ML20043C342

Proposed Tech Specs Re Spent Fuel Pool Storage Capacity

ML20043C342
Person / Time
Site:	FitzPatrick
Issue date:	05/31/1990
From:	POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
To:
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ML20043C339	List: ML20043C338 ML20043C340 ML20043C342 ML20043C362
References
NUDOCS 9006050084
Download: ML20043C342 (14)

Text

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i ATTACHMENT l i

e PROPOSED TECHNICAL SPECIFICATION CHANGES REGARDING SPENT FUEL POOL STORAGE CAPACITY (JPT549435) l l

New York Power Authority I

. JAMES A. FITZPATRICK NUCLEAR POWER PLANT L Docket No. 50 333 DPR 59 90060500s4 90033.1 fDR ADoCK 05o00333 PDC l' . - _ _. .-. . . .

JAFMPP 5.5 (Cont'd)

8. The spent fuel storage pool is Q sed to mantain lg less than 0.95 under all conditions as described in the Authority's appEcations for spent fuel storage modification transmitted to the NRC July 26,1978 and May 31,1990. This 19 value is sabsfied if the maximum, exposure dependent, 'mfinite lattice multiplication factor, ( , of the individual fuel bunde is less than or equal to 1.36. The number of spent fuel assemblies stored in g the spent fuel pool shall not exceed 2,797. -

5.6 SEISMIC DESIGN The reactor building and all engineered safeguards are (ms; greed on a basis of dynamic analysis using acceleration response spectrum curves which are normalized to a ground mobon of 0.08g for the Operating Basis Earthquake and 0.15g for the Design Basis Earthquake.

Amendment No.%)#[JOI 246

te JAFNPP t

5.5.B Bases The spent fuel pool and high density fuel storage racks are ClassI structures %wd to store up to 2/97 fuel bundles.

l The storage racks are designed to mairtain a subenbca! 'A configuration having a multiplicabon factor (kg ) less than 0.95 '

for all possible operabonal and six untie; conditions. The nuclear g wit;cei;ti analysis for the Spent Fuel Recks (References 1 and 3) concludes that fresh fuel bundles with 3.3 w/o U-235 meet the l 0.95 kg limit. This design basis buncts was reenetyced to determine its infinite fathce mulapiscahon factor, k,, when in a reactor core geometry (Reference 2). This k, was obtained under conservative calculabonal assunipikaw and reduced by 2.33 times the stendigd deviabon in the calculabon resulbng in the T*3 (scal Specdk:ation linit of 1.36. -

References:

1) increased Spent Fuel Storage Mvdir.cetion, Stonc &

Webster Engineenng COpuretion, Boston, Mass. March 15,1978.

l 2)

General Electric letter, P. Van Dieman to G. Rorke, FitzPatrick Fuel Storage K-infirwty Conversion, Revision 1, dated July 10,1986.

3) Increased Storage Capacity for FitzPatrick Spent Fuel Pool, Holtec Irdemational Mount Laurel, New Jersey, February,1989.

Amendment No. g 246a-

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ATTACHMENT 11 4

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i SAFETY EVALUATION-FOR PROPd5Y5TEURRICKL EPECIFICATION CHANGE APACITY '

(JPTS-8H35) 1 l

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i New York Power Authority JAMES A. FITZPATRICK NUCLEAR POWER PLANT Docket No. 50 333 DPR 59 t

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l Attachment ll SAFETY EVALUATION Page 1 of 10 ,

Section i DESCRIPTION OF PROPOSED CHANGES The proposed changes to the FitzPatrick plant Technical Specifications, Appendix A, are limited to Specification 95.5.B on page 246 and Bases page 246a. They are:

Page 246, First Paragraph

[a] The first sentence will add the new spent fuel storage amendment application date.  ;

The sentence will read:

B. The spent fuel storage pool is designed to maintain k,n less than 0.95 under all conditions as described in the Authority's applications for spent fuel storage modification transmitted to the NRC July 26,1978 and May 31,1990.

[b] The number of spent fuel assemblies that can be stored in the spent fuel pool (SFP) will be changed from *2,244' to *2,797,*

Bases Page 246a, First Paragraph

[c] The number of spent fuel assemblies stored in the spent fuel pool will be changed from *2,244* to *2,797.* .

[d] The third sentence adds Reference 3, deletes the 3.2% Ak margin, and reads:

The nuclear criticality analysis for the Spent Fuel Racks (References 1 and 3) concludes that fresh fuel bundles with 3.3 w/o U 235 meet l the 0.95 k,g limit.

l Reference 3 reads:

3) Increased Storage Capacity for FitzPatrick Spent Fuel Pool, l Holtec International, Mount Laurel, New Jersey, February

1989.

In addition, the Stone & Webster report date in Reference 1 on page 246a, was inadvertently introduced as March 15,1986. The date should read March 15,1978.

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Attachment ll SAFETY EVALUATION Page 2 of 10 Section 11 PURPOSE OF THE PROPOSED CHANGES t The proposed modification increases the storage capacity of the FitzPatrick plant spent fuel pool to provide sefe storage for up to 2,797 spent fuel assemblies. Thls allows planned power ,

operation until 1997 while maintaining the capability of a full-core offload.

Background

On October 17,1974, the Authority obtained Operating Ucense DPR 59 for the FitzPatrick plant. The plant commenced commercial operation on July 28, 1975. The original storage capacity was 760 fuel assemblies or approximately 1 1/2 cores.

In 1981, the SFP was reracked with high density aluminum alloy racks increasing capacity to 2,244 fuel assembly locations. This provided significant additional storage capacity. With presently available storage, the FitzPatrick plant will lose full core off load capability in 1991.

To provide Increased capacity to store additional spent fuel assemblies, the Authority plans to ,

add five modules containing 553 new locations. This will increase the total SFP capacity to 2,797 l locations. Adding new racks provides the most economical cnd feasible solution to preclude the i potential shortage of storage capacity. <

l This effort is consistent with the objective of the Nuclear Waste Policy Act (NWPA) of 1982.

The NWPA established a federal program to deal with high-level nuclear waste (HLW) and spent fuel Specifically, with regard to spent fuel, the NWPA states in 6131(a)(1):

Persons owning and operating civilian nuclear power reactors +

have the primary responsibility for providing Interim storage of spent nuclear fuel from such reactors, by maximizing, to the extent practical, the effective use of existing storage facilities at the sita of each civillan nuclear power reactor, and by adding new onsite storage capability in a timely manner where practical.

This NWPA was intended to resolve the storage problems created by the elimination of commercial fuel reprocessing and obligated the Department of Energy (DOE) to accept spent fuel i no later than January 31,- 1998. However, delays in the program have occurred and DOE currently estimates that a HLW repository will es be operational before 2010.

Based on the above, there is clearly a need for additional onsite SFP storage capacity to assure continued operation of the FitzPatrick plant.

Attachment 11 SAFETY EVALUATION Page 3 of 10 Section til IMPACT OF THE PROPOSED CHANGES The overall safety criterion for Specification 65.5 is to maintain the spent fuel pool subcritical with a kg ess l than 0.95 under all conditions. The proposed rack expansion will not significantly impact this criterion, operation or safety of the plant.

Changes [a] and [d] are administrative changes. Change [a] adds an amendment application date to supplement the Authority's existing application for amendment dated July 26, 1978.

Change [d] adds the new Ucensing Report Analysis performed by Holtec International, as Reference 3, in support of the increased storage capacity. The Reference 1 report date is corrected to read March 15, 1978. This typographical error was introduced in a previous amendment. ,

The 3.2% Ak margin In the Technical Specifications, was calculated for the existing racks. This .

number will be removed from the Technical Specifications, since it no longer appIles to all the racks in the SFP. It was included for information only and can be found in documents supporting Amendment No. 63, page 246a.

Changes [b] and [c] will allow additional storage capacity by increasing the number of assemblies in the SFP. This proposed change to the Technical Specification is a direct result of adding the new rak modules. 7 The proposed modification to add new rack modules changes the physical configuration of the ,

SFP. The primary effect of adding new spent fuel storage racks will be an increase in the amount of spent fuel assemblies that can be stored in tne pool. This will increase the weight to be supported by the pool floor and increase the amount of decay heat which must be removed by the spent fuel pool cooling system.

The impact of these and other effects has been evaluated in Attachment 111. This report documents the design and analysis performed to demonstrate that the new spent fuel racks satisfy requirements of applicable codes and standards including those of Reference 2. These analyses demonstrate that the consequences of postulated accidents, as a resutt of the proposed changes, remain well within the acceptable limits established in the FitzPatrick plant FSAR and applicable federal regulations.

The safety assessment (Attachment lil) of the rack expansion addresses issues such as l criticality, thermal hydraulics, structural adequacy, and rack Installation safety impact. The

! following is a brief summary of the analyses presented in Attachment lit.

A. Criticality Safety Analysis i

The racks are designed to maintain spent fuel assemblies in a space geometry which l precludes the possibility of criticality during normal and abnormal conditions. Criticality analyses of the spent fuel racks were performed using NRC approved computer codes (Section 4.0 of Attachment 111).

The high density spent fuel racks for the FitzPatrick plant are designed to assure that the neutron multiplication factor (kg ) is less than 0.95 with the racks fully loaded with fuel of the highest anticipated reactivity; and the pool flooded with non borated water at a temperature corresponding to the highest reactivity.

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Attachment il SAFETY EVALUATION Page 4 of 10

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The criticality design basis for the new racks is identical to the basis used for the existing  ;

racks. Briefly stated, Ohls basis conservatively established the infinite multiplication factor '

(k ) of a 3.3 w/o enriched 8 x 8 fuel bundle without gadolinium poison in the reactor core geometry at 1.36. This design basis fuel bundle had been previously shown to meet kg <_

0.95 for the existing spent fuel racks. In this application, the same fuel bundle has been used to calculate suberiticality in the new racks. The new rack vendor has also performed a check calculation of the design basis bundle in the reactor lattice to verify consistency ,

with the previous basis calculations. The results of the check calculation demonstrate that the Technical Specification basis is conservative under the methods used to analyze the new racks.  ;

B. Thermal-Hydraulle Considerations The proposed storage expansion will slightly increase the heat load in the SFP. An analyses has been performed in Attachment lli that describes the design basis, method of ana!ysis, and the numerical results.

As a result of these analyses, the present cooling capacity 41.the SFP cooling system is cufficient to absorb the incremental heat load. Analysis also shows that the maximum -l

, cladding temperature is kept low enough by tne existing SFP cooling system. This will l preclude nucleate bolling or volding of coolant on the surface of the fuel rods and between l the racks.

The increase in radioactivity released to the pool water as a result of the proposed .

modification is insignificant. It will be kept at levels close to which have existed prior to the modification by the SFP cooling and cleanup system. The existing SFP cooling, makeup, and cleanup systems will not require any modifications. Therefore, adequate pool water 3 conditions are maintained. q C. Structural Analysis Racks The new rack modules are solsmic Class I in accordance with the FitzPatrick plant Final Safety Analysis Report (FSAR). A seismic analysis of the spent fuel storage racks was

, performed, as presented in Section 6.0 of Attachment lil, to determine the rack behevlor 1 i and to ensure no loss of function resulting from elther a safe shutdown earthquake (SSE)

! or an operating basis earthquake (OBE). The existing and new racks for the FitzPatrick plant SFP are qualified as non impacting freestanding racks, i.e., each module is freestanding and self supporting and undergoes minimal kinematic displacements during postulated seismic events. Thus, rack to-rack (old and new racks), and rack to-wall impacts are precluded.

Spent Fuel Pool Floor A structural analysis was performed (See attachment Ill) to show that the Integrity of the pool structure is not compromised for postulated loading conditions. The reinforced 1 concrete floor will continue to meet allowable stress limits with the addition of the new racks when fully loaded with fuel. Thus, the safety function of the pool structure is not affected.

Attachment ll SAFETY EVALUATION Page 5 of 10 Rack installation The Authority will not carry any rack directly over the stored spent fuel assemblies. All work in the SFP area will be controlled and performsd in strict accordance with specific written procedures. The Reactor Building crane will be used to bring the racks onto the Reactor Building refueling floor. The crane has been evaluated to meet the guidelines of Section 5.1 of NUREG 0612," Control of Heavy Loads at Nuclear Power Plants,* regarding rack movements. The rack Installation proc 6 dure provides for the safe handling of heavy loads in the vicinity of the SFP. Therefore, the proposed new racks, existing racks, and pool s'tructure will remain functional and withstand any postulated rack installation accident.

Section IV EVALUATION OF SIGNIFICANT HAZARDOUS CONSIDERATIONS The changes proposed to the FitzPatrick plant Technical Specifications as a result of the expansion, will allow an increase in the number of spent fuel assemblies stored in the spent fuel pool from 2,244 to 2,797. Operation of the FitzPatrick plant in accordance with the proposed amendment will not involve significant hazards considerations as defined in 10 CFR 50.92, since it would not:

(1) Involve a significant increase in the probability or consequences of an accident previously evaluated.

The evaluation of the proposed spent fuel pool modification has considered postulated accident scenarios including:

a Criticality (e.g., spent fuel assembly drop in the spent fuel pool, rack movement);

e Thermal Hydraulic (e.g., loss of spent fuel pool system flow, heat load); and a Structural (e.g., seismic event, spent fuel pool structure, rack movement, rack installation);

The effects of these and other potential accident scenarlos have been fully analyzed in attachment lll. These analyses demonstrate that the consequences of postulated l accidents remain within the acceptable limits established in the FitzPatrick plant FSAR and '

L applicable federal regulations.

Criticality analyses of the spent fuel racks werc performed using NRC approved computer codes. Under normal operating conditions, these calculations demonstrate that the maximum effective neutron multiplication factor (ken ) in the fuel pool will be less than 0.95, ,

including all uncertainties. Reactivity effects of abnormal and accident conditions have l also been evaluated to assure that ko n would remain less than 0.95 with the racks fully (

l loaded with fuel. The effective multiplication factor (k,y ) is less than 0.95 for all possible operating and accident conditions. Since the acceptable margin to criticality is not changed, the consequences of reactivity accident are not significantly increased from those previously evaluated.

l m Attachment ll r SAFETY EVALUATION .

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The analyses show that sufficient time is available to provide alternate means of cooling in tha event of a failura in the spent fuel pool cooling system, and unavailability of the Residual Heat Removal heat exchangers. Adequate cooling under both normal and-abnormal conditions is maintained. The existing spent fuel pool cooling, makeup, and .

cleanup systems will not require any modifications. Thus, the consequences of this type of accident are not significantly increased from previously evaluated loss of cooling system flow accidents.

Fuel assembly drop has been analyzed for impact on the racks and presented in Attachment 111. In the unlikely event of a fuel assembly drop on top of a storage rack, the resulting deformation will not distort the racks. Thus, the intended safety function of the racks will continue to be met. Also, the radiological consequences of a fuel assembly drop .

are not changed from those previously analyzed in Section 14.6 of the Final Safety Analysis Report, ,

The new racks are free standing and self supporting as are the existing racks. They are seismic Class I design in accordance with the FitzPatrick plant Final Safety Analysis Report. A seismic analysis of the spent fuel storage racks was performed to determine rack behavior and to ensure that no loss of function would result from a safe shutdown l

earthquake or an operating bases earthquake. The analysis demonstrates that rack to-rack (existing and new racks), and rack to-wallimpacts are precluded. The spent fuel pool structure will continue to meet the allowable stress limits with the addition of the new racks fully loaded with fuel. The integrity of the pool structure is not compromised for ,

postulated loading conditions. The safety function of the racks (new and existing) and pool structure is not affected. Therefore, the consequences of a seismic event are within previously evaluated limits.

(2) create the posalbility of a new or different kind of accident from those previously evaluated.

The proposed rack addition has been reviewed and analyzed for possible types of accidents. The criteria used in the analyses, design, and installation of the new spent fuel racks to account for anticipated loadings and postulated conditions that may be imposed I

upon the structure during their service lifetime are in conformance with established criteria, codes, standards, and specifications acceptable to the NRC.

Factors that could affect the neutron multiplicatbn factor in the spent fuel pool have been addressed conservatively. The Authority concludes that the maximum neutron multiplication in the pool with the addition of the high density racks will not exceed the subcriticality limit of kg = 0.95.

The addition of new racks and spent fuel in the spent fuel pool will produce an incremental heat load. However, analysis has shown that the present cooling capacity in the FitzPatrick plant spent fuel pool is sufficient to absorb the incremental heat load. The peak pool bulk temperature will be maintained below the threshold value to preclude local boiling. The incremental heat load will not alter the safety considerations of spent fuel pool cooling from those which the Authority previously reviewed and found to be acceptable.

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SAFETY EVALUATION Page 7 of 10 I

Rack impact analysis has been performed (Attachment Ill) to investigate pussible impact due to seismic events (i.e., rack to rack and rack to-wall imoacts). The analysis has demonstrated that the proposed storage expansion does not result in rack to rack or i rack to wall impact during postulated seismic events, in conclusion, the analyses demonstrates that there is no increase in the probability of a new or different accident from those previously evaluated. No new failure modes are i created; nor is the safety function or operation of the spent fuel pool coo'ing, makeup, and j structural systems changed by the rack addition. This amendment to the Technical q Specifications will not result in the initlation of any type of accident or plant transient nor affect the plant's safe shutdown capability.

(3) involve a significant reduction in the margin of safety, l i Tho safety function of the spent fuel pool and the racks is to preclude inadvertent criticality in a safe, specially designed, underwater storage location for spent fuel assemblies which  !

require shielding and cooling during storage and handling. The NRC Staff has established that the issue of margin of safety, when applied to spent fuel racking modifications, should address the following criteria; 1) Nuclear criticality considerations; 2) Thermal hydraulic considerations; and 3) Mechanical, material and structural considerations. Using these criteria assures that the new fuel pool structure will withstand specified design conditions without impairment of structuralintegrity or the performance of requimd f.,sfety functions.

The proposed change to add new racks will not significantly reduce the margin of safety for nuclear criticality, since criticality calculations show that the maximum effective l' multiplication factor (k,g ) is within NRC acceptance criterion. Criticality calculations for the new racks indicate a slightly higher k,g (0.937 vs 0.918) than the existing racks for the same criticality design basis. The difference is primary due to the difference in B 10 loading in the BORAL plates. However, the margin of safety applicable to the racks (i.e.,

k,g < 0.95, under all credible conditions given conservative assumptions) remains unchanged.

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L Conservative methods and assumptions are used to calculate the maximum fuel temperature and the increase of the water in the spent fuel pool. The thermal hydraulic evaluation uses a conservative mathematical model to demonstrate that the temperature margins of safety are maintained. The proposed rack addition with the increased fuel inventory will slightly increase the heat load in the spent fuel pool. The evaluation shows that the existing spent fuel pool cooling system will maintain the bulk pool water temperature below 150 for normal discharge and for full core offload. The evaluation also shows that maximum local water temperature along the hottest fuel assembly is well below the nucleate boiling condition. Thus, there is no reduction in the margin of safety for thermal-hydraulic or spent fuel cooling concerns.

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Attachment il SAFETY EVALUATION '

Page 8 of 10 Structural concems have been analyzed for normal and abnormal conditions, including earthquake, spent fuel assembly drop, rack installation, rack.to-rack impact, etc. The new racks are Seismic Category I equipment and designed to remain functional during an  ;

Operating basis earthquake, as specified by NRC Regulatory Guide 1.29. The rack i materials used are compatible with the spent fuel pool and spent fuel assemblies, and degradation due to material combinations (e.g., galvanic corrosion) is not expected. In addition, the spent fuel assemblies remain intact, and no criticality concems exist. Thus, the margin of safety is not reduced by the proposed addition of rack modules.

Similar Examples  !

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The NRC published

• Example of Amendments That Are Considered Not Ukely To involve '

Significant Hazards Considerations" as provided in the final NRC adoption of 10 CFR 50.92 published on page 7751 of the Federal Register Volume 51, No. 44, March 8,1986. Example X of that list is applicable to the proposed amendment to the Technical Specifications.

The example states:

(X) An expansion of the storage capacity of a spent fuel pool when all of the following criteria are satisfied: ]

l Criterion (1)

The storage expansion method consists of either replacing existing racks with a design  !

which allows closer spacing between stored spent fuel assemblies or placing additional racks of the original design on the pool floor if space permits.

l Proposed Amendment: l l The FitzPatrick spent fuel pool storage expansion involves adding racks of a new design where spcce permits. The new racks exceed the structural strength of the old racks, and meets the criticality and thermal hydraulic characteristics of th6 cid racks.

Criterion (2)

The storage expansion method does not involve rod consolidation or double tiering.  ;

Proposed Amendment:

t The proposed racks are not double tiered as all racks will rest on the spent fuel pool floor, j Additionally, the proposed storage expansion does not involve rod consolidation.

l. Criterion (3)

The k,y of the poolis maintained less than or equal to 0.95.

Proposed Amendment:

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i The design of the new spent fuel racks contains a neutron absorber, Boral, to allow close l storage of spent fuel assemblics while ensuring that the k,n remains less than 0.95 under l

all conditions with pure water in the pool.

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Attachment 11 SAFETY EVALUATION 4

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Critorion (4)

No new technology or unproven technology is used in either the construction process or the analytical techniques necessary to justify the expansion. ,

Proposod Amendment:

The NRC has licensed at least ten other plants for rack installation of the same rack design. The construction processes _and analytical techniques remain substantially the same as these other ten rack installations. Thus, no new or unproven technology is used in the construction or analysis of the new high-density spent fuel racks at FitzPatrick. 3 Section V IMPLEMENTATION OF THE PROPOSED CHANGES Implementation of these changes, as proposed, will not Impact the ALARA or Fire Protection Programs at the FitzPatrick plant, nor will the changes impact the environment.

Section VI CONCLUSION The changes, as proposed, do not constitute an unreviewed safety question as defined in 10 CFR 50.59. That is, they:

a) will not increase the probability of occurrence or the consequences of an accident or malfunction of equipment Important to safety previously evaluated in the safety analysis report; b) will not increase the possibility for an accident or malfunction of a type different from any evaluated previous!y in the safety analysis report; 3 c) will not reduce the margin of safety as defined in the basis for any technical specification; and d) involve no significant hazards consideration, as defined in 10 CFR 50.92.

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SAFETY EVALUATION Page 10 of 10 i

Section Vil REFERENCES

1. James A. FitzPatrick Nuclear Power Plant Final Safety Analysis Report (FSAR).

Section 9.3, " Spent Fuel Storage."

Section 9.4, " Fuel Pool Cooling and Cleanup System."

Section 12.5,

• Dynamic (Seismic Analysis).*

Section 12.6, " Analysis of Spent Fuel Storage Pool."

Section 14.6, " Analysis of Design Basis."

2. NRC Standard Moview Plan, NUREG-0000, dated June 1987.

Section 3.7,

• Seismic Design

• Section 9.1.2, " Spent Fuel Storage."

Section 9.1.3, " Spent Fuel Pool Cooling and Cleanup."

3. NRC letter to all Power Reactor Ucensees, "NRC Guidance On Spent Fuel Storage Modifications Review and Acceptance of Spent Fuel Storage and Handling Applications,"

dated April 14,1978, and the January 1979 revision thereto.

4. NRC NUREG-0612, " Control of Heavy Loads at Nuclear Power Plants."

5. Increased Storage Capacity for FitzPatrick Spent Fuel Pool, Holtec International, Mount Laurel, New Jercey, February 1989, i

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