ML20217N424
| ML20217N424 | |
| Person / Time | |
|---|---|
| Site: | Westinghouse |
| Issue date: | 03/31/1998 |
| From: | WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | NRC |
| Shared Package | |
| ML20217N421 | List: |
| References | |
| NUDOCS 9804090098 | |
| Download: ML20217N424 (50) | |
Text
TABLE OF CONTENTS' NUMBER AND TITLE PAGE TABLE OF CONTENTS.......................................................................................
I REVISION RECORD...........................................................................................
CHAPTER 1.0 GENERAL INFORMATION................................................. 1.0 1.1 FACILITY AND PROCESS DESCRIPTION.............................. 1.0 1.2 INSTITUTIONAL INFORMATION......................................... 1.4 1.3 SITE DESCRIPTION............................................................ 1. 5 1.4 TERMS AND DEFINITION S................................................. 1. 8 j
i CHAPTER 2.0.
. M ANAGEMENT ORGANIZATION....................................... 2.0 2.1 ORGANIZATIONAL RESPONSIBILITIES AND AUTH ORITIES.................................................................. 2.0 2.2 S AFETY COMMITTEES...................................................... 2. 8 CHAPTER 3.0 CONDUCT OF OPERATION S............................................... 3.0 3.1 CONFIGURATION M ANAGEMENT...................................... 3.0 3.2 M AINTENANCE................................................................ 3. 2 3.3 QUALITY ASS URANCE...................................................... 3.4
' 3.4 PROCEDURES, TRAINING AND QUALIFICATION................. 3.8 3.5 H UM AN FACTORS............................................................ 3.14 3.6 AUDITS AND SELF-ASSESSMENTS..................................... 3.15 3.7 INCIDENT INVESTIGATIONS............................................. 3.17 3.8 RECORDKEEPING AN D REPORTING.................................. 3.20 i
CHAPTER 4.0 INTEGRATED SAFETY ASSESSMENT.................................... 4.0 CHAPTER 5.0 RADIATION S AFETY........................................................ 5.0 5.1 ALARA (As Low As Reasonably Achievable) POLICY................. 5.0 5.2 -
RADIATION WORK PERMITS (RWP).................................... 5.1 5.3 VENTILATION SYSTEMS.................................................... 5.2 5.4 AIR S AM PLING................................................................. 5.4 5.5 CONTAMINATION CONTROL............................................. 5.5 5.6 EXTERNAL EXPOSURE...................................................... 5.8 5.7.
INTERN AL EXPOSURE....................................................... 5. 8 5.8 RESPIRATORY PROTECTION............................................. 5.10 5.9 INSTRUM ENTATION......................................................... 5.12 5.10 SUMMING INTERNAL AND EXTERNAL EXPOSURES........... 5.12 9004090098 980331 PDR ADOCK 07001151 i
C PDR
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TABLE OF CONTENTS (Cont'd)
' NUMDER AND TITLE PAGE CHAPTER 6.0 N UCLEAR CRITICALITY SAFETY........................................ 6.0 6.1.
PROGRAM ADMINISTRATION............................................ 6.0 6.2 CONTROL METHODOLOGY AND PRINCIPLES..................... 6.8 6.3 ALARM SYSTEM.............................................................. 6.1 1 6.4 CONTROL DOCUMENTS................................................... 6.21 1
6.5 A LARM S YSTEM S............................................................ 6. 22 CHAPTER 7.0 CH EMICAL S AFETY........................................................ 7.0 7.1 CHEMICAL S AFETY PROGRAM.......................................... 7.0 7.2 CHEMICAL SAFETY HAZARD EVALUATIONS..................... 7.0 7.3 CHEMICAL SAFETY PROGRAM STRUCTURE....................... 7.1 7.4 '
ADDITIONAL CHEMICAL SAFETY COMMITMENTS.............. 7.2 i
CHAPTER 8.0 FIRE S AFETY.................................................................... 8.0 8.1 STRUCTURE OF THE FIRE SAFETY PROGRAM..................... 8.0 8.2 FIRE SUPPRESSION SERVICES........................................... 8.10 CHAPTER 9.0 EMERGENCY MANAGEMENT PROGRAM............................ 9.0 9.1 EM ERG ENCY PLAN........................................................... 9.0 9.2 EMERGENCY EQUIPMENT................................................. 9.0 CHAPTER 10.0 ENVIRONM ENTAL PROTECTION....................................... 10.0 10.1 EFFLUENT AIR TREATMENT............................................ 10.0 -
g 10.2' LIQUID WASTE TREATMENT FACILITIES........................... 10.0 10.3
' SOLID WASTE DISPOSAL FACILITIES................................. 10.1
-10.4 PROGRAM DOCUM ENTATION........................................... 10.1 10.5 EVALUATIONS................................................................ 10.2 10.6 OFF-SITE DOS E................................................................ 10.2 CHAPTER 11.0 DECOMMISSIONING........................................................... 11.0 11.1 CONCEPTUAL DECOMMISSIONING PLAN.........................11.0 11.2 DECOMMISSIONING FUNDING. PLAN AND FINANCIAL ASS URANCE.................................................................... 1 1.1 CHAPTER 12.0 AUTHORIZATIONS AND EXEMPTIONS............................... 12.0 12.1 AUTH ORIZATIONS........................................................... 12.0 12.2 EX EM PTION S................................................................... 12.4 g
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License No. SNM-1107 Revision Submittal Date: 31 MAR 98 Revision No. 10.0
REVISION RECORD REVISION DATE OF PAGES NUMBER REVISION REVISED REVISION REASON 1.0 30APR95 All Update to current operations.
2.0 28JUN96 iii, 6.8 Clarify Criticality Safety Basis for the compaction operation.
3.0 30AUG96 iii,1.7,1.9,12.6,12.7 Incorporate Safety Condition S-3 into Application; correct reference to Figure 1.3 instead of 2.3, to reflect expansion of the CAA in order to eliminate need for gate.
4.0 30SEP96 iii, 6.11, 6.12 Clarification of Criticality Safety Basis for the Pellet Stripping System Equipment and Hoods & Containment.
5.0 08NOV96 iii,1.12,3.18, and 3.19 Incorporation of a definition, (Reprinted all document and incident notification i
pages in Microsoft Word criteria, recently approved format) by NRC Staff.
6.0
'5bfAY9.'
6.12 (Reprinted all Clarify Evaluation document pages in Bounding Assumptions Microsoft Word format.)
for Storage of Annular Pellets.
7.0 14JUL97 iii,12.2 and 12.3.
Withdraw an existing authorization, and expand another authorization to enable cement manufacturing with CaF -
2 i
8.0 11 AUG97 iii,2.4 and 8.1 (Reprinted Change emergency exercise all document pages in frequencies for consistency Microsoft Word format.)
with Emergency Plan.
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9.0 23SEP97 iv, Chapter 6 To respond to NRC Staff request for additional information.
To revise table to correlate to CSE organization and clarify discussions regarding margin of safety with respect to normal operations, expected process upsets and credible process upsets.
10.0 31 MAR 98 Table of Contents i-iv To replace Revisions Chapters 3.0 and 6.0.
Numbers 2.0, 4.0, 6.0, and 9.0; and respond to SMIP initiative regarding SNM-1107, Chapter 6.0. (Chapter 3.0 shown with bars & 6.0 Major Rewrite).
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1 CHAPTER 3.0 CONDUCT OF OPERATIONS The basis for total quality conduct of operations at the Columbia Fuel Fabrication Facility (CFFF) will be the Safety Margin Improvement Program (SMIP). This program will be a structured oversight process that maintains management awareness, and enables monitoring, of management-specified regulatory and process improvement activities; and, will be a management decision process for determining where and when resources will be allocated. This program will address, until their logical completion, elements of Environmental Protection Improvement, Criticality Safety Margin Improvement, Occupational Safety Improvement, and General Plant Improvement. A responsible individual will be assigned accountability for each SMIP element initiative. The Safety Margin Improvement Program will not be a commitment tracking system; SMIP commitments will be followed to management-approved completion by the responsible individual specifically assigned accountability for each particular initiative. This program will be a d6tumented demonstration of CFFF Managements' strong commitment to evaluate, on a continuing basis, opportunities to improve the Plant margin of safety - with the understanding that: addition, change, and/or deletion of program elements and/or initiatives; continuation of ongoing program elements and/or initiatives; and/or, additions, deletions and/or changes of program implementation c,chedules - relevant to the Safety Margin Improvement Program - will always be at the discretion of the Plant Manager, as advised by the Engineering, Manufacturing, and Regulatory Components.
3.1' CONFIGURATION MANAGEMENT To assure that design changes will not adversely impact on environmental protection, health, safety, quality, and/or safeguards programs at the Columbia Fuel Fabrication Facility (CFFF), a formal review process will be established to analyze new systems and components, or modification; ta existing systems and components, in order to reliably predict performance under normal operating conditions and potential process upsets.
Structured hazard analyses, as conducted in accordance with Chapter 4.0 of this License Application, will specifically include analysis of verified drawings under configuration management.
3.1.1 CONFIGURATION MANAGEMENT PROGRAM AND PROCEDURE The CFFF Configuration Management Program will embrace an approved procedure for implementation of proposed additions or changes to facility systems. The procedure will define the review and approval process to assure the impacted systems will continue to meet or exceed regulatory specification requirements of baseline safety assessments. The Docket No.
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procedure will specify documentation required to maintain a current record of existing system conditions.
3.1.2 CONFIGURATION MANAGEMENT IMPLEMENTATION The Configuration Management Program will be a major sub-element of the Safety Margin Improvement Program described in the introduction to this Chapter.
Configuration management will not be a substitute for procedures described in Subsection 3.4.1 of this Chapter, but will facilitate continuing compliance with their requirements through responsible facility addition and/or change project reviews.
3.1.3 CONFIGURATION MANAGEMENT PROCESS The following sequence of activities will be utilized for all facility addition and/or change projects.
Complexity of each project, and the issues involved, will determine the magnitude of effort afforded to each activity.
(a)
A project will be formally opened for review by an assigned responsible individual completing a configuration change control form, and enclosing specified project information for the review process.
@)
Manufacturing, Engineering, And Quality Component Reviews Designated Manufacturing, Engineering, and/or-Quality Component Functions will review the project proposal for economics, practicality, and technical merit.
Formal approvals will be documented as part of the review package.
i (c)
Regulatory Component Reviews For Approval Extent and depth of regulatory review of the project will be formally determined
)
by an assigned Regulatory Component Manager.
Designated Regulatory Component Functions will review the project proposal for impact on environmental protection, health, safety, and/or safeguards programs; and, for compliance with applicable regulatory requirements.and conformance to regulatory commitments. Formal approvals will be documented as part of the review package.
.(d)
Ancillary Programs and Procedures
- Ancillary programs and procedures will be activated commensurate with identification of environmental protection, health, safety, and/or safeguards issues. Such programs will range from simple design reviews by cognizant multi-Docket No.70-11f(
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discipline Functions, through structured What-If/ Checklist or Hazards and Operability Analyses. Formal approvals will be documented as part of the review package.
The Regulatory Component may issue conditional, documented approvals for prelhninary and/or detailed project designs as the process advances.
(e)
Specific documents to be updated will be formally identified as the process advances.
(f)
Drawings that are generated, or modified, will be maintained in a "For Construction" state until applicable installation is completed.
Following installation, the "As Built" conditions will be recorded as " Released" drawings that represent actual system configuration.
(g)
A project will be formally closed by the assigned responsible individual signing the configuration change control form, attesting that all required documentation has been updated, all required training has been completed, and the project has been terminated.
3.2 MAINTENANCE The purpose of the maintenance program for safety-related systems and components at the Columbia Fuel Fabrication Facility (CFFF) will be to assure that this equipment is kept in a condition of readiness such that it is likely to perform its desired function when called upon to do so. The maintenance program will embrace three functional activities:
Programmed Maintenance, to include specified frequency calibrations; Periodic Functional Testing; and, Repair or Replacement, for systems and components that fail to perform to required standards.
3.2.1 PROGRAMMED MAINTENANCE OF SAFETY-RELATED SYSTEMS AND COMPONENTS The Manufacturing Component will utilize a suite of maintenance planning and control computer programs to initiate work orders for programmed maintenance, and to record details of the execution of the work orders. The computer programs will include procedures for programmed maintenance of safety-related systems and components -
prepared, reviewed, and approved in accordance with Subsection 3.4.1 of this Chapter.
The following safety-related systems and components will receive programmed maintenance:
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1 Air Compressors; Emergency Electrical Generators; e
Fire Detection and Fire Control; e
Natural Gas Valves; e
Nuclear Criticality Detection; e
Pellet Carts; e
Pressure Relief Valves; Steam Boilers.
Additional safety-related systems and components will be placed under programmed maintenance, as disclosed by the results of Integrated Safety Assessments described in Chapter 4.0 of this License Application. Until a system's Integrated Safety Assessment
'(ISA) is completed, safety-significant controls identified through enhancements of the system's Criticality Safety Analysis (CSA) or Criticality Safety Evaluation (CSE), and/or new controls identified through configuration management reviews of. modifications of the system, will be scheduled, as necessary, for programmed maintenance to assure that the controls are maintained at their original level of availability. Other specified safety-related controls for a system will be placed under programmed maintenance at the.
j discretion of the cognizant Regulatory Engineer Function.
Programmed maintenance of safety-related systems and components will include specified calibration and re-calibration of relevant instruments. Such calibration and re-calibration
- will be initiated and controlled by the maintenance planning and control computer programs. Discrimination between safety-related and non-safety-related calibrations will be by use of an entry on the. electronic instrument calibration card utility within the maintenance planning and control computer programs.
3.2.2 PERIODIC FUNCTIONAL TESTING OF SAFETY-RELATED SYSTEMS AND COMPONENTS The following safety-related systems and components will receive programmed j
maintenance at the frequencies indicated:
Plant-wide Fire Alarm System and Criticality Alarm System - Each working shift, one day per working week;
'l Plant-wide Hazard Warning System - Semiannual; Specified Safety-related Interlocks on Process Equipment - Annual;
)
e Hydrogen and Natural Gas Line Leak Tests Annual.
]
e Additional safety-related systems and components will be placed under periodic functional testing, based on the results of integrated safety assessments described in Chapter 4.0 of this License Application. Until a system's Integrated Safety Assessment l 1
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(ISA) is completed, safety-significant controls identified through enhancements of the system's Criticality Safety Analysis (CSA) or Criticality Safety Evaluation (CSE), and/or -
new controls identified through configuration management reviews of modifications of the system, will be scheduled, as necessary, for periodic functional testing to assure that the controls are maintained at their original level of availability. Other specified safety-related controls for a system will be scheduled for periodic functional testing at the discretion of the cognizant Regulatory Engineer Function.
3.2.3 REPAIR OF SAFETY-RELATED SYSTEMS AND COMPONENTS The maintenance planning and control computer-generated work orders and records will provide documentation of systems and components that have been repaired or replaced.
When a component of a safety-related system is repaired or replaced, the component will be field-tested to assure that it is likely to perform its desired function when called upon to do so.
If the perfonnance of a repaired or replaced safety-related component could be different from that of the original component,' the safety-related system will be field-tested to assure that it is likely to perform its desired function when called upon to do so.
3.3 QUALITY ASSURANCE The purpose of the formal quality assurance (QA) program for safety-significant processing equipment at the Columbia Fuel Fabrication Facility (CFFF) will be to assure that such equipment is designed, installed, operated, and maintained so that it will perform its desired function when called upon to do so. This quality assurance program will be in addition to the quality assurance programs for nuclear components and fuel shipping containers; however, the three programs may share common elements (e.g.,
organization structures, tool and gage control, change management, etc.).
3.3.1 QA PROGRAM STRUCTURE To the maximum extent practicable, the QA program for safety-significant processing equipment will utilize elements of the facility's Process Safety. Management (PSM)
. program (29 CFR 1910.119), stmetured to include licensed radioactive materials. The Engineering Component will maintain a detailed matrix that graphically demonstrates how the PSM program elements will address the following QA program criteria:
(a)
QA Organization;
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(b)
QA Program; (c)
Equipment / System Design Control; (d)
Procurement Documentation Control;.
(e)
Instructions, Procedures, and Drawings; i
(f)
Document Control; (g)
Control of Purchased Materials, Equipment, and Services; (h)
Identification and Control of Materials, Parts, and Components; (i)
Control of Special Processes; 1
(i) ~
Internal Inspections;
.I (k)
Test Control; (1)
Control of Measuring and Test Equipment; (m)
Handling, Storage, and Shipping Controls; (n)
Inspection, Test, and Operating Status; (o)
Control of Nonconforming Materials, Parts, or Components; (p)
Corrective action; (q)
QA Records;' and, (r)
Audits.
The PSM program will then be supplemented, as required, to assure detailed inclusion of all QA criteria.
3.3.2 GRADED APPROACH The " graded approach" will be addressed by performing a systematic and integrated assessment of the hazards at the facility; then, identifying the. safety systems and components that are intended to prevent, or mitigate the consequences of, these hazards; i
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1 then, to apply the programs of assurance which provide the appropriate level of quality.
(Completion of these assessments, as an ancillary supporting process, will be phased-in according to the implementation schedule for the facility's Integrated Safety Assessment.)
l
' Where judgement is required, salient decisions will be documented; when quality requirements are determined not to be necessary, the bases will be documented.
(a)
Quality Level A; Cmcial Safety Systems
]
l These systems are crucial to safety and, therefore, will receive rigorous attention l
to installation, operation, and quality assurance.
They will be defined by controlling the following hazard consequences:
I l
Greater than or equal to 5 rem dose equivalent to an individual offsite; e
and/or, Greater than or equal to 10 milligrams soluble Uranium intake by an e
individual offsite; and/or, Greater than or equal to 25 milligrams HF/m' exposure to an individual e
i offsite.
l Crucial safety systems will require full application of the QA program requirements, where each of the 18 criteria that could apply are specifically addressed. They will be initially qualified when placed into service, and will be requalified as required, using controlled methods and procedures.
(b)
Quality Level B; Important Safety Systems These systems are important to safety and, therefore, will include key aspects that L
require high quality judgement or attention to detail. The key aspects will be l
identified and documented in the hazard assessment. They will be defined by contiolling the following hazard consequences:
Greater than regulatory limits to an individual offsite; e
Death or serious injury to an individual onsite.
Important safety systems will require selected application of the QA program requirements, where elements of the 18 criteria that the Quality Component determines will apply are specifically addressed.
(c)
Quality Level C; Safety Margin Improvemem Systems These systems have safety implications, but are neither crucial nor important to safety. They do not require specified attention to quality assurance, and no Docket No.
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extraordinary level of safety detail is applied.
Safeiy margin improvement systems will be maintained aIxi operated as part of routine and pmdent industry practice.
3.3.3 ADDITIONAL QA PROGRAM COMMITMENTS AND EXCLUSIONS The program will be designed and incorporated, as an ancillary supporting process of the facility's Integrated Safety Assessment, such that it becomes an integral part of routine CFFF operations.
The program will be performance-based. Quality assurance decisions will be based, to the extent practicable, on system performance histories.
The program descriptions will be documented in facility procedures that specify responsibility, authority, and accountability for all program elements. PSM program elements and other facility programs and pixedures important to quality assurance, will be specifically cross-referenced; and, the cross-reference will be maintained by the Quality Component for future audit.
The program elements will be conducted in accordance with approved, written procedures. Training to these procedures will be conducted to ensure the program operates effectively.
The program will require documented records to demonstrate compliance with program requirements.
The program will include a level of checks and balances through functional separation and audit. The program will be developed to incorporate quality-at-the-source concepts.
Routine quality assurance for safety systems may be performed by the functions responsible for operating the systems.
The program will embrace issues identification, remedial actions, and management control elements to ensure that deficiencies, deviations, and defective equipment and components are disclosed, and corrected, in a timely manner.
The program will be forward-fitting upon implementation.
It will be a bounding assumption that existing systems were appropriately designed, installed, and operated in accordance with applicable requirements and acceptable practices. Existing systems will not be back-fhted except for component replacement, system modification, and/or actions I
arising from internal investigations and/or external disclosures such as NRC Information Notices. Such back-fitting will always be at the discretion of the Plant Manager, as advised by the Engineering and Regulatory Components.
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Until a system's Integrated Safety Assessment (ISA) is completed, safety-significant controls identified through enhancements of the system's Criticality Safety Analysis (CSA) or Criticality Safety Evaluation (CSE), and/or new controls identified through configuration management reviews of modifications of the system, will be verified, as necessary, to assure they match the requirements identified in the design criteria. That is, all such' controls will be examined in the "as built" condition, and pre-operationally _
tested, to validate the design and to verify the quality of the installation and the reliability of the controls. Other specified safety-related controls for a system will be scheduled fo'r "as built" examination and pre-operational testing at the discretion of the cognizant Regulatory Engineer Function.
3.4 PROCEDURES, TRAINING AND QUALIFICATION At the Columbia Fuel Fabrication Facility (CFFF), procedures, training and qualification will be integratc j into a combined process to assure that environmental protection, health, safety, quality, and safeguards programs are being conducted in accordance with Westinghouse policies, and in accordance with commitments to Regulatory Agencies.
Elements of this integrated process will be developed by knowledgeable Component staff, will be reviewed and approved by cognizant individuals in affected Components, and will be authorized for implementation by Component Management at a level that is responsible and accountable for the operations covered.
3.4.1 PROCEDURES Operations to assure safe, compliant activities involving nuclear material will be conducted in accordance with approved procedures.
Approved procedures will be maintained and controlled by an Electronic Procedure System. Approved procedures will provide the basis for training of all pasonnel involved in operations with nuclear material at the facility.
Structured hazards analyses, as conducted in accordance with Chaper 4.0 of this License Application, will include human factors analysis of applicable procedures, as described in Section 3.5 of this License Application.
(a)
Regulatory-Significant Procedure Structure CFFF procedures will be classified into three general categories:
(a.1) Category-1 Procedures l
[
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Category-1 procedures will be for use by the Regulatory Component. The salient 1
utility of such procedures will be to provide health, safety, and safeguards training and instructions for Regulatory Functions. They will be prepared, and approved for issuing, by Regulatory Functions assigned by a cognizant Regulatory Component Manager; and, will be reviewed, and approved for issuing, by the cognizant Regulatory Component Manager.
The Category-1 scope will group sets of procedures into such subcategories as:
Administration; Health Physics:
e Nuclear Criticality Safety Environmental Protection e
Safeguards e
Shipment and Transportation; e
Instruments; e
Surveys; e
Dosimetry; e
Bioassay; and, Laboratory Practices e
l Changes to Category-1 Procedures will be prepared, and approved for issuing, by Regulatory Functions assigned by a cognizant Regulatory Component Manager; and will be reviewed, and approved for issuing, by the cognizant Regulatory Component Manager.
(a.2) Category-2 Procedures Category-2 procedures will be for use by individuals outside the Regulatory Component, and deal exclusively with regulatory practices. The salient utilities of such procedures will be to provide health, safety, and safeguards training and instructions for Engineering, Manufacturing, and Quality Functions; and, for use by these Functions in preparing Category-3 Procedures.
They will present regulatory guidance methodology acceptable to the Regulatory Component. They will be prepared, and approved for issuing, by Regulatory Functions assigned by a cognizant Regulatory Component Manager; and, will be reviewed, and approved for issuing, by the cognizant Regulatory Component Manager.
The Category-2 scope will be similar to, and may in many cases overlap, that for Category as applicable to use outside the Regulatory Component.
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Changes to Category-2 Procedures will be prepared, and approved for issuing, by Regulatory Functions assigned by a cognizant Regulatory Component Manager; 1-and, will be reviewed, and approved for issuing, by the cognizant Regulatory Component Manager.m (a.3) ' Category-3 Procedures Category-3 procedures will be for use by responsible individuals outside the Regulatory Component. The salient utility of such procedures will be to provide training and instructions -- including health, safety, and safeguards - for the Operations, Maintenance, Inspection, and Analytical Services Functions. They j
will be prepared, and approved for issuing, by Component Functions assigned by a cognizant Component Manager, based on consideration of applicable Category-2.
Procedures and/or consultation with cognizant Regulatory Component Engineers; I
and, will be reviewed, and approved for issuing, by the cognizant Component I
Manager.
The scope of Category-3 Procedures will be as determined by the cognizant Component Manager.
Changes to Category-3 Procedures will be prepared, and approved for issuing, by Component Functions assigned by a cognizant Component Manager, and will be reviewed, and approved for issuing, by the cogmzant Component Manager.
(b)
Issuance, Approval, and Communication of Contents of Procedures Acceptable practices for environmental protection, health, safety and safeguards activities will be provided to operations Components in documented procedures that are approved, by the Regulatory Component, for electronic issue. Contents of these procedures will be communicated to operations personnel, by Component Management, through incorporation into specified operating and/or quality assurance procedures.
Regulatory-significant practices in operations and quality assurance procedures, and changes to such procedures, will be issued by cognizant Components in accordance with documented policies for procedure preparation, review, and approval. Specifically, Regulatory Component approvals will be required for all regulatory aspects of procedures, and their changes, involving the storage, handling, processing, inspection, and/or transport of nuclear materials.
~ Component Management will be responsible for assuring and documenting that contents of these procedures are communicated to appropriate personnel through Docket No.
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training programs, access to the Electronic Systems, and/or posting of instructions.
(c)
Procedure Review Frequencies 1
Maximum frequencies of reviews-for-updating for regulatory-significant procedures will be:
Annual, for Category-1 and Category-2 Procedures; and, Biennial for Category-3 Procedures.
(d)
Procedure Compliance A formal system will be maintained to enable employees to report inadequate procedures, and/or inability to follow procedures, to their First level Managers for follow-up action.
First level Managers will enable, and require, compliance with all regulatory-significant procedures. This will be accomplished by providing ready employee access to procedures, requiring documented employee procedure review and acknowledgement, then evaluating employee performance with respect to procedure compliance on a continuing baris. Employees will receive additional instruction, if determined necessary by the First Level Manager evaluations; and, if procedures are deliberately or repeatedly violated, disciplinary action will be taken in accordance with established Westinghouse policies.
3.4.2 TRAINING AND QUALIFICATION Training will be provided for every individual in the Columbia Fuel Fabrication Facility (CFFF), commensurate with their duties. Formal training programs will be developed and implemented to enhance and augment procedure review and acknowledgement described in Paragraph 3.4.1(d) of this Chapter, and training responsibilities described in Chapter 2.0 of this License Application. Such training programs will be performance-based; and as such, will incorporate the structured elements of job and task analysis, learning objectives, instructional methodology, implementation, and evaluation and feedback. In addition, training of Nuclear Criticality Safety Function Engineers will include qualification by cognizant Regulatory Component Management that goes beyond the position requirements described in Chapter 2.0 of this License Application. The l
programs will be structured such that specified training and qualification requirements I
will be met prior to safety-significant positions being fully assumed, or covered tasks j
being independently performed. Training records will be maintained in accordance with Section 3.8 of this Chapter.
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i
(a)
General, Topical, and Refresher Training i
All new employees will receive training relative to safety aspects concerning radiation and radioactive materials; risks involved in receiving low level radiation exposure; basic criteria and practices for radiation protection, nuclear criticality safety (based upon selected guidance from ANSI /ANS-8.20-1991, facility operating experience, and area specific requirements), chemical and fire safety, maintaining radiation exposures and radioactivity in effluents As IAw As Reasonably Achievable (ALARA), and material safeguards. Facility visitors will
. either be provided with equivalent training (commensurate with their visit's scope); and/or, will be escorted by trained employees.
Employees or visitors for whom respiratory protection devices might be required,.
within the scope of their work, will receive pre-work training in the proper use of such devices.
Employees designated to take part in emergency response to facility accidents or -
incidents will receive training commensurate with their assigned activities during such response.
Employees who work with nuclear materials will receive regulatory refresher training on a biennial basis. This training will consist of:
Providing each employee with a current revision of the Regulatory Affairs Training Manual; Presenting each employee supplementary videotaped instruction on general regulatory issues; and, Requiring each employee to succes'sfully pass an examination.
e 1
The Training Manual will include such subjects as:
ALARA; General health physics practices; e
Health physics rules and recommendations; e
Area-specific health physics practices; e
General nuclear criticality safety practices; -
Area-specific nuclear criticality safety practices; Industrial safety and hygiene, and fire safety, practices; e
Chemical Area work practices; e
Radiation risks; e
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Emergency planning; and, e
Safeguards.
e Employees who are absent from the facility during scheduled regulatory refresher training will receive such training within one month of their return to work.
(b)
Traidg and Qualification of Nuclear Criticality Safety Function Engineers Nuclear Criticality Safety Function Engineers will develop skills and abilities directed by the cognizant Regulatory Component Manager, who will evaluate fundamental development methodologies for applicability and utilization on a case-by-cases basis. Examples of development methods include:
A nuclear criticality safety short course; Westinghouse auditing certification; e
American Nuclear Society Standards development and review; Facility criticality safety handbook development and review; e
A stmetured hazards analysis course; A structured human factors course; and, Criticality safety calculations certification.
o Demonstrated performance of Nuclear Criticality Safety Function Engineers skills and abilities will be formally reviewed and documented by the cognizant Regulatory Component Manager and the senior Regulatory Component Manager.
Performance evaluated by the Managers, for review on a case-by-case basis, will include:
' Reports of internal audits and inspections conducted; e
Feedback from worker training presented; e
Criticality safety analyses and evaluations performed.
e Qualification of each Nuclear Criticality Safety Function Engineer will be formally documented by the cognizant Regulatory Component Manager and the senior Regulatory Component Manager - prior to the Function position being fully assumed, or crucial tasks being independently performed.
(c)
Training and Qualification of Health Physics Technicians Training and qualification prerequisites for a Health Physics Technician will include, as a minimum, a high school diploma or equivalent.
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Health Physics Technicians will develop skills and abilities, as directed by the cognizant Regulatory Component Manager. Methods evaluated by the cognizant Manager for qualification, on a case-by-case basis, will include:
Documented acknowledgement of applicable procedures; Emergency preparedness training; and/or e
Applicable skills competency training.
e 3.5 HUMAN FACTORS Human factors concepts will be employed at the Columbia Fuel Fabrication Facility (CFFF), in recognition of how the total job environment - areas, equipment, training, and procedures - shapes the expectations, thoughts, and decisions of employees who work with licensed materials. A human factors awareness will be developed at various levels of the organization, and structured human factors analyses will be performed.
Because the operating philosophy of the orga'lization is strongly embodied in procedures, as described in Subsection 3.4.1 of this CiVoter, procedures will receive particular human factors attention.
3.5.1 DEVELOPMENT OF HUMAN FACTORS AWARENESS To enable integration of human factors concepts into facility operations, an initial, formal course - prepared and presented by recognized human factors experts -- will be provided for the Plant Manager; all Engineering, Manufacturing, Regulatory, and Quality Component Managers; and, designated Functions from these Components. The course will address the following elements, including exercises to enhance learned skills:
(a)
Process Safety Management; (b)
Human Factors Concepts; (c)
Performance Shaping Factors For Hardware; (d)
Performance Shaping Factors For Procedures; i
'(c)
Analysis Preparation; i
(f)
Error-Likely Situations;
)
(g)
Procedure Analysis Techniques;
)
o (h)
Worker Self-Checking Techniques; and, I
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(i)
Supervisor Coaching Principles.
3.5.2 STRUCTURED HUMAN FACTORS ANALYSIS A part of the CFFF Integrated Safety Assessments, described in Chapter 4.0 of this License Application, will include a structured human factors analysis of assessed system procedures. These analyses will be led by an individual who has completed a formal human factors course. The analyses will embrace the following:
(a)
Using Procedure-Specific Guide Words For Stmetured Analysis Of Procedures.
(b)
Minimizing Opportunities For Human Errors Of Omission and Commission Related To Procedures.
Results of the structured analyses, including findings and reconunendations for i
improvements, will be documented in formal reports to cogmzant Component Management.
3.6 AUDITS AND SELF-ASSESSMENTS The bases of the Columbia Fuel Fabrication Facility (CFFF) Audits and Self-Assessment program will be the performance-based reporting process described in Section 3.7 of this Chapter, the performance-based internal inspection and audit program, and facility management self-assessment of regulatory program performance.
3.6.1 PERFORMANCE-BASEDINTERNALINSPECTIONS AND AUDITS (a)
INFORMAL INSPECTIONS i
Regulatory Component personnel on duty, including Regulatory Component management, will conduct continuing informal inspections of regulatory program performance in the course of their routine duties. Observed process upsets and procedural inadequacies will be promptly reported to the cognizant First Level Component Manager for remedial action. Repeated upsets and inadequacies will be reported to the cognizant Regulatory Component Manager, who in turn will report them to increasingly higher levels of Component Management until effective remedial action has been taken. Such repeated upsets and inadequacies will be documented in monthly formal audits to assure applicable tracking and resolutions.
1 (b)
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i Cognizant Regulatory Function Engineers will conduct monthly formal audits of regulatory program performance. The auditors will have the technical capability, I
and will be formally directed by Regulatory Component management, to find i
process upsets and procedural inadequacies well beyond those surfaced by simple
' paperwork reviews. The audits will include reviews.of items entered into the performance-based reporting process, and repeated upsets and inadequacies reported to Regulatory Component management, for the areas being audited; and, detailed area walkdowns. Disclosed upsets and inadequacies will be formally documented in a report to cognizant First level Component Managers; and, will be tracked by the audit team leader until appropriately addressed.
3.6.2 FACILITY MANAGEMENT SELF-ASSESSMENT The purpose of the self-assessment program will be to provide a means to assure that deficiencies in regulatory performance are identified and corrected to Westinghouse management standards.
The Plant Manager will document CFFF policy on the purpose and objectives of self-assessment to Component Managers, including aggressive demand for quality assessment performance.
The management self-assessment organization will be the Regulatory Compliance Committee (RCC) described in Chapter 2.0 of this License Application. RCC members will be provided with the Nuclear Regulatory Commission Staff's views concerning self-assessment - rarticularly, that the function of such assessment will be to aggressively disclose and forcefully report identified process upsets and procedural inadequacies before they seJ-reveal and/or Regulatory Agencies find them.
On a semi-annual basis the following assessment parameters will be summarized and trended by the Regulatory Component:
A summary of items documented in the performance-based reporting process; A summary of upsets and inadequacies documented in performance-based internal e
audit reports; l
Facility Collective Dose Equivalent; j
Facility average Total Effective Dose Equivalent; l
l Top 10 facility workers' Total Effective Dose Equivalents; l
Regulatory Agency notifications; i
Ratio of Recordable Incident Rate to SIC code average; e
Lost time accidents per production hour; j
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Results of Special Nuclear Material Physical Inventory (annual);
e Emergency response team activations; e
Radioactive emissions in gaseous effluents; e
Radioactive emissions in liquid e'lluents; e
Radioactive material transportation incidents; and, e
Regulatory Agency violations.
e The summaries and trends will be formally reviewed by the RCC, particularly for need to be addressed by initiatives of the Safety Margin Improvement Program described in Chapter 3.0 of this License Application.
3.7 INCIDENT Ih7ESTIGATIONS At the Columbia Fuel Fabrication Facility (CFFF), the organizational structure described in Chapter 2.0 of this License Application, and procedures in accordance with Subsection 3.4 of this Chapter, will provide for: systematic investigation of abnormal events; making decisions on corrective measures to prevent recurrence of such events; and, follow-up on the implementation of the preventive measurer. Further, the CFFF will have in-place a structured methodology for determining and categorizing the root cause(s) of the failure (s) that led to investigated events.
l 3.7.1 INTERNAL REPORTING OF INCIDENTS l
l
_ A formal system will be maintained to enable employees to report process upsets and procedure inadequacies to their First 12 vel Managers for follow-up action; and, employees will be instructed in its use.
Documentation of this performance-based reporting process will provide for the following information:
Event identification number, date, and time.
Names of the report originatcr and the First level Manager, shift number, and l
event description; Immediate action taken by the First level Managert e
Explanation of ultimate event closure; and, Acknowledgement of closure (and date acknowledged) by the cognizant Engineering Function Engineer, the cognizant Regulatory Function Engineer, the originator's Fir:;t Level Manager, and the originator.
Potential safety-significant reports will be forwarded to the Regulatory Component for evaluation and determination of necessity for action by the incident review committee, as described in Subsection 3.7.2 of uns Chapter. All documentation of the performance-Docket No.
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based reporting process for an area will be reviewed as a part of the formal audits of the area, as described in Paragraph 3.6.1(b) of this Chapter.
3.7.2 STRUCTURED INCIDENT EVALUATION An incident review committee - comprised of the Engineering Component Senior l
Manager, the - Manufacturing Component Senior Manager, and the Regulatory Component Senior Manager - will determine if reported process upsets and/or procedure inadequacies are to undergo structured incident evaluation.
Structured incident evaluations will be maintained by a datapack process. Documentation of this process will provide for the following information:
Results of a Root Cause Analysis, led by an individual with formal training in e
j; l
conducting such an analysis, including iecommendations; Status of corrective action (s) implementation; e
l Regulatory assessment; Notification documentation; e
Training documentation; e
Plant-wide applicability assessment; and, e
Miscellaneous information pertaining to the incident and/or the evaluation.
L 3.7.3 NOTIFICATION OF REGULATORY AGENCIES Cognizant Regulatory Agencies will be promptly notified of major safety incidents in l
accordance with all requirements from 10 CFR Parts 20 and 70. In particular, as points of additional clarification, the NRC Operations Center will be notified of the following types of incidents, within the time limits prescribed:
(a) 1-Hour Notifications (a.1) Any incident for which an Alert or Site Area Emergency has been dec19. red, as prescribed by the Site Emergency Plan described in Chapter 9.0 of this License Application.
'(a.2) Any incident involving Quality level A systems, for which accident controls cannot be initiated, whether or not regulatory limits are exceeded.
g (b) 4-Hour Notifications l
(b.1) Any incident involving Quality Level B systems, for which accident controls cannot be initiated, whether or not regulatory limits are exceeded.
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(b.2) Any nuclear criticality safety incident for which less than double contingency protection remains (multi-parameter control or single-parameter control) and:
Greater than a safe mass is involved and double contingency protection is e
not restored within four (4) hours.
Greater than a safe mass is involved and controls are restored within four (4) hours, but:
- i. Only single contingency protection is restored and more than one of the original controls were modified or replaced.
ii. Double contingency protection is restored but multiple original controls under both contingencies were modified or replaced.
j (b.3) Any determination that a criticality safety analysis or evaluation was deficient and that double contingency protection, in fact, does not exist.
(b.4) Any unanticipated /unanalyzed nuclear criticality safety incident for which the severity and remedy are not readily determined.
(c) 24-hour Notifications (c.1) Any incident for which the work area is unavailable for normal use for an entire day, following a loss of radioactivity contamination control.
i (c.2) Any incident for which Quality level A or B system safety equipment is not performing its intended function.
1 (c.3) Any incident for which an employee, having removable radioactivity contamination, receives medical treatment outside of facility contamination control I
areas.
j i
(c.4) Any incident for which a fire or explosion damages nuclear fuel and its processing equipment or container.
(c.5) Any nuclear criticality safety incident for which less than double contingency protection remains (multi-parameter control or single-parameter control) and:
Less than a safe mass is involved.
e Greater than a safe mass is involved, but a sufficient number of the controls that were lost are restored within four (4) hours such that double i
contingency protection is restored.
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(d)
A procedure will be prepared, maintained, and followed -- in accordance with Subsection 3.4.1 of this Chapter - that details the information to be included in a notification.
Each notification of a nuclear criticality safety incident will include the following information:
Whether the notification is the result of an event, or of a deficient nuclear criticality safety analysis (including the time period for which the deficiency existed);
The significance of the incident; e
Potential criticality pathways involved, including brief scenario (s) of how accidental criticality could occur; Controlled parameters -- mass, moderation, geometry, concentration, etc. --
e involved; Estimated amount, enrichment, and form of licensed material involvec -
including applicable process limits and the percent of worst-case critical mass of the material, in the configuration, involved; A description of the involved failures or deficiencies - including applicable e
nuclear criticality safety controls or control systems; and, Corrective actions to restore safety systems, and when each was implemented.
e 3.8 RECORDKEEPING AND REPORTING The Columbia Fuel Fabrication Facility will identify, maintain, preserve, control, and destroy records - as defined in the records management section of the controller's manual -- in accordance with the guidclines, procedures, and practices set fonh by the Westinghouse Electric Corporation. Such records, specificalle required by applicable regulations, will be maintained in accordance with those re a;itions.
Reporting of c
records data will be as prescribed by applicable regulations.
3.8.1 RECORDS Written procedures, prepared and maintained in accordance with Subsection 3.4.1 of this Chapter, will specify the management program for licensed activity records; including:
(a)
Environmental Surveys; (b)
Radiation And Contamination Surveys; (c)
Personnel Exposures:
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l (d)
Instrument Calibration Results;
-(e)
Nuclear Criticality Safety Evaluations, Analyses and Methodology Validations; (f)
Audit And Inspection Reports; (g)
ALARA Reports; (h)
Regulatorv "ompliance Committee Meeting Minutes; (i)
Employee Training And Re-training Documentation; (j)'
Records Of Plant Alterations Or Additions; (k)
Documentation Of Abnormal Or Atypical Occurrences And Events Associated With Radioactivity Releases; (1)
Decontamination And Decommissioning Files; and, (m)
Other Such Records Required By the Regulations.
These procedures will include Records Flow Schedules, which list:
Record category, Name of record; e
Form numbers; e
Retention period in active files; e
Retention period in the central records bureau; and, e
Retention period in the records center.
e Records of tests, measurements, and surveys required to document compliance with
. conditions of operating licenses and permits will be retained for at least three years,
. unless otherwise specified in the regulations.
Records of nuclear criticality safety analyses will be retained for the lifetime of the facility.
3.8.2 RECORDS RETRIEVAL i
All retained records will be stored, and maintained readily accessible, in order to meet time restraints relative to their use. Retained records will be as complete and detailed as
]
necessary to enable traceability to original source data.
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The records retention system will include the capability to retrieve records within 24-hours for records generated within the past 12-months; and, inside 7-calendar-days for older generation periods.
3.8.3 RECORDS RE-CREATION Prudent measures of protection and redundancy will be afforded such that acts of record alteration or inadvertent destmetion will not foreclose capability for reconstructing a complete and correct set of required records.
In cases where protective measures fail, and a particular record is lost or inadvertently destroyed, a reconstruction may be generated using source data applicable to the time the subject record was originally created. When document is just partially missing, all salvaged portions will be attached to the reconstruction. If source data is not available for re-creating a missing record, the record may be reconstructed using inference to data relative to other documents for similar information and time periods.
l 3.8.4 REPORTS A detailed listing of reports required by NRC regulations will be maintained and followed. This listing will document:
Reference to applicable regulations; Descriptions of the reports required; and, Frequencies at which the reports must be submitted.
I i
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CHAPTER 6.0 NUCLEAR CRITICALITY SAFETY 6.1 CONTROL METHODOLOGY AND PRINCIPLES 6.1.1 GENERAL CONTROL PROGRAM PRACTICES The Double Contingency Principle (ANSI /ANS-8.1) will be the basis for design and operation of processes using special nuclear material (SNM) within the Columbia Fuel Fabrication Facility (CFFF). Whete practicable, all process designs will incorporate sufficient factors of safety to require at least two unlikely, independent, arul concurrent changes in process conditions before a criticality accident is possible. In those instaxes where at least two independent controls are utilized to prevent changes in one process condition, sufficient redundancy and diversity of controls will be utilized. For each significant process within the system, a defense of one or more controlled parameters will be employed and documented within the Criticality Safety Analysis (CSA), Criticality Safety Evaluation (CSE) or Integrated Safety Assessment (ISA). The defense consists of the set of bounding assumptions, criticality safety litrits, and criticality saic!y cWtraints that, as a set, are uniquely sufficient to maintain the minimum suberitical ma: gin against an initiating event.
Criticality Safety Analyses and Evaluations are utilized to identify the specific limits and controls necessary for the safe and effective operation of a process. Types of nuclear criticality safety controls are listed in Subsection 6.1.2. Nuclear criticality safety controls will be included in the process design criteria. These controls will undergo a functional verification process prior to use in any system as described in Subsection 3.3.3 of this License Application, to assure reliability of intended function. A program for routine testing and maintenance, as described in Subsection 3.2.2 and 3.2.3 of this License
)
l Application, will assure continued availability of these controls. Controls which require periodic functional testing will be identified in an implementing procedure.
6.1.2 METfl0DS OF CRITICALITY SAFETY CONTROL I
The relative effectiveness and reliability of controls will be considered during the Criticality Safety Analysis and Evaluation process. Passive engineered controls will be L
preferred over all other tystem controls and will be utilized when available and appropriate. Active engineered controls will be the next preferred method of control.
Administrative controls will be the least preferred.
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l l-(a).
Passive Engineered These will be controls which require no operrcor action or other response to be L
effective when called upon to ensure nuc5 eat criticality safety. Examples of passive engineered controls include favorable geometry equipment such as structurally robust cylinders.
(b)
Active Engineered These will be controls which use a sensed signal or condition to automatically initiate an action to prevent an undesired condition or process from continuing, to ensure nuclear criticality safety. An example of an active engineered control is a shutoff valve actuated by an inline detector signal.
(c)
Administrative These will be controls which rely on operator intervention. While such controls are considered to be necessary, and are acceptable, their use will be limited to i
process systems which, in the judgment of the Nuclear Criticality Safety Function, do not lend themselves to engineered controls. Administrative controls include operator actions which are taken in accordance with a written procedure, operator verificatica of information with the assistance of computer terminals, actions taken in response to process alarms, etc.
6.1.3 CRITICALITY SAFETY CONTROLLED PARAMETERS Nuclear criticality safety will be achieved by controlling one or more parameters of a system within subcritical limits with sufficient factors. of safety as described in Subsection 6.1.1 of this License Application.
The Criticality Safety Analysis or Evaluation processes will be used to identify the significant parameters affected within j
a particular system. All assumptions relating to process / equipment / material theory, function, and operation, including credible upset conditions, will be justified, documented, and independently reviewed.
Specific control parameters that will be considered during the review process are:
(a)
Geometry Geometry control may be used to limit the shape of uranium within specific process operations or vessels, and within storage, transportation, or disposal containers.
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(a.1) Safe geometry defines the characteristic dimension of importance for a single unit of' a specified shape such that nuclear criticality safety will not be dependent on any other parameter except enrichment.
(a.2) Favorable geometry defines the characteristic dimension of importance for a single unit of a specified shape such that criticality safety will be maintained in conjunction with one or more other parameters such as material form, concentration, and/or reflection. Favorable geometry may also be achieved through other means, including level control.
1 (a.3) Geometry control systems will be analyzed. and evaluated for fabrication tolerances and dimensional changes that may occur through corrosion, wear, or mechanical distortion. In addition, these systems will include provisions for periodic inspection if credible conditions exist for changes in the dimensions of the equipment that may result in the inability to meet nuclear criticality safety limits.
(b)
Mass Mass control may be utilized to limit the quantity of uranium within specific i
process operations or vessels, and within storage, transportation, or disposal l
containers. Mass control may be used on its own or in combination with other control methods. Analytical or non-destructive methods will be employed to verify the mass measurements for a specific quantity of material.
Whenever mass control is established for individual rooms or groups of rooms, detailed records will be maintained for mass transfers into and out of these rooms. Establishment of mass limits will involve consideration of potential moderation, reflection, geometry, spacing, and material concentration.
The evaluation will consider normal operations and expected process upsets for determination of the actual mass limit for the system and for the definition of subsequent controls.
(c)
Moderation I
Moderation control may be used for nuclear criticality safety control of systems I
within the facility, and if utilized, will follow the guidelines of ANSI /ANS-8.22 and will incorporate the criteria identified below:
(c.1) Controls will be established to ensure that the interstitial moderator is maintained within the analyzed system defined limits.
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l Two independent controls / measurements or the analysis of two independent samples will be utilized to document system moderator content. The system for collecting, preparing, analyzing, and posting of results pertaining to sample evaluation will be designed to ensure the results obtained are independent.
-(c.2) Controls will be erablished to remove uncontrolled moderator prior to L
returning a system to,roduction and to prevent uncontrolled moderator from entering the system aftu initial loading has occurred. The minimum protection will be that two independent barriers preventing uncontrolled moderator from entering the system must fail before the system can be compromised. Examples include a combination of system containment / roof, system containment / pipes, I
double roofs, etc. A method for detection of failure of the outermost barrier will be established (unless the failure would be promptly self-disclosing), and a l
program to maintain the quality of the barrier will be in place and routinely inspected.
All outermost barriers will be tested for leakage during initial f
installation.
(c.3) The evaluation of a process or system under moderation control will include the establishment of limits for the ratio of maximum moderator to fissile material for both normal operating conditions and credible process upsets and this l
analysis will be supported by parametric studies. Transportation of materials outside of moderation control areas will be proceduralized. Mainunance and inspection programs will be defm' ed and implemented through plant procedures, t
The quality and basis for selection of the barriers will be documented within the CSA, CSE, or ISA process. Controls for the intrchetion and usage of moderating materials within areas that are under moderation control will be defined and approved by the Nuclear Criticality Safety Function.
(d)
Enrichment Enrichment ' control may be utilized to limit the percent "U within a process, 2
vessel, or container, thus providing a method for nuclear criticality safety control.. Active engineered and/or admwmtive controls will be required to verify enrichment and to prevent the introduction of uranium at unacceptable enrichment levels within a defined system.
In cases where enrichment control is not utilized, the maximum credible enrichment will be assumed, l
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(e)
Concentration.
Limiting concentration may be used for nuclear criticality safety control of systems within the facility, and when utilized, will incorporate the following criteria:
-(e.1) Controls will be established to ensure that the concentration level is maintained within the analyzed system defined limits.
Two independent controls / measurements or the analysis of two independent samples will be _ utilized to document concentration level The system for collecting, preparing, analyzing, and posting of results pertaining to sample evaluation will be designed to ensure the results obtained are independent.
(e.2) Controls will be established to prevent concentration withm the system after initial loading has occurred.
Concentration events inch.de evaporation, precipitation, freezing, settling, or chemical phase changes. Each system will have in place controls necessary to detect and/or mitigate the effects of internal concentration within the system.
(e.3). For those systems which utilize concentration as a controlled parameter for i
ensuring nuclear criticality safety, the CSA, CSE, or ISA will demonstrate the solubility limits of the SNM composition, will identify possible precipitating agents to the operators through procedures, will ensure appropriate precautions -
are being taken to assure such agents are not introduced, and will provide a positive means of preventing unwanted transfers (if a possibility exists for-precipitating agents to be transferred via connected processes).
(f)
Reflection Systems are designed and operated with the assumption of either full or partial reflection. Full reflection and partial reflection are defined as 12 inches and 1 inch H O equivalent, respectively. In systems where equipment location or 2
design limits the placement of moderating materials (including humans), near the specific system, partial reflection may be used.
For all system evaluations, the neutron reflection properties of the credible process environment will be analyzed. This will limit the available use of
. partial neutron reflection methodology for specific systems.
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1 (g)
Neutron Absorber -
Neutron absorbing mr.terials may be utilized to provide nuclear criticality safety control for a process, vessel or container. If used, the absorbers will be solid materials such as borosilicate-glass Raschig rings, gadolinium platet, borated stainless steel, or other solid neutron absorbing materials. The use of actron absorbers in this manner will be defined as a passive engineered control.
When Raschig rings are used to control nuclear criticality' safety, their u and maintenance will be in accordance with ANSI /ANS-8.5, with the following
- exceptions for basic solutions:
System PH maintained 511; and System temperature maintained s 60 degrees Celsius.
For other fixed neutron absorbers, the guidelines of ANSI /ANS-8.21 and the following requirements will apply:
l.
' The composition of the absorber will be measured and documented prior to first use; and,
. The presence of the absorber in a process, vessel or container will be verified on a frequency determined in the CSA, CSE, or ISA. The method of verification may take the form of traceability (i.e. serial number,etc.), visual inspection or direct measurement.
(h)
Composition The CSA, CSE, or ISA for each system will determine the effects of material composition within the process being analyzed and will identify the basis for composition selection pertinent to subsequent system modeling activities.
(i)
Heterogeneity The effects of heterogeneity within a system can be significant and will be considered within the CSA, CSE, or ISA when appropriate.
Nuclear criticality safety calculations have demonstrated that for systems where
)
the particle size is less than or equal to 150 microns, the process can be considered homogeneous. For systems where the particle size is greater than or equal to 1000 microns, the system is heterogeneous. The evaluation of systems l
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- where the particle size falls whhm the range of 150 - 1000 microns will take into consideration the effects of heterogeneity appropriate for the process being analyzed.
i (j)
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. When so utilized, this credit will be predicated upon the following requirements:
(j.1)
The bo' Wing assumptions will be defined through the CSA, CSE, or 1SA process and operational limits will be identified within each specific analysis; and, will be communicated, through training and procedures, to appropriate operations personnel.
(j.2)
Utilization of such process and/or material characteristics will be based on established physical or chemical reactions, known scientific principles, and/or facility-spedEc experimental data supported by operational history.
Examples of such an application include:
Conversion and oxidation processes that produce dry powder (s 10 wt.
% H O) as a product of high temperature reactions.
2 Experimental / historical process data demonstrating low moisture pickup (s 3 wt. % H O) from room air in ventilation equipment.
2 Experimental / historical process data. demonstrating uranium oxide powder flow characteristics to be directly proportional to the quantity of moisture present.
(k)
Interaction Nuclear criticalit; safety analyses will consider the potential effects of interaction. Methods utilized are:
(k.1)' Non-Interaction A non-interacting unit is defined as a unit that is under concentration control, contains fixed neutron absorbers, is under moderation control, or is spaced an i
i i
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approved distance from other units such that the multiplication of the subject
. unit is not increased.
1 l
Additionally, units may be considered non-interacting when they are separated by a 12-foot air distance or by 12-inches of full density water equivalent material. - For solid angle interaction analyses, a unit where the contribution to the total solid angle in the array is less than 0.005 steradians will also be l
considered non-interacting (provided the total of all such solid angles neglected l
is less than one half of the total solid angle for the system).
(k.2) Solid Angle Solid angle criteria may be used to determine the total solid angle subtended by each unit in an array, for interaction calculations. The solid angle criteria of TID-7016 (Rev. 2) will apply, as supplemented by reflector conditions of no more than full water reflection on six sides of the array or no more than the equivalent of full concrete reflection on three sides of the array.
The solid angle validation manual will be used for guidance in performing these calculations. Solid angles will be determined by the point-to-plane method. The l
multiplication factor of a single unit will be determined from validated computer calculations.
(k.3) Monte Carlo Individual unit multiplication and array interaction may be evaluated using computer codes (e.g., XSDRN, KENO, MCNP, etc.) for which validations have been documented. When array interaction is evaluated in this manner, the
)
maximum allowed K,,, including all biases and uncertainties at a one-sided 95/95 confidence level, will not exceed 0.95.
6.1.4 MOVABLE NON-FAVORABLE GEOMETRY (NFG) CONTAINERS Movable NFG container usage within the CFFF will be rigorously controlled and utilized only when other practical methods are unavailable. Prior to use of a movable NFO container in the Chemical Manufacturing Area, a comprehensive analysis will be performed. The key components of this analysis follow:
(a)
Operations personnel will provide justification for the use of the movable NFG container.
)
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(b)
Controls will be identified and implemented to provide assurance that the proposed storage and/or movement of an NFG container can be performed safely.
l (c)
A criticality safety analysis will be completed and approved prior to use of the
)
movable NFG container within the facility.
L 6.2 CONTROL DOCUMENTS
. A Criticality Safety Analysis (CSA), Criticality Safety Evaluation (CSE), or Integrated l
Safety Assessment (ISA) is prepared for each new system depending on the complexity j
of the system.
l 6.2.1 CRITICALITY SAFETY ANALYSIS (CSA) i The Criticality Safety Analysis is a subset of the Criticality Safety Evaluation (CSE) defined in Section 6.2.2 of this Chapter. The CSA identifies and documents the basir of nuclear criticality safety for a particular system. An update to an existing CSA will be made, as necessary, for system changes defined via the Configuration Management l
process described in Section 3 of this License Application. The level of detail for a particular CSA will be determined based on the complexity of the system or proposed changes, and will be documented by the Nuclear Criticality Safety Function Engineer and approved by the Nuclear Criticality Safety Function Manager.
Thus, the scope and content of a CSA will be customized to reflect the needs and characteristics of the system being analyzed. A CSA will be formatted similar to a CSE. The following information defm' es the important sections relevant to nuclear criticality safety:
(a)
Safety Analysis The safety analysis documents the results of a comprehensive nuclear criticality safety review of each component within the defined system.
The analysis identifies controlled parameters and establishes bounding assumptions for other i
parameters for the system. Calculations and sensitivity studies are perfonned as j-necessary to identify the margin of suberiticality. The safety analysis then j-demonstrates that Double Contingency Protection exists for the system when l
controls aie applied to the controlled parameters to prevent a contingency from occurring. Reliability of each control will be assessed and potential common mode failures will be considered.
j (b)
Double Contingency Protection Fault Trees i
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a i
i I
This section presents Fault Trees which demonstrate Double Contingency Protection for the analyzed system. Other information discussed in Section 6.0 of Chapter 4.0 of this License Application may also be included.
(c)
License Compliance The criticality safety basis and the bounding assumptions will be reviewed for compliance with Subsection 6.3 of this License Application.
(d)
Appendix 1
This section presents a summary of ancillary information (such as calculations, parametric sensitivity studies, references, etc.) for each defined system.
3 6.2.2 CRITICALITY SAFETY EVALUATION (CSE)
A CSA may be upgraded to a Criticality Safety Evaluation (CSE) with the addition of Sections 1.0, 2.0, 3.0, 4.0, and an entire Section 6.0, Process Hazards Analysis, of an ISA (described in Chapter 4.0 of the license application). (Similarly, the addition of other safety disciplines' analyses will upgrade a CSE to an ISA.) The nuclear criticality safety portion of a CSE or ISA is identical to the content of a CSA.
- 6.2.3 ANALYSIS METHODOLOGIES (a)
K,y Limit Validated computational methods will be used to calculate the k,n for systems with individual vessels or potential vessel interaction. This will include conditions expected to be encountered during routine operations, expected process upsets, and credible accident situations. Based on the results of these calculations, the sensitivity of key parameters with respect to the effect on K,,
will be evaluated for each system to assure that adequate system controls have been defined to demonstrate a sufficient margin of' safety for the analyzed system.
With respect to normal operating conditions and expected process upsets, there will be sufficient margin of safety to ensure that, based on these parameters, the calculated k,n is s 0.95, including all applicable biases and calculated uncertainties.
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l protection, there will be sufficient margin of safety to ensure that, based on these parameters, the calculated k r is < 1.00, including all applicable biases a
and calculated uncertainties.
(b)
Analytical Codes Criticality calculations are performed using the. SCALE system of modules including such codes as NITAWL-II and XSDRNPMS for cross-section -
generation and KENO-Va for reactivity calculations. Other computer codes, particularly Monte Carlo codes like MCNP and KENO-VI, may be used after validation as described in Paragraph (c) below. All methods are benchmarked to various critical experiments to verify their applicability prior to use.
(c)
Validation Techniques I
Nuclear criticality safety analyses will be conducted utilizing computerized 4
methodology which will be validated in accordance with the criteria described in Section 4.3 of ANSI /ANS-8.1.
-(d)
Computer Software and Hardware Configuration Control The configuration of the hardware " calculational platform" used in the support of software for nuclear criticality safety calculations will be maintained such that only authorized system administrators will be allowed to make system changes. System changes will be conducted in accordance with an approved configuration control program that addresses both hardware and software
. qualification. - System operability verification will be performed to alert users to any changes that would impact the operation of " codes" on the. calculational platform.
Software designated for use in nuclear criticality safety calculations on the calculational platform will be compiled into working code versions with executable files that are traceable by length, time, date, and version. Working code versions of compiled software will be qualified on the basis that physical critical experiments were modeled using an established methodology w;th the differences in experiment and analysis being used to calculate bias and j
uncertainty values to be applied to the obtained results.
Modifications to hardware or software, that are essential to the calculational process, will be followed by code operability verification; in which case, 1
selected calculations will be performed to verify identical results from previous analyses. Deviations noted in code verification that might alter the bias or l
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uncertainty will require re-qualification of the code prior to continued use.
-(e)'
Solid Angle Method
- Solid sngle analysis may be utili < uithin the CFFF as described in Subsection 6.1.3, Paragraph (k.2) of this Lix Ntion.
6.2.4 TECHNICAL REVIEW 1
Independent technical reviews of criticality safety analyses, evaluations, or calculations in support of limits specified in a CSA, CSE or ISA will be performed. A qualified reviewer will perform the independent technical review.
The technical reviewer will verify that the proposed calculational geometry model and configuration adequately represent the system being analyzed.
In addition, the reviewer will verify that the proposed material characterizations such as density, concentration, etc., adequately represent the system.
The technical review of the specific calculations and computer models will be performed using one or more 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 a Technical Review Suggestion List for guidance.
This method will include checks of the computer codes used, as well as, evaluations of code input and output.
Verify the calculations with a custom method, and provide detailed information that describes the chosen methodology, i
After the technical review has been completed, the original system analysis and the information provided in the technical review will be approved by the Nuclear Criticality Safety Function Manager.
6.3 TABLE OF PLANT SYSTEMS AND PARAMETRIC CONTROLS Table 6.3 identifies the major areas or systems, and processes or equipment (vessels and containers), within the CFFF. Table entries for each significant item highlight the controlled parameter (s) selected for the Criticality Safety Analysis (CSA), Criticality Docket No.
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4 l
l Safety Evaluation (CSE), or Integrated Safety Assessment (ISA) and for the bounding assumptions relevant to the analysis. An evaluation for unit interaction will be conducted l
and appropriate unit separation distances will be determined for any system containing moderated uranium or for systems where moderating mmr:els are available. Table 4
column definitions are presented below:
I AREA OR SYSTEhi: A defined functional group of processes or pieces of equipment l
that operate as a smgle unit.
PROCESS OR EQUIPAfENT: A defined subgroup of vessels, tanks, process and/or support equipment.
CONTROLLED PARAhiETERS: The parameters listed for a given process / equipment
)
are those upon which double contingency protection is based. (That is, if two parameters are listed, then both are used to provide double contingency protection; if only one parameter is listed, it alone is used to provide double contingency protection).
The following provisions apply for modifications to the Controlled Parameters:
Whe: more than one ccntrolled parameter is listed in the tabl+. the Licensee may mod' the CSA, CSE, or ISA and use fewer of the pararueters for double
- cor,
,ency protection. If the resulting margin of safety is not decreased, a License Amendment is not required. If the resulting margin of safety is decreased, a License Amendment will be obtained prior to implementation.
)
If a new controlled parameter is used in addition to the other controlled parameters listed in the table and the margin of safety is not decreased, a License Amendment is not required. If the margin of safety is decreased, a License Amendment will be obtained prior to implementation.
EVALUATION BOUNDING ASSUh1PTIONS: These are the values used for physical process parameters which are not directly controlled but represent the most reactive i
values that are credible for the system under consideration. As such, the CSA, CSE, or ISA will analyze 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, CSE, or ISA. The approved CSA, CSE, or ISA may limit the operation of the system to levels more conservative than those permitted by the bounding assumptions.
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Table 6.3: Plant Systems and Parametric Controls a
AREA PROCESS CONTROLLED EVALUATION OR OR PARAMETER (S)
BOUNDING SYSTEM EQUIPMENT ASSUMITIONS Storage Pad UF. Cylinders Moderation
= UF. Only a s I wt. % II 0 Equivalent 2
= Full Reflection ADU UF, Cylinder Moderation
= UF. Only Conversion
= s I wt. % 110 Equivalent 2
- Partial Reflection Vaporizer Mass
= liomogeneous UO F 2
= Optimum 110 Moderation Geometry 2
= Partial Reflection (level Control)
Cold Trap System Geometry
= Ilomogeneous UO F, 2
(ADU and IDR)
= Optimum 110 Moderation 2
- Partial Reflection Sump Geometry
= 11omogeneous UO F 2 2 (Level Control)
= Optimum 110 Moderation 2
= Full Concrete Reflection On 5 Sides and Partial Reflection On 1 Side flydrolysis Mass
= llomogeneous UO F:
2
= Optimum 1I 0 Moderation 2
Geometry
= Partial Reflection Nitrate Vessel Mass / Concentration a liomogeneous UO F, 2
- Optimum 110 Moderation 2
Geometry
- Partial Reflection Precipitation Mass / Concentration a liomogeneous UO F:
2
= Optimum II,0 Moderation Geometry
. Partial Reflection Centrifugation Geometry
= liomogeneous UO2
= Optimum II 0 Moderation 2
Centrifuge - Decanter
. Partial Reflection I
Decanter Purge Drain Geometry
= Ilomogeneous UO2 Tank
= Optimum 110 Moderation 2
- Partial Reflection Docket No.
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)
AREA PROCESS CONTROLLED EVALUATION OR OR PARAMETER (S)
BOUNDING SYSTEM EQUIPMENT ASSUMITIONS Drying Geometry
= Homogeneous UO2
= Optimum H O Moderation 2
= Partial Reficction Elevator Geometry
= Ilomogeneous UO2
= Optimum H O Moderation 2
- Partial Reflection Mass Elevator Enclosure
= Homogeneous UO2 Geometry (level
= Optimum H O Moderation 2
Control)
= Partial Reflection Calciner Off-gas Vent Mass
= Homogeneous UO2 Pot Enclosure
= Optimum H O Moderation 2
a Partial Reflection Calciner Tube Geometry
= Homogeneous UO2
= Optimum H O Moderation 2
= Partial Reflection Calciner Combustion Mass
= Homogeneous UO2 Chamber
= Optimum H O Moderation 2
= Partial Reflection Calciner Scrubber Mass (Density)
= Homogeneous UO2 System Slab Tank
= Optimum H O Moderation 2
Geometry
. Partial Reflection Decanter Liquid Mass (Density)
= Homogeneous UO, Discharge Receiver
= Optimum H O Moderation Geometry 2
Tank
= Partial Reflection Calciner Product Hood Mass
= Homogeneous UO2 Moderation
= Optimum H O Moderation 2
= Partial Reflection Fitzmill Enclosure Mass
= Homogeneous UO 2 Moderation
= Optimum 110 Moderation 2
- Partial Reficction Polypack Collection Mass
= Homogeneous UO, Hood
= Optimum H O Moderation Moderation 2
= Partial Reflection Docket No.
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I AREA PROCESS CONTROLLED EVALUATION
{
OR OR PARAMETER (S)
ILOUNDING i
SYSTEM EQUIPMENT ASSUMITIONS Overflow Collection Mass
= Ilomogeneous UO2 llood
= Optimum 110 Moderation Moderation 2
= Partial Reflection Conversion Liquid Concentration a llomogeneous UO2 Efiluent Tanks Neutron Absorber
= s 15 gms. U2" / L ter
= Panial Reflection IDR UF. Cylinder Moderation
= Ilomogeneous UF.
Conversion
= s I wt. % 110 Equivalent 2
= Full Reflection Vaporizer Geometry
= Ilomogeneous UO F 2 2 (Level Control)
- Optimum 110 Moderation 2
- Partial Reflection Sump Geometry
= llomogeneous UO F 2 2 (level Control)
= Optimum 110 Moderation 2
= Full Concrete Reflection On 5 Sides and Panial Reflection On 1 Side DE Chamber / Carbon Moderation
= Ilomogeneous UO2 Filters a s 10 wt. % 110 Equivalent 2
- Partial Reflection Kiln Moderation
= Ilomogeneous UO2 a s 10 wt. % 110 Equivalent
)
2
= Panial Reflection l
4 Check liopper System Geometry
= Ilomogeneous UO2
= Optimum 110 Moderation 1
2
= Panial Reflection Conversion Liquid Geometry
= 11omogeneous UO2 Effluent Tanks
= Optimum 110 Moderation 2
a Panial Reflection Powder Blending Bulk Blender Moderation a liomogeneous UO2
= 5 I wt. % 1I 0 Equivalent 2
= Partial Reflection ADU Bulk Container Moderation
= Homogeneous UO2 Pelleting
= s 10 wt. % 110 Equivalent 2
= Panial Reflection Docket No.
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l AREA PROCESS CONTROLLED EVALUATION OR OR PARAMETER (S)
BOUNDING SYSTEM EQUIPMENT ASSUMITIONS Bulk Enclosure Mass a Homogeneous UO2 Moderation
= Optimum II. :n Moderation a Partial Reflection Powder Transport Geometry
= Homogeneous UO2 Moderation
= Optimum H O Moderation 2
(Powder Lift
. Partial Reflection Enclosure)
Compaction Geometry
= Homogeneous UO 2 (level Control)
= Optimum 110 Moderation 2
Moderation a Partial Reflection Granulation Geometry
= Homogeneous UO2 Moderation
= Optimum H O Moderation 2
= Partial Reflection
(
Pressing Geometry
= Heterogeneous UO
)
2 l
= Optimum H O Moderation 2
a Partial Reflection Sintering Geometry
= Heterogeneous UO2
= Optimum H O Moderation 2
l
= Partial Reflection l
Pellet Grinding Geometry
= Heterogeneous UO 2
= Optimum H O Moderation l
2 a Partial Reflection Rods Loading System Geometry
= Heterogeneous UO2
= Optimum H O Moderation 2
a Partial Reflection j
inspection Geometry
= Heterogeneous UO2
= Optimum H O Moderation 2
= Partial Reflection Final Assembly Fabrication Geometry
= Heterogeneous UO2 Neutron Absorber
= Full Interstitial Moderation
= Full Reflection Inspection Geometry
= Heterogeneous UO2 Neutron Absorber
= Full Interstitial Moderation
= Full Reflection Docket No.
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AREA PROCESS CONTROLLED EVALUATION OR OR PARAMETER (S)
IlOUNDING SYSTEM EQUIPMENT ASSUMITIONS Washing Geometry
= Heterogeneous UO2 Neutron Absorber a Full Interstitial Moderation
- Full Reflection UF Cylinder Washing Cylinder System Mass
- Homogeneous UO F 2 2
- Optimum H O Moderation 2
Moderation
- Partial Reflection Eduction System Geometry a Homogeneous UO F 2 2
- Optimum H O Moderation 2
= Partial Reflection 9
System Storage Geometry
= Homogeneous UO F 2 2 Vessels a Optimum H O Moderation 2
= Partial Reflection Precipitation System Geometry
- Homogeneous UO F 2 2 a Optimum H O Moderation 2
= Partial Reflection Filtration System Geometry
= Homogeneous UO F2 2
- Optimum H O Moderation 2
- Partial Reficction URRS Dissolver Vessels Geometry
= Homogeneous UO2 Dissolver
- Optimum 110 Moderation 2
- Partial Reflection System Filters Geometry
= Homogeneous UO2
- Optimum H O Moderation 2
- Partial Reflection UN Product Tanks Geometry a llomogeneous UN
- s 1000 gms. U / Liter
- Partial Reflection System Support Tanks Geometry a Homogeneous UO2
= Optimum 1I 0 Moderation 2
a Partial Reflection URRS law Level Waste Concentration
- Ilomogeneous UO2 Scrap Processing Processing System Mass a s 15 gms. U2" /1 ter Geometry
- Partial Reflection Incinerator System Mass
- Homogeneous UO2
- Optimum H O Moderation 2
- Partial Reflection Docket No.
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l AREA PROCESS CONTROLLED EVALUATION OR OR PARAMETER (S)
BOUNDING SYSTEM EQUIPMENT ASSUMITIONS Ash Recovery System Geometry
= Homogeneous UO2 Mass
= Optimum H O Moderation 2
a Partial Reflection Liquid Honing System Concentration a Homogeneous UO 2 Mass
- Optimum H O Moderation 2
= Panial Reflection Ultrasonic Cleaning Concentration
= Homogeneous UO2 System Mass
= Optimum H O Moderation 2
= Partial Reflection Shredder System Mass
= Homogeneous UO2
= Optimum H O Moderation 2
a Partial Reflection Mop Water System Geometry
= Homogeneous UO2 Equipment Mass
= Optimum 110 Moderation 2
l
= Partial Reflection Solvent Extraction Dissolver Equipment Mass
= Homogeneous UO2 System
= Optimum H O Moderation 2
= Partial Reflection Adjustment Equipment Mass
= Homogeneous UN Concentration
= Optimum H O Moderation 2
= Panial Reflection l
System Filters Geometry
= Homogeneous UO2 l
= Optimum H O Moderation 2
= Partial Reflection Solvent / Extraction Geometry
= Homogeneous UN Equipment
= s 1000 gms. U / Liter
= Panial Reflection Concentration Geometry
= Homogeneous UN Equipment
- s 1000 gms. U / Liter
= Panial Reflection Fluoride Stripping Concentration
= Homogeneous UO F 2 2 Equipment
= Optimum H O Moderation 2
- Partial Reflection IFBA Mop Water System Geometry
= Homogeneous UO2 Equipment
= Optimum H O Moderation 2
= Panial Reflection Docket No.
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AREA PROCESS CONTROLLED EVALUATION OR OR PARAMETER (S)
BOUNDING SYSTEM EQUIPMENT ASSUMITIONS Pellet Stripping System Geometry
= Heterogeneous UO2 Equipment
= _< 10 wt. % H O Equivalent 2
- Partial Reflection Pellet Coating System Geometry
= Heterogeneous UO2
= Optimum H O Moderation 2
= Partial Reflection Rod Imading System Geometry
= Hetera.rsous UO 2
= Optimum H O Moderation 2
a Panial Reflection Inspection Geometry
= Heterogeneous UO2
= Optimum H O Moderation 2
- Panial Reflection URRS Advanced Wastewater Concentration
= Homogeneous UO2 Waste Treatment Treatment System Mass
= 5 wt. % U
= Partial Reflection f
Storage Tanks Concentration
= Homogeneous UO2 Mass
= 5 wt. % U
= Panial Reflection Ammonia Recovery Concentration
= Homogeneous UO2 System Mass
= 5 wt. % U
= Partial Reflection UN Bulk Storage Tank Concentration
= Homogeneous UN System a s 15 gms. U22$ / liter l
= Panial Reflection Storage Pad Geometry
= Homogeneous UN Concentration
= s 23 gms. U233/ liter
= Full Concrete Reflection On 5 Sides and Partial Reflection On 1 Side Miscellaneous Laboratories Geometry
= Heterogeneous UO2 i
Mass
= Optimum H O Moderation 2
a Partial Reflection i
Ventilation Systems Geometry
= Homogeneous UO2 Mass
= Optimum H O Moderation 2
= Partial Reflection Docket No.
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AREA PROCESS CONTROLLED EVALUATION OR OR PARAMETER (S)
IlOUNDING SYSTEM EQUIPMENT ASSUMITIONS Ventilation Systems Moderation
= Homogeneous UO2 Mass
= s10 wt. % II 0 Equivalent 2
= Partial Reflection Scrubber Systems Geometry
= Homogeneous UO2 Concentration
= Optimum H O Moderation 2
= Partial Reflection Hoods and Geometry
= Hetrogeneous UO2 Containment Mass a s10 wt. % H O Equivalent 2
(licterogeneous
= Partial Reflection Material)
Hoods and Geometry
= Homogeneous UO2 Containment Mass
= Optimum H O Moderation 2
(Homogeneous
= Partial Reflection Material)
Storage Wet Material Geometry
= Homogeneous UO2 Containers
= Optimum H O Moderation 2
(Homogeneous
= Full Reflection Material)
Dry Material Geometry
= Homogeneous UO2 Containers Mo6eration
= s10 wt. % H O Equivalent 2
(Homogeneous
= Partial Reflection Material)
Wet Material Geometry
= Heterogeneous UO2 Containers
= Optimum H O Moderation 2
(Heterogeneous
= Partial Reflection Material)
Dry Material Geometry
= Heterogeneous UO2 Containers Moderation a s10 wt. % H O Equivalent 2
(Heterogeneous
= Partial Reflection Material)
Rod Storage Geometry
= Heterogeneous UO2
= Full Interstitial Moderation
= Full Reflection Fuel Assembly Storage Geometry
= Heterogeneous UO2 Neutron Absorber
= Full Interstitial Moderation
= Full Reflection i
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)
\\
AREA PROCESS CONTROLLED EVALUATION I
OR OR PARAMETER (S)
BOUNDING SYSTEM EQUIPMENT ASSUMITIONS i
Pellet Cabinets Geometry
= Heterogeneous UO 2
Neutron Absorber a Full Interstitial Moderation a Partial Reflection 6.4 PROGRAM ADMINISTRATION 6.4.1 POSTING OF NUCLEAR CRITICALITY SAFETY LIMITS AND CONTROLS Posting refers to the placement of signs or painting of floor areas to summarize key administrative criticality safety limits and controls, to designate approved work and storage areas, and to provide instructions or specific precautions to personnel by supplementing operating procedures.
Appropriate postings will be provided at the entrance to work or storage areas such as hoods, zones, or modules where special nuclear material is handled, processed, or stored. Storage postings will be located in conspicuous places and will include as appropriate; material type, container identification, number of items allowed; and, mass, volume, moderation, and/or spacing limits. Criticality safety precautions or prohibitions related to fire fighting such as use of water, fog nozzles, and high pressure sprays will be posted at the etatrance to the affected area.
Postings will be approved and issued by the Nuclear Criticality Safety Function.
Compliance with the requirements identified on postings will be documented by formal inspections. First level managers will be responsible for assuring that such postings are made available to appropriate personnel.
6.4.2 PROCESS AND PROGRAM REVIEWS Nuclear Criticality Safety process (i.e. Storage Pad, ADU Conversion, etc.) reviews consist of periodic assessments of the conduct of operations within the facility. During the period of time in which the NCS function is developing the ISAs, and upgrading CSEs and CSAs, these activities will satisfy the process review requirement. Once the ISA and CSE/CSA upgrade projects are completed, the process review program will consist of technical reviews of the systems and processes contained in the ISAs and CSEs/CSAs. All portions of the ISAs and CSEs/CSAs will be reviewed at least every five years. Actual frequency of a review will be determined based on such factors as previous findings, Datapacks, Redbook occurrences, NRC inspection activities, independent audits, configuration control activity, and date of last review.
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Nuclear Criticality Safety program (i.e., Control Methodalogy and Principles, Control Documents, etc.) reviews vill be conducted on an annuN basis. The specific portions of the program evaluated during a particular review will oe determined based on previous findings, NRC inspection activities, independent av6ts, current operating conditions, and date oflast review. Each program will be reviewed at least triennially. This frequency provides a mechanism for assessing the effectiveness of all components of the nuclear criticality safety program on a revolving basis and maximizes the utilization of nuclear criticality safety personnel.
6.5 ALARM SYSTEM l
6.5.1 SPECIFICATIONS l
The nuclear criticality alarm system radiation monitoring unit detectors will be located to assure compliance with the requirements of ANSI /ANS-8.3 and 10CFR70.24. The location and spacing of the detectors will be chosen to avoid the effect of shielding by massive equipment or materials. Spacing will be reduced where high density building materials such as brick, concrete, or cinder block shield a potential accident area from the detector. Low density materials of construction such as wooden stud construction walls, asbestos, plaster, or metal-corrugated panels, doors, non-load walls, and steel office partitions will be disregarded in determining the spacing.
Should the nuclear criticality alarm system be out of service for a time period exceeding four hours, all movements of SNM will cease until the alarm service has been restored, or until special monitoring, approved by the Nuclear Criticality Safety Function, has been implemented. Routine testing, calibration and/or maintenance of the system is permitted with no suspension of SNM movements.
6.5.2 OPERATION The nuclear criticality alarm system initiates immediate evacuation of the facility.
Employees will be trained in recognizing the evacuation signal, which is a continuous-sounding siren. This system, and proper response protocol, are detailed in the CFFF Site Emergency Plan and Emergency Procedures. In the event of loss of normal power, emergency power will be automatically supplied to the systetn.
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