ML19208B972

Requests Proposed Change 70 to Section 9.10 of FSAR Re Spent Fuel Pin Storage.Class III Amend Fee Encl

ML19208B972
Person / Time
Site:	Maine Yankee
Issue date:	09/18/1979
From:	Johnson W Maine Yankee
To:	Office of Nuclear Reactor Regulation
References
PC-70-1, WMY-79-97, NUDOCS 7909240234
Download: ML19208B972 (28)

Text

Proposed Change No. 70 i

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. 4 PC-70-1 W>IY 79-97 September 18, 1979 United States Nuclear Regulatory Commission Washington, D.C.

20555 Attention:

Office of Nuclear Reactor Regulation

Reference:

(a) License No. DPR-36 (Docket No. 50-309)

(b) USNRC letter to YAEC Re: MYAPC, dsted June 22, 1979 (c) MYAPC letter to USNRC, dated Novettber 22, 1978 (d) MYAPC letter to USNRC, dated March 27, 1975 (PC No. 28)

(e) MYAPC letter to USNRC, dated June 26, 1975 (f) MYAPC letter to USNRC, dated July 25, 1975 (g) USNRC letter to MYAPC, dated October 31, 1975 (Amendment No. 11)

Subject:

Modified Spent Fuel Pin Storage Pursuant to the directions expressed in Reference (b), and in accordance with Section 50.59 of the Commission Regulations, Maine Yankee Atomic Power Company hereby proposes the following modification.

PROPOSED CilANGE: Reference is made to Section 9.10 of the Final Safety Analysis Report. We propose the follosing:

Increase the existing spent fuel storage capacity from 953 spent fuel assemblies to 1545.

This will be accomplished following a suitable cooling period and involves the disassembly of spent fuel assemblles, and reassembly into consolidated fuel bundles designed to provide a more compact fuel pin array. The consolidated fuel bundles will fit into the storage location of the existing spent fuel racks.

REASON FOR CHANGE:

In 1974, it seemed inevitable that availability of commercial fuel reprocessing would be significantly delayed and that reprocessing could offer no near-term relief for a continuing lack of storage ca pacity. Therefore, Maine Yankee on March 9,1975, forwarded to the NRC a finding under 10CFR 50.59 to allow expansion of storage capacity f rom 318 to 953 assemblies by installation of new high density storage racks.]iU3 eJ U L' 790924023 4 K

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USNRC September 18, 1979 Attn: Office of Nuclear Reactor Regulation Page 2 This modification resulted in a license amendment in October of 1975.

We now believe that neither reprocessing nor waste disposal at a repository or storage at a government facility can be relied upon to correct a lack of storage space d tring the 1980's.

Thus, there 18 a clear need for Maine Yankee to pursue other alternatives to provide the :tecessary storage space.

The projected discharge / refueling schedule for Maine Yankee as shown in Table 6.1, shows that with the current storage pool configuration full core reserve will be lost in 1984 and normal refueling discharge capability will be lost in 1987.

Consolidation of spent fuel assemblies will satisfy the temporary need for more storage at Maine Yankee, resulting in space for an additional 592 assemblies or somewhat more than eight regular discharges of 72 assemblies.

This space is adequate to assure that regular refuelings could continue until 1996 (without the capability for discharge of a full core).

While fuel consolidation will result in the capability to store more spent fuel on site, it will not necessarily extend the time period that fuel is kept on site.

Spent fuel could be sent off-site for reprocessing or to off-site storage after the government makes a firm decision on the reprocessing option and a facility is available to accept the fuel.

DESCRIPTION OF CIIANGE: Attachment A contains a detailed description of the change.

In general, the storage concept, " fuel consolidation", involves existing spent fuel bundle disassembly, and reassembly of only the fuel rods or pins into a modified skeleton cage within an envelope identical to that of a standard fuel assembly.

SAFETY CONSIDERATIONS: The spent fuel storage facility has no effect on plant operation or protection.

The facility serves as a storage facility for spent fuel prior to shipment for reprocessing. The proposed modification only increases the storage capacity of the existing racks within the design limits and analyses of the FSAR and Proposed Change No. 28.

The storage of additional spent fuel as part of the modification is considered permissible within the scope of previously submitted Proposed Change No. 25, which permits the licensee to receive, possess and use at any time, special nuclear material as reactor fuel, in accordance with the limitatioos for storage and amounts required for reactor operation as described in the Final Safety Analysis Report, as supplemented and amended.

Previous in-rfepth technical reviews by your staff of this proposed modification found our supporting evaluations (Reference (c)) to be 3 39 3903M

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USNRC September 18, 1979 Attn: Office of Nuclear Reactor Regulation Page 3 essentially complete and satisfactory.

This proposed change has been reviewed in light of Part 10CFR51 of the Commissions Regulations and we have determined that the quality of the human environment will not be affected, thereby negating any need for an environmental impact appraisal. Our determination has been found to be consistent with your findings as detailed in the Final Generic Environmental Impact Statement on Handling and Storage of Spent Light Water Power Reactor Fuel, NUREG 0575, dated August 1979.

Based on the considerations contained herein, it is concluded that there is reasonable assurance that operation of the Maine Yankee plant within the proposed specifications will not endanger the health and safety of the public.

This proposed change has been reviewed by the Nuclear Safety Audit and Review Committee.

Based on the considerations herein, it is concluded that there is reasonable assurance that operation of the thine Yankee Plant with these Spent Fuel Storage Facility modifications will not endanger the health and safety of the public.

This proposed change has been reviewed by the Nuclear Safety Audit and Review Committee.

FEE DETERMINATION: This propose hange is the result of an NRC letter, d

Reference (b) requiring that Maine Yankee Atomic Power Company amend the MY license in regards to our modified spent fuel pit :torage concept.

This proposed change requires an approval that involves a single safety issue and is deemed not to involve a significant hazards consideration.

For this reason, Maine Yankee Atomic Power Company proposes this change as a Class III Amendment. A payment of $4,000 is enclosed.

Maine Yankee Atomic Power Company believes that the NRC revised fee schedule is illegal. Maine Yankee Atomic Power Company is aware that an appeal of the decision of the U. S. Court of Appeals for the Fifth Circuit is being considered and therefore submits the enclosed fee under protest without waving its right to contest the validity of this fee or the entire NRC fee schedule, and without waving any right to recover, in whole or in part, all fees paid or to be paid under the invalid fee schedule.

SCHEDULE OF CHANGE:

The proposed spent fuel consolidation will be implemented on a phased basis to provide an increasing capacity consistent with the normal refueling cycle. The program of fuel consolidation will be accomplished utilizing written procedures which will ensure exposures are maintained as low as reasonably achievable, and provide maximum safety to personnel.

We respectfully request approval of this proposed change not later than December 3, 1979.

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USNRC September 18, 1979 Attn: Of fice of !!uclear Reactor Regulation iage 4 We trust this information is acceptable to you; however, should you have any questions, please feel free to contact us.

Very truly yours, MAINE YANKEE ATOMIC POWER COMPANY t],n (] /

(l;rb._ L L npgq W. F. Johnson Vice President COMMONWEALTil 0F fiASSACilUSETTS)

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COUNTY OF WORCESTER

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Then personally appeared before me, W. P. Johnson, who being duly sworn did state that he is Vice President of Maine Yankee Atomic Power Company, that he is duly authorized to execute and file the foregoing request in the naiae and on the behalf of Maine Yankee Atomic Power Company, and that the statements therein are true to the best of his knowledge and belief.

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9.10 F0EL hah'DLING SYSTEM i

9.10.1 Ceneral l

The fuel' handling system provides a safe and efficient method to unload 1

and store new fuel assemblies in the fuel building, to trensfer and load new assentlics into the reactor core, to remove spent fuel assemblies f rom l

the reactor core, to transfer spent fuel assemblies from the reactor c antainme.it to storage in the fuel pool and to ship spent. fuel assemblies I

to be processed offsite.

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Equin.ent is provided to exchange control element assemblieu.between fuel assemblies and for underwater close inspection of fuel assemblies and control

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element assemblies by periscope. The refueling equipment arrangement is

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I shown en Figure 9.10-1.

9.10.2 New Fuel Storage The new fuel storage area, located at elevation 44 feet in the fuel building, l

contains facilities to receive, handle, and store new fuel assemblies as they arrive at the site.

New fuel assemblies are delivered by truck to the site in steel containers.

l The ner fuel containers are unloaded and conveyed into the fuel building where the 5 ton crane picks up the fuel assemblies and transports them to storage in the new fuel storage racks.

The new fuel is stored dry in racks that have a center-to-center spacing of 20 inches.

There is provision for the storage of 160 fuel assemblies (2/3 of one core load plus 15 spares)

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in the new fuel stonge area.

For initial core loading, the additional I

fuel assemblics (1/3 of one core load) are stored in the fuel pool.

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place an assembly in the pool, the fuel assemblies are placed in the fuel i

I elevator by the 5 ton crane.

From the fuel elevator, the fuel assemblies are moved to spent fuci storage locations by the movable platform hoist.

Fuel assemblies are also stored on 12 inch centers in the spent fuel pool.

i These dimensions were chosen because a considerable cargin of suberiticality exists even if the fuel pool or new fuel storage area were filled with demineralized water.

During normal refueling, a new fuel assembly is lifted from the new fuel storage rack by the 5 ton crane and placed in the fuel elevator in the spent fuel pool. The fuel elevator lowers the fuel assembly to the bottom of l

the spent fuel pool where the tovable platform hoist lifts it and places it into the fuel assembly upender for transport into the containment.

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9.10.3 Fuel Pool The fuel pool, 37 feet wide by 41 feet long by 38 feet deep, is located l

in the end of the fuel building adjacent to the reactor containment.

The pool is constructed of reinforced concrete with a wall and floor lining l

of 1/4 inch thick stainless steel. A fuel transfer tube connects the spent fuel pool with th'e refuel-cavity located in the containment. Welds in the liner are backed up by test channels which are piped to the fuel building l

sump.

There are 16 aluminum poison curtain storage racks provided for the storage of a maximum of 1545 equivalent fuel assemblies spaced on a minimum of 12 l

inch centers. The spent fuel is stored in borated water.

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ATTACFCIENT A 1.0 PIN STORAGE CONCEPT l

l The pin storage concept, termed " fuel consolidation", involves the disassembly of irradiated fuel assemblies and the reassembly of the fuel l

pins into a skeleton cage similar to the standard cage.

The consolidation is derived from the elimination of CEA guide tubes, rod spacer springs (in i

the egg-crate grid assemblies) and the elimination of poison rods (shims).

l The skeleton cage is depicted in Figure 1.1.

A tighter but still square t

I pitch is achieved with an envelope identical to that of a standard fuel assembly.

l This scheme for increasing storage capacity is very simple, and can be implemented by adapting the procedures and equipment previously used for

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fuel reconstitution with which Maine Yankee and many other utilities are g

f amiliar.

i The potential for increased storage is significant.

The existing racks have spaces for a total of 953 fuel assemblies.

Based on the proposed g

consolidated storage scheme, each rack cell can accommodate 285 fuel pins.

t Since there are a maximum of 176 fuel pins per normal assembly (some bundles l

have up to 16 non-fuel pins), the minimum equivalent number of assemblies which can be accommodated in the 953 cells is:

953 x 285 %1543 compacted assemblies

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or an additional 590 f u21 assemblies. The final number will be slightly i

higher because of the poison shim rods in some assemblies.

These non-fuel rods would be disposed of as radwaste.

This is enough added space for l

somewhat more than eight additional regular discharges of 72 assemblies, and therefore can assure adequate spent fuel storage for an additional eight years or more.

l 1.1 Mechanical Disassembly Maine Yankee has working experience in the disassembly of irradiated fuel j

assemblies. The t!,e required for disassembly is a function'of fuel design and tooling constraints. The earlier fuel types (A, B, C and RF) have a l

lower retention grid and a permanently fastened upper end fitting, and str.

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assemblies of this design have been disassembled at Maine Yankee.

While the disassembly is time consuming, experience has shown it to be straight l

forward.

With the pins being handled singly, the per pin removal time l

averaged ten minutes. This " pin removal time" included the time to position the bundle, remove the upper end fitting and replace the pins.

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More recently, Maine Yankee has accomplished reconstitution of the E type fuel.

(Fuel types D and beyond are reconstitutable by design.)

In the E fuel, 16 burnable poison shim rods per assembly were removed and replaced l

by 16 water filled shim pins.

The operation required about four hours per assembly, three hours was the best time observed.

Removal of the upper l

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l end fitting consumes about 20 minutes.

Based on these experiences, we estimate pin handling time to range from five to ten minutes per pin for l

the overall extraction and storage operation.

Again, this estimate includes l

not only the time to extract a given pin, but the time also to position the assembly, remove the end fittings, extract 2he rods, store the rods, l

and transport the empty cage and end fittings to storage.

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1.2 Mechanical Reassembly Reassenbly operations are similar to the reconstutution techniques proven i'

at Maine Yankee. The storage cage or skeleton depicted in Figure 1.1 maintains the assembled fuel pins in a square array on a fixed pitch which has been determined desirable from mechanical, physics and thermal-hydraulic j

considerations. The fuel consolidation cage, with an envelope identical i

I to the standard assembly, is compatible with the present storage racks and standard spent fuel shipping casks.

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I Rods are withdrawn from the assembly cage and loaded directly into the storage skeleton cage, exactly as in a reconstitution process.

The grid-j defined rod positioning also lends itself to potentially automated disassembly / loading schemes.

For manual rod manipulation, it is a l

configuration similar to that with which plant personnel have experience.

The intended loading scheme is to fill each consolidated cage 1.o capacity before placing it into the rack for storage.

This limits the number of consolidated cages and spent fuel assemblies in a state of partial disassembly.

1.3 Schedule for Fuel Consolidation To accommodate the 72 assembly equilibrium cycle discharge, would require the disassembly and consolidation for storage of approximately two hundred I

assemblies per cycle. This is judged within the capabilities of a well designed, production oriented process for the proven single pin extraction and handling technique.

However, full benefit of consolidated storage would likely require at least limited automation of pin handling due to the already i

accumulated non-consolidated spent fuel assembly backlog.

sao s 2.0 THERMAL / HYDRAULIC ANALYSIS i

2.1 Pool Heat Load The Spent Fuel Pool (SFP) cooling system design is described in the Final Safety Analysis Report (FSAR), Section 9.8.

The pool heat load remains j

unchanged from the Proposed Change No. 28 analysis. Therefore, the capacity of the existing cooling system is adequate to maintain pool temperatures l

below design limits for the consolidated fuel storage concept.

l 2.2 Pin Cooling Analysis To assure that both consolidated storage bundles and fuel assemblies receive adequate local cooling, a detailed thermal / hydraulic analysis was conducted.

l The criteria for adequate cooling is the preclusion of nucleate boiling j

in the fuel rack under the maximum pool heat load conditions.

This analysis utilized the RETRAN computer program to model a selected row of fuel storage cells as shown in Figure 2.1.

Selection of a single row of assemblics, containing either consolidated storage bundles or fuel assemblies, provides i

a conservative thermal evaluation of any possible assembly loading or i

placement in the spent fuci pool.

l In the evaluation of the row of consolidated fuel bundles, it was conservatively assumed that the consolidated storage bundles were composed of fuel pins taken from fuel assemblies which had been removed from the reactor and cooled for 120 days.

(Initially, Maine Yankee does not intend to disassemble any fuel that has been cooled less than three years.) The operating history of these pins while in the reactor was assumed to be infinite.

Additionally, a row of freshly discharged fuel assemblies was also addressed.

These fuel assemblies were conservatively assumed to cool in the reactor for only four days following shutdown prior to being placed

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in the pool. Assembly average exposures of 35,000 MWD /MTU, a conservatively high burnup, were assumed.

The maximum pool heat load and bulk pool stabilization temperature occurs when a full core is discharged just after a shutdown from full power. Using Branch Technical Position APCSB 9-2 guidelines (Rev 1, 11/24/75), cooling times were established consistent with the above " worst case" full core discharge. To bound all future considerations, the full core discharge was assumed to fill the last available spaces in the pool with the assumption l

that all fuel is consolidated once it has cooled for 120 days.

Required cooling times for the full core discharge are based on not exceeding a bulk pool temperature of 1540F or a heat load of 22 x 106 BTU /hr, assuming a primary component cooling water temperature of 850F as described in the FSAR.

In the event that a full core discharge becomes necessary, pool temperatures will be monitored. The bulk pool temperature will be controlled by simply limiting fuel movement from the reactor to the pool.

The RETRAN model nodalization is shown in Figure 2.2.

Each fuel storage cell shown in Figure 2.I was modeled separately.

The upper bulk pool volume, the downcomer region between the edge of the racks and the pool-walls, and the region along the pool floor were addressed.

Conservative fluid friction 39)

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and form losses were assumed for all flow paths.

The channel region between each fuel storage cell was assumed to be heated

'by gamma deposition in the channel walls.

Results indicate that all bundles and assemblies receive adequate flow. 'The outlet: conditions.of the. hottest compacted bundle and the hottest discharged fuel assembly are shown in Table 2.1.

The maximum void fraction in the assemblies and the maximum void fraction in the space between fuel storage cells are zero.

Thus, boiling is not of concern as a source of moderator density variations in reactivity calculations.

2.3 Loss of Forced Pool Circulation lf all forced circulation cooling flow to the pool is lost, the large volume of pool water (363,000 gallons) provides a heat sink which allows time for corrective action. The minimum time for the pool water to reach the saturation temperature is 7.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, for the maximum design condition heat load of 22 x 106 BTU /hr.

In the event that normal cooling flow is lost, on either shell or tube side of the pool heat exchanger, back-up cooling is available. The plant fire protection system, with two 2,500 gpm pumps, can be connected to the pool heat exchanger shell side via two emergency cooling connections (designated ECC on FSAR Figure 9.8-1).

Heat exchanger tube side flow is provided by two (2) 750 gpm pumps.

In the event that one (1) pump is lost, the alternate pump is available to provide flow.

Additionally, emergency cooling connections are provided at the suction and discharge of the pumps to permit connection of a portable pump.

2.4 Loss of Pool Water Loss of pool cooling water as a result of failure of the coolant inlet and outlet lines exterior to the pool is addressed in the response to Question 9.15 in Amendment 18 to the FSAR.

It was determined that it is not possible to drain the fuel pool by siphoning as a result of a rupture of any line normally connected to the fuel pool.

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e 3.0 CRITICALITY ANALYSIS 3.1 Methodology and Assumptions l

Criticality calculations were performed using.rhe methodology described j

and utilized in Proposed Change No. 28 and its Supplements.

Proposed Change No. 28, supporting the conversion of the spent fuel racks at Maine Yankee to the high density, boron poison rack design received NRC approval in October, 1975.

The consolidated fuel storage bundle configuration is comparable to that l

considered in Proposed Change No. 28 with the significant difference being j

the placement of fuel pins in the consolidated cage.

The fuel pin pitch I

established by the storage grid structure was utilized in the criticality l

calculation and represents the nominal case. The criticality calculation g

assumed the mechanical uncertainties to be at " worst case" conditions.

l That is, the minimum water inside the flux trap and the minimum BORAL plate i

thickness are assumed. The nominal case also utilizes the 95/95 confidence level Boron-10 content as described in Brooks and Perkins report No. 540 (0.007235 atoms B-10/bn-cm). Additionally, a number of other conservative f

assumptions were used in this study, including the following:

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Unirradiated (fresh) 3.2 w/o U-235 fuel 11)

No soluble poison in the pool water I

111) No axial or radial leakage from the rack iv)

Calculations were done at the cost reactive pool temperature anticipated (68 F) v)

A calculational uncertainty of 3% was added to all results based on Monte Carlo calculations and the analysis of critical experiments.

The PDQ07 program, utilizing four energy groups with LEOPARD cross section input and an X-Y representation of the racks was used to determined K

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The accuracy of this representation was verified by comparison to KEN Monte Carlo calculations.

3.2 Nominal Conditions The sensitivity of rack Keffective to the storage lattice pitch and the density of the Boron-10 atoms is presented in Figures 3.I and 3.2.

No calculational uncertainty has been added to these values.

The K of the fully loaded nominal consolidated pin rack, including cal $bSSNr$$1 l

uncertainty, is 0.604, well below the value of the same racks loaded with I

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4 The reason for this large difference (0.604 versus standard fuel assemblies.

0.773) is the under moderation of the racks storing consolidated fuel.

is well below the values presented in Proposed Change No. 28 This result j

and/or the existing Facility License.

3.$U"Off-NominalConditions Of f-nominal conditions were reviewed for the consolidated pin storage rack j

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One area which needed to be considered was the KThenominalfuelpf$$his0.480 ef ive concept.

partially loaded cages stored in a rack.

To bound all possible loading configurations, the most reactive fuel pin pitch was determined. Therefore, a pin pitch of 0.65 was used

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I These cross sections were then in calculation of new fuel cross sections.

This configuration i

assigned to the fuel in the nominal 680F calculation.

f 0.760) than the is calculated to be 33.9% more reactive (or KThislimitingcaseforoff-nonfnaIive=dition ef con nominal design.

of the racks storing consolidated fuel under any loading j

the low Keffective configuration.

3.4 Accident Conditions A review of accident conditions for consolidated fuel storage reveals that 28 and/or the existing the information presented in Proposed Change No.

Facility License is bounding and need not be addressed in this submittal.

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4.0 MECHANICAL, MATERIAL, AND STRUCTURAL CONSIDERATIONS l

l Maine Yankee proposed to change the spent fuel pool storage design with Proposed Change No. 28 submitted to the NRC March 27, 1975.

After an

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extensive review, the NRC issued Amendment No. 11 to License No. DPR-36 g

which allowed Maine Yankee to replace the existing storage racks with high I

density poison racks.

l From a mechanical, material and structural standpoint, Maine Yankee has m-evaluated this design to reflect the consolidated fuel storage concept l

and at the same time incorporate the April 14, 1978 NRC guidelines for fuel l

2 storage modifications.,

4.1 Mechanical and Material Considerations l

The design concept, materials, manufacturing techniques, along with the quality assurance and surveillance programs were all presented in detail l

in the original submittal and subsequent supplements.

The consolidated fuel storage concept will not impact these items and, therefore, does not i

require their re-evaluation.

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4.2 Structural Considerations r

i Under the consolidated fuel storage concept the effective weight of a fuel assembly in a storage cell increases. Therefore, the pool and rack structues t

were re-evaluated as described below to ensure their structural integrity l

under design loads.

4.2.1 Spent Fuel Pool The 6 foot thick reinforced concrete floor of the Maine Yankee fuel storage pool rests on bedrock. The increase in weight supported by the floor l

utilizing the pin storage concept would be distributed in such a manner l

to assure allowable floor loadings.

The increase in weight is well within t

the static supporting capacity of the pool floor.

The storage pool will l

be reanalyzed incorporating the consolidated fuel concept to ensure conformance to the original Seismic I design requirements.

4.2.2 Cask Drop All of the conclusions reached concerning a dropped cask accident for the existing !!aine Yankee fuel storage pool are equally applicable for the new g

design.

I If the cask should drop while being handled over the pool, the pool floor

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would not be damaged to the extent that makeup capability could not be g

assured or that resultant flooding could cause critical systems to become I

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" Review and Acceptance of Spent Fuel Storage sud Handling l

Applications", NRC Publication dated April 14, 1978.

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-e inoperable.

Electrical interlocks prevent the yard crane block from passing over spent fuel. The proposed modification will not increase the area j

assigned to fuel storage.

Therefore, if a cask drop should occur, the cask will not contact spent fuel assemblies.

l Our review of the cask drop accident leads us to conclude, therefore, that f

there is no increase in the probability or consequencies of a fuel cask l

drop accident with the consolidated fuel storage design.

4.2.3 Storage Racks l

The original structural design analysis for the high density storage rac s has been re-evaluated incorporating the pin storage concept. The resulting l~

stress levels indicated that the present racks with minor modifications will accommodate the increased weight without exceeding any original dcsign basis.

In addition, Maine Yankee has reviewed the April 14, 1978 NRC guidelines suggested by the NRC concerning the structural analyses for new storage rack submittals.

Following the intent of these guidelines, the racks will be analyzed for seismic and impact loads.

I Seismic excitation will be imposed simultaneously in the N-S, E-W, and vertical directions.

Inertia, rack to rack impact, and fuel undle impact effects will be considered. Responses will be combined and load combinations made in accordance with the guidelines. All stresses shall remain within

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the applicable code limits. Maine Yankee will perform all analyses and make any required rack modification necessary to satisfy the intent of each l

guideline.

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5.0 RADIOLOGICAL EVALUATION

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Radiological considerations of consolidation and increased spent fuel storage 1978 NRC document.

g have been reviewed under the guidelines of the April 14, I'

5.1 Solid Waste The present annual quantity of solid radioactive wastes generated by the k

3 It is expected that the SFP purification system is approximately 20 ft.

increase in solid wastes as a result of the expanded capacity will not be j

significant, i.e., less than 10% increase.

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5.2 Gaseous Waste t

t Data regarding Krypton-85 measured from the fuel building ventilation system Airborne Kr-85 releases from MY vent g

are not available for Maine Yankee.

stack (which would include releases from the fuel building) for 1976 and 1977 are 68.91 curies and 17.81 curies respectively.

5.3 Dosage to Personnel 5.3.1 The principal radionuclides present in detectable concentrations 137Cs.

The concentrations of 58Co, 134Cs, and in the SFP water at MY are each of these detectable nuclides is approximately 3 x 10-3 C1/ml.

Dose rates as measured in the spent fuel pool area at MY are typically 5.3.2 on the order of 1 mrem /hr.

Extensive calculations were performed to bound the maximum expected increase in radionuclide concentrations and resulting The dose rates as a result of storage capacity increases for Yankee Rowe.

expansion ratio for Yankee Rowe was 2.3 compared to 1.6 for Maine Yankee.

The calculated incremental exposure rate due to increased radionuclide concentrations in the YR case was less than 1/10 mrem /hr based on successive Therefore, the incremental discharges of leaking fuel (1% failed fuel).

exposure rate in the SFP area due to the increased storage at MY is not expected to exceed 10% of the present level of approximately 1 mrem /hr.

5.3.3 Routine air sampling in the spent fuel pool area indicates no detectable activity, i.e., < 10-11 pCi/cc.

5.3.4 As stated above, routine airborne activity samples indicate no detectable activity in the SFP area.

Is is expected that any increase in airborne activity in the SFP area as a result of the expansion would not be significant, i.e., less than 10% of present levels.

Extensive fuel pools calculations were performed on the increase in activity in spent in conjunction with the expansion of fuel storage capacity at Yankee Rowe Using an extremely unlikely set of occurrences, i.e.,

and Vermont Yankee.

the successive unloading of full cores of leaking fuel (1%) into the SFP in radionuclide until it was filled to its enlarged capacity, the increment concentrations in the pool water due to the expanded capacity were calculated The associated to be less than 107, of the normal measured concentrations.

incremental exposure rate in the SFP area and at the site boundary would be extremely minimal.

h 3V7 390?;;:t 5.3.5 The changing of the SFP demineralizer resin at MY is accomplished remotely.

It is expected that any increase in man-rem associated with the operation as a result of the expansion will be minimal. The filter associated with the SFP demineralizer has required changing approximately four times since MY began power operation in 1972.

The annual man-rem exposure from the change out of the SFP filter is estimated to be less than 0.050 man-rem.

The increase in annual man-rem burden from more frequent changing of this filter is estimated to be less than 0.005 man-rem.

5.3.6 The buildup of crud products (e.g., 58 o, 60Co) along the sides of C

the spent fuel pool has not created any measurable difference in exposure rates along the sides of the pool.

However, if crud buildup were to become l

an exposure problem, underwater vacuum cleaning could be used to remove the buildup.

Such methods have been used successfully,to clean both the l

sides and bottom of the spent fuel pool at Yankee Rowe.

The cleaning j

operations in the Yankee Rowe SFP were conducted to prepare for in pool j

diving necessary to accomplish certain modifications and not because of an exposura problem at the pool edge.

5.3.7 The expected total annual man-rem received by personnel occupying the fuel pool area during a year which included extensive fuel sipping is l

estimated to be considerably less than 10 man-rem.

Levels of radiation j

in the SFP area at MY are typically on the order of 1 mrem /hr while airborne 5

levels are not detectable. As a result, the spent fuel pool area at MY does not present serious radiation protection problems.

Air and water sampling together with direct radiation surveys are routinely conducted i

on a weekly basis in the SFP area.

During refueling operations and whenever extensive work is done in the SFP area, continuous air sampling is performed in conjunction with more frequent Health Physics surveys.

4 5.3.8 Even with the effective expansion of the spent fuel pool permitted by fuel consolidation, it is not expected that any changes in the radiation 1

protection program at MY will be necessary to maintain exposures as low p

as reasonably achievable.

sa 390r%

ADDENDUM The following questions were transmitted to Yankee in regards to our letter, Reference (c). Answers to these questions contained meaningful information, and although sent as a draf t to the USNRC on March 8,1979, are now fornally submitted with this letter.

A.

The present tech spec restricting fuel handling for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after shutdown was based upon a fuel handling accident involving dropping a fuel assembly and damaging an equivalent number of freshly removed fuel pins as contained in fuel assembly.

There is no discussion of:

1.

Procedures to assure that accidental compaction of two freshly removed assemblies will not occur.

Such a situation would be an accident different from previously evaluated ones and would probably require a tech spec change.

RESPONSE

The entire fuel consolidation program will be executed under precise written and approved procedures.

All fuel movements will be controlled by serial number identification.

The characteristic Cerenkov radiation of fresh spent fuel assemblies makes them distinctly identifiable from fuel removed during a previous reload.

Compaction of freshly removed fuel assemblies in conjunction with a fuel handling accident which results in dropping a fuel assembly constitutes a three-fold accident which we need not address in our analysis.(1) 2.

The assessment of fresh spent fuel failure if a compacted old fuel assembly is dropped on stored fresh spent fuel to establish that the heavier missile will not damage more fuel pins than the i

equivalent of one normal freshly removed fuel assembly.

RESPONSE

The spent fuel assemblies in a storage rack sit well below the top grid structure of the rack itself.

In the unlikely event of a consolidated fuel assembly drop onto the rack, two events are possible.

First, with the rack superstructure above the spent l

fuel bundles, the most likely event is that the falling assembly r

would strike, vertically or at an angle, the rack upper grid I

structure and then topple over onto the grid structure.

The rack l

is designed to accept these loadings.

The second event, and i.ess (1) Thic situation would consist of three accidents: a) inadvertently loading a fresh fuel assembly into the f

consolidated cell, b) inadvertently loading 607. of a second l

fresh fuel assembly into the same consolidated. ell, and c) inadvertently dropping this consolidated fuel assembly.

l suo m i.

likely event, is that the falling consolidated assembly drops exactly centered onto a freshly discharged assembly in the storage rack.

The fuel pins in a standard assembly sit well below the upper tie plate and, therefore, the impact load is first carried by the five control rod guide tubes that provide the structural backbone of the standard fuel assembly, uitigating the loading on the pins.

The consequences of this accident are, therefore, clearly bounded by the " Design Basis Fuel Drop Accident" in which all 176 pins in a freshly discharged spent fuel assembly are conservatively assumed to upture.

3.

The potential exposure of the fuel handling workers to Part 20 limits if the compacted assembly is dropped.

This should include the isotopic inventory released and the procedures necessary to assure meeting Part 20.

RESPONSE

The potential exposure to fuel handling workers as a result of the fuel consolidation process is addressed in item a.4).

The limits of 10CFR20 are not applicable to a fuel drop accident situation.

4.

Occupational dose to the workers performing the compaction has not been discussed as required by Part 20.1(C) "as low as reasonably achievable." The experience to date as described is apparently that gained in reconstituting fuel elements by removing up to sixteen poison pins and substituting water filled pins.

This certainly offers some experience in both the process and the tools. To extrapolate this experience to compaction of 1192 uncompacted assemblies there should be discussion recognizing the following:

a)

Removal of 176 fuel pins per uncompacted assembly certainly increases the probability of individual pin failures when working with 40 foot tools. Pin failure could contribute to significant release of fission product particulates to the pool cooling water.

RESPONSE

The occupational dose to a worker as a result of an individual pin failure has been calculated.

The pertinent assumptions of Regulatory Guide 1.25 were followed with two important exceptions.

Exception 1:

The time between the end of the incore irradiation period and the beginning of compaction shall not be less than 120 days.

bONN k

Exception 2:

The rupture of only one individual pin will be analyzed.

j For conservatism it is assumed that the accident occurs without.

the knowledge of the personnel present, i.e.,

the fuel pin releases l

all the gap activity without fracturing the pin (the pin cracks j

Consequentiv, chere is no evacaation of i

but remains intact).

The doses given are two hour doses, since ventilation j'

the area.

exhaust of the building limits appreciable increase in the dose is assumed that the fuel pool ventilation system l

~

It after 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

i 1s operating with a purge rate from the building of 10,000 CFM.

f The occupational doses are presented in Table II.

Table II t

Occupational Doses from J

Pin Rupture Accident i

[

V 120 Day Decay 3 Year Decay Negli ible (<1x10-10 mrem) f 10 mrem Thyroid Dose Whole Body Gamma Dose 0.1 mrem 5x10- mrem Beta Skin Dose 123 mrem 104 mrea The above analysis does not take into Gecount variations in area dose rates as a result of an increase in particulates in the spent The control of particulates in the spent fuel pool fuel pool.

is effectively controlled by the spent fuel pool cleanup system.

If necessary, portable underwater vacuum equipment containing appropriate filtration can be employed to further reduce particulates in the spent fuel pool.

Equipment such as this has been used successfully at both Vermont Yankee and Yankee Rowefuel to help control particulate levels in their respective spent pools during special operations.

Experience to date has shown fuel pool that with this equipment, augmenting the normal spent cleanup system, the control of spent fuel pool particulate activity has not been a problem.

Removal of each of the spent pins from the uncompacted b) assembly and reassembly would be expected to release considerable activated crud from each pin to the cooling water of the pool.

RESPONSE

See answer to question A.4.c).

Experience would say that the source exposure to the fuel

c. )

handling crew during this process will be significantly above the normal 1 mr/hr water surface reading for normal spent fuel storage unless special cleanup systems are provided.

390"M

- M

i i

1 e

Items 1 and 2 above eili contribute considerably more activity in the water.

The fuel handling tools being moved some 14 feet in and out of the pit fer each fuel pin will cause significant turbulance in the pool suspending the particulates released and producing aerosols above the pools for worker exposure.

Procedures or steps to maintain fuel handling worker exposure as well as spent fuel pit cleanup filter maintenance worker exposure "as low as reasonably achievable" should be discussed.

g

[

RESP 0 NEE:

In the event that activated crud particulates released from removing the fuel pins and/or any associated production of aerosols above the pool should result in an increase in area dose rates, the plant Radiation Control Supervisor will evaluate the situation and recommend appropriate measures to insure that occupational doses are maintained as low as reasonably achievable.

Options would include the following:

1.

Augmenting the spent fuel pool demineralizer system with portable water filtration equipment positioned in close proximity to the source of the particulates.

2.

Providing localized air filtration equipment to augment the normal building ventilation system.

3.

Provide workers with respirators.

4.

Suspend the operations until area levels return to acceptable levels by use of normal systems. 't

B.

The Cost / Benefit Assessment fails to consider the man-rem exposare as a cost in determining the approach to take. We believe this is important in establishing that the radiation dose is "as low as reasonably achievable."

RESPONSE

A radiological cost benefit analysis was performed with the following assumptions:

1.

One fuel assembly could be disassembled and reassembled into a consolidated array in 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> by two men.

2.

It is assumed that one empty fuel cage could be cut up and packaged for disposal in one hour by two men.

3.

It is assumed that the radiation dose rate in the work area will be 1 mrem / hour including contributions f rom crud particulates released to the pool water.

I I

sao.we j

3$b.

l l

The total exposure incurred in consolidating the 1,545 assemblies over a period of about eight (8) years would be approximately 97 man-rem.

At $1000/ man-rem the cost in total radiation exposure is a small fraction of the total project cost.

j C.

Provide a commitment to store not more than 953 fuel assemblies, as l

stated in the SER to Tech. Spec. Amend. 11.

RESPONSE

I Maine Yankee accepts the interpretation in the SER of Tech. Spec.

l Amendment 11 as prohibiting storage in the spent fuel pool of more g

than 953 discharged fuel assemblies or the equivalent in fueled rods I

or weight of heavy metal without NRC action so permitting.

l D.

Define specifically, under all circumstances, how long the spent fuel will " cool" once removed from the reactor, before the compaction process l

is undertaken (i.e., provide clarification of the term " initially" i

on page 3 of letter dated 11/22/78).

I l

RESPONSE

As supported by the submitted analysis, fuel assemblies could be safely l

disassembled ~ ir storage after only 120 days of cooling.

However it l

is the intention of !bine Yankee, under administrative procedures, I

to initial' oonsolidate assemblies that have cooled for not less than l

three year.

until such time as this period of cooling restricts the consolidatio.. process.

I i

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TABLE 2.1 Maximum liaxf.um Void-Outlet Temp.

Enthalpy Fract ion (oF)

(BTU /lbc:)

Hottest Pin Bundle 207 175 0.

Hottest Fuel Assembly 220 188 0.

Saturation Condition at Outlet Height 234 202.4 3CC

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PROJECTED MAINE YANKEE FUEL POOL CAPACITY REQULRDfENTS TABLE 6.1 STORAGE ASSD!BLIES AUGMENTED START-UP ASSD:ELIES CELLS IN STORAGE CYCLE DATE DISCEARCED AVAILABLE STORAGE REQUIRED 3

6/17/77 73 592 361

~

4 9/78 72 520 433 5

3/80 72 448 505 6

6/81 73 375 578 7

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10/83 72 231 722 9

12/84 O)(2) 73 217

~736 59 10 1/86 72 736 131 11 2/87 (3) 72 736 203 12 3/88 73 736 276 13 4/89 72 736 348 14 5/90 72 736-420 j

l, 27 6/5 73 736 1361 1

28 7/6 72 736 1433 f

29 8/7 72 736 1505 I

L 30 9/8(4) 73 0

953 1505

{

(

a"

1) Assumes a thirteen month fuel cycle, inclusi'ee of refueling cutage.

2)

Loss of full core discharge withcut augtented storage.

3 9 0 %,*

3)

Loss of refuel capabilit/ without rugt.en.e:! stcrage.

4)

Current License L::pires October 21, 2008.