ML20148G138

Forwards Info in Response to Ltr from Ma Lamastra to Rj Reja.Response Includes Info Discussed at Mgt Meeting on 970520,including Subsequent Telephone Discussion on 970522 Re Concentration Control

ML20148G138
Person / Time
Site:	07001113
Issue date:	05/27/1997
From:	Reda R GENERAL ELECTRIC CO.
To:	Weber M NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS)
References
TAC-L10079, NUDOCS 9706050195
Download: ML20148G138 (115)

Text

{{#Wiki_filter:O GE Nuclear Energt Geners! Elet mc Compey fN l'O Eb 7lD t%immgun V2lH02 \\ / 'J 910 675 5030 l 4 May 27,1997 Mr. M. F. Weber, Licensing Branch, NMSS U.S. Nuclear Regulatory Commission Mail Stop T 8-D-14 ) Washington, DC 20555-0001

• bject:

License Renewal - Response to Request for Additional Information (TAC No. L10079) f

Reference:

(1) NRC License SNM-1097, Docket 70-1113 (2) License Renewal Application,4/5/96 (3) Submittal, RJ Reda to ED Flack,5/6/96 (4) Submittal, RJ Reda to RC Pierson,5/14/96 (5) Letter, RC Pierson to RJ Reda,7/18/96 (6) Submittal, RJ Reda to RC Pierson,8/30/96 p (7) Submittal, PJ Reda to ED Flack,9/26/96 i (8) Letter, MA Lamastra to RJ Reda,10/2/96 v' (9) Submittal, RJ Reda to MA Lamastra,11/22/96 (10) Application, RJ Reda to MF Weber,12/16/96 ) (11) Letter, MA Lamastra to RJ Reda,12/17/96 (12) Submittal, RJ Reda to MF Weber,2/5/97 (13) Letter, MA Lamastra to RJ Reda,2/10/97 (14) Submittal, RJ Reda to MF Weber,2/19/97 (15) Submittal, RJ Reda to MF Weber,2/25/97 (16) Letter, MA Lamastra to RJ Reda,3/5/97 (17) Submittal, RJ Reda to MF Weber,3/27/97 (18) Submittal, RJ Reda to MF Weber,3/28/97 (19) Letter, MA Lamastra to RJ Reda,5/6/97 (20) Letter, MA Lamastra to RJ Reda,5/14/97 (21) Letter, IU Reda, to MA Lamastra,5/21/97

Dear Mr. Weber:

IJfD'f GE's Nuclear Energy Production (NEP) facility in Wilmington, N.C., hereby transmits the enclosed information in response to the above referenced letter dated 5/6/97. The response includes information discussed at the management meeting on 5/20/97 at NRC Headquarters, ) including a subsequent telephone discussion on 5/22/97 regarding concentration control. This C, information 'is being provided in support of our license renewal request. j PoR'18l5A7J8!??ta. ll!ElEllll,Ol!!lllU,llllll ~S g PDR I

Mr. M. F. Weber - May 27,1997 Page 2 () { l contains (1) a description of the changes made to the license renewal by page and - section, and (2) the page changes to our license renewal application for pages contained in the Table of Contents, Chapter 1.0, Chapter 2.0, Chapter 3.0, Chapter 4.0, Chapter 6.0 and Chapter

7. Each chapter is provided in its entirety for easy replacement. Each page within the chapter that contains a change is indicated with a horizontal line ( I) in the right hand column to show where a change has taken place. All replacement pages contain the date of this submittal (5/27/97) and are shown as revision zero.

i Six copies of this submittal are being provided for your use. i Please contact Charlie Vaughan on (910) 675-5656 or me on (910) 675-5889, if you have any I questions or would like to discuss this matter further. i i Sincerely, GE NUCLEAR ENERGY I s i Ralph J. Reda, Manager Fuels & Facility Licensing /zb ' Attachments cc: RJR-97-065 i L. A. Reyes, Region II Administrator G. L. Troup, NRC-Atlanta M. Fry, State of NC l i L0: o

f Mr. M. F. Weber j May 27,1997 Page1of1-l ATTACHMENT 1 l l ) l I i l l l l Response to Request for Additional Information Contained in l Letter from MA Lamastra to RJ Reda Dated May 6,1997 t I l O 1 I l l I l i 4

O

Mr. M. F. Weber May 27,1997 Page1of13 O. Response to NRC Request for Specific Comments and AdditionalInformation Required for GE-NEP's License Renewal Application Please provide thefollowing information: 1. In an NRC letter dated December 17,1996, the Fuel Cycle Licensing Branch (FCLB) commented that the quahfications ofmostpositions have decreased compared to the existing license. GE was requested to demonstrate or explain why such a decrease in the overall experience ofthe staffwould not adversely afect the safety ofplant operations. GE's response dated February 5,1997, stated in part that GE management was re.sponsible and accountablefor the safe operation ofthe plant, GE had in place a management systemfor identifyingjob fimction andselecting quahfied individuals, and that the minimum requirements are generally consistent with other likefacilities. Accordingly, GE made no changes in Section 2.2.1 ofthe application. FCLB agrees that GE is ultimately responsiblefor the safe operation ofthe plant. However,10 CFR 70.23, "Requirementsfor the Approval ofApplications" states, in part, that an applicationfor a license will be approved, ifthe Commission determines that the applicant is qualified by reason oftraining and experience. GE'sproposed minimum quahfication is basically a B.S. degree and two years experience or a high school diploma with 5 years experiencefor both stafand supervisorpositions. FCLB has also reviewed the current minimum levels of training and experience at otherfuelfabricationfacilities and determined that GE's minimum requirements would be the least and sigmficantly less than those ofthe averagefacility. Accordingly, we see no basisfor the reduction in quahficationsfor staffand supervisorypersonal and request thatfor each safety-sigmficant position (radiation protection, criticality safen', chemical safety, fire protection, environmentalsafety etc.) that the minimum quahfications be n pgraded to at least the requirements ofthe current license or a position by positionjustification. In accord with the RAI, GE has modified Sections 2.2.1.2,2.2.1.3,2.2.1.4, 2.2.1.8 and 2.2.1.9 to be consistent with the comments and discussions regarding minimum qualifications. O

Mr. M. F. Weber i May 27,1997 l Page 2 of13 2. The RAI dated March 5,1997, questioned the definition of" practices" as used in Chapter 3.0 ofthe license application. As describedin the license application, these practices should be maintained, controlled and/or approved in the same manner asprocedures. i GEprovided an acceptable response. However, in order to convey the i information provided in the response ofMarch 27,1997, GE should add a statement to the license application that approvedpolicies, practices and \\ procedures will befollowed. An acceptable statement wouldinclude wording similar to thefollowing: Licensed materialprocessing or activities will be conducted in accordance with properly issued and approvedpractices andprocedures (P&P), plant practices, or operatingprocedures. GE has modified Section 3.9 to include the requested clarification in wording. Q 3. Section 4 ofyour application, should be modified to include a schedulefor completing the ISA for the balance ofplant and a schedulefor submitting a revisedISA summaryfor the DCP. The schedule should include milestonesfor a final completion date andintermediate datesfor those systems most importantfor safety. In addition, a clear commitment to complete theproposedISA summary for each system should be provided. Further, and most importantly, GE should commit to maintaining available and reliable systems equipment and controls that 1 are most important to safety based on the ISA results Enclosure 2 and Comment 4 identifies the t) pes ofinformation that a summary should include. At the management meeting at NRC Headquarters on May 20,1997, RAI's 3 and 4 were discussed in relation to the outline for preparing ISA summaries for the NRC and the content of those summaries. GE also presented a schematic overview of the role of the ISA in the proposed safety program as well as an overview of the type and flow ofinformation generated by the ISA process. As a result of these discussions it was mutually agreed that GE and the NRC both have work to do to fine tune the outline with regard to what constitutes an acceptable ISA summary. This work will start after the license is renewed with working sessions to review the records generated by the ISA process, information needs by the NRC and a critique of the current ISA summary. O-In rdation to the schedule to complete ISAs for the balance of the plant it j was mutually agreed that this summary definition work had to be completed

i Mr. M. F. Weber May 27,1997 l Attachment I j Page 3 of13 on a schedule that will support the schedule GE is committing to for that l work. I i i GE's schedule is keyed to the issuance of a renewed license, clarification of i l the summary detail required and systematic and timely feedback of the NRC's critique ofISA summaries as they are submitted to the NRC. I 4 i GE prefers that the formal commitments for this work be called out in a } letter or a license condition as opposed to modifying Chapter 4 - the reason j ] being that this is a onetime effort and it appears th at it would be better j identified that way. j Based on the above detail the schedule that GE proposes is as discussed on May 20,1997, and is as follows-NRC Issue New Facility License June 1,1997 i l GE Finalizes Master ISA Implementation Plan August 1997 I e l I Resoiution oriSA Summarr issues Juir 1998' O Complete Fabrication and GAD Shop December 1998 l e t { i Complete Uranium Recovery Operations December 1999 e I Complete Balance of Nuclear Operations December 2000 l e 2 Complete NRC Review and Reconciliation July 2001 3 Begin Revalidation of DCP July 2001 'The ISA summary issues must be resolved on a schedule th' t identifies a acceptable performance by GE with sufficient lead time to prepare the required summaries. 2The schedule assumes that the NRC reviews the ISA summaries as they j are submitted and keeps them moving on a fairly levelloaded schedule rather than letting them all bunch up at the end. The commitment in the currently proposed license is to revalidate ISAs 3 at a minimum of once every five years.' ) O GE requests that the schedule commitment be such that small shifts in the schedule are accommodated since the concept is that the work is reasonably welllevelloaded over the period and that it is all finished by July of 2001. i

Mr. M. F. Weber May 27,1997 Attachment I n Page 4 of13 V 4. CRITICALITYCONTROLSAND THEISA in the renewal application, when committing to perform an Integrated Safety Analysis (ISA) and to provide a summary description ofit, GE shouldprovide a commitment to provide thefollowing information in the summary: For each accident scenario (identifled in the ISA) that, withoutpreventative controls or mitigation could result in a nuclear criticality, information should te provided identifying the criticality controls established to prevent that scenario and evaluating their adequacy. Specifically, for each scenario state: 1. The controlsformally established to prevent it; 2. The set ofcontrolsfor the process,for the scenario identified, meet established acceptance criteriafor adequacy (There may be a single generic statement, e.g., "Unless otherwise indicated all controls have been determined to meet acceptance criterion xx ofPractices and Procedures document PP->y, i.e., double contingency. "); and () 3. The measures, such as configuration control, maintenance, and training needed to assure the reliability ofthese controls. The renewal should also contain a commitment to establish and maintain the controls identifled in the ISA and a provide the measures to assure their reliability and conformance to the acceptance criteria. 7here should also be a commitmentfor maintaining the 1SA current as changes to theprocesses are made. Based on our discussions and agreements reached at the Management Meeting in Washington on May 20,1997, the content of the first part of RAI

1. 4 dealing with certain aspects of the ISA is to be considered as a part of resolving the operational details of the ISA.

GE has modified Section 4.1 to include the additional commitments requested in this RAI (also see answer to RAI #3). 5. VALIDA TION OF CRITICALITY EVA L UA TIONS A T ENRICHMENTS EXCEEDING FIVEPERCENT Benchmarks do not existfor the conduct ofcriticality evaluations ofcommitments nL.) in the range of5-10 percent. Accordingly,for each specificprocess where uranium enriched to greater thanfive percent is to be used, provide a validation study including a criticality safety analysis and evaluation. This validation study

- -. - - -. _ ~ - - ~. - -- Mr. M. F. Weber May 27,1997 Attachment I q Page 5 of 13 V 4 should establishfor the specific cases calculated, (1) the case and data used are valid. and (2) that the specific quantitative methodfor setting margins ofKag to accountfor uncertainties and biases. This quantitative methodfor margins should address both normal and accident conditions. The current additional margins ofKr.gless than.97for normal and.95for accident conditions, in the absence ofadditional experiments or information, are inadequate to accountfor the uncertainties in extrapolation much beyond 5 percent enrichment. GE has decided to withdraw our request for processing material enrichments up to 6.0 wt. % U235, at this time. Accordingly, Chapter 6, section 6.3.2.3 of i our license rene41 has been modified to exclude validation justification for l processing higher than 5.0 wt. % U235. Similarly Sections 1.1 and 1.2.2 have been modified to limit the authorized enrichment for production to 5% (currently authorized at 6%). Our future plans willinclude a separate license amendment addressing the 5-10% enrichment range for both GEKENO, GEMER. [ At our Management Meeting of May 20,1997, we also provided an overview j )Q of the direction our business is heading with high burn-up fuel and the i implication for higher enrichments over the next few years. [ 6. TABLE OF PLANTSYSTEAIS AND PARAh!ETER CONTROLS i l Table 6.0, page 6.9 does not appear to be complete. Accordingly, please provide information on missing areas andsystems. Specifically, add to this table the Dry Conversion Process Integration Facility, including transfer corridors. 4 Table 6.0 has been modified to call out the same level of detail for the integration of the DCP to the balance of the fuel manufacturing process as was used for the other process steps. l 7. CRITICALITY CONTROLS FOR TRANSFER CORRIDOR ADJACENT TO LAUNDRY ) Please provide information describing potential criticality scenarios identifiedfor the Dry Conversion Process Integration Facility Transfer Corridor adjacent to l the laundry. Have all scenarios that could introduce water unexpectedly into the

p corridor such as washing machine overflows, pipe breaks, drains plugged, etc.,

V been identijled, controls established, and the likelihood ofcausing afailure of moderation control evaluated to be acceptably low? l A

Mr. M. F. Weber May 27,1997 Page 6 of13 l Yes, all credible external water source ingress pathways have been { considered. The model for the criticality accident is a non-uniform distribution of l moderator in 1000 kilograms of uranium oxide powder enriched to 5 wt. i percent U-235. The powder is assumed to be non-homogeneous with respect to particle sizes up to 1500 microns. Neutron reflection at the boundary of j the model is twelve inches of water to represent the worst case. The non-l uniform moderator limit calculated in CSA 1320.02, Rev. 03, is 9.3 kg. of water. L b l Normal operating condition is uranium oxide powder not exceeding 5 wt. percent enrichment in U-235 that is contained in a water (spray) esistant, i strong stainless steel powder container. The transfer corridor is dry and the } powder container is attended at all times while in the transfer corridor. The l two concurrent contingencies required for a criticality accident to be possible {' in the Integration Transfer Corridor are loss of containment of the uranium j oxide powder and accumulation of moderator. The controls for the Integration Transfer Corridor are categorized as two control systems j Q designed to make the occurrence of either of these two contingencies unlikely. l Loss of containment of the r ranium oxide powder is prevented by controls that make significant water ingress to the container or powder spillage from the container unlikely. The powder containers are built to a specification that requires the container be able to withstand 15 psi pressurization, normal handling stresses, and remain essentially dry when subjected to a moderate pressure water spray. The individual control specifications and requirements include the container fabrication drawing specification, operator procedural requirements, and information management systems that authorize container movements. During movement through the Integration Transfer Corridor, attendance of the powder container by an operator who is capable ofimmediately moving the powder container is required. The normal operating condition is to transfer the container through the transfer corridor without delay. Leaving the container unattended during the transfer, as a result of an actual or simulated emergency that requires immediate building evacuation,is a mitigating circumstance that may result in degradation of a control, but is considered an acceptably low safety risk This condition is temporary and does not represent an immediate risk of a criticality accident. 1 Accumulation of moderator in the transfer corridor is prevented by controls O that make it uniiueir or waier from either spraying er spreading over ihe r floor into the corridor to occur. Credible sources of water are identified within the laundry room, overhead process piping, and rain water from the

i Mr. M. F. Weber May 27,1997 3 Page 7 of13 ,J i I external environment. The operator observes the condition of the Integration Transfer Hallway and is trained not to move the container l through standing water on the floor or a water spray into the area. Characteristics of the Integration Transfer Corridor are summarized as 4 follows: j A single barrier roof over the transfer corridor is an acceptable barrier e because of the requirement to directly transfer the powder container (minimize time present in the transfer corridor). l Process piping that normally contains moderator and passes through the Integration Transfer Corridor is encased in a shroud that drains any 3 leakage outside the transfer corridor. Rooms that normally containing moderator, such as the laundry room, e are separated by a wall or doors to prevent ingress of water spray and j other physical barriers that restrict the movement of the large powder transfer container, Gravity drains in the laundry room are physically at a lower elevation e than the corridor and direct any accumulation of water on the floor from O equipment (or fire protection system) away from the transfer coraidor moderation restricted area. j Drains, walls, and other physical barriers that are identified as important to nuclear criticality safety, are identified and marked to indicate their j importance. These requirements are identified and documented in appropriate nuclear criticality safety analyses. I I 8. The RAIdatedMarch 5,1997, questioned whether use ofchemicalsfollowed the OSHA Process Safety Management Standard (29 CFR 1910.119) in Section 7.1 1 (page 7.1). NRC also stated that elements ofthe Chemical Safety Programfor UF and hydrofluoric acidshould be included in the license application. j 6 i GEprovided a general response, noting that the regulations are implemented through internalprocedures as typified by GE's internal safetyprocedure 303 1 Safety Considerations in Design. However, elements ofthe ChemicalSafety Programfor UFe and hydrofluoric acid were not included or referenced in the l license application. ^ Because UFs is licensed material and is used daily at thefacility and hydrofluoric acid is an ofgas produced by the processing ofthis licensed material, these

O chemicals should be specifically discussed in the Chemical Safety Program.

Release ofthese materials could affect the availability and reliability ofsafety

t i l i Mr. M. F. Weber May 27,1997 Attachment I Page 8 of13 t controls relied onforplant safety. An acceptable approach to resolve the staf]'s concerns would be to include thefollowing language in Chapter 7.0, Section 7.1: i l This chemical safetyprogram is applicable to the chemicals associated with the authorized activities in Chapter 1 and include UFs and hydrofluoric acid i j as well as any other chemicals which may directly or indirectly afect the i nuclear safety ofthese activities. l' The management control elements ofthe chemical safetyprogram of UFs and i hydrofluoric acid should also be included in the license application This means l l that management control elements ofthe GE-Wilmington ChemicalSafety i Program (as described in Section 7.2 ofLicense Application) that apply to UFs l and hydrofluoric acidshould be included in the application. An acceptable i commitment would be asfollows: l l The ChemicalSafety Programfor UFs and hydrofluoric acid utilize the follmving elements: Integrated Safety Analysis and Conduct ofOperations. Exactplacement in the license application is left to the discretion ofthe licensee ,Q but Chapter 7.0 - Chemical Safety appears to be the best chapter. GE has modified Sections 7.1 and 7.2.1 to add the words of clarification identified in RAI #8. 9. In Section 6.2.5.1, page 6.23 ofthe renewal application it is stated that " Structure I and/or neutron absorbers that are not removable constitute aform ofgeometry control... ". Since the use ofthe term " geometry control"forfixedabsorbers has the potentialfor confusion, what measures will be taken to assure the proper assessment and maintenance offixedneutron absorbers? Doplantprocedures mandate compliance with the measures ofANSl'ANS 8.21 and with ANSI /ANS 8.1 section 4.2.3? Section 6.2.5.1 states,"... Favorable geometry is based on limiting dimensions of defined geometrical shapes to established suberitical limits. GE then considers structure and/or neutron absorbers that are not removable constitute a form of geometry control...". This means that for structures which includes neutron absorbers as an integral element (versus removable l elements such as sleeves etc.) of their configuration, we consider the nuclear l poison a subset of geometry control. These structures are included in the ! n periodic verification requirements of section 6.2.5.5. I V GE internal procedures do not mandate verbatim compliance with ANSI /ANS 8.21 and ANSI /ANS 8.1 section 4.2.3, however, our programs and

Mr. M. F. Weber May 27,1997 Page 9 of13 l l procedures conform to the intent of requirements,and guidance expressed in these standards. I I 10. l&t is the technical basisfor the safe batch rule ofsection 6.2.4 embodied in the l equation: kgs UO x 0.88 = kgX

• f wheref= wt. % Uin compoundXandkgs UO is the 2

2 safe batch sizefor UO ? 2 GE acknowledges that there is a problem with the notation, and we have re-l written the equation, for clarity, to read the following: l kgs X = (kgs UO

• 0.88 ) / f 2

where, kgs X = safe batch value of compound 'X' kgs UO2 = safe batch value for UO2 U compound X O sec'i 624 r riice epiic'i h 6ee -edi'ied ccraiair- ~ The safe batch tables in the current SNM-1997 license and in the renewal l application are for uranium dioxide (UO2). The mass includes the oxygen which represents approximately 11.85% of the UO2 mass. Therefore,if we consider another uranium compound and replace the mass contribution from . oxygen by the mass contribution from the non-uranium constituents of the other compound, we will still have a safe batch providing that the other uranium compound is neutronically less reactive. UO2 is the most reactive form of uranium processed at this facility. Therefore,if we apply the equation in question to ammonium diuranate (ADU), we would expect the resulting mass to produce a k-effective less than that resulting from the safe hatch mass of UO2. This is shown in the attached figure for 18.1 Kg of UO2 and 21.545 Kg of ADU, each at 5.00% U235 enrichment. The 21.545 was determined by multiplying 18.1 by 0.88144 and dividing by 0.74049, the respective U-factors for UO2 and ADU. l UO2 is also more reactive than the other uranium compounds characteristic l of our processes [U308, UO2F2, UO2(NO3)2 - 6H20] While uranium metal is not a compound, we recognize that li can be more reactive than UO2. lL LO

. Mr. M. F. Weber May 27,1997 Page 10 of13 SAFE BATCH CONPARISON BETHEEN 002 AND ADU ..,i, LEGEND U COMPOUND + wnTER . se.: es visica x 21.6 5 KG n0Uts) 0.898 f 0.670 A 0.850 K-trF tat O.880 0 010 C) O.770 l 2. se se se 7. weisur reaction warca xi. 8 11, With respect to Table 6.1, " Safe Geometry Values", what is the sigmficance ofthe missing valuesfor cylinder diametersfor Homogeneous Aqueous Solutions? What methods are to be usedin this case, ifnot this table? Inparticular, what method was used to determine the diameter of UFs hydrolysis columns? For cases where the enrichment exceeds 5%, provide additional information showing how these values have been validated and that they incorporate suficient margins to accountfor uncertainties. Are they validated by comparison experiments? For values in Table 6.1 at enrichments less than 5%, a comparison to the most recentlypeer reviewedguidance, Norman L Pruvost and Hugh C. Paxton, LA-12808 Nuclear Safety Guide, Sept.1996 (formerly T1D-7016), shows several values that differ in the non-conservative direction. These are noted below. Pleaseprovide additionaljustificationfor these values or adopt thosefrom LA-12808.

._. - _ _ ~ 4 ~ Mr. M. F. Weber May 27,1997 i Attachment I Page 11 of13 a l Diferences between IA12808 and Table 6.1: I i Enrich. Table 6.1 Ik 12808 Homogeneous UO2 H2O cylinder diam. 2% 16.70 in. 16.50 in. Homogeneous UO2 H2O slabs 2% 8.90 - in. 8.82 in. 3% 6.25 in. 6.10 in. 4% 5.10 in. 4.96 in. 5% 4.45 in. 4.37 in. - Homogeneous aqueous solutions, slabs 4% 6.00 in. 5.94 in. Homogeneous UO2 & water, Kgs UO2 4% 25.7 Kg. 25.5 Kg. 5% 18.1 Kg. 16.0 Kg. Heterogeneous UO2 pellets & water, 3% 38.1 Kg. 36.1 Kg. i Kgs UO2 4% 24.7 Kg. 20.3 Kg. 5% 18.1 Kg. 13.9 Kg. + The values were included in the 1979 version of SNM - 1997 but not in the current version. Currently,while they are acceptable limits, they are O 8 r iir d very ii'*i i e r c rr =' aractie i 'a =*iiize di cree' models in most cases. Some of the older criticality analysis for the facility are based on these values. The values are as follows: wt% U235 Inf Cyl Dia wt.% U235 Inf Cyl Dia 2.00 16.7 in 3.25 12.5 in 2.25 15.0in 3.50 12.1 in 2.50 14.0 in 3.75 11.9 in 2.75 13.3 in 4.00 11.7 in 3.00 12.9 in 5.00 9.5 in These values represent 93% of the minimum critical dimension, and will be added to Table 6.1. The 10-inch Schedule 40 hydrolysis receiver and storage tanks are modeled explicitly and analyzed using the GEKENO Monte Carlo program. The analysis is documented in "CSA - EVALUATION OF HYDROLYSIS AT 5%", performed under Change Request 89.0224. The questions regarding enrichment are discussed under RAI #5. Based on that discussion, the tables are modified by deleting values above 5.00%. The remainder of this item deals with comparison of single parameter limits between Table 6.1 of the GE license, and the recently published LA-12808. 1

Mr. M. F. Weber May 27,1997 i i Page 12 of13 i Specifically, LA-12808 Table 8 (solutions) and Table 9 (homogeneous and { heterogeneous oxides) are based on work originating from H. K. Clark l (Table 8 - NSE vol. 81, pp. 351-378,1982 and Table 9 from DP-1014,1966, i receptively). Both sets of data were obtained using analytical techniques normalized to appropriate critical experiments, however, both include some l degree of uncertainty in the results. The small differences identified in the RAI for dimensional limits appear to be within the uncertainty and are not j believed to be significant. The differences in comparison of safe hatch values is significant and occurs i because the numbers do not represent the same thing. The safe batch values in Table 6.2 represent 45% of the minimum critical mass. The corresponding values in the RAI are obtained by taking 45% of the i suberitical limits reported in Table 9 of LA-12808. These limits were taken from DP-1014 tables of" safe" values which are defined in DP-1014 to l correspond to a k-effective value of 0.98. Significant differences are noted between the safe batch values in Table 6.2 i for Heterogeneous UO2 Pellets & Water Mixtures and the safe batch values iC inferred from LA-12808. Some of the difference can be attributed to the 45% of the minimum critical value (GE) verses the 45% of the subcritical value (LA-12808) explained in the previous paragraph. Another significant difference is due to the data in Table 6.2 being for pellets and the data in LA-2 12808 being for optimum diameter rods. Since the optimum rod diameter l for mass limited systems of 5% enriched UO2 is about 1/8-inch diameter arad i pellets is much larger (1/3 to 2/5-inch diameter) the safe batch table for pellets does not apply for smaller dimensions than the pellet diameter. l In any case, the standard practice at GE is to explicitly model mass limited i ' heterogeneous systems' containing fuel diameters smaller than pellets using j Monte Carlo (e.g., GEMER) analytical methods. l 12. Concerning Section 6.4.1 page 6.36, is the location andspacing ofcriticality ( monitoring alarms such that the system meets the requirements ofeither 10 CFR 70.24 (a)(1) or (a)(2)? Specifically, is coverage ofall areas by two detectors i provided? Are there any areas ofthefacility that will not be covered by detectors j meeting the requirements? Is SNM ever handled, used, or stored in these areas? Yes, GE's criticality monitoring accident alarm system meets the j requirements of 10 CFR 70.24 (a) (1), except for the special authorizations lp %./ stipulated in Section 1.3.11 of the renewal application. i i i 0

i 1 Mr. M. F. Weber j May 27,1997 Attaciunent 1 i o Page 13 of13 V, Outlinefor the ISA Summaryfor GE Balance ofPlant 1. The areas ofreviewfor each system should be listed e.g., radiological safety, criticality safety, chemical safety, fire protection, and industrial safety, etc. 11. For each system, a description ofhow the 1SA teams areformed, t)pe of membership, minimum quahfication ofmembers, how areas ofreview were integrated, and management and QA oversight. 111. A list ofspecific writtenplantprocedures, techniques, and computer based tools used by the ISA teams to perform the ISAfor each system. IV. A list ofthe segments that the system was broken into to perform the ISA and why. V. A description ofhow the sequences ofthe work wasperformed by the ISA team e.g., identify the hazards, determine the consequences andlikelyfrequency, identify the controls whichprohibit or mitigate the consequences, establish a risk ranking (frequency x consequence) unmitigated and mitigated. O VI. A description ofhow the ISA team determined the consequences ofthe event or condition. i V11. Based on the 1SA process, provide a list ofthe most important process segments and the controls relied upon to prevent andfor mitigate an incident. The incident description should include a list ofthe initiating event (internal or external), the unmitigated consequences ofthe resulting accident, and the necessary level of quality and reliability establishedfor each control. VIII. Summary matrix ofaccident sequences plotted by consequence versus probability (qualitative). i O

Mr. M. F. Weber May 271997 Page1of1 .O ATTACHMENT 2 i

1) Description of Revisions to the License Renewal Application by Page and Section

2) License Renewal Page Changes O

O i

Mr. M. F. Weber May 27,1997 Page1of1 l /sV Description of Revisions Pace Section Description 1.1 1.1 Reduced authorized enrichment from 6% to 5%. 1.7 1.2.2 Reduced authorized enrichment from 6% to 5%. 2.3 2.2.1.2 Changed minimum qualifications. 2.4 2.2.1.3 Changed minimum qualifications. I 2.5 2.2.1.3 Changed minimum qualifications. l 2.8 2.2.1.8 Changed minimum qualifications. 2.9 2.2.1.9 Changed minimum qualifications. 3.11 3.9 Provided clarifying words. i 3.12 3.9.2 Corrected footnote references. 4.1 4.1 Added cc.nmitment statement. 6.11 6.2.3 Made changes to the table to better define the integration piece of DCP. i 6.12 6.2.3 Made changes to the table to better define the integration piece of DCP. 6.18 6.2.3 Made changes to the table to better define the integration piece of DCP. 6.19 6.2.3 Made changes to the table to better define the integration piece of DCP. 6.21 6.2.4 Corrected the equation. 6.23 6.2.4 Eliminated values greater than 5%. 6.26 6.2.5.4 Modified definition. 6.33 6.3.2.3 Reduced authorized enrichment from 6% to 5%. 6.36 6.4.1 More clearly call out when decreased detection spacing is required. 7.1 7.1 Provided clarifying wording. 7.1. 7.2.1 Provided clarifying wording.

1 i ~ t 1 TABLE OF CONTENTS l lh ~ Section Title Page i CHAPTER 1.0 GENERAL INFORMATION I e 1.1 Facility and Process Description 1.1 1.2 InstitutionalInformation 1.7 j 1.3 Special Authorizations 1.10 CHAPTER 2.0 4 [ ORGANIZATION AND ADMINISTRATION 2.1 Policy 2.1 i 2.2 Organizational Responsibilities and Authority 2.1 1 2.3 Safety Committees 2.10 CHAPTER 3.0 CONDUCT OF OPERATIONS 1' 3.1 Configuration Management (CM) 3.1 Q 3.2 Maintenance 3.2 3.3 Quality Assurance (QA) 3.4 3.4 Training and Qualification 3.6 3.5 Human Factors 3.7 3.6 Audits and Assessments 3.7 3.7 Incident Investigations 3.9 3.8 Records Management 3.10 3.9 Procedures 3.11 CHAPTER 4.0 INTEGRATED SAFETY ANALYSIS 4.1 Integrated Safety Analysis 4.1 4.2 Site Description 4.1 4.3 Facility Description 4.1 4.4 Process Description 4.2 4.5 Process Safety Information 4.2 LICENSE SNM-1997 DATE 05/27/97 Page DOCKET 70-1113 REVISION 0 1 1

TABLE OF CONTENTS Section Title Page l 4.6 Training and Qualifications of the ISA Team 4.2 4.7 ISA Methods 4.2 3 i 4.8 Results of the ISA 4.3 { 4.9 Controls for Prevention and Mitigation of Accidents 4.4 ] 4.10 Administrative Control of the ISA 4.7 j j CHAPTER 5.0 RADIATION SAFETY l 5.1 ALARA (As Low As is Reasonably Achievable) Policy 5.1 j 5.2 Radiation Safety Procedures and Radiation Work Permits (RWPS) 5.1 5.3 Ventilation Requirements 5.2 5.4 Air Sampling Program 5.3 5.5 Contamination Control 5.5 5.6 External Exposure 5.7 5.7 Internal Exposure 5.7 5.8 Summing Internal and External Exposure 5.9 5.9 Action Levels for Radiation Exposures 5.9 5.10 Respiratory Protection Program 5.9 p) 5.11 Instrumentation 5.10 CHAPTER 6.0 NUCLEAR CRITICALITY SAFETY 6.1 Program Administration 6.1 6.2 Technical Practices 6.5 6.3 Control Documents 6.29 6.4 Criticality Accident Alarm System 6.36 CH APTER 7.0 CIIEMICAL SAFETY 7.1 Chemical Safety Program 7.1 7.2 Contents of Chemical Safety Program 7.1 LICENSE SNM-1097 DATE 05/27/97 Page DOCKET 70-1113 REVISION 0 2

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Section Title Page i-CHAPTER 8.0 l FIRE SAFETY 8.1 Fire Protection Program Responsibdity 8.1 8.2 Fire Protection Program 8.1 8.3 Administrative Controls 8.2 8.4 Building Construction 8.2 ( 8.5 Ventilation Systems 8.3 8.6 _ Process Fire Safety 8.3 8.7 Fire Detection and Alarm Systems 8.3 l 8.8 Fire Suppression Equipment 8.4 8.9 Fire Protection Water System 8.4 8.10 Radiological Contingency and Emergency Plan (RC&EP) 8.5 i 8.11 Emergency Response Team 8.5 CHAPTER 9.0 RADIOLOGICAL CONTINGENCY AND EMERGENCY PLAN 9.1 CHAPTER 10.0 ENVIRONMENTAL PROTECTION 10.1 Air Emuent Controls and Monitoring 10.1 10.2 Liquid Treatment Facilities 10.1 l 10.3 Solid Waste Management Facilities 10.2 l 10.4 Program Documentation 10.2 10.5 Evaluations 10.3 10.6 . Off-site Dose 10.3 10.7 ALARA 10.4 CHAPTER 11.0 DECOMMISSIONING 11.1 LICENSE SNM-1997 DATE 05/27/97 Page DOCKET 70-1113 REVISION 0 3 i

.. _._.._ _ ~.. REVISIONS BY CHAPTER O t Application Application l Page Date Page Date l TABLE OF CONTENTS l l CHAPTER 6.0 l I through 4 05/27/97 l 1 through 36 05/27/97 l l CHAPTER 1.0 l l CHAPTER 7.0 l 1 through 21 05/27/97 l 1 through 3 05/27/97 l l CHAPTER 2.0 l l CHAPTER 8.0 l 1 through 11 05/27/97 l 1 through 5 04/05/96 0 -l CHAPTER 3.0 l l CHAPTER 9.0 l 1 through 12 05/27/97 l 1 02/25/97 l CHAPTER 4.0 l l CHAPTER 10.0 l j 1 through 8 05/27/97 l 1 through 16 04/05/96 l CHAPTER 5.0 l l CHAPTER 11.0 l 1 through 13 OP/30/96 1 04/05/96 LICENSE SNM-1997 DATE 05/27/97 Page DOCKET 70-1113 REVISION 0 4

n CHAPTER 1.0 ('/ GENERAL INFORMATION 1.1 FACILITY AND PROCESS DESCRIPTION The primary purpose of the GE-Nuclear Energy Production facility in Wilmington, North Carolina (identified in this document as GE-Wilmington) is the manufacture of fuel assemblies for commercial nuclear reactors. Nuclear materials enriched to less than or equal to 5 weight percent U-235 are utilized in the product I manufacturing operations authorized by this license. The safety, environmental, quality assurance and emergency preparedness aspects of the manufacturing operations are managed and controlled as described in this license. 1.1.1 SITE DESCRIPTION AND LOCATION GE-Wilmington is situated on a 1,664-acre tract ofland, located on U.S. liighway 117 and approximately six miles north of the City of Wilmington, North Carolina in New Hanover County (refer to Figures 1.1 and 1.2). New Hanover County is situated in the southern coastal plains section of southeastern North Carolina, with the Atlantic Ocean on the east and the Cape Fear River on the west. The Atlantic (] Ocean lies approximately 10 miles east and 26.4 miles south of GE-Wilmington. The surrounding terrain is low-lying, with an average elevation ofless than 40 feet above mean sea level. Castle Hayne, the nearest community, is approximately three miles north of GE-Wilmington. The region around the site is lightly settled with large areas of heavily timbered tracts ofland. Farms, single-family dwellings and light commercial activities are located along U.S. I17. The Wilmington airport is located approximately 3.5 miles southeast of the site. 1.1.2 FACILITY DESCRIPTION The location and arrangement of buildings at the GE-Wilmington site, and their relative distance from the site boundary are shown in Figure 1.3. Located on the GE-Wilmington site are the following major facilities: (1) the GE Aircraft Engine (AE) LICENSE SNM-1097 DATE 05/27/97 Page f] DOCKET 70-1113 REVISION 0 1.1

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I l FIGURE 1.3 (Continued) O on-wituinoros sirs etin tsoexo 1 Fuel Manufacturing Operation (FMO) I 2 : Fuel Components Operation (FCO) l 3 : Aircraft Engine Operation (AE) 4 : Services Components Operation (SCO) 5 : Final Process Basins 6 : Waste Treatment Facility i 7 : Incinerator Building 8 : Filter Facility ) 9 : DA Building 10 : Boiler / URLS 4 i 11 : Office Building 12 : Site Maintenance 13 : Site Warehouse 14 : FMO Storage Building 15 : FMO Maintenance Building 16 : AE Maintenance Building 17 : Waste Treatment Basins i 18 : Fuel Examination Technology 19 : Dry Conversion Process Building (DCP) 20 : Warehouse CaF Storage Warehouse 21 : 2 I i i LICENSE - SNM-1997 DATE 05/27/97 Page DOCKET 70-1113 REVIS!ON 0 1.5

facility which is not involved in the nuclear fuel manufacturing operation, (2) The f) Services Components Operation (SCO) facility where non-radioactive reactor components are manufactured, (3) the Fuel Components Operation (FCO) facility where non-radioactive components for reactor fuel assemblies are manufactured, and (4) the fuels complex containing the fuel manufacturing facility. The fuels complex, which includes the Fuel Manufacturing Operation and Dry Conversion Process (FMO/FMOX & DCP) buildings and supporting facilities, is enclosed by a fence with restricted access. This complex is called the Controlled Access Area (CAA). 1.1.3 FACILITY RESISTANCE TO ENVIRONMENTAL EVENTS In the coastal area of Nonh Carolina, where GE-Wilmington is located, severe weather conditions may result from hurricanes, tornadoes, ice storms, and snow storms. The greatest severe weather threat in this area is due to high winds from hurricanes and possible tornadoes. Facility construction meets or exceeds local codes for strength and in the case of hurricanes, advance notice provides an opportunity for further mitigating actions. Since high winds could impact electrical power, key safety systems are protected with adequate back-up power supplies or fail safe features. Earthquakes are not considered a major threat because this section of the southern Atlantic Seaboard is an area of relatively low seismic activity. The Fuel Manufacturing Operation building in which radioactive materials are f) processed and stored, is designed to provide for containment of material under adverse environmental conditions such as fire, wind, flooding or earthquake to the limits of the building code The roof construction meets Factory Mutual requirements for fire hazard and wind resistance. Detailed information regarding the facility resistance to the effects of potential credible accident events is contained in Chapters 2 and 5 of the Radiological Contingency and Emergency Plan for GE-Wilmington, which is described in Chapter 9.0 of this license, and in Chapters 2 and 6 of the Environmental Report for GE-Wilmington which is described in Chapter 10.0 of this license. 1.1.4 PROCESS DESCRIPTION The product manufacturing operations authorized by this license consist of receiving 1 low-enriched, less than or equal to 6.0 weight percent U-235, uranium hexafluoride; converting the uranium hexafluoride to uranium dioxide powder; and processing the LICENSE ShM-1097 DATE 05/27/97 Page 7 DOCKET 70-1113 REVISION 0 1.6 (Y

uranium dioxide through pelletizing steps, fuel rod loading and sealing, and fuel () assembly fabrication. Two types of processes are used to convest uranium hexafluoride to uranium dioxide powder -- the Ammonium Diuranate (ADU) process, and Dry Conversion Process (DCP). The manufacturing operations are served by support systems such as scrap recovery, waste disposal, laboratory, and manufacturing technology development, which are also descrmed in this license. 1.2 INSTITUTIONAL INFORMATION The GE-Wilmington NRC license number is SNM-1097 (Docket #70-1113). 1.2.1 IDENTITY AND ADDRESS This application for license renewal is filed by the GE-Nuclear Energy Production facility of the General Electric Company, a major corporation with corporate headquarters in Fairfield, Connecticut. General Electric's nuclear energy business, known as GE Nuclear Energy, is headquartered in San Jose, Califomia, with the principal manufacturing facility located in Wilmington, North Carolina. The full address is as follows: GE Nuclear Energy Production, (name of person and g() mail code), P.O. Box 780, Wilmington, NC 28402. 1.2.2 TYPE, QUANTITY, AND FORM OF LICENSED MATERIAL Uranium normally will be used at GE-Wilmington in the Controlled Access Area (CAA) only. Conversion and fabrication is conducted within the fuel manufacturing building (FM0/FMOX & DCP). Small quantities (i.e., less than one safe batch of uranium in a non-dispersible form) may be temporarily moved to other buildings or site locations outside of the CAA for special tests under special authorizations and controls. The following types, maximum quantities, and forms of special nuclear materials are authorized:

1) 50,000 kilograms of U-235 contained in uranium enriched to a maximum l

enrichment ofless than or equal to 5%, in any chemical or physical form except I metal; LICENSE SNM-1097 DATE 05/27/97 Page ] DOCKET 70-1113 REVISION 0 1.7

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2) 500 kilograms of U-235 at enrichments from 6% to <10% contained in uranium

!O co-rouads for use ia iadoratory a=a process deveion= eat overatio s: i

3) 9.649 kilograms of U-235 at enrichments from 10% to <l5% contained in uranium compounds for use in laboratory and process development operations; j

4) 350 grams of U-235 in any form contained in uranium at any enrichment, for use in measurements, detection, research or development activities;

5) Plutonium - 1 milligram in samples for analytical purposes,1 milligram as standards for checking the alpha radiation response of radiation detection instrumentation,20 grams as sealed neutron sources, and in nuclear fuel rods at a 4

2 level ofless than 1 x 10 gram of plutonium per gram of U n,

6) 50 milligrams U-233 for analytical purposes.

1.2.3 ACTIVITY GE-Wilmington complies with applicable parts of Title 10, Code of Federal Regulations, unless specifically amended or exempted by NRC staff. Authorized activities at GE-Wilmington include: 1.2.3.1 Product Processing Operations tO UF Conversion - Conversion of uranium hexafluoride to uranium oxides by e 6 the ADU process, and the Dry Conversion Process. Fuel Manufacture - Fabrication of nuclear reactor fuel (powder, pellets, or assemblies) containing uranium. Scrap Recovery - Reprocessing of unirradiated material from GE-Wilmington and from other cources with nuclear safety characteristics not significantly different from GE-Wilmington in-process materials. Waste Recovery - Recovery of uranium from wet and dry material stored in on-site pits and basins. The recovered uranium will be returned to the fuel processing facility. l i f LICENSE SNM-1997 DATE 05/27/97 Page l O oocK8T va-2"3 nevisio" a '8

s 1 \\ 1.2.3.2 Process Technology Operations r' Development and fabrication of reactor fuel, fuel elements and fuel assemblies of advanced design in small amounts. I Development of scrap recovery processes. e i l Determination ofinteraction between fuel additives and fuel materials. e 1 l Chemical analysis and material testing, including physical and chemical testing and analysis, metallurgical examination and radiography of uranium compounds, alloys and mixtures. Instrument research and calibration, including development, calibration and e functional testing of nuclear instrumentation and measuring devices. 1 A conversion of UF6 to UO and other intermediate compounds using i 2 chemical and dry processes. Other process technology development activities related to, but not limited 8 by, the above. 1.2.3.3 Laboratory Operations Chemical, physical or metallurgical analysis and testing of uranium compounds and m;xtures, including but not limited to, preparation oflaboratory standards. 1.2.3.4 General Services Operations Storage of unirradiated fuel assemblies, uranium compounds and mixtures in areas arranged specifically for maintenance of criticality and radiological safety. Design, fabrication and testing of uranium prototype processing equipment. Maintenance and repair of uranium processing equipment and auxiliary systems. j Storage and nondestructive testing of fuel rods containing trace amounts of j e plutonium as authorized in the license. LICENSE SNM-1097 DATE 05/27/97 Page O oocxer 'a-" ' 3 navisio" a '9 I I

i l 1.2.3.5 Waste Treatment and Disposal m ( ) C Treatment, storage and disposal and/or shipment ofliquid and solid wastes whose discharges are regulated. Decontamination of non-combustible contaminated wastes to reduce uranium contamination levels, and subsequent shipment of such low-level radioactive wastes to licensed burial sites for disposal or as authorized by the NRC. Treatment or disposal of combustible waste and scrap material by incineration pursuant to 10 CFR 20.2002 and 10 CFR 20.2004. 1.2.3.6 Off-site Activities Testing, demonstration, non-destructive modification and storage of materials and devices containing unirradiated uranium, provided that such materials and devices shall be under GE control at all times. 1.3 SPECIAL AUTHORIZATIONS 1.3.1 ACTIVITIES REQUIRING PRIOR NRC AUTHORIZATION BY LICENSE /~G' AMENDMENT 1.3.1.1 Major changes or additions to existing processes which may involve a significant increase in potential or actual environmental impact resulting from utilizing such changes or additions. 1.3.1.2 Major process changes or major additions which involve a new process technology for which the safety demonstration has not been previously subjected to review by the NRC. In determining whether a new process technology requires such prior authorization by license amendment, the fbliowing factors will be considered: (1) type of equipment utilized, (2) chemical reactions involved and (3) potential and/or actual environmentalimpact. l LICENSE SNM-1097 DATE 05/27/97 Page () DOCKET 70-1113 REVISION 0 1.10 l l l

l 1.3.1.3 Proposed activities for which specific application and prior approval are required by () NRC regulations. 1.3.2 CONTAMINATION-FREE ARTICLES l Authorization to use the guidelines, contamination and exposure rate limits specified l at the end of this Section, " Guidelines for Decontamination of Facilities and l Equipment Prior to Release for Unrestricted Use or Termination of Licenses for Byproduct, Source, or Special Nuclear Material," US NRC, April 1993 for decontamination and survey of surfaces or premises and equipment prior to abandonment or release for unrestricted use. 1.3.3 TRANSFER OF CONTAMINATION-FREE LIQUIDS 1.3.3.1 Transfer of Hydrofluoric Acid (HF) for Testing Authorization to transfer test quantities of HF to potential buyers / customers or laboratories for the purpose of analyzing, examination or evaluation, without continuing NRC controls as described in GE-Wilmington's letter to the NRC dated February 26,1996. Test quantities may not contain more than 3 PPM uranium with an enrichment not to exceed 6% U-235. The recipients will be advised that this material is not a nuclear hazard, but will be advised that the material should be handled carefully and in such a manner so as not to be consumed by humans nor used in products used on or in the body or in the food chain. 1.3.3.2 Hydrogen Fluoride Solutions Authorization, pursuant to 10 CFR 70.42(b)(3), to transfer liquid hydrofluoric acid to Brush Wellman, Elmore, Ohio, through the chemical supplier, Consolidated Chemical Company, Kansas City, Missouri, without either company possessing an l NRC or Agreement State license for special nuclear material, provided that the concentration of uranium does not exceed three parts per million by weight of the liquid and the enrichment is less than or eqtal to 6 weight percent U-235. l LICENSE SNM-1097 DATE 05/27/97 Page (~') DOCKET 70-1113 REVISION 0 1.11 v l

l The hydrofluoric acid is transferred and used in such a manner that the minute () quantity of uranium does not enter into any food, beverage, cosmetic, drug or other commodity designated for ingestion or inhalation by, or application to, a human being such that the uranium concentration in these items would exceed that which naturally exists. Additionally, the acid is used in a process which will not release the low levels of radioactivity to the atmosphere as airbome material and whose residues will remain in a wastewater or other treatment system. Prior to shipment, each transfer is sampled and measured to assure that the concentration does not exceed three parts per million of uranium. j GE-Wilmington shall maintain records under this condition oflicense including, as a i l minimum, the date, uranium concentration and quantity of hydrofluoric acid transferred. 1.3.3.3 Nitrate-Bearing Liquids ] Authorization to transfer nitrate-bearing liquids, provided that the uramum concentration does not exceed a 30-day average of 5 parts per million by weight of the liquids and the enrichment is less than or equal to 6 weight percent U-235 by transport to an off-site liquid treatment system located at Federal Paper Board Corporation, Riegelwood, North Carolina, in which decomposition of the nitrates a will occur and from which the denitrified liquids will be discharged in the effluent V from the system. The environmental monitoring program as described in Chapter 10.0 of this license is used to control these activities. 1.3.4 TRANSFER OF CALCIUM FLUORIDE TEST QUANTITIES Authorization to transfer test quantities of calcium fluoride (CaF ) to potential buyers 2 for the purpose of their examination and evaluation as described in GE-Wilmington's letter to the NRC dated September 24,1992. Test quantities may not contain more than 30 pCi per gram on a dry weight basis and are limited to 1 gram U-235 at each off-site 'ocation. Test activities and end uses of the material will be limited to those that do not allow chemical separation of the uranium or entry of the product into the food chain. LICENSE SNM-1097 DATE 05/27/97 Page (] DOCKET 70-1113 REVISION 0 1.12 l l

1.3.5 TRANSFER OF CALCIUM FLUORIDE (CAF ) TO VENDORS FOR 2 (,) BENEFICIAL REUSE Authorization to transfer quantities ofindustrial waste treatment products (primarily CaF ) to Cametco, Inc., Pittsburgh, PA, for the parpose of briquette manufacturing 2 l and use as a steel flux forming material in the production of steel as described in GE-l Wilmington's letter to the NRC dated December 20,1989. Measurements are made using a sample plan to provide at a 95% confidence level that the population mean for each shipment is less than 30pCi of uranium per gram of material on a dry weight basis. Activities and end use of the material will be limited to those that do not allow chemical separation of the uranium or entry of the product into the food chain. 1.3.6 DISPOSAL OF INDUSTRIAL WASTE TREATMENT PRODUCTS l l Notwithstanding any requirements for state or local government agency disposal permits, GE-Wilmington is authorized to dispose ofindustrial waste treatment l products without continuing NRC controls provided that either of the two following conditions are met: t ,,C 1.3.6.1 Free-standing liquid shall be removed prior to shipment. l The uranium concentration in the material shipped for disposal shall not exceed 30 pCi per gram after free-standing liquid has been removed. The 'icensee shall possess authorization from appropriate state officials prior to disposing of the waste material. The authorization shall be available for inspection at the GE-Wilrnington facility. 1.3.6.2 The uranium concentration in the material shipped for disposal only at approved facilities such as Pinewood, South Carolina (licensed by the State of South Carolina), l !~ shall not exceed 250 pCi per gram of uranium activity, of which no more than 100 pCi per gram shall be soluble. l l l LICENSE SNM-1097 DATE 05/27/97 Page [') DOCKET 70-1113 REVISION 0 1.13 v

1.3.7 SANITARY SLUDGE t'/ Dried sanitary sludge is collected and disposed of at approved offsite facilities in accordance with Section 1.3.6. Authorization to store treated sanitary sludge containing trace amounts of uranium in the sanitary sludge land application area l pending final disposal. l 1.3.8 USE OF MATERIALS AT OFF-SITE LOCATIONS 1.3.8.1 Authorization to use up to 15 grams of U-235 at other sites within the limits of the United States and at temporaryjob sites of the licensee anywhere in the United States where the Nuclear Regulatory Commissien maintains jurisdiction for regulating the use oflicensed material. The manager of the radiation safety function shall establish the safety criteria for material being used at off-site locations and shall designate the individual who will be responsible for carrying out these criteria. 1.3.8.2 Authorization to store at nuclear reactor sites, uranium fully packaged for transport in any NRC approved package, in accordance with the conditions of a license O authorizing delivery of such containers to a carrier for NRC approved transport, at locations in the United States providing such locations minimize the severity of potential accident conditions to be no greater than those in the design bases for the containers during transportation. Provisions for compliance with applicable 10 CFR 73 requirements are described in the NRC-approved GE-Wilmington Physical Security Plan as currently revised in accordance with regulatory provisions. Storage at nuclear reactor sites is subject to the financial protection and indemnity provision of10 CFR 140. 1.3.8.3 Authorization to store at nuclear reactor sites, arrays of finished reactor fuel rods and/or assemblies in any of the inner metal containers of the RA-series shipping package described in NRC Certificate of Compliance Number 4986 at locations in the United States providing such locations minimize the severity of potential accident LICENSE SNM-1097 DATE 05/27/97 Page () DOCKET 70-1113 REVISION 0 1.14

.. _ ~ I i conditions to be no greater than those in the design bases for the containers during f) transportation.- Arrays may be constructed without limit to the number of containers so stored, except that each array shall be stacked to the smaller of 4 containers high or the i height demonstrated to comply with the criticality safety requirements of this license. i Each container must be separated by nominal 2-inch wooden studs, with the width i - and length for each array and separation between arrays determined only by container j handling requirements. Provisions for compliance with applicable 10 CFR 73 requirements are described in i the NRC-approved GE Wilmington Physical Security Plan as currently revised in 1-accordance with regulatory provisions. I Storage at nuclear reactor sites is subject to the financial protection and indenmity provision of10 CFW140. 1.3.8.4 Authorization to transfer, possess, use and store unirradiated reactor fuel of GE-Wilmington manufacture or procured to GE specification at nuclear reactor sites, for purposes ofinspection, fuel bundle disassembly and assembly, including fuel rod replacement, provided that the following conditions are met: A valid NRC license has been issued to the reactor licensee, which authorizes i e . O recei t. Possession and sterase of the ruei at the reactor site. GE Wiiminston P possesses the fuel only within the indemnified location. For dry fuel reconstitution, not more than 99 (9x9 lattices or greater) or 88 (8x8 lattices) unassembled fuel rods may be possessed by GE-Wilmington at l any one reactor site at any one time, except when the fuel has been packaged for transport or as described'in Section 1.3.8.3. The fuel rods must be of the types described in NRC Certificate of Compliance Number 4986. For underwater fuel reconstitution, not more than one fuel assembly plus l unassembled fuel rods so that the total number of rods, including the assembly, possessed by GE-Wilmington at any one reactor site at any one time does not exceed 99 (9x9 lattices or greater) or 88 (8x8 lattices), except when the fuel has been packaged for transport or as described in Section 1.3.8.3. The fuel rods must be of the types described in NRC Certificate of Compliance Number 4986. l i i f l LICENSE SNM-1997 DATE 05/27/97 Page Q DOCKET 70-1113 REVISION 0 1.15 r

l l Operations involving the fuel are conducted by or under the direct ( ) supervision of a member of the GE-Wilmington staff who shall be responsible for work on the fuel element assembly. The person shall comply with applicable reactor license and procedure requirements as directed by reactor site representatives, including appropriate actions that are to be taken in the event of emergencies at the site. Loose rods are stored in RA-series inner metal containers. 1 l Fuel is handled in accordance with pertinent provisions of the reactor license, and also in accordance with applicable GE-Wilmington procedures which are l jointly verified for completion by GE-Wilmington and the reactor licensee. Records of the operation, including the GE-Wilmington procedures used, are maintained at the GE-Wilmington facility. 1.3.9 WASTE OXIDATION-REDUCTION FACILITY Authorization to treat waste and scrap material containing special nuclear material by oxidation and reduction. 1.3.10 DILUTION FACTOR FOR AIRBORNE EFFLUENTS nU Authorization to utilize a dilution factor to the measured stack discharges for the purpose of evaluating the airbome radioactivity at the closest site boundary of stack discharges from the uranium processing facilities. For purposes of control, this dilution factor shall be no greater than 100. For other purposes, specific dilution factors, which consider dispersion model parameters, may be calculated and used. 1.3.11 CRITICALITY MONITORING SYSTEM Authorization that it is not necessary to maintain the criticality accident monitoring system requirements of 10 CFR 70.24 when it is demonstrated that a credible criticality risk does not exist for each area in which there is not more than: 1.3.11.1 A quantity of finished reactor fuel rods equal to or less than 45% of a minimum critical number under conditions in which double batching is credible, or equal to or LICENSE SNM-1097 DATE 05/27/97 Page C) DOCKET 70-1113 REVISION 0 1.16 v i l

less than 75% of a minimum critical number under conditions in which double {} batching is not credible, or 1.3.11.2 The quantity of uranium authorized for delivery to a carrier when fully packaged as for transport according to a valid NRC authorization for such packages without limit on the number of such packages, provided storage locations preclude mechanical damage and flooding, or 1.3.11.3 Arrays of finished reactor fuel rods and/or assemblies in any of the inner metal containers of the RA-series shipping package described in NRC Certificate of Compliance Number 4986, under storage conditions described in Section 1.3.8.3, or 1.3.11.4 Unassembled fuel rods under the restrictions and transfer, possession, use and storage conditions in Section 1.3.8.4. 1.3.12 POSTING Authorization to post areas within the Controlled Access Area in which radioactive materials are processed, used, or stored, with a sign stating "Every container in this 'bc area may contain radioactive material" in lieu of the labeling requirements of 10 CFR 20.1904. 1.3.13 EXTREMlTY DOSE DETERMINATION 2 Authorizatior to use a skin thickness of 38 milligrams /cm in the assessment of worker fingatip doses from uranium and for determining compliance to NRC extremite dose limits. 1.3.14 AUTHORIZED WORKPLACE AIR SAMPLING ADJUSTMENTS Authorization to adjust Derived Air Concentration (DAC) limits and Annual Limit of Intake (All) values in process areas to reflect chemical and physical characteristics of the airborne uranium. LICENSE SNM-1097 DATE 05/27/97 Page ( DOCKET 70-1113 REVISION 0 1.17 a

1 The DAC or ALI may be based on current biokinetic and dosimetric models i recommended in publications of recognized national and international organizations such as NCRP and ICRP. Examples of such model changes specifically include i changes in weighting factors for various organs and revisions / changes in the lung i model. 4 t i 4 i i l i O ) l 1 l l l LICENSE SNM-1097 DATE 05/27/97 Page i DOCKET 70-1113 REVISION 0 1.18

O 1 GUIDELINES FOR DECONTAMINATION OF FACILITIES AND EQUIPMENT PRIOR TO RELEASE FOR UNRESTRICTED USE OR TERMINATION OF LICENSES FOR BYPRODUCT, SOURCE, j OR SPECIAL N'UCLEAR MATERIAL j .O i U.S. Nuclear Regulatory Commission Division of Fuel Cycle Safety and Safeguards Washington, DC 20555 April 1993 LICENSE SNM-1997 DATE 05/27/97 Page l DOCKET 70-1113 REVISION 0 1.19

(s The instructions in this guide, in conjunction with Table 1, specify the radionuclides and radiation exposure rate limits which should be used in decontamination and survey of surfaces or premises and equipment prior to abandonment or release for unrestricted use. The limits in Table 1 do not apply to premises, equipment, or scrap containing induced radioactivity for which the radiological considerations pertinent to their use may be different. The release of such facilities or itetus from regulatory control is considered on a case-by-case basis. 1. The licensee shall make a reasonable effort to eliminate residual contamination. 2. Radioactivity on equipment or surfaces shall not be covered by paint, plating, or other covering material unless contamination levels, as determined by a survey and documented, are below the limits specified in Table 1 prior to the application of the covering. A reasonable effort must be made to minimize the contamination prior to use of any covering. 3. The radioactivity on the interior surfaces of pipes, drain lines, or ductwork shall be determined by making measurements at all traps, and other appropriate access points, provided that contamination at these locations is likely to be representative of contamination on the interior of the pipes, drain lines, or ductwork. Surfaces of premises, equipment, or scrap which are likely to be contaminated but are of such size, construction, or location as to make the surface inaccessible for purposes of measurement shall be presumed to be contaminated in excess of the limits. 4. Upon request, the Commission may authorize a licensee to relinquish possession or y1 control of premises, equipment, or scrap having surfaces contaminated with materials in excess of the limits specified. This may include, but would not be limited to, special circumstances such as razing of buildings, transfer of premises to another organization continuing work with radioactive materials, or conversion of facilities to a long-term storage or standby status. Such requests must: a. Provide detailed, specific information describing the premises, equipment or scrap, radioactive contaminants, and the nature, extent, and degree of residual surface contamination. b. Provide a detailed health and safety analysis which reflects that the residual amounts of materials on surface areas, together with other considerations such as prospective use of the premises, equipment, or scrap, are unlikely to result in an unreasonable risk to the health and safety of the public. LICENSE SNM-1097 DATE 05/27/97 Page DOCKET 70-1113 REVISION 0 1.20

i i O i 2-d t 4 5. Prior to release of premises for unrestricted use, the licensee shall make a comprehensive 4 radiation survey which establishes that contamination is within the limits specified in Table 1. A copy of the survey report shall be filed with the Division of Fuel Cycle Safety and Safeguards, U. S. Nuclear Regulatory Commission, Washingten, DC 20555, and also the Administrator of the NRC Regional Office havingjurisdiction. The report should be filed at least 30 days prior to the planned date of abandonment. The survey report shall: a. Identify the premises, b. Show that reasonable effort has been made. to eliminate residual contamination. c. Describe the scope of the survey and general procedures followed, d. State the findings of the survey in units specified in the instruction. Following review of the report, the NRC will consider visiting the facilities to confirm O the survey. ) i i l I I 1 LICENSE SNM-1097 DATE 05/27/97 Page i .O DOCKET 70-1113 REVISION ~ 0 1.21 V

~ TABLEl

O AccEerAatE sUnrAcE coxriMixArios tEvEts i

NUCtIDES' AVERAGE *f MAXIMUM REMOVABLE er 6 bdf b ) 2 2 2 U-nat, U-235, U-238, and 5,000 dpm ot/100 cm 15,000 dpm ot/100 cm 1,000 dpm at/100 cm associated decay products l 2 2 2 1 Transuranics, Ra-226, Ra-100 dpm/100 cm 300 dpm/100 cm 20 dpm/100 cm 228, Th-230, Th-228, Pa-231, Ac-227,1-125.1129 2 2 2 Th-nat, Th-232, Sr-90, Ra-1000 dpm/100 cm 3000 dpm/100 cm 200 dpm/100 cm j i 223, Ra-224, U-232,1-126, l-131,1-133 2 2 2 Beta-gamma emitters 5,000 dpm py/100 cm 15,000 dpm y/100 cm 1,000 dpm py/100 cm l (nuclides with decay modes t other than alpha emission or spontaneous fission) except Sr-90 and others noted above. i i "Where surface contamination by both alpha-and beta-gamma-emitting nuclides exists, the limits established for alpha-and beta-gamma-emitting nuclides should apply independently. b As used in this table, dpm (disintegrations per minute) means the rate of emission by radioactive material as l determined by correcting the counts per minute observed by an appropriate detector for background, efficiency, and i geometric factors associated with the instrumentation. ' Measurements of average contaminant should not be averaged over more than 1 square meter. For objects ofless surface area, the average should be derived for each such object. The maximum contamination level applies to an area of not more than 100 cm, ) d 2 i

• The amount of removable radioactive material per 100 cm of surface area should be determined by wiping that i

2 area with dry filter or soft absorbent paper, applying moderate pressure, and assessing the amount of radioactive material on the wipe with an appropriate instrument of known efficiency. When remavable contamination on objects ofless surface area is determined, the pertinent levels should be reduced proportionally and the entire surface should be wiped. 'The average and maximum radiation levels associated with surface contamination resulting from beta-gamma emitters should not exceed 0.2 mrad /hr at I cm and 1.0 mrad'hr at I cm, respectively, measured through not more than 7 milligrams per square centimeter of total absorber. LICENSE SNM-1997 DATE 05/27/97 Page DOCKET 70-1113 REVISION 0 1.22

m CH APTER 2.0 l V ORGANIZATION AND ADMINISTRATION 2.1 POLICY l The GE-Wilmington policy is to maintain a safe work place for its employees, to protect the environment, and to assure operational compliance within the terms and j conditions of special nuclear material licenses and applicable NRC regulations. 2.2 ORGANIZATIONAL RESPONSIBILITIES AND AUTHORITY 2.2.1 KEY POSITIONS WITH RESPONSIBILITIES IMPORTANT TO SAFETY (FIGURE 2.1) Responsibilities, authorities, and interrelationships among the GE-Wilmington organizational functions with responsibilities important to safety are specified in approved position descriptions and in documented and approved practices. 2.2.1.i GE.wiimingion raciiity Meneger O The GE-Wilmington facility manager is the individual who has overall responsibility i for safety and activities conducted at the GE-Wilmington facility. The GE-Wilmington facility manager directs operations by procedure, or through other management personnel. The activities of the GE-Wilmington facility manager are performed in accordance with GE policies, procedures, and management directives. The GE-Wilmington facility manager provides for safety and control of operations and protection of the environment by delegating and assigning responsibility to qualified Area Managers. The GE-Wilmington facility manager is knowledgeable of the safety program concepts as they apply to the overall safety of a nuclear facility, and has the authority to enforce the shutdown of any process or facility. LICENSE SNM 1097 DATE 05/27/97 Page () DOCKET 70-1113 REVISION 0 2.1

O Figure 2.1 l

GE-Wilmington Organization l GE Nuclear Energy ,e l Vice President s

)

General Manager d s.umn ward GE-Wilmington >} Environment, Health & Safety Fac.lity Manager y i g Function Manager l,I l A: +hns6maucad u mwAwawad Staff Manager 4 (Includes GE Criticairty Safety 1 O' Function Product Line Management) Area Radiation Safety Managers i Environmental i integrated Safety '--"~ Protection Functior Analysis and Configurat60n Management Function I Chemical and Fire Safety Function Site Securny and Emergency Preparedness Function LICENSE SNM-1097 DATE 05/27//97 Page O oocxer 'o.1113 aEvisioN o 2.2

0 - 2.2.1.2 Area Manager The Area Manager is the designated individual who is responsible for ensuring that activities necessary for safe operations and protection of the environment are . conducted properly within their designated area of the facility in which uranium materials are processed, handled or stored. Designated Area Manager j responsibilities include: Assure safe operation, maintenance and control of activities A'ssure safety of the environs as influenced by operations i e Assure performance ofintegrated safety analyses for the assigned facility i e area, as required i t Assure application of assurance elements to safety controls, as appropriate - I e Assure configuration control for safety controls for the assigned facility area, as required Use approved written operating procedures which incorporate safety controls ~ and limits Provide adequate operator training e The minimum qualifications of an Area Manager is a BS or BA degree in a technical ) O-tieid, and two years of experience in manufacturing operation, one of which is in i nuclear fuel manufacturing; or a high school diploma with five years of manufacturing experience, two of which are in nuclear fuel manufacturing. Area Managers shall be knowledgeable of the safety program procedures (including l chemical, radiological, criticality, fire, environmental and industrial safety) and shall have experience in the application of the program controls and requirements, as they ) relate to their areas of responsibility. The assignment ofindividuals to the position of Area Manager is approved by the GE-Wilmington facility manager, and the listing of Area Managers by area of responsibility is maintained current at the facility. .x 1 LICENSE SNM-1997 DATE 05/27//97 Page Q DOCKET-70-1113 REVISION 0 23 e

O 2.2.1.3 Integrated Safety Analysis and Configuration Management Function The integrated safety analysis and configuration management function is administratively part of the fuel production operations at GE-Wilmington. Designated responsibilities include: Establish and maintain the integrated safety analysis program Establish and maintain the assurance program for safety controls Provide advice and counsel to Area Managers on matters of the integrated e safety analysis program Establish and maintain the configuration control system for fuel manufacturing equipment and safety controls, and related record retention Establish and maintain the operating procedure systems e Minimum qualification requirements for the manager of the integrated safety analysis and configuration management function are a BS or BA degree in science or engineering and two years experience in related manufacturing assignments; or a high school diploma with eight years of manufacturing experience. The manager of I the integrated safety analysis and configuration management function shall have experience in the understanding and management of the assigned programs. 2.2.1.4 Criticality Safety Function The criticality safety function is administratively independent of production responsibilities and has the authority to shutdown potentially unsafe operations. Designated responsibilities include: Establish the criticality safety program including design criteria, procedures and training Provide criticality safety support for integrated safety analyses and configuration control e Assess normal and credible abnormal conditions Determine criticality safety limits for controlled parameters i i LICENSE SNM-1997 DATE 05/27//97 Page C DOCKET 70-1113 REVISION 0 2.4 l

h h Perform methods development and validation to support criticality safety e analyses Perform neutronics calculations, write criticality safety analyses and approve e , proposed changes in process conditions or equirnent involving fissionable material Specify criticality safety control requirements and functionality j e Provide advice and counsel to Area Managers on criticality safety control ) e measures, including review and approval of operating procedures l Support emergency response planning and events - e Assess the effectiveness of the criticality safety program through audit Programs 1 The criticality safety function manager shall hold a BS or BA degree in science or engineering, have at least four years experience in assignments involving regulatory I activities, and have experience in the understanding, application and direction of nuclear criticality safety programs. Minimum qualifications for a senior engineer within the criticality safety function are i a BS or BA degree in science or engineering with at least three years of nuclear l Q industry experience in criticality safety. A senior engineer shall have experience in the assigned safety function, and has authority and responsibility to conduct activities assigned to the criticality safety function. Minimum qualifications for an engineer within the criticality safety function are a BS/BA degree in science or engineering. An engineer shall have e:<perience in the l assigned safety function, and has authority and responsibility to conduct activities assigned to the criticality safety function, with the exception ofindependent 4 verification of criticality safety analyses. l 2.2.1.5 Radiation Safety Function The radiation safety function is administratively independent of production responsibilities and has the authority to shutdown potentially unsafe operations. Designated rerponsibilities include: ) 1 4 5 LICENSE SNM-1997 DATE 05/27//97 Page O oocxer 7a-"i3 aEvisios 0 2.s

O Estahiish the radiation erotection and radiation -onitorins erestams Establish the radiation protection design criteria, procedures and training e programs to control contamination and exposure to individuals Evaluate radiation exposures of employees and visitors, and ensure the e maintenance of related records Conduct radiation and contamination monitoring and control programs e Evaluate the integrity and reliability of radiation detection instruments e Provide radiation safety support for integrated safety analyses and e configuration control Provide analysis and approval of proposed changes in process conditions and e process equipment involving radiological safety Provide advice and counsel to Area Managers on matters of radiation safety e Support emergency response planning and events e Assess the effectiveness of the radiation safety program through audit e programs The radiation safety ftmetion manager shall hold a BS or BA degree in science or engineering, have at least two years experience in assignments that include responsibility for radiation safety, and have experience in the understanding, application and direction of radiation safety programs. Minimum qualifications for a senior member of the radiation safety function are a BS or BA degree in science or engineering with at least two years of nuclear industry experience in the assigned function. Alternate minimum experience qualification for a senior member of the radiation safety function is professional certification in health physics. A senior member shall have experience in the assigned safety function, and has authority and responsibility to conduct activities assigned to the radiation safety function. LICENSE SNM-1997 DATE 05/27//97 Page -O ooc'er 7a->>>3 evisio" a 2.6

O 2.2.i.6 Environmental Protection Function The environmental protection function is administratively independent of production responsibilities and has the authority to shutdown operations with potentially uncontrolled environmental conditions. Designated responsibilities include: Identify environmental protection requirements from federal, state and local regulations which govern the GE-Wilmington operation Establish systems and methods to measure and document adherence to regulatory environmental protection requirements and license conditions Provide advice and counsel to Area Managers e Evaluate and approve new, existing or revised equipment, processes and e procedures involving environmental protection activities Provide environmental protection support for integrated safety analyses and configuration control Assure proper federal and state permits, licenses and registrations for non-e radiological discharges from the facilities Minimum qualifications for the manager of the environmental r>rotection function are a BS or BA degree in science or engineering and two years of experience in O ssisameats iavoivias te8uiatorx etivities or eauiv ieat. 2.2.1.7 Chemical and Fire Safety Function The chemical and fire safety function is administratively independent of the production responsibilities and has the authority to shutdown operations with potentially hazardous health and safety conditions. Designated responsibilities include: Identify fire protection requirements from federal, state, and local regulations e which govern GE-Wilmington operations Develop practices regarding non-radiological chemical safety affecting j e nuclear activities j Provide advice and counsel to Area Managers on matters of chemical and fire ) e safety LICENSE SNM-1997 DATE 05/27//97 Page 4 O oocxer 'a->>>> nevisio" a 2.7 4

O Provide censuitation and revie-ef ae. existins or revised e9ui ment, P processes and procedures regarding chemical safety and fire protection Provide chemical and fire safety support for integrated safety analyses and e configuration control Minimum qualifications of the manager of the chemical and fire safety function are a, BS or BA degree in science or engineering and two years of experience in related assignments. 6 2.2.1.8 Site Security and Emergency Preparedness Function The site security and emergency preparedness function is administratively independent of the production responsibilities. Designated responsibilities include: j Provide physical security for the GE-Wilmington facility e Establish and maintain the emergency preparedness program, including. e training and program evaluations Provide advice and counsel to Area Managers on matters of physical security e and emergency preparedness Maintain agreements and pre,mredness with off-site emergency support e groups Minimum qualifications are a BS or BA degree in science or engineering, one year of experience in related assignments, or a high school diploma with eight years of l experience in related assignments. 2.2.I'.9 Environment, Health & Safety (EHS) Function The EHS function is administratively independent of production responsibilities but has the authority to enforce the shutdown of any process or facility in the event that controls for any aspect of safety are not assured. This function has designated overall responsibility to establish the radiation safety, criticality safety, environmental j protection, chemical safety, fire protection and emergency preparedness programs to i ensure compliance with federal, state and local regulations and laws governing operation of a nuclear manufacturing facility. These programs are designed to ensure LICENSE SNM-1997 DATE 05/27//97 Page Q DOCKET 70-1113 REVISION 0 2.8

i o I h ' the health and safety of employees and the public as well as protection of the environment. The managers of the criticality safety, radiation safety, environmental - protection, chemical and fire safety, and site security and emergency preparedness I functions report to the EHS function manager.' l ' The manager of the EHS function must hold a BS or BA degree in science or. engineering and have five years of management experience in assignments involving l regulatory activities. The manager of the EHS function must have appropriate l understanding of health physics, nuclear criticality safety, environmentul protection, and chemical and fire safety programs. ) 2.2.2 MANAGEMENT CONTROLS l 1 Management controls for the conduct and maintenance of the GE-Wilmington health, safety and environment protection programs are contained in documented plant ] practices _ described in Section 3.9.1, which are approved by cognizant management. i Such practices are part of a controlled document system, and appropriately span the organizational structure and major plant activities to control interrelationships,'and to specify program objectives, responsibilities and requirements. Personnel are appropriately trained to the requirements of these management controls, and compliance is monitored through intemal and independent audits and evaluations. c Management controls documented in practices address requirements including: Configuration Management o Integrated Safety Analysis Radiation Safety e e~ Criticality Safety e Environmental Protection Chemical Safety e Fire & Explosion Safety e Emergency Preparedness e Quality Assurance e Training i 1 LICENSE SNM-1997 DATE 05/27//97 Page DOCKET 70-1113 REVISION 0 2.9 i ) j

O() e Procedures e Maintenance e Audits Incident Investigation & Reporting e Fissile Material Accountability and Control e 2.3 -SAFETY COMMITTEES 2.3.1 WILMINGTON SAFETY REVIEW COMMITTEE - The functions of the Wilmington Safety Review Committee include responsibility for j the following: An annual ALARA review which considers: Programs and projects undertaken by the radiation safety function and e the Radiation Safety Committee Performance including, but not limited to, trends in airborne e O

• "**""'i "" ""di "*"v "* ""*' **" '"'*' ""d ""vi' ""*"t" monitoring results Programs for improving the effectiveness of equipment used for e

efiluent and exposure control Review of major changes in authorized plant activities which may affect e nuclear or non-nuclear safety practices Professional _ advice and counsel on environmental protection, and criticality, e radiation, chemical and fire safety issues affecting the nuclear activities. j The committee is responsible to the GE-Wilmington facility manager. Its proceedings, findings and recommendations are reported in writing to the GE-i Wilmington facility manager and to appropriate stafflevel managers responsible for operations which have been reviewed by the committee. Such reports shall be j retained for at least three years. _j l -l j LICENSE SNM-1997 DATE 05/27//97 Page i 'C DOCKET 70-1113 REVISION 0 2.10 1

-.-. ~ i i O The committee holds at least three meetings each calendar year with a maximum interval of 180 days between any two consecutive meetings. 2.3.2 RADIATION SAFETY COMMITTEE The objective of the Radiation Safety Committee is to maintain occupational radiation exposures as low as reasonably achievable (ALARA) through improvements in fuel manufacturing operations. The committee meets monthly to maintain a continual awareness of the status of projects, performance measurement and trends, and the current radiation safety l conditions of shop activities. The maximum interval between meetings does not - j exceed 60 days. A written report of each Radiation Safety Committee meeting is forwarded to cognizant Area Managers and the manager of the EHS function. Records of the committee proceedings are maintained for three years. The committee consists of managers or representatives from key manufacturing functions with activities affecting radiation safety. ,Oy ? 4 i a j ll 4 i c LICENSE SNM-1997 DATE 05/27//97 Page DOCKET 70-1113 REVISION 0 2.11

CHAPTER 3.0 g CONDUCT OF OPERATIONS 3.1 CONFIGURATION MANAGEMENT (CM) 3.1.1 CONFIGURATION MANAGEMENT PROGRAM A formal configuration management process, governed by written, approved practices, ensures that plant design changes do not adversely impact on safety, health, or environmental protection programs at GE-Wilmington. The configuration management program ensures that the information used to operate and maintain safety controls is kept current. Safety controls are systems, structures, components and procedures which prevent and/or mitigate the risk of accidents. The use of current plant information is an integral part of the integrated safety analysis program described in Chapter 4.0. The CM program includes the following activities: Maintenance of the design information for the plant Control ofinformation used to operate and maintain the plant .O Documentation of changes e Assurance of adequate safety reviews for changes Periodic comparison assessment of the conformance of specific safety e controls to the documentation of plant design bases 3.1.2 PLANT DESIGN REQUIREMENTS Written plant practices define the development, application, and maintenance of the design specifications and requirements. Plant design specifications and requirements are maintained as controlled information. The specific content of the information depends on the age of the design and the requirements in place at the time of design. As a minimum the information required for safe operation of the facility is available. LICENSE SNM-1097 DATE 05/27/97 Page h DOCKET 70-1113 REVISION 0 3.1

3.1.3 CHANGE CONTROL ,,() Written plant practices describe the configuration management program for change management, including approval to install and operate facility changes. Facility changes are assessed by a trained and approved safety reviewer to determine if the applicable ISA is impacted, and if further review and approval is required by an ISA team as described in Chapter 4.0. The written plant practices also prescribe controls and define the distinction between types of changes, ranging from replacement with identical designs which are authorized as part of normal maintenance, to new or different designs which require specified review and approval. 3.1.4 DOCUMENT CONTROL Documented plant practices define the control system, including creation, revision, storage, tracking, distribution and retrieval of applicable information including : Operating procedures Drawings Technical specifications and requirements e 3(J Software for safety controls e Calibration instructions e Functional test instructions e The documented plant practices describe the responsibilities and activities which maintain consistency between the facility design, the physical facility, and the documentation. They also describe how the latest approved revisions are made available for operations. 3.2 MAINTENANCE The purpose of planned and scheduled maintenance for safety controls is to assure that systems are kept in a condition of readiness to perform the planned and designed functions when required. Area Managers are responsible to assure the operational readiness of safety controls in their assigned facility areas. For this reason the maintenance function is administratively part of or closely coupled to fuel production I LICENSE SNM-1097 DATE 05/27/97 Page DOCKET 70 1113 REVISION 0 3.2 ( i l l

. ~ _ -,~ i operations. The maintenance function utilizes a systems-based program to plan, O scheduie,1 rack and maintain records for maintenance activities. Maintenance instructions are an integral part of the maintenance system for maintenance activities. Discrimination between specified safety controls and other systems based on integrated safety analyses is maintained in the database. Key maintenance requirements for safety controls such as calibration, functional testing, and j l replacement of specified components are derived from integrated safety analyses described in Chapter 4.0, and the application of the graded approach to assurance elements. Maintenance activities generally fall into the categories described below: 3.2.1 SCHEDULED PREVENTIVE MAINTENANCE Examples of safety controls included for scheduled preventive maintenance are : t e Radiation Measurement Instruments Criticality Detection Devices l Effluent Measurement & Control Devices e Emergency Power Generators i . Fire Detection and Control Systems e O Pressure Relief Valves 'l Air Compressors e Steam Boilers e 3.2.2 PERIODIC FUNCTIONAL TESTING \\ Examples of safety controls included for periodic functional testing include : Criticality Warning System i e Fire Alarm System e Specified Active Engineered Controls on Process Equipment l e l Frequencies and requirements for functional testing of various safety controls are l derived through quality and reliability activities using a graded approach to assurance i i LICENSE SNM-1997 DATE 05/27/97 Page O oocxer va->>>> anvisios a 3.3 4

as described in Section 3.3. The integrated safety analysis is the basis for this , b-implementation. I J i 3.2.3 REPAIR OF SAFETY CONTROLS j 1 The maintenance planning and control system provides documentation and records of ] systems and components which have been repaired or replaced. 1 ) When ' component of specified safety control is repaired or replaced, the component a o is functionally verified to assure that it has the capability to perform its planned and designed function when called upon to do so. If the performance of a repaired or replaced safety control could be different from that or the original component, the change to the safety control is specifically approved under the configuration management program and tested to assure it is likely to perform its desired function when called upon to do so. 3.3 QUALITY ASSURANCE (QA) i The application of assurance measures to safety controls at GE-Wilmington focuses ) on assuring that these controls are designed, installed, operated and maintained such that their planned function is not compromised, j .O. l 3.3.1 ASSURANCE ELEMENTS The following assurance elements are applied to safety controls at GE-Wilmington: Organization e Program e Equipment / System Design Control e Procurement Documentation Control e Instructions, Procedures, and Drawings e Document Control e Control of Purchased Materials, Equipment, and Services l-e f Identification and Control of Materials, Parts, and Components i l LICENSE SNM-1997 DATE 05/27/97 Page O oocxer va->>>> anvisiox a 3.4

1 l i Control of Special Processes e O_ InternalInspections Test Control l e Control of Measuring and Test Equipment Handling, Storage, and Shipping Controls Inspection, Test, and Operating Status j Control of Nonconforming Materials, Parts, or Components e Corrective Action e e Records Audits 1 3.3.2 ASSURANCE ELEMENT APPLICATION TO SAFETY CONTROLS In accordance with documented intemal practices, the assurance elements are applied to safety controls in proportion to their importance to safety, and as an integral part l of the Integrated Safety Assessment program described in Chapter 4.0. This graded l approach segregates safety controls and activities into three categories in applying the l assurance elements: g For safety controls intended to prevent or mitigate the consequences of the l l highest risk category, each of the assurance elements are specifically evaluated and applied to the control, and their application requirements documented as part of the ISA. Justification for each assurance element not i applicable to a control in this category is also documented. For safety controls intended to prevent or mitigate the consequences of the e mid-level risk category, each of the assurance elements is thoroughly i evaluated and applicable assurance elements and their requirements are ) i applied and documented. Safety Controls in the low risk category are operated and maintained as part j e of routine and prudent industry practice, and are controlled by means of normal, established manufacturing assurance systems. No extraordinary assurance element requirements are documented. i l i I LICENSE SNM-1997 DATE ' 05/27/97 Page . O oocxer va->>>> navisio" a 3.5 P

Assurance element requirements and application decisions are based on sound ,,(j engineering practices and judgment. Assurance element descriptions and application, are included in documented practices as part of the GE-Wilmington management system. These practices also specify the requirements for related record retention. 3.4 TRAINING AND QUALIFfCATION Training is provided for each individual at GE-Wilmington, commensurate with assigned duties. Training and qualification requirements are met prior to personnel j fully assuming the duties of safety-significant positions, and before assigned tasks are independently performed. Formal training relative to safety includes radiation and radioactive materials, risks involved in receiving low level radiation exposure in accordance with 10CFR19.12, basic criteria and practices for radiation protection, nuclear criticality safety principles not verbatim, but in general conformance with ANSl/ANS 8.20 guidance, chemical and fire safety, maintaining radiation exposures and radioactivity in effluents As Low As Reasonably Achievable (ALARA), and emergency response. The s) <cm established for maintaining records of training and retraining is described in Section 3.8. O LJ 3.4.1 NUCLEAR SAFETY TIMINING Training policy requires that employees complete formal nuclear safety training prior to unescorted access in the airbome radioactivity controlled area. Methods for evaluating the understanding and effectiveness of the training includes passing an initial examination covering formal training contents and observations of operational activities during scheduled audits and inspections. Such training is performed by trained instructors approved by the manager of the criticality safety function and the manager of the radiation safety function. Training program contents are reviewed on a scheduled basis by the manager of the criticality safety and radiation safety functions to ensure that training program contents are current and adequate. i Previously trained employees who are allowed unescorted access to the airborne radioactivity controlled area are retrained at least every two years. The effectiveness LICENSE SNM-1097 DATE 05/27/97 Page ) DOCKET 70-1113 REVISION 0 3.6

._ _ _._ _ ___.~.___ _ _ _ _ _ _. _ l of the training program is evaluated by an initial training exam.. Visitors are trained lO commensurate with the scoge of their visit andrer escorted e treined empierees. x i 1 3.4.2 OPERATOR TRAINING Operator training is performance based, and incorporates the structured elements of analysis, design, development, implementation, and evaluation. Job-specific training l ir.cludes applicable procedures and safety provisions, and requirements. Emphasis is placed on safety requirements where human actions are important to safety. Operator training and qualification requirements are met prior to process safety-related tasks being independertily performed or before startup following significant changes to safety controls. i 3.5 HUMAN FACTORS Human factors are an integral part of tne management and operational safety l philosophy at GE-Wilmington. The consideration of human factors in relation to l operational safety is included in integrated safety analyses. Human factors concepts are also considered in: Equipment design ! O-Safety control design Operator training e Maintenance e Audits and assessments l e Incident investigations e 3.6 AUDITS AND ASSESSMENTS l Planned and scheduled internal and independent audits are performed to evaluate the l application and effectiveness of management controls and implementation of I programs related to activities significant to plant safety. Written operating procedures are based on GE-Wilmington practices, applicable regulations and license conditions. Audits are performed to assure that operations are conducted in LICENSE SNM-1997 DATE 05/27/97 Page -O oocxer va->>>> anvisio" a 3.7

l accordance with the operating procedures, and to assure that safety programs {} reflected in the operating procedures are maintained. ~ 3.6.1 CRITICALITY, RADIATION, CHEMICAL AND FIRE SAFETY AUDITS Representatives of the criticality safety function, the radiological safety function, and the chemical and fire safety function conduct formal, scheduled safety audits of fuel manufacturing and support areas in accordance with documented, approved practices. 3 These audits are performed to determine that operations conform to criticality, radiation, and chemical and fire safety requirements. Criticality and radiological audits are performed quarterly (at intervals not to exceed 110 days) under the direction of the manager of the criticality safety function and the manager of the radiation safety function. Chemical and fire safety audits are performed under the direction of the chemical and fire safety function manager. Personnel performing audits do not report to the production organization and have no direct responsibility for the function and area being audited. Audit results are communicated in writing to the cognizant Area Manager and to the manager of the environment, health & safety function. Required corrective actions are documented and approved by the Area Manager, reported to the GE-Wilmington facility manager, and tracked to completion by the environment, health & safety function. /7 V i Radiation protection personnel within the radiation safety function conduct weekly l nuclear safety inspections of fuel manufacturing and support areas in accordance with documented procedures. Inspection findings are documented and sent to the affected Area Manager for resolution. Records of the audit or inspection, instructions and procedures, persons conducting the audits or inspections, audit or inspection results, and corrective actions for identified violations oflicense conditions are maintained in accordance with procedural requirements for a minimum period of three years. 3.6.2 ENVIRONMENTAL PROTECTION AUDITS An audit schedule of the environmental protection program is developed by the environmental protection function on an annual basis. Audits are conducted in LICENSE SNM-1097 DATE 05/27/97 Page n DOCKET 70-1113 REVISION 0 3.8 b i l l

.i l accordance with documented practices 60 ensure that operational activities conform l .O.. io documented environmentai re2eirements. 3 Personnel under the direction of the manager of the environmental protection j function perform the environmental protection' audits. Personnel performing the

l audits do_ not report to the production organization and have no direct responsibility.

i for the function and area being audited. j l Audit findings are communicated to the cognizant Area Manager, who is responsible 1 l-for nonconformance corrective action commitments in accordance with documented ) l practices. The manager of the environmental protection function or delegate is responsible for resolution follow-up for identified nonconformances. Audit results in the form of corrective action items are reported to the GE-Wilmington facility manager and staff for monitoring of closure status. j l l 3.6.3 INDEPENDENT AUDITS The GE-Wilmington safety piogram elements (radiation, criticality, ch'emical, fire protection, industrial safety and environmental protection) are audited biennially by appropriately trained and experienced individuals who have a degree of l independence of the GE-Wilmington organization, and are not involved in the routine performance of the work or program being audited. The scope of O' independent audits covers the adequacy of the safety program as well as compliance to requirements. Audit results are reported in writing to the GE-Wilmington facility manager, the Area Managers, the manager of the radiation safety function, and the manager of the criticality safety function, as appropriate. The safety function and/or Area Managers, as appropriate, take necessary response actions in accordance with documented corrective action commitments. Audit results in the form of corrective action items are reported to the GE-Wilmington facility manager and staff for tracking until closure. 3.7 INCIDENT INVESTIGATIONS Unusual events which potentially threaten or lessen the effectiveness of health, safety j or environmental protection are reviewed by the Area Manager and reported to the environment, health & safety function in accordance with documented practices and { i LICENSE SNM-1997 DATE 05/27/97 Page j h. DOCKET 70 1113 REVISION 0-3.9 { L 4 l

methods. Each event is considered in terms of reporting requirements in accordance oC) with applicable regulatory requirements. The depth ofinvestigation relates to the l i severity or potential severity of the event in judgment of such factors as levels of uranium released and/or the degree of potential for exposure of workers, the public or the environment. Documented incident investigation practices provide for: Formal and systematic analyses for determination of root cause(s) i Investigations by independent, qualified teams when warranted i Documented identification and tracking of corrective actions e Documentation and record retention for purposes of application of" lessons e leamed" The environment, health and safety function is responsible for maintaining a list of agencies to be notified, determining if a report to an agency is required, and for notifying the agency when required. This function has the responsibility for continuing communications with government agencies. 3.8 RECORDS MANAGEMENT Records appropriate to integrated safety analyses and the application of appropriate Uq assurance elements to resulting controls, criticality and radiation safety activities, training / retraining, occupational exposure of personnel to radiation, releases of radioactive materials to the environment, and other pertinent safety activities are maintained in such a manner as to demonstrate compliance with license conditions and regulations. Records ofintegrated safety analyses and results are retained during the conduct of the activities analyzed and for six months following cessation of such activities to which they apply or for a minimum of three years. Records of criticality safety analyses are maintained in sufficient detail and form to permit independent review and audit of the method of calculation and results. Such records are retained dming the conduct of the activities and for six months following cessation of such ac uties to which they apply or for a minimum of three years. Records associated with personnel radiation exposures are generated and retained in such a manner as to comply with the relevant requirements of 10 CFR 20. The l LICENSE SNM-1097 DATE 05/27/97 Page () DOCKET 70-1113 REVISION 0 3.10 l

~ l' i L j l l i r following additional radiation protection records will be maintained for at least three p- ,f years: Records of the safety review committee meetings Surveys of equipment for release to unrestricted areas e Instrument calibrations e Safety audits e l Personnel training and retraining l l Radiation work permits e Surface contamination surveys 1 e Concentrations of airborne radioactive material in the facility Radiological safety analyses ) 3 Records associated with the environmental protection activities described in Chapter j 10 are generated and retained :n such a manner as to comply with the relevant { l requirements of 10 CFR 20 and this license. l l i I 3.9 PROCEDURES ) lO Licensed material processing or activities will be conducted in accordance with l-properly issued and approved practices and procedures (P&P), plant practices or l operating procedures. i 3.9.1 PLANT PRACTICES Licensed material activities are conducted in accordance with management control L pograms described in administrative and general plant practices approved and issued l by cognizant management at a level appropriate to the scope of the practice. These documented practices direct and control activities across the manufacturing functions, and assign functional responsibilities and requirements for these activities. Management controls described in Chapter 2.0 are included in these practices. These practices are reviewed for updating at least every two years. LICENSE SNM-1997 DATE 05/27/97 Page h DOCKET 70 1113 REVISION 0 3.11 I t

3.9.2 OPERATING PROCEDURES k>I Area Managers are responsible to assure preparation of written, approved and issued operating procedures incorporating control and limitation requirements established by the criticality safety function, the radiation safety function, the environmental protection function and the chemical and fire safety function. Integrated safety analysis results as described in Chapter 4.0 are used to identify procedures necessary for human actions important to safety. Operating procedures are initiated and controlled within the guidelines of the configuration management system described in Section 3.1. Area Managers assure that operating procedures are made readily available in the work area and that operators are trained to the requirements of the procedures and that conformance is mandatory. Operators are trained to report inadequate procedttres, and/or the inability to follow procedures. Nuclear safety control procedure requirements for workers in uranium processing areas are incorporated into the appropriate operating, maintenance and test procedures in place for uranium processing operations. The safety program design requires the establishment and maintenance of documented procedures for environmental, health and safety limitations and requirements to govern the safety aspects of operations. Requirements for procedure control and approval authorities are documented. Procedure review for updating frequencies are as follows: qO Document Review Reviewing & Approving Frequency Functional Manager i Operating Procedures (ops) When Area Managerand AfTected t { Note: Nuclear Safety changed

• EHS Discipline (Radiation, Release / Requirement (NSR/R)

Criticality, Environmental, limitations and requirements Industrial *, or MC&A) l are incorporated into ops) Operating Procedures (ops) Every 3 Area Manager and Affected Years

• EHS Discipline (Radiation, Criticality, Environmental, Industrial *, or MC&A) l Nuclear Safety Instructions Every 2 Radiation & Criticality Safety (NSIs)

Years

• Environmental Protection Every 2 Environmental Protection Instructions (EPIs)

YearsA l LICENSE SNM-1997 DATE 05/27/97 Page p DOCKET 70-1113 REVISION 0 3.12 o i i

l l l

1) The safety awareness portions of these ops are reviewed and updated by the appropriate environment, health, and safety (EHS) discipline when warranted based on process related facility change requests.

2) Every 2 years means a maximum interval of 26 months.

p

3) Every 3 years means a maximum interval of 39 months

4) EHS Discipline - Industrial means normal worker safety, chemical safety, and fire and explosion protection.

j l Nuclear safety control procedure requirements for workers in uranium processing areas are incorporated into the appropriate operating, maintenance and test j procedures in place for uranium processing operations. t i i l I l I i i t l i t l l l l. l l l I LICENSE SNM-1997 DATE 05/27/97 Page O oocxEr 7o-iii3 aEvisioN o 3.ia e i l l t

l CHAPTER 4.0 l -Q INTEGRATED SAFETY ANALYSIS 4.I INTEGRATED SAFETY ANALYSIS 1 Integrated Safety Analysis (ISA)is the focal point for safety at GE-Wilmington. ISA is a process in which multifunctional teams analyze the hazards at the site to determine accident scenarios and risk, and ensure that controls are in place to prevent and/or mitigate accidents. The risk associated with an accident scenario is used to judge the level of ongoing assurance that is applied to controls which are in place to prevent the accident. The broad scope of the team's analysis includes criticality safety, radiological safety, environmental protection and industrial safety including chemical safety and fire protection. The accidem scenarios identified in the ISA are l reviewed by the appropriate safety functions to ensure that the plant continues to comply with site safety policy and regulatory limits. GE commits to establish and maintain the controls identified in the ISA and to provide an appropriate level of assurance to ensure their reliability. The ISA will be i maintained current through the configuration management process (Section 4.10). This program applies to the Dry Conversion Process (DCP) and other process areas ' {) as they become baselined using the ISA process. l l 4.2 SITE DESCRIPTION A general description of the site is included in Chapter 1.0. More detailed site ) information is included in the Environmental Report described in Chapter 10.0. The j i credible external events which are considered by the ISA teams are defined m an established written practice. j 4.3 FACILITY DESCRIPTION Safety-significant information describing the facility, including arrangement of buildings on the site, location with respect to the site boundary, and the facility's ability to withstand credible external events, is included in drawings and reports maintained under configuration management. LICENSE SNM-1097 DATE 05/27/97 Page l] DOCKET 70-1113 REVISION 0 4.1 I l l

4.4 PROCESS DESCRIPTION Processes covered by this license are summarized in Chapter 1.0. Detailed information concerning these processes is typically included in technical reports, nuclear safety analyses, operating procedures, Process & Instrumentation Drawings l (P&lDs), and other detailed process information, which is maintained under i configuration management. ) 4.5 PROCESS SAFETY INFORMATION Process technology information is gathered and maintained for future use by ISA teams. Technical reports, which typically include process chemistry, intended inventories, and safe upper and lower limits for process variables such as temperature, pressure, flow, and composition, are maintained under configuration management. Process equipment information is maintained in accurate condition through configuration management. Examples include P&lDs, materials of construction, electrical classification, ventilation system design, and safety systems including interlocks, detection, and suppression systems. Hazardous material information, including toxicity, permissible exposure limits, physical data, reactivity data, corrosivity data, and thermal and chemical stability O. data is avaiiadie to emP ovees and iS^ teams in the form of Materiai Saferv Date i Sheets (MSDS's), 4.6 TRAINING AND QUALIFICATIONS OF THE ISA TEAM ISAs are conducted by teams ofindividuals with diverse, pertinent knowledge and experience. The team members are chosen to provide operational and technical expertise in the study area, and appropriate safety expertise based on the hazards that are known to exist in the study area. The composition of the team is dermed in an established plant practice. 4.7 ISA METHODS The hazards in the facility are identified and analyzed using methodology that is widely accepted throughout the chemical industry. Examples of the methodology are 1 l LICENSE SNM-1997 DATE 05/27/97 Page l 'O DOCKET 70-1113 REVISION 0 4.2 G L

described in Guidelines for Hazard Evaluation Procedures, published by the Center f'i for Chemical Process Safety of the American Institute of Chemical Engineers (1992). Hazards are analyzed using established methods, for example: Preliminary Hazards Analysis What If/ Checklist Hazards and Operability Analysis Failure Mode and Effect e Fault Tree e Event Tree Human Reliability Analysis Procedural guidance is provided to the ISA teams in the form of a written plant practice that outlines the special treatment these methods mquire when applied to processes in the nuclear industry. Examples of this special treatment includes the consideration of criticality and radiological hazards. In this procedure, the teams are instructed to consider start-up, shutdown, upsets, and maintenance, in addition to normal operating conditions. Guidance is provided conceming the external events which must be considered in ISAs. The written plant practice also provides guidelines for ranking accident scenarios ( ) according to risk, that is, unmitigated consequence and likelihood. The team then ensures that the controls that prevent or mitigate accidents are of the appropriate quality and reliability. 4.8 RESULTS OF THE JSA The results of the ISA team's analysis are communicated in a summary report to appropriate levels of management. This report summarizes the elements that are I important to safety in the area studied. The lists of hazards and accident scenarios are compiled and maintained by the configuration management function. Guidance to the teams is provided in a written plant practice to ensure comprehensive reports. l ) l 1 ~ LICENSE SNM-1097 DATE 05/27/97 Page l 1 p DOCKET 70-1113 REVISION 0 4.3 v

4.9 CONTROLS FOR PREVENTION AND MITIGATION OF ACCIDENTS ') O Controls which are relied upon to prevent or mitigate serious accidents are maintained in a ready state through the application of a wide range of assurances. Examples of assurances typically used at GE-Wilmington include: configuration management, preventative maintenance, functional tests, quality assurance, purchasing specifications, training, procedures, audits, assessments and inspections. The level of assurance applied is consistent with the level of risk associated with the { specific accident scenario. Responsible risk management requires consideration of the components of risk, specifically consequences and likelihood. Accident scenarios are rated by the ISA teams in terms of unmitigated consequences and likelihood of an initiating event according to criteria defined in written plant practices. The general categories of consequences are defined as follows: the highest category is assigned to accidents that could result in injury to the public located outside the site boundary and to extreme on-site catastrophes. The middle level is assigned to accidents that would result in regulatory violations and/or serious on-site consequences. All other accidents are assigned to the lower level. These categories i are summarized in Table 4.1. o \\ LICENSE SNM-1997 DATE 05/27/97 Page DOCKET 70-1113 REVISION 0 4.4

1 \\, !V l Table 4.1 Consequence Levels Severity Radiological / Environmental / Ranking Criticality Industrial / Chemical 3 e exposure to an individual fatality e member of the public off-site medical treatment for a (5 rem,30 mg intake of U) member of the public off-severe exposure to an site e employee (400 rem internal permanent disability e plus external dose or 230 mg off-site contamination e intake of U) above regulatory standards 2 e exceed regulatory limits for serious injury e employee exposure (5 rem,10 exceed permit limits or e mg U internal) regulatory limits lost time injury e reportable release () I e exceed administrative limits OSHA recordable injury e on daily air samples, lung first aid e counts, bioassays, e exceed internal limits contamination, TLDs spillinside containment 10% of annual exposure limit UIR Accident scenarios are rated according to the likelihood of occurrence. The likelihood is categorized in qualitative terms that can easily be applied by the ISA teams. The highest category oflikelihood is applied to initiating events that could occur at any time in the irnmediate future. The middle category is for events that are likely to occur during the life of the operation. The lowest likelihood category is used for events that are not expected to occur during the life of the facility. In order to provide consistency in ranking, quantitative levels are provided as guidelines to the teams. These levels are summarized in Table 4.2. l LICENSE SNM-1097 DATE 05/27/97 Page ,o(j DOCKET 70-1113 REVISION 0 4.5 1

_~._._ ~_ -. l l O rani 4.2 l Likelihood Levels i 2 I i LEVEL FREOUENCY LIKELIHOOD 4 '3 more frequent than once every likely to occur in the immediate j two years future 2 every two to fifty years likely to occur during the life of the facility I less frequent than once every not likely to occur during the life fifty years of the facility 4 0 incredible likelihood is indistinguishable from zero i .The levels of consequence and likelihood are combined to estimate the level of risk ofinitiating a particular accident. Figure 4,1 demonstrates the risk assignment p matrix. This risk assignment is used by the teams to determine the level of assurance d that will be applied to the controls that protect against that particular accident,. I LICENSE SNM-1097 DATE 05/27/97 Page O oocxer va->>>> nevisio" a 46

Figure 4.1 Risk Assignment Matrix C o 3 Mid-level n Risk S e q 2 Low Risk Mid-level u . Risk e n c 1 Low Risk Low Risk Mid-level e Risk 1 2 3 Likelihood Controls that prevent or mitigate events in the highest risk category receive full evaluation and appropriate application of all assurance elements defined in Chapter 3.0. Appropriate assurance elements are applied to mid-level risk controls. Low risk 1 controls are treated with normal, prudent attention. 4.10 ADMINISTRATIVE CONTROL OF TIIE ISA The ISA is maintained current through a configuration management piogram that ensures that: 1) facility changes receive adequate integrated safety review, and 2) changes are adequately documented. Proposed facility changes are reviewed by a trained and approved integrated safety reviewer to determine if the change impacts the existing ISA. If so, an ISA team is assembled, and the change is analyzed. The results of the ISA and the recommendations of the team are used in approving or rejecting the proposed change. After the change is implemented, the revised ISA becomes a part of the controlled documentation for the facility. The trained and approved integrated safety reviewer possesses the experience, I training and skills to consider criticality, radiological, environmental, chemical, and industrial impact within a predefined set oflimits. The reviewer is approved by the ~ LICENSE SNM-1097 DATE 05/27/97 Page DOCKET 70-1113 REVISION 0 4.7

f manager of the EHS function and reports organizationaily to the manufacturing O greauct iine. This ersanizationai structure sives ewnershig ef ererationei sefety te the manufacturing function. l l 1 I l O' i i 1 1 1 1 l l I LICENSE SNM-1997 DATE 05/27/97 Page ( DOCKET 70-1113 REVISION 0 4.8 i l

. - -.. ~ -.. - ~ ... ~. -. - _ l ) i 7'. CHAPTER 6.0 i ) NUCLEAR CRITICALITY SAFETY. l L 6.1 PROGRAM ADMINISTRATION l i l t 6.1.1 CRITICALITY SAFETY DESIGN PHILOSOPHY l The Double Contingency Principle as identified in nationally recognized American i National Standard ANSI /ANS-8.1 (1983) is the fundamental technical basis for l design and operation of processes within the GE-Wilmington fuel manufacturing ) operations using fissile materials. As such," process designs will incorporate L sufficient margins of safety to require at least two unlikely, independent, and L concurrent changes in process conditions before a criticality accident is possible." For each significant portion of the process, a defense of one or more system parameters is documented in the criticality safety analysis, which is reviewed and l enforced. l The established design criteria and nuclear criticality safety reviews are applicable to: all new processes, facilities or equipment that process, store, transfer or l.O e l etherwise handie rissiie materiais. and l any change in processes, facilities or equipment which may have an impact e on the established basis for nuclear criticality safety. l 6.1.2 EVALUATION OF CRITICALITY SAFETY 6.1.2.1 Changes to Facility As part of the design of new facilities or significant additions or changes in existing facilities, Area Managers provide for the evaluation of nuclear hazards, chemical l-hazards, hydrogenous content of firefighting materials, and mitigation ofinadvertent unsafe acts by individuals. Specifically, when criticality safety considerations are impacted by these hazards, the approval to operate new facilities or make significant L changes, modification, or additions to existing facilities is documented in accord. j l l l LICENSE SNM-1997 DATE 5/27/97 Page DOCKET 70-1113 REVISION 0 6.1 l J I

with established facility practices and conform to configuration management function ' Integrated Safety Analysis' (ISA) require.ments described in Chapter 4.0. Change requests are processed in accordance with configuration management requirements described in Chapter 3.0. Change requests which establish or involve a change in existing criticality safety parameters require a senior engineer who has been approved by the criticality safety function to disposition the proposed change with respect to the need for a criticality safety analysis. If an analysis is required, the change is not placed into operation until the criticality safety analysis is complete and other preoperational requirements are fulfilled in accordance with established configuration management practices. 6.1.2.2 Role of the Criticality Safety Function Qualified personnel as described in Chapter 2 assigned to the criticality safety function determine the basis for safety for processing fissile material. Assessing both normal and credible abnormal conditions, criticality safety personnel specify functional requirements for criticality safety controls commensurate with design criteria and assess control reliability. Responsibilities of the criticality safety function are described in Chspter 2.0. O 6.i.3 orna^rixa raoCsouimS Procedures that govern the handling of enriched uranium are reviewed and approved ) by the criticality safety function. I sach Area Manager is responsible for developing and maintaining operating procedures that incorporate limits and controls established by the criticality safety function. Area Managers assure that appropriate area engineers, operators, and other concerned personnel review and understand these procedures through postings, training programs, and/or other written, electronic or verbal notifications. Documentation of the review, approval and operator orientation process is maintained within the configuration management system. Specific details of this system are described in Chapter 3.0. LICENSE SNM-1097 DATE S/27/97 Page r DOCKET 70-1113 REVISION 0 6.2 (

y a 6.1.4 POSTING AND LABELING i 6.1.4.1 Posting of Limits and Controls i Nuclear criticality safety requirements for each process system that are defined by the criticality safety function are made available to work stations in the form of written i or electronic operating procedures, and/or clear visible postings. Posting may refer to the placement of signs or marking of floor areas to summarize key criticality safety requirements and limits, to designate approved work and storage l~ areas, or to provide instructions or specific precautions to personnel such as: t i Limits on material types and forms. e Allowable quantities by weight or number. e Allowable enrichments. e Required spacing between units. e t Control limits (when applicable) on quantities such as moderation, density, or q presence of additives. Critical control steps in the operation. e ) ) j Storage postings are located in conspicuous places and include as appropriate: O Material type. e Container identification. ] e Number ofitems allowed. i Mass, volume, moderation, and/or spacing limits. e 1 Additionally, when administrative controls or specific actions / decisions by operators I I are involved, postings include pertinent requirements identified within the criticality safety analysis. 6.1.4.2 Labeling Where practical, process containers of fissile material are labeled such that the material type, U-235 enriclunent, and gross weights can be clearly identified or determined. Deviations from this process include: large process vessels, fuel rods, shipping containers, waste boxes / drums, contaminated items, UFe, cylinders l l LICENSE SNM-1997 DATE 5/27/97 Page DOCKET 70-1113 REVISION 0 6.3 O i i

1 1 1 containing heels, cold trap cylinders, samples, containers of I liter volume or less, or O other containers where labeling is not practical. 6.1.5 AUDITS & INSPECTIONS 6.1.5.1 Audits and Inspections Details of the facility criticality safety audit progmm are described in Chapter 3.0. Criticality safety audits are conducted and documented in accordance with a written procedure and personnel approved by the criticality safety function. Findings, recommendations, and observations are reviewed with the Environment, Health & Safety (EHS) function manager to determine if other safety impacts exist. The i findings, recommendations, and observations are then transmitted to Area Managers i for appropriate action. l Routine surveillance inspections of the processes and associated conduct of operations within the facility, including compliance with operating procedures, postings, and administrative guidelines, are also conducted as described in Chapter 3. 6.1.5.2 Independent Audits b7-A nuclear criticality safety program review is conducted on a planned scheduled basis by nuclear criticality safety professionals independent of the GE Wilmington fuel manufacturing organization. This provides a means for independently assessing the effectiveness of the components of the nuclear criticality safety program. The audit team is composed ofindividuals recommended by the manager of the criticality safety function and whose audit qualifications are approved by the GE-Wilmington facility manager or Manager, EHS. Audit results are reported in writing to the manager of the criticality safety function, who disseminates the report to line management. Results in the form of corrective action requests are tracked to closure. LICENSE SNM-1097 DATE 5/27/97 Page (3 DOCKET 70-1113 REVISION 0 6.4 v

6.1.6 CRITICALITY SAFETY PERSONNEL U 6.1.6.1 Qualifications Specific details of the criticality safety function re:;ponsibilities and qualification requirements for manager, senior engineer, and engineer are described in Chapter 2.0. 6.1.6.2 Authority Criticality safety function personnel are specifically authorized to perfonn assigned responsibilities in Chapter 2.0. All nuclear criticality safety function personnel have authority to shutdown potentially unsafe operations. 6.2 TECIINICAL PRACTICES 6.2.1 CONTROL PRACTICES A() Criticality safety analyses identify specific controls necessary for the safe and effective operation of a process. Prior to use in any process, nuclear criticality safety controls are verified against criticality safety analysis criteria. The ISA program described in Chapter 4.0 implement performance based management of process requirements and specifications that are important to nuclear criticality safety. 6.2.1.1 Verification Program The purpose of the verification program is to assure that the controls selected and installed fulfill the requirements identified in the criticality safety analyses. All processes are examined in the "as-built" condition to validate the safety design and to verify the installation. Criticality safety function personnel observe or monitor the performance ofinitial functional tests and conduct pre-operational audits to verify that the controls function as intended and the installed configuration agrees with the criticality safety analysis. 1 LICENSE SNM-1097 DATE 5/27/97 Page DOCKET 70-1113 REVISION 0 6.5 l l l

Operations personnel are responsible for subsequent verification of controls through O" the use of functional testing or verification. When necessary, control calibration and routine maintenance are normally provided by the instrument and calibration and/or maintenance functions. Verification and maintenance activities are performed per established facility practices documented through the use of forms and/or computer tracking systems. Criticality safety function personnel randomly review control verifications and maintenance activities to assure that controls remain effective. 6.2.1.2 Maintenance Program The purpose of the maintenance program is to assure that the effectiveness of criticality safety controls designated for a specific process are maintained at the original level ofintent and functionality. This requires a combination of routine maintenance, functional testing, and verification of design specifications on a periodic basis. Details of the maintenance program are described in Chapter 3.0. 6.2.2 MEANS OF CONTROL The relative cirectiveness and reliability of controls are considered during the criticality safety analysis process. Passive engineered controls are preferred over all other system controls and are utilized when practical and appropriate. Active (l engineered controls are the next preferred method of control followed by administrative controls. A criticality safety control must be capable of preventing a criticality accident independent of the operation or failure of any other criticality control for a given credible initiating event. 6.2.2.1 Passive Engineered Controls These are physical restraints or features that maintain criticality safety in a static manner (i.e., fixed geometry, fixed spacing, fixed size, nuclear poisons, etc.). Passive engineered controls require no action or other response to be effective when called upon to ensure nuclear criticality safety. Assurance is maintained through specific periodic inspections or verification measurement (s) as appropriate. 6.2.2.2 Active Engineered Controls LICENSE SNM-1097 DATE 5/27/97 Page (l DOCKET 70-1113 REVISION 0 6.6 v

.- -.,~ l I l 3 t l A means of criticality control involving active hardware (e.g., electrical, mechanical, !{ j hydraulic) that protect against criticality. These devices act by providing predefined l automatic action or by sensing a process variable important to criticality safety and providing automatic action (e.g., no human intervention required) to secure the L system to a safe condition. Human intervention augmented by warning devices and l [~ ' interlocks that prevent continued operation may be used to sense a process variable. j i Assurance is maintained through specific periodic functional testing as appropriate. l Active engineered controls are fail-safe (e.g., meaning failure of the control results in -{ L a safe condition). i l 6.2.2.3 Administrative Controls x i l Controls that rely for their implementation on actions, judgment, and responsible - l actions ofpeople. Their use is limited to situations where passive and active control ? are not practical. Administrative controls may be proactive (requiring action prior to proceeding) or reactive (proceeding unless action occurs). Proactive administrative controls are preferred. Assurance is maintained through training, experience, and ] audit. [; 'l L 6.2.3 TABLE OF PLANT SYSTEMS AND PARAMETER CONTROLS l-O Table 6 identifies m8j r Process areas or support facility processes within the GE-L~ Wilmington fuel manufacturing complex and support facilities. Table entries for ] L each significant p'rocess item highlight the safety basis selected for the criticality i I safety analysis (CSA) and related worst credible contents (or bounding assumptions). i Table column definitions are presented below: AREA OR SYSTEM: A defined functional group of processes or pieces of equipment that operate as a single unit. PROCESS SUBAREA OR EQUIPMENT: A defined subgroup of vessels, tanks, l process and/or support equipment within an area that operate as a single unit. ( BASIS FOR CRITICALITY SAFETY: The controlled parameters established i within a CSA for nuclear criticality safety for the identified process subarea or equipment. For multiple parameter entries, the basis for nuclear criticality safety established in the CSA may be based on the identified parameter (s), as appropriate, l including the use of ' coupled' parameter control (e.g., mass / moderation). i LICENSE SNM-1997 DATE 5/27/97 Page DOCKET 70-1113 REVISION 0 6.7 i l

m. l l NOTE-To be included as section 1.3.15 infinal License: Changesfrom oneparameter to anotherparameterforprocess subareas or equipment in which multiple (at least two) parameters are controlled are made in accordance with established change control measures and reported to the NRC within 90 days ofcompletion. Changes to single parameter controlledprocesses or equipmentfrom the identspedparameter to a newparameter(s) will require NRC approvalprior to the change being made. A CSA BOUNDING ASSUMPTIONS: These are the values used for physical process parameters which are not directly controlled but represent the most reactive credible j . values for the system, process subarea, or equipment under consideration. As such, ' the CSA is performed to consider all process operations and credible upsets that fall j within this range of assumptions. For items containing no bounding assumptions, all process operations and credible upsets must be analyzed within the CSA. The approved CSA may limit the operation of the system to levels more conservative _ - - than those permitted by the bounding assumptions. i In the following Table 6.0, unless otherwise specified, the enrichment limit for all processes are 5.0 wt. % U235 (or hie), with the exception of conversion lines 1,2, and 4 and related MSG lines 1-6 which are presently analyzed for 4.025 wt. % U235 (or LoE). When pails are used for product,5-gallon cans may be used for LoE enrichments, while 3-gallon containers may be used for hie material. All scrap materialis treated as hie. 1 i O. .1- ) j i 1 LICENSE SNM-1997 DATE 5/27/97 Page D-DOCKET 70-1113 REVISION 0 6.8 d.. !/

i i i t Table 6.0 Plant Systems and Parameter Controls j l AREA PROCESS BASIS FOR CSA l OR SUBAREA OR CRITICALITY BOUNDING ASSUMPTIONS i SYSTEM EQUIPMENT SAFETY Fuel Support: UF Cylinder Receipt Enrichment 99.5 wt. % pure UF 6 6 Storage Pads and Storage s 0.5 wt. % H O equivalent 2 l OptimalInterunit H O 2 l Scrap 3 and 5-gallon Geometry Homogeneous or Heterogeneous UO2 l Container Storage Mass Optimal H O Moderation 2 l Full Reflection RA-Inner and Outer Geometry Heterogeneous UO2 Container Storage Moderation Optimal 110 Moderation 2 l Full Reflection j l Waste Box Container Geometry / Mass Homogeneous UO2 Storage Mass Optimal 110 Moderation 2 l Full Reflection BU-1, BU-7,7A Drum Geometry Homogeneous or Heterog:neous UO2 Storage Man Optimal H O Moderation 2 Moderation Full Reflection Fuel Support: Waste Box Load Mass Heterogeneous UO2 New Decon Optimal H O Moderation 2 Full Reflection Oil Drum Load Mass Homogeneous UO2 Optimal H O Moderation 2 Full Reflection Chemical ADU UF Cylinders Moderation 99.5 wt. % pure UF6 Conversion System 5 0.5 wt. % H O equivalent 2 Full Reflection Autoclave Moderation 99.5 wt. % pure UF6 Vaporization s 0.5 wt. % H O equivalent 2 Full Reflection Cold Trap System Geometry Homogeneous UO2 l Moderation Optimal H O Moderation 2 l Full Reflection [ liydrolysis Receiver, Geometry Homogeneous UO F22 Storage, and Scrubber Concentration Optimal H O Moderation 2 l Tanks Full Reflection l l Sump Geometry Homogeneous UO2 l Mass Optimal H O Moderation 2 Full Reflection Precipitation Tanks Geometry Homogeneous UO2 l (lines 1,2,4) Optimal H O Moderation 2 l Full Reflection

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

l l l LICENSE SNM-1997 DATE 5/27/97 Page DOCKET 70-1113 REVISION 0 6.9 i l i

l l C { a f AREA PROCESS BASIS FOR CSA l OR SUBAREA OR CRITICALITY BOUNDING ASSUMPTIONS SYSTEM EQUIPMENT SAFETY l Precipitation Tanks Geometry Homogeneous UO2 l (Lines 3,5) Mass Optimal H O Moderation [ 2 Full Reflection j Dewatering Geometry Homogeneous ADU or U 0 l 3 Centrifugation Mass Optimal H O Moderation 2 Full Reflection Outside Containment { j-Clarifying Geometry Homogeneous UO2 ? t Optimal H O Moderation l Centrifugation Mass 2 Full Reflection i f Calcination Geometry Homogeneous UO2 Geometry / Mass Optimal H O Moderation 2 Full Reflection l Calciner Scrubber Geometry Homogeneous UO2 l Concentration Optimal H O Moderation 2 Full Reficction 3 or 5-Gallon Product Geometry Homogeneous UO2 Container Mass Optimal H O Moderation i 2 Full Reflection i UO Powder Geometry or Mass Homogeneous UO2 i 2 l Pretreatment: Mill, Moderation Optimal H O Moderation 2 j Slug, Granulate (MSG) Full Reflection l-LoE and hie UO2 Geometry Homogeneous UO2 Powder Blending Mass / Moderation Optimal H O Moderation 2 Full Reflection O LoE Fluoride Efiluent Geometry Homogeneous UO2 Vessels Concentration Optimal H O Moderation 2 Full Reflection Line 3 Geometry Homogeneous UO2 Accumulator / Permeate Concentration Optimal H O Moderation 2 Vessels Full Reflection Nitrate Quarantine Geometry Homogeneous UO2 Efiluent Vessels Concentration OptimalH O Moderation 2 i Full Reflection Powder Pack Geometry Homogeneous UO2 Screener Moderation Optimal H O Moderation 2 Full Reflection Powder Pack Geometry Homogeneous UO2 l Product Container Mass Optimal H O Moderation 2 Full Reflection HVAC: Wet Areas Geometry Homogeneous UO2 i Mass Optimal H O Moderation ] 2 j Full Reflection l LICENSE SNM-1997 DATE 5/27/97 Page ( t i{ j DOCKET 70-1113 REVISION 0 6.10 1 -y- ,,y-23, y w-y,-- 7- .r- -,-3 y-

._m m._ .m AREA PROCESS BASIS FOR CSA OR SUBAREA OR CRITICALITY BOUNDING ASSUMPTIONS f SYSTEM EQUIPMENT SAFETY l HVAC: Dry Areas Mass 11omogeneous UO2 Moderation Optimal H O Moderation 2 Full Reflection Exhaust Scrubber Geometry / Mass llomogeneous UO2 Optimal H O Moderation Mass 2 f. Full Reflection Utilities: Steam, N, Mass Backflow into large supply vessels 2 H2, Dissoc. NH4, H O prevented by backflow prevention 2 Supply measures, physical barriers, and/or l process characteristics. REDCAP: Oxidation Geometry Heterogeneous UO2 Feed Containers Mass Optimal H O Moderation 2 Full Reflection i I REDCAP: Oxidation Geometry lieterogeneous UO2 Furnace Moderation Optimal H O Moderation 2 l Full Reflection REDCAP: Oxidation Geometry Homogeneous UO2 l Output Containers Mass Optimal H O Moderation 2 Full Reflection REDCAP: Oxidation Geometry Homogeneous UO2 Optimal H O Moderation I Off-Gas System Mass 2 Full Reflection l Miscellaneous: 3 and Geometry Homogeneous or Heterogeneous UO2 D 5-Gallon Container Mass Optimal H O Moderation 2 Floor storage Full Reflection Integration Geometry Heterogeneous UO2 OXIDIZE 3 and 5-gal. Mass Optimal H O Moderation 2 Feed Containers Full Reflection Integration Geometry } Heterogeneous UO2 OXIDIZE 3 and 5-gal. Mass f* OptimalInterunit H O Moderation 2 i l Feed Container Storage Moderation Full Reflection l Integration: Geometry Homogeneous or Heterogeneous UO2 l OXIDIZE Mass Optimal H O Moderation 2 ( Feed Hood Full Reflection Integration Geometry Heterogeneous UO2 l OXIDIZE Moderation Optimal H O Moderation 2 Furnace Full Reflection Integration Moderation het:rogeneous UO2 RECYCLE Maximum Credible wt. % H O 2 Powder Outlet Full Reflection 1

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

i LICENSE

• SNM-1097 DATE 5/27/97 Page O

oocx81 'a->>>> xevisio" a 6" L l l

c AREA PROCESS BASIS FOR CSA / J OR SUBAREA OR CRITICALITY BOUNDING ASSUMPTIONS SYSTEM EQUIPMENT SAFETY l Integration Moderation IIeterogeneous UO2 l RECYCLE Maximum Credible wt. % H O 2 l Blender Full Reflection Integration Moderation lieterogeneous UO2 l RECYCLE Mass Maximum Credible wt. % H O 2 l DM-10 Vibromill Full Reflection Integration Moderation Heterogeneous UO2 RECYCLE Unicone Maximum Credible UO Density 2 Container Storage Maximum Credible wt. % H O 2 OptimalInterunit H O 2 Ir.tegration Geometry Heterogeneous UO2 l RECYCLE 3-gal. Mass Optimal Interunit H O Moderation 2 Product Container Moderation Full Reflection Storage Integration Moderation Heterogeneous UO2 RECYCLE Maximum Credible UO Density 2 l Powder Transfer Maximum Credible wt. % H O 2 Corridor Full Reflection Uranium Recovery Unit Fluoride Waste Process Geometry Homogeneous UO2 (URU) System Vessels Concentration Optimal H O Moderation 2 Full Reflection Fluoride Waste Concentration Homogeneous UO2 Surge Vessel Mass Optimal 110 Moderation 2 (V-106) Full Reflection Radwaste Process Geometry Homogeneous UO2 s Optimal H O Moderation Vessels Concentration 2 Full Reflection Nitrate Waste Process Geometry Homogeneous UO2 Vessels Concentration Optimal H O Moderation 2 Full Reflection j Nitrate Waste Concentration Homogeneous UO2 Optimal H O Moderation Surge Vessel Mass 2 (V-103) Full Reflection Oxidation Feed Geometry Heterogeneous UO2 Optimal H O Moderation i Containers Mass 2 l Full Reflection l Oxidation Furnace Geometry Heterogeneous UO2 Optimal H O Moderation 2 Full Reflection l Oxidation Furnace Geometry Heterogeneous UO2 Boat Dump Moderation Optimal H O Moderation 2 l Full Reflection

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

LICENSE SNM-1097 DATE 5/27/97 Page {G DOCKET 70-1113 REVISION 0 6.12 s l

AREA PROCESS BASIS FOR CSA ( OR SUBAREA OR CRITICALITY BOUNDING ASSUMPTIONS V SYSTEM EQUIPMENT SAFETY Oxidation 3-gallon Geometry lieterogeneous UO2 Container Storage Mass Optimal 110 Moderation 2 Moderation Full Reflection Oxidation Off-Gas Geometry Heterogeneous UO2 System Mass Optimal H O Moderation 2 Full Reflection Dissolution: Can Geometry lieterogeneous UO2 Dump Feed Conveyor Mass Optimal 110 Moderation 2 Moderation Full Reflection Dissolution: Geometry Heterogeneous UO2 Dissolvers, Pumps, Concentration Optimal 110 Moderation 2 Sumps, Filters, Piping Full Reflection Oberlin Filter Geometry lieterogeneous UO2 Concentration Optimal H O Moderation 2 Full Reflection Dissolution: NOX Concentration Homogeneous UO2 Scrubber Mass On-Line Density Meter Full Reflection Counter-Current Geometry Heterogeneous UO2 Leaching: Can Dump Mass / Moderation Optimal H O Moderation 2 Full Reflection Counter-Current Geometry Heterogeneous UO2 q Leaching: Leach Concentration Optimal H O Moderation 2 V Troughs, Pumps, Full Reflection Filters, Storage Tanks, Product Containers Utilities: Steam, DI Mass Backflow into large supply vessels H 0, Nitric Acid, prevented by backflow prevention 2 Aluminum Nitrate measures, physical barriers, and/or process characteristics. Head-End Geometry llomogeneous UNH Concentrator Process Concentration Optimal 110 Moderation 2 Full Reflection Solvent Extraction Geometry Homogeneous UO2 Process Concentration Optimal H O Moderation 2 Full Reflection UNH Product Storage Geometry Homogeneous UNH Vessels Concentration Optimal H O Moderation 2 Full Reflection

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

i LICENSE SNM-1097 DATE 5/27/97 Page (] DOCKET 70-1113 REVISION 0 6.13

AREA PROCESS BASIS FOR CSA [] OR SUBAREA OR CRITICALITY BOUNDING ASSUMPTIONS I k SYSTEM EQUIPMENT SAFETY j Waste Solvent Drum Mass Homogeneous UO2 Load Optimal H O Moderation 2 j Full Reflection Uranyl Nitrate UNH LEM Tank Feed Geometry Homogeneous UO2 Conversion (UCON) Tanks Concentration Optimal H O Moderation 2 System Full Reflection UCON: Precipitation Geometry Homogeneous UNH l Tanks Mass Optimal H O Moderation 2 Full Reflection j UCON: Dewatering Geometry Homogeneous ADU or U 0 3 Centrifugation Mass Optimal H O Moderation 2 Full Reflection Outside Containment UCON: Clarifying Geometry Homogeneous UO2 Centrifugation Mass Optimal H O Moderation 2 Full Reflection UCON Process: Geometry Homogeneous UO2 Calcination Geometry / Mass Optimal H O Moderation 2 Full Reflection Waste Treatment Fluoride Waste Concentration Homogeneous UO2 Facility (WTF) Barrens Surge Vessel Mass Optimal H O Moderation 2 (V-108) Full Reflection Nitrate Waste Barrens Concentration Homogeneous UO2 Surp Vessel (V-104) Mass Optimal H O Moderation 2 p Full Reflection d Centrifuge Geometry Homogeneous UO2 Mass Optimal H O Moderation 2 Full Reflection Oberlin Filter Geometry / Mass Homogeneous UO2 Concentration Optimal H O Moderation 2 Full Reflection Uranium Recovery from URLS Process Tanks Concentration Homogeneous UO2 Lagoon Sludge (URLS) Optimal H O Moderation 2 Facility Process Full Reflection URLS Process Non-Geometry /Concent. Homogeneous UO2 Leach Filter Press Concentration Optimal 110 Moderation 2 Full Reflection URLS Process Product Concentration Homogeneous UO2 Waste Container Mass Optimal H O Moderation 2 Full Reflection Waste Oxidation / Incinerator Mass (Box Monitor) Heterogeneous UO2 Reduction (Incineration) Combustible Box Feed Mass (E-Gun) Optimal H O Moderation 2 Facility Containers Full Reflection LICENSE SNM-1097 DATE 5/27/97 Page DOCKET 70-1113 REVISION 0 6.14

I AREA-PROCESS BASIS FOR CSA l OR SUBAREA OR CRITICALITY BOUNDING ASSUMPTIONS f SYSTEM EQUIPMENT SAFETY j ) Incinerator Mass (UPilOLD) Heterogeneous UO2 i l Mass (INHOLD) Optimal H O Moderation 2 Full Reflection incinerator Product 3 Geometry Homogeneous UO2 { or 5-Gallon Containers Mass Optimal H O Moderation 2 Full Reflection l l l Dry Conversion Process UF Cylinder Receipt Enrichment 99.5 wt. % pure UF 6 6 (DCP) Conversion - and Storage s 0.5 wt. % H O equivalent 2 OptimalInterunit H O 2 Vaporization Moderation 99.5 wt. % pure UF6 Autoclave w/UF 's 0.5 wt. % H O equivalent 6 2 Cylinder Full Reflection Vaporization Geometry Homogeneous UO2 Cold Trap System Moderation Optimal H O Moderation 2 Full Reflection Conversion: Moderation Homogeneous UO2 Reactor / Kiln Maximum Credible UO Density 2 Maximum Credible wt. % H O 2 Full Reflection Conversion: Moderation Homogeneous UO2 O-Powder Outlet Box Maximum Credible UO Density 2 Maximum Credible wt. % H O 2 j Full Reflection Powder Outlet: Moderation Homogeneous UO2 l Cooling Hopper Maximum Credible UO Density 2 j Maximum Credible wt % H O 2 Full Reflection Powder Transfer & Moderation Homogeneous UO2 Storage: Normal Maximum Credible UO Density 2 Product Container Maximum Credible wt. % H O i 2 l. i Full Reflection j Powder Transfer & Geometry Homogeneous UO2 Storage: Out-of-Spec Moderation Maximum Credible UO Density 2 Moisture Product Maximum Credible wt. % H O 2 Container Full Reflection Homogenization Moderation Homogeneous UO2 Maximum Credible UO Density 2 Maximum Credible wt. % H O 2 Full Reflection J f LICENSE SNM-1997 DATE 5/27/97 Page l DOCKET 70-1113 REVISION 0 6.15

-a .s . s -a >u. a n uw c-.+ .--.-an+.u = -., cax .s w a s --mm..--.a.u., +n.- AREA PROCESS BASIS FOR CSA O OR SUBAREA OR CRITICALITY BOUNDING ASSUMPTIONS SYSTEM EQUIPMENT SAFETY i

Blending, Moderation Heterogeneous UO2 Precompaction, Maximum Credible UO Density 2

l Granulation - Maximum Credible wt. % H O 2 Full Reflection { Tumbling: Moderation Heterogeneous UO 2 in Powder Container Maximum Credible UO Density ) 2 Maximum Credible wt. % H O 2 l Full Reflection Powder Pack Moderation Heterogeneous UO2 l. Screener Maximum Credible UO Density 2 Maximum Credible wt. % H 0 Full Reflection Powder Pack Geometry Homogeneous UO2 Product Container Mass Optimal H O Moderation 2 Full Reflection j Utilities: N, H, H O Mass Backflow into large supply vessels not 2 2 2 Supply, Refrigerant credible due to backflow prevention j i measures, physical barriers, and/or j process characteristics. l HF Eflluent Recovery Geometry llomogeneous UO2 l and Storage Vessels Mass Optimal H O Moderation 2 Full Reflection IEcycle Blender Moderation Heterogeneous UO2 Maximum Credible UO Density 'O 2 l Maximum Credible wt. % H O 2 l Full Reflection l Recycle Unicone Moderation Heterogeneous UO2 Product Maximum Credible UO Density 2 Container / Storage Maximum Credible Internal wt. % H O 2 Optimal Interunit 110 2 Recycle 3-Gallon Geometry 1 lieterogeneous UO2 Product Container / Mass f* Optimal H O Moderation 2 l Storage Moderation Full Reflection l Press Warehouse Conveyor Storage: Geometry 1 Homogeneous UO2 l Facility Process 3 and 5-gallon Cans Mass J* OptimalInterunit H O Moderation 2 Moderation Full Reflection Powder Dump Transfer Geometry Homogeneous UO2 t I Hopper / Chute Moderation Optimal H O Moderation 2 Full Reflection Pellet Presses Geometry / Mass Heterogeneous UO2 Moderation Optimal H O Moderation 2 Full Reflection

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

LICENSE SNM-1997 DATE 5/27/97 Page i .Il DOCKET 70-1113 REVISION 0 6.16 i %/ i

AREA PROCESS BASIS FOR CSA (") OR SUBAREA OR CRITICALITY BOUNDING ASSUMPTIONS SYSTEM EQUIPMENT SAFETY Press Lubricant Sump Geometry Heterogeneous UO2 Mass OpCmal H O Moderation 2 Full Reflection Press: Green Pellet Geometry Heterogeneous UO2 Boat Product Container Moderation Optimal H O Moderation 2 Full Reflection i 3-gallon Powder Geometry Heterogeneous UO2 Cleanup Container Mass Optimal 110 Moderation { 2 Full Reflection Integration: Moderation Heterogeneous UO2 PWDR-MRA Maximum Credible wt % H O 2 Press Feed Full Reflection integration Geometry / Mass Heterogeneous UO2 PWDR-MRA Moderation Maximum Credible UO Density 2 Container-Storage Maximum Credible wt. % H O 2 Full Reflection Integration Moderation Heterogeneous UO2 PWDR-MRA Maximum Credible UO Density 2 Powder Transfer Maximum Credible wt % H O 2 Corridor Full Reflection Pellet Sintering System Feed / Exit Conveyors Geometry Heterogeneous UO2 Moderation Optimal H O Moderation 2 Full Reflection p Sintering Furnace Geometry Heterogeneous UO2 V Moderation Optimal H O Moderation 2 Full Reflection Pellet Grinding System Feeder Hopper Bowl or Geometry Heterogeneous UO2 Flat Feeder Table Moderation Optimal H O Mod: ration 2 Full Reflection Grinder Geometry Heterogeneous UO2 Moderation Optimal H O Moderation 2 Full Reflection Grinder APITRON Geometry Homogeneous UO2 Filter Moderation Optimal H O Moderation 2 Full Reflection Grinder Swarf 3-Geometry Heterogeneous UO2 Gallon Container Moderation Optimal H O Moderation 2 Full Reflection Grinder Hardscrap 3-Geometry Heterogeneous UO2 Gallon Container Mass Optimal H O Moderation 2 Full Reflection

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

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l l AREA PROCESS BASIS FOR CSA { OR SUBAREA OR CRITICALITY BOUNDING ASSUMPTIONS SYSTEM EQUIPMENT SAFETY Grinder Pellet Product Geometry 1 Heterogeneous UO2 Tray Mass J* Optimal H O Moderation 2 Moderation Full Reflection Pellet Transfer Cart Geometry. Heterogeneous UO l 2 Moderation OptimalInterunit H O Moderation 2 Full Reflection Rod Load, Out-Gassing, Rod Load, Out-Geometry Heterogeneous UO2 and Final Rod Welding Gassing, and Final Rod Moderation Optimal H O Moderation 2 l System Weld Full Reflection l Pellet Storage Cabinet Geometry Heterogeneous UO2 l Moderation Optimal H O Moderation 2 l Full Reflection Rod Storage Cabinet Geometry Heterogeneous UO2 Moderation Optimal H O Moderation 2 Full Reflection Gadolinia Shop Press, Sintering, Similarto UO Shop Similarto UO Shop Above 2 2 l Grinding, Rod Load,. Above Rod Storage, & Outgas Gadolinia 3 and 5-Geometry Homogeneous UO2 l Gallon Feed Containers Mass Optimal H O Moderation 2 Full Reflection l Gadolinia 3 and 5-Geometry 1 Homogeneous UO2 Gallon Feed & Product Mass f* Optimal H O Moderation 2 [- Container Storage Moderation Full Reflection i Gadolinia DM-10 Geometry Heterogeneous UO2 Vibromill(MCA) Moderation Optimal H O Moderation 2 Full Reflection Gadolinia DM-3 Mass Homogeneous UO2 L Vibromill(MCA) Moderation Optimal H O Moderation 2 1 Full Reflection j Pellet Storage: Geometry / Mass Heterogeneous UO2 Ministacker Moderation OptimalH O Moderation 2 Full Reflection l Integration: Mass Homcgeneous UO2 l Gadolinia MEZZ-MRA Moderation Maximum Credible UO Density 2 l Unicone Feed Maximum Credible wt. % H O 2 l Container Full Reflection l Integration Moderation Heterogeneous UO2 Gadolinia MEZZ-MRA Maximum Credible wt. % H O 2 t DM-10 Vibromill Full Reflection

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

l LICENSE SNM-1997 DATE 5/27/97 Page Q DOCKET 70-1113 REVISION 0 6.18

l I l AREA PROCESS BASIS FOR CSA , l ) OR SUBAREA OR CRITICALITY BOUNDING ASSUMPTIONS V SYSTEM EQUIPMENT SAFETY Integration Moderation Heterogeneous UO2 Gadolinia MEZZ-MRA Maximum Credible wt. % H O 2 l Rotary Slugger Full Reflection Integration Moderation Heterogeneous UO2 i Gadolinia MEZZ-MRA Maximum Credible wt. % H O 2 Granulator Full Reflection l Integration: Geometry 1 Homogeneous UO2 Gadolinia MEZZ-MRA Mass f* Optimal H O Moderation 2 3 and 5-Gallon Feed & Moderation Full Reflection Product Container Storage Integration Moderation Heterogeneous UO2 Gadolinia MEZZ-MRA Maximum Credible UO Density 2 Powder Transfer Maximum Credible wt. % H O 2 Corridor Full Reflection Bundle Assembly Rod Trays Geometry Heterogeneous UO2 Mass Optimal Interunit H O Moderation 2 Full Reflection Rod Storage Cabinets Geometry Heterogeneous UO2 Moderation Optimal Interunit H.,0 Moderation Full Reflection Rod Tray Transfer Geometry Heterogeneous UO2 Vehicle:" Big Joe" Moderation OptimalInterunit H O Moderation 2 p Full Reflection U Magnetic and Passive Geometry Heterogeneous UO2 Scanner:" MAPS" Moderation Optimal Interunit H O Moderation 2 Full Reflection Bundle Accumulator: Geometry Heterogeneous UO2 "BACC" Moderation OptimalInterunit H O Moderation 2 Full Reflection Automatic Bundle Geometry Heterogeneous UO2 Assemble Machine: Moderation Optimalinterunit H O Moderation 2 "ABAM" Full Reflection Rod Scanner: Geometry Heterogeneous UO2 " Fat Albert" Moderation Optimal Interunit H O Moderation 2 Full Reflection Assembly Table Geometry Heterogeneous UO2 Moderation OptimalInterunit H O Moderation 2 Full Reflection Upender: Bundle and Geometry Heterogeneous UO2 RA Container Moderation OptimalInterunit H O Moderation 2 Full Reflection

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

l i l LICENSE SNM-1097 DATE 5/27/97 Page !J DOCKET 70-1113 REVISION 0 6.19 l l l

.. -._~ ~ _..._.- -..~.-.. - ~_. -.- h AREA PROCESS BASIS FOR CSA 4 OR SUBAREA OR CRITICALITY BOUNDING ASSUMPTIONS SYSTEM EQUIPMENT SAFETY Inspection Pit Geometry Heterogeneous UO2 Moderation Optimal Interunit H O Moderation 2 Full Reflection ~ Bundle Storage: Geometry Heterogeneous UO2- " Forest" Moderation OptimalInterunit H O Moderation l 2 Full Reflection f RA Container: Geometry Heterogeneous UO2 Transfer Port & RA Moderation OptimalInterunit H O Moderation 2 i Conveyor Full Reflection Rod Scanner: Geometry Heterogeneous UO2 X-Ray-Unit Moderation OptimalInterunit H O Moderation 2 Full Reflection 1 Rod Inspection: Geometry Heterogeneous UO2 Serface-Plate Moderation OptimalInterunit H O Moderation 2 Full Reflection Rod 141ovement: Geometry Heterogeneous UO2 One & Two-Tray Cart Moderation OptimalInterunit H O Moderation 2 Full Reflection Container Storage: Geometry Heterogeneous UO2 RA-Inner / Outer Moderation OptimalInterunit H O Moderation 2 Storage Full Reflection 4 Decontamination & Wash Down Areas, Geometry / Mass Homogeneous UO2 f V-lume Reduction Sumps, Bag Filters Mass Optimal H O Moderation (q 2 / Facility (DVRF) Full Reflection { Dust Hog Mass Homogeneous UO2 Optimal H O Moderation 2 Full Reflection i HVAC Geometry Homogeneous UO2 Optimal H O Moderation j Mass 2 Full Reflection j 3-Gallon Waste Geometry Homogeneous UO2 1 Container Storage Mass Optimal H O Moderation 2 Full Reflection j LICENSE SNM-1997 DATE 5/27/97 Page Q DOCKET 70-1113 REVISION 0 6.20

--- ~ -.. l l l 6.2.4 SPECIFIC PARAMETER LIMITS I }j lk The safe geometry values of Table 6.1 below are specifically licensed for use at the GE-Wilmington. facility. Application of these geometries is limited to situations where the neutron reflection present does not exceed that due to full water reflection. Acceptable geometry margins of safety for units identified in this table are 93% of the minimum critical cylinder diameter, 88% of the minimum critical slab thickness, and l 76% of the minimum critical sphere volume. When cylinders and slabs are not infinite in extent, the dimensional limitations of Table l 6.1 may be increased by means of standard buckling conversion niethods; reactivity ? - formula calculations which incorporate validated K-infinities, migration areas (M ) and extrapolation distances; or explicit stochastic or deterministic modeling methods. l l The safe batch values of Table 6.2 are specifically licensed for use at the GE-l Wilmington facility. Criticality safety may be based on U235 mass limits in either of the j following ways: If double batch is considered credible, the mass of any single accumulation shall not exceed a safe batch, which is defined to be 45% of the minimum critical mass. Table 6.2 lists safe batch limits for homogeneous mixtures of UO and water as a 2 function of U235 enrichment over the range of 1.1% to 15% for uncontrolled geometric configurations. The safe batch sized for UO of specific compounds may l 2 be adjusted when applied to other compounds by the formula: kgs X = (kgs UO

• 0.88 ) / f 2

1 where, kgs X = safe batch value of compound 'X' i kgs UO2 l = safe batch value for UO2 j 0.88 = wt. % U in U0 i 2 f = wt. % U in compound X i l Where engineered controls prevent over batching, a mass of 75% of the minimum i critical mass shall not be exceeded. Subject to provision for adequate protection against precipitation or other circumstances l which may increase concentration, the following safe concentrations are specifically licensed for use at the GE-Wilmington facility: A concentration ofless than or equal to one-half of the minimum critical I concentration. l A system in which the hydrogen to U235 atom ratio (H/U235) is greater than 5200. e I t i LICENSE SNM-io97 DATE 5/27/97 Page O oocxer 7.-i113 anvisioN o 6.2i L i

I Table 6.1 Safe Geometry Values i Homogeneous UO - Weight Percent Infinite Cylinder

• Infinite Slab
• Sphere Volume
• 2

{ H O Mixtures U235 Diameters Thickness 2 (Inches) (Inches) (Liters) f 2.00 16.70 8.90 105.0 3 2.25 14.90 7.90 - 75.5 2.50 13.75 7.20 61.0 a 2.75 12.90 6.65 51.0 4 3.00 12.35 6.25 44.0 3.25 11.70 5.90 38.5 t 3.50 11.20 5.60 34.0 3.75 10.80 5.30 31.0 l 4.00 10.50 5.10 29.0 5.00 9.50 4.45 24.0 4 Homogeneous Weight Percent Infinite Cylinder Infinite Slab Sphere Volume 4 Aqueous U235 Diameters Thickness Solutions (Inches) (Inches) (Liters) 2.00 16.7 9.30 106 4 2.25 15.0 8.40 80.5 i 2.50 14.0 7.80 66.8 2.75 13.3 7.30 56.2 3.00 12.9 7.00 49.7 3.25 12.5 6.70 44.8 1 3.50 12.1 6.50 41.0 { 3.75 11.9 6.30 38.0 4.00 1I.7 6.00 34.9 0 5.00 9.5 4.80 26.0 Heterogeneous Weight Percent Infinite Cylinder Infinite Slab Sphere Volume Mixtures or U235 Diameters Thickness Compounds (Inches) (Inches) (Liters) 2.00 11.10 5.60 35.7 2.25 10.50 5.10 30.7 2.50 10.10 4.80 27.3 2.75 9.70 4.60 24.7 3.00 9.40 4.40 22.6 3.25 9.20 4.30 20.9 3.50 9.00 4.20 19.2 3.75 8.90 4.10 18.2 4.00 8.80 4 00 16.9 5.00 8.30 3.60 13.0

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

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. ~ _ m i i 1' Table 6.2 Safe Batch Values for UO and Water

• 2 l

Nominal Weight Homogeneous Heterogeneous Norninal Weight Homogeneous Heterogeneous J Percent U235 00 Powder & UO Pellets & Percent U235 U0 Powder & UO Pellets & 2 2 2 2 1 Water Water Water Water 4 Mixtures Mixtures Mixtures Mixtures l (Kgs UO ) (Kgs UO ) (Kgs UO ) (Kgs UO ) I 2 2 2 3 l.10 2629.0 510.0 4.00 25.7 24.7 1.20 1391.0 341.0 4.20 23.7 22.9 l.30 833.0 246.0 4.40 21.9 21.4 1.40 583.0 193.0 4.60 20.2 20.0 1.50 404.0 158.0 4.80 19.1 I 8.8 1.60 293.3 135.0 5.00 18.1 18.1 7 1.70 225.0 116.0 1.80 183.0 102.0 l I.90 150.6 90.5 2.00 127.5 81.6 2.10 109.2 73.I 2.20 96.8 66.4 2.30 84.3 61.0 2.40 74.7 56.1 2.50 68.9 52.1 2.60 60.5 48.8 2.70 56.6 45.4 2.80 52.2 42.9 2.90 47.6 40.1 3.00 44.5 38.1 3.20 38.9 34.I 3.40 34.6 31.0 3.60 31.1 28.5 i 3.80 28.3 26.4 1

• NOTE: These values represent 45% of the minimum critical mass. For enrichments not specified, smooth curve interpolation of safe batch values may be used.

t i i 1 J i i 1 5 I l' i e I i j 1 i j LICENSE SNM-1997 DATE 5/27/97 Page )Q DOCKET 70-1113 REVISION 0 6.23 J 4 W

~. . ~ - - - .. -.... -... - - -. - ~. - -. ~. - -. -. - - -. L 3 ~. 6.2.5 CONTROL PARAMETERS 4 . Q Nuclear criticality safety is achieved by controlling one or more parameters of a ' system within' established suberitical limits. The criticality safety review process is - used to identify the significant parameters associated with a particular system. All l assumptions relating to process equipment, material composition, function, and j operation, including upset conditions, arejustified, documented, and independently j reviewed. 4 i identified below are specific control parameters that may be considered during the i j. review process: { 6.2.5.1 Geometry - Geometry may be used for nuclear criticality safety control on its own or l in combination with other control methods. Favorable geometry is based on limiting j dimensions of defined geometrical shapes to established suberitical limits. Structure i and/or neutron absorbers that are not removable constitute a form of geometry. j control. At the GE-Wilmington facility, favorable geometry is developed j conservatively assuming unlimited water or concrete equivalent reflection, optimal [ hydrogenous moderation, worst credible heterogeneity, and maximum credible

~

enrichment to be processed. Examples include cylinder diameters, annular I inner / outer dimensions, slab thickness, and sphere diameters.

ih Geometry control systems are analyzed and evaluated allowing for fabrication i U tolerances and dimensional changes that may likely occ'ur through corrosion, wear, or mechanical distortion. In addition, these systems include provisions for periodic inspection if credible conditions exist for changes in the dimensions of the equipment that may result in the inability to meet established nuclear criticality safety limits.

6.2.5.2 Mass - Mass control may be used for a nuclear criticality safety control on its own or in combination with other control methods. Mass control may be utilized to limit the quantity of uranium within specific process operations or vessels and within storage, transportation, or disposal containers. Analytical or non-destructive methods may be employed to verify the mass measurements for a specific quantity of material. i Establishment of mass limits involves consideration of potential moderation, reflection, geometry, spacing, and material concentration. The criticality safety analysis considers normal operations and credible process upsets in determining actual mass limits for the system and for defining additional controls. When only 1 . LICENSE SNM-1997 DATE 5/27/97 Page ^ DOCKET 70-1113 REVISION 0 6.24 U

administrative controls are used for mass controlled systems, double batching is . f) considered to ensure adequate safety margin. ] v i 6.2.5.3 Moderation - Moderation control may be used for nuclear criticality safety control on its own or in combination with other control methods. When moderation is used in conjunction with other control methods, the area is posted as a ' moderation control area'. When moderation control is the primary design focus and is designated as a the primary criticality safety control parameter, the area is posted ' moderation restricted area'. When moderation is the primary criticality safety control parameter the following graded approach to the design control philosophy is applied in accordance with established facility practices (in decreasing order of restriction): At each enriched uranium interface involving intentional and continuous introduction of moderation (e.g., insertion of superheated steam into reactor), at least three controls are required to assure that the moderation safety factor is not exceeded. At least two of these controls must be active engineered

controls, At enriched uranium interfaces involving intentional but non-continuous e

.i introduction of moderation at least three controls are required to assure that I the moderation safety factor is not exceeded. At least one of these controls O must be aa active easiaeerea coatroi. ueiess a moderatiea serety ractor greater than 3 is demonstrated. For situations where :noderation is not intentionally introduced as part of the process, the requirea number of controls for each credible failure mode must be established in accordance with the double contingency principle. When the maximum credible accident is considered, the safety moderation limit (i.e., % H O or equivalent) must provide sufTicient factor of safety above the process 2 moderation limit. This ' moderation safety factor', which is the ratio of the safety moderation limit to the process moderation limit, will normally be three or higher, but never less than two. The value of the moderation safety factor depends on the likelihood and time required for this system being considered to transition from the process moderation limit to the safety moderation limit. In some cases, as described above, increased depth of protection may be required, but i the minimum protection is never less than the following: two independent controls prevent moderator from entering the system through a defined interface and must fail i I LICENSE SNM-1997 DATE 5/27/97 Page Q DOCKET 70-1113 REVISION 0 6.25

before a criticality accident is possible. The quality and basis for selection of the m() controls is documented in accordance with Integrated Safety Analysis process ~ described in Chapter 4.0. Controls for the introduction and limited usage of moderating materials (e.g. for cleaning or lubrication purposes) within areas in which the primary criticality safety parameter is moderation are approved by the criticality safety function. 6.2.5.4 Concentration (or Density) - Concentration control may be used for nuclear criticality safety control on its own or in combination with other control methods. Concentration controls are established to ensure that the concentration level is maintained within defined limits for the system. When concentration is the only parameter controlled to prevent criticality, concentration may be controlled by two independent combinations of measurement and physical control, each physical control capable of preventing the concentration limit being exceeded in a location where it would be unsafe. The preferred method of attaining independence being that at least one of the two combinations is an active engineered control. Each process relying on concentration control has in place controls necessary to detect and/or mitigate the effects ofinternal concentration within the system (e.g., Dynatrol density meter, Rhonan density meter, etc.), otherwise, the most reactive credible concentration (density) is assumed. U 6.2.5.5 Neutron Absorber - Neutron absorbing meterials may be utilized to provide a j method for nuclear criticality safety control for a process, vessel or container. Stable compounds such as boron carbide fixed in a matrix such as aluminum or polyester resin; elemental cadmium clad in appropriate material; elemental boron alloyed stainless steel, or other solid neutron absorbing materials with an established dimensional relationship to the fissionable material are recommended. The use of neutron absorbers in this manner is defined as part of a passive engineered control. Credit may be taken for neutron absorbers such as gadolinia in completed nuclear fuel bundles (e.g., packaged and stored onsite for shipment) provided the following requirements are met: The presence of the gadolinia absorber in completed fuel rods is documented and verified using non-destructive testing; and the placement of rods in completed fuel bundles is documented in accordance with established quality control practices. LICENSE SNM-1097 DATE 5/27/97 Page O DOCKET 70-1113 REVISION 0 6.26 v

. j i i 1 1-i Credit may be taken for neutron absorbers that are normal constituents of filter media j h (e.g., natural boron) provided the following requirements are met: The failure or loss of the media itself also prevents accumulation of e L . significant quantities of fissile material. The neutron absorber content is certified. i For fixed neutron absorbers used as part of a geometry control, the following 7 requirements apply: e The composition of the absorber are measured and documented prior to first use. Periodic verification of the integrity of the neutron absorber system l-subsequent to installation is performed on a scheduled basis approved by the ' criticality safety function. The method of verification may take the form of i. traceability (i.e. serial number, QA documentation, etc.), visdal inspection or i j-direct measurement. j 62.5.6 Spacing (or Unit Interaction) - Criticality safety controls based on isolation or interacting unit spacing. Units may be considered effectively non-interacting E (isolated) when they are separated by either of the following: 12-inches of full density water equivalent, or e l the larger of 12-foot air distance or the greatest distance across an j orthographic projection of the largest of the fissile accumulations on a plane i perpendicular to the line joining their centers. I. For Solid Angle interaction analyses, a unit where the contribution to the total solid angle in the array is less than 0.005 steradians is also considered non-interacting (provided the total of all such solid angles neglected is less than one half of the total i solid angle for the system). Transfer pipes of 2 inches or less in diameter may be excluded from interaction consideration, provided they are not grouped in close . arrays. 4[ Techniques which produce a calculated effective multiplication factor of the entire system (e.g., validated Monte Carlo or S Discrete Ordinates codes) may be used. Techniques which do not produce a calculated effective multiplication factor for the I entire system but instead compare the system to accepted empirical criteria, (e.g., ' Solid Angle methods) may also be used. In either case, the criticality safety analysis must comply with the requirements of Sections 6.1.1 and 6.3. 1 j. LICENSE SNM-1997 DATE 5/27/97 Page DOCKET .70-1113 REVISION 0 6.27 L k g- ,,-s ,w ..y. r-. -r.

6.2.5.7 Material Composition (or Heterogeneity) - The criticality safety analysis for each O process determines the effects of material composition (e.g., type, chemical form, i physical form) within the process being analyzed and identifies the basis for selection i of compositions used in subsequent system modeling activities. It is important to distinguish l'etween homogeneous and heterogeneous system canditions. Heterogeneous effects within a system can be significant and therefore must be considered within the criticality safety analysis when appropriate. ) Evaluation of systems where the particle size varies take into consideration effects of heterogeneity appropriate for the process being analyzed. 6.2.5.8 Reflection - Most systems are designed and operated with the assumption of 12-inch water or optimum reflection. However, subject to approved controls which limit } reflection, certain system designs may be analyzed, approved, and operated in situations where the analyzed reflection is less than optimum. In criticality safety analysis, the neutron reflection properties of the credible process environment are considered. For example, reflectors more effective than water (e.g., i concrete) are considered when appropriate. 6.2.5.9 Enrichment - Enrichment control may be utilized to limit the percent U-235 within a O-process, vessel, or container, thus providing a method for nuclear criticality safety control. Active engineered or administrative controls are required to verify enrichment and to prevent the introduction of uranium at unacceptable enrichment levels within a defined subsystem within the same area. In cases where enrichment control is not utilized, the maximum credible area enrichment is utilized in the criticality safety analysis. 6.2.5.10 Process Characteristics - Within certain manufacturing operations, credit may be taken for physical and chemical properties of the process and/or materials as nuclear j criticality safety controls. Use of process characteristics is predicated upon the j following requirements: The bounding conditions and operational limits are specifically identified in the criticality safety analysis and, are specifically communicated, through training and procedures, to appropriate operations personnel. LICENSE SNM-1097 DATE 5/27/97 Page t] DOCKET 70-1113 REVISION 0 6.28

1' 1 Bounding conditions for such process and/or material characteristics are C" based on established physical or chemical reactions, known scientific principles, and/or facility-specific experimental data supported by operational history. The devices and/or procedures which maintain the limiting conditions must have the reliability, independence, and other characteristics required of a criticality safety control. Examples of process characteristics which may be used as controls include: Conversion and oxidation processes that produce dry powder as a product of high temperature reactions. Experimental data demonstrating low moisture pickup in or on uranium i materials that have becu ecnditioned by room air ventilation equipment. Experimental / historical process data demonstrating uranium oxide powder e flow characteristics to be directly proportional to the quantity of moisture present. 6.3 CONTROL DOCUMENTS g# j 6.3.1 CRITICALITY SAFETY ANALYSIS (CSA) In accordance with ANSI /ANS-8.19 (1984), the criticality safety analysis is a collection ofinformation that "provides sufficient detail clarity, and lack of ambiguity to allow independent judgment of the results." The CSA documents the physical / safety basis for the establishment of the controls. The CSA is a controlled element of the Integrated Safety Analysis (ISA) defined in Chapter 4.0. The CSA addresses the specific concerns (event sequences) of nuclear criticality safety importance for a particular system. A CSA is prepared or updated for each new or significantly modified unit or process system within the GE-Wilmington facility in accordance with established configuration management control practices defined in Chapter 3.0. The scope and content of any particular CSA reflects the needs and characteristics of the system being analyzed and includes applicable information requirements as follows: LICENSE SNM-1097 DATE 5/27/97 Page {} DOCKET 70-1113 REVISION 0 6.29

i i Scope - This element defines the stated purpose of the analysis. e General Discussion - This element presents an overview of the process that is affected by the proposed change. This section includes as appropriate; process description, flow diagrams, normal operating conditions, system interfaces, and other important to design considerations. Criticality Safety Controls / Bounding Assumptions - This element defines e a minimum of two criticality safety controls that are imposed as a result of the F analysis. This section also clearly presents a summary of the bounding assumptions used in the analysis. Bounding assumptions include; worst credible contents (e.g., material composition, density, enrichment, and i moderation), boundary conditions, interunit water, and a statement on assumed structure. In addition, this section includes a statement which i i summarizes the interface considerations with other units, subareas and/or areas. Model Description - This element presents a narrative description of the e actual model used in the analysis. An identification of both normal and credible upset (accident condition) model filenaming convention is provided. Key input listings and corresponding geometry plot (s) for both normal and l credible upset cases are also provided. Calculational Results - This element identifies how the calculations were performed, what tools or reference documents were used, and when ( appropriate, presents a tabular listing of the calculational result and associated uncertainty (e.g., Keff + 3a) results as a function of the key parameter (s) (e.g., wt. fraction H O). When applicable, the assigned bias of the 2 calculation is also clearly stated and incorporated into both normal and/or accident limit comparisons 4 Safety During Upset Conditions - This element presents a concise summary e of the upset conditions considered credible for the defined unit or process a system. This section include a discussion as to how the established nuclear criticality safety limits are addressed for each credible process upset (accident condition) pathway. Specifications and Requirements for Safety - When applicable, this element presents both the design specifications and the criticality safety requirements for correct implementation of the established controls. These requirements are incorporated into operating procedures, training, LICENSE SNM-1997 DATE 5/27/97 Page O oocke' 'a-"'3 x8visio" a 6.3o /TW . W

maintenance, quality assurance as appropriate to implement the specifications

g and requirements.

Compliance - This element concludes the analysis with pertinent summary statements and includes a statement regarding license compliance. Verification - Each criticality safety analysis is verified in accordance with section 6.3.2.5 by a senior engineer approved by the criticality safety function and who was not involved in the analysis. Appendices - Where necessary, a summary ofinformation ancillary to calculations such as parametric sensitivity studies, references, key inputs, model geometry plots, equipment sketches, useful data, etc., for each defined system is included. 6.3.2 ANALYSIS METHODS 6.3.2.1 KeffLimit Validated computer analytical methods may be used to evaluate individual system units or potential system interaction. When these analytical methods are used, it is required that the effective neutron multiplication factors for credible process upset (d3 (accident) conditions are less than or equal to 0.97 including applicable biases and calculational uncertainties, that is: Keff + 3a - bias s 0.97 (accident conditions). Thus, the established delta-k safety margin used at the GE-Wilmington facility is 0.03. Nonnal operating conditions include maximum credible conditions expected to be encountered when the criticality control systems function properly. Credible process upsets include anticipated off-normal or credible accident conditions and must be demonstrated to be critically safe in all cases in accordance with Section 6.LI. The sensitivity of key parameters with respect to the effect on Keff are evaluated for each system such that adequate criticality safety controls are defined for the analyzed system. LICENSE SNM-1097 DATE 5/27/97 Page []) DOCKET 70-1113 REVISION 0 6.31

6.3.2.2 Analytical Methods (') Methodologies currently employed by the GE-Wilmington criticality safety function include hand calculations utilizing published experimental data (e.g., ARH-600 handbook), Solid Angle methods (e.g., SAC code), and Monte Carlo codes (e.g., GEKENO, GEMER) which utilize stochastic methods to solve the 3D neutron transport equation. Additional Monte Carlo codes (e.g., Keno Va and MCNP) or So Discrete Ordinates codes (e.g., ANISN or XSDRNPM) may be used after validation as described in subparagraph (c) below. GEKENO (Geometry Enhanced KENO) is a multigroup Monte Carlo program which solves the neutron transport equation in 3-dimensional space. The GEKENO criticality program utilizes the 16-energy group Knight-Modified Hansen Roach cross-section data set, and a potential scattering cr resonance correction to p compensate for flux depression at resonance peaks. GEKENO is normally used for homogeneous systems. For infinite systems, K. can be calculated directly from the Hansen Roach cross-sections using the program KINF. GEMER (Geometry Enhanced merit) is a multigroup Monte Carlo program which solves the neutron transport equation in 3-dimensional space. The GEMER criticality program is based on 190-energy group structure to represent the neutron energy spectrum. In addition, GEMER treats resolved resonances explicitly by tracking the neutron energy and solving the single-level Breit-Wigner equation at {~') each collision in the resolved resonance range in regions containing materials whose resolve resonances are explicitly represented. The cross-section treatment in GEMER is especially important for heterogeneous systems since the multigroup treatment does not accurately account for resonance self-shielding. 6.3.2.3 Validation Techniques Experimental critical data or analytical methods which have been validated (benchmarked) by comparison with experimental critical data in accordance with criteria described in section 4.3 of ANSI /ANS 8.1 (1983) are used as the basis for validation. An analytical method is considered validated when the following ere established: the type of systems which can be modeled the range of parameters which may be treated e the bias, if any, which exists in the results produced by the method. i l l LICENSE SNM-1097 DATE 5/27/97 Page l O' DOCKET 70-1113 REVISION 0 6.32

't i /^L Currently GEMER is validated against 123 critical experiments and GEKENO is validated against 56 critical experiments. Both validations produce a bias fit as a function of H/U235 atom ratio. This fit is established against the lower limit of the j 3-sigma confidence band (see Figures 6.1 and 6.2). The bias (Kc ic - 1.0) is applied over its negative range and assigned a value of zero over its positive range. The j range of applicability covers all compounds in use at GE-Wilmington and enrichments up to 5.0 % wt. % U235. 4 FIGURE 6.1 - CEMER BIAS DETERMINATION, PARTICLE NEICHI i 1.19 i LEGEND l 128 DATA SET . PARTICLE HEIGHT i a SRD CPDER FIT OF LIMIT ' 1.98 = t-EFF e 1.0 a L!kEAR FITS ORDERS 2 99.782 CONFIDENCE BAND 1.48 a 1.04 i E Eff 18. ) ~ U 1.., a 1... f i 0.960 -it 28 50 80 110 140 179 HYOR0 GEN-70-0285 Mit' I LICENSE SNM-1997 DATE 5/27/97 Page C DOCKET 70-1113 REVISION 0 6.33

a ~]/ FICURE 6.2 - GEKENO BIAS CALCUIATION 1.10 LEGEND GEKENO VER$10N 80 e 56 DATA POINTS x 8RD CROER FIT OF LIMIT 3... - zerr. i.. LINEAR FITS ORDERS 2 99.70E CONF 20fMCE SAND 1.06 1.04 li K-tFF taf I N "u o l' l Q t s

Q{f' O.900 0

.. u. a 8. 6. 32. 35. MYDR0 GEN-TO-U285 Mit' 6.3.2.4 Computer Software & Ilardware Configuration Control The software and hardware used within the criticality safety calculational system is configured and maintained so that change control is assured through the authorized system administrator. Software changes are conducted in accordance with an approved configuration control program described in Chapter 3.0 that addresses both hardware and software qualification. Software designated for use in nuclear criticality safety are compiled into worxing code versions with executable files that are traceable by length, time, date, and version. Working code versions of compiled software are validated against critical experiments using an established methodology with the differences in experiment LICENSE SNM-1997 DATE 5/27/97 Page O oocx8T 'a->>>> navisio" a 634 I l l

-._ -. - - _~ i [ and analytical methods being used to calculate bias and uncertainty values to be i _O . applied to the calculational results. l Each individual workstation is verified to produce results identical to the l development workstation prior to use of the software for criticality safety I calculations demonstrations on the production workstation. I Modifications to software that may affect the calculational logic require re-validation l of the software. Modifications to hardware or software that do not affect the i calculational logic are followed by code operability verification, in which case, selected calculations are performed to verify identical results from previous analyses. Deviations noted in code verification that might alter the bias or uncertainty requires i re-qualification of the code prior to release for use. 6.3.2.5 Technical Reviews l Independent technical reviews of proposed criticality safety control limits specified + in criticality safety analyses are performed. A senior engineer within the criticality safety function is required to perform the independent technical review. The independent technical review consists of a verification that the neutronics geometry model and configuration used adequately represent the system being i analyzed. In addition, the reviewer verifies that the proposed material h characterizations such as density, concentration, etc., adequately represent the i system. He/She also verifies that the proposed criticality safety controls are i ' adequate. The independent technical review of the specific calculations and computer models j are performed using one of the following methods: Verify the calculations with an alternate computational method. Verify the calculations by performing a comparison to results from a similar e design or to similar previously performed calculations. j Verify the calculations using specific checks of the computer codes used, as e well as, evaluations of code input and output. { Verify the calculations with a custom method. Based on one of these prescribed methods, the independent technical review provides a reasonable measure of assurance that the chosen analysis inethodology and results are correct. i l LICENSE SNM-1997 DATE 5/27/97 Page ' DOCKET 70-1113' REVISION 0 6.35

l 6.4 . CRITICALITY ACCIDENT ALARM SYSTEM 6.4.1 SPECIFICATIONS The criticality accident alarm system radiation monitoring unit detectors are located to assure compliance with appropriate requirements of ANSI /ANS-83 (1986). The location and spacing of the detectors are chosen to avoid the effect of shielding by massive equipment or materials.. Spacing between detectors is reduced where high density building materials such as brick, concrete, or grout-filled cinder block shield l a potential accident area from the detector. Low density materials of construction such as wooden stud construction walls, asbestos, plaster, or metal-corrugated panels, doors, non-load walls, and steel office partitions are disregarded in determining the spacing. 6.4.2. OPERATION The criticality accident alarm system initiates immediate evacuation of the facility. Employees are trained in recognizing the evacuation signal. This system, and proper response protocol, is described in the Radiological Contingency and Emergency Plan for GE-Wilmington.

D 6.43 MAINTENANCE j

i The nuclear criticality alarm system is a safety-significant system and is maintained through routine calibration and scheduled functional tests conducted in accordance with internal procedures. In the event ofloss of normal power, emergency power is automatically supplied to the criticality accident alarm system. LICENSE - SNM-1997 DATE 5/27/97 Page O oocker 7a->>>> xevisio" a 6 36

) l l l 1 l lp CHAPTER 7.0 i V i l CHEMICAL SAFETY j l l ( 7.1 CHEMICAL SAFETY PROGRAM 1 l It is the policy of GE-Wilmington to provide a safe and healthy work place by minimizing the risk of chemical exposure to employees and members of the general public. The chemical safety program is applicable to the chemicals associated with the authorized activities in Chapter 1 and include UF6 and hydrofluoric acid as well as any other chemicals which may directly or indirectly affect the nuclear safety of l these activities. The GE-Wilmington chemical safety program is documented in l written, approved practices that are followed, and ensures that processes and j operations comply with applicable federal and state regulations pertaining to chemical safety. Hazard evaluations are performed on nuclear and non-nuclear operations within the l nuclear manufacturing operations where the potential exists for hazardous chemicals l to be used in such a manner that they could effect the nuclear safety program. This l ensures appropriate controls are in place for adequate protection of the general public l and safe use by employees, and that the use of chemicals does not create potential l conditions that adversely effect the handling oflicensed nuclear materials. Employees using hazardous materials are trained to ensure safe handling, use, and [ disposal. l 7.2 CONTENTS OF CHEMICAL SAFETY PROGRAM The following management control elements are incorporated into GE-Wilmington chemical safety program: + 7.2.1 CHEMICAL SAFETY IN INTEGRATED SAFETY ANALYSIS Considerations of chemical safety for hazardous materials as described in this Chapter are incorporated in GE-Wilmington's Integrated Safety Analysis program. This program includes UF and hydrofluoric acid. GE-Wilmington's Integrated l 6 Safety Analysis Program is explained in detail within Chapter 4.0. l LICENSE SNM-1097 DATE 05/27/97 Page O oocxer va->>>> anvisio" a 7> 1

\\ L 7.2.2 CHEMICAL APPROVAL / EVALUATION O - Prior to new hazardous materials being brought on-site or used in a process, they are approved through the environmental protection function and the chemical and fire safety function. The formal approval process consists of evaluations of the following l potential hazards: i Physical Hazards e Health Hazards Fire / Explosive Hazards l Potential Impact on handling oflicensed nuclear material e The conclusions of this approval process may dictate the following assurance of chemical process safety: New procedures or changes in existing procedures e Maintenance programs for control related equipment e Configuration management Emergency Planning Training l e O l 7.2.3 LABELING & IDENTIFICATION Hazardous materials or conveyance systems are labeled or identified to meet applicable regulations. The proper identification of hazardous materials decreases the likelihood ofimproper use, handling and disposal reducing potential negative l consequences. l 7.2.4 EMPLOYEE TRAINING & AWARENESS Radiation workers receive nuclear safety training and otherjob related training (Chapter 3, Section 3.4) which includes safety information related to chemicals associated with nuclear material and chemicals in the area which could impact the nuclear safety of the process. l LICENSE SNM-1997 DATE 05/27/97 Page C DOCKET 70-1113 REVISION 0 7.2 t i

[ \\ 7.2.5 INCIDENT CLASSIFICATION & INVESTIGATION GE-Wilmington's incident classification and investigation program is discussed in. Chapter 3.0. 7.2.6 CONDUCT OF OPERATIONS Other elements of the chemical safety program are included in Chapter 3.0, " Conduct ofOperations". O 6 LICENSE SNM-1997 DATE 05/27/97 Page OV DOCKET 70-1113 REVISION 0 7.3 ...}}