ML20023D887

Forwards Evaluations Submitted Re NUREG-0612, Control of Heavy Loads at Nuclear Power Plants. Handling Shielded Shipping Casks Over Spent Fuel Pool Will Be Prohibited Until Measures Taken to Reduce Probability of Cask Drop

ML20023D887
Person / Time
Site:	FitzPatrick
Issue date:	06/02/1983
From:	Bayne J POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
To:	Vassallo D Office of Nuclear Reactor Regulation
References
REF-GTECI-A-36, REF-GTECI-SF, RTR-NUREG-0612, RTR-NUREG-612, TASK-A-36, TASK-OR JPN-83-49, NUDOCS 8306060121
Download: ML20023D887 (14)

Text

4 123 Man Street W5r e Plains, New York 10601 914 681 6240 4 NewYorkPower a. esiii'a ar-1# Authority f/J TeO*"'

June 2, 1983 JPN-83-49 Director of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Attention: Mr. Domenic B. Vassallo, Chief Operating Reactors Branch No. 2 Division of Licensing

Subject:

James A. FitzPatrick Nuclear Power Plant Docket No. 50-333 Control of Heavy Loads

References:

1. NRC letter, D.G. Eisenhut to All Operating Reactors, dated December 22, 1980.

2. " Control of Heavy Loads at Nuclear Power Plants," NUREG-0612, dated July 1980.

3. PASNY letter, J.P. Bayne to T.A. Ippolito, dated October 15, 1981 (JPN-81-82).

4. PASNY letter, J.P. Bayne to D. B. Vassallo, dated February 26, 1982 (JPN-82-25).

5. " Control of Heavy Loads," draft Technical Evaluation Report, Franklin Research Center, dated March 25, 1982.

6. Federal Register, Vol. 48, No. 50, dated March 14, 1983, pp. 10772 - 10776.

Dear Sir Ref erence 1 requested that the Autnority review heavy load-handling operations at FitzPatrick and required a two-phase submittal of evaluations of their conformance to the guidelines of NUREG-0612 (Reference 2).

The Authority completed the first phase of the review and submitted an evaluation to the NRC in October 1981 (Reference 3).

In February 1982 the Authority submitted its evaluation for the second phase of the review (Reference 4). This evaluation indicated that the postulated consequences of certain load drops would not, or might not, meet the guidelines of NUREG-0612. Hence, the Authority prohibited d

8306060121 830602 /$08[HE gonApocn05000g

i handling of these loads until a further evaluation could be conducted demonstrating that the likelihood of such drops is acceptably small or, alternatively, that the postulated consequences of such drops are acceptable.

This evaluation has been completed. A summary of its results is included as an attachment to this letter.

In March 1982, Franklin Research Center, under contract to the NRC, completed a draft Technical Evaluation Report of

. the Authority's phase 1 submittal (Reference 5). This report identified a number of items which required further analysis or protective measures. The Authority discussed these items with the NRC and Franklin Research Center in a telephone conference on October 7, 1982. In that  ;

conference call, the NRC requested that a response be provided that would document certain agreements reached during the call. That response will be submitted by June 30, 1983.

As noted in the attachment, the evaluation of postulated drops of the reactor vessel head, steam separator -

assembly, shipping casks, or recirculation pump motor indicate that the probability of such drops, following the initial lift and hold of these loads, is below or comparable to the NRC's current core melt " safety goal" probability of 1.0 X 10 4 per reactor year ( Reference

6). With-the exception of shipping casks, the Authority considers the calculated probabilities of drops of these

' . loads to be suf ficiently low as to preclude the need for analysis or protective measures beyond those discussed in the attachment. Hence, the load handling restrictions imposed by the Authority in Reference 4 for the reactor vessel head,_ steam separator assembly and recirculation pump motor have been removed.

The Authority will prohibit the handling of shielded shipping casks over the spent fuel pool until measures are taken either to further reduce the probability of a cask l

i drop or to acceptably minimize the consequences of a drop.

If you have any questions regarding this letter or the attachment, please contact Mr. J.A. Gray, Jr. of my staf f.

l Very truly yours, l

l. f r / \

l (QJj P Ba)yne x

l

' Exacutive Vice President I Nuclear Generation l

cc: Mr. J. Linville Resident Inspector U.S. Nuclear Regulatory Commission P.O. Box 136 l Lycoming, NY 13093 I

I.-- . . . _ . - __. - _

e e ATTACHMENT TO JPN-83-49 NEW YORK POWER AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT EVALUATION OF HEAVY LOAD HANDLING GPERATIONS l

i i

l t

l l

1 t

l l

1

PABLS Oe' COM PSN P3 INPRODUCTION 1 RSAC POR VESSSL HSAD, S PEAM SCPARAPOR ASSSMBLY 2 AND RECIRCULAPION PUMP MOTOR LIFPS SilIPPING CASK LIPUS 6 MEAN PROBABILIPIES OF DROPS OF 8 R8ACPOR VBSSSL HEAD, S PBAM SSPARAPOR ASSEMBLY AND RECIRCULAPION PUMP MOP 0R MSAN PROBASILITIES OF 9 SHIPPING CASK DROP RdFERdNCSS LO l

l l

l l

l

IMPRODUCULON -

1 In our February 26, 1982 submittal (Reference 1), the Authority _

stated that some lifts of heavy loads at the James A. FitzPatrick i Nuclear Power Plant (JAP) required additional analyses to demonstrate compliance with NUR83-0612 criteria. Interim measures prohibiting lifts of the reactor vessel nead, steam separator assembly, recirculation pump motor and shipping casks, as identified in our letter, were applied until these evaluations could be completed. This. report supplements our prior response and documents the results of our consultant's evaluations.

The evaluations involved control of heavy loads during refueling and maintenance activities. The evaluations involving ref ueling activities addressed the potential drop of the reactor vessel head or the steam separator assembly. The other evaluations, invciving maintenance activities, addressed potential drops of the recirculation pump motor or shipping casks as they would be. carried by the Reactor Building Crane across the operating floor and down l

l the southeast equipment hatch. All of these loads are handled of the Reactor Building Crane, which was evaluated against industry design standards for such cranes and lifting devices and found to be in compliance (Reference 2). Additionally, the procedural requirements of NUREG-0612 for operator training and qualification and for crane inspection, testing and maintenance have been tot for handling of all loads. .

i Page 1 of 10 l

e -, - - - - . . _ _ y . -_----.-- ,,, , .g c , "-

__ . ._. _. _ _ . . - _ _ _ _ _ _ . _ _ _ _ . _ _ _ _ _ . _ _ _ _ _ _ _ _ _ . _ .__m.. . _ . - - _ _ _ _ _ .

As descri ed belos, the' evaluations of tne nandling activities 1

associated with the reactor vessel head, steam separator assembly I and recirculation pamp motor have demonstrated that the li ke l i hoo.1 of dropping these Lords is extremely small. Consequently, these i

Loa:1 drop scenarios .lo not require additional analysis or protective sne2sures. Therefora, the Autnority nas removed interim restrictions )

Ton the handling of tne reactor vessel nead, steam separator assembly m

1 and, recirculation pamp motor.

\

Tne Authority will prohibit the handling of shielded shipping casks over-the-spent fuel pool until measures are taken either to further reduce the probability of a cask drop or to acceptacly minimize the .

l consequences of a drop'. .

4 i

I Reactor Vessel Head, Steam Separator and Recirculation Pump Motor 4

i Lifta ,

Lifts of the reactor vessel head, steam separator assembly and recirculation pam'p motor by the Reactor Building Crane nave been

~

i l analyzed on a probabilistic casis. The study identified and quantitatively analyzed, using fault tree methods, tne potential

. - < i

' mechanisms for , drops o f t hes'e loads. -

The study was performed in i s l 'e, acebr0ance with the following steps:

i s .

s

, t

+

e Review of the Reactor, Building Crane system and associated t,

T testing, maintenance, inspection, training and lift lN[

1 i procedures for r,ymoval and installation of the reactor .

t n

,.,; \ ' head, steam separator and recirculation pump motor.

., e v -

s e .1< s I $' '

, 'Page 2 of 10 l

L,s ~, .

s s I~ , ,,

5

\

v-. - p- ,

f'l

,s

• 1  ? ~

e Event identification and fault tree -

construction-- ,

, c +. .. . e ,

determinati9n of all the ways that thel Reactor Building i

• I Crano system,could fail, including:

I t.

,j

+t '. \ .

?,l-2 9

( 1) .Stcuctural.,f,ailure whil.e a.iojected toinormal Load

't ,

I e- .  ? tondit;i'ons; ' ,

w .

,r i l . 's (2) "

Structucal failure due to excessive load; f.

fg ',

UJ \ y d

i) two-blocking eyent

.c ii) load hangup event

,e .'

t 4

-/ ,

> s

( 3 )' Overspeed event--loss of. hoisting or lowering 4

\ ,-

capability coupled with loss of brakes.

41 t t

?* F y +

. ,+, <

e Qualitystive analysis--find system f ailure modes and

?

(i , r estabi*ish all single failure events leading to system

,4 i ' l failura. l

\> %

e Probabilistic analysis

f, .

s.

a \J,<

g x

~'

i (1) Find sources of dati and determine applicability to JAF

~) 1 f

3 load-handling operations; ,

,,\ ,

.st.s

.,s' }

> 1 - %( 4 . r

';- (2) Compute prooability of the undesired load dr.op event;

.> k /

a c

. st ,

,o *

-.)b4

) l ,

( , '*f i , '(3)

Probabilistically rank basic events and system failure L ' s,t- r< ,,

. f"D,/ modes (i.e., conduct a sensitivity analysis);

i s

• _j/ ( .; -h *

^ /

'! Page 3 of 10  !

% ,, ,) *'

_. s'I

o Develop conclusions, recommendations and results.

The undesired load drop event for the analysis was defined in terms of two individual events:

e Drop during removal e Drop during installation These two events generate identical load drop scenarios, with two exceptions:

e During installation, a two-blocking event would most likely occur above the reactor head laydown area. Hence, this scenario is not considered during installation.

e A reactor head or steam separator load hangup event could only occur during removal. Again, this scenario is not considered during installation.

During removal operations, the reactor head and steam separator are initially lifted only several inches. The lifting rigs are then visually inspected before further lifting. To account for these operations, the analysis was segregated into two types of potential load drops:

o Drop during initial lift o Drop after initial lift Page 4 of 10

Table I summarizes the mean proabilities of a load drop during lifts of the reactor vessel head, steam separator assembly and recirculation pump motor. The results indicate that the dominant failure mechanisms for tuese lifts (excluding the recirculation pamp motor, where the dominant failure mecnanism is relate.1 to overspeed events) are those related to tne occurrence of a todd drop during the initial lift a nd ho ld . Because the initial lift neight is limited to less tnan la inches, the postulated consequences of a load drop at this stage of a lift were found to comply with the evaluation critaria of NOREG-06L2 (i.e., reactor vessel integrity is maintained and no fuel damage will occur). Tne mean procaoilities of a load drop subsequent to the initial lift and hold are shown in Table 1. Considering the inherent conservatism of the model used to calculate these probabilities, these mean values are themselves considered to be conservative.

In addition, the ACRS (Reference 3) and the NRC staff (Reference 4) have recently discussed quantified safety goals in an effort to establish a preliminary total prooanility for a reactor core melt.

Those discussions led to puolication of a preliminary core melt

-4

safety goal probability of 1.0 X 10 per radctor year (Reference

! 5). It should be empnasized tnat the probabilities listed in Taoles l

l I and II are for a load drop onif. Tne consequences of any load drop accident would be considerably less severe than those expected I from a core melt accident.

l l

l l

Page 5 of 10 1

I In summarf, the dominant tailure mode for lifts of tne reactor )

, i vessel head and steam separator assembly were found to occur during l the initial lift and hold stage. The consequences of drops at this stage were found to comply with NURdG-0612 evaluation criteria. For lifts of the reactor vessel head, steam separator assembly and recirculation pump motor, the probabiliry of failure suosequent to the initial Lift and nold stage was determined to be sufficiently

. small as to preclude the need for further analysis.

Shipping Cask Lifts For shipping cask lifts, the reliability analysis described above applies also. In this case, the probabilities of failure are shown in Table 2. Using both systems and structural analyses, the Authority 's consultant has evaluated lifts of the various shipping and spent fuel casks identified for possible frequent use. Based on the evaluations, it has been determined that, by restricting the size of casks to about 35 tons and the lifting height to about 6 l inches, the postulated consequences of load drops onto the refueling deck at Elevation 369' comply with NUREG-0612 evaluation criteria.

! That is, while some local structural damage may occur (e.g. concrete scabbing), no gross structural f ailures are expected. The systems analysis also indicated that safe shutdown capability and core

cooling would still be maintained.

i i Page 6 of 10 I

=

In view of the probabilities of a snipping cask d rop over the spent fuel pool, the Authority will prohinit movement of shipping casks over the pool until additional measures are taken to further reduce the ptobability of a cask drop or to acceptacly minimize the consaquences of suen a drop.

Page 7 of 10

TABLE 1 MEAN PROBABILITY OF LOAD DROP P8R LIPT Load Drop Scenario idean Procanility Drop During Drop After Initial Lift In.itial Lift Reactor Vessel Head 1.8 X 10-4 6.9 X 10-5 Steam Separator Assembly 2.3 X 10-4 6.8 X 10-5 Recirculation Pump' Motor not relevant

• 3.3 X 10-4 1 .

• Probabilistically, the dominant failure mechanism leading to a drop of a recirculation pump motor is an overspeed event.

Consequently, a drop of the motor during the initial lif t and hold stage is not considered.

J Page 8 of 10 t - r---. - - - ..- * -y -,w- +e. , -m,-,- 7 w-

TABLE 2 MEAD PROBABILITY OF SdIPPING CASK DROP PER LIFT (for casks weighing 35 tons or less) 2 Shipping Cask Drop Scenario Mean Procabiti.ty Drop During Drop After Initial Lift Initial Lift Equipment Hatch not relevant

• 3.3 X 10-4 Operating Floor 8.2 X 10'S 6.9 X 10-5 o

• Probabiliatically, the dominant failure mechanism leading to a drop of a shielded shipping cask is an overspeed event.

Consequently, a drop of a shipping cask during the initial lift and hold stage is not considered.

Page 9 of 10

-y,- ,- -- y - . - , _ --

REFCRENCES

1. PASNY letter, J. P. Bayne to Domenic d. Vassalto, February 26,

, 1982 (JPN-82-25), " Response to NURSG-0612, Report No. 2" (nine-month submittal).

2. PASNY letter, J. P. Bayne to Thomas A. Ippolito, October 15, 1981 (JPN-81-82), " Response to NUREG-0612, Report No. 1" (six-month submittal).-

3. U.S. Nuclear Regulatory Commission, An Approach to Quantitative Safety Goals for Nuclear Power Plants, NORSG-0739, October 1980.

4. NRC Office of Policy Evaluation, Discussion Paper on Safety Goals, July 23, 1981.

5. Federal Register, Volume 48, No. 50, March 14, 1983, pp. 10772 -

10776.

I Page 10 of LO

-- y & - - - - - - - -%mwv - - - - - ---wy T c ,e