ML20054H889

Safety Evaluation Re Proposed Facility Mods to Increase Capacity of Spent Fuel Storage Pool.Mods Acceptable Subj to Resolution of Open Items

ML20054H889
Person / Time
Site:	Maine Yankee
Issue date:	06/16/1982
From:	Office of Nuclear Reactor Regulation
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ML20054H885	List: ML20054H884 ML20054H889 ML20054H893
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NUDOCS 8206250143
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION MAINE YANKEE ATOMIC POWER CO W ANY MAINE YANKEE ATOMIC POWER STATION DOCKET NO. 50-309 1.0 Introduction, By letter dated September 18,1979 (Ref.1) as supplemented by references 2 through 9, Maine Yankee Atomic Power Company (MYAPC or the licensee) proposed changes in the spent fuel storage design for the Maine Yankee Atomic Power Plant. The changes consist of reracking the spent fuel pool for a greater number of storage. cells and conducting fuel pin consolidation (compaction) to allow more fuel pins to be stored in each cell.

Maine Yankee currently has the capacity for storing 953 fuel assemblies.

This represents an increase from the'318 assembly capacity, approved when Maine Yankee was licensed, which was accomplished by reracking the spent fuel pool as approved by the NRC in Amendment No.11 (Ref.10) to the operating license. The currently proposed reracking program would increase the number of storage cells from 953 to 1476. This would be. accomplished by installing racks in which storage location center to center spacing is 10.25 inches instead of the 12 inch spacing of the current racks. The licensee proposes-to further increase the storage capacity by disassembling tne spent fuel assemblies and reassembling the fuel pins into bundles having a more consoli-dated fuel pin array. Each consolidated fuel bundle will contain up to i

285 fuel pins as compared to the 176 fuel pins in a standard fuel assembly.

l Both the standard and consolidated fuel assemblies are sized such that they can be stored in the proposed spent fuel racks. By consolidating fuel pins in this manner, the storage capacity will be increased from 1476 standard assemblies to the equivalent of 2038 such assemblies plus the capacity for a full core discharge of 217 assemblies. This is also equivalent to 2390 standard assemblies without full core discharge capability.

The licensee also proposes to use an " emergency" rack' located in the spent fuel cask laydown area for temporary storage of up to 121 spent fuel assem-blies in their "as discharged" configuration. This rack will only be used for the temporary storage of fuel during a full core discharge should it<

be needed. This would allow an additional 121 storage cells in the pemanent racks to be used for standard / consolidated fuel assembly storage while retaining the capacity for a full core discharge.

8206250143 820616 DR ADOCK 05000309 PDR

a The Maine Yankee plant has a rated thermal power of 2630 Mwt. The reactor core consists of 217 fuel assemblies. A normal core discharge for refueling consists of 72 fuel assemblies. With the currently approved capacity for spent fuel storage, Maine Yankee will lose full core reserve capability in 1984 and will be unable to discharge a full reload batch of assemblies in 1987.

We have evaluated the safety considerations associated with the proposed changes to Maine Yankee's spent-fuel storage design. A separate environ-mental impact appraisal addressing these changes has been prepared. Our evaluation is presented as follows:

Section Topic Page 1.0 Introduction 1

2.0 Evaluation 2

2.1 Criticality Considerations 2

2.2 Spent Fuel Pool Cooling and Makeup 5

2.3 Removal and Installation of Storage Racks and Load Handling 8

2.4 Structural Design 11 2.5 Material s 14 2.6 Spent Fuel Pool Cleanup System 16 2.7 Occupational Radiation Exposure 18 2.8 Radioactive Waste Treatment 20 2.9 Radiological Consequences of Cask Drop and Fuel Handling Accidents 21 3.0 Open Items 22 4.0 Technical Specifications 23 5.0 Conclusion 23 6.0 References 24 2.0 Evaluation 2.1 Criticality Considerations 2.1.1 Descripti.on The Maine Yankee spent fuel pool criticality calculations were preformed for the proposed rack design center-to-center spacing of 10.25 inches and a maximum fuel enrichment for storage of 3.5 weight percent U-235. The new racks have been analyzed for storage of both standard Maine Yankee fuel assemblies and for the compacted assen611es.

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2.1.2 Evaluation A list of the conservative assumptions used in the calculation of the effective multiplication factor (keff, a measure of reactivity with criticality at keff = 1.000) is as follows:

Fresh fuel of 3.5 weight percent U-235 No soluble boron in the pool water No axial or radial neutron leakage from the racks 68* F water in the pool (the lowest anticipated temperature of pool)

A value of boron loading in the BORAL plates such that. there is a 951 probability that the boron concentration will be greater than that value, with 95% confidence.

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Worst case values of mechanical parameters including center-to-center spacing, BORAL plate thickness, etc.

Calculations were performed with the KENO-IV code with the 123-group cross-sections prepared with the AMPX/NITAWL code system. These codes were bench-marked against critical experiments performed by Battelle Pacific Northwest Laboratory (BPNL) and by Babcock and Wilcox (B&W). The critical experiments encompassed the enrichment value proposed as a limit for the ' tored fuel s

and included tests with lattices separated by BORAL plates. The benchmarking l

showed a positive reactivity bias (i.e. overprediction of keff) of 0.12 percent reactivity change with a standard deviation of 0.23 percent reactivity change. For conservatism, no credit was taken for the bias and an uncertainty of 0.8 percent reactivity change (the 95/95 value) was assigned to the calculated value.

The.results of the calculations is a value of 0.945 for the effective multiplication factor of the spent fuel storage racks when loaded with standard Maine Yankee fuel assemblies having a fuel enrichment of 3.5 percent U-235. Current Technical Specifications (1.3) limit the fuel used in the. reactor to 3.03 weight percent U-235. The 0.945 value of l

the effective multiplication factor meets our acceptance criterion of 0.95 for this quantity and is therefore, acceptable.r The licensee has described the inspection and testing to be performed by the rack supplier to assure the required presence of neutron poison in the racks.

In addition the spent fuel racks are classified as safety-related and suby ject to the Quality Assurance Program requirecer.ts of Appendix B to 10 CFR l

550. Therefore, adequate assurance is provided that the assumed boron loading l

will be present in the proposed racks.

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. The effect of loading compacted fuel assemblies in the racks is a decrease in reactivity. The licensee proposes to disassemble the fuel assemblies and place the pins in new " cages" having smaller center-to-center rod spacings (0.48 inches instead of 0.58 inches). The outer dimensions of the compacted assembly would be the same as that of the standard assembly but each storage location would hold 285 fuel pins compared to the 176 pins in the standard assembly. This pin compaction results in a decrease in the assembly water to metal ratio and a more undermoderated assembly. Calcula-tions performed by the licensee show that the effective multiplication factor reduction between a standard fuel assembly with 4.1 weight percent U-235 and a compacted assembly (285 pins) with 4.1 weight percent U-235 would be 0.16.

The same reduction was calculated in a comparison of assemblies with 3.2 weight percent U-235. Even if optimum moderation is assumed in the racks of compacted assemblies (due to incomplete filling of cages) the keff value is less than that calculated for storage of the standard Maine Yankee fuel assemblies due to the presence of stainless steel grids rather that the zirconium grids present in the standard assemblies.

The licensee has also examined the effect of having the new 10.25 inch racks adjacent to the existing 12.0 inch racks during the changeover and concludes that this is a less reactive configuration than an infinite array of the 10.25 inch racks. We concur with this conclusion based on the fact that the 12.0 inch racks have a much smaller effective multiplication factor than the 10.25 inch racks (0.76 vs. 0.945).

The licensee has considered the effect of various accidents on the reactivity of the spent fuel pool. These include the effect of dropping an assembly on the racks or alongside the racks and seismic distortion of the racks.

In all of these events credit may be taken for the presence of soluble boron in the pool. The Technical Specifications (1.1 Fuel Storage) require that the spent fuel pool be maintained at the refueling water boron concen-tration. This concentration is in the range of 1700 to 2000 parts per million of boron. At this concentration the effective multiplication factor of the pool is reduced substantially. Accounting for this presence of soluble boron, all the accident configurations analyzed had lower effective multiplication factors than the design value (0.945).

In addition, for certain postulated spent fuel cooling malfunctions boiling mAy occur in the pool.

In all credible cases the voiding due to this boiling would decrease reactivity.

2.1.3 Conclusions We conclude that the design of the proposed spent fuel storage racks is e

acceptable for storage of both standard Maine Yankee fuel, assemblies and compacted fuel assemblies, containing up to 285 fuel pins per assembly, in any combination, with respect to potential criticality in normal usage

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l i and in credible accident configurations. We also conclude that the proposed racks may be used in combination with the existing racks. Our conclusions are based on the following:

1.

State-of-the-art calculation procedures, which have been verified by comparison with experiments perfomed by BPNL and B&W, have been used for the analyses.

2.

Conservative values have been used for the input parameters of these calculations.

3.

The resultant value of the calculated design effective multiplication factor of the proposed racks, meets our acceptance criterion of less than or equal to 0.95.

4.

The effective multiplication factor of the proposed racks in contiination with existing racks is lower than for the proposed racks alone.

5.

The effect of storing compacted assemblies of fuel rods in the proposed racks'is a decrease in the effective multiplication factors.

6.

The effects of credible accidents on criticality have been considered and shown to be bounded by the design calculations.

2.2 Spent Fuel Pool Cooling and Makeup 2.2.1

System Description

There have been no modifications proposed to the existing MYAPC spent fuel pool cooling system (SFPCS).

It is a single loop coolinj system consisting of two parallel pumps, each rated at 750 gpm (3.75 x 100 pounds per hour) and one heat exchanger. The design is such that the rupture of any one line connected to the pool will not result in draining the pool by siphoning. A flow of 850 gpm of primary component cooling water cools the heat exchanger.

The rating of the SFPCS heat exchanger depends upon the pool water tempera-ture. Assuming the component cooling water inlet temperature is 85* F and6 a pool water temperature of 100* F, the heat exchanger rating is 4.72 x 10 BTU /HR When the pool water is at its limiting temperature condition of 154' F_1_/ he heat exchanger is rated at 22 x 100 BTU /HR.

t Three sources of makeup water have been identified:

1) the nomal makeup system is the chemical volume control system (CVCS) which can provide 150 gpm in less than 15 minutes; 2) backup makeup sources of water are three primary grade water hose connections, each capable of supplying 20 gpm in less than 15 ' minutes and 3) an emergency makeup source is the fire main system.

It is capable of providing 150 gpm in less than 20 minutes.

1/This value (154'F) is presented in the FSAR as the pool water temperature which will not be exceeded during storage of 1 and 1/3 cores. Although the cooling system is designed for higher temperatures (250*F), the licensee has chosen to retain 154*F as an administrative limit.

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.' 2.2.2 Spent Fuel Pool Cooling Since it is possible to discharge a full core at a rate such that the total heat load in the pool exceeds the heat removal capability of the SFPCS, MYAPC proposes to administratively control the rate at which spent feel is transferred to the storage pool such that the total heat load in the pool 6 BTU /HR in order to limit the at any point in time will not exceed 22 x 10 pool water temperature to 154' F.

To verify that these ifmits are not exceeded the pool water temperature must be monitored in certain situations.

MYAPC calculations illustrate how the discharge rate of fuel into the storage pool will change as the spent. fuel accumulates in the pool over the life of the plant. MYAPC has calculated the required decay time for a full core discharge, should it occur, when the fifth cycle fuel is discharged and when the thirtieth cycle is discharged.

In order to limit the pool water temperature to 154*F, a 13 day decay time would be required for a full core discharge at the end of th6 fifth cycle, i.e., June 1981. A 19 day decay time will be required for a. full core discharge should it occur at the end-of-pool life (cycle 30 discharge). Our calculations of the decay heat loads and required decay times are in close agreement with the licensee's, hence we conclude that the concept to limit fuel pool temperature by extending times before discharges to the spent fuel pool is acceptable.

MYAPC states typically seven days is required to remove the reactor vessel head and prepare for fuel discharge. Theoretically fuel assemblies can be discharged every half hour thereafter. Therefore, under ideal conditions a tull core could be discharged in approximately 10 days following shutdown.

In order to provide assurance that the total decay heat load in the spent fuel storage pool does not exceed the capacity of the SFPCS during full core discharges, it is necessary that after the first third of the core has been discharged, the licensee monitor the pool's bulk water tempera-ture following the insertion of each additional fuel assembly to verify that the limiting temperature of 154' F is not exceeded. Should the pool's '

bulk water temperature exceed 154' F, recently discharged fuel should be returned to the reactor vessel until the pool water temperature drops to or below 154' F.

l Should one of the two SFPCS pumps become inoperable when the pool contains the maximum heat load of 22 x 106 BTU /HR, and assuming no other action is taken, the bulk' temperature of the pool water will rise to and stabilize at 190' F.

This condition could be maintained for an extended period of time c

i assuming makeup water is provided.

In addition, MYAPC states that, in order to preclude the loss of cooling in the event of the loss of the primary com-ponent cooling water, the shell side of the SFPCS heat exchanger has been provided with emergency cooling connections which will enable the fire protection system pumps to be connected.

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. It has been calculated that should all cool cooling b lost when (1) the cooling system is at the maximum heat load of 22 x 10g BTU /HR, (2) the bulk temperature of the pool water is 154' F, and (3) assuming no heat loss by conduction through the pool walls and floors, pool water boiling would not be reached in less than 6.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. Further, for this case, the maximum boil off rate, i.e., minimum makeup rate, would be 50 gpm.

With regard to the loss of water due to a dropped load which perforates the pool liner, MYAPC has provided the results of an analysis of dropping a consolidated fuel bundle. The consolidated fuel bundle weighs 2099 pounds as compared to the 1280 pound weight of a standard fuel assembly.

In addition, MYAPC provided schematic infonnation regarding the consolidated fuel bundle.

The results of the licensee's analysis indicate that the consolidatc4 fuel bundle could penetrate 2.4 inches into the pool's concrete floor. This penetration would be by the bundle support structure and not the fuel rods.

The pool liner would be penetrated. The staff also calculated the penetration of a dropped consolidated fuel bundle. The results indicate that the liner -

would be penetrated and the depth of penetration into the concrete would be somewhat less that the licensee's calculated 2.4 inches. Therefore we conclude that the licensee's calculated depth of penetration is acceptable.

Due to the liner having been penetrated, pool water leakag'e will occur and its consequences were considered. Test channels behind the steel liner will direct the leakage to the spent resin pit sump for collection. FSAR Volume III, Answer to Question 9.14 indicates that leakage of pool water through the 18 inch penetration of the 6 foot thick reinforced pool structure bottom would be at a rate of 2 to 5 gpm following a 100 ton cask drop. We conclude that the leakage rate following the drop of a 2099 pound consolidated fuel bundle, which is calculated to penetrate 2.4 inches, will not exceed this value. Since the leakage is contained and well within the capability of three sources of makeup water the consequences of this accident are accep-table.

2.2.3 Pin Cooling Analysis An analysis was made to verify that local boiling does not occur in either the fuel assemblies or the consolidated fuel bundles. The analysis conser-vatively assumed the following:

the pool bulk water temperature was 154' l

F, the discharge fuel assemblies had a three day decay time from shutdown, the consolidated fuel bundle had a minimum of 120 days decay time from shutdown (whereas initially MYAPC does not intend to consolidate fuel until it is cooled for three, years), and the fuel assemblies had an average exposure,.

of 44,500 MWD /MTU. All decay heat calculations were in accordance with Branch Technical Position APCSB 9-2.

The results indicate that the maximum outlet temperature of the fuel assemblies and consolidated fuel bundles are well l

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. below (22oF and 130F respectively) the saturation temperature at the cell outlet. While the coolant flow area in the consolidated fuel bundle has been decreased, to a large extent the decrease in its decay heat release rate due to the longer decay time (i.e., to approximately one fourth that of a newly discharged fuel assembly) compensates for the reduced coolant flow area. We conclude that,within the limits of the assumptions, boiling will not occur in either the fuel assemblies or consolidated fuel bundles and the equilibrium void fraction.in the assemblies and the space between storage cells is zero. As ncted in Section 2.2.2, pool boiling could occur due to certain cooling system malfunctions. The effects of pool boiling and the elevated bulk pool temperatures are highly unlikely to cause a degradation of fuel integrity.

In addition the effects of boiling on criticality have been considered (Section 2.1.2).

Therefore, we find the proposed spent fuel storage configuration, from the standpoint of pin cooling, acceptable.

2.2.4 Conclusions 1.

The licensee must have procedures stating that during full core discharges, after the first third of the core has been discharged, pool bulk water temperature will be monitored following the insertion of each additional assembly. Should this temperature exceed 154' F, these procedures must direct that recently discharged fuel be returned 'to the reactor vessel until pool bulk water temperature drops to or below 154* F.

2. ' A limit must be added to the Maine Yankee Technical Specifications which requires at least 120 days decay of fuel from shutdown before it may be consolidated.

3.

Upon satisfaction of Conclusions 1 and 2 above we conclude that suffi-cient measures have been taken to assure the availability of adequate spent fuel pool cooling during normal and off normal conditions; time is available to provide sufficient makeup water to the pool in the unlikely event it is needed; and local boiling will not occur in either the standard fuel assemblies or the consolidated fuel bundles under the most limiting conditions expected.

2.3 Removal and Installation of Storage Racks.and Load Handling 2.3.1 Description The proposed storage racks are similar to those used at the Sequoyah plant.

There are 26 free' standing, fixed poison type, stainless steel storage racks composed of double walled stainless steel canisters with boral plates sealed within the double walls of the canisters. The canisters are welded to a top and bottom grid to fogm the racks. The racks are supported at the four corners by adjustable legs that rest on the pool floor. The size and weight of the racks vary with the size of the storage cell array. The array ranges between 6 x 6 and 8 x 9 and the weight of the individual racks ranges between 10,800 pounds and 21,600 pounds. The four support legs support the rack at least 4.25 inches above the pool floor in order to provide a water plenum for the lateral flow and entry of water into the individual storage cells.

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. The racks have been designed and fabricated in accordance with the applicable portions of Regulatory Guide 1.13,1.29 and 1.92 as well as Standard Review Plan, Sections 2.7, 3.9.4 and 9.1.

MYAPC states that one of the new storage racks is designated as a temporary storage rack and that it will only be used for the temporary storage of fuel during a full core discharge should it be needed.

It will be handled by the Fuel Building Overhead Crane and a four pick sling that meets the guidance of Section 5.1.6(1) of NUREG-0612.

It will be placed in the cask laydown area in the pool. Once the fuel has been removed from the rack and reloaded into the reactor the temporary storage rack will be removed.

2.3.2 Removal and Installation of Storage Racks Reracking of the MYAPC spent fuel pool will involve the removal of the 16 existing anodized aluminum fixed poison storage racks that were installed in the first spent fuel pool expansion program. Their weights range from 9,000 to 17,200 pounds as compared to the new proposed racks which weigh between 10,800 pounds and 21,600 pounds.

It is anticipated, due to cost considerations, that the reracking will take place over a number of refueling cycles. The heavy load handling equipment utilized in reracking has been previously reviewed in the first spent fuel pool expansion. This equipment will consist of the 125 ton Fuel Building Overhead Crane (Yard Crane) and a temporary crane that spans the storage pool. The function of the temporary crane is to move storage racks to areas located outside of the range of travel of the Fuel Building Overhead Crane hook. To minimize the consequences of a load drop when the load is handled by the temporary crane the following precautionary measures will be taken:

(1) fuel will be removed from the storage rack before the rack is moved, (2) the rack to be removed will be disconnected from adjacent racks, (3) a clear travel path will be established for the rack being moved, (4) fuel in the adjacent racks will be moved out of the area of influence should the rack be dropped, (5) using the temporary crane and four point rigging the rack will be lifted clear of the pool floor I

(carrying height not to exceed six inches); and transported to a lay-down i

area within the pool that is within the range of travel of the Fuel Building l

Overhead Crane hook, (6) four point rigging meeting the requirements of Section 5.1.6(1) of NUREG-0612 and the Fuel Building Overhead Crane will be used to lift and transport the storage rack out of the pool and deposit it in the Radiation Control Area (RCA) Building, and (7) the removed storage rack will be placed within a. plastic housing to await disposal.

The new racks will be installed in the pool by reversing the sequence of the steps taken to remove a storage rack. Technical Specification l

3.13 requires that racks may be moved only in accordance with written l

procedures which ensure that no rack modules are moved over fuel assem-l blies.

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Prior to commencing reracking operations, the Fuel Building Overhead Crane will be given a complete visual inspection by a factory representative.

The installation and removal of the temporary crane will be accomplished using the Fuel Building Overhead Crane and a four pick sling that meets the guidance of Section 5.1.6(1) of NUREG-0612.

MYAPC states in its September 18, 1981 submittal that the crane operator training, qualification and conduct will be in compliance with USAS B30.2-1967.

Chapter 2-3 of USAS B30.2-1967 addresses (1) visual, hearing, sight, physical coordination, speed, and emotional stability of the operator, (2) practical operator examination on the equipment being operated and (3) the_ conduct of the operator when operating the overhead crane.

While USAS B30.2-1967 has been superseded by ANSI B30.2-1967, MYAPC states that as it relates to the qualifications of the operator the two standards are essentially the same. We agree that the two standards are equivalent in this area. Therefore, we conclude compliance with Chapter 2-3 of USAS B30.2-1967, as its relates to the load handling operations associated with the spent fuel expansion program, is adequate and therefore, acceptable.

2.3.3 Light Loads In reference to the generic concern relating to the dropping of light loads, (i.e., weighing less than one fuel assembly and its associated handling tool),

over stored spent fuel MYAPC states:

"A fuel-bundle-handling-fixture drop on the fuel pool and. liner results in a kinetic energy of 31 ft.-kips.

In order to exceed this, an item dropped from the fuel handling crane must exceed a weight of 650 pounds. There are no items which exceed 100 pounds that nonnally pass over the liner from crane height."

Based on the above indication of normal practice at Maine Yankee and con-sidering the minimal pool leakage resulting from the dropping of an assembly or a cask (Section 2.2.2), we find that this concern has been adequately addressed.

2.3.4 Conclusions Based on our review of the removal and installation of storage racks and other load handling we conclude that adequate measures will be taken to prevent a load drop during the reracking program and adequate measures will be taken to minimize the consequences in the unlikely event of a load drop. Therefore, in this regard the methods proposed by the licensee for the removal and instal'ation of the storage racks and handling of other loads over the pool are acceptable.

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. 2.4 Structural Design 2.4.1 Description of the Installation The Maine Yankee spent fuel pool is an existing reinforced concrete box structure with six foot thick walls and floor. The inside dimensions are approximately 39 feet deep by 37 feet wide by 41.5 feet long. The pool is lined with a watertight, continuous,1/4 inch thick, stainless steel plate. The pool is founded on bedrock.

The new spent fuel storage racks are to be tonstructed of 300 series stain-less steel with " canned" Boral (boron carbide in an aluminum matrix core and clad with 1100 series aluminum sheet) poison sandwiched between stain-less steel sheets. The racks are vertical " egg-crate" structures, each of which is freestanding on four pads on the floor. An 8 x 8 rack (64 cells) would be approximately 14.8 feet high by 6.8 feet square. The minimum clearance between racks is to be one inch. The minimum clearance between a rack and the pool wall. is to be 3 inches,- the maximum is several feet.

2.4.2 Applicable Codes, Standards and Specifications The design, fabrication, installation and quality assurance standards for the new spent fuel racks were compared with the staff's "0T Position for Review and Acceptance of Spent Fuel Pool Storage and Handling Applications" dated April 1978, including a revision dated January 1979 (to be referred to henceforth as the "0T Position", Reference 11).

The proposed racks are designed in accordance with the requirements of the American Institute of Steel Construction (AISC) Manual which is an acceptable alternative in the OT Position.

The licensee proposes to use stainless steel conforming to ASTM-A666-72, Grade B, for certain portions of the racks. This material specification is not found in the ASME Code. The staff's position is that all rack material should conform to all applicable requirements of Section III, Division 1, Subsection NF of the ASME Code.

The licensee intends' to qualify the rack material in question to ASME Code Subsection NF (material specification SA240) and to obtain valid test results to justify the higher yield stress allowed by ASTM-A666, Grade B.

This is acceptable to the staff provided:

(1) A110wab'le weld and base material stresses in all heat-affected-zones are based on the yield strength for SA240 material.

, (2) Tensile tests indicate that the yield strength of the material used is not greater than 90 ksi.

(3) The licensee can provide objective evidence that stress corro-sion cracking of both base metal and weldement will not occur.

Citing of previous experience would be an acceptable approach.

(4) Complete documentation of the licensee's compliance with the above is developed.

In Reference 9 the licensee addressed these conditions. With regard to the first condition, the licensee indicates that such a reduction is not required since as-welded strength of ASTM-A666 material, as documented by test, is well above the required design yield strengt!.. This agrees with test results we have seen from the Farley Unit 2 application of this material.

In addition, test results of samples of steel to be used in the racks will be monitored during manufacturing to assure that the required minimum mechanical properties of the material are maintained. Thus, we find that the concerns of condition (1) have been adequately addressed.

With regard to the second condition, material certification records indicate the material yield strength as less than 90 ksi thereby satisfying this concern.

With respect to stress corrosion cracking, condition 3, the licensee has cited experience in nuclear spent fuel storage applications and factors such as low carbon content and a lack of sustained high temperatures, which will reduce the susceptibility to stress corrosion. We find that the licensee has satisfied this concern.

The licensee will develop and maintain complete documentation of material properties discussed above thereby satisfying condition four. Therefore, we find the licensee's use of ASTM-A666. Grade B material acceptable.

2.4.3 Seismic and Impact Loads A seismic time history analysis of the pool and its contents was perfomed using time histories which envelop the licensing design basis ground response spectra for Maine Yankee. Damping values were one percent for OBE and two percent for SSE. The dynamic model, consisting of springs, masses, gaps, and damping elements for a double rack system, includes the potential for rack-to-rack interaction, fuel-to-fuel interaction and. floor-to-rack inter-action. The seismic time history analysis considered the minimum and maximum storage configurations. A coefficient of friction between the pool and rack of 0.2 was used in order to define maximum credible sliding. The analysis was also performed using a coefficient of friction of 0.8 in order to define a worst case loading condition. The spacing of the' racks is such that. rack-to-rack impacts may occur in some modes; however, in all cases, stresses are maintained within allowable limits.

. In addition, impact loading of the racks from a fuel bundle drop was combined with dead loads and live loads at suitable themal levels. There are no current plans to use a spent fuel shipping cask in the Maine Yankee spent fuel pool for the foreseeable future. When the option to ship spent fuel off-site becomes available, the applicant has committed to make an evaluation for the cask drop accident on the liner and the concrete slab. Until this evaluation is completed, spent fuel shipping casks will not be lifted over the fuel pool.

Loads and load combinations were compared with the criteria outlined in SRP Section 3.8.4 and found to be acceptable.

2.4.4 Design and Analysis Procedures A dynamic analysis of the rack / pool was conducted using lumped masses, spring elements, gap elements and damping elements to model the system.

Hydrodynamic ef fects were considered.

In addition, a static analysis of the racks, using forces developed from the dynamic analysis, was also accomplished. The racks are not attached to the pool walls and the pool itself is founded on bedrock, therefore, the motion of the pool walls will not directly amplify the rack seismic motions.

The licensee's analysis includes consideration of the loads, acting upward, of a stuck fuel assembly as it is being lifted out of the rack. For this case, no pomanent defomation of the rack is allowed. Seismic loads were imposed simultaneously in three orthogonal directions and peak responses from each direction were combined by the SRSS (Square Root of the Sum of the Squares) method.

2.4.5 Structural Acceptance Criteria The structural acceptance criteria outlined in the licensee's submittal was compared to that outlined in SRP Section 3.8.4.11.5 and was found to be in conformance. The staff is currently guided in the consideration of ductility ratios by the information contained in Appendix A to'SRP Section 3.5.3 (NUREG-0800, Rev. O, July 1981). The licensee has indicated that ductility ratios used in its designs and analyses are in confomance with this position. With the exception (use of ASTM-A666, Grade B material) noted and approved previously, all materials are in accordance with the ASME Code, as are fabrication and inspection procedures. The racks are to be dimensionally verified using dummy fuel bundles. for each cell.

Poison cavities will be leak tested.

2.4.6 Conclusion 1.

We find that the racks will be constructed of acceptable materials and that the procedures for fabrication and inspection are in con-formance with the ASME Code.

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We find that the spent fuel pool, with fuel stored in the manners pra -

posed, complies with the Class I seismic design criteria forthe Maino Yankee plant.

3.

The license must be conditioned to preclude lifting a spent fuel shi > ping cask over the pool until a cask drop analysis is submitted by the li:ensee and approved by the staff.

2.5 Materials 2.5.1 Materials Description We have reviewed the compatibility and chemical stability of'the materials (except the fuel assemolies) in the pool water. The proposed new spent fuel storage racks are fabricated of Type 304 stainless steel with the exception of the adjusting bolts of the rack feat. These bolts are made from Type 17-4 PH stainless steel. The 17-4 PH stainless steel will be heat-treated at 1100*F.

The spent fuel storage pool contains high purity water with approximately 2,000 ppm boron as boric acid present in it. Tight controls are placed on impurities in this water, such as chlorides and fluorides to minimize stress corrosion cracking (SCC).

The new high density fuel rack modules are composed of poison cannisters and a bottom grid. The poison cannisters consist of two concentric stain-less steel tubes with Boral neutron poisonous material in the annulus.

Boral is Boron carbide in an aluminum matrix core, clad with 1100 series aluminum.

2.5.2 Chemical Compatibility Leakage of water into the weld sealed Boral cavity due to weld failure is unlikely, since welds are made in accordance with ASME Code procedures and both inspected and leak checked. Without the presence of water in the cavity, hydrogen gas resulting from the corrosion of aluminum will not be present.

Even if some gas should fopn, the rack design utilizes the inner wall core as the structural member, so that only the outer skin would bow from gas buildup, thereby preventing the fuel bundle, which is inside the canister, from being wedged and causing any mislocation of the Boral.

If isolated cases of leakage should occur, any swelling of the cans would not represent a safety hazard, t

Upon exposure of 'the Boral plates (8 C/A1 matrix) to the spent fuel pool 4

water, galvanic coupling between the aluminum-Boral liner' aluminum binder and the stainless steel shroud could occur. Deterioration of the Boral plates woula be limited to edge attack by general corrosion and pitting corrosion I

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of the aluminum liner and binder in the general area of the leak. The f

84C neutron adsorption particles are inert to the pool water and would j

become embedded in corrosion products preventing loss of the B C particles.

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Thus, this small amount.of deterioration would have no effect on neutron shielding, attenuation properties or criticality considerations (Ref.12).

Boral neutron poison material encapsulated in stainless steel in a borated f

i water coolant environment has been previously reviewed and accepted by us for similar designs in the Salem Nuclear Station and the Zion Nuclear Sta-i tion. These plants have ongoing material surveillance programs which will timewise, lead the operation of the Maine Yankee Spent Fuel Pool.

In the unlikely event that any adverse service experience is noted in these sur-l l

veillance programs there would be sufficient time to initiate corrective action for Maine Yankee.

In addition, the perfonnance of the other materials of which the spent fuel pool is constructed have been proven by experience and tests (Ref.13) to be stable and to operate satisfactorily at both temperatures and radiation levels in excess of those anticipated in the Maine Yankee Spent Fuel Pool. Based on the above, we conclude that a materials surveillance program is not necessary in the case of the Maine Yankee Spent 1.

l Fuel Pool.

The pool liner, rack lattice structure, adjusting bolts and fuel storage cannisters are stainless steel, which is compatible with the storage pool i

environment.

In this environment of oxygen-saturated borated water, the corrosive deterioration of the t;ype 304 stainless steel should not exceed a dept of 6.00 X 10-5 inch in 100 years (Ref.14), which is negligible relative to the initial thickness. Dissimilar metal contact corrosion (galvanic attack) between the stainless steel of the pool liner, rack lattice structure, fuel storage tubes, adjusting bolts and the Inconel and the Zircaloy in the spent i,

fuel assemblies will not be significant because all of these materials are protected by highly passivating oxide films and are therefore at similar potential s.

2.5.3 Conclusions We conclude that the corrosion that will occur in the spent fuel storage pool environment should be of little significance during the remaining life of the plant (Ref. 15). Components in the spent fuel storage pool are tonstructed of alloys which have a high resistance to general corrosion, localized corrosion, and galvanic corrosion. We therefore conclude that the environmental compatibility and stability of the materials used in the spent fuel storage pool is adequate based on test data and actual service experience in operating reactors. We find that the selection e

of appropriate materials of construction by the licensee meets the requirements of"10 CFR Part 50, Appendix A. Criterion 62, preventing criticality by maintaining structural integrity of components and is therefore acceptable.

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2.6 Sp_ent Fuel Pool Cleanup System 2.6.1 System Description i

The spent fuel pool cleanup system consists of a filter detineralizer l

l (precoat filter material and powdered anion and cation resin), filters, and assocated piping, valves, and fittings. The system is designed to remove corrosion products, fission products, and impurities from the pool water. Pool water purity is monitored by monthly chemical and radiochemical analysis. Demineralizer resin will be replaced when pool water samples show reduced decontamination effectiveness. The licensee has indicated that no change or equipment addition to the spent fuel pool cleanup system is necessary to maintain pool water quality for high den-sity fuel storage. The licensee has indicated that portable water filtra-tion equipment could be used to augment the installed cleanup system if deemed necessary to reduce general area dose rates during pin compaction.

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2.6.2 Evaluation We have considered the effects of the proposed increase in quantity of spent fuel in the pool 'and pin compaction on the capability of the cleanup system to perform its function.

The licensee estimates only a small increase in solid radioactive wastes, from change-out of the demineralizer and filters, as a result of the pro-4 posed storage expansion. This increase is due to fuel pin handling for fuel examination, reconstitution and compaction and, to a lesser extent, the increase in quantity of spent fuel stored.

Pool water purity is monitored monthly by chemical and radiochemical analy-sis. Chloride ion concentration is maintained less than 0.1 ppm. Gross gamma activity is maintained less than 1.0 X 10-2 Cf /cc. The decontamina-1 tion factor across the demineralizer is maintained greater than 1.0.

If any of these limits are not met the spent fuel pool demineralizer is nor-mally considered exhausted and a fresh change of resin is installed and the demineralizer is returned to service. The filters normally operate with a differential pressure in the range of 3 to 15 psi. Filters are replaced when the differential pressure exceeds 20 psi. The licensee reports that during a six year period of operation the deminer'alizer resin and filters were changed four times.

Experience at operating reactors shows that the greatest increase in radioactivity and impurities in spent fuel pool water occurs during f

refueling and spent fuel handling. Furthermore, experience at operating reactors and at fuel storage facilities in Morris, Illinois and Zest Valley, New York, shows that there is no significant leakage of fission products from spent fuel stored in pools after the fuel has cooled for

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. several months. Since the refueling frequency and the amount of additional spent fuel stored each refueling will not increase and the leakage of fission products from spent fuel in extended storage is not significant, we conclude that the proposed quantity of spent fuel to be stored in the pool will not significantly increase the impurities to be removed by the cleanup system.

Any increase in pool water impurities due to fuel pin handling can be accommodated by the cleanup system. Since fuel handling is a controlled activity and criteria exist for demineralizer and filter replacement, the effect of any increase in impurities would be more frequent operation of the cleanup system and more frequent filter and demineralizer replacement. The use of portable filtration equipment, which is currently pemitted, is set forth as an option in the unlikely event that area dose rates increase during fuel pin handling operations. That option would allow continued work. Another viable option, as noted by the licensee, is to suspend fuel pin handling operations until the cleanup system reduces area dose rates.

Based on the above, we have detemined that the proposed expansion of the spent fuel pool will not affect the capability of the spent fuel pool cleanup system to perfom its function. Accordingly, no change to the present cleanup syrtem is needed. More frequent replacement of filters or demineralizer resin, required when the differential pressure exceeds 20 psi or decontamination effectiveness is reduced, can offset any potential increase in radioactivity and impurities in the pool water as a result of the proposed ; pent fuel storage expansion.

2.6.3 Conclusion We have detemined that the existing fuel pool cleanup system, (1) provides the capability and capacity of removing radioactive materials, corrosion products, and impurities from the pool in accordance with the requirements of General Design Criterion 61 in Appendix A of 10 CFR Part 50 as it relates to appropriate systems for fuel storage; (2) is capable of reducing occupa-tional exposures to radiation by removing radioactive products from the l

pool water in accordance with the requirements of Section 20.1(c) to 10 CFR Part 20, as it relates to maintaining radiation exposures as low as is reasonably achievable; (3) confines radioactive materials in the pool water to the filters and demineralizers, and thus satisfies Regulatory Position i

C.2.f(2) of Regulatory Guide 8.8, as it relates to reducing the spread of contaminants from the source; and (4) removes suspended impurities from the pool water by filters, and thus satisfies Regulatory Position C.2.f(3) of t

Regulatory Guide 8.8, as it relates to' removing crud from fluids through physical action,.and, therefore, is acceptable for use with the proposed -

expansion of spent fuel storage capability.

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l 2.7 Occupational Radiation Exposure 2.7.1 Description We have reviewed the licensee's plans for modifications of the spent fuel pool (SFP) by reracking with high density racks coupled with consclidation of fuel assemblies with respect to keeping occupational exposures as low as is reasonably achievable (ALARA). To allow flexibility in the modifi-cation plan, the licensee is not specific in the manner in which the modification sequence will be performed. However, based on his analysis of the various tasks to be performed which are described below, we have evaluated a worst case operational plan (i.e., that which will provide the greatest cumulative occupational exposure) which considers removal and disposal of all low density spent fuel storage racks and replacing them with high density racks, followed by consolidation of 200 assemblies per year thereafter.

2.7.2 Increase in Storage of Spent Fuel The increment in dose rate resulting from the proposed increase in stored fuel assemblies will be quite small compared to the dose rate from the present spent fuel pool capacity. The spent fuel assemblies themselves contribute a negligible dose rate to the pool area because of the depth of water shielding in the pool. Calculations made for the dose equivalent rate (mrem /hr) above the surface of a spent fuel pool, similar to the Maine Yankee pool, from 1100 fuel elements stored in high density racks show a radiation level of about 10-8 mrem /hr. Thus, the direct radiation dose rate levels from the fuel assemblies themselves are not an important part of the total dose rate in the spent fuel pool area. The major contribution to dose rate in the spent fuel pool area comes from introduction cf reactor coolant water into the pool area during refueling. Dislodging of crud (activation products) from the surface of an assembly during fuel handling, radioactivity in the reactor coolant water from fuel leaks, and leakage of radioactivity from the stored spent fuel provides the balance of the radioactivity in the pool. ko, g {sotogggCs and ggations indicate that the contribution Gamm concen from the s,

o activities provides the major source of dose rate above the pool. Most of the activation products producing this dose rate come from the water interchange of the primary coolant water during transfer from the reactor during refueling. Many of these products will be removed from the pool by the spent fuel pool clean-up system. Experience has also indicated that there is little radionuclide leakage from spent fuel stored in the pool after the fuel has cooled for I

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l several months since, as stated previously, the radionuclides that are present in the reactor coolant system prior to refueling or crud dislodged l

from the surface of the spent fuel during handling, comprise the activity a

in the SFP water. During and after refueling, the spent fuel pool clean-up system reduces the radioactivity concentration in the water consider-l ably. Thus, the increase in the number of fuel elements in the pool due to the spent fuel pool modifications should not cause a significant increase in the radionuclide concentration and subsequent increase in dose rate.

2.7.3 Rack Removal and Installation 1

i The occupational collective dose equivalent for the removal and disposal 1

of the existing racks and the installation of the higher density racks is a one time exposure that is estimated by the licensee to provide about i;

12 man-rems. The occupational collective dose equivalent for pin compaction, which is an annual event, will range from 12 to 40 man-rems depending upon the work area to be occupied (i.e., work will.be perfomed in areas with dose rates ranging from 1 to 3 mr/hr). The above estimates are based on the licensee's breakdown of occupational exposure for each phase of the pool modification considering the reracking phase and a consolidation phase.

i The licensee considered the number of indi,vidualsperforming a specific job, their occupancy time while performing the job, and the dose rate in the J

area where the job will be performed. No underwater work'is necessary j

l so that there is no need for divers. The existing racks will be decon-taminated upon removal from the pool, moved into the Radiation Control i

Area where they will be dismantled and cut into pieces compatible with shipping containers.

It will then be shipped to a licensed burial site.

An alternative plan for packaging and shipping intact racks will be evaluated in tems of keeping occupational exposure ALARA at the time the racks are to be removed. We find this plan to be acceptable.

i 2.7.4 Fuel Consolidation i

In the consolidation operation, pins will be extracted from the fuel bun-dies and installed in consolidated assemblies.

The empty cages will l

be cut-up under water and packaged in a shipping cask for disposal. The total annual collective dose equivalent from the consolidation program will range from 12 to 40 man-rems. The range is based on compaction of l

200 assemblies in work areas where the dose rate varies from 1 to 3 mr/hr.

l To keep occupational exposure ALARA during fuel handling operations involving e

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i compaction, the licensee will provide, as necessary, a portable augmented fuel pool clean-up system with close capture water filtration units, a localized air filtration system to augment the nomal building ventilation system, and the plant Radiation Control Supervisor will evaluate the need to suspend all operations if the expected dose rates are exceeded.

In summary, the estimated collective dose equivalent for the reracking and consolidation programs is as follows:

12 man-rems for reracking and disposal of all existing fuel racks, and 12 to 40 man-rems for the compaction of 200 assemblies and disposal of the empty fuel cages. The latter operation will be an annual dose. Although the 40 man-rems is a best estimate upper bound collective dose equivalent for the compaction operation, it never-theless remains a small fraction of the collective dose equivalent from all normal plant operations, representing about 13% of the plant's annual everage man-rem for occupational exposure based on operations from 1973 through 1980 (

Reference:

NUREG-0713, Vol. 2). Consequently, the estimated annual maximum collective dose equivalent for the aforemen-tioned operation represents a small increase in the plant's occupational radiation exposure and should not affect the licensee's ability to main-1 tain individual occupational doses at ALARA levels and to be within the limits of 10 CRF Part 20.

Conclusion Based on our evaluation of the proposed modifications to increase the storage capacity of the Maine Yankee spent fuel pool we conclude that the increase in occupational radiation exposure to individuals due to-the consolidation of fuel in the SFP and replacing existing racks with reew higher density racks will be ALARA taking into account the state i

of technology of nodification of spent fuel pools and the economics of maximizing the capacity of the existing pool.

2.8 Radioactive Waste Treatment i

2.8.1 Description and Evaluation The plant contains waste treatment systems designed to collect and pro-cess the gaseous, liquid, and solid wastes that might contain radioactive material. These waste treatment systems were evaluated in the staff's Safety Evaluation dated February 25, 1972, which supported issuing an operating license to Maine Yankee. The licensee has not proposed changes in the plant's radiological effluent technical specifications as part of this storage expansion. There will be no change in the waste treatment systems or in the conclusions of the evaluation of these systems as described in Section 3.1.7 of the earlier Safety Evaluation because of the proposed modification.

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j 2.8.2 Conclusion We conclude that the staff's February 25, 1972 evaluation of Maine Yankee's radioactive waste treatment systems remains valid and therefore the systems designed to collect and process gaseous, liquid and solid radioactive wastes are acceptable.

2.9 Radiological Consequences of Cask Drop and Fuel Handling Accidents 2.9.1 Cask Drop Accidents The licensee has not presented any analysis of the radiological conse-quences of a cask drop accident onto consolidated spent fuel..

In its October 5,1981 submittal, the licensee states that, "there are no current plans to use a spent fuel shipping cask in the Maine Yankee spent fuel pool for the forseeable future." Based upon this assertion, therefore, we have not evaluated the radiological consequences of this accident.

Our accident evaluation has concentrated on the consequences of accidents involving fuel assembly handling (S.R.P.15.7.4).

2.9.2 Fuel Handling Accidents The licensee has proposed to expand the storage capacity of the SFP by a combination of reracking and consolidation of fuel assemblies from 176 pins per assemoly to 285 pins per assembly. The licensee states thdt the consolidated assemblies will initially consist of fuel cooled for at least 3 years. In addition, the licensee has performed calculations for fuel cooled for 120 days prior to being moved. The Safety Evaluation Report for Maine Yankee, issued February 25, 1972, evaluated the potential offsite doses (25 rems to throid and 2.0 rems whole body, at the exclusion area boundary) for fuel handling accidents resulting in rupture of an assembly containing 176 fuel pins, assuming that the accident occurred 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after removal from the reactor. An increase in the number of pins per assembly to a maximum proposed value of 285 would increase the undecayed source term used in that evaluation by an estimated factor of 1.62.

This increase is more than offset by the radioactive decay during the proposed increase in decay time (prior to movement of the assembly) from 3 days (72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />) to at least 120 days. The staff estimates that as a result of the increase in radiation sources and the 120 days of cooling prior to handling consolidated fuel assemblies, sufficient decay of noble gas and iodine isotopes will occur so that the release of activity in these r

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. elements during a postulated accident would be negligible (<.1 rem, S.R.P.15.7.4).

Since no penetration of the spent fuel pool liner by fuel rods is expected as a result of the drop of an assembly (Section 2.2.2),

no overheating of any fuel elements in the pool has been postulated and, hence, no release of other radionuclides is assumed.

2.9.3 Conclusions Based upon the above evaluation, the staff concludes that:

1.

The whole body and thyroid doses resulting from a design basis fuel handling accident are bounded by the SER evaluation of February 25, 1972, and therefore are acceptable provided that the licensee does not move spent fuel assemblies containing greater than 176 fuel pins which have decayed for less than 120 days.

2.

The license must be conditioned to preclude lifting a spent fuel shipping cask over the pool until a cask analysis is submitted by the licensee and approved by the staff.

3.0 Open Items A summary of those items for which additional information and/or licensee actions are necessary are summarized below. The conclusion specified in Section 5.0 of this evaluation and acceptance of the proposed changes are contingent upon resolution of these open items.

1.

The licensee must have procedures stating that during full ccre discharges and after the first third of the core has been dis:harged pool bulk water temperature will be monitored following the ilsertion of each additional assembly. Should this temperature exceed <54*F, these procedures must direct that recently discharged fuel be returned-to the reactor vessel until pool bulk water temperature drops to or below 154*F.

2.

A limit must be added to the Maine Yankee Technical Specifications which requires that fuel decay at least 120 days from shutdown before it may be consolidated.

l 3.

The license must be conditioned to preclude lifting a spent fuel shipping cask over the pool until a cask drop analysis is submitted by the licensee and approved by the staff.

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l 4.0 Technical Specification The itcensee has proposed one change to the Maine Yankee Technical Specification in conjunction with the proposed spent fuel pool modi-fication. This change to Technical Specification 1.1 Fuel Storage will increase the maximum allowable design value of the effective multiplication factor (keff) from 0.90 to 0.95.

This meets the staff criterion that the keff value be no greater than 0.95 as indicated in References 10 and 11 and the Standard Review Plan Section 9.1.2.

Based on our review, contained in Section 2.1 of this safety evaluation, we find the above change accep-table.

5.0 Conclusion Based on our evaluation and the conclusions presented in Section 2 of this safety evaluation and contingent upon resolution of the open items presented in Section 3 of this safety evaluation we find the proposed modifications to the spent fuel pool design and manner in which spent fuel is stored at Maine Yankee acceptable.

We conclude, then, based on our evaluation, 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 con-ducted in compliance with the Commission's regulations, and (3) the proposed license amendment will not be inimical to the common defense and security or to the health and safety of the public.

Dated: JUN 161982 PRINCIPAL CONTRIBUTORS:

W. Brooks F. Clemenson

0. Rothberg B. Turovlin P. Wu S. Block M. Wohl M. Thadani C. Nelson i

24 -

6.0 References 1.

Maine Yankee Atomic Power Company (MYAPCo.) letter to U.S. Nuclear Regulatory Commission (USNRC), dated September 18, 1979, Proposed Change 70.

2.

MYAPCo. letter to USNRC, dated October 18, 1979.

3.

MYAPCo. letter to USNRC, dated September 29, 1980, Proposed Change 70, Supplement No.1.

4.

MYAPCo. letter to USNRC, dated July 28, 1981.

5.

Licensee's Status Report dated September 29, 1981.

6.

MYAPCo. letter to USNRC, dated October 5,1981.

7.

MYAPCo. letter to USNRC, dated October 26, 1981, Proposed Change 70, Supplement No. 2.

8.

MYAPCo. letter to USNRC, dated February 10, 1982.

9.

MYAPCo. letter to USNRC, dated May 7,1982.

10. Amendment No.11 to Facility Operating License No. DPR-36, dated October 31, 1975.

11.

OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications, April 14, 1978.

12. Fuel Storage Racks Corrosion Test Program, Boral-Stainless Steel Xn-NS-TP-009, Exxon Nuclear Company, Inc., October 1978, Richland, Washington.

13. Reactor Handbook, Volume I Materials, Interscience Publishers 1960.

14.

A. B. Johnson, Jr., " Behavior of Spent Nuclear Fuel in Water Pool Storage" BNWL 2256, September 1977.

/

6.0 References (continued) 15.

C. Czajkowski, J. R. Weeks, el. al., " Corrosion of Structural 'and Poison Materiel in Spent Fuel Storage Pools". Paper 163, Corrosion /81, April 6, 1981.

16. USNRC letter to MYAPCo., dated October 5,1979.

17. USNRC letter to MYAPCo., dated November 30, 1979.

18. USNRC letter to MYAPCO., dated January 28, 1980.

19. USNRC letter to MYAPCo., dated March 17, 1981.

20. USNRC letter to MYAPCo., dated September 24, 1981.

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