05000245/LER-1997-013-01, :on 970217,evaluation of Impact Load of MP1 Refueling Platform Fuel Grapple Mast Over Spent Pool Racks & Reactor Vessel Guide Plate Occurred.Caused by Failure to Maintain Design basis.GESTAR-11 Analysis Will Be Reviewed

:on 970217,evaluation of Impact Load of MP1 Refueling Platform Fuel Grapple Mast Over Spent Pool Racks & Reactor Vessel Guide Plate Occurred.Caused by Failure to Maintain Design basis.GESTAR-11 Analysis Will Be Reviewed

ML20147F449
Person / Time
Site:	Millstone
Issue date:	03/19/1997
From:	Robert Walpole NORTHEAST NUCLEAR ENERGY CO.
To:
Shared Package
ML20147F446	List: 05000245/LER-1997-013-01, :on 970217,evaluation of Impact Load of MP1 Refueling Platform Fuel Grapple Mast Over Spent Pool Racks & Reactor Vessel Guide Plate Occurred.Caused by Failure to Maintain Design basis.GESTAR-11 Analysis Will Be Reviewed
References
LER-97-013-01, LER-97-13-1, NUDOCS 9703250133
Download: ML20147F449 (4)

LER-1997-013, on 970217,evaluation of Impact Load of MP1 Refueling Platform Fuel Grapple Mast Over Spent Pool Racks & Reactor Vessel Guide Plate Occurred.Caused by Failure to Maintain Design basis.GESTAR-11 Analysis Will Be Reviewed

Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(viii) 10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(iv), System Actuation 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition
2451997013R01 - NRC Website

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i NRC FDRM 366 U.S. NUCLEAR REUULATORY COMMISSION APPROVED BY OMS NO. 3160 0104 (4-95)

EXPIRES 04/30/98 NoEAr coEEET" N REO ES 60 RS E O TED n^c?io^"!4&"Pa^ Paso'?dl@ J"22'Jo*Ni LICENSEE EVENT REPORT (LER) lS"uT=fa"#st'4"'E #ds"s'#"."EJM" &

i (See reverse for required number of digits / characters for each block)

FACtMTV NAME (1)

DOCKET NUMBER (2)

PAGE (3)

Millstone Nuclear Power Station Unit 1 05000245 1 of 4 1

TITLE (4)

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 (5)

LER NUMBER (6)

REPORT DATE (7)

OTHER FACILITIES INVOLVED (8)

[

j MONTH DAY YEAR YEAR SEQUENTIAL REVISION MONTH DAY YEAR FAcitiTv #4AME DOCKET NUMBER q

NUMBER l

" "^"'

02 17 97 97 013 00 03 19 97 OPERATING THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR 5: (Check one or more) (11)

MODE (9)

N 20.2201(b) 20.2203(a)(2)(v) 50.73(a)(2)(i) 50.73(a)(2)(viii)

POWER LEVEL (10) 000 20.2203(a)(1) 20.2203(a)(3)(i)

X So.73(a)(2)(ii) 50.73(a)(2)(x) 20.2203(a)(2)(i) 20.22o3(a)(3)(ii) 50.73(a)(2)(iii) 73.71 20.2203(a)(2)(ii) 20.2203(a)(4) 50.73(a)(2)(iv)

OTHER 20.2203(a)(2)(iii) 50.36(c)(1) 50.73(a)(2)(v) specifY n Abstract below 6

4 20.2203(all2)(iv) 50.36(c)(2) 50.73(a)(2)(vii)

LICENSEE CONTACT FOR THIS LER (12)

NAME TELEPHONE NUMBER (include Ataa Codel Robert W. Walpole, MP1 Nuclear Licensing Manager (860)440-2191

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COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)

CAUSE

SYSTEM COMPONENT MANUFACTUPER REPORTABLE

CAUSE

SYSTEM COMPONENT MANUFACTURER REPORTABLE TO NPRDs To NPRDs SUPPLEMENTAL REPORT EXPECTED (14)

EXPECTED MONTH DAY YEAR SUBMISSION f YES NO 05 15 97 (if yes, complete EXPECTED SUBMISSION DATE).

ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines) (16)

On February 17,1997, with the plant in COLD SHUTDOWN, it was discovered that the weight of the lower mast j

ssctions of the Refueling Platform fuel grapple were not added to the fuel assembly weight, to determine the total impact load for # 1 design of the spent fuel pool storage racks. In the evt.nt of a single active failure of a hoist component, sucl. as a cable break, the mast would not reach its full extension before an impact occurs with the top of fuel or the storage racks, in the course of reviewing potential interactions between the refueling mast and the j

spsnt fuel storage racks, it was discovered on February 24,1997, that the mast would also be operated over the reactor vessel in a less than fully extended configuration for the fuel handling grapple to clear the top of the guide pt.te. 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 the top of fuel or the reactor vessel internals. These conditions were determined to be outside the design basis of the plant and reported on February 18, 1997, and February 25,1997. The cause of the event has been determined to be the failure to maintain the design basis as stated in the Updated Final Safety Analysis Report (UFSAR) by not including the revised analysis by General Elsctric (GE) in 1987. Restriction has been placed on refueling platform main hoist fuel grapple to prevent fuel movement activities. For the fuel pool storage racks, new impact load analysis will be performed. For the reactor vsssel core area, the latest GESTAR-Il analysis will be reviewed for applicability, if required, a new analysis will be parformed. Changes will be processed for incorporation into the Millstone Unit No.1 UFSAR.

9703250133 970319 PDR ADOCK 05000245 S

PDR

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1 NRC F;RM 316A U.S. NUCLEAR REGULATORY COMMISSION (4 951 LICENSEE EVENT REPORT (LER) i TEXT CONTINUATION FACILITY NAME (1)

DOCKET NUMBER (2)

LER NUMBER (6)

PAGE (3)

YEAR SEQUENTIAL REVISION Millstone Nuclear Power Station Unit 1 05000245 NUMBER NUMBER 2 of 4 97 013 00 TEXT (if more space is required, use additional copies of NRC Form 366A) 117) i 1.

Descrintion of Event On February 17, 1997, with the plant in COLD SHUTDOWN, during a documentation review performed in l

support of a response to the Staff for additional information regarding the impact load for the design of the fuel 4

pool storage racks, it was diccovered that the weight of the lower mast sections of the Refueling Platform fuel j

grapple were not included in the impact load for the design of the spent fuel pool storage racks. It was j

dstermined that a single active failure, such as a cable break, could result in a dropped fuel assembly falling w'th portions of the mast onto fuel stored in the spent fuel storage racks or in an empty stcrage rack cell. The i

cdditional kinetic energy from the falling mast sections was not included in the design of the spent fuel storage i

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 bstween the refueling mast and the spent fuel storage racks, it was discovered that the mast would be operated 4

over the reactor vessel in a less than fully extended configuration for the fuel harming grapple to clear the top of j

the guide plate. 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 strike the top of fuel or the reactor vessel internals. This additional i

kinetic energy from the falling mast sections is not shown in the UFSAR Chapter 15.8 fuel drop accident analysis for Millstone Unit 1. This condition was reported on February 25,1997, pursuant to 10CFR50.72(b)(1)(ii)(B).

The main hoist fuel grapple was tagged out as a compensatory measure to prevent the movement of fuel. There were no automatic or manually initiated safety system responses as a result of this event.

11.

Cause of Event

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 anelysis 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 Licensing Topical Report NEDE-24011-PA, " General Electric Standard Application for Reactor Fuel" (GESTAR-II).

Since the new analysis showed fewer failed rods (104 versus 124), the radiological consequences were bounded by the previous GESTAR-il analysis presented in the UFSAR. The differences between the new accident analysis (drop height, fuel assembly weight, mast weight, total energy, etc.) and the GESTAR-il analysis were incorporated into a revision of the GESTAR-il report. This revised report, should have been incorporated into the UFSAR, Section 15.8 fuel handling accident, after being appropriately evaluated for potential effect on the existing design basis.

Bised 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 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 pstformed for Millstoae Unit No.1.

• U.s. NUCLEAR REGULATORY COMMISSloN LICENSEE EVENT REPORT (LER) l TEXT CONTINUATION FACILITY NAME (1)

DOCKET NUMBER (2)

LER NUMBER (6)

PAGE (3)

YEAR SEQUENTIAL REvlSION Millstone Nuclear Power Station Unit 1 05000245 NUMBER NUMBER 3 of 4 97 013 00 TEXT (If more space is required, use additional copies of NRC form 366A) (17) i 111. Analysis of Event l

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

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

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 in a less than fully extended e

configuration for the fuel handling grapple to clear the top of the guide plate. In the event of a break in the hoist cable, the mast will not reach its full extension before impact occurs, and consequently, if a cable break y

occurred, a portion of the mast would be attached to a dropped fuel assembly which would strike the top of i

the spent fuel pool rack, reactor vessel guide plate, or fuel assemblies.

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

The mechanical drop accident analyses associated with the spent fuel pool reracks for 1976 and 1988 considered j

only a fuel assembly dropped from a specific height above the rack. The 1976 rerack project assumed a 634 pound unchannelled fuel assembly dropped 24 inches above the spent fuel pool rack. The 1988 rerack project used an 800 pound load dropped from a 36 inch height above the spent fuel pool rack. The 800 pound load 4

(

assumed a 634 pound fuel assembly and channel (64 pounds) plus an additional 100 pounds to envelope the largest Boiling Water Reactor fuel assembly used in the industry.

4 A detailed review of the fuel mast and grapple shows that the hoist cable supports the grapple head, fuel assembly 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 forces imparted to the spent fuel pool racks, the top of fuel or the reactor vessel internals. This additional weight i

was not included in the impact analysis performed for the 1976 and 1988 rerack projects. The September 1988 Rsvision to the GESTAR 11 analysis included this additional weight and needs to be reviewed to determine its applicability to Millstone Unit No.1.

1

IV. Corrective Action

1.

Immediate corrective action was a restriction to the refueling platform main hoist fuel grapple mast to prevent fuel movement activities.

2.

Northeast Nuclear Energy Company (NNECO) will perform a new impact load analysis for the fuel pool storage racks that includes the additional weight of the cable supported fuel grapple mast sections prior to core reload.

3.

For the reactor vessel core area, the latest GESTAR-Il analysis will be reviewed to determine if it is applicable to Millstone Unit No.1 and changes processed for incorporation into the Millstone Unit No.1 UFSAR, if appropriate, prior to core reload. If the GESTAR ll analysis is not applicable to Millstone Unit No.1, then a j

new analysis will be performed and changes processed for incorporation into the Millstone Unit No.1 UFSAR prior to core reload.

4.

NNECO will supplement this LER with the results of the impact load analysis. NNECO will supplement this 4

LER by core reload.

1

,U.S. NUCLEAR REGULATORY COMMISSION (4 95) c LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME (1)

DOCKET NUMBER (2)

LER NUMBER (6)

PAGE (3)

YEAR SEQUENTIAL REVISION i

Millstone Nuclear Power Station Unit 1 05000245 NUMBER NUMBER 4of4 j

97 013 00 TEXT (11more space is required, use additional copies of NRC Form 366A) (17)

V.

Additional Information

Similar Events None Manufacturer Data Not Applicable

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i 93RC FORM 366A (4-95)

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