Information Notice 2005-22, Inadequate Criticality Safety Analysis of Ventilation Systems at Fuel Cycle Facilities
| ML051890406 | |
| Person / Time | |
|---|---|
| Issue date: | 07/29/2005 |
| From: | Pierson R NRC/NMSS/FCSS |
| To: | |
| References | |
| IN-05-022 | |
| Download: ML051890406 (5) | |
UNITED STATES
NUCLEAR REGULATORY COMMISSION
OFFICE OF NUCLEAR MATERIAL SAFETY AND SAFEGUARDS
WASHINGTON, D.C. 20555
July 29, 2005 NRC INFORMATION NOTICE 2005-22:
INADEQUATE CRITICALITY SAFETY ANALYSIS
OF VENTILATION SYSTEMS AT FUEL CYCLE
FACILITIES
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 alert
addressees to a safety concern arising from inadequate criticality safety analysis of ventilation
systems at fuel cycle facilities. It is expected that recipients will review the information for
applicability to their facilities and consider actions, as appropriate, to avoid similar problems.
However, suggestions contained in this IN are not new NRC requirements; therefore, no
specific action nor written response is required.
DESCRIPTION OF CIRCUMSTANCES
Recently, two events occurred at NRC-licensed fuel cycle facilities involving the failure to
implement criticality safety controls on process off-gas or ventilation systems when minor
differences between otherwise similar systems, analyzed under a single broad criticality
analysis, were not recognized by criticality safety analysts. The first instance was noted
subsequent to a backflow event in an off-gas line from a uranium dissolver. The licensee used
a single criticality safety analysis for ventilation systems in the facility. The ventilation analysis
took credit for off-gas piping typically having either a siphon break and a drain, or two drains.
However, a concern about off-gas accumulation in an enclosed area led to a design
modification for the off-gas line on the uranium dissolver such that only one drain was in the
system. During preparation of the facility criticality safety analysis, criticality safety analysts
failed to recognize that the design difference defeated the siphon break so that double
contingency was not established.
The second instance was noted when a fuel cycle licensee observed an accumulation of
uranium dioxide powder in a high-efficiency particulate air (HEPA) filter housing where no
uranium was expected. The licensee determined that what criticality safety analysts thought
was a breathing air-ventilation system was also connected to a process off-gas line from a hood
on a uranium oxidation furnace. The licensee identified a design difference in the system in
that ventilation and off-gas lines were connected differently, as they approached the HEPA
filter, than was customary in the remainder of the plant. The licensee had several broad
criticality safety analysis packages related to ventilation and process off-gas, grouping them as
breathing air, dry off-gas, and wet off-gas. The criticality safety analysts failure to recognize
the design difference in duct connections in one part of the plant from other areas of the plant
led to the incorrect determination that the system was breathing air and criticality was not
credible. This incorrect determination resulted in the failure to implement criticality safety
controls typical for off-gas ventilation in the plant.
DISCUSSION
Under 10 CFR Parts 70 and 76, certain licensees processing, storing, or handling critical
masses of fissile material are required to analyze accident scenarios leading to criticality and
provide reliable controls to assure that inadvertent criticality events are highly unlikely. When
processes, systems, equipment, or procedures are repeated in a facility, licensees frequently
elect to combine similar processes, systems, equipment, or procedures into a single criticality
analysis. The safety concern arises when modifications resulting in minor design differences
between otherwise similar systems defeat the credited double-contingency arrangement or non- credibility determination.
In the two events described, the two licensees used a single criticality safety analysis to develop
controls for groups of ventilation and process off-gas systems that were similar in form and
function. While crafting the analyses, developing the criticality safety controls, and
implementing the credited controls, licensee criticality safety analysts failed to recognize design
differences between the systems that defeated some of the assumptions or credited controls
used in some portion of the facility.
In the first instance, a design change occurred, during construction of the system, that involved
placing an additional column into the system that effectively defeated the siphon break for the
uranium dissolver. The criticality safety review for this design change looked at the analysis for
the process, but did not consider the impact that the change would have on off-gas ventilation.
In the second instance, contractors were constructing a new facility, and criticality safety
analysts did not recognize design differences in the ventilation system.
Minor design changes during construction of new processes or facilities are common at fuel
cycle licensees and may have a subtle effect on criticality controls. Licensees should consider
actions, as appropriate, to mitigate this vulnerability. These actions could include reviewing all
criticality safety analyses that group similar systems, to assure that all assumptions regarding
the forms and functions of the systems are valid for all applications. Actions could also include
verifying that the design change review process is adequate to trigger an in-depth criticality
safety review for changes arising during construction.
The Part 70 integrated safety analysis (ISA) and the Part 76 safety analysis report (SAR)
provide an integrated approach to assure that inter-relationships between accident scenarios
and their controls are appropriately evaluated during related design and change activities.
Licensees should consider whether their ISA/SAR provides an adequate integrated review of
ventilation and related systems. This IN requires no specific action nor written response. If you have any questions about the
information in this notice, please contact the technical contact listed 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
301-415-6107 E-mail: dcm@nrc.gov
Attachment: List of Recently Issued NMSS Generic Communications This IN requires no specific action nor written response. If you have any questions about the
information in this notice, please contact the technical contact listed 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
301-415-6107 E-mail: dcm@nrc.gov
Attachment: List of Recently Issued NMSS Generic Communications
ADAMS ACCESSION #: ML051890406 OFC
FCSS/TSG
Tech ED
FCSS/TSG
FCSS
NAME
DMorey:dw
Ekraus: by fax
MGalloway
RPierson
DATE
07/ 08 /05
07/ 12 /05
07/ 20 /05
07/ 29 /05
Attachment Recently Issued NMSS Generic Communications
Date
GC No.
Subject
Addressees
07/13/05 RIS-05-13
NRC Incident Response and
the National Response Plan
All licensees and certificate
holders.
07/11/05 RIS-05-11
Requirements for Power
Reactor Licensees in
Possession of Devices
Subject to the General
License Requirements of 10
CFR 31.5
All holders of operating licenses
for nuclear power reactors and
generally licensed device
vendors.
06/10/05 RIS-05-10
Performance-Based
Approach for Associated
Equipment in 10 CFR 34.20
All industrial radiography
licensees and manufacturers and
distributors of industrial
radiography equipment.
04/18/05 RIS-05-06
Reporting Requirements for
Gauges Damaged at
Temporary Job Sites
All material licensees possessing
portable gauges, regulated under
6/23/05 IN-05-17
Manual Brachytherapy
Source Jamming
All medical licensees authorized
to possess a Mick applicator.
05/17/05 IN-05-013 Potential Non-conservative
Error in Modeling Geometric
Regions in the
Keno-v.a Criticality Code
All licensees using the Keno-V.a
criticality code module in
Standardized Computer Analyses
for Licensing Evaluation (SCALE)
software developed by Oak
Ridge National Laboratory
(ORNL)
05/17/05 IN-05-012
Excessively Large Criticality
Safety Limits Fail to Provide
Double Contingency at Fuel
Cycle Facility
All licensees authorized to
possess a critical mass of special
nuclear material.
Note: NRC generic communications may be found on the NRC public website, http://www.nrc.gov, under Electronic Reading Room/Document Collections.