ML20135F521

Affidavit of H Walker in Support of NRC Staff Response in Support of Licensee Motion for Summary Disposition.* W/Certificate of Svc

ML20135F521
Person / Time
Site:	Oyster Creek
Issue date:	12/06/1996
From:	Walker H NRC (Affiliation Not Assigned), NRC OFFICE OF THE GENERAL COUNSEL (OGC)
To:
Shared Package
ML20135F471	List: ML20135F468 ML20135F521
References
OLA, NUDOCS 9612130096
Download: ML20135F521 (17)

Text

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l A'ITACHMENT 2 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION I

REFORF THE ATOMIC R AFFTY AND 1 TCENRINO ROARD In the Matter of )

)

GENERAL PUBLIC UTILITY NUCLEAR ) Docket No. 50-219-OLA CORPORATION )

)

(Oyster Creek Nuclear Generating )

Station) )

i AFFIDAVIT OF HAROLD WALKER IN SUPPORT OF THE NRC STAFF'S l RESPONSE IN SUPPORT OF LICENSEE'S MOTION FOR SITMMARY DISPOSITION I, Harold Walker being duly sworn, do hereby state as follows-1 l

1. I am employed by the U. S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation. My address and telephone number are:

Harold Walker, Senior Reactor Engineer (systems)

Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington D. C. 20555 (301) 415-2827 l

My professional qualifications statement is attached as Exhibit 1.

2. I am currently assigned to the Plants Systems Branch in the Division of Systems Safety and Analysis. One of my recent assignments was review of Oyster Creek Nuclear Generating Station's application to amend Technical Specification 5.3.1.B.

3. At all nuclear plants, overhead cranes are used to lift heavy objects in the vicinity of spent fuel. If a heavy object such as a spent fuel shipping cask or shielding block were to 9612130096 961206

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- fall onto spent fuel in the storage pool or reactor core during refueling and damage the fuel, there could be a release of radioactivity to the environment. Such an occurrence also has the potential for overexposing plant personnel to radiation. If the dropped object were large and the damaged fuel contained a considerable amount of undecayed fission products, radiation releases to the environment could exceed 10 C.F.R. Part 100 guidelines. With the advent of increased and longer-term storage of spent fuel, the NRC determined that there was a need for a systematic review of requirements, facility designs, and technical specifications (TS) regarding the movement of heavy loads to assess safety margins and improve them where necessary. This item was originally identified in NUREG-0371 (Task Action Plans for Generic Activities (Category A), U.S. Nuclear Regulatory Commission, November 1978) and was later determined to be an unresolved safety issue (USI).

4. In January 1978, the NRC published NUREG-0410 entitled, "NRC Program for the Resolution of Generic Issues Related to Nuclear Power Plants - Report to Congress." As part of this program, the Task Action Plan for Unresolved Safety Issue Task No. A-36, l

" Control of Heavy 1.oads Near Spent Fuel," was issued. l I

5. This USI (Task Action Plan Item A-36) was resolved with the publication of i NUREG-0612 (Control of Heavy Loads at Nuclear Power Plants Resolution of Generic Technical Activity A-36, U.S. Nuclear regulatory Commission, July 1980) and Standard Review Plan (SRP)Section 9.1.5.

6. Task A-36 was established to systematically examine staff licensing criteria and the adequacy of measures in effect at operating plants and to recommend necessary changes to assure the safe handling of heavy loads. The task involved review oflicensee information,

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evaluation of historical data, performance of accident analyses and criticality calculations, development of guidelines for operating plants, and review of licensing criteria.

l' NUREG-0612 provides the results of the NRC staff's review of the handling of the heavy i

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loads and includes the NRC Staff's recommendations on actions to assure safe handling of j heavy loads. These recommendations include: (1) a program to review operating plants against the guidelines developed in TASK A-36; (2) certain interim measures for operating

plants until completion of this review program; (3) changes to certain Standard Review Plans

, and Regulatory Guides to incorporate the guidelines in this report; (4) changes to technical

! specifications after completion of the review; and (5) initiation of a task to establish guidelines for the control of small loads near spent fuel. The guidelines proposed include definition of

{ safe load path, use ofload handling procedures, training of crane operators, guidelines on slings and special lifting devices, periodic inspection and maintenance for crane, as well as various alternatives that include: use of a single failure proof handling system, use of j l

mechanical stops or electrical interlocks to keep heavy loads away from fuel or safe shutdown l

I equipment, or analyzing the consequences of postulated heavy load drops to show these are within acceptable limits. NUREG-0612 completed Task A-36.

7. Mr. John C. Fornicola's Affidavit, offered by the Licensee in support of its Motion is correct in stating that the interim protective measure recommended by l

NUREG-0612, that licensees implement Technical Specifications like the one at issue was not

] adopted in the Generic letters following NUREG-0612. Fornicola Affidavit at i 19.

8. NUREGs are NRC documents issued for guidance and/or for acceptance criteria.

} NUREGs provide guidance that the NRC staff believes should be followed to meet

l requirements. A NUREG is not a substitute for the regulations, and compliance is not a requirement. However, an approach or method different from the guidance contained therein will be -~jt~I only if the substitute approach or method provides a basis for determining that regulatory requirements have been met. NUREGs issued to establish acceptance criteria provide criteria that the NRC staff uses in evaluating whether an applicant / licensee meets applicable requirements. Here again, a NUREG is not a substitute for the regulations, and compliance is not a requirement. However, the use of criteria different from those set forth in j the issued NUREG is acceptable only if the substitute criteria provide a basis for determining that regulatory requirements have been met. NUREG-0612 provided guidance and acceptance criteria; it is not a regulation. (Exhibit 2 is an NRC management directive concerning NUREGs.)

9. Furthermore, as stated in Mr. Fornicola's affidavit, Technical Specification 5.3.1.B, as issued, was adopted more than three years prior to the publication of NUREG-0612 and, therefore, was not in response to NUREG-0612. The Technical Specification, as .

originally issued, reads as follows: " loads greater than the weight of one fuel assembly shall not be moved over stored irradiated fuel in the spent fuel storage facility." It was established at Oyster Creek as part of a 1977 amendment that increased the spent fuel pool storage l l

capacity from 840 to 1800 fuel assemblies by replacing the existing storage racks with more closely spaced storage racks.

10. NRC regulations applicable to this issue are General Design Criteria (GDC) 2, 4, 5 and 61. GDC 2 is applicable as it relates to the ability of structures, equipment, and mechanisms to withstand the effects of earthquakes. Acceptance is based in part on meeting .

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1 position C.1 of Regulatory Guide 1.29 (attached as Exhibit 3 for information, but not applicable for this issue) for safety-related equipment and position C.2 for nonsafety-related  ;

equipment, and positions C.1 and C.6 of Regulatory Guide 1.13 (Exhibit 4). GDC 4 is applicable as it relates to protection of safety-related equipment from the effects of internally generated missiles (i.e. dropped loads). Acceptance is based in part on meeting positions C.3 l

and C.5 of Regulatory Guide 1.13. GDC 5 is applicable as it relates to the sharing of equipment and components important to safety, and GDC 61 is applicable as it Islates to the  ;

safe handling and storage of fuel.

Specific criteria / guidance for meeting the relevant requirements of General Design Criteria 2,4 and 61 are as follows: .

i

a. NUREG-0554 I

b. NUREG-0612 I

c. ANS 57.1/ANS N208

d. ANS 57.2/ANS N210

11. It is important to note that criteria / guidance are not regulations, but rather  ;

represent one acceptable way of meeting the regulations.

12. The NRC staff suggested that GPUN amend the TS to clarify that moving the shield plug onto the top of the dry storage canister in the transfer cask in the cask drop protection system was acceptable. The Staff had always recognized that the shield plug of any shipping or transfer cask would have to be placed on the top of the canister / cask and the TS was not meant to prohibit this. The Staff believed that, because it was not entirely clear that i

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j the movement at issue was permitted by the technical specifications, it was in the Licensee's 4

best interest to amend Technical Specification 5.3.1.B to make it clear that this activity was l permissible.

! 13. The information set forth above is true and correct to the best of my knowledge

and belief.

1

/IIarold Walke7 '

i Sworn and stibscribed to before me this blD y of Decem 1996.

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My commission expires:

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NOTA 7/ ! LEUC D,T2 OF MAD'tAND My Comnricn Expires De: ember 1,1999 l

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i PROFESSIONAL QUALIFICATION OF HAROLD WALKER i

I am a Senior Reactor Engineer in the Special Projects Section of the Plant Systems Branch, l Division of Systems Safety and Analysis, Office of Nuclear Reactor Regulation, United States i Nuclear Regulatory Commission. My duties include serving as a principal reviewer in the area

{ of nuclear plant protection to assure against various hazards and certain aspects of containment, i radioactive waste processing and other support systems assigned to the branch. Prior to this

! assignment, I was a Mechanical Engineer in the Electrical, Instrumentation and Control Systems j Branch, where I reviewed the electrical operability and functional capability of mechanical and electrical equipment, mechanical components, and their supports needed for safe operation and

! safe shutdown of nuclear facilities.

i l Prior to being assigned to the Electrical, Instrumentation and Control Systems Branch, I was a I

Mechanical Engineer in the Equipment Qualification Branch where my duties included j

performance of technical reviews, analyses and evaluations of the adequacy of the environmental i qualification of electrical and mechanical equipment whose failure, due to such environmental

conditions as temperature, humidity, pressure and radiation, could adversely affect the  ;

l performance of safety systems. I was previously a Materials Engineer in the Materials l

Engineering Branch where my duties and responsibilities involved the reviev and evaluation of l materials performance from the standpoint of operability and functional capability and integrity i l under normal, abnormal, and accident loading conditions, and analyzing fracture toughness of l' l reactor vessel materials, including specific data to assure that the materials will behave in a non-i brittle manner.

l

! Prior to my position in the Materials Engineering Branch, I was a Materials Engineer in the l

Engineering Branch, Division of Oper& ting Reactors. My duties and responsibilities included j the review of operating problems to determine whether safety requirements were being satisfied ,

i and to assure that operating problems were corrected, and met with due regard for safety and environmental protection.

i

Prior to my position in the Engineering Branch, I was an ACRS Fellow at the Advisory i Committee on Reactor Safeguards. My duties included collecting and consolidating information J

pertaining to non-destructive testing methods.

i I have a B.E. degree in Mechanical Engineering from the City College of the City University l of New York and I have taken graduate courses at the University of Pittsburgh.

) Prior to joining the NRC, I was an engineer at Westinghouse Research Corporation in l Pittsburgh, Pennsylvania where my duties included the application of state of the art fracture i mechanics as well as the study of structural integrity of materials in various environments and under various loading conditions.

I December 1996 1

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, i EXHIBIT 2 l .. '

U.S. NUCLEAR REGULA TORY. COMMISSION

N' '

i ., ..

DIRECTIVE TRANSMITTAL

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TN: DT-95-03

'Ib: NRC Management Directives Custodians Sub.iect: Thmsmittal of Management Directive 3.7, "Unclassi5ed Staff Publications in the NUREG Series" Purpose Directive and Handbook 3.7 are being revised in their entirety to update information, to include format changes, and to add i

information about references. Specifically, infonnation has been l l added to Directive and Handbook 3.7 to specify that NRC must h

! obtain prior approval from the Institute of Nuclear Power j Operations (UIPO) before referencing INPO documents and to j explain how to reference proprietary reports.

I i Office and '.

{  % Division of Origin: Office of Administration "

l Division of Freedom ofInformation and Publications Services

Contact:

Juanita Beeson,415-7166 j

Date Approved: April 23,1991 (Revised: February 9,1995)

Volume: 3 Information Management '

l Part: 1 Publications, Mail, and Information Disclosure  ;

~

i Directive: 3.7 Unclassified Staff Publications in the NUREG Series Availability: U.S. Government Printing Office, (202) 512-2409  ;

h J

O 4 1

I OFFICE OF ADMINISTRATION '

1

e

' Volume 3, Part 1 - Publications, Mail, and Information Disclosure Unclassified Staff Publications in the NUREG Series ,

i I

Handbook 3.7 Part II . i i < l i ,

! . l Disclaimers (F) m; l

i

Guidance and Acceptance Criteria for Ucensees (1) -

l NRC positions communicated to licensees are not binding

requirements unless they are formally issued as regulations or

included in orders or as part of a permit or a license. Accordingly, reports that provide guidance and acceptance criteria to licensees  ;

j should contain one of the following statements or its equivalent. (a) l For guidance:(b) l b

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These statements must be printe6 in a box that appears on the title page of the report or they may be incorporated into the preface or introduction to the report, if appropriate. If placed on the title page, the DFIPS staff will produce the boxed disclaimer when creating the cover and title page for the report.

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Approved: April 23,1991 8 (Revisad Fahruavv 9. IM

! e EXHIBIT 3 f

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, as Revision 3

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U.S. NUCLEAR REOULATORY COMMISSION g 3,yg

@g) OFFICE OF STANDARDS GUIDE DE REGULATORY GUIDE 1.30 8

SEISt00C DESIGN CLAStrICATION .

l A. INTRODUCTION mares of light-wassr cooled suelear plaats abat l' General Design Criterion 2, " Design Bases for should be designed to whbstand e8Eare. of the l Protection Against Natural Phenomena." of Appen- SSE. The Advisory Committee os Reactor l dix A. " General Design Criteria for Nuclear Power Safegends he boss ceaselend regeding his guide Plants," to 10 CPR Part 50, "rha===eie Licensing and be concesed is te agulasary posidos.

of Pmoucima sad Utilizados Pacinties," mquires that nuclear power plant structures, sysseans, and S. OsSCUSSION coriponents important to safety be designed to with- After revowing a number of appbcesions for con- l stand the effects of earthquakes without loss of caps- permis and M heensa for boiling '

' bility m perform their safety functaoss. and peessorised woest nuclear power plants, abe NRC 3

staff has developed a seismic design classification  ;

Appendix B, " Quality Assurance Criesria for Nu- system for identifying those plant featmos that abould clear Power Plants and Puel Reprocessing Plants," to be designed to wisbotand the ofBeces of the SSE.

10 CFR Part 50 establishes quality assurance re- Dose structures, systems, and components that quirements for the design, construction, and opera- should be designed to remais flametaceal if the SSE i tion of nuclear power plant structures, systems, and occurs have been designaamd as Soissue Cassgogy 1.

components that prevent or mitigase the consequences of postulated accidents that could cause undue risk to C. REGULATORY POSffl0N the health and safety of the public. De portaaset re-

' quirements of Appendix B apply to all activities af- 1. De following structures, systems, and cosopo-facting the safety related functions of those struc- aonts of a seclear power plant, including absir foun-l sures, systems, and components. dations and supports, are designated as Seismic Cate-gory I and should be designed to withstand abe effects i

i Appendix A, " Seismic and Geolo ic Siting of the SSE and rosmala functional. The partineet qual- l i

Criteria for Nuclear Power Plants," to I CPR Part in assurance mquiremens of M B a 10 Cm 100, " Reactor Site Criesria," requires that all au- Pan 50 should be appued w all acth ameting  ;

! clear power plaats be dcaigned so that, if abs Safe he safen-related functions of 6em semcares, sys- '

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Shutdown Eanbquake ($5E) occurs, certain struc. eses,and P . .

j tures, systems, and components reanais faectional.

These plant features are abose necessary to ensure (1) a. De reactor coolant pressure w-a y. i i

the integrity of the reactorcoolant pressure M.=dary, b. De reactor core and reactor vessel internals.

j (2) the capability to shut down the roertor and main-

esia it in a safe shutdown. condition, or (3) abe caps- c. Systems 8 or portions of systems that are re-i bility to prevent or mitigate the coesequences of ac- quired for (1) emergency eare cooling, (2) postacci-

{ cidents that could result in potential offsite exposures

! comparable to the guidelias exposures of 10 CPR .uese teessee est,maseve seneses from previses lesse.

Part 100. ine sysum hm,meny inciman es. pasene er es syssse re-geeses e assempelsh me spectnes sessy passeios and seenessed his guide describes a metbod acceptable to the pipios,,er.se w ) e, i, and

,in,,imenemme,,es armorveh,e peelemas ee NRC staff for identifying and classifying thoes fan-

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l dont containment bes* removal, or (P pontaccident n. The control r m, including its associ:ted t

oostainment atmosparre cleanup (. .. a gen re- equipine::t and all equipment needed to maintain the e

moval system). co

tr:I room withis safe habitability limits for .

i d. Syssess8 or portions of systems eat are re- Pennael W d enhmaW Mts br Wtal quired for (1) reactor shutdown, (2) residual heat re- 84miP emt J

] moval, or (3) cooling the spent fuel storaps pool. o. Primary and secondary reactor containment.

i e. Those portions of the steam systems of boil- '

i ing water reactors extending from the outermost son. P. Syseems,8 seer than m&oactive weste man-l tainment isolation valve up to but not including the asement systems,2 set covered by items 1.a through i

marbios stop valve, and cosaected piping of 2% 1.o above that contain or may contain resoective ma-

! inches or large nominal pipe size up to and including emedal and whose posntiated failwe would result in i to first valve that is either normally closed or cepe, oosservatively calculated pometial oNaim deoes (us-i ble of automatic closure during all modes of normal lag metoomlogy as recomsnended is Regulatory i reactor operation. The turbine stop valve should Guide 1.3, ".C ,'--- Used for Evaluating the l be designed to withstand the SSE and ===i=*=*= its Possetial Ra&ological & _ y:: of a 14ss of I hugrity. Coolant Accident for Boiling Water Reactors," and Regulatory Oside 1.4, "Assumptiosa Used for

f. Those portions of the steam and feedwater Evaluating the peeestial Radiological C--  ; a systems of pressurized water reactors extending hem of a 1. mas of Coolant Accident for Paseurimod Wasst and including the secondary side of seeam generators Reactors") that are more than 0.5 rues to the whole op to and including the outermost contaiaraset isole- body or its equivaleet to any past of the body.

elon valves, and connected piping of 2% inches or larger nominal pipe size up to and including the first q. The Class IE elecanic systems, including the valve (including a safety or relief valve) that is either emailiary systems for the easite electric power normally closed or capable of automatic closure dar- supplses, that provide the emergency electric power ing all modes of normal reactor operation aseded for functioning of plant foamares included in leses 1.s through I.p above.

g. Cooling water, component cooling, and auxiliary feedwater systems 8 or portions of these sys-

, 2. Those portions of strucanos, syseenas, or com-esms, including the intake structures, that are re-ponents whose continued function is not required but quired for (1) emergency core cooling, (2) postacci- whose failure could reduce the Asoctioning of any dont containment heat removal (3) posteccident com- plant festwe included is items 1.s through 1.q above tainment atmosphere cleanup, (4) residual heat re- l to an uw-M ^ N safety level or could result in in- l moval from the reactor, or (5) cooling the speet fuel capacitating "lajory to occupants of the comerol room storage pool, should be designed and constructed so that the SSE  !

h. Cooling water and seal water systems 8 or would not cause such falhare.8 portions of these systems that are required for func.

tioning of reactor coolant system components unpor- 3. Seismic Category I design requiremsets should tant to safety, such as reactor coolant pumps. eM m te But amin retraim es de-Bood boundaries. Those portions of strucesses, sys-

i. Systems' or portions of systems that are re- tems, or pa===** that form interfaces between quired to supply fuel for emergency equipment. Seismic Caessory I and non-Seismic Caesgery I fee-

j. All electric and mechanical devices and cir- tures should be designed to Seismic Category I g,,,,, ,

cuitry between the process and the input terminals of the actuator systems involved in generating signals 4. The partineet quality assurance requirements of that initiate protective action. .

Appendia B to 10 CPR Part 30 should be apphed to

k. Systems' or portions of systems that are re. all activities aNocting the safety-teleend functions of quired for (1) monitoring of systems important to those portions of structures, sysums, and compo-safety and (2) actuation of systems important to nests covered under Regulatory Positions 2 and 3 safety. above.

1. The spentJuni storage pool structure, includ-ing the fuel rocks. , s,,g, ,,w,,,, ,, g,w, ,m ,, ,,,,,,,, o

m. The reactivity control systems, e.g., control rods, control rod drives and boros injection g gf ameld M casu most % sheeW he selw.med w nye system. eased = ihe essem required as emanesse mis pessnery.

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1.29-2

e D. IMPLEMENTA'i10N cant propo .s an acceptable alte native method for L

mmplying wi3 specified poruons of the Commis-i .

The purpose of this section is to provide informa. sion's regulations, the method described herein is tion to applicants regarding the NRC staff's plans for being and will continue to be used in the evaluation

, using this regulatory guide. of submittals for operating license or construction permit applications until this guide is revised as a re-4 This guide reflects current NRC staff practice, suit of suggestions from the public or additional staff Therefore, except in thow cases in which the appli- review.

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) , EXHIBIT 4

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e i U.S. NUCLEAR REGULATORY COMMISSION Resisten 1

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l REGULATORYGU DE

OFFICE OF STANDARDS DEVELOPMENT l

I i

i REGULATORY GUIDE 1.13 i .

i FENT FUEL STORAGE FACILITY DESIGN SASIS a 1

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! A. M RODUCTIM g, g,,, ,g y,,,, g,,,, gg,,,,, p ,g i

l

• General Design Criterion 61, " Fuel Storey and Unises protective snesswes are taken, loss of water j Hand!mg Criteria for Nuclear Power Plants," of Appen.

from a fuel stotap pool ccauld cause ting of the j 6x A. " General Design Criteria for Nuclear Power spent fuel and resultant damer to f dding intes-j Plants," to 10 CFR Part 50, "r__w S of Production rity and could result in release of ti serials to j and Utilization Facilities,', requires that fuel storay and the environment. Natural uakes or handling systems be designed to assure adequate safety

, under normal and postulated accident constions. It also ggg .g g ,

requires that these systems be designed with appropnate by se pnwado: of .

@6 l could also cause stru

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j contamment. confinement, and filtering systems and be gg"., to fall into the p,,g, g g,,,,,,,,,  !

designed to prevent significant reduction in the coolant inventory of the storage facility under accident conds-whout d m @ W ty

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would al i tions. This guide desenbes a method acceptable to the j j NRC staff for implementing this criterion.

! o vy loads, such as a 100 ton fuel cask, '

j B. DISCUSSIM probability, cannot be ruled out in plant j nts where such lands are positioned or nuoved

it is important that fuel handling and storap over es fuel pool. Possible soluuons to eis 4 be designed to

l tential problem include (1) preventing, preferably by l ossip rather than interlocks, hasvy loads from being I l a. Prevent loss of water from the pool lifted over the pool; Q) using a highly rehable handing l 3 would uncover fuel. system desiped to prevent dropping of heavy loads as a I i result of any single failum; or (3) densping the pool to' s b. Protect the fuel from mechanical dannap.

withstand dropping of the load without sipificant l leakap from the pool area in which fuel is stored.

c. Provide the cape ting the potential '

l j offsate exposures se ificant release of  !

i radioactivity fr Even if the menswes described above to prevent loss of lenk tight integrity are fouowed, small leaks may still if s cilities are not located within occur as a result of structural failwe or other unforeseen the pn ntainment or provided with events. For exarapie, equipment faawes in systems adequate etive featwes, redsonctive materials could connected to the pool could result in loss of water from be released e environs as a result of either loss of the pool if such loss is not prevented by desip. A water from the storage pool or mechanical damap to permanent f ';--c' c- '- ^ raskeup system with a fuel within the pool. Inoderate capability, and with suitable redundancy or

.% m. i,,t.ne., ch.asse froen pse. toes iss . backup, could prewat the fuel from being uncowmd if usunc neout. Atony ousoss c- w= . c

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i such leaks should occur. Early detection of po d ienkap sidtably contround duras sofuehag operations. The and fuel damap could be provided by poolwter level desige of the ventdation and futration system should be

! momsors and radiat2on monitors designed to a*.wi,d.

based ou 114.memption that the cladding of at of the locally and in a continuously manned locataan. Tienely

! operation of building filtration systems can be answed fuel rods in one fust bunde miWit be breached. The 8 inventory of resoective meterials avaEable for leakap by actuating these systems by a sipal from local from the bueding should be based on the annunptions

radiation monitors.

j giwa in Regulatory Guide 1.25. "r . c Used for Evaluating the Posential Radiological Consequences of a i

2. haelmnical Densage to Fort Fuel Handhng Accident in the Fuel Handung and Storay Facehty for toiling and Pressustaed Water l The release of radioactive material from fuel may Reactors"(Safety Guide 25).

occw dunng the refuehng process,and at other tunes, as 5. The spent fuel storage faceity should have at least a result of fuel. cladding faGures or mechanical damage one of the fo5owsig provimons with respect to the

{ caused by the dropping of fuel elements or the dropping handung of heavy loeds, inchuling the refuelinf, ;ask:

of objects onto fuel elements.

l a. Cranes capable of carrying heavy loads should i Missues pnerated by high wmds can also be a be pewated, paferably by design rather than by i

j potential cause of mechanical damap to fuel.Desiping interlocks, frans snoving into the wichnity of the pool; or the fuel storep facility to prevent such missues from

! contacting the fuel would eliminate this concern. b. Cranes should be damped to pr6 vide single.

l failure-proof handling of heavy loads, so that a single l

A relatively smaU amount of mechanical damey to failwe wil not sesult in loss of capabihty of the l the fuel might cause significant offsite doses if no dose crane.handhng sysum to perform its safety function; or

reduction features are provided. Use of a controlled j isakage buildmg surrounding the fuel storap pool, with c. The fuel pool should be densned to withstand, 4

associated capability to limit releases of radioactsve without leakage that could uncowr the fuel,theimpact i material resulting from a refueling accident, appears of the heaviest load to be carried by the crane from the j maximum height to which it- can be lifted, if this feasible and would do much to elirninate this concern.

j approach is used, design provisions should be made to f

C. REGULATOftY POSITION P. event the crane, when carrying hoevy loads, from movmg in the vicinity of stored fuel.

4

1. The spent fuel storage facility (including its j structures and equipment except as noted in paragraph 6 6. Drains, permanently connected mechamcal er j below) should be designed to Category I seistruc require-ments. hytirauhc systens, and other features that by analoper.i-j tion or faGute could cause loss of coolant that woukt uncover fuel should not be installed or included in tri:

2. The facihty should be designed (a)to keep tor. desip. Systems for maintaming water quality and l nadic wmds and missiles pnerated by these wmds from quantity should be designed so that any maloperation or caunng sigruficant loss of watertight integnty of the fuel fauure of such systems (including failures resulting from storap pool and (b)to keep missiles pnerated by the Safe Shutdown Earthquake) will not cause fuel to bel

( l tornadic winds from contacting fuel within the pool. uncovered. These systems need not otherwise meet

{ Category I semanuc requhements.

! 3. Interlocks should be provided to prevent cranes 1 from passms over stored fuel (or near stored fuelin a 7. Reliable and frequently tested monitoring equip.

4 mantwr such that if a crane failed, the load could tip rnent should be provided to alarm both locauy and in a over on stored fuel) when fuel handhng is not in continuously manned location if the meter level in the propess. Dunng fuel handhng operations, the interlocks fuel storap pool falls below a predeterminei level or if may be bypassed and adnunistrative control used to high local-radiation levels are experienced. The high-prevent the crane from carrying loads that are not j radiation-level instrunnentation should also actuate the necessary for fuel handhng over the stored fuel or other filtration system. l prohibited areas. The facility should be designed to minmaze the need for bypassing such interlocks. 8. A seismic Category I makeup system should be

' provuled to add coolant to the pool. Appropnate

4. A controlled leakage building should enclose the redunda' icy or a backup system for filhng the pool from fuel pool. The bmiding should be equipped with an a reliable source, such as a lake, river, or onste seismic appropnate ventilation and filtration system to limit the Category I water. storage facility, should be provided. If potential release of radioactive iodine and other radio. a backup system is used,it need not be a permanently ,

active matenals. The building need not be desiped to instdled systern. The capacity of the makeup systems withstand extremely high winds, but leakage should be should be such that water can be supplied at a rate 1.13 2 s l

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desarmined by consideration of the leakage rate that D. IMPLEMENTATION would be expected as the result of damspe to the fuel i

storage pool from the dropping of loads, from earth-quakes, or from massales origmatmg in high winds.* Any of the alternatives in Regulatory Position C.5 of Reva on I sney be applied at the option of applicants 1 for construction permits and operating licenses for all

'The staff u considenng the development of addataonalguidance plants, reprdless of the date of application.

concemmg prossetion agamst masades that might be pnerstad by plant faduns such as turbine faihares. For the prenant, the protecton of the fuel pool against such susales will be evaluated on a case by case bans.

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00CKETED l USMRr '

4 UNITED STATES OF AMERICA l NUCLEAR REGULATORY COMMISSION

'96 DEC -9 P2 :39 REFORF THE ATOMIC RAFFTY AND 1_ICENRING ROARn  !

0FFiCE N "! 'N ETARY DOCM iM ' 'RViCE I In the Matter of ) HH W H  !

)

. GENERAL PUBLIC UTILITY NUCLEAR ) Docket No. 50-219-OLA l CORPORATION ) l

, ) l (Oyster Creek Nuclear Generating ) l Station) )

CRRTIFICATE OF RRRVICE

I hereby certify that copies of "NRC STAFF RESPONSE IN SUPPORT OF LICENSEE'S MOTION FOR

SUMMARY

DISPOSITION" in the above-captioned proceeding have been served on the following through deposit in the Nuclear Regulatory Commission's internal mail system,

, or by deposit in the United States mail, first class, as indicated by an asterisk, or hand-delivered as indicated by a double asterisk or by facsimile or E-mail transmission with a conforming copy,

as indicated by a triple asterisk this 6th day of December,1996:

1 G. Paul Bollwerk, III, Chairman ** Office of the Secretary (2)"

! Administrative Judge U.S. Nuclear Regulatory Commission Atomic Safety and Licensing Board Washington, D.C. 20555  ;

U.S. Nuclear Regulatory Commission Attn: Docketing and Service Branch l 4

Washington, D.C. 20555  ;

I Dr. Peter S. Lam ** Dr. Charles N. Kelber" Administrative Judge Administrative Judge

Atomic Safety and Licensing Board Atomic Safety and Licensing Board

U.S. Nuclear Regulatory Commission U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Washington, D.C. 20555 i

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l Adjudicatory File (2) Ernest L. Blake, Jr.***

Atomic Safety and Licensing Board David R.12wis U.S. Nuclear Regulatory Commission Shaw, Pittman, Potts & Trowbridge Washington, D.C. 20555 2300 N. St., N.W.

Washington, DC 20037-1128 j Paul Gunter*** Michael Laggart*

Nuclear Information and Resource GPU Nuclear Corporation ,

Service 1 Upper Pond Road  !

i 142416th St., N.W., Suite 404 Parsippany, NJ 07054 l Washington, DC 20036

) Atomic Safety and Licensing Board Office of Commission Appellate

Panel (1) Adjudication (1)

U.S. Nuclear Regulatory Commission U. S. Nuclear Regulatory Commission Washington, D.C. 20555 Washington, D.C. 20555 i 1

Deborah Katz* William decamp, Jr.*** I Citizens Awareness Network Oyster Creek Nuclear Watch j P.O. Box 83 P.O. Box 243 3

Shelburne Falls, Massachusetts 01370 Island Heights, NJ 08732 l i

)

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' / (/LtL ,

e cf 'A j Annp.Hodgdon Counsel for NRC Staff i

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