ML20195C253

Responds to NRC Re Violations Noted in Insp Rept 70-1100/86-01.Corrective Actions:Adequate Number of Coat Hangers Installed in Each Licensed Lab in Bldg 5

ML20195C253
Person / Time
Site:	07001100
Issue date:	05/19/1986
From:	Lichtenberger ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY
To:	Martin T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
References
NUDOCS 8605300187
Download: ML20195C253 (5)

Text

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SOMBtf5 TION ENGINEERING l

License SNM-1067 Docket 70-1100 May 19, 1986 l

U. S. Nuclear Regulatory Commission Region I 631 Park Avenue King of Prussia, PA 19406 Attention:

Mr. Thomas T. Martin, Director Division of Radiation Safety and Safeguards

Reference:

1) Letter from Thomas T. Martin, NRC, to H. V. Lichtenberger, CE, Dated April 23, 1986; Combined Inspection No's 70-1100/

86-01 and 30-3754/86-01

2) Letter from H. V. Lichtenberger, CE, to W. T. Crowe, NRC, Dated December 16, 1985

3) Letter from N. Ketzlach, NRC, to H. V. Lichtenberger, CE, Dated February 14, 1986

4) Letter from H. V. Lichtenberger, CE, to N. Ketzlach, NRC, Dated March 18, 1986

Dear Mr. Martin:

This is in reply to the above referenced letter in which you reported that as a result of your inspector's visit to our facility on January 13-17, 1986, certain of our activities were not in full compliance with NRC requirements. Our response to the Notice of Violation and Notice of Deviation, Appendixes A and B, respectively, is as follows:

APPENDIX A, ITEM A Section 4.14. " Posting of Limits", of your NRC-approved license application (Part 1-Criteria), dated June 15, 1984, states, in part, that all work stations and storage areas shall be posted with a nuclear safety limit approved by the Manager, Nuclear Licensing, Safety, Accountability and Security (NLSA&S) or the Nuclear Criticality Specialist.

Contrary to the above, on January 13, 1986, the work station in the bundle storage room, used to store up to two open or closed fuel bundle shipping containers in designated locations, was not posted with nuclear safety limits.

8605300187 860519 PDR ADOCK 07001100 C

PDR Power Systems 1000 Prospect Hdl Road (203) 688 1911 Combustion Engineering. Inc.

Post Office Box 500 Telex: 99297 Windsor, Connecticut 06095-0500

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Mr. Thomas T. Martin May 19, 1986 U. S. Nuclear Regulatory Commission

RESPONSE

At the time of your inspector's visit, the bundle storage room was posted with a nuclear safety limit sign.

However, your inspector was not satisfied with the sign because it did not include a restriction on the number of fully loaded fuel bundle shipping containers in the room even though SNM-1067 allows us a maximum of two containers at one time.

It was explained to your inspector that our normal practice is to limit the number of containers in the room to a maximum of two since placing a third container in the room makes it difficult, if not impossible, to perform routine operations. Your inspector was not satisfied with this explanation and we agreed to modify the safety sign to place a limit of two containers in the room at one time.

The new sign was posted on February 6, 1986.

APPENDIX A, ITEM B Section 2.2.2, " Nuclear Fuel Manufacturing - Windsor", of Part 1 (Criteria) of your NRC-approved license application states that the General Manager delegated to the Production and Material Control Manager and to the Engineering Manager responsibility to assure that all operations involving nuclear materials have been analyzed to establish the required safety limits and controls.

The Manager, NLSA&S or Nuclear Criticality Specialist shall assist the Engineering Manager and the Production and Material Control Manager by performing the analysis required and establishing the appropriate controls.

Contrary to the above, on January 13, 1986, the inspector iden-tified that the following two operations involving nuclear materials had not been analyzed to establish the required safety limits and controls:

(1) in the fuel rod prestacking operation, the configuration of fuel rods was modified to include 20 blank, hollow spacers in the array'of fuel rods in storage trays; and (2) in the pellet press operation, there was an additional tray located underneath the press to pellet boat transfer ramp (screen) which increased the slab thickness from 4.0 to 4.5 inches.

RESPONSE TO ITEM (1)

On December 16, 1985 we submitted a license change request in-cluding appropriate analysis results, Reference 2, which would allow us to stack fuel rods in storage boxes, using hollow bars as spacers, to a maximum height of six inches.

It should be noted that the five and one-half inch slab height presently approved in SNM-1067 was based on a 0.48 water / fuel ratio (Ref. Para. 8.33, Page II.8-17) khich was based on the following conservative assumptions:

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a Mr. Thomas T. Martin May 19, 1986 U. S. Nuclear Regulatory Commission

- the fuel rods in the rod box were stacked in a close packed hexagonal array, and

- the cladding and clad / fuel gap were assumed to be water (UO2 pellet stacks were assumed to be surrounded by an imaginary water film)

Utilizing the 0.48 water / fuel ratio, a critical infinite slah thickness of 11 inches was established for a 4.1% enrichment using Figure 1.E.16 of UKAEA Handbook ANSB 1.

Applying the license safety factor for slabs (Para. 4.2.4, Page 1.4-6) of 1.2 yields an allowable slab thickness of about nine inches which is much higher than the five and one-half inches approved in the original license or the six inches requested in our Reference 2 submittal.

In reply to our Reference 2 submittal, we received Reference 3 requesting clarification and further information regarding our change request. One of the items discussed in Reference 3 was the fact that we had not provided sufficient information regarding the use of hollow spacer bars in the pre-stacking of fuel rods in rod storage boxes.

It was felt that the hollow spacer bars introduced more water into the rod storage box which in turn changed the water / fuel ratin and the critical infinite slab thickness.

In answer to the Reference 3 letter, we submitted Reference 4 which provided additional inforsati.on to clarify our original Reference 2 submittal. This information is further elaborated on as follows:

By introducing the 20 hollow spacer bars into the array of fuel rods in storage trays, the water / fuel ratio was recalculated using the following assumptions:

- the fuel rods in the box were stacked in a close packed l

hexagonal array, and l

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- the cladding and clad / fuel gap were not displaced by water and the neutron absorbing property of the cladding was neglected, and

- the 20 hollow spaccis were assumed to be water which was i

uniformly distributed throughout the rod box.

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Utilization of the noted assuwptions resulted in a new wator/ fuel j

ratio of 0.37 which is less than the 0.48 water / fuel ratio speci-fled in the license. Again, utilizing Table 1.E.16 of UKAEA Handbook AHSB 1, it. was found that the critical infinite slab thickness, utilizing either a 0.48 or 0 37 water / fuel ratio and 3

a 1.2 safety factor, is still about nine inches which is much higher than the presently approved five and one-half inch slab height or the requested change to a six inch slab height.

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A Mr. Thoxas T. Martin May 19, 1986 U. S. Nuclear Regulatory Commission RESPONSE TO ITEM (2)

The area in question within the pellet presses was originally analyzed on the basis of a slab limit. The trays in question have always been located in the position noted since they were considered to be part of the original slab limit analysis.

However, UO, powder sometimes overflows the trays and increases the overall slab height to four and one-half inches. To eliminate this problem we have performed preliminary analysis which shows I?

that the slab height can be increased to five inches without impairing the safety of the operation. A formal request for approval for the new five inch slab limit will be submitted to i

the NRC within the next several weeks.

In the interim we will require the pellet press operator to empty the powder tray on a more frequent basis to eliminate the possibility that the press slab limit will be exceeded.

APPENDIX B As a result of the inspection conducted on January 13-17, 1986, and in accordance with NRC Enforcement Policy (10 CFR 2, Appendix C), the following deviation from standard industry practice was identified:

t Acceptable radiation control procedures do not permit the co-mingling of protective clothing that is potentially contaminated with removable radioactivity with protective clothing that is not contaminated.

Contrary to the above, on January 14, 1986, the inspector observed lab coats (protective clothing) that were potentially contaminated with removable radioactivity on the outside hanging on top of one another in the Building 5 chemistry laboratory such that the inside of one garment was in contact with the outside of the garment beneath. The inspector also observed an individual starting to don one of the garments without conducting a contamination survey to determine whether the inside was contaminated.

RESPONSE

Appendix B indicates that employees working in the Building 5 chemistry laboratory were observed hanging potentially contam-inated lab coats on top of one another. This practice could j'

result in the transfer of radioactive contamination from the 1

outside of the bottom garment to the inside of the top garment.

Results of a subsequent review of the incident revealed that an insufficient number of hangers were available for the number of employees authorized to work in the laboratory. An adequate number of coat hangers have been installed in each of the i

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Mr. Thomas T. Msrtin May 19, 1986 U. S. Nuclear Regulatory Commission licensed laboratories in Building 5.

In addition, laboratory employees have been instructed in the proper procedure for handling potentially contaminated protective clothing. This corrective action was completed within one week of the inspectors visit.

Very truly yours, yk ' t, C

H. V. Lichtenberger Vi,ce President Nuclear Fuel HVL/RES/ sam I

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