ML20132D356
| ML20132D356 | |
| Person / Time | |
|---|---|
| Site: | 07001113 |
| Issue date: | 12/16/1996 |
| From: | Reda R GENERAL ELECTRIC CO. |
| To: | Weber M NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS) |
| References | |
| NUDOCS 9612190341 | |
| Download: ML20132D356 (41) | |
Text
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GENuclear Energy GeneralDectnc Company I
PO Box 780. Wdmmgwn tic 2843?
S10615 SM December 16,1996 Mr. M. F. Weber, Licensing Branch, NMSS U.S. Nuclear Regulatory Commission l
l Mail Stop T 8-D-14 l'
- Washington, D.C. 20555-0001
Subject:
License Renewal Application Revision - Chapter 6 i
References:
- 1) Docket 70-1113, NRC License SNM-1097 i
L
- 2) License Renewal Application, RJ Reda to EQ TenEyck,4/5/96 l
Dear Mr. Weber:
GE's Nuclear Energy Production (NEP) facility in Wilmington, N.C. hereby requests that the above referenced application for renewal of NRC license SNM-1097 be revised. The requested revisions affecting our criticality safety requirements in Chapters 6 are based upon input received from your staff. Chapter 6 is provided.as an attachment to this letter and replaces in its entirety Chapter 6 contained in our license renewal submittal of 4/5/96.
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The major changes include: (1) Section 6.2.3 and Table 6.0, which identify the controlled parameters as requested by the NRC, (2) Section 6.2.4, which incorporates some standard parameter limits for use at the facility, and (3) additions to Section 6.3.2.3, which justify the use of our codes under certain conditions for enrichments up to 6% U235. Other minor changes are editorial in nature.
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Page 2 o'f the Table of Contents has been updated to reflect the above changes. Page 4 of the Table of Contents is the Revisions By Chapter which identifies each chapter, the number of pages in that chapter, and the date of the submittal. Changes are indicated with a vertical bar ( l ) in the right hand margin.
To avoid confusion and facilitate chapter replacement in the binder submitted on 4/5/96, if desired, all pages of Chapter 6 contain the date of this submittal. This method is also easier for us to administratively control. The revision number on all pages remains as zero (0).
Six copies of this submittal are hereby provided for your review.
Please contact Charlie Vaughan on (910) 675-5656 or Rick Foleck on (910) 675-6299 if you require additional information or would like to discuss this matter further.
Sincerely, GE NUCLEAR ENE Y
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9612190341 961216
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PDR ADOCK 07001113 C
PDR R.. Reda, Manager b
Fuels and Fac;lity Licensing j
/zb enclosure I
cc: RJR-96-140 1
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TABLE OF CONTENTS Section Title Page 4
I CHAPTER 1 GENERAL INFORMATION l
1.1 Facility and Process Description 1.1 1.2 InstitutionalInformation 1.7 1.3 Special Authorizations 1.10 CHAPTER 2 ORGANIZATION AND ADMINISTRATION 2.1 Policy 2.1 2.2 Organizational Responsibilities and Authority 2.1 2.3 Safety Committees 2.10 CHAPTER 3 CONDUCT OF OPERATIONS O
3.1 Configuration Management (CM) 3.1 3.2 Maintenance 3.2
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3.3 Quality Assurance (QA) 3.4 3.4 Training and Qualification 3.6 i
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 CHAPTER'4 INTEGRATED SAFETY ANALYSIS 4.1 Integrated Safety Analysis 4.1 4.2 Site Description 4.1 4.3 Facility Description 4.1 4.4 Process Description 4.2 4.5 Process Safety Information 4.2 LICENSE SNM-1097 DATE 12/16/96 Page DOCKET 70-1113 REVISION 0
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i TABLE OF CONTENTS Section Title Page 4.6 Training and Qualifications of the ISA Team 4.2 4.7 ISA Methods 4.2 4.8 Results of theISA 4.3 4.9 Controls for Prevention and Mitigation of Accidents 4.4 4.10 Administrative Control of the ISA 4.7 CHAPTER 5 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 5.7 5.8 Summing Internal and External Exposure 5.9 5.9 Action Levels for Radiation Exposures 5.9 p/
s 5.10 Respiratory Protection Program 5.9 j
5.11 Instrumentation 5.10 NUCLEAR CRITICALITY SAFETY 6.1 Program Administration vi 6.2 Technical Practices 6.5 6.3 Control Documents 6.28 6.4 Criticality Accident Alarm System 6.36 CHAPTER 7 CHEMICAL SAFETY 7.1 Chemical Safety Program 7.1 7.2 Contents of Chemical Safety Program 7.1 LICENSE SNM-1097 DATE 12/16/96 Page DOCKET 70-1113 REVISION 0
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TABLE OF CONTENTS Section Title Page CHAPTER 8 FIRE SAFETY 8.1 Fire Protection Program Responsibility 8.1 8.2 Fire Protection Program 8.1
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8.3 Administrative Controls 8.2 8.4 Building Construction 8.2 8.5 Ventilation Systems 8.3 l
8.6 Process Fire Safety 8.3 i
8.7 Fire Detection and Alarm Systems 8.3 8.8 Fire Suppression Equipment 8.4 l
8.9 Fire Protection Water System 8.4 l
8.10 Radiological Contingency and Emergency Plan (RC&EP) 8.5 8.11 Emergency Response Team 8.5 CHAPTER 9 RADIOLOGICAL CONTINGENCY AND EMERGENCY PLAN 9.1 CHAPTER 10 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 i
CHAPTER 11 DECOMMISSIONING 11.1 l
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l REVISIONS BY CHAPTER l
Application Application Page Date Page Date l
TABLE OF CONTENTS l
l CHAPTER 6 l
l.
I through 4 12/16/96 l
1 through 36 12/16/96 l l
l l
CHAPTER 1 l
l CHAPTER 7 l
1 l
1 through 22 08/30/96 1 through 3 04/05/96 l
CHAPTER 2 l
l CHAPTER 8 l
1 through 11 08/30/96 1 through 5 04/05/96 O
l CHAPTER 3 l
l CHAPTER 9 l
1 through 12 08/30/96 1
04/05/96 l
CHAPTER 4 l
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CHAPTER 10 l
1 through 8 04/05/96 1 through 16 04/05/96 l
CHAPTER 5 l
l CHAPTER 11 l
1 through 13 08/30/96 1
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O cu^erea 6 a NUCLEAR CRITICALITY SAFETY 6.1 PROGRAM ADMINISTRATION 6.1.1 CRITICALITY SAFETY DESIGN PHILOSOPHY 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 suffident 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 j
enforced.
i The established design criteria and nuclear criticality safety reviews are applicable to:
all new processes, facilities or equipment that process, store, transfer or q
V 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 l
6.1.2.1 Changes to Facility i
As pan of the design of new facilities or significant additions or changes in existing i
facilities, Area Managers provide for the evaluation of nuclear hazards, chemical hazards, hydrogenous content of firefighting materials, and m't.igation ofinadvertent i
unsafe acts by individuals. Specifically, when criticality sare": considerations are impacted by these hazards, the approval to operate new faenies or make significant changes, modification, or additions to existing facilities is humented in accord LICENSE SNM-1097 DATE 12/16/96 Page O
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O with established facility practices and conform to configuration management function
' Integrated Safety Analysis' (ISA) requirements described in Chapter 4.0.
j 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.
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-If an analysis is required, the change is not placed into operation until the criticality i
l safety analysis is complete and other preoperational requirements are fulfilled in j
L accordance with established configuration management practices.
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l 6.1.2.2 Role of the Criticality Safety Function Qualified personnel as described in Chapter 2 assigned to the criticality safety l
l 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.
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6.1.3 OPERATING PROCEDURES l
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 l
function. Area Managers assure that appropriate area engineers, operators, and other l
concerned personnel review and understand these procedures through postings, training programs, and/or other written, electronic or verbal notifications.
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.
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l 6.1.4 POSTING AND LABELING 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
.i key criticality safety requirements and limits, to designate approved work and storage i
areas, or to provide instructions or specific precautions to personnel such as:
l Limits on material types and forms.
j Allowable quantities by weight or number.
e Allowable enrichments.
l-Required spacing between units.
l Control limits (when applicable) on quantities such as moderation, density, or l
l presence of additives.
Critical control steps in the operation.
l Storage postings are located in conspicuous places and include as appropriate:
i Material type.
Container identification.
i e' Number ofitems allowed.
Mass, volume, moderation, and/or spacing limits.
e 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, J
t shipping containers, waste boxes / drums, contaminated items, UF6 cylinders containing heels, cold trap cylinders, samples, containers of I liter volume or less, or other containers where labeling is not practical.
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O 6.1.5 AUDITS & INSPECTIONS l
6.1.5.1 Audits and Inspections l
l 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 procedure and personnel approved by the criticality safety function. Findings, l
recommendations, and observations are reviewed with the Environment, Health &
Safety (EHS) function manager to determine if other safety impacts exist. The j
l findings, recommendations, and observations are then transmitted to Area Managers l
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.
1 6.1.5.2 Independent 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 fonn of corrective action requests are tracked to closure.
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6.1.6 CRITICALITY SAFETY PERSONNEL t
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 i
L 2.0.
6.1.6.2 Authority Criticality safety function personnel are specifically authorized to perform assigned l
responsibilities in Chapter 2.0. All nuclear criticality safety function personnel have authority to shutdown potentially unsafe operations.
i 6.2 TECHNICAL PRACTICES 6.2.1 CONTROL PRACTICES r
Criticality safety analyses identify specific controls necessary for the safe and I
effective operation of a process. Prior to use in any process, nuclear criticality safety O
controls are verified against criticality safety analysis criteria The ISA program
-i 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 dinitial functional tests and conduct pre-operational audits to verify j
that the controls function as intended and the installed configuration agrees with the j
criticality safety analysis.
Operations personnel are responsible for subsequent verification of controls through the use of functional testing or verification. When necessary, control calibration and 1
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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
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criticality safety controls designated for a specific process are maintained at the l
original level ofintent and functionality. This requires a combination of routine maintenance, functional testing, and verification of design specifications on a j
j periodic basis. Details of the maintenance program are described in Chapter 3.0.
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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 i
other system controls and are utilized when practical and appropriate. Active engineered controls are the next preferred method of control followed by y
l administrative controls. A criticality safety control must be capable of preventing a i
i criticality accident independent of the operation or failure of any other criticality l
control for a given credible initiating event.
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-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 specific periodic inspections or verification measurement (s) as appropriate.
6.2.2.2 Active Engineered Controls A means of criticality control involving active hardware (e.g., electrical, mechanical, t
j hydraulic) that protect against criticality. These devices act by providing predefined a
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 i
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t system to a safe condition. Human intervention augmented by waming devices and O
interlocks that prevent continued operation may be used to sense a process variable.
1 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 i
a safe condition).
6.2.2.3 Administrative Controls Controls that rely for their implementation on actions, judgment, and responsible I
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 l
proceeding) or reactive (proceeding unless action occurs). Proactive administratvie
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controls are preferred. Assurance is maintained through training, experience, and audit.
l 6.2.3 TABLE OF PLANT SYSTEMS AND PARAMETER CONTROLS j
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:
.s f,
i AREA OR SYSTEM: A defined functional group of processes or pieces of l
equipment that operate as a single unit.
PROCESS SUBAREA OR EQUIPMENT: A defined subgroup of vessels, tanks, process and/or support equipment within an area that operate as a single unit.
BASIS FOR CRITICALITY SAFETY: The controlled parameters established within a CSA for nuclear criticality safety for the identified process subarca or 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).
r NOTE - To be included as section 1.3.15 infinal License: Changesfrom oneparameter to anotherparameterfor process subareas or equipment in which multiple (at least two) parameters are controlled are made in accordance with established change control measures and reported to the NRC within 90 days ofcompletion. Changes to singleparamater controlledprocesses or i
equipmentfrom the identufiedparameter to a new parameter (s) will require NRC approvalprior l
to the change being made.
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OV CSA BOUNDING ASSUMPTIONS: These are the values used for physical process parameters which are not directly controlled but represent the most reactive credible values for the system, process subarea, or equipment under consideration. As such, l
the CSA is performed to consider all process operations and credible upsets that fall within this range of assumptions. For items containing no bounding assumptions, all process operations and credible upsets must be analyzed within the CSA. The i
approved CSA may limit the operation of the system to levels more conservative l
than those permitted by the bounding assumptions.
In the follo ving Table 6.0, unless otherwise specified, the enrichment limit for all processes are 5.0 wt. % U235 (or hie), with the exception of conversion lines 1,2,
j and 4 and related MSG lines 1-6 which are presently analyzed for 4.025 wt. % U235
-l (or LoE). When pails are used for product,5-gallon cans may be used for LoE enrichments, while 3-gallon containers may be used for hie material. All scrap materialis treated as hie.
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l Table 6.0 Plant Systems and Parameter Controls AREA PROCESS BASIS FOR CSA OR SUBAREA OR CRITICALITY BOUNDING ASSUMPTIONS 4
SYSTEM EQUIPMENT SAFETY Fuel Support:
UF6 Cylinder Receipt Enrichment 99.5 wt.% pure UF6 Storage Pads and Storage s 0.5 wt. % H O equivalent 2
Optimal Interunit H O 2
l Scrap 3 and 5-gallon Geometry Homogeneous or Heterogeneous UO2 4
Optimal H O Moderation Container Storage Mass 2
Full Reflection RA-Inner and Outer Geometry Heterogeneous UO2 Optimal H O Moderation Container Storage Moderation 2
Full Reflection Waste Box Container Geometry / Mass Homogeneous UO2 i
Optimal H O Moderation Storage Mass 2
Full Reflection BU-J, BU-7,7A Drum Geometry Homogeneous or Heterogeneous UO2 Optimal H O Moderation Storage Mass 2
Moderation Full Reflection Fuel Support:
Waste Box Load Mass Heterogeneous UO2 Optimal H O Moderation New Decon 2
Full Reflection Oil Drum Load Mass Homogeneous UO2 Optimal H O Moderation 2
Full Reflection Chemical ADU UF6 Cylinders Moderation 99.5 wt.% pure UF.
Conversion System 5 0.5 wt. % H O equivalent 2
Full Reflection Autoclave Moderation 99.5 wt.% pure UF.
Vaporization s 0.5 wt. % H O equivalent 2
Full Reflection Cold Trap System Geometry Homogeneous UO2 Optimal H O Moderation 3
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Moderation 2
Full Reflection Hydrolysis Receiver, Geometry Homogeneous UO F2 2
Optimal H O Moderation Storage, and Scrubber Concentration 2
Tanks Full Reflection i
Sump Geometry Homogeneous UO2 Mass Optimal H2O Moderation Full Reflection l
Precipitation Tanks Geometry Homogeneous UO2 Optimal H O Moderation (Lines 1,2,4) 2 Full Reflection
- two out of any three control parameters required for criticality safety.
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f-AREA PROCESS BASIS FOR CSA l ~
OR SUBAREA OR CRITICALITY BOUNDING ASSUMPTIONS j
SYSTEM EQUIPMENT SAFETY Precipitation Tanks Geometry Homogeneous UO2 j
Optimal H O Moderation (Lines 3,5)
Mass 2
1 Full Reflection -
l Dewatering Geometry Homogeneous ADU or U 0 3
Optimal H O Moderation j
Centrifugation Mass 2
Full Reflection Outside Containment 1
4 j
Clarifying Geometry Homogeneous UO2 OptimalH O Moderation i
Centrifugation Mass 2
i Full Reflection Calcination Geometry Homogeneous UO2 j
Geometry / Mass Optimal H2O Moderation i
Full Reflection I
f Calciner Scrubber Geometry Homogeneous UO2 i
Optimal H O Moderation Concentration 2
I Full Reflection
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3 or 5-Gallon Product Geometry Homogeneous UO2 Optimal H O Moderation Container Mass 2
i Full Reflection UO Powder Geometry or Mass Homogeneous UO2 2
Optimal H O Moderation i
Pretreatment: Mill, Moderation 2
Slug, Granulate (MSG)
Full Reflection LoE and hie UO2 Geometry Homogeneous UO2 Optimal H O Moderation Powder Blending Mass / Moderation 2
Full Reflection j
LoE Fluoride Effluent Geometry Homogeneous UO2 Optimal H O Moderation -
j Vessels Concentration 2
Full Reflection f
Line 3 Geometry Homogeneous UO2 i
Optimal H O Moderation j
Accumulator / Permeate Concentration 2
j Vessels FullReflection Nitrate Quarantine Geometry Homogeneous UO2 Optimal H O Moderation j
Effluent Vessels Concentration 2
Full Reflection j
3 Powder Pack Geometry Homogeneous UO2 Optimal H O Moderation Screener Moderation 2
4 Full Reflection j
j l
Powder Pack Geometry Homogeneous UO2-i Product Container Mass Optimal H2O Moderation j
Full Reflection i
HVAC: Wet Areas Geometry Homogeneous UO2 Optimal H O Moderation Mass 2
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AREA PROCESS BASIS FOR CSA g
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OR SUBAREA OR CRITICALITY BOUNDING ASSUMPTIONS SYSTEM EQUIPMENT SAFETY HVAC: Dry Areas Mass Homogeneous UO2 q
Optimal H O Moderation Moderation 2
Full Reflection j
i Exhaust Scrubber Geometry / Mass Homogeneous UO2 OptimalH O Moderation Mass 2
Full Reflection i
Utilities: Steam, N,
Mass.
Backflow into large supply vessels 2
H, Dissoc. NH4, H O prevented by backflow prevention 2
2 l
Supply measures, physical barriers, and/or process characteristics.
1 REDCAP: Oxidation Geometry Heterogeneous UO2 Optimal H O Moderation f
Feed Containers Mass 2
j Full Reflection j
4 j
REDCAP: Oxidation Geometry Heterogeneous UO2
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OptimalH O Moderation Furnace Moderation 2
j Full Reflection J
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REDCAP: Oxidation Geometry Homogeneous UO2 Optimal H O Moderation j
Output Containers Mass 2
Full Reflection 1
REDCAP: Oxidation Geometry Homogeneous UO2 Optimal H O Moderation j
Off-Gas System Mass 2
Full Reflection i
Miscellaneous: 3 and Geometry Homogeneous or Heterogeneous UO j q 2
5-Gallon Container Mass Optimal H2O Moderation Q
Full Reflection
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Floor storage Uranium Recovery Unit Fluoride Waste Process Geometry Homogeneous UO2 i
Optimal H O Moderation j
(URU) System Vessels Concentration 2
Full Reflection l
Fluoride Waste Concentration Homogeneous UO2 l.
Surge Vessel Mass Optimal H2O Moderation (V 106)
Full Reflection l
Radwaste Process Geometry Homogeneous UO2 Optimal H O Moderation Vessels Concentration 2
Full Reflection
- 4 Nitrate Waste Process Geometry Homogeneous UO2 Optimal H O Moderation Vessels Concentration 2
Full Reflection i
j Nitrate Waste Concentration Homogeneous UO2 1
Optimal H O Moderation Surge Vessel Mass 2
(V 103)
Full Reflection Oxidation Feed Geometry Heterogeneous UO2 Optimal H O Moderation i
Containers Mass 2
i Full Reflection i
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AREA PROCESS BASIS FOR CSA l
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OR SUBAREA OR CRITICALITY BOUNDING ASSUMPTIONS SYSTEM EQUIPMENT SAFETY l
Oxidation Furnace Geometry Heterogeneous UO2 Optimal H O Moderation 2
1 Full Reflection Oxidation Furnace Geometry Heterogeneous UO2 Optimal H O Moderation Boat Dump Moderation 2
FullReflection Oxidation 3 gallon Geometry Heterogeneous UO2 Optimal H O Moderation Container Storage Mass 2
Moderation Full Reflection Oxidation Off-Gas Geometry Heterogeneous UO2 Optimal H O Moderation
.l System Mass 2
Full Reflection j
Dissolution: Can Dump Geometry 1
Heterogeneous UO2 Optimal H O Moderation Feed Conveyor Mass J*
2 Moderation Full Reflecten Dissolution:Dissolvers, Geometry Heterogeneous UO2 Optimal H O Moderation Pumps, Sumps, Filters, Concentration 2
Full Reflection i
Piping l
Oberlin Filter Geometry Heterogeneous UO2 Concentration Optimal H2O Moderation Full Reflection Dissolution: NOX Concentration Homogeneous UO2 Scrubber Mass On-Line Density Meter Full Reflection g
Counter-Current Geometry Heterogeneous UO2 Leaching: Can Dump Mass / Moderation Optimal H2O Moderation Full Reflection
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Counter-Current Geometry Heterogeneous UO2 Optimal H O Moderation Leaching: Leach Concentration 2
Troughs, Pumps, Full Reflection Filters, Storage Tanks, Product Containers Utilities: Steam, DI Mass Backflow into large supply vessels H20, Nitric Acid, prevented by backflow prevention Aluminum Nitrate measures, physical barriers, and/or process characteristics.
Head-End Geometry Homogeneous UNH Optimal H O Moderation Concentrator Process Concentration 2
Full Reflection
- two out of any three control parameters required for criticality safety.
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AREA PROCESS BASIS FOR CSA mj OR SURAREA OR CRITICALITY BOUNDING ASSUMPTIONS SYSTEM EQUIPMENT SAFETY Solvent Extraction Geometry Homogeneous UO2 Optimal H O Moderation Process Concentration 2
Full Reflection f
UNH Product Storage Geometry Homogeneous UNH Optimal H O Moderation Vessels Concentration 2
Full Reflection Waste Solvent Drum Mass Homogeneous UO2 Optimal H O Moderation Load 2
Full Reflection Uranyl Nitrate UNH LEM Tank Feed Geometry Homogeneous UO2 Optimal H O Moderation f
Conversion (UCON)
Tanks Concentration 2
Full Reflection System UCON: Precipitation Geometry Homogeneous UNH Optimal H O Moderation i
Tanks Mass 2
Full Reflection UCON: Dewatering Geometry Homogeneous ADU or U 0 3
Centrifugation Mass Optimal H2O Moderation Full Reflection Outside Containment UCON: Clarifying Geometry Homogeneous UO2 Optimal H O Moderation Centrifugation Mass 2
Full Reflection UCON Process:
Geometry Homogeneous UO2 Optimal H O Moderation Calcination Geometry / Mass 2
(d_)
Full Reflection i
Waste Treatment Fluoride Waste Concentration Homogeneous UO2 Optimal H O Moderation Facility (WTF)
Barrens Surge Vessel Mass 2
Full Reflection (V-108)
Nitrate Waste Barrens Concentration Homogeneous UO2 Optimal H O Moderation Surge Vessel (V-104)
Mass 2
Full Reflection Centrifuge Geometry Homogeneous UO2 Optimal H O Moderation Mass 2
Full Reflection Oberlin Filter Geometry / Mass Homogeneous UO2 Optimal H O Moderation Concentration 2
Full Reflection Uranium Recovery from URLS Process Tanks Concentration Homogeneous UO2 Optimal H O Moderation 2
Lagoon Sludge (URLS)
Full Reflection FacilityProcess URLS Process Non-Geometty/Concent.
Homogeneous UO2 Optimal H O Moderation Leach Filter Press Concentration 2
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OR SUBAREA OR CRITICALITY BOUNDING ASSUMPTIONS SYSTEM EQUIPMENT SAFETY URLS Process Product Concentration Homogeneous UO2 Optimal H O Moderation Waste Container Mass 2
Full Reflection Waste Oxidation /
Incinerator Mass (Box Monitor)
Heterogeneous UO2 Optimal H O Moderation Reduction (Incineration)
Combustible Box Feed Mass (E-Gun) 2 Facility Containers Full Reflection Incinerator Mass (UPHOLD)
Heterogeneous UO2 Optimal H O Moderation Mass (INHOLD) 2 Full Reflection Incinerator Product 3 Geometry Homogeneous UO2 Optimal H O Moderation or 5-Gallon Containers Mass 2
Full Reflection Dry Conversion Process UF6 Cylinder Receipt Enrichment 99.5 wt.% pure UF6 (DCP) Conversion and Storage s 0.5 wt. % H O equivalent 2
OptimalInterunit H O 2
Vaporization Moderation 99.5 wt.% pure UF6 Autoclave w/ UF6 s 0.5 wt. % H O equivalent 2
Cylinder Full Reflection Vaporization Geometry Homogeneous UO2 Optimal H O Moderation Cold Trap System Moderation 2
Full Reflection Conversion:
Moderation Homogeneous UO2 Reactor / Kiln Maximum Credible UO Density 2
(,n)
Maximum Credible wt. % H O 2
Full Reflection b'
Conversion:
Moderation Homogeneous UO2 Powder Outlet Box Maximum Credible UO Density 2
Maximum Credible wt. % H O 2
Full Reflection PowderOutlet:
Moderation Homogeneous UO2 Cooling Hopper Maximum Credible UO Density 2
Maximum Credible wt. % H O 2
Full Reflection Powder Transfer &
Moderation Homogeneous UO2 Storage: Nonnal Maximum Credible UO Density l
2 Product Container Maximum Credible wt. % H O 2
Full Reflection Powder Transfer &
Geometry Homogeneous UO2 Storage: Out-of-Spec Moderation Maximum Credible UO Density 2
Moisture Product Maximum Credible wt. % H O 2
Container Full Reflection 1
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AREA PROCESS BASIS FOR CSA OR SUBAREA OR CRITICALITY BOUNDING ASSUMPTIONS SYSTEM EQUIPMENT SAFETY Homogenization Moderation Homogeneous UO2 Maximum Credible UO Density 2
Maximum Credible wt. % H O j
2 Full Reflection i
f
- Blending, Moderation Heterogeneous UO2 Precompaction, Maximum Credible UO Density 2
Granulation Maximum Credible wt.% H O 2
4 Full Reflection Tumbling:
Moderation Heterogeneous UO2 l
in Powder Container Maximum Credible UO Density l
2 Maximum Credible wt. % H O 2
Full Reflection Powder Pack Moderation Hoterogeneous UO2 i
Screener Maximum Credible UO Density l
2 I
Maximum Credible wt. % H O 2
Full Reflection Powder Pack Geometry Homogeneous UO2 Optimal H O Moderation 4
Product Container Mass 2
Full Reflection Utilities: N, H2, H O Mass Backflow into large supply vessels not 2
2 Supply, Refrigerant credible due to backflow prevention measures, physical barriers, and/or process characteristics.
HF Effluent Recovery Geometry Homogeneous UO2 OptimalH O Moderation and Storage Vessels Mass 2
FullReflection Dry Recycle Facility Feed 3 aid 5-gallon Geometry Heterogeneous UO2 OptimalH O Moderation Facility Containers Mass 2
Full Reflection l
l Feed 3 and 5-gallon Geometry 1
Heterogeneous UO2 Container Storage Mass J*
Optimallnterunit H O Moderation 2
Moderation Full Reflection l
Recycle Furnace Geometry Heterogeneous UO2 f
Optimal H O Moderation Moderation 2
Full Reflection Recycle DM 10 Moderation Heterogeneous UO2 1
Vibromill(MRA)
Maximum Credible wt. % H O i
2 Full Reflection Recycle Screener Geometry Heterogeneous UO2 Optimal H O Moderation l
Moderation 2
Full Reflection j
- two out of any three control parameters required for criticality safety.
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AREA PROCESS BASIS FOR CSA C/
OR SUBAREA OR CRITICALITY BOUNDING ASPUMPTIONS SYSTEM EQUIPMENT SAFETY Recycle Blender Moderation Heterogeneous UO2 MaximumCredibleUO Density 2
Maximum Credible wt. % H O 2
Full Reflection Recycle Unicone Moderation Heterogeneous UO2 Maximum Credible UO Density 2
Product Maximum Credible Internal wt.% H O 2
Container / Storage OptimalInterunit H O 2
Recycle 3-Gallon Geometry 1*
Heterogeneous UO2 Optima.1 H O Moderation Product Container /
Mass f
2 Full R flection Storage Moderation t
Press Warehouse Conveyor Storage:
Geometry 1*
Homopeous UO.
Facinty Process 3 and 5-gallon Cans Mass J
OptimalInterunit H O Moderation 2
Moder-tion Full Reflection Powder Dump Transfer Geometry Homogeneous UO2 Optimal H O Moderation Hopper / Chute Moderation 2
Full Reflection Pellet Presses Geometry / Mass Heterogeneous UO2 Optimal H O Moderation Moderation 2
Full Reflection Press Lubricant Sump Geometry Heterogeneous UO2 Optimal H O Moderation Mass 2
Full Reflection g)
Press: Green Pellet Geometry Heterogeneous UO2
(
Optimal H O Moderation Boat Product Container Moderation 2
'~
Full Reflection 3-gallon Powder Geometry Heterogeneous UO2 OptimalH O Moderation Cleanup Container Mass 2
Full Reflection Pellet Sintering System Feed / Exit Conveyors Geometry Heterogeneous UO2 Optimal H O Moderation Moderation 2
Full Reflection Sintering Furnace Geometry Heterogeneous UO2 Moderation Optimal H2O Moderation Full Reflection Pellet Grinding System Feeder Hopper Bowl or Geometry Heterogeneous UO2 Flat Feeder Table Moderation Optimal H2O Moderation Full Reflection Grinder Geometry Heterogeneous UO2 Moderation Optimal H2O Moderation Full Reflection
- two out of any three control parameters required for criticality safety.
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AREA PROCESS BASIS FOR CSA h
(,/
OR SUBAREA OR CRITICALITY BOUNDING ASSUMPTIONS SYSTEM EQUIPMENT SAFETY Grinder APITRON Geometry Homogeneous UO2 Filter Moderation OptimalH2O Moderation Full Reflection
/
Grinder Swarf 3-Geometry Heterogeneous UO2 Optimal H O Moderation Gallon Container Moderation 2
Full Reflection Grinder Hardscrap 3-Geometry Heterogeneous UO2 Optimal H O Moderation Gallon Container Mass 2
Full Reflection
)
Grinder Pellet Product Geometry
}
Heterogeneous UO2 Optimal H O Moderation Tray Mass f*
2 Moderation Full Reflection l
Pellet Transfer CrA Geometry Heterogeneous UO2 Moderation OptimalInterunit H O Moderation 2
Full Reflection Rod Load, Out-Gassing, Rod Load, Out-Geometry Heterogeneous UO2 Optimal H O Moderation and Final Rod Welding Gassing, and Final Rod Moderation 2
System Weld Full Reflection Pellet Storage Cabinet Geometry Heterogeneous UO2 Optimal H O Moderation Moderation 2
Full Reflection Rod Storage Cabinet Geometiy Heterogeneous UO2 Optimal H O Moderation Moderation 2
Full Reflection Gadolinia Shop
' ress, Sintering, Similar to UO Shop SimilartoUO Shop Above 2
2 ainding, Rod Load, Above Rod Storage, & Outgas Gadolinia 3 and 5-Geometry Homogeneous UO2 Optimal H O Moderation Gallon Feed Containers Mass 2
Full Reflection Gadolinia 3 and 5-Geometry
}
Homogeneous UO2 OptimalH O Moderation Gallon Feed & Product Mass J*
2 Container Storage Moderation Full Reflection Gadolinia Unicone Mass Homogeneous UO2 Feed Container Moderation Maximum Credible UO Density 2
Maximum Credible wt. % H2O Full Reflection Gadolinia DM 10 Geometry Heterogeneous UO2 Vibromill(MCA)
Moderation Optimal H2O Moderation Full Reflection
- two out of any three control parameters required for criticality safety.
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AREA PROCESS BASIS FOR CSA I
OR SUBAREA OR CRITICALITY BOUNDING ASSUMPTIONS SYSTEM EQUIPMENT SAFETY Gadolinia DM-3 Mass Homogeneous UO2 Optimal H O Moderation Vibromill(MCA)
Moderation 2
Full Reflection Gadolinia DM-10 Moderation Heterogeneous UO2 Vibromill(MRA)
Maximum Credible wt. % H O 2
Full Reflection Pellet Storage:
Geometry / Mass Heterogeneous UO2 Optimal H O Moderation
]
Ministacker Moderation 2
Full Rettedon j
Bundle Assembly Rod Trays Geometry Heterogeneous UO2 i
Mass OptimalInterunit H2O Moderation Full Reflection 4
Rod Storage Cabinets Geometry Heterogeneous UO2 2
Moderation OptimalInterunit H2O Moderation Full Reflection Rod Tray Transfer Geometry Heterogeneous UO2 Vehicle:" Big Joe" Moderation OptimalInterunit H O Moderation 2
Full Reflection Magnetic and Passive Geometry Heterogeneous UO2 l
Scanner: " MAPS" Moderation Optimal Interunit H O Moderation 2
Full Reflection Bundle Accumulator:
Geometry Heterogeneous UO2 "BACC" Moderation Optimal interunit H2O Moderation FullReflection Automatic Bundle Geometry Heterogeneous UO2 Assemble Machine:
Moderation OptimalInterunit H2O Moderation "ABAM" Full Reflection Rod Scanner:
Geometry Heterogeneous UO2 1
" Fat Alben" Moderation OptimalInterunit H O Moderation 2
Full Reflection Assembly Table Geometry Heterogeneous UO2 Moderation Optimalinterunit H O Moderation 2
Full Reflection Upender: Bundle and Geometry Heterogeneous UO2 RA Container Moderation OptimalInterunit H O Moderation 2
Full Reflection inspection Pit Geometry Heterogeneous UO2 Moderation Optimalinterunit H O Moderation 2
Full Reflection bu$dih Storage:
Geometry Heterogeneous UO2
" Forest" Moderation Optimal Interunit H O Moderation 2
i Full Reflection 4
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/7 AREA PROCESS BASIS FOR CSA V
OR SUBAREA OR CRITICALITY BOUNDING ASSUMPTIONS SYSTEM EQUIPMENT SAFETY RA Container:
Geometry Heterogeneous UO2 Transfer Port & RA Moderation OptimalInterunit H O Moderation l
2 Conveyor Full Reflection l
Red Scanner:
Geometry Heterogeneous UO2 X-Ray-Unit Moderation OptimalInterunit H O Moderation 2
Full Reflection
~
Rod Inspection:
Geometry Heterogeneous UO2 Surface-Plate Moderation OptimalInterunit H O Moderation 2
Full Reflection i
Rod Movement:
Geometry Heterogeneous UO2 i
One & Two-Tray Cart Moderation OptimalInterunit H O Moderation 2
FullReflection i
Container Storage:
Geometry Heterogeneous UO2 RA-Inner / Outer Moderation OptimalInterunit H O Moderation f
2 Storage Full Reflection i
i Decontamination &
Wash Down Areas, Geometry / Mass Homogeneous UO2 Optimal H O Moderation Volume Reduction Sumps, Bag Filters Mass 2
Full Reflection Facility (DVRF)
Dust Hog Mass Homogeneous UO2 Optimal H O Moderation 2
Full Reflection HVAC Geometry Homogeneous UO2 Optimal H O Moderation Mass 2
-s Full Reflection l
3-Gallon Waste Geometry Homogeneous UO2 Optimal H O Moderation Container Storage Mass 2
Full Reflection 2
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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 geoemtry margins of safety for units identified in this table are 93% of the minimum critical cylinder diameter, 88% of the minimum critical slab thickness, i
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 wbich incorporate validated K-infinities, migration 2
areas (M ) and extrapolation distances; or explicit stochastic or deterministic i
modeling methods.
.l 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 e
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 2
water as a function of U235 enrichment over the range of 1.1% to 15% for i
O uncontrolled geometric configurations. The safe batch sized for UO of specific 2
V compounds may be adjusted when applied to other compounds by the formula:
kgs UO x 0.88 = kg X of 2
where f = wt. % U in compound X Where engineered controls prevent over batching, a mass of 75% of the e
minimum critical mass shall not be exceeded.
Subject to provision for adequate protection against precipitation or other circumstances which may increas concentration, the following safe concentrations are specifically licensed for use at the GE-Wilmington facility:
A concentration of less than or equal to one-half of the minimum critical concentration.
A system in which the hydrogen to U235 atom ratio (H/U235) is greater than 5200.
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Table 6.1 Safe Geometry Values s
Homogeneous UO -
Weight Percent Infinite Cylinder
- Infinite Slab
- Sphere Volume
- 2 H O Mixtures U235 Diameters Thickness 2
(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.M 12.90 6.65 51.0 3.00 12.35 6.25 44.0 3.25 11.70 5.90 38.5 3.50 11.20 5.60 34.0 3.75 10.80 5.30 31.0 4.00 10.50 5.10 29.0 i
5.00 9.50 4.45 24.0 6.00 8.95 4.00 18.5 7.00 8.45 3.75 17.0 Homogeneous Weight Percent infinite Cylinder Infinite Stab Sphere Volume Aqueous U235 Diameters Thickness Solutions (Inches)
(Inches)
(Liters) 2.00 9.30 106.4 2.25 8.40 80.5 1
2.50 7.80 66.8 2.75 7.30 56.2 3.00 7.00 49.7 3.25 6.70 44.8 3.50 6.50 41.0 3.75 6.30 38.0 0s 4.00 6.00 34.9 5.00 4.80 26.0 6.00 4.40 22.5 7.00 4.10 19.5 Heterogeneous Weight Percent Infinite Cylinder Infinite Slab Sphere Volume Mixtures or U235 Diameters Thickness Compounds (loches)
(Inches)
(Liters) 2.00 I l.10 5.60 35.7 2.25 10.50 5.10 30.7 2.50 10.10 4.80 27.3 2.75 9.70 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 6.00 7.90 3.50 11.0 7.6 7.00 7.40
- These salues 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.
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l f'h Table 6.2 Safe Batch Values for UO and Water
- 2 j
l Nominal Weight Homogeneous Heterogeneous NominalWeight l Homogeneous Heterogeneous Percent U235 UO Powder &
UO Pellets &
Percent U235 UO Powder &
UO Pellets &
2 2
Water Water Water Water Mixtures Mixtures Mixtures Mixtures (Kas UO2)
(Kas UO )
(Kas UO )
(Kgs UO )
2 2
2 1.10 2629.0 510.0 4.00 -
25.7 24.7 1
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 18.1 18.I 1.70 225.0 116.0 5.50 15.4 15 4 1.80 183.0 102.0 6.00 13.8 13.8 l.90 150.6 90.5 7.00 8.3 8.3 i
2.00 127.5 81.6 8.00 6.9 6.9 2.10 109.2 73.I 9.00 5.9 5.9 l
2.20
%.8 66.4 10.00 5.1 5.1 j
2.30 84.3 61.0 Il.00 4.4 4.4 2 40 74.7 56.1 12.00 3.9 3.9 l
2.50 68.9 52.1 I3.00 3.5 3.5 2.60 60.5 48.8 14.00 3.3 3.3 2.70 56.6 45.4 15.00 3.0 3.0 2 80 52.2 42.9 2.90 47.6 40.1 3.00 44.5 38.1 l
3.20 38.9 34.1 V
3.40 34.6 31.0 j
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.
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l 6.2.5 CONTROL PARAMETERS Nuclear criticality safety is achieved by controlling one or more parameters of a i
system within established suberitical limits. 'Ihe criticality safety review process is l
l used to identify the significant parameters associated with a particular system. All l
assumptions relating to process equipment, material composition, function, and j
operation, including upset conditions, are justified, documented, and independently i
i reviewed.
Identified below are specific control parameters that may be considered during the
[
review process:
j i.
6.2.5.1 Geometry - Geometry may be used for nuclear criticality safety control on its own or j
in combination with other control methods. Favorable geometry is based on limiting dimensions of defined geometrical shapes to established suberitical limits. Structure j
and/or neutron absorbers that are not removable constitute a form of geometry i
control. At the GE-Wilmington facility, favorable geometry is developed j
l conservatively assuming unlimited water or concrete equivalent reflection, optimal j
j hydrogenous moderation, worst credible heterogeneity, and maximum credible enrichment to be processed. Examples include cylinder diameters, annular i
inner / outer dimensions, slab thickness, and sphere diameters.
Geometry control systems are analyzed and evaluated allowing for fabrication tolerances and dimensional changes that raay likely occur through corrosion, wear, or
)
mechanical distortion. In addition, these systems include provisions for periodic j
[
inspection if credible conditions exist for changes in the dimensions of the equipment i
that may result in the inability to meet established nuclear criticality safety limits.
i' l
'6.2.5.2 Mass - Mass control may be used for a nuclear criticality safety control on its own or in ec.ibination with other control methods. Mass control may be utilized to limit the j.
qu.ratity of uranium within specific process operations or vessels and within storage, j
transportation, or disposal containers. Analytical or non-destructive methods may be i
)
employed to verify the mass measurements for a specific quantity of material.
t J
Establishment of mass limits involves consideration of potential moderation, reflection, geometry, spacing, and material concentration. The criticality safety i
I analysis considers normal operations and credible process upsets in determining l
actual mass limits for the system and for defining additional controls. ' When only administrative controls are used for mass controlled systems, double batching is j
l considered to ensure adequate safety margin.-
i; j
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6.2.5.3 Moderation - Moderation control may be used for nuclear criticality safety control L
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 j
area'. When moderation control is the primary design focus and is designated as a j
the primary criticality safety control parameter, the area is posted ' moderation restricted area'.
l When moderation is the primary criticality safety control parameter the following l
graded approach to the design control philosophy is applied in accordance with l
{
established facility practices (in decreasing order of restriction):
j At each enriched uranium interface involving intentional and continuous l
l e
introduction of moderation (e.g., insertion of superheated steam into reactor),
l
?
at least three controls are required to assure that the moderation safety factor i
j is not exceeded. At least two of these controls must be active engineered j
j controls.
l
]
At enriched uranium interfaces involving intentional but non-continuous e
introduction of moderation at least three controls are required to assure that i
the moderation safety factor is not exceeded. ' At least one of these controls l
must be an active engineered control, unless a moderation safety factor greater than 3 is demonstrated.
l For situations where moderation is not intentionally introduced as part of the e
j' process, the required number of controls for each credible failure mode must be established in accordance with the double contingency principle.
When the maximum credible accident is considered, the safety moderation limit (i.e.,
s
% H O or equivalent) must provide sufficient factor of safety above the process 2
3 l
moderation limit. This *1noderation safety factor', which is the ratio of the safety l
moderation limit to the process moderation limit, will normally be three or higher, l
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 1
process moderation limit to the safety moderation limit.
In some cases, as described above, increased depth of protection may be required, but j
the minimum protection is never less than the following: two independent controls i
[
prevent moderator from entering the system through a defined interface and must fail j
before a criticality accident is possible. The quality' and basis for selection of the i
controls is documented in accordance with Integrated Safety Analysis process described in Chapter 4.0. Controls for the introduction and limited usage of
~j-moderating materials (e.g. for cleaning or lubrication purposes) within areas in which 3
1 L
i i
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}
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,i I
the primary criticality safety parameter is moderation are approved by the criticality safety function.
1 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 -
l maintained within defined limits for the system. When used on its own, two independent controls / measurements or the analysis of two independent samples are utilized to document this compliance. The system for collecting, preparing, analyzing, and posting of results pertaining to sample evaluation are designed to i
ensure the results obtained are independent. Controls are established to prevent l
unacceptable concentration increases within the defined system after initial process operation. Each process relying on concentration control has in place controls i
necessary to detect and/or mitigate the effects ofinternal concentration within the
'l system (e.g., Dynatrol density meter, Rhonan density meter, etc.), otherwise, the most reactive credible concentration (density) is assumed.
J 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 i
I
/O dimensional relationship to the fissionable material are recommended. The use of O
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:
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.
i The neutron absorber content is certified.
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i i
For fixed neutmn absorbers used as part of a geometry control, the following requirements apply:
1 The composition of the absorber are measured and documented prior to first l
e i
use.
Periodic verification of the integrity of the neutron absorber system j
e subsequent to installation is performed on a scheduled basis approved by the j
criticality safety function. The method of verification may take the form of
-l traceability (i.e. serial number, QA documentation, etc.), visual inspection or direct measurement.
l 6.2.5.6 Spacing (or Unit Interaction) - Criticality safety controls based on isolation or l
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 l
j e
the larger of 12-foot air distance or the greatest distance across an i
e orthographic projection of the largest of the fissile accumulations on a plane perpendicular to the line joining their centers.
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 (provided 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.
Techniques which produce a calculated effective multiplication factor of the entire system (e.g., validated Monte Carlo or Sn 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 j
of compositions used in subsequent system medeling activities.
i LICENSE SNM-1097 DATE 12/16/96 Page
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4 e
It is important to distinguish between homogeneous and heterogeneous system j
conditions. Heterogeneous effects within a system can be significant and therefore
]
must be considered within the criticality safety analysis when appropriate.
i Evaluation of systems where the particle size varies take into consideration effects of heterogeneity appropriate for the process being analyzed.
l 1
{
6.2.5.8 Reflection - Most systems are designed and operated with the assumption of 12-inch q
i water or optimum reflection. However, subject to approved controls which limit l
reflection, certain system designs may be analyzed, approved, and operated in j
situations where the analyzed reflection is less thaa optimum.
In criticality safety analysis, the neutron reflection properties of the credible process j
j environment are considered. For example, reflectors more effective than water (e.g.,
j l
concrete) are considered when appropriate.
l 6.2.5.9 Enrichment - Enrichment control may be utilized to limit the percent U-235 within a i
process, vessel, or container, thus providing a method for nuclear criticality safety control. Active engineered or administrative controls are required to verify i
[
enrichment and to prevent the introduction of uranium at unacceptable enrichment j
l levels within a defimed subsystem within the same area.~ In cases where enrichment control is not utilized, the maximum credible area enrichment is utilized in the
- I criticality safety analysis.
- O 6.2.5.10 Process Characteristics - Within certain manufacturing operations, credit may be j
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 i
following requirements:
l The bounding conditions and operational limits are specifically identified in e
the criticality safety analysis and, are specifically communicated, through training and procedures, to appropriate operations personnel.
1 Bounding conditions for such process and/or material characteristics are j
e j
based on established physical or chemical reactions, known scientific l
principles, and/or facility-specific experimental data supported by operational history.
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.
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- C 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 i
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.
J 6.3 CONTROL DOCUMENTS t
6.3.1 CRITICALITY SAFETY ANALYSIS (CSA)
In accordance with ANSI /ANS-8.19 (1984), the criticality :.afety analysis is a collection ofinformation that "provides sufficient detail clarity, and lack of ambiguity to allow independentjudgment of the results." The CSA documents the i
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.
The CSA addresses the specific concerns (event sequences) cf 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 i
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.
I the system being analyzed and includes applicable information requirements as follows:
Scope - This element defines the stated purpose of the analysis.
a i
General Discussion - This element presents an overview of the process that f
e is affected by the proposed change. This section includes as appropriate; process description, flow diagrams, normal operating conditions, system j
interfaces, and other important to design considerations.
j I
Criticality Safety Controls / Bounding Assumptions - This element defines e
a minimum of two criticality safety controls that are imposed as a result of the I
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i analysis. This section also clearly presents a summary of the bounding
- g j\\
assumptions used in the analysis. Bounding assumptions include; worst i
credible contents (e.g., material composition, density, enrichment, and i-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, subarcas and/or J
4 areas 3
Model Description - This element presents a narrative description of the e
[
actual model used in the analysis. An identification of both normal and credible upset (accident condition) model filenaming convention is provided.
l Key input listings and corresponding geometry plot (s) for both normal and
[
credible upset cases are also provided.
1.
Calculational Results - This element identifies how the calculations were l
performed, what tools or reference documents were used, and when j
appropriate, presents a tabular listing of the calculational result and associated uncertainty (e.g., Keff + 30) results as a function of the key parameter (s)
(e.g., wt. fraction H2O). When applicable, the assigned bias of the 4-calculation is also clearly stated and incorporated into both normal and/or -
{
accident limit comparisons l
Safety During Upset Conditions - This element presents a concise summary j
e j (
of the upset conditions considered credible for the defined unit or process
- J' 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.
Specifications and Requirements for Safety - When applicable, this
)
e i
element presetu both the design specifications and the criticality safety i
requirements for correct implementation of the established controls. These j
requirements are incorporated into operating procedures, t aining, maintenance, quality assurance as appropriate to implement the specifications i
and requirements.
Compliance - This element concludes the analysis with pertinent summary j
l e
l statements and includes a statement regarding license compliance.
j Verification - Each criticality safety analysis is verified in accordance with e
i section 6.3.2.5 by a senior engineer approved by the criticality safety function j
j.
and who was not involved in the analysis.
1 i
f T
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I-
.,s Appendices - Where necessary, a summary ofinformation ancillary to
(
e calculations such as parametric sensitivity studies, references, key inputs, model geometry plots, equipment sketches, useful data, etc., for each defined system is included.
I 6.3.2 ANALYSIS METHODS
?
l 6.3.2.1 KeffLimit 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 j
calculational uncertainties, that is:
Keff + 3e - bias s 0.97 (accident conditions).
f Thus, the established delta k safety margin used at the GE-Wilmington facility is 3
0.03.
J Normal operating conditions include maximum credible conditions expected to be encountered when the criticality control systems function properly. Credible process
. O upsets include anticipated off-normal or credible accident conditions and must be
- 'V 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 j
system such that adequate criticality safety controls are defined for the analyzed system.
l 6.3.2.2 Analytical Methods Methodologies currently employed by the GE-Wilmington criticality safety function 1
i include hand calculations utilizing published experimental data (e.g., ARH-600 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 So l
Discrete Ordinates codes (e.g., ANISN or XSDRNPM) may be used after validation as described in subparagraph (c) below.
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,s l
l l
GEKENO (Geometry Enhanced KENO) is a multigroup Monte Carlo program which l
solves the neutron transport equation in 3-dimensional space. The GEKENO criticality program utilizes the 16-energy group Knight-Modified Hansen Roach -
l cross-section data set, and a potential scattering o resonance correction to p
compensate for flux depression at resonance peaks. GEKENO is normally used for
)
homogeneous systems. For infinite systems, K. can be calculated directly from the
.l Hansen Roach cross-sections using the program KINF.
j GEMER (Geometry Enhanced merit) is a multigroup Monte Carlo program which solves the neutron transport equation in 3-dimensional space. The GEMER j
criticality program is based on 190-energy group structure to represent the neutron l
energy spectrum. In addition, GEMER treats resolved resonances explicitly by j
tracking the neutron energy and solving the single-level Breit-Wigner equation at i
each collision in the resolved resonance range in regions containing materials whose l
resolve resonances are explicitly represented. The cross-section treatment in l
GEMER is especially important for heterogeneous systems since the multigroup treatment does not accurately account for resonance self-shielding.
L 6.3.2.3 Validation Techniques Experimental critical data or analytical methods which have been validated (benchmarked) by comparison with experimental critical data in accordance with O
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 established:
the type of systems which can be modeled e
the range of parameters which may be treated
-j e
the bias, if any, which exists in the results produced by the method.
e Currently GEMER is validated against 123 critical experiments and GEKENO is
- i 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-l sigma confidence band (see Figures 6.1 and 6.2). The bias (Ke - 1.0) is applied over l
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 enrichments up to j
6% wt. % U235. The range of applicability through 6 wt. % U235 enrichment is justified as follows:
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e>~
The bias fits used for criticality safety anlaysis at GE-Wilmington are based on the O-hydrogen-to-uranium 235 atom ratio, and an increase in enrichment simply reduces the relative amount of U238. Critical benchmarks used to validate GEKENO range i
from 1.4%-5.0% while GEMER benchmarks range from 1.4%-5.0%, but include 40
)
experiments above 90%.
For a given H/U235 atom ratio, a change in enrichment from 1.4% to 5.0%
l effectively reduces the U238 content by about 72% [or (5.0 - 1.4)/5.0). Likewise, a change in enrichment from 1.4 to 6.0% effectively reduces the U238 content by J
about 77% [ or (6.0 - 1.4)/6.0]. Therefore, this small extrapolation to 6% is l
acceptable for well moderated systems since, for these systems, a wide range of enrichment and moderation is included in the critical benchmarks.
]
)
GEMER's detailed cross section representation allows a single well-behaved bias to be developed over the entire range of enrichment and moderation. GEMER critical j
benchmarks include several low H/U235 atom ratios, thus a reasonable basis exists for extrapolating GEMER up to 6% enrichment for very low H/U235 atom ratios provided no positive biases are applied.
GEKENO which uses the Knight-Modified Hansen-Roach cross sections does not smoothly transition between well moderated and unmoderated systems. GEKENO critical benchmarks include only one low H/U235 atom ratio. Thus for GEKENO, there is no mathematical basis for extrapolation of these data to 6% (i.e., since only one data point is available). However, GEKENO can be used at 6% based on its comparison to GEMER and the extra margin (relative to GEMER) at very low moderation content that results from excluding positive biases (refer Figures 6.1 and i
6.2).
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i 4
O FICURE 6.1 - CENER BIAS DETERNINATION, P6RTICLE NEIDIT 1.10 l
LE$tND 124 DATA Sgt
. PARTICLE W IGHT i
M 3RD ORDER FIT OF LIMIT '
l
- g.gtr a 1,3 1.46 LINEAR FITS OR0f ts 2 99.782 CorFIDENCE DAND l
1.06 1
a 1.04 K-tFF ils 1.02 4
w A'.
1.90 ig.
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O i
f 0.969
-10 to le 40 119 14e 174
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FICURE 6.2 - CEKEHO BIRS CALCUIATION s,se g
LEGEND
(
stKEN0 wa: ION..
e 56 DATA POINTS x 8A0 ORDER FIT OF LIMIT 1...
= KEFF e 1..
LtNEAR FITS OR0tte 2
...F.I CONF 10tNCE.AND
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K-EFF
$30 I o y
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MYDR0 GEN-70-U285 N1."
6.3.2.4 Computer Software & Hardware Configuration Control j
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 approved configuration control program described in Chapter 3.0 that addresses both hardware 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 j
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a e' b e.
3 and analytical methods being used to calculate bias and uncertainty values to be O
applied to the calculational results.
Each individual workstation is verified to produce results identical to the i
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 software. Modifications to hardware or software 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 j
l re-qualification of the code prior to release for use.
63.2.5 Technical Reviews Independent technical reviews of proposed criticality safety control limits specified i
in criticality safety analyses are performed. A senior engineer within the criticality i
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 analyzed. In addition, the reviewer verifies that the proposed material characterizations such as density, concentration, etc., adequately represent the j
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 calculations with an alternate computational method.
Verify the calculations 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.
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Based on one of these prescribed methods, the independent technical review provides
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a reasonable measure of assurance that the chosen analysis methodology and results are correct.
6.4 CRITICALITY ACCIDENT ALARM SYSTEM l
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 l
density building materials such as brick, concrete, or cinder block shield a potential l
accident area from the detector. Low density materials of construction such as 1
wooden stud construction walls, asbestos, plaster, or metal-corrugated panels, doors, i !
non-load walls, and steel office partitions are disregarded in determining the spacing.
6.4.2 OPERATION 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 g
for GE-Wilmington.
j 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.
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