ML19263D596

Responds to NRC Request for Revision to Proposed Tech Specs. Forwards Revision Re Limit on Fuel Assembly Loading Rather than Limit on Fuel Enrichment

ML19263D596
Person / Time
Site:	Zion  File:ZionSolutions icon.png
Issue date:	02/23/1979
From:	Reed C COMMONWEALTH EDISON CO.
To:	Harold Denton Office of Nuclear Reactor Regulation
References
NUDOCS 7904130047
Download: ML19263D596 (21)

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an Commonwealth Edison One f asi Nat.on.v Ptam c.n.c a+ > me..s Address Reply to Post Ofhm Don 7f,NRC PUBilC DOrtn.U'NT ROW Chicago linno s 60690

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February 23, 1979

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2, '.,j,t c.' s Mr. Harold R. Denton, Director N

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Office of Nuclear Reactor Regulation N

9 U.S. Nuclear Regulatory Commission Washington, D.C.

20555 I

Subject:

Zion Station Units 1 and 2 Revision to Proposed Change to Facility Operating License Nos. UPR-39 and DPR-48 NRC Docket Nos. 50-295.and 50-304 References (a) :

April 13, 1978 letter from C.

Reed to Edson G.

Case (b) :

January 24, 1979 letter from W.

F.

Naughton to Harold R.

Denton

Dear Mr. Denton:

Per Reference (a), Commonwealth Edison Company subnitted technical specification changes to modify the spent fuel storage racks at Zion Station.

Subsequent to that submittal, the NRC Staff requested, Reference (b), that Commonwealth Edison revise the proposed technical specification change to provide a limit on fuel assembly loading rather than a limit on enrichment. to this letter contains the requested technical specification change.

In addition, the limit has been increased from 39.4 grams of Uranium-235 per axial centimeter of fuel assembly (~3.1 w/o U-235) to 40.6 grams (~3. 2 w/o U-2 3 5).

The increase to 3.2 w/o U-235 is necessary because the cion Unit I reload fuel to be roccived in August 1979 will be enriched to 3.2 w/o U-235.

In support of this change, Section 3.3 of the M 2 clear Services Corporc'.on (NSC) report submitted with Reference (a) has been modified to reflect an analysis for the 3.2 w/o U-235 fuel.

The revised pas % af that report are included in Attachment 2 which is an Addendum to the NSC Zion Licensing Report.

As was the case in the original report, Reference (a), for normal and abnormal storage and handling of fuel in the storage racks, Keff does not exceed 0.95 even if all uncertainties are applied.

'/90413CC D

e Commonwealth Edison NRC Docket Nos. 50-295/304 Mr. Harold R. Denton: February 22, 1979 The revised technical specification change (Attachment 1) and supporting documentation ( A tta c hme nt 2) have been reviewed and approved by Commonwealth Edison On-Site and Off-Site Review with the conclusion that there are no unreviewed safety questions.

For your information, also included in Attachment 2 are updated pages to the NSC report which reflect minor modifications made in the final design of the storage racks.

These modifications include:

1) Pages 3-4, 3-40, and 3-42 which show that on the south and west end of the pool the rack legs rest directly above a steel beam embedded into con-crete;

2) Page 3-6 which reflects a longer poisoned length in the absorber tube to increase manufacturing ease; and

3) Page 3-7 which reflects the actual rack height used in the thermohydraulic analysis.

Please address any additional questions that you might have to this office.

Three (3) signed originals and thirty-seven (37) copies of this letter are provided for your use.

Very truly yours, L

Cordell Reed Assistant Vice-President attachments SUBSCRIBED and. } 'gto WOR befor(me,this

, day of J f L4U /tALF

, 1979.

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x.s ATTACIDtENT 1 ZION STATION UNITS 1 AND 2 NRC DOCKET NOS. 50-295 AND 50-304 PROPOSED TECilNICAL SPECIPICATION CHANGES The following pages have been revi' sed:

298 and 299

The reactor containment structure for There are three sections of racks with each Zion Unit 2 is essentially identical in station made up of two rows.

The two design and construction to that of Unit 1 parallel rows in each section have a nominal except that it is reoriented.

Numerous center to center spacing of 21 inches and each mechanical and electrical systems section is separated by a distance of 44".

The new fuel storage vault is protected from pe ne t ra te the containment wgg through welded steel penetrations, flooding by its free flood drain.

5.4. 3 Cont a i nment Penetrations Sideways New fuel may also be temporarily stored in the spent fuel pool in preparation for refueling.

All containment penetrations (both The fuel assemblies are stored in racks in electrical and piping) are double Parallel rows, having a nominal center to barrier assemblies consisting of a l center distance of 10.35 inches in both closed sleeve, in most cases, or a directions.

This spacing is sufficient to double gasketed closure for s pec ia l l maintain a K ef fective of less than.95 when pe ne t ra t ions such as the fuel transfer flooded with unborated water, for fuel having tube.

The space between the double a maximum loading of 40.6 gms. U-235 per barriers will be continuously pres-axial centimeter of fuel assembly length (about surized, by the Penetration Pressurization 3.2 weight percent U-235).

System, to a pressure in excess of the containment design pressure. (3) 5.5.2 Spent Fuel Storage Irradiated fue assemblies will be stored References Ps Q

r(Br 30 Qi e shipment in the stainless 7

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g ui l pool which is located in (1) FSAR Section 5.1.1 the fue handl ng building.

Borated sater (2) FSAR Section 5.1.2 f,yg.;cseq.toooed p atio' ys fil the spent fuel storage pit at Q g[h 3$

(3) FSAR Section 5.1.4 I

i to match that used in the re ace"or cavity and refueling canal during 5.5 Fuel Storage refuelino operations.

The fuel is stored in 5.5.1 New Fuel Storage a vertical array with a nominal center to center spacing of 10.35" between assemblies New fuel assemblies are stored to assure a K c f fective of less than 0.95 even in a separate storage vault which if unborated water is used to fill the pit, for is designed to hold 132 new l fuel having a maximum loading of 40.6 gms. U-235 assemblies.

The new fuel storage racks a ccommoda t e 2/3 cf a core.

298

5.5.2 Soent Fuel Storage (Continued) 5.6 Seismic Design per axial centimeter of fuel assembly The structures, mechanical components and length (about 3.2 weight percent Engineered Safeguards Systema vital to U-235).

safe shutdown and containment isolation, or whose failure might cause or increase References the severity of a loss of coolant acciden' are designed pe r the seismic

1. Fuel Pool Modification Report criteria of Design Basis Earthquake (DBE).

Revision 2, dated February 3, Design Basis Earthquake is based on 1978.

ordinary allowable stresses as set forth in applicable codes, plus the additonal require-

2. Addendum to the Fuel Pool ment that a safe shutdown be made during a Modification Report dated horizontal ground acceleration of 0.179 and October 20, 1978 and revised a veritcal acceleration of 0.11g occurring February 1979.

simu ltaneou s ly.

These systems and equip-ment are defined as Seismic Class 1.

Other systems and mechanical components in

- ~' l a support or auxiliary function are A T~u O P D D i designed per the seismic criteria of

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Operational Basis Earthquake (OBE), or n

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per applicable codes.

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v' Ef ' i 4.;., g C ADDENDUM ZION LICENSING REPORT October 20, 1978 Revised February, 1979

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All dirnensions in itg inches FIGURE 3.2-2.

10 X 10 SPENT PJEL PACK ZION UNITS 1 AND 2 3-7

.G nUCLERR SER'nCES CDRPORRilon 3.3 Nuclear Design _

(for 3.2 w/o U235)

R The 10.35 x 10.35 inch spacing of the fuel assemblies is sufficient to maintain k below 0.95 for all normal and abnormal fuel storage eff conditions.

5 The a slysis results are described below.

3.3.1 Sum.ary of Results The value of keff is determined as follows:

k,ff 1 ko + ak 2

2 2

t + ak2 + ak3 + oks + (aks + aks + ak )1/2 7

x where ko = nominal calculated k,f f (2-0 diffusion theory) ak t = transport correction ak2= tube positioning errect ak 3 = rack spacing tolerance effect aks = methods bias aks = uncerta!nty in methods bias (95% confidence level) aks = channel thickness tolerance effect ak7= fuel fabrication tolerance effect The value of k,ff is maintained below 0.95 for all normal and abnormal storage conditions in accordance with Standard Review plan 9.1.2.

Results are sumarized in Table 3.3-1.

3.3.2 Method of Calculation Verification of k,77 is obtained using a two dimensional (X-Y) diffusion theo'y computer code calculation.

The calculational model covers both finite and infinite arrays of stored fuel assemblies.

Fuel neutron cross sections are developed for a four group energy range using the CHEETAH code which is an adaptation of the t.EOPARD-C DDER c0de.

XSORN anich is a one dimensional discrete ordinates spectrsi aversging code, was used for

0.

nUCLCAR SER'llCES CDRPORATICC TABLE 3.3-1 ZION, UNIT #1&2 k,ff RESULTS TERM VALUE METH00 ko 0.92629 Calculated for cell ak

.005 Estimated from similar design t

ak 0.00436 Calculated from sensitivity 2

analysis ak3 0

Since rack pitch errors cannot accumulate, cannot justify nonzero value akg

-0.0046 Critical experiment calculations aks

.0084 Critical experiments results.

957, confidence level 10 aks O

Minimum values of B density and thickness are used for absorber plate (.02 gm/cm )

ak

.0007 Calculated for 0. 01 w/o enrichment 7

increase (aks: + aks2 + sk 2) W k,ff f,ko + akt + ak2 + ak3 + ak

+

7

" 0.93928 3-9

DUCLEAR SERY1 CSS CORPORATiGn TABLE 3.3-4 FUEL DESIGN PARAMETERS Fuel Assembly Array 15 x 15 No. of Fuel Rods 204 No. of Water Rods 21 Rod Pitch 0.563" Fuel Pellet 0.0.

0.3659" Clad 0.0.

0.422" Clad Thickness 0.0243" Clad Material Zircaloy 4 Pellet Density, % T.O.

94.95 T.D.

nrichment, wt % UZ35 3.2 Max. Bundle e Nominal Active Fuel Length 144" 3-14

GUCLEAR $SRVICES CORPORAi!00

" Abnormal" refers to a single operator error.

Included as an abnomal condition is dropping of a fuel assembly.

The effect of missing absorber plates is also considered.

3.3.5 Results of Calculations The nominal results of the k,ff calculations for normal storage and handling are listed as follows:

CONDIT M k,77 1.

Nomal positioning in the spent 0.92629 fuel storage array See Figure 3.3-1, 3.3-2 2.

Eccentric positioning in the spent 0.91700 fuel storage array (clumped in groups of four) with fuel at corners of the channels.

See Figure 3.3-3A 3.

Nomal positioning in the spent 0.93065 fuel storage array with the channel offset 0.15" (.381 cm) in both X and Y See Figure 3.3-38 4

One extra fuel assembly at side of 0.93272 rack See Figure 3.3-4 The value of k,ff is not expected to change significantly under ab-nomal storage situations. Consideration was given to a fuel assembly lying on top of a rack. Due to the small axial neutron leakage reactivity of 0.002 ak, and the separation of the fuel assembly from active fuel in the rack, the increase in reactivity is <0.002.

3-15

DUCT.CRR SERV!CES CORP 0997100 The drop of a fuel assembly onto the rack was considered.

The fuel rack is designed to prevent any plastic deformation in the fuel region for these loads. Therefore, the reactivity effect of these abnormal loading conditions is insignificant.

The fact that the B C in the BCRAL core is not mixed with 4

Al homogeneously is also considered.

The effect is found to be small.

Coupled with the positive bias of our method of calculation, we believe that the slight reactivity effect of this so-called " grain effect" can be neglected without losing conservatism.

The effect of " missing" absorbe plates was evaluated (see Figures 3.3-5, 3.3-6).

The reactivity worth of 1

" missing" absorber plate per 4 tubes (16 plates) is 0.0092 (ak = 0.0079).

The reactivity worth of 1 " missing" absorber plate per 32 plates is o.005:' (6 k - o.0045), see Figure 3.3-7.

The Quality Assurance Program for the manu-facture of the boron plates and the fabrication of the absorber tubes will confirm and document that each absorber tube includes absorber plates of the required boron content.

This rack design is able to satisfy keff < 0.95, even if condition 4 and all other uncertainties shown in Tsble 3.3-1 are applied at the same time (keff < o.94571). Al though Paragraph II. l.2 of the

'iRC " Position for Review and Acceptance of Soent Fuel Storage and Handling Applications" issued April 14, 1978, allows consideration of the presence of soluble boron in the pool water for condition 4, this analysis conservatively does not take credit for the soluble boron.

Sensitivity studies were performed to evaluate the influence on k

of fuel tube spacing and pool temperature.

The results of eff these analyses are surrnarized below.

3-16

nUCt. ERR SCR41CCS CORPOR AT1Gn TABLE 3.3-5 k,ff VERSUS SPACING, 40*F Cell Pitch, in.

k,7f 9.956 x 9.956 0.9814 10.35 x 10.35 0 '263 10.743 x 10.743 0.8849 11.137 x 11.137 0.8538 See Figure 3.3-8 3-18

DUC1.ER.9 SER's:CES CO.100R47100 TABLE 3.3-6 eff Infinite lattice Temoera ture, F 0.92629 40 0.91756 130 0.90566 212 3-19

REFLECTIVE WATER

--- 1.54432cm TUBE 0.5511Scm WATER 0.70778 cm REFLECTIVE FUEL 21.4445cm

- REFLECTIVE.

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0.70778 cm 0.5511Sc:n 0.78232cm RE. LECTIVE FIGURE 3.3-25.

RACK TUBE OFFSET 3-23

1.1 -

1. 0 -

k rr e

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=

0.8 -

i g

-2cm

-icm 0

+1cm

+2cm

+3cm (9.956")

(10.35")

(10.743")

(11.137")

CllANGE Ill PITCil (80 Tit'liOR120tiTAL DIRECTI0 tis) k rr vs. Change in Pitch for 3.2 w/o U235 Figure 3.3.8 e

GUCt. EAR SERY1 CSS CORPORATI0n TABLE 3.4-2 STRESS t.ALVATION FOR MOST CRITICAL RACK ELEMENTS Rack Critical Critical Load Allowable Computed Component Stress Type Combination Stress Stress (ksi)

(ksi) 0+B+E 18.60 9.10 Pm 0 + B + E' 27.90 11.29 Tube Wall 0+B+E 27.90 9.23 Pm+b 0 + B + E' 41.85 11.31 ll) 55.80 (1) 13.83 D+B+Q+To+E n+b+P 0 + B + Q + 4 + E,(2)

(2) e g

0+B+E 18.60 5.29 p

m 0+B+E' 27.90 7.01 D+B+E 27.90 5.37 Base p

Plate m+b 0 + B + E' 41.85 7.09 III 0 + B + 0 + To + E 55.80 5.44 P

0 + B + 0 + To + E,(2)

,,, (2) m+b e

7.32 0+B+E 18.60 9.54 m

Rack 0 + B + E' 27.90 10.88

'89 D+B+E 27.90 15.17 l

pm+b 0 + B + E' 41.85 19.70 Notes:

(1) For ASME Class 2 support structures, evaluation is not required for this load combination. However, evaluation was perfomed conservatively using criteria similar to those for ASME Class 1 support structures.

(2) Evaluation is not required for component support structures designed according to ASME,Section III, Subsection NF.

Hcwever, stresses were computed for comparison purposes and to show additional argin of aafety.

3 40

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3. 5 EVALUATION OF POOL STRUCTURES 3.5.1 Description of Pool Structures The spent fuel pool is a rectangular box-type structure n

open at the top (see Figure 3.1-2).

The structural part of the pool consists of a concrete floor and four thick reinforced concrete walls.

The cask handling area is located at the north west corner of the pool and is separated from the storage area by two thick walls.

The inside of the pool including the cask handling area is 63 feet long in the north-south direction and 33 feet in the east west direction.

Oepth of pool is 41 feet.

Pool floor and walls are lined with 3/16 inch continuously welded stainless steel liner anchored rigidly to the concrete floor.

There are 31 stainless steel plates embedded in the floor concrete, and the proposed racks would be placed on these plates as shown in Figure 3.1-2.

These plates would transfer the loads frcm the racks to the concrete floor slab.

At the south and west ends of the pool, where there are no embedded plates, the pro-posed rack arrangement is such that the legs of racks would rest directly above a steel beam embedded into

oncrete.

3.5.2 Loads, load Combinations and Structural Acceptance Criteria 3.5.2.1 Loads - Fuel pool structures were evaluated for the following loads:

a)

Deadweight (0) -- Weight of pool structure com-ponents including the buoyant weight of the proposed spent fuel racks.

b)

Live load (L) -- Live load on the operating floors which are supported on the pool walls.

3-42

^

0320

UNITED STATES OF AMERICA D1Tra COWsPo NUCLEAR REGULATORY COMMISSION In the Matter of I

)

COMMONWEALTH EDISON COMPANY

)

p (Zion Station Units 1 and 2)

)

Docket Nor.f~ -295 N

)

50-37&p

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Proposed Amendments to

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k

[2 Increase Spent Fuel Storage

)

MAR 91978 >. 1 1

Cacacity

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( 4 '3 F. R. 30938)

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March 2, 1979 D'

9 lxl Edward Luton, Chairman Dr. Forrest J.

Remick Atomic Safety and Licensing 305 East Hamilton Avenue Board Panel State College, Pennsylvania 16801 U.S.

Nuclear Regulatory Commission Washington, DC 20555 T : san N.

Sekuler AJsistant Attorney General D:. Linda W.

Little 138 West Randolph Street Research Triangle Institute Suite 2315 P.O.

Box 12194 Chicago, Illinois 60601 Research Triangle Park, North Carolina 27709 Atomic Safety and Licensing Board Par.el Docketing and Service U.S.

Nuclear Regulatory Commission Section Washington, DC 20555 U.S.

Nuclear Regulatory Commission Richard Konter Washington, DC 20555 617 Piper Lane Lake Villa, Illinois 60046 Richard Goddard Office of the Executive Legal Director U.S.

Nuclear Regulatory Commission Washington, DC 20555 To Members of the Licensing Board and All Parties of Record:

Attached is a copy of a revision to Commonwalth Edison Company's license amendment request in the above matter.

This revision makes a minor change in the enrichment limit for fuel which will be stored in the Zion Spent Fuel Pool.

S in c er,ely,

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PPS/bjn Enclosure