ML20236G477

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Proposed Tech Specs Replacing U-235 Loading Limit for Spent Fuel Storage Pool W/Infinite Multiplication Factor Limit & Clarifying Multiplication Factor Limit for New Fuel Storage Vault.Dj Brager Encl
ML20236G477
Person / Time
Site: Cooper Entergy icon.png
Issue date: 10/20/1987
From:
NEBRASKA PUBLIC POWER DISTRICT
To:
Shared Package
ML20236G471 List:
References
NUDOCS 8711030028
Download: ML20236G477 (11)
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Text

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5.4.C' (cont'd) penetrations shall;be. designed in accordance h standards set forth.in' ,

Section V-2.3.4 of'the SAR. /-

]

15.5 Fuel Storage A. The new fuel storage vault shall be such that the K dry 0.90 and flooded is less than 0.95. These K- limib are'issatisified less.than- by maintainingthemaximum, exposure-dependent (oftheindividualfuelbundles

< 1.29.

B. The spent fuel storage racks are designed and shall be maintained with a nominal 6'9/16 inch center-to-center distance between fuel assemblies placed in the storage racks. K f shall be maintained < 0.95 with the storagej pool filled with unborated watef.f This K- is satisfied by maintaining the ;

maximum, exposure-dependentK,oftheindINdual'fuelbundles<1.29.  ;

C. The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 2366 fuel assemblies.

D. The fuel handling bridge fuel hoist has a load-limit cell set at no more than 1230 pounds.

5.6 Seismic Design The seismic design for Class I structures and equipment is based on dynamic analyses using acceleration. response spectrum curves which are based on a ground motion of 0.1.g. The vertical ground acceleration assumed is equal to of the horizontal ground. acceleration. For the design of Class I structures and equipment, the maximum horizontal and vertical accelerations were con-sidered to occur simultaneously. Where applicable, stresses were added directly.

The combined stresses resulting from dead, live, pressure, thermal and' earthquake having a ground acceleration of 0.2g are such that a safe shut-down can be achieved.

5.7 . Barge Traffic

, Barge traffic on the Missouri River past the site has been analyzed to determine that the present size and cargo materials do not create a hazard to the safe operation of the plant. Contact will be maintained with the Corps of Engineers to determine if and when additional analyses are required

due to changes in barge size or cargo.

i1 0711030028%hhg8 ppg PDR ADOCK P

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i G EN E R AL $ ELECTRIC .

NUCLEAR ENERGY BuslNEss OPERATIONS -

, _ GENERAL ELECTRIC COMPANY

  • 11429 MIRACLE HILLS DRIVE
  • SUITE 304.. OMAHA,' NEBRASKA 68154 * (4o2) 496-6919 G-HP0-7-236' cc: Nebraska Public Power Districte.

' Columbus General Office July 13, -1987 .T. J. 0akes/w. enc.

G.-A. Trevors/w. enc. -

j 'R.!E.-Wilbur/w.? enc.

Cooper Nuclear Station- -?-

P. L. Ballinger/w, enc. :

'G.:R. Horn /w. enc.

.E. M. Mace /w.; enc.

. J. M. Meacham/w. enc.- j

,. 1 i

Mr. K. C . . Walden j Nuclear Licensing and Safety Manager -

Nebraska Public Power District P. O. Box 499 ,

Columbus, Nebraska 68601  ;

I

SUBJECT:

COOPER FUEL STORAGE K-INFINITY CONVERSION,1 REVISED

References:

gl. GESTAR-II, " General Electric Standard Application for

- Reactor Fuel", NEDE-24011-P-A-8.- <

2. NES 81A0472, Rev.1, " Nuclear Design Analysis. Report for the Cooper Nuclear Station Fuel; Expansion Program", ,

May 9, 1978.

3. Task Authorization No. 991 1

Dear Mr. Walden:

l

',. . Enclosed .is a revised- summary of. the Cooper Nuclear Station Fuel Storage ,

. K-Infinity conversion as discussed.

The k-infinity value has been calculated'for the maximum reactivity fuel bundle' l used in the criticality safety analyses of the Cooper spent 1 fuel storage racks.

The k-infinity value was calculated in the- cold uncontrolled' reactor core - '

geometry. The result for.the 2.74 wt% uranium-235 BWR fuel. bundle is:

k = 1.299.1 0.005 _( 26) .

The recommended Tech Spec limit for the high density spent fuel storage rack.

t design is k = 1.29. The Tech Spec limit represents the l lower 95/95% tolerance

limit for.the calculated value. The k-infinity'.1imit applies to alll of the y~

spent fuel storage racks utilizing square aluminum boxes, spaced onJa 6.56. inch l} nominal pitch, and containing Boral po'ison sheets on two sides: of.each storage

. cell (Reference 2). A summary description of the k-infinity calculation:is l1l attached.

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RECEIVED : < <

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. N GENER AL h ELECTRICL 2 Mr. K. C. Walden -

IJuly 13,.1987L '

il The ma'ximum exposure dependent 1_attice k-infinity for each licensed 'GE fuel .

design is . contained in Reference 1. . Th'e current maximum fuel : lattice . . .

k-infinity contained in Reference 1 is? equal to 1.251. The Cooper Station fuel:

bundle. types currently in use (i.e.. .P8DRB265L and P8DRB283) have a maximum >

1 i k-infinity equal to 1.239 and.1.228, respectively. l As a result,; the Cooper : ,

"I Station spent fuel storage racks with a 1.29 fuel k-infinity limit can '

. - accommodate the current fuel ' bundle designs, as well eas any;other GE licensed fuel design.

- If there are any' remaining GE supplied' fuel. storage racks at the~ Cooper. .

Station, the recommended Tech Spec k-infinity.1imits .for the GE fuel storage '

j L racks is' contained'in Reference 1. The'.GE low density new and spent fuel storage. racks have a k-infinity limit of 1.31.

l l The documentation and verification supporting the k-infinity anslyses for th'e  ; 1 Cooper Station spen.t-fuel storage racks is contained in GE Design 1. Record File.

F16-00030-0001.

1 i

If'you require additional information or have'any questions.concerning the -

I analyses or the application of the results, please do not hesitate to call.

Sincerely, j David J. rager-Servicer Project Manager DJB:cj Enclosure 1

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DRF-F16-00030-0001

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'JUL$ 1987. .

s tit" . GENERAL ELECTRIC COMPANY.

SUBJECTE ~ Cooper-Station Fuel 1 Storage k-infinity'Conversihn . c

SUMMARY

.The infinite. neutron multiplicationLfactor.(k-infinity)1has L been

'. calculated for the' design basis 1 fuel: bundle used in theinuclear safety analysis- (Reference 1) for ,ths Cooper Station;high .c density spent-fuel storage' racks. The' Reference 1' criticality 1 safety analyses. was . performed _ with ' two different fuel- bundle '

types. 'The. fuel bundle' type which_resultedLin the'highesti storage rack.k-effective.(< 0.95) was used for this, analyses'.

The calculation ~ consisted of an infinite array of.~ design basis ,

fuel in the cold uncontrolled reactor, core geometry.- The resulting k-infinity,. including the critical benchmark bias,'is:-

t STORAGE RACK k-infinity ( 2 d) . d a

L HIGH DENSITY 1.299' ~0.005:

9 I.

': t 7

Assuming a Normal distribution, a lower bound value'can'be ll calculated using 95/95% statistics._ ItLis' recommended that the 95/95% lower tolerance limit value of k = 1.29 be.used as the Technical Specification limits in place of the existing.

uranium-235 enrichment limit.

4 The maximum exposure dependent lattice:k-infinity for each licensed GE fuel design is contained in' Reference 2. The '

4 current maximum fuel lattice ~k-infinity. contained.in. Reference 2: u h;

is equal to'1.251. The Cooper Station specific fuel bundle' i

t. A E

f i -

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  • iDRF-F16-00030-0001.= 4

. JULY;1987 typesLP8DRB265L and P8DRB283 have'a maximum lattice k-infihity? '

Y equal to 1.239 and t 1.228, respectively'.1 JTherefore'mthe current Cooper Station fuel ~ bundle types,;as welltas'all'GE fuel' bundle types,fcan be'storednin the Cooper'. Station spent fuel' storage-

- , racks with.a 1.29 fuel bundle k-infinity: limit. '

ANALYTICAL METHOD The calculation was performed,withLthe GeneralLElectric MERIT-

~

monte carlo neutron transport computer program. 'The' MERIT' program:is a monte carlo program for' solving.the' linear' neutron transport equation as a fixed source or an eigenvalue problem in three space dimensions.- The cross! sections-in MERIT.are-processed from the ENDF/B library in the.multigroup and.

  • resonance parameter, formats. Thermal. scattering'in. water is represented by the Haywood kernel obtained from1thej ENDF/B library. The MERIT. program utilizes 190. full: spectrum.. cross section energy groups. The types of reactions considered in MERIT are fission,: elastic, inelastic land-(n,2n) reactions.

Absorptions are implicitly treated by applying the non-absorption probability to neutron weightscon eachl collision.-

QUALIFICATION The MERIT program has been thoroughly verified'for programming, sampling procedure,. particle tracking, rand;.L number generation,-

fission source distribution, statistical evaluation,' resonance ~ <

cross section evaluation, edits'and other functions of the program. The overall performance.of MERITiand the crossfsection.

i .

data was evaluated by comparison against critical experiments which include:

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1. c CSEWG.thermalreactbrh'nchmarkiprobl' ce ems:

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< TRX-1,TRX-2,ORNL-1,ORNL-2,PNL-1,PNL-2 1- t?

'2. Babcock and Wilcox, Small^: Lattice -- Facility..

e s Jersey.; Central' Gamma' Scan Experiments.

'3.- ic

4. BWR Gadolinia. Critical Experiments ,

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5. Battelle Critical 1 Experiments'with. Fixed Neutron'_ Poi' sons-

% j 6.- Nippon Atomic. Industrial Group (NIAG'7 ) Critical--

Experiments with BWR Control: Rods )

W=

The analyses of these_. benchmark calcula'tions indicate?that MERITc >

- based on the ENDF/B-IV~ cross sections under-predicts'k-effechive. [

by approximately 0.5% for thermal reactor criticals.H-The.CSEWG ,

y evaluation of'the ENDF/B-IV data concludedithat;theDexperiment is under-predicted by 0.5%'for the high moderator-to-fuelbratio in water moderated uranium latticesE(Referencef3). .;The=MERITL results confirm the biases'supportingLthe<CSEWG conclusions 1 (Reference 4)'. The MERIT result' reported in this' summary-includes the delta-k bias correction of_O.005't 0.002: (ld)..

. CALCULATIONS  ;

h An infinite lattice neutron multiplication'~ factor was calcul'ated j '

for'the fuel bundle used in the. Cooper > Station:highidensity.

spent fuel storage rack nuclearfsafety analysis. LTheffuel bundle parameters used in the spent' fuel storage"analysesJare i contained-in Table-1.- Since the fuel pell't e dimensions:and' q s

i .(

,. , - t < t .

4

- DRF-F16-00030-0001

- JULY 1987 density were not" contained in the' Reference 1 report, the_ actual  ;

isotopic atom densities contained in the criticality' report were  !

used-(Table 2). The enrichment distribution of the fuel rods is

- shown.in Table 3. The fuel bundle k-infinity was calculated' ,

using the reactor core geometry consistent with'the cooper Station without the presence of control rods. Since an unirradiated high enrichment lattice without burnable' poisons is-overmoderated in the reactor core geometry, the calculation was .

performed at 20 c, resulting in the minimum design basis' k-infinity. The' calculation.was performed with 55 batches'of 1000 neutrons, starting with a uniform fissi'on~ source distribution. The first 5 batches were discarded.to. eliminate- -.

the bias resulting from the initial source distribution. The ]

J final result was based on the remaining 50 batches or.50,000 j neutron histories. The resulting k-infinity value is:

1 FUEL RACK k-INFINITY (t 26)

HIGH DENSITY 1.299 1 005 i

i REFERENCES

1. NES 81A0472, Rev. 1, Nuclear Design Analysis Report for the i Cooper Nuclear Station Fuel Expansion Program, May 9, 1978.
2. GESTAR-II, " General Electric Standard Application for-Reactor Fuel", NEDE-24011-P-A-8.

I

3. E. M. Bohn et.al. (Ed), " Benchmark Testing of ENDF/B-IV",

ENDF-230, Vol. I, March 1976. ')

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DRF-F16-00030-0001' , .

' JULY'1987J l

4. C. M. Kang and E.-C. Hansen, "ENDF/B-IV Benchmark Analyses With Full Spectrum Three Dimensional Monte' Carlo:Models,". .

paper presented in November 1977'atlthe winter San'Franciso'-  ?!

meeting of the American Nuclear... Society, Vol. 22, p.891=.'

(

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Prepared by: , 4 d

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.P.Evan Diemen' Reviewed by: 'l Senior Engineer J..K. arrett'  :.

'1 Reactor Systems Design ' Senior Engineer ]

j I l-l (408) 925-6160 :Engrg.; Analysis Sycs j

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Approved by: g'[

W. A.:Pitt, Manager Reactor Systems Design u

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JULY 1987:  ;

.: .< .. e TABLE l'. ., BWR Fuel:- Bundle . Parameters -

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PARAMETER. VALUE

____________________________ _____________________~.

1 Fuel Rod Array' 8'x 8

-Fuel Rod O.D.' (inch), .O.501 Fuel-Rod ~I.D. (inch)- O.429 Fuel Rod Pitch (inch)- .O.642' f Average U-235 Enrichment (%) 2.74.

-Rod Enrichment Distribution distributed-Burnable Poisons 'none. -

)

Number Inert Rods 4 s.

Channel Thickness .(inch)- 0.080' t

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DRF-F16-00030-0001 4- ' JULY 1987; Table 2' Input! Material-Isotepi's-c ,

'e MATERIAL' 'At/bn-cm

-~~----------------- ----~~--~~----

4 1.. UO 2

(3.00 wt.%)

U-235 6.5782E-04.

U-238 2;1002E-02 ,

OXYGEN 4.3319E-02'

2. UO 2

(2.38 wt.%)-

L U-235 5.2187E-04 U-238 2.1136E-02.

I OXYGEN 4.3316E >

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3. UO 2 (1.64 wt.%)

U-235 3.5963E-04 .

U-238 2.1296E-02 OXYGEN 4.3312E-02

4. ZIRCONIUM 4.2900E ( 5. MODERATOR (20 C) b HYDROGEN 6.6531E-02 l

l OXYGEN 3.3265E-02 l

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1 Table 3 Fuel-Bundle Enrichment: Distribution .'

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A' .B C' D '- E F. Gi -H' l Al 3; '3 2 2 '2i 2 ,

~2. 3-

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u- Bl 3 2 1 l' 1- ' 1. * ' l~- '2  % -1 p ,

Cl 2 1 l' 1 1 1~ l' 1 Dj 2 1 1 4i _4 l1 l' 'l El 2 1 1 4' 4 l'.. :1 1 F] 2 1 1 :1 1' 1 -1. l ..

'l-L -Gl 2 .~ 1 1 1 1' 1-1 . . .

Hl- -3' .2- 1 l' 1 1 -- 1- 2..

I ROD NUMBER TYPE '!

l' 'UO 2 (5.00 wt.%)

2 -UO 2

-(2.38 wt. %) ,

3' UO (1. 64 :; wt. % ) -

2 4 . INERT'(ZIRC) i:

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