ML20059K587
| ML20059K587 | |
| Person / Time | |
|---|---|
| Site: | Limerick  |
| Issue date: | 01/14/1994 |
| From: | Hunger G PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20059K588 | List: |
| References | |
| NUDOCS 9402020171 | |
| Download: ML20059K587 (24) | |
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- LPECO ENERGYi ;==gre Wayne, PA 19087-5691 WB 10 CFR 50.90 '
s January 14,:1994 Docket Nos. 50-352 50-353 n License Nos. NPF-39 NPF-85
' U.S. Nuclear Regulatory Commission L
Attn: Document Control Desk Washington, DC 20555 m -
Subject:
Limerick Generating Station, Units 1 and 2
. Technical Specifications Change Request No. 93-19-0 c
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Gentlemsn:
y PECO Energy Company (PECo), formarly Philadelphia Electric Company, is submitting Technical Specfications (TC,' Change Request No. 93-19-0, in accordance with 10 CFR 50.90, reques.n.ig an amendment to theTS (i.e.,
Appendix A) of Facility Operating License Nos. NPF-39 and NPF-85 for Limerick -
gC Generating Station (LGS), Units 1 and 2, respectively. This propsed TS change 6 involves revising TS Sections 5.5.1.1, " Criticality,'_ and 5.5.3, " Capacity," to support implementation of a modification to install new hif, desnsity spent fuel i storage racks in each of the spent fuel pools at LGS. Installation of the new spent fuel storage racks will increase the' spent fuel storage capacity in each
. spent fuel pool from 2040 h,l assemblies to 4117 fuel assemblies. Information supporting this TS Chant,e Request is contained in Attachments 1 and 2 to this letter, and the proposer; replacemert pages for the LGS, Units 1 and 2, TS are contained in Attachment 3.
- LWe request.that, if approved, the amendments be issued and effective by June 15,1994, to facilitate implementation of the modifications at LGS, Units 1 and 2.
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January 14,1994 i Page 2 If you have any questions or require additional information, please do not
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Very truly yours,
. N. ,
G. A. Hunger, Jr.
Director Licensing Section Attachments f'
cc: T. T. Martin, Administrator, Region I, USNRC (w/ attachments)
N. S. Perry, USNRC Senior Resident inspector, LGS (w/ attachments)
W. P. Dornsife, Director, PA Bureau of Radiological Protection ,
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COMMONWEALTH OF PENNSYLVANIA
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-t D. R. Helwig, beinig first duly sworn, deposes and says: -
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That he is Vice President of PECO Energy Company (forrnerly Philadelphia Electric -
Company); the Applicant herein; that he has read the foregoing Technical Specifications -l Change Request No. 93-19-0 for Limerick Generating Station, Units 1 and 2, Facility _ ;
Operating License Nos. NPF-39 and NPF-85, to increase the spent fuel storage capacity l i
l in the spent fuel pools, and knows the contents thereof; and that the statements and matters set forth therein are true and correct to the best of his knowledge, information, j and belief. 1 h
Vice Presid -
i Subscribed and sworn to 1 i
' before me this/ day !
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b ATI'ACHMENT 1 LIMERICK GENERATING STATION t
UNITS 1 AND 2 Docket Nos. 50-352 <
50-353 License Nos. NPF-39 NPF-85 TECHNICAL SPECIFICATIONS CHANGE REQUEST No. 93-19-0
- Revise Technical Specifications to increase the Spent Fuel Storage Capacity in Each Spent Fuel Pool from 2040 Fuel Assemblies to 4117 Fuel Assemblies" i i
l Supporting Information for Changes - 20 Pages
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1 Attachment 1 Page 1 I
PECO Energy Company (PECo), formerly Philadelphia Electric Company, Licensee under l Facility Operating License Nos. NPF-39 and NPF-85 for Limerick Generating Station >
(LGS), Units 1 and 2, respectively, requests that the Technical Specifications (TS) contained in Appendix A to the Operating Ucenses be amended as proposed herein, to l revise TS Sections 5.5.1.1, " Criticality," and 5.5.3, " Capacity," to facilitate an increase in i the spent fuel pool storage capacity. The proposed TS changes are necessary to j support implementation of a modification to install new high density spent fuel storage racks in each of the spent fuel pools (SFPs) at LGS. Installation of the new spent fuel ,
storage racks will increase the SFP storage capacity in each spent fuel pool from 2040 l fuel assemblies to 4117 fuel assemblies. The proposed changes to the TS are indicated -l by a vertical bar in the margin of TS page 5-8. The TS pages identifying the proposed ,
changes are contained in Attachment 3.
We request that the NRC review the TS changes proposed herein and, if approved, issue ,
the amendments, effective upon issuance, by June 15,1994, to facilitate implementation of the modification.
t 7 :S Change Request provides a discussion and description of the prooosed TS !
cnanges, a safety assessment of the proposed TS changes, information supporting a ,
finding of No Significant Hazards Consideration, and information supporting an l Environmental Assessment. '
Discussion and Description of the Proposed Chances [
Currently, the Limerick Generating Station (LGS), Units 1 and 2, Technical Specifications (TS) limit the amount of spent fuel that can be stored in each of the spent fuel pools l (SFPs) to 2040 fuel assemblies. The proposed TS changes involve revising TS Sections 5.5.1.1, " Criticality," and 5.5.3, " Capacity," to support implementation of a modification to install new high density spent fuel storage racks in the SFPs to increase the spent fuel 'l storage capacity. Installation of this modification is designed to increase the spent fuel pool storage capacity in each SFP from 2040 fuel assemblies to 4117 fuel assemblies. ;
The new high density spent fuel storage racks were designed by Holtec International and utilize the detuned honeycomb technology developed by Holtec International. These racks are free-standing and self-supporting and have been utilized in over a dozen SFPs (e.g., the LaSalle Station, Units 1 and 2). Holtec International's Safety Analysis Report is i provided in Attachment 2 and documents the design and analysis performed to ;
demonstrate that the new spent fuel storage racks and intermediate configurations utilizing the existing racks satisfy all applicable regulatory requirements, codes, and !
standards. The Holtec Safety Analysis Report also considers the effects of a 24-month refueling cycle and a power rerate to 105% (i.e., increase in rated core thermal power from 3293 MWt to 3458 MWt) in the design of the new spent fuel storage racks.
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Attachment 1 l Page 2 Current plans are to install the new high density racks in the Unit 2 SFP first. This will involve moving some of the existing Unit 2 racks to the Unit 1 SFP. This will allow for ;
the maximum storage capacity in the Unit 2 SFP and an interim increase in the storage capacity in the Unit 1 SFP.- Both SFPs have been analyzed for installation of the new high density spent fuel storage racks, and the intent of this proposed TS change is to license both SFPs for the maximum storage capacity of 4117 fuel assemblies per SFP. Several intermediate SFP configurations will be utilized during the process of reracking both SFPs. l These intermediate configurations are bounded by the complete reracking of both SFPs with the new high density spent fuel storage racks as described in Holtec's Safety i Analysis Report (Attachment 2).
i
. Safety Assessment l The spent fuel storage facility at Limerick Generating Station (LGS), Units 1 and 2, 1 provides specially designed underwater storage space for the new and spent fuel ;
assemblies. The facility is located in the refueling area which is common for both units. i A description of the Spent Fuel Pool (SFP) design, existing fuel storage racks, and fuel pool cooling capability, and new fuel storage racks is provided below.
i SFP Desion !
The Spent Fuel Pools (SFPs) for LGS, Units 1 and 2, are elevated reinforced concrete structures with post-tensioned girders flanking the north and south ;
extremities of the 72-inch thick reinforced concrete slab. Each of the two (2) LGS i SFPs are currently licensed to store no more than 2040 fuel assemblies. The SFPs :
have a volume of approximateiy 46,000 ft and are filled with demineralized water to a normal depth of 38 feet 3 inches (38'-3"). This provides approximately 23 feet !
of water above the tops of the stored fuel assemblies.
The SFPs are lined with stainless steel plate to minimize leakage and reduce corrosion product formation. A leakage collection system is provided to permit l expedient detection of leaks through the stainless steel liner plate and to prevent '
the uncontrolled loss of pool water to areas below the pool. Drainage paths are formed in the floor slab below the floor liner, and are designed to permit free gravity flow. The design of the drainage system is described in Section 9.1 of the l LGS Updated Final Safety Analysis Report-(UFSAR). Pool leakage is routed !
through a piping system, provided at the base of the pool wall, via one (1) of three (3) dirty radwaste funnels. Leakage from each of seven (7) segments of the leak collection system is routed through separate piping to enable identification of the area of the liner that is leaking.
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. ' Page 3 Leakage is detected by observation of water flowing out of the piping into the dirty, .
radwaste funnel or by low level indication in the SFP skimmer surge tank or the SFP itself. Flow into the funnels is observed during periodic operator inspections.
Skimmer surge tank low level alarms and trips are also provided as described in Section 9.1.3.5 of the LGS UFSAR. -
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To ensure that the SFP water levelis not lowered by a malfunction of the Fuel Pool Cooling and Cleanup (FPCC) system, the system takes suction from the pool near :
the normal water level via the skimmer surge tanks. The system return lines enter the SFP from above the normal water level and are provided with siphon breaker -
holes near the normal water level to preclude the possibility of siphoning the pool. ,
The SFP structures are designed in accordance with seismic Category I requirements as specified in Section 3.2 of the LGS UFSAR. The components and !
supporting structures of any system, equipment, or structure that is not seismic I Category I and whose collapse could result in loss of a required function of the ,
spent fuel storage facility are analytically checked to determine that they will not !
collapse when subjected to seismic loading resulting from the Safe Shutdown i Earthquake (SSE) (i.e., seismic Category llA).
t Liner leakage detection system piping, the FPCC' system piping, and the wave i suppression scupper piping are all seismic Category llA. The only other piping 3 attached to or in the SFP is from the Residual Heat Removal (RHR) and i Emergency Service Water (ESW) systems which provide a backup source of water -
for SFP cooling and makeup. This piping is seismic Category 1. .
Loss of any of the seismic Category llA piping would not affect the ability to :
maintain spent fuel cooling or to maintain adequate submergence of the fuel. _
Accidental dropping of movable heavy objects into the SFP is precluded by the use :
of administrative procedures, electrical interlocks to limit the load travel over the l spent fuel pool, and the use of guardrails and curbs around the pools and the .
reactor wells to prevent fuel handling and servicing equipment from falling into the pools. The electrical interlocks and administrative procedures are described in Section 9.1.4 of the LGS UFSAR. In addition, heavy load handling in the vicinity of the SFPs is accomplished in accordance with the guidance delineated in .
NUREG-0612, " Control of Heavy Loads at Nuclear Power Plants," such that the likelihood of a heavy load drop is precluded. 3 Existina Fuel Storage Racks Currently, each SFP at LGS can contain up to 23 high density spent fuel storage I racks. The maximum analyzed fuel storage capacity of each SFP is 2862 fuel assemblies with a presently licensed capacity of 2040 fuel assemblies. The spent fuel storage racks are modular, freestanding, top entry racks designed to maintain
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J Attachment 1 Page 4 ;
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the spent and new fuel in a space geometry whereby each fuel assembly has a - ,
neutron poisoning material between it and any adjoining fuel at semblies. . This ;
precludes the possibility of criticality under normal and abnormal conoWons. The -
only point of contact between the spent fuel rack and the SFP structure is with the l bottom liner plate. The existing spent fuel rack moduies consist of six (6) basic .l structural components: top grid casting, bottom grid casing, poison cans, side -
plates, corner angle clips, and adjustable foot assemblies. The top and bottom.
cast aluminum grids sandwich the square cross-section poison cans into pockets ';
in a checkerboard arrangement. The design of the existing SFP storage racks are described in Section 9.1 of the LGS UFSAR. The grids are held in place by.
aluminum side plates and corner angles bolted and riveted with aluminum bolts o and rivets. The rack modules are individually leveled with adjustable foot 'l assemblies at the four (4) corners of the bottom grid. The adjustable foot ,
assemblies consist of a 304 stainless steel bearing plate, a volumetrically captured :
1/4-inch thick ABS plastic insulator, and an aluminum threaded section for height ,
4 adjustment. The insulator provides protection from galvanic corrosion between the -
stainless and aluminum surfaces. All aluminum components are anodized individually.
There are three (3) sizes of rack modules in use at LGS (i.e.,10 feet x 11 feet,10 feet x 12 feet, and 11 feet x 12 feet). TM 10x11 modules have 55 poison cans, -
the 10x12 modules have 60 poison cms, and the 11x12 modules have 66 poison ;
cans. '
The poison cans consist of two (2) concentric square aluminum tubes, with four j '
(4) plates of Boral (i.e., Boron carbide in an aluminum composite matrix) in the annular gaps. The Boral is so positioned that it overlaps the fuel pellet stack j length in the fuel assemblies by one (1) inch at the top and bottom. The outer ,
concentric tube is folded into the inner tube at both ends and totally seal-welded. ,
Each poison can is pressure and vacuum leak tested and then plug-welded to isolate the Boral from the pool water. The poison cans are then anodized. The poison cans are not vented.
The top and bottom grid castings hold the fuel assemblies in a vertical position.
The weight of the assemblies is supported by the lower grid casting and it, in turn, -
is supported by the four (4) adjustable foot assemblies that allow adjustment for variations in SFP floor level. To maintain a flat, uniform contact area, the leveling screw bearing pads are free to pivot. Each hole in a casting has adequate clearance for inserting or withdrawing a fuel assembly, either channeled or unchanneled. Sufficient guidance is provided to preclude damage to the fuel assemblies. The nominal center-to-center spacing between fuel assemblies in a module is 6.625 inches. The nominal center-to-center spacing between fuel assemblies in adjacent modules is 9.375 inches. -
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- Attachment 1 !
Page 5 The spent fuel storage racks are installed in the SFPs in such a manner as to ;
ensure that there is a Boral plate between each adjoining fuel storage position. ,
Each storage module is level with each other module at the top. There are 7.25 1 adequa earance for cooing wa e t en e e c fu l ce a d 51r gh n tura -
convection, keep each fuel assembly cool.
q The rack materials have no significant degradation due to the total radiation doses )
expected in the SFP over the design life. The racks are designed to withstand various loading conditions such as dead and live loads; ioads experienced by a jammed fuel assembly or dropped fuel assembly; and loads experienced during :
seismic events (e.g., Operating Basis Earthquake). .
Fuel Pool Cooling and Cleanuo The Fuel Pool Cooling and Cleanup (FPCC) system is designed to remove the decay heat generated by the spent fuel assemblies stored in the SFP and to maintain the pool water at a clarity and purity suitable .both for underwater operations and for the protection of personnel in the refueling area. There is a' FPCC system for each SFP. The FPCC system consists primarily of the pool water collection equipment, including wave suppression scupper and skimmer surge tanks, a cooling train with two (2) heat exchangers, two pumps, a cleanup loop,. ;
and the discharge diffusers in the SFP. A backup heat exchanger and a backup pump are also included in the system. The FPCC system has no function related ;
to the safe shutdown of the plant.
The FPCC system piping is designed so that operator error.or a loss of piping i integrity cannot result in the' draining of the SFP so that stored fuel would be i uncovered, and provides a source of makeup water to ensure the maintenance of - ,
the SFP waterlevel. All piping and components of the FPCC system that form part. ,
of the flow path for makeup water from the Emergency' Service Water (ESW) -
system, Residual Heat Removal Service Water (RHRSW) system, and the cooling ,
water to and from the Residual Heat Removal (RHR) system are c'esigned to remain functional following a Safe Shutdown Earthquake (SSE) event. The FPCC-system is designed to maintain the bulk water temperature in the SFP at or below t 140 F under normal opertha conditions, with a normal decay heat load .of 7
1.632x10 Btu /hr, with two e ; 400 pumps and two (2) FPCC heat exchangers-in operation. This is based on th: tormal heat load discharge history as described 3 in Section 9.1 o 1 s LGS UFSAR. The FPCC system is designed l
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.I to permit the RHR system to be used, through a cross-tie, to maintain the bulk. I water temperature in the spent fuel pool at or below 140 F, with a maximum i 7
anticipated decay heat load of 3.64x10 Btu /hr. This is based on one (1) full core . l unloaded from the reactor ten (10) days after shutdown to fill the pool, plus the. :
previous normal refueling loads from 18-month refuelings as described in Section -
9.1 of the LGS UFSAR. However, the analysis documented in Holtec's Safety. .
Analysis Report considers the heat load generated from a 24-month refueling. '
cycle.
Water from the SFPs flows through weirs and a wave suppression scupper at the $
pool surface into two (2) skimmer surge tanks adjacent to the pool. Water in the a skimmer surge tanks flows by gravity through the fuel pool heat exchangers to the ,
suctions of the fuel pool cooling pumps. From the pumps, water is returned to the SFP through two (2) diffusers located at the bottom of the pool. A portion of the ,
discharge flow from the pumps can be diverted through the cleanup loop before ' ,
being returned to the pool. Heat is removed from the fuel pool heat exchangers :
by the Service Water (SW) system. [
During normal plant operation, the FPCC system serves only the SFP. During refueling operations, however, when the reactor well, dryer / separator pool, and/or .;
cask loading pit are filled, with water, the FPCC system _can be aligned to l recirculate and process the water in all these cavities.' . Water from the refueling-1 water storage tank is used to fill the refueling area cavities. The refueling water j pumps fill the cask loading pit through its drain line and fill the reactor well and the ;
dryer / separator pool through diffusers in the reactor well. After refueling activities j are completed, the refueling water pumps transfer water from the refueling area '
cavities back to the refueling water storage tank via a condensate filter /demineralizer if additional-cleanup is required. Gravity draining of the- ;
refueling water directly to the refueling water storage tank is also possible. ;
As the heat load in the SFP changes, the number of operating fuel pool cooling :
pumps and heat exchangers is adjusted to maintain the desired water temperature.
The FPCC system has sufficient cooling capacity to maintain the SFP water at a temperature at or below 140 F, with a normal decay heat load of 1.632x107 Btu /hr, t with two (2) pumps and two (2) heat exchangers operating. !
If an abnormally large heat load is placed in the SFP, a cooling train of the RHR ;
system, consisting of an RHR pump and heat exchanger, can be substituted for the FPCC pumps and heat exchangers for cooling the SFP. water. . A cross-connection between the drain line from the skimmer surge tanks and the RHR-j system allows one (1) RHR pump to take suction from the skimmer surge tanks and pump SFP water through an RHR heat exchanger before returning it to the SFP via diffusers at the bottom of the SFP provided specifically for use with the i
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i Attachment 1--
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i RHP. system. The interconnecting piping between the RHR system and FPCC system is accomplished by.use of a spool piece (i.e., one (1) blind flange for normal operations or one (1) open spool for wheri the intertie is required).
Administrative controls prevent the use of the RHR system intertie unless the l associated reactor is shut down and is in the refueling mode. -
t The RHR system alone is capable of cooling the SFP water under the conditions when a full core of irradiated fuelis offloaded into the SFP. The RHR system has :
sufficient heat removal capacity to maintain the SFP water at a temperature at or :
below 140 F, with a maximum anticipated decay heat load of 3.64x107 Btu /hr. The j RHR system may also be used for cooling if the FPCC system should be unavailable.
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if normal fuel pool cooling should be lost as a result of a pipe break in the seismic -
Category llA portion of the system, the quantity of water released would be limited to the inventory in the SFP above the overflow weirs, the skimmer surge tanks, and ,
the pump suction piping. The flood height and environmental conditions resulting from this break would not prevent personnel from making the necessary RHR system crosstie connection which requires manual action to' establish. The maximum temperature (i.e.,150 F) and pressure (i.e.,31 psig) of the water in the line are not high enough to significantly affect the temperature, pressure, or . '
humidity conditions in the area where the crosstie is 'made. The released fluid would not be highly radioactive. The maximum flood height in the area resulting from this break is conservatively calculated to be about one (1) foot. However, if the floor drains in the area are functioning, the flood water height would be much lower, and the water would drain out of the room at approximately the same rate as it flowed in from the break. .l
.. If there is a Loss of Offsite Power (LOOP), the Class IE buses are powered by the l L emergency diesel generators (EDGs), and the'two (2) FPCC pumps that receive' ;
Class 1E power can be restarted. Since normal SW is not available in this case, the FPCC heat exchangers can be cooled by the Reactor Enclosure Cooling Water '*
(RECW) system, which is cooled by the ESW system, by interconnecting piping, after installation of normally removed spool pieces. However,.other cooling :
L methods would also be available as described below.
If there is a complete loss of capability to remove heat from the SFP using heat' exchangers, heat can be removed by allowing the pool to boil and adding makeup
- l. water to maintain the SFP water level. Makeup water is normally supplied to the j- skimmer surge tanks from the demineralizer water makeup system by manipulating i a remote manually operated valve. If makeup water from this source is not available, makeup can be provided from the ultimale heat sink (i.e., Spray Pond) by one (1) of two (2) flow paths. The first of these backup makeup sources is a ,
L loop of the ESW system via a cross-connecting line to one (1) of the RHR system I
diffusers in the SFP. The two (2) ESW pumps in the ESW loop provide ,
I Attachment 1 ,
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redundancy in motive power for this source of makeup supply. The manual valves that must be opened to initiate makeup from the ESW system are located in the ;
control structure and are accessible after an accident th,at would render the reactor .;
and refueling secondary containments inaccessible ~ The second of these backup _
. i makeup sources is a loop of the RHRSW system via the piping of one (1) RHR ;
system loop and the cross-connecting piping leading to the RHR diffusers in the SFP. The two (2) RHRSW pumps in the RHRSW loop provide redundancy in motive power for this source of makeup supply. These backup sources of makeup water provide substantial flow rates to ensure adequate makeup capability. The Spray Pond is designed with sufficient water volume in order to provide a source of makeup water for the SFP for 30 days, without makeup to the pond during which time the cooling function of the FPCC system or RHR system can be established or an alternate makeup water supply can be established. ,
As described above, the SFP is provided with redundant seismic Category 1-makeup capability to ensure an adequate supply of makeup water to the SFP under conditions of maximum anticipated evaporation associated with fuel pool :
boiling. The radiological consequences of a boiling SFP are discussed in Section ,
9.1.3.6 of the LGS UFSAR. Makeup water to the SFP is supplied from the Spray Pond using either the ESW system or RHRSW system. Redundant pumps, capable of being powered by the associated EDGs, in each loop of the ESW and {
RHRSW systems provide assurance of the availability of motive power for pumping the makeup water.
New Soent Fuel Storaoe Racks i
The proposed Technical Specifications (TS) changes involve revising the TS to-support implementation of a modification to install new high density spent fuel -
storage racks in each of the SFPs at LGS. -Installation of the new spent fuel ,
storage racks will increase the spent fuel storage capacity in each SFP from 2040 j fuel assemblies to 4117 fuel assemblies. Installation of the new spent fuel storage -;
racks will extend the " full core reserve" capability from the year 1998 to 2013. The increase in storage capacity in each SFP that results from a reduction in the rack layout pitch would be rather modest, since the pitch in the new rack modules is only slightly smaller than that in the existing ones (i.e., 6.244 inches vs. 6.625 inches). Rather, the increase in the capacity is mainly derived from' the more efficient utilization of SFP floor space. This will permit equipping the LGS SFPs with a maximum density storage configuration consistent with the current industry practice.
The new spent fuel storage racks were designed by Holtec International and utilize the detuned honeycomb technology which was developed by Holtec. The new 6 spent fuel storage racks are free-standing and self-supporting. The principal construction materials for the new racks are American Society of' Mechanical Engineers (ASME) SA-240-Type 304L stainless steel sheet and plate stock, and SA
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Attachment 1- !
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-i 564 (i.e., precipitation hardened stainless steel) for the adjustable support spindles. i The only non-stainless steel material in the rack is the neutron absorber material, -i which is a boron carbide aluminum cermet (i.e., Boral*). l The new racks are designed and analyzed in accordance with ASME Boiler and l Pressure Vessel (B&PV) Code, Section 111, Division 1, Subsection NF. The material procurement and fabrication of the rack modules meet the applicable requirements of 10 CFR 50, Appendix B. The test coupon program identified by the :
manufacturer of the new high density spent fuel storage racks (i.e., Holtec International) is slightly different than the existing program described in Section' ;
9.1.2.4 of the LGS UFSAR. The test coupon program for the new racks incorporates neutron attenuation type testing. :
f Attachment 2 of this letter provides Holtec International's Safety Analysis Report ,
I which documents the design and analysis performed to demonstrate that the new high density spent fuel storage racks satisfy all applicable requirements, codes, and standards, in particular, the NRC's "OT Position for Review and Acceptance ,
of Spent Fuel Storage and Handling Applications (1978)." In addition; Holtec's l Safety Analysis Report demonstrates that under the analyzed conditions the existing racks can be used to facilitate reracking operations of SFPs. .
Holtec International's Safety Analysis Report includes a discussion regarding the ;
thermal-hydraulic, criticality, and structural adequacy of the new racks. Thermal-hydraulic adequacy requires that fuel cladding will not fait due to excessive thermal stress, and that the steady state pool bulk temperature will remain within the limits -
prescribed for the SFP. Demonstration of structural adequacy primarily involves analysis showing the stability of free-standing rack modules under the postulated '
seismic events, and that the primary stresses in the module structure will remain below the ASME Code allowables. Similar analyses for the existing racks are also _ :
included in the Safety Analysis Report. The structural qualification includes analytical demonstration that the subcriticality of the stored fuel will be maintained -
under accident scenarios such as fuel assembly drop, and accidental ,
misplacement of the fuel outside a rack. ;
The criticality analysis presented in Holtec's Safety Analysis Report demonstrates that the neutron multiplication factor (Kg ) is less than 0.95 with the racks fully loaded, with fuel of the highest anticipated reactivity, and the SFP flooded with pure water at a temperature corresponding the highest reactivity. The Safety Analysis l Report also considers the effects of a 24-month refueling cycle and power rerate (i.e., increase in rated core thermal power from 3293 MWt to 3458 MWt). j i
Our current plans are to install the new high density racks in the Unit 2 SFP first. !
This will involve moving some of the existing Unit 2 racks to the Unit 1 SFP. This '
l will allow for the maximum storage capacity in the Unit 2 SFP and an interim increase in the storage capacity for the Unit 1 SFP. Both SFPs have been
ee q ttachment 1 ~ !
m Page 10 l analyzed for installation of the new maximum high density spent fuel storage racks - ;
and the intent is to license both the LGS SFPs for a', maximum spent fuel pool storage capacity of 4117 fuel assemblies per SFP. Several intermediate SFP ;
configurations will be utilized during the process of reracking both SFPs. 'These - {
intermediate configurations are bounded by the analysis provided in Holtec's !
Safety Analysis Report (Attachment 2). ;;
The existing racks in the Unit 1 and Unit 2 SFPs are identical. Therefore, placing . i existing Unit 2 rack assemblies in the Unit 1 SFP does not introduce any new ;
criticality or radiological concerns beyond those previously evaluated. However, j the addition of the existing Unit 2 rack assemblies in the Unit 1 SFP does alter the SFP thermal-hydraulic and multi-body fluid coupling effects in the Unit 1 SFP.
These two (2) considerations are evaluated further in Holtec's Safety Analysis ;
Report. f j
To facilitate installation of the new high density spent fuel rack in the Unit 2 SFP, the Unit 2 SFP cooling RHR piping, which provides a backup cooling source for i the FPCC system will be modified to minimize the number of storage cells which may encounter accessibility difficulties. This piping modification will consist of removing the majority of RHR piping inside the SFP and installing an elbow to t maintain proper cooling water flow in the SFP. Also, the storage location of the SFP gates will be relocated from the North and South walls to the East wall to avoid interference with the new racks. Unit 1 SFP work to support installation of the new high density spent fuel pool racks will be performed at a later dated and will be addressed in a separate evaluation. This piping modification will not prevent -
the RHR system from providing a backup source of cooling water for the SFP.
The installation of the new maximum density racks in the Unit 2 SFP and placement of additional racks in the Unit 1 SFP will not interfere with the ability of the FPCC systems from adequately cooling their respective SFPs. Increasing the spent fuel storage capacity will result in a small increase in the maximum normal i decay heat load from 16.32 x 108 Btu /hr to 18.05 x108 Btu /hr. This increase can be attributed to 1) the maximum storage capacity of 4117 fuel assemblies,2) a 5%
power rerate consideration (i.e., increase in rated core thermal power from 3293 MWt to 3458 MWt),3) a reduction in the minimum in-core decay time prior to fuel ,
movements, and 4) the refueling cycle has been increased from 18-months to 24-months. :
Section 9.1.3, " Spent Fuel Pool Cooling and Cleanup System," of NUREG-0800, ,
" Standard Review Plan for the Review of Safety Analysis, Reports for Nuclear Power Plants," recommends that the bulk SFP temperature be maintained at or below ;
140 F. Holtec's Safety Analysis Report indicates that with the increase in spent fuel storage capacity the temperature that two (2) trains of fuel pool cooling can :
maintain SFP temperature will increase to 143 F. The time period that two (2) -
trains of fuel pooling cooling can not maintain the SFP temperature b*s 140 F
l Attachment 1 ;
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is 2.5 days and SFP temperature will exceed 140 F approximately 160 hrs after I' plant shutdown. This slight increase (i.e.,140' F to 143' F) in SFP temperature is -
considered acceptable since it is for a short duration (i.e., 2.5 days) and the fuel pool cooling mode of RHR will be available for additional SFP cooling if necessary.
The maximum decay heat load, assuming full core discharge and remaining cells ;
filled, will increase from 36.4x108 Btu /hr to 37.6 x108 Btu /hr; however; the RHR system is still be capable of maintaining SFP temperature less 140 F as described in Section 9.1.3.2.3 of the LGS UFSAR and Holtec's Safety Analysis Report.
The attached Safety Analysis Report demonstrates that the new high density spent .
fuel storage racks possess wide margins of safety in the areas of thermal--
hydraulic, criticality, structural, and radiological considerations. In addition, the Safety Analysis Report provides an evaluation of heavy load considerations for the proposed reracking operations. If the proposed TS changes are approved the analysis provided by Holtec will supersede the analysis currently provided in the >
applicable sections of LGS UFSAR.
Information Supportina a Findina of No Slanificant Hazards Consideration -
We have concluded that the proposed changes to the Limericts Generating System (LGS),-
Units 1 and 2, Technical Specifications (TS) to increase the spent fuel storage capacity in each spent fuel pool (SFP) from 2040 fuel assemblies to 4117 fuel assemblies do not involve a Significant Hazards Consideration. In support of this deterrnination, an.
evaluation of each of the three (3) standards set forth in 10 CFR 50.92 is provided below.
- 1. The orocosed Technical Soecifications (TS) changes do not involve a significant increase in the orobability or consecuences of an accident nt_eviously evaluated.
increasing the spent fuel storage capacity in each Spent Fuel Pool (SFP) to 4117 fuel assemblies does not increase the probability of occurrence of an accid _ent. Since all fuel handling activities will be performed using approved procedures and compatible equipment, the probability of a fuel handling accident occurring is unchanged.
The intermediate configuration involving the installation of the new maximum density racks in the Unit 2 SFP and placement of additional existing racks in the Unit 1 SFP will not prevent the ability of the Fuel Pool Cooling and Cleanup (FPCC) systems from adequately cool lng their respective SFP.
The backup cooling and makeup systems (i.e., Residual Heat Removal (RHR), Emergency Service Water (ESW), and Residual Heat Removal Service Water (RHRSW) systems) will continue to function as designed to provide an alternate source of cooling and makeup water to ensure SFP cooling is maintained. Increasing the spent fuel storage capacity in each
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5 Attachment 1 Page 12 -
r SFP will result in a slight increase in the maximum normal decay heat load 8 8 from 16.32 x 10 Btu /hr to 18.05 x 10 Btu /hr. This increase is due to 1) the heat load associated with a maximum storage capacity of 4117 fuel .i assemblies, 2) a 5% power rerate consideration (i.e., the effects of f increasing the rated core thermal power from 3293 MWt to 3458 MWt),3) l a reduction in the minimum decay time until fuel movements begin, and 4) the effects of increasing our refueling cycles from 18-months to 24-months. :
Section 9.1.3, " Spent Fuel Pool Cooling and Cleanup," of NUREG-0800, t
" Standard Review Plan for the Review of Safety Analysis Reports for Nuclear i Power Plants," recommends that the SFP temperature be maintained at or below 140 F. However, due to the increase in the maximum normal decay ;
heat load, and with two (2) trains of fuel pool cooling operating, the temperature that the SFP can be maintained will increase from 140 F to ;
143 F. The time period that two (2) trains of fuel pool cooling can not ;
maintain the pool temperature below 140 F is 2.5 days and the SFP '
temperature will exceed 140 F approximately 160 hours0.00185 days <br />0.0444 hours <br />2.645503e-4 weeks <br />6.088e-5 months <br /> after plant shutdown. This slight increase in SFP temperature (i.e.,140"F to 143*F) is considered acceptable since the increase is small (i.e., 3 F), and the duration in which the temperature exceeds 140 F is short (i.e., 2.5 days).
In addition, during this period the RHR system will be available for operation to maintain the desired SFP temperature. The maximum decay heat load, y assuming full core discharge and remaining cells filled, will increase from 36.4 x10 Btu /hr to 37.6 x 10 8Btu /hr; however, the RHR system is still be capable of maintaining SFP temperature less than 140 F as described in LGS Updated Final Safety Analysis Report (UFSAR) and supporting Safety l Analysis Report provided in Attachment 2. This increase in temperature will I not increase the probability of a loss of fuel pool cooling accident or adversely affect the Refuel Floor ventilation system.
The proposed piping modifications to the RHR system piping inside the Unit 2 SFP will not interfere with the RHR system's ability to adequately cool the i SFP or to prevent siphoning of the SFP water.
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Movement of the Unit 2 SFP gates to the new storage location and installation of the new fuel storage racks will be accomplished in-accordance with the guidance specified in NUREG-0612, " Control of Heavy j Loads at Nuclear Power Plants." Approved procedures, safe load p'aths, :
and single failure proof rigging will be used. Therefore, the probability of a heavy load drop is unchanged.
The consequences of a Fuel Handling Acciden't as described in the LGS l UFSAR is not increased since the number of fuel assemblies stored in a i SFP is not an input to the initial conditions of the accident evaluation. This i accident evaluates the dropping of a spent fuel assembly and the fuel i grapple assembly into the reactor core during refueling operations. A drop I
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Attachment 1 Page 13-height of 32 feet for the spent fuel assembly and 47 feet for the fuel grapple assembly are assumed and will produce the. largest number of failed fuel rods. The tops of the new spent fuel racks are at the same level as the existing spent fuel racks. Since the maximum possible height a fuel assembly can be dropped over the SFP.does not exceed 32 feet, the -
consequences of a Fuel Handling Accident will not be increased by increasing the number of fuel storage cells. The increase in dose estimates '
presented in the Safety Analysis Report are within 10 CFR 100 limits and are ,
the result of increased fuel enrichment for power rerate and 24-month -
refueling cycles, and not as a result of an increase in the number of fuel storage cells. These other changes are the subject of separate TS Change Requests that have already been submitted to the NRC for approval.
The consequences of a loss of fuel pool cooling as described in Section 9.1.3.6 of the LGS UFSAR will not be increased. The event described in the UFSAR assumes that the iodine in the fuel from past refuelings is negligible, due to the long decay time. lodine is the major contributor to thyroid dose. ,
Since the iodine in the fuel from past refuelings is negligible, due to the long l decay time, increasing the number of fuel storage cells will not increase the ,
dose due to the release of iodine in the SFP water resulting from boiling and therefore, the consequences are not increased. The time to boil of 13.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> currently specified in UFSAR bounds the time to boil of 9.15 hrs presented in the supporting Safety Analysis Report since the 13.5 hrs is for 21 days after reactor shutdown and the 9.15 hrs'is for 7.25 days after reactor shutdown, and the decay heat from the newly discharged fuel decreases exponentially with time after plant shutdown.
The new maximum density storage racks have been designed and analyzed to maintain K,n 50.95. The supporting Safety Analysis Report includes the effects of various anomalies such as a fuel assembly drop event,.
manufacturing tolerance variations, and abnormal location of a fuel assembly. Since a K,3 of 50.95 with a confidence factor of 95% is maintained, the consequences of an event that would affect criticality control will not increase. The planned interim configuration of the Unit 1 poo! is i bounded by the current analyses in the UFSAR, since the rack design is unchanged.
The new maximum density storage racks have been designed and analyzed to seismic Category I criteria and are capable of remaining functional during the event of a fuel assembly and fuel grapple assembly impacting the rack l from a height of 36 inches, as described in the attached Safety Analysis Report. Since the new maximum density storage racks are capable of ~ ,
. withstanding an impact from a height of 36 inches, the consequences of the l events described in the LGS UFSAR which use a drop height of 16 inches, j are not increased. '
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- Attachment 1 t Page 14 Increasing the on-site storage capacity by installing additional storage ce::s will not increase the probability of a malfunction of the stored spent fuel based on the thermal-hydraulic ana!ysis presented in the supporting Safety Analysis Report which concludes that sufficient cooling exists with 4117 fuel assemblies in a SFP. As for fuel criticality, this determination is based on-the criticality analysis documented in the supporting Safety Analysis' Report ,
which confirms that the stored fuel assemblies will remain sub-critical under .
normal and abnormal conditions. ,
increasing the on-site storage capacity by installing additional storage cells will not increase the probability of a malfunction of the SFP liner based upon the SFP structural analysis as documented in the supporting Safety Analysis Report which indicates that adequate margin exists to prevent overstressing of the SFP liner, increasing the on-site storage capacity by installing addition storage cells will not increase the probability of a malfunction of the SFP structure. This is based upon the SFP structural analysis as documented in the supporting _ .
Safety Analysis Report which confirms that the SFP structure still has adequate margin to prevent overstressing and meets the code requirements for the LGS. .
Increasing the on-site storage capacity by installing additional storage cells will not increase the probability of a malfunction of the spent fuel storage racks based on the seismic / structural analysis documented in the supporting Safety Analysis Report which concludes that interaction of racks during a seismic event will not result in loss of the spent fuel storage racks' ability to function. The planned relocating the storage location of the SFP gates will not increase the probability of a malfunction of the SFP gates since, while being stored, the SFP gates do not perform a safety function.
The hangers used to secure the SFP gates will be designed / installed to the same requirements as the existing hangers, increasing the on-site spent fuel storage capacity will not increase the probability of a malfunction of the Fuel Pool Cooling and Cleanup (FPCC).
system. The only impact on the FPCC system of increasing the spent fuel storage capacity will be a slight increase in fluid temperature (i.e.,140 F to 143 F) which is within the design temperature of the system (i.e.150 F) as described in the LGS UFSAR.
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Attachment 1 '
Page -15 Modifying the RHR piping in the Unit 2 SFP such that it will not interfere with -
increased fuel storage will not increase the probability of a malfunction of '
the RHR system since the RHR. system's ability to cool the.SFP and to ;!
prevent siphoning of the SFP water will remain unchanged. Only the RHR ;
discharge piping inside the SFP will be modified. The proper flow pattern will be maintained and net positive suction head requirements will be - :
unaffected.
The probability of a malfunction of fuel handling equipment will not be !
increased since increasing the on-site storage capacity does not affect fuel "
handling equipment.
Increasing the on-site spent fuel storage capacity does not iricrease the !
consequences of a spent fuel assembly failure since the failure of one assembly will not result in additional spent fuel assembly failures.
Increasing the on-site spent fuel storage capacity does not increase the ;
consequences of a loss of fuel pool cooling as described in Section 9.1.3.6 l of the LGS UFSAR which evaluated the radiological affects due to thyroid i dose. lodine is the major contributor to thyroid dose. The iodine in the fuel i from past refuelings is negligible, due to the long decay time. Since the j release of iodine resulting from the SFP water boliing is entirely due to the 'l freshly discharged fuel, the consequences of reracking _the SFPs are unchanged from that previously evaluated. The evaporation rate will i increase due to higher decay heat load. However, since the time to boil is 9.15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br />, as discussed previously, adequate time exists to align the i alternate makeup water sources (e.g., RHR, Emergency Service Water - !
(ESW), and Residual Heat Removal Service. Water (RHRSW) systems) to. ;
maintain SFP water level and therefore, the : consequences are not' ,
increased.
jl Increasing the on-site storage capacity will not increase the consequences -
of spent fuel storage rack failure, since both the new maximum density I racks and the existing racks have been designed / qualified to limit the ,
consequences of a failure. A failure of or damage to one (1) storage rack will not result in failure or damage to another storage rack. '
increasing the on-site storage capacity will not increase the consequences _ -
of a failure of the SFP gates or SFP liner since the design of the SFP to ;
maintain adequate water level and the available makeup capacity are i unaffected. l J
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/ - Attachment 1 Page 16 Increasing the on-site storage capacity will not increase the consequences ~ .,
of the failure of fuel handling equipment since the maximum expected "
number of fuel rods damaged by a fuel handling equipment failure remains as evaluated in the LGS UFSAR. 1 Therefore, the proposed TS changes do not ' involve an increase in the
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probability or consequences of an accident previously evaluated.
- 2. The orocosed TS chances do not create the oossibility of a new or different kind of accident from any accident oreviously evaluated.
Increasing the spent fuel storage capacity in each of the SFPs at LGS to a maxirnum of 4117 fuel assemblies-as analyzed in the attached Safety Analysis Report will not create the possibility of an accident of a different type. The SFP configurations have been analyzed for reactivity / criticality effects, thermal / seismic-structural effects, radiological effects, and thermal-hydraulic effects. Since the increase in storage capacity is achieved by the installation of additional storage racks which are passive components, the possibility of creating a new accident does not exist. .
No new operating schemes or active equipment types will be required to -!
store additional fuel bundles in the SFP. Therefore, the possibility of a
- different type of malfunction occurring is not created. -
Therefore, the proposed TS changes do not create possibility of a new or -
different kind of accident from any previously evaluated.
- 3. The crocosed TS changes do not involve a significant reduction in a margin ,
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of safety.
Since the existing TS limits for fuel handling interlocks, heavy loads restrictions, water coverage over irradiated fuel, and fuel sub-criticality will be maintained, the margin of safety will not be reduced.
Therefore, the proposed TS changes do not involve a reduction in a margin of safety.
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Attachment 1 1 Page 17 i
information Supporting an Environmental Assessment l The proposed TS changes have been evaluated against the criteria in 10 CFR 51.21 for identification of licensing and regulatory actions requiring an environmental assessment.
We have concluded that the proposed TS changes do not meet the criteria for categorical ,
exclusion as defined in 10 CFR 51.22(c)(9). Therefore, in accordance with the requirements of 10 CFR 51.30, the following information is provided to support an Environmental Assessment.
Installation of the new high density spent fuel storage racks in the Limerick Generating 1 Station (LGS), Units 1 and 2, spent fuel pools (SFPs) will be accomplished by keeping l radiation exposure as low as reasonably achievable (ALARA). Shielding from the spent fuel assemblies will be assured by maintaining the water level in the SFP at or above the minimum water level.
- 1) Need for increased Soent Fuel Storace Cacacity The SFPs at LGS, Units 1 and 2, are currently licensed to store 2040 fuel assemblies. Since all the spent fuel generated so far from operating the LGS facility is stored onsite in the SFPs, the SFPs are approaching their maximum storage capacity. Increasing the spent fuel storage capacity from 2040 fuel ,
. assemblies to 4117 fuel assemblies will enable us to defer the ' loss-of-full-core-reserve" year from 1998 to 2013. Since the current operating licenses for LGS, !
Units 1 and 2, expire in the year 2024 and 2029, respectively, it is evident that the LGS facility will be nearly capable of meeting its operating life storage requirement with some modest adjustments such as fuel consolidation or other palliative technologies. If the proposed TS changes are not approved, the spent fuel .
storage capacity at LGS, Units 1 and 2, will be. exhausted in the near future and could prevent continued plant operation.
- 2) Alternatives and Alternative Use of Resources
- Reprocessing of spent fuel has not developed as originally anticipated. In 1975, i l the NRC performed a Generic Environmental Impact Statement (GEIS) to evaluate alternatives for the handling and storage of spent fuel. The GEIS was to consider alternate methods for spent fuel storage as we!I the possible restrictions on termination of the generation of spent fuel through reactor shutdown.
f in 1979, the NRC issued NUREG-0575, " Final Generic Environmental Impact ,
Statement (FGEIS) on Handling and Storage of Spent Light Water Reactor Fuel."
The FGEIS indicated that the environmental costs of interim storage are essentially i negligible, regardless of where the spent fuel is stored. The FGEIS also determined that it would be uneconomical to shutdown a reactor before the end of its normal lifetime if the existing spent fuel storage capacity was filled. i 9
Attachment 1:
Page 18 Shipment of spent fuel to a high-level radioactive storage facility _is an alternative to increasing the onsite spent fuel storage _ capacity. However, the U.S.
Department of Energy's-(DOE's) high-level radioact!ve waste repository is not expected to begin receiving spent fuel until approximately 2010, at the earliest.
The existing SFPs at LGS willloose full core offload capability in 1998. Therefore,-
shipping spent fuel to the DOE's repository is not considered an alternative to -
Increased onsite spent fuel storage capacity.
Reprocessing of spent fuel from the LGS facility is no.t a viable alternative .since' there are no operating commercial reprocessing facilities in the United States.
Therefore, spent fuel would have to be shipped to an overseas facility for reprocessing.
Reducing the amount of spent fuel generated by improving usage of fuel and/or operation at a reduced power level would extend the life of the fuel in the reactor.
In the case of extended burnup of fuel assemblies, the fuel cycle would be extended, and fewer offloads would be necessary. We have already increased our refueling cycles from 18-months to 24-months by taking advantage of new fuel design technology. However, full-core offload capability will be lost in the near future. Operating the plants at a reduced power level would not make effective use of available resources, and would cause unnecessary economic hardship on us and our customers. Therefore, reducing the amount of spent fuel generated is not considered a practical alternative.
Reracking the SFPs was considered to be the most practical method for increasing the spent fuel storage capacity at LGS to satisfy our interim spent fuel storage needs. Expanding the onsite spent fuel storage capacity by modifying the existing SFPs is an acceptable alternative as described in the ,FGEIS. Other alternatives were also evaluated (e.g., in-Pool Rod Consolidation, Cask Storage, and Vaults) and are discussed in the attached Safety Analysis Report.
- 3) Environmentalimoact of the Prooosed Action The approval of this proposed TS Change Request will result in no significant effect on the human environment. Environmental and radiological considerations regarding this proposed TS change are evaluated in the supporting Safety Analysis -
Report.
The waste treatment systems for LGS, Units 1 and 2, are designed _to collect and process gaseous, liquid, and solid waste that may contain radioactive material.
The proposed TS changes to support implementation of the modification to install new high density spent fuel storage racks in each SFP at LGS will not impact the ability of the waste treatment systems to perform their intended design functions.
Increasing the spent fuel storage capacity as proposed will result in additional heat
1 Attachmsnt 1 I Page;19 ,
i load due to the increased spent fuelinventory. The anticipated maximum bulk SFP '
temperature _is approximately 143 F. The total heat load under worst case )
conditions is less than 37.6 million BTU /hr, which is less than 0.04% of the total l heat released to the environment due to plant operation, and well within the J
1 capability of the plant cooling systems (i.e., Fuel Pool Cooling and Cleanup (FPCC) and Residual Heat Removal (RHR) systems).
The increased bulk pool temperature will result in an increased SFP water evaporation rate. This has been calculated to increase Refuel Floor relative' 1 humidity as evaluated in the supporting Safety _ Analysis Report; however, this- )'
increase is within the capacity of the existing LGS Heating, Ventilation and Air .
Conditioning (HVAC) systems and does not necessitate any - hardware l modifications to the HVAC systems. The environmentalimpact resulting from the ;
increased heat load and water vapor emission are considered negligible. 1 i
All operations involved in reracking the SFPs will utilize detailed approved procedures with full consideration of ALARA principles. Similar operations have j been performed at a number of other facilities in the past, and there is reasonable -
assurance that the reracking operations at LGS can be accomplished safely and . '
efficiently, with minimum radiation exposure to personnel. The existing radiation protection program in place at LGS is adequate for the reracking operations.
Work, personnel traffic, and the movement of equipment will be monitored and. I controlled to minimize contamination and to assure that exposures are maintained ALARA.
During the reracking operation, existing storage _ racks will be' removed, as-necessary, and washed down in preparation for packaging and. shipment.
Estimates of person-rem exposures associated with'washdown and readying the ;
old racks for shipment are included in the supporting Safety Analysis Report. I Shipping containers and procedures will conform to Department of Transportation 1 (DOT) regulations and to the requirements of any state through which the shipment may pass.
- 4) Conclusion )
l We have concluded, that the NRC does not need to prepare a supplemental environmental impact statement in connection with approving this TS Change Request, and that a finding of no significant impact is supported by the information provided above. ,
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ll Conclusion l l
The Plant Operations Review Committee and the Nuclear Review Board have reviewed I this proposed change to the LGS, Units 1 and 2, TS and have concluded that they do not i involve an unreviewed safety question, and will not endanger the health and safety of the .j public.
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