ST-HL-AE-1957, Forwards marked-up FSAR Sections 4.3.2.6 & 9.1.3 & Response to NRC Question 220.32N Reflecting Revs Resulting from Planned Replacement of Current Spent Fuel Racks w/high- Density Spent Fuel Racks

Forwards marked-up FSAR Sections 4.3.2.6 & 9.1.3 & Response to NRC Question 220.32N Reflecting Revs Resulting from Planned Replacement of Current Spent Fuel Racks w/high- Density Spent Fuel Racks

ML20205G866
Person / Time
Site:	South Texas
Issue date:	03/26/1987
From:	Wisenburg M HOUSTON LIGHTING & POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
ST-HL-AE-1957, NUDOCS 8703310564
Download: ML20205G866 (26)

Text

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The Light NE Uf Ilouston I.ighting & Power Ito. Box 1700 Ilouston, Texas 77001 (713) 228-9211

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March 26, 1987 ST-HL-AE-1957 File No.: G9.1 10CFR50 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555 South Texas Project Units 1 and 2 Docket Nos. STN 50-498, STN 50-499 Annotated FSAR Revisions Concerning the Incorporation of High Density Spent Fuel Racks Enclosed for your review are mark-ups of FSAR sections 4.3.2.6 and 9.1.3 and the response to NRC Question 220.32N which reflect revisions resulting from the planned replacement of our current spent fuel racks with high-density spent fuel racks. These revisions will be incorporated into a future FSAR amendment and are provided for the Staff's immediate use in the preparation of the Safety Evaluation Report.

Four rack modules wit.h 14 in. center-to-center spacing are currently installed in Unit 1.

These racks provide sufficient storage space to accept new fuel deliveries. These racks are also available to accept spent fuel during low-power testing and the early part of Cycle 1 should any or all fuel need to be discharged. These racks will be removed from Unit 1 and replaced with Region 1 high-density racks during Cycle 1.

Prior to installation of the high-density racks, the results of seismic and criticality analyses demonstrating that these racks meet the requirements contained in Standard Review Plan section 3.8.4 Appendix D will be submitted for NRC review.

We do not plan to install the 14 in, rackr in Unit 2.

Instead, for Unit 2 only, we plan to install and use the Region 1 high-density racks for both new fuel dry storage and spent fuel wet storage.

8703310564 870326 DR ADOCK 0500 8

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ST-HL-AE-1957 Houston Ughting 8e Power Cannpany File No.: G9.1 Page 2 The high-density spent fuel racks will consist of two separate regions in each unit's spent fuel pool. Region 1 will accept up to 288 spent fuel assemblies (1 1/2 cores) at the highest enrichment anticipated for use in STP.

This region has removable Boraflex assemblies in water boxes between each cell. This allows a 10.8 in. center-to-center spacing without taking credit for fuel burnup.

Figure 1 shows a representation of the design of these racks.

Region 2 will accept up to 1681 spent fuel assemblies (8 2/3 cores) at the highest enrichment anticipated for use in STP.

In this **E R

fuel burnup is taken.

This region has a fixed Boraflex sheet between each rack cell which allows a 9.15 in. center-to-center spacing.

Figure 2 shows a representation of the design of these racks.

Placement of assemblies in Region 2 racks will be administrative 1y controlled. A change to Technical Specification 3/4.9.1 and the Design Features Technical Specification will be revised to include the following statement:

The spent fuel racks are designed and shall be maintained so that the calculated effective multiplication factor is no greater than 0.95 (including all known uncertainties) when flooded with demineralized water.

Storage in Region 1 of the racks is restricted to fuel assemblies having initial unrichment less than or equal to the design enrichment for the racks (4.5 weight percent U-235).

Storage in Region 2 is restricted to those assemblies in the " acceptable" domain of Figure X.

(This figure will show storage acceptance criteria: minimum required burnup as a function of each assembly's initial enrichment and will be provided once the criticality analysis is completed.) A complete record of the analyses employed in making the selections will be kept so long as the assemblies remain in the Region 2 racks.

If you should have any questions on this matter, please contact Mr.

M. E. Powell at (713) 993-1328.

s W

M. R. Wis nburg Deputy Project Manager MEP/KDS/yd

Attachment:

(1) Annotated FSAR Revisions Concerning High Density Spent Fuel Racks (2)

Figure 1 (3)

Figure 2 Ll/NRC/bd

ST-HL-AE-1957 File No.: C9.1 Houston Lighting & Power Company Page 3 cc:

Regional Administrator, Region IV M.B. Lee /J.E. Malaski Nuclear Regulatory Commission City of Austin 611 Ryan Plaza Drive, Suite 1000 P.O. Box 1088 Arlington, TX 76011 Austin, TX 78767-8814 N. Prasad Kadambi, Project Manager A. von Rosenberg/M.T. Hardt U.S. Nuclear Regulatory Commission City Public Service Board 7920 Norfolk Avenue P.O. Box 1771 Bethesda, MD 20814 San Antonio, TX 78296 Robert L. Perch, Project Manager Advisory Committee on Reactor Safeguards U.S. Nuclear Regulatory Commission U.S. Nuclear Regulatory Commission 7920 Norfolk Avenue 1717 H Street Bethesda, MD 20814 Washington, DC 20555 Dan R. Carpenter Senior Resident Inspector / Operations e/o U.S. Nuclear Regulatory Commission P.O. Box 910 Bay City, TX 77414 Claude E. Johnson Senior Resident Inspector /STP c/o U.S. Nuclear Regulatory Commission P.O. Box 910 Bay City, TX 77414 M.D. Schwarz, Jr., Esquire Baker & Botts one Shell Plaza Houston, TX 77002 J.R. Newman, Esquire Newman & Holtzinger, P.C.

1615 L Street, N.W.

Washington, DC 20036 T.V. Shockley/R.L. Range Central Power & Light Company P. O. Box 2121 Corpus Christi, TX 78403 Ll/NRC/bd Revised 2/3/87

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ATTACHMENT /

STP FSAR ST-HL AE.1951 PAGE l OF /9 A representative calculation for two banks of control rods withdrawn simultaneously (rod withdrawal accident) is given on Figure 4.3-37.

)

Calculation of control rod reactivity worth versu.s time following reactor trip involves both control rod velocity and differential reactivity worth.

The rod position versus time of travel after rod release assumed is given on figure 4.3-38.

For nuclear design purposes, the reactivity worth ver-cus rod position is calculated by a series of steady state calculations et various control rod positions assuming all rods out of the core as the initial position in order to minimize the initial reactivity insertion rate. Also to be conservative, the rod of highest worth is assumed stuck out of the core and the flux distribution (and thus reactivity impor-tance) is assumed to be skewed to the bottom of the core. The result of these calculations is shown on Figure 4.3-39.

The shutdown groups provide additional negative reactivity to assure an adequate shutdown margin. Shutdown margin is defined as the amount by which the core would be subcritical at hot shutdown if all RCCA's are tripped, but assuming that the highest worth assembly remains fully with-drawn and no changes in menon or boron take place. The loss of control rod worth due to the material irradiation is negligible since only bank D may be in the core under normal operating conditions.

The values given in Table 4.3-3 show that the available reactivity in withdrawn RCCA's provides the design bases minimum shutdown margin allow-ing for the highest worth cluster to be at its fully withdrawn position.

An allowance for the uncertainty in the calculated worth of N-1 rods is made before determination of the shutdown margin.

}

4.3.2.6 Criticality of the Reactor During Refueling and Criticality of Fuel Assemblies. The basis for maintaining the reactor suberitical during refueling is presented in Subsection 4.3.1.5 and a discussion of how control requirements are met is given in Subsections 4.3.2.4 and 4.3.2.5.

Criticality of fuel assemblies.outside the reactor is precluded by ade-i quate design of fuel transfer'and fuel storage facilities and by admin-istrative control procedures. This section identifies those criteria j

43,s. 6.plim ortant to criticality safety analyses.

i l.

New Fuel Storage,

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fu1isstoredin21-in. center-to-centerracksinthenew' fuel 5torage facilities in a dr condition. Prior to initial core loading, new fuel can be stored dey-the 14 in. h center-to-center spent fuel racks.

l7 For the flooded condition (with unborated water assuming new fuel of the highest cnticipated enrichment (3.5 w/o 2 U-235] in the new or spent fuel racks) the l7 effective multiplication factor does not exceed 0.95.

For the normally 6

dry condition the effective ~ multiplication factor does not exceed 0.98 (with fuel of the highest anticipated enrichment in place and assuming..

possible sources of moderation such as aqueous foam or mist).

/pse.In A

..I 4.3-30 Amendment 7,7/16/79

ATTACFDAE T l ST HL AE-1 B'l PAGEst 0 //

Insert A to Section 4.3.2.6.1 For Unit 2, new fuel is stored in 21-in. center-to-center racks 15 the new fuel storage facilities in a dry condition.

Prior to initial core loading, new fuel can be stored dry in the 10.8-in. center-to-center high density spent fuel racks using alternate cells in a checkerboard fashion. For the normally dry or flooded condition (with unborated water assuming new fuel of the highest anticipated enrichment [3.5 w/o % U-235] in the new or [3.0 w/o % U-235] in the high density spent fuel racks), the effective multiplication factor does not exceed 0.95.

8192N:0302N

ATTACHMENT I STP FSAR ST-HL AE 195 7 l PAGE 3 0F/g In the analysis for the storage facilities, the fuel assemblies are I

assumed to be in their most reactive condition, namely fresh or I

undepleted and with no control rods or removable neutron absorb 1rs present. Credit is taken for the inherent neutron-absorbing effect of the construction materials of the racks. Assemblies cannot be closer 6

together than the design separation provided by the storage facility, except in special cases such as in fuel shipping containers where analyses are carried out to establish the acceptability of the design.

The mechanical integrity of the fuel assembly is assn =at 4'3.2 6 2.

16 +ha-14 '"

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sve.nt %*t SP*at M.shv 15

1. + Spent Fuel StorageUda+ 1 (2 nteem D4ssjn)

• m'rd PN *' +b "f Nn' 3.2.6. 2. I AThe followinR descrihoc uptufuel storage in the spent fuel poolk Un-I83b &,

borated water of 1.0 g/cc is assumed in the analysis. Over the range of 10.8in,hi3h S

water densities of interest (corresponding to 60*F through 212*F), full dangdy spu4 density water is a conservative assumption since a decrease in water den-Qg sity will cause the effective multiplication factor (k,fg);of the system j

to decrease. I: he_13 5: - ;'::i :d the: i:ill ; i: ::. :::ittr' *^

r:d:: ac.y J:: ret::: r The design basis for wet fuel storage criticality analysis is that, con-l sidering possible variations, there is a 95 percent confidence level that the effective multiplication factor (k of the fuel storage array will be less than 0.95 per ANSI Standard N1Ib-)1973.

i The possible variations in the criticality analyses are in three categories: 1) calculational uncertainties, 2) fuel rack fabrication uncertainties, and 3) transport effects.

The results of comparing standard mentsassummarizedinTable4.3-4g1 ions with 101 critical experi-indicate that:

1.

The average difference between the calculations and experimental results or bias in the computations, was 0.1 percent ok which is denoted as the calculational bias, and 2.

The standard deviation in the difference between the calculations and experimental results was 0.86 percent Ak. Multiplying the stan-dard deviation by the appropriate one-sided upper tolerance factor results in a calculational uncertainty valid at the 95 percent confidence level.

Reactivity uncertainties corresponding to fuel rack fabrication uncer-tainties are defined as follows:

1.

The clearance spacing to permit insertion of the fuel assemblies into the rack is equivalent to a clearance reactivity.

2.

The reactivity associated with the tolerance on the position of full length structural support members within a given cell is denoted as cell reactivity.

4.3-31 Amendment 6, 6/11/79

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ATTACHM i

ST HL-AE

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STP FSAR

--FAGE 4/

3.

The reactivity associated with the tolerance on center to center spacing l

between cells is denoted as center-to-center (c-c) reactivity. _

5 Comparing corresponding transport and diffusion results at several center to center spacing allows a transport bias to be calculated by the following equation.

ggg(Transport)-k,gg(Diffusion)

Transport Bias =

K,gg (Transport) x 100 l

The three types of uncertainties described above are statistically combined in the following equation:

1 Total Uncertainty - Transport Bias + Calculational Bias +

{(CalculationalUncertainty)2,g,,,

Reactivity)2 + (Cell Reactivity)2,

(Clearance Reactivity)2) 1/2 In evaluating the above expression for an infinite array of 17x17 assemblies l

enriched to 3.5 w/o uranium-235, the total uncertainty is calculated to be 3.6 Ak Subtracting 0.036 from 0.95 gives an effective multiplication percent factorof0*II4whichresultsinacentertocenterspacingof14.0in.

1 Accordingly, as 14.0-in. center-to-center rack spacing corresponds to at least l

95 percent of the time k will not exceed 0.95 at a 95 percent confidence j

level. Therefore,the1%g(n. center-to-centerdesignconservativelymeetsthe 6

-E specification of ANSI Standard N18.2-1973.

Verification that the design criteria for fuel storage are met is achieved I

through the use of standard Westinghouse design methods such as the LEOPARD and FDQ codes. It should be noted that on the basis on evaluations of UREC coupled-core experiments [4.3-15] it has been concluded that there is no need l

to apply a transport bias to diffusion theory results obtained using the LEOPARD /FDQ computational system. However, for additional conservatism, Vestinghouse will continue to include a transport bias in the calculation of total uncerta gty.

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4.3.2.7 Stability l

l 4.3.2.7.1

==

Introduction:==

The stability of the FUR cores against zenon-induced spatial oscillations and the control of such transients are discussed extensively in References 4.3-8, 43.-16, 4.3-17 and 4.3-18.

A summary of these reports is given in the following discussion and the design bases are give in Section 4.3.1.6.

In a large reactor core, menon-induced oscillations can take place with no l

corresponding change in the total power of the core. The oscillation may be caused by a power shift in the core which occurs rapidly by comparison with the menon-iodine time constants. Such a power shift occurs in the axial direction when a plant load change is made by control rod notf.on and results in a change in the moderator density and fuel temperature distributions.

Such a power shift could occur in the diametral plane of the core as a result of abnormal control action.

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• 4.3-32 Amen'dment 52 m.

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Af f ACHMENT /

ST-HL AE /457 PAGE 6 OF/f Insert B 4.3.2.6.2.2 Unit 1 and Unit 2 (Final Design): The following describes wet,

spent fuel storage in the spent fuel pool in both Unit 2 and Unit 7 after the replacement of the 14-in. racks. Two storage regions are provided in the spent fuel pool including a small region for non-credit for burnup storage.

Region 1 Design The Region 1 racks have a 10.8 in. center-to-center spacing with locked removable poison assemblies between the cells.

This region is conservatively designed to accommodate unirradiated fuel at enrichments to 4.5 weight percent. The criticality analysis of the racks is performed with the state-of-the-art LEOPARD /PDQ-7 code package. These codes have been benchmarked against critical experiments in the range of pellet diameters, water-to-fuel ratios and enrichments that encompasses the STP design. This benchmarking led to the conclusion that the calculational model is capable of determining the multiplication factor of the Region 1 racks to within 1-percent reactivity change with a 95-percent probability at the 95-percent confidence level.

The effects of various mechanical and thermal uncertainties will be analyzed using this model. These include pool-water temperature, placement of assemblies in storage boxes, stainless steel thickness variation, and UO2 pellet density. Combining these uncertainties with the calculational uncertainty yields a multiplication factor for the Region 1 racks, when loaded with fuel assemblies of 4.5 weight percent U-235, of less than 0.95, thus meeting the requirement of GDC 62. Therefore, any number of fuel assemblies of the Westinghouse 17 x 17 XLR design having enrichments no greater than 4.5 weight percent U-235 may be stored in Region 1 of the racks.

Region 2 Design The Region 2 racks have a 9.15 in. center-to-center spacing with fixed poison material surrounding each cell. This region is designed to accommodate fuel that has attained the required burnup. The same code package was used for the multiplication factor calculations as was used in the Region 1 analysis. The ability to calculate the isotopic composition of burned fuel was verified by comparing the LEOPARD calculation to the measured results of irradiations performed on U02 fuel in the Yankee-Rowe reactor and on mixed oxide fuel in Saxton.

Similar evidence was used to assess the fission product bui.ldup uncertainty and its reactivity effect and the reactivity effect of the transuranium isotopes.

The various uncertainties ar,e summed to obtain a total calculational uncertainty of 2.17 percent reactivity change with a 95-percent probability at the 95-percent confidence level.

In addition, sensitivity studies were performed to obtain the thermal and mechanical uncertainties.

Combining these uncertainties with the calculational uncertainty will yield a value for the multiplication factor for Region 2 racks when loaded with irradiated Westinghouse 17 x 17 XLR fuel assemblies.

In order to establish 8191N:0302N/l

Af fACHMENT /

ST-HL ' E l'i5 7 PAGE 1 OF/8 burnup criteria for fuel storage in Region 2, the infinite multiplication factor (at 68"F with minimum postshutdown fission product inventory) of a fuel assembly as a function of burnup is obtained for a number of initial j

enrichments. The multiplication factor of the storage racks is next j

calculated as a function of the infinite multiplication factor of the stored assemblies for the various initial enrichments.

The assembly infinite i

multiplication factor which produces a rack multiplication factor of 0.95 less all uncertainties is obtained from these curves as a function of initial i

enrichment. Finally, a curve is drawn of the burnup required to obtain the specified rack multiplication factor as a function of initial fuel enrichment.

Any number of Westinghouse 17 x 17 XLR fuel assemblies with burnups in excess of the minimum acceptable burnup may be stored in Region 2 of the spent-fuel storage racks.

The decision as to whether a particular assembly is to be placed in Region 2 of the racks is made under administrative control. After the assemblies to be discharged from the core have been placed in the Region 1 racks, an analysis of the burnup of each assembly will be made. This analysis will make use of (1) the records which show the location of each assembly at all times since its arrival on site and (2) core operating histories and power distributions while the assembly was in the reactor. The re:Ord: Of fue! ::: mbly lec: tier 07: ;;;;r:t:d 5; f:11:w',g r-itten, previce:!y :pprcred precedure:.

The burnup history Of c ch :::cebly is integrated tc crriv: et it total burnup.

The burnup value is then compared to the storage acceptance criterion in Section 5.6 of the Technical Specifications in order to permit or deny storage in Region 2.

8191N:0302N/l 1

AITACHM Tl ST hLE / 57 7O M STP FSAR

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i 9.1 FUEL STORAGE AND HANDLING Facilities for the receipt and storage of new fuel and the storate and tran' -

s for of spent fuel are housed in the Fuel-Handling Building (FHB). A separate and independent FHB is provided for each unit of the STP. Each FHB is designed as a controlled-leakage seismic Category I structure. The design of i

Tthe FHB Heating, Ventilating and Air Conditioning System is discussed in Section 9.4.2.

The structural design considerations are described in Section

,3.8.4.

9.1.1 New Fuel Storage 9.1.1.1 Design Bases. The new fuel storage pit is a reinforced concrete pit and an integral part of each seismic Category I FHB. This pit provides tamporary dry storage for approximately one-third core (66 fuel assemblies) of new fuel. The fuel is stored in racks (Figure 9.1.1-1) composed of individual vertical cells fastened together to form three 2 x 11 modules which may be bolted to anchors in the floor and walls of the new fuel storage pit. The new fuel racks are classified as seismic Category I components, as defined by Regulatory Guide (RC) 1.29, and American Nuclear Society (ANS) Safety Class (SC) 3 (see Section 3.2).

l j

The new fuel racks are designed with a center-to center spacing of 21 in.

This spacing provides a minimum of 12 in, between adjacent fuel assemblies.

This separation is sufficient to maintain a suberitical array assuming optimum moderation.

Space between storage positions is blocked to prevent insertion of fuel. All rack surfaces that come into contact with the fuel assemblies

(

are made of annealed authentic stainless steel, and the support structure is painted carbon steel.

The racks are designed to withstand normal operating loads, as well as to remain functional with the occurrence of a Safe Shutdown Earthquake (SSE).

l The new fuel racks'are designed to withstand a maximum uplift force of 5,000 pounds and to meet the requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code,Section III, Appendix XVII.

The new fuel storage pit access hatch is a three-section cover. This cover will minimize the introduction of dust and debris into the pit. The cover is designed to withstand the impact force of a new fuel assembly dropped from the maximum elevation allowed by the 2-ton hoist of the FHB overhead crane.

In addition, space is provided for the storage of fuel during refueling inside the Reactor Containment Building (RCB). See Section 9.1.2.1 for a description 53 of the racks.

9.1.1.2 Facilities Description. The IRB abuts the south side of the RCB l53 and is adjacent to the west side of the Mechanical-Electrical Auxiliaries Building (MEAB) of each unit. The locations of the two IHBs are shown in the station plot plan on Figure 1.2-3.

For general arrangement of the new fuel storage facilities, refer to Figures 1.2-39 through 1.2-48.

l53 New fuel assemblies are received in the receiving area of each FNB and tempo.

rarily stored in the shipping containers in the new fuel handling area. In

./

the new fuel handling area, each new fuel assembly is removed from its ship-ping container and inspected visually to confirm the assembly has not been 9.1-1

'A'sendment 53 go 0.kaM

AFTACHM i

ST-HL AE-PAGE_J O STP FSAR damaged during shipment. The new fuel assemblies ars transported from the

, inspection area to the new fuel storage pit or to the new fuel elevator by the 15/2-ton, dual-service FHB crane. The 2-ton hoist of this crane is designed to handle new fuel assemblies. New fuel handling is discussed in detail in Section 9.1.4.

Use of the 2-ton hoist of the 15/2-ton crane or of the fuel-handling machine to handle new fuel ensures that the design uplift of c1w racks will not be exceeded.

The new fuel storage pit is situated in the approximate center of each FHB.

he floor of the new fuel storage pit is at El. 50 ft 3 in. H e new fuel storage pit access hatch is provided with a three-section protective cover at El. 68 ft.

The fuel assemblies are loaded into the new fuel storage racks through the top and stored vertically.

9.1.1.3 Safety Evaluation. Units 1 and 2 of the STP are each provided with separate and independent fuel handling facilities.

Fload protection of each FHB is discussed in Section 3.4.1.

Flooding of the new fuel storage pit from fluid sources inside either FHB is not considered credible since all fluid systems components are located well below the eleva-tion of the new fuel storage pit access hatch. A floor drain is provided in the new fuel storage pit to minimise ecliection of water.

S e applicable design codes and the ability of the FHB to withstand various i

external loads and forces are discussed in Section 3.8.4 Details of the seismic design and testing are presented in Section 3.7.

Missile protection of the FHBs is discussed in Section 3.5.

Failure of nonseismic systems or structures will not decrease the degree of subcriticality provided in the new fuel storage pit.

In accordance with ANSI N18.2, the design of the normally dry new fuel stora&*

2 racks is such that the effective multiplication factor will not exceed 0.98 with fuel of the highest anticipated enrichment in place, assuming optimum Q10.

moderation (under dry or fogged conditions). For the unborated flooded condi-7 tion, assuming new fuel of the highest anticipated enrichment in place, the effective multiplication factor does not exceed 0.95.

Credit may be taken for the inherent neutron - absorbing effect of the materials of Construction.

The new fuel assemblies are stored dry, the 21-in. spacing ensuring a safe geometric array. Under these~ conditions, a criticality accident during refu-eling and storage is not considered credible.

safety analysis is discussed in Section 4.3.

Consideration of criticality Design of the facility in accordance with RC 1.13 ensures adequate safety under both normal and postulated accident conditions. The new fuel stora&*

racks also meet the requirements of General Design Criterion (CDC) 62.

l44 9.1.2 Spent Fuel Storage 9.1.2.1 Design Bases. He spent fuel pool is a stainless steel-lined reinforced concrete pool:and is an integral part of each FHB. All spent fuel racks are classified as seismic Category I, as defined by RC 1.29, an! as ANS 23 SC 3.

53 9.1 2

[uliendment 53 Ao M

EkdHENI /

t/nir I (Idem. DamA GTSL e '9D9 i

mWp STP FSAR p.,4,'i&%1 roelkg anJ 4es+mp et> Tae W<. per os erje 4 i

g Afne hundred ninety-six (196) storage spaces are provided in the IRB spent, fuel 1

53 pool in four 7 x 7 modules. These modules have 14-inch center-to-center j

spacings (Figure 9.1.2-lb) and are " free standing" in that they rest on vertical shear pins attached to adapter pla'tes that are bolted to the spent fuel pit floor. Spaces between storage cells are blocked to prevent improper

insertion of fuel.

i I

1. 7 th the 14-in. and 16 in. (described belowT spacings provide sufficient 23 g sE 6. separation between fuel assemblies to maintain a suberiticaL. array rack

- surfaces that cose into contact witn fuel assemblies are made of anneale

(, [ norm"al and emergency water quality conditions.*]These z - ti; stainless steel.

-A-CThe racks are designed to withstand normal operating loads as well as to remain functional with the occurrence of an SSE. The racks are designed with adequate energy absorption capabilities to withstand the impact of a dropped I'l,W-spent fuel assembly from the maximum lift height of the spent fuel pit bridge

• "O hoist.

The racks are designed to withstand a maximum volift force eaual en l Mo~/*'-

the uplift force of the bridge hoist. The{racksalso et the requirements of ASME Code,Section III #ppedy D 4 5AF 3 9 V Appendix XVII.Tht. kn daws's 3PgAtr Fup4.(E4 Cad 5 MEE.7 THE cetTael4 cF Shielding for the spent fuel pool is adequate to protect plant personnel from exposure to radiation in excess of published guideline values as stated in Section 12.1.

A depth of approximately 10 ft of water over the top of the 53 spent fuel assemblies will limit direct radiation to 2.5 mE/hr (surface dose rate).

(

h e FHB Ventilation Exhaust System is designed to limit the offsite dose in the event of a significant release of radioactivity from the fuel, as dis-cussed in Sections 12.3.3, 15.7.4 and 9.4.2.

The FHB is designed to prevent missiles from contacting the fuel. A more j

detailed discussion on missile protection is given in Section 3.5.

In addition, space is provided for storage of fuel during refueling inside the RCB for 64 fuel assemblies in four 4~x 4 modules having 16-inch center-to-cen-i 53 i

ter spacing (Figure 9.1.2-la).

These modules are firmly bolted in the floor.

I i

9.1.2.2 Facilities Description. The FHB abuts the south side of the RCB and is adjacent to the west side of the MEAR of each unit. The locations of i

the two IMBs are shown in the station plot plan on Figure 1.2-3.

For general i

arrangement of the spent fuel storage facilities, refer to Figures 1.2-39 through 1.2-48.

853 I

i i

h e spent fuel storage facilities are designed for the underwater storage of spent fuel assemblies and control rods after their removal from the reactor vessel, h e spent fuel is transferred to the FHB and handled and stored in 16 s

l the spent fuel pool underwater. The fuel is stored to permit some decay, then l7 trannferred offsite. For a detailed discussion of spent fuel handling, see l6 Section 9.1.4.

He spent fuel pool is located in the northwest quadrant of esc. IHB. The floor of the pool is at El. 21 ft 11 in., with normal water level at El. 66 ft 6 in. The top of a fuel assen51y in a storage rac g not extend above the

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topofthestoragerackwhichisEl.39ft.3in{

' dlin th: ::=egW l6 105 2 M 1-essemblic: i: 2e.oemmodeced i.. i:.; p ;1.

The fuel assemblies 53 7....

are loaded into the spent fuel racks through the top and are stored vertical jy i

ly.

16 9.1 3 Amendment 56 k

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Af TACHMENT l ST bL AE 1957 PAGL /0 OF /8' Insert C to FSAR Section 9.1.2.1 Unit-1-and Unit 2 (Final Design)

The spent-fuel-storage facility provides storage capacity for 1969 high density poison spent fuel racks in a honeycomb array in each unit. Two storage regions are provided in the spent fuel pool with a small region for non-credit for burnup storage. There are 288 non-credit for burnup storage cells (Region 1) and 1681 credit for burnup storage cells (Region 2). Figure 9.1.2-2 shows the pool layout for both Units 1 and 2.

The four Region I rack modules are located in the northwest corner of the spent fuel pool.

The Region 1 racks have a 10.8 in. center-to-center spacing with removable poison assemblies between the cells. This region is conservatively designed to accomodate unirradiated fuel at enrichments to 4.5 weight percent.

Region 1 storage cells are each bounded on four sides by a water bgx except on the periphery of the pool. The neutron poison material (Boraflex K)j$

located in these water boxes. This is accomplished by capturing two sheets of the poison material on two outside opposite faces of a thin-walled rectangular box, which is welded to structure under a lead-in guide. The poison sheets are captured under thin stainless steel sheets which are intermittently welded all around to the thin-walled rectangular box. A locking device engages the structure under the lead-in guide to hold the assembly in place. Special tools are provided for unlocking, removing, re-installing and locking this poison assembly. The axial location of the poison with respect to the active fuel region is provided and maintained by this welded assembly structure.

Figure 9.1.2-3 shows a typical Region 1 spent fuel rack.

During refueling, the fuel assemblies which have been removed from the reactor are placed in Region 1 where they are inventoried. The history of each fuel assembly is reviewed, and calculations are performed to determine the burnup of each assembly. Once it has been determined that a fuel assembly has attained the required burnup, it is added to the list of assemblies designated for movement to Region 2; otherwise it remains in Region 1.

The Region 2 racks have a 9.15 in. center-to-center spacing with fixed poison material surrounding each cell. This region is designed to accomodate fuel that has attained the required burnup. A sheet of neutron poison material is captured between the side walls of all adjacent boxes. To provide space for l

the poison sheet between boxes, a double row of matching flat round raised areas are coined into the side walls of all boxes. The raised dimension of these locally formed areas on each box wall is half the thickness of the poison sheet. The boxes are welded together at all these local areas. The poison sheets are scalloped along their edges to clear these areas.

Figure 9.1.2-4 shows a typical Region 2 spent fuel rack.

The axial location of the poison with respect to the active fuel region is provided and maintained by the structure of each box. At the outside periphery of each rack, a sheet of poison material is captured under thin stainless sheets which are intermittently welded all around to the box.

8190N:0302N

7 Al l ACHME.N T I STMLM /9 STP FSAR PAGL /l OF 9.1.2.3 Safety Evaluation. Units 1 and 2 of the STP are each provided with separate and independent fuel handling facilities.

Flood protection of each FHB is discussed in Section 3.4.1.

A detailed discussion of missile protection is provided in Section 3.5.

Th[applicabledesigncodesandthevariousexternalloadsandforcesconsid-cred in the design of the FHB are discussed in Section 3.8.4.

Details of the coismic design and testing are prosented in Section 3.7.

Design of this storage facility in accordance with CDC 62 and RC 1.13 ansures o safe condition under normal and postulated accident conditions. The DEM6M CF M easter-te=cenese> die ; m-e Mr:n n",:::n spent fuel assemblies is sufficient to ensure K

<0.95 oven if unborated water is used to fill the spent fuel pool. The $$$ign of the spent fuel storage rack is such that it is impossible to insert the spent fuel assemblies in other than prescribed locations, j

thereby preventing any possibility of accidental criticality. Consideration I

cf criticality safety analysis is discussed in Section 4.3.

l l

The spent fuel pool is designed to maintain leaktight integrity. To ensure l

cuch integrity, the pool is lined with stainless steel plate, and plate welds cre backed with channels to detect and locate leakage. Laskage entering these channels is directed to the Liquid Waste Processing System (LUPS) via the FHB cump. Should a leak be detected, either by a low-level alarm (setpoint: 6 in. below normal water level) or by the fuel pool liner channel leak detection method, the operator would initiate makeup to the spent fuel pool. Makeup capability is provided by permanently installed connections to:

(1) the Domineralized Water System (DWS), (2) the Reactor Makeup Water System (RMUS),

cnd (3) the refueling water storage tank (RWST) in the Emergency Core Cooling i

System (ECCS).

A complete loss of spent fuel pool cooling is not considered a credible event 4

oince the components involved are designed to SC 3 seismic Category I require-ments and could be powered from redundant Engineered Safety Features (ESF) 3 power supplies. Further, the systems providing cooling are redundant.

Therefore, no single failure would result in a complete loss of fuel pool l

cooling.

For a more detailed discussion of spent fuel pool cooling, refer to Section 9.1.3.

9.1.3 Spent Fuel Fool Cooling and Cleanup System The Spent Fuel Fool Cooling and Cleanup System (SFPCCS) is designed to remove the decay heat generated by spent fuel assemblies stored in the spent fuel pool and/or the in-Containment stora8e area. A second function of the system is to maintain visual clarity and purity of the spent fuel cooling water and the refueling water.

!44 h.1.3.1 Design Bases. The SFPCCS design heat loads are given in Table 2

l 9.1 1.

System capabilities to withstand natural phenomena and piping rupture Q10.8 i

I are addressed in Chapter 3.

The spent fuel pool cooling portions of the SFFCCS are designed to seismic Category I requirements, and are located in the FHB, a seismic category I building. The spent fuel pool water purifiestion 44 portions of the SFPCCS are not required for safety functions and are.not I

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The Fuel-Handling Building contains a spent fuel pool. A copy of " Minimum Requirements for Design of Spent Fuel Racks" is enclosed (Attachment 3).

Provide the information as required and discuss your compliance with this

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g HL&P evaluat the icng term need,for increased spent fu torage at STP through the use of higher density spent fuel racks ( +

decided to 32 purchase higher density racks, these new racks will comply with Appendix D C sehetwfee, the prer::: rech; vill be an:1 :ed 'a' Arr-a m D cc pli:nce, 7

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The present racks will be used only for the initial fuel delivery, low-power testing and the early part of Cycle 1.

The analysis of these 14" center to center spent fuel racks was performed using the load combinations and' acceptance limits outlined in Table 1 (attached).

These load

' combinations and acceptance limits are taken from the paper

'"OT Position for. Review and Acceptance of Spent Fuel Storage and Handling. Applications," dated April 14, 1978 with modifications dated January 18, 1979, and have been used consistently by Westinghouse for the evaluation and recent license amendments for spent fuel racks at other plants.

The load combinations and acceptance limits for the seismic and thermal loads are from the table on page IV-6 of the January 18, 1979 modifications.

The load combinations for the stuck fuel. incident and the fuel drop accident are taken from the text of the paper.

Although these load combinations and acceptance limits are not exactly the same e.s those stated in Appendix D of Standard Review Plan 3.8.4, the intent of Appendix D has been met.

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[OSERT A CP 2 4a) g ATTACHMENT I TABLE 1 CE[ O 8

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STORAGE RACK LOADS AND LOAD COMllhA110N5 Lead Combination Atteotence Limit O+L Normal Limits of NF 8t31.la D+L+Pf Normal Limits of NF 8231.14 0+L+E Normal Limits of NT 3231.14 0+L+T lesser of 25y or s stress range o

u D+L+To+t Lesser of 25 or Sg stress range y

D+L+Te+t' Lesser of 25y or $g stress range 0+L+To + pf Lesser of 2S or s stress range y

u 0 + L + Ta + E' Faulted aondition limits of NF 3231.1c (see hete 3) 0+L*F6 The functional capability of the fuel racks shall be demonstrated l

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The abbreviations in the table above are those used in Standard Review Plan (5RP) Section 3.4.4 where each term is defined except for T,, which is defined here as the highest temperature associated with the postulated abnormal design conditions. Fg is the force caused by the accidental drop of the heaviest lead from the maximum possible height, and Pg is the upward force on the racks caused by a postulated stuck fuel assembly.

2.

The provisions of NF-3231.1 of ASME section 111, Division I, shall be amended by the requirements of Paragraphs c.2, 4 and 4 of Regulatory Evide 1.124. entitled " Design Limits and Load Combiantions for Class A Linear-Type Component Supports.'

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For the faulted load combination, thermal loads were neglected when they are secondary and self-limiting in nature and the material is duttile.

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