Information Notice 1997-39, Inadequate 10 CFR 72.48 Safety Evaluations of Independent Spent Fuel Storage Installations

Inadequate 10 CFR 72.48 Safety Evaluations of Independent Spent Fuel Storage Installations

ML031050510
Person / Time
Site:	Beaver Valley, Millstone, Hatch, Monticello, Calvert Cliffs, Dresden, Davis Besse, Peach Bottom, Browns Ferry, Salem, Oconee, Mcguire, Nine Mile Point, Palisades, Palo Verde, Perry, Indian Point, Fermi, Kewaunee, Catawba, Harris, Wolf Creek, Saint Lucie, Point Beach, Oyster Creek, Watts Bar, Hope Creek, Grand Gulf, Cooper, Sequoyah, Byron, Pilgrim, Arkansas Nuclear, Three Mile Island, Braidwood, Susquehanna, Summer, Prairie Island, Columbia, Seabrook, Brunswick, Surry, Limerick, North Anna, Turkey Point, River Bend, Vermont Yankee, Crystal River, Haddam Neck, Ginna, Diablo Canyon, Callaway, Vogtle, Waterford, Duane Arnold, Farley, Robinson, Clinton, South Texas, San Onofre, Cook, Comanche Peak, Yankee Rowe, Maine Yankee, Quad Cities, Humboldt Bay, La Crosse, Big Rock Point, Rancho Seco, Zion, Midland, Bellefonte, Fort Calhoun, FitzPatrick, McGuire, LaSalle, Fort Saint Vrain, Shoreham, Satsop, Trojan, Atlantic Nuclear Power Plant
Issue date:	06/26/1997
From:	Kane W, Slosson M NRC/NMSS/SFPO, Office of Nuclear Reactor Regulation
To:
References
IN-97-039, NUDOCS 9706260174
Download: ML031050510 (11)

UNITED STATES

NUCLEAR REGULATORY COMMISSION

OFFICE OF NUCLEAR MATERIAL SAFETY AND SAFEGUARDS

OFFICE OF NUCLEAR REACTOR REGULATION

WASHINGTON, D.C. 20555-0001 June 26, 1997 NRC INFORMATION NOTICES 97-39: INADEQUATE 10 CFR 72.48 SAFETY

EVALUATIONS OF INDEPENDENT SPENT FUEL

STORAGE INSTALLATIONS

Addressees

All holders of operating licenses or construction permits for nuclear power reactors. All

holders of licenses for independent spent fuel storage installations (ISFSls).

Purpose

The U.S. Nuclear Regulatory Commission (NRC) is issuing this information notice to alert

addressees to inadequate safety evaluations performed under Section 72.48 of Title 10 of the

Code of Federal Regulations (10 CPR 72.48). It is expected that the recipients will review

this information notice for applicability to their facilities and consider actions, as appropriate.

However, suggestions contained in this information notice are not NRC requirements;

therefore, no specific action or written response is required.

Background

Section 72.48, "Changes, tests, and experiments," states that a holder of an ISFSI license

may make changes in the ISFSI described in the safety analysis report (SAR), may make

changes in the procedures described in the SAR, or may conduct tests or experiments not

described in the SAR, without prior NRC approval, unless the proposed change, test, or

experiment involves a change in the license conditions incorporated in the license, an

unreviewed safety question, a significant increase in occupational exposure, or a significant

unreviewed environmental impact. A proposed change is deemed to involve an unreviewed

safety question if the probability of occurrence or the consequences of an accident or

malfunction of equipment important to safety previously evaluated in the SAR may be

increased, if a possibility for an accident or malfunction of a different type than any evaluated

previously in the SAR may be created, or if the margin of safety as defined in the basis for

any technical specification is reduced. The licensee is required to maintain records of

changes made to the ISFSI, which include a written safety evaluation that provides the bases

for the determination that each change, test, or experiment does not involve an unreviewed

safety question.

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IN 97-39 June 26, 1997

Description of Circumstances

The NRC performs inspections of ISFSI licensees to verify adequate implementation of

10 CFR 72.48 requirements. Recent NRC inspections at several ISFSI licensees have

identified violations of these requirements. These violations involve failure by the

licensees to perform 10 CFR 72.48 safety evaluations when applicable and failure to

document sufficient technical justification to support safety evaluation conclusions.

A summary of inspection findings follows.

The NRC issued a violation to Arkansas Nuclear One (ANO), owned and operated by Entergy

Operations, Inc. The violation was based on an inspection conducted at ANO (Inspection

Report Nos. 50-313/96-25; 72-13/96-02). The licensee has a Part 72 general license to store

spent fuel at an ISFSI and uses the Sierra Nuclear Corporation ventilated storage cask, model VSC-24. The inspection identified 12 nonconformances that were not evaluated in

accordance with 10 CFR 72.48. The licensee considered nonconformances as one-time

changes to its SAR that did not require a safety evaluation because they were not permanent

design changes. The licensee accepted tile nunconformances as "use as-is" without

addressing 10 CFR 72.48 requirements. The inspectors concluded that the licensee did not

adequately evaluate the conditions to ensure that an unreviewed safety question did not

exist. The licensee reevaluated the identified nonconformances under 10 CFR 72.48 requirements and did not identify any unreviewed safety questions. The licensee also

reevaluated nine other nonconformances that had originally been resolved as "use as-is" and

did not identify any unreviewed safety questions. In addition, the licensee revised its

procedures to require an engineering evaluation for a "use as-is" determination, which

requires a 10 CFR 72.48 review, when applicable.

The NRC issued a violation to Point Beach Nuclear Plant (PBNP), owned and operated by

Wisconsin Electric Power Company. The violation was based on inspections conducted at

PBNP (Inspection Report Nos. 50-266/301-95008; 50-266/301-95014). The licensee has a

Part 72 general license to store spent fuel at an ISFSI and uses the VSC-24 cask. The

inspections identified several one-time dimensional changes and other changes that either

should have received a 10 CFR 72.48 safety evaluation or that did not receive an adequate

safety evaluation. In one such safety evaluation, the licensee did not adequately discuss the

geometric effects on criticality when reducing the thickness of spent fuel guide sleeves in the

VSC-24 fuel basket. The licensee subsequently revised its safety evaluation to address the

geometric effects on criticality. In another case, the licensee did not perform a safety

evaluation to allow the VSC-24 to be reflooded with a flow rate between 34 Umin (9 gpm)

and 42 Umin (11 gpm) during loading and unloading. This oversight was of concern

because the pressure transient calculation for the cask pressure was only valid for a flow rate

of 38 Umin (10 gpm) or less. The licensee subsequently revised reflooding procedures to

ensure that a flow rate of 38 Umin (10 gpm) is not exceeded. In response to the inspection

findings, the licensee screened all other design and procedure changes made to its ISFSI

and performed full 10 CFR 72.48 safety evaluations, as necessary. The licensee did not

identify any unreviewed safety questions.

IN 97-39 June 26, 1997 Another violation was issued to PBNP, in part, because the licensee failed to perform

adequate safety evaluations for two procedures. The violation was based on an augmented

inspection team inspection at PBNP (Enforcement Action [EA] 96-273 and Inspection Report

No. 50-266/301-96005). The licensee did not perform a 10 CFR 72.48 safety evaluation for a

lifting evolution that created a potential for dropping the VSC-24 multi-assembly sealed

basket (MSB) transfer cask off the top of the ventilated concrete cask, an accident not

described in the SAR. The licensee also did not provide sufficient technical justification to

support conclusions in a safety evaluation for an MSB weighing procedure. The safety

evaluation did not include supporting information that ensured that the shield lid would not be

inadvertently removed from the MSB and expose the workers to spent fuel. Before lifting the

lid, the licensee developed a new weighing procedure that did not create the potential to

inadvertently remove the lid.

The NRC issued a violation to Prairie Island Nuclear Plant (PI), owned and operated by

Northern States Power Company. The violation was based on an inspection conducted at Pi

(Inspection Report Nos. 50-282/95002; 50-306195002; 72-10/95002). The licensee has a

Part 72 site-specific license to store spent fuel in the Transnuclear, Inc., metal storage cask, model TN-40. Upon initial inspection, the inspectors determined that the licensee did not

have a procedure in place for conducting 10 CFR 72.48 safety evaluations. The licensee

subsequently revised its procedures to incorporate 10 CFR 72.48 safety evaluations into its

existing 10 CFR 50.59 review process. The inspectors then reviewed a sample safety

evaluation and found that it was not adequate. The sample did not address 10 CFR 72.48 requirements on a TN-40 lifting beam, which is described in its ISFSI SAR. The licensee

believed that only a 10 CFR 50.59 safety evaluation was required because the cask-handling

equipment was used in the Auxiliary Building. The inspectors discussed with the licensee the

need to conduct a 10 CFR 72.48 safety evaluation if changes were made to any equipment

that was described in the ISFSI SAR, independent of whether the equipment was used to

handle the cask in the Auxiliary Building. The licensee subsequently performed a 10 CFR

72.48 safety evaluation on the TN-40 lifting beam.

Discussion

The NRC inspects licensees to assess the adequacy of their programs to perform safety

evaluations in accordance with 10 CFR 72.48. Holders of an ISFSI license are required by

10 CFR 72.48 to perform a safety evaluation when changing a component or a procedure

described in the ISFSI SAR and, in the case of a general licensee, the cask SAR. The

licensee's determination that a modification to an ISFSI does not involve an unreviewed

safety question provides confidence that the bases on which the NRC issued a license to

operate an ISFSI are preserved.

An NRC-licensed ISFSI that uses dry-storage casks is a passive system in which its

structural, criticality-control, thermal, and shielding performances depend on the detailed

drawings and descriptions provided as the design bases in the SAR. It is important that the

licensee perform an adequate 10 CFR 72.48 safety evaluation of any modification in the

ISFSI SAR or cask SAR, including any changes in dimensions, materials, and procedures. In

addition, one-time changes to the ISFSI SAR or the cask SAR constitute a modification that

also requires an adequate 10 CFR 72.48 safety evaluation.

IN 97-39 June 26, 1997 Related Generic Communications

IN 95-29, "Oversight of Design and Fabrication Activities for Metal Components Used in

Spent Fuel Dry Storage Systems," dated June 7, 1995.

IN 96-34, "Hydrogen Gas Ignition During Closure Welding of a VSC-24 Multi-Assembly

Sealed Basket," dated May 31, 1996.

This information notice requires no specific action or written response. If you have any

questions about the information in this notice, please contact one of the technical contacts

listed below or the appropriate Office of Nuclear Material Safety and Safeguards (NMSS)

project manager.

Marylee M. Slosson, Acting Director iliam F. Kane, Director

Division of Reactor Program Management Spent Fuel Project Office

Office of Nuclear Reactor Regulation Office of Nuclear Material

Safety and Safeguards

Technical contacts: M. Waters, NMSS

301-415-3875 E-mail: mdwl@nrc.gov

V. Hodge, NRR

301-415-1861 E-mail: cvh@nrc.gov

Attachments:

1. List of Recently Issued NMSS Information Notices

2. List of ecently Issued NRC Information Notices

k 'd4/4 J4 A

Attachment I

IN 97-39 June 26, 1997 Page 1 of I

LIST OF RECENTLY ISSUED

NMSS INFORMATION NOTICES

Information Date of

Notice No. Subject Issuance Issued to

96-53, Retrofit to Amersham 660 06/23/97 All industrial radiography

Supp. 1 Posilock Radiography licensees

Camera to Correct Incon- sistency in 10 CFR

Part 34 Compatibility

97-35 Retrofit to Industrial 06/18197 All industrial radiography

Nuclear Company (INC) licensees

IR100 Radiography Camera

to Correct'inconsistency

in 10 CFR Part 34 Compatibility

97-30 Control of Licensed 06/03/97 All material and fuel

Material During Reorgani- cycle licensees

zations, Employee- Management Disagreements, and Financial Crises

97-24 Failure of Packing Nuts 05/08/97 All U.S. Nuclear Regulatory

on One-Inch Uranium Commission licensees and

Hexafluoride Cylinder certificates authorized to

Valves handle uranium hexafluoride

in 30- and 48-inch diameter

cylinders

97-23 Evaluation and Reporting 05/07/97 All fuel cycle conversion, of Fires and Unplanned enrichment, and fabrication

Chemical Reaction Events facilities

at Fuel Cycle Facilities

97-20 Identification of Certain 04117/97 Registered users of trans- Uranium Hexafluoride portation packages for

Cylinders that do not uranium hexafluoride

Comply with ANSI N14.1 Fabrication Standards

Attachment 2 IN 97-39 June 26, 1997 LIST OF RECENTLY ISSUED

NRC INFORMATION NOTICES

Information Date of

Notice No. Subject Issuance Issued to

97-38 Level-Sensing System 06/24/97 All holders of OLs or CPs

Initiates Common-Mode for nuclear power reactors

Failure of High-Pressure- Injection Pumps

96-53, Retrofit to Amersham 660 06/23/97 All industrial radiography

Supp. 1 Posilock Radiography licensees

Camera to Correct Incon- sistency in 10 CFR Part 34 Compatibility

97-37 Main Transformer Fault 06/20/97 All holders of OLs or CPs

with Ensuing Oil Spill for nuclear power reactors

into Turbine Building

97-36 Unplanned Intakes by 06/20/97 All holders of OLs and CPs

Worker of Transuranic permits. All licensees of

Airborne Radioactive of nuclear power reactors

Materials and External in the decommissioning

Exposure Due to Inadequate stage and fuel cycle

Control of Work

97-35 Retrofit to Industrial 06/18/97 All industrial radiography

Nuclear Company (INC) licensees

IR100 Radiography Camera

to Correct Inconsistency

in 10 CFR Part 34 Compatibility

OL = Operating License

CP = Construction Permit

IN 97-39 June 26, 1997 Related Generic Communications

IN 95-29, "Oversight of Design and Fabrication Activities for Metal Components Used in

Spent Fuel Dry Storage Systems," dated June 7, 1995.

IN 96-34, "Hydrogen Gas Ignition During Closure Welding of a VSC-24 Mulffi-Assembly

Sealed Basket," dated May 31, 1996.

This information notice requires no specific action or written response. If you have any

questions about the information in this notice, please contact one of the technical contacts

listed below or the appropriate Office of Nuclear Material Safety and Safeguards (NMSS)

project manager.

original signed by S.H. Weiss for original signed by

Marylee M. Slosson, Acting Director William F. Kane, Director

Division of Reactor Program Management Spent Fuel Project Office

Office of Nuclear Reactor Regulation Office of Nuclear Material

Safety and Safeguards

Technical contacts: M. Waters, NMSS

301-415-3875 E-mail: mdwl enrc.gov

V. Hodge, NRR

301-415-1861 E-mail: cvh@nrc.gov

Attachments:

1. List of Recently Issued NMSS Information Notices

2. List of Recently Issued NRC Information Notices

I

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OFFICIAL RECORD COPY 97-39. IN

IN 97-xx

June xx, 1997 Related Generic Communications

IN 95-29, "Oversight of Design and Fabrication Activities for Metal Components Used in

Spent Fuel Dry Storage Systems," dated June 7, 1995.

IN 96-34, "Hydrogen Gas Ignition During Closure Welding of a VSC-24 Multi-Assembly

Sealed Basket," dated May 31, 1996.

This information notice requires no specific action or written response. If you have any

questions about the Information in this notice, please contact one of the technical contacts

listed below or the appropriate Office of Nuclear Material Safety and Safeguards (NMSS)

project manager.

Marylee M. Slosson, Acting Director William F. Kane, Director

Division of Reactor Program Management Spent Fuel Project Office

Office of Nuclear Reactor Regulation Office of Nuclear Material

Safety and Safeguards

Technical contacts: M. Waters, NMSS

301-415-3875 E-mail: mdwlnrc.gov

V. Hodge, NRR

301-415-1861 E-mail: cvh@nrc.gov

Attachments:

1. List of Recently Issued NMSS Information Notices

2. List of Recently Issued NRC Information Notices

OFC SFPO E SFPO C:PECB E D:DRPM E

NAME MWaters 5/08/97* WKane* AChaffee* MSlosson

VI-Iodge 5/23/97 _

DATE 5/97 05/28/97 06/03/97 I /97 l

C - COVER E = COVER & ENCLOSURE N = NO COPY

OFFICIAL RECORD COPY G:XNF97-XX.WP5

4t;441 A 0

IN 97-xx

June xx, 1997 Related Generic Communications

IN 96-34, "Hydrogen Gas Ignition During Closure Welding of a VSC-24 Multi-Assembly

Sealed Basket," dated May 31, 1996.

IN 95-29, "Oversight of Design and Fabrication Activities for Metal Components Used in

Spent Fuel Dry Storage Systems," dated June 7, 1995.

This information notice requires no specific action or written response. If you have any

questions about the information in this notice, please contact one of the technical contacts

listed below or the appropriate Office of Nuclear Material Safety and Safeguards (NMSS)

project manager.

Marylee Slosson, Acting Director William F. Kane, Director

Division of Reactor Program Management Spent Fuel Project Office

Office of Nuclear Reactor Regulation Office of Nuclear Material

Safety and Safeguards

Technical contacts: M. Waters, NMSS

(301) 415-3875 E-mail: mdwl@nrc.gov

V. Hodge, NRR

301-415-1861 E-mail: cvh@nrc.gov

Attachment: List of Recently Issued NRC Information Notices.

OFC E SFPO E C:PECB lE D:RPM

NAME MWaters 5108197- WKane* AChaffee L MSlosson

VHodge 5/23/97 */

DATE 5197 05/28/97 4 /3/97 1 /97 C = COVER E - COVER & ENCLOSURE N - NO COPY

OFFICIAL RECORD COPY G:NINF97-XX.WP5 r

IN 97-xx

May xx, 1997 Related Generic Communications

IN 96-34, "Hydrogen Gas Ignition During Closure Welding of a VSC-24 Multi-Assembly

Sealed Basket," dated May 31, 1996.

IN 95-29, "Oversight of Design and Fabrication Activities for Metal Components ed in

Spent Fuel Dry Storage Systems," dated June 7, 1995.

This information notice requires no specific action or written response. If u have any

questions about the Information in this notice, please contact one of the echnical contacts'

listed below or the appropriate Office of Nuclear Material Safety and feguards (NMSS)

project manager.

Marylee Slosson, Acting Director William .Kane, Director Division of

Reactor Program Management Spent uel Project Office

Office of Nuclear Reactor Regulation OffI of Nuclear Material

fety and Safeguards

Technical contacts: M. Waters, NMSS

(301) 415-3875 E-mail: mdwl@nrc.gov

V. Hodge, NRR

301-415-1861 E-Mail: cvh@nr ovV]:

OFC SFPO l PP0E l E lQPECB E D:DRPM i

NAME MWaters 5/08/97 WKane* AChaffee MSlosson

VHodge 5/23/97 /

DATE /97 05/28/97 I /97 i I /97 C -C9VER E - COVER & ENCLOSURE N - NO COPY

OFFICIAL RECORD COPY G:\INF97-XX.WP5

/

111_

IN 97-xx

April xx, 1997 Related Generic Communications

IN 96-34, "Hydrogen Gas Ignition During Closure Welding of a VSC-24 Multi-Assembly

Sealed Basket," dated May 31, 1996.

IN 95-29, "Oversight of Design and Fabrication Activities for Metal Components Used in

Spent Fuel Dry Storage Systems," dated June 7, 1995.

This information notice requires no specific action or written response. If you have any

questions about the information in this notice, please contact the technical contact listed

below or the appropriate Office of Nuclear Material Safety and Safeguards (NMSS) project

manager.

Marylee Slosson, Acting Director William F. Kane, Director Division of

Reactor Program Management Spent Fuel Project Office

Office of Nuclear Reactor Regulation Office of Nuclear Material

Safety and Safeguards

Technical contact: M. Waters, NMSS

(301) 415-3875 E-mail: mdwl@nrc.gov

EFCIU

Distribution:

NRC File Center PUBLIC NMSS R/F SFPO R/F LPittiglio, NMSS

EHaston V I farpe SSnanKman I-Sturz ELeeds LHaughney

SFPO Q E C:PECB_ _ hAI

SF___ E C:PECB IsE D:DRPM I _

NAME MWaters:dd:vt VHodge WaneAChaffee MSlosson

DATE 517/F/97 "//97 / /97/ / /97 1 197 _ / 97 C = COVER E = COVER & ENCLOSURE N = NO COPY

OFFICIAL RECORD COPY G:\INF97-XX.WP5