Text

Summary of Proposed Mods to Spent Fuel Storage Pool Associated W/Increasing Storage Capacity. Affidavits of RW Calder & Hs Mckay Encl
	ML19224D390
Person / Time
Site:	North Anna
Issue date:	04/30/1978
From:	; VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:	;
Shared Package
ML19224D386	List: ML19224D385; ML19224D387; ML19224D390;
References
	NUDOCS 7907120027
	Download: ML19224D390 (76)
	v • d • e

{{#Wiki_filter:" Os 3 g%V E1H " "" ac runuc Docu: -A (5\\ qhh,nf/ 5~

SUMMARY

OF f;; / u x s ',% j j, j \\# PROPOSED MODIFICATIONS TO TIIS SPENT FU2.L STORAGE POOL ASSOCIATED WITH INCREASING STOPAGE CAPACITY FOR NORTH ANNA POWER STATION UNIT NOS. 1 AND 2 DOCKET NOS. $0-338 50-339 LICENSE NOS. NPF-4 SNM-1801 APRIL, 1978 n q,- -, g... ') J Z too VIRGINIA ELECTRIC AND POWER COMPANY ggon200A7

~ TABLE OF CONTENTS 1.0 Introduction 2.0 Reason for Modification 3.0 Proposed Action and Schedule 4.0 Alternatives 5.0 Description of Existing Facilities and Operating Experienc3 at Surry Power Station 6.0 Design of New Spent Fuel Storage Racks 7.0 Analys.ts of Exist.tng Facilities and Systems Affected by the Proposed Modification 8.0 Installation and Removal of the Spent Puel Storage Racks 9.0 Analysis of the Safety aplications of the Proposed Modificaticn 10.0 En rironmental Impact of the Proposed Modification 11.0 Conclusions e 352 287 .1

PREFACE ~ The purpose of this document is to su:cnariz e the proposed changes to the spent fuel storage pool at the North Anna Power Station, Unit Nos. 1 and 2, associated with increasing the storage capacity of the pool. It is intended that the document fully describe and analyze the proposed modification, as well as describe the installed systems which may be affected by the modification. Operating experience and data are presented where appropriate to substantiate the conclusions made regarding systems performance and environmental impact. The information contained berein is also intended to provide pertinent technical details associated with the modification and its environmental impact., as well as to identify the reasons for the change. This document should provide the Nuclear Regulatory Commission Starf with sufficient information to review the proposed change in accordance with regulatory requirements. o O m 352 209 ii

1.0 INTRODUCTION

~ Virginia Electric and Power Company's North Anna Power Station, Unit No. 1 was issued Operating License No. NPF-4 on November 26, 1977. The unit will have a thermal generation rate of 2775 megawatts. The operating license for Unit No. 2 is expected in Dectmber of 1978, with commercial operation tentatively scheduled for March 1979. The initial fuel cycle for Unit No. 1 is presently scheduled to end in November 1979. After each fuel

cycle, at an average design burn-up of 33,000 1:WD/MTU, approximately one-third of the fuel elements are discharged permanently from the core and are stored in the spent fuel storage pool.

The present storage capacity of the spent fuel pool is 400 fuel assemblies, or approximately 2 1/2 cores. It is prudent engineering practice and the policy of the Virginia Electric and Power Company (Vepco) to reserve storage space in the spent fuel pool to receive an entire reactor core (157 fuel assemblies), i.e., full core off-load, should unloading of the core be necessary or desirable because of operational considerations. Virginia Electric and Power Company has a policy and con:nitment to its customers to supply reliable and economic electric service, which requires tha t its nuclear power stations be operated reliably and continuously. To ensure continued operation, the ability to discharge spent fuel must be maintained. This, together with the tact that alternative spent fuel

storage, reprocessing, or permanent disposal facilities cannot assuredly be available to Vepco when first needed leads to the conclusion that an increase in the spent fuel storage capability f.s necessary to accommodate both subsequent spent fuel discharges and to maintain the entire core off-load capability.

Vepco has evaluated th'e available alternatives for ensuring that the above objectives are realized. The decision to install additional spent fuel storage capacity provides the most favorable solution considering commitment of resources and availability. It is intended to be a near term solution which will provide storage capacity until 1987, at which time it is planned to ship the spent fuel to an offsite storage facility and then to a reprocessing or permanent disposal f acility suoject to the implementation of the Department of Energy's (DOE) spent fuel management policy. To accommodate bo th subsequent spent 'uel discharges until 1987, and maintain the full core off-load capability, a modific.' tion is planned to increase the spent ruel storage capacity by installing new spent fuel storage racks with a reduced center-to-center spacing of the f uel assemblies while maintaining suberiticality under all conditiono. The planned modirication will result in a maximum storage capacity of 966 fuel assemblies. 352 289 1

In summary, to avoid unnecessary' unit shutdowns when an entire core off-load capability is not available for normal and emergency operating conditions, and to provide for additional spent fuel discharge storage capacity until 1987, given the uncertainty of alternative capabilities, an increase in the spent fuel storage capacity at the North Anna Power Station, Unit Nos. I and 2, is necessary. Figure 1-1 diagrammatically shows the buildup of spent fuel assemblies in the spent fuel pool. 352 290 2

7 ... _ _. - ~. _ _ _ _. _.. _..._.-.-_y_ .e .t* r 6. C C _.'.m i Q Q- ..__.__._.m._ _m p_._..._..____......___. m... g i { _.6 . 4). .J t 't -..._k4._.. ~_ ._.___.-___.2,..__._a.. -___J _ _ ___. 3 F_._..._.___..--.._..__.__. .-.- g .w -m._. g. _. _. =. ,g -c . g :.---- -... ; .._.7._. g. . _ l. , + _. % ._..._____.. _ _. _. Q... Q m tp F,'. _...-*--N ' : h Lu' _...______'=.I' - i . -.. l _2 k4. b ._.J $N 1 Q __..,' M_s : c. _'- _. _ _ _ _ _.. _J. -e mm .[.__.____.____ ..._7_... 73.,. - _-' ] l- 'r._ _ ^ - ' -. -h... 3 OQ%%%%-

-q-L

^--'-.-..

L __'W__ O_O.O.C _ _ _. - - ~ - u J;r, g. ~ m: _.._ -

4. m..

[___..,_., l -. _. _._.. g y rej u _g - -

_. _ _.. _. - -_._ _. _.__-[ x - : L.

__ - f _{j. . g (;) c.=_ .. = ;; x; n y., gy y _ _.._ ,[~ -- r - : g : _ - - ~- ].. r- _- - l - __ r -} f:. _ : =_... - - =^ _ : - -:. :~ -- _ ' :_z. l==L. 2.. b ~2: 5 -~~_ - _5 y n o y ;. m._- p Q q, y ,4-N - ' - -- ' 3 ' -.p_ ~ F - -?El [ n.E -

- y..

j_--- _-f._;;g2f:- 1 r _; V-i-- - =, f.,.;./,-,*_.. i- ' r yIl

f..-.:

....T:. _..? % _ " :1.. - WWW4 W 'A M ' ~ - ? __. _: =: L -== - l L. . - _: T ' L ^ .? G 1: -- - -- _ _. 51-- -- Z.- [ -- - ~ % -OMr-- =. :-.

== " N N N : N d d N N.-- > '_ -2:$ 'J--- [ c. _.. = = =. =. r _ - = _ _ =- - a.= y - - g= - - p=- _== = u=; _

m

~

.#.g(.,_==.__=__=_=_-.=__.-=._..=.:_.=_=._,=_.l,.___. - t cl -- t :t :In - F i c g....

=-I-.- =_. _ _ - yCWWm-m O-'O'7-T. .. _. =~n-d -= = = +. c-r - r: r-w ^ m w n _:== =- ._N. _ _.= : =.r - v_ l. N r-i N -4 r-t Ch H m - -- 6 .: - -._ ! :. _. c -t_ i 1-1 I _r-1 . H - ..... _ _.._.. : _ '. l r g -. _. _. _ _ _ __-.M._.-.....:-~- _ .10 ' (s 7; h.=.5, n._== E_ =. ri.m_ _ =..:. =;=- -:. : _=. lE.=.__m..: p... _=. w __ : :-_=1-i. :..... :....... _._:: -_ .g s y.. =, ._ ~~.5:- ' ---.: =- :5 ~ h_':. l?--- '" 7~l-. T.- - ^.. :-? -- T:1-~ z':~;_ _ _- -] .M ? ^ -. ~ Z=- l- -4 ."- ~ ~ [T u._T[ _, - . ' _ ~ --Z---.-_.'.- _~ - -_:. :_~_:.-_'--'--_-}*T:_.{

' ^..: _ :[:lTqj.--l'; T 3: q: - -

'^..': ^'--~'_T~~-"._ . ~ -'.: T.:. 4 (~~$ -== ? , : - :_'-W. i.E _ L: IT;; -- l ~ ' ~ = ~ ^3 ~ ~ : } - ' =~, ~ - ' T _ L

P ="

~ ~>.. _. s _r_ .__________.2. ._?' l _-. I CD ._g . ch _ _ _ __ ____h______.__._-._ _.__.__._.l__ l. __.. - _ r-i ' :U..---'. __ )_ [=

:'-. ' T [

{_ _._.. ^ ---5.'~~5-~~~.^._'__ V. __r__ __--_=:_ - - - - - - - . - - - - - - p_7 - - - - - _ w_l_~~_.:__'_'..T_.-'...-^:_ _ _ _ __. _ _ _ _ _ -_.:..=^'__..&.-._.. - % 5C h. 0 j_._.: - -,, ; 'r Q l_ ], f ] {g __,;;_7i.f..__4 7- _. u . Q { )g_7,,p __7 7.__ _ _. k _. _ _ _ g3_ _ _.. p m.T r- _ _. p 1 .,._.g m _. _ _ _ _. _ __ : 1; .-r-4 5 ~. g _ - - _. = _: ,----r. '-- A;+. T:"T ~ .: g C

(l :__--. = - : -. -- '- :. : -- '. ; $ -'? V' -

y =--7-' y v - - : =_= - r -_ - ^ -.2. .. _.z _- - =

b. I

.. _. -.-_..--7'.__ ..(._.- 2 =-7_:r --'- - g - _z -- ___; __3 Q:.. _ z -_ g3____ r T--. .5. .g l .. - - - ' ~ ' s-- ---_ _7 * ._.g l_ . x, }y -r-- - - G } _ ___d. _. l _ _ _ _~: --__ _ _ _k-- . _ L _. : _. __ _. ..__g S _ q. g _.._ _ _._.. _. __._p 9 N lM ._.M7l. lM r*' y-ta g.. _ _. _ _ [g. p 1 -Q. "4 ,Lt ~ b CO l."tD_. p S. _ _ _ _ __ _ _ _. h ' 1 _ _. _ _ _ _ __ l. G.. -H 3* M _..'_ ___ _ _.. _ __... _. _.. @8. l. ]..._ _ ) ]. _ - ' q _. _. _...___... l M _. .._,a _.__.,..a.. o c-3i ~ ___:q y .l.c. [. a _ _._ _ _ _ _ _.,a. _, __ e...- i - c_ Q Cn Q l X l*__ l @. .O.. .o_, _._.go____ _ __vm J c, i .a r c D U l'O. l _ _.__m -._ .g U 1 = __ _. I q - t _. . M Q l3 H M 0 i l _ __ _ _. 3_ l.. p_. l. 3 _) _ p _ __ _ g ____ _.

1. __. g '. ! (.

[3 ( 1 -m__.--_. ,l : .g. lV _... _ _. -V V 5 i .e i I to a i t , q, l l M H l 8 w c. a o. O, c. a. o - - - - - O O O O C O N o m w v N. H r.-e b N U b N E UI d2 b d3 $ U SC 05 { .,) a l_ L 3

2.0 REASONS FOR THE MODIFICATION The basic reason for the planned modification is to maintain full core off-load capability and to provide additional onsite spent fuel storage capacity until' such a time when adequate offsite interim

storage, reprocessing, or disposal facilities are operational and available to the U.S. commercial nuclear power industry.

The existing design of the pool was predicated on being able to ship spent fuel offsite for processing after about 150 days resident time in the pool for decay of the short lived radioactive fission products.

However, the President has recently suggested that commercial reprocessing of spent fuel will be indefinitely deferred until the international problems associated with the proli feration of nuclear weapons are resolved.

Consequently, spent nuclear fuel will have to be disposed of as wastes or stored until reprocessing o-an alternative use of spent fuel is implemented. The has recently announced its spent fuel management policy t a eby the federal government will take title to utilities' spent fuel and provide interim storage facilities by 1983 and radioactive waste disposal facilities by 1988. However, the capacity and location of these facilities have yet to be defined. Furthermore, a policy has yet to be implemented on a firm basis through the enacti m of legislation and other necessary actions. In addition, considering the anticipated large ' accumulation of spent fuel ir.ventory in the U.S. by 1983, it is not known whr Vepco would be able to ship spent fuel to the facility cce in operatio. t. This condition necessitates that the ma3arity of spent fuel discharged from operating reactors be stored until at least the late 1980's. With the present number of fuel racks available at the North Anna fuel

pool, Vepco anticipates ~ loss of full core discharc a in the Winter of 1981.

The ability to discharge one third ccre for a normal refueling will be lost in the Fall of 1983. In order to avoid a problem with spent fuel storage capacity at North Anna Power Station Units 1 and 2 (which results from insuificient storage, reprocessing, or disposal f acilities), it is desirable to replace the existing fuel racks with the high density fuel racks before any spent fuel is stored. This will minimize radiation exposure to installation personnel. Therefore, Vepco plans to coa plete the high density fuel rack installation as outlined in this report. Therefore, the most prudent course of action to satisfy the necessity for storage of our spent fuel until the late 1980's is to complete the modification as described herein. JJL L /,_ 4

1 3.0 PROPOSED ACTION AND SCHEDULE The Virginia Electric and Power Company has contracted with NUS Corporation to design, engineer, and manufacture the new spent fuel storaga racks. The following are key dates associated with the proposed modification: Date Descriotion July, 1976 Placed order with vendor April, 1978 C Aait report to NRC October, 1978 Receive NRC approval October, 1978 Commence rack installation November, 1978 Complete rack installation In

summary, the additional storage capacity will provide VEPCO with additional operating flexibility which is desirable even if adequate offsite storage facilities hereafter becomu

'railable. 5

5-11-79 1 4.0 ALTERNATIVES TO THE PROPOSED MODIFICATION The proposed modification has been chosen after an evaluatio_n of the possible alternatives to delay a possible shortage of spent fuel _orage. capacity. The following alternatives were considered: 1. Expand the spent fuel storage capacity at the nuclear station 2. Ship the spent fuel to reprocessors or commercial storage facilities. 3. Reduce the unit rating by operating them at less than 100 percent power 4. Shutdown the operating units when existing storage capi. city is depleted. 5. Build new storage pools 6. Ship fuel to Vepco's Surry Power Station 7. Store spent fuel at Department of Energy (DOE) Facilities 8. Ship fuel to other utilities 9. Physical Expansion of Existing Pool 30. Storage at North Anna Units 3 and 4 Each of the above alternatives is discussed below. 4.1 Increase Storage Caacity of the North Anna 1 and 2 S'ent Fuel Pool The storage canacity of the spent fuel pool can be increased by replacing the existing racks with racks of reduced center-to-center spacing resulting in an increased storage for a total storage of 966 fuel asse;nblies. This modification-wi.11 require a minimum of physical changes to the pool structure. he racks will be manuf actured of fsite anc will be shipped to che station for installing; thereby reducing disruption of no. mal operation. The proposed modification will not alter the external physical geometry of the spent fuel pool or require additional modifications to the spent fuel pool cooling or purification system. The proposed modification does not affect in any manner the quanti ty of uranium fuel utilized in the reactor over the anticipated operating life of the facility. The rate of spent fuel generation and the total quat.tity of spent fuel generated during the anticipated operating lifetime of the sta tion and stored in the spent fuel pool remains unchanged as a result of the proposed. expansion. The modifica tion will increase the gq 70d JC '/' 6

2 number of spent fuel assemblies stc red in the spent fuel pool as well as th: T.ength of time that some of the fuel assemblies will be stored. -he pool. The approximate cost of the spent fuel racks is $2,600,000, exclusive of installation. The cost of installing the racks is estimated to be 3100,000, which yields an estimated total cost of the modification to be $ 2,700,000. Based on the increased storage capacity of the spent fuel storage pool from 400 to 966 fuel assemblies, the approximate cost of the modifica tion per added fuel assembly is $4,770. No additional operating costs will be incurred as a result of the modification. Based en eccnanic and operational considerations, as well as existing conditions, the expansica of scent fuel storage capacity of the existing pcol provide: the most feasible alternative to delay a pcssib~ shortage of spent fuel storage capacity. 'Iherefore, this altemative was selected. 4.2 Shinment to Ret;rocessors or Commercial Storace, Facilities The shipment of spent fuel trom North Anna Units 1 and 2 to a commercial-fuel reprocessing or storage f acility cannot be relied on as ci viable alternative at this time. On April 7,

1977, the President issued a statement outlining his policy on continued develooment.of nuclear energy in the United States.

The President stated that: "We will defer indefinitely the conrnercial reprocessing and recycling of the plutonium produced in the U.S. Nuclear Power Programs." At a direct result of the current policy on reprocessing and recycling as stated above, these services are not expected to be available to Vepco in t ' near future. Presently, three private coreanies have unused storage space. Allied-General Nuclear Services (AGNS) has an unused storage space of approximately 400 MTU as spent fuel at its Barnwell Fuel Reprocessing Elant in Barnwell, South Carolina, and Nuclear Fuel Serva.ces (NFS) has an unused storage space of approrimately 95 MTU as spent fuel at its plant in West

Valley, New York.

However, NFS is not willing to accept additional quantities of spent fuel for

storage, and, under existing conditions, AGNS cannot be considered realistically to be a source of spent fuel storage capability at this time.

Even if NFS and AGNS would receive spent fuel for storage, the associated shipments would be uneconomical because of additional handling and. shipping ~ costs. Assuung appropriate shipping vehicles are available, it is estimated tha t the cost of shipping the additional S66 assemblies, which could be stored at North Anna with the use or high density racks, to the Barnwell Plant would be in excess of $6,500,000 (in 1977 dollars) which is equivalent 7 E ') no-J J /_. L/) 7

3 to $11,485 per added fuel assembly. This cost is based on preliminary cost data with no firm contractual arrangements. Charges for storage at Barnwell and subs equent trans portation from the AGNS plant to a potential Federal storage f acility or another commercial facility would be additional. The third company, the General Electric Company (G. E.), also cannot reasonably be expected to provide commercial storage. G. E.'s Midwest Fuel Recovery Plant near Morris, Illinois, is presently capable of storing approximately 700 MTU as spent fuel with an application before the NRC to expand this capacity to approximately 1,800 MTU.

However, all of this space is considered by G.E. to be reserved for other utilities.

Even if such storage space were made available to

Vepco, additional handling and shipping costs make this alternative economically unacceptable.

A similar estimate of the cost of snipping 550 assemblies from North Anna to

Morris, Illinois, based on preliminary cost data with no firm contractual arrangements, results in costs exceeding

$9,500,000 (in 1977 dollars) or $16,785 per added fuel assembly. Therefore, these alternatives are unacceptable because of

economic, operational, and availability considerati ons.

However, should storage be required beyond the time allowed by using nigh censity

racks, any of these alterna tives may be implemented.

4.3 Reduce Outout of the Two Units The amount of spent fuel to be shipped could be reduced by lowering the units' output, thereby extending the life of the fuel. The obvious discrepancy in this alternative is that the unit could not be operated to the extent possible and the amount of electrical power generated would be reduced. This alternative is not viable because it does not effectively use the resources available and would result in a significant increase in fuel costs to Vepco customers for replacement power from either 'Vepco owned f acilities or electrical purchases from other utilities. 4.4 Shutdown of the Units Assum ng that the storage capac:.ty of the pool remains the same and no offsite shipments are made, the units would have to be shutdown in late-1983. This is clearly not a viable or practical alternative. The generation prov ad by the nuclear units is necessary to supply customcr reau rements at the lowest cost possible. An economic evaluation would show that in a ma tter of several days, at approximately 3250,000 per unit per day, the replacement cost of North Anna generation would exceed that or the proposed modification. 7E 70 JJ2" L/b 8

~ 4 4.5 Build New Storace Pools, Additional storage capacity could be made available by building a new storage pool, either on or offsite. A detailed evaluation has not been performed because of the obvious large cost associated with such a facility. The current estimate is approximately $25,000,000 (in 1977 dollars) or about $22,007 per added fuel assembly. Another cost resulting from this alternative is the cost associated with double handling the fuel. Also such a facility would require four to six years to

design, license, and construct; therefore, it does not satisfy our near term requirement of completing the modification before the first refueling of Unit 1 which would result in spent fuel being stored in the pool or our intermediate term requirement of avoiding a

possible shutdown of the units in 1983. This alternative is unacceptable because of economic, operational, and availability considerations.

However, should the completion of the aforementioned DOE and commercial storage programs be significantly

delayed, this alternative may constitute a possible long term solution.

4.6 Shipment of Scent Fuel to Surry Power Station Unit 1 of Vepco's Surry Power

Station, located approximately 110 miles (road mileage) SE of North Anna, has been in operation since 1972.

Surry Units 1 S 2 will be equippel with Fl h density racks,- with a capacity of 1,044 spent fuel 7 ar.emblies. Shipment of spent fuel between North Anna and Surry could be carried out;

however, it

.has several distinct disadvantages. Shipment of spent fuel from North Anna to Surry would be a very short-term solution. Msuming a full-core discharge capability is maintained and the spent fuel from North Anna 1 and 2 (in excess of its existing design capacity) and Surry 1 and 2 is s tored in the Surry pool, the full-core discharge capability of the pool would be exceeded in 1983. If full-core discharge capability is not maintained, the pool 's capacity would be exceeded in 1934. Thus, the storage capacity of the Surry pool would be exceeded before the expected availability date of DOE 8s aforementioned storage facility. Taking into considera tion the critical need for these facilities at the time they become available, as well as the possibility of additional delays, it is not prudent to depend primarily on intra-system shipments to alleviate the North Anna storage problem. Therefore, although it is a viable alternative in the very near term, intra-system shipment of spent fuel is not advantageous for the reasons outlined above. In the event of an emerg en cy, intra-system ship,ents could be made. 352 297 9

5-11-79 5 4.7 shioment to and Storace of Scent Fuel at DOE Facilities As previously

stated, the DOE's spent fuel management policy calls for the DOE to provide interim storage capacity to utilities.

The DOE could do so by utilizing any of the three following alternatives: (1) present commercial storage and reprocessing facilities; (2) present domestic Federal facilities; and (3) new Federal facilities. However, the use of any of these facilities would involve substantial transportation and double handling costs thereby making this alternative economically unattractive at this ti:ne. In addition, it does not seem prudent at this time for Vepco to rely solely on the availability of DOE facilities because of the preli ninary nature of the DOE spent fuel program. 4.8 Shipment to and Storace at Other Utility Storace Facilities This alternative is mentioned only for completeness and is not considered to be a viable alternative. Because of the increasing lack of alternative

storage, reprocessing, or permanent disposal capabilities, all utilities are f aced with the same storage

. problem. Even if other utility pools were available, the economics of such shipments would be unfavorable. Again double handling would be required and would be similar to the other altermtives discussed. 4.9 Physical Expansion of Existing Pool The alternative of physical expansioa of the spent fuel pool would cousist of removing one wall of the fuel pool and expanding the fuel pool in the direction of the removed wall. At present the fuel pool is bounded on four sides by existing structures necessary for the operation of Units 1 and 2 at " orth Anna. Those structures are en the north side the auxiliary building, on the west side the unit two containment, on the east side the unit one containment, and on the south side the decontamination building, the waste solidification area, the waste gas decay tanks and the primary water storage tanks. The work, time, and money involved including the moving of the structures on any side of the expansion of the fuel pool would be in excess of building a new fuel pool, which is discussed in Section 4.5. This would have to be done with no spent fuel in the pc l and the power o station would have to shut down during the construction. 10 %2 298

5-11-79 6 4 10 Storage at North Anna Units 3 and 4 North Anna Units 3 and 4 are expected to be complete in the mid to late 1980's. This is too late to avert a loss of full core discharge capability in 1981 or a loss of refueling discharge capability in 1983. It is difficult to accelerate the completion of the fuel building because of its early stage of construction and because of dependence on the service wate. and component cooling water systems which will run throughout the facility. Licensing and conotruction problems make this solution unreliable at best. 4.11 Summary Based on a review of the possible alternatives, the proposed modification is the best and has been selected to be implemented pending NRC approval. 10A 352 299

5-11-79 5.0 EXISTING FACILITIES The ey:isting spent fuel storage pool is common to both units and has a total storage capac_ity of 400 fuel elements. Figures 5-1A and 5-1B show the arrangement of the building. 5.1 Fuel Euilding The fuel building is a Class I seismic structure and is supported by a reinforced concrete mat on bedrock. The spent fuel pool is a reinforced concrete structure with a 1/4 in. stainless steel (304) liner. The pool is designed for the underwater storage of agent fuel assemblies and control rod assemblies after their removal from the reactor. It is designed to acconcuodate a total of 400 fuel assemblies and a spent fuel shipping cask. These assemblies are stored in 25 fuel racks, each containing 16 fuel assemblies each. In the currently installed racks, fuel assemblies are placed in vertical

cells, grot: ped in parallel rows having a

minimu:a center-to-center distance of 21 in. in both directions. Restraining

clips, which are welded to the floor embedment pads, prevent lateral motion of the spent fuel storage racks.

5.2 Fuel Pool Coolina and Purification System The spent fuel pool is equipped with a spent fuel pool ecoling system to remove decay heat and a purification system for maintaining fuel pool water quality. These systems are shown schematically in Figure 5-2. 5.2.1 Design Basis The fuel pool cooling and purification system is designed to: a. Remove the residual heat produced by one-third of an irradiated core 150 hr ~ after mer.or shutdown while maintaining the spen fuel pit water temperature at or below 140 F with one fuel pit cooler and associated puT.p with 105 F component cooling water (i.e., normal condition). b. Remove the residual heat produced by one irradiated core 150 hr after shutdown and one-third irradiated core 45 days after shutdown, while maintaining the spent fuel pit water at a temperature of 170 F or less with one pump and two coolers with 113.2 F l component cooling water (i. e., abnormal condician). c. Remove soluble and particulate impurities from the water in the spent tuel

pit, yc-d 1

e. r.s e. e 'I y I Il b-

l....,.I s :.d

_5 s a 5, ~4

sy 3,

.a

1. L '.

{ s r, IE g i I ( i /- { gM~.- ~ .. n u '5 3 - 8 Eg p% e 3 g

..l 1

y Et y/ p :,:'?! i 5 z 'i ., ;' ? s, ~9 b h* M. 8 -i, .; z 5-c . ; . j ao- ./. n t/._.

s. o

%; e. 'c a* e r ./ t g J.z. t M ;. te v ", s..! F w e-v.a_ __g- "b L[. l --- a s v;y. - n o s,.i 2g5 @4-s, ._ z m - t a a s -i (t h d.! 4--- y- .a.... .I h. b' -77 4. a, p +; l N:y. ["b g r >... l f 4 t-g._. MI= =_.,iLb._ 4, - hl".lilCf$b< y4 -4e 3 k, '*w ~ 4 4) ! r-- -._? f^ .~y I .'. L,,, ;_ _ m;--r-- _- ~A is l.u. N q. . I f_ ' !! M. > ': to b.ii% -w +i C i - 4 :. - J. n ter

c. - -

u o p: 1< L W. -M (Ze,- - 5-ps_t -N ,2-_. .. -.-.,>J m" J s -- n "r=== ,,"1.-- ---4___.--m.,- -..,=s,

i

.m_ 9.. ,-4 g_ y -n g / i,#

f~

.l

y.. L +\\,,. n,f b,w.,[~ 'Ql;,,

"I f l-w= n I }, y2 z-v. j' L - 9 - i 1 e - 7 4 m,., 5 I 4

s-e; a

a a-ii. p. , a. - - -5.5-3,. I!1 $..s t a x. ~~ I. . 1' .G ~ _j- ~ tf .s uJi -6 3 _. a r-g_ gi a -dCi 3 r s le t H

t g

,U. - @*,t ! h; ' i. S

L CE h

h I i M / cu,_--- - - - s.t 1 2 -, b! I.is 3- --= !..w e m.. - Ef y]s._.s,h 45 a 4 a c;, I g, -- - . p. l p 2" 3 ;u i, .e of I... i _ e T.' ..r_ - !/,._s p "y, A ~, m. a.--- ---- p 'g yg - M,. ad,_._,g g P '!9-1 '- o i- "? T i sd O7 I-: - ~M"C#',c cf*t'v4 -5 5 2f = r A [" q. l. i2: ' r.* r W ~ F' N J 4 ' ! M. : 1 }- .I Ed 4!' N t .j-1, Lj ,.:e. p r--*l1 ~ .g o 3.,"-*s* N ii.# 6 N h ! Eif - s 3 Q JT-a SH l j ff 'l M l N1 :.#.!l [ ! .'J'C"Ml. ~s 7-f " bi {lklie D b-I ' . l. ,l f l$ [ ' ---- 5 Fi . b m. c. d.j p li4 i.. e n: q:L; LvM %.r_:=m d - ef-ll C.. g A ._-c-g* 9-l. I 'l

d. 3,. " 'v "- nt. -.

,7-N I '- T h. ~. M....f , :_. rm 'c.,,,a ni. u- --.,s_i i L_. =4, rf 1 it it *y A ^ ~~ ~.M.;. Q i../..i l

s

._U.

e e --

i t.> ~.2. >';i.__.

,7,

C-- ^ i.^- d e- $b. i .-i o..,- u o -, q

...e

..,,o ,m i... ,h .I 3J' N-b- b ' h ly-

l i s i,H ' M G.

h y

' E.*.35 $ U N ni I 2 . d' 1 N- :q.:s"' _,... M:.' d I ; m '?in' + O' ' ',. h idi ,'p Y .s n. m., ,.3 om, e :.o.o i p- . c: ::

a

=.....,_,------y .> e i W.~r..% p Ei.h _."10.lb _.i %, . -01 4;W-N i S .--m L l tf m -1 a o, - t ff ... f l 7._2 _ t - L - i; t< 3 r', 9w,,_~ g _ ;) g W2.r "' w _,_j4.._. p ' 4, (..u... 11 a ,1 l 1: L-. c- = -- %; I

.....,5

-r e 4 4 i .r -min,-.i. t i .-' I ,3 - i..- / .-V i i SC l l;J-g g,;. je .h .1 h 4 3 1 s = = - - -e t p ~ ! NP I 'D 2 'r' - : L l u IIl l l M)h }7Q * ' ' : N I..' I i' I . n m 3 sl

c

-u i g .s ! l N, b 3

C

- - t Li ; ..r om -L a v. 3, a t l' a i s s'N. 1 L d.*-Y-M. 1-oi dyl '{ 2 ~ - ~ d,}%4-2 j. 2 =% a i s'

j;. -

"[_} $_. s.- =- i ,i!.c w..'.-,. M, - - - -, ' 1 8 % \\ .r ,e # S ,i i

-i w

l J h 7 1. ;lu. e, h YM'; 4 . ;i - g. 'i' h I { lN l n .9r s .J r: ar .i ~.....m .~a I' h-l %-47."M. ; d ! ha li U p.g,/ \\ j!M i .-q i in .s-c. '" i * !. ",.;g d. a .-s L.. - t n, s. l 3, i I_. f. 2..,, h. j .e - - b '"-L y-1

(

=~ = -2 8 aw .a I s. h y-_ 4,.. _. _ .C _,. 1q -

..: , a.,

se % a ~ 5.. i...- .h r ! ;. q F-r ..s-I

i

-m-3 I L '2 !. O p> h U 1 id l: D i G; '! $. l r-

f4 1

d i d i "d l rr-

p. p;m_E _ --T.-

. l ~ m2 J 1 l _3 i W ; %. ~.~,_- - i 1 .m .a. e =m .,..l*, - f ~,: a w J 4 .f.- ..J. ! ; i ) .y ..- 4 3 [-J di g %~- --' j 4 T* 6 i. = I! l It is 2 I, I t= t' e (.., V b. 3 'v. 2) I 6 g

g l

]) j.. * .h n e. - ~ ~

r L.'

l' T6 2. _l' _. _. 3" _ 7 J ME' 7qj ,.; L. 1 ~9

3 __..P i o --= -W,, g c,_ rl-- L.- p =W m L' q-8 Q{.*fs T

l h'h, p_

M e e c e.; 8 i a ca : ..s.. 3 -w a. A b - -h I' I I{ 2 8 .-C E Sw i,_-.. a q -/ .? e$ - - -. r_ _-- fa h' q 8

. g

_y, y / p^- __ -a .J _. - 8 Syp ? 5 =; y I 12 _ 7 - - - ---r [ a-- c, t,7.4i, o - 4 . n z o 2

we l

l G G W, g .'l g. n iMr w$E f l h j! :. 6 a I 'l j @, --= i sl f -* ~ j !- 4 l <s Q E. [-[- - - '--- [.~_.. f, I e-i d ,_=u.-- 2 I in g h--O73 ,,4 3.,3 l [I T-T l r-! 3 %E~ A:; 3 [ n. 7 $< y ' l* ! ('l b_h. f f _.- __ _ /,_,2 . II I l .Cb -ht-Y { I g ,e I t-g - + y .u g p d .r4 l 'l

y I 4

t. .e { - } - t g

.___=.

s' s !,Ti l1. ? t----- l -1 c-l: p 4 1. P: .e - i :# 2, a2 4 ;- _.

m.. _,.4. g.I p i S.

a 34 1-j o i h-a c. L i< '7 hg l irl s ! % = =: q y l/ * 'j. Ets i h; y/

- .J [I ~.

t.# ^^ i fll )s -h* i 4 l . !f, ;/ 3, {- ?;1 P *.__..- M L - f I cr=- - T i 3 Y r l. M ++ .g 39.t_+ _ m^.-... 3 4,, : - -- - L i.; '1 *.4 ia 4 n2 4 r Ja 2

G t2-

~ ~ - ~ a

M, W -

i ljiN M,. t c'w --) l N r t i ? v 3 10 N,fa 9 I S e I 3.- JV l' y' = r. e :.1 jI p 5 S. }' w e g-l$f 'l W,I P - g 3 f.

l a

/ 4 I, !y , Im ,h, 5 - T a s s-a3-(? / j

8 l b; ' !y

.b, 0 (P,-

b\\

m s I: N% 4 -- e !.l-c! 4 'i N - c ,1 d1

ij !

2-4 1 5; %r $ !}S v 8 'd -c . [h M<-i, %_. 8 1C _... h. m 42 --#. - 'a g .. T_._. _._. _ -]

f..

t-i.t.r-- ...._... J.. H ea '. y*

-*c--@\\_ LJ 3

I y ' '":;"" " N 8 i ~ I l* -',3 'a^- Tl i. hU-.J!}!'!;I,l.. v. l r.i ' hi." 3' ji II ', "j'-i.' 's -<= s s r

. s,

e y,, e , i, ?- --- -- ul 9: i. [Q',M jif * ? fi_jL 1 * * .'I U,f f / % 57' f' C I,,'; 2 C !"-..l d 1' U ,7 Is f. i s. 1.-

-T ja I

j. ! s' - --' O r h "id U! l! i w6 ' '" d. .. [J - -, -. O.= N'! Eg 2 E' y.. u m- .2* ~ T: 2 _,__sr, s, u 1. l l- = ,9. i ?- r:., ,.i 7 t e-a qw, .-,.n--_- m- '_ ~r w T a 'N .. - n x*':' __ *- ,y ;' I

?O -~..

t, s gt :. r- e u r' i :.' ' = ,r [ u-L I

p.I E

;* *. i ;-

T_~_%__ ~ k 3: y s I t1 ~-

e...e'<'

e

3-il

_!,' 'r g, 4.2 s,

- q

= i*i ll -3 1 i*- a -y n 3 4* ' f r-- u t m _., f i k' [/ 4 3-ll .Vw 1 I' I $f p l L i 7 l ( .I; 11 9 lI i l I=3 -.= +. l' af 'f *l b - ' M. / H l gs b_- ug---(. _ _ _ -_ _ - NT-.. -.. Wll dC P l. e l -r 9 -;

l t

~ j. - _.-i - -- Wp, ; -Q,.., ;; c -u It'- S-y 4 -1' F-r- -._ r. 3,s..g r. adj ^\\ ) I.

~.1<,i., l.

.t. q o s r N 4 i *'

}/

2 e i a. /.46(?('.p '-hg [ II s"-' / t i ~w a E \\/ ,g ( "i f 1

- g

{ ~ ~ ' s I4

11) h@ hh u bid it l*'

-[G / JJ/ Jv.

T b [ ..... _ '.. -. _ _ - _ _ _..r 3 ',[ s { =.,,.s. I (? 9 2 {i . = - .,.-6 z.t 4 .d h o ,1 e "* 9 oo -r s a 3 s o i ,1., ,3 l' t j-i.". r";*7.",l. p' 1 -, e--Q' l - } y at u 7, __) H ' +-;f M j 3p i:f y. -.i. t e w 't i i i - i 30 5 I r as 1.- g. 3 i. o c. 2 8 i to a ; !g %.s, O .!{ i i e e r4 I v -.,5 z i - i o [* n .t,..r-e. ' 1_

s. t f s:

8 tw-ad_ v-3 s I t .s. j ^_g! - si 4 g ,2 e " g g3 T to 3s t e't .J 3: ,.). r. q/;. .i Owiz 4 p i

u. x 2 0 i

3 l Q3 t fe il i j N 3, L t. _ _. -... - -. - - - - -(_/' w. a + y i I ,m, vs k I' h

1. E-(

i il .y o 3 d4, s-;

n+t -;

se*h, v g, 'I t

g
4,.

'; P - ,;) i;b 3 )H tk., %'d ~{ i / -4 e Er 2- 'i G1 s 4 i;j--------Hr-------- q 4 aju- ? 3

j p

a i; s35 ; 4 n r -t fal j f -dA t* n. ~J > - t i. .t.$ [$, 4 t -. <. i e. q i j e, n. n. s i

4. !...,'

m i

r 1

xs .t i i = , l i; ., *. r. g. e -i i., m p? . rr 4 gu. . T ;. t.s

p wa--

s y 'si,_z: S ! :' *W f / IN e . ' % I: 1.e 3' p' i. , n.1 st =t < /.- e g, t i [ s 2 g 5 ie { 3 i j j' l! N .h _ \\,1 j i. r-i. 4 t,,/ y :( g* n, s % t._.,.,r m'. 'a -.2 e n n. 2

s. ;;

e e 1 -( 3 M i. P-.

g;

til,n 3.L. > : -d,,, W.4 1; .' I ' y".1 % b'te...u. ;., Ed. / i -l t i s .. 5 -- n, -l 3 R: I* i ~ N,. % h, p. Y {.g W

4..'.' T - S. Q I.. < '...

H= M! ; b E d., 3 .l-M6 < -Q',.. - Il.1 t I. ! l ",'53*n' 44 t/;i. i: %,y

e. t !

i; ce.

[

6 i.i,

- e.

lI,.,,5,: 'S! f r. i 1 .{.i t f.- , b. .2

3 o

..c -s ,t ..; n ' i' :sj i s f M,,. + 1' :M M-- . -- w w'. n,J.['. q, A p!' (s; m -m Al s:. pl

I.M ;p. / --) [ { - Ti s F. p> M], 't :.

~ n.o a. l> !.I ' k,! ( $ y, "j ) M ; F,.;j 2 q,1 - i- - A -.- 1. /. V R. A . il ' ' ;_w: :d i l.-; f 14 .V ; l' ' 1' m'. L ' ' I'" $ )0:i b r i. ] '- "- ? ? '. / ' 3 M Is I N ' i -- ' t p } a i, 4 i; t'h -n 2 5 i i C O M j'.21 r. i l' a .ti )'~ ? !,Q Er il . p): i i d' $8 a;' 'y .8, m -Q2 : } _,

r- -

'" 1 : t 5 1

d. 2 th

[ F v..,y, N ~.. i.'* ,a. 9 y l 5

n....,, t i g.:_(( J i i

e. --,r,..l N

o--,._-_w., F .u i - ;-- f , -~ -t -.l.,,,- g & e'

- -%/~p. 3i j ct.Vl '.
- * }.

. ~r ~

[@

-s i z.

i h L.: s,.a e lyk-L 3 i d 6 i c .5.,.#_r7 "t 5 r u, I ' I,. c rp p.t 3# ; 2-99 1 5 .t. } r~

s
'. *.,l

. I <, i u e s : _ =a - i.. n.,

3 I-

-[1 Q'i s i..J. d ;.,.pg.n..i2 !'M "l iI t, s ,( l ..n 6,.. s : "m.. !.: .te / 2, 3 -, ~ y'} a 's;is I A 9 N C. 3 _.y ' ~ .f. P ,i. :

g.

-c-q,.. _4 .i 4 ) ;, i : , i)]- a w.; s .a -. t' ' E

_----._----_p

.. t,; y int- .f .- e.* y- .} j m,

j p

,pr..gir, n p i, yLlb 3 :q ? .i 2 d; 4 ?' = i n 3 3

i. 1,5 i

.i. h 9*; r,, v, - -b. ,l i 2-i: - o 3, {Ii...yi: I:

, ;;.:.. - b p.'

1 st 3 l, T _ l b, - )- 2. I, I g{; n". ,j, H =; , ~. e i ?,- 3 ;;s. (( -

I

,ll ',0 b

a

.T3 ?y .-l 2 { na e i

q i r

4, d h'.. 2q ji 3 i j 1 j 2 i

e..

e< N 3 L .,M ti !i, -.._...-__-__J..____._---_.l ?.^ e 7 1' l e m. '-t a..' A a' 4' et

9 r-)

e J/

5-11-79 either reactor refueling cavity, and either refueling water storage 'ank, to maintain the cavity water optically clear and radiation levels within acceptable limits. 5.2.2 Descriotion The fuel pool cooling and curification avsten has two shell and tube heat exchangerc, tsto circulating p=ps, and tbrec KO rercent capacity purification pumps, all located in the fuel building. The heat exchangers and pumps are arranged for cross-connected operation. The heat exchangers are cooled with component cooling water with service water available as an emergency supply of cooling water. The purification pumus take suction at the outlet of the fuel pool coolers and pump pool water to a demineralizer and filters located in the auxiliary building The demineralizer or the filters can be bypassed if not required. The water returns to the fuel pool at the end of the pool opposite the suction point to assure mixing. The purification system is run independently of the cooling system whenever purification is required. The surface or the water is kept clear of floating matter by two permanently installed skimmers connected to the suction of the spent fuel pit cooling purns. 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. The spent fuel pool cooling and purification system is designed as a Class I seismic system. All piping, valves, and comconents which come in contact with the fuel pit water are austenitic stainless steel. Design data for components of the system are sumnarized in Table 5-1. 5.2.3 Svstem oneration The existing spent fuel pool cooling system is equipped with tuo circulating pumps and two fuel pool heat l exchangers. Redundant piping is provided from the fuel pit through the pumps and coolers to the main return header located above pool water level. The 400 gpt filtering rate of the purification system results in a clean-up half life of 27 hr and maintains suspended solids at a low concentration for optical clarity. The skinner 1 }g/ yp/ 15

TABLE 5-1 FUEL PIT COOLIllG SYSTEM COMPO! TENT DESIGli DATA Fuel Pit Coolers Number 2 Design duty, Btufar each 56,800,000 (with tube inlet 210 F and shell inlet 105 F) Shell Tube Fluid flowing Component cooling Fuel pit water water or service water Design pressure, psig 150 100 Design temperature, F 150 212 Operating pressure, psig 110 45 Material Carbon steel Stainless steel type 304 Design code ASME VIII, ASME VIII, Div. 1-1968 Div. 1-1968 Soent Fuel Pit Pumos Number 2 Type Horizontal centrifugal Motor horsepower, hp 100 Seals Mechanical Capacity, gpm 2,700 Head at rated capacity, ft 80 Design pressure, psig 125 Design temperature, F 250 1. 16 )

TABLE 5-1 (Cont'd) ~ Materials Pump casing Stainless steel type 316 Shaft Stainless steel type 316 Impeller Stainless steel type 316 Refuelin' _ Purification Pumps NumLar 3 Type Vertical centrifugal Motor horsepow_r, hp 20 Pump capacity, gpm 400 Seals Mechanical Head at rated capacity, ft 99 Design pressure, psig 185 Design temperature, F 200 Materials Pump casing Stainless steel type 316 Shaft Stainless steel type 316 Impeller Stainless steel type 316 Refuelina Purification Filter Number 2 Retention size, microns 3 Filter element capacity, gpm at 5 psid, normal / max. 400/440 Material Stainless steel type 304 Design pressure, psig 150 Design temperature, F 250 2. 352 300 1,

TABLE S-1 (Cont'd) Refuelina Purification Ion Exchancer Number 1 Active volume, cu ft 45 Design pressure, psig 200 Design temperature, F 250 Demineralizer resin 50/50 cation-anion Materials Stainless steel type 316L Design flow rate, gpm 200 Design code ASME VIII, Div. 1-1968 Ski:nmer Asse:r.41ies Number 4-2 in spent fuel pit, 1 in each reactor cavity Debris-basket' 1/8 in. x 1/4 in. openings Design temperature, F 210 Plow rate, gpm each 25 (min) to 55 (max) Material Housing, base, and deck High impact grade cycolac Plate frame (Acrylonitrile-Butadi en-Styrene) Debris basket Polypropylene Gasket Buna-N Screws and springs stainless steel 300 Series 18-8 Fuel Pit Coolina Piping and Valves Materials Austenitic stainless steel Design code ANSI B31.7-1969 and ANSI B31.1-1967 3. 18 352 307

filter rcmoves particles which f all and float on the water surface thus reducing the amount of impurities entering the water and reducing surface refraction. 5.3 Fuel Building Ventilation Svstem The fuel building is equipped with a ventilation system to provide high-efficiency filtrstion, heating to inhibit the buildup of condensation, and exce s exhaust flow to maintain a negative pressure in the Lailding to prevent outward leakage. 5.3.1 Descriotion The fuel building 'rentilation system has two supply fans, one to serve the spent fuel pit area and one for the remote equipment space at El. 249.33. Both take succion from a common plenum fitted with a combination roll and high-efficiency filter (95 percent atmospheric dust spot efficiency) and steam coils for air tempering and space heating. The exhaust fans discharge through the ventilation vent and are arranged for selective bypass through the auxiliary building filter bank. The area of the ramote equipment room subject to radioactive contamination is exhausted by ct branch from the decontamination building ee.haust system. The design provides (1) sufficient air at a tem-perature that will inhibit condensation on the overhead structure to uvoid drippage into the pool, (2) high-ef ficiency supply air filtration 'to mininine dust clouding of the surface, and (3) supply air distribution to avoid ruffling 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 th;ough

he ventilation: vent, with the capability to bypass through the auxiliary building iodine filter bank.

The exhaust is filtered continuously during irradiated fuel-handling operations to prevent the spread of any possible airborne contamination through the exhaust air system. The frel building exhaust also discharges air entering the fuel building from the tunnel between the fuel building and the waste discosal building-5.u Instrumentation Anolications Instrumentation provided.gives local indication in the fuel building and the auxiliary building and remote 5 08 19

indications and alarms in the main control room. Unit 1 control board indication and alarms include: 1. Fuel pit temperature indication 2. Spent fuel pit temperature at >140 F and >170 F 3. Spent fuel pit high/ low water level with the low-level alarm 6 in. below normal water level (El. 289.33) 4. Start /stop switch for spent fuel pit cooling pumps with run indication on both Units 1 and 2 main control boards 5. High differential pressure alarm for the refueling purification filters Local indications include various flows, temperatures, pressures, and differential pressures. The system instrum-atation, including the spent fuel pit level and tt - g erature instrumentatimn, are calibrated on a periodic basis. 352 309 20

5.5 Ooeratina Eroerience at Surry Power station Based on the operating experience VEPCO has received at Surry Power Station, Unit Nos. 1 and 2, similar results are expected at North Anna since both are of similar design. The information that follows has been extracted from information from Surry Power Station. The operating history of Unit Nos. 1 and 2, Surry Power Station, has oeen reviewed in light of the fuel pool cooling and purification system, ventilation system, personnel e::posures, etc. The purpose of tuis review was to confirm the satisfactory operation of the systems and to provide baseline data for estimating the effect the proposed modification may have. Unit Nos. 1 and 2 have operated for a bout 45,480 and 42,360 hr, respectively, and have generated a gross electrical output of approximately 43,179,352 megawatt-hours as of February 28, 1978. The operating experience at Surry Power Station has been grouped into four general areas for purposes of discussion: 1. Performance of the spent fuel pool cooling and pur-ification system 2. Environmental conditions in the fuel building 3. Ventilation system 4. Radiation exposure 5.5.1 Performance of Scent Puel Pool Cooline and Pur-ification System at Surrv Power Station Operating experience with the spent fuel pool cooling system has been excellent to date. As of March 1,

1977, 292 fuel assemblies were stored in the pool.

The pool water temperature has been maintained at about 95 F all year round. This temperature has been maintained using only one train of cooling, i.e., one pump and one heat exchanger. During normal operation, both trains have not been operated sbmultaneously since one train maintains proper temperature. This system has performed satisf actorily. The normal flow rate for the purification syst :m is about 110 gpm and remains in operation con-tinuously_ to maintain a clean and clear pool. The maximum allowable differential pressure across the filter is 15 psi. The maximu:a allowable differential pressure across the dealinerali2 Or is 25 psi. If the pressure drop across the filter or domineralizer exceeds the 3 7 E ') 7 1 ')

allowable

value, the filter is replaced or the resin is replenished, respectively.

The radiation levels at the demineralizer are usually from 1 P/hr to 4 R/hr. The filters are normally changed because of high pressure drop and usually have radiation levels of about 100 mR/hr. The filters are normally change prior to each refueling, i.e., twice per yee assuming two units are operating. In changing filters an individual receives an exposure of about 150 mR. In replacing the resins in the demineralizer, approximately 55 mR is received. This exposure is divided among three individuals. 5.5.2 Pool Environment Conditions The spent fuel pool purification system removes both radioactive and nonradioactive particulates from the pool water. The purity o1 the pool water is normally maintained between 0 to 0.3 ppm, with a maximum particulate concentra-tion of about 0.4 ppm. This purity level provides sufficient optical clarity for refueling operations. Based on samples taken since station start-up, the following major isotopes have been detected in the pool water in the approximate concentra-tions indicated below. Concentration (micro Curies /ml) Isotone I;o m a l Maximma Minimum Cs-134 0 1.2 x 10-* 0 Cs-137 10-* to 10-5 1.3 x 10-* O Co-58 10-3 to 10-4 1.5 x 10-3 6.6 x 10-* Co-60 10-3 to 10-* 1.1 x 10-3 4.8 x 10 -* I-131 0 6.5 x 10-5 0 Gross 10-3 to 10-5 1.1 x 10-3 5.0 x 10-5 Activity Crud buildup along the sides of the spent fuel pool has not significantly affected the radiation levels on the edge of the pool. S.nea r samples from the sides of the pool had the fol-lowing activities : }a') L ~&

Isotooe Concentration (micro Curies /cm2) Cs-134 1.53 x 10-* Cs-137 5.50 x 10-10 Co-58 1.47 x 10-3 Co-60 2.54 x 10-3 The crud buildup on the sides of the pool is removed with hydrogen peroxide (H202). 5.5.3 Ventilation system Exoerience at surry The ventilation system in the fuel building has maintained the levels of radioisotopes in the atmosphere in the building at acceptable concentrations. During normal station operation, i.e., refueling is not in progress, the gross activity above the pool water is about 1 x 10-11 to 1 x 10-10 micro Ci/ml. The principle isotopes noted are Co-58, Co-60, Cs-134, and Cs-137. During refueling operations, I-131 levels of 5 x 10-11 to 5 x 10-20 micro Ci/ml have been noted. Other isotopes, Co-58, Co-60, Cs-134, and Cs-137, have been noted in the concentration of 1 x 10-10 to 1 x 10-9 micro Ci/ml. Tritium H-3 and Kr-85 have not been detected in the fuel building. As stated, 1-131 has only been detected in the fuel building during refueling cperations. During refueling, the fuel building ventilation is directed through the auxiliary building charcoal filters, with a decontamination f actor of about 100. The fuel building exhaust contributes about one-half of one percent (0.5%) of the total I-131 released from the ventilation vent during the refueling period. 5.5.4 Radiation Excosure at Surry Based on 208 fuel assemblies stored in the fuel pool, individuals would receive the following exposures for the locations indicated: Exposure (mR/hr) Location 3.5 to 5.0 She in. above surface of water 0.7 to 1.5 Waist level a'~ the edge of the pool 0.5 to 1.0 Fuel bridge 352 ?-I? 23

Based on working in the fuel building for a 10-hr day, an individual would receive approxi-mately 15 mR. 5.5.5 Radioactive Waste Exterience at Surry The spent fuel purification system generates certain radioactive waste wnich must be shipped off-site for disposal. When the filters are changed, they are placed in 55 gal drums for shipment as " low specific activity solid waste." Two 55 gal drums of solid radioactive waste (approximately 15 f t3) are generated annually. Skimmer filter changes on the average produce an of solid waste annually. The additional 15 ft3 spent fuel pool ion exchanger resin is replaced twice a year producing about 90 ft3 of solid radioactive waste. Thus, the total radioactive waste generated annually by the spent fuel purification system is app:oximately 120 ft3 Approximately 60 ft3 of solid radioactive waste is associated with spent fuel storage for each refueling. A typical refueling outage produces 'about 1,000 to 1,500 ft3 of radioactive waste. 352 313 24

6.0 DESIGN OF NEW HIGH DENSITY SPENT MIEL STQMGE RACKS 6.1 Introduction This section presents a detailed description of the modified fuel racks and comprehensive analyses of these racks which show their capability to safely store spent fuel for an extended period of tLme. b.2 General Descrintion 6.2.1 Present Design The spent fuel pool currently has a capacity or 400 fuel assemblies of the Westinghouse 15 by 15 or 17 by 17 array design. The present storage racks consist of vertical cells on a center-to-center spacing of 21 in., which are fastened together to form 4 by 4 array racks. The racks are designad for peak site earthquake accelerations of.24g in the horizontal direction and ? .16g in the rertical direction. The spacing of fuel bundles in the spent fuel storage pool is designed to maintain keff less than 0.90, even unuer accident conditions. 6.2.2 Hich Density Storage Rack Desian 6.2.2.1 Desian Basis The following design bases apply to the high density spent fuel storage rack design. Other design bases for the spent fuel storage facility remain unchanged. a. Puol storage space .s provided in the spent fuel storage pool for the 966 fuel assemblies. b. The center-to-center spacing between stored fuel assemblies in a fully loaded rack is sufficient to maintain a keff equal to or less than 0.95 for the normal wet condition and for all abnormal and accident conditions. This design basis is met even with fresh fuel of up to an equivalent 3.50 w/o U-235 enrichment (44.2 grams of U-235 per axial cm. ), a conservative water temperature of 68 F, and no credit for either fixed poison in the fuel assembly or soluble boron in the fuel pool water (this design basis is discussed in Section 6.4). c. The modified rack design precludes storage of a fuel assemoly other than where intended, in the racks. d. The modified racks are classified seismic Category I and are designed to witnstand tne effects of the Design Basis Earthquake (DBE) an' t yet remain functional and maintain subcriticality. 71 JU/ Jl'p 7 ~ y 25

e. The modified racks are designed to withstand either a dropped fuel assembly or the upward force of a stuck assembly without loss of function (structural analysis is discussed in Section 6.5 and accident analysis in Section 6.7). f. The racks are designed to allow adequate cooling of the stored spent fuel assemblies (this is discussed in Section 6.6). 6.2.2.2 Desian Descriotion The high density spent fuel rack design utilizes a center-to-center spacing between storage cells of 14 in. to increase the storage capacity. Each storage location consists of an austenitic, Type 304, stainless steel square

tube, which is separated trc m adjacent tubes and held firmly in place by steel plates welded to the sides of the tubes at several elevations.

Tae tubes are grouped in this manner to form upper rack assemblies, which are welded to a pre-assembled welded base. The base consists of beamn and plates with an opening at each storage location to receive the fuel assembly nozzle and allow cooling water to enter the assembly. The racks are free-standing but restrained from lateral motion under seismic excitation by reu tcaints welded tc embedment plates in the pool floor. The modif ied rack design and spent fuel pool layout of the revised racks are shown in Figures 6.}-1 and 6.3-2, respectively. The mechanic 21 design is discucced more fully in Section 6.3. 6.3 Mechanical Design 6.3.1 Fuel Rack Description The modified spent fuel storage racks, shown in Figure 6.3-1, cons.i.st of square stainless steel tubes of 1/3 in. thick Type 304 austenitic stainless steel. They are spaced at 14 in. center-to-center by Type 304 stainless steel plates. The plates, which are also 1/8 in.

thick, are welded to the sides of the square storage tubes at f our elevations.

The tubes are flared at the top to permit easy storage and retrieval of the stored fuel assemblies, and to be compatible with the fuel handling equipment. Three different rack cell arrays are utilized to maximize use of the available fuel stcrage space in the pool. The upper rack structure is welded to an elevated base which is a system of welded beams and stiffeners. The ba m.e serves to support the weight of the fuel assemblies and to distribute the load on the pool floor. The base, which contains an opening at each fuel assembly storage location to permit coolant flow, accommodates the fuel assembly bottom nozzle. Natural circulation of pool water flows down between the storage tubes and up throt - h the bottom nozzle and the f uel assembly to renova j decay heat. The storage cells are designed to provide lateral support for stored assemblies of the Westinghouse 15 by 15 or 17 by 17 array design and other assemblies with the same external dimensions and similar lower noncle design. ~C' 7jc )J JsJ 26

75 / \\ / \\ / \\ / \\ ,/ 1 i i i i .e p p e .i 4 e g g g j l t i e e g g g e i I e e g e q I g I I I f 9 I i a i e i s I i 4 ,B i s 9 8 8 f g g l i a i l i i i s i 1 l l a e l 185.50 i i i i l l i. e i i 8 g g i 1 .i g i. .i i .i i l i l l l l .i i i i i. I - -. -. _ _ _ - =... - hl. ei .li li l_ #

1 i' ei e :J I Jl i

d l' Figure 6.3-1 Pack Elevatica 27

T A '- t *. II .n u..., s s g _2:2 _L'. _. ':,; c,733,, ([3) 4r i r*4 g e,.,,. i.,;. c /@ \\ n... _, t __\\ _. 3... _ b _. _ __w.,, < ~ ~u c ~ 1__7 l l _ _ _... L / l / \\ / L'// -@/,_____/__. i,/Z.. N ,./-- ~~. /// ,l a , /l /,,.' f,/,,-( n C / -=s-a n / s n ' /, ' 's,. l' '$,'/bh y <~ --( . \\. . (. l ^ c A A A A m _-. =.._..___-.;.___., _..._-3 4 ,.,,s tO p.a. w.w.... j co i l C ^ C A A A A l l l ... u. I n u. c.1 l J 1, =_= m _m.; L.- .._:----.. _ i I j 'lll f] El LJ E 1 . l f_l L.f.]l El f' q "l 11 i ^ A A C em C A .. lL lIL: i g ll~] l u__ n. -.i. u(i n_.n. _,.I_, Llt il l i i i ._....__J' . J l. _ _,, o _. _.. Im I .J ..e,,,,,,.<.. ,,3,n. _ _.... .___...__.___.__._-____.,,s L .... e -.~... o ua% _._ _. -. _ _. _ _.. em~ s ~ e L _. u__ ._.) Figuro 6.3-2 - SPENT FUEL POOL ARRANGEN.ENT

To prevent lateral movement of the racks, particularly from seismic excitation, the rack legs are restrained on their outer faces by right-angle plates which are welded to ambedment plates in the spent fuel pool floor. The angle plates are positioned to allow for thermal expansion. The legs, which raise the racks off the floor to allow natural circulation

flow, are square with gusse ts for reinforcement.

Shims are used to level the ra ks during installation. Each corner of the one 6 by 4 row storage rack has a floor bolt which restrains the rack tipping motion during seismic excitation. The bolts are positioned to aliow the rack to slide horizontally and to be engaged only if a tipping motion occurs. During lateral movement, the rack legs will make contacu with the aforementioned angle plates rather than engaging the bolts. All of the other racks are larger (6 by 6 or 6 by 7 rows) and do not require floor bolts. The

racks, including the base structure, the angle plates, embedment plates, and spent fuel pool liner are all stainless steel.

Thus, galvanic corrosion is not a problem.

Stainless steel has also been shown to be compatible with spent fuel pool water and the stored assemblies. 6.3.2 Modified Fuel Rack Codes and Standards The design and fabrication of the high density spent fuel storage racks wi be performed in accordance with applicable portions of the following codes and standards. 6.3.2.1 Design Codes a. AISC Manual of Steel Construction, 7th Edition, 19/0 b. AS E Bo'iler and Pressure Vessel

Code, S $ction III, Nuclear Power Plants Components, 1974, includiuq Addanda through Winter, 1974 (Used for allowable stress values, coefficients of thermal expansion, and moduli of elasticity) c.

AISC " Specification for the Design, Fabrication, and Erection of Structural % eel for Buildings," February 10, 1969, and Supplements 1 through 3 (Supplement 3 was effective 6/12/74.) 6.3.2.2 Material Snecifications a. ASMS Specification SA-240, Specification for Stainless and Heat-Resisting Chromium and Chromium-Nickel Steel Plate Sheet and Strip for Fusion Welded Unfired Prc sure Vessels b. ASME Specification SA-320, Specification for Alloy Steel Bolting Materials for Low Temperature Service 352 3i8 29

ASME Specification SFA-5.9, Corrosion-Resisting Chromium c. and Chromium-Nickel Steel Welding Rods and Base Electrodes 6.3.2.3 .Weldina Codes a. ASME Boiler and Pressure Vessel Code, Section IX-1974, Welding and Brazing Qualifications b. Regulatory Guide 1.31, Rev. 1, " Control of 5taialess Steel Welding," June 1973 (as moiified by Branch Technical Position MTEB 5-1, November %, 1975) 6.3.2.4 Quality Assurance, Cleanliness and Packa2in' Requirements as Acolicable to the Sceat Fuel Racks a. SSW and VEPCO requirements b. Regulatory Guide 1.38, " Quality Assurance Requirements for Packaging,

Shipping, Receiving,

Storage, and Handling of Items for Water-Coole(.

Nuclear Power Plants," March 1973 c. ANSI N45.2.1-1973, " Cleaning of Fluid Systems and Associated Components During Construction" d. ANSI N45.2.2-1972, " Packaging, -Shipping, Receiving, Storage and Handling of Items for Nuclear Power Plani.s In

addition, the racks are designed in accordance with the aiglicable portions of the following:

ANSI N210-1976 " Design Objectives f or Light Water Reactor S,oent Fuel St orage Facilities at Nuclear Power Stations" 6.4 Criticality Analyses In order to verify that the design basis presented in Section 6.2 pertaining tc criticality is met, analyses were performed for the normal, wet storage of spent fuel and for accidents pertaining to spent fuel storage. The accident ana]yses are presented in Se ction 6.7. The analysis of normal, wet storage of fuel assumes nominal conditions plus geometric material and calculational uncertainties. A parametric study is also presented to show the ~ sensitivity of the results to - varia tions in the assumed parcmeters. The analyses are based on storage of Westinghouse 17 by 17 fuel. In addition, an adjustment has been made to account for a higher reactivity due to the presence of Westinghouse 15 by 15 fuel in the e7ent such fuel should be transferred to the North Anna Power Station for interim storage. 6.4.1 Assu$ notions The calculations were based on the following conservative assumptions: 7L~ 0 Jab Ii 30

a. Fresh fuel of an equivalent enrichment of 3.50 w/o U-235 (44.2 grams of U-235 per axial em) b. Water temperature of 68 F c. Fucl racks are infinite in three dimensions. d. Fiyad neutron poisons in the fuel assembly are neglec-ted. e. No credit is taken for neutron absorption by structural materials in the rack system, except the stainless steel tubes. f. "No soluble neutron poison in the pool water A uniform array of 3.50 w/o enriched fu21 is also assumed, although calculations have shown that this assumption is conservative compared to the more realistic one of distributed enrichments within the array. 6.4.2 Methods of Analvsis The majority of the calculations were performed with methods commonly used in light water reactor design; i.e., 4-group diffusion theory cell calculations using PDg-07. Cross sections for these calculations are generated with NUMICE, the NUS version of 160PAPD. This code uses the same cross-section library tapa and calculational techniques as LEOPARD. Selected cases were checked and the final design multiplication factors were verified with Monte Carlo calculations using KENO with 123-group-cross-sections. The 123-arouo-cross-section library is generated f rom a basic GA'G-THEP310S library using two subroutines, NITA*C and XSDENFM in the AMIT code package. Both the PDQ code with cross-sections based on tha LEOPARD

library, and the KENO code using the 123-group-cross-section librarv have been benchmarked.

The geometry of the fuel rack used in the calculations is shown in Figures 6.4-1 and 6.4-2. 6.4.3 Results of the Analyses Under nom!.nal conditions, a rack assembly as shown in Figure 6.4-1 with a center-to-center spacing of 14 in. between two adjacent storage locations and 1/8 in. Type 304 stainless steel storage tuber, results in a keff of 0.889 based on KENO an alyse:.. Considering maximum variation in the position of fuel assembl..es within the starage

rack, engineering tolerances, seismic induced deflections, and

-calculational uncertainty resulteil in a keff of 0.923 with a confidence level of 95 perce nt. This ' ef f still meets the design basis to maintain r keff equal to or less than 0.95 for the normal wet conditions. The alx. ve uncertainties are discussed further below and summarized in Table 6.4-1. 352 320 31

r 9" SQ. 8.432" SQ. J=0 a i V!AT E R r 7 T T i~ i T- -] ~1 T-1 r ~i- -] T T 1-1 l-h -b 4 +- -f - t- -l --i - I-4 4-d I ' k - -!- - l-h - --I I-- l- -{ --!- -l - l- -{- 1 - L -l t-- O L- -t 4 - + -l h l-l 04 -I- + OM -l--4 -l-4 y h -t - ?- 4 O }_ _4 _ 4_.,__g _ y.4 _l_ }_ L g _; _ p _j O l-P -h -4 1-d - F d -! F-i- -l-i-- -[ -[- d h + -l O-i -l-4 04 -l-l Ot--4-+O-y - + -l 9 - 1 F L -l-4 -l- -{ - l-- 4 - t-l h y - p l -. l-J e a i 1 2 h -!- :- -l- -t - i-- i- --4 ' -l 4 -l- {-- 1. y e a E l-- . Y - I-S l O ;. _ g _ ;_ O_ _ _ ;_ 04 __q _ i_O.4. _ p y I i i - l-- 4 J- - F l- [- + F.-j- -l-- ! - L l - -4 2 I J- --l- + - !-- -t -l-1-- -t- + -l - 4. - h 4 -l-y i I h.b-IO_;_;__4 O ;_4 _ p_O _ 4 _;_O _ __ _q p_ 4 4 _) l-t- -i 4 - h- -h H 4- -i - h -f - H -? -l-- -{ r l-- f-.4_ O._, _ p _; _;_ p_ _;_ _j _ _ _n _; p;__ _t __j l 0 _ l _ i_0 _ l _ _ O _ i _ ;_ _; ___; _ q i-l- ! -l - 4 4 4 F 4-L. J L. 4 --l - F --l-- 4. -1. L-t _J L _1 -_1 -.; _ i_- L I __ l.. L.'_ }_L -.]-;__.L __l _q _1_ __! I '\\ 1/S" SS C NJ %TER g J=0 14" S PA C I." ~ I \\\\ AN. w., A F 1.ra u o 6. /.- 1 Storage lattice Cell 352 32! 32

m INSIDE SQUARE. AT ALL CROSS SECTIONS i S.O O >d5c .12 5. .h T i i,, a '} '" SQUARE if 8.750 f >ff M!N. FREE PATH l j. / ) J I -l f t.. ,J, Ficu m 6.L-2 7 "j 'L -) g-bJl Stainless Steel Tube 33

6.4.3.1 Calculational Une rtaintien In

general, the 4-group PDO diffusion calculatio. > produce kef f values comparable to the Monte Carlo calculations.

Calculational uncertainties in the use of PDQ with cross section: based on the LEOPARD library have been obtained by comparing the results of a series of benchmark calculations with critical experiments. These comparisons (Reference 1) have shown tha t the average difference bctween the calculations and experimental results was 0.009 delta-k. Thus, the LEOPARD-PDQ calculation can be used with confidence in predicting reactivity ef fects in this type of lattice.

However, to establish the absolute value of k, the KENO code was used to establish the uncertainty appropriate for this method.

The EENO code using the 123-group GAM-TIIERMOS cross--section library has been benchmarked. For a series of ten experiments reported in Reference 2, the average keff as calculated using KEMO and 1,3-group-cross-sections was 0.9914 1 0020. Using the same

method, NUS has performed another benchmark on one of the Yankee critical experiments (Ref erence 3, page 82) with Ag-Cd cruciform control.ods banked at 26.37 cm from the bottom of the fuel.

The calculated keff was 1.0077 1 0.0034. On the basis of the above comparisons with criticals a calculational uncer tainty of +.0086 delta-k was assigned to the KENO calculations. n statistical analysis of the Monte Carlo runs results in a standara deviation of 10.0055 giving an uncertainty in k-infinity of 0.0110 delta-k at a 95 percent confidence level. Adding this to the previously identified uncertainty of 0.008 5 in the KENO value of k results in 0.0196 delta-k a.s a conservative uncertainty to be applied to the KENO results. References 1. WCAP-3269-25, " Calculation of Lattice Parameters and Criticality for Uranium Water Moderated Lattices," by L.E. Strawbridge, Westinghouse Electric Corporation, September 1963. 2. " Validation of Monte Carlo Calculations of Shipping Cask Systems," by L.M. Petric and P.G.

McCarty, OPSL,

CONF 731101-14, 1973. 3. YAEC-94 " Yankee Critical Experiments" by P.W. Davison et al., Westinghouse Electric Corporation, April 1959. (. 4.3.2 Geometric and Material Variations The most adverse geo.netric and material ;riticality condition was obt. Tined by uaing the maximum tolerances for the positioning of the fuel assemblies within the storage can as well as the relative can-to-can positioning. The racks are'toleranced on an overall rack width basis, such that cumulative tolerances between can-to-can positioning are not possible.

Thus, simultaneous 352

??< -cJ 34

inclusion of all these tolerances provides added conservatism to the analyses. The tolerances considered in the analysis are for center-to-center spacing between storage locations, storage can inside dimension and wall thickness, and fuel assemoly location in the storage can. These tolerances are: a reduction in ce nte r--to-center spacing from 14.00 to 13.94 in., a maximum positive tolerance on the inside dimension of the stainless steel can of 0.06 in., and a minimum wall thickness of 0.120 in. vs. the nominal 0.125 in. With respect to assembly

location, every four. assemblies were assumed to shift such that they are clustered in a

common

corner, giving the highest potential reactivity increase from fuel assembly location in the rack.

Adding the above reactivity effects gives a reactivity uncertainty due to mechanical spaci ng and tolerances of 0.0083 delta-k. Puel enrichment uncertainty has a direct effect on reactivity. Assuming a very conservative uncertaAnty in enrichmen t of +2 percent, i.e. 3.57 w/o rather than 3.50 w/o, results in a reactivity increase of +0.0029 delta-k. Variation in stainless steel composition also affects reactivity. An arbitrary reduction in the Fe, Cr and Ni content of the 304 stainless steel in the storage tubes to the lowest limit allowed by ASME Specification SA-240 would cause an increase in reactivity ct +0.0031. Spent f uel storage pool temperature variations are expected to be above 68 F. However, for conservatism, the reactivity increase due to decrease in pool temperature to 39 F (4 C), the temperature of maximum water

density, is included in the uncertainues.

This reactivity is +0.00 04 delta-k. 6.4.3.3 Sensitivity Studv The preceding sections discussed the reactivity additions due to worst-case tolerancer and calculational uncertaintj es which are included in the criticality analysis of the racks. These analyse s indicate to some extent the sensitivity of the multiplication factor to these parameters. This section further explores the sensitivity of keff to variations in parameters, in pa rticular, spent fuel pool water temperature, center-to-center spacing, and storage can wall thickness. a. Variation in Pool Water Temnerature The water temperature or 68 F used in the criticality analyses is lower than expected in the storage pool and is ennsidered conservative. Using PDQ, analysis of the latuice of storage cella results in th e tollowP.g temperature reactivity effects: 352 324 35

Water Temoerature, F delta-k 39 +0.0004 68 (Base case) 0 100 -0.0012 212 (no void) -0.0090 b. Variation in Center-to-center Soacina In the discussion of uncertai:-ties, a possible 0.06 in. reduction in center-to-center spacing was included. To further investigate the effect of varying center-to-center spacing, a 10.125 in. variation was considered. Using PDQ, the results are as follows: Center-to-Center Soacina, In. delta-k 13.875 +0.0034 14.0 (base case) 0 14.125 -0.0032 c. Variation in Wall Thickness of Storace Can The minimum thickness of 0.120 in. was considered in the uncertainty discussions. As seen below, the. effect of variation in thickness as calculated using PLQ is the same magnitude for 10.005 in. variations. Wa?1 Thickness, In. delra-k 0.120 +0.0014 0.125 (base case) ,0 0.130 -0.0014 6.4.3.4 Summary of Results The results are su:cnarized in Table 6.4-1. The most adverse reactivity increase due to geometric and material variations is c.0147 delta-k, which, added to the calculational uncertainty of 0.0196 and the no:dnal keff of 0.8893, results in a maximum keff of 0.9236 for the normal storage condition. It should be_noted that an algebraic sum overestimates the effcct of combining the reactirity effects of geometric and material variation. The root mean square of all effects excluding spent fuel pool water temperature, plus the pool water temoerature effect, is more appropriate. This yields a combined reactivity for these effects of 0.0044 rather than 0.0147, and a maximum kef f of 0.9133 rather than 0.923b. 65 Structural Analvsis 6.5.1 Loadr and Loading Criteria 'Ihe spent fuel storage racks are classified Seismic Catecory I. Structural integrity of the fuel racks when subjected to nomal, abnocal, 352 y 36

TABLE 6.4-1 SU:O*u'J1Y OF CRITICALITY *-NAIYSES Nominal Conditions Enrichment: 3.50 w/o Mechanical Spacing: 14 in. Pool Temperature: 68 F keff 0.8865 15 by 15 in Place of 17 by 17 Fuel (del ta-k) 0.0028 Calculational Uncertainties KENO Benchmarks (delta-k) 0.0086 Statis tics 0.0110 Total, Calculational Uncertainties (delta-k) 0.0196 Geometric and Material Variations Variation of Enrichment from 3.5 to 3.57 w/o (delta-k) 0.0029 Mechanical Spacing and Tolerance (delta-k) 0.0083 Possible Variation in S. Composition (delta-k) 0.0031 Pool Temperature (delta-T=29 F) (delta-k) 0.0004 Total, Geometric and Material Uncertainties 0.0147 Maximen keff Nominal keff 0.0893 Calculational Uncertaintice 0.0196 Geometric and Material Variations 0.0147 Maximum keff 0.9236 352 326 37

and seismic loads is demonstrated with, respect to the MRC Standard Review Plan Section 3.8.4. In accordance with the Re. view

Plan, the following

loads, load comb! "lat. ions,

lnd structural acceptance criteria are considered. 6.5.1.1 Loads a. Normal Loads i. Dead Loads - deadweight of rack and fuel assem-blies and buoyancy ii. Live Loads - eff ect of lif ting empty rack during installation iii. Thermal Loads - thermal gradient between adjacent storage locations of 39 F b. Severe Environmental Load - Operating Basis Earthquake (OBE) c. Extreme Environmental Load - Design Basis Earthquake (DBE) d. Accidental drop in water O m speuu fuel assembly from height consistent with fuel handling operations which is .3 f t-1 in. above the tcp of the vant fuel racks. Postulated stuck fuel assembly which causes an upward e. force on the rack of 4,000 lb, which is the refueling bridge crane limit. 6.5.1.2 Load Combinations The fuel racks are analyzed using elastic working stress design rathods for the following applied loads: a. Dead Loads Plus Live Loads b.. Dead Loads Plus OBE c. Dead Loads Plus Thermal Loa ds Plus OBE d. Dead Loads Plus Thermal Loais Plus DBE e. Dead Loads Plus Thermal Load.c Plus Fuel Assembly Drop f. Dead Loads Plus TlIrmal Loads Plus Stuck Fuel Assembly Live loads are not included in load combination b. through f., since the only live load on the rack is that due to lif ting, and lif ting of the racks is performed with the racks empty. 7;9 797 JJL JL. c

6.5.*.3 Structural Accentance Criteria The following are the strength limits for each of the above load combinations: Load Combination Strength Limit a. 1.0 S b. 1.0 S c. 1.5 S d. 1.6 S e. 1.6 S (except as noted below) f. 1.6 S where S is the required section strength based on the elastic design methods and the allowable stresses defined in Part 1 of the AISC "Seecification for the Design, Fabrication and Erection of Structural Steel for Buildings," February 12, 1969, including Supplement Numbers 1, 2 and 3. (Supplement 3 was effective June 12, 1974). For load combination e.., local stresses might exceed the

limits, provided there is no loss of function of the fuel rack..

6. 5. ~2 Seismic Analvnis The seismic loading of a fuel rack module is determined from a res;cnce spectrum =cdai dynamic analysis in which lim stiffness of the fuel assembly is neglected. However, the mass of the fuel assemblies and an effective mass of water are considered to be uniformly distributed along the storage tubes. The appropriate response spectra for the OBE and the DBE as taken from structural analysis presented in the FSAR, Section 3.7.2 are employed. The STARDYNE computer program is used to perform the structural analysis of the racks. Racks are modeled in detail using beam and plate finite elements. The three-dimensional finite element model for a spent fuel rack is shown in Figure 6.5-1. To determine the earthquake response, STARDYME is first run to determine the natural frequencies and participation factors. For frequencies with significant modal participation, mode shapes and modsl loads are calculated. Closely spaced modes are combined directly and then combined in an SRSS manner as defined in Regulatory Guide 1.92 with other significant modes. The results of horisontal and vertical earthquakes are combined in an SRSS fashion. In the general seismic / structural analysis of the fuel racks, the mass of a fuel assembly is assumed to be uniformly distributed along the length of each of the fuel storage cans. This assumption is conservative in th& c lower rack fundamental frequencies are calculated

which, due tc the relatively <itif f rack desi a, result in higher seismic amplified acceleration J

loading on the rack. To account for the submergence of the racks in the analysis, the mass of a volume of liquid was added to the 352 328 39

mass of a structure giving a total " virtual" mass. The structure can then be analyzed as though it stood in air. This approach is discussed in Section 6.4 of Fundamentals of Earthquake Enaineerinc by Newmark and F.osenbleuth. The added mass or water outside of the can was calculated based on a cylinder with a diameter of the width of the can (9.25 in.) and length, that of the can (170.25 in.). The added mass of water inside the can was based on the inside volume of the can minus the water displaced by the fuel. The fundamental frequencies of a typical (6 by 6) rack in the two lateral directions, north-south and east-west, and in the vertical direction are calculated to be 15.2 Hz, 14.1 Hz, and 70.7 Hz, respectively. The high frequency (>40 Hz) in the vertical direction allowed for a static analysis for this direction. Since a gap on the order of 1/4 in. e:<ists between the sides of a fuel assembly and the can, the fuel will actually move within the can during a seismic event and cause impact loads to be transmitted to the fuel rack. The effects of this fuel-can interaction are analyzed by utilizing the ANSYS computer program. A nonlinear dynamic analysts of a single can and fuel assembly is performed to determine the shear force and bending moment which may occur at critical sections of the can as a result of the fuel assembly impacting the can at the maximum velocity. The can and fuel ass e.mily are nodeled by finite elements separated by nonlinear gap elements as shown in Figure 6.5-2. The can has stiffness characteristics representative of a can within a rach.. The fuel, which is assumed to be pinned at its base (by friction), is given an initial velocity relative to the can. This initial velocity is equal to the SRSS of the floor velocity and the velocity of the rack with respect to the floor. 6.5.3 Structural Adeauacy Using the previously listed loads and load combinati' s, stresses are calculated at critical sections of the racks. Tae results of the structural and seismic analyses demonstrate that the fuel racks are structurally adequate and meet the design criteria. Critical stresses together with locations and margins to allowable are given in Table 6.5-1. 6.5.4 Pool Floor Loadinc Tne structural adequacy of the pool for floor loading is discussed in Section 7.1. 6.6 Stored Fuel Assembly Thermal - Hydraulic Analysis The' fuel rack base is elevated above the floor to assure adequate flow under the rack to each fuel assembly. The .m acing of the fuel assemblies also pernts adequate downflow within the rack to each storage location. Analyses have been performed and show 7 c ') 7 ') Q 40 J J L-

TABLE 6.5-1 SU?@!ARY OF CPITICAL STRESS RESULTS Limiting Calculated Allowable Load Stress Stress Location CombinationC1) (Psi) (Psi) Marcin Support Foot Welds d 22,660 27,600 1.22 Puel Can to Base Uelds d 9,7CJ 18,400 1.90 Fuel Cans d 12,740 27,600 2.17 Shear Plate to Fuel Can Welds d 16,990 18,400 1.08 Base Ecam Welds b 7,580 11,500 1.52 (1) See Section 6.5.1.2 for definitions. I JJL s-41

e a e e 'N l / h 'V l 'N ' Y }' / 1 'I, ? \\,l // /li lNirW,{I i i. r :, fx Wl. ~1Ix) i d. l1,y d ' U !/ l>L'! N di l N i . t, o ',% / ~ i i x 1. w

f ll f N If f,%

/\\ 1 x V{'\\! !! ;~f!.!'! ll:$ Illh:',ysuYtN[ ,/ o r. s p l NI nl.ai,.! w,/ m,,: m m. f hN ,) - l l ll. - 'N 'y sl .l' l 'l' o 5 ~, ..l' kj ..m

a i x!,l/'

~ l (,1 r ili ',4 !.i w:l,u.,,;. 'j*y!lI,,.R. i t . m;: i,.x.,%..s.cl t / s L l ilm !~,. ! ./ i,

; '.: ~

T /iN ,I 't .s ',t:+o:',/(ll /fl e: ruv i ; :~.:f.p/ j ' '.-lr l,l. ' : N. '?,sll. ll!.'j N,.'1.1 . j',.i ' lN > 's -

.s, y N

.,, ~. t A*4e (# w r, g-r I Yl!P, ljl,.~ 'y.,;!; !:i ',. ' ',. !x 't~ I 1 . x J ! j:...c ?sl/l 1 .;; p ). -% l l* W lf '1. f. f ]'N'I ~!~. N z Nh.'}I'l t g! N'O. \\s ,l. </- !l l ?h.[1.;:'hk' / ~f,'b'Q-lb'.',;,$!! .(. -4 t d \\ s , y ll ,ll,[:,. l Y ll . ~.3,..' ** Nl# t c i n%:%. ,. :n.w' n: A. * 'M :s, cllf'~/r3lt' . ~.:. u I ' ] ' - l I /. f I*' l. y% Mz. :x' N.;,4 ;iy j';Q r els r, -- m l~l ,,. j :ll;s,',s':: _ r - 'N.,. :,' :- x: <; . z., ' :<s,...,..,, u, :.q

g..
s 1,

l.; .i ;ls/y. ,~y / r:, /N/ f 4 .. s,N}, 'i f ,}'O (lQl,'.sIi)oW!Il:l'f# Q. :. i ;*'l Ii.)fl A ';'S':N' u /N'* J : N'k. , , l 1. < < '. ul->.. di .i W {l,q,I I f/lQ.h . s% '. t.N*: / ^ ~ ; ;f,.., ,1 I.s. - i' n'A .1 ?. * .', N i ? .h, \\ l4 r. %. e

')

N, ? ll;,<f, s'f..':s i s;% ,,s. ', % ~; iN..s &,r o{m':q.,, N t k {,':j ', 0 'e \\' N; rs.',f' .s,N., ) I ?); .s w$ g ,,x N-f,/r' 's s., f ., / * - 6 = w .s.N '.^ s,,/ / s //I j f '. %, ..l ! s.f ll, l l' ., l/ * .l0 s, s. ..t s / ,a r; 's ,fo, 'r_ l %* $ '*. ; '.e, . %.l ?p', f, 4, ,f . s ".u. ~..,.h

E

!, f.{ s,.,. *%lh . {, - r '-. x ; ? j .,.,s ,s%^<. ,%'i l 'e%l,., y sl - Ill / f,)l,' [l f l,l lJc I V,< ?e .9 y e r l l fI,f j' l ' N'h'f. ll{[~.'),~ N'$! (,{li,O $'&l'20['i'll l k kNif; l.?. '6 ;;:l lll{ l ll % i;:';': l: - x l Y

l

,,Y QA f. l {' < ' ~ ? ! l:. M l M ~ g; %.i.

.l.l,*.ll/Jh: '.li;%Ili !

~ l l 1/ t ' ' Q,? '\\,: ~'/ ' s: t'y : ^ '.f I / ! i 6l/fNlu,iy .UW's;y:'n/ 'e N !j / N \\- /,}j ' /4 ', s/ ',dl % N :/. / A-f f A' / \\' . /. d l ?k l' ' l 11 ; ~sy\\s;l, '. r' l r\\ l jI,.S.Gire+,:},z,:ia: t s U'. < t ,n s ,f,

: ~:

\\ Y' q,f;lc.ls's.>,;';.fl.l'll2:fj'.!; y:'!f, 'fll// j/] I W i ,s / w%

:. + ': w.

~: m%,l y y i j '%, l f. l ls / :,,! ,'.,..x, */l \\ ', ' l, 'r..,/jsQ<' ,7;,, ', f l3'* f. ),4 ,,'w_, .L / 1 ? = % ;..z. io w, r </ '\\ r f f I' e ?g e jKl I,l t ) },'\\ f '['Fl/7'l,<s!\\(.l I/ ' l, ' N tQ:; jsNyfN.,'n,,>'N;,'N,.[X / .e.. 'S - s s

s

/x. ( q,',(* 'N' *b' s. / I /% se s TN ,'/ N. /s, ,s i s* / (

,'N.\\+

.I e J e y, N.' / fg % i N,il lR'# .\\s ,j/ l8 ~~ ; f'N'f'l, h'l%l l fl) ~':j 4, Nl,l N,/l<'y, / l ,A / ,..t N N !. N v. y*n,(! .v, / //, / 'N,, / N,. N/ ixv s. nf j s 'N,/ s ,:r N,, <i ,N~i/,f.Nx !!j. t., s s g \\// , /, e' [ ,e j / 'N ' N, s <l l -/ ,\\ l - N/ \\ l/ l' / j / 'N:.'v[N[l[' N / N/ \\' ,/ ~, 'N'. c../ I Ficun o 6.5-1 FINITE ELEMP:T MODEL 7 c -) .,, f JJL jj 42

aid 9/5 'M O D E L_ A / I (e -o) 20 '.- 4 o j m t*. lH T: IB lh 12 gi-y p 3, s, -r: t- 'n ' Vd.' CC t~r/ 6" MOD AL go qp 1; 5 pac.tM C,

E

, H 3 s' W--I F M 13 lh,33 Fu s.L (e,c )-47. IHS. } o g l \\ t <--i F 31 Kwe cAq m 1 t_, w Y \\ \\ s 9 H 'q 19 S -WM H 7.8 / O (Go=o') 8 l h} zs / 4 1 G;& / l l \\ "l F ;7 7 lO" NobAL SPAQhJG 5 ,b 2S t (%= 0) a

'r 2.4 3

',F D cu e/sv.mt>Ms.c Rpc K. SAsq \\ T FFNiSS g ip a N / / L_. ' uk Y $3 q' \\ / \\ P1 /\\ f\\ 43 4+ 45-9 ro si s. x4s 3 I//// '//// a l. 5- " Fi re 6.5-2 7 I -) J 2 759 .sw 43

that sufficient flow is induced by natural convection to preclude local boiling in the hottest storage location. This location is es tablished by a 15 by 15 Westinghouse fr. element which was determined to be worse than for a 17 by 17 fuut element. The analyses were based on the following assumptions: a. The element inlet temperature is the mixed hot temperature of the pool. b. A hot assembly peaking f actor of 1.55 is applied to the core average assembly energy release rate of 2.00 x los Btu /hr, (corresponds to 150 hr af ter shutdown). The maximum local peaking factor is 2.22, giving c. a maximum local heat flux of 1,642 Btu /hr - ft2 d. A film coefficient of 35 Btu /hr - ft2 - F is based on pure conduction through a stagnant boundary layer at the fuel rod surface. The downcomer region is established by 40 percent of e. the flow area between the corner assemblies in the rack since these have the greatest hydraulic resistance. f-One Timensional fluid flow analysis used. g. The Rohsenow pool boiling correlation is used to indicate no local boiling. h. The maximum local heat flux is conservatively applied at the exit of the hot channel. During full-core offload with the bulk pool temperature at 170 F, the maximum temperature of the water exiting from th e hottest storage location, i.e., with a 15 by 15 Westinghouse assembly, is less than 197 F. This is 44 F below the local saturation temperature of 241 F. With a less restrictive 17 'y 17 o Westinghouse assembly in which the hot spot temperature is ~1ower because of' more heat transfer

area, the maximum bulk water temperature exiting from the hottest assembly is no greater than 198 F.

Under design operating conditions, with the bulk pool tempe ature at 140 F, the fuel rod surface temperature calculated on the basis of the heat flux and film coefficient defined above 10 below the local caturation temperature and thus precludes local boiling. Asc,uming a maxi:mim bulk pool temperature of 170 F, the fuel rod surface temperature is at least 4 F below the nucleate boiling temperature evaluated usi:.a the Rohsenow pool boiling correla tion and therefore no local boiling is predicted. 3 5..2 33 44

6.7 Safety Evaluation 6.7.1 Potential Criticality Accidents 'Ib determine the potential for criticality during an accident, two possible but highly unlikely restulated events were analyzed using the analytical methods descr11ed in Section 6.4.2. The first case is a fuel asserbly, in transport in a vertical pssition, which accidentally drops into the water channel between a rack and the pcol wall. The second case is a fuel assmbly which drcps and falls to a horizontal position on tcp of a loaded storage rack. The analyses include the assumations of all the calculational un-certainties and cecr.etrical and raterial variations discussed in Section 6.4.3. The design enric5 ment of 3.50 w/o is assumed for both the dropped and stored fuel assemblies and all fuel asse-blies are ass m ed to be fresh asse-blies. In the first case, the concern is assumed to be une dropping or accidental lowering of a fuel assembly so that it is parallel to and at the same level as the stored fuel in rack assemblies (see Figure 6.7-1). The resulting keff includina uncertainties is 0.9239 which is only slightly higher than the maximum keff for the nom.al storage situation (Section 6.4.3) and is less than the design basis value of 0.95. Mechanical restriction will be provided to prevent an unprotected fuel assembly from being brought closer than 5 in. to the side of any rack assembly in side water channcis. For a dropped assembly assumed to lay across the top of a fuel rack, the fuel assembly end fittings on the top of each fuel assembly provide a spacing between the dropoed assembly and the active fuel in the storage racks of approxima, sly 10 in. This is significantly less reactive condition than t.aat analyzed above a in which the fuel assembly is much closer to the fuel in the rack. Thus, this accidenu has not been analyzed further. ..t 45

J=0 l I I 3 .I -o> l !=.o o a Oc l i~7 3 C I IX i a - J.: x u =. i I F N : t !"' t t. : "t.. 'm Y r>- A w. w m p "u g i ". O U T.C.e r,,n*,'l l t I T

26. 25 " water channe!

lC Il i l e .i l w J=0 Figre 6.7-1 GEC.'.!ETRY CF THE WORST ACCIDE::T CASE M M

I4 J

s3J a 46

7.0 AMt LYSIS OF EXISTING FACILITZES AND SYSTD S AFFECTED LY THE PROPOSED IM;I FICATION The proposed modification does not change the physical configuration vt the spent fuel pool and only requires the addition of two iloor pada. The existing spent fuel pool cooling and purification system and fuel building ventilation system will not require any modificatlons. The primary c.ffect of installing new spent fuel storage racks uill be to increase the amount of spent fuel which may be stored in the pool; thereby, increasing the weight to be supported by the pool floor. The additional spent fuel stored in the pool will riightly increase the amount of decay heat which must be removed by the spent fuel cooling nystem. The effect of the proposed modification on the purification system will be minor. These and other effects are discussed below. 7.1 Structural Considerations The speit fnel pool structure has been Enalyzed to determine the effect of the additional weight of the new racks and the stored fuel will have on the structure under static and dynamic conditions. The analysis also included consideration o:- the thermal effects due to increasing the storage capacity. The existing structure has sufficient design margin which permits the installation of new storage racks without any additional structural modifications to the pool concrete or st i rstructure except for the addition of two new floor pads required to accommodate a high-density rack in an area where presently no rack is located. In the original design of the spent fuel storage racks, the seis-mic response of the rack system was calculated taking into acccunt 2% da. ping due to rack sub.ergence in vater. 1.dditicral respcnse calculations were rado eli-inating the additioral darping and the resulting lcad on the floor e:ded.ent pads were well within the calculated allowable values. 7.2 Fuel Pool Coolim Svstem The installed spent fuel pool cooling system was analyzed in view of the expanded fuel storage capacity. Table 7-1 sum:r.arines the cooling system performance for both the normal and abnormal (full core discharge) conditiona. The design basis heat load was determined using the following assumptions: 1. The irradiation times used were 272, 544, and 816 EFPD which correspond to'a one,

two, and three year fuel

cycle, respectively, with a load factor of 85 percent and an annual fa5 day refueling outage.

2. Back to back refuelings 45 days apart. 3. Uranium decay heat from URC Branch Technical Position 9-2. 4. All fuel to be moved into the pool is done instantaneously " 50 hr af ter shutdmm except for the full cora discharge case when fuel is moved from the 47 352 336

5-11-79 reactor to the pool at a rate of.20 min per assembly starting 150 hr after shutdown. 5. Stretch rating of 2900 MW is used for full power. l 6. Maximum of 966 storage locations in pool. 7. Service water at its design maximum of 110 F yielding a com-ponent cooling water temperature of 113.2 F. The resulting fuel pool temperatures are found to be within the limits of 140 deg for the normal case and 170 deg for the abnormal case if one fuel pool cooling system pump and two coolers are utilized. These fuel pool temperatures are calculated based on very conservative and worst case assumptions and are valid for establishing a design basis. Actual operating temperature experienced by Surry Power Station, Unit Hos. 1 and 2, which is of similar design have been significantly lower than the calculated temperature, and in fact has been maintained at about 95 deg F during winter and summer conditions utilizing only one pump and one cooler. The additional heat load due to the increased storage capacity is compared in Table 7-2 to the design duty of the component cooling water system, the design duty of the service water system, and the total station thermal discharge to the environment. A failure analysis of the spent fuel pool cooling system is su=marized in Table 7-3. The analysis confirm that boiling is prevented even in the event of a postulated failure. The analyses presented in the FSAR related to spent fuel pool makeup capanility is not affected by the proposed modification and is not discussed herein. 7.3 Fuel Pool Purification System As previously mentioned in Section 5.5.2, based on the experience of Surry Power Station, no significant effect on the system is expected due to prolonged storage of spent fuel assemblies. The maximum load on the purification system occurs during refueling operations when fuel is being moved. Therefore, there will be no significant increase on the purification system load due to the modification since the number and frequency of refueling operations will not change. 7.4 Puel liuilding ventilation Svstem Since the added fuel storage represents longer teru storage of well-cooled fuel, the escape of gaseous or volatile ficcion products from even defective fuel is expected to be negligible. Much of the iodines and the xenon has decayed atter 100 days cooling true. Sinc-most of the tritium in the water is 352 2

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 particulare fission producte to become

airh, ne.

Because of its long half life, Kr-85 levels remain in older fuel. However, the thermal driving force required to cause its diffusion in defective fuel is not present. Samples fri m the ventilation filter area at Surry do not show Kr-85 at (.3tectable levels, and it is not expected to become significant as fuel storage increases. Therefore, increased tuel storage will have essentially no impact on concentrations of radioactivity in the air of the fuel building. Since the FSAR pool temperature limits of 140 F (normal case) and 170 F (abnormal case) will not change with the modifica tion, there will be no effect on the design evaporation rate of the pool. W

9 JJL

))'

5-11-79 TABLE 7-1 SPENT FUEL POOL COOLING SYSTEM HEAT LOAD AND OPERAT11:G 'T'EMPERATURE WITH 'fifB INCREASED STORAGE CAPACITY Fuel Pool Temperature, Deg F Decay Heat 1 Train 2 Train MBtu/hr 1P-1Clr IP-2Clr 2P-2Clr Normal 19.4 147.8 135.4 130.4 Abnormal 35.9 176.9 154.2 144.9 4 4 h

2-22-79 TABLE 7-2 E 'OAD COMPARISON Heat Load, Mbtu/Hr Present Svrtem With !A>dification (400 Assemblies) (966 Assemblies) Fuel Pool Cooling System Design Duty Normal Case 13.8 19.4 Abnormal Case 33.2 35.9 Component Cooling Water System Design Duty (1) Normal Operation (2P-2HXR) 103.1 108.7 Abnormal Operation g (4P-4 HXR) (2) 402.1 407.6 i Service Water System Design Duty (1) Normal Operation (2 Pumpe) 103.1 108.7 Abnormal Operation (4 Pumps) (2) 402.1 607.6 Total Heat Discharged to the Environment (1) 13,713 13,719 Normal Operation NOTES: 1. Total for both units 2. Maximum abnormal operating duty for the component cooling water system and the service vater system occurs dt_ ring simultaneous fast cooldown of both reactor units. This does not occur concurrent with the maximum abncrmal fuel pool heat load which is the off loading of one full core starting 150 hr after shutdo'n. 352 3c0 51

5-11-79 TABLE 7-3 SPENT FUEL POOL COOLING SYSTEM MALFUNCTION ANALYSES _ Component Malfunction Co:r:nents and Consequences Spent Fuel Pit Pump fails to start The standby pump will be Cooling Pings or fails during started manually. operation If the operating pimp should stop, over 1 hr exists to start the standby pump before the pit heats up 10 F at the maximum abnormal heat load. Pool temper Lure will not exceed 170 F with only one pump and two coolers in service for the maximint abnormal heat load. Fuel Pit Coolers Loss of Function The standby exchanger will be used. More than 1 hr exis'.3 to reallyli Lite piping system because of the slow heatup rate of the pool. The re-alignment is effected by operating manual valves. During the design ba.;is conditions with only one pump and one cooler in service the maximum calculated tecperature is 176.9 F which is above the administrative limit of 170 F but is still below the temperature at which the structural analysis was performed. [) Jc 'L 7 ['1 'l l J 52

8.0 INSTALLATIO_N _AND REMOVAL OF SP2MT FUEL STORAGE PTtCKS The modification will be limited to the confines of the fuel building and will not involve any. changes in the safety-related systems which are necessary for the station. The fuel cool cooling and purification equipment is located in the area of the fuel building which is physically isolated by concrete walls and ceilings and is not subject to damage while the old racks are removed and the new racks installed. Since the old spent fuel storage racks have not been exposed to radiation, and therefore are not contaminated, it is at this time planned to dispose of them as scrap. 352 3/12 53

9.0 MU1YSIS OF THE SAFETY IMPLICATIONS OF THE PROPOSED MODIFICATION The proposed modification will not change the safety analyses which have been perfornted and reported in the Final Satety galysis

Reoort, Section 15.

The proposed expansion of the spent ruei storage capacity could affect the otfsite radiological consequences of an incident because of the additional increment of long-lived radioactive fission products stored ia the cool. The ef fect 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 in Section 9.4. The following discussion sum:rarizes the potential effects w' ich the proposed modification nny have on the safety of a the station and the public. 9.1 Iass_of 3 cent Fuel Pool Coolina Caoability M discussed in Section 7.2, the proposed modific> tion uill 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 to remove the additional heat load when uranium fuel is stored in tne pool. As indicated by tne railure analysis presented in Section 7.2, cooling capacity could be restored quickly in the event of a component failure. In the unlikely event that the spent fuel pool cooling system was to become completely iroperable, installed station systems could provide sufficient makeup water to cool the fuel and to n.aintain sufficient water shielding over the pool. There are several sources.or makeup water readily available in the event it is required. These sources are: 1. prircarf grade water systen 2. fire protection system 3. toron recovery system 4. refueling water storage tank These sources could be utilized by either changing v/dve lineups or implementing certain temporary measures, such as the use 91. temporary punps or hoses. 352

/; 3 54

In su:.Lary, suf ficient heat' tcmoval capacity is installed to assure that the pool temperatura remains below the boiling point. As additional backup a number of instr 11ed station systems could provide makeup and cooling water if required. 9.2 Fuel Pool Leakace Control and Shieldina The proposed modification will not affect the leakage and shielding requirements contained in the FSAR. The lowest level of pipe penetration through the fuel pool structure is at El. 285 ft-9 in., which provides a minimum water level of over 21 f t of water above the stored fuel to provide shielding and cooling. The proposed modification will not require any additional piping penetrations; therefore, ere are no safety implications associated with spent fuel p 201 leakage control or shielding. 9.3 Earthcuake and Tornado Protection The proposed modification will not requira any structural changes; therefore, it will noc affect the ability >f the structure to withstand the effects or an earthquake or tornado as rtated in the FSAR. The new spent fuel storage racks and pool structure have been analyzed to ensure that the racks can be accommocated by tae structure during a seismic event. These analyses are discussed in detail in other sections of this report. In su. nary, the seismic and tornadic provisions stated in the FSAR are not changed as a result of the proposed modification; therefore, there are no safety implications associated therewith. 9.4 Puel Handlinc Accidents Section 15 of the FSAR describes the fuel handling accidents which have been analyzed, including the case where a fuel assemuly is dropped onto the floor of the spent fuel pool. The proposed modification will not affect the consequences of the accidents analyzed in the FSAR because the analysis assumes that only one fuel assezM y, the one being in stalled, is damaged.

Thus, the consequences of the accident are independent of the number of spent fuel elements stored in the pool.

The high density spent fuel racks have been reviewed in regard to: 1. dropping a fuel assembty on the racks 2. a fuel asseml>1y becomes stuck in tne spent fuel rack 7C7 7-1 aJL J00 3 55

3. dropping a fuel assembly next t.o the rack While minor damage may be incurred to the rack if an element is dropped on it, the stored fuel will not be affected and suberiticality will be maintained. The amount of torce applied to a stuck fuel assembly is limited by the capacity of the crane. Unile damage may be incurred by the stuck fuel assemoly, the weight of the fuel rack is sufficient to prevent any motion of the rack itself. The surrounding stored fuel assemblies will not be damaged and subcriticality will be maintained. Mechanical barriers are provided on the outside of the rack to prevent a dropped fuel assembly from being brought too closs to the rack in order to maintain suberiticality as discussed in Section 6. In

summary, the safety implications of the proposed modification as related to 'uel handling accidents remain the same as previously analyzed in the FSAR.

9.5 Personnel Radiation Excosure Storing additional spent fuel in the pool will increase the amount 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 raol. Dependent upon the disposition of the reprocessing capability, the fuel coulc ce storea in the pool far about 10 years. During the storage of spent fuel under water, both volatile and nonvolatile radioactive nuclides may be released to the water from the surface of the assemblies or from defects in the fuel cladding. Most of tha material released from the surface of the assemblies consists of activated corrosion p vducts such as Co-58, Co-6d, Fe-59 and Mn-54 which are not volat ile. The radionuclides released through defects in the

cladding, such as Cs-134, Cs-137, 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 fuel pool would be exposed.

As noted in Section 5, the four primary isotopes noted in the pool water at Surry have been Cs-134, Cs-137, Co--5 8 ar d Co-6 0. Based on measured data at Surry Power Station, an individual continuously working around the pool would receive absut 1.5 mR/hr, based on approximately 208 fuel assemblies stored in the pool. This exposure will probably slightly increase then additional fuel assemblies are stored; however, because of the current storage

pattern, the increase is not expacted to be significant.

Even if the exposure is doubled, about 3 mR/hr, tnis exposure is a relatively minor contribu*.. ion to the overall exposura at the station. 7cn dJL . /j j 56

The installed purification ' system described in Section 5.2 will be used to remove the nonvolatile corrosion and fission product nucliden. The removal of theJe nuclides will assure that the radiation exposure to percor.n el will be maintained at low levels. The volatile fission product nuclifes of most concern that might be released through defects :.n the fuel cladding are the noble gases (Xenon and Krypton), tritium and iodine isotopes. Since short-lived noble gates will decay to negligible

amounts, the only significant noble gas isotope which could remain in the spent fuel pool and attrinutable to storing 1dditional assemblies for a

longer period of time would be Krypton-85. It is not expected that increasing the spen ~ fuel storage capacity will increaue the Krypton-a5 release rate, since the fuel discharge wil] cantinue on a 1/3 core per year per unit rate and the release of Krypton-85 is most likely to occur during the initial year of storage. Iodine 131 releases will not be significantly increased by the expansion of the teel 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 water. The pool wat -r temperature will be maintained below the current design temperat tro; therefore, it is not expe cted that there will be any significant change in evaporation rates and the release or tritium. Operating experience at Surry to date has not Indic ated the presence of tritium in the fuel building. As discussed in Section 5.5, the purification filters are normally changed becanse of high differential presstre; therefore, it is not expected that the proposed modification will significantly increase personnel radiation exposure during filter changes. nased on experience at Surry Power S ta tion, the radiation exposure is relatively

low, approximately 140 mR/hr.

The demineralizer resins are currently changed about twice a year resulting in personnel exposure of about 110 mR/hr. The proposed modification is not expected to significantly increase this value. In

summary, the proposed modification will not significantly increase personnel radiation exposure during normal and refueling operations.

'7 4 57 'JL ') Q b

10.0 I NVIRONMENTAL IMPACT OF THE PRONSED MODIFICATION The proposed mc dification would inc^9ase the amount of decay heat produced in the spent fuel pool, in_ aas e the amount of radioactivity stored in the pool. and result in a small commitment of metal resources. The environmental impact of the proposed modification is insignificant. The environmental impact has been reviewed in light of the current Final Environmental Statement, the Final Safety Analysis

Recort, 10CFR50, 40CFR1500.6 and guidance contained in Federal Reaister Notice (40 F.R. 42801) dated September 16, 1975.

Based on this review it has been concluded that the proposed modification will not significantly affect the quality of the hu' nan environment. 10.1 Federal Recister Notice (4 0 F.R. 42801) The Federal Register (FR) Notice directs that in consideration of lic e.n sing actions to ameliorate a possible shortage of spent fuel storage capacity, five specific factors should be applied, balanced and weighed in the context of the required environmental appraisal. Each of the five specific factors are addressed below. 10.1.1 Indeoendence of the Action The FR Notice states: "Is it likely that the licensing action here proposed would have a utility that id independent of the utility of other licensing actions designed to ameliora~te a possible shortage of spent fuel capacity?" Eased on the Infor:ution contained in this report, it has been concluded that a need for additional spent fuel storage capacity exists at the North Anna Power Station, Unit Nos. 1 and 2, whien is independent or the utility of other licensing actions designeu to delay a possible shortage of spent fuel storage capacity. 10.1.2 Commit.ent of Resources The FR Notice st..tes: "Is it not likely that the taking of the action here proposed prior to the preparation of the generic statement would constitute a com.titment of resources that would tena to significantly foreclose the alternatives availcole with respect to any other licensing actions designed to delay a possible shortage of spent f uel storage capacity?" Tne p ; posed modification will require the utilizacion of 3.22 x los lo of stainless steel. The amount of stainless steel used annually in the United States is about 2.8.: 1011 lb. The amount of stainless steel required f or the racks is a small annualjy]pinthyfnated percenta g e of this resource consumed P

States and is insignificant. No other. significant material rescurcet will be required becau::e the design of the fuel pool will remain uncnanged. The land area now used for the fuel pool will ha used more erficiently by reducing the spacing among fuel asse:ablies. The existin% installed spent fuel racks will be disposed of as scrap since they have not been exposed to radiation. This should pose no problem. The proposed expansion of the storage capability of the existing fuel pool is only a measure to allow for continued operation and to provide operational flexibility at the station and will not affect similar licensing actions at other nuclear power stations. It is concluded that the increase of the spent fuel storage capacity at North Anna prior to the preparation of the generic statement does not constitute a ccanitment of einner material or nonmaterial resources that would tend to significantly foreclorc the alternatiws available with respect to any other individual licensing action Qnigned to delay a possible shortage of spent fuel storage capacity. 10.1.3 Cumulative Envimnmental Effects The FR. Notice states: "Can the environmental impacts associated with the licensing action here proposed be adequately addressed within the context of the present application without overloading any ethlativ e envuvumental ef fects?" The additional capacity of the spent fuel pool is proposed for North Anna Power Station, Unit Nos. 1 and 2, only; therefore, the environmental impacts can be assessed within the context of the application. Based on the information contained

herein, it has 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 impacts associated with the expansion of the spent fuel pool will not result in radioactive ef fluent releases nor occupational radiation e:gosure nor thermal ef fluent releaces that significantly affect the quality of the_ human environment during either normal operation of the expanded fuel pool or under postulated fuel handling accidents. 10.1.4 Technical Issues The FR Notice states: "Have all technical issues which have risen auring the review of this application Leen resolved within that context?" Le technical issues associated with the proposed mod.iric tion are addressed i r. th is r c po rt. There is reascnable assirance that the proposed modif :a tion can be carried out as described herein with nc adverse e"fects on the health and safety of the public. yc9 ,.N

0 -

JJL ) 59

10.1.5 Heed for the Action The FR Notice states: "Would a deferral or severe restriction on this licensing action result in substantial harm to the public interest?" As stated in Section 4.0, a number of alternatives have been considered. The modification described herein provides the most economically f easible solution to mneliorate tne potential shortage of spent fuel storage capacity. If the propcsed moaification is not implemented, the alternative of ceasing operation of the facility would be much more expensive than the proposed action because of the need to provido fossil fuel replacement power. Deferral or severe restriction of the proposed modification would result in substential harm to the public interest. 10.2 Final Environmental Statement The proposed modification will not significantly alter the evaluations contained in the Final Environmental Statements. The proposed modification will create a slight additional heat load on the station water system; however, because of the minor amount of additional heat to be added to the service water system, no discernable temperature difference in the thermal ef fluent from the station is expected. 10.1 Final Saretv Analysis Renort The descriptive information contained herein is intended to supplement the materl.1 contained in the FSAR. The design criteria specified in the FSAR have been used as the basis for the propased modification and have be en supplemented as appropriate for the new spent fuel storage racks. The modification does not substantially change the analyses and descriptions in the FSA2. 352 74o J / 60

11.0 CONCLUSIous ~ Based on the information co:w.ained herein, it is concluded t. hat: o 1. The proposed modification to increase th 2 storage capacity of the spent fuel pool is necessary to maintain .the capability of a full core discharge and to arsure adequate storage space for normal refuelings. 2. The use f fuel racks with reduced center-to-center spacing trovides the most economical and feasible alternative to delay the potential shortage of storage capacity. 3. The installed fuel pool cooling and purification system and the fuel building ventilation system ar0 adequate withcat any modifications. 4. The fuel building structure is adequate and does not require modification. 5. The availability of alterna te storage, reprocessing, or permanent disposal capabilities cannot be relied upon in the near future to preclude the necessity of the proposed moditication. 6. ' The proposed modification will not af f ect *ha kaalth and safety of the general public. 7. The proposed modification will not significantly affect the quality of the environment. 357- .xn0, 61

~

L- / NRC l'UBLE hE' /

5U$

a tf ,@3]: [j 0 L. Q d ;s F,'? f s ~

S 9

a";g q UNITED STATES OF AMERICA IN si NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSIt'G BOARD In the Matter of ) ) Doc. Nos. 50-338SP 50-339SP VIRGINIA ELECTRIC AND POWER COMPANY ) Procosed Atrendment to jtingLicense (North Anna Power Station, Units 1 and ) ) AFFIDAVIT OF ROBERT W. CALDER My name is Robert W. Calder. A true statement of my professior 1 qualifications is attached to this affidavit. I do uit expect that storing 966 instead of 400 spent fuel assemblics in the North Anna 1 and 2 spent fuel pool will materially increase the corrosion of the fuel cladding, the spent fuel storage racks, or the pool liner. The amount of additional radiation to which these caterials would be exposed in the pool is insignificant compared to the levels in the reactor core during powcr operation. The materials (stainless steel a- .ircaloy) were chosen because of their low susceptability to corrosive attack in a nu sar environ-ment, t'la t is, under exposure to high temperature, high pres-sure, water and radiation. Because the fuel pool water temper-ature of 140 F (normal condition) and 170 F (abnormal condi-tion) will still be maintained, I would not expect the cor-rosion or stress on the fuel, the racks, or the liner to materially increase due to heat. Mr A. B.

Johnson, Jr,,

Staff Scientist, Corrosion Re-352 3!!

.2_ search and Engineering, Battelle Pacific Northwest Labora-tories, has reported that " fuel handling experience in the U. S., going back to 1959, has not revealed any instance where Zircaloy-clad uranium oxide fuel has undergone observ-able corrosion or other chemical degradation in pool stor-age. Thic favorable experience is corroborated by experi-ence in other countries with the following maximum pool residet.cc time for Zircalov-clad fcc? as of late 1977: Canada, 14 years; United Kingdom, 11 years; Belgium (MOL), 10 years; Japan, 9 years; Norway, 9 years; Karlsruhe, Ger-r.any (WAK', 7 years, Sweden, 5 years." Finally, I would not expect the additional storage capacit" to make the eventua' remo.al from the pool of the spent fuel assemblies any more d3fficult due to corrosion o f the fuel cladding. bl;!&g Y& Robert W. Calder DATED: May 11, 1979 Signed and sworn to before me by Robert W Calder this i , day of May, 1979. / Not1ry Public My commission expires p .)/ 0 ") Jf a

M STATEMENT OF PROFESSIONAL QUALIFICATIONS OF ROBERT W. CALDER SUPERVISOR - ENGINEERING SERVICES Vepco Mr. Calder's technical experience includes 9 years in the field of materials engineering, of which 7 years have dealt with nuclear power plants. Mr, Calder joined Martin Marietta Research Institute for Advanced Studies in 1969 and conducted basic research in the areas of physical and mechanical metallurgy of high tempera-ture alloys. Mr. Calder joined Westinghouse Bettis Atomic Power Laboratory in 1973. As a metallurgical engineer he worked in the construction of Naval Nuclear Power Plants. His work included destructive and nondestructive evaluation of zircaloy-clad nuclear fuel elements. In 1975, he transferred to the Naval Reactors Facility where he qualified as a nuclear plant engineer and was responsible for plant testing and maintenance. Mr. Calder joined Virginia Electric and Power Company in 1977 as an engineer in the Nuclear Engineering Services Depart-ment. His present title is Supervisor - Engineering Services. EDUCATION B.S.c Metallurgical Engineering, Grove City College M.S., Material Science, University of Maryland M.B.A., Business Administration, University of Pittsburgh 352 353

N}, N " C PUBLIC DOCU:.'ENT ROW c) o U Q h ?5 h $s dV # p6 6/ gif UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of ) ) Doc. Nos. 50-338SP 50-339SP VIRGINIA ELECTRIC AND POWER COMPANY Proposed Amendment to (North. Anna Power Station, Units ) Operating License NPF-4 1 and 2) ) AFFIDAVIT OF H. STEPHEN McKAY My name is H. .ephen McKay. An accurate statement of my professional qualifications is attached to this Affidavit. I am the Project Engineer responsible for the design and in-stallation of the high-density spent fuel racks for North Anna 1 and 2. I am familiar with the design and technical analyses of those racks that have been done. Attached to this Affidavit is a copy of the document that Vepco has submitted to the NRC in support of its appli-cation to install and use the high-density racks. It is en-titled " Summary of Proposed Modifications to the Spent Fuel Storage Pool Associated with Increasing Storaga Capacity." Certain portions of this Summary have recently been amended; the amended portions are indicated by a vertical line in the right-hand margin. I am familiar with the contents of this Summary. It is true and correct to the best of my knowledge 352 354

and belief. I should like to supplement that Summary in the following respects. All references below are to the paragraph numbers of "Vepco's Statement of Material Facts as to which There is No Genuine Issue to be Heard." The fuel pool heat exchangers and circulating pumps are located in the fuel building (1 3). Each of the spent fuel pit pumps is connected to a separ-ate, independent Class I power supply (1 7) It will require a maximum of 12 gp:n cf evaporation to dissipate the additional heat discharged to the environment because of the proposed modification (1 15). To determine the design basis heat load, a stretch rat-ing of 2900 MWt for full power was assumed (1 18e). The failure analysis of the spent fuel cooling system confirms that boiling and any adverse effects are prevented even in the event of a postulated failure of a spent fuel cooling pump or a spent fuel pool heat exchanger (1 27). The component cooling water temperature could get as high as 113.2 F in the unlikely event of a LOCA in Unit 3 or 4, but the pool temperature would still be less t12n 177.5'F, the temperature that was used for the structural analysis of the spent fuel pool (1 33). Unit 1 control board instruments and alarms include spent fuel pit temperature alarms at greater than 140'F and grenter than 170 F (1 38b) The spent fuel pool heat exchangers transfer the heat from the spent fuel pool water to the component cooling water (or, in an emergency, to the service water), and the component cooling water transfers its heat to the service water (1 12) The service water, in turn, goes to the Service Water Reservoir, where the heat is transferred to the atmosphere (1 13) 352 355 Acceptable and appropriate engineering techniques wert used to calculate the fuel pool temperatures (Y 19) The second heat exchangc: would be required for only a period of 4-5 days, and only if a highly unlikely sequence of events were to occur-a. Unit 1 refueled 45 days before the event; b. Unit 2 just defueled; c. Unit 3 or 4 loss-of-coolant accident (LOCA); and d. Other unit cooldown (f 22). The spent fuel pit is a reinforced concrete, seismic Class I structure lined with stainless steel plate a minimum of 1/4 inch thick. If the integrity of the 1/4 inch thick stainless steel liner were violated, water could enter the channels behind the liner. These channels are connected to a common drain point, which is the fuel building sump. In the event of a leak into one of these channels, water would rise in the fuel building sump, the sump pump would go on, and an alarm would sound in the control room. If the punc-ture were at a point other than the channels and the fuel pit water were somehow to pass through the liner and rein-forced concrete, it would reach the foundation material be-Icv, which is virtually impenetrable (11 43-47). c.'alysis of the thermal-hydraulic characteristics of

he high-,3nsity racks have been performed using techniques that are generally accepted in the engineering community (1 50)

During the full core offload (abnormal case) with the bulk pool temperature at 170 F, the maximum temperature of the water exiting from a storage location is less than 197 F, which is 44 F below the local saturation temperature (boiling point) of 241 F (f 53). Because of the long half-life of Krypton -85 (10.76 years), Kr-85 levels remain in older fuel; however, the thermal driv-ing force required to cause its diffusion in defective fuel is greatly reduced (5 67) 352 356

_4_ Three 1007 capacit; purification pumps take suction at two pernanently installed skimmers and pump water to a demineralizer and filters located in the auxiliary building (1 72). The radiation levels of the demineralizers (which are shielded in cubicles) are usually from 1 to 4 R/hr (1 77) The filters (which are similarly shielded) are normally changed because of high pressure drop and usually have radi-ation levels of about 100 mR/hr (1 78). Any increase in the liquid or gaseous radioactive emis-sions from North Anna 1 and 2 resulting from the proposed modification are expected to be negligible (1 85). They will not violate NRC regulations, either during normal operation of the expanded fuel pool or under postulated fuel handling accidents (1 86). Vepco's seismic analysis of the new spent fuel storage racks and the pool structure shows that the racks can be accommodated by the structure during a seismic event (1 92). The techniques used by Vepco for analyzing the struct-ural integrity of the ~uel racks under normal, abnormal, and seismic loads are generally accepted in the engineering com-munity (1 97), Criticality calculations show subcriticality mainteined with a fuel assembly lying across the top of a rack or next to a rack (1 108). With the normal concentration of boric acid in the pool water, criticality cannot be attained with any possible array of fuel assemblies (1 109). ~ The accident defined as the dropping of a spent fuel as-sembly onto the spent fuel pit floor and the resultant rup-ture of the cladding of all the fuel rods in the assembly has been analyzed in 1 15.4.5.1 of the FSAR (1 111) The analy-sis was done in accordance with NRC Safety Guide 25 (1 112) The analysis of the fuel-drop accident shows that the accident would not result in excessive radiation exposure at ?r .>a2-3;7 the site boundary, that is, in exposures exceeding the guide-lines of 10 CFR Part 100 (f 113). Assuming as a worst case that the cladding of all rods in one entire fuel assembly fails, the offsite doses would not exceed the limits of 10 CFR Part 100 (1 114). An analysis of the effect of a small tornado missile has been performed in 5 15.4.5.2.4 and S 9.1.2 of the FSAR using accepted engineering techniques (T 115) Stored fuel in the spent fuel pit is protected from horizontal missiles by the thick reinforced concrete walls of the pit, which extend 20 feet, 10 inches above grade (1 116) The building geometry protects the fuel elements from dii,ct impact of missiles with angles of approach up to approximately 45 F above the horizontal (1 117). An analysis of t. _isk of turbine missiles has been done and is described in S 10.2.1 of the FSAR (1123) The FSAR turbine missile analysis was done with appropriate and sound calculational techniques (1 124). The FSAR analysis shows that the risk of unacceptable damage to the fuel build- -13 ing is 0 for low-traj ectory turbine missiles, 1.3139 x 10 per unit per year for high-traj ectory missiles at design over- -10 speed, and 1.3235 x 10 per unit per year for high-traj ec, tory missiles at destructive overspeed (i 125) The turbine missile analysis is not changed by the proposed modification (1 126). The amount of additional radiation to which the fuel would be exposed is insignificant in comparison to the levels in the reactor core during power operation (1 128) The zircalov cf the fuel and the 304 stainless of the fuel racks and pool liner are the same material no matter whether the high-density racks or the low-density racks are used (1 132). The proposed modification is not expected to make the eventual removal from the pool of the spent fuel assemblies any more difficult; to the contrary, af ter exter.ded 352 358

. storage the radiation levels and therefore the heat generated will have decayed to lower levels, so the handling and ship-ment of the assemblies will be easier (1 133). The spent fuel pool purification system is adequate to remove any,otential incremental impurities resulting from the proposed modification; most of the impurities are released during refueling, and the number of fuel movements will be no greater with the high-density racks than with the low-density racks (1 134) The exposure to radiation at Surry will probably slight-ly increase when additional fuel assemblies are stored; how-ever, because of the age of the earlier-stored fuel by the time the additional fuel reaches the pool, the increase should not be significant (1 142) Even if the exposure were doubled, to about 3 mR/hr, the exposure is a relatively minor contributor to the overall exposure at the station (1 143). Iodine-131 releases will not be significantly increased by the expansion of the fuel storage capacity, because the inventory of the I-131 (which has a half-life of 8 days) 2 the fuel will decay to negligible levels (1 149) Experience ac Surry indicates negligible levels of I-131 in the pool water; I-131 is noted only during refueling (1 150). Based on experience at the Surry Power Station, the radi-acion exposure is relatively low, approximately 150 mR for a filter change (1 153). There have been no overexposures associated with the Surry fuel pool (1 156). The pool walls and the pool liner cantict be reworked with spent fuel in the pool (1 171) There is not enough time to expand the spent fuel pool before the first refueling, and so the spent fuel would have to be transferred to other storage until the work was done (1 172) This would require finding 352 359

another licensed storage facility and double handling the fuel (1 173) The North Anna 3 and 4 fuel pool would have to be licensed by the NRC before it could be used to store spent fuel (1 178). The high-density fuel racks have already been fabricated and are at North Anna waiting to be installed (1 173) In the event of a leak from the spent fuel pool, several station systems are capable of providing make water until the leak could be repaired. Because the wate: vel in the pool would be maintained despite the leak, there would be no temperature increase in the pool. Instead, there would pro-bably be a decrease due to the addition of cooler make-up uater-The building geometry protects the fuel elements from direct impact of missiles with angles of approach up to cp-proximately 45 above the horizontal (1 117). According to technical papers by D. R. Miller, W. A. Williams, and T. L. Dcan, large missiles such as utility poles and automobiles (which are the design tornado missiles for North Anna 1 and

2) lack sufficient lift or velocity to clear a height of 25 feet (1 118)

These could not, therefore, strike the fuel elements (1 119). The spent fuel elements would be protected from ligb:er missiles by the water covering the storage racks in the pool (1 120) According to the paper by T. L. Doan, small fast-moving missiles trmveling downwards would impact only one fuel assembly (1 121) A tornado missile impacting the spent fuel would not result in radiation doses that exceed the limits of 10 CFR Part 100 (1 122), d rQ v H. Stephen /McKay / DATED: May 11, 1979 352 360

Signed and sworn to before me by H. Stephen McKay this i day of May, 1979. y/ / s Notary Public My commission expires ',.),, /'/ 352 301

STATEMENT OF PROFESSIONAL QUALIFICArIONS OF H. STEPHEN McKAY My name is H. Stephen McKa" My business address is P.O. Box 26666, Richmond, Virginia 23261. I am an Associ-ate Engineer with Virginia Electric and Power Company. I am presently Proj ect Engineer for the North Anna 1 and 2 high-density spent fuel storage rack proj ect. Formerly I was the Proj ect Engineer for the high-density spent fuel storage pro-ject at the Surry Power Station, Units 1 and 2. I hold a B.S. degree in physics from Moravian College, Bethlehem, Pennsylvania (1972), an M.S. degree in physics from Indiana University of Pennsylvania, Indiana, Pennsylvania (1974), and a Master of Engineering degree in nuclear engin-eering from the University of Virginia, Charlottesville, Vir-ginia (1976) 352 362}}

ML19224D390

Text

SUMMARY

1.0 INTRODUCTION

Navigation menu

Search