ML20009A005

Forwards Evaluations of SEP Topics XV-16 Re Radiological Consequences of Failure of Small Lines Carrying Primary Coolant Outside containment,XV-18 Re Consequences of Main Steam Line Failure & XV-19 Re LOCA from Piping Breaks

ML20009A005
Person / Time
Site:	Big Rock Point File:Consumers Energy icon.png
Issue date:	07/01/1981
From:	Hoffman D CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:	Crutchfield D Office of Nuclear Reactor Regulation
References
TASK-15-16, TASK-15-18, TASK-15-19, TASK-RR NUDOCS 8107070355
Download: ML20009A005 (44)

Text

{{#Wiki_filter:f Consumers i power Company l General offices: 212 West Michigan Avenue, Jackson, Michigan 49201 + (517) 7884550 July 1, 1981 l Director, Nuclear Reactor Regulation Att Mr Dennis M Crutchfield, Chief [ Operating Reactors Branch No 5 US Nuclear Regulatory Commission Weshington, DC 20555 l l i DOCKET 50-155 - LICENSE DPR BIG ROCK POINT PLANT - SEP TOPICS XV-16, RADIOLOGIF CONSEQUENCES OF FAILURE OF SMALL LINES CARRYING PRIMARY COOLANT OUTSIDE CO., .INMENT; XV-18, RADIOLOGICAL CONSEQUENCES OF A MAIN STEAM LINE FAILURE OUTSIDE CONTAINMENT (BWR); XV-19, LOSS OF CCOLANT ACCIDE"TS RESULTING FROM SPECTRUM OF POSTULATED PIPING BREAKS WITHIN i THE REACTOR COOLANT PRESSURE BOUNDARY (RADIOLOGICAL PORTION) Attached are the Consumers Power Company evaluations of SEP Topics XV-16, XV-18 and XV-19 (radiological portion). ^ David P Hofften Nuclear Licensing Administrator CC Director, Region III, USNRC NRC Resident Inspector - Big Rock Pofnt 41< ' h 7 8107070355 810701 1 PDRADOCKOSOOOg; p s

s o O ATTACHMENT 1 Topic XV-16: Radiological Consequences of Failure of Ema11 Lines Carrying Primary Coolant Outside Containment, Evaluation The Big Rock Point Plant drawings and design docu=ents were reviewed to deter-mine if any lines carrying primary coolant outside containment if ruptured, could release significant quantities of radioactive material. A number of lines were found. These are Main Steam Line, Feedvater Line, Resin Sluice Line and Core Spray Suction Line. The radiological effects of a ruy.'ure of the Main Steam Line is considered under Topic XVL18. The Feedvater Line connects the feedvater pumps and the Steam Drum. There is a check valve and a sving check valve in series inside containment quite near the Steam Drum on the Feedvater Line. There is also an air-operated valve upstream of the High Pressure Heater. Testing of these valves is covered in Consumers Power Company letter dated February 13, 1976 from Ralph B Sevell, Nuclear Licensing Administrator, CPCo, to Director of NRR, NRC, and in a letter dated December 27, 1979 frem D P Hoffman, CPCo to D L Ziemann, NRC. There vill be no significant out-of-containment leakage due to forward flow of feedwater. Radiological releases due to a break in the Feedvater Line .11 be negligible as a result of the above considerations. The Resin Sluice Line connects the Demineralized Water Tank (part of the reactor cleanup system) with the Resin Disposal Tank. There are two remote manual, normally closed isolation valves (one inside and one outside containment) in this line. In addition, there is a pressure switch which controls the inboard valve. The switch inhibits opening when the pressure is greater than 100 psi. Testing of these valves is discussed in the Dece=ber 27 and February 13 letters referenced above. As a result, radiological releases due to a break in the Resin Sluice Line, vill be negligible. The Core Spray Pump Suction line connects the containment sump suction strainers to the core spray pumps. This l'.ne is underground and covered with concrete over its entire length. There are ne isolation valves on this line. This line, however, only experiences very low pressures and, as a result, failure is highly improbable. See also the February 13 letter and the Fault Trees of the Big Rock Point Probabilistic Risk Assessment (submitted by letter dated March 31, 1981 from D P Hoffman, CPCo, to D M Crutchfield, NRC). Breaks in lines carrying primary coolant outside containment vill not result in significant leakage with the exception of the main steam line due to double valves, low pressure or a combination of these. As a result, the radiological consequences are negligible. It is therefore concluded that Big Rock Point is consistent with current NRC criteria (SRP 15.6.2) for failures outside containment of mall lines which carry primary coolant.

s ATTACHMENT 2 Topic XV-18: Radiological Consequences of a Main Steam Line Failure Outside Containment (3WR) The case considered was a double-ended rupture of the Main Steam Line outside containment. Blowdown data was taken from an analysis of a break in the Main Steam Line inside containment. Reactor trip occurs at 10 seconds due to low reactor vessel water level (actually high void fraction which appears as low water level to the level instrumentation). The MSIV's start to close and are fully closed 60 seconds later (t=TO seconds) Blowdown mass release is obtained assu=ing that the 53IV closes instantly at t=Th seconds rather than ramping closed over 60 seconds as is actually the case. The core remains covered for both cases and therefore no fuel failures occur. Ihe primarr coolant activity was taken from the -Technical Specification and is 35u Ci/ml (STP conditions) iodine and noble gas concentration such that the off-gas release is less than 0.47/E, where E is the average energy of the released gases. The design of Big Rock Point Plant is such that the steam line is in the pipe tunnel. Pressurization of the pipe tunnel vill result in the rupturing of the blowout panel specially designed to rupture prior to the overpressurization of the pipe tunnel. The release vill therefore be a ground level release. No decontamination factor was used. Plume depletion and decoy was not considered. The whole body dose was assumed to be due to a semi-infinite cloud of ga=ma radiatiog/sec. and an infinite cloud of beta radiation. Breathing rate was taken to be 3.hTE-hm Iodine conversion factors were per TID-lh8hh. The total mass release over.the Th seconds was 8.0TE+4 lb. It is assumed that the plant will be cooled using the E=ergency Condenser and Shutdown Cooling System so that no further releases will occur after MSIV closure. These assumptions are consistent with the criteria of SRP 15.6.4. Based on the present Technical Specification limit for radiolodine in the primary coolant (35 uc/ml), the results of the analysis indicate that the thyroid dose is 92 re=, This is greater than 10% of 10CFR 100 limits. At Big Rock Point, however, radioactive halogens have always been well below 10 pc/ml. In fact, levels of 35 pc/ml vould preclude plant operation because staca release limits could not be met, and because in-plant radiation levels would prevent personnel access for many routine operations. In addition, many of the general area radiation monitors would be alarming even at levels of 10 pc/ml which would further restrict activities. A much more realistic limiting value for radiciodines would therefore be 11 pc/ml, which would in turn cause calculated thyroid dose to fall within the guideline of 10% of 10CFR 100 limits. Conclusion It is concluded that the intent of current NRC criteria for a main steam line failure outside containment is met in that the resulting off-site thyroid doses would be less than 10% of 10CFR 100. The calculated dose in excess of this guideline is a result of an imposed assumption about primary coolant radiciodine levels which is not realistic for the plant.

s ATTACHMENT 3 Topic XV-19: Loss of Coolant Accidents Resulting from Spectrum of Postulated Piping Breaks Within the Reactor Coolant Pressure Boundary - Radiological Portion The results of the Loss of Coolant Accident is presented in the Final Hazards Summary Report Section 13 (attached). The event presented is an instantaneous double-ended rupture of the recirculation pump discharge pipe. The containment is assumed to leak at the limits set in the Technical Specifications (0.5% per day at dasign pressure). Post incident system and core spray both operate to reduce pressure. Fission product inventory is based on steady state operation at 2h0 Mwt. The quantity of fission products is consistent with ORNL-2127. It was assumed that 100% of the noble gases, 50% of the halogens,15% of the volitile solids (Te, Se, Ru, Cs) and 0.3% of all other solid fission products are released to the containment atmosphere. Depositien of iodine at a rate of 10-3 per second was assumed. Was out and deposition of the solid fission j products occurred at a rate of 3x10 per second. The containment leakage rate is based on the fission product inventory and containment pressure. Figure 13.3 and 13.h in the FHSR present the leakage rates for noble gases, halogens, and solids. Dose calculations were performed for four types of weather conditions: inversion with wind speed of im/s, inversion with vind speed of 5m/sec, neutral with vind speed of Sm/s and unstable with vind speed of 5m/sec. Doses were calculated for direct shine from containment, external dose from the passing cloud, and ground deposition and internal dose to the thyroid, lung and bones. See attached Section 13-of the Final Hazards Summary Report for details. Leakage of the Post Incident System components was addressed in a letter dated December 27, 1979 from D P Hoffman, CPCo, to D L Ziemann, NRC, concerning actions taken in response to TMI (Section 2.1.6a). Leakage of these components will contribute negligibly to the total radioactivity release as the components are periodically tested and mo;.itored for leakage and any leakage which occurs will be contained either in the core spray pump room or in the radvaste vault. Conclusion It is therefore concluded that the existing analysis of the radiological effects of LOCA events for Big Rock Point are consistent with current NRC criteria -(SRP 15.6.5), and the results are within the limit of 10CFR 100. e 6

. EXCERPT. FINAL HAZARDS

SUMMARY

REPORT - BIG ROCK POINT O sz c rio s 13 MAXIMUM CREDIBLE ACCIDENT

13. 1 GENERAL CONSIDERATIONS O-

13. 1. 1 The safety ana1yses up to this point have dealt with measures designed to protect against release of dangerous amounts of O

reaieactive m >ter ~ s <- m .. system end have shown the very high degree w" ...ans.wovided by the plant design against accidents tuat 'could create radiation hazards. In the 0 extreme 1x uniixeir case thee. in seite of a11 the greceutions. an accident releasing hazardous amounts of radioactive materials from the reactor should occur, the reluased O materials would be containe I within the containm'ent vessel housing the reactor. O

13. 1. 2 The " maximum credib1e eccident" (MCA) fer the Big Reca Point reactor is defined as an instantaneous rupture of the primary coolant system, creating an opening of 7. 06 square O

fe e t.. This is equivalert to twica the flow area of the largest coo 1 ant pipe, namely, re.sirculation pump discharge 1ine, A connected to the reactor ves ?l. 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 potentia 1 effects of such an accident, it is assumed that the enclosure leaks at O

the maximum permissibic rate (O. 5% per Cay at the design pressure).

13. 1. 4 The postulated accident =houM. result in.~adiological conse-quences of litise 4_JIlu~ev ' r ause thic core spray system is designed to prevent significant fore damage in.such

( an event, thus limiting fission pror'act release to a small fraction of the fission product ins entory in the core. However, Q in order to illustrate the effecti$ eness of the various barriers against release of fission proriacts, the MCA analysis includes i examples of varying degreer of effectiveness of the core spray. Q The spectrum of consequences analyzed inciudes those in which: (a) core spray is fully effective; (b) core spray prevents damage to 90% of the.fue.; andJ(c) core spray is assumed to be unavail-( Q able to cool the core, and the c' ore melts down from decay heat. g, 13.1. 5 Although a specific event is described to facilitate analysis of Q the MCA, the assumed accident conditions should not be inter-l l preted as a specific weakness of design. Rather, the conditions chosen for analysis are judged to be at least severe enough to Q constitute an upper boundary for credible operating-accident severities and, accordingly, are suitable for demonstrating the effectiveness of containment. 1 LO l

i. O

l Ol Section 13 p,

13. 1. 6 With respect to the case within the above indicated spectrum of consequences analyzed, in which the core spray system is Q.

fully effective, there would. be only a very small fract.on of the core fission products released to the containment vesset. Q The result of such release would have negligible effect on plant personnel and environs. Thus, this case is discussed no further. It is a measure of the strength of the Big Rock Q Point plant design that only a MCA, which is compounded by a second failure, such as reduced effectiveness of core t' spray, need receive further analysis in order to develop Q circumstances by which the total hazard to the health and safety of the public can be judged. O

13. 2 BASIS FOR "MCA" LIMITATIONS O

13. 2. 1 Power Boiler Statist _ics

13. 2. 1. 1 Statistics on reactor vessels are not available because of h

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 h reactor vessels. It is conservatively estimated that there are 400-500 boilers in the United. States designed to ope rate at a prer sure over 600 psia, and that they represent not h less than 4000 boiler-years of operating exporience. The first such boiler was designed over 30 years ago. An ex-amination of the Hartford Steam Boiler Inspection and h Insurance Company power boiler statistics shows no failure of steam drums in powe boilers designed to operate over 600 psia. In no case have the materials used in th ese O drums been superior to those used in fabrication of reactor vessels. Statistics on all types of boilers and unfired pressure vessels reveal the following relevant points: l l

13. 2. 1. 2 A number of boiler explosions have occurred in small low l

pressure heating boilers. The lack of such failures in h 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. l

13. 4. 1. 4 The boiler explosion statistics indicate a large reduction in incidence of failure over the years, with the trend con-g tinuing in the direction of 'mprovement.

V l

13. 2. 2 Brittle Failure e

l V

13. 2. 2. 1 The following considerations regarding design, fabrication, and operating conditions for the reactor vessel provide g

reasonable assurance that the vessel will not fail in a brittle manne r. O O

O .O Section 13 Page 3 O

13. 2. 2. 2 The vessei., tee 11n the high fiux region ogresite the core j

has a nil ductility transition (NDT) temperature less than i 10

• F.

This low NDT temperature provides considerable h margin for radiation effects. Over the lifetime of the i reactor, parts of the vessel may receive integrated j neutron exposure of about 1019 neutrons per sq. cm., iO which at room temperature would not de exeected to raise the NDT over about 140 to 200*F depending on ,j the neutron energies considered. Since irradiation takes j, O a giace at etevated temeeratures. radimtion effects ere expected to be even less. O

13. 2. 2. 3 The vessel steel is iow in notch sensitivity, and vesset notches are kept to a minimum by careful design, mate-rial selection, quality control, and inspections.

13. 2. 2. 4 The nessel is built in accordance with requirements of O

the ASME Boiler and Pressure Vessel Code as modified for nuclear reactor vessels.

13. 2. 2. 5 O

High loadings are avoided while metal temperatures are in the nonductile range. (This is an inherent characteristic of boiling water reactors controlled on saturation conditions. )

13. 2. 3 Ductile Failure

13. 2. 3. 1 O

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 from 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 shock waves from a rapid nuclear Q excursion or chemical reaction.

13. 2. 3. 2 The safety valves are sized to prevent reactor pressures in O

excess f those allowed by code. The reactor safety system and reactor characteristics preclude any large nuclear ex-cursion or chemical reaction. Therefore, a ductile failure Q is not considered credible.

13. 2. 4 Nuclear Excursion O

i 13.2.4.1 The MCA is not assumed to involve a significant coincident j energy contribution from a nuclear excursion for the follow-- O ins re sons:

13. 2. 4. 2 The very low probability of an excursion, considering the O

many safesu rde incorgerated in the desian of the re ctor and safety systems to prevent such an accident, including: O Control of reactivity insertion ratee; j Multiple scram circuitry; I, Inherent negative coefficients of reactivity; and i Availability of strong procedural control. 1 0

O !. g[ Section 13 Page 4

13. 2. 4. 3 If, nevertheless, a nuclear excursion did occur, none of the excursions postulated for this. reactor are beli( ved capable I

of initiating a primary system rupture. Thus, simultaneous occurrence of an excursion and a pipe break is not considered to be credible.

13. 2. 5 Metal-Water d

'en f ' i Q The primary system rupture accident is not considered to involve a significant coincident energy contribution from a metal-water reaction for the following reasons: O

13. 2. 5. 1 Metal-water reactions of safeguards interest with the Big i

Rock Point reactor are those involving zirconium, which is . Q contained in the zircaloy fuel channels, and the stainless i steel fuel cladcing. There is no experimental evidence which suggests a probability of a reaction involving stainless steel O nd w ter. In the event that stainless ' steel fuel 41 adding is melted j by decay heat as a result of the postulated MCA, the process would be relatively slow and it is not expected that any appre-g O ci die mouat of eiemeat i tro wouid be produced i= g rticies i 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 h

cannot occur when the metal temperature is below the melting i point. The zircaloy channels would not reach melting tem-peratures until sometime after the failure of fuel cladding h and subsequent temperature rise in UO 2 uel in the event of f the postulated.MC A.. E3. e.rimental studies indicate that j this reaction canne. prot.ced329tdly unless the molten metal i O is disgersed into garticlee with e 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 O' heating up to the melting temperature. 1

13. 2. 5. 3 In view of the above considerations, it is concluded that there

,j g i4 would be no significant energy contribution from a metal - l} water reaction in the event of the postulated MCA. Any energy !j g contribution that might be made would not affect the peak i enclosure pressure because of the several minutes aftel-I blowdown before core temperMares reach melting levels. l 3 Thereafter, the small amount of energy that might be con- %=r tributed by a metal-water reaction would have negligible effect on the enclosure pressure transient. Io

13. 3 B ASIS FOR CONTAINMENT DESIGN 13.3.1 The basis for containment design was taken as that operating j

condition where the internal energy of the coolant is a maxi-mum. This is the case at " hot standby" where some of the Q i s, e -e

O i Section 13 Page 5 t i h primary system volume, normally occupied by steam during ope ration, would be filled wi_th hot water which has a greater energy content per unit of volume.

13. 3. 2 The analysis-of the " maximum credible accident," on the m

other hand, is based on reactor conditions where the fission %s' product burden of the core is at a maximum. Since this condition only exists during full power operation, total A system energy at the time of the accident will be less than V that used as a basis for containment design. Thus, the l peak enclosure pressure resulting from a primary system O. 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 0-

• • '**"*'"****""'"***8""'***"'*******^"**"*

~- 27 psig at an early stage in plant design. 4. O

13. 4 "MCA" PRESSURE CALCULATIONS f

l Q

13. 4.1 Peak Pressure Calcula; ons

13. 4.1.1 The largest process line to the reactor,..which would give

'l $ the most rapid release of total system energy to the con-tainment vessel, is assumed to suffer a complete instan- ,i taneous circumferential break with all primary coolant l h issuing from both sides of the break. Accordingly, the i 20-inch recirculating pump discharge line is assumed to break at a point near the bottom of the reactor vessel. i 13.4.1.2 Pressurized hot water partially choked by steam formation initially flows from the break, followed by steam after fh 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. Oq 13.4.1.3 No energy contribution from a nuclear excursion is included on the basis of considerations given in paragraph 13. 2.4

13. 4.1. 4 No energy contribution from a metal-water reaction is includea on the basis of considerations given in paragraph Og 13, 2. 5.

13.4.1.5 fp) The conditions in the containment vessel free space, prior C to the accident, are assumed to be 100*F temperature and 100% relative humidity and atmospheric pressure, h 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. 5 13.4.1.7 The free volume (9.4 x 10 cu. ft. ) assumed in the contain-O' ment vessel was that occupied by atmospheric air prior to -3 the accident. O -~

i t Q Section 13 Page 6

13. 4.1. 8 Heat is added to the free space by cooling of the fuel during the tirne interval between the break and peak pressure.

i

13. 4. 1. 9 Heat transfer to tie cold masses within the containment vessel is assumed :o occur using heat transfer coefficients of 700 d

j and 240 Btu /hr-ft

• F.for steel and concrete, respectively.

Heat transfer to these cold masses is assumed to occur l from the time of the accident until these surfaces are in j thermal equilibrium with the containment vessel atmosphere. I Heat losses to the outside of the containment vessel are O "**"=d"** c c " ' " " 'i ' " ' ' '" * " *" ' " = * " ' " ' ' '

• 2 1

pressure has reached its peak and started to decline. 1 Q

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 Q

psig, and occurs approximately 16 seconds after the break. and is shown in Figure 13.1. i O 2342 Po=* Accide"* Pre = ="re aed"ctio"-

13. 4. 2.1 After the initial pressure peak in the " maximum cred ble Q

accident," pressure and tempe rature in the enclosure would undergo changes with time as a function of the following competing mechanisms: O (a) There would be heat losses from the enclosure atmos-phere to the enclosure shell, to the solid structures in O .the enclosure (which, on the average, would be initially i i at a lower tempe rature than the vapor space), and to the environment outside. i O (b) There would b heat gain from radioactive decay, and l from cooling of the primary system metal masses. O

13. 4. 2. 2 The calculated enclosure pressure and temperature transient curves are shown in Figure 13.1. As indicated by these O

curves. ogeration of tue gost incident egrav system and core spray system results in containment vessel temperature and pressure subsidence sooner than would be the case in the O event neither operated. soth of these seter systems wouid 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 O

sjstems operate to bring the pressure down to slightly above atmospheric pressure after about 6 hours following the rupture. It was assumed that after about 5 hours the core spray system O would begin recirculating water from the containment vessel through the system's heat exchanger. The water level in the containtnent vessel would have risen to about elevation 587 0 feet, which would be sufficient to provide suction 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 O

I O I . p Section 13 Page 8 protection system for about the first 5 hours, followed by water recirculat;on through the core spray system while the post inci-dent system continues to spray fire protection system water for Q approxin ately an additional hour. ~ This method 'of operation would have used approximately 1/3 of the available capacity in the containment vW for water accumulation under the Q given conditions. Thus, intermittent operation of the post incident spray. system after the.2nitial 6 hour period would allow control oPthe containment vessel pressure for a longer Q period of time without recirculating containment vessel water, which may be highly contaminated with fission products, through the post incident spray system.

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

! h damage rather than affect the containment vessel pressure transient. Operation of the core spray system would be auto-matica11y initiated when the reactor system pressure dropped O to zoo is. -hich wou1a occur no"t 15 eco a <ter the v - e tem rupture. The additional cooling provided by the core spray system (in comparison to non-operation of the core spray 0 v te m ). wouie net red =ce the co t i m ent ve e1 ere ure untii after the post incident sprays operate, because this initial core spray cooling capacity would be effectively used O in removina heat tored in the core end erim rv met i m eece. In the event that core spray system cooling was unavailable to the reactor core, the stored heat in the core and primary h metal masses would be relt.ased to the containment vessel atmosphere ata slower rate, thiswate was found to be dependent on the transit time through the ruptured system and through O the equipment insulation. This transit time was calculated to be in the range of about 100 sc conds. O

13. 5 CORE HEATUP TRANSIENT O

rigure 13.1 shows the effect on ti;e arg Rech Point reector core l frons the postulated loss of coolant accident. In order to develop I circumstances by which a spectrum of consequences could be analyzed, assumptions were.made as indicated in paragraph

13. 1. 4.

This included the assumption that cooling was unavail-i able to the reactor core which would result in core meltdown, l anu the assumption that cooling was partially effective so as to limit the extent of fis+ ion product release to 10 percent of the core meltdown case. It is noted that in connection with the assump-O. tion of unavailable cooling for the core that 'he core spray system t is assumed io be operable in all other respects; these conditions O would r.esult only in the event of a highly improbable simultaneous 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 result in little or no cif site radiological effect. l 0

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Section 13 page 9 (

13. 5. 1 Cooling Unavailable to the Core

(

13. 5. 1. 1 When the reactor cooling water is lost from the core through I

the primary system rupture, the fuel rods would become h 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 h 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 pv locations with a typical fuel bundle in each of five core radial re gions. This model was utilized during the initial seconds a of the transient. After about 20 seconds, all core nodes (,) had a temperature ' drop across the fuel pellet of less than 200*F, and the temperature drop across the cladding was n ne gligible. With this redistribution of t:.mperatures apparent V 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 p node at the surface of the cladding. The two-node model was V employed untWall.cf the. cladding reached perforation temperature (1600*F), at which time, the surface node was removed from A the model, and the single node.of fuel and clad temperature v was used to the completion of fuel melting.

13. 5.1. 2 The core heat was based on reactor shutdown in four secoads 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 p

heat is used twice in this evaluation: once, assumed to v leave the fuel to increase co ntainment vessel pressure, and here, retained to heat the fuel. ) 13.5.1.3 The above analvtical model gave results which are plotted on Figure 13.1 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.

13. 5. 1. 4 The above analysis is believed to be conservative. It is expected that the melting of UOg fuel would stop short of a complete meltdown as a result of heat losses by radiation to the vessel wall and.ttien to the containment vessel atmos-phere, in addi6ori"MTeatlossos by conduction from fuel pellets which would }%,e dropped out of the core to the Q

bottom of the reactdr vessel. Analytically, no credit was taken for either such radiant heat lor.s:=s or such fuel pellets which would be free to drop once their temperature g reached about 2550*F, the clad melting temperature.

13. 5. 2 With Core Spray Cooling O

13. 5. 2. 1 In the Big Rock Point reactor, the core spray system starts automatically on simultaneous signals indicating ' low reac-O tor pressure and low reactor water level. Water (at an O

nJ O Section 13 page 10 assumed temperature of 50*F) wo21d be pumped into the reactor at about 400 gallons per minute and 200 psig, to O 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 240 Mwt decay heat rate.

13. 5. 2. 2 Although it has been conventional to analyze the " maxi-O mum credible accident" (MCA) on the basis of 100% core meltdown, the most prob-ble "MCA" is considered to be represented by the included 10% " melt-O down" case. Considerations bearing on this conclusion include: (a) even with a partially effective core spray, the total melting would be significantly less than 100%,

O (b) radiant heat losses from the core would preclude some outer rows of fuel rods from ever reaching melting tem pera-tures even wi~thout any core spray cooling, and (c) when O 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-tures for those pellets. ~ {

13. 6 EMISSION OF FISSION PRODUCTS TO CONTAINMENT VESSEL Fission products in the reactor fuel becorne 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 Q analysis was made.

13. 6. 1 Fissien Product Release From Fuel O

13. 6'.1.1 As indicated earlier, the fission product inventory of the reactor core was based on long-term oper ation at a steady O

st te 9 -er te-et f 24o ther= i mes - tes "9 te the time of the "MCA" rupture. The quantity of the various fission products in terms of curies per megawatt is consistent h with ORNL-2127 information.

13. 6.1. 2 The initial release of fission products would occur when th'e 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 0

na bromine. The desisn of the fuei c1>ddins en the b sis of experimental work at VAL, provides adequate capacity to withstand the f.nternal pressures generated by a 50% O retease of nobie sases and heiosens at the end of fuel iife (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 h 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 O totei core inventerv ef nobie s>ses eed heiesens womid be i released at the time of clad perforation. O i

h Section 13 Page 11 0

13. 6. 1. 3 As shown in rigure 13.1, onse of feet me1 tina, 5= r, occurs with completion of perforation of all the fuel rods.

O Although it is believed that some fission products would continue to be releesed during the heat-up interval to 5000*F, the analytical model considered a two-step release: O initially, 20$nf the gaseous fission products, and finally, the remaininC 8C% of.neble gases and halogen groups, 50% of the volati1. solids (Ru, Gs, Te, Se), and 1% of all oth'er 4 O solids were a.sumed 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, ORN L-C F-6 0 14.

13. 6. 2 Fission Products Available in Containment Vessel O

13. 6. 2.1 Certein frectiens of the verious fissien preaucts wi11 meve from the fuel to the containment vessel free volume depend-O 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 g the assumption that 50% of the halogens and 70% of the a' solid fission product groups would be plated out. Thus, fission product transport from the fuel to an airborne con-O. dition in the containment vessel free volume was in accord- ~ ance with the following percentages: 100% of the noble gases (Xe, Kr) 50% of the halogens (I. Br) 15% of the;..yojshle_4olids.(Te., Se, Ru, Cs) g 0.3% of all other solid fission products

13. 6. 2. 2 The inventory of fission products which remain in an air-borne condition in the containment vessel free volume would be a function of the deposition and washout which would occur in the enclosure as a result of the turbuler.t and Q

water-condensing atmosphere, ivhich would prevail. This, analysis considered the reduction of free volume halogen fission product inventory by deposition as a removal of Q 1 x 10- fraction per second. This deposition rate was estimated on a haaie af.= e=+imated 1 cm/sec deposition velocity and an average 10 meter distance to a deposition surface. The expected natural condensation of water vapor and rainfall by the post incident sprays would provide many " sticky" surfaces for adherence by the halogens and subse-Q quent washdown into the water sump of the contair. ment vescel. The analysis also considered the effect of different concen-trations of the water-soluble iodine in the water phase in I 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 Q ' ratio of iodine concentration in air to that in the water was i n v-i

O Section 13 Page 12 O' found to be greater than 10-4 As long as this ratio is greater than 10-4, io' dine removai from air would be expected to con-tinue on the ba' sis.,of.idformnti.on given in reference ( AECL-0 1130, " lodine Containment by Dousing in NPD-II, " by L. C. Watson, A. R. Bancroft and C. W. Hoelke, October 27, 1960) O

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 evaluation considered the reduction of free volume solid fissiongfraction roducts (volatile and all other solids) as a removal of 3 x 10-per second from the combined effect of deposition and washout.

13. 6. 2. 4 The fission product inventory in the containment vessel free Q

volume also is reduced with time due to radioactive decay, which has been factored into the evaluation. Q

13. 6. 2. 5 The resulting fission product inventory in the enclosure vapor space (or containment vessel free volume), is given in Figure 13. 2 in connection with examples of 100% and 10%

Q fission product release from the core. O

13. 7 LEAKAGE FROM CONTAINMENT VESSEL uJ

13. 7. 1 The leakage rate of the various fission product groups was Q

determined based on enclosure free volume fission product inventory as outlined above, and the leakage rate variability due to enclosure residual pressure reduedon. The leakage O r tes t v rious times fter the accident Are shown for the noble gases and halogens on Figure 13. ~,, and for the solid groups of fission products on Figure 13. 4. The leakage rates h for noble gases and halogens are seen to be in'ignificant for s about the first few minutes and then rise rapidly as a result I of their release from the fuel cladding. The leakage of the solid O ti to= rroa=ct sroue i i==is tric =t-c me r tiveir ror 1 the first 15 minutes as a result of their later in-time release from the fuel. The leakage of noble gases and halogens is j h the most significant due to the larger fraction escaping from l fuel, with leakage of noble gases continuing as long as posi-tive containment vessel pressure prevails since radioactive ,- 3 (cg) decay provides the only mechanism by which noble gas in-ventory is reduced. h

13. 7. 2 Mass leak rates of air, water vapor, and airborne fission prodacts from the containment vessel were calculated in accordance with the orifice flow equation given by reference O

(" Orifice Meters with Segercriticei Comgressibie Fiow," ASME Paper No. 50- A-45, by R. G. Cunningham,1950). It was assumed that the minute imperfections that might exist j 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 aoproximation as most any ! O otaer seometricei shage of an eseiveie'nt cress-eectie# 1 ogenins would result in lower leakage rates in connection with driving l l pressures in the range of 1 psig to 5 psig. l 0

O , Section 13 Page13 Figure 13. 2 i O 3 10 0 ,1,,,, I,,,, ,,,I,,,, ,,i,,,, = ,, i,,,=i g g 2 O O 2 10 10' 2 O 2 O N = s D'\\ O I O MALOCENS O - O rT 0 a 10 / ) _ 10-'j 9"'y 1 .- =- i h l ~_' 3 E 0,0 s 5 f \\ _ iO-2! -i 5 f O~ OTHER i SOLIDS i \\ 0n-2 ? ( \\ i 10-3 O C O.) I I fl 1 11 1 I,I l t i t t ii!lt Ni I l t i ti lItti lOM ft l i02 3 " '"' 10 W4 "' ** ~ 103 106 NNE MONT"lO7 O l flME AFTER ACCIDENT. SECOND$ 1 l CONTAINMENT VE5SEL YAPOR SPACE FISSION PRODUCT INVENT 0RT O 1 A

O. 6 Section 13 Page 14 i 'O Figure 13. 3 t t i O t 1 I I l lill I i Il1Ill I i l llill I l l l lIl l I i l llill = 4 N06LE CASE $ (10M MELT CASE) ha A / a J O i MALOCENS I Con uf LT CASE)

l. O

- m. / \\ \\. NCBLE CASES /' j g% - ~ o n cAit) g / g ( /m N q r% ,/ \\ \\ s_ R / \\ N, $10-2 \\ I HALOCENS o n cAsEn \\ h. ~ \\ \\ i \\ \\ i 3 l! O \\ i g ~ \\ ~ i! O : \\ t { l \\ \\ 1i Q \\ \\ l in-4 t I O E = \\ \\ t-g \\ 1 ,1 \\ I l tif t I I.I lttet i \\ t l, t t.; I il l t ttt' t,, lltt I i 102 3

• E" ""'

10 M4 NE DAY"lO5 ONE WEEKJ 106 %NE MONTH 7 h Tiut ArTEn AccrocwT,sEcoNo$ uiuce uns anoste cisEs a utocrus ? o@ k

O Section 13 Page 15 Figure 13. 4 l l = I I i l Illi i l I l lill i i l l Ill i i i i l 1811 i i i g lig g = 4 O : 1004 CORE MELT EX AMPLE ~ 104 CCRE stLT EXAMPLE O0 10 O ~ O O-l A 2 O gtgete O i 8, io-2 N, >\\ \\ u 0 i / \\ I t, / \\ Eff', / r \\ I \\ 1 \\ I 3 I 5 1 } Z l \\ h Z I \\ ~ osuen r scuos g ~ \\ k 4 l [/ \\ s 5 \\ \\ ~ 7 \\ \\ O / s \\ I \\ l \\ ~ ilI I i ll tit \\ l %._ _ f f \\l t i t_ f I I!!!i! I I i !!!!! I ! I 102 3'"~"" 4 '"E 10 W '#~105 ONE WEEnd 106 %NE MONTH 7 O . <.... C........C... " " ^ " " ~ ' ' ~ "' ~ '" "'' " O O \\ l

i O Saction 13 Pade 16 i O

13. 8

SUMMARY

OF RADIOLOGICAL EFFECTS O

13. 8.1 The r diotoaicet effects efthe "m xim= = credisie c cident "

of interest at off-site locations a re of two forms: firs t, the direct radiation from the fission products contained in O the free volume sgece of the re ctor enciosure; end. second, possib:le leahage of..asnialldractiorreof these fission products from the enclosure. In the latter case, the radiological O effects wouta de due to direct redietien from the gessing " cloud, " direct radiation from radioactive materials deposited O on the ground, internal exposure due to inhalation of radio-ceive meteriels es the c1oud gasses the goint of e:<gosure. and possible contamination of a ricultural produce. o O

13. 8. 2 zxemgie redieienicel effects ere shown for twe sesic essemg-tions invoiving the consequences of both 100% and 10% core melting. As the 10% example obviously has a higher proba-O' bility of more closely approaching an estimate of possi' ole con-sequences of the initiating event, references in this text O

are to this e:: ample. Consequences of the less procaole 100% example may oe read directly from the figures for this section, which appear immediately following paragraph 13.17 (pages 30 through 40, inclusive).

13. 8. 3 The off-site radiological effects are general % insigificant O

and would not cause any concern for pu'aiic 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 O.

case at a distance of 1/2 mile from the center of the reactor enclosure is estimated to be 0.1-roentgens in the lii st two hours after the initiation of the accident, and an estimated O. 6 roentgens for the entire course of the accident.

13. 6. 3. 2 The direct raaiation on the center line of the passing cloud 1/2 mile fro:n the reactor enclosure for the 10% core melt case is generally insignificant. In the first two hours after the accident, the estimated dose will range from less than 1 mrad O'

for lapse diffusion conditions and average wind speed, to an estimated 10 mrads under inversion conditions with low wind speed. Durinj the entire course of the accident, the passing g'a. cloud dose will ran;,e from about e mrads for lapse conditions Nr with average wind speed, to an estimated 60 mrads for inversion conditions and low wind speed. CX C

13. 6. 3. 3 Raaiation dose received from fallout on the ground at a distance of 1/ 2 mile from the reactor enclosure for the 10% core melt case will generally be insignificant., At this point on the cloud center line during the first two hours after. the accident, the dose from fallout is estimated to be about 1 mr under inversion g

coa,iitions with low wind speed, and less than 1 mr durin;, any c::am, ale of octter ciffusion conditions. During the entire course of the acci ant, the dose fro.n fallout will range from about 10 mr Q fur lapse ciffusion conditions and average wind speed, to an e s tirc.a ta a 35 mr Guring inversion conditions with low wind speed. ' O

• 0 Ssetion 13 Page 11

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 fiom 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 lapse diffusion con dtions with average wind speed, to an estimated 2 AL/ rems during inversion conditions with low wind speed. During the entire course of the accident, the lifetime dose to the thyroid is es-Om timated to range from about 30 mrems during lapse diffusion con-d.tions with average wind speed to an estimated 4 rems da' 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 passing cloud at a O

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 O auring lapse diffusion with average wind speed, to an ertimated 0. 3 rems during inversion conditions with low wind speed. During the entire course of the accident, the lifetime dose to lung will range O~ 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.

3. 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 lai<e after passage over only a small area of land a significant portion of the time.

13. 9 DIRECT EXTERNAL GAMMA RADIATION FROM ENCLOSURE Q 13. 9.1 The quantities of fission products available for direct external gamma radiation from t,he enclosure were based on considerations discussed in paragraph 13. 6.

The direct radiation is a sensitive function of the l @ gamma energy levels of the radioisotopes present, because of the variable shielding effect for different gamma energies of the large thickness of air available between the enclosure and the site boundary. M Therefore, the evaluation was made oy calculating the exposure con-tri.>ution from each gamma radiation level from each isotope in the l noble gas, halogen, and volatile solid fission product categories to-l ( (h' gethe - with their appropriate daughter fission products, and by con-sidering shielding and buildup factors for both air and the steel enclosure wall. g 1 v

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 Figures 13. 5 l

g) and 13. 6, at distances of one-half and one mile. At the one-half mile g distance, the maximum dose rate of abrut 10 mr/ hour occurs in the one to two-hour period af ter the accident, and rapidly diminishes thereafter. i.h 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-h ment of nearby inhabitants appears necessary. (This appears to be a j reasonable conclusion for the 100% example accident, also. ) L

--_~.-.._n .: l Section 13 Page 18 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. O These are the poor diffusion conditions caused by inversion (stable),. typical of warm weather nights, at vind speeds of both one and five meters per second; and the better diffusion O 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 O radiological effects at other wind speeds may be approxi-meted by considering the effect to be inversely proportional to wind speed. A limitation should be appreciated at dis-Q 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 Q in the case of deposition on the ground which, under constant diffusion conditions, is largely independent of wind speed. Q 13.10. 2 The atmospheric diffusion methods of Sutton were used for the neutral and unstable cases. Due to the errpirically in-dicated inadequacies of :he Sutton method for inversion con-Q ditions, calculation methods oased on Hanford diffusion results, as outlined in Report HW-54128*, were used for the inversion cases. O 13.10.3 Weather Conditions 0 This evalu tion ssume d th t the we ther conditions involved no precipitation and that the incident occurred during hot summer weather. Precipitation would deposit more con-O <>mi= tio= c e e te the 91 et the= tai evetu tio= i=dic te. thus reducing contamination levels further awa.. If the incident occurred in cooler weather, the fission product O teakese from the encioe=re woeid se 1e s the# i=diceted due to more favorable heat transfer and, consequently, more rapid reduction of the enclosure post-accident pressure. O 13.10. 4 Elevation of Release O teawese from the encioeure is censider a to occur near the ground level. This appears reasonable at.nost enclosure penetradons are near grade. If the postulated leakage O"3 occurred at some significantly different height, such as by emission from the stack, the off, ! ant deposition, and possi-ble inhalation would occur at greater distances than this O~ evaluation shows, but their magnitude would be vastly reduced. ~

• HW-54128, "CaicuJations on Environmental Consequences of Reactor Acci-dents, " Interim Report, by J. W. Healy, December 11, 1957.

O 9 O

Section 13 1 Page 1.9

13. 10. 5 Initial Dilution by Building Wake 1

l @%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.that i Q the effective wake cross section is of the order of one half. V of the verti 11 cross section of the enclosure structure. No additional immediate dilution by other nearby structures or other ground cover !.s considered. This effective wake has been equat<.e e u.c=cle 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 enclosure sphere radius. Q To obtain an estimate of this initial dilution of the leakage, the radius of 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: O Symbo1* Diffusion Wind Speed Virtual Source Distance Q I-1 Invers;on 1 m/s 250 meters I-5 Inve rsion 5m/s 370 N-5 Neutral 5 m/s 200 U-5 Unstable 5 m/s 80 These estimates of the virtual source locations agree generally h with the methods of Holland *for the neutral and unstable cases and are more conservative for the inversion case. h 13.10. 6 Effect of Distance on Diffusion The uownwinu effects, such as passing cloud dose, ground h ueposition, and inhalation exposure, are a function principally of the integrated air concentration at any point. This integrated h concentration subsides with distance due to turo' ulent diffusion in the atmosphere, and depletion of the contaminat ed cloud by deposition on the ground and on ground cover. The magnitude of this effect for the example distances and diffusions is: TABLE 13.1 Unit Integrated Air Concentration (uc-sec/cc/ curie released) Distance Diffusion Gases Halogens Particulates 1/2 mile I-1 4 x 10-4 3 x 10-4 4 x 10-4 I-5 1 x 10-4 7 x 10-D 1 x 10-4 N-5 2 x 10-5 9 x 10-6 2 x 10. 5 U-5 5 x 10-6 3 x 10-6 5 x 10-6 m b) These symbols are used throughout the evaluation to indicate the four diffusion examples illustrated.

• • As indicated in " Meteorology and Atomic Energy. "

j m -- m.

Q.- Section 13 Page 20 v Rev 1 (3/19/62) Table 13.1 (Contd) Unit Integrated Air Concentration O (uc-eec/cc/ curie reieased) Distance Diffusion Gases Halogens Particulates I mile I-l 2 x 10-4 1 x 10-4 2 x 10-4 I-5 6 x 10-5 3 x 10-5 6 x 10-5 N-5 7 x 10-6 3 x 10-6 7 x 10-6 O U-5 2 x 10-6 8 x 10-7 2 x 10-6 O 3 mues 1-1 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 O-U-5 3 x 10'7 2 x 10-7 3 x 10 ~ O 23 20 7 Prod dititv er v rio== wi a s9eea a oirectio - 13.10.7.1 O Meteorological data taken at the site during the pc.st year generally substantiate the five-year data from the Charlevoix Coast Guard Station quoted in the previous hazards summary g reports. Wind data taken at the 32-foot.leve1 indicate that about two-thirds of the time, mclosure 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 Q Direction Azimuths of Wind Movement From Plant No. I 280 -360 -20 O Blowing inland, sparsely populated / o 2 30 -50 B1 ewing intena. tewerd Cherievoix (3 mi1ee) 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 to shore h 6 Any Calm (less than 4 mph) 13.10.7.3 O 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 in parentheses. h Directions I and 2 are totalled, since they represent the total probability of the leakage path being over land adjacent to the plant. Similarly, the over water directions (No. 3, 4, 0

5) are summed.

d 10

O Section 13 Page 21 O TAB II 13. 2 Wind Direction Frequency, Percent; and Average Wind Speed, (mph); 32 Foot Level, Big Rock Point, Michigan O Direction # 1 2 3 4 5 6 1, 2 3,4,5 Nov. 60 2 4(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 4 9(19) 50(17) Jan. 61 4 2(17) 3(09) 36(08) 9(18) 7(14) 3 4 5(16) 5 2(11) Feb. 61 1 7(14) 1 (11 ) 39(09) 11(13 ) 2 5(13 ) 7 18(14) 7 5(11) Mar. 61 33(15) 5(13) 3 0(10) 6(17) 23(14) 3 3 8(15) 59(12) Apr. 61 3 2(13 ) 4(11) 20(08) 16(09) 20(14) 8 3 6(13 ) 56(10) O 33(12) 61(10) May 61 32(12) 1(09) 27(09) 19(12) 15(10) Jun. 61 15(12) 2(10) 32(09) 18(13 ) 2 4(11) 9 17(12) 7 4 (11 ) O 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) O Total Period 29(14) 3(12) 33(09) 13(14) 16(13 ) 7 3 2(14) 61(11) Overall Wind Speed, 11 mph. O Q

13. 10. 7. 4 The overall wind speed of 11 mph corresponds to the 5 m/s O

wind speed used in three of the four diffusion examples i 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 i C 1% to 5%) of the time. O Cl

13. 10. 7. 5 The probability of leakage movement toward Charlevoix (Direction #2) is only 3% of the time, with the three mile di" *^"c" * " ' * " "

• d" c i " 8 '" * " ' 8 "t '"" * * ' " " * "

O 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 O"

minimized by the eleven mile distance available for atmos-pheric diffusion. 13.10. 8 Wind Direction Diversity 13.10.8.1 The accompanying graphs for dose rate from the passing b) cloud, the rate of deposition of fall-out on the ground, and the rate of inhalation of air-borne contaminants are based O

,h 0 Secggn 13 Page 2 2 O on no wind direction change, since such effects are expressed in units of instantaneous rate functions. However, to inte-Q grate an effect such as any of these over a period of sever-a1 hours would be unrealistically conservative, as minor winu airection changes.are occurring continuously, even () during periods $nen the reported genera 1 wind direction represented by4 45* angic is apparent 1y remaining constant. I h

13. 10. 8. 2 The standaru deviation of c1oud width is a measure of the zone occupied by the postu1ated contaminant plume at various distances and diffusion conditions. For the e:: amp 1es O

ev iuated, this is: O stenderd Deviation of cious width, meters Distance I-1 1-3 N-a U-o 0 1/2 miie 30 24 60 11s 1 mile 42 36 10 0 205 0 3 mi1e s 72 32 210 460

13. 10. 8. 3 With the expected Gaussian distribution of cioud concentration in the vertical and cross-wino directions, a centerline effect 0

is reuuceu by a factor of about 10, if the point of exposure is two standard ueviations of cloud width from the center 1ine. Consiuering the over-all precision of the eva1uation, such a O "* 3"c t' " =" Y '* '* "" "^ r ' "'i"* ' e d " " ""i ci * "t t consiuer the effectnonadditive. Thus, wind diversity fac-tors may be related to the wind direction change which will O: causa the piume c.,nterline to rotate an angle equal to twice the standard devia: Ion at the distance and diffusion of interest. l Such angles for the examples evaluated are: l O Angle of Two Standard Deviations degrees @V l Dis tance 1-1 14 N-> U-o l 1/2 mi1e

3. 4

2. 8

6. 8 18.3 9

C 1 mile

2. 4 2.1

5. 7 11.6 3

mi1e s 1.6 1.2

4. 8 10,4 1%

v

13. 10. 8. 4 "Meteoro1ogy and Atomic Energy," pages 66-67, provides

["s) a basis for the degree of wind direction v triation expected l '~' for the diffusion cases eva1uated. The wind direction of I typical inversion periods may be expected to vary by about l t*) 10*. During periods of typicai neutral lapse rate conditions, a winu direction variation of about 20* may be expected, and a variation of the order of 50* may be anticipated during O tvgicei e veime =# t sie c #ditio==- There is seed erod-I ability that the actual variation wou1d be greater than assumed O

O . O Section 13 Page 23 O here during a period of interest of several hours; there is a small pon;sibility that during some periods, the direction O varietion wou1d not be this ierse. 13.10. 8. 5 From such considerations, it is reasonable to conclude O that intesreted effects, coneideri18 wind direction veri-ation over a few hours, may be reduced from those ef-fects calculated for unidirectional winds for the examples O used. byfacters a: O Reduction rector Frem wind Direction oiversitv Distance I-1 1-5 N-5 U-s O u2 mue 1-v 2 2 1-u z 1-v 2 1 mile 2 2-1/ 2 2 2 0 3 mues 3 4 2 2-u2 13.10. 8. 6 Such reduction factors are for a period where the wind is remaining in one nominal direction. When a significant g wind direction. change occurs, the possibility of any additive 47 effect at any one location for the periods before and after the change is completely terminated.

13. 10. 9 Travel Time O

The eva1uation includes the effects of wind trave 1 time from the enclosure to distances of interest off-site. While some additiona1 radioactive decay occurs during such travel periods, p'M 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 a perioci of time is available for evasive action to occur, if it is determWed that such may be desirable as a precautionary measure. ] 13. 11 EXTERNAL RADIATION DOSE FROM PASSING CLOUD 13. 11. 1 Evaluation of effects of passing cloud air concentrations down-h 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 (h dose for he 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-O tration required in an infinite cloud to produce a certain dose O. J

O Section 13 Page 24 0 was evaluated for the radioac'tive decay periods of interest 0 in the post-acciuent pe riod. 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 beometry are: Air Concentrations (uc/cc) Giving One mrad /hr Ibse Rate Decay Time Noble Gases Halobens Volatile Solids 1 Hour 1.6x10-6 0.78x10-6 1.4x10-6 4 Hours 2.3x10 ^ 0.79x10-6

2. 5x10-6 8 Hours

3. I x10- 6 0.88x10-6

2. 5x10 -6 16 Hours

4. 2 x10-6 1,o xio-6

2. 6x10 -6 O

13. 11. 2 The dose from the passing cloud based on uniform concentration O and infinite cloud considerations was then corrected for the finite ctoud size end Geussian distributie:> ef cioud concentration. For the various diffusions evaluated, and for cloud sizes calculated O at the distances illustrated, the ratios of finite cloud to infinite cloud effect are: Finite Cloud Correction Factor Distance I-1 I-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 i

3 (V

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 passi.'q cloud effects. This correction is actually of a small magnitude since most of the passing cloud effects are due to noble gases, j

9

13. 11. 4 The instantaneous dose rates from ihe passing cloud at the one-half mile distance and on the centerline of the postulated plume 9

^re ' -" i" ria"re i3 7-The = *i= "= a e r te cc"r= ^h "* l-1/2 hours after the accident, and is of the order' of 5 mcails per hour; by the end of one day, it has subsided to less th'a'n 1 mrad /hr.

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 ' f 60 i o mrads. These calculated doses assume that the receptor is on [ the center of the cloud path and that r.o incidental shielding, I h such as that provided by housing, is available. [ u

O' w Section 13 rage 25 13.12 EXTERNAL RAD'ATION DOSE FROhi GROUND DEPOSITION 13.12.1 The fall-out concentrationt if radioactive materials were determined on the basis of jarticle settling by eddy diffusion only, since settling by gravity is expected to be negligible in this case. It is expected that the particulate radioac.tive material which might, leak from the enclosure will be only a few microns in diameter. If the material were of a G significantly larger diameter, it would be deposited at a much fastor 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 enclosure since the leakage that may occur is expected to be restricted to that which could pass through minute imperfections in the Q wall or through penetration seals.

13. 12. 2 The extent of halogen and solid fission product deposition on h

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 O used in this evaluation were based on British results cited in HW-54128, and are: 6 O negosition vetocity ca recteristics h Ratio of Deposition Velocity to Deposition Velocity, Wind Wind Velocity cm/sec 0 Diffusien veiocity vertic1e s neio8 ens verticies ne1esens Inversion 1 m/s

2. 2x10-4

3. 4x10-3 0.022 0.34 Inversion 5 m/s 2.2x10-4

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

0

13. 12. 3 The evaluation provides for correction due to radioactive l

decay after the materialis deposited on the ground. As the Q amount of deposition is a function of air concentration, and as the air concentration is depleted by prior deposition at locations closer to the source, correction for this depletion has been made for deposition at the distance illustratd. In h addition, the dose rate from the deposited material has been corrected for the finite size of the deposited source. This h correction is a function of the standard deviation of cloud width, and for the example df(fusions and distances is: O O n

! O. h Section 13 Page 24 Finite Deposition Pattern Correction Factor Distance I-1 I-5 N-5 U-5 0 1/2 mile

0. 41 0.38 0.49 0.54 1

mile 0.45 0.43 0.53 0.60 3 miles 0.50 0.48 0.60 0.67 G

13. 12. 4 The conversion from deposition on the ground to gamma radia-tion dose rate at the conventional one meter above the ground C

was made considering the gamma energies present from the halogen and volatile solid fission products of which the deposited g material, is composed. The conversion is dependent on the age U of the fission products present as follows: Decay, days mr/hr at one meter / curie / meter" 0.1

1. O x 104 O

i 9 1 x io 3 10

7. 2 x 10 O

2o0 3 6 x to O 13 12 5 oree d degesitie at the exemgle diffusie s u der the ssume-tion of unidirectional wind, is shown on Figures 13. 9 to 13.12. The period of significant deposition is completed about six 7 1 hours after the accident, as the leakage after that time is largely noble gases. A principal significant component of the O deposition is Iodine-131, which is shown in addition to the total deposition concentration on this series of figures. 13.12.6 g The possible integrated doses from radioactive material on the o ground, using the previously established appropriate wind diversity factors, at the example diffusions resulting from 0 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 a person continuously occupied a point one-half mile from the Q) point of leakage which has been subjected to the least favorable diffusion condition, a dose of the order of 35 mr would be a received. This assumes no reduction of deposited material Q) during the extended period by mechanisms other than radio-active decay. In all probability, actual integrated doses would a) be reduced by the effects of rain and other weathering action. (_ The doses illustrated are the maxima applying under the leakage plume centerline, with any incidental shielding such as housing pd not considered. s t O . O.

1 nD, i 'O e,c e tion 13 Page 27 0

3, 33 GROUND DEPOSITION OF IODINE-131

< O > > 33,1 Due t the radioisotopic composition of the postulated leakage, the most significant individual effect from contamination of off-sit land would be the possibl. control of milk production Q from azing cattle because of the Iodiae-131 food change relationship from vegetation to cattle to milk to man. Follow-ing the Windscale incident, the British used a limit of 20 rads O to the thyrota or chitdren - criterion for controt or miix. They indicated that such a dose would result trom milk con-taining 0.1 microcuries per liter, and that this concentration 1 O resulted from c itle 8rezins on P ture cent inina bout one microcurie of Iodine-131 deposited per square meter. O 13.13. 2 usins the grevieusiv established wind diversity factore, the immediate maximum Iodine-131 deposition concentrations would be: Immeaiate Maximum Iodine-131 Deposition, pc/m2 Distance I-1 I-5 N-5 U-5 1/2 mile 28 30 9 5 1 mile 8 10 2 1 3 miles 2. 2 41 <1

13. 13. 3 0.

Under the poorest diffusion conditions 'a zone with a length ~ of about four miles downwind and containing an area of less than one square mile might be above the proposed control 2 limit of I ic/ meter. No evacuation of humans or domestic J animals would appear necessary; the only possibly required control would be non-use of milk products from the area 0 (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: O Iodine-131 Deposition After One Month Deca /, pc/m Dis tance El I-5 N-5 U-5 1/2 mile 2 2 41 <1 0 1 miie <1 <1 <1 <1 3 miles <1 41 <1 <1 0 13.13. 4 After this period, the area of the zone would be significantly Q 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 O iimit within a few weeks (Similarly for the 100% example acci-dent, this conclusion appears reasonable after a period of about a month).

n.._,,.-- h* 1 0 Page 28 .,..., r. : ) INTERNAL DOSE TO THYROID Internal exposure to the thyroid gland from inhalation of the h ,,4,; fission product mixture in the passing cloud is primarily due to iodine radioisotopas. This exposure was evaluated con-h sidering the dose from thyroid deposition of Iodine-131,133 and 135. Other iodine radioisotopes of half lives of 2. 3 hours or less were not included, considering their lo~w rem per h microcurie ratio for lifetime dosage considerations, arid i because of the estimated 3 to 6 hour thyroid uptake time af ter the material is inhaled. The lifetime thyroid dose 0 wa c va1uetea for the three isotoges considerins e dreath-ing rate of 30 liters per minute, and a thyroid deposition of 15% of that which was inhaled. ,1

, ;4. 2 The total radioiodine deposition rate in the thyroid on the leakage plume centerline for the example diffusion con-ditions and at a distance of one-half mile, is shown in y

Figure 13.14.

).14. 3 Using the previously established wind diversity factors, the total lifetime dose to the thyroid from inhalation during the first two hours afte.r the accident is

g i Lifetime Thyroid Dose, rems, First Two Hour Exposure f Distance _I-1 1-5 N-5 U-5 i ) 1/2 mile 2 0, 4 0,07 0.02 1 Q 1 mile O. 4

0. 2

', b

13. 14. 4 Similarly, for inhalation during the entire course of the pos-v tulated significant halogen leakage, the lifetime thyroid dose is:

l Lifetime Thyroid Dose, rems Continuous Exposure ( ~ Distance _I-1 1-5 N-5 U-5 1/2 mile 4

0. 7 0.1 0.03

{, 1 mile 1

0. 2 t

h The evaluation assumes no. reduction in the concentration of air inhaled such as would be the case if the receptor were l within a structure. O 13.15 INTERNAL DOSE TO LUNG 13.15.1 Dose to the lungs was evaluated considering that all volatile and other solid fiss' ion products inhaled wsre insoluble, and I ~ by use of conventional standard man metabolic factors. O

__= 0 O Secdon 13 Page 29 0

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 Q one-half mile, is shown in Figure 13.15. The calculations indicate that lifetime dose to lung irom inhalation during the entire course of the postulated significant volatile and other h solids Icakage during the poorest diffusion condition would be about 0. 5 rems O

13. 16 INTERNAL DOSE TO BONE Q

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, bariu.n and cerium, together O

-ith their ergrogri te d us ter fi to= groduct - a

13. 16. 2 Due to the low fractions of these fission products in the post-O ui te21eewese. the

= 1 i i=-icete tuet the tifetime dose to 7 bone at any off-site location, and for any exposure time or diffusion, is always much less than 0.1 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 0' at any distance beyond the plant boundaries are not of a hazard-ous level, and ir fact are quite insignificant at the distance of ~s any small populacion center. The radiological effect cal-J culations of the example where it is assumed that all of the core actually melts show levels auout ten times greater n than those for the partial core melt example; even in chis case, U the calculated doses at any distance, one-half mile or greater l frons the enclosure, for both the first few hours after the n incident and for the entire course of the postulated release, U do not exceed those levels suggested currently as reasonable for site and plant safety evaluation. p .Ov q G Lj i 10

m O O Section 13 Page 30 Figure 13. 5 o io* = i i pio , i i1i.o ,I,,,, ,I,,,, ,l,,,, ~ O O ~ O N E SOLIO CURVES.100% COR E MELT EX AMPLE BR0xEN CURVES.10% CORE MELT EXAkPLE = O 1/2 alLE) 10, -:f N g i E / \\ a N 5 h 100 s n -!( a ! 7 l = i 8 r Oi \\ \\ I LO*l. j( w...LE \\ 1 O \\ s e a) ) \\ O l APPROXIMATE .g \\ \\ MAGNITUDE l g% 3 \\ k !5CrcRo"u'u'O f "^ i \\# l e

/

\\ I b' 10'3 -- t i I l t tit I,1 l I I II lt tI I t l t tes !,i I l tiri i r I i ( 102 3 NE HOURJ 10 d} ONE DAY 'lOS cNE wEexa 106 %NE MONTH 7 Time ArTER ACCU,Eur. sEconos DIRECT CAMMA RADIATION DOSE RATE FROM ENCLOSURE AT 1/2 AND 1 M!LE \\>O..

O O Section 13 Page 31 Figure 13. 6 iO4 : I I i l Illi i i ilIlli i I l lIlli i i l l l l 11 i l l sili-O O SCLID CURVE 5 100% CORE kELT EX AMPLE 8#0 KEN CURVES.10% CORE MELT EX AMPLE lO3 O 7 O ( O 'o * / yl O 5 10 l Q! / / 5 / 5 / ~ h- / ~_ 3 8 / h! 10 o / 3 r, h b / E l l lf i mite g O f) /. W ,/ / ~ 10 = l = / l / / 0 / / ~ v / O / / ~ / ~ pd 10-2 i ' i ll' I I!' i r i t i i I I II'1 I I I IIti I.' 'It l 102

• E 5

ONE WEEKJ 6 %NE MONTH 7 103 "E" ""' @4 10 0 0 1ME arrER accioEnt. sEconos l DitECT CAWtA RADIATION INTEGRATED DOSE FRO 4 ENCLO5URE AT t/2 AND 1 MILE O

) 0-14 Page 32 , Q Figure 13. 7 h l l 1lIllt i I l l1111 I il l Illi i iljilil l l 1lll i l si BROKEN CURVES CCREu EX u LE l el 8 INVE Ril0N, I W5 WIND SPE ED o

.::=:":=:::::

I U 5

• UNST ASLE,5 m/5 WIND SPEED I

b o E %,/ U.S i. N l5 \\ f/ ? O f f N'I e-s ,f U.S \\ ~ /[T M N I' / \\

9

/ / x s \\s 'Q \\ \\ e-l( /// l \\\\ // w \\ c i N N. \\\\ II \\ ' 0: 'N Ni N I// I y x \\ \\\\g' .I l 1 =::" 0 ii l N \\i

tate E O

Y I \\ \\ = l \\\\q l-e t U i \\,\\ l( l 1

O jfl l

u l O

• i l i i,i

,,,I,,,, ,, I,,,., ,, i iii, ,.,,i,,,, l h 103 NE HOURJ 4 ONE DAY"lOS ONE WEEKd 6 %NE MONTH "[ 7tuE AFTER ACCfDENT. 5 ECON 05 00$E RATE FROM PASSING CLOUD At 1/2 MILE DowitwlND 1

w. _ h h 0 Q Q f i e i E' l i I ~lTir7 1 r-rTir-I T p i r~r' r ~ rj1ittj - ; ~ ~; I j. ig i i 10*. E X A ntPL E S F O R SOLID CUf4vE S l*0" CODE wtiTtsawait BROKEN CURVE 5 - 10% CCRE utLT E R AwPLE N 5 AND U 5 LE55 THAN 10 MRADS l 1 z INVERSION,1 M/S wit 3D SPEED I-5 = INVER$10N,5 M/5 WIND SPEED N-5: NEUTRAL, 5 M/S WIND SPEED U-5 = UNSTABLE,5 M/S WIND ' PEED 3 I; 10 i 1 u i 5 t 8 e a G g . i -5 i a i 10 2 r I-I i t N-5 / ~ o / Rt 1 / a w j / . -s 35

1. ,~

,7 - --""' U-5 w I I I l l i ll I i l i-1 1 1 lill M l i ll I I I lill 10 onE noun J out oar ONI WEEK J b ON E MOMin 2 4 5 6 go7 10 10 10 10 10 TIME AFTER ACCIDENT - SECOt4DS INTEGRATED DOSE F. ? ' SING CLOUD AT 1/2 MILE DISTANCE

~ .,, :;on 13 Page 34 e Figure 13. 9 p j l l l l ll l l l l l ll l l l llll l l l l l ll 5 j l lll} 'O O 5 >,= f, % TOTAL C 1004 CORE MELT EIAMPLE IODINE 131 ~ ,,. = = - { / ~ N \\ <\\ 'h x x = [ 10% CORE MELT EXAuPLE l h. I- / g \\ i % TOTAL ( 60 CINE.131

• \\

g j f-- \\ j / \\\\ \\ ~ / / \\ 9 3, ' l'/ \\ \\1 l 0 \\ \\ Q g \\ \\ ~ 'Ir \\4 ii I II \\ N o. \\_ 0 I e-lI \\- h \\ 0 ir \\ 9 E l L (E } 2 l i __- g I h { I I \\ ' ! !! fl. 1 it l t t il l l f i t,I I I I l fitt I t ,f I tli,tt I I i 103 NEHOUR 8 04 "E "'~ 105 ONE WEEKJ 6 %NE MONTH 7 TIME ArTEn acc 0EuT.sEcow0s ~. E,o,,,,m1 m,u o,,,

e.....,,.. u,e,.m1,..< -

,,u o o,..mu,mo. 0 .=7 h

'~ ..:.va 13 Page 35 Figure 13.10 e . 1 0 l l l 0 l ll I l I I l ll I l l llll l l l lllII' ) I l llll 0 C p~ .....L* t 4L t. t s i A. i: 0 2 IC0lN(.j}$ -? ~- ~ r \\ / / h f [ f ,0% COR E ME L T t ' AMPLE C / ,,,A< . 00....,, / 0 ( \\ ~ 0: f NN \\ I,f / / w 1 0. '. g il // \\\\j i g' [ p O-ji \\ ll \\\\ 0 -' h \\ N \\1 \\ k-I l 6-l '2 I g O il \\ 7. I g* ~ ,[ I t-- { r 1 0 i, i ! f ill II l!Illt I,1 l t I ll i I I l t I!! I i I I I l lill I .g2 3 cNE HouaJ 06 'oNE MONTH "[ d ~' 105 oNE WEEK 4 "E l0 10 Tikt AFTER ACCIDENT. SECOND$ m,os.,,o,...,n._,1 ,so,,, ..,o,m,,owiuiso seE ro or s ure ws-

o

C4 N Section 13 Page 36 Figure 13.11 i A go4 i I I l l I l li I I l l I lli i l llIlli i l l a lii i v iil sit _i =... O O O'" 5 s M O TOTAL 100% CORE MELT EIAMPLE 102 t0DtNE e 135 e 9 x / g ff-m N Ot 10% CORE MELT EI AMPLE l / . TOTAL 10 DINE. 83 8 p Ou 1 a \\ j / f ~ ' N' Oi I // N \\ o ,e 1: / r\\ l,/ \\ \\ =s '/ i o 'y [ ll Q \\ jp f ,I Q 10 y 7 l : l[' \\- \\- ) i i.,, ,, i,,,., , I i,,,, ,,i,,,, ,,.,i,,,, 10-< i 106 %NE MQHTH 3 ONE HOURJ 4 ONEDAY"lOS ONE WEEXJ IO7 lQ2 IO fikE AFTER ACCIDE NT. SECONDS GROUND D: POSITION AT 1/2 MILE DIS'ANCE, HEUTRAL, UNIDIRECTIONAL WIND SPEED OF 5 METERS /5ECOND O

O O O D O O O O O O O O O O O O O O O O O'.O 4 10 [, j g g g g gg g g l 3 g gg l l l llll l l l l i ll l l l lllp p ~ g SOLID CURVES - 100'.' CORE MELT EXAMPLE BROKEN CURVES. 105. CORE MELT EXAMPLE l.1 = INVERSION, I W5 WIND SPEED 10% EXAMPLES FOR U-5 AND 1-5 = INVERSION,5 M/S WIND SPEED ~ N 5 CASE 5 LESS THAN 10 MR N-5= NEUTRAL, 5 M/S WIND SPEED U-5 = UN5 TABLE,5 M/S WIND SPEED l 1 m h103 e-5 Z ~ l e s i y 1.i g g 1-5 O i 2 l M 30 / r 3 Z l Z g u.5 ,. i ~ 15 g'l ,$= 5'O ~ l D s-us oy // m // w* ,,1 I I I l i ll i I I IIII I I Ill! // I I ! i ll I I I I!ll

_. w onenova J g

'g, oac car on e -er J L o as ra 2 3 g7 TIME AFTER ACCIDENT - SECONDS CONTINUOUS OCCUPANCY INTEGRATED DO!E FROM GROUND DEPOSITION AT 1/2 MILE DOWNWIND M

\\/ Seedon13 . O, Page 37 I'igu re 13.12 l l l l llll l l l IIll l l l1llll l l l I lll l l jl I l l [_I ! m MIS O O E O E IO I TOTAL 100% CORE kELT EI AMPLE d IODINE 831 O;s= _/ ~~~ ,g w N'3 O i,0, / / N s 10% CORE MELT EXAkPLE ' @0 / \\ \\ ~ \\ fIODINE 138 \\ / TOTAL 4 / / Oi ll/ '~ 'Q \\ ~ ~ / / $\\ \\ h 10 0 As I f. / si g \\\\ 0 ? l ~ \\\\ l 1 l l l \\ t O lO'I \\ I 1 5 h k5 \\l 1 _- Q l' \\ t O \\, ,e.m ,, i,,,, ,,.,i,,,, ,,i,,,, ,, i,, ;,, 4 O 'o* 0 * ""* io4 "' ' '"i os ~ ' " ~ 3 io*

• '""'"iv TIME AFTER ACCIDENT. 5ECONOS CLOUND DEPOSITION AT 1/2 MILE DISTANCE, UNSTABLE, UNIDIRECTIONAL WIND 5 PEED 0F 5 METER $/SECOND g

[ 0 k h

O r o ['h Section 13 Page 39 Figure 13.14

10. l i

l l l# l l lIlli i l l l lll i i l l 1111 i i l ltill O 1 SOLIO CURyt1 1001 Cott MELT (I AMPLE se0 KEN CultV(1 105 CCRf MELT EN AMPLE O l.1

• INVf R$lO'd, I W$ WINO $PE R O l.5 8 tNVf R$10N,3 M/5 WIND SPCEO N-l a NgytR AL, 3 M/$ WIND SPf f 0

~ U-5 a UN1T ASLE, $ W$ WIND $ pef D Oo 2 I. ~ ~ 1.s 3 / O 5 ~~ 5 ] i 11 i O p N.s O b f,.L \\_- s ) h. , y 3 / / I 1 v.s ) _~ h ( 1 \\ N*I \\t g ,y 0 \\1.

2.,

1 (% 4 ~_ j { t hl \\ il r 'l ) ~_ {% I 3 Z l i 1 J-I' I Illit I 'I l ' III I IllM i I i I I'll l' ' ! ' t ti f f j -, 't 3 "'"*" ^#"105 ONE WEEKd 1 10 104 10G 'cNr uonrHg7 Time APTen ACC30ENT. $tCON01 IODfME DEP058 TION RATE IN STANDARD THYRol0 FROM INHALATION ON LEAKACE PLUME j CENTERLINE AT 1/2 MILE DISTANCE s

• 1

~-

bo o

• Q Section 13 Page 40 Figure 13.15 10-2 I

I l I I ll l l l $ lll I l I lllI I l l l l ll I jl IIII $0L10 CURVES

• 100% CORE WELT ER AuPLE

~ '.ROK EN CURVES

• IL% CORE WELT EN AMPLE 1 1
• INVER$10N. I M/S WINO SPEED O

I 5 a INVE R510N. 5 M/$ WIND SPEE D N.S a NEUTR AL 5 W1 WINO SPEED ~ ~ U-5 a UNST ABLE,5 W$ WIND SPEED 11 g. io-a A =_ = O 8., O R

• Og'A 5

s N., 5 s, 01 m -s / A\\\\ . io E['N \\M i O! I Os t g / U., ',*s ' I,,' q 05 s '.e j \\ l /' 3 Q ~ I \\ ) \\.i a 'y v i Q go-7 \\ I l q t f I a t I iliin i i i liiii i., i l ie u l i ii, e i i i l iiii i i.i ! o

.r i

i i o ~< ove> e o~<eA "ios e~< e<<> e e~< o~'"ior ioa se io flut AFTER ACCIDENT + SECOND$ O DEPOSITICH RATE IN STANDARD LUNCS OF IH50LUBLE RETAINED FRACTION FROM INHALATION i ON LFtKAGE PLL'ME CENTERLINE AT t/2 MILE DISTANCE j 1 i O ~}}