ML20071A962

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Forwards Suppl 9 to Rev 1 of Joseph Oat Corp Licensing Rept on High Density Spent Fuel Racks for Quad Cities Unit 1 & 2. Changes Include Reduction in Number of Storage Cells to Be Installed & Rev to Spacing Arrangement
ML20071A962
Person / Time
Site: Quad Cities  
Issue date: 02/18/1983
From: Rybak B
COMMONWEALTH EDISON CO.
To: Harold Denton
Office of Nuclear Reactor Regulation
References
6025N, NUDOCS 8302250255
Download: ML20071A962 (17)
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Text

{{#Wiki_filter:' / Commonwealth Edison f C- ) one First National Plaz?, Chicago, Ilknois ' \\ C 7 Address R; ply to: Post Offica Box 767 N / Chicago, Illinois 60690 February 18, 1983 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555

Subject:

Quad Cities Station Units 1 and 2 Transmittal of Supplement 9 of Revision 1~to the Licensing Report on High Density Fuel Racks NRC Docket Nos. 50-254-and 50-265-Reference (a): R. F. Janecek letter to H. R. Denton dated March 26, 1981.

Dear Mr.' Denton:

Enclosed is Supplement 9 to Revision 1 of the report prepared by Joseph Oat Corporation for Commonwealth Edison entitled " Licensing Report on High Density Spent Fuel Racks for Quad Cities Units 1 and 2." The primary changes are the reduction in the number of storage cells to be installed and the related effect on the thermal and radiological analysis and revision to the minimum spacing between rack modules, the fuel pool wall and fuel pool wall fixtures. The revised spacing arrangement is consistent with our seismic and thermal-hydraulic analyses approved by the NRC. Additionally, minor corrections were incorporated. The following corrected sheets are attached: Pages 1-3, 2-1, 2-2, 2-3, 3-4, 3-5, 5-4, 5-11, 8-2, 10-2, 11-5, 11-6, Table 1.1, Figure 2.1, Figure 2.2, and Figure 3-7. Please address any questions you may have concerning this matter to this office. One (1) signed original and forty (40) copies of this transmittal (with enclosure) are provided for your use. Very truly yours, 8 B. y ak Nuclear Licens Administrator 1m Enclosure cc: Region III Inspector - Quad Cities k 0 R. Bevan - NRR 0 I 6025N 8302250255 830218 PDR ADOCK 05000254 P PDR

racks of the present design, full core discharge and refueling dis-charge capabilities would be lost after the 1984 and 1986 refueling outages, respectively. No further expansion of the Quad Cities spent fuel storage capacity is possible using the presently approved spent fuel storage rack design. In

contrast, high-density spent fuel storage racks have a capacity of 7554 fuel assemblies.

Therefore, 9 full core discharge is possible until the refueling outage of 2001 is Ir completed. Refuel discharge capability would be lost after the_ refueling outage of the year 2003. ~ Commonwealth' Edison ' Company, in ~its - function as operator, pro-poses to increase the spent fuel storage - capacity by replacing the-present spent fuel storage racks with new, high-density storage racks. This modification will include the use of a neutron absorber material in the racks, at an increase of k fr m 0.90 to 0.95. The March 26, eff

1981, letter to the NRC' requests a modification to Quad Cities l

Technical Specification 5.5B, " Fuel Storage," to impl'ement this change in keff* i The specification for design, construction, and quality assurance. of the high-density storage racks was prepared. by Quadrer, a-San Jose based company. The mechanical

design, seismic
analysis, thermohydraulic analysis, and other related calculations as well as the f abrication-of the hardware will be performed by Joseph Oat Corporation.

Joseph Oat Corporation, based in Camden, N.J., possesses ASME Code stamps for Section III, Classes 1, 2, and 3, and MC pressure ' j vessels and components. Southern Science Applications, Inc., of n Dunedin, Florida, is serving as a consultant to Joseph Oat Corporation in the areas of criticality analysis and other radionuclide. evaluations. l Consulting support on the overall - ef fort-is provided by NUS Corporation of Gaithersburg, Maryland. i 1 1-3 l t i

-{ 2. GENERAL ARRANGEMENT The high-density spent fuel racks consist of individual cells with a 6-inch-square cross

section, each of which accommodates a

single BWR fuel assembly. .The cell walls consist of a neutron absorber sandwiched between sheets of stainless steel. The cells are arranged in modules of varying numbers of cells with a 6.22-inch center-to-center spacing, The high-density racks are engineered to achieve. the dual objec-tive of maximum protection against structural loadings (such as ground motion) and the maximization of available storage locations. In

general, a greater.vidth to;- height.' aspect ' ratio provides~ ' greater margin against rigid body _ tipping.

Hence, the, modules are made - as - wide as possible within the constraints of transportation and site - handling capabilities. The high-density spent fuel racks will be installed in the Unit 1 and Unit 2 spent fuel pools, each of'which is 33 feet wide by 41 feet long. The Quad Cities Unit 1~ pool will contain 19 high-densi ty, ifuel racks in 9 different module sizes. The module types are labelled A ~ through K in Figure 2.1, which also shows their relative placement...9s There will be a total of 3657 storage locations. in; the t Quad Cities.L Unit 1 pool. The Quad C i t i e's, U n i t 2 pool will contain '20 ' high-density fuel racks in 9 d if fe rent module sizes. The module types are labelled A~ through K in Figure 2.2, which also shows their relative placement. 9 There will be a total of 3897 storage locations in the Quad Cities Unit 2 pool. Table 2.1 gives the detailed module data (e.g., weight, quantity, and number of storage locations). l The spent fuel rack modules are not anchored to the pool f'oor or connected to the pool walls. The minimum gap between any two spent fuel rack moduels will be 3.0 inches along the top edge. The minimum 9 gap between the fuel pool wall and spent fuel rack modules 2-1 I

will be 6-3/4 inches. The minimum gap between the top of the spent fuel rack modules and pool wall fixtures will be 1 inches. Adequate 9 clearance from other existing pool hardware will also be provided. Due to the gaps provided, the possibility of interrack impact, or rack collision with pool walls or other pool hardware during the postulated ground motion events will be precluded. 4 e l 2-3

i Table 2.1 Module Data Approximate Number of Weight T_yvpoe Quantity Cells / Module Array Size Lbs/ Module' A ~8 210. 14 x_15 .25,100.. B 6 196 14 x 14 23,500 C 8 182 14 x 13 21,850 D 4 135 9 x 15 16,400 l E 3 224 14 x 16 26,750 g F 2 256 16 x 16 30,500 G 1 192 12 x 16 23,050 H 4 195 13 x 15 23,400 J l 208 13 x 16 24,900 K 2 169 13 x 13 20,350 i l l i e 2-2

c applicable portions of Revision 15 of topical report CE-1-A, Commonwealth Edison Company Quality Assurance Program for Nuclear Generating Stations. Revision 15 of this report, dated January 2, 1981, was approved by the NRC in February 1981. V. Other References (a) NRC Regulatory Guides, Division 1, Regulatory Guides 1.13, 1.29, 1.71, 1.85, 1.92, and 1.124 (Revisions effective as of April 1980). (b) General Design Criteria for Nuclear Power Plants, Code of Federal-Regulations., Title.10, Part 50.,. Appendix A (GDC. Nos. 1, 2, 61, and'63). (c) NRC Standard Review Plan,. Sections 3.8.3 and 3.8.4. (d) NRC Standard Review Plan, Section 9.1.2 (as applicable to spent fuel racks). (e) "NRC Position for Review and Acceptance of Spent Fuel Storage and Handling Applications," dated April 14,

1978, and the modifications to this document of January 18,.1979.

3.3 INSTALLATION AND LEVELING The new spent fuel storage racks will ' arrive at the site by

truck, packaged on their
sides, and secured to shipping rigs.

~ Unloading of the packaged racks will be conducted by. station personnel. The racks will ' be brought through the. reactor building receiving bey-equipment air

lock, uprighted vertida11y using the

~ cradle shown on Figure 3.7, and lifted to the spent fuel pool operat-ing floor elevation using the overhead crane. Procedures and specifications will be used to control all opera-tions required to remove existing and install new spent fuel racks. A sequencing system will be employed for relocation of spent fuel within the pools. Initially, several existing racks will be emptied of spent fuel and removed from the pool, thereby creating the required space for the first new racks to be installed. Relocation of fuel to the new racks will then allow additional existing racks to be removed. No 3-4

old racks or new racks will be lifted over stored fuel or near enough to fuel so that any postulated lifting fig f ailure would result in any fuel damage. A diver will assist in leveling the new racks with shims during the installation project. This. will necessitate maintaining separation between the diver and the spent fuel stored in nearby racks. Initial washdown of the existing racks will be performed at the central decontamination area on the fuel handling floor. The racks will then be shredded, the pieces put into 55 gallon drums, and sent to a burial site. 9 The new racks will' be lift.ed and transported to the decontamina-tion. area using the lif ting f rame and rigging assembly shown on Figure 3.8. Four sets of holes allow the frame to accommodate all seven new rack configurations. The lifting rods and plugs shown on Figure 3.8 will extend-and thread into the leg portion of each rack. This assem-bly will also be used to lower the racks into final pool positions. In addition to the procedures which will be developed for rack

handling, other areas which will be addressed are:

acceptance procedures; equipment and specifications for removal of existing rack supports where necessary; interim fuel pool liner repair guidelines; and controls for final disposal of existing racks. l l 3-5

f. T ble 1.1).

This is when the inven' tory of fuel in the pool will be at its maximum resulting in an upper bound on the computed decay l heat rate.* 9 In the

past, Quad Cities reactors have operated on what is commonly referred to as "18 month cycle. "

Quite often, system plan-ning requires ex tended reactor coastdown operation (sometimes to 40% of rated -power).af ter-the end of. full power reactivity (19000 MND/STU) has been reached. The batch average discharge burn-up of current fuel batches is approximately 25000 MWD /STU. In the future, due to present lack of spent fuel reprocessing in the U.S., it is conceivable that the average discharge exposure can approach 30,000 MWD /STU due to higher initial enrichments and longer coastdowns. A longer coastdown period implies a greater value of t in. the foregoing equation, it g also implies' a smaller value of P. It can be shown that an exposure g

period, t

equal to 4.5 years (3-18 month refueling cycles) along o, with the. rated reactor power produces an upper bound on the value of P. This is due to the fact that f (to,ts) is a weak monotonically increasing f unc tion of t.

Hence, the reactor operating time is o

8 assumed to be 4.5 years (t =1.42x10 secs). g Having determined the heat dissipation rate, the nex t task is to evaluate the time temperature history of the pool water. Table 5.1.1 identifies the loading cases examined. The pool bulk temperature time history is determined using the first law of thermodynamics (conserva-tion of heat). The system to be analyzed is shown in Figure 5.1.1. A number of-simplifying assumptions are made to render the analysis conservative. The principal ones are: 1. The cooling water temperature in the fuel pool cooler i and the RHR heat exchangers are based on the maximum I postulated values given in the FSAR. Eecause of the elimination of 130 cells from the originally pro-l tosed expanded storage

capacity, additional conservatism is 9

l introduced into the computed results presented in this section. l 5-4

actual rack floor space is drawn. It is further assumed that the cylinder with this circle as its base is packed with fuel assemblies at the nominal pitch of 6.22 inches (see Figure 5.2.1). c. 'The.downcomer space. around the rack module group varies, as shown 'in Figures 2.' l and 2.2. The nominal downcomer gap 9 available in the pool is assumed to be the total gap avail-able around the idealized cylindrical rack;

thus, the maximum resistance to downward flow is incorporated into the
analysis, d.

No downcomer flow is assumed to exist between the rack modula. In this manner, a ' conservative idealized model for the rack assemblage is devised. The water flow is axisymmetric about the vertical axis of the circular rack assemblage, and thus, the flow is .two-dimensional (axisymmetric three-dimensional). The governing equation to characterize the flow field in the pool can now be written. The resulting integral equation can be solved for the lower plenum velocity field (in the radial direction) and axial velocity (in-cell velocity field), by using the method of collocation. It should be added here that the hydrodynamic loss coefficients which enter into the formulation of the integral-equation are also takne 4 so'rces and wherever discrepancies in reported from well-recognized u values exist, the conservative values are consistently used. l l After the axial velocity field is evaluated, it is a straight-forward matter to compute the fuel assembly cladding temperature. The 1 knowledge of the overall flow field enables pinpointing the storage location with the minimum axial flow (i.e., maximum water outlet temp-erature). This is called the most " choked" location. It is recog-nized that some storage locations, where rack module supports are located, have some additional hydraulic resistance not encountered in 5-11

The radiological consequences of storing the additional quantity of aged fuel have been evaluated. To ensure a conservative evaluation of the storage of failed fuel, it was assumed that the spent fuel storage pool is entirely filled with high-burnup spent fuel (28,500 Mwd /MtU burnup), ranging from newly removed fuel (1 core load of 724 fuel assemblies) to aged f uel with a cooling time of approximately 18 _ years. The maximum. fission-product inventory in the stored fuel in each pool whould result from an idealized fuel cycle in which approxi-mately 181 spent fuel elements were removed from the core and placed in the pool annually. With this fuel cycle, the expanded storage pool capacity, when completely filled, would contain the following: (1) For currently' authorized 724 newly removed assemblies storage capcity (full core load) and 4 refuel-ing discharges of 181 assem-blies with storage periods of 1, 2, 3, and 4 years, respec-tively. (2) Aged fuel in expanded

  • 13 refueling discharges of 181 l9 storage capacity assemblies with storage periods of 5 to 17 years and any remaining capacity (up to 170 assemblies),

containing fuel stored for 18 years. t l Reduced. fuel burnup or increased. cycle. length would result in a lower fission-product inventory or longer storage (decay) periods. Thus, the assumed storage pool composition should result in a conser-vative estimate of any additional radioingical impact due to the expanded storage capacity. Because of the elimination of 130 cells from the originally proposed expanded storage capacity, additional conservatism has 9 l been introduced through the use of these values to calculate fission-product inventory. t 8-2 1

and suspended from the spent fuel pool wall. Eighteen test samples are to be fabricated in accordance with Figure 10.1 and installed in the pool when the racks are installed. The procedure for fabrication and testing of samples shall be as follows: 4, a. Samples shall be cut to size and caref ully weighed in milli-grams. b. Length, width, and average thickness of each specimen to be measured and recorded. c. Samples shall be fabricated in, accordance. with Figure 10.1 and installed in'the pool. d. Twoi samples shall' be removed ' at each time interval 'per the schedule shown in Table 10.1. 10.4 specimen Evaluation After removal of the jacketed poison specimen f rom the fuel pool at...the designated time, a careful evaluation of that specimen will be made to determine its actual condition as well as its apparent durabili ty for continued function. Separation of the poison from the stainless steel specimen jacket must be performed carefully to avoid mechanically damaging. the poison specimen. Immediately upon removal, the specimen and jacket section should be visually examined for any effects of environmental exposure. Specific attention should be directed ~ to the examination.of,the. stainless-steel jacket for evidence i of physical degradation. Functional evaluation of the poison material is accomplished by the following measurements: a. 9 l i b. Nuetron attenuation measurements will allow evaluation of the continuing nuclear effectiveness of the poison. Con-sideration must be given in the analysis of the attenuation measurements for the level of accuracy of such measurements 10-2

area will not change. Therefore, personnel exposure will not significantly change because of the expansion. As explained in Section 8.2, there is also expected to be little to no change in gaseous radioactive release because of the expansion.

Thus, there will be very little to no change in the release of radioactivity and subsequent personnel exposure as a result of expanding the spent fuel
storage capacity of.the Quad-Cities spent fuel pools.

o Chemical Discharges The only chemical discharge that could be affected by the proposed expansion of the spent fuel pools is the powdered ion exchange resins used in the two filters demineralizers. As explained in Section 8 of this report, the frequency of resin replacement is determined primarily by the need for water clarity. As the particulate material that must be removed to maintain water clarity enters the. water during.refuelings and is removed well before the next refueling, the frequency of resin replacement is independent of the number of spent fuel assemblies stored in the pools. Therefore, there should be no change in the amount of spent fuel pool purification system filter resin discharged from the

plant, because of this modification.

o Heat Dissipation The two Quad-Cities spent fuel pool cooling system heat 6 exchangers are designed ;to transfer a total of.7. 3 x.10 .B tu/h r. It is estimated that, the increased storage will result in about a 10%* l9 increase in heat release when the pool is filled, or an increase of 5 approximately 7.3 x 10 Btu /hr* based on the heat exchanger design. 9 When compared to the over 5 x 10 Btu /hr discharged into the environment by each unit, this increase is seen to have a negligible effect on the environment. Because of the elimination of 130 cells from the originally 9 proposed expanded storage capacity, these values will actually be somewhat lower, and are therefore conservative. 11-5

Table 11.1 Quad Cities Station: Projection for Loss of Full Core Discharge Capability (FCDC) and Reload Discharge Capability (RDC) Currently Available Spent Fuel Racks Capacity 2280 = 1556 Capacity with FCDC* = Capacity with RDC 2080 = Lose FCDC 9/81 3/83 Lose RDC Currently "On Site" Spent Fuel Racks ** Capacity 1716 2440 = Capacity with FCDC = Capacity with RDC 2240 = Lose FCDC 12/82 Lose RDC 3/84 Currently Licensed Spent Fuel Racks Capacity 2920 = Capacity with FCDC 2196 = Capacity with RDC 2720 = Lose FCDC 3/84 5/86 Lose RDC High-Density Spent Fuel Racks Capacity 7554 = Capacity with FCDC 6830 = g capacity with RDC 7314 = Lose FCDC 3/01 3/03 Lose RDC Full core capacity = 724

    • Eight (8) racks (160 spaces) on site, not in pool, but available for repair and use 11-6

/ ( Table 1.1 Quad Ci Total Discharg Discharge Assemblies Assemblies In P Year Unit 1 Unit 2 Followin Refu 1974 64 144 208 1975 0 4 212 1976 156 164 532 1977 184 0 716 1978 0 180 896 1979 192 180 1268 1980 224 0 1492 1981 0 224 1716 1982 224 0 1940 1983 0 192 2132 1984 184 184 2500 1985 192 0 2692 1986 0 204 2896 1987 200 200 3296 1988 200 0 3496 1989 0 200 3696 1990 200 200 4096 1991 200 0 4296 1992 0 200 4496 1993 200 200 4896 1994 200 0 5096 1995 0 200 5296 1996 200 200 5696 1997 200 0 5896 1998 0 200 6096 1999 200 200 6496 2000 200 0 6696 2001 0 200 6896 2002 200 200 7296 2003 200 0 7496 2004 0 200 7696

  • The number of locations available after the completion of 4

IL

f \\ k ies Station, Units 1 and 2 imbly Discharges d Remaining Storage Capacity

  • ol With Additional High-Density M

Existing Licensed Racks Racks 2072 2068 1748 1564 1384 1012 788 564 1204 340 980 5614 148 788 5422 0 420 5054 228 4862 24 4658 0 4258 4058 3858 3458 9 3258 3058 2658 2458 2258 1858 1658 1458 1058 858 658 258 58 0 that year's scheduled refueling outage.

l lI 1

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/ ~ / m / s l." I N k R. ~\\ T N / 7. x / / / s J g\\ s / N \\ ~ l l ~ ( / l l l l ( I i l l FIGURE 0-7 -SPENT FUEL RACK UPLIFTING CRADLE l

racks of the present design, full core discharge and refueling dis-l charge capabilities would be lost after the 1984 and 1986 refueling outages, respectively. No further expansion of the Quad Cities spent fuel storage capacity is possible using the presently approved spent-fuel storage rack design. In

contrast, high-density spent fuel storage racks have a capacity of 7554 fuel assemblies.

Therefore, 9 full core discharge is possible until the refueling outage of 2001 is completed. Refuel discharge capability would be' lost .a f te r - the. refueling outage of the year 2003. Commonwealth Edison Company, -in-its-function as. ope r a to r, pro-poses to increase the. spent fuel storage capacity. by replacing 'the present spent fuel storage racks with new, high-density storage racks. This modification will include the use of a neutron absorber material in the racks, at an increase of k,gg from 0.90 to 0.95. The March 26,

1981, letter to the NRC' requests a

modification to Quad Cities Technical Specification 5.5B, " Fuel Storage," to implement this change in keff* The specification for design, construction, and quality assurance - of the high-density storage racks was prepared by Guadrer,.a san Jose - based company. The mechanical

design, seismic
analysis, thermohydraulic analysis, and other related calculations as well as the fabrication of_ the hardware will be performed by Joseph Oat Corporation.

Joseph Oat Corporation, based in Camden, N.J., possesses ASME Code stamps for Section III, Classes 1, 2, and 3, and MC pressure vessels and components. Southern Science Applications, Inc., of Dunedin, Florida, is serving as a consultant to Joseph Oat Corporation in the areas of criticality analysis and other radionuclide evaluations. Consulting support on the overall effort is provided by NUS Corporation of Gaithersburg, Maryland. 1-3

2. GENERAL ARRANGEMENT The high-density spent fuel racks consist of individual cells with a 6-inch-square cross

section, each of which accommodates a

single BWR fuel assembly. The cell walls consist of a neutron absorber sandwiched between sheets of stainless steel. The cells are arranged in modules of varying numbers of cells with a 6.22-inch center-to-center spacing. 1 The high-density racks are engineered to achieve the dual objec, tive of maximum protection against structural loadings (such as ground motion) and the maximization' of available storage locations. In

general, a greater ~ wid th to height aspect; ratio provides greater margin against rigid
  • body tipping.

Hence, ' the modules are made as wide as possible within the c_istraints of transpottation and site-handling capabilities. The high-density spent fuel racks will be installed in the Unit 1 and Unit 2 spent fuel pools, each of which is 33 feet wide by 41 feet long. The Quad Cities ' Unit 1 pool will. contain 19 high-densi ty - fuel racks in 9 different module sizes. The module types are labelled A through K in Figure 2.1, which also shows their relative placement.. 9 There will be a total of 3657 storage. Iocations in the - Quad Cities.4 Unit 1 pool. The Quad Cities Unit 2 pool will contain 20 high-density fuel racks in 9 different module sizes. The module type s, are labelled A through K in Figure 2.2, which also shows their relative placement. 9 There will be a total of 3897 storage locations in the Quad Cities Unit 2 pool. Table 2.1 gives the detailed module data (e g., weight, quantity, and number of storage locations). The spent fuel rack modules are not anchored to the pool floor or connected to the pool walls. The minimum gap between any two spent fuel rack moduels will be 3.0 inches along the top edge. The minimum 9 gap between the fuel pool wall and spent fuel rack modules 2-1

l Table 2.1 Module Data f Approximate Number of Weight Type Quantity cells / Module Array size Lbs/ Module A 8 210. ~14'x.15-25,100 B 6 196 14 x 14 23,500 C 8 182 '14 x 13 21,850 D 4 135 9 x 15 16,400 E 3 224 14 x 16 26,750 9 F 2 256 16 x 16 30,500 G 1 192 12 x 16 23,050 H 4 195 13 x 15 23,400 J l 208 13 x 16 24,900 K 2 169 13 x 13 20,350 i l l I 2-2 l l

will be 6-3/4 inches. The minimum gap between the top of the spent fuel rack modules and pool wall fixtures will be 1% inches. Adequate 9 clearance from other existing pool hardware will also be provided. Due to the gaps provided, the possibility of interrack impact, or rack collision with pool walls or other pool hardware during the postulated ground motion events will be precluded. l l W + 2-3

a s applicable portions of Revision 15 of topical report CE-1-A, Commonwealth Edison Company Quality Assurance Program for Nuclear Generating Stations. Revision 15 of this report, dated January 2, 1981, was approved by the NRC in February 1981. V. Other References (a) NRC Regulatory Guides, Division 1, Regulatory Guides 1.13, 1.29, 1.71, 1.85, 1.92, and 1.124 (Revisions ef fective as of April 1980). (b) General Design Criteria for Nuclear Power Plants, Code of Federal Regulations, Title 10, Part 50, Appendix A (GDC Nos. 1, 2, 61, and 63). (c) NRC Standard Review Plan, Sections 3.8.3 and 3.8.4. (d) NRC Stand ard Review Plan, Sec tion 9.1.2 (as applicable to spent fuel racks). (e) "NRC Position for Review and Acceptance of Spent Fuel Storage and Handling Applications," dated April 14, 1978,- and the modifications to this document of January 18, 1979. 3.3 INSTALLATION AND LEVELING The new spent fuel storage racks will' arrive at the site by

truck, packaged on their
sides, and secured to shipping rigs.

Unloading of the packaged racks will be conducted by station l personnel. The racks will be brought through the, reactor building-receiving bay equipment air

lock, uprighted vertically using the cradle shown on Figure 3.7, and lifted to the spent fuel pool operat-ing floor elevation using the overhead crane.

Procedures and specifications will be used to control all opera-tions required to remove existing and install new spent fuel racks. A sequencing system will be employed for relocation of spent f uel within-the pools. Initially, several existing racks will be emptied of spent fuel and removed from the pool, thereby creating the required space for the first new racks to be installed. Relocation of fuel to the new racks will then allow additional existing racks to be removed. No 3-4

old racks or new racks will be lifted over stored fuel or near enough to fuel so that any postulated lifting fig f ailure would result in any fuel damage. A diver will assist in leveling the new racks with shims during the installation project. .This will necessitate maintaining separation between the diver and the spent fuel stored in nearby racks. Initial washdown of "the. existing racks will be performed at the central decontamination area on the fuel handling floor. The racks will then be shredded, the pieces put into 55 gallon drums, and sent l to a burial site. 9 The new racks will be lifted and transported.to the decontamina-tion area using the -lif' ing frame and rigging assembly shown on Figure t 3.8. Four sets of holes allow the frame to accommodate all seven new rack configurations. The lifting rods and plugs shown on Figure 3.' 8 will extend. and thread into.the leg portion of.each rack. This assem-bly will also be used to lower the racks into fi'nal. pool positions. In addition to the procedures which will be developed for rack

handling, other areas which will be addressed are:

acceptance procedures; equipment and specifications for removal of existing rack supports where necessary; interim fuel pool liner repair guidelines; and controls for final disposal of existing racks. t 3-5

(ref. Table 1.1). This is when the inventory of fuel in the pool will be at its maximum resulting in an upper bound on the computed decay l heat rate.* 9 In the

past, Quad Cities reactors have operated on what is l

commonly referred to as "18 manth cycle. " Quite often, system plan-ning requires ex tended reac t6r coastdown operation (sometimes to 40% .of rated power) after the end of full-power reactivity (19000 MWD /STU) has been reached. The batch average discharge burn-up of current fuel batches is approximately 25000 MWD /STU. In the future, due to present lack of spent fuel reprocessing in the U.S., it is conceivable that the average discharge exposure can approach 30,000 MWD /STU due to higher initial enrichments and longer coastdowns. A longer coastdown period implies a greater value of t in the f orego i'ng equation, it g also implies a smaller value of P. It can be shown that an exposure g

period, t

equal to 4.5 years (3-18 month refueling cycles) along o, with the ra ted reactor power produces an upper bound on the value of 'P.- This is due to the fact that f (to,ts) is a weak monotonically increasing function of t

Hence, the reactor operating time is g.

8 assumed to be 4.5 years (t =1.42x10 secs). g Having determined the heat dissipation rate, the nex t task is to evaluate the time temperature history of the pool water. Table 5.1.1 identifies the loading cases examined. The pool bulk temperature time history is determined using the first law of thermodynamics (conserva-tion of heat). The system to be analyzed is shown in Figure 5.1.1. A number of simplifying assumptions are made to render the analysis conservative. The principal ones are: 1. The cooling water temperature in the fuel pool cooler and the RHR heat exchangers are based on the maximum postulated values given in the FSAR. Because of the elimination of 130 cells from the originally pro-posed expanded storage

capacity, additional conservatism is 9

introduced into the computed results presented in this section. 5-4

actual rack floor space is drawn. It is further assumed that the cylinder with this circle as its base is packed with fuel assemblies at the nominal pitch of 6.22 inches (see Figure 5.2.1). c.' The downcomer space. around the rack. module group varies, as shown in Figures 2.1 and 2.2. The nominal downcomer gap 9 available in the pool is assumed to be the total gap avail-able around the idealized cylindrical rack;

thus, the maximum resistance to downward flow is incorporated into the analysis.

d.- No down' comer flow is assumed to exist between the rack modules. In ' this manner, a conservative idealized model for the rack assemblage is devised. The water flow is axisymmetric about the vertical axis of the circular rack assemblage, and thus, the flow is two-dimensional (axisymmetric three-dimensional). The governing equation to characterize the flow field in the pool can now be written. The resulting integral equation can be solved for the lower plenum velocity field (in the radial direction) and axial velocity (in-cell velocity field), by using the method of collocation. It should be added here that the hydrodynamic loss coefficients which enter into the formulation of the integral. equation are also takne 4 from well-recognized sources and wherever discrepancies in reported l values exist, the conservative values are consistently used. l l 1 l After the axial velocity field is evaluated, it is a straight-1 forward matter to compute the fuel assembly cladding temperature. The knowledge of the overall flow field enables pinpointing the storage location with the minimum axial flow (i.e., maximum water outlet temp-erature). This is called the most " choked" location. It is recog-nized that some storage locations, where rack module supports are located, have some additional hydraulic resistance not encountered in 5-11

The radiological consequences of storing the additional quantity of aged f uel have been evaluated. To ensure a conservative evaluation of the storage of failed fuel, it was assumed that the spent fuel storage pool is entirely filled with high-burnup spent fuel (28,500 Mwd /MtU burnup), ranging from newly removed fuel (1 core load of 724 fuel assemblies) to aged fuel'with a cooling time of approximately 18 years. The maximum fission-product inventory in the stored fuel in each pool whould result from-an idealized fuel cycle in which approxi-mately 181 spent f uel elements were removed from the core and placed in the pool annually. With this fuel cycle, the expanded storage pool capacity, when completely filled, would contain the following: (1) For currently authorized 724 newly ' removed assemblies storage capcity .(full core load) and 4 refuel- 'ing discharges of 181 assem-blies with storage periods of 1, 2, 3, and 4 years, respec-tively. (2) Aged fuel in expanded

  • 13 refueling discharges of 181 l9 storage capacity assynblies with storage periods of 5 to 17 years and any remaining capacity (up to 170 assemblies),

containing fuel stored for 18 years. Reduced fuel burnup or. increased cycle. length would result in a lower fission-product inventory or longer storage (decay) periods. Thus, the assumed storage pool composition should result in a conser-vative estimate of any additional radiological impact due to the expanded storage capacity. Because of the elimination of 130 cells from the originally proposed expanded storage capacity, additional conservatism has 9 been introduced through the use of these values to calculate fission-product inventory. 8-2

and suspended from the spent fuel pool wall. Eighteen test samples are to be fabricated in accordance with Figure 10.1 and installed in the pool when the racks are installed. The procedure for fabrication and testing of samples shall be as follows: a. Samples shall be cut to size and carefully weighed in milli-grams. b. Length, wid th, and average thickness of each specimen to be measured and recorded. c. Samples shall be fabricated in accordance with Figure 10.1 and installed in the pool. d. Two samples shall be removed at each time interval per the schedule shown in Table 10.1. 10.4' specimen Evaluation After removal of the jacketed poison specimen from the fuel pool at the designated time, a careful evaluation of that specimen will be l made to determine its actual condition as well as its apparent l durability for continued function. Geparation of the poison from the stainless steel specimen jacket must be performed carefully to avoid mechanically damaging. the poison specimen. Immediately upon removal, the specimen and jacket section should be visually examined for any effects of environmental exposure. Specific attention should be directed to thec examination of the stainless steel jacket for evidence of physical degradation. Functional evaluation of the poison material is accomplished by the following measurements: 9 a. b. Nuetron attenuation measurements will allow evaluation of the continuing nuclear effectiveness of the poison. Con-sideration must be given in the analysis of the attenuation measurements for the level of accuracy of such measurements 10-2

area will not change. Therefore, personnel exposure will not significantly change because of the expansion. As explained in Section 8.2, there is also expected to be little to no change in gaseous radioactive release because of the expansion.

Thus, there will be very little to no change in the release of radioactivity and subsequent personnel exposure as a result of expanding the spent fuel stor. age capacity of the Quad-Cities spent fuel pools.

o Chemical Discharges The only chemical discharge that could be affected by the proposed expansion of the spent fuel pools is the powdered ion exchange resins used in the two filters demineralizers. As explained in Section 8 of ' this report, the 'f requency of resin replacement is determined primarily by the need for water clarity. As the particulate material that must be removed to maintain water clarity enters the water during refuelings and is removed well before the next refueling, the frequency of resin replacement is independent of the number of spent fuel assemblies stored in the pools. Therefore, there should be no change in the amount of spent fuel pool purification system filter resin discharged from the

plant, because of this modification.

o Heat Dissipation The two Quad-Cities spent fuel pool cooling system heat 6 exchangers are designed to-transfer a total of 7.3 x 10 Btu /hr. It is estimated that, the increased storage will result in about a 10 %* l9 increase in heat release when the pool is filled, or an increase of 5 approximately 7.3 x 10 Btu /hr* based on the heat exchanger design. ] 9 When compared to the over 5 x 10 Btu /hr discharged into the environment by each unit, this increase is seen to have a negligible effect on the environment. Because of the elimination of 130 cells from the originally 9 proposed expanded storage capacity, these values will actually be somewhat lower, and are therefore conservative. 11-5 L

Table 11.1 Quad Cities Station: Prof *ction for Loss of Full Core Discharge Capability (FCDC) and Reload Dischstge Capability (RDC) Currently Available Spent Fuel Racks 2280 capacity = Capacity with FCDC* 1556 = Capacity with RDC 2080 = Lose FCDC 9/81 Lose RDC 3/83 Currently "On Site" Spent Fuel Racks ** Capacity ' = 2440 1716 Capacity with FCDC = Capacity with RDC 2240 = Lose FCDC 12/82 Lose RDC 3/84 Currently ~ Licensed Spent Fuel Racks Capacity 2920 = Capacity with FCDC 2196 = Capacity with RDC 2720 = Lose FCDC 3/84 Lose RDC 5/86 High-Density Spent Fuel Racks capacity 7554 = Capacity with FCDC 6830 = g Capacity with RDC 7314 = 3/01 Lose.FCDC Lose RDC 3/03 Full core capacity = 724

    • Eight (8) racks (160 spaces) on site, not in pool, but available for repair and use 11-6

r Table 1.1 Quad Cit I sse l Total Discharge Assemblies In Po L Discharge Assemblies Following Refuel Unit 1 Unit 2 i Year D'- 208 144 212 ~ 64 1974 4 532 4 0 1975 164 716 156 1976 0 896 184 1977 180 1268 0 1978 180 1492 192 1979 0 1716 224 1980 224 1940 0 1981 0 2132 224 1982 192 2500 0 1983 184 2692 184 1984 0 2896 192 1985 204 3296 0 1986 200 3496 ~ 200 1987 0 3696 200 1988 200 4096 0 1989 200 4296 200 1990 0 4496 200 1991 200 4896 0 1992 200 5096 200 1993 0 5296 200 1994 200 5696 0 1995 200 5896 200 1996 0 6096 200 ~ 1997 200 6496 0 1998 200 6696 200 1999 0 6896 200 2000 200 7296 0 2001 200 7496 200 2002 0 7696 200 2003 200 0 2004 the completion of t of locations available after

  • The number

....L C--___......,,,,

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  1. '. # i.. -

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5. MI "2

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===== 9 / 3 / Wm.K ~~ ./ t A JN FIGURE 3-7-SPENT FUEL RACK UPLIFTING CRADLE

l -racks of the present design, full core discharge and refueling dis-charge capabilities. would be lost after the 1984 and 1986 - refueling i outages, respectively. No further expansion of the Quad Cities spent fuel storage capacity is possible using the presently approved spent fuel storage rack design. In contrast, high-density spent fuel storage racks have a capacity of 7554 fuel assemblies. Therefore, 9 full core discharge is possible until the refueling outage of 2001 is completed. Refuel discharge capability would be lost after the refueling outage of the year 2003. l e Commonwealth Edison Company, in -its function as

  • operator / pro ~

poses to increase the spent fuel storage capacity by replacing ; the i present spent fuel storage racks with new, high-density storage' racks.~ This modification will include the use of a neutron absorber material in the racks, at an increase of k,gg from 0.90 to 0.95. The March 26,

1981, letter to the NRC' requests a modification to Quad Cities Technical Specification 5.5B, " Fuel Storage," to implement this change-in k,gg.

The specification for design, construction, and quality assurance of the. high-density storage racks was prepared. by; Quadrer, a-san Jose- - ~ based company. The mechanical

design, seismic
analysis, thermohydraulic analysis, and other related calculations as well ' as -

the fabrication of the hardware will. be performed ' by _ Joseph Oat. Corporation. Joseph Oat Corporation, based in Camden, N.J., possesses. ASME Code stamps for Section III, Classes 1, 2, and 3, and MC pressure-vessels.and components. Southern Science Applications, Inc., of Dunedin, Florida, is serving as a consultant to Joseph Oat Corporation. in the areas of criticality analysis and other radionuclide' evaluations. Consulting support on the overall - ef fort is provided by NUS Corporation of Gaithersburg, Maryland. 1-3

2. GENERAL ARRANGEMENT The high-density spent fuel racks consist of individual cells with a 6-inch-square cross

section, each of which accommodates a

single BWR fuel assembly. The cell walls consist of a neutron absorber sandwiched between sheets of stainless steel. The cells are- ~ arranged in modules of varying numbers of cells with a 6.22-inch center-to-center spacing. i The high-density racks are engineered to achieve the duel objec-- tive of maximum protection against structural loadings (such as ground i motion) and the maximization of' available storage locations. In

general, a greater. 'wid th' to ' height aspect-ratio provides.' greater.-

margin against rigid body' tipping.:

Hence, the module' are made. as' s

wide as possible within the constraints of transportation and~ site-handling capabilities. The high-density spent fuel racks will be installed in the Unit 1 and Unit 2 spent fuel pools, each of which is 33 feet wide by 41 feet long. The Quad Cities Unit 1 pool will contain: 19. high-density ' fuel racks in 9 different module sizes. The module types are labelled A through K in Figure 2.1, which also shows their rela.tive placement;'_ 9 There will" be a total.of 3657 storage c locations in the, Quad Citiesa Unit 1 pool. The Quad Cities Unit 2 pool will contain 20 high-density fuel racks in 9 different module sizes. The mo'dule types are labelled A. through. K in Figure 2.2, which also shows their relative. placement. 9l There will be a total of 3897 storage locations in the Quad Cities ^^ Unit 2 pool. Table 2.1 gives the detailed module data (e.g., weight, quantit_y,; and number of storage locations). ~ The spent fuel rack modules are not anchored to the pool floor or connected to the pool walls. The minimum gap between any two spent fuel rack moduels will be 3.0 inches along the top edge. The minimum g; gap between the fuel pool wall and spent fuel rack modules 2-1 j

Table 2.1 Module Data Approximate Number of Weight Type Quantity Cells / Module Array size Lbs/ Module A 8 210.' 14 x 15 25,100 B 6 1961 14 x 14 23,500 C 8 182 14 x 13 21,850' D 4 135 9 x 15 16,400 E 3 224 14 x 16 26,750 9 F 2 256 16 x 16 30,500 G 1 192 12 x 16 23,050 H 4 195 13 x 15 23,400' J' l-208 13 x 16 .24,900 K 2 169 13 x 13 20,350 r e 2-2

will ba 6-3/4 inchas. The minimum gcp betwsen tha top of ths epsnt fuel rack modules and pool wall fixtures will be 1 inches. Adequate 9 clearance from other existing pool hardware will also be provided. Due to the gaps provided, the possibility of interrack impact, or rack collision with pool walls or other pool hardware during the postulated ground motion events will be precluded. k 'e e 2-3

i applicable portions of Revision 15 of topical report CE-1-A, Commonwealth Edison Company Quality Assurance Program for Nuclear Generating Stations. Revision 15 of this report, dated January 2, 1981, was approved by the NRC in February 1981. V. Other References (a) NRC Regulatory Guides, Division 1, Regulatory Guides 1.13, 1.29, 1.71, 1.85, 1.92, and 1.124 (Revisions effective as of April 1980). (b) General Design Criteria for Nuclear Power Plants, Code of Federal Regulations, Title 10, Part 50,. Appendix A (GDC, Nos._ 1, 2, 61, and 63)'. (c) NRC Standard Review Plan, Sections-3.8.3 and-3.8.4.- (d) NRC Standard Review Plan, Section 9.1.2 (as applicable to spent fuel racks). (e) "NRC Position for Review and Acceptance of Spent Fuel Storage and Handling Applications," dated April 14,

1978, and the modifications to this document of January 18, 1979._

3.3 INSTALLATION AND LEVELING The new spent fuel storage racks will arrive at the site by

truck, packaged on their
sides, and secured to shipping rigs.

Unloading of the packaged racks will be conducted by station personnel. T h e' ~r a c k s will be brought through the reactor building' receiving bay equipment air

lock, uprighted vertically using th~e' cradle shown on Figure 1.h and lifted to the spent fuel pool operat-ing floor elevation c'u rv, ete overhead crane.

Procedures a r.) speciffcations will be used to control all opera-tions required to remove existing and install new spent fuel racks.- A sequencing system will be employed for relocation of spent fuel within the pools. Initially, several existing racks will be emptied of spent fuel and removed from the pool, thereby creating the required space for the first new racks to be installed. Relocation of fuel to the new racks will then allow additional existing racks to be removed. No 3-4

old racks or new racks will be lif ted over stored fuel or near enough to fuel.so that any postulated. lif ting fig f ailure would result in any fuel damage. A diver will assist in leveling the new racks with shims during the installation project. This will necessitate maintaining _ separation between the diver and the spent fuel stored in nearby racks. Initial washdown of the existing racks will be performed at the central decontamination area on the fuel handling floor. The racks will then be shredded, the pieces put into 55 -gallon drums, and sent to a burial site. 9 The new racks will, be lif ted and transported to the decontamina-tion area using the lif ting frame and rigging assembly shown on Figure 3.8. Four sets of holes allow the frame to accommodate all seven new rack configurations. .The-lif ting rods and plugs shown on Figure 3.8 . will extend and thread into'the. leg portion of each rack. This assem-bly will'also be used to lower the racks into final pool positions. In addition to the 1 procedures which will be developed for rack

handling, other areas which will be addressed are:

acceptance procedures; equipment and specifications for removal of existing rack supports where necessary; interim fuel pool liner repair guidelines; and controls _for final disposal of existing racks. t 3-5

(ref. Table 1.1). This is when the inventory of fuel in the pool will be at its maximum resulting in an upper bound on the computed decay heat rate.* 9 In the

past, Quad Cities reac to rs have operated on what is commonly referred to as "18 month cycle. "

Quite often, sys tem plan-ning requires ex tended reactor coastdown operation (sometimes to 40% .of = rated power) - af ter the end of full power. reactivity (19000 MWD /STU) has been reached. The batch average discharge burn-up of current fuel l batches is approximately 25000 MWD /STU. In the future, due to present lack of spent fuel reprocessing in the U.S., it is conceivable that the average discharge exposure can approach 30,000 MWD /STU due to higher initial. enrichments and longer coas tdowns. A longer coas tdown period implies a greater value of 't in the foregoing equation, it g also implies a smaller value of~P. It'can be shown that an exposure g

period, t

equal to 4.5 years (3-18 month refueling cycles) along o, with the rated reac tor power produces an upper bound on the value of P. This is ~due to the fact that f (to,ts) is a weak monotonically inmaasing function of t

Hence, the reactor operating time is o.

8 assumed to be 4.5 years (tg=1.42x10 secs). Having determined the heat dissipation rate, the nex t task is to evaluate the time temperature history of the pool water. Table 5.1.1 identifies the loading cases examined. The pool bulk temperature time history is determined using the first law of thermodynamics (conserva-tion of heat). - The system to be analyzed is shown in Figure 5.1.1. 'A number of simplifying assumptions.are made to render the analysis conservative. The principal ones are: 1. The cooling water temperature in the fuel pool cooler and the RHR heat exchangers are based on the maximum postulated values given in the FSAR. Because of the elimination of 130 cells from the originally pro-posed expanded storage

capacity, additional conservatism is 9

introduced into the computed results presented in this section. 5-4

actual. rack floor space is drawn. It is further assumed that -the cylinder with this circle as its base is packed with fuel assemblies at the nominal pitch of 6.22 inches (see Figure 5.2.1). The downcomer space.. around the. rack module group varies, as c. shown ini Figures 2.1 and 2.2. The nominal downcomer gap 9 available in the pool is assumed to be the total gap avail-able around the idealized cylindrical rack;

thus, the

. maximum resistance to downward flow is incorporated into the analysis. 4 d. No downcomer-flow is assumed to exist ' between the rack modules. In ' this ' manner, a " conservative idealized model for the rack assemblage is devised. The water flow is axisymmetric about the vertical axis of. the. cir'cular rack assembl. age, and thus, the flow is

.two-dimensional (axisymmetric three-dimensional).

The governing equation to characterize the flow field in the pool can now be written. The resulting integral equation can be solved for the lower plenum velocity field. (in the radial direction) and axial velocity (in-cell velocity field), by using the method of collocation. It - should be added here that the hydrodynamic loss coefficients which H enter -into ^ the formulation of the integral equation are also takne 4 'f rom = well-recognized. sources and "wherever discrepancies in reported values exist, the conservative values are consistently used. I l After the axial velocity field is evaluated, it is a straight-forward matter to compute the fuel assembly cladding temperature. The knowledge of the overall flow field enables pinpointing the storage location with the minimum axial flow (i.e., maximum water outlet temp-erature). This is called the most " choked" location. It is recog-nized that some storage locations, where rack module supports are located, have some additional hydraulic resistance not encountered in 5-11

The radiological consequences of storing the additional quantity of aged fuel have been evaluated. To ensure a conservative evaluation of the storage of failed fuel, it was assumed that the spent fuel storage pool is entirely filled with high-burnup spent fuel (28,500 Mwd /MtU burnup), ranging from newly removed fuel (1 core load of 724 fuel assemblies) to aged. fuel with a cooling time of approximately 18 years. The maximum. ~ fission-product inventory ~in. the stored fuel in each poo1 whould result' f rom an idealized fuel cycle in which approxi- ~ mately 181 spent fuel elements were removed from the core and placed in the pool annually. With this fuel cycle, the expanded storage pool capacity, when completely filled, would contain the following: (1),For currently authorized 7 24 newly removed assemblies storage capcity (full core load) and 4 refuel-ing discharges of 181 assem-blies with storage periods of 1, 2, 3, and 4 years, respec- + tively. (2) Aged fuel in expanded

  • 13 refueling discharges of 181 l9 storage capacity assemblies with storage periods of 5 to 17 years and any remaining capacity (up to 170 assemblies),

containing fuel stored for 18 years. j Reduced. fuel burnup or. increased cycle. length would. result in a lower fission-product inventory or longer storage (decay) periods. Thus, the assumed storage pool composition should result in a conser-vative estimate of any additional radiological impact due to the expanded storage capacity. Because of the elimination of 130 cells from the originally proposed expanded storage capacity, additional conserva"tism has 9 been introduced through the use of these values to calculate fission-product inventory. l 8-2 i

and suspended from the spent fuel pool wall. Eighteen test samples are to be fabricated in accordance ~ with Figure 10.1 and ^ installed in the pool when the racks are installed. The procedure for fabrication and testing of samples shall be as follows: a. Samples shall'~be' cut to size ~ and carefully weighed in milli-grams. b. Length, width, and average thickness of each specimen to be measured and recorded. c. Samples shall be fabricated in accordance with Figure 10.1 and installed in the pool. d. .Two samples shall be removed at each~ - time interval per the j schedule shown in Table 10.1. .10.4' Specimen Evaluation After removal of the jacketed poison specimen from the fuel pool at' the designated time, a careful evaluation of that specimen will be made to determine its actual condition as well as its apparent durability for continued function. Separation of the poison from the stainless uteel specimen jacket must be performed carefully to avoid mechanically damaging. the poison specimen. Immediately upon removal, _the specinen and jacket section should be visually examined for any . effects of environmental exposure. 9pecific attention should be , directed to the ~ examination of the. stainless steel. jacket for evidence of physical degradation. ' Functional evaluation of the poison material is accomplished by the following measurements: a. 9 b. Nuetron attenuation measurements will allow evaluation of the continuing nuclear effectiveness of the poison. Con-sideration must be given in the analysis of the attenuation measurements for the level of accuracy of such measurements 10-2 __ ------ J

area will not change. Therefore, personnel exposure will not significantly change because of the expansion. As explained in Section 8.2, there is also expected to be little to no change in gaseous radioactive release because of the expansion.

Thus, there will be very little to no change in the release of radioactivity and subsequent personnel exposure as a result of expanding the spent fuel storc e capacity of the Quad-Cities spent fuel pools.

o Chemical Discharges The only chemical discharge that could be affected by the proposed expansion of the spent fuel pools is_ the powdered ion exchange resins used in the two filters deminerali~zers. As explained in Section 81of tihis report, the f requency. of resin replacement is determined primarily by the need for water clarity. As the

  • particulate material that must be removed to maintain water clarity

. enters the water.during 'refuelings and is removed well before the next refueling, the frequency of resin replacement is independent of the number of. spent fuel assemblies stored in the pools. Therefore, there should be no change 'in the amount of spent fuel pool purification system filter resin discharged from the

plant, because of this

. modification.. o Heat Dissipation The two Quad-Cities. spent fuel pool cooling system heat 6 exchangers - are '. designed to transfer a ~ total. of.7.3 x 10 . Btu /hr. It is es timated that, the increased storage will result in about a 10%* l9 increase in heat release when the pool is filled, or an increase of 5 approximately 7.3 x 10 Btu /hr* based on the heat exchanger design. c 9 When compared to the over 5 x 10 Btu /hr discharged into the environment by each unit, this increase is seen to have a negligible effect on the environment. Because of the elimination of 130 cells from the originally 9 proposed expanded storage capacity, these values will actually be somewhat lower, and are therefore conservative. 11-5

Table 11.1 Quad Cities Station: Projection for Loss of Full Core Discharge Capability (FCDC) and Reload Discharge Capability (RDC) Currently Available Spe'nt Fuel Racks Capacity' 2280 = 1556 Capacity with'.FCDC* -=- 2080 Capacity with RDC = Lose FCDC 9/81 Lose RDC 3/83 Currently "On Site" Spent Fuel Racks ** 2440 Capacity = Capacity with FCDC. = - 1716 Capacity with RDC 2240 = 12/82 Lose FCDC-Lose RDC '3/84 ' Currently Licensed Spent Fuel Racks Capacity 2920 = Capacity with FCDC 2196 = Capacity with RDC 2720 = 3/84 Lose FCDC 5/86 Lose RDC High-Density Spent Fuel Racks capacity 7554 = Capacity with FCDC 6830 '= 9 Capacity with RDC 7314 = . Lose FCDC _3/01 Lose RDC 3/03 t 1 Full core capacity = 724 Eight (8) racks (160 spaces) on site, not in pool, but'available for repair and use 11-6

Table 1.1 Quad Cities Station, Units 1 and 2 Fuel Assembly Discharges r J Total Discharged Discharge Assemblies Assemblies In Pool Remaining Storage Capacity

  • Year Unit 1 Unit 2 Following Refueling' With Additional High-Density j-Existing Licensed Racks Racks 1974 64 144 208 2072 1975 0

4 212 2063 1976 156 164 532 1748 1977 184 0 716 1564 1978 0 18% 896 ~ 1384 i 1979 192 180 1268 1012 1980 224 0 1492 788 1981 0 224 1716 564 1204 4 i 1982 224 0 1940 340 980 5614 1983 0 192 2132 148 788 - 5422 3 1984 184 184 2500 0 420 5054 1985 192 0 2692 228 4862 1986 0 204 2896 24 4658 1987 200 200 3296 0 4258 1988 200 0 3496 4058 1989 0 200 '3696 3858 1990 200 200 4096 3458 9 1991 200 0 4296 3258 1992 0 200 4496 3058 1993 200 200 4896-2658 1994 200 0 5096 2458 5 1995 0 200 5296 2258 1996 200 200 5696 1858 1997 200 0 5896 1658 1998 0 200 6096 1458 j 1999 200 200 6496 1058 2000 200 0 6696 858 2001 0 200 6896 '658 2002 200 200 7296 258 2003 200 0 7496 58 2004 0 200 7696 0

  • The number of locations available af ter the completion of that year's scheduled refueling outage.

t t 6 k a = + w

AX = W I n n_ er w- - f Table 1.1 Quad Cit ' e = ~+ men-M- p Total Discharge & g Assemblies In Po g Discharge Assemblies -eA Unit 1 Unit 2 Fo p ~ Year ~ -- 208 W 144 64 212 h h 'E 1974 4 0 532 1975 164 716 E!!r-- 156 1976 0 k I-134 895 h 1977 180 1268

== 0 1978 180 1492 e 192 1979 0 224 1716 E T 1980 224 h d. 0 1940 1981 0 E V 1982 224 2132 0 192 2500 8 1983 184 184 2692 P 1984 0 I 192 2896 1985 204 3296 E ~ = =- 0 1986 200 3496 2 200 1987 0 200 3696 F 1988 200 4096 L 4 0 1989 200 4296 C v 1990 200 1991 200 4496 7-C 0 0 200 4896 h 1992 200 200 5096 E R 1993 0 200 5296 imr N 1994 200 0 5t96 r 1995 200 1996 200 5896 0 E [ 1997 200 6096 0 200 6496 1998 200 1999 200 6696 = 0 F = 200 6896 2000 200 S' O 0 7296 2001 200 7496 5-m 200 2002 0 200 7696 2003 200 f 0 2004 q 3 the completion of ^ number of locations available after h

  • The F

~ _ _ w = Y- _5 \\ o __m M

es Station, Units 1 and 2 ably Discharges il Remaining Storage Capacity

  • nq With Additional High-Density 1

Existing Licensed Racks i Racks 2072 ~ 2068 1748 [ 1564 1384 1012 x 788 564 1204 340 980 148 5614 - ~ 788 0 5422 420 5054 228 4862 24 4658 0 4258 4058 3858 2458 9 3258 3058 2658 2458 2258 1858 i 1658 l 1458 1058 858 658 258 58 0 tat year's s c h'.d u led refueling outage. e ? t

e W- -c i p f.,' -f , p f / f ,=. N t /_ ~L-. .= 6 k FIGURE 3-7 -SPENT FUEL RACK UPLIFTING CRADLE

9 1 "5 0 7 8 6 8 { i 1 o 0 1 \\ 4 7 [E.\\ A Ro "8 "3 1 s A T 8 1 1 1 1 K ~0 IN 2 F E G s 1 S 1 \\C U 0 A1 1 1 S / E IT I i C = DA "9 "0 N U Q 1 2

5. M I

1 2 3 4 9 2 1 B B B C R 8 p O F S = L E E L S = UE "9 N U} F/IN m o IM 1 OL D S 1 H 9 2 1 MC 8 "0 ML 8 RA 1 2 3 1 2 OM C C 0 C K E E R KC "2 8 F 3 TP U C 7 9 - AE G A 5 4 R6 LR I PP F 3 FO( = = TN "0 E 2 3 4 2 5 D A A H M E 5 9 4 G 5 NA R RA D E L T T A TPI AR "0 N N E E I E 1 1 2 1 5 M G N 5 A M A D A A H E P 9 TI R G NU O OQT CES "9 \\ = 3 "0 N 0 I 0

5. M "4

"9 "9 ) 1 1 S W C 4 1 1 8 E 1 9 2 7 9 O A 5 8 "6 8 8 "2 L L P 9

1. E B

3 7* 2 4 H ( L T b F E ~R F V O O E TL N U R C E E T P A IP W l I ll ll

\\l( lll ,llI l lj l lI l 1 9 "8 "5 6 9 9 3 7 3 "8 1 6 1 9 8 L 8 1 l 2 0 9 1 8 8 0 1 = l!j 2 = T I N "8 U 8 1 2 3 S 1 J E E p E 0 I 1 T IC D AU Q R "9 0N

5. M "2

O I 2 F 1 8 4 S 6 1 9 C B B B L S f4 8 H E C L A U) LEM D S N 2 OL NP 2 = E ML N A R E E I H M C R KC P 1 L U C 7 6 7 ME G A 8 2 5 9 K C 0P C C RU R8 2 I 8

0. Y OF F

3 "2 3T F F TK AC O( 9 LA 4 = T~ PR = ll + N' E M E "0 G 4 7 8 4 N 5 H A A D A 5 R 9 D R ET T A A TP I L = N N E A I E M G A M R A E P TI R NU O N "0 OQT E CES G 5 3 5 6 3 5 H A A D 9 l 4 lt 5 = ) SE W C H O A L L T E P ^ l B i 2 O "N '4 "6 "5 N ( 0 L 6

5. I F E 0

2 1M F V 1 9 O E -TL U R C E E T P A !P W l}}

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