ML17254B025

Safety Evaluation Supporting Amend 65 to License DPR-18

ML17254B025
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Site:	Ginna
Issue date:	11/14/1984
From:	Office of Nuclear Reactor Regulation
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Download: ML17254B025 (26)

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UNITED STATES NUCLEAR REGULATORY COMMISSION

'lVASHINGTON,D. C. 20555 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AIIENDNENT NO.

65 TO PROVISIONAL OPERATING LICENSE NO.

DPR-18 ROCHESTER GAS AND ELECTRIC CORPORATION R.

E.

GINNA NUCLEAR POllER PLANT DOCKET NO. 50-244

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'R."Agocg'05000244,

'.'OR

'ated:

November 14, 1984

TABLE OF CONTENTS

1. 0 INTRODUCTION 2.0 DISCUSSION AND EVALUATION 2.1 Criticality Considerations 2.2 Spent Fuel Pool Cooling and Makeup 2.3 Rack Modification and Load Handling 2.4 Structural Design 2.5 'aterials 2.6 Occupational Radiation Exposure 2.7 Radioactive Waste Treatment 2.8 Radiological Consequences of Accidents Involving Postulated Mechanical Damage to Spent Fuel 3.0 OVERALL CONCLUSION 4.0 ACVNOWLEDGEMENT

1.0 INTRODUCTION

By letter dated April 2, 1984, as supplemented June 12, 1984, Rochester Gas and Electric Corporation (RGSE, the licensee) submitted an application to increase the storage'apacity of the spent fuel pool (SFP) by modifying the.

six west-most rack modules in the spent fuel pool.

By letters dated July 6, July 31, August 10, August 13, August 27, September 27, and October 23, 1984 the licensee provided additional clarification in response to the Nuclear Regulatory Commission (NRC) staff's requests for additional information.

This would be the second rerack for Ginna, the first being authorized by Amendment No.

11 on November 15, 1976 which increased the capacity of the SFP from its original capacity of 210 to 595 fuel elements.

The present amendment would authorize the licensee to increase the storage capacity of the SFP from the current capacity of 595 fuel assemblies to 1016 fuel assemblies with average planar enrichments no greater than 4.25 weight percent U-235.

At the present time, there are 332 spent fuel assemblies in the SFP.

The licensee also has 81 fuel assemblies stored at what was formerly the NSF at West Valley, New York.

These, assemblies will be transferred to the Ginna SFP by September 1985.

The licensee estimates that full-core reserve in the SFP would be lost following the 1987 refueling.

Since this date is earlier than the date a federal depository should be available for spent fuel [1998-Nuclear Waste Policy Act of 1982, Section 302(a)(5)j additional spent fuel capacity is needed.

I RG8E proposes to increase the storage capacity of the R.

E. Ginna storage pool by modifying six of the nine existing "flux trap" type storage racks currently in the storage pool to high density "fixed poison" type storage racks.

This change will double the storage capacity of the six modified racks from 420 to 840 storage cells.

The storage capacity of the three remaining "flux trap" type storage racks (176 storage cells). will remain unchanged.

Therefore, the total storage capacity of the pool will be increased from 595 to 1016 storage cells.

Since the pool will contain two different types of storage racks, it will be divided into two regions.

Region 1 will consist of the three "flux trap" type storage racks and Region 2 will consist'of the six modified "fixed poison" type storage racks.

Previously, RG&E proposed and received NRC staff approval for a possible increase in the U-235 enrichment of the fuel assemblies from 3.5 to 4.25 weight percent.

The licensee also received approval for the use and storage of the Westinghouse Optimized Fuel Assemblies (OFA).

Table 2-1 of the 11censee's submittal of April 2, 1984 shows that the Region 1 storage racks are capable of safely storing the previously existing R.

E. Ginna fuel assemblies as well as the Westinghouse OFA.

However, prior to storing fuel assemblies in the new fixed poison (Region 2) storage

racks, the fuel assemblies must meet the following conditions:

l.

60 days must have elapsed since the reactor reached hot shutdown.

2.

The combination of the assembly average burnup and the initial U-235 enrichment must be such that the point identified by the two parameters on Figure 5.4-2 of the April 2, 1984 submittal is above the line applic-able for the particular fuel assembly design.

This will assure that k ff for the stored fuel is equal to or less than 0.95.

To assure that the burnup has been properly established, the licensee indicates that the burnup of each assembly will be established using the Nuclear Fuel Accountability Code that was started in 1970.

This code establishes the isotopic content of the fuel and other parameters such as burnup.

This information along with the curves on Figure 5.4-2 of the submittal will be used t'o determine if an assembly can acceptably be stored in Region 2.

The seismic analysis of the modified spent fuel storage racks incorporated higher loadings which would be expected for the case of rod consolidation.

However, the licensee request of April 2, 1984 as supple-mented June 12, 1984 requested approval only for storage of unconsolidated fuel.

A Notice of Consideration of Issuance of Amendment to License and Proposed No Significant Hazards Consideration Determination and Opportunity for Hearing related to the requested action was published in the Federal Reoister on July 27, 1984 (49 FR 30261).

Ho requests for hear>ng and no pou P>c comments were received.

2.0 DISCUSSION AND EVALUATION 2.1 Critica 1it Considerations The storage racks have been analyzed for two groups of fuel assembly designs.

The first group consists of all fuel delivered prior to 1984 and incorporates all Exxon and Westinghouse HIPAR designs used at Ginna containing no more than 39.0 gm U-235 per axial cm (3.5 weight percent U-235).

The second group consists of the Westinghouse OFA design delivered to Ginna beginning in February 1984 containing no more than 41.9 gm U-235 per axial cm (4.25 weight percent U-235).

The Region 2 design consists of six racks, each containing 140 stain-less steel cells for a total of 840 fuel assembly storage locations.

There is a 8.43 inch center-to-center spacing between assemblies and a

neutron absorbing material, Boraflex (Ref. I), is attached to the stainless steel.walls of each storage cell.

Boraflex consists of boron carbide powder in a rubber-like silicone polymeric matrix.

The minimum boron-10 density in the Boraflex is 0.020 gm/cc.

The design is intended to contain any of the Exxon or Westinghouse HIPAR or OFA 14x14 fuel assemblies used in Ginna with an initial enrichment of up to 4.25 weight percent U-235 at an assembly average exposure of 30,000 Hl4D/tlTU.

For lower initial enrichments, the amount of exposure required for storage in Region 2 will be reduced.

For 3.00 weight percent U-235, for example, it is 15,960 tND/NTU and for 1.75 weight percent U-235, even fresh fuel can be stored in Region 2 as seen from Technical Specification Figure 5.4-2.

The criticality aspects of the storage of Westinghouse and Exxon fuel assembly designs used at Ginna in the burnup-dependent region (Region 2) of the spent fuel storage pool have been analyzed using the PDg-7 computer program for reactivity

-determination with four energy group neutron cross sections generated by the LEOPARD program as modified by Pickard, Lowe and Garrick, Incorporated (PLG).

These codes have been bench-mar ked against both Westinghouse and Battelle Pacific Northwest Laboratories critical experiments with pellet diameters, water-to-fuel ratios and U-235 enrichments encompassing those in the Ginna fuel rack design.

In addition, a series of Pu0 UO critical experiments were analyzed to determine the abcurac$ of calculations of systems containing significant amounts of plutonium mixed with UOZ and, therefore, the accuracy of reactivity calculations for irradiated fuel.

These latter results led to the conclusion that the calculational model is capable of determining k

of the Ginna spent fuel racks with a combined LEOPARD/PDD-7 ms' bias of +0.0031 and a 0.0186 ak uncertainty corresponding to a

95 percent probability at a 95 percent confidence level (95/95).

In order to establish burnup criteria for storage in Region 2, a constant storage rack infinite multipTication factor (with minimum post-shutdown fission product, inventory) contour is constructed as a function of burnup and initial enrichment using LEOPARD and PDg-7.

Since the calculations use the basic cell to calculate 'the reactivity of an infinite array of uniform spent fuel racks and axial leakage is not accounted for, the calculated multiplication factor is, in reality, K~, which will be larger (more conservative) than k ff This contour is based on a high enrichment endpoint of 4.23 >veight percent U-235 and 30,000 hlWD/HTU as shown in the attached Figure 5.4-2 from the Ginna Technical Specifications.

This is representative of the Westinghouse OFA fuel delivered after January 1,

1984.

A similar curve for Exxon and Westinghouse HIPAR fuel delivered prior to 1984 is also shown.

2.1.2 S ent Fuel Rack Stora e

The basic rack cell at 20'C, 4.25 weight percent U-235, and 30,000 HWD/MTU results in a reactivity of 0.9072.

Including all the appropriate calculational biases and 95/95 uncertainties results in a maximum reactivity change of 0.0390, giving a maximum reactivity of 0.9462, which meets the staff acceptance criterion of less than or equal to 0.95.

For lower enrichments with the same computed multiplication factor, the amount of exposure will be reduced, reducing the reactivity uncertainties due to depletion of fuel and buildup of fission products.

The total uncertainty is, therefore, reduced making the rack cell reactivity calculated at 4.25 weight percent U-235 and 30,000 t1WD/NTU conservative for all lower enr ichments.

For additional conservatism, a constant multiplication of 0.9050 is used to

.generate the final burnup verus enrichment curves in the Technical Specifications.

2.1.3 Accident Anal ses Since the maximum possible reactivity of the Region 2 spent fuel rack is based on infinite array calculations both laterally and vertically, the effect of a dropped fuel assembly on top of the rack would not exceed the calculated maximum reactivity value.

In addition, the racks are designed to prevent a dropped fuel assembly from occupying a position other than a normal fuel storage location.

Procedures exist to assure that assemblies discharged from the core are first moved to Region l.

After the refueling operation is complete and the suitability of each spent fuel assembly for movement and storage into Region 2 is verified, this fuel will be moved into Region 2.

Therefore, administrative procedures exist to help preclude a fuel misloading event.

However, even if it occurs, the spent fuel storage pit is filled with borated water at a concentration sufficient to maintain k

< below 0.95.

NRC review policy permits credit for this 5Vron.

2.1.4 Technical S ecifications The staff concludes that the modifications to the Ginna Technical Specifications submitted by licensee letters dated april 2,.1984, and June 12, 1984 are acceptable to allow operation with the proposed expansion of SFP storage capacity.

2.1. 5 Concl us ions The staff concludes that the proposed storage racks meet the requirements of General Design Criterion 62 as regards critical'ity.

This conclusion is based on the following considerations:

(1)

Acceptable calculation methods which have been verified by comparison with experiment have been used.

(2)

Conservative assumptions have been made about the enrichment of the fuel to be stored and the pool conditions.

.(3)

Credible accidents have been considered.

(4)

Suitable uncertainties have been considered in arriving at the final value of the multiplication factor.

(5)

The final effective multiplication factor value meets the staff acceptance criterion, 2.2 S ent Fuel Pool Coolino and Nakeu 2.2. 1 Deca Heat Load and Spent Fuel Pool.Coolin S stem In 1981, the staff reviewed and approved a proposed SFP cooling system modification for Ginna (Ref. 2).

This modification will be implemented in 1986, and will consist of the addition of a new cooling loop in parallel with the existing loop which is sized to accommodate the maximum normal and abnormal heat loads should the storage capacity be increased to 1360 fuel assemblies at some future date.

Since the present proposal calls for an increase in the total storage capacity of the pool to 1016 fuel assemblies, the staff concludes that the previously approved SFP cooling system will acceptably handle the maximum normal and abnormal heat loads for this'proposed expansion.'he modified SFP cooling system could accommodate the full core discharge and normal refueling heat loads through the year 2010.

On those occasions where a full core discharge takes

place, the licensee has committed to incrementally increase the decay time in the reactor vessel from 8 days in the year 1981 to 14 days in the year 2010 in order to assure that the maximum pool water temperature will not exceed the Technical Specification limit of 150'F.

The licensee has also indicated that fuel consolidation may be proposed in the future, however this is not included in the currently proposed fuel pool expansion and is not a part of the staff review of the SFP cooling system adequacy.

Based on the above, the staff concludes that the maximum normal and abnormal heat loads resulting from the proposed expansion will not exceed the anticipated heat loads used in the previous evaluation of the SFP cooling system modifications and, therefore, the SFP cooling system is acceptable.

2.2.2 Boiloff Rate and Makeu S stems

-As indicated in the SFP cooling system discussion

above, the decay heat loads will not exceed those previously considered and approved during the pool cooling system modification review in 1981.

Therefore, the staff concludes that the associated boiloff rate also will not exceed that which was previously accepted.

Similarly, the staff concludes that the demands on the pool water makeup system will not exceed those previously reviewed and approved and, therefore the makeup capability is acceptable.

2.2.3 ~21 2

At the time of the previous storage rack expansion review, the licensee provided an analysis to determine the difference in temperature of the water exiting from the top of the storage cells with respect to the corresponding water saturation temperature.

It was assumed in this analysis that a recently discharged batch of fuel assemblies were grouped together in

.the original storage cells (Region 1 arrangement) in a location as far away from the cooling system cold water inlet as possible.

Under these conditions, it was found that the temperature of the water exiting from the hottest fuel assembly is less than 155'F and the corresponding saturation temperature is over 235'F.

There is therefore a margin of about 80'F to prevent local boiling from occurring.

2.2.4 In the case of the modified storage racks (Region 2 arrangement),

fuel will not be moved into these storage racks until at least 60 days of decay has taken place.

Therefore, the decay heat load would have decreased to about 60 percent of that of recently discharged fuel.

This combined with the enlarging of the flow holes in the former water boxes indicates that the exit "f

temperature of the water from the Region 2 storage cells will be less than previously reviewed and approved for the Region I storage cells.

Therefore, the staff concludes that adequate margin to local boiling has been demonstrated for the Region 2

storage racks and they are therefore acceptable in this regard.

C'onclusion The staff has reviewed the spent fuel cooling and makeup as it

-relates to the second SFP expansion program for R.

E. Ginna and concludes the following:

(1)

The resulting decay heat loads in the pool are less than those assumed in the proposed SFP cooling system modification which was approved by the staff in 1981.

(2)

The boiloff rate assuming the loss of-all pool cooling is less than that assumed in the staff's 1981 review, and therefore the makeup systems previously approved by the staff will provide assurance that the fuel will not be uncovered.

(3)

The margin between the temperature of the water exiting from the Region 2 storage cells will be approximately 80'F less than the corresponding pool water saturation temperature, thus providing adequate assurance that local boiling will not occur.

In summary, based on its review, the staff concludes that the R.

E. Ginna proposed second SFP expansion meets the guidelines of SRP Section 9.1.3, and is therefore, acceptable.

2.3 Rack Modification and Load Handlin The steps and procedures required to accomplish reracking the SFP will be developed so as to eliminate the need for carrying loads over stored spent fuel and will ensure that reasonable protective measures will be taken to preclude load drops during reracking.

Modified Storage Racks RG&E engaged US Tool 8 Die to perform the mechanical, structural and material analysis of the modifications to the existing Hachter storage rack's.

The nuclear analysis was performed by

Pickard, Lowe, and Garrick Inc.

The rack modification program will consist of sequentially removing and modifying one storage rack at a time.

The steps involved in the modifications will be as follows:

(2)

(3)

(5)

(6)

The 332 fuel assemblies presently in the popl will be moved as far as practical from the rack to be removed.

A diver will remove the four mounting bolts that attach the storage rack to its support base.

Using the lifting rig, the storage rack will be raised clear of the pool surface and partially decontaminated using high pressure water before it is moved to the decontamination pit..

Following additional decontamination in the decontamination pit, the guide funnels and guide angles will be cut free of -the storage rack.

The existing lifting attachments will be removed, and four modified bottom plates with the new lifting slots will be installed.

The flow holes in the bottom plates will be enlarged and k inch bottom plates will be installed in the former water boxes.

(io)

(11)

(~2)

The right-angled poison assemblies will be installed and welded in place nn each storage cell.

Divers will install appropriate shims at the four corners of the support base in the pool.

The existing jack screws on the racks will be retracted so that the weight of the rack will bear in the support base shims.

The modified rack will be lifted, transported and lowered onto the support base shims.

The above steps will be r epeated for the remainirg five, storage racks to be modified.

Al'1 seismic support between the rack bases will be removed.

The right-angled poison assemblies to be installed in the storage cells will consist of a nominal 0.062 inch thick preformed sheet of stainless steel and two nominal 0.075 inch thick-by-7 -5/8 inch wide strips of Boraflex are'andwiched between the cell walls and the preformed stainless sheets.

This installation will reduce the internal dimensions of the storage cells from a nominal 8.280 x 8.280 to 8.143 x 8.143 inches.

During the 1977 spent fuel expansion when the "flux trap" type storage racks were installed, the staff determined that the racks met seismic Category l criteria.

Since RGSE proposes in this submittal to convert these racks to provide, twice the number of storage cells, the effective weight of the stored fuel in a given rack will be doubled.

Further, RGIIE has

-requested the staff to evaluate the adequacy of the storage racks if at sometime in the future they decide to implement a

~

rod consolidation program.

This will effectively increase the weight of the stored fuel assemblies over that previously approved by the staff during the 1977 review.

A separate structural evaluation of the seismic design capability of the racks which accounts for the increased weight of the stored fuel is reported in Section 2.4 of this Safety Evaluation.

Following the installation of the right-angled poison assemblies in each storage cell, a gauge will be inserted into the cell to verify that the fuel assemblies will not experience unacceptable frictional forces during their insertion or withdrawal.

llestinghouse guidance in this regard states that a drag force of 50 pounds is not to be exceeded.

Further, based on previous experience, the licensee stated that a drag force of approximate-ly 400 pounds is required before damage to the fuel assemblies will occur.

RGINE has committed to evaluate all drag forces in excess of 50 pounds on a case-by-case basis.

The licensee has stated that in no case will the developed drag force be accepted if it is sufficient to threaten the integrity of a fuel assembly.

The modified storage racks will have an estimated weight of 28,000 pounds each.

From this, and the sturdiness of the rack construction, the staff concludes that the vertical frictional force of 400 pounds exerted by the fuel handling crane will not cause damage to the storage rack.

Further, as a result of having removed the lead-in funnels on the storage cells, the licensee has committed to provide a portable lead-in funnel to aid the operator in properly aligning fuel assemblies during their insertion in the Region 2 storage cells.

The gaps between the storage racks are a small fraction of the cross sectional

'dimensions of a fuel assembly.

Therefore, the staff concludes that a fuel assembly cannot inadvertently be placed in any location other than the designated storaoe areas within the lattice array of the racks.

In RG&E's letter of June 12, 1984 the licensee indicated that fuel rod consolidation may be proposed at some time in the future.

However, the licensee requested that the staff not consider consolidation as part of the rack modification and load handling review.

Itith regard to a postulated vertically dropped fuel assembly

accident, the licensee states that if the assembly were to drop 14 feet onto a flat surface, the resulting impact stresses would be acceptably low and no significant damage would be expected in any fuel rods.

If the dropped assembly were to strike a sharp object, the licensee conservatively assumed that one row of fuel rods would fail.

In the case of a tipped fuel assembly

drop, the resultant kinetic energy would be much less than for the vertical drop.

Therefore, aside from the postulated damage to a

-row of fuel rods, the licensee concluded that the crush strength of the storage cells will protect the stored fuel from damage from dropped fuel assemblies.

Based on the above, the staff concludes that the Region 2

storage racks will adequately support and protect the stored fuel assemblies and are, therefore, acceptable.

2.3.2

~22 II d11 There are currently 332 fuel assemblies stored in the pool.

The licensee has indicated that for each reracking operation, the stored spent fuel assemblies will be moved away from the area where the load handling operations are to take place in order to minimize the consequences should a load drop occur and minimize the radiological exposure to the divers who attach the lifting device to the storage racks.

The load handling operations associated 'with reracking will be conducted in accordance with Section 5.1.1 of NUREG-0612, "Control of Heavy Loads at Nuclear Power Plants" as it relates to safe load paths, procedures, crane operator training and qualifications, and crane inspection and maintenance.

Further, the special 'lifting device interposed between the storage racks and the crane hook wil1 consist of two redundant spreader

bars, slings, and vertical lifting adapters.

Both spreader bars are located such that the center of gravity of the storage rack is directly below the crane'ook.

'Therefore, should a failure occur in one of the spreader

bars, the load will remain stable and would not swing.

The calculated stresses for the special lifting device are also less than that prescribed in the guide-lines of ANSI 14.6.

As in the previous SFP expansion, the auxiliary building crane will be used for handling the storage racks.

This crane was procured to EOCI-61 specifications.

Based on the manufacturer's (Whiting Corporation) evaluation of the crane provided by the licensee in response to the criteria

/

of NUREG-0612, the staff concludes that the crane meets the intent of Guideline 7 of NUREG-0612, Section 5.1.1.

The range of travel of the crane is such that the hook cannot be placed directly over the center of gravity of the two most westward storage racks.

To enable these storage racks to be 1'ifted vertically without encountering mechanical interference with the adjacent storage

racks, a chainfall or cable winch will be attached to the main hook block.

The chainfall ot winch will

-be anchored by means of a temporary holding beam attached to three auxiliary building columns in a fashion'similar to that previously done during the 1976 reracking operations.

The licensee acknowledged that this operation may cause some accelerated wear of the auxiliary building crane cable drum.

However, due to the limited time of use for conducting this operation, the wear should not become significant.

Further, the licensee states that the cable drum is due to be replaced as part of the overall upgrade of the crane to satisfy the criteria of NUREG-0554.

Based on the above, the staff concludes that the reracking operations will be performed in accordance with the guidelines of NUREG-0612 as applicable and that all reasonable measures will be taken to preclude unacceptable consequences in the unlikely event of a load drop.

Therefore, the described reracking operations are acceptable.

2.3.3 Conclusion The staff has reviewed the proposed modification of the SFP racks and load handling as it relates to the SFP expansion program for R.

E. Ginna and concludes that the modified racks are designed such that:

(1)

The maximum uplift friction forces developed by the crane will not damage the storage racks.

(2)

The postulated dropping of a fuel assembly v:ill not lead to unacceptable consequences.

(3)

The a'rrangement of the storage racks within the pool is such that it is not possible to =inadvertently insert a fuel assembly into a nondesignated space

-within the storage rack array.

(4)

The racks satisfy seismic Category I criteria for

~

unconsolidated fuel.

(5)

Adequate load handling precautions will be taken during the reracking operations.

In summary, based on its review, the staff concludes that Ginna proposed SFP expansion meets the guidelines of SRP Sections 9.1.2,

9. 1.4, and 9.1.5, and is therefore, acceptable.

2.4 Structural Desi n

The Safe'ty Evaluation (SE) of structural aspects of the proposed modification is based on a review performed by NRC's consultant, Franklin Research Center (FRC).

The FRC Technical Evaluation Report-(TER)

C5506-531 is appended to this SER as Appendix A.

2.4.1 Descri tion of the S ent Fuel Pool and Racks The spent fuel storage pool is designed for the underwater storage of spent fuel assemblies, failed fuel cans and control rods after their removal from the reactor.

The pool is-constructed of reinforced concrete having thick walls and is Class I seismic design.

The slab of the pool is founded on bedrock.

In addition, the entire interior basin face is lined with stainless steel plate, The racks are stainless steel egg-crate structures.

Original design of the racks is composed of three major components.

(1)

The rack modules, which are rectangular arrays of cells of which one out of two are storage cells.

The others are water boxes.

(2)

The support bases, on which the rack modules rest, are rectangular construction of I beams.

(3)

Seismic support between the bases and the pool walls provides a means to transmit horizontal loads from the racks to the walls.

Structural modifications for the proposed amendment are as fol 1 ows:

(l)

Using a special cutting machine remove guide tunnels and guide angles over the water bases so that spent fuel assemblies can be stored.

(2)

Remove all (both Region 2 and Region

1) seismic supports between the rack base and the pool walls.

The seismic analysis was performed for both the standard and consolidated fuels.

2.4.2 A

licable Codes, Standards and S ecifications

-Load combinations and acceptance criteria were compared with those found in the "Staff Position for Review'and Acceptance of Spent Fuel Storage and Handling Applications" dated April 14, 1978 and amended January 18, 1979.

The existing concrete pool structure was evaluated for the new loads in accordance with the requirements in the Ginna FSAR.

2.4.3 Loads and Load Combinations Loads and load combinations for the racks and the pool structure were reviewed and found to be in agreement with the applicable portions of the Staff Position.

2.4.4 Seismic Loads Seismic loads for the rack design are based on the original design floor acceleration response spectra calculated for the plant at the licensing stage.

The seismic loads, were applied to the model in three orthogonal dir'ections.

Damping values for the seimsic analysis of the racks were taken in accordance with the Regulatory Guide 1.61.

Rack/fuel bundle interactions wer e-considered in the structural analysis.

2.4.5 Desi n and Anal sis Procedures (1)

Design and Anal sis of the Racks - Horizontal seismic ana ysss was pervorme using t e time history method.

This accounts for the non-linearities inherent in the spent fuel storage racks which include fuel-to-rack wall impacts, rack sliding, and vertical impact due to rack tipping.

The vertical seismic analysis was performed using spectra method.

The vertical reaction loads were combined with the horizontal seismic loads using the square root of sum of the squares method.

Calculated stresses for the rack components were found to be well within the allowable limit.

The racks were found to have adequate margins against sliding and tipping.

An analysis was conducted to assess the potential effects of a dropped fuel bundle on the racks and results were considered satisfactory.

An analysis was conducted to assess the potential effects of a stuck fuel assembly causing an uplift load on the racks and a corresponding downward load on the 1'ifting device as well as a tension load in the fuel assembly.

Resulting stresses were found to be within acceptance limits.

-(2)

Anal sis of the Pool Structure - The floor of the ss a sta>n ess stee one

, 3-foot thick, reinforced concrete slab.

The slab is founded on bedrock.

The structure of the pool was evaluated for the original FSAR and again for the floor loads associated with subsequent rack replacement.

Because the rack will be modified to a free-standing

design, only the increased concrete bearing stresses on the floor were evaluated.

These were found to be acceptable.

2.4.6 Conclusion The staff concludes that the proposed racg installation will satisfy the requirements of 10 CFR Part 50, Appendix A, General Design Criteria 2, 4, 61, and 62, as applicable to structures, and is therefore acceptable.

2.5 Yaterials The staff has reviewed the compatibility and chemical stability of the materials (except the fuel assemblies) wetted by the pool water.

The only new material or components to be added during the proposed modification are the nuclear absorber strips..

The existing spent fuel racks to be adapted in the proposed expansion are constructed entirely of Type 304 stainless

steel, except for the nuclear poison material; The existing SFP liner is constructed of stainless steel.

The high density spent fuel storage racks will utilize Boraflex sheets as a

neutron absorber.

The spent fuel storage rack configuration is composed of individual storage cel.ls interconnected to form an integral structure.

The major components of the assembly are the fuel assembly cells, the Boraflex material, and the L-shaped stainless steel sheaths.

During modification, the flow holes in the bottom plates of the existing fuel storage cells will be enlarged and additional bottom plates will be added to the former water boxes.

Each cell will contain an. insert consisting of two Boraflex sheets at right angles to one another and an L-shaped stainless steel insert to hold them in place.

The Boraflex absorber will not be sealed within the storage cell and f

vent paths for any gas generated during exposure will be available to the pool.

The pool contains oxygen saturated demineralized water containing boric. acid.

The water chemistry control of the SFP has been evaluated and reported in the SER supporting Amendment No.

11 to the operating license and found to meet NRC recommendations.

The increased storage capacity of the pool does not change this evaluation.

2.5.1 Evaluation

-The pool liner, rack lattice structure and fuel storage tubes are stainless

steel, which is compatible with the storage pool environment.

Boraflex has undergone extensive testing to study the effects of gamma irradiation in various environments, and to verify its structural integrity and suitability as a neutron absorbing material.

The evaluation tests have shown that the Boraflex is unaffected by the pool water environment and will not be degraded by corrosion.

Tests wer e performed at the University of Michigan (Ref. 3), exposing Boraflex to 1.103 x 10 rads of gamma radiation with substantial concurrent neutron flux in borated water.

These materials are being used in many operating

SFPs, The licensee committed to monitor the SFP surveillance program at Point Beach, which the staff has found acceptable.

The materials in the Point Beach program are identical to the materials in this SFP and thus the monitoring of this surveillance is acceptable to meet the surveillance program requirement.

2.5.2 Conclusion From the evaluation as discussed

above, the staff concludes that the corrosion that will occur in the spent fuel storage pool environment should be of little significance during the life of the plant.

Components in the spent fuel storage pool are constructed of alloys which have a low differential galvanic potential between them and have a high resistance to general corrosion, localized corrosion, and galvanic corrosion.

Tests under irradiation and at elevated temperatures in borated water indicate that the Boraflex material will not undergo significant degradation during the expected service life.

The staff further concludes that the environmental compatibility and stability of the materials used in the expanded spent fuel storage pool is adequate based on the test data cited above and actual service experience in operating reactors.

The staff has reviewed the surveillance programs at the reactors cited by the licensee and concludes that the monitoring of materials in these spent fuel storage pools will provide eeason-able assurance that the Boraflex material will continue to perform its function for the design life of the SFP.

The materials surveillance program in these cited units will reveal any instances of deterioration of the Boraflex that might 1'ead to the loss of neutron absorbing power well before comparable radiation exposures have been reached in the licensee's spent fuel racks.

The staff does not anticipate,

-however, that such deterioration will occur.

The monitoring program will ensure that in the unlikely situation that the Boraflex will deteriorate in the SFP environments, the licensee and the NRC will be aware of it in sufficient time to take corrective action.

The staff, therefore, finds that the commitment to follow the monitoring program at the other PMR SFPs and the selection of appropriate materials of construction by the licensee meets the requirements of 10 CFR Part 50, Appendix A, Criterion 61, having a capability to permit appropriate periodic inspection and testing of components.

The staff also finds that the licensee meets Criterion 62, preventing criticality by maintaining structural integrity of components and of the boron poison.

The staff therefore concludes that the materials to be used in the proposed modification are acceptable.

2.6 Occu ational Radiation Ex osure The staff has reviewed the licensee's plan for the modification of the Ginna SFP racks with respect to occupational radiation exposure.

The licensee estimates that the exposure for this operation will be approxi-mately 78 man-rems.

This estimate is based on the licensee's detailed breakdown of occupational exposure for each phase of the modification.

The licensee considered the number of individuals performing a specific job, their occupancy time while performing this job, and the average dose rate in the area where the job is being performed.

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

Evaluation One potential source of radiation is radioactive activation of corrosion products, termed "crud."

Crud may be released to the pool water because of fuel movement during the proposed SFP rack modifications.

This could increase radiation levels in the vicinity of the pool.

During refuelings, when the spent fuel is first moved into the fuel pool, the addition of crud to the pool water from the fuel assembly and from the introduction of primary coolant to the pool water is greatest.

However, the

licensee, based on experience from plant's performing similar modifications, does not expect to have significant releases of crud to the pool water during modification of the SFP racks.

In addition, the purification system for the pool, which has maintained radiation levels in the vicinity of the pool at low levels during normal operations, will be operating during the modification of the SFP racks.

The staff has evaluated the licensee's proposed crud reduction program in the SFP and finds it acceptable.

-The presently installed racks will be individually lifted from the SFP and while suspended over the SFP, will be rinsed using high pressure water to remove any loose radioactivity.

The racks will then be moved to a receiving area for modification.

The licensee has proposed decontaminating most of the components removed from the racks during the modification and then disposing the clean material as industrial waste.

Material that cannot be decontaminated will be packed into drums and disposed of as'ormal radioactive waste.

The disposal methods used will ollow ALARA guidelines.

Divers will be used during the SFP rack modification. "The licensee has developed specific procedures using the recommendations of Regulatory Guide 8.8 to ensure that doses to the divers will be within the requirements of 10 CFR Part 20 and ALARA guidelines.

The ALARA procedures for divers include:

reshuffling of the spent fuel to provide zones around the divers'ork areas where no fuel will be stored; radiation survey after the fuel is reshuffled to map radiation zones; instruction to divers on their travel limits within the pool; and constant monitoring of 'divers'adiation dose by the use of remote readout dosimetry.

2.6.2 Conclusion The staff's evaluation of Ginna's proposed SFP rack modification included a review of the manner in which the licensee will perform..the modification, the radiation protection

program, including the use of area and airborne radioactivity monitoring, and the use of relevant experience from other operating reactors that have performed similar SFP modifications.

Based on this review, the staff concludes that the Ginna SFP rack modification can be performed in a manner that will ensure as low as is reasonably achievable (ALARA) exposures to workers.

The staff has, estimated the increment in onsite occupational dose during normal operations after the pool modification resulting from the proposed increase in stored fuel assemblies.

This estimate is based on information supplied by the licensee for occupancy times and for dose rates in the spent fuel area from rad>onuclide concentrations in the SFP water.

The spent fuel assemblies themselves contribute a

negligible amount to dose rates in the pool area because of the depth of water shielding the fuel.

Based on present and projected operations in the SFP area, the staff estimates that the proposed modification should add less than one percent to the total annual occupational radiation exposure at the plant.

The small increase in radiation exposure should not affect the licensee's ability to maintain individual occupational dose to ALARA levels and within the limits of 10 CFR Part ZO.

Thus,

-the staff concludes that storing additional fuel in the SFP will not result in any significant increase in dose received by workers..

2.7 Radioactive Waste Treatment The plant contains radioactive waste treatment systems designed to collect and process the gaseous, liquid, and solid wastes'that might contain radioactive material.

The radioactive waste treatment systems were evaluated in the SER for the full-term operating license dated October 1983 (NUREG-0944), in support of the issuance of Operating License No. OPR-18.

There wi'11 be no change in the radioactive waste treatment systems or in the conclusions given regarding the evaluation of these systems because of the proposed modification of the SFP racks.

The staff evaluation of the radiological considerations supports the conclusion that the proposed installation of new spent fuel storage racks. at Ginna is acceptable because the conclusions of the evaluation of the radioactive waste treatment

systems, as found in the Ginna SER for the full-term operating license, are unchanged by the modification of spent fuel storage racks.

2.8 Radiolo ical Conse uences of Accidents Involvin Postulated Mechanical amaae to ent ue For evaluation of accidents involving the SFP, three types of accidents were considered; a cask drop or tip, a tornado missile impact, and a

fuel assembly drop while handling fuel.

2.8.1 Cask Dro /Ti Accidents Technical Specification

3. 11.6 states that "The spent'fuel shipping cask shall not be carried by the auxiliary building

'crane, pending the evaluation of the spent fuel cask drop accident,and the crane design by RGSE, and HRC review and approval."

Since the shipping cask cannot presently be 2.8.2 carried by the auxiliary building crane by this administrative

control, because the crane design evaluation has not yet been completed by the staff, a cask drop/tip accident is precluded for the proposed Technical Specification amendment.

Tornado Missile Accidents The design values for tornado wind speed and missile characteristics were those established in the staff review of Systematic Evaluation Program (SEP) Topics III-2, Mind and Tornado Loadings, and III-4.A, Tornado Missiles.

The design missile is stated to be a 1490 lb wooden pole, 35 feet in 1'ength and 13.5 inches in diameter, which could impact the racks with a vertical velocity of 70 ft/sec.

The staff judges that the worst position for impact of this missile would be

-that centered on a fuel storage location where, because of the 13.5 inch missile diameter compared to a diagonal dimension of the spent fuel storage box of 11.9 inches, a total of nine fuel storage cells could be damaged in the reracked six sections of the SFP.

This relative impact orientation of missile and storage cell configuration would have a low likelihood of occurrence, however.

It is thus judged that a conservative estimate of damage to stored spent fuel assemblies from impact of the design missile is sufficient damage to nine assemblies in reracked pool sections, or two assemblies in the unreracked sections to result in the release of their concomitant volatile gap activities.

In performing the accident radiological consequence analysis, it is assumed that the fuel has been discharged from the reactor after operation at a steady-state power 'leve'I of 1551 IW for an extended period of time.

The assumptions in the staR analysis are 'listed in Table 1 below.

The calculated (0-2 hr) offsite accident radiological consequences are estimated to be 63 rem thyroid and Oel whole body at the Exclusion Area Boundary,'or impact with unreracked assemblies.

For impact with reracked assemblies, the corresponding 'offsite radiological consequences are 2 rem thyroid and Oe 1 rem whole body.

Both sets of consequences are well within the guidelines of 10 CFR Part 100.

Table 1:

Assumptions in Staff Offsite Radiological Consequence Analysis of Postulated Tornado Missile Accident Unreracked Section Reracked Section Reactor Power Level Effective Pool Decontamination Factor or Iodine 1551 M1Wth 100 1551 MMth 100 Radial Power Peaking Factor 1.2 1.2 Unreracked Section Reracked Section Fuel Exposure for Impacted Spent Fuel Assembly Number of Equivalent Impacted Spent Fuel Assemblies Cooldown time for Impacted Spent Fuel Assembly 30,000 MWD/MTU 100 hr 30,000 MWD/NTU 60 d Diffusion and Transport Atmospheric Relative Concentration, 0-2 hours, 9 Exclusion Area Boundary 2.2 x 10 sec/m 2.2 x 10 sec/m Filter s 2.8.3 Fuel Handlin Accident none assumed operational none assumed operational In performing the radiological consequence analysis for the fuel handling accident, it was assumed that the fuel has been discharged from the reactor after operation at a steady-state power level of 1551 MW for an extended period of time.

The assumptions in the staI'I'nalysis are listed in Table 2 be1ow.

The calculated (0-2 hr) offsite accident radiological consequences are estimated to be 44 rem thyroid and 0.1 rem whole body at the Exclusion Area Boundary, well within the guidelines of 10 CFR Part 100.

Table 2:

Assumptions in Staff Offsite Radiological Consequences Analysis of Postulated Fuel Handling Accident Unreracked Sec'tion Reracked Section Reactor Power Level 1551 MWth 1551 MWth Effective Pool Decontamination Factor for Iodine 100 100 Radial Power Peaking Factor Fuel Exposure for Impacted Spent Fuel Assembly Number of Equivalent Impacted Spent Fuel Assembly Cooldown Time for Impacted Spent Fuel Assembly 1.

65'0,000 NWD/MTU 100 hr 1.65 30,000 MWD/MTU 60 d

Unreracked Section Reracked Section Diffusion and Transport Atmospheric Relative Concentration, 0-2 hours,

~ 9 Exclusion Area Boundary Filters 2.8.4 Conclusion 2.2 x 10 sec/m none

assumed, operational 2.2 x 10 sec/m.

~ fj none assumed operatianal Since the spent fuel shipping cask may not be carried by the auxiliary bui'1ding crane, cask drop/tip accidents need not be considered.

..The staff also concludes that a tornado missile accident resulting in damage to either two 30,000 NND/NTU spent fuel

-assemblies in the unreracked pool section, or nine similar assemblies in the reracked sections, with at least 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> or 60 days of cooldown time, respectively, will result in atmospheric radionuclide releases with consequences which are well within the guidelines of 10 CFR Part 100.

Additionally, the. staff concludes that a fuel handling accident resulting in damage to either a recently discharged 30,000 NHD/MTU spent fuel assembly in the unreracked pool area, or a more substantially decayed assembly in the reracked

area, F11 result in atmospheric radionuclide releases which are well within the guidelines of 10 CFR Part 100.

The staff therefore concludes that the proposed modifications as acceptable.

3.0 OVERALL CONCLUSION Based on the review, the staff concludes that the licensee's proposed SFP modification to incr ease the storage capacity of the SFP to 1016 fuel assemblies is acceptable.

In addition, the proposed Technical Specifications are acceptable.

The staff concludes, based on the considerations discussed above, that:

(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in'the proposed

manner, and (2) such activities will'e conducted in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.

4. 0 ACKNOMLEDGEMENT This Safety Eva'luation was prepared by the following NRC staff:

M. Hohl, Accident Evaluation Branch J. Lee, Meteorology and Effluent Treatment Branch D. Swift, Radiological Assessment Branch C. Hinson, Radiological Assessment Branch M. Lamastra, Radiological Assessment Branch S.

Kim, Structural and Geotechnical Engineering Branch B. Turovlin, Chemical Engineering Branch

. F. Clemenson, Auxiliary Systems Branch L. Kopp, Core Performance Branch

23

5.0 REFERENCES

1.

J.

S. Anderson, "Boraflex Neutron Shielding Material - Product Performance Gate,".Brand Industries, Inc., Report 748-30-1, (August 1979).

2.

D.

M. Crutchfield (NRC) to J.

E. Maier (RG&E),

SUBJECT:

Spent Fuel Pool Cooling System Modifications (Ginna), dated November 3, 1981.

3.

J.

S. Anderson, "Irradiation Study of Boraflex Neutron Shielding Materials," Branch Industries, Inc., Report 748-10-1, (August 1981).