L-82-393, Forwards Revised Response to Question 410.10 Re Handling of Light Loads.Probability of Unacceptable Consequences Resulting from Drop of Light Load Over Open Reactor Vessel Is Very Low

Forwards Revised Response to Question 410.10 Re Handling of Light Loads.Probability of Unacceptable Consequences Resulting from Drop of Light Load Over Open Reactor Vessel Is Very Low

ML17213A413
Person / Time
Site:	Saint Lucie
Issue date:	09/03/1982
From:	Robert E. Uhrig FLORIDA POWER & LIGHT CO.
To:	Eisenhut D Office of Nuclear Reactor Regulation
References
L-82-393, NUDOCS 8209100098
Download: ML17213A413 (13)

Text

REGULATORY OI,ATION DISTRIBUTION SY M

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AOCESSION NBR;8209100098 DOC ~ DATE: 82/09/03 NOTARI'ZEO; NO FACIL:50-389 St ~ Lucie P lant'niit 2i Flor ida Power 8 Light Co.

AUTH ~ NAME AUTHOR AFFILIATION UHRIGiR,GB Florida Power 8, Light Co "RECIP ~ NAME RECIPIENT AFFILIATION EISENHUTEDBG ~

Division of Licensing DOCKET 05000389

SUBJECT:

Forwards -revised response to Question 410 '0 re handling of light loads, Probability of unacceptable consequences resulting from drop of light load over open reactor vessel is very low, DISTRIBUTION CODE:

BOOIS COPIES RECEIVED:LTR, ENCL g SIZE:,

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'ITLE:

Licensing 'Submittal:

.PSAR/FSAR Amdts 8, Related Correspondence'OTES:

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P. X,JIJNO BEACH, FL 33408 FLORIDAPOWER & LIGHTCOMPANY September 3,

1982 L-82-393 Office of Nuclear Reactor Regulation Attn:

Mr. Darrell G. Eisenhut, Director Division of Licensing U.

S. Nuclear Regulatory Commission Washington, D.

C; 20555

Reference:

St. Lucie Unit P2 Docket No. 50-389 Attached please find Florida Power and Light Company's revised response to NRC question 410.10 regarding the handling of light loads at St. Lucie Unit 2.

If you have any questions, please contact us accordingly.

Sincerely, Robert E. Uhri 9 Vice President Advanced Systems and Technology REU/cab

'P DO]

8209i00098 820903 PDR aDaCX 0S000S8S IIII PDR PEOPLE... SERVING PEOPLE

V

St. Lucie Unit 2

Response

to NRC Question 410.10 Li ht Loads Question 410.10:

Describe, discuss and verify that the maximum potential kinetic energy contained in all objects of less weight than a spent fuel assembly which will be handled over spent fuel will not exceed the effects of the fuel

~

handling accident in section 15.7.4 of the FSAR.

Response

FPL has reviewed the potential for an inadvertent drop of light objects onto spent tuel'at St. Lucie Unit 2.

The objects that were considered were all lighter than a fuel asse'mbly and i.ts associated lifting tool and thus had not been previously considered in the heavy loads analysis.

A postulated drop may occur in either of two locations; over the spent fuel pool or over an open reactor vessel.

Any object that might be lifted over either area was considered in this review.

The conclusions of *this review are presented below:

A)

Li ht loads lifted over the s ent fuel ool'.

All objects that might be lifted over the spent fuel pool were reviewed for a potential drop.

Attachment 1 presents the list of items considered in this review.

Since each fuel assembly in the spent fuel pool is stored in a separate module of the spent fuel rack (see FSAR section

9. 1.2) it is not considered con-'eivable that more than one fuel assembly could be damaged.

Since the analysis contained in section 15.7.4 assumed that all the pins in one full assembly were damaged, and that all the activity in the pin gas gap was released, the: drop of a light load would have results that are no more limiting than the previously analyzed accident.

A dropped light load in the spent fuel pool cannot impact more than one assembly and the offsite dose cannot be greater than that identified in section 15.7.4.

8)

Li ht loads lifted over the o en reactor vessel:

In the process of analyzing this potential accident, the follow-ing steps were taken:

1) 2)

3) determine determine vessel.

determine item ~

calculate what could be dropped.

when these items could impact fuel in the the damage associated with each dropped the consequences of this damage.

The following details each of the steps:

I Li ht Loads Identification The maintenance and operations departments of St. Lucie 1, a near duplicate plant to St. Lucie 2, provided a list of potential items to be considered in this ligEit loads drop review.

This list included any item that would be lifted by a load handling system over the reactor vessel.

This list of items.'is shown on attachment 2.

Since St. Lucie 1 has been refueled four times, the compiled list has been developed with considerable experience on what may be or has been lifted over open reactor vessels.

It should be noted that these items are lifted rarely and do not~consti-

~

'tute a significant percentage of lifts within containment.

Time of Im act The time period during which an impact could occur was reviewed.

An impact of dropped items with spent fuel in the reactor vessel may only occur wE>en tE>e reactor vessel head and the reactor internals upper guide structure are removed from the vessel.

This period only occurs during a refueling of the reactor.

Additionally, release from damaged fuel will not result'n any appreciable offsite dose unless the con-tainment integrity is not established; that is, the 'containment equip-ment hatch must be open to effect appreciable release to the atmosphere.

This is never the case during refueling periods when the reactor is open except for extremely limited time intervals when it is essential to move equipment.into or out of containment.

Core alterations are not allowed when containment integrity is not established.

Further, during these periods lifting activities over the open vessel are not normally. conducted.

Potential Dama e

The damage associated with any individually dropped item was considered.

Several items were identified as potentialIy serious contributors to fuel damage.

An in depth review of these items indicated that a very sophisticated analysis would be required to determine the extent of any fuel damage.

Rather than embarking on a very lengthy and state-of-the-art calcula-tion to assess the precise damage that might be caused by dropped light

loads, a simplified calculation was performed that indi,cated the number of fuel pins that could be damaged and not cause the site boundary dose

'to exceed a small part of 10 CFR 100.limits.

The number of pins to cause site limits to be reached is dependent on the decay time allowed prior to release.

The results of this calculation can be shown in Figure l.

As indicated on the figure, it will take only 6 days of decay'-.""

to allow a release from one entire assembly without exceeding a small

'raction of site limits.

The time of decay is quite significant to this light loads concern.

First, available gas curie content per pin decreases with time; i.e.,

tE>e fission gas decay.

Second, the minimum time tEiat must pass before tEie fuel in the reactor is exposed to potential drop damage is several days.

To date, the reactor vessel head and upper 2

~ ~

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41 guide structure has not been removed in less than 10 days; and this corres-ponds to decay'time of sufficient duration to maintain site boundary limits even if 360 pins are damaged.

As can be seen from these values, this corres-ponds to significant damage to 1.41 fuel assemblies.

Therefore, from a quali-tative standpoint, it is apparent that a dropped light 'load would need'o cause considerable damage before any site boundary limits are exceeded.

The calculations that have been performed to assess boundary doses conserva-tively assume that all released fi.ssion gases are immediately released from containment.

This assumption is very conservative since the actual release path would be out through the containment equipment hatch which is several levels below the refueling canal water level, the level at which the gas is released.

In the event of an actual fission gas release from the refueling canal, the fission gases would mix with the containment air in the upper containment atmosphere and allow considerable time to isolate the equippent hatch.

It is reasonable to assume that the equipment hatch can be closed in 45 minutes following a gas release.

Once the containment hatch is closed, all fission products inside containment will be isolated from the outside atmosphere.

I FPL administrative procedures allow cranes to be operated only by trained personnel.

Current training programs identify the reactor vessel as an area over which loads should not be handled unless specifically required.

There are rare occasions when loads must'e handled over the open reactor vessel,

however, crane operators are instructed in appropriate procedures (such as ANSI 8 30.2-76) to minimize the possibility of accidents.

In order to further minimize the adverse con-sequences of a potential light load drop, FPL will revise its administrative procedures to restrict lifts over the open reactor vessel when the upper guide structure is not in place and containment integrity is not established.

In conclusion, the review of a dropped light load event has lead us to several fiidings.

First, the probability of damage to fuel appears quite low because o'. the large number of pins that would have to be damaged by a dropped item after a reasonable decay time.

Next, the probability's further 'reduced be-cause containment is usually secured when the reactor vessel is open and there-fore an offsite release would not likely occur.

Finally, since FPL does not normally lift items over the open vessel, the incidence of a dropped load is extremely remote.

'Therefore, FPL feels that the probability of unacceptable consequences result-ing from the drop of a light load over an open reactor vessel is very low.

With the incorporation of more definitive information in the crane operator training program to further assure against fuel damage it is felt that this probability is acceptably low and further action does not appear warranted.

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'heet 1 of 1

Attachment 1

List of Items that May be Lifted Over the S ent Fuel Pool b

a ).oad ))andi)n' stem Item Utilgiy Basket TV camera portable CEA removal tool Removable handrail, 1 section Transfer 'canal seal plate A

roximate Wei ht 300II 20II 100II 150II 500/I

C

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Sheet 1 of 1

Attachment

.2 List of Items that May be Lifted Over the Reac tor Vessel b

a Load Handlin S stem Item A

roximate Wei ht Incore Instrumentation adapter torque tool 3 Lifting bolt torque tools 2 Extension shaft handling tools 60 CEA extension shaft protective sleeves Cavity seal ring extension torque wrench ill CEA extension shaft Upper Guide Structure liftrig deck plate(s)

Utility basket TV Camera portable TV Camera console Upper Guide Structure lift rig tripod Reactor cavity ladder cage Reactor cavity upper ladder section 8 Incore detector Bullet noses CEA removal tool 3 Pressurizer safety valves 2 Power-operated relief valves 2

HEPA filter ventilators 4

RCP seals 501/

501/

758 40i/

258 1008 40i/

3008 20 I'I 50//

800/I 250!/

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Figure l NUIIBER OF FAILED FUEL RODS REQUIRED 'TO YIELD 30 REM TO THE THYROI:D AT THE EAB t

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