ML19281B028

SER Re Mod of Spent Fuel Storage Pool

ML19281B028
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Site:	Zion  File:ZionSolutions icon.png
Issue date:	03/29/1979
From:	Office of Nuclear Reactor Regulation
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ML19281B027	List: ML19281B026 ML19281B028 ML19281B029
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NUDOCS 7904200161
Download: ML19281B028 (16)

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i SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATING TO THE MODIFICATION OF THE SPENT FUEL STORAGE POOL FACILITY OPERATING LICENSE NOS. OPR-39 AND OPR-48 COMMONWEALTH EDISON COMPANY ZION NUCLEAR POWER STATION, UNIT NOS. 1 AND 2 DOCKET NOS. 50-295 AND 50-304 March 29,1979 o

780420016l

1.0 INTRODUCTION

By letter' dated April 13, 1978 and as supplemented October 24, November 8 and 29, 1978, January 24 and 26, February 23, March 7, and March 19, 1979, Commonwealth Edison Company (the licensee) requested amendments to Facility Operating License Nos. OPR-39 and OPR-48 for the Zion Station, Units 1 and 2, respectively.

The request was made to obtain authorization to provide additional storage capacity in the common Zion Station, Units 1 and 2, spent fuel pool (SFP).

The proposed modification would increase the capacity of the spent fuel pool from the present design capacity of 868 fuel assemblies to a capacity of 2,112 fuel assemblies.

The increased spent fuel pool capacity would be achievad by installing new racks with a decreased spacing between fuel storage cavities.

The present rack design has a nominal center-to-center spacing between fuel storage cavities of 21 inches.

The proposed new spent fuel racks would be modular stainless steel structures with individual storage cavities to provide a nominal center-to-center spacing of 10.35 inches.

Each stainless steel wall of the individual cavities would contain sheets of Boral (boron carbide in an aluminum matrix) to provide for neutron absorption.

The spent fuel poci is located in a separate fuel handling building adjacent to the reactor containment buildings. The general arrangement and details of the proposed new spent fuel storage racks are shown in Figures 3.1-2, 3.2-1, and 3.2-2 of the spent fuel rack modification report of February 3,1978 proposed by the licensee's consultant, Nuclear Services Corporation (NSC).

This report was forwarded as an enclosure to the licensee's letter of April 13, 1978.

The expanded storage capacity would allow the Zion units to operate until about 1993, or until about 1992 while still maintaining the capability for a full core discharge.

The major safety considerations associated with the proposed expansion of the spent fuel pool storage capacity for Zion Station are addressed below.

A separate environmental impact appraisal has been prepared for this proposed action.

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2.0 DISCUSSION AND EVALUATION 2.1 Criticality Considerations The proposed spent fuel storage racks will be an assemblage of open-ended double-walled stainless steel boxes with storage space for one fuel assembly in the cavity of each box.

These boxes will be about 14 feet long and will have a square cross section with an inner dimension of 8.94 incnes.

The nominal distance between the centers of the stored fuel assemblies, i.e., the lattice pitch, will be 10.35 inches.

The effective side dimension of the square fuel assembly, which was used in the criticality considerations, is 8.443 inches.

This results in an overall fuel region volume fraction of 0.665 in the nominal storage lattice cell.

Boral (boron carbide and alumin ~um) plates are to be press-fitted and seal-welded in the cavities between the double stainless steel walls.

The minimum homogeneous concentra-tion of the Boral plates will be 0.020 gram of the boron-10 (B-10) isotope per square centimeter of plate.

In this full array of storage boxes, there will be two Boral plates between adjacent fuel assemblies.

This makes the minimum areal density of boron between fuel assemblies 2.40 x 1021 B-10 atoms per square centimeter.

The fuel criticality calculations using the proposed new speWN3el racks are based on unirradiated fuel assemblies with no burnable poison and a fuel loading of 40.6 grams of uranium-235 (U-235) isotope per axial centimeter of fuel assembly.

These calculations were made by the Nuclear Services Corporation (NSC) for the licensee.

The basic method was to use the CHEETAH and XSDRN computer programs to generate four energy group cross sections for use in the CITATION diffusion program.

XSDRN, which is a one-dimensional, discrete-ordinates, spectral-averaging program, was used to calculate the cross sections for the Boral core regions.

Also, the internal black boundary condition in the CITATION program was used to calculate the neutron flux in the thermal energy group in tt)e Boral plates.

NSC checked the accuracy of this calculational method by using it to calculate two critical experiments which had Boral plates in them.

As shown in Table 3.3-3 of NSC's February 3,1978 report, the resulting neutron multiplication factors were more than 1 percent higher than'the experimentally determined values.

Thus, NSC assumed that this calculational method gives conservative results for the neutron multiplication factor Keff* in the spent fuel pool.

. NSC first used these programs to calculate a Keff of 0.92 for the nominal proposed storage rack lattice while assuming that the density

^Keff, effective neutron multiplication factor, is the ratio of neutrons from fissions in each generation to the total number lost by absorption and leakage in the preceding generations.

To achieve criticality in a finite system, Keff must equal 1.0.

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of the boron-10 in the Boral was at its minimum value of 0.02 grams per square centimeter of Boral plate and the pool temperature was 40 F.

NSC then calculated the changes in the neutron multiplication factor for various parameter changes, which we have summarized below.

Parameter Chance Chance in Keff 1.

A 0.01 increase in the 3.1 percent uranium-235 enrichment

+.0007 2.

An ir:rease in pool temperature to 212 F

.02 3.

The worst mechanical tolerances plus eccentric positioning of fuel assemblies in the racks

+.003 4.

One extra fuel assembly at the side of a filled rack

+.006 5.

One out of every 16 Boral plates missing

+.008 These considerations result in a maximum calculated Keff of less than or equal to 0.938.

In order to verify that the neutron multiplication factor in the spent fuel pool'will not increase due to a loss of boron from the racks, the licensee has provided a " Neutron Absorber Sampling Plan" that it has committed to implement.

2.1.1 Evaluation A cc.nparison of the above results with the results of other calcula-tions which were made for high density, spent fuel, and storage lattices with boron plates, shows them to be acceptably accurate.

By assuming new, unirradiated fuel with no burnable poison or control rces, these calculations yield the maximum neutron multiplication factor that could be obtained throughout the life of the fuel assemblies.

This includes the effect of the plutonium which is generated during the fuel cycle.

- The NRC acceptance criteria for the criticality aspects of high density fuel storage racks is that the neutron multiplication factor in spent fuel pools shall be less than or equal to 0.95, including all uncertainties, under all conditions throughout the life of the racks.

This 0.95 acceptance criterion is based on the overal.1 uncertainties associated with the calculational methods, and it is our judgment that this provides sufficient margin to preclude criticality in fuel pools.

Accordingly, there is a Technical Specification which limits the neutron multiplication f actor,-Kef f, 2-2

in spent fuel pools, including Zion Station, to 0.95.

Since the neutron multiplication factor in spent fuel pools is not a quantity which is measured with good accuracy, the only available value is a calculated one.

To preclude any unreviewed increase, or increased uncertainty, in the calculated value of the neutron multiplication factor which could raise the actual Keff in the fuel pool above 0.95 without being detected, a limit on the maximum fuel loading is also required.

Accordingly, we find that the Zion Station proposed high density storage racks will meet the NRC criteria when the fuel loading in the assemblies described in these submittals is limited to 40.6 grams or less of uranium-235 per axial centimeter of fuel assembly.

In its letter of February 23, 1979, the licensee proposed such a Technical Specification limit.

We find that the licensee's proposed baron surveillance program is satisfactory for monitoring the condition of the Boral plates.

In its response to our request for additional information, the licensee stated that in addition to the usual quality assurance program, a neutron poison verification test will be conducted at the Zion plant after the racks are installed in the pool.

This will be a qualitative test to statistically show with 95 percent confidence that the boron is not missing from more than 1 out of every 16 S tes.

We find that this will not cause the neutron multiplication factor in the fuel pool to increase above 0.95.

However, in this test, if any Boral plates are found to be missing, the NRC shall be notified and a complete test on every storage location shall be performed.

2.1.2 Conclusion We find that when any number of the fuel assemblies, which the licensee described in these submittals, which has no more than 40.6 grams of uranium-235 per axial centimeter of fuel assembly, is loaded into the proposed racks, the Keff in the fuel pool will be less than the 0.95 limit.

We also find that, in order to preclude the possibility of the Keff in the fuel pool from exceeding this 0.95 limit without being detected, the licensee's proposed Technical Specification limit is necessary to prohibit the use of these high density stcrage racks for fuel assemblies that contain more than 40.6 grams of uranium-235 per axial centimeter of fuel assembly.

On the basis of the information submitted, and the Keff and fuel loading limits stated above, we

onclude that the use of the proposed racks will not result in a

' :riticality.

2.2 Soent Fuel Cooling The licensed thermal power for each unit of the Zion Station is 3,250 ffdt.

The licensee plans to refuel both Units 1 and 2 annually.

This will require the replacement of about 64 of the 193 fuel assemblies in each core every year.

Thus, normal refuelings will take place at 6-month intervals.

To calculate the maximum heat load, 2-3

the licensee assumed that it would take 10 days after the reactor was shut down to complete the transfer of both 1/3 of a core in a normal refueling and full core in a full core offload.

With these delay times, the licensee used the method given in American National Standard 5.1 to calculate 20.4 x 106 Btu per hour as the maximum heat load for an annual refueling and 35.0 x 106 Btu per hour as the maximum heat load for a full core offload.

The spent fuel pool cooling system as described in Chapter 9 of the Final Safety Analysis Report consists of two pumps and two heat exchangers.

Each pump is designed to pump 2,300 gpm (1.15 x 106 pounds per hour), and each heat exchanger is designed to transfer 14.9 x 106 Btu per hour from 120 F fuel pool water to 95 F component cooling water, which is flowing through the heat exchanger at a rate of 1.49 x 106 pounds per hour.

As shown in Chapters 6 and 9 of the Zion Final Safety Analysis Report, there are seismic Category I sources of makeup water for the spent fuel pool.

These are the refueling water storage tanks.

There is one of these stainless steel lined, reinforced concrete, Class 1 structures for each unit, and each one holds 389,000 gallons of water.

2.2.1 Evaluation Using the method given on pages 9.2.5-8 through 9.2.5-14 of the NRC Standard Review Plan dated November 24, 1975, with the uncertainty factor K equal to 0.1 for decay times longer than 107 seconds, we calculatedthatthemaximumpeakheatloadduringBtuperhourand the 33rd refueling (the one that fills the pool) could be 22.2 x 10 that the maximum peak heat load for a full core offload that essentially fills the pool could be 41.4 x 106 Btu per hour.

This full core offload was conservatively assumed to take place 6 months after the 30th refueling.

We also deter.ained that the maximum iricre-mental heat load that could be added by increasing the number of spent fuel assemblies in the pool from 868 to 2,112 is 5.4 x 106 Btu per hour.

This is the difference in peak heat loads for full core offloads that essentially fill the present and the modified pools.

We calculated that with one pump operating with one heat exchanger, the spent fuel pool cooling system can maintain the fuel pool outlet

. sater temperature below 120 F for the normal refueling.

In the nighly unlikely event that both spent fuel pool cooling systems were to faii at the time when there was a peak heat load from a full core in the pool and the water was at its maximum temperature, we calculate that boiling could commence in about 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />.

We also calculate that after boiling commences, the required water makeup rat.e will be less than 85 gallons per minute.

We find that 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> will be sufficient time to establish an 85 gallons per minute makeup rate.

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2.2.2 Conclusion We find that the present cooling capacity for the spent fuel pools at the Zion Nuclear Power Plant will be sufficient to handle the incremental heat load that will be added by the proposed modification.

We also find that this incremental heat load will not alter the safety considerations of spent fuel cooling from that which we previously reviewed and found to be ecceptable.

2.3 Installp"..on of Racks and Fuel Handling The h el building crane, which will be used to remove the present racks and install the new ones, is rated for a 125 ton load.

The lif ting hook of the crane is dye penetrant tested prior to each rafueling outage.

The heaviest rack will weigh about 20 tons.

The licensee states that the racks will not be carried over stored fuel assemblies.

To ensure this, the licensee plans to move the fuel assemblies, which are now in the pool, as far away as possible.from the location whera the racks are being changed, i.e., to the other end of the pool.

The entire transfer operation will be supervised by the licensee's fuel handling foreman.

After the 1979 spring refueling, there will be about 372 fuel assemblies in the Zion spent fuel pool.

Since the present capacity is 868, the fuel assemblies can be removed from over 1/2 of the pool.

We conclude that this will provide enough room to allow the replacement of the racks without having to move them over fuel assemblies.

After the racks are installed in the pool, the fuel handling procedures in and around the pool will be the same as those procedures that were in effect prior to the proposed modifications.

These were previously reviewed and found acceptable by the NRC.

By letter dated April 8,1976, the licensee committed to notify t,he.

Commission in advance should it become necessary to handle heavy loads in the vicinity of the fuel storage pool.

The commitment was made to preclude such movements pending completion of our review of a postulatea accident involving a spent fuel assembly shipping cask.

Subsequent to this commitment, the staff has expanded the scope by initiating a generic review of all load handling operations in the vicinity of the spent fuel pools to determine the likelihood of a heavy load impacting fuel in the spent fuel pool and, if necessary,

- the radiological consequences of such an event.

The design basis fuel handling accident considered by the staff in its Safety Evaluation Report for Zion Station dated October 6, 1972 involves all the fuel rods (204) of one spent fuel assambly.

Pending the completion of this generic review, a Technical Specification restriction will be required for Zion Station to pieclude the handling of any loads of greater weight than a single fuel assembly plus the spent fuel handling tooi.

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In response to a staff question, the licensee identified four loads lighter than a single fuel assembly that are handled over stored fuel assemblies in the spent fuel pool.

These loads are the spent fuel handling tool (referred to above), the burnable poison tool, the rod cluster control (RCC) changing fixture, and the thimble plug.

Although lighter than a single fuel assembly, these four loads could develop greater kinetic energy should they be dropped because of greater potential drop heights.

This larger kinetic energy could theoretically cause more damage to stored fuel assemblies than that calculated assuming a single dropped fuel assembly. The licensee has therefore examined the use of these loads and has provided the infor-mation presented in the enclosed Table 2.3-1.

As indicated, the maximum kinetic energy of an unloaded rod cluster control changing fixture is approximately five times as great as that of a single fuel assembly plus the spent fuel handling tool.

The other loads identified, when carried at their maximum height, although not as great could also generate kinetic energy greater than a single fuel assembly plus the spent fuel handling tool.

To preclude any potential of an accident for which the consequences could be more significant than the design basis fuel handling accident for Zion Station, the licensee has agreed to a Technical Specification change which requires, in effect, that none of the loads identified above will be transported at a height greater than 2 feet over the spent fuel storage racks, the height at which a spent fuel assembly is normally transported. All other loads, ott)er than a fuel assembly, will be prohibited from being transported over spent fuel.

As also indicated in Table 2.3-1, the breaking strength of the double wire rope reeving system on the hook, results in substantial design factors of the wire rope in the unloaded and loaded conditions for each of the loads identified.

We believe these design factors are adequate to conclude that the probability of a wire rope failure accident when handling these loads is extremely small.

In addition,'

the licensee has described the handling procedures for these tools which require that the tools be picked up and secured to the crane hook in an area of the pool where no fuel storage racks are located.

The accidental release of a tool during the attachment process would therefore not involve stored fuel.

2.3.1 Conclusion

~ The consequences of fuel handling accidents in the spent fuel pool area are not changed from those presented in the Safety Evaluation Report dated October 6, 1972.

The design basis accident is ir. dependent of the number of fuel assemblies in the pool and is defined for fuel with the least decay after shutdown for refueling.

The accident is conservatively assumed to occur 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after reactor shutdown.

Technical Specificctions restrict the mvvement of irradiated fuel from the reactor cc/e until at least 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> after reactor shutdown.

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TABLE 2.3-1 SFIIT BPT RCC TP

!!ax. Drop of Empty Tool Over Storage Racks (ft) 15.3 14.2 13.2 16.5 Empty Tool Wt. (lbs) 352 000 975 235 flax. Kinetic Energy at Impact (ft-lbs) 5350 11360 12070 3076 tia x. Drop IIcight Loaded Tool.Over Racks (ft) 2 2

2 2

tiax. Weight Loaded Tool (lbs) 1932 040 1125 275 Fla x. Kinetic Energy a,t Impact (ft-lbs) 3064 1600 2250 550 Unloaded Tool Wire Rope Design Factor 352/43600 000/43600 975/43600 235/43600 Loaded Tool Wiro Hope Design Factor 1932/43600 040/43600 1125/43600 275/43600 Design Factor of Tool Connection (loaded) 5:1 5:1 5:1 5:1 Design Factor of Iloisting System (unloaded) 20:1 12.5:1 10.25:1 43:1 Design Factor of IIolsting System (loaded) 5:1 12:1 9:1 36:1 SFilT - Spent Fuel IIandling Tool llPT

- liurnable Poison Tool ItCC

- ItCC Change Fixtpro TP

- Thimble Plug

In view of the design factors present for the movement of loads over spent fuel in the spent fuel pool and the Technical Specifications Table 2.3.1 that will be issued to limit the maximum weight and maximum height of any load carried over spent fuel, we conclude that the consequences of the design basis accident for Zion Station fuel handling operations will not be exceeded.

2.4 Structural and Mechanical Desian The proposed modification of spent fuel storage capacity will involve the replacement of ex.isting spent fuel storage racks with new higher density racks.

The new storage rack base plates and legs will.be constructed from structural stainless steel, Type 304, and are desigited to seismic Category I criteria.

The new racks consist of a rectangular array of tubes which encapsulate Boral, a neutren absorbing material.

The Boral is a composite panel of bcron carbide (d C)/ aluminum matrix 4

clad with aluminum.

The tubes are interwelded along both their entire lengths and at the bottom to a base plate which is elevated by legs approximately 6 inches above the pool floor.

The racks will sit on existing embedments and leveling plates on the pool floor, with the exception of a few locations in the south side of the pool, where plates are welded to the rack legs to distribute the loads.

The racks will sit in the pool with no rigid attachments to either the adjoining racks or the pool floor.

2.4.1 Evaluation Structural and Mechanical The new spent fuel storage rack designs were reviewed for the following in accordance with the applicable parts of Sections 3.7 and 3.8 of the Standard Review Plan dated November 1975 and Branch Technical Position (BTP) on spent fuel storage:* structural design, material aspects, and analysis procedures for all loads including seismic and impact loadings; supporting arrangements for the racks; loading.

combinations and structural acceptance criteria; and quality control for fabrication and installation.

The seismic analysis performed on the racks used a response spectrum modal analysis, with the operating basis earthquake and safe shutdown earthquake pool floor acceleration values, for each direction.

The seismic model uti zed was a lumped mass stick model with equivalent properties determined, by comparison with a detailed finite element model of the racks.

The hydrodynamic mass effects and the fuel / cell interaction we e also included in this

- onalysis.

The resulting modal responses were comoined in accordance

<ith Regulatory Guide 1.92.

Both stability and sliding analyses were performed to insure that the racks would not overturn or slide into each other or the pool walls under seismic (safe shutdown earthquake)

'BIP "0T Position for Review and Acceptance of Spent Fuel Storage and Handling Applications" (forwarded to all licensees in April 1978).

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excitation.

The sliding analysis incorporated a model equivalent to the seismic model with a conservative value for the coef ficient of friction, established by test data.

A thermai analysis was also performed which considered both the increase in pool temperature and the thermal gradient resulting from a " hot" fuel bundle with adjacent empty cells.

In August 1978, the staff was made aware of a problem at the Monticello facility that had been identified with regard to spent fuel storage racks similar in design to those proposed for use at Zion Station.

The problem involved the in-leakage of water into the stainless steel cans, such that hydrogen gas was generated due to oxidation of the exposed aluminum material.

This gas caused a pressure buildup and resultant swelling of the stainless steel c:nc such that the removal of a fuel assembly, if located at an affected storage location, could not be performed.The Zion fuel rack tubes will be vented to release y off gases from the Boral.

The venting holes will be through nonstructural parts of the rack so the structural integrity is not compromised.

Material considerat?cas resulting from interaction be, ween the water ud structural components of the racks are discussed below.

The poni

'ructure was reanalyzed for loads resulting from the increased pool temperature, maximum thermal gradient, and the operating basis earthquake and safe shutdown earthquake events.

Th1 analyses were done in accordance with ACI Building Code 318-71 and ACI Standard 307-69.

The resulting loads were evaluated in accordance with the applicable parts of Section 3.8.4 of the Standard Review Plan, dated November 1975.

Material The Type 304 stainless steel used in the new storage racks is compatible with the storage pool environment, which is oxygen-saturated, high purity demineralized water containing boron as boric acid and controlled to a maximum 120 F temperature.

In this pool water environment, the corrosive deterioration of the 304 alloy should not exceed a depth of 5.96 x 10 5 inches in 100 years, which is minute relative to the initial thickness.

Dissimilar alloy interaction (electrolytic or galvanic corrosion) between the 304 stainless steel storage racks, Inconel and Zircaloy in the spent fuel assemblies, and the 304L stainless steel pool liner will be of no significance because of the minute electrical potential differential.

The aluminum M the Ecral. neutron absorber plates is more reactive than stainless steel and it will experience galvanic corrosion with the stainless steel tubes encapsu-lating the Boral being vented to the pool water environment.

Carolinas-Virginia Nuclear Power Associates, o '., anc Exxor Nuclear corrosion tests of Boral with a leak in the stainless s. eel covering nave shown a corrosion rate of 1.8 x 10 4 to 3.*

x 10 4 inc es per year for the aluminum in the Boral composite plates.

Tne ceteriora-tion was in the form of pitting and edge attack confi'ec in the area 2-3

of the leak path.

Pitting had no effect on the dislodgement of the B C particles in the Boral core.

In fact, the B C particles are 4

4 inert to pool water environment and galvanic corrosion and become embedded in the aluminum oxide corrosion product which forms on the edges of the Boral plate.

The more noble stainless steel showed no attack by the galvanic coupling.

Although galvanic corrosion does occur in ti unanodized aluminum of the Boral plates, it should not have any significant effect on the neutron absorption capability of the Boral, and certainly no effect on storage rack structural integrity for a period far in excess of 40 years.

The stainless steel pool liner, storage racks and base plates, an'd spent fuel assemblies would not be affected by interaction with the aluminum in the Boral plates for the following reasons:

(1) Stainless steel, Inconel and Zircaloy are more noble than aluminum and will not suffer galvanic or electrolytic corrosion.

(2) The Boral plates are completely encapsulated in the stainless steel tubes of the storage rack module, thus isolating them from the pool liner.

The stainless steel storage rack base forms a further protective layer between the Boral plates and the floor of the pool.

(3) The spacing between the storage racks (containing the Boral) and the pool liner is sufficient enough to cause electrical discon-tinuity.

(t ) The high purity pool water is not a sufficiently strong electrolytic solution to provide a conducting path which would allow galvanic or electrolytic corrosion to occur between any of the components in the modified pool which are not in actual physical contact with each other.

2.4.2 Evaluation Summary Analyses, design, fabrication and installation of the proposed racks are in accordance with accepted criteria and are in conformance with Branch Technical Position "0T Position for Review and

• Acceptance of Spent Fuel Storage and Handling Applications," cited previously.

The

- new storage racks were designed as Category I equipment.

The effects of the additional loads on the existing pool structure due to the high density storage assemblies have been examined to ensure pool structural integrity and leaktightness under the new loading conditions.

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Although acknowledgement has been made that corrosion will occur in the Zion spent fuel storage pool environment, it will be of no signif-icance for at least 40 years.

All the components in the Zion spent fuel storage pool, excluding the aluminum in the Boral neutron absorber plates, are constructed of alloys with the same electrical potential (or a minute differential) that have a high resistance to ger.eral chemical corrosion, electrolytic corrosion, and galvanic corrosion.

The only spent fuel pool components of concern are the storage rack modules, which have a galvanic coupling between the stainless steel tubes and the unanodized aluminum in the Boral.

The deterioration of the aluminum in the Boral by galvanic corrosion, however, would not be of such significance as to affect neutron shielding properties of the Boral.

The B C neutron absorber particles are inert to the p. col 4

water environment.

To aid in verifying the above conclusions, the licensee has committed to conduct a long-term fuel storage surveillance program to verify that the spent fuel storage cell r 2tains the material suitability and mechanical integrity over the life of the spent fuel storage racks under actual spent fuel pool service conditions.

Sample flat plate sandwich coupons and short fuel storage cell sections will be placed adjacent to the fuel storage racks and removed for visual and weight analysis on a schedule extending over a 40 year period.

2.4.3 Conclusion Based on the evaluation presented above, we find that the new proposed Zion spent fuel storage racks, the design materials used and the analyses performed for the racks, support frames, and pool are in conformance with established criteria, codes and standards specified in the staff position for acceptance of spent fuel storage and handling applications.

Therefore, we find that the structural, mechanical and material systems of modification proposed by the licensee are accept-able, and satisfy the applicable requirements of General Design Criteria 2, 4, 61 and 62 of 10 CFR Part 50, Appendix A.

2.5 Occupational Radiation Exposure We have reviewed the licensee's plan for the removal and disposal of the present racks and installation of the high density racks with respect to occupational radiation exposure.

The removed racks will be crated intact and shipped to a licensed burial site.

The occupa-

~ tional radiation exposure for the entire operation is estimated by the licensee to be from 2 to 5 man-rem for both units.

We consider this to be a reasonable estimate based on the dose received by per-sonnel during the 1976 modification.

This operation represents a-small fraction of the total man-rem burden from occupational exposure at the station.

Based on our review, we conclude that the exposures from this operation should be as low as reasonably achievable (ALARA).

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We have estimated the increment in onsite occupational dose resulting from the proposed increase in stored fuel assemblies on the oasis of information supplied by the licensee and by utilizing relevant assump-tions for occupancy times and for dose rates in the spent fuel area from rteionuclide concentrations in the spent fuel pool water.

The spent fuel assemblies themselves contribute a negligible amount to dose rates in the pool area because of the depth of water shielding the fuel.

The occupational radiation exposure resulting from the proposed action represents a negligible burden.

Based on present and projected operations in the spent fuel pool area, we estimate that the proposed modification should add less than 1 percent to the total annual occupational radiation exposure burden at this facility.

The small increase in radiation exposure will not affect the licensee's ability to maintain individual occupational doses to as low as is' reasonably achievable and within the limits of 10 CFR Part 20.

Thus, we conclude that storing additional fuel in the spent fuel pool will not result in any significant increase in doses received by occupational workers.

2. 6 Radioactive Waste Treatment The Zion Station contains waste treatment systems designed to collect and process the gaseous, liquid and solid wastes that might contain radioactive material.

The waste treatment systems were evaluated in the Safety Evaluation Report dated October 1972.

There will be no change in '.ne waste treatment systems or in the conclusions of the evaluation of these systems, as described in Section 11.0 of the Safety Evaluation Report, because of the proposed modification.

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3.0

SUMMARY

Our evaluation supports the conclusion that the proposed modification to the Zion Station spent fuel pool is acceptable because:

(1) The increase in occupational radiation exposure to individuals due to the storage of additional fuel in the spent fuel pool would be negligible.

(2) The installation and use of the new fuel racks does not alter the potential consequences of the design basis accident for the spent fuel pool, i.e., the rupture of a single fuel assembly and the subsequent release of the assembly's radioactive invento,ry within the gap.

(3) Tre likelihood of an accident involving heavy loads in the vicinity of the spent fuel pool is sufficiently small that no additional restrictions on load movement are necessary while our generic review of the issues is underway.

(4) The physical design of the new storage racks will preclude criticality for any credible moderating condition with the limits to be stated in the Technical Specifications.

(5) The spent fuel pool has adequate cooling with existing systems.

(6) The structural design and the materials of construction are adequate to assure safe storage of fuel in the pool environment for the duration of plant lifetime ar.d to withstand the seismic loading of the design earthquakes.

e e

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4.0 CONCLUSION

We have concluded, based on the considerations discussed above, that:

(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (2) such activities will be conducted in compliance with the Commission's regulations and that the proposed action to permit installation and use of high density spent fuel storage racks in the spent fuel pool at the Zion Nuclear Power Station will not be inimical to the common defense and security or to the health and safety of the public.

Date: March 29, 1979 O

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