ML20141B081

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Rev 1 to Guidelines for Decontamination of Facilities & Equipment Prior to Release for Unrestricted Use or Termination of Licenses for Byproduct,Source,Or Snm
ML20141B081
Person / Time
Site: 07001113
Issue date: 04/30/1993
From:
NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS)
To:
Shared Package
ML20141A212 List:
References
PROC-930430-02, PROC-930430-2, NUDOCS 9706230286
Download: ML20141B081 (39)
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GUIDELINES FOR DECONTAMINATION OF FACILITIES AND EQUIPMENT PRIOR TO RELEASE FOR UNRESTRICTED USE OR TERMINATION OF LICENSES FOR BYPRODUCT, SOURCE, OR SPECIAL NUCLEAR MATERIAL l

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U.S. Nuclear Regulatory Commission Division of Fuel Cycle Safety and Safeguards Washington, DC 20555 I

April 1993 l

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O The instructions in this guide, in conjunction with Table 1, specify the radionuclides and radiation exposure rate limits which should be used in decontamination and survey of surfaces or i premises and equipment prior to abandonment or release for unrestricted use. The limits in Table l 1 do not apply to premises, equipment, or scrap containing induced radioactivity for which the l radiological considerations pertinent to their use may be different. The release of such facilities j or items from regulatory control is considered on a case-by-case basis.

1. The licensee shall make a reasonable effort to eliminate residual contamimtion.
2. Radioactivity on equipment or surfaces shall not be covered by paint, plating, or other covering material unless contamination levels, as determined by a survey and documented, are below the limits specified in Table 1 prior to the application c,f the covering. A reasonable effort must be made to minimize the contamination prior to use of any covering.
3. The radioactivity on the interior surfaces of pipes, drain lines, or ductwork shall be determined by making measurements at all traps, and other appropriate access points, provided that contamination at these locations is likely to be representative of contamination on the interior of the pipes, drain lines, or ductwork. Surfaces of premises, equipment, or scrap which are likely to be contaminated but are of such size, construction, or location as to make the surface inaccessible for purposes of measurement shall be presumed to be contaminated in excess of the limits.
4. Upon request, the Commission may authorize a licensee to relinquish possession or O

control of premises, equipment, or scrap having surfaces contaminated with materials in excess of the limits specified. This may include, but would not be limited to, special circumstances such as razing of buildings, transfer of premises to another organization continuing work with radioactive materials, or conversion of facilities to a long-term

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storage or standby status. Such requests must:

a. Provide detailed, specific information describing the premises, equipment or scrap, radioactive contaminants, and the nature, extent, and degree of residual l

surface contamination.

b. Provide a detailed health and safety analysis which reflects that the residual amounts of materials on surface areas, together with other considerations such as prospective use of the premises, equipment, or scrap, are unlikely to result in an unreasonable risk to the health and safety of the public.

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s., g, O 5. Prior to release of premises for unrestricted use, the licensee shall make a comprehensive radiation survey which establishes that contamination is within the limits specified in Table 1. A copy of the survey report shall be filed with the Division of Fuel Cycle Safety -

and Safeguards, U. S. Nuclear Regulatory Commission, Washington, DC 20555, and also the Administrator of the NRC Regional Office having jurisdiction. The report should be filed at least 30 days prior to the planned date of abandonment. The survey report shall:

a. Identify the premises.
b. Show that reasonable effort has been made to eliminate residual contamination.
c. Describe the scope of the survey and general procedures followed.
d. State the findings of the survey in units specified in the instruction.

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Following review of the report, the NRC will consider visiting the facilities to confirm O me semer.

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TABLE 1 O AccEerA8tE suaExcE couriMixArion tEvEts k

I NUCLIDES* AVERAGE' MAXIMUM # REMOVABLE *f U-nat, U-235, U-238, and 5,000 dpm a/100 cm 2

15,000 dpm et/100 cm 2 1,000 dpm a/100 cm2 l associated decay products

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Transuranics, Ra-226, Ra- 100 dpm/100 cm 300 dpm/100 cm 2 20 dpm/100 cm2 j 228, Th-230, Th-228, Pa- 1 231, Ac-227,1-125, I-129 2

Th-nat, Th-232, Sr-90, Ra- 1000 dpm/100 cm 3000 dpm/100 cm2 200 dpm/100 cm 2

l 223, Ra-224, U-232,1 126, 1-131,I 133

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2 Beta-gamma emitters 5,000 dpm py/100 cm 15,000 dpm py/100 cm 2 1,000 dpm py/100 cm 2 l (nuclides with decay modes i other than alpha emission or spontaneous fission) except Sr-90 and others noted above.

"Where surface contamination by both alpha- and beta-gamma-emitting nuclides exists, the limits established for alpha- and beta-gamma-emitting nuclides should apply independently.

D As used in this table, dpm (disintegrations per minute) means the rate of emission by radioactive material as determined by correcting the counts per minute observed by an appropriate detector for background, efliciency, and geometric factors associated with the instrumentation.

" Measurements of average contaminant should not be averaged over more than I square meter. For objects ofless surface area, the average should be derived for each such object.

d 2 The maximum contamination level applies to an area of not more than 100 cm ,

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'The amount of removable radioactive material per 100 cm of surface area should be determined by wiping that area with dry filter or soft absorbent paper, applying moderate pressure, and assessing the amount of radioactive material on the wipe with an appropriate instrument of known efficiency. When removable contamination on objects ofless surface area is determined,6e pertinent levels should be reduced proportionally and the entire surface should be wiped.

khe average and maximum radiation le vels associated with surface contamination resulting from beta-gamma emitters should not exceed 0.2 mrad /hr et I cm and 1.0 mrad /hr at I cm, respectively, measured through not more than 7 milligrams per square centimeter of total absorber.. l l

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O NUCLEAR CRITICALITY SAFETY 6.1 PROGRAM ADMINISTRATION 6.1.1 CRIflCALITY 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 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.

l The established design criteria and nuclear criticality safety reviews are applicable to:

e all new processes, facilities or equipment that process, store, transfer or O otherwise handie fissiie materiais. and e any change in processes, facilities or equipment which may have an impact on the established basis for nuclear criticality safety.

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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 impacted by these hazards, the approval to operate new facilities or make significant changes, modification, or additions to existing facilities is documented in accord LICENSE SNM-1097 DATE 06/11/97 Page l

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L. with established facility practices and conform to configuration managenient function t

O ' 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 i 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 reouirements are fulfilled in accordance with established configuration management practices.

-6.1.2.2 Role of the Criticality Safety Function l

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.

'O 6. o oesaArmo eaoCsouass erecedures that govem 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 I concerned personnel review and understand these procedures through postings,  ;

training programs, and/or other written, electronic or verbal notifications. j 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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6.1.4.1 Posting of Limits and Controls l

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 areas, or to provide instructions or specific precautions to personnel such as:

I e Limits on material types and forms.

  • Allowable quantities by weight or numbet. I

. Allowable enrichments.

  • Required spacing between units.

Control limits (when applicable) on quantities such as moderation, density, or presence of additives, e

j Critical control steps in the operation.

Storage postings are located in conspicuous places and include as appropriate:

  • Material type.
  • Container identification.

. Number ofitems allowed.

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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 i,

material type, U-235 enrichment, and gross weights can be clearly identified or detennined. Deviations from this process include: large process vessels, fuel rods, shipping containers, waste boxes / drums, contaminated items, UF6 cylinders LICENSE SNM-1997 DATE 06/11/97

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containing heels, cold trap cylinders, samples, containers of 1 liter volume or less, or Q other containers where labeling is not practical.

6.1.5 - AUDITS & INSPECTIONS 4

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 procedure and personnel approved by the criticality safety function. Findings, 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 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 OC-Wilmington facility manager or Manager, EHS. Audit results are reported in s.Mng 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.

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6.1.6 CRITICALITY SAFETY PERSONNEL O

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 2 authority to shutdown potentially unsafe operations.

6.2 TECHNICAL PRACTICES f

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 that the controls function as intended and the installed configuration agrees with the criticality safety analysis.

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Operations personnel are responsible for subsequent verification of controls through l lC l

the use of functional testing or verification. When necessary, control calibration and routine maintenance are normally provided by the instrument and calibration and/or

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j l maintenance functions. Verification and maintenance activities are performed per

. established facility practices documented through the use of forms and/or computer L trackiag systems. Criticality safety function personnel randomly review control verifications and maintenance activities to assure that controls remain effective.

l 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 l The relative effectiveness and reliability of controls are considered during the criticality safety analysis process. Passive engineered controls are preferred over all l other system controls and are utilized when practical and appropriate. Active engineered controls are the next preferred method of control followed by

" -Q 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 ,

l manner (i.e., fixed geometry, fixed spacing, fixed size, nuclear poisons, etc.). l

! Pesive 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. )

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i 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 e process variable important to criticality safety and .

providing automatic action (e.g., no human intervention required) to secure the L 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. I 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).

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L 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 i L are not practical. Administrative controls may be proactive (requiring action prior to  ;

proceeding) or reactis e (proceeding unless action occurs). Proactive administrative l controls are preferred. Assurance is maintained through training, experience, and i audit.  ;

l: i 6.2.3 TABLE OF PLANT SYSTEMS AND PARAMETER CONTROLS  !

l. Table 6.0 identifies major process areas or support facility processes within the GE-L '!

Wilmington fuel manufacturing complex and support facilities. Table entries for  !

each significant process item highlight the safety basis selected for the criticality i safety analysis (CSA) and related worst credible contents (or bounding assumptions). 1 Table column definitions are presented below:

l AREA OR SYSTEM: A defined functional group ofprocesses or pieces of equipment that operate as a single unit.

PROCESS SUBAREA OR EQUIPMENT: A defimed subgroup of vessels, tanks, process and/or support equipment within an area thac operate as a single unit.

BASIS FOR CRITICALITY SAFETY: The controlled parameters established l within a CSA for nuclear criticality safety for the identified process subarea or L equipment. For multiple parameter entries, the basis for nuclear criticality safety l

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established in the CSA may be based on the identified parameter (s), as appropriate, l.O including the use of' coupled' parameter control (e.g., mass / moderation). .;

t I i 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, 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 approved CSA may limit the operation of the system to levels more conservative than those permitted by the bounding assumptions.

In the following 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, 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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g, 1 Table 6.0 Plant Systems and Parameter Controls AREA PROCESS BASIS FOR CSA OR SUBAREA OR CRITICALITY BOUNDING ASSUMPTIONS SYSTEM EQUIPMENT SAFETY Fuel Support: UF6Cylinder Receipt Enrichment 99.5 wt. % pure UF6 Storage Pads and Storage s 0.5 wt. % H 2O equivalent OptimalInterunit H 2O Scrap 3 and 5-gallon Geometry Homogeneous or Heterogeneous UO2 Container Storage Mass Optimal H 2O Moderation Full Reflection RA-Inner and Outer Geometry Heterogeneous UO2 Container Storage Moderation Optimal H2O Moderation Full Reflection Waste Box Container Geometry / Mass Homogeneous UO2 Storage Mass Optimal H 2O Moderation Full Reflection BU-J, BU-7,7A Drum Geometry Homogeneous or Heterogeneous UO2 Storage Mass

  • Optimal H 2O Moderation Moderation Full Reflection Fuel Support: Waste Box Load Mass Heterogeneous UO2 New Decon Optimal H 2O Moderation Full Reflection Oil Drum Load Mass Homogeneous UO2 Optimal H 2O Moderation Full Reflection Chemical ADU UF6 Cylinders Moderation 99.5 wt. % pure UF6 l Conversion System s 0.5 wt. % H 2O equivalent J ~

Full Reflection

~ Autoclave Moderation 99.5 wt. % pure UF6 Vaporization s 0.5 wt. % H 2O equivalent Full Reflection l

Cold Trap System Geometry Homogeneous UO2 Moderation Optimal H 2O Moderation Full Reflection Hydrolysis Receiver, Geometry Homogeneous UO F22 l Storage, and Scrubber Concentration Optimal H 2O Moderation j Tanks Full Reflection l

Sump Geometry Homogeneous UO2 Mass Optimal H 2O Moderation Full Reflection Precipitation Tanks Geometry Homogeneous UO2 (Lines 1,2,4) Optimal H 2O Moderation j Full Reflection

  • two out of any three control parameters required for criticality safety.

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AREA PROCESS BASIS FOR CSA

-OR SUBAREA OR CRITICALITY BOUNDING ASSUMPTIONS SYSTEM EQUIPMENT SAFETY Precipitation Tanks Geometry Homogeneous UO2 (Lines 3,5) Mass Optimal H 2O Moderation Full Reflection Dewatering Geometry Homogeneous ADU or U3 0 Centrifugation Mass Optimal H 2O Moderation Full Reflection Outside Containment Clarifying Geometry Homogeneous UO2 Centrifugation Mass Optimal H 2O Moderation Full Reflection Calcination Geometry Homogeneous UO2 Geometry / Mass Optimal H2O Moderation Full Reflection Calciner Scrubber Geometry Homogeneous UO2 Concentration Optimal H2O Moderation Full Reflection 3 or 5-Gallon Product Geometry Homogeneous UO2 Container Mass Optimal H 2O Moderation Full Reflection UO2 Powder Geometry or Mass Homogeneous UO2 Pretreatment: Mill, Moderation Optimal H 2O Moderation Slug, Granulate (MSG) Full Reflection

. LoE and hie UO 2 Geometry Homogeneous UO2 l Powder Blending Mass / Moderation Optimal H2O Moderation l Full Reflection I LoE Fluoride Effluent Geometry Homogeneous UO2 l Vessels Concentration Optimal H 2O Moderation Full Reflection Line 3 Geometry Homogeneous UO2 Accumulator / Permeate Concentration Optimal H 2O Moderation k Vessels Full Reflection Nitrate Quarantine Geometry Homogeneous UO2 Effluent Vessels Concentration Optimal H 2O Moderation Full Reflection Powder Pack Geometry Homogeneous UO2 ,

Screener Moderation Optimal H2O Moderation I Full Reflection Powder Pack Geometry Homogeneous UO2 Product Container Mass Optimal H 2O Moderation Full Reflection HVAC: Wet Areas Geometry Homogeneous UO2 l Mass Optimal H 2O Moderation Full Reflection l

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  • AREA PROCESS BASIS FOR CSA O OR SUBAREA OR CRITICALITY BOUNDING ASSUMPTIONS SYSTEM EQUIPMENT SAFETY HVAC: Dry Areas Mass Homogeneous UO2 Moderation Optimal H 2O Moderation Full Reflection Exhaust Scrubber Geometry / Mass Homogeneous UO2 Mass Optimal H 2O Moderation Full Reflection Utilities: Steam , N2, Mass Backflow into large supply vessels H2, Dissoc. NH4, H2O prevented by backflow prevention Supply measures, physical barriers, and/or l process characteristics.

REDCAP: Oxidation Geometry Heterogeneous UO2 I Feed Containers Mass Optimal H 2O Moderation i Full Reflection )

REDCAP: Oxidation Geometry Heterogeneous UO2 l Furnace Moderation Optimal H 2O Moderation Full Reflection REDCAP: Oxidation Geometry Homogeneous UO2 Output Containers Mass Optirnal H2O Moderation Full Reflection REDCAP: Oxidation Geometry Homogeneous UO2 Off-Gas System Mass Optimal H 2O Moderation Full Reflection l

Miscellaneous: 3 and Geometry Homogeneous or Heterogeneous UO2 5-Oallon Container Mass Optimal H 2O Moderation i Floor storage Full Reflection Integration Geometry Heterogeneous UO2 OXIDIZE 3 and 5-gal. Mass Optimal H 2O Moderation Feed Containers Full Reflection Integration Geometry 1 Heterogeneous UO2 OXIDIZE 3 and 5-gal.

Feed Container Storage Mass Moderation J* Optimal Interunit H 2O Moderation Full Reflection Integration: Geometry Homogeneous or Heterogeneous UO2 OXIDIZE Mass Optimal H 2O Moderation Feed Hood Full Reflection Integration Geometry Heterogeneous UO2 OX1DIZE Moderation Optimal H 2O Moderation Furnace Full Reflection Integration Moderation heterogeneous UO2 RECYCLE Maximum Credible wt. % H2O Powder Outlet Full Reflection

  • two out of any three control parameters required for criticality safety.

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g, te l AREA PROCESS BASIS FOR CSA OR SUBAREA OR CRITICALITY BOUNDING ASSUMPTIONS l

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EQUIPMENT l Integration Moderation Heterogeneous UO2 1 i RECYCLE Maximum Credible wt. % H2 O  !

l Blender Full Reflection l Integration Moderation Heterogeneous UO2 l l RECYCLE Mass Maximum Credible wt. % H2 O I j

DM-10 Vibromill Full Reflection l Integration Moderation Heterogeneous UO2 RECYCLE Unicone Maximum Credible UO2 Density Container Storage Maximum Credible wt. % H2 O l

Optimal Interunit H2O l- Integration Geometry Heterogeneous UO2 RECYCLE 3-gal. Mass

  • Optimal Interunit H 2O Moderation Product Container Moderation Full Reflection Storage l Integration Moderation Heterogeneous UO2 RECYCLE Maximum Credib{e UO2 Density Powder Transfer Maximum Credible wt % H2 O ,

Corridor Full Reflection j l Uranium Recovery Unit Fluoride Waste Process Geometry Homogeneous UO2 l l (URU) System Vessels Concentration Optimal H 2O Moderation i l

Full Reflection I Fluoride Waste Concentration Homogeneous UO2 Surge Vessel Mass Optimal 1102 Moderation p (V-106) Full Reflection v Radwaste Process Geometry Homogeneous UO2 Vessels Concentration Optimal H 2O Moderation Full Reflection Nitrate Waste Process Geometry Homogeneous UO2 i l Vessels Concentration Optimal H 2O Moderation \

Full Reflection Nitrate Waste Concentration Homogeneous UO2 Surge Vessel Mass Optimal H 2O Moderation (V-103) Full Reflection l Oxidation Feed Geometry Heterogeneous UO2 Containers Mass Optimal H 2O Moderation Full Reflection Oxidation Furnace Geometry Heterogeneous UO2 Optimal H 2O Moderation i Full Reflection

[ Oxidation Furnace Geometry Heterogeneous UO2

! Boat Dump Moderation Optimal H 2O Moderation Full Reflection

  • two out of any three control parameters required for criticality safety.

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AREA PROCESS BASIS FOR CSA OR SUBAREA OR CRITICALITY BOUNDING ASSUMPTIONS SYSTEM EQUIPMENT SAFETY Oxidation 3-gallon Geometry Heterogeneous UO2 Container Storage Mass . Optimal H 2O Moderation Moderation Full Reflection Oxidation Off-Gas Geometry Heterogeneous UO2 l System Mass ' Optimal H 2O Moderation

Full Reflection l

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Dissolutiorc Can Geometry Heterogeneous UO2 i Dump Feed Conveyor Mass Optimal H2O Moderation  ;

Moderation Full Reflection 1 Dissolution: Geometry Heterogeneous UO2 l Dissolvers, Pumps, Concentration Optimal 1102 Moderation 1 Sumps, Filters, Piping Full Reflection Oberlin Filter Geometry Heterogeneous UO2 Concentration Optimal H 2O Moderation Full Reflection Dissolution: NOX Concentration Homogeneous UO2 ,

Scrubber Mass On-Line Density Meter l Full Reflection Counter-Current ' Geometry Heterogeneous UO2 Leaching: Can Dump Mass / Moderation Optimal H 2O Moderation Full Reflection Counter-Current Geometry Heterogenecus UO2 p/

s Leaching: Leach Troughs, Pumps, Concentration Optimal 1102 Moderation Full Reflection Filters, Storage Tanks, Product Containers Utilities; Steam, Di Mass Backflow into large supply vessels  ;

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Nitric Acid, prevented by backflow prevention Aluminum Nitrate measures, physical barriers, and/or i process characteristics.

Head-End Geometry Homogeneous UNH Concentrator Process Concentration Optimal H 2O Moderation Full Reflection Solvent Extraction Geometry Homogeneous UO2 Process Concentration Optimal H 2O Moderation Full Reflection l UNH Product Storage Geometry Homogeneous UNH l Vessels Concentration Optimal H 2O Moderation j Full Reflection

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AREA PROCESS BASIS FOR CSA OR SUBAREA OR CRITICALITY BOUNDING ASSUMPTIONS SYSTEM EQUIPMENT SAFETY Waste Solvent Drum Mass Homogeneous UO2 Load Optimal H2O Moderation Full Reflection Uranyl Nitrate UNH LEM Tank Feed Geometry Homogeneous UO2 Conversion (UCON) Tanks Concentration Optimal H2O Moderation System Full Reflection UCON: Precipitation Geometry Homogeneous UNH

! Tanks Mass Optimalll O2 Moderation Full Reflection UCON: Dewatering Geometry Homogeneous ADU or U3 0:

l Centrifugation Mass Optimal H2O Moderation j Full Reflection Outside Containment UCON: Clarifying Geometry Homogeneous UO2 Centrifugation Mass Optimal H2O Moderation Full Reflection UCON Process: Geometry Homogeneous UO2 Calcination Geometry / Mass Optimal H2O Moderation Full Reflection Waste Treatment Fluoride Waste Concentration Homogeneous UO2 Facility (WTF) Barrens Surge Vessel Mass Optimal H 2O Moderation (V-108) Full Reflection Nitrate Waste Barrens Concentration Homogeneous UO2 l Surge Vessel (V-104) Mass Optimal H 2O Moderation

/"' Full Reflection Centrifuge Geometry Homogeneous UO2 Mass Optimal H 2O Moderation Full Reflection \

Oberlin Filter Geometry / Mass Homogeneous UO2 l Concentration Optimal H2O Moderation ,

Full Reflection Uranium Recovery from URLS Process Tanks Concentration Homogeneous UO2 Lagoon Sludge (URLS) Optimal H 2O Moderation Facility Process Full Reflection URLS Process Non- Geometry /Concent. Homogeneous UO2 Leach Filter Press Concentration Optimal H2O Moderation Full Reflection URLS Process Product Concentration Homogeneous UO2 Waste Container Mass Optimal H 2O Moderation Full Reflection Waste Oxidation / Incinerator Mass (Box Monitor) Heterogeneous UO2 Reduction (Incineration) Combustible Box Feed Mass (E-Gun) Optimal H 2O Moderation Facility Containers Full Reflection LICENSE SNM-1097 DATE 06/11/97 Page l

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or 5-Gallon Containers Mass Optimal H 2O Moderation l Full Reflection I

l Dry Conversion Process UF6Cylinder Receipt Enrichment 99.5 wt % pure UF6 (DCP) Conversion and Storage s 0.5 wt. % H 2O equivalent Optimal Interunit H2O l Vaporization Moderation 99.5 wt. % pure UF6 l Autoclave w/UF6 s 0.5 wt % H 2O equivalent Cylinder Full P.cflection l Vaporization Geometry Homogeneous UO2 Cold Trap System Moderation Optimal H 2O Moderation

Full Reflection l Conversion
Moderation Homogeneous UO2 l i Reactor / Kiln Maximum Credible UO2 Density j Maximum Credible wt. % H2O Full Reflection Conversion: Moderation Homogeneous UO2 l p Powder Outlet Box Maximum Credible UO2 Density i Maximum Credible wt. % H2O Full Reflection Powder Outlet: Moderation Homogeneous UO2 Cooling Hopper Maximum Credible UO2 Censity Maximum Credible wt. % H2O Full Reflection l Powder Transfer & Moderation Homogeneous UO2 Storage: Normal Maximum Credible UO2 Density Product Container Maximum Credible wt % H2O Full Reflection Powder Transfer & Geometry Homogeneous UO2 l Storage: Out-of- Spec Moderation Maximum Credible UO2 Density Moisture Product Maximum Credible wt. % H2O Container Full Reflection

! Homogenization Moderation Homogeneous UO2 Maximum Credible UO2 Density Maximum Credible wt % H2O l Full Reflection l

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!' SYSTEM EQUIPMENT SAFETY

! Blending, Moderation Heterogeneous UO2 l Precompaction, Maximum Credible UO2 Density l Gran, .lation Maximum Credible wt. % H2 O Full Reflection Tumbling: Moderation Heterogeneous UO2 in Powder Container Maximum Credible UO2 Density Maximum Credible wt. % H2O Full Reflection Powder Pack Moderation Heterogeneous UO2 Screener Maximum Credible UO2 Density Maximum Credible wt. % H2 O Full Reflection Powder Pack Geometry Homogeneous UO2 Product Container Mass Optimal H 2O Moderation Full Reflection Utilities: N 2, H2 , H 2O Mass Backflow into large supply vessels not j Supply, Refrigerant credible due to backflow prevention measures, physical barriers, and/or process characteristics. 1 HF Efiluent Recovery Geometry Homogeneous UO2 and Storage Vessels Mass Optimal H 2O Moderation Full Reflection Recycle Blender Moderation Heterogeneous UO2 q Maximum Credible UO2 Density V Maximum Credible wt. % H2O Full Reflection Recycle Unicone Moderation Heterogeneous UO2 Product Maximum Credible UO2 Density Container / Storage Maximum Credible Internal wt. % H2 O OptimalInterunit H 2O Recycle 3-Gallon Geometry 1 Heterogeneous UO2 Product Container / Mass Optimal H 2O Moderation f* Full Reflection Storage Moderation Press Warehouse Conveyor Storage: Geometry 1 Homogeneous UO2 Facility Process 3 and 5-gallon Cans Mass OptimalInterunit H 2O Moderation Moderation f* Full Reflection Powder Dump Transfer Geometry Homogeneous UO2 Hopper / Chute Moderation Optimal H 2O Moderation Full Reflection Pellet Presses Geometry / Mass Heterogeneous UO2 Moderation Optimal H 2O Moderation 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 SYSTEM EQUIPMENT SAFETY Press Lubricant Sump Geometry Heterogeneous UO7 Mass Optimal H 2O Moderation Full Reflection Press: Green Pellet Geometry Heterogeneous UO2 Boat Product Container Moderation Optimal H 2O Moderation Full Reflection 3-gallon Powder Geometry Heterogeneous UO2 Cleanup Container Mass Optimal H 2O Moderation Full Reflection Integration: Moderation Heterogeneous UO2 I PWDR-MRA Maximum Credible wt. % 110 2

Press Feed Full Reflection ,

Integration Geometry / Mass Heterogeneous UO2  !

PWDR-MRA Moderation Maximum Credible UO2 Density Container-Storage Maximum Credible wt. % H2 O Full Reflection Integration Moderation Heterogeneous UO2 PWDR-MRA Maximum Credible UO2 Density Powder Transfer Maximum Credible wt. % H2 O Corridor Full Reflection Pellet Sintering System Feed / Exit Conveyors Geometry Heterogeneous UO2 Moderation Optimal H 2O Moderation Full Reflection p Sintering Furnace Geometry Heterogeneous UO2 d Moderation Optimal H 2O Moderation Full Reflection Pellet Grinding System Feeder llopper Bowl or Geometry Heterogeneous UO2 Flat Feeder Table Moderation Optimal H2O Moderation Full Reflection i Grinder Geometry Heterogeneous UO2 Moderation Optimal H 2O Moderation Full Reflection Grinder APITRON Geometry Homogeneous UO2 l Filter Moderation Optimal H2O Moderation Full Reflection 4 Grinder Swarf 3- Geometry Heterogeneous UO2 Gallon Container Moderation Optimal H2O Moderation Full Reflection Grinder Hardscrap 3- Geometry Heterogeneous UO2 Gallon Container Mass Optimal H 2O Moderation Full Reflection

  • two out of any three control parameters required for criticality safety. ,

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AREA PROCESS BASIS FOR CSA OR SUBAREA OR CRITICALITY BOUNDING ASSUMPTIONS SYSTEM EQUIPMENT SAFETY i Grinder Pellet Product Geometry Heterogeneous UO2 Tray Mass Moderation J

h. Optimal H 2O Moderation Full Reflection Pellet Transfer Cart Geometry Heterogeneous UO2 Moderation OptimalInterunit H 2O Moderation Full Reflection Rod Load, Out-Gassing, Rod Load, Out- Geometry Heterogeneous UO2 and Final Rod Welding Gassing, and Final Rod Moderation Optimal H2O Moderation System Weld Full Reflection Pellet Storage Cabinet Geometry Heterogeneous UO2 Moderation Optimal H2O Moderation Full Reflection Rod Storage Cabinet Geometry Heterogeneous UO2 Moderation Optimaill O2 Moderation Full Reflection Gadolinia Shop Press, Sintering, Similar to UO2 Shop Similarto UO 2Shop Above j Grinding, Rod Load, Above Rod Storage, & Outgas Gadolinia 3 and 5- Geometry Homogeneous UO2 -

Gallon Feed Containers Mass Optimal H 2O Moderation ,

Full Reflection Gadolinia 3 and 5- Geometry } Homogeneous UO2

(^ Gallon Feed & Product Container Storage Mass Moderation f* Optimal H 2O Moderation Full Reflection Gadolinia DM 10 Geometry Heterogeneous UO2 Vibromill(MCA) Moderation Optimal H2O Moderation Full Reflection j Gadolinia DM-3 Mass Homogeneous UO 2 )

Vibromill(MCA) Moderation Optimal H 2O Moderation Full Reflection Pellet Storage: Geometry / Mass Heterogeneous UO2 Ministacker Moderation Optimal H2O Moderation Full Reflection Integration: Mass Homogeneous UO2 Gadolinia MEZZ-MRA Moderation Maximum Credible UO2 Density Unicone Feed Maximum Credible wt. % H2 O I Container Full Reflection l

Integration Moderation Heterogeneous UO2 l Gadolinia MEZZ-MRA Maximum Credible wt. % H2 O

! DM-10 Vibromill Full Reflection

  • two out of any three control parameters required for criticality safety, i

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> OR SUBAREA OR CRITICALITY BOUNDING ASSUMPTIONS SYSTEM EQUIPMENT SAFETY Integration Moderation Heterogeneous UO2 Gadolinia MEZZ-MRA Maximum Credible wt. % H2O Rotary Slugger Full Reflection Integration Moderation Heterogeneous UO2 Gadolinia MEZZ-MRA Maximum Credible wt. % H2O Granulator Full Reflection Integration: Geometry 1 Homogeneous UO2 Gadolinia MEZZ-MRA Mass Optimal H 2O Moderation Moderation f* Full Reflection 3 and 5-Gallon Feed &

Product Container .

Storage Integration Moderation Heterogeneous UO2 Gadolinia MEZZ-MRA Maximum Credible UO2 Density Powder Transfer Maximum Credible wt, % H2O Corridor Full Reflection Bundle Assembly Rod Trays Geometry Heterogeneous UOz Mass OptimalInterunit H 2O Moderation Full Reflection Rod Storage Cabinets Geometry Heterogeneous UO2 Moderation OptimalInterunit H 2O Moderation Full Reflection Rod Tray Transfer Geometry Heterogeneous UO2 Vehicle:" Big Joe" Moderation Optimal Interunit H 2O Moderation p Full Reflection

\.. ' Magnetic and Passive Geometry Heterogeneous UO2 Scanner:" MAPS" Moderation Optimal Interunit H2O Moderation Full Reflection Bundle Accumulator: Geometry Heterogeneous UO2 "BACC" Moderation OptimalInterunit H 2O Moderation Full Reflection Automatic Bundle Geometry Heterogeneous UO2 Assemble Machine: Moderation OptimalInterunit H 2O Moderation "ABAM" Full Reflection Rod Scanner: Geometry Heterogeneous UO2

" Fat Albert" Moderation OptimalInterun t H 2O Moderation Full Reflection Assembly Table Geometry Heterogeneous UO2 Moderation OptimalIntetunit H 2O Moderation Full Reflection l Upender: Bundle and Geometry Heterogeneous UO2 RA Container Moderation Optimal Interunit 110 2 Moderation Full Reflection l

  • two out of any three control parameters required for criticality safety.

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AREA PROCESS BASIS FOR CSA (9 OR SYSTEM SUBAREA OR CRITICALITY SAFETY BOUNDING ASSUMPTIONS EQUIPMENT Inspection Pit Geometry Heterogeneous UO2 Moderation OptimalInterunit H 2O Moderation i Full Reflection Bundle Storage: Geometry Heterogeneous UO2

" Forest" Moderation Optimal Interunit H 2O Moderation Full Reflection l RA Container: Geometry Heterogeneous UO2 Transfer Port & RA Moderation Optimal Interunit H 2O Moderation Conveyor Full Reflection Rod Scanner: Geometry Heterogeneous UO2 l X-Ray-Unit Moderation OptimalInterunit 110 2 Moderation l Full Reflection 1 Rod Inspection: Geometry Heterogeneous UO2 Surface-Plate Moderation OptimalInterunit H 2O Moderation l Full Reflection Rod Movement: Geometry Heterogeneous UO2 One & Two-Tray Cart Moderation OptimalInterunit H 2O Moderation Full Reflectjon Container Storage: Geometry Heterogeneous UO:

RA Inner / Outer Moderation Optimal Interunit H 2O Moderation Storage Full Reflection Decontamination & Wash Down Areas, Geometry / Mass Homogeneous UO2 l r Volume Reduction Facility (DVRF)

Sumps, Bag Filters Mass Optimal H 2O Moderation Full Reflection j

l Dust Hog Mass Homogeneous UO2 i Optimal H 2O Moderation l Full Reflection HVAC Geometry Homogeneous UO2 Mass Optimal H 2O Moderation Full Reflection 3-Gallon Waste Geometry Homogeneous UO2 Container Storage Mass Optimal H 2O Moderation Full Reflection l

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A.. 1 6.2.4 SPECIFIC PARAMETER LIMITS The safe geometry values of Table 6.1 below are specifically licensed for use at the O 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; reacthity formula calculations which incorporate validated K-infinities, migration areas (M 2) and 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 UO2 and water as a function of U235 enrichment over the range of 1.1% to 5% for uncontrolled l l geometric configurations. The safe batch sized for UO2 of specific compounds may be adjusted when applied to otbar compounds by the formula:

kgs X = (kgs UO2 e 0.88 ?,/ f where, kgs X = safe batch value of compound 'X' kgs UO2 = safe batch value for UO2 O.88 = wt. % U in UO 2

f = wt. % U in compound X
  • Where engineered controls prevent over batching, a mass of 75% of the minimum  ;

critical mass shall not be exceeded. j Subject to provision for adequate protection against precipitation or other circumstances which may increase concentration, the following safe concentrations are specifically licensed for use at the GE-Wilmington facility:

. A concentration ofless than or equal to one-half of the minimum critical l concentration.

i l

  • A system in which the hydrogen to U235 atom ratio (H/U235) is greater than 5200. l LICENSE SNM-1097 DATE 06/11/97 Page i

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  • Infinite Slab
  • Sphere Volume
  • 110 2 Mixtures U235 Diameters Thickness (Inches) (Inches) (Liters) l 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 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 5.00 9.50 4.45 24.6 Homogeneous Weight Percent Infinite Cylinder Infinite Slab Sphere Volume Aqueous U235 Diameters Thickness Solutions (Inches) (Inches) (Liters) l 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 0 5.00 9.5 4.80 26.0 lieterogeneous Weight Percent Infinite Cylinder Infinite Slab Sphere Volume Mixtures or U235 Diameters Thickness Compounds (Inches) (Inthes) (Liters) 2.00 11.10 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 l

  • These values represent 93%,88% and 76% of the minimum critical cp . der diameter, slab thickness, and sphere volume, respectively. For enrichments not specified, smooth curve interpolation may be used.

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Table 6.2 Safe Batch Values for UO2 and Water

  • Nominal Weight Homogeneous Heterogeneous Nominal Weight Homogeneous Heterogeneous Percent U235 UO2 Powder & UO2 Pellets & Percent U235 UO2 Powder & UO2 Pellets &

Water Water Water Water Mixtures Mixtures Mixtures Mixtures (Kas UO2 ) (Kgs UO ) 2 (Kgs UO2 ) (Kgs UO2 )

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 t 1.40 583.0 193.0 4.60 20.2 20.0 l 1.50 404.0 158.0 4.80 19.1 18.8 l 1.60 293.3 135.0 5.00 18.1 18.1 1.70 225.0 116.0 1.80 183.0 102.0 1.90 150.6 90.5 q 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 3.20 38.9 34.I 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.

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I l-l 6.2.5 CONTROL PARAMETERS i

O Nuclear criticality safety is achieve 6 by controlling one or more parameters of a i

!- 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 i operation, including upset conditions, are justified, documented, and independently i 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 . l control. At the GE-Wilmington facility, favorable geometry is developed l 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 l

l mechanical distortion. In addition, these systems include provisions for periodic l- 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. (

1 6.2.5.2 Mass - Mass control may be used for a nuclear criticality safety control on its own or 1 in combination with other control methods. Mass control may be utilized to limit the l quantity of uranium within specific process operations or vessels and within storage, j transportation, or disposal containers. Analytical or non-destructive methods may be )

employed to verify the mass measurements for a specific quantity of material. l 1

l Establishment of mass limits involves consideration of potential moderation, )

reflection, geometry, spacing, and material concentration. The criticality safety  ;

analysis considers normal operations and credible prccess upsets in determining  !

actual mass limits for the system and for defining additional controls. When only l j

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4. 1 administrative controls are used for mass controlled systems, double batching is

-O coasidered to e=sure ae9uate safety - r8 ia.

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 established facility practices (in decreasing order of restriction): l

  • At each enriched uranium interface involving intentional and continuous 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 controls, e At enriched uranium interfaces involving intentional but non-continuous introduction of moderation at least three controls are required to assure that the moderation safety factor is not exceeded.' At least one of these controls O muet be an active ensineered controi, uniess e moderation safetx fector greater than 3 is demonstrated.

. For situations where moderation is not intentionally introduced as part of the 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 2O or equivalent) must provide sufficient factor of safety above the process 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.

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 l prevent moderator from entering the system through a defined interface and must fail 1

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l before a criticality accident is possible. The quality and basis for selection of the

O controis is decemented in accordance with intestated safe 17 ^naixsis erocess

' 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 l-safety function.

6.2.5.4 _ Concentration (or Density) - Concentration control may be used for nuclear l 1

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 l parameter controlled to prevent criticality, concentration may be controlled by two independent combinations of measurement and physical control, each physical i

control capable of preventing the concentration limit being exceeded in a location I where it would be unsafe. The preferred method of attaining independence being j that at least one of the two combinations is an active engineered control. Each  !

l process relying on concentration control has in place controls necessary to detect l and/or mitigate the effects ofintemal concentration within the system (e.g., Dynatrol '

density meter, Rhonan density meter, etc.), otherwise, the most reactive credible 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 l 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 1 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 l 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.

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Credit may be taken for neutron absorbers that are normal constituents of filter media Q (e.g., natural boron) provided the following requirements are met:

e The failure or loss of the media itself also prevents accumulation of significant quantities of fissile meterial.

  • 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, e 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. 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:

e 12-inches of full density water equivalent, or e the larger of 12-foot air distance or the greatest distance across an 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 dbmeter may be excluded from interaction consideration, provided they are not grouped in close arrays. I j- Techniques which produce a calculated effective multiplication factor of the entire system (e.g., validated Monte Carlo or So 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.,

l Solid Angle methods) may also be used. In either case, the criticality safety analysis l must comply with the requirements of Sections 6.1.1 and 6.3.

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6.2.5.7 Material Composition (or Heterogeneity) - The criticality safety analysis for each O 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.

I 6.2.5.9 Enrichment - Enrichment control may be utilized to limit the percent U-235 within a O 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 credible area enrichment is utilized in the

criticality safety analysis.

l 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 following requirements:

  • The bounding conditions and operational limits are specifically identified in l the criticality safety analysis and, are specifically communicated, through training and procedures, to appropriate operations personnel.

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. Bounding conditions for such process and/or material characteristics are h based on established physical or chemical reactions, known scientific principles, and/or facility-specific experimental data supported by operational history.

l e The devices and/or procedures which maintain the limiting conditions must 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 tempere.ure 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.

6.3 CONTROL DOCUMENTS i

iO 6.3.1 CRITICALITY SAFETY ANALYSIS (CSA) l In accordance with ANSI /ANS-8.19 (1984), the criticality safety analysis is a l collection ofinformation that "provides sufficient detail clarity, and lack of 4 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.

!- The CSA addresses the specific concems (event sequences) of nuclear criticality i 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 l

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:

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. Scope - This element defines the stated purpose of the analysis.

V e 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.

e 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 i

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 i 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 actual model used in the analysis. An identification of both normal and credible upset (accident condition) model filenaming convention is provided.

Key input listings and corresponding geometry plot (s) for both nonnal 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 bq 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) l (e.g., wt fraction H2O). When applicable, the assigned bias of the j'

calculation is also clearly stated and incorporated into both normal and/or accident limit comparisons e Safety During Upset Conditions - This element presents a concise summary  ;

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. l e Specifications and Requirements for Safety - When applicable, this .

element presents both the design specifications and the criticality safety l requirements for correct implementation of the established controls. These  !

requirements are incorporated into operating procedures, training, j 1

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maintenance, quality assurance as appropriate to implement the specifications O 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 METHODS 6.3.2.1 Keff Limit 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 ,

Q calculational uncertainties, that is:

Keff + 3a - bias s 0.97 (accident conditions).

Thus, the established delta-k safety margin used at the GE-Wilmington facility is 0.03.

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 system.

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6.3.2.2 Analytical Methods n

G Methodologies currently employed by the GDWilmington criticality safety function include hand calculations utilizing pubiished experimental data (e.g.,' ARH 600 handbook), Solid Angle methods (e.g., SAC code), and Monte Carlo codes (e.g.,

l GEKENO, GEMER) which utilize stochastic methods to solve the 3D neutron l 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.

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 l cross-section data set, and a potential scattering op 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.

l GEMER (Geometry Enhanced merit) is a multigroup Monte Carlo program which solves the neutron transport equaticn in 3-dimensional space. The GEMER criticality program is based on 190- nergy group structure to represent the neutron 'f l energy spectrum. In addition, GEMER treats resolved resonances explicitly by tracking the neutron energy and solving the single-level Breit-Wigner equation at lQ l

each collision in the resolved resonance range in regions containir.g 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.

h 6.3.2.3 Validation Techniques l Experimental critical data or anrlytical methods which have been validated l (benchmarked) by comparison with experimental critical data in accordance with criteria described in section 4.3 cf ANSI /ANS 8.1 (1983) are used as the basis for L

validation. An analytical method is considered validated when the following are i established:

e the type of systems which can 1 c modeled e the range of parameters which may be treated e the bias, if any, which exists in the results produced by the method.

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m V 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 j 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 (Kwe - 1.0) is applied l 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 enrichments up to 5.0 % wt. % U235.

l FIGURE 6.1 - CEMER BIRS DETERMINATION, PARTICLE REICHT 1.18 LEGEND t

123 CATA $ET I . PARTICLE WE!GHT x 8R0 ORDER FIT OF LIMIT '

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FIGURE 6.2 - GEKENO BIAS CALCULRTION LEGEND l

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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 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 1 ersions 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 methodolcty with the differences in expenment l LICENSE SNM-1097 DATE 06/11/97 Page

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and analytical methods being used to calculate bias and uncertainty values to be applied to the calculational results.

(v)

Each individual workstation is verified to produce results identical to the l development workstation prior to use of the software for criticality safety l calculations demonstrations on the production workstation.

Modifications to software that may affect the cc.lculational 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 l re-qualification of the code prior to release for use.

I 6.3.2.5 Technical Reviews j 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. j The independent technical review consists of a verification that the neutronics geometry inodel and configuration used adequately represent the system being y 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 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.

. Vedfy the calculations with a custom method.

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.

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