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CONNECTICUT YANKEE AT O MIC POWER COMPANYv

BERLIN, C O N N E CTIC U T

P O. BOX 270 HARTFORD C.:.N N ECTICUT 0610t

g ;';;o;;,, August 31, 1981

-(fDocket No. 50-213 s

E10283 [ g X'' --D (b N5 s >,

ATTN: Mr. Dennis M. Crutchfield, Chief, [' ~(!Director of Nuclear Reactor ".egulation -

'J'- .-.E/[/Operating Reactors Branch 65 T

U. S. Nuclear Regulatory Commission ',,"

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Washington, D. C. 20555 %x /'<'i,, , ,

References: (1) D. G. Eisenhut letter to All SEP Licensees,dated July 7, 1981.

(2) W. G. Counsil letter to D. G. Eisenhut,dated July 29, 1981.

Gentlemen:

Haddam Neck PlantSEP Topic IX-1, Fuel Storage

Reference (1) requested to SEP licensees to cocnit additionalresources devoted '9 completion of the SEP. In Reference (2)Connecticut Yankee AtoMc Power Company (CYAPCO) committed todevelop Safety Assessmen - Reports (SAR's) for certain SEPtopics which would be suLaitted for Staff review. In accord-ance with this commitment, CYAPCO hereby provides the SafetyAssessment Report for SEP Topic IX-1, Fuel Storage, which isincluded as Attachnent 1.

The information included in Attachment I related to the

|seismic design of the spent fuel pool should also be con-

i sidered part of SEP Topic III-6, Seismic Design Considerations.|

We trust the Staff will appropriately use this information to

| develop a Safety Evaluaticn Report for this SEP topic.1

Very truly yours,

| CONNECTICUT YANKEE ATOMIC POWER COMPANY

64:5<

W. G. Counsil

h35Senior Vice Presidents

i/I9109040150 0108312DR ADOCK 050002131

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Docket No. 50-213

Att.schment 1

Safety Assessment Report

SEP Topic IX-1, Fuel Storage

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Haddam Neck PlantSafety Assessment Report

Topic IX-1, Fuel Storage

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1.0 INTRODUCTION

The objective of this topic is to assure that fuel is storedsafely with respect to criticality (Keff <0.95) , coolingcapability (outlet temperature <150*F), shielding, and struc-

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tural capability..

2.0 CRITERIA,

This topic has been reviewed for compliance to the require-ments of ANSI N210-1976/ANS 57.2, Design Objectives for LightWater Reactor Spent Fuel Storage Facilities at Nuclear PowerStations.

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3.0 DISCUSSION

On June 8, 1976, the Commission issued Amendment No. 7 toFacility License No. DPR-61 for the Haddam Neck Plant. Thisamendment permitted changes in the design of the spent fuelstorage racks allowing fuel storage capacity to increase from336 to 1,172 fuel assemblies. The amendment also changed theamount of U-235 the_ licensee may possess from 6,500 kg to thatamount limited to storage capacity and amounts required forreactor operation.

CYAPCO is continuing to implement this increase in two phases.During Phase I, which was completed in 1978, the original rackswere removed and new racks, capable of storing 672 assemblies,were installed. During Phase II, the remaining new rackscapable of storing 496 assemblies will be installed which willincrease the total storage capability of 1,168 assemblies. -

Installation of the new racks is currently scheduled for thelatter part of 1983.

The new racks consist of 4 x 4 and 3 x 4 arrays with 10.75inches center to center spacing of fuel assemblies. The newdesign encases each stored assembly with a neutron absorbingmaterial (B C) to ensure subcriticality under all conditions.4The new racks are seismic Category I and are restrained at thebase against the pool walls to prevent movement of the racksunder postulated seismic accelerations. The effect of theincrease in pool heat load due to the increased number of spent

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fuel assemblies has been accommodated by modifying thespent fuel pool cooling system.

The NRC safety evaluation performed for Amendment No. 7 tothe Facility Operating License consisted of a detailedanalysis of each area in which potential safety considerationswere involved. The following has been excerpied from thisreview.

e Criticality Analysis

The center to center spacing of assemblics in the spentfuel pool was decreased from 18 inches to 10.75 inches.This reduced spacing tends to cause an increase in theeffective neutron multiplication factor, Keff, of thearray. This increase is compensated, however, by theuse of a neutron absorbing material, B4C, which encaseseach fuel assembly. The licensee's analysis of the fuelstorage configuration was reviewed to determine if thecriterion for Keff <0.95 is met. Fresh fuel of 4.0 w/oU-235 enrichment was assumed to be in a pool filled withunborated water at 68'F. Design calculations were per-formed by four group diffusion theory (PDQ-07) and veri-fication of the design was done by a Monte Carlo (KENO)calculation. The KENO code was verified by comparisonto critical experiments. The nominal value of Keff forthe storage facility is 0.905. When calculationaluncertainties and worst case tolerances (enrichment,spacing, poison uncertainties, and pool temperature)are added, the maximum Keff is 0.947 which meets thecriterion of 0.95. Abnormal fuel distributions wereconsidered and it was shown that the increase in Keff wasnegligible. The effect of creation of voids in the pooldue to loss of all cooling capability has also beenexamined by the licensee. The calculation shows that amaximum void volume fraction of 0.42 in the hot channelof the hot bundle is possible assuming the most adverseassumptions. Under these conditions, the reactivitydecreases as a function of voids up to 0.42 void fraction.The criticality aspects of the proposed modifications wereconcluded as being acceptable.

e Thermal Considerations

The original spent fuel pool cooling system was designedto maintain sufficient cooling capacity to keep the spentfuel pool water temperature to less than 170 F (FDSAdesign basis) for 336 fuel assemblies. The licenseeadded a pump and heat exchanger designed to the snmecodes and standards as the original equipment, in parallel

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with the existing pump and heat exchanger. The heatremoval capacity for the parallel components are threetimes greater than for the existing system. The licenseeprovided an evaluation of this modified system and deter-minted that it had the capability to cool the pool to lessthan the original design temperature, considering therevised heat load due to the increased storage capacity.NRC reviewed the system and agreed with the licensee'sconclusion that this system is adequate.

The Staff analyzed the spent fuel pool heatup time in theevent that the spent fuel pool cooling system fails fromloss of offsite power. The minimum time to reach boilingfrom a pool water temperature of 140 F was eight hourrunder the most adverse conditions. Either the exis.Ingor additional spent fuel pool cooling pump can be poweredfrom either emergency diesel generator in the event ofloss of offsite power. The pump bus can be manually con-nected to the diesel generator bus well within the eighthours it would take for the pool to heat up to 212'F.Sources of available makeup water include the primarywater storage tank (150,000 gallons), the refueling waterstorage tank (250,000 gallons), and as a last resort, thefire protection system which takes suction from theConnecticut River. Thus, it was concluded there issufficient time for the operator to effect a repair orconnect to additional cooling.

The proposed modifications to the spent fuel pool coolingsystem were deemed acceptable.

e Rack Structural, Mechanical, and Material Review

The supporting arrangement for the racks including theirrestraints; design, fabrication, and installationprocedures; structural analysis for all loads includingseismic and impact loadings; load combinations; structuralacceptance criteria; quality assurance requirements fordesign and installation; and applicable industry codes;in accordance with the criteria described in Sections 3.7and 3.8 of the NRC Standard Review Plan, were reviewed.The analysis, design, fabrication, and installation ofthe new fuel racks were found in accordance with acceptedcriteria.

The stainless steel material for the fuel racks; density,mechanical properties, capacity to withstand temperatureand irradiation of the B4C material; inservice inspectionrequirements to ensure performance, were all reviewed,consistent with the requirements identified in Section9.1.2 of the Standard Review Plan. The licensee qualified

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the B4C material through a test program in order toensure that densification, settling, or stratification ofthis material will not occur during service. The resultsof the qualification program were reviewed and cenfirm thesuitability of the B4C plates for their intended use. Thelicensee also provided for continued assurance of struc-tural integrity of the neutron absorber by placing controlsamples and a fuel box can assembly containing boroncarbide material in a readily accessible location to per-mit easy removal and inspection. The structural, mechanical,and material features of the rack structure were deemedacceptable.

e Radiation Levels

The increment in onsite occupational dose resulting fromthe proposed increase in stored fuel assemblies on thebasis of information supplied by the licensee and byutilizing realistic assumptions for water cleanup periodsand for occupancy times was estimated. This analysisindicates that the occupational radiation exposure resultingfrom this proposed action represents less than one percentof the present total annual occupational burden at thisfacility. The small increase in radiation exposure willnot affect the licensee's ability to maintain individualoccupational doses within the limits of 10 CFR 20. Theoffsite doses associated with the spent fuel pool rackexpansion were also evaluated. Assuming that an individualresides at the nearest boundary for a year, the estimatedincrease in annual dose is less than 6 x 10-4 millirem / yearabove that expected from the existing storage scheme. Thisdose is insignificant when compared to the dose of greaterthan 100 millirem / year that the same individual wouldreceive from natural background radiation and therefore,does not represent a meaningful impact. Thus, no signi-ficant dose to either occupational workers or the generalpublic will result from normal operation of the expandedspent fuel pool.

The above evaluations completed in 1976 have been reviewed todetermine if any discrepancies exist with current criteria. Thecriticality analysis has not been repeated; however, based oncurrent poison rack designs for similar operations, it isobvious that the original analysis was overly conservative.The criterion for Kef f < 0.95 is met.As noted above, modifications were made to the fuel poolcooling system to deal with the increased heat load. Theoriginal system maximum heat load was increased from 5.43 x106 Btu / hour to 18.9 x 106 Btu / hour. A new plate type heat

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exchanger, capable of handling 20.0 x 106 Btu / hour was addedalong with an additional heat exchanger pump. In the presentarrangement, either pump or both pumps can supply either heatexchanger.

Under normal conditions, the maximum pool temperature is 106 F.With a full core off-load, the maximum temperature is 143 F.With a single active component failure in the worst case, themaximum temperature is 170 F. The last two temperatures fallunder those specified in ANSI 210.

Finally, the structural capability of all components of thepool was reviewed. The racks and all appurtenances weredesigned for seismic accelerations of 0.17g (ZPA). RegulatoryGuide 1.60 ground response spectra were normalised to 0.17ghorizontal and 0.12g vertical and treated as the appropriateSSE floor response spectra. The spent fuel storage pit isconstructed of five foot thick reinforced concrete foundationon top of bedrock with six foot thick reinforced concretewalls. The seismic ground accelerations are transmitted frombedrock through the fuel pool floor to the fuel rack supports.Conservative spent fuel pool floor frequency calculations yieldedvery high vertical and lateral fundamental frequencies (>>33 Hz).Hence, amplication of ground accelerations by this extremelyrigid structure is negligible and use of the regulatory guidespectra as floor response spectra is appropriate.

Although the racks, seismic supports, and all appurtenanceswere designed in accordance with current criteria, the spentfuel pit was designed according to the original plant designcriteria. Specifically, a generalized envelope responsespectrum developed by Dr. G. W. Housner from four Californiaearthquakes was employed. This spectra does not envelope theregulatory guide spectra. The generalized spectra was normalizedto a zero period acceleration of 0.17g.

As partially discussed above, the spent fuel storage pit is ofreinforced concrete with walls six feet thick, lined with

| welded stainless steel plate, and filled with borated water.| The structure may be considered to follow the actual ground

acceleration without relative displacement. Since the funda-'

mental frequency is so high, the discrepancy in response spectradoes not affect the design input of 0.17g. Hence, the seismic

; criteria used to design the spent fuel storage pit and liner| were adequate.

3 .1, Associated SEP Topics

e III-6 Seismic Considerations

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4.0 CONCLUSIONS

Based on this review, CYAPCO concludes that fuel is safelystored with respect to criticality, cooling capability,shielding, and structural capability at the Haddam Neck site.

5.0 REFERENCES

1. ANSI N210-1076/ANS 57.2, Design Objectives for LightWater Reactor Spent Fuel Storage Facilities at NuclearPower Stations.

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