ML20112J865

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Affidavit.* Affidavit of D Katz Re Effects of Experimental Transfer of Fuel at Plant While Reactor Operational Into Dry Cask Storage
ML20112J865
Person / Time
Site: Oyster Creek
Issue date: 06/05/1996
From: Katz D
CITIZENS AWARENESS NETWORK, NUCLEAR INFORMATION & RESOURCE SERVICE
To:
Shared Package
ML20112J781 List:
References
OLA, NUDOCS 9606210120
Download: ML20112J865 (5)
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t AllislayitJ2Llkirrah%atz i

1 1, Deborah Kat; bemg Ady swom, appear and state the following I

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' l. Ilive at 80 Davenpmt Road in Rowe, Mawachusetts. My home lies witlin 17 niles of  !

the Verm<.mt Yankco Nuclear Power Station (VYNPS) 1 l

2. I am a member of Nuclear Infimnarkm and Rewurce Service aru! the '

Cairens Awanmees Network (CAN) . i

3. I am concemed about the effects of the caperimental runsfer of fid at the Oywar .i Creek Nuclear Power Staten winle the tensor is operadonalinto dry cask storage. nc #

fuel wE be transfered over the aperanng stactnr vensel. It will have a & rect effect on

  • my heakh and safety. Since Vermont Yankee is a Msrk 1 hoiling water reactor l diet that will he set bv the pmccas at Oynter Creek the Oyster Creek reactor, the y j can dweetly effect me. In particular, I have the followmg concerm.

a)I travel rcgularly to Greenficki , MA and Brattlehom, VT which are wahin the 10 mile tidius of the Vermont Yankee Nuclear Power Stati am concemed phnt dwmid a finel -handimg accident secur, I and my children would be exposed to unsafe radiation kws. i B) I am also concerned about the potentially devastating offects on my and )

my children's health and stafety by a radiological accident at VYPS due to  ;

the movement ofirradiated fuelin contammet We live in the efBuent  !

pathway of the tractor. Our fanu'ly and our environment could be permanently contandnated by surJ1 an x9 C)I am aim concamed about the novement of fact at VYPS since thl Nuclear Engmoer of Vermont, Mr Winimm Sherman, announced at a town l meeting in Bacidend, M A that VYPS wondd bepn dry cask storage of their l

fuelwahm 5 years.  !

I am willsag to have Nuc1 car Intionaten and Resxrtcc Service represent my intercats at .

the hearing and during the intervesenn amcre S'gaad a Date

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Attachment Affidavit For Jean Burnette Oyster Creek Nuclear Watch 715 Chesapeake Drive j

. Forked River, NJ 08731 l t

Notarized and postmarked to NRC by First Class Mail on June 5,1996 l r

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- Amdavit for Shirley R. Schmidt

Oyster Creek Nuclear Watch 291 Wells Mill Road Waretown, NJ 08758 609/971-6162 Notarized and postmarked to NRC by First Class Mail on June 5,1996 ,

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Attrchment Affidavit of Maria Szczech Ocean Township Committeewoman Ocean Township 50 Railroad Avenue Waretown, NJ 08758 609/971-1905 609/693-3302 Notarized and postmarked to NRC by First Class Mail on June 6,1996 .

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OM'B No. 3150-0012 NRC3 96-02 '

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. UNITED STATES NUCLEAR REGULATORY COMMISSION 0FFICE OF NUCLEAR REACTOR REGULATION WASHINGTOM 0.C. 20555-0001 April 11. 1996 1 MOVEMENT OF HEAVY LOADS 0VER SPENT FUEL, OVER FUEL NRC BULLETIN 96-02: IN THE REACTOR CORE, OR OVER SAFETY-RELATED EQUIPMENT Addressegi All holders of boiling-water reactor (54) and pressurized-water reactor (PWR) ;

operating licenses for nuclear power reactors. }

puroose The U.S. Nuclear Regulatory Commission (NRC) is issuing this bulletin to accomplish the following:

4 Alert, addressees to the importance of complying with existing regulatory .

(1) guidelines associated with the control and handling of heavy loads at l nuclear power plants while the plant is operating (in all modes other l than cold shutdown, refueling, and defueled) and remind addressees of l their responsibilities for ensuring that heavy load activities carried I out under their license are performed safely and within the requirements specified under Title 10 of the Code of Federal Regulations.

(2)

Request that addressees review their plans and capabilities for hand heavy loads (e.g., spent fuel dry storage casks, reactor cavity ,

biological shield blocks) in accordance with existing' regulatory I guidelines [specifically NUREG-0612 (Phase 1 final safety analysis' report (FSAR).

(3)

Require addressees to report to the NRC wnether and to what exten have complied with the requested actions contained in this bulletin.

Although this bulletin is particularly concerned with heavy load movement while the plant is operating (i.e., in all modes other than cold shutdown,

. refueling, and defueled), the staff is considering further generic actions on the issue of handling all heavy loads both while the plant is operating and 1

during shutdown.

Backaround l

There are a number of heavy loads being handled in various areas of nu 4

" power plants, especially over safety-related equipment, when the i

J 0a0802-59 i

NRCB 96-02 April 11, 1996

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Some licensees have moved or are planning to move heavy loads such  ;

as spent ruel shipping casks, transfer casks, and reactor cavity biologicalIf th) snield blocks during plant operations.

movement or are dropped on safety-related equipment, the equipment may be l.

unable to perform its function.

Guidelines regarding the movement of these and other heavy loads are provided in a number.of documents that in combination make up the framework The most for the existing regulatory position on heavy load handling and control.

important guidelines are contained in the following three documents:

NUREG-0612, " Control of Heavy Loads at Power Plants," Resolution of (1) Generic Technical Activity A-36, issued July 1980 Unnumbered generic lettei dated December 22, 1980, " Control of Heavy (2)

Loads" GL 85-11

" Completion of Phase !! of Control of Heavy Loads at Nuclear (3) Power Plants, NUREG-C612," dated June 28, 1985 '

is to (1) ensure the safe handling of heavy load:  ;

NUREG-0612 provides guidelt incontrolled movement of heavy loads or load l

The

-(2) reduce the potential fc aquences of dropping a heavy load.

drops, and (3) limit the et Some i guidelines were supported by historical data and fault tree analyses.  ;

l portions of the guidelines were generic to all plants, while others were specific to plant type and location (e.g., the PWR containment building).

The guidelines consider the handling of heavy loads while the reactor is at power and provide a methodology to do so safely.

22, 1980 requested that licensees '

The unnumbered generic letter of December implement the heavy load control guidelines in NUNG-0612 and ji problems that they encountered.

implementation of some. interim actions (safe load paths, crane design an inspection, operator training, and procedures), a 6-month followup response o the status of the implementation of Section 5.1.1 of NUREG-0612 (Phase I),

a 9-month followup response on the status of the implementationsingle- of the l remaining applicable portions of Section 51 of NUREG-0612 (Phase II:

failure-proof cranes, stops / interlocks, or load-drop analyses).

All affected licensees implemented the interim actions and Phase I of theThe sta generic-letter and submitted a response for Phase II. implem that the actions taken by the licensees had significantly decreased th l potential for a heavy load drop. remaining Phase II submittals and did not concerns a sociated with the control of eavy loads.

Subsequently, the staff issued GL 85-11, which informed licensees that ,

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' implementation of Phase II was not necessary but j GL S5-11 relieved licensees from performng the actions requested under )

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NRCB 96-02 April 11,1996 Page 3 of 9 J However, GL 85-11 did not grant l Phase 11 of the previous generic letter.blan'et basis for NRC approval for a))

heavy load

- nor did it authorize licensees to exceed their design transfer. f l

i Although the generic letter stated that the NRC st '

i time, it did not preclude the possible future need for the staff to reviewl additional heavy load handling concerns and to require, as appropriate, further actions by licensees.

i Descriotion of Circumstances i In 1996, GPU Nuclear (GPUN) Corporation, the licensee for the Oyste  !

Nuclear Power Plant, is scheduled to begin GPUN moving heavy is planning loads involvinl to load '

7 storage casks within the Oyster Creek facility. The loaded spent casks, fuel f +

4 placed in an independent spent fuel storage installation. '

each weighing 100 tons, must be moved over safety-related equipmen The licensee's plans involve loading and moving the casks i this process.

during power operation because performing these activities during  ;

outage would significantly increase the outage time. l

' The licensee prepared an initial evaluation pursuant to 10 CFR 50.5  ;

the planned activities for handling the dry storage casks, including l th, of the non-single-failure-proof reactor building To crane to reduce the transfer spenl

' to the dry cask storage facility during plant operation. ,

probability of a load drop, GPUN modified its crane; proposed tol '

pad along part of the load path; and proposed However, tol

cask-handling procedures specific to this evolution and development.

i during two portions of the proposed cask movement inside the re:

a cask drop could damage both isolation condensers and This the torus creating an unisolable loss-of-coolant accident outside containment.

drop could occur in those areas near the spen A cask dropped from either of these i the 119-foot level is not installed. locations on thestated The licensee 119-foot that leve into the torus, damaging all equipment in its path. l core cooling could be maintained by steaming to the condenser using feedwater system and providing makeup from the condensate storag fire water systems by way of the core spray system.

the probability of dropping the cask, the staff was concerned tha casks are heavier than previously considered in the FSAR, a cask result in higher consequences than those previously analyzed. '

As a result of concerns raised by tne staff k ndand GPUN's to minimize the efforts to i

. efficiency of handling the spent fuel storage cas s a 4

probability of a cask drop, GPUN updated its 10 CFR 50.59 evalu e include a number of improvements applicable to Phase I.

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NRCB 36-02 April-11, 1996 l

, Page 4 of 9-

, I the reactor building crane (but not to the level of a single-failure-proof l crane as defined in NUREG-0554, " Single-Failure-Proof The Cranes fixed links for provide Nuclear Power i ,

Plants") ty. installing a fixed link support system. redundant rigging fo It uses horizontal especially in the area _over the isolation condensers. ,

support beams attached to the cask-lif ting yoke and vertical tie-rods  ;

connected to the crane trolley to support the cask in the event of a failure l of a crane hoist component.

f GpVN evaluated postulated load drops while the cask is in the reactor building equipment hatchway (from the 119-foot elevation to the 23-foot elevation) andi at the laydown area on tha 119-foot elevation where the fixed links are not  !

engaged and concluded that if a cask is dropped in either Consequently, of these areas, thel the pressura cask could damage the torus, causing it to drain. The  !

suppression function of the primary containment cou

scram decay heat would have to be removed.would not be aff makeup would not be required immediately. i be damaged, for example, one set of containment spray pumps anl:

ment spray heat exchanger. The ,

any event since GPUN has assumed no water would be present in the torus. '

isolation condenser system would be available to provide long-term h removal from the reactor vessel. If needed, a j could be accomplished remotely by using' condensate transfer.

reactoe building entry to establish shell-side makeup could be per approximately I hour.

be safely shut down following a drop of the cask and that the offsite j consequences of a load drop are bounded by hign-energy line break evaluation The licensee determined that releases resulting from damage to the 52 fuel  !

assemblies in the cask would not exceed 25 percent of the limits set out in  !

10 CFR Part 100 because the fuel assemblies will be more than 10 years old.

GPUN's 10 CFR 50.59 evaluation concludes that no unreviewed safety questions are involved, that movement of the casks can be accomplished in a safe manner because of GPUN's reduction of the probability of dropping the load, and that all license requirements would be satisfied. GPUN based this conclusion on its completion of the Phase I guidelines (Section The 5.1.1 staff states of NUREG-0612) in Gl. 85-11 for the control of heavy loads at nuclear power plants.

that "our review has indicated that satisfaction of the Phase I guidelines This assures that the potential for a load drop is extremely small ."

conclusion is further based on GPUN's evaluation that (1) the fixed li provide redundant load support for the transfer cask, equivalent to a single-failure-proof crane for nearly the entire travel path; (2) safe shutdown can be= achieved where the fixed link support system does not provide protection; and (3) although a' postulated load dropThe could damage safety-related equipmen licensee also noted that the the probability of a drop is extremely ow.

only load drop previously evaluated in the plant safety analysis report (SAR) is the drop of ~a 100-ton fuel shipping cask in the vicinity of the fuel pool.

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NRCB 96-02 I April 11,.1996 Page 5 of 9 l .Diseussio- ,

In 10 CFF 30.59(a)(1), it is stated that "the holder of a license authorizing operation :f a production or utilization facility may (i) make changes in the facility is described in the safety analysis report, (ii) make changes in the procedures as described in the safety analysis report, and (iii) conduct tests 3

or exper ents not described in the safety analys1s report, without prior-j Commissi: approval, unless the proposed change, test _or experiment involves a change ir :ne technical specifications incorporated in the license or an unreviewe: safety question." Section 50.59(a)(2) states that "a proposed

- change, :sst, or experiment shall be deemed to involve an unreviewed safety ,

question f) if the probability of occurrence or the consequences of an '

accident :- malfunction of equipment important to safety previously evaluated in the sa#ety analysis report may be increased; or (ii) if a possibility for l an accide-t or malfunction of a dif ferent type than any evaluated previously in the sa# tty analysis report may be created; or (iii) if the margin of safety j

as define: in the basis for any technical specification is reduced."

The NRC s:aff audited both the initial and updated 10 CFR 50.59 evaluations performec cy the licensee and determined that the propose the NRC f: review and approval pursuant to the requirements of 10 CFR 50.59 and 50.90. The staff based its determination on the fact that, as noted by the licensie, the activity involves movement of loads heavier than those previousl analyzed in the FSAR (except over the cask drop protection system in the fusi pool, where a 100-ton cask drop had been previously analyzed).

This dete .ination is also based on the fact that the load drop had not been previousi evaluated along the remainder of the load path, and on the possibili:/ that a load drop in the reactor building while the reactor is at power cou'd result in consequences that are greater than those previously Therefore, although the licensee had reduced the <

postulate: in the FSAR.

probabili:/ of dropping the cask, the staff was concerned Accordingly, that a load drop as could resdt in an increase in the' potential consequences.

' defined in 10 CFR 50.59(c), if an activity is found to involve an unreviewed safety question, an application for a license amendment must be filed with the Commission pursuant to 10 CFR 50.90.

dased on ne NRC staff's audit of GPUN's 10 CFR 50.59 evaluation, the staff is concerned that other licensees may believe that their heavy load operations are in co711ance with the regulations because they have completed phase I o 22, 1980, and the closecut of Phase II by the generic letter of DecemberGL 85-11 did not relieve licensees of their resi GL 85-11.

~ 10 CFR 50.59 to evaluate new activities with respect to the SAR and the l Technical Specifications to determine whether the activity involves an In l unreviewed safety question or a change in the risks Technical associated Specifications.

with' damage to safety-I

  • .,e addition, GL 85-11 concluded taat I related systems are relatively sma,i because (1) nearly all load paths avoid this equipment, (2) most equipment is protected by an intervening floor, l (3) there is redundancy of components, and (4) crane failure probabil ,

, generally independent of safety-related systems. Creek's preco  ;

d hRCB 95-02

April 11. 1996 Page 5 of 9 Therefore, the staff has concluded that although some licensees have under-taken efforts to further reduce the probabilityEof an accident

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I, ifinvolving the loads heavy

_ loads beyond that previously accepted for NUREG-0612, Phase are heavier and the load paths and potential consequences.of a load drop are

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i different than those previcusly considered.in the FSAR, the probability of an occurrence or the consequences of an accident may be increased.

j Reauested Actions To ensure that the handling of heavy loads is performed safely and within the conditions and requirements specified under Title 10 of the Code of Federal

)

Regulations, all- addressees are requested to take the following actions:

  • Review plans and capabilities for handling heavy loads while the reactor

" is at power (in all modes other than cold shutdown, refueling,Determine and defueled) in accordance with existing regulatory guidelines.

whether the activities are within the licensing basis and, if necessary, 1

j submit a license amendment request. Determine whether changes to the handling Technical Specifications will be required in order to allo,4 of heavy loads (e.g., the dry storage canister shield plug .nd associated lifting devices) over fuel assemblies in the spent fuel cool. .

l Reauired Resconse Pursuant to Section 182a, the Atomic Energy Act of 1954, as amended, and 10 CFR 50.54(f), all addressees must submit the following written information:

For licensees planning to implement activities involving the handling of (1) l heavy loads over spent fuel, fuel in the reactor core,

' the following:

  • A report, within 30 days of the date of this bulletin, that i

i addresses the licensee's review of its plans and capabilities to handle heavy loads while the reactor is at power (in all modes other than cold shutdown, refueling, and defueled) in accordance with existing regulatory guidelines. The report should also indicate -

whether the activities are within the licensing basis and should include, if necessary, a schedule for submission of a licenseAd amendment request.

changes to Technical Specifications will be required.

For licensees planning to perform activities involving the handling of (2) heavy loads over spent fuel, fuel in the reactor core, or safety-related j

equipment while the reactor is at power (in all modes accident that-has not previously been evaluated in the FSAR, submit a license amendment request in advance (6-9 months) of the planned movement i- of the loads so as to afford the staff sufficient time to perform an appropriate review.

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, NRC3 96-02 April 11, 1996 Page 7 of 9 dry storage casks over spent . fuel, fuel in

~(3) For licensees planning to movethe reactor core, or; safety-related equipm power (in all modes other than cold shutdown, refueling, and defueled)  ;

include in item 2 above, a statement of the capability of performing the actions necessary for safe shutdown in the presence of radiological source term that may result from a breach of the dry storage cask, damage to the fuel, and damage to safety-related equipment as a result. of a load l drop'inside the facility. I For licensees planning to perform activities involving the handling of i (4) heavy loads over. spent fuel, fuel in the reactor cora, or safety-related  !

equipment while the reactor is at power (in all modes other than cold l shutdown, refueling, and defuelad), determine whether changes to  !

Technical Specifications will be required in order to allow the handling of heavy loads (e.g., the dry storage canister shield plug) over fuel assemblies in the spent fuel pool and submit the appropriate information in advance (6-9 months) of the' planned movement of the loads for NRC j review and approval.

Address the required written report (s) to the U.S. Nuclear Rt.gulatory i Commission, ATTN: Document Control Desk, Washington. 0.C. 2J555-0001, under l cath or affirmation under the provisions In of Sectionsubmit addition, 182a,aAtomic copy ofEnergy the Act of '

1954, as amended, and 10 CFR 50.54(f).

report to the appropriate regional administrator.

f Related Generic Communications NUREG-0612 " Control of Heavy Loads at Power Plants," Resolution of Generic Technical Activity A-36, issued in July 1980 Unnumbered generic letter dated December 22,1980, " Control of Heavy l

  • )

Loads" i l

Power Plants NUREG-0612," June 28,1985 Backfit Discussion This bulletin is an information request made pursuant to 10 CFR 50.54(f). '

f The objective of the actions requested in this bulletin i I i

with respect to the proper handling and control of heavy loads at nuclear

[~

power plants when the plant is operating (in all modes other th down, refueling, and defueled).the basis of the need.to ensure complia with respect to the weight of the navy loads being moved over spent fuel, over fuel in the reactor core, er over safety-related equipment, and the potentially. severe consequences that can result if a load is dropped.

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NRCB 96-02 April 11, 1996 Page 3 of 9 i 4

Pacerwork Reduction Act Statement This bulletin contains information collections that are subject to the f

l Paperwork Reduction Act of 1995 (44 U.S.C. 3501 et seq.). These information collections were approved by the Office of Management and Budget (OMB),

approval number 3150-0012, which expires June 30, 1997.

The public reporting burden for this collection of information is estimated to

[ average 600 hours0.00694 days <br />0.167 hours <br />9.920635e-4 weeks <br />2.283e-4 months <br /> per response, including the time for reviewing instructions.

searching existing data sources, gatnering and maintaining the The.NRC data needed, is and completing and reviewing the collection of information.

seeking public comment on the potential impact of the collection of infor-mation contained in the generic bulletin and on the following issues:

(1) is the proposed-collection of information necessary for the proper performance of the functions of the NRC, including whether the information will have practical utility?

(2) is the estimate of burden accurate?

(3)

Is-there a way to enhance the quality, utility, and clarity of the

.information to be collected?

f How can the burden of the collection of information be minimized,

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3 including the use of automated collection techniques?

f i Send comments on any aspect of this collection of information, including suggestions for reducing the burden, to the Information and Records Manage- '

ment Branch (T-6 F33), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by Internet electronic mail at bjs10nrc. gov; and to the Desk a

Officer, Office of Information and Regulatory Affairs, NE08-10202 (3150-0012),

j Office of Management and Budget, Washington, DC 20503. '

The NRC may not conduct or sponsor, and a person is not required to respond l to, a collection of information unless it displays a currently valid OMB e

control number.

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. NRCB 96-02

' April 11, 1996 Page 9 of 9 If you have any questions about this matter, please contact the technical [

cor. tact listed below or the appropriate Office of Nuclear Reactor Regulation l (NRR)< project manager.

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'nr[ik M ru h( 'e ctor ,

' O Division of Reactore Program Management l Office of Nuclear Reactor Regulation  ;

l Technical contact: Brian E. Thomas,,NRR (301) 415-1210 Internet: bet @re . gov i

Lead Project Manager: Kevin A. Connaughton, NRR -

(301)'415-3018 Internet: kac0nrc. gov

Attachment:

List of Recently Issued NRC Bulletins -

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OPU Nuoleer Carpere6en

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Post Office Box 368

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gf Route ' South Forked Rrver.New Jersey 08731-0388 609 971-4000  ;

Writer's Deset Dw Number:

May 13. 199f, 6730-96-2160 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555

Dear Sir:

Subject:

Oyster Creek Nuclear Generating Station Docket No. 50 219 ,

l Response to NRC Bulletin 9642.

On April 11,1996 NRC Bulletin 9642, ' Movement of Heavy Lands over Spent Fuel. Over Fuel l i

in the Reac:or Core, or over Safety - Related Equipment," was issued. The Bulletin contained a 30 day reporting requirement for nuclear power licensees.

After reviewing the referenmd bulletin. GPL' :. i: lear treted and was granted by telecon a 30 day extension to submit the response. Additioi '. time is required to perform the rwresury analyses regarding lifted loads and to develop plans with regard to any license amendmerus which may be required. We expect to submit the response to the subject bulletin by June 10. 1996.

If any additio.ial information is required. please contact Mr. Joseph Andrescavage of my staff at l

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609 971 4862. l l

Very truly yours.

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kof Michael S. Roche Vice President & Director j

oysier Creek 200001 MBR/JFA/gl cc: Administrator, Region i NRC Project Manager NRC Resident inspector

-4603200454- 9605 t 3 n, I PDR ADOCK 05000219 ,ll g G PDR n n , , .~,. c , ... , . _ . . , . . . . .s...,.,,. . .

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A UNITED STATES l

-NUCLEAR REGULAllRY COMMISSION '

1 OfflCE OF NUCLEAR R MCTOR REGULATION WASHINGTON, D.C. 20555 April 30, ]996'.

i NRC INFORMATION NOTICE 96-26: RECENT pro 6LEMS WITH OVERHEAD CRANES i

Addressee 1 i All holders of operating licenses or cons truction permits for nuclear power .l

reactors.

J Purnose .

L lhe U.S. Nuclear Regulatory Commission (hRC) is issuing this infomation notice to alert addressees to recent protless with overhead cranes. It is  ;

expected that recipients will review the information for applicability to l

4 their facilities and consider actions, as appropri::te, to avoid similar i Problems, liowever, suggestions contained in this information notice are not NRC requirements; therefore,'no specific action or written response is 3

required. .

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Descrintion of Circ-etances failure of Overhead Crane Bridge Rail s

At.the Trojan Nuclear Plant on July 7,1975, a section of the reactor building polar crane bridge rail failed. The rail had a crack across the top of the

  • top flange and a piece of the flange had been dispiaced. :The and of one section of the rail had failed through th: plane of the rail joint bar bolts extending up through the top flange. Visiaal and metallographic examination of
  • the failure plane indicated that much of the failure was preexisting. Rust on t- the failure surfaces and " peening" of soma areas indicated that the initial crack could extend back to the plant's construction. ,

L The licensee research of construction ricords determined that a nonconformance ,

report, dated July. 26,1972,' noted that t!.e rails were not' slotted for bolts in accordance with the drawings. The coriective action recommended was to

" burn the slots in the field." The 11centee detemined the cause of the ,

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!- failure to be torsional shear and bending at the stress risers from the flame-cut holes. Flame cutting the slots left tesidual stresses in the material  ;

because of the lack of careful preheating and controlled cooling. Also, sharp ,

- notches, noted in the i in of the flame cttting, concentrated the stresses.  ;

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The' inappropriate use of a cutting torch created an untempered martensitic .

heat-affected zone in the high-carbon ste(1 rail. This zone was especially- l i

sensitive' to hydrogen cracking and subsectent brittle crack propagation. The.

4 crack inoucing and propagating loading we primartly due to bending of the ,

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lH 96-26  !

' April 30, 1996 [

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The licensee l rail head io the outside during episodrs of rail misalignment.  !

hau observed rail misalignment to be a continuing problem that had caused or i contributed to 19 bridge truck wheel bearing failures pver 23 years of ,

i operatfon.  !

The root cause of the failure was the inappropriate useThis of apractice cutting torch to created  ;

enlarge drilled holes to slots in the web of the rail.

an untempered, martensitic, heat-affected zone in the rail material that was ,

sensitive.to hydrogen' cracking and subsequent brittle crack propagatico.  :

Actuation of Overheno Crane Safety Sys:em i i

At the Prairie Island Nuclear Generating Plant on May 13, 1995, while lifting a loaded spent fuel storage cask from the spent fuel pool for transfer to the transport bay, the single-failure-proof overhead crane handinng system l automatically stopped on overload, app-oximately 13 cm (5 inches] from the i The bottom of the cask was above the water high hook point (peak lift point).but approximately 8 cm [3 inches] belov pool. Upon investigation of the event, the licensee, Northern States Power j Company-(NSP) determined that the causs of the event was premature actuation l of the erane overload-sensing system. The setpoint on the overload-sensing  ;

system was set too low. Upon actuatics of the overload-sensing system, l control power is automatically removed from the hoist motor and theSubsequeni:

conventional holding brakes are activated.

May 13, the cask remained in the hoisted position until a safety evaluation l was made that supported bypassing the sensing system and resuming the cask  :

lift.

The lift was resumed about 16 haurs after it was stopped, NSPandinitiated the cask a was placed in the decontamination area of the transport bay. The conclusion of j root-cause analysis to ioentify the caJse of the actuation. i this analysis was that the overload-sensing system was inaccurately 1 calibrated.

This event raises a concern for similarly designed overload-sensing As noted systems in the analysis associated with single-failure-proof cranes.

reports, this event was a "fiuisance trip that resulted from inaccurateImproved load c  ;

initial calibration during load cell setting adjustment. l accuracy can help to reduce any unbalanced loading condition in the system.

Additional deta'ils of these events can be found in inspection Report No. 50-344/95-06 issued on September 18, 1995, and. Inspection Report i No. 50-282/95-06 issued on June 27, 1995.  !

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IN 96-26 '

April 30. 1996 .

Page 3 of 3 [

] If ;

' This _information notice requires no specific action or written response.

you have any questions about the infor.sation in this notice, piease contact I

' one of the technical contacts listed b+1ew.

n l

Brian K. Grimes, Acting Director Disisjon of Reactor Program Management

' Office of Nuclear Reactor Regulation i Technical contacts:

Robert J. Pete Region IV Brsan E. Thomas, NRR '

(Slo) 975-0246 (311) 415-1210 In ternet: bet?nrc. gov Internet:rjp19nrc. gov David B. Pereira, Region IV Eric J. Benner, NRR i (510) 975-0307 (301.) 415-1171 Internet:dbpenrc. gov In ternet:ejbl9nrc. gov ,

i t Russell L. Bywater, Region !!!

(612) 388-8209 Internet:rlb33nrc. gov

Attachment:

List of Recently Issued N tc Information Notices

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Headquarters Daily Report MAY 08, 1996 Consolidated Edison Co. Of'N.Y. MR Number: 1-96-0047

-Indian Point 2 Date: 05/08/96 Buchanan;New York- SRI /RI PC Dockets: 50-247 PWR/W-4-LP .

Subject:

UNIT 1 CONTAINER DROP Discussion:

On May 7, 1996, con Edison was in the process of moving a metal transportation container in the Unit 1, fuel handling floor. The container measured 8 ft. X8 ft. X 20 ft. and weighed approximately 5000 lbs. Four nylon slings were used for the lift of the container. Operators attached the slings to a hook on an overhead crane. The hook had a springAs

~

loaded keeper installed to prevent the container was being lifted, the slings from sliding off.

the lighter end lifted off the floor first and caused the container to rotate. The light end of the container was lifted up approximately 18 inches when two of the slings slipped off the hook, damaging the keeper, and the container dropped to the floor.

All lifting operations on the fuel handling floor have been stopped )

pending review of the event. l i

con Edison has determined that because the slings used were too short for this lift, the angle of the slings from the container to .

' the hook was approximately 24 degrees. Posted guidance was to have a minimum of a 30 degrees angle to accomplish a lift. As a result, as the container rotated during the lift, the slings also rotated until they slipped off the hook. No personnel injuries or other Con Edison is continuing equipment damage resulted from the drop.

to review the causes and correceive actions for this event.

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Oate: 12-30-94 PRELIMINARY NOTIFICATION OF EVENT OR UNUSUAL OCCURRENCE PNO-II-94-055 This preliminary notification constitutes EARLY notice of events of POSSIBLE safety or public interest significanc.e. The information is as initially received without verification or evaluatior., and is basically all that is known by the Region II staff on this date.

FACILITY: Licensee Emergency Classification:

Georgia Power Company Notification of Unusual Event Hatch -1 Alert Baxleyy Georgia Site Area Emergency General Emergency DocketiNo.-- 50-321

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- _X_ Net Applicable (EN 28194 Info Only)

SUBJECT:

HATCH UNIT 1 SPENT FUEL P0OL STEEL LINER PUNCTURED WHEN CORE SHROU BOLT OROPPED 1

On Decemb& 28,1994, at 8:53 p.m., Hatch Unit I tore a three inch diameter gash in the stainless steel liner of their spent fuel pool when a 350 pound core shroud bolt was dropped from one foot above the pool water surface. A steel cable sling failed as the bolt was being removed from the pool for shipment offsite. Seven shroud bolts had been removed from the reactor during the September 1994 refueling outage and stored in the spent fuel pool awaiting shipment offsite. Leakage through the liner gash has been contained in the annulus between the liner and the concrete outer structure of the spent fuel pool. Pool level has been restored via the normal makeup system and the falling bolt did not contact any spent fuel. The licensee is monitoring leakage of l l

approximately 0.7 gpm which is occurring through system penetrations in the concrete structure and is being collected in the reactor building sump. There l

I has been no release of radioactivity offsite. A contingency plug is available to insert in the gash if leakage from the concrete structure increases significantly. Contract divers are expected onsite Friday, December 30 and will i j

assist in removal of the impacted bolt from the liner and installation of a temporary weighted gasket plug. Permanent underwater welding repairs are ~

i expected to commence Monday, January 2, 1995, at the earliest. )

The Seior Resident Inspector responded onsite to monitor the licensee response at 12:15 a.m. December 29 when notified of the occurrence. The resident staff wi'41 continue to monitor licensee activities to repair the damage through the weekend.

The licensee informed the state of Georgia.

The NRC Emergency Response Center received initial notification of this event by telephone from the licensee at 2:33 a.m. (ET) on December 29, 1994.

This infnrmation is current as of 10:0^ a.m. on 12/30/94.

CONTACT: S. J. Cahill - (404) 331-4198

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1996 / Notiuts Federal Register / Vol. 61. No. 90 / Wedne3 day, May 8, ..

20842 .

proposed determination for each Polfstown Public I,ibrary, .iun 18;gh amendment request is shown below.

determination on the isum of no Street, Pottstow n, Pmms31 tama 19M4.

signifit;mt hiuards consideratiori. The The ConunM,on is seeking public gg,,d a Hm kvn WW nat. tha 3rd 4,y toi .ents on this proposed final detenninalb i will serve to detido og my joyg O' termination. Any comtnents received when the hearing is held.

N die Nut le ,ir Regulaton Commiwion with,m 30 days af ter the date of if the final determination is that the -

amendment request mvolves no . d" "" . '

publication of this notico will be significant hazards consideration, the comidered in making any final Cornmission may issue the amendmont Iml"'

"" "" "IN'"'Wf' N""'I"#P'"F

""7" #" Ih"

"" "0"#

'" Id'

" f detennination.

and make it immediately effective, N"*" d" N" Normally, the Commission will not notwithstanding tho request for a hearing. Any hearing held would take IN N' *" U"d #^* # ""d issue the amendment until the expiration of the 30-day notice period.

8*uwa coor noo45-P tiowever, should circumstances change placo after issuance of the amendment.

during the notice period such that If the final detennination is that the failure to act in a timely way would amendment rupiest involves a Biweekly Notice; Applications and Amendments to Facility Operating result, for example, in derating or significant hazards consideration, any hearing held would take place before WensesinvoMng no Ankant shutdow'n of the facility,the tbc issuance of cny amendment. Comruission may issue the license Hazards Considerations amendment before the expiration of the A mquest for a hearing or a petition for leave to intervene must be filed with I,11ackground 30-day notice period, provided that its the Secretary of the Commission, U.S. Pursuant to Public Lave 97--415, the final determination is that the Nuclear Regulatory Commission, amendment involves no significant U.S. Nuclear Regulatory Commission Washington, DC 20555-0001, Atfontion: hazards consideration. The final Dockering and Services pram.h. or may (the Commission or NRC staf0 is determination will consider all public be delivered to the Commisslun's Public publishing this regular hiweekly and notice.Stato comments received before Document Room, the Gelman Building. Public law 97-415 revised section in9 action is taken. Should the Conunission 2120 L Street, NW., Washington, DC, by of the Atomic Energy Act of 1954, as take this action, it will publish in the the above date. Whero petitions are filed amended (the Act), to require the federal Register a notice ofissuance Commission to publish notice of any and provide for opportunity for a during the last to days of tho notico period,it is requested that the petitioner amendments issued, or proposed heneing afterto issuance.

be The Cornmission promptly so inform the Commission by issued, under a now provision of section expects that the need to take this action a toll-free telephone call to Western 189 of the Act. This provision grants the will occur very infrequently.

Commission the authority to issue and Written comments may be submitted Union at 1-(800) 248-5100 (in Missouri make immediately effective any by mail to the Rules Review and 1-(800) 342-6700). Thn Western Union amendment to an operating license Directives Branch, Division of Freedom operator should be given Datagram upon a determination by the of Infonnation and Publications .

identification Number N1023 and the Services, Office of Administration, U.S.

following message addressed to John F- Commission that such amendment involves no significant hazards Nuclear Regulatory Commission, Stolz: petitioner's name and telephone number, dato petition was mailed, plant consideration, notwithstanding the Washington, DC 20555-0001, and name, and publication date and pago pendency before the Commission of a should cite the publication date and number of this Federal Register notice. request for a hearing from any person. page number of this Federal Register A copy of the petition should also be This hiweekly notice includes all notice. Written comments may also be n tices f amendments issued, or delivered to Room GD22, Two White sent to the Office of the General Counsel, U.S. Nuclear Regulatory pmposed to be issued from Aprd 13, Flint North,11545 Rockville Pike, Commission. Washington, DC 20555- Rockville, Maryland from 7:30 a.m. to 1996, through April 26,1996. 'Ihe last

' 0001, and to J.W. Durham, Sr., Esquire, inweeMy n tice was published on April 4:15 p.m. Federal workdays. Copies of Sr. V.P. and General Counsel, 24,1996 (61 FR 18162). written comments received may be Philadelphia Electric Company,2301 examined at the NRC Public Document Market Street, Philadelphia, Notice of Consideration ofIssuance of Room, the Gelman fluilding,2120 L Amendments To Facility Operating Street, NW., Washington, DC. The filing Pennsylvania 19101, attornay for the Licenses, Proposed no Significant licensee. of requests for a hearing and petitions Nontimely filings of petitions for llazards Consideration Determination, for leave to intervene is discussed leave to intervene, amended petitions. and Opportunity for a llearing below.

supplemental p9titions and/or requests By June 7,1996, the licensee may file '

The Commission has made a a request for a hearing with respect to for hearing will not be entertained absent a detonnination by the proposed determination that the issuance of the amendrnent to the following amendment requests involve subject facility operating license and Commission, the presiding officer or the no significant hazards consideration. l any person whose interest may be presiding Atomic Safety and Licensing Under the Commissh 's reguhtions in 4

Board that the petitirn and/or request 10 CFR 50.92, this mus tha; operation affected by this proceeding and who should be granted based upon a wishes to participate as a party in the of the facility in accordance with the balancing of the factors specified in 10 proceeding must file a written request proposed amendment would not (1) for a hearing and a petition for leave to CFR 2.714(a)(1)(i)-(v) and 2.714(d). involve a significant increase in the For further details with respect to this intervene. Requests for a hearing and a action, see the application for probability or conwquences of an petition for leave to intervene shall be amendment dated April 25,1996, which accident previously evaluated; or filed (2) in accordance with the create ao possibility of a new or is available for public inspection at the different kind of accident from any Commission's " Rules of Practice for Commission's Public Document Room, Domestic Licensing Proceedings" in 10 accident previously evaluated; or (3) CFR Part 2. Interested persons should the Gelman Hullding,2120 L Street, involve a significant reduction in a NW., Washington, DC, and at the local consuh a current copy of to CFR 2.714 margin of safety. The basis for this public document room located et the

J ,

1 20843 Federal Register / Vol. 61, No. 90 / Wednesday, May 8,1990 _ / Notices Nontimely filings of petitions for must provide sufficient information to .,

which is available at the Commission's show that a genuine dispute exists with leave to intervene, amended petitions, ,;

Public Document Room, the Gelman supplemental petitions and/or requests the applicant on a material issue of law for a hearing will not be entertained ilullding,2120 L Street, NW., n Wcshington, DC and at ' e local public or fact. Contentions shall be limited to absent a determination by the matters within the scope of the document room for the particular Commission, the presiding officer or the facility involved. If a request for a amendment under consideration. The Atomic Safety and 1.icensing Board that contention must he one which, if hearing or petition for leave to interverm proven, would entitle the petitioner the petition to and/or request should be h filed by the above date, the relief. A petilloner who fails to file such granted based upon a balancing of Commission or an Atomic Safety and factors specified in 10 CFR 1.icensing Board, designated by the a supplement which satisfies these Commission or by the Chairman of the requirements with respect to at jeast one c or2.714(a)(1)(ij-(v) further details withand 2.714(d).

respect to this contention will not be permitted to Atomic Safety and Licensing floard action, see the application for panel, will rule on the request and/or participate as a party. 3 Those permitted to intervene become amendment which is available for petition; and the Secretary or the designuted Atomic Safety and Licensing parties to the proceeding, subject to any publ'ic limitations in the order granting leave to Public Document Roem, the Gelman inspec C

Board will issue a notice of a hearing or intervene, and have the opportunity to Ilullding,2120 L Street, NW.,

a, appropriate order, participate fully in the conduct of the Washington, DC, and at the local public As required by 10 CFR 2.714, a hearing, including the opportunity to document room for the particular '..

patition for leave to intervene shall set facility involved.

forth with particularity the interest of present evidence and cross-examino t witnesses. Baltimore 60s and Elecin. c Co pany, the petitioner in the proceeding, and If a hearing is requested, the how that interest may be affected by the Commission will make a final Docket Nos. 5&317 and 543 8. Calvert results of the proceeding.The petitmn Cliffs Nuclear Power Plont, Ulit Nos. I should specifically explain the reasons determination on the issue of no and 2, Calvert County, Mary and ,

significant hazards consideration. The why intervention should be permitted Date of omendments reg est: March with particular reference to the final determination will serve to decide following facton (1) the nature of the when the hearing is held. 28,1996.

Description of amendm nts request: [

if the final detennination is that the petitioner's right under the Act to lm amendment request involves no Pursuant to 10 CFR 50.90 the Haltimore

mide a party to the proceeding
(2) the Gas and Electric Compar (BGE) hereby significant hazards consideration, the nature and extent of the petitioner's Commission may issue the amendment requests ca amendment o Operating ,

property, financial, or other interest in and make it immediately effective. License Nos. DPR-53 a d DPR-69 to the proceeding; and (3) the possible notwithstanding the request for a reduce the moderator t mperaturo effect of any order which may be hearing. Any hearing held would take coefficient (MTC)limi shown on i entered in the pmceeding on the Technical Specificati i Figure 3.1.1-1.

place after !ssuance of the amendment. y petitioner's interest. The petition should .If the finaldetermination is that *th" This proposed chang is necessary to also identify the specific aspect (s) of the ammuument request involves a sup ort changes in t e safety analyses subject matter of the proceeding as to significant hazards consideration, any ma to accommod, e a larger numtmr which petitioner wishes to intervene. hearing held would take place before of plugged steam ge crator (SG) tubes Any person who has filed a petition for the issuance of any amendment. .

for future operatin cycles.The leave to intervene or who has been A request for a hearing or a petition proposed limit wi be more restrictive admitted as a party may amend the for leave to interveno must bo, filed with than the existing I mit to match the petition without requesting leave of the analytical assum ions. In addition, the (

the Secretary of the Commission, U.S. J lloard up to 15 days prior to the first Nuclear Regulatory Commission' licensee provide information to clarify prehearing conference scheduled in the Washington, DC 20555-0001, Attention: the relationship f the MTC to an proceeding, but such an amended Docketing and Services Branch.,or may Anticipated Tra sient Without scram i petition must satisfy the specificity be dehvered to the Commission s Public event in its lice sin basis.

requirements described above. Document Room, the Gelman Buildmg, Basisforpro os no significant l Not later than 15 days prior to the first 2120 L Street, NW., Washington DC, by 3 hazards consic emtion determination:

prehearing conference scheduled in the the above date. Where petitions are filed As required b to CFR 50.91(a), the d t-proceeding, a petitioner shall file a during the last to days of the notice licensee has ovided its analysis of the sur plement to the petition to intervene /

period,it IP requested that the petitioner issue of no si ificant hazards which must include a list of the promptly so inform the Commission by consideratio , which is presented contentions which are sought to be yl a toll free telephone call to Western below; 4 litigited in the matter Each contention must consist of a specific statement of Union at 1-(800) 248-5100 (in Missouri 1. W uld n t involve a significant increase 1-(800) 342-6700).The Western Union mba or the issue of law or fact to be raised or ",Ihe Qujnces of an ,

controverted. In addition, the petitioner operator should be given Datagram g , 3, Identification Number N102' and th" The safety analyses for the current fuel i

mil provide a brief explanation of the bases of the contention and a concise statement of the alleged facts or expert following message addresse tn (Project cycles assu e 500 tubes per steam generator Director): petitioner's name and ISG) are ph ;ged and the maximum beginning- f-cycle moderator temperamre f

y tele hone number,date petition was y' opinion which support the contention cocmcient MTC)is assumed to follow the and on which the petitioner intends to mai ed, plant name, and publication curve in T chnical Specifiution Figure p date and page number of this Federal 3.1.1..t. F ir the fuel cycle to be installed in 4 rely in proving the. contention at the >

hearing. The petitioner must also Register notice. A copy of the petition 3 81

  • I""j"[*8d",d provide references to those a meific should also le sent to the Office of the $"$1 y General Counsel, U.S. Nuclear analyse that more SG tubes are plugged than sources and documents of w tich the :5e curr nt limit, and it is necessary to credit i petitloner is aware and on which the Regulatory Commission Washington. j DC 20555-0001, and to the attorney for a more festrictive (less positive) limit on the petit,ioner intends to rely to establish maximum positive MTC to mitigate the those facts or expert opinion. Petitioner the licensee.

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Federal Register / Vol. 61, No. 90 / Wednesday, May 8,1996 / Notices - -

20848 -

referent.ss in Technical Specifications as credible event. Therefore. it does not inrn use used in determining compliance with the Testing of 1 D(Ps during power operation will not afhu.t th<s ,ivailabihty or operation of the probabihty of or e onsequences regula* oflimits.an \

a< r.ident. The references, as pmposnt to tm included any offsite mun e of pown. In addiHon. the

2. State the basis for the determination that in section 5.14 of the Technicai ElX; being testcd remains capabl<e of meeting the activity does or does not create the its intended design functiont Therefore the Specifications, have previously (men ,

possibdity of an acadent or malfunction of reviewed and approved by the NRC for proponed change to Ihe Technical a different ty[m than any previously generic applicability to PWRs (Pressurized Jperification Surveillance Rerguirement identified in the SAR lsafety analysis reporti, Water Reactors]. The reports identified in the

I 8.1.14 will not result in a reituction in a This activity will not create the possibility Proposed Change have been accepted by the margm of safety. of a new or differont type of accident than NRC for referencing in plant hcensing I The NHC staff has reviewed the previously evaluated in the SAR because the applications. , , .

licensee s analysis and, based on ibis pnipmed heavy load handling exception Since the references Usted in the Proposed 3

review,it appears that the three does not create a new credible accident Change have previously been found to meet scenario. Dropping the shield plug on a the conditions of to CFR 50.46 and to CFR -

standards of to CFR 50.92(c) are loaded DSC and damaging s ent fuel APPemlix K, and that the plant specific satisfied. Therefore, the NRC staff proposes to determine that 'he assemblies is not considere[a safety credible analysisevent.

anged i acceptance limits have not nm f hu s thew

3. State the basis for the detennination that g amendment request involve
  • no the margin of safety is not reduced, i significant hazarda consideration, This activity will not involve a significant ih h N W b i nr n 'Y " consistent with prior plant speci_fic and Local Public Docurnent Hoom n i s industry requirements and practices.

a e Location: Judge Coorge W. Armstrong el puc Therefore, we have concluded that the.

Library,220 S. Commerce Street, not create a credible accident scenario.

Proposed Change will not result in a 1 Natchez, MS 39120.

4 The NRC staff has reviewed the significant increase in the probability or '-

Attorneyfor licensee: Nicholas S. licensee's analysis and, based on this consequences of an accident previour,1y Reynolds. Esquire, Winston and Strawn, review,it appears that the three "Vah ted.

1400 L Street, N.W.,12th Floor, standards of 10 CFR 50.92(c) are 2.

Washingtmy DC 20005-3502. satisfied. 'Iherefore, the NRC staff the possibility for a new or different kind of NHC Pmject Director: Willinm n. proposes to determme that the omdent?

The Proposed Changes introduce no new 11cckner. amendment request involvea no mode of plant operation; do not involve the CPU Nuclear Corporation, et al., Docket significant hazards consideralmn. physical modification of any structure.

No. SN19, Oyster Creek Nuclear local Public Document Room system or component; do not effect the Generatmg Station, Ocean County, New location: Ocean County Library, function, operation or survei!!ance for any Reference Department,101 Washington equipment necessary for safe opuadon or l##8#f Street. Toms River, NJ 08753. shutdown of the plant; and, do not involve Date of amendment request: April 15, Attorneyfor licensee: Ernest L. Blake, "8 P

[d Changes 1996 (TSCR No. 244). Jr., Esquire. Shaw, Pittman, Potts & hring Iers'"TYPr Description of amendment request: are administrutive in nature only.

The proposed amendment would revise Trowbridge,2300 N Street, NW., Therefore, we have concluded that the Specification 5.3.1.D of the Oyster Creek Washington, DC 20037. Proposed Change cannot result in the Technical Specifications. The current NRC Pmiect Dimctor: John F. Stolz. possibility of a new or different kind of M m. e YanAN AfoNC {owerComPany, 3.accident from that previously evaluated.

specification prohlhits handling a load Does the Pmposed Amendment involve greater in weight than one fuel assembly Docket No. 5N09, Mome Yankee a si nificant reduction in a margin of safety?

over irradiated fnel in the spent fuel Atomic Power Station, Lmcoln County, T e Pmposed Changes am a&nMstraum storage facility. The proposed change M ine in nature, consistent with the guidance of will facilitate the off load of spent fuel Date of amendment request: April 19. Generic Letter 88-12. and have been to the Oyster Creek Independent Spent reviewed pmviously by the NRC and found 1996. acceptable with regard to the requirements of Fuel Storage Installation (ISFSI). Description of amendment request: to CFR 50.46 and to CFR Appendix K.

Specifically, the shield plug for the dry The pmposed amendment would revise Additionally, the plant specitic safety shield canister (DSC) and the associated Technical Specification 5.14 to add the analysis acceptance criteria has not changed lifting hardware will be moved over appropriate references identifying the from that used in the latest core reload irradiated fuel which is contained in the detailed methodology and conditions analysis.

DSC within the transfer cask located in for analyzing the Small Break Loss-of_ weha the Cask Drop Protection System p[, yf g ;econjudo<l 9 hatthe Coolant Accident (SBLOCA) to the list significant reduction in a margin of safety.

(CDPS,). of the approved Core Operatm, g Limits Rasts for proposed no significant Report methods. The NRC staff has reviewed the hazards considemtion determination: Basis forpmposed no significant licensee s analysts and, based on this As required by to CFR 50.91(a), the hazards considemtion detennination: review,it appears that the three licensee has provided its analysis of the As required by 10 CFR Sa 91(a), the standards of 10 CFR 50.92(c) are issue of no significant hazards licensee has provided its nalysis of the satisfied. Therefore, the N'tC staff cons'deration, which is presented issue of no significant hazards consideration, which is presented PmPoses to determine that the below: amendment request involves no

1. State the basis for the determination that bel *.* significant hazards consideration.

the proposed activity will or will not increase 1. Does the Proposed Amendment involve Local Public Document Room the probability of occurrence or a significant increase m the probabihty or location Winswt Public UM Hi&

. consequences of an accident. consequences of an accident previously Street

  • P.O. Box 367* Wiscasset' MF*

'I he es n features and capacity of the evalua:ed? 04578.

reactor but ding crane provide a significant These Proposed Changes are administrative Attorneyforlicensee: Mary Ann safety factor, in addition personnel training in nature and are consistent with the Lynch, Esquire, Maine Yankee Atomic and other udministrative controls further guidar:ce set forth in the NRC Generic Letter Power Company,329 Bath Road, reduce risk.Thus, the dropping of the DSC 88-16 identifying the requirements for the shield plug onto a loaded DSC and causing Brunswick, ME 04011.

inclusion of analytical methodology damage to the spimt fuel assembhes is not a 3

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