ML19326C144

Fuel Storage Rack Replacement Safety Analysis.

ML19326C144
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Site:	Arkansas Nuclear
Issue date:	10/07/1976
From:	ARKANSAS POWER & LIGHT CO.
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ML19326C135	List: ML19326C134 ML19326C140 ML19326C144
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Text

O ATTACfBIENT III ARKANSAS NUCLEAR ONE-UNIT I FUEL STORAGE RACK REPLACEffENT SAFETY ANALYSIS

r-TABLE OF CONTENTS Page A. INTRODUCTION 1

1. Ilistory and Need for Replacement 1

2. General Description 3 B. ENVIRONNENTAL ASPECTS S C. SAFETY ANALYSIS 10

1. Criticality Considerations '

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a. Calculational Methods 10

b. Results 11

c. Systematic Uncertainties 12

d. Independent Audit of Calculations 12

2. Fuel llandling Considerations 13

3. Cask Drop Considerations 13

4. Material Considerations 14

5. Thermal Considerations 14

a. Fuel Assembly !! cat Removal 14

b. Fuel Pool Cooling 17

6. Installation Considerations 19

7. Mechanical Considerations 20 TABLE B Spent Fuel Pool Activity 21 TABLE B Gaseous Radioactivity Releases from Arkansas Nuc1 car One-Unit 1 22 .

TABLE Cl Assumptions Used in Criticality Analysis 23 TABLE Cl Theory-Experiment Correlations 24 TABLE C5 Thermal 11ydraulic Parameters for 37 KW and 100 KN Bundle lleat Load in Proposed AP6L Spent Fuel Rack 25 TABLE C5-2 26

' TABLE C5-3 27 FIGURE A2 Fuel- Storage Cell FIGURE A2 Typical liigh Density Spent Fuel Storage Rack for PWR's.

FIGURE A2 ANO-1 Spent Fuel Pool Arrangement '  ;

FIGURE B-l ' - Fuel. llandling Dose Rates l

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A. INTRODUCTION

1. llistory and Need for Replacement The Arkansas Nuclear One-Unit 1 spent fuel storage pool was designed on the assumption that a yearly fuel cycle wou'Id be in existence that would require the storage of a single batch of spent fuel for less than one year in the spent fuel pool. Therefore, a pool storage capacity of 1-1/3 cores was considered idequate. This would allow the complete unloading of the reactor for maintenance or inspection, even if one batch (1/3 core) were in the pool.

Currently, spent fuel is not being reprocessed on a commercial basis in the United States. In addition, spent fuel storage off-site is not available at the present time. It is therefore desirable to modify tne existing spent fuel storage facility to allow storage j -

l of additional spent fuel on the Arkansas Nuclear One-Unit 1 site.

I l Arkansas Power and Light Company has evaluated several alternatives 1

to increasing the storage capacity of the spent fuel pool. These alternatives include:  ;

a .- Shipment of fuel to another reactor site

, b. Shipment. of fuel to a reprocessing facility l

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c. Storage in existing racks All of these alternatives were determined to be unsatisfactor; because

.of the following:

. a. ~Because the fuel reprocessing and high level waste disposal problems are generic to the nuclear industry, it is~ not logical to attempt to store our fuel at another reactcr site.

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b. Currently, both the Nuclear Fuel Services (NFS) and the General Electric Company's reprocessing plants are in a decommissioned condition. Their fuel storage pools are available on a very limited capacity to only a few of their original customers -

Arkansas Power 6 1.ight Company does not have access to this storage. Reprocessing of the first six batches of fuel from ANO-1 is contracted to AGNS. Ilowever, the Allied General Nuclear Services (AGNS) plant is not licensed to operate and cannot be depended upon for receipt of spent fuel until all of the ~ issues

'1 relating to GES510, spent fuel shipment and waste disposal have l been settled. Therefore, shipment of spent fuel to a reprocessing plant is not an available alternative for several more years,

c. Storage in the existing racks is possible but only for a short period of time. The first batch of spent fuel will be discharged .

in January 1977, and an additional batch of spent fuel will be discharged annually thereafter. Therefore, after discharge of the second spent fuel batch in January 1978, the reactor will be operating without a' full . ore discharge capability. The fuel could not be discharged from the core in case of an emergency  ;

shutdown for inspection of the vessel or removal of the core internals. In addition, we would be unable to discharge the normal batch of spent- fuel by January 1981 because the existing j racks would aircady be filled with spent fue l . In either of the above two situations the purchase of replacement power would be approximately $450,000 for each day the reactor was not operating. Therefore, due to this high differential cost of power, this alternative is not acceptable.

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Based on the evaluation of these alternatives, Arkansas Power &

Light has concluded that increased on-site storage must be provided.

In addition, this modification to the racks must be completed before the first batch of spent fuel is discharged from the reactor in January 1977. This will allow drainage of the spent fuel pool so that the modifications and basic installation of some of the racks can be made under essentially clean conditions without spent fuel present.

2. General Description The Arkansas Power 6 Light Company has entered into a contract with the Exxon Nuclear Company, Inc., of Richland, Washington for the design, analysis, and fabrication of replacement spent fael storage racks for 590 fuel assemblies. These replacement spent fuel storage racks will provide storage capacity for approximately.3-1/3 cores through 1987. Therefore, 10 annual discharges may be accommod2ted or 7 annual discharges may be accom-modated while still maintaining the capability for a full core dis-charge until 1984. The contract price for the design and fabrication of the replacement racks is approximately $1,100,000 including estimated I freight charges. The current estimate for removal of the existing racks and ingtallation of the new racks is $200,000. This gives a total construction cost of $1,300,000 for the spent fuel rack modification.

. The replacement spent fuel storage racks wi si m , cicated from 504 stainless steel, totaling approximately 200,000 pcun<. . The racks do not use a poison material such as boron-impregnatea . stainless 3

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e steel or aluminum. The individual fuel assemblies will be stored in square fuel guide tubes fabricated from minimum .115 inch thick stainless steel. The 9-1/16 inch storage fnel guide tubes are mounted in the rack structure on a center-to-center spacing of 13-1/2 inches.

The replacement spent fuel storage racks consist of frames sup-porting storage receptacles (fuel guide tubes ) for spent fuel.

The basic structural function of these storage racks is to main-tain safe geometric spacing between spent fbel assemblies during and after all applicable loading combinations and t ansients. A typical fuel guide tube is shown in Figure A2-1. A fuel rack module consis'ing of an array of fuel guide tubes is shown in Figure A2-2. The rack. arrangement in the Arkansas Nuclear One-Unit-1 spent fuci pool is shown in Figure A2-3. .

The fuel storage modules will be supported on a gridwork of fabricated I-beams oriented in both North-South and East-West directions. This gridwork of beams is designed to provide uniform support to the racks and to transmit the weight of the racks to t:1e existing floor embedments. Provision for shimming under all support points is made to ensure a level support for the modules. Each module is located on the floor beams by dowel pins to assure the precise location of the modules required for proper operation of the fuel handling equipment.

1 The dowel pins are also designed to transmit hori: ental-siesmic loads to the floor beams. These loads are then transmitted to the pool walls by restraint arms bearing on the pool walls at the ends of all beams.

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e s B. ENVIRON 5fENTAL ASPECTS On September 16, 1975, the Nuclear Regulatory Commission announced its intent to prepare a generic environmental impact statement on handling and storage of spent fuel from light water power reactors. Arkansas Power and Light Company (APSL) is reque' sting a license amendment to allow modification of the Arkansas Nuclear One-Unit 1 spent fuel storage pool in advance of this environmental impact statement. The modification would give AP6L greater operating flexibility which would be desirable even if adequate off-site storage facilities should-late. secome available. Also, it is not likely that this modification would constitute a commitment of resources that would affect the alter-natives available to other. nuclear power plants or future actions taken by the industry to alleviate fuel storage problems. l The proposed modifications will require custom-made racks made of stain-less steel. This material is readily available in abundant supply. This material requirement is insignificant and does not represent an irrevers-l ible commitment of natural resources. l AP6L has assessed the environmental impacts of this modification. There are no potential effects on the environs outside of the auxiliary l building that will result from the proposed consturction work. With-in this building, the impacts are expected to be limited to those normally associated with metal-working activities. In addition, there are .

1 no adverse effects that will occur on-site or in the st crounding

. l environs that can be associated with an increase in the number of fuel assemblics stored in the pool.

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Since irradiated fuel has only been stored in the ANO-1 spent fuel pool for a short period during removal of the surveillance specimen holder tubes, it is expected that the existing fuel storage racks have a very low level of contamination. It is anticipated that very little effort wi. be required to bring the existing contamination levels down to those acceptable to allow disposing of the original fuel racks as ordinary scrap or low level waste. A final decision on the ultimate disposition of the existing racks will be made after their decontamination.

Coatinuous water purification is used to remove liquid wastes from the spent fuel pool water. This same filtration and demineralization will be used after the rack modification to maintain the quality of the water at the same high level as originally planned. There fore, there will be no increase in radiation levels inside the auxiliary building.

The total volume of water in the spent fuel pool, cask pit and tilt pit at normal pool level is 368,000 gallons. The spent fuel pool purification loop utilizes a filter and a 20ft3 nonregenerative mixed bed demineralizer with a flow rate of 180 gpm. One volumo of the spent fuel pool can thus be processed in approximately 34 hours3.935185e-4 days <br />0.00944 hours <br />5.621693e-5 weeks <br />1.2937e-5 months <br />.

4 There will be no increase in filter replacement frequency or the total radioactivity on the filter as a result of this modification.

The filter is presently designed to be replaced once every year with l

-capacity to remove corrosion products (crud burst) released by 1/3 of a ' fuel' core. Since the spent fuel rack modification increases only the storage capacity and not the frequency or the amount of fuel 6

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4 to be replaced for each fuel cycle, the amount of corrosion products released into the pool during any year will be the same regardless of the storage capacity of the pool. Thus, the present arrangement of replacing the filter on high differential pressure at a frequency of about once per year remains unchanged as a result of this rack modification.

The radioactivity of the spent fuel pool purification system demineralizer resin increases only by a small factor due to the increase in storage capacity. The domineralizer resin, like the filter, is expected to be changed based on differential pressure increase rather than on loss of capacity to remove radioactive contaminants. As a result, the resin replacement frequency will not be significantly altered by the increase in storage capacity.

In view of the above, it is anticipated that the radioactive solid waste generated from the facility will not increase as a result of

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this modification.

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- Table B-1 presents the maximum concentrations of radionuclides in the

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spent fuel pool. These concentrations are the sane as those used in the initial design of the spent fuel pool and are highly conservative since they are based on one percent failed fuel and a crud burst model.

They are also appropriate for the incr:ased storage capacity since the primcry contribution to the activity comes from the curd burst assumption which is not affected by the increased storage capacity.

Figure B-1 shows the dose contribution from fuel as a function of the distance from the top of the active fuel through water. Note that no 7

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significant contribution to dose is made by the fuel covered to a depth of 24 feet with water. Fuel being transferred is the controlling con-tributor to the basic dose rates, not the stored fuel.

Thus, there should be no significant increase in annual man-rem expo-sure over that which would be experienced with the existing spent fuel storage facility.

Radioactive gases may be released from the spent fuel pool directly into the atmosphere of the auxiliary building. This air is exhausted through particulate and charcoal filters. The major radioactive gas that may be released during fuel storage is Kr-85 with a half-life of 10.76 years.

The design of the ANO-1 facility does not permit measurement of radio-active gases released from individual ventilation systems but data is available for releases from the overall plant. Data for part of 1974, all of 1975, and the first half of 1976 are presented in detail in the

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Semi-Annual Reports for ANO-1. For easy reference, the data for Kr-85, I-131 and tritium are provided in Table B-2, As shown in Table 3.8 of the ANO-1 Final Environmental Statement, the initial estimate of Kr-85 release from the auxiliary building was 6 curies / year. Indreasing the fuel storage capacity by a factor of 2.33 will not necessarily increase the Kr-85 release rate. The fuel discharge from the reactor will continue on a 1/3 core per year rate and the release of Kr-85 is most likely to occur during the initial handling and the first year'of storage. Nevertheless, a conservative approach is to assume that the Kr-85 yearly release will increase by a factor of 2.33.

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- Therefore, .the maximum Kr-85 release from the auxiliary building is

'14' curies / year, an~ increase of 8 curies / year. The total plant release of .Kr-85 initially projected was 710 curies / year; thus,- the maximum percentage increase due to fuel storage expansion would be less than 1.' 13 *. .

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• 4 C. SAFETY ANALYSIS

1. Criticality Considerations -

An analysis was performed of the potential maximum reactivity of the fuel stored in the proposed fuel assembly storage facility.

This analysis considered the minimum possible spacing under normal and earthquake conditions, the maximum fuel enrichment level, the most reactive conditions of fuel density, and the most reactive water temperature. The parasitic neutron char-acteristics of the structure were also considered in the analysis.

No credit was taken for any boron presea+. in the storage pool water. The limiting conditions assumed for these varior: para-meters are given in Table Cl-1.

a. Calculational Methods The KENO-II Monte Carlo model was utilized to calculate the reactivity of the Arkansas Nuclear One-Unit 1 fuel storage array. Multigrour cross section data (18 energy groups) utilized in.these calculations were averaged using the CCELL2, -

BRT-1, 3 and GAMTEC-114 codes. Specifically, the cross section data for vari,ous regions within the storage array were obtained as follows:

CCELL - Utilized to obtain cell averaged multigroup cross section data for fuel rod-water lattices. Such calculations included _ both the bundle averaged call parameters and the actual lattice cell parameters. In addition, CCELL was used to (1) examine the effects of UO2 Pellet density, 10

,- - M r

• moderator temperature, and fuel temperature on the infinite

. media multiplication factor of the fuel assembly, and (2) calculate epithermal multigroup cross section data for

-stainless steel (e<0.683 ev) averaged in a neutron energy spectrum characteristic of.the water regions within a fuel assembly.

BRT Thermal group (0.683 ev) cross section data for the stainless' steel fuel guides were averaged using the Bate 11e Revised TilERMOS code. Such data were averaged assuming a 0.115 inch thick region of stainless steel GAbfrEC-II - Multigroup cross section data for water were averaged over a neutron energy spectra characteristic of an infinite media.

In addition to the codes identified above, the XMC Monte Carlo code was utili:cd to verify the accuracy of CCELL calculated -

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values of k, for rod-water lattices. The XMC code is a pseudo-point energy Monte Carlo code (approximately 90 energy groups) ,

which permits the discrete representation of the entire fuel assembly.

b. Results The reactivity of storage arrays were computed with both the nominally spaced guide tubes (13.5" center-to-center) and minimum spaced tubes ( 13.125" center-to-center) . The infinite multiplication factors f, r the arrays were calculated to be 11

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.913 1004 and 0.925+.004, respectively.

(The uncertainty factors refer to the Sionte Carlo calculational statistics at the 95% confidence level.)'

In addition to the above, a calculation was performed for a condition of pool water and fuel assembly temperature to bound the limiting effects of pool temperature variations. Specifically, the reactivity of the array was computed assuming the fuel assembly and associated moderator were at 200C while the water between adjacent fuel assemblies was assumed to be at 100 0C. Such assumptions maximize both the reactivity of the fuel assembly and the interaction between adjacent assemblies. For this boundary case, the reactivity was calculated to be 0.925+.004.

c. Systematic Uncertainties Theory-experiment comparisons have been made for small water-

' moderated critical arrays of fuel rods. Such critical experiments have been evaluated using the KEN 0 bbnte Carlo code with cross section data averaged as for this criticality safety evaluation.

The results of these calculations are shown in Table Cl-2. In-spections of the results indicate that the calculational method yicids conservative results relative to the experimental data.

In addition, the KENO calculated reactivities agree with the previ-ously performed DTF-IV S transport theory calculations within the statistical uncertainty of the >fonte Carlo Calculations.

d. Independent Audit of Calculations

~ An independent review of the criticality safety calculations 12

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is- being performed by Battelle Northwest to confirm the adequacy of the calculations. The results of this review will be reported in the October 18, 1976 submittal . '

2. Fuel llandling ' Considerations ,

An analysis was performed of the consequences of a fuel handling accident in the Final Safety Analysis Report for Arkansas Nuclear One-Unit 1. The Nuclear Regulatory Commission's Safety Evaluation Report for ANO-1 concluded that the analysis was acceptable. The modification proposed for the spent fuel racks would not affect the consequences or probability of that accident nor introduce a different or more severe accident.

An examination was made of the possible positions which a fuel assembly could assume if it were . inadvertently placed horizontally on the top of the filled racks. This examination showed that the reactivity of the array could not exceed the reactivity of the .

normal bundle array as ~ determined in the criticality analysis.

3. Cask Drop Consequences -

The Nuclear Regulatory Commission concluded in its Safety Evaluation Report for ANO-1 that the' spent fuel shipping cask storage area had been designed to minimize the consequences of an . accidental drop of a spent fuel shipping cask and was acceptable. The proposed spent fuel rack modification does not involve the spent fuel shipping cask' area. Therefore, the proposed modification does not affect the original cask drop evaluation.

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4. Material Considerations All permanent structural material used in the fabrication of the new spent fuel storage racks is 300 series stainless steel, mostly 304. This material was chosen for compatibility with the spent fuel pool water which contains boric acid at a nominal concentration of 1800 ppm boron.

At the normal operating temperature of 120F, there is no deteri-oration or corrosion of stainless steel in this environment. There is also no corrosion problem at temperatures up to and including pool boiling. All other structural components in the spent fuel pool system, such as the pool liner, cooling ystem pipe connections, etc., are made of stainless steel.

No poison material such as boron :arbide has been used in the de-sign of the fuel storage racks. The 304 stainless steel utilized at fabrication of the racks shall comply with ASIN specifications .

A276-71 or A167-74. All weld electrodes shall be 308 or 308L stainless steel.

In summary, the material used in the new spent fuel storage racks is similar t,o present components and does not affect or alter previous evaluations.

5. Thermal Considerations

a. Fuel' Assembly IIcat Removal The decay heat from a spent fuel assembly discharged from the A -kansas Nuclear One-Unit 1 reactor has been determined using 14

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ANS Standard 5.1. The heat generation six days after shutdown is approximately 37 KW. This value (37 KN) and a much more con-sen ative value (100 KN) has been used in analysis of heat removal from a single fuel bundle.

In the new fuel rack design, heat is removed from the stored-fuel by the natural convective flow of water up through the fuel guide. The rack utilizes a .115 in. thick stainless steel shroud with a square cross-sectional shape as the vertical fuel guide. The shroud acts as a chimney. Coolant is. heated within the chimney, expands slightly, and rises from buoyancy.

Upon leaving the shroud, the coolant mixes with water in the region above, is cooled, and then returns to the bottom of the shroud by flowing between the open spaces between the fuel shrouds. In the design, flow is provided to the fuel

' assembly by four orifices in the shroud - - one in each side of the shroud near the bottom. The combined flow area of these four orifices is 18 square inches.

Analysis was conducted to determine the adequacy of natural con- ,

vection cooling in maintaining fuel rod clad temperatures at accept-able levels. The analysis was ccaducted for two different conditions:

(a) normal operation with an-assumed pool discharge temperature and shroud inlet temperature of 1400F; (b) total loss of external cooling with a pool surface temperature of 2120F and a shroud inlet temperature of 240.90F. This last condition, item (b) is based on the conservative assumption that the coolant surrounding

. the fuel racks and entering the fuel assembly shroud is at saturation 15

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temperature corresponding to the hydrostatic pressure present at the top of the racks. The analysis was based on bundle flow resistance datr supplied by the fuel supplier. All calculations were made using a todified version of the COBRA computer code.

The results of the analysis are presr.ited in Table C51. As indicated there, the maximum clad temperature at the maximum postulated shroud inlet temperature is 254F. Cladding integrity is not degraded at these temperatures; thus natural convection cooling provides adequate cooling for the new rack design.

As noted in Table C5-1, boiling occurs in the fuel assemblies under the conditions postulated in item (b). Under these hypothetical conditions and the maximum calculated heat generation rate (37 KW) the flow is 40,500 pounds per hour, the quality is

.003'5, and the discharge void fraction is

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.218*.. The Martinelli calculation technique was used in the calculation of pressure drops within the fuel assembly for the boiling cases.

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b. Fuel Pool Cooling

' An evaluation was made to determine the adequacy of the existing spent fuel _ pool cooling system for the increased capacity of

-590 storage spaces. For purpose of the evaluation, ANS Standard 5.1 was used for decay heat rate calculations and the following worst case conditions were esaluated:

Case A - Normal refueling which accumulates a total of 10 batches (3-1/3 cores) of spent fuel with the exposure and cooling times listed in Table C5-2.

Case B - Complete core unload with seven batches of' spent fuel stored in the spent fuel pool with the fuel exposure and cooling times listed in Table C5-3.

Following is- a summary of the results of the evaluation made for these two worst cases:

Case A - Normal Refueling: (Table C5-2)

Assumptions: -(1) 930 days irradiation on all 10 fuel ba*.dtes ~

(2) 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> cooling time on last batch discharged Results: (1) Heat generation rate is 12.0 MBTU/HR.

-(2) The spent fuel pool cooling system will maintain the pool water temperature at approximately 120F.

Use of only one pump and one cooler will maintain pool temperature at approximately 145F.

(3) The pool will reach 212F in approximately 19 hours2.199074e-4 days <br />0.00528 hours <br />3.141534e-5 weeks <br />7.2295e-6 months <br /> if all cooling to the pool were lost.

Case B - Core Discharge: . (Tabic C5-3)

Assumptions: (1) Seven fuel batches stored in pool at time of discharge.

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(2) Irradiation time on core is 100 days in:o the eighth cycle.

(3) 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> cooling time when the core is dis-charged.

Results: (1) lleat generation rate is 28.5 MBTU/HR.

(2) The spent fuel pool cooling system will maintain the pool water temperature at approximately 150F.

(3) The pool will reach 212F in approximately 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> if all cooling to the pool is lost.

It is concluded that the existing spent fuel pool cooling system is adequate for the proposed increase in storage capacity since it can maintain the pool temperature at the original

. design temperatures.

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6. Installation Considerations Arkansas Nuclear One is currently operating in its first fuel cycle so there is no spent fuel currently being stored in the racks.

However, the irradiated first core fuel was temporarily stored in the spent fuel racks this past spring during the surveillance holder tube repair outage. Therefore, the pool water and the old racks have been exposed to irradiated fuel.

. The spent fuel pool has been drained and decontamination of the-pool and the existing racks has begun. The racks will be removed beginning November 1,1976 and will be disposed of under a separate contract.

Exxon Nuclear will handle the installation of the floor beam support network as well as installation of the new spent fuel ,

racks. Installation procedures will be required and will be reviewed and approved by the Plant Safety Committee Before instal-4 lation begins.

The first 'six rack modules (approximately 200 storage spaces)  :

- will be installed in early January while the pool is dry. These modules will be placed at the south end of the spent fuel pool.

The remaining twelve modules will be installed later during the first quarter of 1977 with the pool flooded. Since the first six modules will be installed in the south end, the remaining twelve modules can be brought up through the equipment hatch and installed in the pol without traversing over any of the installed modules.

The spent fuel handling bridge will also have to be reindexed for the new spent fuel rack spacings. <

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7. Mechanical Considerations This section will be formally submitted on October 18, 1976.

I G. E. Whitesides and N. F. Cross, " Keno - A Multigroup Monte Carlo Criticality Program," CTC-5, Union Carbide Corporation Nuclear Division, September 1969.

2 W. W. Porath, "CCELL Users Guide," BNW/JN-86, Pacific Northwest Laboratories, February 1972.

3 C. L. Bennett and W. L. Purcell, "BRT-1" Bate 11e Revised TilERMOS," BNWL-1434, 1 Pacific Northwest Laboratories, June 1970 )

l 4 L. L. Carter, C. R. Richey, and L. E. Ilushey, "GAMTEC-II: A Code for Gen- l erating Consistent Multigroup Constants Utilized in Diffusion and Transport  !

Theory Calculations," BNWL-35, Pacific Northwest Laboratories, March 1965. I 1

5 K. D. Lathrop, "DTF-IV - A FORTRAN-1V Program for Solving the Multigroup Trans-port Equation with Anisotropic Scattering," LA-3373, Los Alamos Scientific Laboratory, July 1965

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1 TABLE B-1 SPENT FUEL POOL ACTIVITY Isotope Activity (uCi/cc)

Sr-89 6:39(-10)

Sr-90 5.77(-11)

Sr-91 4.52(-12)

'Y-8%f 6.39(-14)

Y-90 6.51(-9)

Y-91 8.33(-10)

Y-9Di 2.92(-12)

Mo-99 3. 40 (-8)

I-131 4.06 (-8)

'I-133 5. 49 (-9)

I-135 2.59(-11)

Cs-134 9. 26 (-4)

Cs-136 3. 87 (-4)

-Cs-137 9.15(-3)

Da-140 9.08(-10)

La-140 7.86(-10)

_Pr-143- 4.52(-11)

'Ce-144 4.52(-11)

Cs-135 3.31(-12)

Fe-59 4.99(-6)

Mn-54 6.68(-6) .

Cr-51 8.63(-4)

Zr-95 2. 22 (-7)

Co-58 1.11 (-3)

Co-60 1.27(-4) -

Nb-95 1. 28 (-8)

Rb-95M 1.91(-9) 21

4 TABLE B-2 Gaseous Radioactivity Releases From Arkansas Nuclear One-Unit 1 Quantity Released (Ci) .

Nuclide 1974*: 1975 1976**

Kr-85 5.6E-2 3.15E0 2.23E0 1-131 5.3E-2 <7.2SE-3 3.44E-2

' H-3 3.0E-2 5.2E-1 3.4E0

• ANO-1 began operation in August 1974. Data reflects the releases from August 1974 through December 1974.

• • Through. June 30, 1976.

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TABLE Cl-1 ASSUMPTIONS USED IN CRITICALITY ANALYSIS Parameter Value Fuel Bundle Design 208 Fuel Rods (15 x 15 array) standard design used for Arkansas Nuclear One-Unit 1 Enrichment All fuel rods 3.5% U235 unirradiated UO2 Density 92.5's of Theoretical Maximum Temperature 20F for fuel and moderator

, Field Bundle Spacing 13.125 inches

• center-to-center for four bundle cluster Stainless Steel Thickness 0.115 inches surrounding each bundle

• Nominal spacing 13.5 inches, minimum spacing with adjacent bundles in corners of guide tubes 13.25 inches, maximum vibratory deflection of spacer during SSE 0.05 inches. The reactivity calculation is for a four-bundle cluster with 13.125 inches center-to-center spacing.

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TABLE Cl-2 TIIEORY-EXPERDIENT CORRELATIONS Exp'tl.

.Results Moderator- Cylin. CCELL-DTF-IV CCELL-DENO-II Fuel Cladding Square Core Calculated Density to-Fuel Calculated Exp't. Wt % Pellet' Thick. Lattice Volume Radius Reactivity Reactivity No. (g/cm3) 235U Dia. (in)- Mat'l. (in) Spacing (in). Ratio em (keff) (keff) 1 10.18 2.70 0.300. 304 SS 0.0161 0.435 1.405 26.820 1.016 1.008+.006

')

2 10.18 2.70 0.300 304 SS 0.0161 0.470 1.853 24.294 1.015 1.0141,005 3 10.18 2.70 0.300 304 SS 0.0161 0.573 3.357 23.600 1.011 1. 003,1l. 005 4 10.18 2.70 0.300 304 SS 0.0161 0.615 4.078 24.771 1.009 1.0101,.006

5 10.18 2.70 0.300 0.0161 0.665 4.984 304 SS . 27.172 1.005 1.005+.005 24

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.IABLE C5-1 .

Thermal flydraulic Parameters for 37 Kw and 100 Kw Bundle IIcat Load in Proposed AP4L Spent Fuel Rack Parameter 37 KW Assembly 100 KW Assembly 4

• Subcooled Pool Boiling Pool Subcooled Pool 1 Boiling Pool Outlet Temp. Surfacg Temp. Outlet Tgmp.

140 F 212 F Surfacg Temp.

155.7 F 212 F Coolant Mass Flow Rate Ib/hr. 12,000 40,500 19,200 60,000 Bundle Bulk Inlet Temperature 140 240.9 155.7 240.9 Bundle Bulk Discharge Temp. OF 150.5 240.9 173.4 240.9 fuel Pin Film Temp. Drop 0F 15 <10 0 31 <10 Fuel Pin Peak Cladding Temp. F 165.5 253 204.4 254 Equilibrium Quality

• 0 .003 0 .006 Void Fraction

• 0 .218 0 t.

.398

• At top of Assembly 25

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TABLE C5-2 4

i-Time of Time of Core irradiation Cooling i i Batch Fraction in-days . In years  !

j 1 1/3 (or 59 FA) 930 9

'2 1/3 930 8

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9- 1/3 930 1

.10 .1/3 930 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br />. .

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TABLE C5-3 Time of Time of Core irradiation Cooling.

Batch Year Fraction in days . in years

1. 1975' 1/3 (59 FA) 930 6 + 100 days

2. 1976 1/3 930 5 + 100 days 3 1977- 1/3 930 4 + 100 days 4 1978 1/3 930 3 + 100 days-5 1979 1/3 930 2 + 100 days 6 1980- 1/3 930 1 + 100 days 7 1981 1/3 930 100 days 8 -1981 1/3 720 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> 9 1981 1/3 410 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> 10 1981 1/3 100 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br />  ;

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