05000245/LER-1997-013, :on 970217,discovered That Weight of Lower Mast Sections of Refueling Platform Fuel Grapple Were Not Added to Fuel Assembly Weight.Caused by Failure to Maintain Design Basis.Performed New Impact Load Analysis.With

:on 970217,discovered That Weight of Lower Mast Sections of Refueling Platform Fuel Grapple Were Not Added to Fuel Assembly Weight.Caused by Failure to Maintain Design Basis.Performed New Impact Load Analysis.With

ML20196H223
Person / Time
Site:	Millstone
Issue date:	12/03/1998
From:	Harkness E, Willoughby P NORTHEAST NUCLEAR ENERGY CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
B17064, LER-97-013, LER-97-13, NUDOCS 9812090017
Download: ML20196H223 (6)

LER-1997-013, on 970217,discovered That Weight of Lower Mast Sections of Refueling Platform Fuel Grapple Were Not Added to Fuel Assembly Weight.Caused by Failure to Maintain Design Basis.Performed New Impact Load Analysis.With

Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(2)(ii) 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition
2451997013R00 - NRC Website

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Northeast noi. Fony adanoute is6),waa-w Oe8s j

Nuclear Energy xai. tone N-iear ro, station i

Northeast Nuclear Fargy Company l

P.O. Box 128 Tatedord, CT 0638s-0128 (860) 447-1791 Fax (860) 444-4277 I

The Northeast Utilities System DEC - 31998 i

Docket No. 50-245 I

B17064

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Re: 10CFR50.7S(a)(2)(ii)

U. S. Nuclear Regulatory Commission l

Attention: Document Control Desk

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Washington, DC 20555 j

s This letter forwards supplemental Licensee Event Report (LER) 97-013-01 1

(Attachment 1), documenting a condition that was discovered at Millstone Unit i

No.1 on February 17,1997, and initially reported on March 19,1997. This LER is submitted pursuant to 10CFR50.73(a)(2)(ii).

i There are no regulatory commitments contained within this letter.

i i

Very truly yours, 1

NORTHEAST NUCLEAR ENERGY COMPANY nL E. J.44a'rkness Director-Unit Operations

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cc:

H. J. Miller, Region i Administrator y/ gg

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j L. L. Wheeler, NRC Project Manager, Millstone Unit No.1 9912090017 901203 9 PDR ADOCK 05000245(

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PMd l

083422-5 RLY 12-95. '

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l Docket No. 50-245 B17064 Millstone Nuclear Power Station, Unit No.1 Licensee Event Report 98-013-01 i

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I December 1998

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NHC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3160-0104 (4-95)

EXPfRES 04/30/98 UCENSEE EVENT REPORT (LER)

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PAGE (3) l Millstone Nuclear Power Station Unit 1 05000245 1OF4 TITLE 14)

Evaluation of impact Load of the MP1 Refueling Platform Fuel Grapple Mast Over Spent Fuel Pool Racks and the Reactor Vessel Guide Plate EVENT DATE (s)

LER NUMBER (6)

REPORT DATE (7)

OTHER FACILITIES INVOLVED (8)

SE AL RE N

MONTH DAY YEAR YEAR MONTH DAY YEAR NU NU R

02 17 97 97

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CAUSE

SYSTEM COMPONENT MANUFACTURER

CAUSE

SYSTEM COMPONENT MANUFACTURER PRDS F DS l

SUPPLEMENTAL REPORT EXPECTED (14)

EXPECTED MONTH DAY YEAR YES SUBMISSION O (If yes, complete EXPECTED SUBMISSION DATE).

@ NO DATE (1 s)

ABSTRACT (Limrt to 1400 spaces. i.e.. approximately 15 single-spaced typewnttenlines) (16)

On February 17,1997, with the plant in COLD SHUTDOWN, it was discovered during a document review that the weight of the lower mast sections of the Refueling f:atform fuel grapple were not added to the fuel assembly wright, to determine the total potential impact load for the design of the spent fuel pool storage racks. When moving fusi in the fuel pool, in the event of a single active failure of a hoist component, such as a cable break, could result in a dropped fuel assembly falling with portions of the mast onto fuel stored in the spent fuel storage racks or in an empty storage rack cell. In the course of reviewing potentialinteractions between the refueling mast and the spent fu:1 storage racks, it was also discovered on February 24,1997, that a similar interaction may occur when the mast l

I would be operated over the reactor vessel. In the event of a break in the hoist cable, the mast would not reach its full extension and consequently, a portion of the mast could remain attached to a dropped fuel assembly and strike th3 top of the fuel or the reactor vessel internals. These conditions were determined to be outside the design basis of l

tha plant and reported on February 18,1997, and February 25,1997, in accordance with 10CFR50.72(b)(1)(ii)(B).

Th3 cause of the event has been determined to be the failure to maintain the design basis as stated in the Updated l

Finil Safety Analysis Report (UFSAR) based on a revised analysis by General Electric (GE) in 1987.

The revised l

an: lyses were subsequently performed for postulated drops over the fuel pool storage racks and reactor vessel core crea. Changes were processed for incorporation of the revised analysis into the Millstone Unit No.1 UFSAR.

NRC FORM 366 (4-95)

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Nf1C FORM 366A u.S. NUCLEAR REGULATORY CoMMISSloN (4-95)

LICENSEE EVENT REPORT (LER)

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TEXT CONTINUATION i

FACILITY NAME (1)

DOCKET LER NUMBER (6)

PAGE (3)

SEQUENTIAL REVISION A

NUMBER NUMBER 2OF4 Millstone Nuclear Power Station Unit 1 05000245 97 001 l

l TEXT (If more spaceis required, use additionalcopies of NRC Form 366A) (11l l

1.

Description of Event

On February 17,1997, with the plant in COLD SHUTDOWN, during a documentation review performed in support of a response to the NRC for additionalinformation regarding the impact load for the design of the fuel pool storage racks, it was discovered that the weight of the lower mast sections of the Refueling Platform fuel grapple were not included in the impact load for the design of the spent fuel pool storage racks. It was determined that a single active failure, such as a cable break, could result in a dropped fuel assembly falling with portions of the mast onto fuel stored in the spent fuel storage racks or in en empty storage rack cell. The additional kinetic energy from the falling mast sections was not included in the design of the spent fuel storage racks. This condition was determined to be reportable on February 18,1997, and was reported pursuant to 10CFR50.72(b)(1)(ii)(B) as a condition that is outside the design basis of the plant.

On February 24,1997, with the plant in COLD SHUTDOWN, in the course of reviewing potential interactions between the refueling mast and the spent fuel storage racks, it was discovered that a similar interaction may occur when the mast would be operated over the reactor vessel. In the event of a break in the hoist cable, the mast would not reach its full extension, and consequently, a portion of the mast will be attached to the assembly as it strikes the top of the fuel or the reactor vessel internals. This additional kinetic energy from the falling mast sections is not shown in the UFSAR Chapter 15.8 fuel handling accident analysis for Millstone Unit

1. This condition was reported on February 25,1997, pursuant to 10CFR50.72(b)(1)(ii)(B) as a condition that is outside the design basis of the plant.

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II. Cause of Event

l The cause of the event has been determined to be the failure to maintain the design basis as stated in the UFSAR based on a revised analysis by GE in 1987.

The refueling accident analysis of a main hoist cable break is documented in a letter from GE to NRC on November 13,1987. In this analysis, the radiological results of this refueling accident were compared to the l

Licensing Topical Report NEDE-24011-PA, " General Electric Standard Application for Reactor Fuel" (GESTAR-l 11). Since the new analysis showed fewer failed rods (104 versus 124), the radiological consequences were bounded by the previous GESTAR-ll analysis presented in the UFSAR. The differences between the new l'

accident analysis (drop height, fuel assembly weight, mast weight, total energy, etc.) and the GESTAR-il analysis. vere incorporated into a revision of the GESTAR-Il report. This revised report, should have been i

incorporated into the UFSAR, Section 15.8 fuel handling accident, after being appropriately evaluated for l

potential effect on the existing design basis.

1 Based on the discovery of the new accident analysis presented in the GE letter, the fuel grapple mast is not redundant or single failure proof. In order to operate the equipment with this condition, physical equipment j

modifications (redundant cable and drum, etc.) or load drop analyses are required to justify the operation of the mast for fuel movement in the reactor vessel and the spent fuel pool. However, these options were never performed for Millstone Unit No.1.

CC FORM 366A (4 95) i i

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.U.S. NUCLEAR REGULATORY COMMISSION (4-93)

UCENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME (1)

DOCKET LER NUMBER (6)

PAGE (3) l SEQUENTIAL REVISION YEAR NUMBER NUMBER 3OF4 Millstone Nuclear Power Station Unit 1 05000245 97 ~ 13 ~

001 TEXT (11more space is reqwred, use additionalcopies of NRC Form 366A) (11) lli. Analysis of Event i

This report is pursuant to 10CFR50.73(a)(2)(ii)(B) as a condition that was outside the design basis of the plant, i

since:

the weight of the lower mast sections of refueling platform were not included in the weight summation of i

the impact load for the design of the spent fuel storage racks.

it has been discovered that the mast must be operated over the reactor vessel and in the event of a break i

e in the hoist cable, the mast will not reach its full extension before impact occurs, consequently, a portion of the mast would be attached to a dropped fuel assembly which would strike the top of the reactor vessel guide plate, or fuel assemblies, j

t A detailed review of the fuel mast and grapple shows that the hoist cab!e supports the grapple head, fuel l

Essembly and the telescoping mast sections that are not fully extended. In the event of a break in the hoist cable, the weight of the cable supported mast sections would provide a significant contribution to the impact l

. forces imparted to the spent fuel pool racks, the top of fuel or the reactor vessel internals. This additional weight was not included in the impact analysis performed for the 1976 and 1988 rerack projects.

Analyses have subsequently been performed to demonstrate that the Boraflex (Holtec) and Tetrabor (NUS) fuel

. storage racks would remain functional after shallow drop (straight drop onto the top of a rack) and deep drop (straight drop through an individual cell all the way to the bottom of the rack cell onto the baseplate) accident i

scenarios. The impact force includes the combined weight of the fuel assembly, grapple head and telescoping

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mast sections. The structural analysis performed concludes that rack cell deformation does not extend to the 1

active fuel region of the cell. Thermal-hydraulic analysis for the deformed cell resulting in a reduction in the flow area available for the upflow of water coolant, was performed to demonstrate that the fuel is maintained in a coolable condition within the rack cell. Deep drop analysis performed for the racks show that the l

baseplate material does not fail and that the liner is not impacted. Criticality analysis was also performed to demonstrate that in the event a dropped assembly accident were to occur, the resulting configurction would not compromise criticality safety.

The September 1988 Revision to the GESTAR ll analysis included the additional weight of the mast and was reviewed to determine its applicability to Millstone Unit No.1.

This review determined that the GESTAR 11 enalysis is not completely applicable due to differences in the weight of the fuel handling mast.

Therefore, new load drop analyses have also been performed for the reactor vessel core area. The impact force includes the combined weight of the fuel assembly, grapple head and telescoping mast sections falling onto the top of the core and impacting a group of four assemblies. Revised analyses have been performed for 7 x 7,8 x 8 and 9 x 9 fuel assemblies. Although the results demonstrate that the 7 x 7 fuel design remains the limiting cccident scenario, this fuel is no longer in use and not applicable to the current evaluation. The analysis for I

the current vintage 8 x 8 and 9 x 9 fuel results in 1.92 failed fuel assemblies (119 fuel pins for the 8 x 8 assembly and 142 fuel pins for the 9 x 9 assembly).

These results (and associated offsite radiological dose) remain bounded by the existing UFSAR analysis which assumes 2.0 failed fuel assemblies (124 fuel pins for the 8 x 8 assemblies and 158 fuel pins for the 9 x 9 assemblies).

There were no actual safety consquences as a result of this event.

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LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME (1)

DOCKET LER NUMBER (6)

PAGE (3)

SEQUENTIAL REVISION YEAR NUMBER NUMBER 4OF4 Millstone Nuclear Power Station Unit 1 05000245 97 ~ 13 ~

001 TEKT (11more space is required, use additionalcopies of NRC Form 366A) (17)

IV. Corrective Action

1. Immediate corrective action was a restriction to the refueling platform main hoist fuel grapple mast to prevent fuct movement activities. These restrictions were subsequently removed based on the revised analyses performed for commitment number 2.

2. Northeast Nuclear Energy Company (NNECO) has performed a new impact load analysis for the fuel pool storage racks that includes the additional weight of the cable supported fuel grapple rnast sections.

Changes have been processed for incorporation of the results into the Millstone Unit No.1 UFSAR.

j

3. For the reactor vessel core area, the latest GESTAR-ll analysis will be reviewed to determine if it is i

applicable to Millstone Unit No.1 and changes processed for incorporation into the Millstone Unit No.1 UFSAR, if appropriate. If the GESTAR ll analysis is not applicable to Millstone Unit No.1, then a new analysis will be performed and changes processed for incorporation into the Millstone Unit No.1 UFSAR.

This analysis and the appropriate UFSAR update will be completed prior to fuel movement in the fuel pool.

i This review has been completed. The GESTAR-il analysis is not completely applicable to Millstone IJnit No.

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1 due to design differences in the fuel handling mast (weight). Similar analyses have been performed using j

specific Millstone Unit No.1 data. Changes have been processed for incorporation of these results into the i

Millstone Unit No.1 UFSAR.

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4. NNECO will supplement this LER with the results of the impact load analysis. NNECO will supplement this LER before moving fuelin the spent fuel pool. This LER Supplement fulfills this commitment.

l V. Additionat information

Similar Events

l None Manufacturer Data l

l Not Applicable l

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