Information Notice 2007-32, Out of Service Equipment Connected to in Service Process Line Results in Fissile Solution Spill at Fuel Cycle Facility

Out of Service Equipment Connected to in Service Process Line Results in Fissile Solution Spill at Fuel Cycle Facility

ML072530077
Person / Time
Issue date:	10/15/2007
From:	Pierson R NRC/NMSS/FCSS
To:
References
IN-07-032
Download: ML072530077 (10)

UNITED STATES

NUCLEAR REGULATORY COMMISSION

OFFICE OF NUCLEAR MATERIAL SAFETY AND SAFEGUARDS

WASHINGTON, D.C. 20555 October 15, 2007 NRC INFORMATION NOTICE 2007-32: OUT-OF-SERVICE EQUIPMENT CONNECTED

TO IN-SERVICE PROCESS LINE RESULTS IN

FISSILE SOLUTION SPILL AT FUEL CYCLE

FACILITY

ADDRESSEES

All licensees authorized to possess a critical mass of special nuclear material.

PURPOSE

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

addressees of a criticality safety concern regarding failure to fully disconnect out-of-service

equipment from operational equipment at fuel cycle facilities. NRC expects that licensees will

review this information and consider actions, as appropriate, to avoid similar problems.

Suggestions contained in this IN are not NRC requirements; therefore, no specific action or

written response is required.

DESCRIPTION OF CIRCUMSTANCES

An NRC licensee that operates a fuel cycle facility with high-enriched uranium (HEU)

processes, constructed a new facility to convert HEU compounds into concentrated uranyl

nitrate (UN) solution, and process the UN solution through a solvent extraction system, for final

purification. The new solvent extraction system included a tray dissolver and filter system in

two large gloveboxes. The tray dissolver glovebox system, as shown in Figure 1, was designed

to recover small amounts of uranium from various sources and return filtered UN solution

directly to the solvent extraction system.

Delays in the construction of the new facility placed schedule pressure on the licensee. The

tray dissolver glovebox system, installed during facility construction, was considered non- essential to process startup, and finalization of equipment installation was delayed. The tray

dissolver glovebox system was connected to a main HEU solution transfer line which was

connected to the solvent extraction system collection vessel. The tray dissolver glovebox

system was flagged as out-of-service but was not isolated from the main HEU solution

transfer line.

Before initiating operation of the new HEU process system, the licensee conducted reviews of

the newly constructed facility, including requiring process engineers to compare installed piping

and valves to design drawings. During this process, an engineer mistook a diverter valve in the

tray dissolver filter glovebox for a block valve, and the associated as-built drawing was changed

to reflect the error. The incorrect drawing gave the impression that the tray dissolver filter

glovebox was isolated from other process lines and the error was not detected during

subsequent readiness reviews. The diverter valve directed flow towards a sample point, and

could not have impeded solution flow in the process line.

Figure 1 Tray Dissolver Filter Glovebox Equipment Arrangement

Licensee startup procedures included hydrostatic testing of process equipment with natural

(non-enriched) UN solution. On several occasions after system startup, operators observed and

reported yellow solution in the tray dissolver filter glovebox, and these instances were attributed

to the hydrostatic testing. The solutions in the filter glovebox were assumed to be natural UN

solution and were recovered, but not sampled. After a few years of operation, the licensee decided to move the unused tray dissolver glovebox

system to a new location in the facility. UN solution was found in the system filters and

operators drained the system without a specific work procedure and thus did not sample the

solution in the filters, restore the original valve line-up, or fully re-tighten the filter cover bolts.

The next day a large HEU solution transfer took place through the transfer line to the solvent

extraction system collection vessel. Approximately 37 liters of concentrated HEU solution

spilled into the filter glovebox, through the glovebox drains, and to the floor of the facility.

Licensee operators observed the spill as it passed under a door, investigated, observed solution

spraying from the tray dissolver filters, and took corrective actions that terminated the event.

The tray dissolver filter glovebox was an unfavorable geometry configuration. To protect

against criticality in its gloveboxes, the licensee normally installed two glovebox drains

consisting of 1-inch or greater diameter tubes, and implemented controls to assure that the

drains were not blocked during operation. The tray dissolver glovebox was constructed with the

drains, but because the system was considered out-of-service, no controls were implemented to

prevent blockage of the drains. Subsequent to the spill event, the licensee discovered that tools

and cleaning material had been stored in the tray dissolver filter glovebox and had partially

obstructed one of the drains.

To protect against criticality during spill events, the licensee had surveyed the facility floor and

eliminated or directed flow away from solution collection points. During investigations

conducted immediately after the event, the licensee discovered an elevator pit, near the path of

the solution flow, which was not protected against solution ingress. The elevator pit was an

unfavorable geometry configuration.

SUBCRITICAL MARGIN

NRC considers this spill to be significant because the mass involved exceeded the minimum

critical mass for the fissile solution. A minimum critical mass is a theoretical value used to

analyze safety margin for events involving fissile material. For a solution, the minimum critical

mass is determined based on a critical reflected sphere of the solution. An approximate value

for a critical sphere can be taken from Figure 2 which is Atlantic Richfield Hanford Company

Criticality Handbook (ARH 600), Volume 2, Figure III.B.6(100)-1. The as-found solution can be

represented by UN solution containing 170 grams uranium-235 (U-235) per liter. Figure 2 shows that UN solution at 170 grams U-235 per liter has a minimum critical mass of 1.4 kilograms. 37 liters of UN solution at 170 grams U-235 per liter results in approximately 6.5 kilograms of U-235 which exceeds the minimum critical value of 1.4 kilograms.

Although the volume of spilled solution would not have attained the critical slab height at the

collection points, the mass spilled in the event would have been sufficient to sustain criticality in

a slab configuration of sufficient thickness. An approximate value for a critical slab can be taken

from Figure 3 which is ARH 600, Volume 2, Figure III.B.5(100)-1. The most likely way to reach

the critical slab would have been with additional process solution since 200 liters were available

in the transfer that led to the spill. Critical slab heights were estimated based on the actual

process solution by filling both collection points with process solution until the critical height was

determined. The estimated critical slab height in the glovebox is approximately 4.1 inches and Figure 2 Spherical Critical Mass for UN Solution

is attained with 130 liters of process solution. The estimated critical slab height in the elevator

pit is approximately 3 inches and is attained with 100 liters of solution. A critical slab height can

be determined based on the mass actually spilled by converting that mass into a concentration

based on the two estimated critical volumes.

The mass in the 37 liters spilled can be converted from 170 grams U-235 per liter in 37 liters to

48 grams U-235 per liter in 130 liters or 63 grams U-235 per liter in 100 liters. Figure 3 shows

that the minimum critical slab height at 48 grams U-235 per liter is approximately 4.2 inches and

that the minimum critical slab height at 63 grams U-235 per liter is approximately 3.4 inches.

This analysis demonstrates that the height of a critical slab at the collection points would not be

affected significantly by reducing the uranium concentration of the process solution. This

supports the conclusion that sufficient mass was available during the event to attain criticality if

a suitable geometry had been reached. Figure 3 Critical Slab Height versus Solution Concentration

DISCUSSION

When handling licensed material, licensees must completely understand the material flowpath.

In the above spill, the licensee lost control of configuration and did not clearly understand the

flowpath of HEU solution. No safety controls existed to preclude inadvertent criticality.

Criticality did not occur because the spilled solution did not assume a favorable configuration.

NRC is concerned that fuel cycle licensees have configuration management and start-up

procedures that detect and preclude starting a process with out-of-service equipment cross- connected to in-service equipment. NRC is also concerned that licensees use formal work

processes such as written procedures to accomplish work related to licensed activities.

The failure to develop, maintain, and fully integrate management measures, operating

procedures, and criticality, radiological, fire, and chemical controls can lead to uncontrolled

process operations as in the above spill event. NRC safety inspections typically include review of the licensee safety audit program, to ensure that analytical assumptions are regularly

reviewed in all areas. NRC safety inspections also routinely review licensee configuration

management programs, to ensure that plant changes are controlled, and design and as-built

information are updated to accurately reflect criticality safety assumptions and controls.

This information notice does not require any specific action or written response. Please direct

any questions about this matter to the technical contact below.

/RA/

Robert C. Pierson, Director

Division of Fuel Cycle Safety

and Safeguards

Office of Nuclear Material Safety

and Safeguards

Dennis Morey, NMSS

301-492-3112 E-mail: dcm@nrc.gov

Enclosure:

List of Recently Issued FSME/NMSS Generic Communications of the licensee safety audit program, to ensure that analytical assumptions are regularly

reviewed in all areas. NRC safety inspections also routinely review licensee configuration

management programs, to ensure that plant changes are controlled, and design and as-built

information are updated to accurately reflect criticality safety assumptions and controls.

This information notice does not require any specific action or written response. Please direct

any questions about this matter to the technical contact below.

/RA/

Robert C. Pierson, Director

Division of Fuel Cycle Safety

and Safeguards

Office of Nuclear Material Safety

and Safeguards

Dennis Morey, NMSS

301-492-3112 E-mail: dcm@nrc.gov

Enclosure:

List of Recently Issued FSME/NMSS Generic Communications

ML072530077 OFC FSCC/TSB Tech ED FSME FCSS/TSB FCSS FCSS

NAME D.Morey E.Kraus: fax A.McIntosh D.Jackson J.Giitter R.Pierson

DATE 9/ 20 /07 9/ 12 /07 9/ 11 /07 9/ 25 /07 10/15/07 10/15/07 OFFICIAL RECORD COPY Recently Issued FSME/NMSS Generic Communications

Date GC No. Subject

Addressees

02/02/07 IN-07-03 Reportable Medical Events Involving All U.S. Nuclear Regulatory Commission

Patients Receiving Dosages of medical use licensees and NRC Master

Sodium Iodide Iodine-131 less than Materials Licensees. All Agreement State

the Prescribed Dosage Because of Radiation Control Program Directors and

Capsules Remaining in Vials after State Liaison Officers.

Administration

02/28/07 IN-07-08 Potential Vulnerabilities of Time- All U. S. Nuclear Regulatory Commission

reliant Computer-based Systems licensees and all Agreement State

Due to Change in Daylight Saving Radiation Control Program Directors and

Time Dates State Liaison Officers.

03/13/07 IN-07-10 Yttrium-90 Theraspheres and All U.S. Nuclear Regulatory Commission

Sirspheres Impurities (NRC) Medical Licensees and NRC Master

Materials Licensees. All Agreement State

Radiation Control Program Directors and

State Liaison Officers.

04/04/07 IN-07-13 Use of As-Found Conditions to All licensees authorized to possess a

Evaluate Criticality-related Process critical mass of special nuclear material.

Upsets at Fuel Cycle Facilities

05/02/07 IN-07-16 Common Violations of the Increased All licensees who are implementing the

Controls Requirements and Related U.S. Nuclear Regulatory Commission

Guidance Documents (NRC) Order Imposing Increased Controls

(EA-05-090), issued November 14, 2005 and December 22, 2005.

05/21/07 IN-07-19 Fire Protection Equipment Recalls All holders of operating licenses for nuclear

and Counterfeit Notices power reactors and fuel cycle facilities;

except those licensees for reactors that

have permanently ceased operations and

who have certified that fuel has been

permanently removed from the reactor

vessel; and except those licensees for

decommissioned fuel cycle facilities.

06/11/07 IN-07-20 Use of Blank Ammunition All power reactors, Category I fuel cycle

facilities, independent spent fuel storage

installations, conversion facility, and

gaseous diffusion plants. Date GC No. Subject

Addressees

IN-07-23 Inadvertent Discharge of Halon All holders of operating licenses for nuclear

1301Fire-suppression System from power reactors, except those who have

Incorrect and/or Out-of-date permanently ended operations and have

Procedures certified that fuel has been permanently

removed from the reactor vessel. All

holders of licenses for fuel cycle facilities.

07/19/07 IN-07-25 Suggestions from the Advisory All U.S. Nuclear Regulatory Commission

Committee on the Medical Use of (NRC) medical-use licensees and NRC

Isotopes For Consideration to Master Materials Licensees. All Agreement

Improve Compliance With Sodium State Radiation Control Program Directors

Iodide I-131 Written Directive and State Liaison Officers.

Requirements in 10 CFR 35.40 and

Supervision Requirements in 10

CFR 35.27

08/13/07 IN-07-26 Combustibility of Epoxy Floor All holders of operating licenses for nuclear

Coatings at Commercial Nuclear power reactors and fuel cycle facilities

Power Plants except licensees for reactors that have

permanently ceased operations and who

have certified that fuel has been

permanently removed from the reactor

vessel.

03/01/07 RIS-07-03 Ionizing Radiation Warning Symbol All U.S. Nuclear Regulatory Commission

licensees and certificate holders. All

Radiation Control Program Directors and

State Liaison Officers

03/09/07 RIS-07-04 Personally Identifiable Information All holders of operating licenses for nuclear

Submitted to the U.S. Nuclear power reactors and holders of and

Regulatory Commission applicants for certificates for reactor

designs. All licensees, certificate holders, applicants, and other entities subject to

regulation by the U.S. Nuclear Regulatory

Commission (NRC) of the use of source, byproduct, and special nuclear material

03/20/07 RIS-07-05 Status and Plans for Implementation All NRC materials licensees, Radiation

of NRC Regulatory Authority for Control Program Directors, State Liaison

Certain Naturally-occurring and Officers, and NRCs Advisory Committee

Accelerator-produced Radioactive on the Medical Uses of Isotopes

Material

04/05/07 RIS-07-07 Clarification of Increased Controls All U.S. Nuclear Regulatory Commission

for Licensees That Possess (NRC) licensees issued NRCs Order

Collocated Radioactive Material Imposing Increased Controls and all

During Transportation Activities Radiation Control Program Directors and

State Liaison Officers Date GC No. Subject

Addressees

05/04/07 RIS-07-09 Examples of Recurring Requests for All holders of, and applicants for, a: (1) 10

Additional Information (RAIs) for 10 CFR Part 71 certificate of compliance

CFR Part 71 and 72 Applications (CoC) for a radioactive material

transportation package; (2) 10 CFR Part 72 CoC for a spent fuel storage cask; and (3)

10 CFR Part 72 specific license for an

independent spent fuel storage installation

(ISFSI).

06/27/07 RIS-06-27, Availability of NRC 313A Series of All U.S. Nuclear Regulatory Commission

Suppl. 1 Forms and Guidance for Their (NRC) medical-use licensees and NRC

Completion Master Materials licensees. All Radiation

Control Program Directors and State

Liaison Officers.

05/15/07 RIS-07-10 Subscriptions To New List Server

For Automatic Notifications Of All U.S. Nuclear Regulatory Commission

Medical-Related Generic (NRC) medical-use licensees and NRC

Communications, Federal Register Master Materials licensees. All Radiation

Notices And Newsletters Control Program Directors and State

Liaison Officers.

Note: A full listing of generic communications may be viewed at the NRC public website at the following address:

http://www.nrc.gov/Electronic Reading Room/Document Collections/Generic Communications