ML20211P035
| ML20211P035 | |
| Person / Time | |
|---|---|
| Site: | 07001113 |
| Issue date: | 09/09/1999 |
| From: | Vaughan C GENERAL ELECTRIC CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM), NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS) |
| References | |
| CMV-99-045, NUDOCS 9909130125 | |
| Download: ML20211P035 (60) | |
Text
,
l GENuclearEnergy GeneralDectnc Campany PO Ba 780 Wilnunginn NC:8402 9106lb b000 September 9,1999 Director Office of Nuclear Material Safety & Safeguards U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555 0001
Subject:
License Revision to Chap
- er 6.0 - Criticality Safety
Reference:
NRC License SNM-1097, Decket 70-1113
Dear Sir:
GE's Nuclear Energy Production (NEP) facility in Wilmington, N.C. hereby submits a revision to the criticality safety program requirements in Chapter 6.0 and Table 6.0 of our Consolidated Application, currently Revision 2. This submittal is made in accordance with Section 1.3.1.2 of the Application.
There are three types of changes being made in this submittal:
(1) Modifying the nomenclature for Process Area and Process Subarea consistent with current configuration descriptions.
(2) Changing one of the parameter controls for IIEPA filters in U 0s ystems from " mass" to " geometry" 3
s (effective 6/10/99).
(3) Rearrangement of selected REDCAP equipment to facilitate Gad Scrap Recycle (GSR)- a subarea of the Uranium Recovery Unit.
An Integrated Safety Analysis (ISA) was conducted for items 2 and 3. A copy of the ISA Summaries is included in this submittal.
l Attachment I provides a description of the changes by section and page. provides the Revisions By Chapter, Table of Contents and Chapter 6.0 in its entirety. All page changes in Chapter 6.0 of this revision have been dated 9/9/99 and identified as Revision 3.
Vertical lines ( l ) in the right hand column indicate where changes have been made. provides an Integrated Safety Analysis (ISA) of the Dry Scrap Recycle U 0s HEPA Filter 3
Modification, i provides an Integrated Safety Analysis (ISA) of the GAD Scrap Recycle.
..m J fJ Po '
f 1
99o,13o12s 99o9o, I
PDR ADOCK o7oo1113 C
PDR b
f-E.
Director, NMSS September 9,1999 Page 2 of 2 l
Six copies of this submittal are being provided for your use.
Please contact me on (910) 675-5656 or Rick Foleck on (910) 675-6299, if you have any questions or would like to discuss this matter further.
Sincerely, GENUCLEAR ENE GY c $$f.
M Gem Charles M.Va ghan Manager Facility Licensing
/zb Enclosure cc:
CMV-99-045 H. M. Astwood, NRC-HQ D. A. Ayres, NRC-Atlanta i
1
)
i f
l
)
Director, NMSS j
Septembcr 9,1999 Page 1 of 2 ATTACllMENT 1 Description of Revisions Page(s)
Section(s)
Description 1
Revisions by Chapter Changed the date on the Revisions By Chapter, Table of l
Contents and Chapter 6.0 to 9/9/99.
2 Table of Contents Changed the page number for Section 6.3 from 6.29 to 6.31 and Section 6.4 from 6.36 to 6.39.
6.8 6.2.3 Modified to be consistent with the factory configuration post ADU and Fluoride Waste process shutdown.
Pages 6.9 - 6.22 Table 6.0 Changed areas and subarcas in all of Table 6.0 to be consistent with the current facility con 0guration management designation.
6.9 Table 6.0 Included new decon sort table and sump trench. These operations have always been there, but are added for clarity.
6.10 Table 6.0 Corrected "liomogeneous UNil" to " Homogeneous ADU".
6.11 Table 6.0 The HVAC Wet Area and all DVRF areas (last four blocks on this page) were moved from pages 6.10 and 6.20 respectively of Revision 2 of Chapter 6.0.
6.12 Table 6.0 In Revision 2, some of the URU subareas were included in the Chemical Area and some in URU. For purposes of clarity, the subareas for URU are all identiDed in one location.
As the result of Reportable 99-01 made June 2,1999, an investigation (root cause and ISA) was performed and corrective actions identiDed. Results of this change are reDected where a stronger control," geometry", is substituted for " mass" in the DSR-U30 llEPA system.
6.14 Table 6.0 Removed the word " Surge" from Nitrate Waste Surge Vessel. Also, added ilVAC-Wet Areas and HVAC-Dry Areas to URU for clarity.
t l
l L
h, Director, NMSS September 9,1999 1
Page 2 of 2 I
ATTACHMENT 1 Description of Revisions Page(s)
S_ection(s)
Description I
6.15 Table 6.0
.Added to UF Cylinder Receipt and Storage the wording 6
" Geometry Single Planer Array" since we limit storage to a single planer layer. Revised H O equivalent to be 2
expressed in terms of HF, which is consistent with the J
neutronics model.
Added " General: HVAC" for clarity.
6.17 Table 6.0 Added new Area titled " Fabrication". Added new subarea,
" General: HVAC-Dry Areas". The subareas did not change from the Revision 2 version on page 6.17. The change is for clarity.
6.20 Table 6.0 Added " General: HVAC - Dry Areas" for clarity.
The " Ground: Offline Granulator" subarca is a new location for the MSG previously operated in the Chemical Area. There are no changes in operation.
The rotary slugger has been broken into two sections (upper and lubricant sump) to clarify the controls. See the rotary slugger found on page 6.19 of Revision 2. Also, the granulator was on page 6.19 of Revision 2.
6.21 Table 6.0 Removed the 5-gallon feed product container from MEZZ-MRA on page 6.19 of Revision 2.
1 In the first and third box, corrected " Heterogeneous UO "
2 to " Homogeneous UO ".
2 The last seven blocks on this page are a combination of new entries and a move from page 6.11, Revision 2 of some REDCAP items. The existing REDCAP equipment has been reconfigured to better integrate into the Gad Scrap l
Recycle (GSR) process.
e
F.
Director, NMSS September 9,1999 Page1ofI l
ATTACHMENT 2 Table of Contents, Revisions By Chapter and Chapter 6.0 in its entirety Changes in this chapter are indicated with a vertical line ( { ) in the right hand column.
i l
I I
REVISIONS BY CHAPTliR Application Application Page Date Page Date l
TABLE OF CONTENTS l
l CHAPTER 6.0 l
1 through 3 09/09/99 l
1 through 39 09/09/99 l
l REVISIONS BY CHAPTER l CHAPTER 7.0 l
1 09/09/99 l
1 through 3 06/05/97 l
CHAPTER 1.0 l
l CHAPTER 8.0 l
1 through 22 05/13/98 1 through 5 06/05/97 l
CHAPTER 2.0 l
l CHAPTER 9.0 l
1 through 11 03/10/98 1
06/05/97 l
CHAPTER 3.0 l
l CHAPTER 10.0 l
1 through 12 06/05/97 1 through 16 06/05/97 l
CHAPTER 4.0 l
l CHAPTER 11.0 l
1 through 8 06/05/97 1
06/05/97 l
CHAPTER 5.0 l
l APPENDIX l
1 through 13 08/06/99 1
06/05/97 LICENSE SNM-1997 DATE 09/09/99 Page DOCKET 70-1113 REVISION 8
1
TABLE OF CONTENTS Section Title Page REVISIONS BY CHAPTER 1
CHAPTER 1.0 GENERAL INFORMATION 1.1 Facility and Process Description 1.1 1.2 InstitutionalInformation 1.7 1.3 Special Authorizations and Exemptions 1.10 CHAPTER 2.0 ORGANIZATION AND ADMINISTRATION 2.1 Policy 2.1 2.2 Organizational Responsibilities and Authority 2.1 1
2.3 Safety Committees 2.10 CH APTER 3.0
]
CONDUCT OF OPERATIONS 3.1 Configuration Management (CM) 3.1 3.2 Maintenance 3.2 3.3 Quality Assurance (QA) 3.4 3.4 Training and Qualification 3.6 3.5 Human Factors 3.7 3.6 Audits and Assessments 3.7 3.7 Incident Investigations 3.9 3.8 Records Management 3.10 3.9 Procedures 3.11 CH APTER 4.0 INTEGRATED SAFETY ANALYSIS 4.1 Integrated Safety Analysis 4.1 4.2 Site Description 4.1 4.3 Facility Description 4.1 LICENSE SNM-1097 DATE 09/09/99 Page DOCKET 70-1113 REVISION 2
1 I
TAHLE OF CONTENTS Section Title Page 4.4 Process Description 4.2 4.5 Process Safety Information 4.2 4.6 Training and Qualifications of the ISA Team 4.2 4.7 ISA Methods 4.2 4.8 Results of the ISA 4.3 4.9 Controls for Prevention and Mitigation of Accidents 4.4 4.10 Administrative Control of the ISA 4.7 i
CHAPTER 5.0 -
RADIATION SAFETY 5.1 ALARA (As Low As is Reasonably Achievable) Policy 5.1 5.2 Radiation Safety Procedures and Radiation Work Permits (RWPS) 5.1 5.3 Ventilation Requirements 5.2 5.4 Air Sampling Program 5.3 5.5 Contamination Control 5.5 5.6 External Exposure 5.7 5.7 Internal Exposure 5.8 5.8 Summing Internal and External Exposure 5.9 5.9 Action Levels for Radiation Exposures 5.9 5.10 Respiratory Protection Program 5.9 5.11 Instrumentation 5.10 CHAPTER 6.0 NUCLEAR CRITICALITY SAFETY 6.1 Program Administration 6.1 6.2 Technical Practices 6.5 6.3 Control Documents 6.31 6.4 Criticality Accident Alarm System 6.39 CHAPTER 7.0 CHEMICAL SAFETY 7.1 Chemical Safety Program 7.1 LICENSE SNM-1097 DATE 09/09/99 Page DOCKET 70-1113 REVISION 2
2 f
TABLE OF CONTENTS Section Title Page 7.2 Contents of Chemical Safety Program 7.1 CH APTER 8.0 FIRE SAFETY 8.1 Fire Protection Program Responsibility 8.1 8.2 Fire Protection Program 8.1 8.3 Administrative Controls 8.2 8.4 Building Construction 8.2 8.5 Ventilation Systems 8.3 8.6 Process Fire Safety 8.3 8.7 Fire Detection and Alarm Systems 8.3 8.8 Fire Suppression Equipment 8.4 8.9 Fire Protection Water System 8.4 8.10 Radiological Contingency and Emergency Plan (RC&EP) 8.5 8.11 Emergency Response Team 8.5 CHAPTER 9.0 RADIOLOGICAL CONTINGENCY AND EMERGENCY PLAN 9.1 CHAPTER 10.0 ENVIRONMENTAL PROTECTION 10.1 Air Effluent Controls and Monitoring 10.1 10.2 Liquid Treatment Facilities 10.1 10.3 Solid Waste Management Facilities 10.2 10.4 Program Documentation 10.2 10.5 Evaluations 10.3 10.6 Off-site Dose 10.3 10.7 ALARA 10.4 CHAPTER 11.0 DECOMMISSIONING 11.1 APPENDIX A.1 LICENSE SNM-1997 DATE 09/09/99 Page DOCKET 70-1113 REVISION 2
3
l
(
CHAPTER 6.0 NUCLEAR CRITICALITY SAFETY 6.1 PROGRAM ADMINISTRATION 6.1.1 CRITICALITY SAFETY DESIGN PHILOSOPHY l
The Double Contingency Principle as identified in nationally recognized American National Standard ANSI /ANS-8.1 (1983) is the fundamental technical basis for design and operation of processes within the GE-Wilmington fuel manufacturing operations using fissile materials. As such," process designs will incorporate sufficient margins of safety to require at least two unlikely, independent, and concurrent changes in process conditions before a criticality accident is possible."
For each significant portion of the process, a defense of one or more system parameters is documented in the criticality safety analysis, which is reviewed and enforced.
The established design criteria and nuclear criticality safety reviews are applicable to:
all new processes, facilities or equipment that process, store, transfer or e
otherwise handle fissile materials, and any change in processes, facilities or equipment which may have an impact e
on the established basis for nuclear criticality safety.
6.1.2 EVALUATION OF CRITICALITY SAFETY 6.1.2.1 Changes to Facility As part of the design of new facilities or significant additions or changes in existing facilities, Area Managers provide for the evaluation of nuclear hazards, chemical hazards, hydrogenous content of firefighting materials, and mitigation ofinadvertent unsafe acts by individuals. Specifically, when criticality safety considerations are i
impacted by these hazards, the approval to operate new facilities or make significant
{
changes, modification, or additions to existing facilities is documented in accord l
l i
LICENSE SNM-1097 DATE 09/09/99 Page l
DOCKET 70-1113 REVISION 3
6.1 i
l l
l l
a with established facility practices and conform to configuration management function ' Integrated Safety Analysis' (ISA) requirements described in Chapter 4.0.
Change requests are processed in accordance with configuration management requirements described in Chapter 3.0 Change requests which establish or involve a change in existing criticality safety parameters require a senior engineer who has been approved by the criticality safety function to disposition the proposed change with respect to the need for a criticality safety analysis.
If an analysis is required, the change is not placed into operation until the criticality safety analysis is complete and other preoperational requirements are fulfilled in accordance with established configuration management practices.
6.1.2.2 Role of the Criticality Safety Function Qualified personnel as described in Chapter 2 assigned to the criticality safety function determine the basis for safety for processing fissile material. Assessing both normal and credible abnormal conditions, criticality safety personnel specify functional requirements for criticality safety controls commensurate with design criteria and assess control reliability. Responsibilities of the criticality safety function are described in Chapter 2.0.
6.1.3 OPERATING PROCEDURES Procedures that govern the handling of enriched uranium are reviewed and approved by the criticality safety function.
Each Area Manager is responsible for developing and maintaining operating procedures that incorporate limits and controls established by the criticality safety function. Area Managers assure that appropriate area engineers, operators, and other concerned personnel review and understand these procedures through postings, training programs, and/or other written, electronic or verbal notifications.
1 Documentation of the review, approval and operator orientation process is maintained within the configuration management system. Specific details of this system are described in Chapter 3.0.
i
~
LICENSE SNM-1997 DATE 09/09/99 Page DOCKET 70-1113 REVISION 3
6.2
n 1
1 6.1.4 POSTING AND LABELING l
6.1.4.1 Posting of Limits and Controls Nuclear criticality safety requirements for each process system that are defined by the criticality safety function are made available to work stations in the form of written or electronic operating procedures, and/or clear visible postings.
Posting may refer to the placement of signs or marking of floor areas to summarize key criticality safety requirements and limits, to designate approved work and storage areas, or to provide instructions or specific precautions to personnel such as:
Limits on material types and fonus.
Allowable quantities by weight or number.
Allowable enrichments.
Required spacing between units.
Control limits (when applicable) on quantities such as moderation, density, or presence of additives.
Critical control steps in the operation.
Storage postings are located in conspicuous places and include as appropriate:
Material type.
Container identification, Number ofitems allowed.
e Mass, volume, moderation, and/or spacing limits.
Additionally, when administrative controls or specific actions / decisions by operators are involved, postings include pertinent requirements identified within the criticality safety analysis.
6.1.4.2 Labeling Where practical, process containers of fissile material are labeled such that the material type, U-235 enrichment, and gross weights can be clearly identified or determined. Deviations from this process include: large process vessels, fuel rods, LICENSE SNM-1997 DATE 09/09/99 Page DOCKET 70-1113 REVISION 3
6.3
I
)
l l
shipping containers, waste boxes / drums, contaminated items, UF cylinders 6
containing heels, cold trap cylinders, samples, containers of 1 liter volume or less, or other containers where labeling is not practical.
6.1.5 AUDITS & INSPECTIONS l
6.1.5.1 Audits and Inspections Details of the facility criticality safety audit program are described in Chapter 3.0.
Criticality safety audits are conducted and documented in accordance with a written i
procedure and personnel approved by the criticality safety function. Findings, I
recommendations, and observations are reviewed with the Environment, Health &
Safety (EHS) function manager to determine if other safety impacts exist. The findings, recommendations, and observations are then transmitted to Area Managers for appropriate action.
Routine surveillance inspections of the processes and associated conduct of operations within the facility, including compliance with operating procedures, postings, and administrative guidelines, are also conducted as described in Chapter 3.
6.1.5.2 hidependent Audits A nuclear criticality safety program review is conducted on a planned scheduled basis by nuclear criticality safety professionals independent of the GE-Wilmington fuel manufacturing organization. This provides a means for independently assessing the effectiveness of the components of the nuclear criticality safety program.
The audit team is composed ofindividuals recommended by the manager of the criticality safety function and whose audit qualifications are approved by the GE-Wilmington facility manager or Manager, EHS. Audit results are reported in writing to the manager of the criticality safety function, who disseminates the report to line management. Results in the form of corrective action requests are tracked to closure.
LICENSE SNM-1097 DATE 09/09/99 Page DOCKET 70-1113 REVISION 3
6.4
o 1
i l
6.1.6 CRITICALITY SAFETY PERSONNEL l
6.1.6.1 Qualifications Specific details of the criticality safety function responsibilities and qualification requirements for manager, senior engineer, and engineer are described in Chapter 2.0.
6.1.6.2 Authority Criticality safety function personnel are specifically authorized to perform assigned responsibilities in Chapter 2.0. All nuclear criticality safety function personnel have authority to shutdown potentially unsafe operations.
1 I
6.2 TECHNICAL PRACTICES 6.2.1 CONTROL PRACTICES Criticality safety analyses identify specific controls necessary for the safe and effective operation of a process. Prior to use in any process, nuclear criticality safety controls are verified against criticality safety analysis criteria. The ISA program described in Chapter 4.0 implement performance based management of process requirements and specifications that are important to nuclear criticality safety.
6.2.1.1 Verification Program The purpose of the verification program is to assure that the controls selected and installed fulfill the requirements identified in the criticality safety analyses. All processes are examined in the "as-built" condition to validate the safety design and to verify the installation. Criticality safety function personnel observe or monitor the performance ofinitial functional tests and conduct pre-operational audits to verify LICENSE SNM-1097 DATE 09/09/99 Page DOCKET 70-1113 REVISION 3
6.5 1
that the controls function as intended and the installed configuration agrees with the
{
criticality safety analysis.
I Operations personnel are responsible for subsequent verification of controls through the use of functional testing or verification. When necessary, control calibration and i
routine maintenance are normally provided by the instrument and calibration and/or maintenance functions. Verification and maintenance activities are performed per established facility practices documented through the use of forms and/or computer tracking systems. Criticality safety function personnel randomly review control verifications and maintenance activities to assure that controls remain effective.
l 6.2.1.2 Maintenance Program The purpose of the maintenance program is to assure that the effectiveness of criticality safety controls designated for a specific process are maintained at the original level ofintent and functionality. This requires a combination of routine maintenance, functional testing, and verification of design specifications on a periodic basis. Details of the maintenance program are described in Chapter 3.0.
6.2.2 MEANS OF CONTROL The relative effectiveness and reliability of controls are considered during the criticality safety analysis process. Passive engineered controls are preferred over all other system controls and are utilized when practical and approp-iate. Active engineered controls are the next preferred method of control followed by administrative controls. A criticality safety control must be capable of preventing a criticality accident independent of the operation or failure of any other criticality control for a given credible initiating event.
6.2.2.1 Passive Engineered Controls These are physical restraints or features that maintain criticality safety in a static manner (i.e., fixed geometry, fixed spacing, fixed size, nuclear poisons, etc.).
Passive engineered controls require no action or other response to be effective when called upon to ensure nuclear criticality safety. Assurance is maintained through i
specific periodic inspections or verification measurement (s) as appropriate.
I LICENSE SNM-1997 DATE 09/09/99 Page l
DOCKET 70-1113 REVISION 3
6.6 l
l
6.2.2.2 Active Engineered Controls A means of criticality control involving active hardware (e.g., electrical, mechanical, hydraulic) that protect against criticality. These devices act by providing predefined automatic action or by sensing a process variable important to criticality safety and providing automatic action (e.g., no human intervention required) to secure the system to a safe condition. Human intervention augmented by warning devices and interlocks that prevent continued operation may be used to sense a process variable.
Assurance is maintained through specific periodic functional testing as appropriate.
Active engineered controls are fail-safe (e.g., meaning failure of the control results in a safe condition).
6.2.2.3 Administrative Controls Controls that rely for their implementation on actions, judgment, and responsible actions of people. Their use is limited to situations where passive and active control are not practical. Administrative controls may be proactive (requiring action prior to proceeding) or reactive (proceeding unless action occurs). Proactive administrative controls are preferred. Assurance is maintained through training, experience, and audit.
6.2.3 TABLE OF PLANT SYSTEMS AND PARAMETER CONTROLS Table 6.0 identifies major process areas or support facility processes within the GE-Wilmington fuel manufacturing complex and support facilities. Table entries for each significant process item highlight the safety basis selected for the criticality safety analysis (CSA) and related worst credible contents (or bounding assumptions).
Table column definitions are presented below:
AREA OR SYSTEM: A defined functional group of processes or pieces of equipment that operate as a single un,it.
PROCESS SUBAREA OR EQUIPMENT: A defined subgroup of vessels, tanks, process and/or support equipment within an area that operate as a single unit.
I l
BASIS FOR CRITICALITY SAFETY: The controlled parameters established l
within a CSA for nuclear criticality safety for the identified process subarea or I
LICENSE SNM-1097 DATE 09/09/99 Page l
DOCKET 70-1113 REVISION 3
6.7 1
I i
r e
equipment. For multiple parameter entries, the basis for nuclear criticality safety established in the CSA may be based on the identified parameter (s), as appropriate, including the use of' coupled' parameter control (e.g., mass / moderation).
CSA BOUNDING ASSUMPTIONS: These are the values used for physical process parameters which are not directly controlled but represent the most reactive credible l
l values for the system, process subarea, or equipment under consideration. As such, i
the CSA is performed to consider all process operations and credible upsets that fall l
within this range of assumptions. For items containing no bounding assumptions, all l
process operations and credible upsets must be analyzed within the CSA. The approved CSA may limit the operation of the system to levels more conservative than those permitted by the bounding assumptions.
In Table 6.0, unless otherwise specified, the enrichment limit for all processes is 5.0 wt. % U235. When pails are used for product,3 or 5-gallon cans may be used for LoE enrichments (s 4.025 wt.% U235), while 3-gallon containers must be used for hie (> 4.025 wt.% and s 5.00 wt.% U235) material. All scrap material, with the exception ofincinerator ash, is treated as hie.
l I
i LICENSE SNM-1997 DATE 09/09/99 Page DOCKET 70-1113 REVISION 3
6.8 l
t
r Tcble 6.0 Plzt Systems end Pn meter Ccitrels AREA PROCESS SUBAREA BASIS FOR CSA OR OR EQUIPMENT CRITICALITY BOUNDING ASSUMPTIONS
(
SYSTEM SAFETY l
Fuel Support Pads: UF, Cylinder Enrichment 99.5 wt. % pure UF.
Receipt and Storage Geometry (single s 0.5 wt. % llF planar array)
OptimalInterunit H O 2
Pads: 3 and 5-gallon Geometry Homogeneous or Heterogeneous UO2 Container Storage Mass Optimalll O Moderation 2
Full Reflection Pads: Waste Box Geometry / Mass Heterogeneous UO2 Container Storage Mass Optimalll O Moderation 2
Full Reflection New Decon:
Mass Heterogeneous UO2 Waste Box Load Optimal 110 Moderation 2
Full Reflection New Decon:
Mass Homogeneous UO2 Oil Drum Load Optimal 110 Moderation 2
Full Reflection New Decon:
Mass Heterogeneous UO2 Sort Table Optimal H O Moderation 2
Full Reflection New Decon:
Geometry lieterogeneous UO2 Sump Trench Mass Optimal H O Moderation 2
Full Reflection Waste Treatment Concentration llomogeneous UO2 Facility (WTF):
Mass Optimal H O Moderation 2
Nitrate Waste Vessel Full Reflection (V-104)
WTF:
Geometry Homogeneous UO2 Centrifuge Mass Optimal 110 Moderation 2
Full Reflection
[
WTF:
Geometry / Mass Homogeneous UO2 Oberlin Filter Mass Optimal H O Moderation 2
Full Reflection WTF:
Concentration /
Homogeneous UO2 Lagoons Geometry Optimal 110 Moderation 2
Full Reflection
(
Incinerator:
Mass (Box Monitor)
Heterogeneous UO
{
2 l
Combustible Box Feed Mass (E-Gun)
Optimal H O Moderation 2
Containers Full Reflection Incinerator:
Mass (UPHOLD)
Heterogeneous UO2 l
Operation Mass (INHOLD)
Optimal 110 Moderation 2
Full Reflection l
Incinerator:
Geometry Homogeneous UO2 l
3 or 5-Gallon Ash Mass Optimal 1I 0 Moderation 2
Product Containers Full Reflection LICENSE SNM-1997 DATE 09/09/99 Page DOCKET 70-1113 REVISION 3
6.9 t.
AREA PROCESS SUBAREA BASIS FOR CSA OR OR EQUIPMENT CRITICALITY BOUNDING ASSUMPTIONS SYSTEM SAFETY j
Support (cont.'d)
Uranium Recovery of Concentration Homogeneous UO2 Lagoon Sludge Optimal H O Moderation 2
(URLS): Process Tanks Full Reflection URLS: Non-Leach Mass Homogeneous UO2 Filter Press Optimal H O Moderation 2
Full Reflection URLS: Product Waste Mass Homogeneous UO2 Container Storage Concentration Optimal H O Moderation 2
Full Reflection Chemical Cold Trap System Geometry 99.5 wt. % pure UF.
Moderation 5; 0.5 wt. % HF OptimalInterunit H O 2
General: 3 or 5-Gallon Geometry Homogeneous UO2 Product Container Wet Mass Optimal H,0 Moderation
]
Storage Full Reflection General: 3 or 5-Gallon Geometry Homogeneous UO2 Product Container Dry Mass Optimal H O Moderation 2
Storage Moderation Full Reflection Radwaste:
Geometry Homogeneous UO2 Sump Mass Optimal H O Moderation 2
Full Reflection Radwaste:
Geometry Homogeneous UO2 Slabtank Concentration Optimal II 0 Moderation 2
Full Reflection Uranium Conversion Geometry Homogeneous UO2 (UCON): UNH Feed Concentration Optimal H O Moderation 2
Tanks Full Reflection UCON: Line 5 Geometry Homogeneous ADU Precipitation Tanks Mass Optimal H O Moderation 2
Full Reflection UCON: Dewatering Geometry Homogeneous ADU or U 0, 3
Centrifugation Mass Optimal 110 Moderation 2
Full Reflection Outside Containment UCON: Clarifying Geometry Homogeneous UO2 Centrifugation Mass Optimal H O Moderation 2
Full Reflection UCON Process:
Geometry Homogeneous UO2 Calcination Geometry / Mass Optimal H O Moderation 2
Full Reflection UCON: Calciner Geometry Homogeneous UO2 Scrubber Concentration Optimal H O Moderation 2
Full Reflection
- two of any three control parameters are required for criticality safety.
l l
I LlCENSE SNM-1097 DATE 09/09/99 Page DOCKET 70-1113 REVISION 3
6.10 t
. AREA PROCESS SUBAREA BASIS FOR CSA OR OR EQUIPMENT CRITICALITY BOUNDING ASSUMPTIONS SYSTEM SAFETY Chemical (cont.'d)
UCON: UO Powder Geometry or Mass liomogeneous UO2 Mill, Slug, Granulate Moderation Optimal 110 Moderation 2
(MSG)
Full Reflection Nitrate Quarantine Geometry 11omogeneous UO 2 Efiluent Vessels Concentration Optimal 110 Moderation 2
Full Reflection General:
Geometry liomogeneous UO2 HVAC-Wet Areas Mass Optimal 110 Moderation 2
Full Reflection General:
Mass llomogeneous UO2 IIVAC - Dry Areas Moderation OptimalII 0 Moderation 2
Full Reflection Utilities: Steam, N,
Mass Backflow into large supply vessels 2
11, Dissoc. NH4, H O prevented by backflow prevention 2
2 Supply measures, physical barriers, and/or process characteristics.
Decon Volume Geometry / Mass Homogeneous UO2 Reduction Facility Mass Optimal 110 Moderation 2
(DVRF): Wash Down Full Reflection Areas, Sumps, Bag Filters DVRF:
Mass Homogeneous UO2 Dust 11og Optimal 110 Moderation 2
Full Reflection DVRF:
Geometry Homogeneous UO2 IIVAC Mass Optimal H O Moderation 2
Full Reflection DVRF:
Geometry llomogeneous UO2 3-Gallon Container Mass Optimal 110 Moderation 2
Storage Full Reflection LICENSE SNM-1997 DATE 09/09/99 Page DOCKET 70-1113 REVISION 3
6.11
r l
AREA PROCESS SUBAREA BASIS FOR CSA OR OR EQUIPMENT CRITICALITY BOUNDING ASSUMPTIONS SYSTEM SAFETY Uranium Recovery Dry Scrap Recycle Geometry 3
lieterogeneous or llomogeneous UO2 Unit (URU)
(DSR): feed container Mass J*
Optimal Interunit 110 Moderation 2
storage Moderation Full Redection DSR:
Geometry lieterogeneous UO2 Feedflood Moderation Optimal 110 Moderation 2
Full Reflection DSR:
Geometry lieterogeneous UO2 l
Furnace Moderation Optimal 110 Moderation 2
l Full Reucction DSR:
Geometry lieterogeneous UO2 Screener Moderation Optimal 1-10 Moderation 2
Full Reucction l
DSR: 3-gal. product Geometry lieterogeneous UO2 l
container storage Mass Optimal Interunit 110 Moderation 2
(oversize particle)
Moderation Full Redection DSR:
Moderation lieterogeneous UO2 Powder Outlet Maximum Credible wt. % 110 2
l (Unicone Fill)
Full Reflection DSR:
Moderation lieterogeneous UO2 Blender Maximum Credible wt % 110 2
Full Reflection DSR:
Moderation / Mass lieterogeneous UO2 DM-10 Vibromill Maximum Credible wt. % 110 2
Full Reucction DSR:
Moderation lieterogeneous UO2 Unicone Container Maximum Credible UO Density 2
Storage Maximum Credible wt % ll O 2
Optimal Interunit II 0 2
DSR:
Moderation lieterogeneous UO2 Powder Transfer Maximum Credible UO Density 2
Corridor Maximum Credible wt. % }{20 Full Reucction DSR:
Mass llomogeneous UO2 IfVAC - UO2 Moderation Optimal 110 Moderation 2
Full Reflection DSR:
Geometry llomogeneous UO2 IIVAC-U308 Moderation Optimal 110 Moderation 2
Full Reflection
- two of any three control parameters are required for critkality safety, l
LICENSE SNM-1997 DATE 09/09/99 Page j
DOCKET 70-1113 REVISION 3
6.12
]
c
=
AREA PROCESS SUBAREA BASIS FOP.
CSA OR OR EQUIPMENT CRITICALITY BOUNDING ASSUMPTIONS SYSTEM SAFETY URU (cont.'d)
Oxidize:
Geometry Heterogeneous UO2 Feed Containers Mass Optimal 110 Moderation 2
l Full Reflection Oxidize:
Geometry Heterogeneous UO2 Furnace Optimal 110 Moderation 2
Full Reflection l
Oxidize:
Geometry lieterogeneous UO2 l
Furnace Boat Dump Moderation Optimal 110 Moderation 2
Full Reflection Oxidize:
Geometry lieterogeneous UO f.
l 3-gallon Container Mass Optimal H O Moderation 2
Storage Moderation Full Reflection Oxidize:
Geometry Heterogeneous UO2 oft-Gas System Mass Optimal 110 Moderation 2
l Full Reflection SX-Stor:
Geometry Homogeneous UNil llead-End Concentrator Concentration Optimal 110 Moderation 2
Full Reflectien Dissolution:
Geometry Heterogeneous UO2 Can Feed Conveyor Mass i
Optimal H O Moderation 2
Moderation f*
Full Reflection Dissolution:
Geometry Heterogeneous UO2 Dissolvers, Pumps, Concentration / Mass Optimal H O Moderation 2
Sumps, Fiiters, Piping Full Reflection Dissolution:
Geometry Heterogeneous UO2 Counter-Current Mass / Moderation Optimal 110 Moderation 2
Can Dump Full Reflection Dissolution:
Geometry Heterogeneous UO2 Counter Current Concentration Optimal H O Moderation 2
i Process Tanks Full Reflection Dissolution:
Geometry Heterogeneous UO2 l
Oberlin Filter Concentration Optimal H O Moderation 2
Full Reflection l
Dissolution: NOX Concentration Homogeneous UO2 Scrubber Mass Optimal 110 Moderation 2
Full Reflection l
i
- two of any three control parameters are required for criticality safety.
l l
i l
LICENSE SNM-1997 DATE 09/09/99 Page DOCKET 70-1113 REVISION 3
6.13
I
\\
l l-AREA PROCESS SUBAREA BASIS FOR CSA l
OR OR EQUIPMENT CRITICALITY BOUNDING ASSUMPTIONS l
SYSTEM SAFETY j
URU (cont.'d)
Solvent Extraction:
Geometry llomogeneous UO2 Process Tanks Concentration Optimal H O Moderation 2
l Full Reflection UNH Product Storage Geometry Homogeneous UNH Vessels Concentration Optimal H O Moderation 2
Full Reflection Waste Solvent Drur.
Mass Homogeneous UO2 Load Optimal 110 Moderation 2
Full Reflection Radwaste Process Geometry Homogeneous UO2 Vessels Concentration Optimal H O Moderation 2
Full Reflection Nitrate Waste Process Geometry Homogeneous UO2 Vessels Concentration Optimal H O Moderation 2
Full Reflection Cross Flow Filters Geometry llomogeneous UO2 (CFF)
Optimal H O Moderation 2
Full Reflection Centrifugation Geometry llomogeneous UO2 Mass Optimal H O Moderation 2
Full Reflection Outside Containment Nitrate Waste Concentration Homogeneous UO2 Vessel Mass Optimal H O Moderation 2
(V-103)
Full Reflection Utilities: Steam, DI Mass Backflow into large supply vessels H 0 Nitric Acid, prevented by backflow prevention 2
Aluminum Nitrate, measures, physical barriers, and/or Lime process characteristics.
General:
Geometry Homogeneous UO2 HVAC-Wet Areas Mass Optimal 110 Moderation 2
Full Reflection General:
Mass Homogeneous UO2 HVAC-Dry Areas Moderation Optimal H O Moderation 2
Full Reflection General:
Geometry llomogeneous UO2 Sumps Concentration / Mass Optimal H O Moderation 2
Full Reflection General:
Geometry / Mass Homogeneous UO2 FMOX Exhaust Mass Optimal H O Moderation 2
Scrubber Full Reflection LICENSE SNM-1997 DATE 09/09/99 Page DOCKET 70-1113 REVISION 3
6.14
1 AREA PROCESS SUBAREA BASIS FOR CSA OR OR EQUIPMENT CRITICALITY BOUNDING ASSUMPTIONS SYSTEM SAFETY Dry Conversion General:
Enrichment 99.5 wt. % pre UF.
Process (DCP)
UF. Cylinder Receipt Geometry s 0.5 wt. % HF and Storage OptimalInterunit 110 2
General:
Moderation liomogeneous UO2 IIVAC Maximum Credible UO Density 2
Maximum Credible wt. % 110 2
Full Reflection Convert:
Moderation 99.5 wt. % pure UF.
Vaporization Autoclave s 0.5 wt. % 110 equivalent 2
w/UF. Cylinder Full Reflection Convert:
Geometry liomogeneous UO2 Cold Trap System Moderation Optimal 110 Moderation 2
Full Reflection Convert:
Moderation liomogeneous UO2 Reactor / Kiln Maximum Credible UO Density 2
Maximum Credible wt. % l-10 2
Full Reflection Convert:
Moderation liomogeneous UO2 Recycle Maximum Credible UO Density 2
Maximum Credible wt % 110 2
Full Reflection Convert:
Moderation liomogeneous UO2 Powder Outlet Maximum Credible UO Density 2
Outlet Box Maximum Credible wt. % 110 2
Full Reflection Convert:
Moderation liomogeneous UO2 Powder Outlet Maximum Credible UO Density 2
Cooling flopper Maximum Credible wt. % 110 2
Full Reflection Blend / Pack / Granulate Moderation llomogeneous UO2 (BPG):
Maximum Credible UO Density 2
llomogenization, Maximum Credible wt. % 110 2
Sifter Full Reflection BPG:
Moderation lieterogeneous UO2
- Blend, Maximum Credible UO Density 2
Precompact, Maximum Credible wt. % 110 2
Granulate Full Reflection i
BPG:
Moderation Hemogeneous UO2 Tumble Maximum Credible UO Density 2
Maximum Credible wt. % 110 2
Full Reflection LICENSE SNM-1097 DATE 09/09/99 Page DOCKET 70 1113 REVISION 3
6.15
AREA PROCESS SUBAREA BASIS FOR CSA OR OR EQUIPMENT CRITICALITY BOUNDING ASSUMPTIONS SYSTEM SAFETY DCP (cont.'d)
Pwdr-Transfer:
Moderation liomogeneous UO Powder Container Maximum Credible UO Density 2
Maximum Credible wt. % 110 2
Full Reflection Pwdr Transfer:
Moderation liomogeneous UO2 Container Storage Geometry Maximum Credible UO Density 2
Array Maximum Credible wt. % 110 2
1 Full Reflection Pwdr Transfer:
Moderation llomogeneous UO2 Powder Pack Maximum Credible UO Density 2
Maximum Credible wt %II 0 2
Full Reflection Pwdr-Transfer:
Geometry 11omogeneous UO2 Pack 3-gallon Mass Optimal 110 Moderation 2
Product Container Moderation Full Reflection Utilities: N.11. II 0 Mass Backflow into large supply vessels not 2 2 2
Supply, Refrigerant credible due to backflow prevention measures, physical barriers, and'or process characteristics.
IlF:
Particulate:
llomogeneous UO i
2 ilF Effluent Recovery Geometry Optimal 110 Moderation 2
and Storage Vessels Mass Full Reflection Unreacted UF. Gas:
Mass
- two of any three control parameters are required for criticality safety.
LICENSE SNM-1997 DATE 09/09/99 Page DOCKET 70-1113 REVISION 3
6.f 6
i AREA PROCESS SUBAREA BASIS FOR CSA OR OR EQUIPMENT CRITICALITY BOUNDING ASSUMPTIONS SYSTEM SAFETY Fabrication General:
Mass Homogeneous UO2 IlVAC-Dry Areas Moderation Optimal 110 Moderation 2
Full Reuection
]
Powder Moderation Moderation liomogeneous UO l
2 i
Restriction Area Maximum Credible UO Density 2
(PWDR-MRA):
Maximum Credible wt. % 110 2
Powder Transfer Full Reflection Corridor PWDR MRA:
Geometry / Mass 11omogeneous UO2 Container Storage Moderation Maximum Credible UO Denrity 2
Maximum Credible wt. % 110 2
)
Full Reflection PWDR-MRA:
Moderation liomogeneous UO2 Press Feed Maximum Credible wt. % 110 2
Full Reflection Powder:
Geometry 1
liomogeneous UO2 3 or 5-Gallon Product Mass J*
Optimal Interunit 110 Moderation 2
Container Storage Moderation Full Reflection Powder:
Geometry liomogeneous UO2 Dump liood Moderation Optimal 110 Moderation 2
Full Reflection Press:
Geometry / Mass lieterogeneous UO2 Rotary-Press Moderation Optimal 110 Moderation 2
Full Reflection Press:
Geometry lieterogeneous UO 2
Lubricant Sump Mass Optimal 110 Moderation 2
Full Reflection Press:
Geometry lieterogeneous UO, Green Pellet Boat Moderation Optimaill O Moderation 2
Full Reflection Press: 3-gallon scrap Geometry lieterogeneous UO2 container Mass Optimalil O Moderation 2
Full Reflectio'.1 Furnace:
Geometry lieterogeneot s UO2 Feed / Exit Conveyor Moderation Optimalll O Moderation 2
Storage Full Reflection Furnace:
Geometry lieterogenemis UO, Sintering Furnace Moderation Optimal 110 Mc.ieration 2
Full Reflection
- two of any thret control parameters are required for criticality safety.
LICENSE SNM-1997 DATE 09/09/99 Page DOCKET 70-1113 REVISION 3
6.17
(
l AREA PROCESS SUHAREA BASIS FOR CSA OR OR EQUIPMENT CRITICALITY BOUNDING ASSUMPTIONS SYSTEM SAFETY Fabrication (cont.'d)
Grind:
Geometry lieterogeneous UO2 Feeder Bowl / Table Moderation Optimal 110 Moderation 2
Full Reflection Grind:
Geometry lieterogeneous UO, Grinder Moderation Optimal 110 Moderation 2
Full Reflection i
Grind:
Geometry liomogeneous UO l
2 APITRON Moderation Optimal 110 Moderation 2
Filter Full Reflection Grind:
Geometry Heterogeneous UO2 Scrap 3-Gallon Moderation Optimal 110 Moderation 2
Container (swarf)
Full Reflection Grind:
Geometry lieterogeneous UO i
2 l
Scrap 3-Gallon Mass Optimal H O Moderation 2
Container (hardscrap)
Full Reflection Grind:
Geometry lieterogeneous UO2 f,
Pellet Tray Mass Optimal H O Moderation 2
Moderation Full Reflection Grind:
Geometry Heterogeneous UO Pellets 2
Pellet Tray Moderation OptimalInterunit H O Moderation 2
Transfer Cart Full Reflection Grind:
Geometry lieterogeneous UO Pellets 2
Beaker Cart Moderation Optimal Intertmit H O Moderation 2
Full Reflection Rod Load:
Geometry lieterogeneous UO, Pellets / Rods Rod Load, Out-Gas, Moderation Optimal Interunit H O Moderatica 2
Final Weld Full Reflection Rod Load:
Geometry Heterogeneous UO Pellets 2
Pellet Storage Cabinet Moderation Optimal interunit 110 Moderation 2
Full Reflection Rod Load:
Geometry Heterogeneous UO Rods 2
Rod Storage Cabinet Moderation Optimal Interunit H O Moderation 2
Full Reflection Assembly:
Geometry Heterogeneous UO Rods
~
2 Rod Trays Mass Optimal H O Moderation 2
Full Reflection Assembly:
Geometry lieterogeneous UO Rods 2
Rod Storage Cabinets Moderation Optimal Interunit H O Moderation 2
Full Reflection
- two of any three control parameters are required for criticality safety.
l i
LICENSE SNM-1097 DATE 09/09/99 Page l
DOCKET 70-1113 REVISION 3
6.18 L
1 1
AREA PROCESS SUBAREA BASIS FOR CSA OR OR EQUIPMENT CRITICALITY BOUNDING ASSUMPTIONS SYSTEM SAFETY Fabrication (cont'd)
Assembly:
Geometry Heterogeneous UO Rods 2
Rod Tray Transfer Moderation Optimal Interunit H O Moderation 2
j
" Big Joe" Full Reflection j
Assembly:
Geometry Heterogeneous UO Rods J
2 Magnetic and Passive Moderation Optimal Interunit 110 Moderation
)
2 Scanner " MAPS" Full ReDection Assembly:
Geometry lieterogeneous UO Rods
)
2 Bundle Accumulator:
Moderation OptimalInterunit H O Moderation 2
"BACC" Full Reflection Assembly: Automatic Geometry Heterogeneous UO Rods 2
Bundle Assemble Moderation OptimalInterunit H O Moderation 2
Machine "ABAM" Full ReDection Assembly:
Geometry Heterogeneous UO Rods 2
Rod Scanner:
Moderation 0, timal Interunit H O Moderation 2
X-Ray-Unit Full ReDection Assembly:
Geometry Heterogeneous UO Rods 2
Rod Scanner Moderation Optimal Interunit H O Moderation 2
" Fat Albert" Full ReDection Assembly:
Geometry lieterogeneous UO Rods 2
Assembly Table Moderation OptimalInterunit H O Moderation 2
Full RcDection Assembly:
Geometry Heterogeneous UO Bundle RA/ Bundle Upender Moderation Optimal Interunit H O Moderation 2
Full RcDection Assembly:
Geometry Heterogeneous UO Bundle 2
Inspection Pit Moderation OptimalInterunit H O Moderation 2
Full Reflection Assembly:
Geometry Heterogeneous UO Bundle 2
Fuel Bundle Moderation OptimalInterunit H O Moderation 2
Storage " Forest" Full ReDection Assembly:
Geometry Heterogeneous 00 7 die 2
RA Transfer Moderation OptimalInterunit H O Moderation 2
Conveyor Full Reflection Rod Inspection:
Geometry Heterogeneous UO Rods 2
Surface-Plate Moderation Optimalinterunit H O Moderation 2
Full ReDection Assembly:
Geometry Heterogeneous UO Rods Rod Tray Cart Moderation Optimal Interunit 110 Moderation 2
[
Full ReDection l
l LICENSE SNM-1997 DATE 09/09/99 Page DOCKET 70-1113 REVISION 6.19
{L
p
)
AREA PROCESS SUBAREA HASIS FOR CSA OR OR EQUIPMENT CRITICALITY HOUNDING ASSUMPTIONS SYSTEM SAFETY Gadolinia Ground: Press, Similar to UO Shop Similar to UO Shop Above 2
2 Sintering, Grinding, Above Rod Load, Rod Storage General:
Mass liomogeneous UO2 IIVAC-Dry Areas Moderation Optimalil O Moderation 2
Full Reflection Ground:
Geometry lieterogeneous UO2 3-Gallon Container Mass Optimal 110 Moderation 2
Floor Storage Full Reflection Ground:
Geometry
)
lieterogeneous UO2 3-Gallon Conveyor Mass f.
Optimal 110 Moderation 2
Storage Moderation Full Reflection l
Ground:
Geometry lieterogeneous UO2 I
Offline Granulator Moderation Optimal 110 Moderation 2
Full Reflection l
Ground:
Geometry / Mass lieterogeneous UO Pellets 2
i Pellet Ministacker Moderation Optimal II 0 Moderation 2
Storage Full Reflection Mezz:
Geometry liomogeneous UO, DM-10 Vibromill Moderation Optimal 110 Moderation 2
Moderation Controlled Stand-Off Reflection I. Area (MCA) l Mezz:
Mass llomogeneous UO2 DM-3 Vibromill Moderation Optimal 110 Moderation 2
(MCA)
Full Reflection Mezz:
Geometry / Mass lieterogeneous UO2 Rotary Slugger: Upper Moderation Maximum Credible wt. % 110 2
Full Reflection Rotary Slugger:
Geometry / Mass lieterogeneous UO2 Lubricant Sump Optimalil O Moderation 2
Full Reflection Mezz:
Geometry lieterogeneous UO2 Granulator Mass Maximum Credible wt. % 110 2
Moderation Full Reflection
- two of any three control parameters are required for criticality safety.
l LICENSE SNM-1097 DATE 09/09/99 Page DOCKET 70-1113 REVISION 3
6.20
AREA PROCESS SUBAREA BASIS FOR CSA OR OR EQUIPMENT CRITICALITY BOUNDING ASSUMPTIONS SYSTEM SAFETY Gadolinia (cont.'d)
Mezz-MRA Moderation liomogeneous UO2 Powder Transfer Maximum Credible UO Density 2
Corridor
. Maximum Credible wt. % H O 2
Full Reflection Mezz-MRA Moderation Homogeneous UO2 Unicone Feed Maximum Credible UO Density 2
Container Maximum Credible wt. % H O 2
Full Reflection Mezz-MRA Moderation Homogeneous UO2 DM-10 Vibromill Maximum Credible wt. % H O 2
Full Reflection Annular Geometry Mezz-MRA Geometry q
Homogeneous UO2 3-gallon feed / product Mass Optimal H O Moderation
)
2 containers Moderation Full Reflection Recycle:
Geometry Heterogeneous UO2 Oxidation Feed Mass Optimal H O Moderation 2
Container Storage Full Reflection Recycle:
Geometry Heterogeneous UO2 Oxidation Furnace Moderation Optimal H O Moderation 2
Full Reflection Recycle:
Geometry Homogeneous UO2 Oxidation Product 3-Mass Optimal H O Moderation 2
gallon Containers Moderation Full Reflection Recycle:
Geometry Homogeneous UO2 DM-10 Vibromill Moderation Optimal H O Moderation 2
StandoffReflection Recycle:
Geometry Homogeneous UO2 HVAC - U308 Moderation Optimal H O Moderation 2
Full Reflection 421-Warehouse:
Geometry / Mass Homogeneous UO2 Exhaust Scrubber Mass Optimal H O Moderation 2
Full Reflection 421-Warehouse:
Geometry 1 Homogeneous UO2 3-Gallon Conveyor Mass f*
Optimal H O Moderation 2
Storage Moderation Full Reflection
- two of any three control parameters are required for criticality safety.
LICENSE SNM-1997 DATE 09/09/99 Page DOCKET 70-1113 REVISION 3
6.21
r l
AREA PROCESS SUBAREA BASIS FOR CSA OR OR EQUIPMENT CRITICALITY BOUNDING ASSUMPTIONS l
SYSTEM SAFETY Shipping Pads:
Geometry lieterogeneous UO2 RA Inner / Outer Storage Moderation OptimalInterunit 110 Moderation 2
Full Reflection Warehouse:
Geometry lieterogeneous UO2 Product Container Moderation Optimal Interunit 110 Moderation 2
Full Reflection Chemet-Lab General Geometry llomogeneous or lieterogeneous UO 2 Mass Optimal H O Moderation 2
Full Reflection
\\
i
{
LlCENSE SNM-1997 DATE 09/09/99 Page DOCKET 70-1113 REVISION 3
6.22 1
l l
\\.
6.2.4 SPECIFIC PARAMETER LIMITS The safe geometry values of Table 6.1 below are specifically licensed for use at the GE-Wilmington facility. Application of these geometries is limited to situations where the neutron reflection present does not exceed that due to full water reflection.
Acceptable geometry margins of safety for units identified in this table are 93% of the minimum critical cylinder diameter, 88% of the minimum critical slab thickness, and 76% of the minimum critical sphere volume.
When cylinders and slabs are not infinite in extent, the dimensional limitations of Table 6.1 may be increased by means of standard buckling conversion methods; reactivity formula calculations which incorporate validated K-infinities, migration areas (M ) and 2
extrapolation distances; or explicit stochastic or deterministic modeling methods.
The safe batch values of Table 6.2 are specifically licensed for use at the GE-Wilmington facility. Criticality safety may be based on U235 mass limits in either of the following ways:
If double batch is considered credible, the mass of any single accumulation shall not exceed a safe batch, which is defined to be 45% of the minimum critical mass.
Table 6.2 lists safe batch limits for homogeneous mixtures of UO and water as a 2
function of U235 enrichment over the range of 1.1% to 5% for uncontrolled geometric c.onfigurations. The safe batch sized for UO of specific compounds may 2
be adjusted when applied to other compounds by the formula:
kgs X -(kgs UO, e 0.88 ) / f where, kgs X
= safe batch valuc of compound 'X' kgs UO,
= safe batch value for UO, 0.88
= wt. % U in UO, f
= wt. % U in compound X Where engineered controls prevent over batching, a mass of 75% of the minimum gritical mass shall not be exceeded.
Sabject to provision for adequate protection against precipitation or other circumstances wish ms.y increase concentration, the following safe concentrations are specifically i
licensed for use at the GE-Wilmington facility:
A concentration ofless than or equal to one-half of the minimum critical concentration.
A system in which the hydrogen to U235 atom ratio (II/U235) is greater than 5200.
LICENSE SNM-1997 DAT' 09/09/99 Page DOCKET 70-1113 REVISION 3
6.23
Fi, l
Table 6.1 Safe Geometry Values Homogeneous UO -
Weight Percent Infinite Cylinder
- Infinite Slab
- Sphere Volume
- 2 H 0 Mixtures U235 Diameters Thickness 3
(Inches)
(Inches)
(Liters) 2.00 16.70 8.90 105.0 2.25 14.90 7.90 75.5 2.50 13.75 7.20 61.0 2.75 12.90 6.65 51.0 3.00 12.35 6.25 44.0 1
3.25 11.70 5.90 38.5 3.50 11.20 5.60 34.0 i
3.75 10.80 5.30 31.0 4.00 10.50 5.10 29.0 l
5.00 9.50 4.45 24.0 Homogeneous Weight Percent infinite Cylinder Infinite Slab Sphere Volume Aqueous U235 Diameters Thickness Solutions (Inches)
(Inches)
(Liters) 2.00 16.7 9.30 106.4 2.25 15.0 8.40 80.5 2.50 14.0 7.80 66.8 2.75 13.3 7.30 56.2 3.00 12.9 7.00 49.7 3.25 12.5 6.70 44.8 3.50 12.1 6.50 41.0 3.75 11.9 6.30 38.0 4.00 11.7 6.00 34.9 5.00 9.5 4.80 26.0 Heterogeneous Weight Percent Infinite Cylinder Infinite Slab Sphere Volume Mixtures or U235 Diameters Thickness Compounds (Inches)
(Inches)
(Liters) 2.00 11.10 5.60 35.7 1
2.25 10.50 5.10 30.7 2.50 10.10 4.80 27.3 l
2.75
- 9. M 4.60 24.7 3.00 9.40 4.40 22.6 3.25 9.20 4.30 20.9 3.50 9.00 4.20 19.2 3.75 8.90 4.10 18.2 4.00 8.80 4.00 16.9 5.00 8.30 3.60 13.0
- These values represent 93%,88% and 76% of the minimum critical cylinder diameter, slab thickness, and sphere volume, respectively. For enrichments not specified, smooth curve interpolation may be used.
l l
LICENSE SNM-1997 DATE 09/09/99 Page DOCKET 70-1113 REVISION 3
6.24 t
E i
Table 6.2 Safe Batch Values for UO, and Water
- j l
Nominal Weight Homogeneous Heterogeneous Nominal Weight Homogeneous Heterogeneous Percent U235 UO, Powder &
UO Pellets &
Percent U235 UO Powder &
UO Pellets &
3 3
3 l
Water Water Water Water Mixtures Mixtures Mixtures Mixtures J
(Kgs UO,)
(Kgs UO )
(Kgs UO )
(Kgs UO )
3 2
1.10 2629.0 510.0 4.00 25.7 24.7 1.20 1391.0 341.0 4.20 23.7 22.9 1.30 833.0 246.0 4.40 21.9 21.4 1.40 583.0 193.0 4.60 20.2 20.0 1.50 404.0 158.0 4.80 19.1 18.8 1.60 293.3 135.0 5.00 18.I I 8. I 1.70 225.0 116.0 1.80 183.0 102.0 1.90 150.6 90.5 2.00 127.5 81.6 2.10 109.2 73.I 2.20 96.8 66.4 2.30 84.3 61.0 2.40 74.7 56.1 2.50 68.9 52.1 2.60 60.5 48.8 2.70 56.6 45.4 2.80 52.2 42.9 2.90 47.6 40.1 3.00 44.5 38.1 3.20 38.9 34.1 3.40 34.6 31.0 3.60 31.1 28.5 3.80 28.3 26.4
- NOTE: These values represent 45% of the minimum critical mass. For enrichments not specified, smooth curve interpolation of safe batch values may be used.
LICENSE SNM-1997 DATE 09/09/99 Page DOCKET 70-1113 REVISION 3
6.25
6.2.5 CONTROL PARAMETERS Nuclear criticality safety is achieved by controlling one or more parameters of a system within established subcritical limits. The criticality safety review process is used to identify the significant parameters associated with a particular system. All assumptions relating to process equipment, material composition, function, and operation, including upset conditions, are justified, documented, and independently reviewed.
Identified below are specific control parameters that may be considered during the review process:
6.2.5.1 Geometry - Geometry may be used for nuclear criticality safety control on its own or in combination with other control methods. Favorable geometry is based on limiting dimensions of defined geometrical shapes to established subcritical limits. Structure and/or neutron absorbers that are not removable constitute a form of geometry control. At the GE-Wilmington facility, favorable geometry is developed conservatively assuming unlimited water or concrete equivalent reflection, optimal hydrogenous moderation, worst credible heterogeneity, and maximum credible enrichment to be processed. Examples include cylinder diameters, annular inner / outer dimensions, slab thickness, and sphere diameters.
Geometry control systems are analyzed and evaluated allowing for fabrication tolerances and dimensional changes that may likely occur through corrosion, wear, or mechanical distortion. In addition, these systems include provisions for periodic inspection if credible conditions exist for changes in the dimensions of the equipment that may result in the inability to meet established nuclear criticality safety limits.
6.2.5.2 Mass - Mass control may be used for a nuclear criticality safety control on its own or in combination with other control methods. Mass control may be utilized to limit the quantity of uranium within specific process operations or vessels and within storage, transportation, or disposal containers. Analytical or non-destructive methods may be employed to verify the mass measurements for a specific quantity of material.
Establishment of mass limits involves consideration of potential moderation, reflection, geometry, spacing, and material concentration. The criticality safety analysis considers normal operations and credible process upsets in determining LICENSE SNM-1097 DATE 09/09/99 Page DOCKET 70-1113 REVISION 3
6.26 J
actual mass limits for the system and for defining additional controls. When only administrative controls are used for mass controlled systems, double batching is considered to ensure adequate safety margin.
6.2.5.3 Moderation - Moderation control may be used for nuclear criticality safety control on its own or in combination with other control methods. When moderation is used in conjunction with other control methods, the area is posted as a ' moderation control area'. When moderation control is the primary design focus and is designated as a the primary criticality safety control parameter, the area is posted ' moderation restricted area'.
When moderation is the primary criticality safety control parameter the following graded approach to the design control philosophy is applied in accordance with 1
established facility practices (in decreasing order of restriction):
I At each enriched uranium interface involving intentional and continuous e
introduction of moderation (e.g., insertion of superheated steam into reactor),
at least three controls are required to assure that the moderation safety factor is not exceeded. At least two of these controls must be active engineered i
controls.
At enriched uranium interfaces involving intentional but non-continuous i
introduction of moderation at least three controls are required to assure that l
the moderation safety factor is not exceeded. At least one of these controls must be an active engineered control, unless a moderation safety factor greater than 3 is demonstrated.
i For situations where moderation is not intentionally introduced as part of the e
process, the required number of controls for each credible failure mode must be established in accordance with the double contingency principle.
l When the maximum credible accident is considered, the safety moderation limit (i.e.,
% H O or equivalent) must provide sufficient factor of safety above the process 2
moderation limit. This ' moderation safety factor', which is the ratio of the safety moderation limit to the process moderation limit, will normally be three or higher, but never less than two. The value of the moderation safety factor depends on the likelihood and time required for this system being considered to transition from the process moderation limit to the safety moderation limit.
LICENSE SNM-1097 DATE 09/09/99 Page DOCKET 70-1113 REVISION 3
6.27
In some cases, as described above, increased depth of protection may be required, but the minimum protection is never less than the following: two independent controls prevent moderator from entering the system through a defined interface and must fail before a criticality accident is possible. The quality and basis for selection of the controls is documented in accordance with Integrated Safety Analysis process described in Chapter 4.0. Controls for the introduction and limited usage of moderating materials (e.g. for cleaning or lubrication purposes) within areas in which the primary criticality safety parameter is moderation are approved by the criticality safety function.
6.2.5.4 Concentration (or Density) - Concentration control may be used for nuclear criticality safety control on its own or in combination with other control methods.
Concentration controls are established to ensure that the concentration level is maintained within defined limits for the system. When concentration is the only parameter controlled to prevent criticality, concentration may be controlled by two independent combinations of measurement and physical control, each physical control capable of preventing the concentration limit being exceeded in a location where it would be unsafe. The preferred method of attaining independence being that at least one of the two combinations is an active engineered control. Each process relying on concentration control has in place controls necessary to detect and/or mitigate the effects ofinternal concentration within the system (e.g., Dynatrol density meter, Rhonan density meter, etc.), otherwise, the most reactive credible i
concentration (density) is assumed.
6.2.5.5 Neutron Absorber - Neutron absorbing materials may be utilized to provide a method for nuclear criticality safety control for a process, vessel or container. Stable compounds such as boron carbide fixed in a matrix such as aluminum or polyester resin; elemental cadmium clad in appropriate material; elemental boron alloyed stainless steel, or other solid neutron absorbing materials with an established dimensional relationship to the fissionable material are recommended. The use of neutron absorbers in this manner is defined as part of a passive engineered control.
Credit may be taken for neutron absorbers such as gadolinia in completed nuclear fuel bundles (e.g., packaged and stored onsite for shipment) provided the following requirements are met:
LICENSE SNM-1997 DATE 09/09/99 Page DOCKET 70-1113 REVISION 3
6.28
The presence of the gadolinia absorber in completed fuel rods is documented and verified using non-destructive testing; and the placement of rods in completed fuel bundles is documented in accordance with established quality control practices.
Credit may be taken for neutron absorbers that are normal constituents of filter media (e.g., natural boron) provided the following requirements are met:
The failure or loss of the media itself also prevents accumulation of significant quantities of fissile material.
The neutron absorber content is certified.
For fixed neutron absorbers used as part of a geometry control, the following requirements apply:
The composition of the absorber are measured and documented prior to first use.
Periodic verification of the integrity of the neutron absorber system subsequent to installation is performed on a scheduled basis approved by the criticality safety function. The method of verification may take the form of traceability (i.e. serial number, QA documentation, etc.), visual inspection or direct measurement.
6.2.5.6 Spacing (or Unit Interaction)- Criticality safety controls based on isolation or interacting unit spacing. Units may be considered effectively non-interacting (isolated) when they are separated by either of the following:
12-inches of full density water equivalent, or e
the larger of 12-foot air distance or the greatest distance across an e
orthographic projection of the largest of the fissile accumulations on a plane perpendicular to the linejoining their centers.
i For Solid Angle interaction analyses, a unit where the contribution to the total solid angle in the array is less than 0.005 steradians is also considered non-interacting (pmvided the total of all such solid angles neglected is less than one half of the total solid angle for the system). Transfer pipes of 2 inches or less in diameter may be excluded from interaction consideration, provided they are not grouped in close arrays.
i l
LICENSE SNM-1997 DATE 09/09/99 Page DOCKET 70-1113 REVISION 3
6.29
]
r i
I Techniques which produce a calculated effective multiplication factor of the entire system (e.g., validated Monte Carlo or S, Discrete Ordinates codes) may be used.
Techniques which do not produce a calculated effective multiplication factor for the entire system but instead compare the system to accepted empirical criteria,(e.g.,
Solid Angle methods) may also be used. In either case, the criticality safety analysis must comply with the requirements of Sections 6.1.1 and 6.3.
6.2.5.7 Material Composition (or Heterogeneity)- The criticality safety analysis for each process determines the effects of material composition (e.g., type, chemical form, physical form) within the process being analyzed and identifies the basis for selection of compositions used in subsequent system modeling activities.
It is important to distinguish between homogeneous and heterogeneous system conditions. Heterogeneous effects within a system can be significant and therefore must be considered within the criticality safety analysis when appropriate.
Evaluation of systems where the particle size varies take into consideration effects of heterogeneity appropriate for the process being analyzed.
6.2.5.8 Reflection - Most systems are designed and operated with the assumption of 12-inch water or optimum reflection. However, subject to approved controls which limit reflection, certain system designs may be analyzed, approved, and operated in situations where the analyzed reflection is less than optimum.
In criticality safety analysis, the neutron reflection properties of the credible process environment are considered. For example, reflectors more effective than water (e.g.,
concrete) are considered when appropriate.
6.2.5.9 Enrichment - Enrichment control may be utilized to limit the percent U-235 within a process, vessel, or container, thus providing a method for nuclear criticality safety control. Active engineered or administrative controls are required to verify enrichment and to prevent the introduction of uranium at unacceptable enrichment levels within a defined subsystem within the same area. In cases where enrichment control is not utilized, the maximum credib'e area enrichment is utilized in the criticality safety analysis.
I i
LICENSE ENM-1997 DATE 09/09/99 Page DOCKET 70-1113 REVISION 3
6.30 i
3 e
t
(
6.2.5.10 Process Characteristics - Within certain manufacturing operations, credit may be taken for physical and chemical properties of the process and/or materials as nuclear criticality safety controls. Use of process characteristics is predicated upon the l
following requirements:
The bounding conditions and operational limits are specifically identified in the criticality safety analysis and, are specifically communicated, through training and procedures, to appropriate operations personnel.
Bounding conditions for such process and/or material characteristics are based on established physical or chemical reactions, known scientific principles, and/or facility-specific experimental data supported by operational l
- history, i
The devices and/or procedures which maintain the limiting conditions must e
have the reliability, independence, and other characteristics required of a criticality safety control.
Examples of process characteristics which may be used as controls include:
Conversion and oxidation processes that produce dry powder as a product of high temperature reactions.
Experimental data demonstrating low moisture pickup in or on uranium materials that have been conditioned by room air ventilation equipment.
Experimental / historical process data demonstrating uranium oxide powder flow characteristics to be directly proportional to the quantity of moisture present.
l 6.3 CONTROL DOCUMENTS 6.3.1 CRITICALITY SAFETY ANALYSIS (CSA)
In accordance with ANSI /ANS-8.19 (1984), the criticality safety analysis is a collection ofinformation that "provides sufficient detail clarity, and lack of ambiguity to allow independentjudgment of the results." The CSA documents the physical / safety basis for the establishment of the controls. The CSA is a controlled element of the Integrated Safety Analysis (ISA) defined in Chapter 4.0.
l LICENSE SNM-1997 DATE 09/09/99 Page DOCKET 70-1113 REVISION 3
6.31
i The CSA addresses the specific concerns (event sequences) of nuclear criticality safety importance for a particular system. A CSA is prepared or updated for each
{
new or significantly modified unit or process system within the GE-Wilmington facility in accordance with established configuration management control practices defined in Chapter 3.0.
The scope and content of any particular CSA reflects the needs and characteristics of the system being analyzed and includes applicable information requirements as follows:
Scope - This element defines the stated purpose of the analysis.
General Discussion - This element presents an overview of the process that is affected by the proposed change. This section includes as appropriate; process description, flow diagrams, normal operating conditions, system interfaces, and other important to design considerations.
Criticality Safety Controls / Bounding Assumptions - This element defines a minimum of two criticality safety controls that are imposed as a result of the analysis. This section also clearly presents a summary of the bounding assumptions used in the analysis. Bounding assumptions include; worst credible contents (e.g., material composition, density, enrichment, and moderation), boundary conditions, interunit water, and a statement on assumed structure. In addition, this section includes a statement which summarizes the interface considerations with other units, subareas and/or areas.
Model Description - This element presents a narrative description of the e
actual model used in the analysis. An identification of both nonnal and credible upset (accident condition) model filenaming convention is provided.
Key input listings and corresponding geometry plot (s) for both normal and credible upset cases are also provided.
Calculational Results - This element identifies how the calculations were performed, what tools or reference documents were used, and when appropriate, presents a tabular listing of the calculational result and associated uncertainty (e.g., Keff + 3a) results as a function of the key parameter (s)
(e.g., wt. fraction H O). When applicable, the assigned bias of the 2
calculation is also clearly stated and incorporated into both normal and/or accident limit comparisons l
LICENSE SNM-1997 DATE 09/09/99 Page
~
DOCKET 70-1113 REVISION 3
6.32
l
\\
i Safety During Upset Conditions - This element presents a concise summary
{
of the upset conditions considered credible for the defined unit or process l
system. This section include a discussion as to how the established nuclear criticality safety limits are addressed for each credible process upset (accident condition) pathway, i
Specifications and Requirements for Safety - When applicable, this I
e clement presents both the design specifications and the criticality safety seguirements for correct implementation of the established controls. These requirements are incorporated into operating procedures, training, maintenance, quality assurance as appropriate to implement the specifications and requirements.
Compliance - This element concludes the analysis with pertinent summary statements and includes a statement regarding license compliance.
Verification - Each criticality safety analysis is verified in accordance with section 6.3.2.5 by a senior engineer approved by the criticality safety function and who was not involved in the analysis.
Appendices - Where necessary, a summary ofinformation ancillary to calculations such as parametric sensitivity studies, references, key inputs, model geometry plots, equipment sketches, useful data, etc., for each defined system is included.
6.3.2 ANALYSIS METilODS 6.3.2.1 KeffLimit i
Validated computer analytical methods may be used to evaluate individual system units or potential system interaction. When these analytical methods are used, it is required that the effective neutron multiplication factors for credible process upset (accident) conditions are less than or equal to 0.97 including applicable biases and calculational uncertainties, that is:
Keff + 3o - bias s 0.97 (accident conditions).
Thus, the established delta-k safety margin used at the GE-Wilmington facility is 0.03.
LICENSE SNM-1997 DATE 09/09/99 Page DOCKET 70-1113 REVISION 3
6.33 1
I Normal operating conditions include maximum credible conditions expected to be
~
encountered when the criticality control systems function properly. Credible process upsets include anticipated off-normal or credible accident conditions and must be demonstrated to be critically safe in all cases in accordance with Section 6.1.1. The sensitivity of key parameters with respect to the effect on Keff are evaluated for each
)
system such that adequate criticality safety controls are defined for the analyzed i
system.
6.3.2.2 Analytical Methods Methodologies currently employed by the GE-Wilmington criticality safety function include hand calculations utilizing published experimental data (e.g., ARH-600 4
handbook), Solid Angle methods (e.g., SAC code), and Monte Carlo codes (e.g.,
GEKENO, GEMER) which utilize stochastic methods to solve the 3D neutron transport equation. Additional Monte Carlo codes (e.g., Keno Va and MCNP) or S, Discrete Ordinates codes (e.g., ANISN or XSDRNPM) may be used after validation as described in subparagraph (c) below.
GEKENO (Geometry Enhanced KENO) is a multigroup Monte Carlo program which solves the neutron transport equation in 3-dimensional space. The GEKENO criticality program utilizes the 16-energy group Knight-Modified Hansen Roach i
cross-section data set, and a potential scattering o, resonance correction to compensate for flux depression at resonance peaks. GEKENO is normally used for homogeneous systems. For infinite systems, K. can be calculated directly from the Hansen Roach cross-sections using the program KINF.
GEMER (Geometry Enhanced merit) is a multigroup Monte Carlo program which solves the neutron transport equation in 3-dimensional space. The GEMER criticality program is based on 190-energy group structure to represent the neutron energy spectrum. In addition, GEMER treats resolved resonances explicitly by tracking the neutron energy and solving the single-level Breit-Wigner equation at each collision in the resolved resonance range in regions containing materials whose resolve resonances are explicitly represented. The cross-section treatment in GEMER is especially important for heterogeneous systems since the multigroup treatment does not accurately account for resonance self-shielding.
6.3.2.3 Validation Techniques LICENSE SNM-1997 DATE 09/09/99 Page DOCKET 70-1113 REVISION 3
6.34
)
Experimental critical data or analytical methods which have been validated i
(benchmarked) by comparison with experimental critical data in accordance with criteria described in section 4.3 of ANSI /ANS 8.1 (1983) are used as the basis for validation. An analytical method is considered validated when the following are i
established:
the type of systems which can be modeled e
the range of parameters which may be treated e
the bias, if any, which exists in the results produced by the method.
e Currently GEMER is validated against l'23 critical experiments and GEKENO is validated against 56 critical experiments. Both validations produce a bias fit as a function of H/U235 atom ratio. This fit is established against the lower limit of the 3-sigma confidence band (see Figures 6.1 and 6.2). The bias (K,ac - 1.0) is applied over its negative range and assigned a value of zero over its positive range. The range of applicability covers all compounds in use at GE-Wilmington and l
enrichments up to 5.0 % wt. % U235.
l l
l 1
i l
LICENSE SNM-1997 DATE 09/09/99 Page DOCKET 70-1113 REVISION 3
6.35 l
n-FIGURE 6.1 - CEMER BIRS DETERMINATION, PARilCLE NEICHT 1.10 LEGEND 128 CATA SET PARTICLE WEIGHT x 3RD ORDER FIT OF LIMIT '
1.00
- -Err. 1.0 L! HEAR FITe ORDERS 2 90.T32 CONFIDENCE SAND 1.06 1.04 K-EFF 13e 1.02 4
s
.w.
1.0.
g 8
0.980 0.960 att 20 50 et 110 140 1T0 HY0kOGEN-70-U245 Mit LICENSE SNM-1097 DATE 09/09/99 Page DOCKET 70-1113 REVISION 3
6.36
y FICURE 6.1 - CENER BIAS DETERMINATION, PARTICLE WEICHT 1.10 LEGEND 128 DATA SET
. PARTICLE WEIGHT u 4RD ORDER FIT OF LIMIT 1.06
= g-EFF e 1.3 LINEAR FITS ORDERe 2 99.733 CONFIDENCE SAND 1.06 1.0, K-EFF 139 1.02 4
ku.
j 0.969
-se 2e se se e
sie are MYOR0 GEN-TO-U235 Nit"I LICENSE SNM-1097 DATE 09/09/99 Page DOCKET 70-1113 REVISION 3
6.36 1
l l
i
I c
1 i
l FIGURE 6.2 - CEKENO BIAS CALCULATION 1
i.i.
LEGEND GEKENO VERSION..
e 5. DATA Po!NTS l
M 3RD ORDER FIT OF LIMIT E,,
LINEAR FITS ORDERm 2
...T.
c0N,10ENet..ND
)
i...
i li
\\,
K-EFF ist pN q
1=
i Nb r,e
%;y j
a.
6.
32.
35.
1
- I HYDROGEN-TO-U285 X1.
6.3.2.4 Computer Software & Hardware Configuration Control The software and hardware used within the criticality safety calculational system is configured and maintained so that change control is assured through the authorized system administrator. Software changes are conducted in accordance with an soproved configuration control program described in Chapter 3.0 that addresses both hedware and software qualification.
Software designated for use in nuclear criticality safety are compiled into working code versions with executable files that are traceable by length, time, date, and version. Working code versions of compiled software are validated against critical experiments using an established methodology with the differences in experiment LICENSE SNM-1097 DATE 09/09/99 Page DOCKET 70-1113 REVISION 3
6.37 l
l
D and analytical methods being used to calculate bias and uncertainty values to be applied to the calculational results.
Each individual workstation is verified to produce results identical to the development workstation prior to use of the software for criticality safety calculations demonstrations on the production workstation.
Modifications to software that may affect the calculational logic require re-validation
. of the sonware. Modifications to hardware or sonware that do not affect the calculational logic are followed by code operability verification, in which case, selected calculations are performed to verify identical results from previous analyses.
Deviations noted in code verification that might alter the bias or uncertainty requires re-qualification of the code prior to release for use.
6.3.2.5 Technical Reviews Independent technical reviews of proposed criticality safety control limits specified in criticality safety analyses are performed. A senior engineer within the criticality safety function is required to perform the independent technical review.
The independent technical review consists of a verification that the neutronics geometry model and configuration used adequately represent the system being i
analyzed. In addition, the reviewer verifies that the proposed material characterizations such as density, concentration, etc., adequately represent the system. He/She also verifies that the proposed criticality safety controls are adequate.
The independent technical review of the specific calculations and computer models are performed using one of the following methods:
Verify the ca::ulations with an alternate computational method.
e Verify the calculad;ns by performing a comparison to results from a similar design or to similar previously performed calculations.
Verify the calculations using specific checks of the computer codes used, as well as, evaluations of code input and output, Verify the calculations with a custom method.
e LICENSE SNM-1997 DATE 09/09/99 Page DOCKET
'70-1113 REVISION 3
6.38
1 m
Based on one of these prescribed methods, the independent technical review provides a reasonable measure of assurance that the chosen analysis methodology and results are correct.
6.4 CRITICALITY ACCIDENT ALARM SYSTEM
- 6.4.1 -
SPECIFICATIONS The criticality accident alarm system radiation monitoring unit detectors are located to assure compliance with appropriate requirements of ANSI /ANS-8.3 (1986). The location and spacing of the detectors are chosen to avoid the effect of shielding by massive equipment or materials. Spacing between detectors is reduced where high density building materials such as brick, concrete, or grout-filled cinder block shield -
a potential accident area from the detector. Low density materials of construction such as wooden stud construction walls, asbestos, plaster, or metal-corrugated panels, doors, non-load walls, and steel office partitions are disregarded in determining the spacing.
i j
6.4.2 OPERATION i
The criticality accident alarm system initiates immediate evacuation of the facility.
Employees are trained in recognizing the evacuation signal. This system, and proper response protocol, is described in the Radiological Contingency and Emergency Plan for GE-Wilmington.
6.4.3 MAINTENANCE The nuclear criticality alarm system is a safety-significant system and is maintained through routine calibration and scheduled functional tests conducted in accordance with internal procedures. In the event ofloss of normal power, emergency power is automatically supplied to the criticality accident alarm system.
l l
l l
LICENSE SNM-1997 DATE 09/09/99 Page DOCKET 70-1113 REVISION 3
6.39
g Director,NMSS Sept mber 9,1999 Page1ofI l
1 ATTACHMENT 3 l
4 Integrated Safety Analysis (ISA) of the Dry Scrap Recycle U 0s HEPA Filter Modification 3
1 4
1 l
1 I
t I
i
e a
ISA Summary GE Wilmington Dry Scrap Recycle U 0s HEPA Filter Modification 3
Introduction An integrated safety analysis (ISA) was performed to support the modification of the Dry Scrap Recycle (DSR) facility at the GE Wilmington site. The modification studied was the change from a Mass criticality control to a Geometry criticality control at the U 0 HEPA filter system. This was required due to U 0s 3
3 particle loading behavior and its impact on the delta-pressure across the HEPA housing. This report summarizes the important results of the safety study for this DSR modification. These changes were also generically implemented in the other U 0s HEPA housings in a separate Gad Scrap Recycle (GSR) facility.
3 Detailed records are retained on-site.
ISA Team The ISA review was performed June 7,1999 by a team which included expertise in criticality safety, radiological safety, industrial and environmental safety, process and HVAC engineering, and a configuration management facilitator.
Procedures, Techniques, and Tools The ISA team used the "What If" method to study the modifications effects. The studies were conducted in accordance with License SNM-1097, Chapter 4, Section 4.7, and GE Wilmington's internal Practices and Procedures (P/P) 10-20, Integrated Safety Analysis.
Background
This modification to DSR was made due to an unexpected accumulation of approximately 50 kg of U 0 3
powder in the HEPA filter housing used for the furnace exhaust and feed gate valves (Exhaust Unit 562-X).
The system was originally designed with Mass and Moderation criticality controls. Mass control was claimed due to external monitoring of a delta-pressure gauge measurement across the llEPA filter housing.
Extensive data correlation had indicated that the holdup would not exceed 25 kg at 4" delta-P. 25 kg is a safe mass of UO at 5.00% U235 enrichment. In this event, the mass control was compromised by the failure 2
of the delta-pressure control sensor to reach its high-level limit of 4" H O even with 50 kg of U 0, in the 2
3 housing (prefilter + primary HEPA cartridge mass).
This failure mode was not discussed in the original DSR HAZOP review conducted in March 1997, nor was it anticipated.
As a result, the parameter control scheme for U 0 HEPA applications was modified. The mass parameter 3
control was replaced with a geometry parameter control, via instailation of new HEPA filter housing (s) with geometrically safe 5-7/8" thick primary filter cartridge with no pre-Olter.
An additional concem was noted in the ISA meeting. This was the plan for removal of uranium powder from any HEPA filter that had accumulated more than 25 kg.
Page1of2
r L
t.
l Hazard Considerations ACCIDENT: An unsafe mass of Uranium is transferred to the new furnace HEPA filter PROTECTION: The furnace llEPA filter housings are geometrically safe 5-7/8" slabs. No pre-filters are used in these U 0s applications. Periodic non-destructive assay gamma scan surveillance ensures uranium 3
accumulation remains low. Separate periodic examination of the filter by HVAC maintenance also assures powder levels are not excessive.
ACCIDENT: Moderation controls in DSR fail PROTECTION: The DSR recycle-feed subarea is designated as a Moderation Controlled Area (MCA). As such, moderators and process piping are strictly controlled. In addition, only restricted material types of known low moisture levels are allowed to the furnace feed system by the active computer assisted controls on logical move transactions. Only Hard Scrap and swarf are procedurally allowed in DSR furnace feed stream, and these are controlled to less than 50,000 ppm H 0. There is no process piping containing 2
moderators in the furnace oft-gas system.
ACCIDENT: An overweight HEPA filter is transported out of the MRA area.
PROTECTION: Procedural control prohibits movement of more than 25 kg net weight out of an MCA (or MRA).- A limited group of highly trained maintenance personnel are responsible for filter changes.
j Maintenance procedures and Nuclear Safety Release / Requirements (NSR/Rs) have been modified to reflect proper control and handling of heavy filters. Only authorized HVAC personnel handle these HEPA cartridges. Scales have been identified in the DSR area for maintenance personnel to check accumulation of mass.
Page 2 of 2
p-Director, NMSS September 9,1999 Page1ofI ATTACHMENT 4 Integrated Safety Analysis (ISA) of the GAD Scrap Recycle
r 1
l 1
i ISA Summary l
GE Wilmington GAD Scrap Recycle l
l Introduction An integrated safety analysis (ISA) was performed to support the gadolinia shop RECYCI.E subarca at the GE Wilmington site. This GSR process reconfigures the l
former REDCAP UO pellet oxidation process to enable dry recycle of UO pellets 2
2 containing gadolinia (Gd 0 ), in a manner similar to the Dry Scrap Recycle (DSR) 2 3 facility The analysis was conducted in accordance with Chapter 4, Integrated Safety Analysis, of SNM License 1097, submitted April 5,1996, as amended. The ISA documents the results of the evaluation performed and the controls in relation to the risks that they mitigate to insure that the proper levels of assurance measures are applied. This report summarizes the important results of the safety study for the GAD Scrap Recycle (GSR) facility. Detailed records are retained on-site.
ISA Team The ISA was performed by a team of people who systematically analyzed the hazards in a focused meeting environment. The team included expertise in operations, maintenance, criticality safety, radiological safety, and industrial safety and environmental protection.
Also included were process engineers, equipment engineers, operation management, and a configuration management facilitator.
Procedures, Techniques, and Tools i
The ISA team adopted the "What If' method to perform the analysis in view of the GSR process being a reconfiguration of an existing scrap oxidation process (REDCAP), and it's similarity to our existing DSR process. The proceedings of the studies were captured using a personal computer tool based on Microsoft Access.
This tool facilitated consensus between the team members and permitted efficient publication of reports.
Guidance was provided by Guidelines for liazard Evaluation Procedures'. The studies were conducted in accordance with License SNM-1097, Chapter 4, Section 4.7, and GE Wilmington's internal Practices and Procedures (P/P) 10-20, Integrated Safety Analysis.
' GuideP;csfor Ha:ard Evaluation Procedures. Second Edition with Worked Examples, Center for Che.acal Proce:s Safety of the American Institute of Chemical Engineers, New York,1992.
E D
L l.
GSR ISA Summary
[-
GE Nuclear Energy L
September 1999 l:
Process Overview The Gad Scrap Recpcle (GSR) process oxidizes UO2 pellets containing gadolinia and prep:nc; (U GdhOs generated from clean hard scrap and swarf generated in the GAD Fuel Fabrication Operation. This oxidized product is then added back inta the UO2 powder stream in the existirig GAD MRA facility to produce (U, Gd)O. The process 2
consists of the following major sta !ons:
Scrap collection station in the GAD shop.
L
' Static furnace with an active grinder, sifler, can fill at the exit end.
e Mechanical can lift and dump station into a commercial Vibromill with can discharge.
e Laboratory moisture analyzer for samples.
Processed can warehouse storap area.
- Can lift elevator to MRA level.
GAD blending Moderation Restricted Area (MRA) transfer station.
The UO2 scrap generated in the Gad Fabrication area will be oxidized to (U,Gd)30s in the static furnace, milled and blended, and finally added back to (UO )Gd blends.
2 All areas of the GSR process are withiri hsignated Moderation Controlled Area (MCA) areas of the facility. All material logical station movements in the GSR area are controlled by computer transactions that verify container identification and material type.
Methodolony Each process was divided into sub-processes, and the team members postulated "What If" questions to identify accident scenarios. For each "What If" postulated, the consequence and the safeguards in place to prevent or mitigate the "What If" were :de.ufied.
The seve:ity of the consequence was ranked both v sthout cor.trols, S, and with controls i
in place, S2 Similarly, the likelihood of the cowauence was ranked without controls, L, and with controls, L.
i 2
The risk without controls indicated the relative importance of the controls that prevent the consequence in this "What If".
Risk (withoo controis> = S x Li i
The risk with the controls in place was used to judge the adequacy of the controls, where Risk <nna) = S x L2 2
Page 2 0f 9
G 1
c.
\\
o GSR ISA Summary GE Nuclear Energy September 1?99 I
)
If the final risk was judged by the team to be too hi;;h, recommendations for improv1.i.st were captured and reviewed with management. Recommendations that correct higli O 4 risk ' situations are tracked to_ completion using management tracking systems and techniques.
This assessment is consistent with License SNM-1097, Chapter 4, Section 4.9 " Controls fc,r Prevention and Mitigation of Accidents" which provides additional details on the ri i rnking system.
The facility has fully implemented the double contingency principal for criticality safety control. In performing ISA analyses,' where criticality initiating events are involved, risk is assessed based on the fact that only one of the unlikely conditions is present. In rare cases where a singular process upset can render both controls ineffective, the team gives these situations special consideration.
I i
l l
Page 3 of 9 L
r %,.
i GSR ISA Summary GE Nuclear Energy September 1999 Hazard Considerations The following hazard considerations were determined to be the highest unmitigated risks.
ACCIDENT: Can cards are mixed up and wrong can passes into the MRA receipt hood.
PROTECTION: Active computer assisted entry gate interlock controls can entry into the receipt hood.~ Only ~ cans with the correct material type code, tamper safe seal ID, and moisture sample result < 4000 ppm H O are permitted to enter the receipt hood through 2
gate #1.
'A second computer assisted. active control then requires a local sample confirmation that the moisture content < 4000 ppm H2O before the second exit gate #2 interlock to be opened and the can enter the MRA. The material type code, tamper safe seal, and local moisture confirmation prevent wrong material from entering the MRA.
ACCIDENT: Cans with excessive moisture or moderator entering the MRA.
U 0s feed material-move computer transactions into the MRA are not PROTECTION:
3 allowed anywhere except through the receipt hood. The receipt hood contains two c'omputer controlled interlocking gates. Prior to cans entering the first gate, computer interfaces verify both the can and tamper safe seal identification and initial moisture assay result.
Once the past the first interlock gate, the operator is required to analyze a second moisture
-assay on the U30s material feed can. A second computer assisted active control then r.: quires a local sample confirmation that the moisture content < 4000 ppm H2O before the second interlock with exit gate #2 to open and the can enter the MRA.
The MRA exit conveyor has a one directional gate and a non-reversing conveyor.
Operators are trained to only allow GSR material types into the transfer hood and that only empty cans are allowed entry into the MRA without a computer move transaction.
The GSR cans are unique due to their tamper safe seals, and the black U 0s powder is 3
visibly different from UO2 Powder.
r-Page 4 of 9
m GSR ISA Summary GE Nuclear Energy Septemt'er 1999 ACCIDENT: Loss of mass control due to excessive accumulation of U30s powder in furnace exhaust HEPA filter.
PROTECTION: The furnace HEPA filter housings are geometrically safe 5-7/8" slabs.
No pre-filters are used in these U 0s pplications. Periodic non-destructive assay gamma q
3 a
scan surveillance ensures uranium accumulation remain low.
Separate periodic examination of the filter also assures powder levels are not unreasonable.
The HEPA filter housing is physically isolated from overhead sprinklers to assure moderation control. In addition, only restricted material types of known low moisture levels are allowed in the furnace by the required computer move transactions.
I Active controls auure that the furnace exhaust temperatures are not excessive, and additionally, a passive spoiler design ensures exhaust gas temperatures will not damage the filter.
ACCIDENT: An overweight HEPA filter is transported out of the MCA area.
PROTECTION: Procedural control prohibits movement of more than 25 Kg's net weight out of an MCA or MRA. A limited group of trained maintelance personnel are l
responsible for filter changes. Maintenance procedures reflect proper handling of the filters, and training of these authorized personnel has been conducted. Scales have been identified in the GSR area for maintenance personnel to check accumulation mass.
' ACCIDENT: Overfill of accumulation 3 gallon can in GAD Fabrication area.
PROTECTION: Collection of hard scrap and swarfis done in a safe geometry container j
and weighed on an accountability scale prior to movement to the GSR area. A process scale is automatically interlocked with a visible alarm to indicate the 25 kg maximum gross weight is met.
ACCIDENT: Too many beakers are stored in the scrap accumulation transfer hood.'
PROTECTION: Administrative controls limit the hood to a single feed container at a time (one beaker, or one pellet boat, or one 3-gallon can). The discharge only permits a single 3-gallon container. ' A sign is also posted at the hood as an operator aid.
1 i
Page 5 of 9 l
~, %
i GSR ISA Summary
. GE Nuclear Energy Septemt>er 1999 6CCIDENT: Oily pellets are placed in the scrap accumulation can.
PROTECTION: Computer-assisted administrative controls ensure that only grinder swarf, or clean-_hard scrap from the grinder, rod load and sintering furnace are accumulated for transfer into the GSR area. The hood is labeled to indicate sinteied pellets only; no green pellets. Operators are also trained that not pellets with visible moisture or oil are permitted as feed to this hood.
ACCIDENT: Loss of moderation control due to an overhead water leak:: ar fire system sprinkler activation at the fumace.
PRGIECflON: Pellet feed, furnace muffle, and oxidized product boat conveyors'are equipped with passive engineered covers that shed water. Selective sprinkler heads have -
been removed over the furnace exit and sifter. Cans are covered with positive closure lids.- Pellet boats are designed to drain water.
Operators are trained to repart any roof leaks. Cleaning protocol (mop pail use and storage) is administratively controlled.
Also, chemical combustibbs restricted in area as a fire prevention measure.
ACCIDENT: Exhaust from furnace is too hot and damages the exhaust HEPA filter.
PROTECTION: An active '.emperature interlock is used in the furnace exhaust system filter housing transition which shuts down the furnace if exhaust temperatures become excessive, 'and provides an alarm both locally and at a remote control console.
Additionally, the exhaust is passively cooled with room dilution air to reduce temperatures.
ACCIDENT: Boat jams in furnace feed or discharge area.
PROTECTION: An active contre! on the pusher motor prevents movement of the
- conveyor system or muffle in the event of a boat jam. In addition, index logic built into the control design uses discharge sensors to alarm boat jams to the operator.
~Adminirtrative controls prohibit operators from stacking boats in this system.
l l
Page 6 of 9 l
rc GSR ISA Summary GE Nuclear Energy September 1999 l ACCIDENT: ' Boats with Uranium are stack (d in Furnace feed / discharge hood.
I PROTECTION: Operating procedure does not allow stacked boats in either the feed or discharge hoods. The furnace muffle is designed such that only one layer of boats can move through the furnace.
l-ACCIDENT: Fuel height is exceeded in furnice boats.
PROTECTION: Furnace boats dinv, vins prevent overfilling. Operators volumetrically fill the boats with a single layer of pellets, or equivalent powder. Passively engineered height limiting bars at the furt,r.ce feed limit the geometry of material in a boat to safe slab height.
ACCIDENT: Heterogeneous material generated due to furnace temperature control excursion' or feed rate failures.
s PROTECTION: - Temperature profile of furnace zones automatically controlled, with alarms provided to the operator. The stoke rate of the furnace is fixed and not adjustable by the operator. The sifter at the exit of the furnace utilizes a perforated plate to insure only. homogeneous material can pass, and sorts any large particles to a collection container for reprocessing.
- ACCIDENT: Material accumulation under feed or discharge of the furnace conveyor.
. PROTECTION: Operating procedures require weekly inspection and log of findings and clean-outs. ' A monthly scrap height report is issued. Nuclear Safety has modeled this operation to allow for material build up.
- ACCIDENT: Operator gets too close to hot surfaces.
PROTECTION: Personnel are trained in burn hazards in their areas and are provided with insulated gloves for use in abnormal conditions. The furnace is enclosed and L
insulated to reduce surface temperatures to a safe level. Boats travel through a cool'down conveyor prior to being loaded into the grinder hopper.
I I
i l
Page 7 oi 9 l
i l
)
ei.
o GSRISA Summary GE Nuclear Energy September 1999
.- ACCIDENT: Multiple boats u dumped into furnace discharge grinder.
. PROTECTION:,The lid on the grinder dump mechanism is interlocked with the mill
. motor to insure rnaterial is not mounded its hopper. The operating procedure instructs to 1 only process one boat at a time through the mill.
ACCIDENT:' A screen break occurs in the furnace sifter.
PROTEC110N: The sifter utilizes a fine mesh screen for process quality control only. It is assembled with a perforated metal plate underneath the screen with a minimum mesh required to assure homogeneity. The operating procedure requires periodic inspection of the process screen to insure integrity.
ACCIDENT: Furnace sifier screen is not installed properly.
PROTECTION: The sifier is installed in a ventilated hood to minimize radiological
' concerns to the operator. Operator observation of this area is natural due to the proximity j
of the discharge collection can. The operating procedure requires shift supervision to j
. verify proper screen installatica whenever the screen is assembled.
ACCIDENT: Furnace sifier output can is overfilled.
PROTECTION: Output alarm and weight indicator warns the operator when each can is full. Procedures allow only one boat of material to be processed at a time through the grinder, sifter, and can fill system.
i l
ACCIDENT: Vibromill feed hopper lid is not closed on the can dump prior to mill
- startup.
PROTECTION: An active control interlocks the lid'down position with the mill run-command.
l ACCIDENT: Heterogeneous material is feed to the Vibromill due to a sifter screen break.
. PROTECTION: The sifier is assembled with a perforated metal plate underneath the process screen that assures material homogeneity. In addition, the Vibromill blend time ensures homogeneity ofits powder output. The operator is instructed to closely monitor screen integrity at the sifter, and make repairs immediately.
Page 8 of 9 s
r o4
)
GSR ISA Summary GE Nuclear Energy September 1999 ACCIDENT: Wrong material type is put into the Vibromill.
PROTECTION: Active computer interlocks control the material types allowed into the Vibromill room. Additional computer interlocks are placed on movement of material to the Vibromill dump station. Operators are instructed to visibly inspect the material they dump into the Vibromill. Also, green scrap (non-oxidized material) generated from the are physically separate from the sintered (U,Gd)O2 pellet /swarf hard scrap accumulation area.
ACCIDENT: Operator takes moisture sample from the wrong can prior to can movement to warehouse storage.
PROTECTION: Operators are required to sample each blend in order to assign a moisture content. The blend time on the Vibromill insures homogeneity of every can.
Each Vibromill output can is provided with a tamper safe seal that is uniquely identified, if any U30 can seal is broken, it must be re-sampled for moisture content. The moisture content of each can is also verified at the receipt hood prior to entry into the MRA.
ACCIDENT: Moisture gets into cans in warehouse storage prior to entering MRA.
PROTECTION: The designated storage area is designed with a secondary roof to prevent moisture during storage. All cans in the storage area have a secured lid with a tamper safe seal. Only cans with tamper safe seals are allowed into the MRA transfer sample hood. An additional moisture sample is verified from every can prior to entry into the MRA.
ACCIDENT: The can elevator to GAD Fabrications MRA level fails.
PROTECTION: The can lift is limited to one 3 gallon can and an engineered can restraint cage provides protection. Structural integrity of the elevator has been verified.
ACCIDENT: Moisture level is higher than indicated by instruments due to common mode failure on one of the two laboratory moisture analyzers.
PROTECTION: Separate calibration of each analyzer should signify any common mode failure-s. Also, each analyzer is independently verified with known standards at specified frequency.
Page 9 of 9 L-
,