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 nonessential 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. 

ML072530077
IN 2007-32 

 Page 2 of 6 

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. 
 IN 2007-32 

 Page 3 of 6 

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 
IN 2007-32 

 Page 4 of 6 

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. 
IN 2007-32 

 Page 5 of 6 

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 crossconnected 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 
IN 2007-32 

 Page 6 of 6 

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 

Technical Contact: Dennis Morey, NMSS 

 301-492-3112 

 E-mail: dcm@nrc.gov

Enclosure: 

List of Recently Issued FSME/NMSS Generic Communications 
IN 2007-32 

 Page 6 of 6 

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 

Technical Contact: 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
 Enclosure 

 IN 2007-32 

Page 1 of 3

Recently Issued FSME/NMSS Generic Communications

Date GC No. Subject Addressees 

02/02/07 IN-07-03 Reportable Medical Events Involving 

Patients Receiving Dosages of 

Sodium Iodide Iodine-131 less than 

the Prescribed Dosage Because of 

Capsules Remaining in Vials after 

Administration 

All U.S. Nuclear Regulatory Commission 

medical use licensees and NRC Master 

Materials Licensees. All Agreement State 

Radiation Control Program Directors and 

State Liaison Officers. 

02/28/07 IN-07-08 Potential Vulnerabilities of Timereliant Computer-based Systems 

Due to Change in Daylight Saving 

Time Dates 

All U. S. Nuclear Regulatory Commission 

licensees and all Agreement State 

Radiation Control Program Directors and 

State Liaison Officers. 

03/13/07 IN-07-10 Yttrium-90 Theraspheres® and 

Sirspheres®

 Impurities 

All U.S. Nuclear Regulatory Commission 

(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 

Evaluate Criticality-related Process 

Upsets at Fuel Cycle Facilities 

All licensees authorized to possess a 

critical mass of special nuclear material. 

05/02/07 IN-07-16 Common Violations of the Increased 

Controls Requirements and Related 

Guidance Documents 

All licensees who are implementing the 

U.S. Nuclear Regulatory Commission 

(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 

and Counterfeit Notices 

All holders of operating licenses for nuclear 

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. 
 Enclosure 

 IN 2007-32 

Page 2 of 3

Date GC No. Subject Addressees 

 IN-07-23 Inadvertent Discharge of Halon 

1301Fire-suppression System from 

Incorrect and/or Out-of-date 

Procedures 

All holders of operating licenses for nuclear 

power reactors, except those who have 

permanently ended operations and have 

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 

Committee on the Medical Use of 

Isotopes For Consideration to 

Improve Compliance With Sodium 

Iodide I-131 Written Directive 

Requirements in 10 CFR 35.40 and 

Supervision Requirements in 10 

CFR 35.27 

All U.S. Nuclear Regulatory Commission 

(NRC) medical-use licensees and NRC 

Master Materials Licensees. All Agreement 

State Radiation Control Program Directors 

and State Liaison Officers. 

08/13/07 IN-07-26 Combustibility of Epoxy Floor 

Coatings at Commercial Nuclear 

Power Plants 

All holders of operating licenses for nuclear 

power reactors and fuel cycle facilities 

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 

Submitted to the U.S. Nuclear 

Regulatory Commission 

All holders of operating licenses for nuclear 

power reactors and holders of and 

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 

of NRC Regulatory Authority for 

Certain Naturally-occurring and 

Accelerator-produced Radioactive 

Material 

All NRC materials licensees, Radiation 

Control Program Directors, State Liaison 

Officers, and NRC’s Advisory Committee 

on the Medical Uses of Isotopes 

04/05/07 RIS-07-07 Clarification of Increased Controls 

for Licensees That Possess 

Collocated Radioactive Material 

During Transportation Activities 

All U.S. Nuclear Regulatory Commission 

(NRC) licensees issued NRC’s Order 

Imposing Increased Controls and all 

Radiation Control Program Directors and 

State Liaison Officers 
 

 Enclosure 

 IN 2007-32 

Page 3 of 3

Date GC No. Subject Addressees 

05/04/07 RIS-07-09 Examples of Recurring Requests for 

Additional Information (RAIs) for 10 

CFR Part 71 and 72 Applications 

All holders of, and applicants for, a: (1) 10 

CFR Part 71 certificate of compliance 

(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, 

Suppl. 1 

Availability of NRC 313A Series of 

Forms and Guidance for Their 

Completion 

All U.S. Nuclear Regulatory Commission 

(NRC) medical-use licensees and NRC 

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 

Medical-Related Generic 

Communications, Federal Register 

Notices And Newsletters 

All U.S. Nuclear Regulatory Commission 

(NRC) medical-use licensees and NRC 

Master Materials licensees. All Radiation 

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