gases and halogens would be released at the time of clad perforation.
Section 13 Page 11
- 13. 6.1. 3 As shown in Figure 13.1, onset of fuel melting, 5000*F, occurs with completion of perforation of all the fuel rods.
Although it is believed that scme fission products would continue to'be released during the heat-up interval to 5000*F, the analytical model considered a two-step release:
initially, 20% of the gaseous fission products, and finally, the remaining 80% of noble gases and halogen groups, 50%
of the volatile solids (Ru, Cs, Te, Se), and 1% of all other solids were assumed to be released as fuel temperatures reach and exceed 5000*F. These release fractions are reasonably consistent with recent information given in a memo report, ORNL-CF 60-12-14.
- 13. 6. 2 Fission Products Available in Containment Vessel
- 13. 6. 2. 1 Certain fractions of the various fission products will move from the fuel to the containment vessel free volume depend-ing on their physical characteristics.
Experimental results have indicated that significant portions of all gassified fission products, except noble gases, would be plated out on colder surf aces of the primary system metal masses (reactor vessel and associated equipment) and the inside of the containment vessel. This analysis was made on the assumption that 50% of the halogens and 70% of the solid fission product groups would be plated out. Thus, fission product transport from the fuel to an airborne con-dition in the containment vessel free volume was in accord-ance with the following percenti.ges:
100% of the noble gases (Xe, Kr) 50% of the halogens (I, Br) 15% of the volatile solids (Te, Se, Ru, Cs)
O.3% of all other solid fission products i
1 13, 6. 2. 2 The inventory of fission products which remain in an air-borne condition in the containment vessel free volume would i
be a function of the deposition and washout which would 4
occur in the enclosure as a result of the turbulent and water-condensing atmosphere, which would prevail. This analysis considered t2.e reduction of free volume halogen fission product inventor, 'y deposition as a removal of 1 x 10 ~3 fraction per second. This deposition rate was.
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estimated on a basis of an estimated 1 cm/sec deposition velocity and an average 10 meter distance to a deposition surface. The expected natural condensation of water vapor f
and rainfall by the post incident sprays would provide many
" sticky" surfaces for adherence by the halogens and subse-quent washdown into the water sump of the containment vessel.
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The analysis also considered the effect of.different concen-trations of the water-soluble iodine in the water phase in relation to its concentration in the vapor phase. During the initial six hour period of the accident, which encompasses all of the significant leakage of halogens that occurs, the ratio of iodine concentration in air to that in the water was s
Section 13 Page 12 found to be greater than 10-4 As long as this ratio is greater than 10-4, iodine removal from air would be expected to con-tinue.on the.ba' isAf..informntion given.in, rde rancq ( AEC L-s 1130, " Iodine Containment by Dousing in NPD-II," by L. C.
Watson, A. R. Bancroft and C. W. Hoelke, October 27, 1960)
- 13. 6. 2. 3 The inventory of the solid groups of fission products which remain in an airborne condition in the containment vessel free volume is also a function of deposition and washout occurring in the enclosure atmosphere. This e Gaibnoducts considered the reduction of free volume solid fissiongf raction (volatile and all other solids) as a removal of 3 x 10-per second from the combined effect of deposition and aashout.
- 13. 6. 2. 4 The fission product inventory in the containment vessel free volume also is reduced with time due to radioactive decay, 4
which has been factorea bto the evaluation.
- 13. 6. 2. 5 The resulting fission product inventory in the enclosure vapor space (or containment vessel free volume), is give n in Figure 13. 2 in connection with examples of 100% and 10%
fission product release from the core.
- 13. 7 LEAKAGE FROM CONTAINMENT VESSEL
- 13. 7. 1 The leakage rate of the various fission product groups was determined based or enclosure free volume fission product inventory as outlined above, and the leakage rate variability due to enclosure residual pressure reduction. The leakage rates at various times after the accident are shown for the i'
noble gases and halogens on Figure 13. 3, and for the solid groups of fission products on Eigure 13. 4.
The leakage rates for noble gases and halogens are seen to be insignificant for
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about the first few minutes and then rise rapidly as a result of their release from the fuel cladding. The leakage of the solid fission product groups is insignificant, comparatively, for f
the first 15 minutes as a result of their later in-time release from the fuel. The leakage of noble gases and halogens is the most significant due to the larger fraction escaping from fuel, with leakage of noble gases continuing as long as posi-tive containment vessel pressure prevails since radioactive decay provides the only mechanism by which noble gas in-t ventory is reduced.
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- 13. 7. 2 Mass leak rates of air, water vapor, and airborne fission products from the containment vessel were calculated in accordance with the orifice flow equation given by reference i
(" Orifice Meters With Supercritical Compressible Flow,"
ASME Paper No. 50-A-45, by R. G. Cunningham,1950).
It was assumed tl'at the minute imperfections that might exist
(
in the penetrations to the containment vessel causing the postu-lated leakage can be represented by a single circular orifice.
This is believed to be a conservative approximation as most any l
other geometrical shape of an equivalent cross-sectional opening would result in lower leakago rates in connection with driving pressures in the range of 1 psig to 5 psig.
Section 13 Page 13 Figure 13,2
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Section 13 Pade 16
- 13. 8
SUMMARY
OF RADIOLOGICAL EFFECTS 13.8.1 The radiological effects of the " maximum credible accident" of interest at off-site locations are of two forms: firs t,
the direct radiation from the fission products contained in the free volume space of the reactor enclosure; and, second, r-jiossible leakage gLamnfalliractiorcof these fission products from the enclosure. In the latter case, the radiological effects would be due to direct radiation from the passing
" cloud," direct radiation from radioactive materials deposited on the ground, internal exposure due to inhalation of radio-active materials as the cloud passes the point of exposure, and possible contamination of agricultural produce.
- 13. 8. 2 Example radiological effects are shown for two basic assump -
tions involving the consequences of both 100% and 10% core meltinb. As the 10% example obviously has a higher proba-bility of more closely approaching an estimate of possible con-sequences of the initiating event, references in this text are to this example. Consequences of the less prooable 100%
example may de read directly from the figures for this section, which appear immediately following paragraph 13.17, (pages 30 through 40, inclusive).
- 13. 6. 3 The off-site radiological effects are generall'g insi ificant d
and would not cause any concern for public health and safety.
This conclusion applies specifically to the 10% core melt example, and also applies generally to the 100% core melt example.
- 13. 8. 3. 1 The direct radiation from the enclosure for the 10% core melt case at a distance of 1/2 mile from the center of the reactor enclosure is estimated to be 0.1 roentgens in the first two hours after the initiation of the accident, and an estimated O. 6 roentgens for the entire corurse of the accident.
- 13. 6. 3. 2 The direct raaiation on the center line of the passing cloud 1/2 mile from the reactor enclosure for the 10% core melt case is generally insignificant. In the firsc two hours after the accident, the estimated dose will range from less than 1 rcrad I.
for lapse diffusion conditions and average wind speed, to an estimated 10 mrads under inversior. conditions with low wind speed. During the entire course of the accident, the passing cloud dose will range from about 4 mrads for lapse conditions with average wind speed, to an estimated 60 mrads for inversion conditions and low wind opeed.
(
- 13. 6. 3. 3 Raaiation dose received from fallo.it on the ground at a distance of 1/ 2 mile from the reactor encionure for the 10% core melt case will generally be insignificant., At this point on the cloud I
center line during the first two hours after the accident, the dose from fallout is estimated to be about 1 mr under inversion conditions with low wind speed, an:1 less than 1 mr during any a
\\
example of oetter diffusion conditicans. During the ent re course i
of the acciuent, the dose fro;a fallcut will rande from about 10 mr for lapse diffusion conditions and average wind speed, to an estimatal 35 mr during inversion conditions viith low wind speed.
1 w
~--
.s.,-
y e
Section 13 Page 17 l
- 13. 8. 3. 4 The lifetime dose to the thyroid gland from inhalation of radioiodines on the cloud center line at a distance of 1/2 mile from the reactor en-closure for the 10% core melt example generally will be insignificant.
During the first two hours after the accident, the lifetime dose to the thyroid at this distance is estimated.to range from about 20 mrems during I,
lapse diffusion conditions with average wind speed, to an estimated 2 rems during inversion conditions with low wind speed. During the entire course of the accident, the lifetime dose to the thyroid is es-timated to range from about 30 mrems during lapse diffusion con-d.tions with average wind speed to an estimated 4 rems du' ring inversion conditions with low wind speed.
- 13. 8. 3. 5 The lifetime dose to the lung from the inhalation of fission products assumed to be insoluble on the center line of the pasv._g cloud at a distance of 1/2 mile from the reactor enclosure for the 10% core melt e:: ample will be insignificant. During the first two hours after the accident, the lifetime dose to lung will range from less than 10 mrems auring lapse diffusion with average wind speed, to an estimated 0. 3 rems during inversion conditions with low wind speed. During the entire course of the accident, the lifetime dose to lung will range from less than 10 mrems during lapse conditions with average wind speed, to an estimated 0. 6 rems during inversion conditions with low wind speed.
- 13. 8. 4 The radiological effect calculations of the example where it is assumed that all of the core actually melts show levels about 10 times greater than those above for the partial core melt example.
- n. 8. 5 Due to the low levels from " fallout," no evacuation of off-site inhabitants would be necessary. Some minor control of dairy cattle products may be appropriate in the vicinity. The probability of even this measure being necessary is small, as the wmd patterns shown by the meteorologi-cal studies would be expected to convey any airborne contaminants during poor diffusion conditions in directions out over the lake after passage over only a small area of, land a significant portion of the time.
- 13. 9 DIRECT EXTERNAL GAMMA RADIATION FROM ENCLOSURE
- 13. 9.1 The quantities of fission products available for direct external gamma radiation from the enclosure were based on considerations discussed in paragraph 13. 6.
The direct radiation is a sensitive function of the gamma energy levels of the radioisotopes present, because of the g
variable shielding effect for different gamma energies of the large thickness of air available between the enclosure and the site boundary.
Tharefore, the evaluation was made by calculating the exposure con-(
tribution from each gamma radiation level from each isotope in the noble gas, halogen, and volatile solid fission product categories, to-(
gether with their appropriate daughter fission products, and by con-sidering shielding and buildup factors for both air and the steel enclosure wall.
A
- 13. 9. 2 The results of the evaluations of direct radiation from the enclosure, in terms of both dose rate and integrated dose, are shown on Figure s 13. 5 and 13. 6, at distances of one-half and one mile. At the one-half mile distance, the maximum dose rate of abrut 10 mr/ hour occurs in the one to two-hour period after the accident, and rapidly diminishes thereafter.
It is noted that at distances of one mile and greater, the integrated dose is less than that received from natural background in one day. No move-ment of nearby inhabitants appears necessary. (This appears to be a reasonable conclusion for the 100% example accident, also. )
(
~~
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i Section 13 Page 18 i
13.10 METEOROLOGICAL DIFFUSION EVALUATION METHODS 13.10.1 The radiological effects of leakage were evaluated at four selected points in the atmospheric diffusion spectrum, which encompass the conditions encountered at the reactor site.
These are the poor diffusion conditions caused by inversion (stable), typical of warm weather nights, at wind speeds of both one and five meters per second; and the better diffusion conditions, typical of daytime, and represented by neutral (isothermal) and unstable (lapse) diffusion, both at wind speeds of five meters per second. Estimations of most radiological effects at other wind speeds may be approxi-(
mated by considering_the effect to be inversely proportional to wind speed. A limitation should be appreciated at dis-tances where the change in wind speed will cause a signifi-cant change in travel time, so that the amount of radioactive decay occurring becomes important. An exception exists s
in the case of deposition on the ground which, under constant diffusion conditions, is largely independent of wind speed.
13.10. 2 The atmospheric diffusion methods of Sutton were used for the neutral and unstable cases. Due to the errpirically in-j dicated inadequacies of the Sutton method.for ir. version con-ditions, calculation methods based on Hanford diffusion results, as outlined in Report HW-54128*, were used for the inversion cases.
I 13.10, 3 Weather Conditions This evaluation assume d that the weather conditions involved t
no precipitation ind that the incident occurred during hot summer weathe:. Precipitation would deposit.more con-tamination close to the plant than this evaluation indicates, f
thus reducing contamination levels further away. If the incident occurred in cooler weather, the fission product
(
leakage from the enclosure would be less than indicated due j
to more favorable heat transfer and, consequently, more rapid reduction of the enclosure post-accident pressure.
(
13.10. 4 Elevation of Release
(
Leakage from the enclosure is considered to occur near the ground level. This appears reasonable as most enclosure penetrations a re near grade. If the postulated leakage
(
occurred at some significantly different height, such as by emission from the stack, the off-plant deposition, and possi-ble inhalation would occur at greater utstances-than this
(
evaluation shows, but their magnitude would be vastly reduced.
HW-54128, " Calculations on Environmental Consequences of Reactor Acci-(
dents," Interim Report, by J. W. Healy, December 11, 1957.
(
~
(
t i
Section 13 Page 19
- 13. 10. 5 Initial Dilution by Building Wake 1
F This evaluation recognizes that initial immediate dilution of the leakage will occur due to the turbulent wake of the enclosure structure produced by the passing wind. It is estimated.tha :
the effective wake cross section is of the order of one half.
of the vertical cross section of the enclosure structure. No additional immediate dilution by other nearby structures or other ground cover is considered. This effective "ake has been equated to a semicircle of equivalent area centered at ground level. Centering the initially diluted leakage at some greater height would reduce the off-plant effects of leakage from those evaluated. It is noted that the radius of the equivalent semicircle is less than the anclosure sphere radius.
To obtain an estimate of this initial dilution of the leakage, the radiustof the equivalent semicircle was estimated to represent about 1-1/2 standard deviations of cloud width. From these considerations, virtual source points were calculated equiv-alent to various upward distances, dependent upon the diffusion condition, and are:
Symbol ^
Diffusion Wind Speed Virtual Source Distance I-l Inve rsion 1 m/s 250 meters I-5 Inversion S in/ s 370 N-5 Neutral 5m/s 200 U-5 Unstable Sm/s 80 These estimates of the virtual source locations agree generally with the methods of Holland *for the neutral and unstable cases and are more conservative for the inversion case.
- 13. 10. 6 Effect of Distance on Diffusio The uownwinu effects, such as passing cloud dose, ground ucposition, and inhalation exposure, are a function principally of the integrated air concentration at any point. This integrated concentration subsides with' distance due to turbulent diffusion i
in the atmosphere, and depletion of the contaminatied cloud by deposition on the ground and on ground cover.- The magnitude 2
of this effect for the example distances and diffusions is:
i T AB LE 13.1 Ii Unit Integrated Air Concentration (uc-sec/cc/ curie released)
Distance Diffusion Gases Halogens Pa rticula te s i
4 3 x 10-4 4 x 10-4 1/2 mile-I-1 4 x 10 4 7 x 10-5 1 x 10-4 4
I-5 1 x 10-(
N-5 2 x 10 6 9 x 10 6 2 x 10.-5 5
U-5 5 x 10-3 x 10-6 5 x 10 6 These symbols are used throughout the evaluation to indicate the four i
'(
~ diffusion examples inustrated.
(
" Meteorology and Atomic Energy. "
- As indicated in l
S2cti n 13 Page 20 Rev 1 (3/19/62) t Table 13.1 (Contd) r Unit Integrated Air Concentration d
(Jac-sec/cc/ curie released)
Distance Diffusion Gases Halogens Par ticulate s 1 mile I-l 2 x 10-4 1 x 10-4 2 x 10-4 5
5 6 x 10 6 3 x 10-5 6 x 10 6 I-5 N-5 7 x 10-3 x 10-6 7 x 10-U-5 2 x 10-6 8 x 10-7 2 x 10-6 3 miles I-I 6 x 10-5 3 x 10-5 6 x 10-5 I-5 3 x 10-5 8 x 10" 3 x 10-5 N-5 2 x 10-6 6 x 10-7 2 x 10-6
-7 U-5 3 x 10 2 x 10-7 3 x 10 13.10.7 Probability of Various Wind Speeds and Directions 13.10.7.1 Meteorological data taken at the site during the pact year generally substantiate the five-year data from the Charlevoix Coast Guard Station quoted in the previous hazards summary reports. Wind data taken at the 32-foot level indicate that about two-thirds of the time, enclosure leakage would be conveyed over the lake rather than over adjacent land.
13.10.7.2 The wind data have been summarized and grouped into directions of various significance as follows:
Wind Wind From Significance of Direction Direction Azimuths of Wind Movement From Plant No. I 280 -360 -20 Blowing inland, sparsely populated 2
30 -50 Blowing inland, toward Charlevoix (3 miles) 3 100 -230 Blowing over open lake 4
240 -270 Over lake, toward Petoskey and Harbor Springs (11 miles) 5 60 -90 Over lake, parallel tc shore l{
6 Any Calm (less than 4 mph) 13.10.7.3 Table 13.2 summarizes the 32-foot level winds in this manner,
with the percent of time numbers shown not in parentheses,
and the average mph wind speed shown i1 parentheses.
Directions I and 2 are totalled, since they represent the
^t total probability of the leakage path being over land adjacent to the plant. Similarly, the over water directions (No. 3, 4, t
- 5) are summed.
(
Section 13 Page 21 TAB LE 13. 2 Wind Direction Frequency, Percent; and Average Wind Speed, (mph);
32 Foot Level Big Rock Point, Michigan Direction #
1 2
3 4
5 6
1, 2 3,4,5 Nov. 60 24(19) 2(12) 50(12) 11( 2 2 )
9(18) 4 26(19) 70(14)
Dec. 60 4 2(19) 7(16) 31(12) 12(28) 7(22) 1 49(19) 50(17)
Jan. 61 4 2(17) 3(09) 36(08) 9(18) 7(14) 3 45(16) 5 2(11)
Feb. 61 1 7(14) 1 (11 )
39(09) 11(13 )
25(13) 7 18(14) 7 5(11)
Mar. 61 33(15) 5(13) 3 0(10) 6(17) 23(14) 3 38(15) 59(12)
Apr. 61 3 2(13 )
4(11) 20(08) 16(09) 20(14) 8 3 6(13 )
56(10)
May 61 32(12) 1(09) 27(09) 19(12) 15(10) 6 33(12) 61(10)
Jun. 61 15(12) 2(10) 32(09) 18(13 )
2 4(11) 9 17(12) 7 4(11)
Jul.
61 20(09) 3(07) 26(07) 18(09) 15(10) 18 23(09) 59(08)
Aug. 61 3 0(10) 2(10) 35(09) 9(12) 11(1 0 )
13 3 2(10) 55(10)
Total Pericxi 29(14) 3(12) 33(09) 13(14) 16(13) 7 3 2(14) 61(11)
Overall Wind Speed, 11 mph.
- 13. 10. 7. 4 The overall wind speed of 11 fnph corresponds to the 5 m/s wind speed used in three of the four diffusion examples illustrated in this analysis. The data indicate that the most severe diffusion example illustrated (inversion,1 m/s wind speed) probably is applicable only a few percent (perhaps 1% to 5%) of the time.
13.10. 7. 5 The probability of leakage movement toward Charlevoix (Direction #2) is only 3% of the time, with the three mile distance to the outskirts reducing the significance of even this low probability direction.
13.10. 7. 6 The probability of leakage movement over the lake in the direction (#4) of Harbor Springs and Petoskey is indicated to be about 13% of the time. The significance of this is minimized by the eleven mile distance available for atmos-pheric caffusion.
13.10, 8 Wind Direction Diversity 13.10. 8.1 The accompanying graphs for dose rate from the passing cloud, the rate of deposition of fall-out on the ground, and the rate of inhalation of air-borne contaminants are based t
Section 13 Page 2 2 on no wind direction change, since such effects are expressed in units of instantaneous rate functions. However, to inte-grate an effect such as any of these over a period of sever-al hours would be unrealistically conservative, as minor wina oirection changes are occurring continuoucly, even during periods when the reported general wind direction representeu by a 45* angle is apparently remaining constant.
13.10. 8. 2 The standara deviation of cloud width is a measure of the zone occupied by the postulated contaminant plun,e at various distances and diffusion conditions. For the examples evaluated, this is:
1 Standard Deviation of Cloud Width, meters Distance 1-1 1-5 N-5 U-5 1/2 mile 30 24 60 115 1
mile 42 36 100 2 05 3
mile s 72 52 210 460 13.10. U. 3 W%h the expected Gaussian distribution of cloud concentration in the vertical and cross-winu directions, a centerline effect is reuucca by a factor of about 10, if the point of exposure is two standard ueviations of cloud width from the centerline.
Consiuering the over-all precision of the evaluation, such a reduction may be reasonably considered as sufficient to consiuer, the effect nonadditive. Thus, wind diversity fac.-
tors may be related to the wind direction change which will cai se the plume centerline t'o rotate an angle equal to twice the standard deviation at the distance and diffusion of interest.
Such angles for the examples evaluated are:
Angle of Two Standara Deviations, degrees Distance 7_- 1 1-9 N-5 U-5 4
1/2 mile
- 3. 4
- 2. 8
- 6. 8 18.3 1
mile
- 2. 4 2.1
- 5. 7 11.6 3
mile s 1.6 1.2
- 4. 8 10.4
- 13. 10, 8. 4
" Meteorology and Atomic Energy," pages 66-67, provides a basis for the degree of wind direction variation expected for the diffusion cases evaluated. The wind direction of typical inversion periods may be expected to vary by about 10*.
During periods of typical neutral lapse rate conditions, a wind direction variation of about 20* may be expected, and a variation of the order of 50* may be anticipated during typical daytime unstable conditions. There is a good prob-ability that the actual variation would be greater than assumed i
Section 13 Page 23 here during a period of interest of several hours; there is a small possibility that during some periods, the direction variation would not be this large.
13.10, 8. 5 From such considerations, it is reasonable to conclude that integrated effects, considering wind direction vari-ation over a few hours, may be reduced from those ef-fects calculated for unidirectional winds for the examples used, by factors of:
Reduction Factor From Wind Direction Diversity Distance 1-1 I-5 N-5 U-5 1/2 mile 1-l/ 2 2
1-1/ 2 1-1/2 1
mile 2
2-1/ 2 2
2 3
mile s 3
4 2
2-1/ 2 13.10. 8. 6 Such reduction factors are for a period where the wind is remaining in one nominal direction. When a significant wina direction change occurs, the possibility of any additive effect at any one location for the periods before and after the change is completely terminated.
13.10.9 Travel Time I
~
The evaluation includes the effects of wind travel time from the enclosure to distances of interest off-site. While some additional radioactive decay, occurs during such travel periods, the important feature is that during the poorest diffusion con-ditions, no effect at all occurs at distances beyond a few miles for several hours after the initiation of the accident. Thus, a period of time is available for evasive action to occur, if it is determheu that such may be desirable as a precautionary
- measure, i
- 13. 11 EXTERNAL RADIATION DOSE FROM PASSING CLOUD
- 13. 11. 1 Evaluation of effects of passing cloud air concentrations down-wind were estimated using the Sutton and Hanford methods as outlined above. Particular emphasis was taken in this eval-uation in the conversion from hir concentration to integrated dose for the passing cloud effect. Due to the radioactive decay of the equilibrium fission product mixture which occurs during the post-accident period, the conversion from con-centration to dose becomes more favorable in reducing dose as the decay period available increases. For the noble gas, halogen and volatile solid fission product groups, the concen-1
(,
tration required in an infinite cloud to produce a certain dose
(
i Section 13 Page 24 f-was evaluated for the radioactive decay periods of interest in the post-acciuent period. Selected. values of the air concentration in an infinite cloud. in units of microcuries per cc, which will produce a dose of one mrad per hour with hemispherical geometry are:
Air Concentrations (uc/cc) Giving One mrad /hr Ibse Rate
('
Decay Time Noble Gases Halogens Volatile Solids 1 Hour 1.6x10-6 0.78x10-6 1.4x10-6 7
4 ' Hours 2.3x10-6 0.79x10-6
- 2. 5x10 6 8 Hours 3.1 x10-0 0.88x10-6
- 2. 5x10-6 16 Hours
- 4. 2 x10 6 1,o xio 6
- 2. 6x10 -6 13. 11. 2 The dose from the passing cloud based _on uniform concentration and infinite cloud considerations was then corrected for the finite cloud' size and Gaussian distribution of cloud concentration. For the various diffusions evaluated, and for cloud sizes calculated at the distances illustrated, the ratios of finite cloud to infinite cloud effect are:
Finite Cloud Correction Factor Distance I-1 1-5 N-5 U-5 1/2 mile 0,16 0,13
- 0. 29 0.43 1
mile 0,22 0,20 0.40 0.60 3
mile s 0.33 0,26 0.60
- 0. 83
- 13. 11. 3 The reduction of cloud concentration at the distances evaluated because of prior deposition on the ground of halogens and solids, was factored into the passing cloud effects. This correction is actually of a small magnitude since most of the passing cloud effects are due to noble gases.
I 13. 11. 4 The instantaneous dose rates from the passing cloud at the one-half mile distance and on the centerline of the postulated plume are shown in Figure 13. 7.
The maximum dose rate occurs about 1-1/2 hours af ter the accident, and is of.the order' of 5 mravis per hour; by the end of one day, it has subsided to lead th'a'n 1 mrad /hr.
t 13. 11. 5 Examples of possible integrated doses from the passing cloud are shown in Figure 13. 8.
With a continuing wind direction, the maximum integrated dose at 1/2 mile is of the. order bf 60 mrads. These calculated doses assume that the receptor is on the center of the cloud path and that no incidental shielding, such as that provided by housing, is available.
L
Section 13 Page 25 EXTERNAL RADIATION DOSE FROM GROUND DEPOSITION 13.12 13.12.1 The fall-out concentrations of radioactive materials were determined on the basis of particle settling-by eddy diffusion only, since settling by gravity is expected to be negligible in this case. It is expected that the particulate radioactive i
material which might leak from the enclosure will be only a few microns in diameter. If the material were of a significantly larger diameter, it would be deposited at a much faster rate within the sphere and thus would not be available for leakage. Also, if the particles were larger, they might not be able to escape from the encloi.a; e since the leakage that may occur is expected to be restricted to that which could pass through minute imperfections in the wall or through penetration seals.
13.12. 2 The extent of halogen and solid fission product deposition on the ground is a function of the apparent deposition velocity.
The deposition velocity is considered to be a function of the diffusion condition and wind speed.
Deposition velocities used in this evaluation were based on British results cited in HW-54128, and are:
Deposition Velocity Characteristics Ratio of Deposition Velocity 16 Deposition Velocity, Wind Wind Velocity cm/sec Diffusion Velocity Particle s Halo gens Particles Halogens Inversion 1 m/s
- 2. 2x10-4
- 3. 4x10-3, O.022 0.34 Inve rsion 5 m/s
- 2. 2:d0-4
- 3. 4x10-3
- 0. 11
- 1. 7 Neutral 5 m/s 3
x10-4 4.6x10-3 0.15
- 2. 3 Unstable 5 m/s 6
x10-4 8
x10-3
- 0. 3 4
i The evaluation provides for correction due to radioactive 13.12.3 decay after the material is deposited on the ground. As the amount of deposition is a function of air concentration, and as the air concentration is depleted by prior deposition at 1
locations closer to the source, correction for this depletion
(
has been made for deposition at the distance illustrat.:d. In addition, the dose rate from the deposited material has been corrected for the finite size of the deposited source. This correction is a function of the standard deviation of cloud
(
width, and for the example diffusions and distances is:
(
(
I
. Section 13 Page 26 i
Finite Deposition Pattern Correction Factor
- Dis tanc e I-1 I-5 N-5 U-5 4
1/2 mile 0,41 0.38 0.49
- 0. 54
(
1 mile 0.45 0.43 0.53 0,60 3
mile s 0.50 0.48 0,60 0.67 13.12. 4 The conversion from deposition on the ground to gamma radia-tion dose rate at the conventional one meter above the ground was made considering the gamma energies present from the halogen and volatile solid fission products of which the deposited material. is composed. The conversion is dependent on the age g
of the fission products present as follows:
Decay, days mr/hr at one meter / curie / meter" 4
0.I
- 1. O x 10 3
1 9.1 x 10 3
10
- 7. 2 x 10 3
100
- 3. 6 x 10
- 13. 12. 5 Ground deposition at the example diffusions under the assump-tion of unidirectional wind, is shown on Figures 13. 9 to 13.12.
The period of significant deposition is completed about six i
hours after the accident, as the leakage after that time is largely noble gases. A principal significant component of the deposition is Iodine-131, which is shown in addition to the total deposition concentration on this series of figures.
- 13. 12. 6 The possible integrated doses from radioactive material on the
(
ground, using the previously established appropriate wind diversity factors, at the example diffusions resulting from the postulated leakage, are shown on Figure 13.13. Due to
(
the composition of the mixture deposited, the infinite time dose is largely delivered in a period of several months. If a person continuously occupied a point one-half mile from the point of leakage which has been subjected to the least favorable diffusion condition, a dose of the order of 35 mr would be received. This assumes no reduction of deposited material
(
during the extended period by mechanisms other than radio-active decay. In all probability, actual integrated doses would be reduced by the effects of rain and other umathering action.
The doses illustrated are the maxima applying under the leakage plume centerline, with any incidental shielding such as housing not considered.
.(
i.
(.-
i.
I Section 13 Page 27 i
- 13. 13 GROUND DEPOSITION OF IODINE-131 i
13.13.1 Due to the radioisotopic composition of the postulated leakage, the most significant individual effect from contamination of off-site land would be the possible control of milk production from grazing cattle because of the Iodine-131 food change i
relationship from vegetation to cattle to milk to man. Follow-ing the Windscale incident, the British used a limit of 20 rads to the thyroid of children as a criterion for control of milk.
They indicated that such a dose would result from milk con-taining 0.1 microcuries per liter,-and that this concentration resulted from cattle grazing on pasture containing about one microcurie of Iodine-131 deposited per square meter.
- 13. 13. 2 Using the previously established wind diversity factors, the immediate maximum Iodine-131 deposition concentrations would be:
2 In.meaiate Maximum Iodine-131 Deposition, pc/m I
~
U-5
~
Distance I-l I-5 N-5 1/ 2 mile 28 30 9
5 1
mile 8
10 2
1 3
miles 2.
2 41
< 1 l
- 13. 13. 3 Under the poorest diffusion conditions a zone with a length of about four miles ' downwind and containing an arer. of less than one square mile might be above the proposed control 2
limit of 1 ac/ meter. No eva'cuation of humans or domestic J
animals would appear necessary; the only possibly required control would be non-use of milk products from the area (This appears to be a reasonable conclusion for the 100%
example also). Due to radioactive decay alone, the deposition concentrations one month later would be reduced to:
Iodine-131 Deposition After One Month Decay, pc/m Distance 1-1 I-5 N-5 U-5 1/2 mile 2
2 41
< 1
'i 1
mile 41 41
<1
< 1 3
mile s
<1 41
<1
< 1
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13.13. 4 After this period, the area of the zone would be significantly reduced due to decay alone. Considering the probable simul-taneous reduction action by precipitation, it appears probable that all off-site areas would be within the proposed control
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limit within a few weeks (Similarly for the 100% example acci-dent,3his conclusion appears reasonable after a period of i
about a month).
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' Section 13 Page 28 l
13.14 INTERNAL DOSE TO THYROID 13.'14.1 Internal exposure to the thyroid gland from inhalation of the fission product mixture in the passing cloud is primarily due to iodine radioisotopes. This exposure was evaluated con-siderir,g the dose from thyroid deposition of Iodine-131,133 i
and 135. Other iodine radioisotopes of half lives of 2. 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> or less were not included, considering their low rem per microcurie ratio for lifetime dosage considerations, and because of the estimated 3 to 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> thyroid uptake time after the material is inhaled. The lifetime thyroid dose t
was evaluated for the three isotopes considering a breath-ing rate of 30 liters per minute, and a thyroid deposition of 15% of that which was inhaled.
13.14, 2 The total radioiodine deposition rate in the thyroid on the leakage plume centerline for the example diffusion cun-ditions and at a distance of one-half mile, is shown in Figure 13.14.
13.14. 3 Using the previously established wind diversity factors, the 4
total lifetime dose to the thyroid from inhalation during the first two hours after the accident is:
Lifetime Thyroid Dose, rems, First Two Hour Exposure Distance 11 I-5 N-5 U-5 1/ 2 mile 2
- 0. 4 0,07 0.02 1
mile
- 0. 4
- 0. 2 i
13.14. 4 Similarly, for inhalation during the entire course of the pos-tulated significant halogen leakage, the lifetime thyroid dose is:
(
Lifetime Thyroid Dose, rems, Continuous Exposure Distance I-1 I-5 N-5 U-5 t
1/ 2 mile 4
- 0. 7 0.1 0.03 l
t 1
mile 1
- 0. 2 i
The evaluation assumes no. reduction in the concentration of air inhalc.d such as would be the case if the receptor were I
within a structure.
l 13.15 INTERNAL DOSE TO LUNG i
13.15.1 Dose to the lungs was evaluated considering that all volatile and other solid fission products inhaled were insoluble, and
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by use of conventional standard man metabolic factors.
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-i Secdon13 Page 29 i'
13.15. 2 Considering the composition of the postulated leakage, the analysis indicates that essentially all of the lifetime dose to the lungs is due to the longer-lived radioisotopes of cesium, ruthenium and tellurium. The rate of deposition in the lung under the example diffusion conditions, and at a distance of one-half mile, is shown in Figure 13.15. The calculations
(
indicate that lifetime dose to lung from inhalation during the entire course of the postulated significant volatile and other
(
solids leakage during the poorest diffusion condition would be about 0. 5 rems
, (
- 13. 16 INTERNAL DOSE TO BONE t
13.16.1 The analysis of bone dose indicates that essentially all of the contribution is due to the longer-lived radioisotopes of strontium, yttrium, zirconium, barium and cerium, together with their al.propriate daughter fission products.
13.16. 2 Due to the low fractions of these fission products in the post-ulatea leakage, the analysis inuicatt.s that the lifetime dose to bone at any off-site location, and for any exposure time or diffusion, is always much less than O. I rems.
13.17 GENERAL
SUMMARY
13.17.1 Thus, it can be seen that for the postulated accident where it is assumed that as much as ten percent of the core melts, doses at any distance beyond the plant boundaries are not of a hazard-ous level, and in fact are quite insignificant at the distance of any small population center, e The radiological effect cal-culations of the example where it is assumed that all of the core actually melts show levels aaout ten times greater than those for the partial core melt example; even in snis case, the calculated doses at any distance, one-half mile or greater from the enclosure, for both the first few hours after the incident and for the entire course of the postulated release, do not exceed those levels suggested currently as reasonable j (
for site and plant safety svaluation.
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Section 13 Page 39
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Section 13 Page 40
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