ML19345E394

Forwards Request for Change to Tech Specs of License DPR-6. Proposed Change 8 Will Permit Licensee to Refuel Reactor w/zircaloy-clad Reload Fuel During Next Refueling Outage

ML19345E394
Person / Time
Site:	Big Rock Point File:Consumers Energy icon.png
Issue date:	12/23/1965
From:	Haueter R, Wall H CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:	Boyd R, Doan R US ATOMIC ENERGY COMMISSION (AEC)
References
NUDOCS 8101160255
Download: ML19345E394 (44)

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00llSilliiGIS D(G)!. C0!ilpDHy Funci-i c.c...

e ufos m Wcr t Michsgan Avenue.,Mmon, Mechigan WOi. Area CoJe SG NeOwO December 23, 1965 Dr. R. L. Doan, Director Re:

Docke United States Atomic Energy M,)

Division of Reactor Licensing f

Commission Washington, D. C.

20545

Dear Dr. Doan:

Att:

Mr. Roger S. Boyd Transmitted herewith are three (3) executed and nine-teen (19) conformed copies of a request for a change to the Technical Specifications of License DPR-6, Docket No. 50-155, issued to Consumers Power Company on May 1, 1964, for the Big Rock Point Nuclear Plant.

The proposed change (No. 8) vill enable Consumers Power Company to refuel the reactor at Big Rock Point with zirealoy-clad re-load fuel during the next refueling outage.

Since this outage is sched-uled for March 1966, we would appreciate your early consideration of this request.

Yours very truly, g

RLH/wf/d b Robert L. Raueter Attach.

Assistant Electric Production Superintendent W

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i CONSUMERS POWER COMPANY Docket No. 50-155 gle C07!/

Request for Change to the Technical Specifications License No. DPR-6 For the reasons hereinafter set forth, it is requested that the Technical Specifications of License DPR-6 issued to Consumers

. Power Company on May 1, 1964, for the Big Rock Point Nuclear Plant, be changed as follows:

I.

Section 5 A.

In Section 5 1 5, delete the entire section and replace it with the following:

"5 1 5 General Core Composition The data in this section present general design features of the original fuel, research and development fuel and reload fuel that shall make up the physical composition of the core.

(a) Enrichment of Fuel, approximate weight percent U-235 from 2.6 to 4 5, inclusive.

(b) General Core Data:

Number of Fuel Bundles in Core, A Maximum of 86 Total Nominal Weight UO in 8h Bundles, Lb 29,300 2

Moderator-to-Fuel Volume Ratio 2.65 Equivalent Core Diameter (Approximate), Inches 77 (c) Fuel Bundles:

The general dimensions and configuration of the three types of fuel bundles shall be as shovn'in Figures 5 2, 5 3 and 8.1 of these specifice' tons.

Principal design features shall be essentially as follows-g '~

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Fuel Bundles Original Reload Research.& Development General Gecmetry, Fuel Rod Array 12 x 12 11 x 11 11 x 11 Rod Pitch, Inches 0 533 0 577 0 580 Standard Fuel Rods per Bundle 132 109 109 Special Fuel Rods per Bundle 12*

12 12 Spacers per Bundle 3

5 7

Fuel Rod Cladding Material 304 SS Zr-2 304 SS, Zr-2 Inconel 600 and/orIncoloy800 Standard Red Tube Wall, Inches 0.019 0.03h 0.010 to 0.030, Inclusive Special Rod Tube Wall, Inches 0.031 0.031 0.010 to 0.030, In clusive Fuel Rods Standard Rod Diameter, Inches 0 388 0.hh9 0.425 Special Rod Diameter, Inches 0 350 0 3h4 0 320 UO Density, Percent Theoretical 94

• 1 94

• 1 90 to 95, Inclusive 2

Active Fuel Length, Inches Standard TO 70 68 to 70, Inclusive Corner 59 Fill Gas Helium Helium Helium -

• (h Special Fuel Rods at Bundle Corners Are Segmented)

(d) Channels:

Number of 30h SS and/or Zr-2 88 Wall Thickness, Inches:

30h SS 0.075 Zr-2 0.100 Inside Width, Inches:

30h SS 6.57 Zr-2 6.5h

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Length, Inches:

304 SS' 79-5/8 Zr-2 79-3/4

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.(e) Total Weight' Supported by Core Support Plate:

86 Fuel: Bundles @ 440 Lb/ Bundles, Lb 37,840

'88 Support-Tube-and-Channel Assemblies-

@ 100 Lb/ Assembly, Lb

-8,800' 86 Orifices @.10 Lb/ Orifice, Lb 860

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2ChannelPlugs@10Lb/ Plug,Lb-20

.1 Flow Distributor Assembly, Lb 2,500 Total Wei6ht, Lb 50,020 B.

Add a New Section - 5 1 7:

"5 1 7 Zr-Cr Alloy Test Bundle One of the reload' fuel bundles may contain up to 18 rods (2 dummy and 16 fuel rods) clad with an annealed Zr + 1.15

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'v/o Cr alloy and up to 8 rods (2 dummy and 6 fuel rods) clad with annealed special Zr-2.

The remaining fuel rods j

in the bundle (maximum of 95) shall be clad with cold-worked standard Zr-2."

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Section 7-In'Section 7 5 7, delete the entire sectio hndreplace -

it with'the following:

"7 5 7 It'shall be permissible.to remove a control' rod' drive from the reactor vessel when the reactor is in the shut-

-down' condition and the mode selector switch is locked in the " Shutdown" position.

The core shutdown margin of 0 3%a6tk,ff/keff vith.the strongest control rod out of the core shall have been met prior to the control rod drive removal; and in addition, the equipment shall.be properly tagged. The control rod drive that was removed shall without delay be replaced by a spare control rod drive or the original control-rod drive shall be reinstalled. One control rod drive pwnp shall be operating during removal and reinsertion and during the 4

time the control rod drive is outside the reactor vessel."

III. Discussion - Reload Fuel The above proposed change to Section 5 1 5 will etnole Consumers Power Company to-refuel the reactor at Big-Rock Point with-zircaloy-clad reload fuel.

The objective of the proposed change is to take advantage of the superior performance characteristics of Zr-2 fuel cladding.

Insertion of the zircaloy-clad fuel into the reactor will be on a batch basis over the next several years.

It is expected that-be-tween 10 and 20 reload bundles.will be loaded into the reactor during the

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next refueling (currently scheduled for March 1, 1965).

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The proposed addition of Section 5 1 7 vill permit insertion l

of a special Zr-Cr cl&d developmental fuel bundle into the reactor.

T1.9 use of the special Zr-Cr alloy cladding for one of two reload bundles (the other of which employed cold-worked standard Zr-2 cladding) was approved previously by the Commission (Change No. 5, dated March 12,1965).

Au-thority for use of these two developmental fuel bundles was discontinued by Change No. 7, dated July 9,1965 4

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Design Features of Reload Fuel Theprinc!p,Ldesignfeaturesofthereloadfuel(thbulated below) are verg similar;to the zircaloy-2 clad research and' development fuel now operating in the reactor. Each bundle utilizes'two-enrichments as well as reduced-size corner rods to minimize local power peaking. The higher enrichment fuel

. rods are located in the center of the bundle. Fuel r'ods are held in position by five wire and spring type ' spacers located

-along the length of the bundle. These spacers serve to mini-mize deflection and vibration of the fuel rods.

The above de-

. sign details are shown in Figure'5 3 Fuel Outside Rods Inside Rods Large Small

. Rod OD, Inches 0.449 0.449 0 344 Pellet.0D, Inches 0 373 0 373 0.275

• Zr-2 Cladding-Thickness, Inches c.034 0.034 0.031.

UO Gr und UO Gr und UO Ground Fuel Material 2

2 2

. Pellets Pellets Pellets Nominal Cladding-UO Gap, Inches 0.008 0.008-0.007 2

.INiel Enrichment, Percent h.2 2.6 2.6 UO Density, Percent 94.

94 94 2

j Numbe icg5cnts per Rod 1

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. Active

.el Iength, Inches 70 70 70 f

Weight of UO per Rod, Lb 2.846 2.8ho 1 547 2

Water-to-Fuel Ratio 2.68 2.68 2.68

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Number of Rods per Bundle 37 72 12 Weight of UO Per Bundle, Lb 328.8 2

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Principal Calculated Nuclear Characteristics of Reload Fuel For reference purposes, the nuclear characteristics of the reload fuel bundles are presented below:

(a) Reactivity (koo)

Temperature koo 68 F, Zr Channels 1.275 572 F, Zr Channels 1 303 572 F, Zr Channels +

20% Void 1.296 (b) Moderator Temperature Coefficient (A k

/k Per F) eff eff

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-3 End of Cycle

-0.14 x 10 (c) Void Coefficient (Ak

/k per Unit Void Within the Channel) eff eff 68 F 572 F

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-0.09 (d) Doppler Coefficient The Doppler Coefficient is dependent on moderator-to-fuel volume ratio. Since the reload fuel and original fuel have essentially the same moderator-to-fuel volume ratio, the Doppler Coefficient for the reload fuel vill be the same.

IV.

Hazards Considerations A.

General Considerations The reload bundles described above and shown in Figure 5 3 vill utilize the same hardware as the present Phase I and Phase II develop-mental bundles, except for the spacers. Eight Phase I bundles utilizing this hardware have been in the reactor since April 1963, and have exposures ofapproximatelyh000Mvd/T. Fifteen Phase II bundles utilizing this same hardware have been in the reactor since May 1964, and have exposures of approximately 2800 Mwd /T. Experience with this hardware has been excellent.

The decision to utilize it for planned reload bundles is a result of this excellent performance.

The spacers shown in Figure 5 3 are of the wire and spring type design currently being offered by General Electric for their commercial fuel. The springs provide the necessary lateral pressure to minimize fretting wear of the zircaloy cladding.

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ite dual enrici. ment utilized in these bundles is ex-pected to pro ide a significant improvement in core peaking factors at a slight enrichment penalty. The effect upon other nuelsar and' thermal-hydraulic factors is not significant.

Based on the above' considerations, and a thorough review of the nuclear and thermal-hydraulic aspects.of the use of zirealoy-clad reload fuel, we have concluded that its use presents no safety problems related to normal operation of the reactol The safety considerations t

i related to the consequences of using zirealoy cladding under off-normal conditions are discussed in the following cactions:

B.

Accident Analysis The use of a 100% zircaloy-clad core in the reactor at Big Rock Point modifies a portion of the analysis and results shown in the Final Hazards Summary Report (FHSR), dated November 14, 1961. An analysis has been made of the zircaloy core heatup during the loss-of-l l

coolant accident and the effects of the subsequent pressure rise in the containment. The entire analysis has been made using a series of digital computer programs developed by General Electric Company.

1.

Metal-Water Reaction i

The loss-of-coolant accident as described in the FHSR has been re-analyzed with the addition of an assumed zirconium-water reaction. All other asswmptions were reviewed and updated based upon current knowledge. The details of the calculations are included herein as Appendices I, II and III.

In the calculation of the zirconium-vater reaction, the reactor core was subdivided into radial and axial nodes.

The fuel bundles were further divided into zones of fuel rods.

This fuel-bundle subdivision allowed the analysis to be conducted with a refine? specification of power distribution throughout the reactor core, radially and axially, as well as throughout the fuel bundles and fuel rods. With the same degree of subdivision, the temperature distribution throughout the reactor core could be determined precisely throughout the course of the temperature transient.- Although radiation heat transfer from fuel rod to fuel rod and to the channel was accounted for in the reogram, no' credit tas taken for the radiative heat loss from the overall

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reactor core itself. The convective heat transfer from the fuel rods to any coolant passing through the core vss controlled according to the case being au 'Yzed.

The heat trans'ler coefficient used in the code during the blowdown phase vr.s arrived at from correlations determined experimentally by General Electric Company.

The tests indicated that a high boiling coefficient exists from J.o secondr, after initiation until h seconds after initiation (dependent on the fuel power) and then diminishes to 0 in about 12 seconds. The fuel channel material was analyzed seIerately from the fuel cladding and fuel pellet material.

Because of the emphasis placed on the existence of any metal-water reaction, the digital computer program a?vo involved a continuous calcula-tion of the extent and current rate of the metal-mter reaction for all zircaloy surfaces within the reactor core on the above-described subdivision basis. The metal-water reaction is de-fined in the computer program by the expressions of " Baker" and "Just" which define the rate of reaction as a function of both metal temperature and the current extent of the reaction.

In the calculation, the reaction rate was not limited by any deficiency of water for reaction.

The more important results are shown graphically on Figure 1 for the case where the core spray system functions as designed.

This accident was assumed to consist of the instan-taneous severance of a main coolant recirculation line vitt the subsequent loss of coolant from the reactor vessel. It was assumed that the core spray came on in 15 seconds (see also Figure 3) and functioned as designed. Figure 1 shows the ex-tent of the metal-water reaction with time. The cladding temper-ature carves show that none of the fuel cladding reaches the mel1.ing point of 3350 F.

Figure 1 indicates that most of the fuel cladding would still be above about 740 F after 40 minutes (2400 seconds).

Core spray experiments, conducted by General Baker, Louis & Just, Louis C., " Studies cf Metal-Watar Reactions at High Temperature, III.

Experimental and Theoretical Studies of the Zirconium Water Reaction," MTL-6548, May 1962.

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thus ending the-temperature excursion.

The heat transfer.coef-ficient applied to this case was taken from core spray experimental correlations (on full-scale bur.dles) determined-at the General Electric test facilities..

The results showed that clad melting did not occur on any o'f the fuel ~ rods; and that of 'all the zirconium in the s'ystem, j

only 0.6% reacted _vith water. -This reaction produced only'3 3~

-pounds of hydrogen and 207,000 Btu of energy.

The response of ther containment would be unaffected by the metal-water reaction. There-fore, the conclusions as presented in Section 13 1.6 of the FHSR remain unchanged for the case in which core spray is effective.

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That is, the radiological. effects of this accident are negligible.

If it were assumed that no core cooling followed the loss-of-coolant _ accident, the core heatup results would change (these are shown graphically on Figure 2).

This. accident is the same as the accident described above except that there is no core cooling and no' credit for convective heat transfer is taken after blowdown. The effect of the chemical' energy addition to the core from the metal-water ceaction was to accelerate the rise in clad temperature. The total reaction occurring (see Figure 2) in the ccre was 22 9% of the zirconium in the system. To establish the maximun extent of reaction, it was ascumed that the cladding,which l

melts before reacting in the core, drips into the water remaining'at I

the bottom of _the reactor vessel.

Some water is expected to be 4

present at the bottom of the vessel from the control rod drive cooling water, the core spray, or the feed-water system. The smallest potential drop sizes of molten metal which might occur from the fuel bundles have been determined by using a surface tension model and fuel end-plate dimensions.

(This has been shown experimentally to give very good results. Molten zircaloy drops on the order of 0 350" to 0.400" were predicted from analytical considerations. The drop sizes obtained in the tests ranged from a minimum of 0.135" to a maximum of 0.477". ) Dnploying the model

Figuro 2 BIG ROCK POINT CORE HEAT-UP No Core Cooling Case In-Core Metal Water Reaction %

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. balance giving' a' total of 26.8% reaction for the case of no core spray.

2.

Containment Pressure Response The. pressure' response of the containment in the realistic

. case where the. core spray functions as designed is unchanged by the introduction of the 100% zirealoy-clad core.

'In the case of no core spray, it has been shown that a maximum metal-water reaction of 26.8% may occur. The additional mass of gas.and chemical energy added to the containment will change slightly the pressure response of the containment -from that -shown in-the FHSR. The energy and gas released from the metal-water reaction were assumed to be liberated to the containment system in a uniform

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manner over the first 900 seconds after the-initiation of the accident.

1 Figure 2 shows that the assumed 900 second period is less than the core heat up calculation would indicate and is therefore conservative.

The modified pressure. response is shown on Figure 3 along with the pressure response for no metal-water reaction. The chemical energy released is responsible for the majority of.the pressure rise between hO seconds and 900 oeconds, the partial pr;s: _ze of the added hydro-gen gas being only 0.6 psi at 900 seconds.

The peak pressure cal-culated was 35.6 psia. This is well below the h2 psia design pressure of the containment system.

. 3 Hydrogen Flammability Considerations The introduction of hydrogen into the air in the containment raises the possibility of a reaction of the hydrogen with the oxygen in the air. For the no core spray case, the maximum extent of the metal-water reaction of 26.3% results in the production of 72.h pound-moles of hydrogen. The containment confines 2142 pound-moles of air.

Therefore, the upper limit mole fraction of hydrogen (considering-the mixture to be dry) is 3 3%.

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. Therefore,-it is concluded that a' burnable mixture would not be 8

produced -in the containment even for the no core spray case.

. 4.' Radiological Effects The radiological effects which result from th' postulated accident were comprehensively reviewed. It was foundf that the difference between the radiological effects reported in the FHSR and those calculated for the zircaloy-clad core case were different by only 1%.

This small difference is due basically to the small increase in the area under the containment pressure vs time curve, which slightly enhanced the leakage from the con -

tainment. It can be seen on Figure 3 that this small increase in enclosure pressure operates only ove-a short time interval, from about 100 seconds (1.7 minutes) to 1000 seconds (16.7 minutes),

which is relatively insignificant even for the two-hour dose calculation.

C.

Conclusions The calculations and considerations discussed above show that the addition of an assumed zirconium-water reaction to the loss-of-coolant accident results in only a very small-increase in the leakage of radioactive fission-products from the containment sphere.

This increase is considered to be insignificant, and we have concluded that the requested change permitting the use of zircaloy-clad fuel does not present significant hazards considerations not described or tmplicit in the Final Hazards Sum-s mary Report.

V.

Discussion - Control Rod Drive Maintenance The proposed change to Section-7 5 7 is intended to clarify the procedures used when work is done on control rod drives while the reactor vessel contains fuel.

4 Control rod drives may be removed from the reactor only when j

the nuclear steam supply system water temperature is below 125 F and the system is depressurized. The removal. procedure basically is as follows:

Riggs, C.- 0. & Watcher, John, " Hydrogen Release Hazards in the PM-3A Containment. Vessels," MND-M3A-3108 - Revised January 28, 1964

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12 A control rod drive pump is placed in operation and the control rod drive to be removed is withdrawn to the fully withdrawn *posi-tion (Notch 23). The control rod is then unlatched from the control rod drive index tube and a withdrawal signal is applied to the drive, moving the index tube to its overtravel position. At this point, the insert and withdraw isolation valves on the drive to be removed are closed and tagged. This information is immediately reported to the control room where the reactor mode selector switch is placed in the'

" Shutdown" position. It should be emphasized that, up to this point j.

In time, any.or all of the control rod drives can be scrammed, in-cluding the drive to be removed - up to the point that it is valved out.

r With the mode selector switch in the " Shutdown" posi-

' tion, each control rod drive, except the one being worked on, is in the scram condition. The control rod drive pump maintains a minimum I

pressure (about 500 psig) on the insert piston, which guarantees that-all drives' vill remain in the full in position.

It should also be neted that the poison injection system is kept operative at all times during the control rod drive work. Standard operating procedures call for use of the liquid poison system whenever suberiticality cannot be assured by the control rods.

CONSUhERS POWER CO SANY By Vice President Date: Dacember 23, 1965 Sworn and s

=cribed to bef ore me this 23rd day of December 1965

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W Notary Public, Jackson County, Michigan My commission expires February 16, 1968

APPENDIX I

CALCUIATION OF THE ZIRCONIUM-WATER REACTION The detailed calculations for the 100% core melt case are:

A.

Mass of Zirconium In Core The total mass of zirconium in the core was determined from the. values given in the Final Hazards Summary Report (FHSR). Those values are as follows:

Mass of Zr in Fuel Cladding - 96.7 Lb/ Bundle Mass of Zr in Channels

- 50.0 Lb/ Bundle For a full core loading containing 84 zircaloy-clad fuel bundles and zircaloy channels, the total mass would be:

(96.7 + 50) 84 = 12,320 Ub B.

Core Heatup and Metal-Water Reaction The TACT-5 computer heatup calculation for uncooled core gave a reaction of 2822 lb of zirconium in the core. The analysis of po-tential drop sizes of molten zirconium indicates an additional 5% reaction of the unreacted zirconium as it melts from the core for a total reaction c" 26.8% or 3297 lb of zirconium.

Figure ' indicates that the core melts down in over 1000 seconds. To be conservative, it is assumed that the core melts down in 900 seconds, generating all the hydrogen and chemical energy uniformly during this period. The assumpuion of uniform hydrogen and energy release is convenient and materially does not affect the result since the result is not strongly "9+.e dependent.

C.

Hydrogen Generation and Energy Release Calculation The results reported in Ref 2, Figure 131, can be modified simply by adding the hydrogen generated to the inventory of the containment and adding the chemical energy to the system. Note that the entire system is assumed to be at the same temIerature.

The added hydrogen is:

Pounds of Hydrogen = 3237 tb Zr x 4 Lb H/91.2 Lb Zr

= 144.8 Ub H2 i

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2 A P P'E N-D I X I_(Contd)'

The added chemical enercy is:

Chemical Energy = (2800 Btu /Lb Zr) 3297 Lb Zr

= 9 23 x lo Btu The mass of water in_the containment (with core spray on) at approximately 900 seconds would be:

MT = 103,000 + 37,500 = 140,500 Lb 3

The containment free volume is 940,000 ft,

Therefore, the specific volume of the containment.is:

3 v = 940,000/140,500 = -6.7 Ft,

Ref 3 shows that the specific internal energy at 210 F is:

u=h50 Btu /Lb The incremental specific internal energy caused by the chemical energy is:

6 u = 9 23 x lo /140,500 = 66 Btu /Lb q

Thus, Efter the' chemical energy has been released, the total specific internal energy is:

u = 516 Btu /Lb Note that during the latter portion of the 900 second melt 4

}

down process, Figure 13 1 of Ref 2 indicates that the specific internal

]

energy of the system is constant except for chemical energy; i.e.,' the tem-f-

perature is constant at 210 F.

Ref 3 shows that with the addition of this chemical energy, the t

j pressure vill rise to:

P = 35 Psia and the containment temperature. to:

T = 239 F The partial pressure of the added hydrogen vill be:

p

_ R*(T + 460) x K,

_(1545) (239 + h60) x 144.8 2

• H xV "2

~

2 (144) x 940,000 H

2 i

i i

~

d 1

3 APPEN' DIX I(Contd)

P

=.-v.58 Psi

~

-I

.k Therefore, the total pressure after all-the hydrogen and chemical energy has been released will be:

P ' = 35 + o.58 = 35.6 Psia T

at

_t = 900 seconds Since the release of energy and hydrogen is assumed unilorm, a reasonable representation of the pressure can be made by adding the prec-sure increment uniformly over the first 900 seconds.

D.

Hydrogen Mole Fraction and Flammability Considerations With the original temperature of air in the containment at:

' loo F, and 100% relative humidity, the partial pres::ure of the water vapor -

would be:

P = -o.949 Psi Partial Pressure'of the Water Vapor

=

The mass of air in the containment would therefore be:

6

=~ E = 13.697 x o.94 x 10 x 14h 62'1h4 Pounds M

=

Air RT 53 3 (460 + loo) ass 6

2142 Pound Moles Air Moles of Air

=

=

3 3%

Hydrogen Mole Fraction

=

=

214 2.4 22 4

Therefore, the resulting mole fraction is less than the 7

flammability mole fraction of approximately 5%. This value is conserva-tively evaluated neglecting the water vapor which is present and would suppress flammability.

1 i

l

~

l c

APPENDIX II CALCULATION OF THE FISSION PRODUCT RELEASE FROM FUEL AND REACTOR i

In reexamining the accident analysis in the FHSR to investi -

gate the consequences of inserting the zircaloy-clad fuel in the core (at Big Rock Point), it was decided to update the assumptions made in the original accident analysis in order to take advantage of all the developments in haz--

ards analysis made over the past few years.

An attempt is made in this and

'the following Appendix (III).to update not only the accident analysis but also to relate the new calculation results to previous results.

'It was assumed, conservatively, that 100% of the fuel is heated to 3000 F within 10 minutes after the loss of water from the re-4

-actor.

Experimental data (Ref 6) indicate that the fission gas release i

from UO fuel r ds is relatively small except forthese portions of the fuel 2

above the UO r rystallization temperature of about 3000 F.

While some 2

. earlier release would occur between the cladding failure and recrystalliza-tion temperatures, this release would be small in comparison to the-assumed release at 3000 F.

The following fractions of the fission products are assumed to be released from the overheated portion of the fuel when it f--

reaches 3000 F.

Percent Noble Gases (Kr, Xe) 100 Halogens (Br, I) 50 Volatile Solids (Te, Se, Cs, Ru) 50 Other Solids 1

The fission product inventory is based on a reactor having been operated at 230 Mwt for 1000 days before the accident. This assumption is different from the one made in the FHSR. It is, however, more conserva-tiv9 and is used here in line with current GE practices in analyzing fission product releases from fuel.

Since organic halogens do not fall out as rapidly as inorganic i

halogens, a conservatively large fraction of halogens was assumed to be organic. Fuel melting experiments (Ref 7, 8) have shown that 0.1% to 3%

.of the released halogens are of the organic form. For the loss-of-coolant

~,

--. ~_-

t c.

2^

i

APPENDIX II (Contd) analysis,- 10% of the halogens released from the' fuel are assumed to'be organic form. This' assumption is thus conservative by a' factor of 3 to 100.

The fallout and plateout of fission products.within the're-actor vessel and piping reduce the amount of fission products available for transport.to containment.

All organic halogens are assumed to escape both:

fallout-and pla'teout. Of the remaining 90% (which are inorganic),.50% plate out on metal surfaces. The fallout and plateout'in the reactor vessel.and piping are:

Percent Noble Gases 0

Halogens: Organic 0

Inorganic-50 T

Volatile Solids 70 Other' Solids TO A.

Fallout in the Containment The fission products will fall out and plate out in the containment. All organic halogens are assumed to remain airborne. An effective-half-life for fallout and plateout of inorganic halogens of 30

~

minutes is estimated until 10% of them remain airborne throughout.the ac-

]

cident (Ref 9, 10, 11).

An effective half-life for fallout of volatile and other solids of 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> is estimated. This is based on fallout coefficients for solid i

fission products (simulated by Y 0 reported in MSA-TR-58, " Simulation of 23 Major Pipe Rupture in a Pressurized Water Type Plant").

B.

Containment Leakage The containment pressure is reduced by the containment cooling system. The turbulent (rough passage) equation (Ref 12) is used to calculate 2

leakage rate, with O.5%/ day at 27 psig as a base.

t s

3 APP'ENDIX II (Contd)

The calculated leakage. rate of fission products from the containmentis:

TABLE A FISSION PRODUCT LEAKAGE RATE FROM 'CoNTAINhENT l

(Curies /Second)

-Time Noble Gases

-Halogens Volatile Solids other Solids.

10 Min

- 2.2 0.88 1.6

.o.12 30 Min 17 o.48 o.69 0.088-1 Hr -

13 0.19 0 37

.0.062-3 Hr-0 75 0.095 0.14 0.030

-3 l.

-10 Hr-0 34 0.034 0.021 5.4 x 10

-3 1 Day-0.41 0.038 39x10 9.8 x lo-

-6 7

3 Days.

o.23 0.021 29x10 8 x 10 lo Days o

o o

o 5

j a

..g.

y

?

X APPENDIX III RADI0IDGICAL EFFECTS OF 100% CORE MELT DUE TO ' IDS _S OF-C00IANT ACCIDENT FOR BIG ROCK POINT A.

Calculation Results The radiological effects which resulted from this postulated accident vere comprehensively.revfewed. They are summarized in Thble B and Thble C for two different methods of calculation. The methods yielding Thble B are those recommended by General Electric Company although both methods yield results which satisfy'the required criteria. Sections B, I

.C and D present further discussion of the methods and derivation of these

~

.results.

.A direct comparison of current results with those results published in th FHSR is only approximate because of the evolution of analytical methods and the complexity of relationships.

Tuble D summa-rizes this comparison for two cases, thyroid dose for tvo hours, and for i

the total a cident. Three comparisons are made for each of these two cases, (I) the 100% Zr-H O accident by the methods of Table B, (II) the 2

100% Zr-H O accident by the methods of Table C, and (III) the accident 2

with no Zr-H 0 reaction by the methods of Thble C.

2 Comparison of II and III shows the radiological effect of the Zr-H O reaction to.be insignificant (calculated to be less than 2

1% different from the no Zr-H O reaction case).

The total curies released 2

are now calculated to be substantially larger than the total presented in the FHSR. The two main reasons are:

1.

Assuming fuel exposure at full power for 1,000 days rather than assuming an average fuel exposure of 10,000 Mvd/T, as in thi FHSR. This affects the two-hour release due to the different fission product inventory in the fuel. This effect is much less for the total accident where only a slight dif-ference results due to saturation effects.

1 1

APP'ENDIX III (Contd) 2 TABLE B (1 RADI0IOGICAL EFFECTS BIG ROCK POINT 100% CORE MELT Dict nce-

' First Two Hours Total Accident (Miles)

VS-2 MS-2 N-2' N-10 U-2 U-10 VS-2 MS-2 N-2 N-10 U-2

.U-10 Passing Cloud Whole Body Dose (rad) 1/2 0.09 0.07-0.04 ( 0.01

,0.01 ( 0.01 0.4 0.2 0.1 0.03-0.09 0.01-1 0.07 0.04 0.02 <0.01 <0.01 <0.01 03 0.1 0.04 0.01 0.03 <0.01 3.

0.03 0.02 <0.01 <0.01 <0.01 <0.01 0.2 0.09 0.02 <0.01 <0.01 <0.01 5

-(2)

- <0.01

- <0. 01 0.2 0.04 <0.01 <0.01 <0.01 <0.01

- <0. 01

- <0. 01 0.1 0.03 <0.01 <0.01 <0.01 <0.01

!10 Life Time Thyroid Dose (rem) 1/2 84 35 13 3-5

<1 280 1c0 47 8

19 3-1 32.

13 h-

<1 1

<1 133 49 15-3 7

1

'3 24

'3.

<1 -

<1

<1

<1 40 12 3

<1 1

<1

<1

<1 2h 8

1

-41

<1

<1 5

1 12 4

<1

<1

<1

<1 10-

<1 LifeTimeLungDose(rej 1/2 8

3 1

<1

<1

'l

-l 4

1

<1

<1

<1 1

3 1

<1

<1

<1

<1

<1 i

5

<1

<1

<1

'l

<1

<1 10

<1

<1

<1

1

'l (1

4 t? hole Body Fallo _ut Dose (roentgen) 1/2 3

1 1

<1

<1

<1 1

1

<1

<1

'l K1 (1

3

<1

<1

-C1

'l

<1

<1 5

<1 el

'El

<1

<1

<1 10

<1

<l

-'1

<1

<1

<1 Using Methods in Journal of Applied Meteorology Two Hour Dose Zero Since Time of Cloud Travel Is Greater Than Two Hours w

~

^-A P'P E N D I X III (Contd) 3 TABLE C RADIOLOGICAL' EFFECTS

' BIG ROCK' POINT 1004 CORE MELT

. Dict:nce First Two Hours Tetal Accident

.(Miles)

VS-2

.MS N-2 N-lo U-2 U-lo VS-2 MS-2 N-2 N-lo U-2

'U-lo Paesing Cloud Whole' Body Dose (rad) 1/2 0.09 :0;o7 0.04 <Co.01 0.ol <Co.ol 1

0.6 03 0.1 0 '3 0.03 1

0.07-0.oh o.oe <[o.ol <Co.ol <Co.01 09 o.4-o.1 0.03 0.1 <o.01 3

<Co.ol <o.ol <Co.ol<Co.ol <Co.ol <o.01 0.6 03 o.06<Co.01 0.03 <o.01 5

-(2)

- <C o. ol -

- < o.01 05 0.1 0.03 <o.ol <o ol <Co.01

- <Co.01

- <Co.01 03 0.09 <Co.ol <Co.ol <o.ol <o.01

- 'lo Life Time Thyroid Dose (rem) l1/2 160 60 21 4

5 1

1210 450 160' 35 39 11-1 45 28 5

3

<1

<C1 385 210 41 16 5

3 3

11 5

<;l

<1

<C1

<C1 87 41-7 1

1

<1

<C1'

<1 35 20

'3

<1

<1

<1 5-

<C1

<C1 8

7

<1

<C1

<1

<C1 lo

-Life Time Lung Dose (rem)

.1/2' 13 3

2 1-l' 1

1 7

3 1

1; 1

1 3

2 1

1 1

1 1

5 1

1 1

1 1

1 lo 1

1 1

1 1

1 Whole Body Fallout Dose (roentgen) 1/2 11 h

3 3

1 2

1 6

3

<1

<[1

<1

<[1 3'

2

<C1

<C1

<C1

<C1

<C1 5

. <C1

<C1

<C 1

<C1

<C1

<C1

'10

<C1

<C1

- <C1

<[1

<1

<Cl 1) 2)Uning Methods in HW-SA-2809-Two Hour Dose Zero Since Time of Cloud Travel Is Greater Than Two Hours l

_y.

?.

4' 4

APPENDIX III(Contd)

TABLE D COMPARISON OF RADIOLOGICAL EFFECTS 100% CORE MELT Normalized to FHSR' Values

• Two Hour

• fotal' Accident-FHSR I

II III

'FHSR I

II III Curies Released-Exposure Assumptions 1.0 1 36 ~1 36 1 36 1.0 1.06 1.06 1.06 4

Airborne Fission Products 1.0 1 35-1 35 1 35 1.0 8 75 -8 75 8.75 Zr-H O Effects 1.0 1.01 1.01 1.00 1.0 1.01 1.01 1.00 2

Breathing Rate 1.0

.46

.46-

.46 1.0

.46

.46-

.46 Thyroid Uptake 11.0 -1 53 1 53 1 53 1.0 1 53-1 53 1 53 Air Concentration Effect 1.0 1 30 2 30 2 30 1.0 1 30 -2 30 2 30 Wind Diversity Effect-1.0 1 50 1 50 1 50 1.0 50 1 50.1'50

Dose per Unit Exposure 1.0 1.28 1.28 1.28 1.0 1.28 1.28 1.28 Roundoff and Uncertainty 1.0 1.29 1 39 1 39 1.0 1.27 1.04 1.04

'l

--Total Result 1.0 4.2 8.0 79 1.0 70 30.2 30.0 4

Comparative 1/2 Mile Thyroid I

Dose - ( rem) 20 84 160 158 40 280 '1210 1200 I - Zr-H 0, Method of Table B 2

)

II - Zr-H 0, feethod of Table C 2

III - No Zr-H 0, Method of Table 'C 2

i

c.

y,

~

5

-APPENDIX III (Contd) 2.

The organic halogens are now assumed to remain airborne and, hence, are available to leak from the enclosure throughout the pressure transient..In the FHSR, the hr.logens were assumed to fall out'well before the pressure transient ter-minated. This, of course, is most dramatic for the total accident case and accounts for most of the difference between the FHSR and present methods.

B.. Analytical Techniques The sources of radiation considered in this accident' analysis were (a) tihe noble gases and their external whole body dose effect, (b) the halogens and'the resulting' thyroid dose from inhalation and (c) volatile solids-(cesium, rubidum, molybdenum, antimony, arsenic, selenium and tellerium) resulting in lung dose from inhal.ation.

Various meterological conditions were examined to give a spectrum of radiological effects during the poor diffusion conditions of inversion and the better diffusion conditions of lapse or unstable.- Six points in the meteorological spectrum were examined; these were (a) very stable' and moderately stable, each at a wind speed of two mph, (b) neutral conditions at wind speeds of 2 and 10 mph, and (c) unstable ' conditions at wind speeds ~ of 2 and 10 mph.

Wind direction persistence and variability of direction were considered in the radiological effects analysis. Persistence of direction was assumed for 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br />' duration. Various values of the diffusion para-meter (CTU), the product of standard deviation of wind direction fluctua-tion and average wind speed were assumed to exist for the entire 15-hour period of persistent direction. This information was used in a recently I

reported. diffusion calculational technique. A comparison between this type of diffusion calculation and a more common analysis method, where wind direction variability is not considered (vind persistence was constant for entire period of accident), was also made.

1.

General Radiological effects of the loss-of-ccolant accident (LOC $ core melt) were evaluated at distances from the plant of one-half mile

, _ - ~ _ _ -, _. -,.

6-APPENDIX III (Contd)

(plant site boundary), three miles, and five miles (center of nearest city - Charlevoix, Michigan, population 2751 per 1960 census).

It was assumed that airborne materials were released at ground level and diluted into the building vake, as described in the FHSR, Section 12.

C.

Meteorological Diffusion Eealuation Methods The radiological effect a of containment leakage at ground level should include the least favorable conditions to be encountered at the reactor site. These are the poor diffusion conditions caused by in-version (stable), at a vind speed of about two miles per hour, typical of varm weather nights, for both very stable and moderately stable conditions, and the better diffusion conditions, typical of daytime, represented by neutral and unstable (lapse) diffusion, both at vind speeds of 2 and 10 miles per hour. The atmosphere diffusion methods recently reported in the, journal of Applied Meteorology verc used.

1.

Diffusion Conditions In using the referenced diffusion methods, an important para-meter chosen was the product of wind speed and wind direction variability over the period of interest, given as & u.

Here cg was the star.dard deviation of the horizontal viui direction fluctuations and u was the average vind velocity.

Combined with the stability condition assumed, specification of mgu permitted calculation of air concentrations at various distances ficm the source.

A summary of data taken at several sites around the country shows that the value of mgu equal to 20 degree-mph (or 0.16 radian-meters /sec) is a conservative value for the parameter for a 1-hour period of time (observed about 1% of the time). This value was used to describe the horizontal spreading of the plume for the two mph wind speed cases.

Prediction of Environmental Exposures From Sources Near the Ground, Based on Hanford Experimental Data, J. J. Fuquay, C. L. Simpson and W. T. Hinds, Journal of Applied Meteorology, Volume 3, No. 6, December 1964

t 7

APPENDIX 'III (Contd)

A value of 130' degree-mph (or 1.0 radian-meters /sec) for.cg was U

j chosen to evaluate the effects during 10 mph conditions. This value of 'g u is about an average value for this parameter and would be an g

example of relatively. favorable diffusion canditions.

2.

Wind' Direction Persistence Inherent in the choice of the parameter y n ic the restriction that.

g vind direction.be unvarying within some specific direction increment.

(for example,-I 22-1/2, or 45 )..Since the values of CgiI chosen-j.

were for 1-hour periods of time, and since larger values of g u would g

be applicable to longer periods, in general, a choice of persistent vind direction (number of continuous hours) was made during which cgu values chosen could be expected to apply.

Wind persistence studies from site data (1} and by the U.S. Weather Bureau (see Table 7) give che frequency of unvarying vind direction i

for several sites including some coastal locations. The data are for

~

C a 22-1/2 direction increment (except for site data which is for 20 );

i.e., persistent direction within plus or minus 11-1/4. From the data it can be seen that the site is not greatly different from the others and that there is about a 99% - 99 9% probability that the wind will' not blow in any single direction (22-1/2 sector) for a period greater j

than 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br />. For the other sites, tabulations are based on the most 1

persistent directicn, all vind speeds included,and for all stability conditions, etc, combined. If the value of Cgu and the persistence i

of wind direction vere totally independent, assuming a value of 20 degree-mph for a 15-hour period is very conservative. Such independ-ence obviously does not exist since persistent vinds indicate relatively l

However, the persistence data consist of standard j

unvarying conditions.

weath(r bureau observations, with 5 - 10 minutes sampling per hour of a continuous vind direction recording. The average direction during this 5 - 10 minute period was recorded and the persistence of this aver-age direction vithin a variation of 22-1/2 was tabulated. It is quite Big Rock Point Meteorological Progress Report No. 2 l

i I:

,.,m

-... - = - - -

8 i

i i

APPENDIX III (Contd) apparent that a wide range of CTgor 7gu could exis~t and still yield d

tabulations indicating high persistence, such as the weather bureau summary. Thus, it is concluded that although some degree of depend-ence may exist between persistent winds and the diffusing parameter 7 g V, significant independence does occur.

It is estimated that a value of 7gu of 20 degree-mph has a probability of between 0.14 and 1

i O.01% for a 15-hour period. Assuming wind direction persistence for 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br />, a value of ggu of 20 degree-mph for a 2 mph wind' speed and poor diffusion conditions, is most conservative.

For a 10 mph wind, a value of CTgu of 130 degree-mph corresponds to 7g of 13 degrees which is quite similar to the~ value of 10 de-grees for the 2 mph case.

Thus, approximately, the same amount of wind variability is being considered and the 15-hour persistence assumption appeared equally applicable.

3 Application to Radiological Effects Calculation The diffusion and sind direction persistence conditions determine the method of application to a certain extent. Since 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> of

~

vind persistence was considered, and since the coolant loss accident' postulated leakage for a much longer period, the 15-hour time in-crement where the maximum integrated leakage occurs was the period 1.

of interest in determining total accident effects '(deses).

In the l

case of the 2-hour dose, the first 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of integrated leakage i

were used to calculate the dose assuming persistent vind during the entire period.

The 15-hour period of maximum leakage in the coolant loss analysis is the first 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> after onset of the accident. During this time, 30% of the noble gases, 3h% of the halogens, and 100% of the volatile and other solids are released. Thus, the quantity of curies released in the first 15-hour period was taken as the amount transported in one general direction and the dose therefrom was calculated.

The remraining fraction, in each case, was assumed to be spread around the site rather uniformly in different directions accompanied 3

4.

i 9

APPENDIX III (Contd) by the more frequently occurrin6 hL7hly variable vind patterns common to the Big Rock Point location. It was recognized that a portion of the remaining leakage, assumed to be videly spread, might diffuse in the same direction as that during which the 15-hour persistent vind occurred. However, reduced leakage, increased horizontal spreading and vertical spreading due to stability changes and greater direction variability during the remainder of the leakage period all combine to make any such added incremental transport of material (and dose) small compared to the doses calculated by the methods described, k.

Cloud Dispersion Calculations The above diffusion methods were used in calculating cloud disper-sion. In these methods, horizontal cloud growth, as expressed by the standard deviation of vidth & is given by At - AG + AQ e G

(1)y

=

y where 13 + 232 5 (g li)

(2) A

=

g

^

(3) and a = 2(crg )2 a

time after release and is g, where x is downwind.

t

=

Vertical cloud growth, as defined by the standard deviation of vidth CF,is given by g

-k t ) + bt

• stable case (

(2) cf

= a 1-e b

d 2 2-n 7

(5)Crf=C

• neutral and unstable case (2) g Prediction of Environmental Exposures From Sources Near the Ground, Based on Hanford Experimental Data, J. J. Fuquay, C. L. Simpson and W. T. Hinds, Journal of Applied Meteorology, Volume 3, No. 6, December 196h

( Environmental Radioactive ',7ntamination As a Factor in Nuclear Plant Siting Criteria, E. C. Watson, C. C. Gamertsfelder, February lb, 1963 - HW-SA-3809

10 APPENDIX III (Contd)

The values of the constants in Equations (h) and (5) which were used in each case are given below:

TABLE 1 (1 Diffusion Constants Used Wind 2

C a

b Y

g Speed Stability (Mph)

(m )

(m /Sec)

(Sec~)

(m "! )

n 4

Very Stable 2

3h 0.025 8.8 x 10 Moderately Stable 2

97 0 33 2 5 x 10-Neutral 2

0.17 0.25 Neutral 10 0.14 0.25 Unstable 2

0 35 0.20 Unstable 10 0 30 0.20 The calculated values for CF and CT were used in the Gaussian j

z equation to calculate concentrations in air at various downwind distances:

Q/Q l2)

-1/2 7-(6) X/Q

=

e 2

NFz" Py ;

where X/Q integrated air concentration (X) per unit activity release

=

g (Q )

g distance from center line crosswind (since plume center y

=

line used, y = o Q/Q correction for depletion (halogens and particulates

=

only

).

Environmental Radioactive Contamination As a Factor jn Nuclear Plant Siting Criteria, E. C. Watson, C. C. Gamertsfelder, February 14, 1963-HW-SA-2809

11 i

APPENDIX III (Contd)

The diffusion parameterCTgu, which was chosen for the first 2-hour period of the accident, vaa also applied to the calculation of doses for the longer time period of the total accident. This is quite conservative because larger values af this parameter obviously are appropriate for the longer time period; i.e., the values used as dis-cussed previously were for the 1-hour periods and thus are somewhat conservative vhen applied to the 2-hour period dose calculations and are markedly conservative when applied to the total accident (15-hour period) calculation.

5 Calculated Air Concentrations The methods described above were used to calculate integrated air concentrations (pe-sec/ce) from a unit release of 1 curie. The following table shows the values caleniated for the six different meteorological conditions assumed.

i i

12 A P P E N D'I X III (Contd)

TABLE 2.

' UNIT INTEGRATED AIR CONCENTRATIONS (pe-sec/cc Curies Released)

(By Methods Described in Journal of Applied Meteorology)

~

' Distance 7

(Miles)

'*VS-2 MS-2 N -2 N-10 U-2 U-10 1/2 NG 5 1 x 10 1 9 x 10

• 8.5 x 10~

1 5 x lo 3 4 x 10-5 5 5 x lo-4

-5 g,5 x ig-6 Ha1.

3 9 x lo 1.6 x 10 6.5 x 10-5^ 1.; x 10-5 2.8 x 10 Part. '5 0 x lo 1 9 x 10 8.0 x 10-5 1.4 x 10-5 3 3 x 10-5 55x10

-5 1,7 x 1g% -

-5 2 7 x 10-5 4.8 x 10-6 1,1 x 1g

'l NG -

2 5 x 10 9 0 x 10

-6 6

Ha1.

1 5 x 10 6.4 x 10-5 2.0 x lo > 3 5 x 10 8 5 x 10 1 3 x 10 4 7 x 10 1.0 x 10-5 1.6 x 10 Part. 1.4 x 10 8 9 x 10-5 2.6 x 10-5

-5

-7

-6

-7 3

'NG 7 3 x 10 2.2 x 10 4.4 x lo 8.8 x 10 1 7 x lo 3,g x 1g 0

-5 3 1 x'lo'# 5 5 x 10 1 3 x 10-2.6 x 10-7

-7 Ha1.

2 7 x 10 1 3 x 10 6'.9 x 10 2.1 x 10 4.3 x 10 8.5 x 10 1.6 x 10 2 9 x 10-7

-5

-7 Part.

-5

-5

-7

-8 5

NG 4.4 x 10 1 3 x 10 2.2 x lo-4.1 x 10-7 4.2 x 10 7 5 x 10

'Ha1.

1.1 x 10 6.5 x lo 1 5 x lo 2 5 x 10 3 1 x lo-I 5 0 x 10

-5

-7

.Part.

4.0 x 10 1.2 x 10 2.1 x lo 4'.o x 10-7 4.1 x 10-7

-5 7 0 x'lo j.

-5

-7 1.6 x 10-I 2 5 x lo 10 NG 2.2 x 10 6.1 x 10-1.1 x lo-2.1 x 10

-7

-7 1.8 x 10-0 Ha1.

3 1 x :.0-3 0 x 10 4.0 x 10-7 1.1 x 10 1.2 x 10

-5

-7

-7

-7

-8 Part.

2.0 x 10 5 9 x 10 8 5 x 10 1 9 x 10 15x10 2.4 x 10 f

~

1. Symbols refer to stability and vind speed conditions:

i.e., VS, MS, N and U mean very stable,. moderately stable, neutral and unstable, resIectively; 2 and 10 mean 2 miles per hgor and 10 miles per hour, respectively. The diffusion parameter CT T assumed i

- is 20 -mph (0.16 rad-m/sec) for the 2 mph cases and 130 -mph (1.0 rad-m/se'fl6) for the

'10 mph cases. NG _- Noble Gases; Ha1. - Halogens; Part. - Particulates.

t 0

r -

, +,

w

-,.n--

i.

13 APPENDIX III (Contd) 6.

Comparison to othar Calculational Methods As a point of comparison, the technique of calculating cloud growth (or spreading), in terms of its standard deviation of width (i.e., & and 7 ) described in E'f-SA-2809 were also used. This technique does not c( asider wind direction varia-bility (assumes persistent wind direction for entire period of release) and uses diffusion parameters for calculating cloud spreading derived from experimental work involving short term (10 minute) release and sampling times. Presumably this tech-nique is more appropriate for near instantaneous or puff re-leases rather than th longer time periods of interest in the hypothetical accidents described. Values of g,and Of cal-culated by this technique were used in the Gaussian equation for air concentration determination. The integrated air con-centrations pit unit amoant released (pc-sec/cc per curie released) calculated by this technique are shown in Tabit 3 It should be noted that unit air concentratione calculated by this technique are quite cimilar to those calculated by the method described previously. The tables (B and C of Appendix II) of radiological effecte use the two-cloud die-persion calculational techniques discussed here. These tables show the difference in dose calculated using the two different methods.

Environmental Radioactive Contamination As a Factor in Nuclear Plant Siting Criteria, E. C. Watson, C. C. Gamertsfelder, February 14, 1963 - hW-SA-2809

t.

14 A~P P E N D I'X

~ III -(Contd).

~

TABLE 3

. UNIT. INTEGRATED AIR CONCENTRATIONS (pe-sec/cc per Curie Released)

(Calculated by. Methods in IN-SA-3809)

Distance (Miles)

• VS-2 MS-2 N-2 N-lo U-2 U-lo 4

1/2 NG 9 2 x 10 34x10 1.2 x 10 2.4 x 10-5

-5 2.4 x lo 3 1 x 10 Ha1.

6.9 x lo 2.8 x'10 9 5 x.10-5 '2'.2 x 10-5 2.6 x 10-5 6.8 x 10-6 4

4

-5 Part. 9 1 x 10 2.4 x lo

'1.2 x 10 2.4 x 10-5 3 0 x 10 8.2 x.lo

-5

-5 1 5 x 10.5 2 5 x lo

~

1~

NG 5 2 x lo 2.0 x lo 31x10 1 5 x 10 Ha1.

2 3 x lo 1 3 x lo 25x10-5 1.0 x 10-5 3 0 x lo 2.0 x lo

-5

-6 Part. 5 0 x lo 1 9 x lo

' 3 0 x 10 9 5 x lo

-1.4 x 10-5 2.4 x 10

-5

'3 NG 1.4 x lo 4 3 x 10 6.0 x lo 1.4 x 10 1.2 x lo 3 1 x 10-7 Ha1.

5 3 x 10-5 2.6 x 10-5 4.1 x lo 9 5 x 10 8.8 x 10-7

-7

-7 2.2 x 10 Part. 1 3 x lo 4.2 x 10-5

-5 5 9 x 10 1.4 x lo 1.1 x 1.0 3 0 x lo-I~

5 NG 7 9 x 10 2 3 x 10-5 2.6 x 10-6 6.2 x'10-7 4.8 x 10-7 4

-3 1.3 x 10

-5

-5

-7 4

Ha1.

2.0 x 10 1.2 x 10 1.8 x 10 3 9 x 10-7 3.6 x 10 94x10

-5

-5 Part. 7 3 x-10 2.2 x 10 2 5 x lo 6.1 x 10-7

-7 1.2 x 10-7 4 7 x 10

-5

-6

-6

-7

-7

-8 10 NG 3 2 x 10 2.2 x 10 1,g x 1g 2.0 x 10 15x10 4.0 x 10 iHa1.

5 0 x lo 4 3 x lo 4.0 x lo-I

-7

-I 1.1 x 10 1.0 x lo 2 5 x lo

-6

-6

-7 3 0 x 10-5 8.0 x 10 1,g x 1g 1,9 x 1g 1,g x 1g-7

-8 3 8 x-lo Part.

• Symbols refer to stability and wind speed conditions:

1.e., VS, MS, N and U mean very stable, moderately stable, neutral and unstable, respectively; 2 and 10 mean 2 miles per hour and 10 miles per hour, respectively. NG - Noble Gases; Hal. - Halogens;

.Part. - Particulates.

• /

15 A P P E N l'I X III (Contd)

D.

Radiological Effects Calculation Methods The downwind effects, such as ground deposition and inhalation exposure, are principally a function of the integrated air concentration at any point. This integrated concentration decreases with distance due to turbulent diffusion in the atmosphere, and depletion of the containment cloud by deposition on the ground and on the ground cover.

The magnitude of this effect is shown in Table 2.

1.

External Radiation Dose From Passing Cloud The air concentrations downwind were estimated using the methods described previously. The conversion from air concentration to integrated dose from the passing cloud is time dependent due to the radioactive decay of the equilibrium fission product mixture.

For the noble gases, halogens and volatile solids, the concentra-tion required in an infinite cloud to produce a certain dose was evaluated for the radioactive decay periods of interest in the post ccident period.

The air concentrations in an infinite cloud 2

(pc/cc) which vill produce a whcle body dose rate of one trad per hour with hemispherical geometry are:

TABLE h Air Concentrations (ue/ce)

Giving One crad/Hr-Whole Body Dose Rate Decay Time (Minutes)

Noble Gases Halogens Volatile Solids 10 1.2 x 10' O.79 x 10~

1.7 x 10~

-6

-6 10 2.0 x 10~

0 79 x 10 2.0 x 10 3

10 3.h x 10~

O.90 x 10 2 5 x 10~

~

-6

-6 10 7 0 x 10~

1 90 x 10 2.h x 10

-6

~

~

10 h.5 x 10 1 90 x 10 2.2 x 10 2.

External Radiation Dose From Ground Deposition The fallout concentrations of radioactive materials were deter-mined on the bases of particle settling by eddy diffusion only, since settling by gravity is expected to be negligible in this l

case.

m 16

- i

~

c

. APPBNDIX

.III.(Contd).

The extent of halogen and solid fission product deposition on the ground is a function of the apparent deposition velocity 3

which, 'in turn, is considered to be a function of the diffusion condition and vind. speed.- Deposition velccities used in this:

evaluation were based on British results cited in HW-SA-2809 i

and are:

I TAElE_5 Ratio of Deposition Deposition Velocity t'

Velocity to Wind Velocity em/Sec Meteorology Wind Velocity Particles Halogenc Particles Halogens Very Stable. -1 M/S (2 Mph) 1 5 x lo' 2.4 x 10-3 0.015 0.24 Moderately Stable 1 M/S (2 Mph) 2.2 x lo"4 3.h x 10

<o.022 0 34

-3 Neutral 1M/S(2 Mph) 3.o x lo 4.6 x 10-3 0.03 0.46

-Neutral-5 M/S (lo Mph) 3 0 x lo 4.6 x 10-3 o.15 23

-3 Unstable 1 M/S (2 Mph) 6.o x lo" 8.0 x 10 0.06 o.8

[

Unstable 5 M/S (lo Mph) 6.0 x lo 8.0 x 10-3

~

o.3 4.0 1

j The evaluation provides for correction due to radioactive decay after the material is deposited on the ground. As the amount of deposition is a function of air concentration, and as the air' con-centration is depleted by prior deposition at locations closer to the source, correction for this depletion has been made for de-position at the distances illustrated.

In addition, the' dose l 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 is:

TABM 6 Finite Deposition Pattern Correction Factor Distance

.(Miles)

VS-2 MS-2 N-2 N-10 U-2 U-lo 1/2 0.82 0.82 0.84 0.81 0 92 0 91 3

o.90 0 90 5

0 95 0 95 t

-.y.-

1-,

e m

.v.

?,-

17 APPENDIX IIX (Contd)

The. conversion from deposition on the ground to gamma radia--

tion dose rate, at one meter above the grcund, res made~ con-sidering the gamma energies present from the halogen and solid fission products of which the deposited material is compored.

The conversion is dependent on the age of fission products present as follows:

mr/Er at 1 Meter Above Decay, Days -

a Curie / Meter Surface 0.1 1.0 x 10 h

1.0 9.1 x 10 3

10.0 7.2 x 10 3

100.0 3.,6 x 10 3

Exposurc Due to Inhalation 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 considering the dose from thyroid deposition of Iodine-131, 133 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, c "eidering their low rem per microcurie ratio for life time dosage and because of the estimated three to six hour thyroid uptake time after the material is inhaled.

The life time thyroid dose was evaluated for the three isotopes, considering a breathing rate of 230 ce/see as given by ICRP, and a thyroid deposition of 23% of that which was inhaled which was soluble as given by ICRP.

Dose to the lungs is primarily from the volatile solids and was evaluated by considering that all volatile and other solid fission products inhaled were insoluble and by the use of conventional standard-man metabolic factors.

___.._______.m._.

~

4 s

APPENDIX III(Contd)

Table 7

. Continuity of Wind : Direction Hours (One Sector 1/2 )

Longest-Longest No. Hours ( }

Station Direction 50%

10%

g 0.1%

No. Hours In Any Direction Augusta, Georgia W

2 3

8 13 18 W

18 Birmingham, Alabama S

2 4

9 16 16;

-SSE 20 Chicago, Illinois SSW 2

5 12 21 22 NNE 25 Little Rock, Arkansas SSW 2

4 9

17 28 sSW 28 Ihoenix, Arizona E

2 3

6 9

12 E.

12 ENE Rochester, New York WSW 2

6 13 23 28 WSW 28 Salt Lake City, Utah SSE 2

4 7

13 15 S-17 San Diego,. California NW 2

6 12 16 17 WSW

-33 S

Tampa, Florida ENE 2

3 7

13 14

.SSW 18-Yakima, Washington W

2 5

9 14 17-Average 2

4 9

16

. WNW 19 Big Rock Point (3) 14 WSW 2-7 13 NE 3

8 14 18 18 -

Direction-examined is the one showing. greatest frequency of persistent vinds. Directions are from N, etc. -

Longest number of hours observed may not be same direction as direction showing most frequency of persistent vinds.

(3) Based on Record from February 1961 to January 1962.

Data are for 20 increments.:

m e

=

+

19 APPENDIX III(Contd)

REFERENCES 1

1.

Consumers Power Company BRP-2 Reload Fuel Data Sheet, August 5,1965 2.

" Final Hazards Summary Report for Big Rock Point," Vol I, November 14, 1961.

3 Moody, F.

J., "GEAP-3515, Enclosure Pressure and Temperature Corresponding to Additions and/or Extractions of Water Phases and Heat," August 22, 1960.

4.

Baker, Louis and Louis C. Just, " Studies of Metal-Water Reactions at High Temrerature, III. Experimental and Theoretical Studies of the Zirconium Water Reaction," ANL-6548, May 1%2.

5 Riggs, C. O. and John Watcher, " Hydrogen Release Hazards in the PM-3A Containment Vessels," MND-M3A-3108 - Revised January 28, 1964.

6.

Reynolds, M.

B.,

"The Measurement of Free Fission Gas Pressure in Operating Beactor Fuel Elements," GEAP 4135, January 1963 7

Collins, D. A., et al,

" International Symposium on Fission Product Release and Transport Under Accident Conditions," 1965, Oak Ridge, Tennessee, Parer 59 8.

Collins, R. D. and Hillary, " International Symposium on Fission Product Release and Transport Under Accident Conditions," 1965, Oak Ridge, Tennessee, Parer 44.

9 Croft, Iles, AEEW - R-72,1962.

10.

Croft, Iles and Davis, AEEW - R-265,1%3 11.

Keilholtz and Barton, " Behavior of Iodine in Reactor Containment Systems,"

ORNL-NSNC 4, February 1965 12.

Maccary, R. R., et al, " Leakage Characteristics of Steel Containment Vessels and the Analysis of Leakage Rate Determinations," TID-20583, May 1964.

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