ML20063B032

Summary of Info in Support of Increasing Spent Fuel Storage Capacity of North Anna Power Station Units 1 & 2

ML20063B032
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Site:	North Anna
Issue date:	08/31/1982
From:	VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
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ML20063B001	List: ML20063A999 ML20063B012 ML20063B032
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NUDOCS 8208250162
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SUMMARY

OF INFORMATION IN SUPPORT OF INCREASING THE l SPENT FUEL STORAGE CAPACITY l !O AT l NORTH ANNA POWER STATION UNITS 1 AND 2 .O 0 DOCKET NOS. 50-338 50-339 LICENSE NOS. NPF-4 NPF-7 O O AUGUST 1982 VIRGINIA ELECTRIC AND POWER COMPANY O O P

O AB 0F COMENTS

O Page 1.0 I'ntroduction....................................................... 4 O

2.0 Reasons For The Modification....................................... 8 3.0 Proposed Action and Schedule...................................... 11 4.0 Alternatives To The Proposed Modification......................... 12 0 5.0 Existing Facilities............................................... 24 6.0 Design Of Neutron Absorber Spent Fuel Storage Racks................................................... 51 7.0 Analysis Of Existing Facilities And Systems Affected O By The Proposed Modifftstions................................... 79 8.0 Neutron Absorber Spent Fuel Rack Installation and Removal of High Density Racks...................................... 86 9.0 Analysis Of The Safety Implications Of The Proposed O Modification.................................................... 87 10.0 Environmental Impact Of The Proposed Modification................. 94 11.0 Conclusions....................................................... 99 O

O O

.o 2 0

O Preface e The purpose of this document is to summarize the proposed changes to the spent fuel storage pool at the North Anna Power Station Unit Nos. I and 2, 3 associated with increasing the storage capacity of the pool. It is intended that the document will describe the proposed modification, as well as describe the installed systems which may be affected by the modification. Operating 3 experience and data is presented where appropriate to substantiate the conclusions made regarding system perfomance and environmental impact. O O O O O ( 3 g i

'O l.0 INTRODUCTION 4

O Virginia Electric and Power Company's North Anna Power Station, Unit Nos.1 and 2, were issued License Nos. NPF-4 on April 1,1978, and o

NPF-7 on August 21, 1980, respectively. The units commenced commercial operation on August 21, 1978, and December 14, 1980, respectively. Both units were licensed to operate at a full power of 2775 megawatts thermal o with a net generation of about 907 megawatts electrical. North Anna Unit No. 1 is presently refueling prior to its fourth fuel cycle, and Unit No. 2 is presently operating in its second fuel cycle. As of o June 1,1982, no spent fuel has been shipped from the site resulting in 237 fuel assemblies presently in storage in the spent fuel pool. O The initial licensed capacity of the North Anna Unit Nos.1 and 2 spent fuel pool was 400 fuel assemblies. In 1979 the fuel pool was reracked with High Density Spent Fuel Racks of the " flux trap" design o with a storage capacity of 966 fuel assemblies. This reracking allows the two North Anna units to operate until 1989 without the loss of Full Core Discharge (FCD) capability. This storage capacity allows North 'O Anna Unit Nos. 1 and 2 to operate until 1991 and 1990 respectively, after which they will be forced to shutdown due to a lack of the necessary storage space to support refueling. O Vepco's policy, as well as its legal duty, is to supply reliable i electric service to its customers at reasonable rates. Fulfilling this lO duty requires that the Company's nuclear power stations operate reliably and continuously. To assure continued operation, Vepco must maintain the ability to discharge spent fuel whenever necessary. 'O 4

) Whenever Vepco refuels one of its reactors, replacing about forty ) percent of the fuel assemblies in the reactor core, it must have room to store the spent fuel that is removed from the reactor. For each reactor these refuelings occur at intervals of approximately every 18 months. ) In addition, Vepco believes that if it is to fulfill its public service obligation it must maintain the ability to discharge the full core in a particular reactor at any time. This " full core discharge" capability ) is essential whenever inspections or repairs necessary for continued operation require the offloading of the entire core from the reactor. f ) Yepco has evaluated the available alternatives for assuring that l the above needs and objectives are realized. The decision to install additional spent fuel storage capacity provides the most favorable ) solution considering commitment of available resources. The modification which is planned for the North Anna Unit Nos. 1 ) and 2 spent fuel pool (to increase the storage capacity) is to replace the presently installed high density fuel racks with neutron absorber fuel racks. The neutron absorber fuel racks have a reduced ) center-to-center spacing of fuel assemblies but still maintain subcriticality under all conditions. The planned modification will result in a storage capacity of 1737 fuel assemblies. ) In evaluating its other nuclear facilities Vepco has found that no additional fuel over its present licensed capacity may be stored in the 3 Surry Unit Nos.1 and 2 spent fuel pool without exceeding structural b 5 t

O design criteria. An option which Vepco is actively pursuing for Surry O is to store spent fuel from Surry at North Anna. On July 13, 1982, Vepco made application to the NRC for a license amendment to permit the storage of up to 500 spent fuel assemblies from Surry Units 1 and 2 in 'O the North Anna 1 and 2 spent fuel pool. Pursual of this option increases the need for additional spent fuel storage at North Anna Unit Nos. 1 and 2. O i The addition of neutron absorber spent fuel racks in the North Anna 1 and 2 spent fuel pool will provide adequate storage for North Anna

O Unit Nos. I and 2 until 1998.

If 500 Surry spent fuel assemblies are stored at North Anna, this solution will provide for adequate spent fuel storage capacity for North Anna Unit Nos. 1 and 2 until 1993. .O 4 The " proposed modification" for the purpose of this summary is defined as:

O The installation of neutron absorber spent fuel storage racks in the North Anna Unit Nos. 1 and 2 spent fuel pool which would O

increase the spent fuel storage capacity from the present 966 assemblies to 1737 assemblies while maintaining subcriticality under all postulated conditions.

O The license change required would be to modify the Technical Specifications such that:

O O

6

O

1. The maximum allowable spent fuel storage capacity design feature 3

at North Anna Unit Nos. I and 2 is increased from 966 to 1737 fuel assemblies. 3

2. The center-to-center design feature at North Anna Unit Nos. I and 2 is decreased from 14 inches to 10 9/16 inches while maintaining K,7f less than or equal to 0.95.

O In summary, to avoid future unit shutdowns due to lack of spent fuel storage space, given the uncertainty of fuel reprocessing or a O permanent solution to the spent fuel problem, an increase in the spent fuel storage capacity at the North Anna Power Station Unit Nos.1 and 2, is necessary. O O O O O 7 .g

O I 2.0 REASONS FOR THE MODIFICATION .O The basic reason for the planned modification is to provide additional spent fuel storage capacity to prevent future unit shutdowns O of Vepco's nuclear units due to a lack of spent fuel storage space.

2.1 BACKGROUND

20 The original storage capacity of the North Anna Unit Nos. 1 and 2 spent fuel pool was based upon providing storage for two and one half O cores of the two units combined. This design capacity was predicated on being able to ship spent fuel offsite to a reprocessing facility. Due j to delays in the early 1970's of bringing reprocessing facilities on O line, Vepco undertook to increase the storage capacity of the North Anna Unit Nos.1 and 2 spent fuel pool from 400 storage locations to 966 storage locations. The increased capacity was installed in 1979. 'O In April 1977, President Carter's concern over the spread of nuclear weapons and the possible threat of weapons-grade plutonium being O stolen from commercial reprocessing plants caused him to " indefinitely defer" commercial reprocessing of spent nuclear fuel in the United States. He proposed instead to provide interim Federal storage capacity 'O for spent nuclear fuel until a permanent Federal repository became available. O The Department of Energy announced in October 1977 that it would provide interim storage space for utilities' spent fuel by buying or O g

O leasing some of the fuel pool storage space at the inoperative commercial reprocessing facilities or by building new "away-from-reactor" storage, g facilities. Congress, however, failed to provide the necessary authorizing legislation or funding. I O When Congress had not yet acted after a year's time in response to i the Administration's proposal to provide interim Federal storage, Vepco fa ed the possibility that neither reprocessing nor Federal interim .O storage would be available in time to meet Vepco's needs. Between 1979 and late 1981, while the nuclear industry continued to

g press for government action to provide interim Federal storage capacity for spent nuclear fuel, Vepco investigated other possible storage alternatives.

O These alternatives included building another wet storage facility at Surry, North Anna or elsewhere; reracking the North Anna spent fuel g I pool a second time; dry storage techniques; and redesigning and expanding the storage facility at the North Anna Unit No. 3 which is currently under construction. These alternatives and others are O discussed in more detail in Section 4 of this summary, i In March 1981, the Reagan Administration directed that utilities O take care of their own interim spent fuel storage problems and eliminated any funding for interim spent fuel storage from the Federal

g budget.

In October 1981, the President also lifted the ban on commercial reprocessing. ~O 9

D The lifting of the commercial reprocessing ban does nothing to diminish Vepco's immediate storage problem. No company has expressed an sJ interest in building or operating a reprocessing plant, and, even if it did, it would likely take a minimum of 10 years to put the plant into operation. 3 Currently, Congress is considering several legislative proposals which would provide some limited interim storage capacity for commercial ) spent fuel on an emergency basis. While Vepco supports these legislative efforts, it is impossible now to say what exact form this legislation might take, or even when or if it will be enacted. These 3 proposals face difficult political obstacles and some of them would deny federal storage space to utilities that can resolve their own storage problems. Therefore Vepco cannot rely on the availability of federal 3 interim storage space. In light of the uncertainties surrounding reprocessing and the g unavailability of federal interim storage space, and in light of the long lead times associated with a permanent repository, Vepco must now move swiftly to solve its near term spent fuel storage space problem g using its own facilities. Vepco believes that the most prudent course of action is to rerack the North Anna Unit Nos. I and 2 spent fuel pool with neutron absorber spent fuel racks. 3 D 10 J

O 3.0 PROPOSED ACTION AND SCHEDULE ,0 The proposed action will increase the existing spent fuel storage capacity by approximately 80 percent to a storage capacity of 1737

O storage locations. This will be accomplished with racks which can store spent fuel assemblies closer together by use of a neutron absorber material. The racks are therefore referred to as neutron absorber spent

.O fuel storage racks. The Virginia Electric and Power Company has contracted with Nuclear 10 Energy Services, Inc. of Danbury, Connecticut to design, engineer, and manufacture the new spent fuel storage racks. The installer, if other than Vepco, has not been selected at this time. Vepco personnel are

O managing the overall project.

The following are key dates associated with the proposed .O modification. Date Description LO March 1982 Contract for the Purchase of Neutron Absorber Spent Fuel Racks August 1982 Submit Application to NRC O January 1984 Commence Rack Installation April 1984 Complete Rack Installation O 11 O

lO 4.0 ALTERNATIVES TO THE PROPOSED MODIFICATION O The proposed modification has been chosen after an evaluation of the paisible alternatives to ameliorate a shortage of spent fuel storage J apacity. The following af ternatives were considered. O 1. Expand the spent fuel storage capacity at North Anna Power Station Unit Nos. 1 and 2. O 2. Ship spent fuel from North Anna to Surry. .O 3. Ship spent fuel from North Anna Unit Nos.1 and 2 to the North Anna Unit No. 3 fuel storage pool.

O 4.

Build a new separate spent fuel storage pool. 1

O 5.

Utilize dry casks or drywell storage techniques. 6. Ship spent fuel from North Anna to a reprocessing facility. .O 7. Ship fuel to Federal interim storage facilities. 8. Ship spent fuel to a Federal permanent repository. 10 9. Improve fuel utilization. .O

10.. Operate North Anna at a reduced power level to reduce spent fuel production.

O 12

O

11. Cease reactor operations when spent fuel capacity is expended.

D

12. Ship spent fuel to other utility fuel storage pools.

Each of the above alternatives is discussed below. 4.1 Increase The Storage Capacity Of The North Anna Unit Nos. 1 And 2 Spent Fuel Pool The storage capacity of the spent fuel pool can be increased by replacing the existing racks with racks of reduced center-to-center g spacing which utilize a neutron absorber material resulting in an increased storage capacity of 1737 fuel assemblies. This modification will require no physical changes to the pool g structure. The racks will be manufactured offsite for installation, thereby reducing disruption of normal station operations. The proposed modification will not alter the external physical geometry of the spent fuel pool or require modifications to the 3 spent fuel pool cooling or purification systems. The proposed modification does not affect, in any manner, the quantity of uranium fuel utilized in the reactor over the anticipated operating life of the facility and thus in no way affects the spent fuel generation. The rate of spent fuel generation and the total g quantity of spent fuel generated during the anticipated operating lifetime of the station and stored in the spent fuel pool remains O 13

1 3 unchanged as a result of the proposed expansion. The modification 3 will increase the number of spent fuel assemblies that can be stored in the spent fuel pool. 3 The total cost of the spent fuel racks is approximately 3.0 million dollars, exclusive of installation. Installation of the spent fuel racks is estimated to be approximately $400,000 resulting in an 3 estimated total cost of the modification to be 3.4 million dollars. No additional operating costs will be incurred as a result of the modification. D Based on economics, operational considerations and existing conditions, the expansion of the spent fuel storage capacity of the g existing spent fuel pool provides the most feasible means of ameliorating the shortage of storage capacity. 3 4.2 shipment Of Spent Fuel To Surry Power Station Surry Power Station Unit Nos. I and 2 is located about 125 miles g southeast of North Anna. Shipment of spent fuel from North Anna to Surry could be carried out; however, it has several distinct disadvantages. For example, the Surry 1 and 2 spent fuel pool g presently has a storage capability of 1044 spent fuel assemblies. As of August 1982 there are presently 644 spent fuel assemblies stored in the pool. Based on current projections the Surry 1 and 2 O Spent fuel Pool will lose full core discharge capability in the fall of 1984 and the two units would have to be shutdown in 1987 due to insufficient spent fuel storage capability to conduct a 3 normal refueling. 14

's i O ~ Shipment of spent fuel from North Anna to Surry would only cause an g even earlier loss of storage capability and shutdown of the two Surry units. -g-The reverse of this option, that is, the transfer and storage of spent fuel from Surry to North' Anna is presently under consideration as a means to ameliorate the problem of dwindling

g spent fuel storage capacity of,Surry Power Station Unit Nos.1 and r

2. _s, ~ g Although shipment of fuel from North Anna to Surry is v'f able in the. } very near term, it is not advantageous for'the reasons cited. ~ 1 g 4.3 Shipment Of Spent Fuel From North Anna 1 And [To North Anna Unit 31 Vepco's fifth nuclear power unit, North ' A' na Unit No. 3, is now n O under design and construction and is expected to be completed in It is being deitgned with an expanded spent fuel pool to ' 1989. provide more than adequate spent fuel storage capacity.for the g entire operating ~1ife of North Anna Unit -No. 3. The additional ~ storage capacity will be available to accept spent fuel from. Yepco's other nuclear units after the proposed neutron absorber g racks for North Anna 1 and 2 are filled and until connercial reprocessing or a Federal repository becomes available. b" h-s s / 15 O

O i However, it is not Vepco's intent to complete the expanded North 9 Anna 3 pool without also expanding the capacity of the North Anna 1 and 2 pool. It is more attractive economically to increase the storage capability of the North Anna 1 and 2 spent fuel pool as !g much as possible prior to completion of the expanded North Anna 3 spent fuel pool for the following reasons.

g 1.

The shipment of 771 spent fuel assemblies (the proposed increase for North Anna 1 and 2) from North Anna 1 and 2 to North Anna 3 would cost approximately $1.7 million in 1982 lC dollars. e 2. The cost of the proposed modification of 3.4 million

g dollars on

a. per storage location cost of additional j

storage is approximately $4350. The cost of expanding the North Anna 3 spent fuel pool on a per storage location cost O of additional storage is approximately $8600. i 3. Perfonning the proposed modification to the North Anna 1 0i and 2 spent fuel pool is therefore approximately 60% less expensive for these 771 spent fuel assemblies than providing this additional storage capacity at North Anna ~ o, Unit 3 plus the cost of shipping the spent fuel from North Anna 1 and 2 to North Anna 3 of $1.7 million. 'O [ g 16

O The total additional cost of shipping 771 spent fuel assemblies from North Anna 1 and 2 and storing them in the North Anna 3 pool g would be $5 million more than the cost of the proposed modification. Therefore, the proposed modification is more attractive than shipping the 771 assemblies to North Anna 3 on an 3 economic basis. 4.4 Build A New Independent Spent Fuel Storage Pool 3 Additional spent fuel storage capacity could be made avafiable by building a new spent fuel storage pool, either at North Anna or on 3 another site. 1.ike existing facilities the structure would be a pool built of reinforced concrete with a stainless steel liner with the same type of support facilities as presently existing at the 3 North Anna 1 and 2 spent fuel pool. Present cost estimates for such a facility are in the range of approximately 100-125 million dollars. In a'dition it would take approximately 8 years to ,~) design, license, and construct such a facility. 1his alternative is unacceptable in comparison to the proposed 3 modification on both an economic and a timeliness basis. 4.5 Utilize Dry Casks or Drywell Storage Techniques g In the dry cask system, spent fuel would be stored in steel and lead, or cast iron casks similar to those now used for the shipment g 17 3

"O of spent fuel assemblies. These casks would be stored in o facilities at the plant site. It is estimated that it would cost approximately $19 million in 1982 dollars to store 771 spent fuel assemblies (the proposed increase for North Anna 1 and 2). Dry o storage casks have not been used by any domestic utilities, but have been used in Europe. ,0 Currently, no dry cask storage systems have been licensed for use in the United States by the NRC but at least one cask type is being reviewed by the NRC. O In the drywell system, spent fuel would be stored in steel canisters buried underground.

Cost, schedule, and long term iO technical feasibility need to be better understood and demonstrated before this option can be further considered.

O The dry cask alternative is inferior to the proposed action, both in economic terms and because of the licensing uncertainties involved with dry cask storage. It is therefore less certain that 'o this alternative would be available in a timely manner to meet the spent fuel storage needs at North Anna Units 1 and 2. lO s.6 Ship Spent Fuel to a Reprocessing Facility There is no operating commercial reprocessing facility in the O United States. None are licensed; none are in licensing proceedings. Neither government nor private enterprise has O 18

\\ .O expressed any significant interest in building or operating a O reprocessing facility. There

are, in

short, no domestic reprocessing facilities available and no prospect that a'ny such facilities will be available in the foreseeable future.

O There are presently in operation foreign reprocessing facilities in the United Kingdom, France, Germany, and Japan.

However, the shipment of any domestic spent fuel to any foreign country for O

storage or disposal is considered to be in conflict with U. 3. nonproliferation objectives. Section 82 of the Atomic Energy Act

O f 1954, as amended by the Nuclear Nonproliferation Policy Act of 1978, states that the NRC is authorized to distribute by-product material (such as spent fuel) provided that such distribution would n t be inimical to the connon defense and security of the United O

States. Under current administration policies, such distribution would almost definitely be considered inimical to the United States. .O Due to conflicts with government policies, shipping spent fuel to a foreign reprocessor is not a viable alternative. O Reprocessing, then, is not a viable alternative. l 4.7 Ship Spent Fuel to Federal Storage Facilities O Of all the potential alternatives, Yepco favors interim storage in lO a federal storage facility. This is precisely what the Federal l l 19 O l

O government promised in 1977 to provide. But authorizing legislation has not been adopted, and the current administration .O has now reversed the earlier commitment. Despite the position of the current Administration that utilities O should be responsible for their own interim spent fuel storage, there are several legislative proposals before the Congress that could make Federal interim storage available. It is unlikely, 'O however, that these provisions, if passed, will provide a near-term solution to Vepco's storage needs. .O First, any legislative proposal to provide interim federal storage will face serious political obstacles during Congressional consideration. Second, even if the proposals that currently appear .g to have the best chance of passing should pass, it is unlikely that they would meet Yepco's spent fuel storage needs in the near term. While these proposals vary in several respects, they would require O a showing by a utility desiring to use federal interim storage space that it was unable to solve its own fuel storage problem by transshipping between its own reactors or building a new storage O facility at an existing power station. Thus, the present prospect is that even if legislation is adopted, it is not likely to provide any relief for Vepco. Third, even if the legislation were adopted O without any requirement that Vepco exhaust the possibility of transshipment or building a new facility, there simply can be no certainty today that the Federal Government can make the storage

g O

20 l

-O space available by the time Vepco will need it. Thus, this alternative is inferior to the proposed action on the ground that g its capacity to solve Vepco's problem is far less certain than that of the proposed action. .O Ship Spent Fuel to a Federal Permanent Repository 4.8 Current legislative proposals in both houses of Congress would O establish schedules for the Federal Government for selecting repository sites and authorizing their construction. These in the very Proposals all contemplate construction authorizations O late 1980's or early 1990's. If such legislation were adopted, no Federal permanent repository could be expected to be in place prior to the mid-to-late 1990's. This is thus not a viable alternative O for solving Vepco's near-term storage needs. 4.9 Improve Fuel Utilization

O Vepco is presently participating in a Department of Energy program to extend the allowable burnup of fuel assemblies at both its North O

Anna 1 and 2 and Surry 1 and 2 Power Stations. In this program, increased energy is derived from each individual fuel assembly, decreasing the number of spent fuel assemblies discharged. .g However, the impact of this program on near-term fuel storage requirements is negligible and the present problem of lack of spent fuel storage capacity remains. iO l 21 lO

O. 4.10 Operate North Anna 1 and 2 at a Reduced Power Level to Reduce Spent Fuel Production j The amount of spent fuel generated could be reduced by lowering the unit's output, thereby extending the life of the fuel. The obvious disadvantage in this alternative is that the unit could not be operated to the extent possible and the amount of electricity generated would be reduced. This alternative is not viable because I it does not effectively use the resources available and would I result in significant economic penalties to Vepco's customers I through increased dependence on more costly coal and oil fired .O generation of electricity. l 4.11 Cease Reactor Operations When Spent Fuel Capacity is Expent'ed lO l Assuming that.the storage capacity of the pool remains the saae and no offsite shipments are made, the units would have to be shutdown

O i

in early 1991 and late 1990 respectively. If spent fuel is shipped I from Surry to North Anna, with North Anna 1 and 2's storage i capacity remaining the same, all four of Vepco's nuclear units would be shutdown in 1989. This is clearly not a viable or practical alternative. The generation provided by the nuclear units is necessary to supply customer needs at the lowest possible cost. Economic studies show that in a matter of several days, the replacement cost of nuclear unit generation would exceed that of i the proposed modification. l l

O 22

i f ,0 4.12 Ship Spent Fuel To Other Utilities 9 In 1979, Vepco inquired of 10 neighboring utilities as to whether they would be willing to store spent fuel from Surry in their spent fuel storage pools. The utilities that responded uniformly .O rejected this proposal. That is understandable, since many of these utilities face spent fuel storage space shortages of their w"* O As with several of the other alternatives reviewed, this one simply 3 is not realistic. 4.13 Summary i

O Based on a review of the available alternatives, the proposed modification is the best available and has been selected for O

impi mentation.

O O

.O 23 0

D 4

EXISTING FACILITIES 5.0 The spent fuel pool for North Anna Unit Nos.1 and 2 is common

O units and presently has a storage capacity of 966 fuel assemblies.

JO 5.1 Fuel Building The Fuel Building is a Class I seismic structure and is supported The arrangement of the lO by a reinforced concrete mat on bedrock. Fuel Building is shown in Figures 5-1 and 5-2. The Fuel Building is designed to handle new fuel, spent fuel and a .O The building is sized for spent fuel cask and related equipment. The structure is approximately 136 ft-0 in. Iong by 41 two units. The top of the foundation mat is approximately 21 g f t-0 in. wide. The main roof area is approximately 48 ft-0 ft-8 in. below grade. in. above finish grade, with the roof of the trolley bay The spent fuel storage area has clear approximately 20 ft higher. in, wide by 72 f t-6 in. inside dimensions approximately 29 ft-3 Narrow canals connect the spent fuel 1 ng by 42 f t-6 in. deep. New fuel racks O storage areas to the Units 1 and 2 containments. are mounted in the new fuel area above the slab at E1 The lowest This area will be accessible to the platform crane. The level slab supports the fuel pool coolers and cooling pumps. g is stored vertically in stainless steel racks, which ment fuel pro 'ide separation to preclude criticality. lg 1 24 ~O . 1

O The spent fuel pool contains a 3 ft-6 in, reinforced concrete wall, i

O extending from the foundation mat to the top of the pool, which separates the spent fuel cask handling area from the spent fuel racks and is designed to prevent a spent fuel cask from impacting

!O the spent fuel storage racks. The Fuel Building structure is supported by a concrete mat founded n bedrock. The walls of the spent fuel storage pool are 6 ft

O thick reinforced concrete for biological shielding.

Exterior and interior walls enclosing the fuel pool coolers are of concrete for lO missile shielding. Exterior walls above the concrete work are covered with insulated metal siding on structural steel framing. A large tee-shaped rolling steel door permits moving the trolley and ,g spent fuel cask through the door opening. Another similar rolling steel door is provided for bringing new fuel into the structure. Passage doors are hollow metal type. O The superstructure walls and the roof are supported on steel i framing. The roof is covered with insulated metal decking and O asphalt and gravel roofing. Intermediate platforms in the new fuel area are concrete slabs on steel framing. Stairs have steel framing with grating treads and grating platforms. O Movable gates between the spent fuel pool and each canal permit dewatering the canals for access to the fuel transfer mechanisms g without dewatering the entire pool. 25 g

O Rails embedded in the concrete are provided for operation of the

!O motor driven platform with hoists for transferring fuel. The spent fuel pool is lined with stainless steel plate, a minimum 0. of 1/4 inch thick, and is designed for the underwater storage of spent fuel assemblies and control rod, burnable poison rod and thimble plugging assemblies. -O The fuel building is also described in the North Anna 1 and 2 FSAR ) Sections 9.1.2 and 3.8.1. l i-

O-5.2 Spent Fuel Storage Racks iO The presently installed spent fuel storage racks provide a storage capacity of 966 spent fuel assemblies.

The spent fuel storage racks are classified seismic Category I and are designed to 10 withstand the effects of the Design Basis Earthquake (DBE) and yet remain functional and maintain subcriticality. Details of the seismic design criteria are presented in Section 3.7 of the North

O Anna 1 and 2 FSAR.

j The presently installed spent fuel storage racks consist of square stainless steel cells of 1/8 in. thick Type 304 austenitic lO stainless steel. They are spaced at 14 in. center-to-center by Type 304 stainless steel plates. The plates, which are also 1/8 lg in thick, are welded to the sides of the square storage cells at four elevations. The cells are flared at the top to permit easy i storage and retrieval of the spent fuel assemblies and to be O compatible with the fuel handling equipment. The base of the 26

3 spent fuel rack serves to support the weight of the spent fuei 3 assemblie and to distribute the load to the spent fuel pool floor. The base, which contains an opening at each fuel assembly location to permit coolant flow, accommodates the fuel assembly 3 bottom nozzle. Natural circulation of pool water flows down between the storage cells and up through the bottom nozzle to remove decay heat. The storage cells are designed to provide 3 lateral support for the Westinghouse 15 x 15 or 17 x 17 assembly array design. 3 Three different rack cell arrays are used to maximize use of the available storage space in the pool. The three arrays are: 3 No. Of Arrays Type of Array Capacity of Array 18 6 x 6 racks (36 assemblies) 7 6 x 7 racks (42 assemblies) 3 1 6 x 4 rack (24 assemblies) The spent fuel racks have been conservatively designed to be able 3 to store either Surry or North Anna fuel assembifes. The present spent fuel racks are also described in the North Anna 1 and 2 FSAR Section 9.1.2. D 5.3 Fuel Pool Cooling and Purification System D The spent fuel pool is equipped with a spent fuel pool cooling system to remove decay heat and a purification system for maintaining fuel pool water quality. These systems are shown D schematically in figure 5-3. 27

r O { 5.3.1 Design Basis

O

~ The Fuel Pool Cooling and Refueling Purification System is designed to: to 1. Remove the residual heat produced by one-third of an irradiated core 150 hr after reactor shutdown r O while maintaining the spent fuel pool water j temperature at or below 140 F with two fuel 0 pool coolers and one associated pump. !O 1 2. Remove the residual heat produced by one irradiated core 150 hr after shutdown plus O ne-third irradiated core 45 days after shutdown, r while maintaining the spent fuel pool water at a 4 0 temperature of 170 F or less with two fuel pool Coolers and one pump. 3. Remove soluble and particulate impurities from

O the water in the spent fuel pool, either reactor refueling cavity and refueling water storage tank, to maintain the cavity water optically O

clear and radiation levels within acceptable limits. .O 28 O

!O ~ 4. Provide a path for make-up and boration of the i -O spent fuel pool water from both refueling water storage tanks. 10 5. Maintain a minimum water level in the spent fuel pool to provide adequate radiation protection from irradiated fuel. O 5.3.2 System Description

O The portion of the Fuel Pool Cooling and Refueling Purification System used to cool the spent fuel pool water has two shell and tube coolers and two circulating pumps

O located in the Fuel Building.

The coolers and pumps are arranged for cross-connected operation, if necessary. The coolers are cooled by component cooling water, with service O water available as an emergency supply of cooling water. All fuel pool piping penetrations are located so that at 'O least 24 ft.1 in. of water would remain above the active portions of the spent fuel assemblies stored in the pool even if the water should drain through the penetrations,

O thus ensuring adequate shielding for the spent fuel assemblies.

4 g The system also includes three refueling purification pumps, two filters, and an ion exchanger.

O 29

D Spent fuel pool water can be purified, if required, by pumping a portion of the fuel pool cooling loop flow g through the refueling purification filters and ion exchanger with the refueling ptrification pumps. The refueling purification pumps take suction from the cooling g system loop flow downstream of the coolers and return it to the pool. The filters and ion exchanger are operated in series as a filter-ion exchanger-filter arrangement or 3 either of the filters may be used alone. Flow through the ion exchanger and filters provides adequate purification of the water to permit access to the working area and to maintain optical clarity of the pool water. The refueling purification pumps can be run to purify the pool water independently of the cooling pump operation. The water surface of the spent fuel pool is maintained free of floating material by two permanently installed skimers connected to the suction of the spent fuel pool cooling pumps. The fuel pool skimmers are also provided with a pump which allows the skimmers to be operated when the fuel pool cooling pumps are not in operation. Make-up water, borated and unborated, is supplied to the spent fuel pool and the refueling water storage tanks from the boric acid blender in either of the Chemical and Volume Control Systems. Make-up water can also be supplied to the O spent fuel pool from the Fire Protection System. 30

!O. All parts of equipment and piping in contact with water iO which has been borated to refueling water concentration are constructed of austenitic stainless steel. 10 The design data for the Fuel Pool Cooling and Refueling Purification System components is given in Table 5-1. i IO The Fuel Pool Cooling and Refueling Purification System is also described in the North Anna 1 and 2 FSAR Section 9.1.3. 1 !O L 0 !O O. iO-j O f

O 31

0-TABLE 5-1 9 FUEL POOL COOLING AND REFUELING PURIFICATION SYSTEM DESIGN DATA O Fuel Pool Coolers o Number 2 O Design duty, Btu per hr, each 56,800,000 0 (with tube inlet 210 F and shell inlet 105 ) O Shell Tube O Fluid Flowing Component cooling Fuel pool water or service water O Design pressure, psig 150 100 0 Design temperature,0F 150 212 O l l

1. 3 32 1

O 4 f TABLE 5-1 (Cont'd) !O Fuel Pool Coolers (Cont'd) i Operating pressure, g j psig 110 45 i }' Material Carbon steel Stainless Steel, !O. j Type 304 Design code ASME VIII, Div. ASME VIII, Div. O I 1-1968 1-1968 I

O j

Spent Fuel Pool Cooling Pumps I Number 2 g Type Horizontal centrifugal l0 Motor horsepower, hp 100 I 1 Seals Mechanical g Capacity, gpm, each 2,750 i O Head at rated capacity, ft 80 .i i .O 33

20 1 TABLE 5-1 (Cont'd) IO Spent Fuel Pool Cooling Pumps (Cont'd) 125 Design pressure, psig 250 Design Temperature, F lO Materials Stainless steel, Pump casing O Type 316 Stainless steel, Shaft O Type 316 Stainless steel, ge er !O Type 316 Refueling Purification Pumps O 3 Number O Vertical centrifugal Type 20 Motor horsepower, hp lg 400 Pump capacity, gpm, each 'O 34

!O TABLE 5-1 (Cont'd) iO Refueling Purification Pumps (Cont'd) Seals !O Mechanical i Head at rated capacity, ft 99 !O Design pressure, psig 185 O Design temperature, F 200 j. Materials !O Punp casing Stainless steel, Type 316 l0 i Shaft Stainless steel, i Type 316

O Impeller Stainless steel, 1

Type 316 i ^O lO f a g 35

i J TABLE 5-1 (Cont'd) e Refueling Purification Filters 2 Number 3 3 Retention size, microns O 400/440 Filter element capacity, gpa at 5 psig,' normal / max 0 Stainless steel, Material Type 304 3 150 Design pressure, psig 250 Design temperature,0F g Refueling Purification Ion Exchanger 3 1 Number i O 45 Active resin volume, cu ft 200 Design pressure, psig 3 250 Design temperature, F O 36

O TABLE 5-1 (Cont'd) D Refueling Purification Ion Exchanger (Cont'd) Ion exchange resin 50/50 anion-cation ,J Material Stainless steel, Type 316L ,J Design flow rate, gpm 200 Design code ASE VIII, Div.1-1968 O Skimmer Assemblies Number 4 - 2 in spent fuel pool, 1 in each reactor cavity Debris Basket 1/8" x 1/4" openings ,J Design temperature,UF 210 D Flow rate, gpm (approx.), each 25 (min) to 55 (max) l 0 O 37

O 5.4 Fuel Building Ventilation System e The fuel building is equipped with a ventilation system to provide high-efficiency filtration, heating to inhibit the buildup of condensation, and excess exhaust flow to maintain a negative g pressure in the building to prevent outward leakage. The Fuel Buiding Ventilation System is also described in the North Anna 1 and 2 FSAR Section 9.4.5. g 5.4.1 Description 3 The Fuel Building Ventilation System has two supply fans, one to serve the spent fuel pool area and one for the 3 remote equipment space at E1. 249.33. Both take suction from a common plenum fitted with a combination roll and high-efficiency filter (95 percent atmospheric dust spot 3 efficiency) and steam coils for air tempering and space heating. The exhaust fans discharge through the ventilation vent and are arranged for selective bypass 3 through the auxiliary building filter bank. The area of the remote equipment room subject to radioactive contamination is exhausted by a branch from the 3 Decontamination Building Exhaust System. The design provides (1) sufficient air at a temperature that will inhibit condensation on the overhead structure to 3 avoid drippage into the pool, (2) high-efficiency supply air filtration (3) supply air distribution to avoid 3 38

O rippling the surface. The dual exhaust combined with two-speed supply fan arrangement provides step capacity control and protection against a single failure. The exhaust is continuously vented through the ventilation 3 vent, with the capability to bypass through the auxiliary building iodine filter bank. The exhaust is filtered continuously during irradiated fuel-handling operations to g prevent the spread of any possible airborne contamination through the exhaust air system. 3 The Fuel Bufiding exhaust also discharges air entering the Fuel Building from the tunnel between the Fuel Building and the Waste Disposal Building. O 5.5 Fuel Handling Shielding O Fuel handling shielding is designed to facilitate the transfer of spent fuel assemblies from the reactor into the spent fuel storage racks and between the spent fuel storage racks and spent fuel 3 shipping casks. It is designed to protect personnel against the radiation emitted from the spent fuel and control rod, burnable poison and thimble plugging assemblies. O The spent fuel pool in the Fuel Building is permanently flooded to provide approximately 9 feet of water above a fuel assembly while g it is being transferred. Under these conditions, the dose rate is less than 30 mrem per hour at the water surface. The water height O 39

D above stored fuel assemblies is a minimum of 26 feet. The sides 3 of the spent fuel pool, three of which also form part of the fuel building exterior walls, are 6 foot thick concrete to ensure a dose rate of no more than 2.5 mrea per hour outside the butiding. D Fuel Building shielding is also discussed in the North Anna 1 and 2 FSAR Section 12.1.2.5. G 5.6 Fuel Building Instrumentation 7 Instrumentation provided gives local indication in the Fuel Building and the Auxiliary Building and remote indications and alarms in the main control room. Unit I control board indication and alams include: 1. Fuel pool temperature indication S 2. Spent fuel pool temperature alarms at 140 F and 0 170*F S 3. Spent fuel pool high/ low water level with the low-level alarm 6 in, below normal water level (E1. 3 289.33) 4. Start /stop switch for spent fuel pool cooling pumps 3 with run indication on both Units 1 and 2 main control boards 3 40

J 5. High differential pressure alarm for the refueling g purification filters Local indications include various flows, temperatures, pressures, and differential pressures. The system instrumentation, including the spent fuel pool level 3 and temperature instrumentation, are calibrated on a periodic basis. Instrumentation associated with the spent fuel pool is also described in the North Anna 1 and 2 FSAR Section 9.1.3.5. O 5.7 Spent Fuel Pool Water Leakage Control 3 No means exist for completely draining the spent fuel pool using installed systems and equipment. The water level could be lowered to E1. 285.75, which is 4 ft-1 in. below the normal water level 3 and 24 f t-1 in, above the fuel by incorrect operation of, or a failure in, the Fuel Pool Cooling and Refueling Purification System. In this instance, an adequate water level would exist 9 over the fuel to provide for cooling and radiation protection. The spent fuel pool level could be lowered to E1. 264.33 which is 9 2 ft-8 in above the stored fuel during refueling, by incorrect operation of the reactor cavity drain and the gate valve on the fuel transfer tube. This tube connects the reactor refueling 9 cavity to the spent fuel pool. The operating procedures used during refueling ensure that the fuel transfer tube gate valve is g 41

l closed before draining of the reactor refueling cavity commences. In addition, the procedures require placing the bolted blank 3 flange on the fuel transfer tube as soon as the reactor refueling cavity is drained. If the spent fuel pool level were inadvertently lowered via the reactor refueling cavity drain, this 3 condition would be detected before the level reached El. 264.33 either by excessive refueling water storage tank level, if the water level were lowered by pumping, or by a containment sump 3 level alarm if the water level were lowered by draining. The operator would then take action to correct the condition. D The solid rock foundation and reinforced concrete structure of the spent fuel pool will prevent leakage from the pool should a heavy 3 object, such as a spent fuel cask be dropped in the pool; however, the integrity of the stainless steel liner could be violated. Water could then enter the channels behind the liner seams which 3 were used for testing during construction. These channels are interconnected in four zones with a 1/2 in. line from each zone to the fuel building sump. If a leak occurs in the fuel pool liner 3 and the sump fills, the event alarm will occur on the Unit No.1 Control Board. To date no leakage h s ever been detected. j 1.eakage control for the spent fuel pool is also described in the North Anna 1 and 2 FSAR Section 9.1.3.3.3. J ) 42

O 5.8 Operating Experience O The operating history of the North Anna Power Station Unit Nos.1 and 2 spent fuel pool has been reviewed in 41ight of the fuel pool { O cooling and purification system, ventilation system, and. personnel exposures.- The purpose of the review was to confirm the satisfactory operation of th systems and to provide baseline data O for estimating the effect,the proposed modification 'may..have. In addit _fon, operating experience from the Surry Power Station Unit Nos. 1 and' 2 sphnt fuel p'ool have been added where deemed ~ O appropriate to be used in est'imat$1ng; the effects of increased spent

• fuel storage capacity as Surry.gresently has more spent fuel stored than North Anna.

r[_. O ^ l The operating experience has been hrouped into our (4) general areas for the purpose of this discussion: O l J. sPerformance of the spent fuel pool cooling and . i e I r>F purification' system .s 9 ~. / 2. Environmental conditions in the fuel buf1 ding ,/ g 3. Ventilation system 4. Radiation exposure 0 Q ,f .fa, 43 O t ^ 'y r..-

4 O 5.8.1 Performance of the Spent Fue; Pool Cooling and Purification O E System Operating experience with the North Anna fuel pool cooling o, system to date has been excellent. At the present time 237 spent fuel assemblies are being stored in the North Anna 1 and 2 spent fuel storage pool. The highest temperature O' which has been recorded in the spent fuel pool was 103 F which was after a total core offload. At other times the fuel pool temperature remains at less than 100 F o. utilizing one pump and one heat exchanger. Once every month the pump and heat exchanger in use is rotated with the other pump and heat exchanger, and a monthly periodic o test is performed on the pump. Both trains of cooling capability have not been operated simultaneously since one i / ~ cooling train maintains proper temperature. This system o has been essentially trouble free and only routine maintenance has been required. The operating experience at Surry has been similar to North ,o pv T-Anna. At the present time 644 spent fuel assemblies are. I y being stored in the Surry fuel pool. The highest o temperature which has been recorded in the spent fuel pool was 110 F which was after-a full core offload. At other times the fuel pool temperature at Surry remains below d i_ 100 F using one pump and one heat exchanger. No problems have been experienced with this system. 44 L

O The fuel pool purification system has also performed g satisfactorily and does an outstanding job of maintaining the optical clarity of the fuel pool water. The fuel pool purification system normally remains in operation O continuously, except periodically it diverts to either the Refueling Water Storage Tank (RWST) system during post refueling or the reactor cavity during refueling. O Demineralizer resins and filter changes have been performed on an average of once every 18 months. The basis for the l frequency upon which resin and filters are changed out is O 1 w dec ntaminatfor. factor, high differential pressure, or high radiation level. If the levels of decontamination factor, differential pressure, or radiation level vary from g predetermined allowable values prior to the normal 18 month change out, the filter or resin is changed. 5.8.2 Fuel Pool Environmental Conditions O Based on samples taken in the fuel pool areas at both Surry O and North Anna, isotopes have been detected in the fuel pool water in the approximate concentrations indicated in Table 5-2. O The ventilation system in the Fuel Buildings at both North Anna and Surry have maintained the levels of radioisotopes O at 8CC8Ptable levels. During normal operations i.e., g_ 45

O refueling not in progress, the gross activities above the O fuel Pool water have been maintained at levels from below 10'II to 1.0 x 10-10 detectable amounts to 1.0 x mCf/ml. A tabulation of airborne radioisotope levels in g the fuel buildings at both Surry and North Anna is presented in Table 5-3. Tritium (H-3) and Kr-85 have never been detected in either the Surry or North Anna Fuel Buildings. O 5.8.3 Radiation Exposure O The overall radiation exposure received by station workers due to the storage of spent fuel accounts for only a small O fraction of total personnel exposure for the stations (Surry and North Anna). Exposure records for the year ending December 31, 1981 showed a total personnel exposure 'O of 9.796 man-rem for Surry Power Station and 2.4 man-rem for North Anna Power Station for fuel pool related activities. These activities include fuel handling-0 (refueling), filter / resin changes and pool cleaning. Table 5-4 is a tabulation of exposure rates for key areas in the Fuel Building.

O l

l O 46

O

3 l 5.8.4 Radioactive Waste 3 The spent fuel pool operation activities generate radioactive wastes in the form of filters, demineralizer 3 resins, protective clothing, etc. which must be shipped offsite for disposal. The estimate of solid radioactive waste attributable to spent fuel pool operations is 22.05 3 f t.3 for the year end December 31, 1981 for North Anna Power Station. The average annual quantity of so11d radioactive waste for Surry Power Station which can be 3 attributed to the spent fuel pool operations is approximately 60 ft.3 At both North Anna and Surry the solid waste contribution from the spent fuel pool is less 3 than 10% of the station's total solid radioactive waste. O o 10 O

o 47

J TABLE 5 - 2 3 FUEL POOL WATER IS0 TOPIC CONCENTRATIONS SURRY AND NORTH ANNA POWER STATIONS O North Anna ) Isotope Concentration mCf/mi Mn-54 2.44 E-5 Co-58 1.82 E-4 g Co-60 6.97 E-5 Sb-122 3.07 E-6 Cs-134 2.32 E-5 3 Cs-137 7.10 E-5 3 Surry Concentrations (mC1/mi) 3 Isotope Refueling Normal Xe-133 1.0 E-5 1.0 E-5 I -131 5.0 E-6 3 C0-58 4.0 E-4 5.0 E-5 Co-60 5.0 E-4 5.0 E-4 Cs-134 5.0 E-5 5.0 E-5 3 Cs-137 1.0 E-4 1.0 E-4 3 48

,0 Table 5 - 3 (g FUEL BUILDING AIRBORNE RADI0 ISOTOPE CONCENTRATIONS SURRY AND NORTH ANNA POWER STATIONS O North Anna g Air samples taken daring refueling and nonrefueling historically show less than minimum detectable activity for important radionuclides. O Surry g Concentrations (mci /ml) Isotope Refueling Normal I -131 1.0 E-10 4.0 E-11 O Co-58 4.0 E-ll 4.0 E-11 Co-60 6.0 E-11 6.0 E-11 Cs-134 4.0 E-ll O Cs-137 6.0 E-11 6.0 E-11 O .O 49 O

) TABLE 5 - 4 I ) FUEL BUILDING AREA EXPOSURE RATES NORTH ANNA AND SURRY POWER STATIONS ) Surry Dose Rate (mR/hr) ) location Refueling Normal Pool Water Surface 20 10 (6-12 inches above) ) Edge of Pool (waist level) 10 5 ) Fuel Bridge (waist level) 8 3 ) North Anna Dose Rate (mR/hr) ) Location Refueling /Nonaal* Pool Water Surface 0.5 - 0.6 (6-12 inches above) s ) Edge of Pool (waist level) 0.15 - 0.4 ) Fuel Bridge (waist level) 0.4

• Normal is defined as the time period between refuelings.

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O 6.0 DESIGN OF NEUTRON ABSORBER SPENT FUEL STORAGE RACKS O

6.1 Design Basis The neutron absorber spent fuel storage racks provide storage O locations for up to 1,737 fuel assemblies and two failed fuel canisters, and are designed to maintain the stored fuel having an O equivalent uranium enrichment of 4.3 weight percent U-235 in a safe, coolable, and subcritical configuration during normal and abnonnal conditions. O 6.2 Storage Rack Description O The free-standing neutron absorber spent fuel storage racks are designed to provide a maximum storage capacity of 1,737 locations in the spent fuel pool. The fuel storage rack arrangement will O contain several types of storage racks with array sizes ranging from 9x9 to 11x11 storage locations, as shown in Figure 6-1. O Each rack array consists of a welded assembly of individual storage cells. The storage cells are comprised of double wall Type 304 stainless steel boxes welded to each other with tie P ates to maintain the cell pitch of 10-9/16 inches. l O Each storage cell has an~ interior height of 168 inches to ensure =9 that the top nozzle and core components do not extend above the O 51

O top of the spent fuel rack when the fuel assembly is fully 3 inserted. The double wall construction of the storage cells provides four vented (open to the pool) compartments in which BC neutron absorber elements are placed for criticality 4 3 control. The neutron absorber elements are positioned on each side of the storage cell at an elevation corresponding to the fuel region of an assembly placed within the cell. A cross section 3 view of the storage cell is shown in Figure 6-2. The bottom opening of each storage cell sits on and is welded to the rack base which is comprised of the fuel assembly support plate and 3 support feet. The fuel assembly support plate provides the level seating surface required for each fuel assembly and also contains the openings necessary for adequate cooling flow. Figure 6-2 3 shows a schematic drawing of a 10x11 rack and a typical storage cell. 3 The rack support feet raise the racks above the pool floor to the height required to provide an adequately sized cooling water supply plenum. The rack design is based on natural circulation 3 flow within the spent fuel pool. Consequently, no credit is taken for coolant flow directed by diffuser systems. The support feet contain remotely adjustable jackscrews (accessible from the top of g the spent fuel rack) to facilitate and ensure proper support for the vertical loads and achieve the required leve1 ness. 3 52 g

The storage cell structure, acting in concert with the rack base and the rack support feet, provides the structural strength and stiffness characteristics required for the rack to accommodate the applicable seismic accelerations presented for the North Anna p Power Station, Units 1 and 2. No wall bracing or attachments are required to support the fuel racks under any design condition. Sufficient space is provided between adjacent spent fuel racks to preclude impact / collision in the event that sliding occurs during a seismic event. l 3 All structural members of the spent fuel racks are Type 304 Stainless Steel. The neutron absorber material is Boraflex with 10 2 an areal density of.025 gms B /cm and is fully compatible 3 with the gamma radiation field and water / stainless steel environment of the spent fuel pool. It should be noted that the compartments containing the Boraflex are not watertight; 3 therefore, the potential for a pressure buildup within the compartment (i.e., due to radiolysis of entrained water vapor) is not possible. D 6.3 Storage Rack Evaluation 3 6.3.1 Structural and Seismic Analyses The North Anna Power Station, Units 1 and 2, neutron h absorber spent fuel storage racks have been designed to meet the requirements for Seismic Category I J 53

O structures. Detailed structural and seismic analyses of 0 the neutron absorber spent fuel storage racks have been performed to verify the adequacy of the design to withstand the loadings encountered during installation, normal

o operation, the severe and extreme environmental conditions of the operating basis and safe shutdown earthquakes and the abnormal loading condition of an accidental fuel O

assembly drop event. 6.3.1.1 Application Codes, Standards and Specifications a

O The following design codes, regulatory guides and specifications have been used in the design / analysis O

of the spent fuel storage racks. 1. AISC Manual of Steel Construction, Eighth ~O Edition, 1980 2. AISI Stainless Steel Cold-Formed Structural Design Manual, 1974 -O 3. USNRC SRP 3.8.4 "Other Category I Structures" 4. USNRC Regulatory Guide 1.92, " Combination of Modes and Spatial Components in Seismic Response O Analysis", Rev. 1, February, 1976 5. USNRC Reg. Guide 1.122, " Development of Floor Design Response Spectra", 1978 O 54 C) 1 I

O 1 6.3.1.2 Loads and Load Combinations l l D The following load cases and load combinations have been considered in the analysis in accordance with the 3 requirements of USNRC Standard Review Plan, Sections 3.8.4 and the USNRC Position Paper. 3 Load Case 1 - Dead Weight of Rack, D + L (Normal Load) Under normal operating conditions, the rack is g subjected to the deadweight loading of the rack structure itself plus the loads resulting from the storage cells and a full complement of fuel assemblies 3 stored in the cells. Load Case 2 - Dead Weight of Rack Plus Installation 3 Load, D + I.L. (Normal Load) During installation the rack is subjected to the J 3 loading resulting from a 2g vertical load resulting from a suddenly applied crane load. 3 Load Case 3 - Operating Basis Earthquake, E (Severe Environmental Load) l 3 The rack, fuel assemblies, and virtual water mass react to the three dimensional loading of the horizontal and vertical components of the seismic ) 55 l--

D response acceleration spectra specified for the 3 Operating Basis Earthquake in the North Anna Power Station Units 1 and 2 Seismic Design Specifications. The seismic analysis is performed for the fully loaded 3 rack since this loading condition results in higher stresses and higher reaction loads. The effects of fuel assembly impact during a seismic event are taken into account. 3 Load Case 4 - Safe Shutdown Earthquake, E' (Extreme Environmental Load) 3 Same as Load Case 3 except that the seismic response acceleration spectra corresponding to the Safe 3 Shutdown Earthquake was used in the analysis. 3 Load Case 5 - Uplift Load, U.L. (Abnormal Load) The possibility of the fuel handling bridge fuel hoist 3 grapple getting hooked on a fuel storage cell was considered. The axial upward force considered for this load case was 4,000 pounds. 3 J j 56

O Fuel Assembly Drop Impact Load, FD Load Case 6 (Abnomal Load) l The possibility of dropping a fuel assembly on the rack from the highest possible elevation during spent

-O fuel handling was considered. A 1778 pound weight (110% of spent fuel assembly weight) was postulated to O drop on the rack from a height of 59 inches above the top of the rack. Three cases were considered: (1) a direct drop on the top of a cell; (2) a subsequent O tipping of the assembly onto the surrounding storage cells; (3) a straight drop through the storage cell and impact onto the rack base plate.

O Load Case 7 - Thermal Loading, T (Normal Load)

The stresses and reaction loads due to thermal O loadings are insignificant since clearances are provided to allow unrestrained growth of the racks for O maximum pool temperatures. O O 57 'O =

3-Load Combinations: D (a) For service load conditions, the following load combinations are considered. (1) D+L 3 (2) D + IL (3) D+L+E (4) D+L+T g (5) D+L+E+T (b) For factored load conditions, the following 3 load combinations are considered. I (6) D + L + T + E' 3 (7) D + L + UL (8) D + L + T + FD 3 6.3.1.3 Structural Acceptance Criteria 3_ The following allowable stress limits constitute the structural acceptance criteria used for each of the load combinations presented in Section 6.3.1.2. D D 58

O Load Combinations Limit *

'O 1,2,3,4,5 S 6, 7 1.6S or Fy (whichever is less) 20 8 Where S is the required section strength based on the ,o elastic design methods and the allowable stresses defined in ASME Code, Section III, Subsection NF,1980 Edition. 0

• The acceptance criteria are based on the applicable

'O sections of the NRC Position Paper on Fuel Storage Racks, SRP 3.8.4, and the ASME Code, Section III, Subsection NF. O

• • The acceptance criteria for Load Combination 8, the accidental spent fuel assembly drop on the rack, is O

that the resulting impact will not adversely affect the overall structural integrity of the rack, the leak-tightness integrity of the fuel pool floor and O liner plate and that the defomation of the impacted storage cells will not adversely affect the value of X,ff or ability to cool adjacent fuel elements. .O O-59

,0 6.3.1.4 Method of Analysis O The response of the rack structure to specified static loading conditions has been evaluated by means of linear-elastic analysis using the finite element O method. The rack was mathematically modeled as a three-dimensional finite-element structure consisting f discrete three-dimensional elastic beams and plates lO interconnected at a finite number of nodal points. The stiffness characteristics of the structural members are related to the plate thickness, cross O sectional area, effective shear area and moment of inertia of the element section. Six degrees of freed m (three translations and three rotations) were

O

. permitted at each nodal point. Appropriate boundary conditions were assumed for each load case. ^O The response of the rack structure to specified seismic loading conditions has been evaluated by O mathematically modeling the storage rack as a lumped

mass, multi-degree-of-freedom system.

Masses are lumped so as to represent the dynamic characteristics f the storage racks. The eigenvalues and O eigenvectors (frequency and mode shapes of vibration) of the lumped mass model have been calculated using O the Householder-QR technique. 60 0 \\

O The seismic response analyses are then performed using

.g response spectrum modal superposition methods of dynamic analysis, using the North Anna Power Station, Units 1 and 2 Amplified

Response

Spectra and I g appropriate damping for welded steel structures. Individual modal responses of the system are combined in accordance with Regulatory Guide 1.92. The maximum O response of the system for each of the three orthogonal spatial components (two horizontal and one vertical) of an earthquake have been combined on a

g square root of the sums of the square (SRSS) basis (Regulatory Guide 1.92).

g For the static deadweight and live load analysis, Load Case 1, all external loadings are applied as pressure on the face of the appropriate plate elements. The O deadweight of the rack is accounted for by specifying the appropriate density for the plate elements resulting in a distribution of the deadweight g throughout the rack structure. Similarly, for Load Case 2, rack installation and removal analysis, the deadweight of the rack is distributed throughout the O structure. The effect of a suddenly applied crane load is considered by applying a 2G vertical load. .O For Load Case 5 a net vertical uplift load of 4,000 pounds is applied at the worst location of the storage

rack,

'O 61

D The sloshing effects of water on the fuel racks have been evaluated using the analytical methods given in g USAEC's TID 7024 " Nuclear Reactors and Earthquakes". The " rattling" effects of the fuel inside the cell have been accounted for by using suitable impact g factors. The stability analysis of the free-standing spent fuel g storage rack and the structural damage resulting from a fuel assembly drop onto a rack (Load Case 6) was performed using energy-balance methods. g The structural adequacy of the rack design was further verified for the storage of consolidated fuel assuming g a compaction ratio of 2:1 and the utilization of non-compacting loose pin cannisters. D 6.3.1.5 Results of Analysis The results of the seismic and structural analysis 3 indicate that the stresses in the rack structure resulting from the specified load cases are within allowable stress limits for Seismic Category I structures. The fundamental frequency of vibration of the fuel 3 storage rack is 19.38 cps. 62

-O Sloshing of pool water in a seismic event will have insignificant effects on the fuel storage racks. lO The analysis of the accidental fuel assembly drop condition indicates acceptable local structural damage O to the storage cells with no buckling or collapse, and no puncturing of the stainless steel liner. Therefore, no significant changes in the value of-O K,ff will occur and the leak-tightness of the fuel pool will be maintained. O It is concluded that the design of the North Anna Power Station Units 1 and 2 high density neutron absorber spent fuel storage racks is adequate to

0.

withstand the conditions of the various load cases. 'O 6.3.2 Rack Sliding Analysis 6.3.2.1 Method of Analysis O A detailed non-linear time history seismic analysis was performed to evaluate the maximum sliding of the O spent fuel racks and to determine the maximum frictional resistance load transmitted by the spent fuel racks to the pool floor liner plate during the SSE.

O

.O-63

O The North Anna Power Station, Units 1 and 2 spent fuel pool floor absolute acceleration time history was used o to evaluate the sliding response of the storage rack structure. This 16-second duration acceleration time history represents the 2% damping spectra for the SSE.

o In order to perform the non-linear time history seismic analysis of the spent fuel ssembly/ storage O

cell structure, an 11x11 spent fuel rack and the stored fuel assemblies were represented by a two dimensional lumped mass finite element model. The 'O model consists basically of two coincident finite element cantilever beams: one representing the storage cells and the other stored fuel assemblies O attached to a " floor" mass by means of a non-linear sliding element. O The fuel element cantilever beam consists of masses lumped at the nodal points interconnected by discrete beam elements. Each lumped mass represents the

o tributary weight of the fuel element mass.

The stiffness characteristics of the beam elements are related to the effective flexural rigidity of the fuel o assemblies. I O l

O 64

\\

O The spent fuel cantilever beam similarly consists of lumped masses interconnected by discrete elastic beam

O elements.

Each lumped mass represents the tributory weight of the storage cells, water trapped inside the cells and the virtual water mass to account for the O ~ hydrodynamic effects. The stiffness characteristics of the storage rack beam elements are related to the dynamic characteristics (fundamental frequency of O vibration) of the storage rack as determined by the Householder QR Modal Extraction method of analysis. O In order to account for fuel assembly impact, adjacent masses of the fuel assembly beam and the spent fuel rack beam are laterally coupled by means of non-linear 'O spring / gap elements. The non-linear spring / gap elements permit the adjacent masses to impact each other whenever the gap closes during a seismic event. lC The stiffness of the non-linear spring is taken as the stiffness value for each spacer grid. An initial gap reflecting the lateral gap between the fuel assembly

O and the storage cell

wall, is provided.

The non-linear spring / gap elements are effective for fuel assembly impact on either side of the storage cell.

O

O

.O 65

I O i The two cantilever beams representing the storage cells and fuel assemblies are attached to the pool g floor mass by means of the non-linear sliding element to best represent the rack standing freely on the pool floor. The sliding of the rack is initiated when the g lateral force in the sliding element exceeds the frictional resistance force which is equal to the coefficient of friction times the vertical weight of O the rack. The effective vertical weight is taken as the vertical weight of the storage rack less the uplift loads due to the vertical component of the SSE. g Since the spent fuel racks with 12x10, 12x9, 11x10, and 9x9 arrays consist of similar honeycomb modules g with dynamic (frequency) characteristics essentially similar to the 11x11 storage rack, the lateral seismic inertia load and the frictional resistance load for the various spent fuel rack sliding response, which depends upon the relative magnitude of the lateral 3 seismic load and the frictional resistance load, will therefore be identical for the 12x10, 12x9, llxil, 11x10, and 9x9 spent fuel racks. O 6.3.2.2 Results of Analysis o The non-linear time history seismic analysis of the North Anna Power Station, Units 1 and 2 free-standing neutron absorber spent fuel storage racks was Q performed with the ANSYS computer code. 66

lO The maximum accumulated sliding displacement of the O spent fuel rack relative to its initial floor location has been calculated to be 0.32 inches for the coefficient of friction value of 0.20. This maximum O displacement value represents the accumuisted spent fuel rack sliding response during the 16 seconds of I the applied time history. O The maximum rack top deflection was calculated to be 0.18 inches for a coefficient of friction of 1.5, O which precludes any s11 ding between the rack and the pool floor. O The maximum rack cell displacement (rack sliding plus cell top flexural deflection) is therefore 0.50 inches at top level. This is less than the two inch O clo w e provided between adjacent racks and the ' he ance between the pool walls and racks. O It has therefore been concluded that the gaps provided between storage racks and between pool walls and adjacent storage racts are sufficient to preclude any O collision of adjacent structures during an SSE. It was assumed that the adjacent racks slide towards one another simultaneously. O 'O 67

.a, O 6.3.3 Nuclear Analysis ] O A detailed nuclear analysis has been perfomed to demonstrate that for all anticipated. normal and abnomal O configurations of fuel assemblies within the spent fuel storage racks, the K effective of the system is below 0.95. Certain conservative assumptions about the fuel assemblies and racks have been used in the" calculations and O theseassumptionsaredescribedinSectiop6.3.3.1. ( O The calculation method used for the criticality analysis was transport theory using the Monte-Carlo code KENO-IV. Input cross sections for, KEN 0 were determined with the lO NITAWL code. A detailed description of the calculation i method and codes is presented in Section 6.3.3.3, together with a description of thefbenchmark studies of KENO-IV. lo The results of the critiediit'y analysis are presented in ) l Section 6.3.3.4. 'o 6.3.3.1 ' Design' Criteria and Assumptions The criticality design criterion established for the O North Anna Power Station, Units 1 and 2 is that the multiplication

constant, K,ff, calculated by transport theory shall be less than 0.95 for all o

normal and abnormal configurations. The normal and O 68

O abnormal configurations analyzed are described in 6.3.3.2. The following conservative assumptions have O been used in the criticality calculations: 1. The pool water has no soluble poison. O 2. The fuel assemblies have no burnable poison. 3. The fuel is fresh and of an enrichment as high or O higher than that of any fuel available. 4. The rack is represented as an infinite repeating array. 5. No credit is taken for structural materials other O< than the fuel cells.

O 6.3.3.2 Configurations Analyzed j

The various configurations of fuel within racks that lO are possible are classed as either normal or abnormal configurations. Normal configurattons include the reference configurations and those variations in rack O dimensions, fuel parameters, and fuel location permitted by fabrication tolerances. Abnormal configurations are typically results of accidents or h malfunctions such as seismic events, malfunction of the fuel pool cooling system, etc. O O 69

O Noma 1 Configuration O The first reference configuration represents an infinite array of Westinghouse 17x17 fuel assemblies O centrally located in fuel cells of nominal dimensions. The second reference configuration is similar, but with 15x15 Westinghouse fuel. Table 6-1 O shows the important fuel and cell parameters used in these studies. O Other normal configurations included variations in storage cell ID, wall thickness, neutron absorber 10 material width and B

loading, storage cell

!O center-to-center

pitch, and eccentrically located fuel. Variations were studied with 17x17 fuel.

O l O O O O 70

O Table 6-1 i

O Fuel and Cell Parameters for Reference Configuration Westinghouse Fuel Type 17x17 15x15 Fuel Enrichment, w/o 4.3 4.3 Fuel Rod OD, inches 0.374 0.422 Fuel Rod ID, inches 0.329 0.3734

O Fuel Rod Pitch, inches 0.496 0.563 Number of Fuel Rods 264 204 Cell Pitch, inches 10 9/16 10 9/16 i

Cell ID, inches 8 7/8 8 7/8 Cell Wall Thickness, inches 0.090 0.090 4 Neutron Absorber Mg Boraflex Boraflex Neutron Absorber B'gerial

O Neutron Absorber Thickness,g, gms/cc2 Loadin 0.025 0.025 inches 0.085 0.085 Neutron Absorber Width, inches 7.5 7.5 Neutron Absorber Length, inches 138 138 Neutron Absorber Chamber Width, inches 0.095 0.095

O Cover Sheet Thickness, inches 0.029 0.029

O lO O

'O lO 1 71 ,0 i

O Abnormal Configuration

O' The abnormal configuration analyzed was a variation of 0

pool temperature from 68 F to 212 F. Inadvertent P acement of a fuel assembly adjacent to a fuel rack l O was not analyzed since a structure is provided around the periphery of the racks where required to prevent this abnormal configuration. Crushed fuel in a O storage cell resulting from a dropped fuel assembly was also examined and found to cause a reduction in K O eff* Fuel assembly displacements within a rack that would result from a seismic event are random in nature. O This case was therefore not analyzed because it is bounded by the configuration analyzed for eccentrically positioned fuel which is discussed l O. below. 6.3.3.3 Methods of Analysis O The reference configuration was analyzed by representing a rack, infinite in extent, with a 10 O 9/16 inch cell pitch, 4.3 w/o fuel assemblies, and 68 F water temperature by a single fuel cell with reflecting boundary conditions. Effects of pitch, O wall thickness, and cell ID variations were determined by separately varying each of these parameters. The O. 72

J effect of variation in pool temperature was determined 3 by running the reference configuration with reduced water density and changed cross sections to represent 212 F. 3 The configuration for eccentrically positioned fuel was studied by assuming an assembly to be located in a !3 corner of the storage cell. This is a conservative representation because it actually represents the entire rack array having groups of four adjoining

3 assemblies placed as closely as possible to one l

another. 3 Cross sections for all normal and abnormal cases were determined with NITAWL. 3 Codes The Monte Carlo code KENO-IV was used to evalu-g ate K,fffer the criticality configurations considered in this analysis. KENO-IV is a three dimensional, multi-group criticality code that solves a the Boltzmann transport equation to determine K,7f values. Working cross sections were generated using NITAWL and the 123 group XSDRN library. 3 J 73

O Uncertainty and Benchmark Calculations. O The uncertainties in Monte Carlo criticality calculations can be divided into two categories: O 1. Uncertainty due to the statistical nature of Monte Carlo calculations (the variance of K,ff values over many generations) O 2. Uncertainty due to bias in the calculational techniques. .O The variance of K,ff values for the many neutron generations is directly determined in the KENO-IV l calculations. Its magnitude can be reduced by

O increasing the number of ' neutrons tracked.

For fuel l rack criticality calculations, the number of neutrons tracked is selected to reduce the variance in K,ff O to less than 0.005. This variance is accounted for in the final result by adding it to the variations due to .o other normal variations in K,ff. The second class of uncertainty is accounted for by O benchmarking the calculational method against experimental results. t ,0 74 o

O KENO-IV has been extensively benchmarked for light water lattices. It has been shown to be typically O accurate to within 1% in K,ff by the Babcock & I Wilcox study performed for D.O.E. using 123 group Cross sections. Similar benchmark calculations have been performed by NES on KENO-IV for critical experiments on light water lattices carried out at Battelle Northwest Laboratories. Results confirm that g the KENO-IV code, with 123 group cross sections, yields conservative results within 1% of the experimental value. No credit has been taken for this O bias in order to perserve the conservative nature of l the results. .O 6.3.3.4 Results of Analysis 2-Normal Configurations l '[L - The following results were obtained for the normal t l) configurations: g \\ Configuration Keff. k Reference 0.9350 0 Pitch Variation, ition, + 1/16" + 1/16" + 0.004 Poison Width Vari T 0.001 Cell ID Variation, + 3/6T" T 0.009 Wall Thickness Variitions, +.005" T 0.005 Eccentric Location,.1" displacement T 0.004 Poison Cover Thickness, + 0.005" + 0.002 ,O KENO Uncertainty T 0.003 Meutron Absorber B10 Loading i 0.000 (minimum loading use) Fuel Enrichment + 0.000 (maxiiiium loading use) O 75

O The worst case normal configuration K,ff is obtained O by statistically combining the effects of the normal variations: 2 2 2 2 (.004 + .001 + .009 + .005 lo K,ff = .0042 +.0022 +.003 )1/2 =.012 2 g The worst case normal configuration K,ff equals 0.935 + 0.012 = 0.947. I O Abnormal configuration l The only abnormal configuration actually analyzed was increased pool temperature, since other abnormal O i occurrences either decrease K,ff (crushed fuel) or are prevented by the rack structure (misplaced fuel O assemblies). A KEN 0 problem was run for the reference configuration -O with decreased water density and cross sections changed to represent a pool temperature of 212 F. The X,ff decreased with increased temperature. O Haximum K,77 Value g Since the abnomal configuration does not in-K,ff, the final value is the worst case crease riomal, 0.947. O. 76

O (O

6.3.4 Thermal-Hydraulic Analysis The adequacy of natural circulation flow to cool the spent O fuel assemblies in the rack matrix was verified by establishing, for the North-South rack row with the maximum 1 ) number of assemblies (see Figure 6-1) a thermal-hydraulic O balance between the driving head provided by decay heat generation and the pressure losses existing in the natural circulation flow path. Pressure losses in the downcomers, O in the rack inlet plenum and within the fuel assemblies were explicitly considered in the analysis. Cross-flows in the inlet plenum area were conservatively neglected. The O closing of flow paths due to rack sliding was also censidered. O The results of the thermal-hydraulic analyses indicate that even with the most conservative assumptions, the natural circulation in the spent fuel pool is adequate to preclude O local boiling. The maximum temperature increase in the assembly with the minimum flow for the normal case (pool 0 bulk temperature of 140 F) was calculated to be 52 F. O The maximum temperature increase for the abnormal case 0 (pool bulk temperature of 170 F) was evaluated to be 50 F. In both

cases, maximum outlet temperatures 0

0 .O (192 F and 220 F respectively) are still below the pool 0 water saturation temperature of 241 F at the highest assembly evaluation. O 77

.O 6.3.5 References =O 1. Nuclear Energy Services, Inc., " Criticality Analysis Design Report for the North Anna Power Station, Units 1 &2 Neutron Absorber Spent Fuel Storage Racks", O 81A0875 o 2. Nuclear Energy Services, Inc., " Structural Analysis Design Report for the North Anna Power Station Units, 1 &2 Neutron Absorber Spent Fuel Storage Racks", !O 81A0876 3. Nuclear Energy

Services, Inc.,

" Seismic Sliding

O Analysis Design Report for the North Anna Power Station Units, 1 & 2 Neutron Absorber Spent Fuel Storage Racks", 81A0877 0

4. Nuclear Energy

Services, Inc.,

" Thermal-Hydraulic Analysis Design Report for the North Anna Power o Station Units, 1 & 2 Neutron Absorber Spent Fuel Storage Racks", 81A0878

O o

l l O 78

] U 1 6 ER '3 U G '9 F r I V 2 T I f 6 ,1 , 7 e { = U I 1 hl 1 9 2 1 X 1 X X 9 1 0 ,01 1 1 U ,7 1 9 X 9 I 1 T 2 8 X X N O 10 0 E 1 M U E S S G L L ,1 N L L ,6 E E A C C + 1 il R 2 E E X X R G G 0 1 R R ~6 1 0 'O A A A U "4 9 5 O O '7 K T T C S S 4 X 9 A L L R E E S U U K F F C E 2 1 1 1 X X A G T D N 0 0 R A E U .,4 E O F 1 1 E N R P IL S A 2 1 1 E X W T 7 3 2 9 T S 17 E B L 21 1 1 P E Y X X U 4 1 0 T U 0

• 2 F

1 = c ,6 2 1 c' 1 1 X X U 1 0 O" ,2 E 1 I 5 e _0 0 c 6)L 1 , 6 '6 E U 5-N ,l o 1 '4 F N 1 O y DTI R 'l 2 EA 1 LC I 0 AO FL 4 8 U 1 1 8

lllll 4' l )S UL R L B H E (L 0 0 O 0 0 2 T C T 8 8 I 6 1 N E W 5 2 6 0 8 6 G 6 6 3 3 5 E U I R C U 0 ' O A K 3 3 3 3 2 R 5 P T G f C O A I E T E T R F 8 S s [ Y P t l 6 5 1 3 T 1 Ym Q T S 6 6 K l M 1 8 1/1 C I 8 1/ C _Ll I I = D / 16 1/ 16 9 A 5 1 6 6 l I f l i 1 2 1 2 R RNI l MN Y B l 1 1 1 I -1 T i u 6 6 E l I 8 / /1 G 1 i / 9 9 5 5 D_Enl OOnL RI D A 8 6 5 5 9 9 A 1 0 0 R 1 1 1 O ZE ZEE; O H W Y R T E A C W S ON l OI l A X X X X X E T I eI R 2 1 2 P V R l 1 I 1 1 9 P L 1 R l J I l j O C I I 0 P O O 9 9 U I 5 I l K t L F[ F l L I 0 T A 1 R N O F l l

DX-l I F S I-l NN I l l j IO I I I P ,h{ = G T E = G IN TN ~ RE A LD OM L R L EA P E V P O OO , gT P C E U OO L ^ E L e S L S P F s W[f i f _ / i J L i W E I V E D i S I L I s4 Al 7 n 2 7 1 0 4 8 11 8 l

'O - 7.0 ANALYSIS OF EXISTING FACILITIES AND SYSTEMS AFFECTED BY THE PROPOSED MODIFICATION O The proposed modification does not change the physical configuration of the spent fuel pool or require any structural changes. The existing O spent fuel pool cooling and purification system and fuel building ventilation system will not require any modifications. The primary effect of the proposed modification will be to increase the amount of o spent fuel which may be stored in the pool, thereby increasing the weight 'to be supported by the floor of the pool. The additional spent fuel stored in the pool will slightly increase the amount of decay heat O which must be removed by the spent fuel pool cooling system. The effect on the spent fuel pool purification system will be minor. These and other effects are discussed below. O 7.1 Structural Considerations O The spent fuel pool structure has been analyzed to determine the effect the additional weight of the new racks and the stored fuel will have on the structure under static and dynamic conditions. O The existing structure has sufficient design margin which permits the installation of new storage racks without any structural O modifications to the pool concrete or superstructure. To verify the adequacy of the neutron absorber fuel racks a O dynamic model representing the fuel building structure and the - subgrade was prepared. This model was used to generate amplified response O 79

'O spectra (ARS) due to the design basis earthquake for the site. !O ARS were generated for both the safe shutdown earthquake and the operational basis earthquake at the mat surface, the top of the concrete structure and the roof of the steel superstructure. The ) J

o response spectra of the design earthquakes used are consistent with the requirements set forth in the North Anna Units 1 and 2 j

FSAR. The structural adequacy of the neutron absorber spent fuel racks is considered in section 6.3.1. jO 7.2 Fuel Pool Cooling System LO The installed North Anna 1 and 2 spent fuel pool cooling system has been analyzed in view of the Surry spent fuel assemblies which O are being considered for shipment to North Anna. Table 7-1 i summarizes the cooling system perfomance for both normal and abnormal (full core discharge) conditions.

O The design basis heat load was determined using the following conservative assumptions:

O 1. The irradiation times used for North Anna fuel were 272, 544, and. 816 Effective Full Power Days which correspond to

O a one, two, and three year fuel cycle, respectively, with a load factor of 85 percent and an annual 45 day refueling outage.

.O O gg l

O 2. Decay heat rates calculated using NRC Branch Technical O Position 9-2. 3. All fuel to be moved into the pool is done instantaneously O 150 hours after shutdown except for the full core discharge case when fuel is moved from the reactor to the fuel pool at a rate of 20 minutes per assembly starting 150 hours O after reactor shutdown. 4. Stretch rating of 2900 MW is used for full power for North O Anna. i 5. Maximum storage capacity of 1737 fuel assemblies. 10 6. Service water temperature at its design maximum of 110 F. Component cooling water temperature O of 117.8 F and 2800 GPM per cooier. 7. Two discharge schemes have been considered: a) all O refuelings are normal refuelings; b) all refuelings are normal refuelings except the final refueling which is a full core discharge. O The currently installed spent fuel pool cooling system has sufficient cooling capacity to maintain the pool water temperature O at or below the FSAR design criteria of 140 F for the normal case and 170 F for the abnormal case with one fuel pool cooling system pump and two coolers in operation. 81

lO i l These fuel pool temperatures are ' calculated based on very l O_ c nservative worst case assumptions and are valid for establishing a design basis for the proposed action. O Actual operating temperatures experienced by both North Anna and Surry have shown maximum temperatures for full core discharge U cases to be 103 F and 110 F respectively with one pump and one O cooler operating and normal operating temperatures less than 100 F. O The calculated temperatures and heat loads are summarized in Table 7-1. O Possible malfunctions and corrective actions are summarized in Table 7-2. O 7.3 Fuel Pool Purification System No significant effect on the Fuel Pool Purification System is O expected due to the prolonged storage of additional spent fuel assemblies. The maximum load on the fuel pool purification system occurs during refueling operations when spent fuel is being O moved. Therefore, there will be no significant increase on the purification system since the number and frequency of refueling operations will not change. O O-82

O 7.4 Fuel Buf1 ding Ventilation System O Since the added fuel storage capacity represents longer term storage of spent cooled fuel, the escape of gaseous or volatile fission produ::ts from even defective fuel is expected to be O ] negligible. Much of the iodine and xenon has decayed after 100 i days cooling time. Since most of the tritium in the water is

g formed primarily as a product of the neutron irradiation of boron in the primary coolant, the contribution of fission product tritium is minor.

There is no mechanism for particulate fission i'g products to become airborne. Becai:se of its long half life, Kr-85 remains even in older spent fuel. However, the thermal driving force required to cause its diffusion in defective fuel is not g present. Air samples taken in the fuel buildings at both Surry and North Anna do not show Kr-85 in detectable levels, and it is not expected to become significant as fuel storage increases.

O Therefore, increased fuel storage will have essentially no impact on concentrations of radioactivity in the fuel building atmosphere.

1 i 0 g Since the FSAR pool temperature limits of 140 F (normal case) I and 170 F (abnormal case) will not change with the modification, l there~will be no effect on the design evaporation rate of the pool. 9 The effects on the Fuel Building Ventilation System are therefore insignificant. .O O 83

D TABLE 7 - 1 3 SPENT FUEL P0OL COOLING SYSTEM HEAT LOAD AND OPERATING TEMPERATURES WITH STORAGE OF SURRY SPENT FUEL IN THE NORTH ANNA 1 AND 2 FUEL POOL 3 3 Decay Heat Fuel Pool Temperature, F MBTU/HR 1P2C 2P2C 3 Current High Density Racks (966) Normal 19.4 135.4 130.4 3 Abnormal 35.9 154.2 144.9 3 Proposed Neutron Absorber Racks (1737) Normal 23.1 137.5 133.5 g Abnormal 40.1 160.0 148.5 D 1P2C - 1 Pump 2 Coolers 2P2C - 2 Pumps 2 Coolers O O 84

. _ =

O TABLE 7 - 2 i

f lO. SPENT FUEL POOL COOLING SYSTEM MALFUNCTION ANALYSIS I I component Malfunction comments and consequences O i' i Spent Fuel Pool Pump fails to start The standby pump will be i Cooling Pumps or fails during started manually. ~0 operation 1 i_ - Fuel Pool Loss of Function Sufficient makeup water O could be made available from Coolers other station systems to cool i the fuel and maintain !'O sufficient water shielding i i, 1 over the fuel. More than 1 hr exists to realign the l'O piping system because of the 1 slow heatup rate of the i e pool. The realignment could i ,Q be effected by either changing valve lineups or implementing certain tem-lO-porary measures, such as the use of temporary pumps or hoses. ? O-85 n n

l O j 8.0 NEUTRON ABSORBER SPENT FUEL RACK INSTALLATION AND REMOVAL OF HIGH DENSITY RACKS 0 The installation of the neutron absorber spent fuel racks can be made while both units are either operating or shutdown without affecting the c ntinued safe operation of the station. The new racks will be O installed with about 357 spent fuel assemblies in the spent fuel pool. The new neutron absorber spent fuel racks will be shipped to the site by O truck. The Fuel Building crane will be used to lift the racks off of the truck and bring them into the Fuel Building. Movement of the racks into position will either be done with a special temporary crane or O utilizing special rigging on the Movable Platform With Hoist. A special lifting rig is being provided by the rack manufacturer to facilitate the installation and positioning of the new fuel racks in the fuel pool. O The rig features remotely actuated positive capture devices, which All movement of preclude accidently dropping a rack during handling. spent fuel and spent fuel racks will be controlled by written O administrative procedures which will prohibit movement of the spent fuel racks over locations in the pool where fuel is stored. O The high density spent fuel racks will be removed from the pool, decontaminated, and cut up for disposal offsite. O Measures will be taken to ensure that personnel exposure will be kept as low as reasonably achievable (ALARA) for the installation of the neutron absorber spent fuel racks. g O. 86

O 9.0 ANALYSIS OF THE SAFETY IMPLICATIONS OF THE PROPOSED H0DIFICATION ,O The proposed modification will not change the safety analyses which have been performed and reported in the Final Safety Analysis Report, Section i O 15. The proposed expansion of the spent fuel storage capacity could affect the offsite radiological consequences of an incident because of the additional increment of long-lived radioactive fission products o stored in the pool. The effect of this amount of additional radioactive products on normal station operation is discussed in Section 9.5 of this report and its effect on the spent fuel handling accident is discussed

O in Section 9.4.

The following discussion summarizes the potential effects which the o proposed modification may have on the safety of the station and the

public, o

9.1 Loss of Spent Fuel Pool Cooling Capability As discussed in Section 7.2, the proposed modification will O increase the amount of heat energy which is added to the pool water which must be removed by the spent fuel pool cooling system. The existing cooling system has sufficient design margin .o to remove the additional heat load when irradiated fuel is stored in the pool as shown by Table 7-1. As indicated by the failure analysis presented in Section 7.2, cooling capacity could be o restored quickly in the event of a pump failure. O 87

i O-In the unlikely event that the spent fuel pool cooling system was lQ to become completely inoperable, installed station systems could provide sufficient makeup water to cool the fuel and to maintain sufficient water shielding over the fuel. There are several o sources of makeup water readily available in the event it is l re quired. These sources are: o 1. Primary grade water system 2. Fire protection system 3. Boron recovery system 'O 4. Refueling water storage tank 4 These sources could be utilized by either changing valve lineups ]o or implementing certain temporary measures, such as the use of i temporary pumps or hoses. >g In summary, sufficient heat removal capacity is installed to assure that the pool temperature remains below the boiling point. As additional backup, a number of installed station systems could 1 O provide makeup and cooling water if required. I 9.2 Fuel Pool Leakage Control and Shielding l The proposed modification will not affect the leakage and shielding requirements contained in the FSAR. The lowest level of (O ~ pipe' penetration through the fuel pool structure is at El. 285 ft-9 in., which provides a minimum water level of over 24 feet of L ~ water above the stored fuel to provide shielding and cooling. O 88

(O^ The proposed modification will not require any additional' piping fO Penetrations; therefore, there are no safety implications associated with spent fuel pool Teakage control or shielding. A 9.3 Earthquake and Tornado Protection

O The proposed modification will not require any structural changes;

!O therefore, it will not affect the ability of the structure to withstand the effects of an earthquake or tornado as stated in the FSAR Sections 9.1.2, 3.7, and 3.3. The new spent fuel storage racks and pool structure have been analyzed to ensure that the jO-racks can be accommodated by the structure during a seismic event. These analyses are discussed in detail in other sections of this report. .O In sumary, the seismic and tornado design provisions stated in the FSAR are not changed as a result of the proposed modification. O 9.4 Fuel Handling Accidents O Section 15 of the FSAR describes the fuel handling accidents which have been analyzed, including the case where a fuel assembly is

O dropped onto the floor of the spent fuel pool.

The proposed modification will not affect the consequences of the accidents i h analyzed in the FSAR because the analysis assumes that' only one fuel assembly, the one being installed, is damaged. Thus, the 0 l consequences of the accident are independent of the number of spent fuel assemblies stored in the pool. 89

O The neutron absorber spent fuel racks have been reviewed in regard to: O I 1. Dropping a fuel assembly on the racks 2. A fuel assembly becoming stuck in the spent fuel rack O 3. Dropping a fuel assembly next to the rack 4 While minor damage may be incurred to the rack if a fuel assembly 1

g is dropped on it, the stored fuel will not be affected and subcriticality will be maintained. The amount of force applied to i

a stuck fuel assembly is limited by the capacity of the crane. O While damage may be incurred by the stuck fuel assembly, the weight of the fuel rack is sufficient to prevent any motion of the rack itself. The surrounding stored fuel assemblies will not be O damaged and subcriticality will be maintained. In summary, the safety implications of the proposed modification ~O as related to fuel handling accidents remain the same as previously analyzed in the FSAR. 'O 9.5 Personnel Radiation Exposure Storing additional spent fuel in the pool will increase the amount g of corrosion and fission product nuclides introduced into the pool water. The proposed modification will approximately double the amount of fuel to be stored in the pool. g O 90

.O. During the storage of spent fuel under water, both volatile and lo nonvolatile radioactive nuclides may be released to the water from the surface of the assemblies or from defects in the fuel cladding. Most of the material released from the surface of the o assemblies consists of activated corrosion products such as Co-58, Co-60, Fe-59 and Mn-54 which are not volatile. The radionuclides released through defects in the cladding, such as Cs-134, Cs-137, o Sr-89 and Sr-90, are predominantly nonvolatile and, as with the activated corrosion product nuclides, the primary effect is their contribution to radiation levels to which workers near the spent .o fuel pool would be exposed. As noted in Section 5, the four primary isotopes noted in the pool water at Surry and North Anna have been Cs-134, Cs-13), Co-58 and Co-60.

o Based on the experience from both Surry and North Anna, the contribution of personnel exposure attributable to the fuel pool 10 has been only a small percentage of the total station exposure.

l Even if this exposure were doubled it would still be a relatively ^ minor contribution to the overall personnel exposure at the

O station.

1 The installed purification system described in Section 5.2 will be O used to remove the nonvolatile corrosion and fission product nuclides. The removal of these nuclides will assure that the radiation exposure to personnel will be maintained at low levels.

O O

91 ,-n

q a The volatile fission product nuclides of most concern that might be released through defects in the fuel cladding are the noble J 3 (Xenon and Krypton), tritium and iodine isotopes. Since gases short-lived noble gases will decay to negligible amounts, the only significant noble gas isotope which could remain in the spent fuel g pool and attributable to storing additional assemblies for a It is not expected longer period of time would be Krypton-85. that increasing the spent fuel storage capacity will increase the 3 Krypton-85 release rate, since the fuel discharge will continue on approximately 1/3 core per year per unit rate and the release of Krypton-85 is most likely to occur during the initial year of 3 storage. increased by the Iodine 131 releases will not be significantly 3 expansion of the fuel storage capacity since the I-131 inventory in the fuel will decay to negligible levels. Operation experience at Surry to date indicates negligible levels of I-131 in the pool 3 water. The pool water temperature will be maintained below the current 3 design temperature; therefore, it is not expected that there will be any significant change in evaporation rates and the release of Operating experience at Surry to date has not indicated tritium. g the presence of tritium in the fuel building. i I 92 O I

O As discussed in Section 5.5, the purification filters are normally () changed because of high differential pressure; therefore, it is not expected that the proposed modification will significantly increase personnel radiation exposure during filter changes. O Based on experience at Surry Power Station, the radiation exposure is relatively low, approximately 140 mR/hr. The demineralizer j resins are currently changed about twice a year resulting in

O Personnel exposure of about 55 mR.

The proposed modification is not expected to significantly increase this value. O In sumary, the proposed modification will not significantly l increase personnel radiation exposure during normal and refueling l operations. 'O O .O O O

O 93

s O { 10.0 ENVIRONMENTAL IMPACT OF THE PROPOSED MODIFICATION O The proposed modification would increase the amount of decay heat produced in the spent fuel pool, and result in a small commitment of metal resources. ,g The environmental impact of the proposed modification is insignificant. The environmental impact has been reviewed in light of the current Final O Environmental Statement, the Final Safety Analysis Report, and 10CFR50. Based on this review it has been concluded that the proposed modification will not significantly affect the quality of the human

g environment.

l 10.1 Independence of the Action f0 This licensing action would clearly have a utility that is independent of other licensing actions designed to ameliorate a O possible shortage of North Anna spent fuel storage capacity. As discussed in Sections 2.0 and 4.0, a need for additional spent O fuel storage capacity has been established as an interim solution l to allow North Anna Unit Nos.1 and 2 to continue to operate when present storage capacity is expended. As a long-term solution for .g spent fuel

storage, the federal government is committed to encouraging commercial reprocessing and/or providing a repository fr spent fuel.

While the proposed action may not completely O cover the time period until these disposition alternatives are 94 O

O expected to be available, there are a number of alternatives which

O Vepco is actively pursuing to alleviate the problems associated with the storage of spent fuel. The alternatives include:

lO 1. Extending fuel burnup through high burnup R&D programs to decrease the amount of spent fuel generated, l O 2. Design and construction of independent spent fuel storage installations utilizing both dry and wet i techniques. O 3. Expansion of the fuel storage facilities for North Anna Unit No. 3 which is presently under construction. "O. The proposed modification will allow Vepco additional flexibility, which is desirable even if adequate offsite storage facilities 'O become available prior to when the proposed additional storage capacity is expended. O It has therefore been concluded that a need for additional spent fuel storage capacity at North Anna 1 and 2 exists and that this need is independent of the utility of other licensing actions O designed to ameliorate a possible shortage of spent fuel storage capacity. O O 95

\\ \\p 10.2 Commitment of Resources l The proposed modification will require the utilization of 4.9 x 5 101bs of stainless steel. The amount of stainless steel used O annually in the United States is about 1.1 million tons. The amount of stainless steel required for the racks is a small percentage of this resource consumed annually in the United States and is insignificant. No other significant material resources will be required because the design of the fuel pool will remain unchanged. The land area now used for the fuel pool will be used O more efficiently by reducing the spacing among fuel assemblies. It has been determined that the proposed action will allow for the continued operation and provide operational flexibility for North 3 Anna Power Station Unit Nos.1 and 2 and will not affect similar licensing actions at other nuclear power stations. E 10.3 Cumulative Environmental Effects The additional capacity of the spent fuel pool is proposed for O North Anna Power Station, Unit Nos.1 and 2, only; therefore, the environmental impacts can be assessed within tne context of the application. Based on the information contained herein, it has O been shown that the environmental impact due to the installation and operation of an expanded spent fuel pool storage capacity is insignificant. It is concluded that the cumulative environmental O impacts associated with the expansion of the spent fuel pool will not result in radioactive effluent

releases, occupational O

96 l

O radiation exposure or thermal effluent releases that significantly ] affect the qu

y of the human environment during either normal operation of the expanded fuel pool or under postulated fuel handling accidents.

J 10.4 Technical Issues J The technical issues associated with the proposed modification are addressed in this report. There is reasonable assurance that the proposed modification can be carried out as described herein with O no adverse effects on the health and safety of the public. 10.5 Need for the Action O As. stated in Section 4.0, a number of alternatives have been considered. The modification described herein provides the most O economically feasible solution to ameliorate the need for additional spent fuel storage capability described in Section 2.0. If the proposed modification is not implemented, the O alternative of ceasing operation of the facility would eventually result. This would result in an additional cost to Vepco's customers which is estimated to be approximately $350 million O dollars per year (in 1982 dollars) for purchase of alternate power, if available, and maintenance of the station in a shutdown condition. Deferral or severe restriction of the proposed O modification would result in substantial harm to the public interest. O 97

O 10.6 Final Environmental Statement O

The proposed modification will not significantly alter the evaluations contained in the Final Environmental Statement. The proposed modification will create a slight additional heat load on the service water system; however, no discernible temperature difference in the thennal effluent from the station is expected. l 'O l 10.7 North Anna 1 and 2 Final Safety Analysis Report The descriptive information contained herein is intended to O supplement the material contained in the FSAR. The design criteria specified in the FSAR have been used as the basis for the proposed modification and have been supplemented as appropriate for the new spent fuel storage racks. The modification does not substantially change the analyses and descriptions in the FSAR. O O O O g 98

O

11.0 CONCLUSION

S O Based on the infomation contained herein, Vepco has concluded that: lO 1. The proposed action is necessary to maintain the capability of a full core discharge and to assure adequate storage space for normal refuelings at North Anna Power Station O through 1998. 2. The proposed action provides the most economical of the O feasible alternatives to ameliorate the potential shortage of storage capacity, and it is the one most certain to avoid the early forced shutdown of one or both North Anna units. 3. The North Anna Power Station Unit Nos.1 and 2 spent fuel pool is adequate for the installation of the proposed neutron absorber spent fuel racks without modification. O 4. Spent fuel from both North Anna Unit Nos.1 and 2 and Surry Unit Nos. 1 and 2 can be stored safely in the proposed neutron _ absorber spent fuel racks. O 5. The proposed action will not affect the health and safety of the general public. O 6. The proposed action will not significantly affect the quality of the human environment. O 99 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _}}