ML19249F121

Spent Fuel Pool Mod,Description & Safety Analysis.

ML19249F121
Person / Time
Site:	Three Mile Island
Issue date:	01/31/1977
From:	METROPOLITAN EDISON CO.
To:
Shared Package
ML19249F122	List: ML19249F121 ML19249F123
References
1134, PROC-770131, NUDOCS 7910100462
Download: ML19249F121 (45)

Text

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t METROPOLITAN EDISON COMPANY THREE MILE ISLAND NUCLEAR STATION UNIT 1 SPENT FUEL POOL MODIFICATION DESCRIPTION AND SAFETY ANALYSIS 99 Deckst # 501 o,....i#ll3 DheNS 19 of Cret: ment

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TABLE OF CONTENTS Section Title Page 1.0 Introduction 1-1 2.0 General Description 1-1 3.0 Mechanical Design 3-1 4.0 Criticality Considerations 4-1 5.0 Structural Analysis 5-1 6.0 Cooling Considerations 6-1 7.0 Radiological Considerations 7-1 8.0 Conclusion 8-1 1407 352 P

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1.0 INTRODUCTION

Because of the uicertainties in the future availability of fuel reprocessing facilities, Metropolitan Edison Company (Met-Ed) plans to increase the storage capacity of Three Mile Island (TMI)

Nuclear Station Unit 1 spent fuel pool "B" to prevent a shortage of space for storing spent fuel. The proposed method of accom-plishing th is increase is to install new storage racks with fuel assembly center-to-center spacing smaller than that of the present racks. These high density racks will utilize square tubes, made of stainless steel, to maintain the required shutdown margin and support the fuel assemblies.

The original unit design assumed a viable fuel reprocessing in-dustry in the United States by the time the unit commenced oper-ations. Therefore, the TMI-l "A" and "B" spent fuel pools were sized to accommodate 2-1/3 cores which was thought to be a con-servative approach. The assumption made during the design stage was that the one-third of a core discharged each year would be shipped to a reprocessing f acility in a timely manner. Therefore, the pools would always have the capability to accept an entire core offload. However, the ability to reprocess fuel does not and will not exist for some time. Because of this situation, the spent fuel generated as a result of reactor operation cannot be disposed of and must be stored.

When it became apparent that fuel reprocessing capability would not be available for many years, Met-Ed decided to expand the storage capacity of the "B" pool. In preparation for this ex-pansion the original spent fuel racks were removed from the "B" pool. This reduced the storage capacity f rom 430 assemblies to 256 assemblies which is the capacity of the " A" pool only.

At present 56 assemblies are being stored in the "A" pool. The unit is scheduled for shutdown for refueling in March 1977. This will increase total number of assemblies in storage to 104.

1407 353 1-1

This situation will preclude the unloading of the entire core (177 assemblies) should it become necessary. Met-Ed finds this condition to be unacceptable and requests the approval of the NRC to increase the storage capacity of the "B" spent fuel pool to 496 elements. This increase allows the storage of spent fuel until 1986 and allows Met-Ed to retain the capability to offload an entire core up to that time. This report discusses in detail the various design features incorporated in this modification and demonstrates that these design features will have no detrimental effect on the health and saf ety of the public.

1407 354 1-2

2.0 GENERAL DESCRIPTION 2.1 Present Desion Three Mile Island Nuclear Station Unit 1 is a 2535-MWt PWR (B&W) with a total of 177 fuel assemblies in the core. Its spent fuel

, storage complex consists of two Pools, "A" and "B", connected to each other by a canal and sliding gate, and a spent fuel pool cooling system. Water in the system contains approximately 1800 ppm boron. The cooling system includes two coolant pumps, two coolers, one borated water recirculation pump, and associated piping, valves, etc. The spent fuel cask loading pit is adjacent to Pool "B", and Pool "A" is connected to the reactor building fuel transfer canal by two fuel transfer tubes.

The major equipment components of the cooling system are located on the west side of the the Fuel Handling Building, a Class I structure hardened to withstand hypothetical aircraft impact as described in the TMI-l FSAR. Part of the cooling system piping extends into the Reactor Building and into the Auxiliary Building.

Both these structures are also Class I and hardened to with:tand hypothetical aircraft impact .

. The cooling system is designed to maintain 135 F in the pools with a heat load based on decay heat from one-third of a fully irra-diated core that has been cooled for 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br />, the postulated normal time between shutdown and removal of fuel from the core.

This can be acccmplished with one pump and one cooler. After an entire core of fload with an additional one-third of a core already in the pool f rom a ref ueling 100 days earlier, the pool can be maintained at 153 0 F by using both pumps and both coolers. The design capacity of the cooling system is 9.5 x 10 6 Stu/hr during a normal refueling and 28.0 x 10 6 Btu /hr during an Entire Core Of fload condition. The worst case heat generacion rate will cause the spent fuel pools to heat up at a rate of 5.2 0F/hr should all cooling be lost. During this Entire Core Of fload condition, 1407 555 2-

sufficient time would exist to activate the Reclaimed Water Sys-tem as an additional water source or to restore service to one of the spent fuel pool cooling chains. A purification loop is provided within the Radioactive Liquid Waste Disposal system for removing fission products and other contaminants from the water.

A small flow from the spent fuel cooling pumps is diverted to e a radiation monitor. The spent fuel cooling system is designed so that a line rupture will not cause a serious lowering of pool water level.

The present TMI-l fuel storage capacity consists of:

a. 253 Wet fuel locations in Pool "A"

b. 3 Wet failed fuel locations in Pool "A"

c. 63 Wet fuel locations in the Reactor Building Transfer Canal (rack temporarily removed but available for

, reinstallation)

d. 1 Wet failed fuel detection location in the Reactor Building Transfer Canal (temporarily removed but avall-able for reinstallation)

e. 66 Dry new fuel locations in New Fuel Storage Pcol

f. Pool "B" is now empty, but was originally designed for wet storage of 171 assemblies and 3 failed fuel assemblier.

The spent and new fuel assemblies are stored in racks in parallel

, rows having a center to center distance of 21.125" in both direc-tions. Control rod assemblies requiring removal from the reactor are stored in the spent fuel assemblies.

At present, Pool "A" contains spent fuel stored in already exist-ing racks. Pool "B", on the other hand, has never been used,

'contains neither water not spent fuel racks, and is free of radio-active contaminants. The proposeo modification, discussed in the next section, is for new spent fuel racks to be installed in Pool B.

2-2 1407 356~'

2.2 Proposed Modification The proposed f uel rack modifications, which conform in all re-spects to Safety Guide 13 (USNRC RG 1.13) , will involve install-ing high density storage racks in the empty "B" pool.

A rack assembly consists of a rectangular array of storage cell.s with a 13.625" center-to-center spacing. Each storage cell con-sists of a 9.12" I.D. square stainless steel cell having a wall thickness of 0.187". The array size of each rack was chosen to maximize use of pool space as shown in Figure 2-1. The expanded storage capacity of Pool "B" is 496 elements. The new racks con-tain no materials installed purely for neutron absorption capabil-ity. Recctivity calculations do consider the nuclear properties of the stainless steel cells and water but do not take credit for the 1800 ppm boron in the pool water.

The Spenn Fuel Pool Cooling System will maintain the fuel pools at a maximum of 135 F during Normal Refueling with one pump and one cooler, and 1470 F following an Entire Core Offload with two pumps and two coolers in operation.

As the installation will be made in a dry uncontaminated pool, no radiological problems are anticipated. The installation will not require movement of the new racks over the spent fuel in the "A" pool or over the new fuel storage area.

2.3 Schedule for Procosed Modification The schedule for the proposed installation of spent fuel racks is presented in Table 2-1. In order to maintain an Entire Core Offload storage capability, the racks must be available follow-ing the 1977 refueling outage that is scheduled for completion in May 1977. In order for rack procurement and construction to be-gin in a timely manner, initial NRC review and comments will be necessary by March 11, 1977 with final approval by May 1, 1977.

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TABLE 2-1 SCHEDULE FOR PROPOSED MODIFICATION Item Date Submittal of Safety Analysis February 4, 1977 Report and Environmental Impact Evaluation Initial NRC Review and Comments March 11, 1977 Final NRC Approval May 1, 1977 Rack Installation July-September 1977 1407 359 2-5

3.0 MECHANICAL DESIGN 3.1 Scent Fuel Storage Cells Each fuel assembly is stored in a stainless steel cell 13.5' long and having a square cross section with a 9.12" interior dimension.

These storage cells are fabricated of 0.187" type 304 stainless steel sheet and are of an all welded construction. The cells are flared at the top to an 11" square cross section to facilitate insertion and removal of fuel assemblies. Each cell is partially closed at the bottom by welding two 0.187" stainless steel bars across the bottom of the cell on opposite sides. These bars provide two 2" wide support ledges for the fuel assembly. This method of support leaves a 45 square inch rectangular opening at the bottom of the cell. This opening allows cooling water to flow upward through the fuel assembly to provide for removal of decay heat. The cell manufacturing and rack assembly process are controlled to ensure that there will be no binding during insertion or removal of a fuel element.

3.2 Fuel Rack Assemblies The individual cells are assembled into racks having a 13.625" pitch between adjacent cells by being welded to two lattices of structural stainless cteel channels (see Figure 3.1) . One lat-tice is located near the top of the rack and one near the bottom.

Each lattice consists of two sets of continuous channels, one set laid over the other and running the length or width of the rack.

The channels are welded together at the bottom support points to provide structural rigidity.

Lifting lugs are attached at four intersection points of the upper lattice of each tack. To provide structural rigidity, the overlap'-

ping channels are welded to each other under the lif ting points

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Each rack is supported at the bottom by four individually adjust-able leveling legs. The legs are located between cells at the intersections of two channels and are welded directly to the channels. Gussets are welded between the channels and the legs to provide additional strength. The bearing pads of the support legs are sized to ensure that stresses on the pool liner and under-lying concrete are within acceptable limits.

At the periphery of each rack, where one rack touches another, there is a system of fixed or adjustable bearing plates attached to the upper and lower lattice structure to transmit thermal and seismic loads between adjacent racks. Where the racks f ace the pool wall, compression-type seismic restraint devices are pro-vided on both the upper and lower lattice structure. These seis-mic restraint devices (see Fig. 3. 2) are adjustable and a gap will be provided for thermal growth of the racks resulting from expec-ted temperature variations in the pool. The bearing pads of the seismic restraint devices are sized to ensure that wall and liner stresses are within acceptable limits. ,

There are three different sizes of rack, each having been chosen to maximize the storage capacity of the pool. The rack types are as follows:

Tvee Arrav A 5x5 B 5x4 C 8x2 Within each type the racks are identical except for the array size and the arrangement of fixed and adjustable restraints, which depend on the rack's specific location within the pool.

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-4 1408 003

3.3 Codes, Standards, and Practices for Fuel Assembly Rack Design, Construction, and Assembly The following are the codes, standards, and practices to which the fuel assembly tacks will be designed, constructed, and assem-bled (revisions utilized are those in eff ect as of November 1, 1976). Other applicable codes, standards, and regulatory guides are identified elsewhere in this document.

1. Design Codes

a. AISC Specifica~ tion for the Design, Fabrication and Erection of Structural Steel for Buildings, 1969, including Supplements 1, 2 and 3.

b. ASME Boiler and Pressure Vessel Code,Section III, Nuclear Power Planc Components (Tables I-7.0 and I-8.0 are used for yield values for materials of

, construction).

2. Material Codes

a. ASTM Specification A-240-74a, Specification for Stain-less and Heat-Resisting Chromium and Chromium-Nickel Steel Plate Sheet and Strip for Fusion-Welded Unfired Pressure Vessels,

b. ASTM Specification A-320-74, Specification for Alloy Steel Bolting Materials for Low Temperature Service.

c. AWS Specification A-5.9-69, Corrosion-Resisting Chromium and Chromium-Nickel Steel Welding Rods and Bare Electrodes.

d. ASTM Specification A-276-73, Specification for Stain-less and Heat-Resisting Steel Bars and Shapes.

3-5 1408 004

e. AWS Specification A-5.4-69, Specification for Corrosion-Resisting Chromium and Chromium-Nickel Steel Covered Welding Electrodes.

3. Welding Codes

a. ASME Boiler and Pressure Vessel Code, Section IX-1974, Welding and Brazing Qualifications.

4.  ; *.li ty Assurance, Cleanliness, and Packaging Recuirements O'

a, 10CFR50 Appendix B, Quality Assurance Criteria for Nuclear Power Plants

b. RG 1.28 Quality Assurance Program Requirements -

Design and Construction (Safety Guide 28) , 6/7/72

c. RG 1.37 Quality Assurance Requirements for Cleaning of Fluid Systems and Associated Components of Water-Cooled Nuclear Power Plants, 3/16/73

d. RG 1.38 Quality Assurance Requirements for Packaging, Shipping, Receiving, Storage, and Handling of Items for Water-Cooled Nuclear Power Plants, 3/16/73

e. RG 1.64 Quality Assurance Requirements for the Design of Nuclear Power Plants, Rev. 2, 6/76

f. RG 1.88 Collection, Storage, and Maintenance of Nuclear Power Plant Quality Assurance Records, Pev.

2, 10/76 3-6 1408 005

3.4 Fuel Rack Installation Procedure The proposed racks will be installed in Pool "B" which at present is dry and has nc. ca.s:s installed. The racks are however designed for either wet et dry installation. This pool has never been filled with water having any radioactive contamination and there-fore the installation will involve no potential health physics problems. The installation procedures will preclude handling the racks over Pool "A" which has spent fuel in it, or over the new fuel storage area.

3.5 Material Compatibility Because the replacement racks, their associated hardware, and the seismic restraints are of all stainless steel construction, as is the spent fuel pool liner, there is no potential for galvanic corrosion. Material compatibility between the fuel assemblies and the new storage racks is also not a problem as stainless steel has been shown to be compatible with both f uel assemblies and with borated water in the pool.

3-7 1408 006

4.0 CRITICALITY CONSIDERATIONS The racks are designed for a .3.625" center-to-center spacing between storage cells. The results of the criticality analyses are as follows:

1. The center-to-center spacing of 13.625" results in a k, of 0.892 under nominal conditions.

2. The s3rst case situations, considering maximum varia-t i e r..; in the position of fuel assemblies wi thin the storage rack, variations in cell dimensions, the most reactive temperature, calculational uncertainties, and worst case accidents result in a k, of 0.934 with a confidence level of 95%.

4.1 Assumotions and Method of Analysis The referenced set of calculations were based upon the follow-ing assumptions:

1. New fuel of 3.50 wt% 235 U nominal average enrichment equivalent to 45.90 grams of 235 U per centimeter of height.

2. Water temperature of 68 0F.

3. No credit taken for soluble poison.

4. Fuel racks are infinite in three dimensions.

5. Control rods and other fixed poisons are not present in the fuel assembly.

The majority of the calculations were performed with methods com-monly employed in light water reactor design, i.e., four-group 4-1 1408 007

dif f usion theory cell calculations using PDQ-07. The cross sections for these calculations are generated with NUMICE, the NUS version of LEOPARD. This code uses the same cross section library tape and calculational techniques as LEOPARD.

Selected cases were checked and the final design multiplication factors were verified with Monte Carlo criticality calculations using KENO with 123-group crcss sections. The 123-group cross section library is generated f rom the basic GAM-THERMOS library using XSDRN (P3'3)* 8 4.2 Results of Analysis Figure 4-1 shows the geometry of the reference storage cell used in the design calculations. Section 4.3 summarizes the results of both the dif f usion theory and Monte Carlo calculations. In general, the four-group PDQ 3iffusion calculations produce k, values about 0.015 lower than the Monte Carlo calculations.

Calculational uncertainties in the use of PDQ with cross sections based on the LEOPARD library have been obtained by comparing the results of a series of benchmark calculations with critical ex-periments. These comparisonsII) have shown that the average dif-f erence between the calculations and experimental results was 0.009 ak..

The KENO code, using the 123-group GAM-THERMOS cross section library, has been extensively benchmarked also. For a series of ten critical experiments reported (2) the average k,gg as (1) WCAP-3269-25, " Calculation of Lattice Pat' meters and Criti-cality for Uranium Water Moderated Lattices," L. E. Strawbridge, Westinghouse Electric Corporation, September 1963.

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" Validation of Monte Carlo Calculations of Shipping Cask Sys-tems," L. M. Petrie and P. G. McCarty, ORNL, CONF-731101-14, 1973. /

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calculated using KENO and 123-group cross sections was 0.9914 1 0.0020. Using the same method, NUS has performed another bench-mark on one of the Yankee critical experiments (3) with Ag-In-Cd cruciform contr-1 rods banked at 26.37 cm from the bottom of the fuel.

The calcmiated keff was 1.008 1 0.006. On the basis of the above comparisons with criticals, a calculational uncertainty of 0.008 Ak ,was assigned to the KENO calculations.

Statistical analysis of the Monte Carlo results shows a standard deviation of 1 0.004, giving a 2a uncertainty of 0.008 ak,. Thus, an additional 0.008 Ak, uncertainty is assigned to the KENO calcu-lations.

The worst-case criticality condition was obtained by using the maximum tolerances for the positioning of the fuel assemblies w'ithin the storage cell as well as the relative cell-to-cell positioning and cell dimensions.

4.3 Worst-Case Analysis of Tolerances and Calculational Uncertainties The following are the results of the KENO analysis of the worst-case tolerances and calculational uncertaintics:

Nominal Conditions, k_ 0.892 Enrichment, 3.5 wt% -

Mechanical Spacing, 13.625" Pool Temperature, 68 F

3) " Yankee Critical Experiments - Measurements on Lattices of Stainless Steel Clad Slightly Enriched Uranium Dioxide Fuel Rods in Light Water," P. W. Davison et al., YAEC-94, April 1959, page 32.

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Worst Tolerances, Ak, Enrichment, 102% of nominal 0.003 Most Reactive Temperature 0.003 SS Composition 0.002 Eccentric Fuel Loading in Cell 0.007 Mechanical Design 0.009 TOTAL 0.024 Calculational Uncertainties, ak, KENO Benchmark 0.008 Statistics (2a) 0.008 Total Calculational Uncertainties 0.016 Maximum, k, Nominal, k, 0.892 Worst Tolerances 0.024 Calculational Uncertainties 0.016 0.040 MAXIMUM, k, (without accident case) 0.932 MAXIMUM, k, (with accident case - 0.934 see Section 4.5) 4.4 Parametric Studies The ba a case , as es cablished in the preceding sections, refers to the rack design with 13.625" spacing, 3.50 wt% nominal enrich-ment and 68 0F pool water temperature. The .k of the base case is 0.892 based on the 123-group KENO calculation. Parametric stud-ies were performed to determine the eff ect on k of varying the base case conditions one at a time. The results are presented below:

4-5 1408 011

1. k, vs. Center-to-Center Spacing (PDQ) 13.375 +0.007 ak, (Nominal) 13.625" (Base) 13.875" -0.0067 ak,

2. k,vs. Enrichment (PDQ) 3.40 wtt U-235 -0.0052 ak.

(Nominal) 3.50 wt% U-235 (Base) 3.60 wt% U-235 +0.0045 ak,

3. k,vs. Cell Wall Thickness (PDQ) 0.173" +0.0023 ak.

(Nominal) 0.187" (Base) 0.201" -0.0021 ak ,

4. k ,vs. Water Temperature (PDQ)

• 40 0 F -0.0022 ak.

(Nominal) 680F (Base) 1000 F +0.0011 ak.

1530 F +0.0019 ak.

2'120F +0.0019 2k ,

4.5 Fuel Handlina Accident Analysis The worst accident during spent fuel handling involves dropping a fuel assembly that would Land horizontally on top of the storage racks. Inadvertent positier.ing of an assembly vertically next to the rack is not possible for the following two reasons. First, all the racks will be installed in the pool before fuel storage commences and thus there will not be any open water region except for the gaps between the racks and the pool walls. Second, a permanent barrier will be installed in each gap between the racks 4-6 1408 012

and the pool walls, as necessary, to prevent insertion of an assembly. In any assembly drop accident the minimum distance between the active fuel and the accident assembly is greater than 12". Calculations show that the multiplication factor is not penalized more than 0.002 ak, if this separation is 5" or more.

Therefore, this situation has an insignificant effect on the pool criticality.

4-7 1408 013

5.0 STRUCTURAL ANALYSIS 5.1 Loads and Loading Criteria In accordance with Regulatory Guide 1.29, the spent fuel storage racks were designated Seismic Category I. Structural integrity of the fuel racks when subjected to normal and abnormal loads, as well as the OBE and DBE, has been demonstrated with respect to the NRC Standard Review Plan Sect 3 on 3.8.4. In accordarca with the Review Plan, the following loads, load combinations, and structura; acceptance criteria were considered:

5.1.1 Loads

a. Normal Loads

i. Dead Loads - dead weight of rack and fuel assem-blies and hydrostatic loads

11. Live Loads - effect of lifting empty rack during installation iii. Thermal Loads - uniform thermal expansion of racks due to change in average pool temperature from 70 to 147 F and a thermal gradient between adjacent storage boxes of 20 0F.

b. Severe Environmental Load - Operating Basis Earthquake (OBE)

c. Extreme Environmental Load - Design Basis Earthquake (DBE)

d. Acci. dental drop of a spent fuel asscmbly from a height of 2.67~,' above the top of the racks, which is conserva-tive with regard to fuel handling operations.

5-1 1408 014

e. Postulate 1 stuck fuel assembly which causes an upward force of 300 lb, cr a downward force of 350 lb.

5.1.2 Load Combinations The spent f uel storage racks were analyzed using elastic working stress design methods for the following applied loads:

a. Dead Loads Plus Live Loads

b. Dead Loads Plus Impact Loads Plus OBE

c. Dead Loads Plus Thermal Loads Plus Impact Loads Plus OBE

d. Dead Loads Plus Thermal Loads Plus Impact Loads Plus DBE (SSE)

e. Dead Loads Plus Fuel Assembly Drop

f. Dead Loads Plus Stuck Fuel Assembly Live loads were not included in load combinations b through f, since the only live load on the rack was that due to lif ting, and li f ting of the racks is performed with the racks empty.

5.1.3 Structural Acceptance Criteria The following were the strength limits for each of the above load combinations:

Load Combination Strength Limit a 1.0 S b 1.0 S c 1.5 S d 1.6 S e 1.6 S (except as noted below) f 1.6 S (except as noted below) where S is the required section strength based on the elastic design methods and the allowable stresses 'e fined in Part 1 of 5-2 1408 015

the AISC " Specification for the Design, Fabrication and Erection of Structural Steel for Buildings," February 12, 1969, including Supplement Numbers 1, 2 and 3. (Supplement 3 was effective June 12, 1974.) Yields for major structural members were obtained from ASME Boiler and Pressure Vessel Code Section III. For load combinations e and f, local stresses might exceed the limits, pro-vided there was no loss of function of the fuel rack. In addition the fuel cell was checked for buckling to ensure that no collapse occurs due to compressive loading.

5.2 Seismic Analysis The individual fuel racks described in Section 3.0 will be of all-welded construction. The racks will rest on the floor and butt against one another at the top and bottom. At the perimeter of the pool there will be clearance between the pool wall and the upper and lower seismic restraints sufficient to allow for thermal expansion of the racks.

The seismic loading of a typical fuel rack was determined from a response spectrum modaf. dynamic analysis in which the stiff-ness of the fuel assembly was neglected. However, the mass of the fuel assemblies ano an effective mass of water were consid-ered to be uniformly distributed along the storage cells.

The response spectrum modal analysis assumed that all gaps were closed by thermal expansion. The assumption that the restraint was in contact with the wall (no gap) was necessary in order to consider the worst case of fuel / cell interaction (see discussion below). The assumption used in tha analysis of the rack was cra-sistent with this requirement and provided a basis for combining the results of the modal analysis and the fuel / cell interaction analysis. A consequence of the closed gaps is that each rack is touching the one next to it and the whole array must be modeled.

The detailed modal rack in Figure 5-1 is part of an array of racks supported at the walls. The analysis considered the entire pool 5-3 1408 016

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GRID SUPP0flT FOOT Figure 5-1 TMI 5 x 5 Rack Model 5-' 1408 017

- WALL DETAILEO 5 x 5 "A" RACK o

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CENTER OF "B" RACKS CENTE R OF "A" RACKS ao og,

Xi GRIO sE AMS v

%e V V V Figure 5-2 Entire Pool in I.ateral Vibration (Plan View)

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CRIGINAL GECMETRY

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MODE 1 (ELEVATION VIEW, X. AXIS)

MODE 3iELEVATION VIEW, Xj AXIS) o

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MODE 2(ELEVATION VIEW X AXIS)

MODE 4 (ELEVATION VIEW, X AXIS)

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MODE 5 (ELEVATION VIEW. X3 A XIS)

Figure 5-3 Mode Shapes 5-5 1408 018

in lateral vibration, as shown in Figure 5-2. Thus, when the rack vibrated in the Xt or X2 direction, the detailed 5 x 5 "A" rack portion of the model showed the cell bending and cell / grid interaction. The portion of the model with only grid beams and concentr.ited weights (at the positions shown as centers of "A" or "B" racks in Figure 5-2) showed the spring effect of all the racks acting :ogether through the grios. The detailed portion was sup-ported in the vertical dirc.cion at the feet.

The appropriate response spectra for the OBE and DBE, respectively, were employed. To obtain the appropriate response spectra, the response spectra at elevations 302.5' and 329' were linearly interpolated to obtain the maximum response spectrum which occurs at the upper seismic restraint location. Th is spectrum was then used at each frequency of interest. These frequency response spectra were developed using the methods reported in the TMI-l FSAR. The damping factors of 0.5% and 2% for the OBE and DBE, respectively, utilized for the initial unit design were increased to 2% and 4% to account for the additional damping afforded sub-merged structures.

The STARDYNE computer program was used to perform the structural analysis of the racks. Storage racks were modeled in detail using beam and plate finite elements. The three-dimensional finite ele-ment model for the 5 x 5 rack is shown in Figure 5-1. This is considered representative of the rack design.

To determine the earthquake modal response, STARDYNE was "irst run to determine the natural frequencies, mode shapes, and parti-cipation factors. For the five analyzed modes the significant frequencies, participation factors, and corresponding accelera-tions are given in Table 5-1. Figure 5-3 provides a sketch of modes 1, 2, 3, 4, and 5. Modes 3 and 4 are similar to modes 1 and 2, respectively, but are in the X 2 direction.

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TABLE 5-1 MODAL DATA

__ Participation Factors Hori- Hori- Horizontal Mode Frequency zontal zontal Vertical Azceleration (c)

Shape 'HJ X, X X OBE DBE 7 7 1 11.5 1.56 0 0 0.85 1.27 2 13.9 0.23 0 0 0.44 0.65 3 15.6 0 1.84 0 0.32 0.51 4 19.7 0 0.17 0 0.39 0.62 5 29.1 0.95 0 0 0.20 0.34 The seismic loads were then determined using the above modal data. Since the static vertical direction had no participation below 33 Hz, a static seismic analysis was run for the vertical direction using a "g" load of 0.089 (OBE) and 0.199 (DBE). The results for the horizontal earthquakes were determined using a response spectrum analysis. Since no modes were closely spaced (as defined by USNRC RG 1.92) , the results of each mode were com-bined with the other modes in an SRSS fashion. The results of the three directions of earthquake were then combined in an SRSS f ashion , per USNRC RG 1. 92.

In the general . seismic / structural analysis of the fuel racks, the mass of a fuel assembly was assumed to be uniformly distributed along the length of each cf the fuel storage cells. This assump-tion was conservative in that lower rack fundamental f requencies were calculated which, due to the relatively stiff rack design, result in higher seismic amplified acceleration loading on the rack. Since a gap on the order of 0.6" will exist between the sides of a fuel assembly and the cell (fuel leaning on one side of cell) , the fuel will move within the cell during a seismic event. The ef f ect of this motion, termed fuel-cell interaction, 5-7 1408 020

was analyzed using the ANSYS computer program. A nonlinear dynamic analysis of a single cell and fuel assembly was performed to determine the maximum shear force and bending moment that might occur at critical sections of the cell as a result of the fuel assembly impacting the cell at maximum velocity. The cell and fuel assembly were modeled by beam finita elements and are separated by nonlinear gap elements. The cell was restrained at the upper and lower grid elevations by a spring that represents the grid stiffness. The fuel, which was assumed to be pinned at its base, was given an initial velocity relative to the cell.

The impact velocity is taken to be the SRSS sum of the maximum rack velocity and the maximum floor velocity. This approach is conservative in that the rack and fuel vibration must be out of phase if impact is to occur. Impact loads were determined as a function of time and were included in load combinations b, c, and d.

5.3 Structural Adecuacy Using the previously listed loads and load combinations, stresses s

were calculated at critical sections of the rack. The results of the structural and seismic analyses demonstrate that the spent fuel racks are structurally adequace and will meet the design criteria. Critical stresses together with locations and margins to allowable are gis n in Table 5-2.

5.4 Pool Wall and Floor Loading The ability of the fuel pool floor and walls to withstand the loads imposed by the modified f uel racks will be analyzed in ac-cordance with Section 5.4 of the TMI-l FSAR. All loads, includ-ing the thermal and water sloshing loads will be ccmbined in com-pliance with NRC Standard Review Plan 3.8.4.

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TABLE 5-2

SUMMARY

OF CRITICAL STRESS RESULTS Limiting Calculated Allowable Load Stress Stress Location Combination * (esi) (osi) Margin Cell to d 14,000 17,600 1.26 grid welds Grid beams d 23,700 26,400 1.11 Leg to gusset b 9,770 11,000 1.13 welds Seismic restraint b 9,670 11,000 1.14 bearing pad to grid weld Lifting lug a 12,800 16,500 1.29

• See Section 5.1.2 for definition.

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6.0 COOLING CONSIDERATIONS 6.1 General Description The Spent Fuel Pool (SFP) Cooling System removes decay heat gen-erated by spent fuel stored in the TMI-l Pools "A" and "B". The system consists of two cooling water pumps, two heat exchangers, a borated water recirculation water pump, and associated piping that connect this system to the Decay Heat Removal and the Radio-active Liquid Waste Disposal (RLWD) Systems. The Cooling System is redundant in that either of the two pump / cooler combinations or both pumps and coolers can be used to cool Pool "A", Pool "B",

or both pools. The system is further described in Section 9.4 of the TMI-l FSAR.

6.1.1 Normal Refueling The SFP Cooling System maintains Pools "A" and "B" at a maximum temperature of 135 F under a Normal Refueling condition with one pump and one cooler in operation. A Normal Refueling condition is defined as the decay heat generated from eleven (11) yearly refuelings with the eleventh being placed in the spent fuel pool within 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> of reactor shutdown.

6.1.2 Entire Core Offload The SFP Cooling System has the additional capacity to maintain the spent f uel pools at a _ maximum temperature of 147 0F under an Entire Core Offload condition with two pumps and two coolers in operation .

An Entire Core Offload condition is defined as the decay heat generated from eleven (11) yearly refuelings and an entire core of fload with the entire core being placed in the spent fuel pool within 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> of reactor shutdown. Following the Entire Core Offload, both "A" and "B" spent fuel pools will be full, resulting in the worst case heat generation condition.

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6.2 Cooling System Performance The adequacy of the Spent Fuel Pool Cooling System has been ana-lyzed in view of the expanded fuel storage capacity. Table 6-1 summarizes the cooling system performance for the Normal Refuel-ing and Entire Core Offload conditions.

TABLE 6-1 SPENT FUEL POOL COOLING HEAT LOADS AND OPERATING TEMPERATURES Heat Load (Btu /hr) Pool Temperature (OF)

Normal Refueling 9.7 x 10 6 135 Entire Core Offload 2.57 x 10 7 147 The decay heat loads were calculated using the computer code ORIGEN developed ac Oak Ridge National Laboratory. ORIGEN is a point depletion code that solves the equations of radioactive buildup and decay for large numbers of isotopes with arbitrary coupling. The methods of analysis used in this evaluation are considerably more advanced than those used in the TMI-l FSAR decay heat analysis. That analysis was based on infinite irra-diation time of fuel assemblies whereas this analysis is based on finite irradiation time and the actual projected refueling schedule.

The average design burnup of spent fuel is 30,120 MWD /MTU. This burnup is an average of fuel discharged af ter 3 and 4 cycles of core residence.

The temperature of the Nuclear Services Closed Cycle Cooling Water going into the Spent Fuel Pool (SFP) Cooling Heat Exchangers 6-2 1408 024

is the controlling factor in establishing the heat transfer cap-ability of the Spent Fuel Pool Cooling System. A review of recorded temperatures since unit startup in 1974 shows that using a temperature of 95 F conservatively assumes worst case heat ex-changer cooling conditions.

6.2.1 Decay Heat Loads In evaluating the impact of increasing the storage capacity of TMI-l spent fuel pools, two decay heat loads have been analyzed:

1. Normal Refueling

a. 521 spent fuel assemblies in storage accumulated from ten successive yearly refueling outages,

b. 53 spent fuel assemblies discharged to the storage pools during the eleventh TMI-l refueling outage scheduled in 1986. It was conservatively assumed that this discharge is completed 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> follow-ing reactor shutdown from full power.

c. It was conservatively assumed that all eleven re-fuelings occurred following reactor operation for 272 effective full power days for an entire year.

The results of this analysis are as follows:

Pool Heat Generation Rate: 9.7 x 10 6 Btu /hr Maximum Pool Temperature: 135 F

2. Entire Core Offload

a. 574 spent fuel assemblies in storage accumulated from eleven successive yearly refueling cutages.

5-3 1408 025

6

b. 177 fuel assemblies (entire core) discharged to the storage pools dur,ing the twelf th TMI-l ref ueling outage scheduled in 1987. It was conservatively assumed that this discharge is completed 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> following reactor shutdown from full power.

c. It was conservatively assumed that all eleven refuel-ings occurred following reactor operation for 272 effective full power days for an entire year.

The results of this analysis are as follows:

Pool Heat Generation Rate: 2.57 x 10 7 Btu /hr Maximum Pool Temperature: 147 F0 6.2.2 Failure Analysis Two conditions of f ailure analysis have been analyzed:

1. Single-failure analysis Although single failure is unlikely, Table 6-2 is a single-failure analysis of the TMI-l Spent Fuel Pool Cooling System. -

2. Lous of all spent fuel pool cooling It is highly unlikely that a complete loss of spent fuel pool cooling will occur due to the following reasons:

a. Seismic Class I System,

b. Redundant cooling pumps. -

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TABLE 6-2 SPENT FURL POOL COOLING (SFPC) SYST124 SINGLE ACTIVE FAILURE ANALYSIS Cay .equence NormalRgfueling EntireCorgOffload Component Failure Mode (9.7

• 10 Btu /nr) (2.57 x 10 Btu /hr)

Spent Fuel Mechanical 1. Redundant pump is operational. 1. Redundant pump is operational.

Pool Cooling No effect on SFPC. 2. Inventory of spares is avall-Pump able for rapid repairs.

3. Supplemental cooling methods available as discussed in Section 6.2.3.

4. Maximum pool temperatures 199 F.

Nuclear Services Mechanical 1. No (ffect on SPFC. 1. No effect on SFPC.

Closed Cycle 2. Redundant NSCCCW pump is 2. Redundant ESCCCW pump is Cooling Water available for 1004 system available for 100% system (NSCCCW) Pump capacity. capacity.

(cools SPEC heat m exchangers) i River Water (RW) Mechanical 1. No effect on SFPC. 1. No effect on Sl'PC.

Pump (cools 2. Redundant RW pump is available 2. Redundant RW pump is avull-NSCCCW heat for 100% system capacity. able for 1004 system enchangers) capacity.

Offsite Power Electrical 1. Energency power is available. 1. Emergency power is available.

2. Manual starting of SPIC, 2. Manual starting of SFPC, NSCCCW, and RW pumps is NSCCCW, and RW pumps is possible. possible.

L No effect on SFPC. 3. No effect on SFPC.

SFIC Air-Oper- Ioss of 1. No effect on SFPC. 1. No effect on SFPC.

ated Valves Air and/or 2. Valve position remains "as is" 2. Valve position remair.s "as it" (suction of Electrical upon failure, upon failure, pumps amt out- Control let of heat Power

" exchangers)

O NSCCCW Motor- Ioss of 1. No effect on SFFC. 1. No effect on SPPC.

CO Operated Valves Electrical 2. Valve position remains "as is* 2. Valve position remains "as is" (NS-V16A and Power upon failure. upon failure.

Q NS-V16B at inlet g of Srnt heat q exchangers)

c. Cooling pumps are supplied from separate electrical sources, each with the ability of being powered by separate emergency diesels.

d. Redundant heat exchangers.

e. The systems that provide the ultimate heat sink for the spent f uel pool cooling heat exchangers are Seismic Class I, redundant systems.

Since the heat generation rates as a res it of this modification will be essentially the same as those reported in Section 9.4 of the TMI-l FSAR, the heatup rates resulting from a loss of all cooling will be essentially unchanged. Therefore, commencing with a pool temperature of 1*7 0F the niinimum time for the pool (s) to reach 2120 F will be 12.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />.

6.2.3 Eupplemental Cooling There are four supplemental means of providing for cooling the spent fuel pools in addition to the Spent Fuel Pool Cooling System,

l. The Decay Heat Removal System can be used to cool the pools.

2. The forced ventilation system can be used to improve the cooling effects of pool surface evaporation.

3. Reclaimed water can be used for pool water makeup as well as for its cooling effect.

4. Time delays can be imposed on the transf er of fuel as-semblies into the fuel pool. The decay heat analysis very conservatively assumes tha t an entire core will be discharged into the spent f uel pools within 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> 6-5 1403 028

of reactor shutdown. The first refueling outage at TMI-l required an entire core offload. From reactor shutdown to removal of the entire core to the spent fuel pool, 436.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> (18.17 days) elapsed. Subtracting delays and learning enperience, a realistic minimum time for this operation is on the order of 240 hours0.00278 days <br />0.0667 hours <br />3.968254e-4 weeks <br />9.132e-5 months <br /> (10 days). Moni-toring the spent fuel pool water temperature and impos-ing time delays is undesirable, but as an emergency method it is simple and reliable.

6.3 Fuel Element Heat Transfer The bottom of it fuel cell is elevated above the floor to assure adequate flow tnder the rack ea each fuel assembly. There is also sufficient space between the rack complex and the pool walls to provide adequate downcomer clearance. Analyses have been per-formed which show that sufficient ficw is induced by natural con-vection to preclude.. local boiling in the hottest storage location.

The analyses were based on the following assumptions:

1. The element inlet temperature is the mixed hot tempera-ture of the pool. This temperature is 1470F and applies to the thermally limiting condition of a no f ailure full-core offload.

2. A hot assembly peaking factor of 1.78 is applied to a limiting batch average assemb'y energy release rate.

The average assembly energy release rate is 1.43 x 105 Btu /hr corresponding to 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> after shutdown.

3. The maximum local peaking factor is 2.67 giving a maxi-mum local heat flux of 1360 Stu/hr-ft2, 1408 029 6-7

4. A film coefficient of 36 Btu /hr-ft _oF 2

is based on pure conduction through a stagnant boundary layer at the fuel rod surface.

5. A one-dimensional fluid flow analysis is used.

6. For the hottest storage location the downcomer region on the periphery of the pool feeds ten assemblies in a row, each assumed to be generating the maximum heat rate de-fined in Assumption 2.

With the spent fuel pool cooling system operational, the local o

coolant temparature will not exceed 170 F for an Entire Core Off-load. For this condition the maximum surf ace temperature of a fuel rod is less than 207 F0 providing more than 31 F margin to local boiling. The margin to bulk boiling is greater than 68 F.

This represents the limiting thermal condition in the pool. Un-der the preceding conditions the hottest fuel rod surface tem-perature is below the local saturation temperature of 23S F O and thus precludes local boiling. During normal refueling the ther-mal conditions in the pool are less limiting than those given above.

6.4 Spent Fuel ?ool Chemistry Control Water purity and clarity are maintained by the Spent Fuel Pool Cooling (SFPC) System and the Radioactive Liquid Waste Disposal (RLWD) System. The borated water recirculttion pump takes a suc-tion on the skimmers located at the top of Pool "A" or Pool "B" and discharges to the RLWD Systems. Water passes through one of two precoat filters in the RLWD System. The filter uses stain-less steel wire wound elements with a powdered resin. When the pressure drop across the filter becomes too great, it is back-washed and recoated. In this manner , floating debris is removed fecm the pool surface and water quality is maintained. The vol-ume of Pools "A" and "B" is 649,000 gallons. The borated water 6-a 1408 03C

recirculation pump has a capacity of 180 gpm. This system, there-fore, has the capacity to filter the pools every 60 hours6.944444e-4 days <br />0.0167 hours <br />9.920635e-5 weeks <br />2.283e-5 months <br />.

Radioactive contaminant levels in the fuel pools are primarily a function of failed fuel fraction and reactor operating level with the highest levels during and shortly following refuelings. The RLWD System additionally processes pool water to control radioac-tive contaminant levels. Besides the precoat filters which uti-lize ion exchange as well as a filterirg mechanism, the RLWD Sys-tem contains demineralizers and waste eraporators for process-ing pool water. Besides these means, natural radioactive decay helps reduce levels of radioactivity. Operating experience has demonstrated that the SFPC and RLWD Systems are highly eff ective in maintaining water purity and clarity during storage periods and refueling operations. Throughout 1976, the water in Pool "A" was circulated through the RLWD cleanup system for a total of 177 hours0.00205 days <br />0.0492 hours <br />2.926587e-4 weeks <br />6.73485e-5 months <br />, 104 hours0.0012 days <br />0.0289 hours <br />1.719577e-4 weeks <br />3.9572e-5 months <br /> of this cleanup occurred prior to placing any fuel in the spent fuel pool; 73 hours8.449074e-4 days <br />0.0203 hours <br />1.207011e-4 weeks <br />2.77765e-5 months <br /> of cleanup occurred while an entire core was temporarily stored in Pool "A". That cleanup proved more than satisf actory to maintain pool chemistry conditions and minimize pool radioactive contaminant levels. For the last eight months of 1976, there has been no need to circu-late pool water through the cleanup system, due to the thorough-ness of cleanup during the refueling outage.

6_9 1408 031

7.0 RADIOLOGICAL CONSIDERATIONS 7.1 Fuel Handling Building Dose Rates The additional spent fuel assemblies in Pool "B" resulting from this rack modification will have in insignificant impact on the radiological eff ects discussed in the TMI-l FSAR. A QAD computer code was utilized for this analysis. Calculations indicate that the dose rates in the fuel handling building as reported in Sec-tions 11.3.1 and 11.3.2.6 of the TMI-l FSAR will be essentially unchanged. Levels will normally be less than 1.5 mR/hr with cer-t&in refueling manipulations causing short term levels in excess of 1.5 mR/hr. As stated in the TMI-l FSAR, radiation levels will be closely monitored during refueling operations to establish the allowable exposure times for unit personnel in order not to exceed the integrated doses specified in 10 CFR 20.

Additionally, a QAD'ccmputer analysis indicates that the radia-tion levels discussed in Section 11.3.2.6 of the TMI-l FSAR will remain unchanged. That is, during spent fuel transfer, the dose rate at the pool surface with a minimum of 7 feet of water shield-ing between the top of an assembly and the surf ace will result in approximately 15 mR/hr. The dose contribution attributed to the increased fuel storage is negligible.

The radiological consequences of a fuel handling incident are discussed in Section 14 of the TMI-l FSAR. The bases for the analysis are unaffected by enlarging the capacity of the spent fuel storage pool and therefore the analysis and results are still applicable.

7.2 Cask Handling The spent f uel cask drop analysis was filed with the NRC on February 14, 1976. This submittal is currently under NRC review.

7-408 032

This fuel rack modification has no effect on the evaluation, since the possibility of a cask drop in the spent fuel pool has been minimized.

7-2 1408 033

8.0 CONCLUSION

Based on the above analyses and description, Met-Ed concludes that the described modification can be accomplished without undue hazard to the health and saf ety of the public and that it con-forms to applicable regulations. -

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