ML20198K261
| ML20198K261 | |
| Person / Time | |
|---|---|
| Site: | Westinghouse |
| Issue date: | 09/23/1997 |
| From: | WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML20198K258 | List: |
| References | |
| NUDOCS 9710230245 | |
| Download: ML20198K261 (31) | |
Text
......
l TABLE OF CONTENTS { PRIVATE }
c NUMBER AND TITLE PAGE TAB LE O F CONTENTS......................................................................................
REVI S I O N RECORD.......................................................................................
CHAPTER 1.0 G EN ERAL IN FORM ATION.................................................. 1.0 1.1 FACILITY AND PROCESS DESCRIPTION.............................. 1.0 1.2 IN STITUTION AL IN FORM ATION......................................... 1.4 1.3 SITE DESCRIPTION............................................................ 1.5 1.4 TERM S AN D D EFINITION S................................................. 1. 8
\\
CHAPTER 2.0 MAN AG EM ENT ORG ANIZATION........................................ 2,0 2.1 ORGANIZATIONAL RESPONSIBILITIES AND -
A UTi l ORITI ES................................................................... 2. 0 2.2 SAFETY COMMITI'EES...................................................... 2. 8 CHAPTER 3.0 CON DUCT OF OPERATIONS............................................... 3.0 3.1 CON FIG URATION M AN AG EM ENT...................................... 3.0 3.2 M AI NTEN AN C E................................................................ 3.2 3.3 Q UA liTY ASSURANCE...................................................... 3.4 3.4 PR0CEDURES, TRAINING AND QUALIFICATION.......,......... 3.7 3.5 H UM AN FACTORS............................................................ 3.13 3.6 AUDITS AN D SELF-ASSESSM ENTS..................................... 3.14 3.7 INCIDENT INVESTIG ATIONS............................................. 3.16 3.8 RECORDKEEPING AND REPORTING.................................. 3.19 CHAIYI'ER 4.0 INTEG RATED S AFETY ASSESSM ENT.................................... 4.0 CHAPTER 5.0 RADI ATION S AFETY......................................................... 5.0 5.1 ALARA (As low As Reasonably Achievable) POLICY.................
5.0
- 5.2 RADI ATION WORK PERMITS (RWP).................................... 5.1 5.3 VENTILATION SYSTEMS.................................................... 5.2
' 5.4 AI R S AM PLING................................................................ 5.4 5.5 CONTAMIN ATION CONTROL............................................. 5.5 5.6 EXTERN AL EXPOSURE...................................................... 5. 8 5.7 INTERN A; ?.XPOS URE....................................................... 5. 8 5.8 RESPIRATORY PROTECTION............................................. 5.10 5.9 IN STRUM ENTATION......................................................... 5.12 5.10 SUMMING INTERNAL AND EXTERNAL EXPOSURES........... 5.12 9710230245 971006 PDR ADOCK 07001131 C
PDR t-Docket No.
70 1151 Initial Submittal Date:
_30APR90 Page No.
i
. License No. SNM Il07 Revision Submittal Date: 23SEP97 Revision No. 9.0
K TABLE OF CONTENTS (Cont'd)
NUMBER AND TITLE PAGE C11 APTER 6.0 N UCLEAR CRITICALITY S AFETY........................................ 6.0 6.1 PROG RAM A DMINISTRATION............................................ 6.0 6.2 CONTROL METilODOLOGY AND PRINCIPLES..................... 6.2 6.3 A LA RM S YSTEM.............................................................. 6. 20 6.4 CO NTROL DOC UM ENTS................................................... 6.21 ClIAPTER 7.0 C11 EMICAL S AFETY.......................................................... 7.0 7.1 CilEMICAL SAFETY PROG RAM.......................................... 7.0 7.2 CilEMICAL SAFETY IIAZARD EVALUATIONS...................... 7.0 7.3.
CllEMICAL SAFETY PROGRAM STRUCTURE....................... 7.1 7.4
-ADDITIONAL CllEMICAL SAFETY COMMITMENTS.............. 7.2 CIIAFTER 8.0 FI RE S A FETY.................................................................... 8. 0 8.1 STRUCTURE OF TIIE FIRE SAFETY PROGRAM.....................
8.0 8.2 FIRE S U PPRESSION SERVICES........................................... 8.10 CilAPTER 9.0 EMERGENCY MANAGEMENT PROGRAM............................ 9.0 9.1 EM ERG ENC Y PLAN........................................................... 9.0 9.2
'M ERG ENCY EQ U1PM ENT................................................. 9.0 CilAPTER 10.0 ENVIRON M ENTAL PROTECTION.......................................,10.0 10.1 EFFLUENT AIR TREATM ENT............................................ 10.0 10.2 LIQUID WASTE TREATM ENT FACILITIES........................... 10.0 10.3 SOLID WASTE DISPOSAL FACILITIES................................. 10.1 10.4 PROG RAM DOC UM ENTATION........................................... 10.1 10.5 EVA LU ATI ON S................................................................ 10.2 10.6 O FF-S ITE DO S E................................................................ 10. 2 CIIAPTER 11.0 DECOMMISSIONING.....................................................,,.... 1 1,0 11.1 CONCEPTUAL DECOMMISSIONING PLAN.........................11.0 11.2 DECOMMISSIONING FUNDING PLAN AND FINANCIAL ASS URA N C E.................................................................... 1 1.1 CHAPTER 12.0 AUTilORIZATIONS AND EXEMPTIONS............................... 12.0 12.1 A UTHORIZATIONS........................................................... 12.0 12.2 EX EM PTlON S................................................................... 12.4 Docket No.
70-1151 Initial Submittal Date:
30AP'WO Page No.
il License No. SNM 1107 Revision Submittal Date: 23SEP97 Revision No. 9.0
REVISION RECORD e
REVISION DATE OF PAGES i
NUMBER REVISION REVISED
_ REVISION REASON 1.0 30APR95 All Update tu current operations.
2.0 28JUN%
ill, 6.8 Clarify Criticality Safety Basis for the compaction operation.
3.0 30AUG%
ill,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 ill, 6.11, 6.12 Clarification of Criticality Safety Basis for the Pellet Stripping System Equipment and floods & Containment.
5.0 08NOV%
ill,1.12,3.18, and 3.19 Incorporation of a definition, (Reprinted all document and incident notification pages in Microsoft Word criteria, recently approved format) by NRC Staff.
6.0 05MAY97 6.12 (Reprinted all Clarify Evaluation document pages in Bounding Assumptions Microsoft Word format.)
for Storage of Annular Pellets.
7,0 14JUL97 ill,12.2 and 12.3.
Withdraw ~
an existing authorization, and expand another-author!zation to enable cement manufacturing with CaF.
3 8.0 11AUG97 iii,2.4 and 8.1 (Reprinted Change emergency exeirise all document pages in frequencies for consistency Microsoft Word fornut.)
with Emergency Plan.
j Docket No.
70-1151 Initial Submittal Date:
30APR90 Page No.
iii License No. SNM 1107 Revision Submittal Date: 23SEP97 Revision No. 9.0
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 nonnal operations, expected process upsets and credible process upsets.
4 i
i Docket No.
70-1151 Initial Submittal Date:
30APR90 Page No.
iv-
- License No. - SNM-1107 Revision Submittal Date: _23SEP97 Revision No. 9.0
.o CHAPTER 6.0 NUCLEAR CRITICALITY SAFETY 6.1
-PROGRAM ADMINISTRATION 6.1.1 FACILITY PROCEDURES (a)
Plant Operating ProcMures Procedures, and procedure changes, impacting nuclear criticality safety will be reviewed and approved by the Nuclear Criticality Safety Function. First level managers will be responsible for assuring that such nrocedures are made available -
to appropriate personnel; through posting of limits, talning programs, anxi/or other written, electronic or verbal notifications. Documentation of the process of review, approval, and operator signoff wi!! be maintained electronically within the -
facility procedure control system. Specific details of the system are described in Chapter 3.0, Subsection 3.4.1 of this License Application.
(b)
Regulatory Affairs Guidance Procedures Regulatory Significant procedures define the policies of. the Regulatory Component, including nuclear criticality safety, and identify the requirenients for implementation of applicable _ NRC regulations and license conditions. These guidance procedures will be issued by the Regulatory Component and approved --
by Regulatory Component Managers. Criticality safety issues will be addressed and communicated to first level managers for incorporation into plant operating procedures. Regulatory Component approvals will be required for all regulatory-significant procedures, anxi their changes, involving the handling, processing,-
storage, inspection, and/or movement of special nuclear materials. Specific details of the system are described in Chapter 3.0, Subsection 3.4.1 of this License Applieuton.
(c)
Posting of Limits and Controls
- Posting refers to the placement of signs or painting of floor areas to summarize key criticality safety requirements and limits, to designate approved work and storage areas, or to provide instructions or specific precautions to personnel.
Posting will occur at the entrance to work or storage areas such as hoods, zones, or modules where special nuclear material is handled, processed, or stored.
Docket No.
70-1151 Initial Submittal Date:
,30APR90 Page No.
6.0 License No.. SNM-1107 Revision Submittal Date: 23SEP97
) evision No. 9.0
9
.f 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. Additionally, when administrative controls or specific actions / decisions by operators are involved, postings _will include pertinent requirements identified within the configuration control and/or criticality safety evaluation.
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 entrance 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 docunwnted by formal inspections and audits.
6.1.2 INSPECTIONS & AUDITS inspections and audits will be performed to document that plant operations are conducted in accordance with applicable license conditions, company policies, and written procedures. This program incorporates process, procedure, and program reviews as tools to evaluate the effectiveness of the criticality safety program. All such inspxtions and audits will be conducted and documented in accordance with a written procedure.
Personnel involved in these reviews will be knowledgeable in nuclear criticality safety and the findings, recommendations, and observations will be reviewed by management within the Regulatory Component, After the review by the Regulatory Component has 3
been completed, the findings, recommendations, and observations will be tansmitted to appropriate managers. Specific details of the facility inspection and audit program are
. described in Chapter 3.0,' Subsection 3.6.1 of this License Application.
Process reviews include inspections and audits of the conduct of operations within the facility and will be conducted on an annual frequency. The specific areas of interest for
- the annual review will be selected based on previous findings, trend analyses, and current operating conditions.
Procedure reviews will be conducted on a frequency corresponding to the appropriate procedure category. The steps within the procedure will be compared with the covered operation to demonstrate accuracy knd representativeness. Specific details of the facility procedure review program frequency are described in Chapter 3.0, Subsection 3.4.1, Paragraph (c) of this License Application.
--A Nuclear Criticality Safety program review will be conducted on an annual basis. The specific portions of the program evaluated during a particular review will be detennined Docket No.
70-1151-Initial Submittal Date: ;30APR90 Page No.
6.1 License No.. SNM-1107 --
Revision Submittal Date: 23SEP97 Revision No. 9.0 i -
9
.~
based on previous findings, NRC inspection activities, current operating conditions, and date of last review. All portions of the program will be reviewed at least triennially. This provides a mechanism for t.sressing the effectiveness of all components of the nuclear criticality sa':ty program on a revolving basis and maximizes the utilization of nuclear criticality safety personnel.
Nonroutine inspections and audits may be conducted at the discretion of the Nuclear Criticality Safety Function or may be performed as the result of an operation upset, floor observations, or external investigations.
6.1.3 NUCLEAR CRITICALITY SAFETY PERSONNEL The specific details of the Regulatory Component personnel position accountability and requirements are described in Chapter 2.0, Subsection 2.1.3, Paragraph (c) of this License _ Application. Key features specific to nuclear _ criticality safety are identified below:
(a)
Requirements The minimum qualifications for a Nuclear Criticality Safety Function Engineer or Manager will be a baccalaureate degree, with physical science or engineering emphasis; and, two years of expericNe in the nuclear industry. This will include at least one year of nuclear criticality safety experience. A Nuclear Criticality Safety Function Engineer will have demonstrated proficiency in nuclear criticality safety, and in the prformance of the assigned position function. Specialized training and testing will be used as appropriate to document this experience and proficiency. Specific details are provided in Chapter 3.0, Subsection 3.4.2, Paragraph (b) of this License Application.
(b)
Authority Nuclear Criticality Safety Function Engineers and Managers will be empowered to review and approve facility procedures, perform validated nuclear criticality safety evaluations, set safety limits, and approve temporary operations. All Nuclear Criticality Safety personnel have authority to shutdown potentially unsafe operations. Nuclear Criticality Safety Function Engineers and Managers have authority to allow operations to restart once criticality safety issues have been resolved.
6.2 CONTROL METHODOLOGY AND PRINCIPLES 6.2.1 GENERAL CONTROL PROGRAM PRACTICF3 Docket No.
70-1151 Initial Submittal Date:
30APR90 Page No.
6.2 License No. SNM-1107 Revision Submittal Date: 23SEP97 Revision No. 9.0
/
1
.s The Double Contingency Principle (ANSI /ANS-8.1-1983 (R 1988)) will be the basis for design and operation of processes within the Columbia Fuel Fabrication Facility (CFFF) using special nuclear materials. Where practicable, all process designs will incorporate sufficient factors of safety to require at least two unlikely, independent, rad concurrent changes in process conditions before a criticality accident is possible. In those instances 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 portion of the process, a defense of one or more system parameters will be employed and documented within the Criticality _ Safety Evaluatior.. The defense is comprised of the set of bounding assumptions, criticality safety limits, and criticality safety constraint:: that, as a set, are uniquely sufficient to maintain the minimum subcritical margin against an initiating event.
Criticality Safety Evaluations (CSE) are utilized to identify the specific controls necessary for the safe and effective operation of a process. Nuclear criticality safety controls will be incorporated into the process design criteria documentation. Prior to use in any process.
- these controls will be undergo a functional verification process. A program for routine maintenance and testing will assure continued compliance.
(c)
Verification Program The purpose of the verification program will be to assure that the controls selected and installed match the requirements identified in the design criteria. All equipment will be examined in the "as-bulh" condition to validate the design and to verify the quality of the installation. In addition, a functional test will be
- performed to verify that the controls function as intended.
-(b)'
Maintenance Program The purpose of the maintenance program will be to assure that the controls designated for a specific process are maintained at the_ original level of implementation. This requires a combination of routine maintenance, functional
- testing, and verification of design specifications on a periodic basis.-Specific details of this program are described in Chapter 3.0, Section 3.2 of this License Application.
' Operations personnel will be responsible for the verification of controls through the use of functional tests. Assistance will be provided by Instrument Technician Functions as required. Control calibration and routine maintenance will normally be provided by the Instrument Technician Function. All verification and maintenance activities will be performed per detailed facility procedures and documented through the use of forms Docket No.
70 1151 Initial Submittal Date:
30APR90 Page No.
6.3 License No. SNM-1107 Revision Submittal Date: 23SEP97 Revision No. - 9.0
=
m__,_.,,,._
,o and/or computer systems. Nuclear Critica'ity Safety Function personnel review all control verifications and maintenance activities and utilize a facility computer tracking sy.;m as the mechanism for tracking problems and documenting that corrective actions have been taken.
6.2.2 METHODS OF CRITICALITY SAFETY CONTROL The relative effectiveness or reliability of controls will be considered -during the Criticality Safety Evaluation process. Passive engineered controls will be preferred over all otaer system controls and will be utilized when available and appropriate. Active engineered controls will be the next preferred method of control and admini::trative controls will be the least preferred.
(a)
Passive Engineering These will be controls which require no action or other response to be effective when called upon to_ ensure nuclear criticality safety. Examples of passive engineered controls include safe geometry equipment such as structurally robust -
cylinders.
(b)
Active Engineering These will be controls which require an external signal _ and/or an electronic / mechanical action / operation to occur when called upon 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 user intervention and do not Imve the same level of reliability as engineered controls and will be least preferred. While such controls may be necessary, and hence acceptable, their use will be limited to process systems which do not lend themselves to engineered controls.
Administrative controls include operator actions which are taken in accordance with a written procedure, operator verification of information with the assistance-of computer terminals, actions taken in response to process alarms, etc.
6.2.3 ' TABLE OF PLANT SYSTEMS & PARAMETRIC CONTROLS This table identifies the major area or system equipment, vessels, and containers within the CFFF. Table entries for each significant ic highlight the safety basis selected for the Criticality Safety Evaluation (CSE) or Critics y Safety Analysis (CSA) and for the Docket No.
70-1151 Initial Submittal Date:
30APR90 Page No.
6.4
' License No. SNM-1107 Revision Submittal Date: 23SEP97 Revision No. 9.0
.o bounding assumptions relevant to the analysis. An evaluation for unit interaction will be conducted and appropriate unit separation distances will be determined for any system containing moderated uranium or in cases where moderating materials are available.
Table column definitions are presented below:
AREA OR SYSTEht: A defined functional group of processes or pieces of equipment that operate as a single unit.
PROCESS OR EQUIPhlENT: A defined subgroup of vessels, tanks, process and/or support equipment.
CRITICALITY SAFETY HASIS: At the time of issue of this table, 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). Double contingency protection is specified and described in the CSE/CSA.
The following provisions apply for modifications to the Criticality Safety Basis:
Where more than one parameter is listed in the table, the Licensee may modify the CSE and use only one of the parameters for double contingency protection, if the resulting margin of safety is not decreased, no License amendment is required. If the resulting margin of safety is decreased, the Licensee will seek a License amendment prior to implementation.
If a parameter not specified in the table is used to provide double contingency protection, the Licensee will seek a License amendment prior to implementatien.
EVALUATION BOUNDING ASSUMPTIONS: These are the values used for physical process parameters which are not directly controlled but represent the most reactive values that are credible for the system under consideration. As such, the CSE or CSA will analyze all process operaticns 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 CSE or CSA. The approved CSE or CSA may limit the operation of the system to levels more conservative than those permitted by the bounding assumptions.
Docket No.
70-1151 Initial Submittal Date:
30APR90 Page No.
6.5 License No. SNM-1107 Revision Submittal Date: 23SEP97 Revision No. 9.0
Y AREA PROCESS CRITICALITY EVALUATION OR OR SAFETY HOUNDING SYSTEM IX)UIPMENT BASIS ASSUMPTIONS Storage Pad U13. Cylinders Moderation
- UP. Only
- s I wt. % 110 Equivalent 2
- Full Reflection ADU UP. Cylinder Moderation
- UF. Only Conversion
- s 1 wl. % 110 Equivalent 3
- Partial Reflection Vaporizer Mass a llomogeneous UO,F
- Optimum II,0 Moderation hv[ frol)
= Partial Reflection Cold Trap System Geometry
- Homogeneous UO F:
2 (ADU and IDR)
- Optimum 110 Moderation 3
- Partial Reflection Sump Geometry a llomogeneous UO,F (level Control)
- Optimum 110 Moderation 3
- Full Concrete Reflection On 5 Sides and Partial *tellection On i Side flydrolysis Mass e llomogeneous UO F 2
- Optimum 110 Moderation 2
Geometry
- Partial Reflection Nitrate Vessel Mass / Concentration
= llomogeneous UO,F,
- Optimum 110 Moderation 2
Geometry
- Partial Reflection P.ccipitation Mass / Concentration a llomogeneous UO,F
- Optimum 110 Moderation 3
Geometry
- Partial Reflection Centrifugation Geometry e llomogeneous UO2
= Optimum 110 Moderation 3
Centrifuge. Decanter
. Partial Reflection Decanter Purge Drain Geometry
- lic,mogen.wus UO, Tank
= Optimum q0 Moderation
- Partial Reflection Drying Geometry
= Homogeneous UO2
- Optimum 110 Moderation 3
- Partial Reflection Docket No. 1151 Initial Submittal Date:
30APR90 Page No.
6.6 License No. SNM-1107 Revision Submittal Date: 23SEP97 Revision No. 9.0
O y
6.2.4 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 6.2.1 of this License Application. The Criticality Safety Evaluation process will be used to identify the significant parameters affected within a particular system.
All assumptions relating to process / equipment / material theory, function, and operation, including credible upset conditions, will be justified, docummted, and independently reviewed.
Identified below are the specific control parameters that will be considered during the review process:
(a)
Geometry Safe geometry defines the characteristic dimension of importance for a single unit of a specific geometric shape such that nuclear criticality safety will not be dependent on any other parameter except enrichment (e.g., optimal moderation, reflector thickness, or concentration).
Geometry defines the characteristic dimenslor of importance for a single unit of a specific geometric shape such that criticality safety will be maimained in conjunction with one or more other parameters such as material form, concentration, and/or reflection.
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 process operations or vessels and within storage, transportation, or disposal containers. Mass control may be used on its own or in combination with other control methods. Analytical or non-destmetive methods will be employed to verify the mass measurements for a specific quantity of material.
Whenever mtss control is established for individual rooms or groups of rooms, 8
detailed records will be maintained for mass transfers into and out of these Docket No.
70-1151 Initial Submittal Date:
30APR90 Page No.
6.14 License No. SNM-1107 Revision Submittal Date: 23SEP97 Revision No. 9.0
3
-9 AREA PROCESS CRITICALITY EVALUATION OR OR SAFETY llOUNDING SYSTEM EQUIPMENT BASIS ASSUMI'TIONS Elevator Geometry a llomogeneous UO3
- Optimum II,0 Moderation
- Partial Reflection Elevator Enclosure Mars
- Homogeneous UO,
- Optimum 11,0 Moderation 3
Geometry
- Partial Reft ction (Level)
Calciner Off gas Vent Mass a llomogeneous U03 Pot Enclosure
- Optimum ll O Moderation 2
- Partial Reflection Calciner Tube Ocometry
- llomogeneous UO
- Optimum H O Moderation 2
- Partial Reflection Calciner Combustion Mas:
- Homogeneous UO2 Chamber
- Optircum H 0 Moderation 3
- Partial Reflection Calciner Scrubber Mass (Dcnsity)
- Homogeneous UO, System Slab Tank
- Optimum 110 Moderation 3
Geometry
- Partial Reflection Decanter Liquid Mass (Density)
- Homogeneous UO Discharge Receiver
- Optimum H 0 Meration Geometry 3
Tank
- Partial Reflection Calciner Product Mass
- Homogeneous UO2 Hood
- Optimum 110 Moderation Moderator 2
- Partial Reflection Fitzmill Enclos: ire Mass
- Homogeneous UO3 Moderator
- Optimum H O Moderation 2
- Partial Reflection Polypack Collection Mass
- Homogeneous UO, HW
- Optinmm H O Moderation Moderator 2
- Partial Reflection Docket No.
70-1151 Initial Subntittal Date:
30APR90 Page No.
6.7 License No. SNM-1107 Revision Submittal Date: 23SEP97 Revision No. 9.0
=
AREA PROCESS CRITICALITY EVALUATION OP.
OR SAFETY BOUNDING SYSTEM EQUIPMENT BASIS ASSUMITIONS Overflow Collection Mass a llomogeneous UO2 Hood
- Of timum H O Netation Moderator 2
- Partial Reflection Conversion Liquid Concentration a llomogeneoas UO, Effluent Tanks Neutron Absorber
- s 15 gms. U2" / Liter
- Partial Reflection IDR UP. Cylinder Moderation a llomogeneous UP.
Conversion a s I wt. % 110 Equivalent 3
- Full Reflection Vaporizer Geometry
- Homogeneous UO F 33 (Level Control)
- Optimum H O Moderation 2
- Partial Reflection Sump Ocometry a Homogeneous UO F 2 2 (level Control)
- Optimum H O Moderation 2
- Full Concrete Reflection On 5 Sides and Partial Reflection On 1 Side DE Chamber ! Carbon Moderation a llomogeneous UO2 Filters a s 10 wt. % H O Equivalent 2
- Partial Reflection Kiln Moderation
- Homogeneous UO3
- s 10 wt. % II 0 Equivalent 2
- Partial Reucction Check Hopper System Geometry a llomogeneous UO2
- Optimum 110 Moderation 2
- Partial Reflection Conversion Liquid Geometry a llomogeneous UO3 Effluent Tanks
- Optimum 110 Moderation 2
- Partial Reflection Powder Blending Bulk Blender Moderation a llomogeneous UO2
= s I wt. % 110 Equivalent 2
- Partial Redection ADU Bulk Container Moderation a llomogeneous UO3 Pelleting a s 10 wt. % 110 Equivalent 2
- Partial Reflection Docket No.
70-1151 Initial Submittal Date:
30APR90 Page No.
6.8 License No. SNM-1107 Revision Submittal Date: 23SEP97 Revision No. 9.0
V AREA PROCESS CRITICALITY EVALUATION OR OR SAFE'IY llOUNDING SYSTEM EQUIPMENT llASIS ASSUMITIONS Bulk Enclosure Mass a llomogeneous UO3 Moderator
- Optimum 110 Moderation 3
- Partial Redection Powder Trataport Geometry a llomogeneous UO, Moderation
- Optimum 11,0 Moderation (Powder Lift
. Partial Redection Enclosure)
Compaction Geometry (Level
- llomogeneous UO 3 Control)
- Optimum II,0 Moderation Moderation
= Partial Reucction Granulation Geometry a llomogeneous UO Moderation
- Optimum 11,0 Moderation
- Partial Reucction Pressing Geometry a lieterogeneous UO 3
- Optimum 110 Moderation 3
- Partial Renection Sintering Geometry a tieterogeneous UO2
- Optimum 110 Moderation 2
- Partial Renection Pellet Grinding Geometry a tieterogeneous UO,
- Optimum 110 Moderation 2
- Partial Reflection Rods leading System Geometry a tieterogeneous UO3
- Optimum II,0 Moderation
- Partial Renection inspection Geometry a tieterogeneous UO2
- Optimum 110 Moderation 3
- Partial ReDection Final Assembly Fabrication Geometry
= lieterogeneous UO, Neutron Absorber
- Full Interstitial Moderation
- Full Reflection luspection Geometry a tieterogeneous UO, Neutron Absorber
- Full Interstitial Moderation a Full Reucction Docket No.
70-1151 Initial Submittal Date:
30APR90 Page No.
6.9 License No. SNM-1107 Revision Submittal Date: 23SEP97 Revision No. 9.0 l
4 6
AREA PROCESS CRITICALITY EVALUATION OR OR SAITTY BOUNDING SYFITat EQUIPMENT HASIS ASSUMl"rIONS Washing Geometry a lieteroger.cous UO, 4
Neutron Absorber a Full Interstitial Moderation
- Full Reflection 4
UF, Cylinder Washing Cylinder System Mass
- llomogeneous UO,P
- Optimum 11,0 Moderation M"*'8f
- Partial Reflection Eduction System Geometry a llomogeneous UO F:
i a
2
- Optimum 11,0 Moderation
- Partial Reflection System Storage Geometry a llomogeneous UO,F, Vessels
- Optimum 1I 0 Moderation 2
- Partial Reflection Precipitation System -
Geometry a llomogeneous UO,F
- Optimum 11,0 Moderation j
- Partial Reflection Filtration System Geometry a llomogeneous UO F 2
- Optimum 11,0 Moderation
- Partial Reflection URRS Dissolver Vessel Geometry a llomogeneous UO, Dissolver
- Optimum 110 Moderation 2
- Partial Reflection j
System Filters Geometry
= Ilomogeneous UO
- Optimum II,0 Moderation
- Partial Reflection UN Product Tanks Geometry a llomogeneous UN
- s 1000 gms. U / Liter
- Partial Reflection System Support Tanks Geometry
- llomogeneous 002
- Optimum 110 Moderation 3
- Partial Reflection URRS lew level Waste Concentration a llomogeneous UO Scrap Processing Processing System Mass a s 15 gms. U238 / liter Geometry
- Partial Reficction incinerator System Mass
- llomogeneous UO,
- Optimum 110 Moderation 2
- Partial Reflection 4
Docket No.
70-1151 _
Initial Submittal Date:
30APR90 l' age No.
6.10 License No. SNM-1107 Revision Submittal Date: 23SEP97 Revision No, 9_e
~
A AREA PROCESS CRITICALITY EVALUATION OR OR SAllTY llOUNDING SYFIT.M EQUIPMENT llASIS ASSUMITIONS Ash Recovery System Geometry a llomogeneous UO, Mass
- Optimum 11,0 Moderation -
- Partial Reflection Liquid lioning System Concentration a llomogeneous UO3 Mass
- Optimum 110 Moderation 2
- Partial Reflection Ultrasonic cleaning concentration a llomogeneous UO, System Mass
- Optimum 110 Moderation 2
- Panial Reflection Shredder System Mass e llomogeneous UO,
- Optimum 11,0 Moderation
- Partial Reflection Mop Water System Geometry a llomogeneous UO, Equipment Mass
- Optimum 110 Moderation 3
- Partial Reflection Solvent Extraction Dissolver Equipment Mass e llomogeneous UO, System
- Optimum 11,0 Moderation 4
- Partial Reflection Adjustment Equipment Mass a llomogenmus UN Concentration
- Optimum 110 Moderation
- Partial Reflection p
System Filters Geometry a llomogeneous UO2
- Optimum 110 Moderation 3
- Partial Reflection Solvent / Extraction Geometry a llomogeneous UN Equipment
- s 1000 gms. U / Liter
- Partial Reflection Concentration Geometry a llomogeneous UN Equipment a s 1000 gms. U / Liter
- Partial Reflection Fluoride Stripping Concentration a llomogeneous UO F:
2 Equipment
- Optimum 110 Moderation 2
- Partial Reflection IFBA Mop Water System Geometry
- llomogeneous UO, Equipment
- Optimum 110 Moderation 3
- Partial Reflection Docket No.
70-1151 Initial Submittal Date:
30APR90 Page No.
6.11 Lice.nse No. SNM 1107 Revision Submittal Date: 23SEP97 Revisio, No. 9.0
4 AREA PROCLSS CRITICALITY EVALUATION OR OR SAIPETY BOUNDING SYS'IDt EQUIPMENT llASIS ASSUMITIONS Pellet Stripping Geometry e lieterogeneous UO, System Equipment
- < 10 wt. % II,0 Equivalent
- Partial Renection Pellet Coating System Ocometry a tieterogeneous UO,
- Optimum 11,0 Modera:lon
- Partial Reflection Rod leading System Geometry a lieterogeneous UO,
- Optimum 11,0 Moderation
- Partial Reucction Inspection Geometry
- lieterogeneous UO,
- Optimum 11,0 Moderation
- Partial Renection URRS Advanced Wastewater Concentration a llomogeneous 00, Waste Treatment Treatment System Mass e 5 wt. % U
- Partial Renection Storage Tanks Concentratien a llomogeneous UO, Mass
- 5 wt. % U
- Partial Reflection Ammonia Recovery Concentration a llomogeneous UO, System Mass
- 5 wt. % U
- Partial Renection UN 11ulk Storage Tank Concentration a llomogeneous UN System a s 15 gms. U2" / liter
- Partial Reucction Storage Pad Geometry a llomogeneous UN Concentration
- 5 23 gms. U2"/ liter e Full Concrete Renection On 5 Sides and Partial Reucction On 1 Side Miscellaneous Laboratories Geometry a lieterogeneous UO, Mass
- Optimum 11,0 Moderation
- Partial Reflection Ventilation Systems Gecmetry e llomogeneous UO, Mas:
- Optimum 11,0 Moderation
- Partial Renection Docket No.
70-1151 Initial Submittal Date:
30APR90 Page No.
6.12 License No. SNM-1107 Revision Submittal Date: 23SEP97 Revision No. 9.0 I
s AREA PROCl3S CRITICALITY EVAL,UATION OR OR SAITlY llOUNDING SYSTEM EQUIPMENI' IlASIS ASSUMPTIONS Ventilation Systems Moderation a llomogeneous UO, Mass
- 510 wt. % 11,0 Equivalent
- Partial Reflection Scrubber Systems Geometry.
- llomogeneous UO, Concentration
- Optimum II,0 Mo(eration
- Partial Reflection lloods & Containment Geometry
- 11etrogeneous UO, lieterogeneous Mat'I.
Mass
- $10 wt. % 11,0 Equivalent e Partial Reflection lloods & Containment Geometry a llomogeneous UO, llomogeneous Mat'I.
Mass
- Optimum 110 Moderation 2
- Partial Reflection -
Storage Wet Material Geometry
- llomogeneous UO, Containers
- Optimum 11,0 Moderation
- Full Reflection Dry Material Geometry
- llomogeneous UO, Containers Moderation
- s10 wt % II,0 Equivalent
- Partial Reflection Wet Material Geometry
- lieterogeneous UO, Contdners
- Optimum 11,0 Moderation
- Partial Reflection Dry Material Geometry a tieterogeneous UO, Containers Moderation
- s10 wt. % 11,0 Equivalent
- Partial Reflection Rod Storage Geometry
- 11eterogeneous UO,
- FullInterstitial Moderation
- Full Reficction Fuel Assembly Geometry a lieterogeneous UO, Storage Neutron Absorber
- Full Interstitial Moderation
- Full',ellection Pellet Cabinets Geometry
- lieterogeneous UO, Neutron Absorber a Full Interstitial Moderation
- Partial Reflection Docket No.
70-1151 Initial Submittal Date:
30APR90 Page No.
6.13 License No. SNM-1107 Revision Submittal Date: 23SEP97 Revision No 9.0
rooms. Establishment of nuss limits will involve consideration of potential moderation, reflection, geometry, spacing, and material concentration. The evaluation will consider normal operations and expected process upsets-for determliation of the actual mass limit for the system and for the definition of subsequent controls. When only administrative controls are used for mass contielled systems, double batching is assumed to be the worst credible upset condition.
(c)
Moderation Moderation control may be used for nuclear criticality safety control of systems within the facility, and if utilized, 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.
Two independent controls / measurements or the analysis of two independent samples will be utilized to document this compliance. 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 established to prevent moderator from entering the system after initial loading has occurred. The minimum protection will be that two independent barriers preventing moderator from entering the system must fall before the system-can be compromised. Examples include a combhution of system containment / roof, system containment / pipes, double roofs, etc. A method for detection of failure of the outermost barrier will be established and a program to maintain the quality of the barrier will be in place and routinely
-inspected. All barriers will be tested for leakage during initial 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 analysis will be supported by parametric studies. Transportation of materials outside of moderation control areas will be proceduralized. Maintenance and inspection programs will be defined and implemented through plant procedures.
The quality and basis for selection of the barriers will be documented within the
- CSE process. Controls for the introduction and usage of moderating materials within areas that are under moderation control will be defined and approved by.
the Nuclear Criticality Safety Function.
(d)
Interaction Docket No.
70-1151 Initial Submittal Date:
30APR90 Page No.
6.15 License No SNM 1107 Revision Submittal Date: 23SEP97-Revision No. 9.0
-e Nuclear criticality safety evaluations will consider the potential effects of interaction. Methods utilized are:
(d.1) Non Interaction Methoa 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 approved distance from other units such that the multiplication of the subject unit is not increased.
Additionally, units may be considered non interacting when they are separated by a 12 fet air distance or by 12 in:hes of full density water equivalent. 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 considered non-interacting (provided the total of all such solid angles neglected is less than one half of the total solid angle for the system).
Transfer pipes no greater than a 2-inch diameter and ventilation lines may be excluded from interaction considerations provided that an evaluation is performed to document the basis for treatment and application as a-non-interacting unit.
e (d.2) Solid Angle Method 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 reficction 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 deteimined by the point to-plane method. The multiplication factor of a single unit will be determined from validated computer calculations, (d.3) Monte Carlo Method Individual unit multiplication and erray 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 Docket No. L70-1151 Initial Submittal Date:
30APR90 Page No.
6.16 License No. SNM-1107 Revision Submittal Date: 23SEP97 Revision No. 9.0
95/95 confidence level, will not exceed 0.95.
-(e)
Enrichment Enrichment control may be utilized to limit the percent U2" within a process, vessel, or container, thus providing a method for nuclear criticality safety control. Active engineered and/or administrative 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.
(f)
Concentration & Density Concentration control may be used for nuclear criticality safety control of systems within the facility, and if utilized, will incorporate the following criteria:
(f.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 this compliance. The system for collecting, preparing, analyzing, and posting of results pertaining to sample evaluation will be designed to ensure the results obtained are independent.
(f.2)
Controls will be established to prevent concentration within the system after
'nitial loading has occurred.- - These events _' might: _ include _ 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.
(f.3) -For those systems which utilize concentration as a controlled parameter for-
- ensuring nuclear criticality safety, the CSE's will demonstrate the solubility.
limits of the SNM composition, identify possible precipitating agents to the 4
operators through procedures and ensure appropriate precautions are being taken to ensure such agents are not introduced, and provide a positive means of preventing unwanted transfers if a possibility exists for precipitating agents to be transferred via connected processes.--
.(g)
. Reficction Docket No.
70-115L Initial Submittal Date:
30APR90 Page No.
6.17
- License No. SNM 1107 Revision Submittal Date: 23SEP97 Revision No. 9.0
Many systems are designed and operated with the assumption of full reflection, liowever, certain system designs will be analyzed, approved, and operated in situations where reflection is less than full. Two such methods (partial and bare reflection) may be employed. Partial reflection is defined as 1 inch 110 3
equivalent in systems where equipment locr. tion or design limits the placement of moderating materials, including humans, near the specific system. Bare reflection (no moderator considered) will be utilized in systems where equipment location or design impose physical _ limits on the ability to place moderating materials, including humans, near the specific system.
For all system evaluations, the neutron reflection properties of the credible process environment will be analyzed. This may limit the available use of partial or bare neutron reficction methodology for specific systems.
(h)
Neutron Absorber Neutron absorbing materials may be utilized to provide a method for nuclear criticality safety control for a process, vessel or container. If used, these will take the form of solid materials such as borosilicate-glass Raschig rings, gadolinium plates, borated stainless steel, or other solid neutron absorbing materials. The use of neutron absorbers in this manner will be defined as a passive engineered control.
When Raschig rings are used to control nuclear criticality safety, their use and maintenance will be in-accordance with ANSI /ANS-8,51986, with the following exceptions for basic solutions:
System Pli maintained s 11; and System temperature maintained s 60 degrees centigrade.
For other fixed neutron absorbers, the following requirements apply:
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 an annual basis. The method of verification may take the form of traceability (i.e. serial number,etc.), visual inspection or direct measurement.
Docket No. 1 0-1151 Initial Submittal Date:
30APR90 Page No.
6.18 License No. SNM 1107 Revision Submittal Date: 23SEP97 Revision No. 9.0
(i)
Composition The CSE for each system will determine the effects of material composition within the process beinf, analyzed and identify the basis for composition
= selection pertinent to subsequent system moucling activities.
0)
Heterogeneity The effects of heterogeneity within a system can be significant and will be considered within the CSE when appropriate. (It is therefore important to distinguish between homogeneous and heterogenous systems, especially at lower enrichments, where lumping of the uranium can have an adverse impact on the nuclear criticality saf:ty of a system.)
The primary concern is the determination of the particle size that will be permitted in a uniform system such that the system will still be considered homogeneous from the viewpoint of nuclear criticality safety. A review of nuclear criticality safety calculations has 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 purely heterogeneous. The evaluation of systems where the particle size falls within the range of 150 - 1000 microns will take into consideration the effects of heterogeneity appropriate for the process being analyzed.
(k)
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:
(k.1) The bounding assumptions will be defined through the CSE pro:ess and
. operational limits will be identified within the Criticality Safety Analysis (CSA);
and, will be specifically communicated, through training and procedures, to appropriate manufacturing personnel.
(k.2) Utilization of such process and/or material characterist cs will be based on i
established physical or chemical reactions, known scientific principles, and/or facility specific experimental data supported by operational history.
Examples of such an application include:
Docket No. 1151 Initial Submittal Date:
30APR90 Page No.
6.19 License No. SNM-1107 Revision Submittal Date: 23SFP97 Revision No. 9.0
Conversion and oxidation processes that produce dry powder (s 10 wt.
N i
% H O) as a product of high t:mperature reactions.
2 Experimental data demonstrating low moisture pickup (s 3 wt. % H 0) from room air i ventilation equipment.
D Experimental /nistorical process data demonstrating uranium oxide T
powder flow characteristics to be directly proportional to the quantity of moisture present.
6.2.5 MOVABLE NON-FAVORABLE GEOMETRY (NFG) CONTAINERS Movable NFG container us;:ge within the CFFF will be rigorously controlied and utilized only when other practical methods are unavailable. Prior to use of a movable NFG container in the Chunical Manuiacturing Area, a comprehensive analysis will be performe<l. The key components of this analysis follow:
(a)
Manufacturing personnel will provide justification for the use of the movable 3
NFG c entainer.
(b)
Controls will be identified and implemented to provide assurance that proposed movements of an NFG container can be performed safely.
(c)
A Criticality Safety Analysis will be completed and approved prior to use of the movable NFG container within Se facility.
6.3 ALARM SYSTEM 4
6.3.1 SPECIFICATIONS The nuclear criticality alarm system radiation modtoring units detectors will be located to assure compliance with the requirements of ANSI /ANS-8.3-1986. 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 orick, concrete, or cinde' 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 spem be ou s f service for a time period m
Docket No.
70-1151 Initial Submittal Date:
30APR90 Page No.
6.20 License No. SNM-1107 Revision Submittal Date: 23SEP97 Revision No. 9.0
,V
.1
._ exceeding four hours, all movements of SNM will cease until the alarm service has been restored-or special monitoring, approved by the Nuclear Criticality Safety Function, will be implemented. Routine testing, call'aration and/or maintenance of the systerriwill be permitted with no suspen:,lon of SNM movements.
6.3.2 OPERATION -
The nuclear criticality alarm system initiates immediate evacuation of the facility.
Employees will be trained in recognizing the evacuation signal, which is a continucus-sounding siren. This system, and proper response protocol, is detailed in the CFI<F Site Emergency Plan.
6.3.3 MAINTENANCE The nuclear criticality alarm system will be a safety-related system t id will be maintained through routine calibration and sched-led functional tests conducted in the manner described in Subsection 6.2.1 of this License Application. In the event of loss
'of normal power, emergency power will be automatically connected to the system, 6.4 :. CONTROL DOCUMENTS 6.4.1 CRITICALITY SAFETY EVALUATION (CSE)
. The. Criticality Safety -Evaluation is essentially a subset-of the Integrated Safety Assessment (ISA) defined in Chapter 4.0 of this License Application. The CSE identifies and documents the basis of nuclear criticality safety for a particular system.
A CSE is prepared or updated for each new or significantly modified system within the CFFF. The level of detail for a particular CSE 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. The same process will be used to identify the CSE requirements for an existing system modification. Therefore, the scope and' content of any particular CSE reflects the needs and characteristics of the system being analyzed and includes appropriate infortnation from the following:
(a)
. Process Descripon This section presents a precise narrative definition of normal operation as it relates to each defined system. 'It also provides a schematic representation rd
. the system and a narrative outline of the system transfer interconnections, with
- text references that detail normal operating boundaries (e.g., composition, concentrations,' flows, and sampling). References are previded to all relevant
? Docket No. 70-1151 Initial Submittal Date:
30APR90 Page No.
6.21 License No. SNM-1107 Revision Submittal Date: 23SEP97 Revision No. 9.0
. + _
L.
y I,e' drawings and procedures; and, photographs, diagrams, tables, and charts
> depicting crucial system and subsystem equipment.
(b)
Proc ss Th-ory This tection presents a narrative description of the normal process operating parar eters,-including the' ranges of conditions expected, for each defined system.
It also provides descriptions. of upset conditions which have the potential for exceeding established safety limits.
References are provided documenting the sources of the process theory.
(c)
Process Design and Equipment This section presents the dimensions, construction materials, and design configuration of process equipment, vessels, and transfer lines of each defined system. It also provides a precise narrative &finition of subsystem equipment controls and features, as related to the defined system, and a tabulation of.
relevant reference drawings.
(d)
Drawings and Operating Procedures
~ This section presents a complete reference listing,f all ' documents used in performing the formal Process Hazarda Analysis, and describes their-relationship to the evalaation prccess, for each uefined system. It also provides phatographs of the system / subsystem equipment : hat had relevance to (and were used during) thi analysis process. (All oth.i de.uments collected for review and/or information purposes will be retained as part of the Data Pack for the system's Criticality Safety Evaluation.);
(e).
Safety Analysis This section constitutes the results of a comprehensive nuclear criticality safety review for each component within the defined system. The evaluation reviews the criticality safety ' control reliability through documentation of the. bases identified for each control. Reliability will be systematically assessed through identification of the relative strengths and weaknesses of each con:rol including the consideration for potential common mode failures. Included in the evaluation.will be a summary of facility practices to ensure the quality of safety signifi: ant measurements for tha system, as well as, summary documentation of the controls within each component of the system.
Docket No.
70-1151 Initial Submittal Date:
30APR90 Page No.
6.22 License No. SNM-1107 Revision Submittal Date: 23SEP97 Revision No. 9.0
,i v-w'_
The Safety Evaluation incorporates portions of the Criticality Safety Analyses (CSA) which will be included within the CSS Appendix. The CSA defines the safety _ limits for each component within the system.
~
(f);
Process Hazards Analysis This section presents each defined system's Hazards and Operability Analysis
- (HAZOP) Table or What If/ Checklist Results, Fault Trees, and Event Trees.
(This will constitute an integrated analysis that includes Nuclear Criticality Safety.) Essential elements of the HAZOP Table or What if/ Checklists Results will include a listing of each process / system upset and deviation disclosed in the analysis, the significant causes and consequences of each such process / system upset or deviation, and the controls in place to prevent each cause and/or
- mitigate each consequence _ The complete Analyses Report, providing the detailed results of the analysis in a narrative format, will be retained as part of the documentation for each system's Criticality Safety Evaluation.
'(g)
License Compliance The criticality safety basis and the bounding assumptions generated during the CSE process will be reviewed for compliance with Subsection 6.2.3 of this License Application.
(h)
Appendix 3
This section presents a summary of ancillary information (such as calculations, parametric sensitivity studies, references, etc.) for each defined system.
6.4.2 CRITICALITY SAFETY ANALYSIS (CSA)
The CSA is comprised of a review of criticality safety controls to identify the minimum requirements necessary to ensure nuclear criticality safety. Expert determination is t
utilized to evaluate system reliability and to document the adequacy and effectiveness of each control.
6.4.3' ANALYSES METHODOLOGIES (a)
K,J imit Validated computer analytical methods will be used to calculate the k,y for systems with.' individual vessels or potential-vessel -interaction. conditions include conditions expected to - be encoiatered during routine operations, Docket No.
70-1151 Initial Submittal Date:
39APR90 Page No.
6.23 4
License No. SNM-1107 Revision Submittal Date: 23SEP97 Revision Nc. 9.0
fdi y
.-g.
process upsets, and credible accident situations. TBased on the results of these calculations, the sensitivity of key parameters with respect to' the effect on K,,-
will be evaluated for each system such that adequate system controls are defined for the analyzed f k, will be evaluated to ensure that controls defined for the analyzed system demonstrate a sufficient margin of safety.
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,, s 0.95, including applicable biases and calculated uncertainties.
- With respect to credible process upsets that could lead to a single contingency, there will be ' sufficient _ margin of safety to ensure that, based-on these parameters, the calculated k,,<1.00, including all applicable biases and calculated uncertainties.
(b)
- Analytical Analysis
- Criticality calculations are performed using the SCALE system of modules.
including such codes as NITAWL-II and XSDRNPMS for cross-section generation-and KENOMCNP-Va for reactivity calcu!ations. 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 berchmarked to various critical experiments to verify their applicability prior to use.-
-(c)
Validation Techniques Nuclear criticality _ safety. analyses conducted utilizing computerized -
methodology will be validated in accordance.with the criteria. described in Section 4.3 of ANSI /ANS-8.1-1983.
--(d) _
' Computer Software & 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.
- Docket No. ' 70-1151 Initial Submittal Date:
_30APR90 Page No.
6.24 License No. SNM-1107 Revision Submittal Date: 23SEP97 Revision No. 9.0
\\.
g_
v y
Software designated for use in. nuclear-criticality _ safety calculations on the
. calculational platform will be compiled into 1 working code versions with executable files that ere 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 with the differences in. experiment and analysis being used to calculate bias - and uncertainty va'.ues to be applied to the obtained results.
Modifications to hardivare or software that is essential to the calculational-process will: be ' followed by code. operability verification, in which case, selected calculations will be performed to verify identical results from previous analyses. Deviations noted in code verification _that might alter the bias or uncertainty will require re-qualification of the code prior to continued use.
'(e)-
Technical Review 2
Independent technical reviews of criticality safety assessments, criticality safety evaluations, or calculations in support of limits specified in CSA's or CSE's will be performed. The Nuclear Criticality Safety Function Manager will aHign a qualified reviewer the task of performing the independent technical review.
The technical reviewer will verify that the proposed _' calculational geometry model and configuration adequate'y 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 of the follovcing 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 specific checks of the computer codes used, as well as, evaluations of code input and output.
Verify the calculations with 'a custom method. Provide detailed
- ~
information that describes the chosen methodology.
After the technical review has been completed, the original system analysis and
_ Docket No. 1151 -
Initial Submitut! Date:
30APR90 Page No.
6.25 License No. SNM-1107 Revision Submittal Date: 23SEP97 Revision No. 9.0 3
p
<8 e
- the information provided in the technical review will be approved by the Nuclear Criticality Safety Function Manager.
(f)
Solid Angle Method Solid angle analysis methods may be utilized within the CFFF as described in Subsection 6.2.4, Paragraph (d.2) of this License Application.
1 Docket No.
70-1151 Initial Submittal Date:
30APR90 Page No.
6.26 License No. SNM-1107 Revision Submittal Date: 23SEP97 Revision No. 9.0
,