ML20049J993
| ML20049J993 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 03/31/1982 |
| From: | ALABAMA POWER CO. |
| To: | |
| Shared Package | |
| ML20049J990 | List: |
| References | |
| TAC-48219, NUDOCS 8203290302 | |
| Download: ML20049J993 (100) | |
Text
_ _ . . - _ - . _ _ _ _ _ _ _ _ _ _ _ _. __.
1 I
JOSEPH M. FARLEY NUCLEAR PLANT UNIT 1 DOCKET NO. 50-348 SPENT FUEL POOL MODIFICATION l
O i.
March 1982 l
O ,
8203290302 820319 PDR ADOCK 05000348 PDR P
TABLE OF CONTENTS
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I. Introduction and Conclusions II. Overall Description III. Nuclear and Thermal-Hydraulic Considerations .
- 1. Neutron-Multiplication Factor 1.1 Normal Storage 1.2 Postulated Accidents 1.3 Calculation Methods 1.4 Rack Modification 1.5 Acceptance Criteria for Criticality (1) Neutron Absorber Verification (2) Decay Heat Calculations for the Spent Fuel ;
(3) Thermal-Hydraulic Analyses for Spent Fuel l Cooling . .
t (4) Potential Fuel and Rack Handling Accidents (5) Technical Specifications IV. Mechanical, Material and Structural Considerations .
O (1) Description of the Spent Fuel Pool and Racks (a) Support of Spent Fuel Racks ,
O' (2)
(b) Fuel Handling Applicable Codes, Standards and. Specifications (3) Seismic and Impact Loads (4) Loads, Load Combinations, and Structural Acceptance Criteria (5) Design and Analyses Procedures (6) Structural Acceptance Criteria (7) Materials, Quality Control and Special Construction Techniques j
(8) Testing and Inservice Surveillance l V.1. Cost / Benefit Assessment j V.2. Radiological Evaluation j V.3. Accident Evaluation h
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I. INTRODUCTION f"T !
V This r'eport provides information required by the Nuclear Regulatory Commission (NRC) in support of an application for installation of high-density poison spent fuel storage racks at the Joseph M. Earley Nuclear Plant Unit 1. The report has been prepared using the guidance of the NRC position paper entitled "OT Position for Review and Acceptance of Spent Fuel Pool Storage and Handling Applications" dated April 14, 1978, as amended by NRC letter dated January 18, 1979. Sections III through V of the report are consistent with the section/ subsection format and content of the NRC position pc'er, sections III through V: The extent of information provided involves an overall description of the spent fuel rack system and addresses disciplines such as nuclear and thermal hydraulics, mechanical, material, structural, and environmental.
The overall description (section II) provides a detailed description of the high-density poison racks and their location relative to other systems.
The nuclear and thermal hydraulic aspects of the report (section III) address the neutron multiplication factor, considering normal storage and handling of spent fuel as well as postulated accidents with respect to criticality
,/,
and the ability of the spent fuel pool cooling system to maintain sufficient cooling.
Mechanical, material, and structural aspects (section IV) involve the capability of the fuel assembly, storage racks, and spent fuel pool system to withstand effects of natural phenomena and other service loading conditions.
The environmental aspects of the report (secticn V) concern the thermal and radiological release from the facility under normal and accident conditions. This section also addresses the occupational radiation exposures, generation of radioactive waste, need for expansion, commitment of material and nonmaterial resources, and a cost-benefit assessment.
In conclusion, this report documents the compliance of the Earley Unit 1 high-density poison spent fuel rack installation with the NRC requirements, as specified in the position paper referenced above.
O I-1
II. OVERALL DESCRIPTION
(~h s_/ Low-density spent fuel racks designed by Westinghouse, with a large pitch (13 inches center-to-center), are presently installed in the Farley Unit 1 spent fuel pool.
Due to an increased demand for spent fuel assembly storage space Alabama Power Company has recognized a need to more efficiently utilize the available space in the existing pool. This can be accomplished by reracking the Unit 1 pool with high-density poison racks, designed by par Systems, with a smaller pitch (10.75 inches center-to-center). This will increase storage capacity from 675 to 1407 cells.
Plan and elevation views of the containment and auxiliary building, containing the spent fuel pool, are shown in figures II-1 and II-2. The plan view and general arrangement of the high-density storage system are shown in figure II-3, and storage racks details are illustrated in figures II-4 and II-5.
The spent fuel rack modules are free-standing and free to move on the pool liner floor during a seismic event. The module is composed of poison canisters with a bottom grid. 3 Except for the neutron absorber (vented boraflex) and threaded foot (17-4 PH alloy), all other rack materials are fabricated using 304 stainless steel alloy.
'l Three basic module configurations are planned. Their dimens' ions are 6x7, 7x7, and 7 x 8. The combined e
capacity will provide 1407 cells with the following breakdown:
Module Total Individual Configuration Cacacity guantity Cavities Weight (lbs) 6x7 42 2 84 13,500 7x7 49 19 931 15,750 7x8 56 7 392 18,000 Total: 1,407 All Deld techniques and processes will produce clean, spatter-free welds with good penetration and no slag formation. The outer canister wrapper will be gas tungsten arc spot welded (TIG) together and also TIG spot-welded onto the canister. The individual cavities will be welded on special fixtures to maintain required squareness
- w. with dimensional tolerance. The module cavities and y ) bottom grid are then welded together using a special II-l
/~ fixture to assure the assembly is square, properly
(_)T aligned, and the center-to-center spacing is accurate.
After functional testing (dummy bundle drag test), the racks will be cleaned and completely wrapped with reinforced plastic and skid mounted. The racks will be covered with tarp and shipped to the plant site. The spent fuel racks are designed to withstand shipping, handling, normal impact loads (impact and dead loads of fuel assemblies) as well as safe shutdown earthquake (SSE) and operation basis earthquake (OBE) seismic loads meeting American Seclety of Mechanical Engineers (ASME)Section III, Appendix XVII which is equiva. lent to American Institute of Steel Construction (AISC) Section 5 requirements. The racks are also designed to meet Category I seismic requirements of Regulatory Guide 1.13.
In summary, spent fuel will be stored in the spent fuel pool in 10.75 inch center-to-genter racks. The racks are composed of vented boraflex between individual austenitic stainless steel canisters, which are fastened together in a free-standing module. The module, maintaining a 10.7,5 inch center-to-center cell spacing with the neutron absorbing material, is sufficient to maintain a suberitical array.
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i
rN III. NUCLEAR AND THERMAL-HYDRAULIC CONSIDERATIONS N-]
III.1. Neutron Multiplication Considerations The effective neutron multiplicatior. factor (kef f) is calculated, assuming the following conditions in the spent fuel pool.
III.1.1 Normal Storage III.1.1.a The Farley high-density spent fuel racks are designed to store Westinghouse 17 x 17 fuel assemblies with a maximum U-235 enrichment of 4.3 weight percent without control rods or any noncontained burnable poison. The fuel is assumed to be fresh and is at the most reactive point of its life.
III.1.1.b The moderator used in the criticality analysis is pure water at the temperature within the spent fuel pool limits which yields the maximum reactivity.
III.1.1.c In the nominal case analysis, the rack array is
(~)
LJ assumed to be infinite in lateral extent.
III.1.1.d Sensitivity studies have been performed to obtain the reactivity effect of the mechanical tolerances.
III.1.1.e The analysis has taken credit for the neutron absorption by the control poison material (Boraflex), the stainless steel structural material of the spent fuel racks, and some of the structural materials of the fuel assemblies.
III.1.2 Postulated Accidents The double contingency principle of American National Standards Institute (ANSI) N16.1-1975, which stipulates that two unlikely, independent, concurrent events are required to produce a criticality accident. is used.
III.1.2.(1) a) Single assembly dropped on top of rack No adverse .eactivity effect is expected from dropping a (s)
fuel assembly on top of a fully loaded storage rack during fuel handling because of the large water thickness-(about III-1
10 inches) existing between the top of the assemblies f~/')
N-already inside the cavities and the dropped assembly resting on top of the rack. Moreover, the calculational model assumes an infinite fuel length in the axial direction.
(
b) Single assembly next to rack The dropping of an assembly outside the rack is a possible
' event because of the unobstructed water area existing between the periphery of the storage racks and the side walls of the poo). A conservative analysis was performed to evaluate this situation. The results indicate that with the presence of soluble boron in the pool water, as 4 permitted by the double contingency rule, the dropping of fuel assembly next to the rack proper does not raise the k egg value of the racks, with all uncertainties and biases included, to above 0.950.
III.1.2.(2)
Protection against a cask drop is assured by the Selsmic Category 1, Crane Manufacturers Association of American (CMAA)
Specification No. 70, Class A1, single failure-proof outdoor spent fuel cask crane, by the single failure-proof lifting device, and by the interlocks and administrative controls t
described in the Farley FSAR subsection 9.1.4.
A cask drop or tip into the spent fuel pool is also prevented by permanently installed rail stops and mechanical bumpers which prohibit cask crane travel over or into the vicinity of the spent fuel pool. The cask crane hook approach to the spent fuel pool, as shown in Farley FSAR figures 1.2-2 and 1.2-10, is l limited to approximately 12 feet. Since the cask will not be i
handled in the vicinity of the spent fuel pool, the consequences of the cask drop are not affected by the increased storage of the spent fuel pool.
The spent fuel bridge crane, located inside the spent fuel pool l
building, is used for refueling operations. The spent fuel bridge crane is the only crane capable of handling objects over the opent fuel pool. When fuel assembies are stored in the spent fuel pool, the size of the load that can be handled over the spent fuel pool is limited to 3,000 pounds by Farley Unit 1 l Technical Specification, Section 3.9.7.1. Use of the spent fuel bridge crane is discussed in Farley FSAR subsections 9.1.2 and 9.1.4.
Subsection III.1.1.(1) of this submittal discusses the dropp;ng of a fuel element on top of the racks or any other achievable abnormal location of a fuel assembly in the pool.
O) x_
i III-2
e .
III.l.2.(3) tO
(_) The exterior walls and roof slab of the spent fuel area are designed for tornado wind, differential pressure, and missile loadings on the basis that the area is fully enclosed. Farley FSAR subsection 3.3.2.1 specifies tornado design criteria for fully enclosed Category I structures. A tornado will not have an effect on the deformation and relative position of the fuel racks since the building structure will not be modified for the rerack program, and all existing building analyses will remain valid and unchanged.
The design of the racks is seismically qualified, therefore, the earthquake effect on criticality is of no concern due to the structural acceptance of the rack.
III.l.2.(4)
Loss of all cooling systems will not result in criticality. In addition, the spent fuel pool cooling system is Seismic Category I and meets the single failure criteria.
III. 1.3 Calculation Methods The criticality analysis employed two independent models to p)
( validate the results of the nominal geometry calculations.
The base calculational method employs the KENO-IV/AMPX model.
The basic neutron cross-section data comes from the master library of AMPX - a 123 group GAM-THERMOS neutron library prepared from ENDF/B version II data. The NITAWL model of the AMPX program is used to perform a Nordheim integral treatment of the U-238 resonsances accounting for self-shielding effect.
The working library produced by this process retains the 123 group energy structure and is used directly by KENO-IV. The KENO-IV/AMPX model has been benchmarked against the critical experiment data measured by Battelle Pacific Northwest Laboratories.
The diffusion theory model, which uses three codes, namely, CHEETAH-P, CORC-BLADE, and PDQ-7 is used for the validation process and is also used to perform all the sensitivity calculations. The model has been extensively tested through benchmarking calculations of measured criticals, as well as through core physics calculations for several operating reactors.
The final k eff value f r the Earley spent fuel racks is j obtained by summing the k eff value calculated by the KENO-IV/AMPX model, the calculational bias which was obtained from the benchmark work and the total uncertainty which was obtained y,) by a statistical combination of the calculational and III-3
4 mechanical uncertainties. The calculational uncertainity is such that the true multiplication factor (keff) will be less than the calculated value with a 95 percent probability at a 95 percent confidence level.
4 III.1.4 Rack Modification The spent fuel storage racks being supplied to Alabama Power Company are of new construction with no modifications to the existing racks or pool liner being required. The existing racks will be partially removed prior to installation of the new racks as described in section III.1.5.(4).
?
) III.1.4.(a) The overall fuel assembly parameters for the
,' Farley pressurized-water reactor (PWR) 17 x 17 fuel assemblies are ns follows.
Fuel assembly dimension 8.426 in. x 8.426 in.
s Storage cell pitch 10.75 in.
Percent of total cell area 61.4 occupied by a fuel assembly 1
III.l.4.(b) This section of the NRC position paper is not i
applicable since the new high-density racks utilize a neutron j
absorbing poison rather than stainless steel flux traps.
III.1.4.(c) Refer to III.1.4.(b) above.
I III.1.4.d.(1) There are two poison plates separating every two adjacent fuel assemblies stored in the racks. The poison plates have a B-lO loading of 0.015 g/cm 2.
i i
III.1.4.d.(2).a The analysis is based on an average enrichment
-of 4.3 weight percent of U-235. The fuel loading is calculated to be 54.25 gram of U-235 per axial centimeter of fuel assembly. The reactivity sensitivity of enrichment _is calculated to be 0.0031 delta k per gram U-235 change around the nominal fuel loading.
III.l.4.3.(2).b The nominal storage lattice pitch is 10.75 in.
The pitch reactivity sensitivity is calculated to be 0.010 delta k per 0.1 in. change in pitch around the nominal value.
l III-4
_ .. .- _ _ _ _ . - _ - _ _ _ . _ . . . ,_ _ . _ - - . _ . _ , . _~ _-.___.,,2
III.1.4.d.(2).c The B-10 loading in the poison plates is 0.015
/)
\_/
g/cm2 The reactivity sensitivity due to B-10 loading variation is calculated to be 0.0109 delta k per 0.005 g/cm 2 change around the nominal loading.
III.1.5 Acceptance Criteria For Criticality The acceptance criteria for criticality calculations is that keff be less than or equal to 0.95 including all uncertainties.
III.1.5(1) Neutron Absorber Verification par Systems, operating under a Quality Assurance Program which meets 10 CFR 50 Appendix B, requires the poison manufacturer to produce his product under a program which also meets 10 CFR 50 Appendix B. A detailed specification is part of the purchase order. This specification covers the neutron absorber sheet requirements, material requirements, quality assurance program requirements, documentation requirements, etc. par requires the manufacturer to submit his quality assurance program manual and operating procedures for approval before the start of production. par also audits the quality assurance program at the manufacturing facility at least once a year (or before the first order). After receipt of material, par reviews all documentation for conformance before incorporating the poison
('- into the spent fuel racks. par maintains traceability of the poison material throughout the rack manufacturing process.
Alabama Power Company or its agent is committed to periodically perform quality audits and inspection of the above described quality program.
III.1.5(2) Decay Heat Calculation for the Spent Fuel The calculations for the amount of thermal energy that will have to be removed by the spent fuel pool cooling system are made in accordance with Branch Technical Position APCSB 9-2 entitled, " Residual Decay Energy for Light Water Reactors for Long Term Cooling." This Branch Technical Position is part of the Standard Review Plan (NUREG 78/087).
III.1.5.(3) Thermal Hydraulic Analysis of Spent Fuel Cooling The computer code HPOOL is used to analyze the natural circulation cooling of the spent fuel under normal cooling conditions. HPOOL is a proprietary program of Nuclear Associates Incorporated (NAI). HPOOL calculates the precsure loss through a fuel assembly for a given flowrate. Thi.=
pressure loss is compared with the buoyant head resulting from the difference between the average density of the fluid in the p)s
(_ fuel channel and the average density of the fluid in the III-5
downcomer. The downcomer is the space between the wall of the pool and the racks. If the density difference results in a buoyant head greater than the pressure loss, the flowrate through the fuel assembly is increased and a new average density of the fluid is determined. This iterative process is continued until the buoyant head and pressure loss in the fuel assembly are cqual. Using this flowrate, HPOOL determines the fuel temperature.
HPOOL was used to determine the bulk spent fuel pool water temperature for the normal refueling case and the emergency full-core offload case. The full-core offload case represents the worst case heat loading condition which provides the maximum bulk spent fuel pool water temperature. The preliminary analysis and assumptions results for these two cases are presented below.
A. Normal Refueling:
Assumptions:
- 1. All storage cells filled.
- 2. 28 years - 1/3 of the core removed each year.
29th year - 0.4G of the core removed.
,, ( ) (Envelope's largest normal refueling discharge).
- 3. 105'F component cooling water temperature (Influent to the spent fuel pool heat exchangers.)
Results:
Total Spent Fuel No. of Cooling Hours After Heat Lead Pool Bulk Trains Operational Shutdown (108 Btu /hr) Temp. ( F) l 1 100 19.755 139
- 2 100 19.755 122 l 1 150 17.579 135 2 150 17.579 120 B. Full-Core Offload
l Assumptions:
- 1. All storage cells filled.
- 2. 25 years - 1/3 of core removed each year. 26th year - O.40 of core removed. Emergency full-core offload is necessary 180 hr after normal
'O refueling in 26th year. Full core is discharged III-6
to the spent fuel pool 10 days after the (O emergency shutdown.
- 3. 105*F component cooling water temperature.
Results:
No. of Total Spent Fuel Cooling Trains Heat Load Pool Bulk Operation (106 Btu /hr) Temnerature (*F) 1 30.384 158 2 ,
30.384 131 The computer code BPOOL is used to analyze the natural circulation cooling of the spent fuel in the event of a loss of all external means of cooling for the spent fuel pool. BPOOL is a proprietary program of NAI. The code is based on.the assumption that boiling takes place near the top of the fuel channel. BPOOL evaluates the saturation pr6petLies of the
~
, coolant on the basis of the static pressure at the top of the storage racks. These properties include water density, temperature, and steam density. The steam is assumed to separate and flow out of the pool. The water at the saturation temperature corresponding to the pressure at the top of the racks flows downward to the inlet of the storage racks. The
( sp/ static pressure at this location is higher than the pressure at the top of the storage racks and as a result the fluid is subcooled as it enters the fuel assembly. The fluid becomes less dense as it passes up the fuel channel. Near the top of the fuel channel the fluid reaches saturation conditions and net boiling occurs. The computer code, BPOOL, assumes a loss of all external means of cooling, but it should be noted that the Farley spent fuel pool cooling system is redundant and single failure-proof.
Voiding in the space between fuel assemblies is not possible since these spaces contain poison plates.
III.1.5.(4) Potential Fuel and Rack Handling Accidents The hign-density poison racks are of a free-standing design, utilizing bottom support pads, resting on the floor of the spent fuel pool. The installation of the high-density racks will include removal of the existing 13-inch center storage racks. The high-density racks will be installed wet since there is spent fuel in the storage pool.
The following is a sequence of events for installing the high-density poison racks.
Phase I Install and test a temporary crane for handling the l(')
existing racks and the high-density racks. The III-7
I spent fuel bridge crane is a 4,000-pound capacity crane and is not of adequate capacity for the re-rack modification.
Phase II Remove and decontaminate a portion of the existing 13-in. center racks, leaving enough racks intact for existing spent fuel assemblies and one emergency full-core offload and those fuel assemblies, which are stored in the pool from three previous refuelings.
At no time during this phase of work will the 13-in. spent fuel rack modules be moved over spent fuel assemblies.
Phase III Remove all adapter plates.
Phase IV Install the high-density poison racks into the pool areas vacated by the removal of the 13-inch center-to-center racks. This work will be done systematically to assure that at no time will the new spent fuel racks be moved over spent fuel assemblies. Storage capability for one emergency full-core offload will be available at all times during the spent fuel pool rerack program.
() Phase V When one full-core offload storage capacity is achieved with new racks and existin9 fuel assemblies are transferred to new racks, then remove any remaining 13-inch center-to-center racks 'and adapter plates fran the spent fuel pool.
Phase VI Install the balance of high-density racks into the spent fuel pool to complete rack installation.
Phase VII Remove temporary crane from the spent fuel pool l area.
Throughout the above phases of work the spent fuel assemblies
! in the spent fuel pool will be moved and located so that no heavy loads will be carried over them. Strict administrative procedures will be imposed throughout the rerack program to assure protection against a rack module dropping on the spent
- fuel assemblies.
l These phases of work will require support work; i.e. leveling of new racks, testing, etc.; to complete the reracking program.
l The outdoor spent fuel cask crane will be used to bring the high-density poison racks from the delivery vehicle into the j spent fuel cask area. Physical stops prohibit this crane from
' O carrying loads over the spent fuel pool. The racks will then be moved from the spent fuel cask area, by the temporary crane, i into the spent fuel pool. The reverse sequence will be III-8
-s performed to remove the existing 13-inch center storage racks v
from the spent fuel pool.
The installation of the high-density poison racks will not increase the potential for a fuel and rack handling accident for the following reasons:
e No heavy loads will be carried over any spent fuel assemblies during the rerack program. The spent fuel assemblies will be moved and located so that no heavy loads will be carried over them.
e The temporary crane, as with the spent fuel bridge crane, can carry loads over the spent fuel cask area and the spent fuel pool only. There is no safe shutdown equipment located in these areas. Therefore, there will not be any damage to safe shutdown equipment should a rack drop into these areas.
e The spent fuel cask crane, used to bring the racks into and out of the spent fuel cask area, is a single failure proof crane as described in subsection III.1.2.(2), Physical stops prohibit this crane from carrying loads over the spent fuel pool.
Protection against a rack drop is assured since the cask crane
(^'; is single failure-proof, and a dual point attachment will be
' j used between the spent fuel pool cask crane main hook and the lifted spent fuel rack module.
III.l.5.(5) Technical Specifications To ensure against criticality, the following technical specifications are proposed in figure III-1 on spent fuel storage in the high-density poison racks.
III.1.5.(5).1 Paragraph 5.6.1.1 of the proposed revision to the Farley Unit 1 Technical Specifications requires that the spent fuel storage racks be designed and maintained such that the neutron multiplication factor (kegf) in the fuel pool shall be less than or equal to 0.95 when flooded with unborated water. This represents the most conservative pool condition from a criticality standpoint.
III.1.5.(5).2 In addition, paragraph 5.6.1.1 of the proposed revision to the Earley Unit 1 Technical Specifications also specifies a maximum enrichment of 4.3 weight percent U-235 (which equates to 54.25 grams per axial centimeter of the fuel assembly) for fuel loading in the fuel assemblies. This limit
(~_ }'
is consistent with the design of the high-density poison racks N
to preclude criticality in the fuel pool.
III-9
-)
DESIGN FEATURES 5.6 FUEL STORAGE CRITICALITY 5.6.1.1 The spent fuel storage racks are designed and shall be mair ained wi th:
- a. A Kerr eqi.ivalent to less than or equal to 0.55 when flooded with unborated water, which includes c'Jnservative allowances for uncer-tainties and biases based on a maximum enrichment of 4.3 weight percent U-235.
- b. A nominal 10.75 inch center-to-center distance between fuel assemblies placed in the storage racks.
5.6.1.2 The new fuel pit storage racks are designed and shall be maintained with a nominal 21 inch center-to-center distance between new fuel assemblies such that Kerr will not exceed 0.98. based on a maximum enrichment of 3.5 weight percent U-235, assuming aqueous foam moderation.
h
%/
DRAINAGE 5.6.2 The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 149.
CAPACITY 5.'6.3 The spent fuel storage pool is designed and shall be maintained with a storage capacity limited +.o no more than 1407 fuel assemblies.
5.7 COMPONENT CYCLIC OR TRANSIENT LIMIT 5.7.1 The components identified in Table 5.7-1 are de<igned and shall be r maintained within the cyclic or transient limits of Table 5.7-1.
l l
l i
- FARLEY-UNIT 1 5-7 l
l b JOSEPH M. FARLEY PROPOSED TECHNICAL
""" " # " "*"I SPECIFICATION CHANGES
/ abarva Power unit i SPENT FUEL POOL s
MODIFIC ATION FIGURE III-l 444n-3
i
- IV. MECHANICAL, MATERIAL, AND STRUCTURAL CONSIDERATIONS
- 1 j IV.(1) Description of Spent Fuel Pool and Racks 4
Alabama Power Company has purchased high-density spent fuel storage racks for the Joseph M. Farley Nuclear Plant Unit No. 1. The spent fuel pool can contain sufficient high-density racks to safely store 1407 spent fuel assemblies. Figures IV-1 through IV-7.show the plans and sections of the spent fuel pool with figure II-3 showing the arrangement of the spent fuel i storage racks. Details of the spent fuel pool structure and liner are shown in figures IV-8 through IV-20.
The high-density (poison) spent fuel storage racks are of stainless steel welded construction. They consist of three basic components. (See figures II-4 and II-5.)
- 1. Bottom grid.
I
- 2. Neutron absorber canister (poison cans).
- 3. Adjustable foot assembly.
T The neutron absorber canister, hereafter called poison can, consists of a stainless steel wrapper holding the Boraflex
'() firmly against the inner can. The neutron absorber (Boraflex) is comprised of a polymeric silicone encapsulant entraining and
, fixing fine particles of boron carbide in a homogeneous stable mixture. The outer wrapper is comprised of two L-shaped sheets which are firmly pressed against the Boraflex and inner can and i then spot welded along the canister length at diagonally opposite corners, i
Only the inner tubes of the poison cans act as structural elements. The upper ends of the inner tubes are expanded by die forming to provide lead in surfaces for the fuel. The
- upper ends of adjacent cans are welded together to form the j module top grid. The lower die formed ends of the inner tubes 4
are also welded to the bottom grid. The fuel support surfaces and rack support feet are integral to the bottom grid. Large leveling screws are located at the bottom grid feet to adjust
, for variations in pool floor level.
The maximum overall floor loading associated with a seismic event (D + L+T1 + E2) is 3731 lb/ft 2 with the allowable being 3800 lb/ft2 The maximum local foot load of 164,300 lb results in a bearing stress of 4949 lb/in.with the allowable stress being I 5950 psi. The overall pool floor loading associated with a fuel assembly drop (worst case drop accident) onto the storage racks (D & L & T & I) is less than the loading associated with j a seismic event. The drop accident associated with a fuel j assembly is discussed in Section IV of the apent fuel pool 4
modification report.
IV-1
73 The Farley Plant Technical Specifications prohibit the spent
() fuel pool bridge crane from transporting any loads greater than 3000 pounds over the spent fuel pool. Therefore, no heavy loads will be moved over the pool. The heaviest loads that the bridge crane can carry over the pool is a fuel assembly or spent fuel pool gate. Due to the physical limitations between the racks and wall it is impossible to drop a fuel assembly onto the pool floor. With the new high density storage racks it is improbable that the gate will drop freely onto the pool liner. This improbability is due to the thickness of the gate, the extension of the gate hangers (hooks), the close proximity of the new racks to the pool wall and drag afforded by the water in the pool. ..., _
IV.(1).a Support of Spe_nt Fuel Racks The spent fuel storage racks are freestanding and are thus free to slide or rock on the pool floor during a seismic event. The only interface with the pool floor are the four stainless steel pads (per module), attached to the rack leveling screws. In areas where interference with a pool floor weld seam occurs, a stainless steel plate will be placed to bridge the obstruction.
This plate will be grooved for placement over a weld seam.
The only floor obstructions that require bridging are two weld
(^T seams. The remaining weld seams and all embedment bolts are
( s/ far enough away from the rack feet to eliminate interference.
The floor bridging plate shown in figure IV-21 reduces loads on floor weld seams for following reasons:
- 1. Coefficient of friction between the machined foot and 1/2-in. plate is less than the coefficient of friction between the 1/2-in. plate and liner plate. Therefore, the rack foot should slide on the 1/2 -i n . plate rather than the 1/2-in. plate sliding on the liner plate.
- 2. If the 1/2-in. plate were to slide with the foot it would take an incremental amount of force to exist between the 1/2-in. plate and liner plate to allow the foot to slide on the 1/2-in. plate. This incremental force would be insignificant and not affect the weld seam.
This method of bridging will allow a smooth, unobstructed surface for the sliding of the rack foot stainless steel pad.
This type of restraint system has the following advantages.
- 1. Uplift loads are eliminated from any pool floor embedments.
f'}
s-
- 2. Horizontal loads are eliminated from pool walls.
IV-2
i J
i '
- 3. -Horizontal forces on the pool floor are reduced
. relative to a vertically restrained rack.
- 4. Each rack is individually'self-supporting and can be removed or installed with minimal. effort.
The racks are individually installed with the bottom grids of I
adjacent racks butting to one another leaving a nominal 5/8-in.
gap at the top. Since the racks are not tied together there is potential for rack interaction along with sliding and rocking.
These three concerns are all considered in the analysis and discussed further in paragraph IV.(S).
i The pool layout (figure II-3) leaves ample clearance for
' seismic displacement and coolant. flow-between the perimeter j racks and the spent fuel pool walls.
IV.(1).b Fuel Handling 1
4
' The spent fuel storage rack is designed to withstand the following fuel handling and heavy load impact conditions.
- 1. Fuel bundle drop-from 42 in. above the rack impacting on the middle of the top grid.
- 2. Fuel bundle drop from 42 in. above the rack impacting on the-corner of the top grid.
4
- 3. Fuel bundle drop from 42 in. above the rack free falling through an empty cavity and impacting the bottom grid.
b
. 4. Inclined fuel bundle drop on' top of the rack.
l 5 Gate drop from 9 in, above the rack impacting on the top of the rack.
i.
Refer to figures IV-7 and IV-22 for details of the fuel handling system.
IV.(2) Applicable Codes, Standards, and Specifications The spent fuel storage racks are a welded structure consisting of materials of United States origin. The following chart
, presents material specifications and alloys used in the rack i assembly.
, s-
' IV-3
. Material Spec. Description Alloy (l
ASTM-A240 Bottom 304SS Gri d ASTM-A240 Outer Wrapper 304SS ASTM-A666 Inner 304SS Gr. B Tube ASTM-A564 Threaded Foot Type 630 H-1100 ASME-SA351 Bottom Grid Foot CF-8C The ASTM-A666 material is equivalently qualified to the ASME Code, Subsection NF (SA240 material) with the exception of the higher yield strength (Grade B) which is documented by certified material test reports.
The allowable stresses in the heat affected zones of the weld and base material, utilizing ASTM-A666 in storage rack construction, are higher than the allowable stresses of ASME-SA240 steel. Neither the ASME Code or the ASTM Code requires an assumed reduction in yield strength in heat affected zones of the weld and base material. Tests have been performed by an independent testing laboratory on samples of
(- welded A666, Grade B material. These tests indicate an average yield stress to be 17.9 percent higher than the minimum yield stress specified by ASTM for unwelded A666 Grade B material and 76.9 percent higher than the minimum yield stress specified for SA240 material. These welded samples failed in the heat affected zone.
Further A666 material is currently in use in a comparable spent
> fuel rack design. This design was licensed by the NRC and has design stresses in the heat affected zones based upon the ASTM Code allowable stresses for the A666 material. The design stresses for the proposed Farley Nuclear Plant Unit 1 racks are comparable to those currently licensed.
The material certification records document the yield strength of the material to be less than 90,000 lb/in In this particular design the stress corrosion cracking problem is not a concern due'to several factors. Stress corrosion cracking may occur where extensive intergranular precipitation is found in combination with tensile stresses. While some residual stresses cannot be avoided in this design, the amount of heat input from welding is small and the rate of cooling is rapid. Low heat input, a maximum interpass temperature of 350 F for all welds, and rapid cooling due to the configuration of intergranular precipitation occurs when a high temperature
[) is held for a long period of time; i.e., a typical sensitizing IV-4
treatment would be 1200 F for 2 hr; which will not be the
(()
case.
In addition to the above factors all welds will be made with 308L filler metal and the heaviest weldment (the foot corner casting) will have a 0.03-percent maximum carbon content.
These factors also greatly reduce the level of susceptibility to stress corrosion.
, par Systems has manufactured many spent fuel storage modules and other equipment of welded Type 304 stainless steel for nuclear applications with no indication of stress corrosion cracking problems.
The design of the spent fuel storage rack is in accordance with
.Section 5 of tne AISC Steel Construction Manual with fabrication welding' in compliance with ASME Section IX.
In as much as this is not a code stamp project the certification requirements of NF4000 of the ASME Code do not apply. This is also true for the design requirements of NF3000 and the marking requirements of NF8000. With these exceptions, GCA/ par's quality and fabrication programs meet ASME Section III Subsection NF.
The racks were also designed and fabricated to meet and utilize ps the applicable portions of the following regulatory guides,
( safety review plan sections, published standards, and computer programs.
- 1. United States Nuclear Regulatory Commission (USNRC)
- a. Reg. Guide 1.13 Spent Fuel Storage Facility Design Basis, Rev. 1, Dec.
1975.
- b. Reg. Guide 1.29 Seismic Design Class, Rev. 2, Feb. 1976.
- c. Reg. Guide 1.92 Combination of Modes in Seismic Analysis, Rev. 1, Feb.
1976.
- d. Reg. Guide 1.38 "QA Requirements for Packaging, Shipping, Receiving, Storage, and Handling of Items for Water Cooled Nuclear Fower Plants",
Rev. 2, 1977.
- e. Reg. Guide 1.60 Design Response Spectra for Seismic Deisgn of Nuclear Power Plants, Rev. 1, Dec.
O' 1973.
IV-5
, s f. Reg. Guide 1.61 Damping Values for Seismic Design of Nuclear Power Plants, Oct. 1973.
- g. Reg. Guide 1.31 Control of Ferrite Content in Stainless Steel Weld Metal, Rev. 3. April 1978.
- n. Reg. Guide 1.124 Service Limits and Loading Combinations for Class 1 Linear-Type Component Supports, Rev. 1, January 1978.
- i. SRP 3.7 Seismic Dec1gn, 1975.
- j. ERP 3.8.4 Seismic Category I Structures, 1975.
- k. SRP 9.1.2 Spent Fuel Storage, 1975.
- 1. NRC Guidance on Spent Fuel Pool Modifications, Review and Acceptance of Spent Fuel Storage and
}{andling_ Applications (April 14, 1978 revised Jan.
18, 1979).
i 2. Industry Codes and Standards
- a. ASME Boiler & Pressure Vessel Code Section IX and Section III, Appendix I, XVII, and Article NF-4000, 1980 Edition (American Society of Mechanical Engrs.).
, 8th Edition, Dec. 1980 (American Institute of Steel Construction).
l c. ACI 318-71 Building Code Requirements for Reinforced Concrete.
(American Concrete Institute.)
- e. ANSI N45.2 " Quality Assurance Program Requirements for Nuclear Facilities," 1977.
- f. ANSI N45.2.2 " Packaging and Shipping, Receiving Storage and Handling of items for Nuclear Power IV-6
- 1. , ,
, Plants," 1972, except Para.
- j O 2.4 and 2.6.
- g. ANSI N210 Des 2 gn Objectives for liigh Water Reactors Spent Fuel Storage Facilities at Nuclear Power Stations, 1976".
- h. ANSI N45.2.10 " Quality Assurance Terms and Definigion," 1973.
- i. ANSI N18.2 Nuclear Safety Criteria for the Design of Stationary '
Pressurized Water Reactor Plants, 1973.
- j. SNT-TC-1A Recommended Practice for Personnel Qualification and Certification in Nondestructive Testing, American Society for Nondestructive Testing, 1975.
- k. ANSI N]6.9 Validation of Calculation Methods for Nuclear Criticality Safety," 1975.
(_/ ' 3. Federal Specifications (Standards)
- a. 10 CFR 10 Code of Federal Regulations, Title 10, Part 50 (Appendix A and B).
- b. 10 CFR 73.55 Requirements for Physical Protection of Licensing Activities in Nuclear Power Reactor Against Industrial Sabotage.
- c. 10 CFR 20 Standards for Protection Against Radiation.
- d. 10 CFR 21 Reporting of Defects and Nonconformances.
- 4. Computer Programs
- a. ANSYS Computer Program " Engineering Analysis System" Swanson Analysis Systems, Inc.
,, b. SIMQKE Computer Program, digitized
( : time histories are generated 5/
artifically using "SIMQKE IV-7
f"' which uns developed under the auspices of the National Science Foundation.
t c. RCKN 1 Computer program for considering nonlinear rocking sliding motion of cubmerged eccentrically loaded fuel racks. It also considers simultaneous horizontal and vertical time history acceleration.
- d. SPECT Computer Program, a subroutine of SIMQKE for the computation of spectra from time histories digitized at equal time intervals.
- e. CHEETAH-P/CORC-B/ Computer Programs for PDQ-7 Criticality Analysis (diffusion theory model).
- f. KENO-IV-AMPX-123 Computer Programs for-Criticality Analysis (Monte Carlo model).
f)
%/ g. HPOOL Computer Program for Spent Fuel Pool cooling waterflow and heat transfer under normal condition.
h< BPOOL Computer Program for Spent Fuel Pool cooling waterflow
' and heat transfer under boiling condition.
IV.(3) Seismic and Impact Loads I The floor response spectra and damping values are 2 percent (OBE) and 5 percent (SSE). The design response spectra for seismic design Category 1 structures are discussed in the Parley FSAR under design response spectra (paragraph 3.7.1.1) and design response spectra deviation (paragraph 3.7.1.2). The damping values for seismic design for Category 1 structures are discussed in paragraph 3.7.1.3, critical damping values. These values were reviewed and approved by the NRC staff during the operating license review.
m IV-8
In addition, the NRC position paper states that it is
(()
acceptable to use different damping values from Reg. Guide 1.61 if the response spectrum curves were developed for the particular plant and not from Reg. Guide 1.60; therefore, the 2-percent and 5-percent values are acceptable.
The computer model. used in the time history analysis for the new spent fuel storage racks are three dimensional. This model was excited simultaneously with the three orthogonal earthquakes.
The resulting stresses are actual stresses with no wethod of combining required.
The seismic analysis was performed on the spent fuel pool structure for each horizontal direction and vertical direction.
The loads from the analysis for the vertical direction were combined with loads from each horizontal direction separately; i.e., loading 1 is N-S and vertical, loading 2 is E-W and vertical. This procedure is in accordance with FSAR paragraph 3.7.2.1 which was reviewed and approved by the NRC staff as part of the operating license review.
The response spectrums used for the seismic rack analysis were conservatively developed from an enveloped response spectrum of both the OBE (2 percent) and DBE (5 percent) curves. The computer program "SIMOKE" was then used to generate the time
. histories for each of the 3 envelope curves (N-S, E-W, Vert.).
(( ) The spent fuel pool floor response spectra are provided in Farley FSAR section 3.5 and subsections 3.7.1.1 and 3.7.1.2 which were reviewed and approved by the NRC staff as part of the operating license review. There are no rooms below the spent fuel pool.
The pool floor rests on concrete columns and earth as shown in figures IV-4 and IV-7. The pool is part of the Auxiliary Building. Floor response spectra for auxiliary building elevation 121 ft O in. are shown in table IV-1.
For a detailed description of analysis method and the parameters involved see paragraph IV.(5).
IV.(4) Loads, Load Combinations, and Structural Acceptance Criteria The following contains the loads, load combinations, and design allowable stress to which the racks are designed.
A. Load Definitions ,
D = Dead load of racks L = Live load due to the weight of fuel assemblies which shall be considered as varying from zero to full load and loadings corresponding to i(]} varying placement of the fuel assemblies in IV-9
the rack chall be considered so that the most critical' loads are obtained.
t( )
T = Thermal loads for water temperature of 150 F.
The minimum water temperature is 40 F.
P = Lifting force of 4000 pounds applied to the top of the rack at any fuel bundle location.
H = Horizontal force of 1000 pounds applied to the top of the rack at any fuel bundle location.
E = Loads generated by the OBE resulting from ground surface horizontal acceleration and vertical ground surface acceleration acting simultaneously.
El = Loads generated by the SSE resulting from ground surface horizontal acceleration and vertical ground surface acceleration acting simultaneously.
T3 = Thermal loads for loss-of-coolant condition corresponding to pool surface temperature of 212 F (240'E at rack elevation).
I = Impact load resulting from the follot ng
(_ conditions:
Condition 1 -
fuel drop from 42 in, above the rack impacting on the middle of the top grid.
Condition 2 -
fuel drop from 42 in. above the rack impacting on the corner of the top grid.
Condition 3 -
fuel drop from 42 in. above the racks free falling through an empty cavity and impacting the bottom grid.
Condition 4 -
inclined fuel bundle drop on the top of the rack.
Condition 5 -
gate drop from 9 in. above the rack impacting en the top of the rack.
IV-10
('}
B. Load Combinations i V The following load combinations shall be satisifed:
Load Combinations Stress Limit
- 1. D+L+T+P Fs
- 2. D+L+T+H Fs
- 3. D+L+T+E Es
- 4. D+L+T+ I Condition 1 1.6 Fs (1)
Condition 2 1.6 Fs (1)
Condition 3 1. 6 Fs (1)
Condition 4 1.6 Fs (1)
~
Condition 5 1. 6 Fs (1)
- 5. D+L+T2 + E2 1.6 F s NOTE:
(~)
Local failure of the fuel support or the top grid impact interface is allowed. However, overall member stresses shall be limted to 1.60 Es and resulting rack deformation shall not cause the fuel configuration to reach a keff of 0.95.
C. Design Allowable Stresses
= Allowable working stress l Es f s' = Calculated stress Fy = Yiel.; stress D. Allowable S,resses (for Stainless)
The allowable stresses are in accordance with ASME Boiler and Pressure Vessel Code Section III Appendix XVII. This is interpreted as being identical to the AISC Steel Construction Manual (Section 5).
The one-third increase in allowable stress for emergency condition is not allowed. The increase in allowable stress is defined by the preceeding
() paragraph B.
IV-11 l
. The following table summarizes the maximum stresses of
( (-)
the rack components for each pertinent load combination. The following calculated seismic stresses are based on the cumulative conservations associated with bounding the OBE and DBE response spectra and using the 2 percent damping values.
Figures II-4 and II-5 of Section II are structural drawings for the assembly of the spent fuel modules.
7 x 8 RACK MAXIMUM STRESS INTE2 ACTIONS (ACTUAL /
ALLOWABLE) (a)
.135 .120 Load Comb. Cans Cans Grid Cruciform (b) Foot D+L+T+P -
.27 .38 .25 .32 D+L+T+H x
.23 .41 .26 .32
,- D + L + T + Hz -
.25 .40 .26 .31 (c) D+L+T+Is -
.92 .58 .5 .47 (c) D+L+T+ Ia .54 .44 .46 .45 .35 (c) D+L+T+ I 3
.27 .63 .62 .39 D+L+T+E .71 .90 .51 .28 .38 D+L+T2 + E2 .51 .65 .37 .20 .27 O
6 x 7 RACK MAXIMUM STRESS INTERACTIONS (ACTUAL /
ALLOWABLE) (d)
.135 .120 Load Comb. Cans Cans Grid Cruciform (b) Foot D+L+T+E .66 .88 .44 .28 .21 D+L+T2 + E2 .47 .63 .32 .20 .15 NOTES:
(a.) 7 x 7 rack bounded by the 7 x 8 rack.
(b.) Fuel support cruciform.
(c.) I, & Is are bounded by these drop conditions.
(d.) Static load cases are bounded by the 7 x 8 racks.
IV.(5) Design and Analysis Procedures The following is a brief description of the methods used to structurally analyze the spent fuel storage rack design. This freestanding rack design was structurally qualified by a detailed time history and static analysis.
IV-12
Simplified time history analysis was done at both 0.2 and 0.8 coefficients of friction (u) conditions with 0, 1/4, 1/2, f ('}
N- 3/4, and full eccentric fuel loading conditions. The low coefficient (0.2) was used to define maximum credible sliding displacement, and the higher coefficient (0.8) was used to define the worst loading condition on a rack. These simplified analyses were done using par proprietary computer program RCKN1 and are further explained in paragraph A of this section.
A detailed time history analysis using the ANSYS computer code was then performed for the worst loading condition (previously defined) on a three-dimensional double rack model (6 x 7 and 7 x 8). This detailed analysis defines the combined dead, live, and seismic stresses on a 6 x 7 and 7 x 8 rack. The model and analysis method are further explained in paragraph B of this section.
A static analysis was done on each detailed ANSYS finite element model of a 6 x 7 and 7 x 8 rack for the following static load cases:
- a. Rack dead load ("D" loading).
- b. Fuel load ("L" loading).
- c. Impact load ("I" loading).
t () d. Fuel handling ("P", "H" loading).
This model and method of analysis are further described in paragraph C of this section.
The stresses resulting from the static analysis were combined as in load combinations 1, 2, and 4 of paragraph IV. (4) B and stresses from the time history analysis were combined as in load combinations 3 and 5 of paragraph IV. (4) D.
The resultant member stresses from the above load combinations satisfy the allowable stress limits as stated in paragraph IV.(4).
A. RCKN1 Time History Model Description This model (figure IV-23) was used to determine the worst loading conditions based on the afo.rementioned eccentric fuel loading conditions and varying coefficients of friction. Fuel rattling and rack / rack interaction were not considered in this analysis. In this program the rack is idealized as a vertical beam connecting to a riding base via a torsional spring.
This spring / beam is sized to match the lowest horizontal rack frequency. In addition, this beam may
.~ be located eccentrically on the base to account for h(_)5 eccentric fuel loading. At each corner of this base, IV-13
7 -
(~T vertical gap springs are located. These springs take
\msI only compression loads allowing for rack uplift and rocking. At the lower left corner of the base a horizontal slider spring is located. This element allows for sliding when the horizontal force exceeds y times the total ger spring force. Fluid coupling forces are also included in the equations motion. The fluid coupling forces are assumed to be either in-phase or out-of-phase with the support motion ~. The model is excited using simultaneous horizontal and vertical support accelerations.
All the springs have dashpots associated with them to represent structural damping.
This model has been benchmarked against a comparable ANSYS computer model.
B. ANSYS Time History Model Description To consider the effects of module rocking, sliding, and interaction, the double. rack three-dimensional ANSYS model was used and is shown on figure IV-24.
For illustration purposes this model is a beam representation (Section (1h of a 6 x 7 and 7 x 8 rack.
However, this mr_el was generated directly from the
() detailed rack models, described later in paragraph C, using the super element capabilities of ANSYS. This time history model was then verified by comparing its fundamental natural frequencies to the detailed finite element models frequencies.
Section No.(h of this model represents the mass and stiffness of all the fuel assemblies extending the height of the rack. It is pinned at the bottom of the I
rack and is allowed to impact at the top, top-quarter, and midpoint. A gap at the top-grid, top-quarter, and
- midpoint represents fuel-to-can clearance. These clearances are represented in the model by ANSYS gap spring elements. Note, also, that this model conservatively assumes that all fuel assemblies are in phase and move together at all times.
l The rack is connected to the floor by use of three-l dimensional interface elements. This element
- represents two plane surfaces which may maintain or l break physical contact and slide relative to each other. At each time step, the program determines if tensile forces exist in the element; e.g. allowing for uplift and rocking. In addition, the program checks the horizontal force in the interface element to see gg if sliding occurs.
O IV-14 l
The double rack model includes module interaction or
(( ) potential for banging with other racks in the pool.
Gap springs are located at the top grid elevation and initially have a 0.625-in. gap. This model assumes that the largest interaction occurs for a pair of racks because their rocking motion away from each other is unconfined by no adjacent modules.
A. single vertical degree of freedom spring represents the pool floor under the racks. The spring rate is calculated based on the floor stiffness and using the mass of both racks.
A structural damping of 5 percent (SSE) and 2 percent (OBE) for welded steel frame structures was used. All internal water entrapped within the rack envelope is added to the horizontal mass. The external hydrodynamic water mass determination is based upon a paper by R. J. Fritz entitled, "The Effects of Liquids on the Dynamic Motions of Immersed Solids," Journal of Engineering for Industry, February 1972. Both racks for this analysis are assumed to be full of fuel.
The model accounts for in-phase fluid coupling with the pool water by use of a fluid dynamic coupling element. This element is used to represent dynamic coupling between two points. The coupling is based on
% the dynamic response of two points connected by a constrained mass of fluid, as described in the aforementioned paper by R. J. Fritz.
The effects of sloshing on the spent fuel pool wall were predicted based on the method of analysis presented in Chapter 6 AEC Publication TID 7024,
. " Nuclear Reactors and Earthquakes." The analysis for SSE conditions indicates a maximum water surface displacement of 11 in. and maximum convective pressure of 40 lb/ft 2 due to sloshing. Normal water surface level in the pool is approximately 16 in.
below the top of the pool walls. Since the minimum thickness of the concrete walls for the pool is 42 in., the convective pressure on the pool walls due to sloshing will be negligible.
The impact damping of 10 percent was used for all gap elements. The repetitive impacting of these elements dissipates substantial amounts of energy.
Consequently, there is a higher damping within the structure than would exist if there were no gaps.
C. Finite Element Model
['N The computer program ANSZ3 was used to analyze the t>d detail finite element model.
IV-15
s Figure IV-25 delineates the computer model. The spent
) fuel rack is idealized as a three-dimensional detall
~'
finite element model of nodal points, consisting of chell and beam elements representing the poison cans and bottom grid.
The poison cans are flared at the top and welded to each other to form the top grid and fuel lead-in surfaces. The cans are also welded into a bottom grid. The super element capability of ANSYS is used to model the cans and grid. (See figure IV-26.) Each of the poison cans are modeled with the same super element. The can super element consists of over 80 quadrilateral shell elements. This allows proper modeling of the flared can section and weld connections. The bottom grid super element is modeled with beam, solid, and shell elements.
The rack static model is then made by assembling the can and grid super elements, by coupling the top can nodes to each other, and by coupling the bottom can nodes to the bottom grid.
IV.(6) Structural Acceptance "riteria
- Refer to paragraph IV.(4) for the acceptance ariteria used for m) the design of the racks. This criteria is consistent with the applicable sections of the Standard Review Plan 3.8.4 which is to be used in conjunction with AISC code.
IV.(7) Materials, Quality Control, and Special Construction Techniques The materials used in the fabrication of the spent fuel storage racks are per the chart in paragraph IV.(2). These materials have Celtified Material Test Reports which include actual chemical and physical test results. The weld filler materials also have, as a minimum, certificates of compliance and ferrite data.
The fabrication of the racks is in accordance with the following QA/QC procedures:
e Manufacturing and inspection plan.
e Cleaning procedure.
e Liquid penetrant inspection procedure.
e Visual weld inspection procedure.
I, _ )
'- e Welding procedures.
IV-16
e Packaging, shipping, and handling procedure.
(,_
(_/' e Final inspection / test procedure.
e Procurement specification for Boraflex.
e Procurement specification for castings.
e Weld repair procedure.
e QA documentation checklist.
e Fabrication and inspection procedure.
The high-density racks will be installed wet since there is upent fuel in the storage pool. The spent fuel racks to be installed are independently self supporting and can be installed in any sequence. Each rack is equipped with a leveling screw (11.0-in. adjustment) in each corner which allows a rack to be individually leveled after installation.
This method of adjustment compensates for variations in the pool floor. A storage location system will also be installed around the perimeter of the rack layout which allows for individua' cell identification.
s IV.(8) Testing and Inservice Inspection
\
l Refer to BISCO reports 748-10-1 and 748-30-1 which contain irradiation studies and performance data for the neutron absorbing material Boraflex.
Poison test coupons are supplied for an inservice surveillance program of the neutron absorber (Boraflex) used in the high-density fuel storage racks. The coupons duplicate the condition of the Boraflex which is encased in the poison canisters. Ten coupons are provided; they are to be removed and analyzed at intervals of 1, 2, 4, 7, 11, 15, 20, 25, 30, and 35 years after installation.
The poison coupons are to be hung alongside the high-density fuel racks, to be subjected to the maximum neutrons, gamma and heat flux.
The procedure for fabrication and testing of the coupons is as follows.
- 1. Boraflex samples are cut to cice and conditaaned in normal atmosphere at 20 C to 30 C and 30 percent to 70 percent relative humidity for 3 days.
p
( )
%j IV-17
- 2. Each sample has the following measurements taken at predetermined points on the samples:
- 1. Dimensions including thickness, length, and width.
2-. Hardness on the Shore A scale.
- 3. Neutron Attenuation at 0.06 eV.
- 3. Each sample is then fabricated into a coupon and this coupon is installed in the pool.
- 4. A coupon is to be removed per the above schedule.
- 5. -Carefully open the coupon without damaging the neutron absorber (Boraflex) sample. Condition the samples in accordance with Step 1 of this procedure.
- 6. Remeasure the samples as indicated in Step 2, assuring that the measurements are made in the same
-location and by the same procedure as originally performed.
- 7. Visually examine the surface for change. Take photographs of the surface and any suspect areas.
- 8. Prepare report of sample test results'and observations.
'9. Should any adverse conditions be detected, the samples may be subject to further neutron transmission studies.
10 '. Retain samples.
NOTE:
A reference neutron absorber sample will be provided for each coupon. This reference sample is from the same strip of neutron absorber as the coupon sample. This reference sample can be used for future comparison to-the installed sheets in the test coupons.
O IV-18
TABLE IV-1 PERCENTAGE OF CRITICAL DAMPING FACTORS 1/2 Safe Safe-Shutdown Shutdown Earthquake Earthquake (E')
0.05g Ground 0.10g Ground Type of Structure Acceleration Acceleration Vital Piping 0.50 1.00 -
Welded Steel Plate 1.00 2.00 Assemblies Welded Steel Frame 2.00 5.00 Structures Bolted and Riveted 3.00 5.00 Steel Reinforced Concrete 2.00 5.00 Structures and Equipment Supports Prestressed Concrete 2.00 5.00 Structures Soil Damping 4.00 7.00 l
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SPENT FUEL POOL moOlFICATION FIGURE IV-1 i
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.c.4c JOSEPH M. FARLEY SPENT FUEL POOL LINER NUCLEAR PL ANT SECTIONS G,H,J&K AUXILIARY BUILDING
""'" I t\lalXinia POWCf SPENT FUEL POOL 8 MODIFIC ATION FIGURE IV-20 I
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3 V COST / BENEFIT ASSESSMENT V.1 Environmental Assessment Following is the information required for the environment cost / benefit assessment. V.1.1 Sp_ec_ial Needs for Increased Storage Capacity The present spent fuel storage facility for Farley Unit 1 was designed for temporary storage,of. spent fuel until the fuel heat output had decayed sufficiently for transport to a reprocessing facility. The absence of activity in the construction of fuel reprocessing facilities and the cessation of operation of existing reprocessing facilities have created the need for increased onsite storage of spent fuel to permit continued power plant operation. V.3.1(a) Alabama Power Company currently has no contracts for reprocessing. V.1.1(b) The [nticipated spent fuel discharge schedule for Farley Unit 1 is described in table V-1. The existing spent (~x g ,) fuel pool has a capacity of 675 storage locations. When that capacity is filled to less than 157 storage locations remaining then full-core storage reserve capability is lost. Under the schedule described in table V-1, full-core storage reserve capability is lost with the 1990 discharge and all storage capacity would be expended with the 1993 discharge. V.1.1(c) At present 136 spent fuel assemblies are stored in the Farley Unit 1 spent fuel pool. V.1.1.(d) At the end of the current refueling outage, the following items will be stored in the Farley Unit 1 spent fuel pool: '
- 1. Surveillance capsule rack (one). ,
- 2. Insert holders (three).
V.1.1(e) Under the proposed expansion, the spent fuel pool capacity would be increased by 732 storage locations which would provide approximately 14 years of additional storage capacity, based on the schedule shown in gtable V-1. 1 [s_ ) , i V-1 ' o
^g V.1.1(f) With the expanded capacity, full-core storage reserve
( (V' capability would be lost with the 2004 discharge and all storage capacity would be expended with the 2007 discharge. (See table V-1.) V.l.2. Total Construction Associated with Proposed Modification Alabama Power Company will be removing the existing non-poison
- racks in Farley Unit I and replacing them with the new poison racks which are being fabricated by the vendor's facility. A temporary crane will be installed for this purpose. There will be no modifcations to the spent fuel pool liner plate. The spent fuel pool lighting system will be slightly modified to accommodate the new racks.
Listed below are the costs associated with procuring fabricated racks, removing the existing racks, and licensing and installing the new racks. e_, Progress payment U Nine months - $1,320,695 Thirteen months - 708,135 Seventeen months - 236,045 Total of progress payments $2,264,875 Cost of money for progress payments at 20-percent interest. A i normal payment date of 11 months after placement of the order is used. l l Nine months -$ 44,024 l Thirteen months - (23,605) Seventeen months - (23,609) Net cost of money $ (3,190) for progress payments Total for manufacturing racks $2,261,685 Estimated cost of engineering and $ 247,000 licensing (other than that provided by the rack supplier) r) (b) V-2
d Estimated cost of installation $2,100,000 (including installation and testing
' / of-new racks, removal, decontamination, and disposal of existing racks)
I' Estimated-total costs associated $4,608,685 with reracking The existing racks in Farley Unit 1 will be' decontaminated an'd shipped to a burial site as radwaste. Alabama Power Company presently has no plans to fully decontaminate the racks for scrap. i r+ The progress payments are firm and will be made in late 1982 and early 1983. The engineering costs will be incurred in early 1982 and are estimated in 1981 dollars. The installation costs will be incurred in late 1983. They are also estimated in 1981 dollars. A V.1.3 Alternatives to Increased Storage Capacity
! Several alternatives to the expansion of the Farley Unit 1
< spent fuel pool to alleviate the spent fuel storage space shortage were considered. i () V.1.3.(a) There are no commercial spent fuel reprocessing facilities in operation in the United States. Therefore shipment to a reprovassing facility is not a viable alternative . to the expansion of the Earley Unit 1 spent fuel pool. V.I.3(b) Spent fuel storage at a private or government l operated independent spent fuel storage facility to serve Farley Unit 1 is not currently available. The alternative of l constructing a facility to serve Farley Unit I would not be
- economically viable. Table V-2 shows some selected alternative storage facilities to spent fuel pool expansion. The lowest
- unit cost alternative is the 10,000 MTU open field drywell i
storage facility at $48/KgU in construction costs. A facility si ed down for the storage needs of Farley Unit 1 would be expected to cost more per kilogram. These costs are j significantly higher than the estimated cost of the increased storage capacity which will be obtained by expanding the present spent fuel pool (approximately $14/KgU construction
- costs).
, V.1.3.(c) The only other reactor site within Alabama Power Company which could store Farley Unit 1 fuel is Farley Unit 2. Farley Unit 2 is expected to exhaust its present spe'nt fuel 1 storage capacity by about 1994. Under its proposed spent fuel 1 pool modification (Joseph M. Farley Nuclear Plant, Unit 2, V-3 I
Docket No. 50-364 - Spent Fuel Pool Modification, December
'()
1981) Unit 2 would be expected to exhaust its expanded storage capacity by about 2009. Storing Unit 1 fuel in the Unit 2 pool would advance the date of loss of storage capacity at Unit 2. Storing Farley Unit 1 fuel at the expense of the spent fuel pool capacity for Farley Unit 2 is not a viable alternative. V.1.3.(d) The shutting down of the Farley Nuclear Plant-Unit I would require either purchase of replacement power outside of the Alabama Power Company system or construction of new generating capacity'. The installation of the new racks will allow an additional 14 years of operation at a cost of
$3,582,577 (approximately $17,000,000 in 1994 dollars). The purchase of replacement power is not a viable option as no long-term commitment may be obtained for the purchase of 863 MW from either within or outside of the Southern Company system.
All current' Alabama Power Company long-term generation expansion planning uses standardized 818-MW fossil units. The costs (1995 dollars) of placing an 818-MW size fossil plant in operation in 1994 (old fuel racks filled to capacity in 1993) are as follows:
- 1. Capital Cost $2,156,760,000
.. 2. Scrubber Capital Cost S 358,200,000
'( ) 3. Fixed o & M $ 23,510,880/ year
- 4. Variable O & M $ 10,655,100/ year
- 5. Environmental Cost $ 21,100,000/ year
- 6. Fuel Cost $ 287,645,500/ year It is readily seen that the cost of the increased spent fuel pool storage is insignificant compared to the replacement cost of the Farley Unit 1 generation. These figures do not include any capital (fixed) cost dollars for Farley Nuclear Plant Unit I that still would have to be amortized whether the plant is operating or not. Also, not included is the cost of maintaining the plant in a shutdown condition and maintaining site security.
V.1.4 Resource Committment The relatively small quantities of material resources that would be committed to the proposed modification would not significantly foreclose the alternatives available with respect to any other licensing actions designed to ameliorate a possible shortage of spent fuel storage capacity. The material ( resources that would be comsumed by the proposed modification are listed below. V-4 o N
I
.I Material Quantity (1b) Dollars /lb l g/^')
l 1 304 Stainless 4.5 x 105 $1.72 steel
, x 10, Boraflex ~.9 $2.31 i
V.1.5 Maximum Water Temperature in the Spent Fuel pool f i ! Scoping calculations have shown that the heat load will f increase in the spent fuel pool as a result of the increased storage capacity of the spenti4uel pool. Maximum temperature
- in the pool for worst-case heat loading conditions will be less
! than the design basis temperature of the spent fuel pool, with one cooling train in operation. Section III.1.5.(3) of this submittal provides additional information on the heat load of i the spent fuel pool. i f i l ) i I l C\ v l O V-5 i l
-_n_ , _ _ _
~V^"Y~ " ' - ~ ~ ' '
-~s TABLE V-1
( ) ESTIMATED SPENT FUEL DISCHARGE SCHEDULE FARLEY UNIT 1 Annual Discharge Schedule Cumulative Discharge Year (No. of Assemblies) (No. of Assemblies) 1979 52 52 1980 52 104 1981 32 136 1983 52 188 1984 52 240 1985 52 292 1986 52 344 1987 52 396 1988 52 448 1989 52 500 (1) 1990 52 552 1991 52 604 1992 52 656 , (2) 1993 52 708 1994 52 760 1995 52 812 T'N 1996 52 864 (_) 1997 52 916 1998 52 968 1999 52 1020 2000 52 1072 2001 52 1124 2002 52 1176 2003 52 1228 (3) 2004 52 1280 2005 52 1332 2006 52 1384 (4) 2007 52 1436 l
- 1. Existing storage capacity - loss of full-core reserve.
- 2. Existing storage capacity - filled.
- 3. Expanded storage capacity - loss of full-core reserve.
- 4. Expanded storage capacity - filled.
1
TABLE V-2 3
SUMMARY
OF ALTERNATIVE SPENT FUEL STORAGE FACILITIES Description Capital Cost $/KgU
- 1. 500 MTU Water Basin Storage $265,000,000 53 Facility (DOE /SR-0006 " Department of Energy Report on Fee for .
Spent Nuclear Fuel Storage and Disposal Services," October, 1980).
- 2. 10,000 MTU Open Field Dry'w 11 Storage Facility $477,000,000 48 (pp 5-24,25; WI:JBW:81-176
" Conceptual Design and Evaluation Study for an Interim Off-Site Spent Fuel Storage Installation," performed by Westinghouse Electric Corp.
for Edison Electric Institute, Tennessee Valley Authority, and U.S. Department of Energy; August, 1981).
- 3. 10,000 MTU Tunnel Drywell
/~h Storage Facility $1,313,900,000 131 kY (pp. 5-41, 42; Ibid.).
O
i l 2 V.2 Radiological Evaluation . (1) V.2.1 The present annual quantity of solid radioactive wastes generated by the spent fuel pool purification system is approximately 70 ft3 The spent fuel pool domineralizer resin is replaced once per year. The domineralizer resin is solidified and shipped in a 50 ft3 container. This container also contains the spent fuel pool cartridge filter which (p,also changed once per year. , No significant increase is expected in the wastes generated by the spent fuel pool cleanup system over those generated for less dense storage racks. Most of the crud is released soon i after fuel is removed from the reactor. Fuel stored for an additional number of years will not contribute significantly to crud loading in the water. Operating plant experience with
- ) high-density fuel storage has not indicated any noticeable i increase in the solid radioactive wastes generated by the increased fuel storage. However, since the rate of waste production in the future will be related to the number of fuel cladding defects, and since the opportunity for the storage of defective fuel assemblies will increase, it can be assumed that 1
the probability of waste production will increase. ? V.2.2 i ! Listed below are data regarding krypton-85 taken from the Unit 1 plant vent stack for the last 2 years. 4 1979 - 21.7 ci 1 1980 - 16.3 ci The above figures are taken from all sources of the plant. Due
- to plant operating procedures, it is estimated that over 99.9
- percent of the above krypton is released from the waste gas i
decay tanks during plant shutdown and startup. Operating plant experience (non-Farley) has not indicated any significant release of krypton-85 from spent fuel pools regardless of storage density. 2 V.2.3.(a) Table V-3 below shows the most recent isotopic analysis of spent fuel pool water. The sampline was taken on December 23, 1981 at 0740 hours. The counting began cn December 23, 1981 at 0749 hours with an elapsed time of 1.500E-01 hours. V-6 J 7 --7 -,-.,-y .,-y _v.,_--- .._.-___-m-.c,- ,,wm- y- . . . - , , -- . . _ - r._ . . .,., .--,-.
(s (_) Table V-3 Quantitative Analysis Standard Percent Percent Isotone guantity Deviation Rel_ Error of Total Co-57 -6.56E-06 pCi/M1 1.25E-06 pCi/M1 19.066 0.35 Ba-140 6.24E-05 pCi/M1 1.70E-05 pCi/M1 27.287 3.3 Cs-137 2.58E-05 pCi/M1. 4.15E-06 pCi/M1 16.060 1.39 Co-58 1.48E-03 pCi/M1 1.88E-05 pCi/M1 1.270 79.6 Mn-54 2.97E-05 pCi/M1 6.18E-06 pCi/M1 20.780 1.6 Co-60 2.53E-04 pCi/M1 1,.,Q7E-05 pCi/M1 5.405 13.6 V.2.3.(b) The external doses rates include 6 mr/hr at the surface of water over the center of the pool and 1.5 mr/hr for the general area over the pool. These dose rates increased to about 12 mr/hr over the center and about 5 mr/hr for the general area immediately after refueling. Experience has shown that these dose rates can increase as much as a factor of ten with the purification system out of service for approximately 10 days. V.2.3.(c) The only airborne radionuclide experience to date has been tritium. This is expected to change as fuel defects are encountered. Since the number of fuel assemblies which can be stored has increased, the expected number of fuel defects will also increase. V.2.3.(d) Operating plant experience with dense fuel storage has shown no noticeable increases in airborne radioactivity above the spent fuel pool or at the site boundary. V.2.3.(e) As stated in the response to subsection V.2.1 of this submittal and based on operating experience with dense fuel storage racks, there is no increase in the radwaste generated by the spent fuel pool clean up system. This is because operating experience has shown that with high density storage racks there is no significant increase in the radioactivity levels in the spent fuel pool water. Operating experience with high-density storage racks has shown no increase in_the annual man-rem due [D
\/ to the increased fuel storage including the changing of spent fuel pool cooling system resins and filters.
V-7 t__
V.2.3.(f) (~~ (_ Operating experience with dense storage of spent fuel has shown that there is no increase in dose rates anywhere above the pool due to increased storage capacity of the spent fuel pool. As stated in subsection V.2.1 of this submittal, most of the crud released to the pool and cleanup system is released soon after fuel is removed from the reactor,.not as a result of the increased storage capacity. V.2.3.(g) ,y i There is no access around the sides or underneath the spent fuel pool. The depth of the water above the fuel is sufficient so there will be no measurable increase in dose rates above due to radiation emitting directly from the fuel. As stated in subsections V.2.1 and V.2.3.(e) of this submittal, operating experience with high-density racks has shown no increase in the processing of radioactive waste or the man-rem associated with it. The response to subsection V.2.3.(f) of this submittal indicated that operating experience with dense racks has shown no increase in radioactivity levels in the water or dose rates above the pool. As discussed in subsections V.2.3.(c) and () V.2.3.(d), operating experience has shown no increased airborne radioactivity for high-density storage racks. Operating experience has also shown no increase in the man-rem due to the increased fuel storage with high-density racks. Therefore, no increase in the annual man-rem is expected at Farley as a result of the increased storage capacity of the spent fuel pool. V.2.4 The spent fuel storage rack modules that will be removed from the spent fuel pool wei h approximately 16,000 pounds each. The existing racks will be decontaminated, cut up, crated, and shipped offsite for burial as radioactive waste. V-8
V.3 Accident Evaluation V.3.1.(a) Protection against a cask drop is assured by the Seismic Category I, single failure-proof, outdoor spent fuel cask crane, by the single failure-proof lifting device, and by the interlocks and administrative controls described in the Farley FSAR subsection 9.1.4. A cask tip or drop into the spent fuel pool is also prevented by permanently installed rai(.ptops and mechanical bumpers which prevent cask crane travel over or into the vicinity of the spent fuel pool. The cask crane approach to the spent fuel pool, as shown in Farley FSAR figures 1.2-2 and 1.2-10, is limited to approximately 12 feet. Since the cask will not be handled in the vicinity of the spent fuel pool, the consequences of the cask drop are not affected by the increased storage capacity of the spent fuel pool. The rerack program will not alter the cask handling procedures described in Farley FSAR subsection 9.1.4.2.3. V.3.1.(b) I The spent fuel cask crane is not physically capable of traveling over or into the vicinity of the spent fuel pool. j Permanently installed rail stops and mechanical bumpers prevent
- the cask crane from traveling over the spent fuel pool. Since
! the cask will not be handled over or in the vicinity of the spent fuel pool, the consequences of the cask drop are not affected by the increased storage capacity of the spent fuel l pool. A complete cask crane component description, cask i handling description, and cask crane design evaluation are described in Farley FSAR subsections 9.1.4.2.2, 9.1.4.2.3, and' 9.1.4.3.2, respectively. V.3.2 The only crane capable of cask movement is the spent fuel cask crane. The spent fuel cask crane is a Seismic Category I, single failure-proof outdoor crane. A cask drop or tip into the spent fuel pool is prevented by-permanently installed rail stops and mechanical bumpers which prohibit cask travel over or 4 into the vicinity of the spent fuel pool. The accident aspects of review establish acceptability with respect to subsections V.3.1.(a) and V.3.1.(b) of this submittal. lO V-9 l
V.3.3.(1).(a) The accident aspects of review establish acceptability with respect to subsection V.3.1.(b). Also, the design evaluation of the spent fuel cask crane, as discussed in Farley FSAR paragraph 9.1.4.3.2, will not be affected as a result of the rerack program. V.3.3.(1).(b) See response to V.3.3.(1).(a). V.3.3.(2) See response to V.3.3.(1).(a). V.3.4 The cask drop /tip analysis is provided in subsection 3.1.(a). V.3.5 / ', The maximum weight of loads which may be transported over the
. s' spent fuel is not substantially in excess of a fuel assembly.
Earley Unit 1 Technical Specification Section 3.9.7.1 limits the size of the load that can be handled over spent fuel to 3,000 pounds. V.3.6 Since the spent fuel cask will not be handled over or into the vicinity of the spent fuel pool, the proposed modification does not significantly change or impact any previous determinations of Farley Unit 1 Safety Evaluation Reports.
- Since there will be a negligible change in radiological conditions due to the increased storage capacity of the spent fuel pool, no change is anticipated in the radiation protection program. In addition, the environmental consequences of a postulated fuel handling accident in the spent fuel pool, described in FSAR subsection 15.4.5, remain unchanged.
Therefore, there will be no change or impact to any previous determinations of Farley Unit 1 Final Environmental Statements. o
-s a
V-10}}