ML20055C785

Forwards Proposed Changes Incorporated Into Application for Renewal of License SNM-1168.Changes Include 43 Conditions Presented to B&W on 900618

ML20055C785
Person / Time
Site:	07001201
Issue date:	06/21/1990
From:	Lester K BABCOCK & WILCOX CO.
To:	Soong S NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS)
References
NUDOCS 9006250116
Download: ML20055C785 (55)

Text

{{#Wiki_filter:7_. f , B&W FuelCompany An Amerzen company wnh P.O. Box 11646 Lynchburg. VA 24506 1646 (804) 522 6000 June 21, 1990 !bh(( Mr. Sean Soong Uranium Fuel Section Fuel Cycle Safety Branch Division of Industrial and Medical Safety, NMSS Nuclear Regulatory Commission Washington, D.C. 20555

REFERENCE:

SNM-1168 License Renewal Application, Docket 70-1201

Dear Sean:

I am forwarding to you the proposed changes that 1 incorporated into the above referenced license. The changes include the list of the 43 conditions presented to B&W Fuel Company on the 18th of June, the conditions presented by Amar Detta per our telephone conversation on June 21, 1990 and some minor changes resulting from conversations that I had with you and Dave McCaughy prior to his resignation. The 43 conditions and Amar's conditions are all addressed in Attachment I. The effected pages of the license are also included, Attachment II. The paragraph that was modified has i been aide barred and marked with the appropriate condition number I for easy identification and clarification. Attachment III provides the explanation of the other changea. The effected paragrapha have also been provided and side barred. Attachment IV providos the internal and external exposure data that you requested. Please note that the pages have been stamped at draft. Upon your concurrence, the entire document will be submitted. Should you have any comments or questions, please feel free to call me, (804) 522-6202. I will '>e out of town until July 5, 1990. During this time period you may contact either Richcrd f Alto, (804) 5??-5315 or Ernie Coppola, (804) 522-5654. j Ill Sincerely, B&W FUEL COMPANY ] COMMERCIAL NUCLEAR FUEL PLANT b *b$ Hathryn S. Lester 9006250116 900621 PDR ADOCK 07001201 q C PDC

O@N ATTACHMENT I 1 -10) B&W accepta conditions 1 10. 11) The plant manager shall have, as a minimum, a bachelor *a degree in science or engineering, 10 years of experience in the nuclear industry, and 5 years experience in management positions. The incumbent plant manager la deemed to have the equivalent of these requirementa. Section 2.2.1, page 2-3 deleted "In absence of a Technical Degree, 18 years experience in the nuclear industry and 5 years experience in management will suffice." Section 1.7.5, page 1-6 was added which states "The incumbent plant manager is deemed to have the equivalent of the requirements outlined in section 2.2.1." l 12) The Health and Safety Monitora shall have as a minimum a high school diploma with 6 months of experience as a Health and Safety Monitor trainee. Section 2.2.5, page 2-4 was changed to state "The Health-Safety Monitora shall have as minimum, a high school diploma or GED equivalent with six months of experience as a radiation monitor. They may fulfill the experience requirements on the job as a Health-Safety Monitor Trainee." 13) The Operational Area Supervisora shall have, an a minimum, a bachelor's degree in science or engineering, followed by 2 years experience in the nuclear industry. In our discuacions on June 18, 1990, BWFC and NRC concurred that these requirements were for the production managara not the operational supervisora. To clarify this matter, fiqure 2.1, page 2-1 was updated to illustrate the Organization at the CNFP. Also, section 2.1.2, page 2-2 was added to state the following: " production Managers report directly to the plant manager. They are responalble for managing operational supervisora and are responalble for production functions. The production Managera shall have, as a minimum, a bachelor #m degree in science or engineering, followed by two years experience in the nuclear industry." l

@N1 14) The individual performing the independent review of the nuclear criticality safety evaluation shall have, ma a minimum, two years experience as a Nuclear Criticality Safety Specialist. Section 4.1.2, page 4-1 was changed to include "All nuclear criticality safety evaluations shall be independently reviewed by an individual meeting the quellfications of nuclear criticality specialist as defined above with two years experience as a Nuclear Criticality Specialist." 15) The Safety Review Board chairman shall have, as a minimum, the qualifications of the Manager, Quality and Safety. Section 2.3, page 2-5 deleted "or eight years experience" to make the qualifications identical to those for the Manager, Quality and Safety. "The Board shall be chaired by the Manager, Quality and Safety" ( was also deleted. j Qualified designee was added to the glossary of terma, section 1.6, page 1-4 for clarification purposes. 16) The Safety Review Board members shall have, as a minimum, a bachelor's degree in science or engineering, followed by four years experience in the nuclear i industry. BWFC and NRC concurred that the experience can be reduced to two years. Section 2.3, page 2-5 was changed to read "The permanent membership of the Board shall consist of representatives from production management, Quality and Safety and others as deemed necessary by the chairman." The qualifications established in section 2.1 and section 2.2 for production management and Quality and Safety meet those . outlined here: 17) Prior to implementation, new or revised operating procedures shall be approved in writing by the cognizant Operating Area Supervisor (changed to Production Manager) and the Manger, Quality and Safety. Section 2.6, page 2-8 deleted "representativea of plant management." and added " production management and Manager, Quality and Safety." 18) This condition was changed to read: Operating procedures shall be reviewed, at least every two years, by the cognizant Production Manager and the Manager, Quality and Safety.

-( Section 2.6, page 2-8 waa' changed to include " Operating proceduren which involve SNM shall be reviewed at least every two years by the appropriate production manager and Manager, Quality and Safety." 19) This condition was changed to read: The licensee shall evaluate the effectiveness of the training program at least every two years. BWFC decided that this should be the responalbility of the Safety Review Board and accordingly, section 2.3, page 2-6 "At least every two years, the Safety Review Board shall evaluate the effectiveness of the radiation / nuclear safety training program" was added. 20) Notwithstanding Section 3.1.3 of the application, the RWP shall be used only for non-routine operations, reviewed for industrial safety and be approved by the Health Physicist. A RWP previously approved may be j reimaued and approved by the Health Safety Foreman. It was agree that "used only for non-routine operations could be reeoved fror this condition. Section 3.1.3, page 3-1 was changed to read "The RWP shall be reviewed for industrial safety and approved by the Health Physicist. An RWp previously approved may be reissued and approved by the Health-Safety Foreman." 21) The annual ALARA report shall be reviewed by the Safety Review Board to determine: 1) If there are any upward trenda developing in personnel exposures (internal and external) for identifiable categories of workers, types of operation, or effluent releases; 2) if exposures and releases might be lowered in accordance with the ALARA concept; and 3) if equipment for effluent and exposure controla is being properly used, maintained, and inspected. The Board shall report the results of its review and recommendations to the Plant Manager. Section 2.3, page 2-6 added "An annual ALARA report shall be prepared under the direction of the Manager, Quality and Safety. The report shall be submitted to the Safety Review Board in which they will review to determine: 1) If there are any upward trenda developing in personnel exposures (internal and external) for identifiable categories of workers, typen of operation, or effluent releases; 2) if exposures and releases might be lowered in accordance with the ALARA concept; and 3) if equipment for effluent and exposure controla is being properly used, maintained, and inspected. A copy of the report shall be cent to the Plant Manager along with the resulta and recommendations."

QEbh 22) The license shall take corrective action when the face velocity of the ventilated hood la below 100 LFM. The record of the investigation shall be kept for 2 years. Our SWN License 1168 already states the 100 LFM limit. It is BWFC policy that anything in violation of are license is not allowed. Therefore, we do not see the need to include this as part of the license. It will, however, be proceduralized. 23) The breathing zone air in a Radiation Control Zone shall be sampled continuously and analyzed for airborne concentrations of radioactivity after each operational shift. If an individual's internal exposure exceeda 8 r MpC-hr per operational shift, an investigation of the cause shall be conducted and corrective action shall be taken and documented. Section 3.3, page 3-16 added " Fixed air samples shall be used to sample air continuously. Analysea for airborne concentrations of radioactivity shall be conducted after each operational shift." The Health-Safety procedures dictate that an ivestigation be conducted if an individual receives 10 MpC-bra in any seven day period. We will include that an investigation be conducted for MPC-hr exceeding 2 for an individus1 on a daily basia. 24) When beta gamma contamination above action levela is identified, cleanup operation shall commence no later than at the start of the next work shift. Section 3.3, page 3-10 added " Clean-up operations shall commence no later than the start of the next work shift." 25) After each operational day, the change area for the beta gamma radiation work areas shall be surveyed for removal contamination. The action level for cleanup will be'1,000 dpm/100cm2 for removal contamination. Section 3.3, page 3-15 included "Tho change room shall l be surveyed each operational day and shall not exceed the limita established in Table 2. 26) Notwithstanding Section 10.4.2. of the application, solid waste material containing less than 0.002ue/gm L ahall be treated as radioactive material and be disposed at a licensed burial alte. l l

i cun hk l The paragraph in question waa deleted. Section 10.4.2, page 10-9 was changed to read " Uncontrolled disposal of solid wastem or equipment is authorized when contamination levela do not exceed the levela defined i in section 1.7.4 and under the concept of ALARA. l 27) Procesa designa shall incorporate sufficient factors of safety to require at least two unlikely, independent and concurrent changea in process conditions before a criticality accident la possible. I Section 4.1.1, page 4-1 added "procesa designa shall incorporate sufficient factora of safety to require at least two unlikely, independent and concurrent changes in procesa conditions before a criticality accident la possible." i 28) Notwithstanding the statements in Section 4.1.3., all nuclear criticality safety evaluations and i interpretations shall be performed by the nuclear criticality safety specialist. Section 4.1.3.3, page 4-4 was modified to read "The i proposal shall be reviewed by the Safety Review Board Chairman or hia qualified designee for content, completeneas, and conformance with previously evaluated conditions." Section 4'.1.2, page 4-2 also replaced the word " interpretations" with "conformance". 29) Prior to initiation of a new or changed procesa, the individual who performed either the nuclear criticality i safety evaluation or the independent review shall participate in the pre-operational inspection. Section 4.1.3.3, page 4-4 added "The individual who performed either the nuclear criticality safety evaluation or the independent review shall participate in the pre-operational inspection." 30) The number of fuel assemblies in the fuel assembly storage area shall not exceed 100. Section 4.2.4.6, page 4-17 added "No stream sources or aprinklera shall be located near the atorage array. Restriction that prohibits any significant quantity of moderating material auch as paper, plastic or oil and etc. within the fuel anaembly shall be prominently posted in the boundries of the anaembly storage array. l

i-3 f( Chu "b t 31)- The license shall submit semiannual effluent resulta to f the NRC as required by 10 CFR 70.59, in accordance with Section 5 of Regulatory Guide 4.16 dated December 1985. i Saction 5.1, page 5-1 added "A report providing the affluent results in accordance with section 5 of Regulatory Guide date December 1985 shall be submitted to the NRC semiannually." 32) If quarterly averaged concentration of radioactivity of the liquid effluent discharging from retention tank exceeda 2.5 percent of the MPC (based on a gross alpha count of uranium) in 10 CFR 20, Appendix B, Table 11 Vol. 2, the licensee shall conduct an investigation of the source and take corrective action, if necessary. Section 5.1.2, page 5-1 added " Utilizing this procean for the management of liquid effluents, the discharge from the retention tanks shell not exceed 2.5% MPC for uranium. An investigation shall be conducted for levels exceeding 2.54 MPC." 33) The licensee shall inform the NRC within 30 days if the State permitting agency revokes, supersedes, conditions, modifies, or otherwise nullifies the effectiveness of the State-issued NPDS permit for the discharge of liquid effluents. j BWFC chooses to leave this as a condition and requesta that the reports be allowed to be submitted semi-annually along with the semiannual effluent report. 34) The licensee shall conduct a characterizatior, survey and develop an action plan for the cleanup of the contamination soil from the wet-weather stream and l aubmit the plan for NRC review within 9 months from I the date of license renewal. BWFC choosen to leave this as a condition. 35) The license shall conduct environmental monitoring in accordance with approved procedures, which require the generated data be evaluated against an internal action level. 1 Section 5.2, page 5-2 included " Environmental monitoring shall be in accordance with approved procedures which require the generated data to be evaluated against internal action levels." 36) ADatta LC. Per our telephone conversation on June 21, 1990, the t m

4 .s0bN. following changes were incorporated into cur application: Section 2.3, page 2-4 added " maintenance of fire gcargy" to be reviewed quarterly by the Safety Review efe Section 2.5, page 2-6 added " Chemical and fire safety" to the indoctrination of personnel training topics. Section 2.5, page 2-7 added " Documentation of formal training and retraining shall be maintained by Health-Safety and retained for at least two years." Section 2.7, page 2-9 added acnthly safety inspections + to include fire safety. Section 2.7, page 2-10 added fire safety to our semiannual independent audita. 37) The licensee la hereby exempted from the labeling requirementa in 10 CFR 20.203(e) and (f) provided a sign bearing the words " Caution Radioactive Material. Any Area or Container Within This Area May Contain Radioactive Material" or equivalent worda la posted outside the specific work areas. It was agreed that the postinga may still uma the word " Plant" versua " area". Section 1.7.1 was changed to read "The intent of this section la met by posting areas which house or temporarily store radioactive material..." 38) Notwithstanding the statement in Section 1.7.4 Part I of the application dated June 21, 1989, release of equipment and material from the plant to offaite for unrestricted use or from contaminated to clean area onsite shall be in accordance with the attached, " Guidelines for Decontamination of Facilities an Equipment Prior to Release for Unrestricted Use or ) Termination of license for Byproduct, Source or Special Nuclear Material," August 1987. Section 1.7.4, page 1-6 added "B&W Fuel Company la authorized to release equipment and material from the controlled areas to the uncontrolled areas in accordance with "Guidelinem for Decontamination of Facilities and Equipment Prior to Release for Unrestricted Use or Termination of Licenaea for By-Product, Source or Special Nuclear Material", USNRC August 1987, Exhibit A." The definitiona of uncontrolled and controlled areas was added to the glomaary of terms, section 1.6.

"[ 39) The licensee is hereby granted the exemptions and special authorizations in Sections 1.7.2 and 1.7.3, Chapter 1, of the application. Per our discussion on June 18, 1990, 1.7.3 was changed to include " Fuel assemblies stored in accordance with NRC licensed shipping containers Certificate of Compliance No.6206 are exempt from the criticality 1 monitoring requirements of 10 CFR 70.24 provided: -The loaded container array has a minimum separation of 12 feet." 40) At not more than 2-year intervals from July 31, 1990 the licensee shall update the demonstration sections of the renewal application to reflect the licensee's current operations. The updates to the application shall,as a minimum, include information for the health and safety section of the application as required oy 10 CFR Part 70.22(a) through 70.22(f), and 70.22(1) and operational data or environmental releases as required by 70.21. In lieu of an update at the end of the 10 year renewal period the licensee shall file an application for renewal on or before July 31 1999. BWFC accepta this sa a condition. 41) On or before July 31, 1995, the license shall be in compliance with the financial assurance for decomm-lasioning provisions of 70.35 of 10 CFR Part 70, 40.36 (d) and (e) of 10 CFR Part 40, and 30.35 (e) and (f) of 10 CFR Part 30. BWFC accepts this as a condition. 42) On or before July 31, 1990, the licensee shall be in full compliance with the emergency plan requirements of Section 70.22(1) of 10 CFR Part 70, 40.31(j) of 10 CFR Part 40, and 30.32(1) of 10 CFR 30, as appropriate. BWFC accepts this as a condition. 43) On or before July 31 1991, the licensee shall evaluate his safety program against the " Guidance on Management Controla/ Quality Amaurance, Requiremunta for Operation, Chemical Safety, and Fire Protection for Fuel Cycle Facilities", Federal Register, March 21, 1989, and shall propose license conditions, na appropriate, to modify the license in accordance with the Guidance...." ] BWFC accepta this se a condition.

DRAM ATTACll!1Elli 11 B&J FUEL COMPANY, COMMERCIAL NUCLEAR FUEL PLANT USNRC LICENSE SNM-1168, DOCKET 70-1201 PART I - CHAPTER 1.0 - STANDARD CONDITIONS & SPECIAL AUTHORIZATIONS 1.6 Glossary of Terms 1.6.4 CNFP - Commercial Nuclear Fuel Plant of the B&W Fuel Company. 1.6.5 Controlled Area - An area that has the potential W for radiocontaminants above the unrestricted release g{$ limits specified in 1.7.4 of this document. 1.6.6 Fuel Assembly / Assembly - A grouping of fuel rods into a rigidly restrained configuration suitable for use in reactors. 1.6.7 Fuel Pellet / Pellet - right circular cylinders of uranium oxide that have been compacted. They may be sintered or unsintered. 1.6.8 Fuel Rod / Rod - Tubing that has been loaded with fuel pellets. 1.6.9 Geometrically Safe container - A container with a dimension of 9 5/8 inches ID x 11 inches high, or other cylinder demonstrated to be safe, authorized for uranium in any form. 1.6.10 Qualified Designee-An individual who at the g minimum possess the qualifications of the position that they are filling in for in the absence of the ($ normally assigned individual. 1.6.11 Radiation Worker - An employee whose job has a significant potential for exposure to radiation or who.is required to handle SNM. 1.6.12 Removed from Service - Process equipment which is removed from production for replacement repairs or modifications which may affect safety related controls, or which is in " storage" for a period of 4 months or more whether or not modifications have been undertaken. " Removed from service" is not taken to include equipment which remains on line but is infrequently used, or which is temporarily bypassed as a result of routine process requirements. 1.6.13 Safe Mass - The maximum safe mass, independent of geometry, degree of water moderation, and reflection is defined as 850 grams of U-235. PAGE: 1-4 DATE: 6-22-90 REV.: O SUPERSEDES: PAGE: DATE: REV.

B&W FUEL COMPANY, COMMERCIAL NUCLEAR FUEL PLANT USNRC LICENSE SNM-1168, DOCKET 70-1201 PART I - CHAPTER 1.0 - STANDARD CONDITIONS & SPECIAL AUTHORIZATIONS 1.6 Glossary of Terms 1.6.14 Safe Volume - Defined as 14-liter maximum capacity for 4.10 w/o uranium in any form. 1.6.15 Tube - Zircaloy or stainless steel tubing used as fuel pellet cladding. 1.6.16 Uncontrolled Area - An area where the contamination levels exceed the limits specificed in 3k3 1.7.4. 1.7 Specific Exemptions and Special Authorizations 1.7.1 Postings A continued exemption is requested from the labeling and posting requirements of 10 CFR 20.203(e) and 20.203 (f) because of the nature of our operation. M The intent of this section is met by posting areas which house or temporarily store radioactive material 74 with signs incorporating the radiation symbol and the 'p following warning: CAUTION RADIOACTIVE MATERIAL ANY AREA OR CONTAINER WITHIN THIS PLANT MAY CONTAIN RADIOACTIVE MATERIAL. This exemption is based on practicality and/or experience and has been applied effectively at CNFP for the past 20 years. 1.7.2 Storage of-UF Cylinders g UF cylinders may be stored outside the CNFP main g building in a planar array. This storage area is exempt from the criticality monitoring requirements of 10 CFR 70.24. 1.7.3 Storage of Fuel Assemblies Within Model B Shipping containers Fuel assemblies stored in accordance with NRC i licensed shipping containers certificate of Compliance No. 6206 are exempt from the criticality tF monitoring requirrments of 10 CFR 70.24 provided: [j) PAGE: 1-5 DATE: 6-22-90 REV.: 0 SUPERSEDES: PAGE: DATE: REV.:

DRAfl B&O FUEL COMPANY, COMMERCIAL NUCLEAR FUEL PLANT USNRC LICENSE SNM-1168, DOCKET 70-1201 PART I - CHAPTER 1.0 - STANDARD CONDITIONS & SPECIAL AUTHORIZATIONS 1.7 Specific Exemptions and Special Authorizations j The containers and contents have been subject to i the inspections and determinations required by 10 CFR 71, Subpart D. The containers are sealed, properly identified, i and meet shipment requirements. A minimum of 38" edge-to-edge separation is maintained between the loaded container array and other accumulations of SNM. d The loaded container array has a minimum 2fp separation of 12 feet. 1.7.4 Free-Release Limits B&W Fuel Company is authorized to release equipment and material from the controlled areas to the uncontrolled areas in accordance with " Guideline for I Decontamination of Facilities and Equipment Prior to Release for Unrestricted Use or Termination of 5(h Licenses for By-product, Source or Special Nuclear Material", USNRC, August 1987, Exhibit A. j 1.7.5 Personnel The incumbent plant manager is deemed to have the S\\\\ equivalent of the requirements outlined in section 2.2.1. ~ i 1 I PAGE: 1-6 DATE: 6-22-90 REV.: 0 SUPERSEDES: PAGE: DATE: REV.:

DRAFT. B&J FUEL COMPANY, COMMERCIAL NUCLEAR FUEL PIANT USNRC LICENSE SNM-1168, DOCKET 70-1201 PART I - CHAPTER 1.0 - STANDARD CONDITIONS & SPECIAL AUTHORIZATIONS EXHIBIT A PAGE 1 OF 4 GUIDELINES FOR DECONTAMINATION OF FACILITIES AND EQU!PMENT PRIOR TO RELEASE FOR UNRESTRICTED USE OR TERMINATION OF LICENSES FOR BYPRODUCT. SOURCE, OR $PECIAL NUCLEAR MATERIAL U.S. Nuclear Regulatory Consission Division of Industrial and Medical Nuclear Safety Washington, DC 20555 ~ August 1987 PAGE: 1-7 DATE: 6-22-90 REV.: 0 SUPERSEDES: PAGE: DATE: REV.:

B&W FUEL COMPANY, COMMERCIAL NUCLEAR FUEL PLANT USNRC LICENSE SNM-1168, DOCKET 70-1201 PART I - CHAPTER 1.0 - STANDARD CONDITIONS & SPECIAL AUTHORIZATIONS EXHIBIT A PAGE 2 OF 4 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 equigaent prior to abandonment or release for unrestricted use. The limits in Table 1 do not apply to premises. equipment, or screo containing induced radioactivity for which the rediological considerations pertinent to their use may be different. The release of such facilities or items free 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, es detemined 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 minimite the contamination prior to use of any covering. 3. The radioactivity on the interior surfaces of pipes, drain lines, or I ductwork shall be determined by making measurements at all traps, and q other appropriate access points, provided that contamination at these locations is likely to be representative of contamination on the interior j 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 Comnission may authorire a licensee to relinquish possession or control of premises, equipment, or scrap having surfaces contaminated with materials in excess of the limits specified. This may i include, but would not be limited to, special circumstances such as rating i of buildings. transfer of promises to another organization continuing work with radioactive meterials, or conversion of facilities to a long-tem storage or standby status. Suca requests must: ~ ~ a. Provide detailed. specific infomation describing the premises, J equipment or scrap, radioactive conteminants, 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. l PAGE: 1-8 DATE: 6-22-90 REV.: 0 SUPERSEDES: PAGE: DATE: REV.:

D B&W FUEL COMPANY, COMMERCIAL NUCLEAR FUEL PLANT USNRC LICENSE SNM-1168, DOCKET 70-1201 J PART I - CHAPTER 1.0 - STANDARD CONDITIONS & SPECIAL AUTHORIZATIONS EXHIBIT A PAGE 3 OF 4 5. Prior to release of premises for unrestricted use, the licensee shall make a comprehensive radiation survey which establist,as that contemination is within the limits specified in Table 1. A copy of the survey report shall be filed with the Divisten cf Industrial and Medicel Nuclear $4fety. l U. S. Nuclear Regulatory Conahston. Washington DC 20$55, and also th2 Admiristrator of the NRC Regional Office having jurisdiction. The report l should be filed at least 30 days prior to the planred date of abandownt. The survey report shall: 4. 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. Followirg review of the report. the NRC will consider visiting the facilities to confim the survey. PAGE: 1-9 DATE: 6-22-90 REV.: 0 SUPERSEDES: PAGE: DATE: REV.:

i B&W FUEL COMPANY, COMMERCIAL NUCLEAR FUEL PLANT USNRC LICENSE SNM-116Br DOCKET 70-1201 j PART I - CHAPTER 1.0 - STANDARD CONDITIONS & SPECIAL AUTHORIZATIONS EXHIBTT A PAGE 4 OF 4 1ABLt 1 ACC(PTABLE $URf Att (DNTAMINAtl0N LtV[L$ NUCL10(18 AV(hAGEb c f Mhtigab d f PINOVAOLtb e f U-nat. U 235. L' 238. and t 15.000 dpa e/100 ce! 1.000 dem e/100 ce! associated decay products 5.000 den e/100 ce fransuranics. Ra.!!6. ks.!!B. Th !30. Thotti. Pa 231 100 dpm/103 ce! 300 dpm/100 ce! 20 dom /100 cet Ac !!!. l.Ill.1 129 Th net. Th 232. $r.g0, h.!!3. Ra !N. O232.1 124 1000 dpa/lD0 ce! 3000 dpm/100 cmI 200 dpa/100 cmI 1 131. l.133 h ts.gesse eettlers (nuclides tith decer modes other than alpha emission or spontaneous 5N dpa ev/100 ca! 16.000 dem ev/100 ce! 1000 dpa 05/l00 cm2 fission) encept St go and others noted above. 6Where surface coatselnation by both alpha. and beta-gasme emitting nuclides entsts, the limits established for alpha. and beta.genau-emitting nuclides should apply indepeneently, bas used in this table, dpa (dtsintegrations per minute) means the rate of emission by radioactive uterial as detersthed by correcting the counts per minute observed by as appropriate detector for background ef ficiency. and geometric factort associated alth the instrumentation. RNuasurements of average contaminant should not be averaged over more than I square seter. For objects of less surface area, the average should be derived for each such object. dihe matsum contaetnation level applies to an area of not more than 100 cat, 'The amount of removable radioactive material per 100 cm2 of surface area should be determined by wiping that area with dry filter or sof t ahteeMat ptMr( applying moderate pressure, and assessing the amount of radioactive material on the wipe with an appropriate instrument of known efficica y. ep removable contamination on objects of less surf ace area is determined, the pertinent levels should be reduced proportionally and the entire s#fa;a should be w.tped. Ifhe average and maataan radiation levels associated with surface contamination resulting from beta.pasma emitters should not exceed 0.2 arad/hr at I cs and 1.0 mead /hr at I cm. respectively, esasured through not sore than ! milligrams per sware centimeter of total absorber. PAGE: 1-10 DATE: 6-22-90 REV,t 0 SUPERSEDESI PAGE: DATE: REV.:

B&W FUEL COMPANY, COMMERCIAL NUCLEAR FUEL PLANT 1RAET-USNRC LICENSE SNM-1168, DOCKET 70-1201 CHAPTER 2.0 ORGANIZATION AND ADMINISTRATIVE PART I e 2.1 Organizational Responsibilities and Authority 2.1.1 Management It is the responsibility of the Plant Manager to assure the safety of the operation and compliance with license conditions. Control shall be established by: - designation of responsibility to qualified personnel - review and approval of Health-Safety procedures review of program effectiveness a - prompt correction of nonconforming conditions The CNFP management structure is as shown in Figure 2.1. FIGURE 2.1 ORGAN!!ATION CHANI g,,,% w f...........g l-. ev,...it, .I l letet, avvian bare Ow. l l Presortin ansps l l namese. ensitt,4 sekt, l ansinistratie menspo l t l osuretimal susnism l l nye..nsatin p* vites a tse f IIEALTH4AFETf l nessin setety samm l - 5(CTICil l imelth 6 fot, neitm I i l

• (e.g.. Manuf acturing. $pecialty Hfg.. Field Operations) l
  • (e.g.. Accounting. Personnel. Purchasing. Data Processing)
    • (not directly responsible for production functions)

PAGE: 2-1 DATE: 6-22-90 REV.: 0 SUPERSEDES: PAGE: DATE: REV.:

F B&O FUEL COMPANY, COMMERCIAL NUCLEAR FUEL PLANT DRAET. USNRC LICENSE SNM-1168, DOCKET 70-1201 CHAPTER 2.0 ORGANIZATION AND ADMINISTRATIVE PART I 1 2.1 Organizational Responsibilities and Authority 2.1.2 Production Managers Production Managers report directly to the plant manager. They are responsible for managing operational (F area supervisors and are responsible for production functions. gg The Production Managers shall have at, a minimum, a bachelor's degree in science or engineering, followed by two years experience in the nuclear industry. 2.1.3 operational Area Supervision Operational area supervision in that supervision directly responsible for the control of materials, personnel, equipment, and activities in specific areas. Those responsibilities include assuring that approved control procedures developed by Health-Safety shall be available in writing to operators and other concerned personnel and shall be adhered to. Minimum qualifications of operational area supervision shall includes (a) A high school education e.nd a minimum of 2 years experience in the nuclear industry. Experience shall include the practical application of criticality control techniques and a familiarity with the applicable specific limitations imposed on CNFP operations. 2.1.3 The Health-Safety Section The llealth-Safety Section shall be responsible to interpret the license conditions, provide monitoring facilities, develop safe operation guidelines, maintain training programs, and review and approve operating procedures to assure safe operation and license compliance. These responsibilities include maintenance of nuclear sofety and radiation safety with the approval authority limited to authorized specific or general license conditions. The Health-Safety section shall not be directly responsible for the performance of manufacturing operations. PAGE: 2-2 DATE: 6-22-90 REV.: 0 SUPERSEDES: PAGE: DATE: REV.:

f: B C FUEL COMPANY, COMMERCIAL NUCLEAR FUEL PLANT DRAFI USNRC LICENSE SNM-ll68, DOCKET 70-1201 PART I ORGANIZATION AND ADMINISTRATIVE CHAPTER 2.0 2.1 Organizational Responsibilities and Authority The Manager, Quality and Safety or his qualified designee shall be responsible to provide management with assurance of the effectiveness of the safety program by maintaining an audit program that includes periodic inspection of controls and operations, reports to management, follow-up of nonconforming conditions and necessary documentation (see Audits, Section 2.7). 2.2 Personnel Education and Experience Requirements 2.2.1 Plant Manager The Plant Manager shall have a Bachelor's Degree in SI Science or Engineering, a minimum of 10 years experience in the nuclear industry, and 5 years experience in ll, management. 2.2.2 Manager, Quality & Safety The manager to whom the section reports shall have a Bachelor's Degree and a minimum of five years experience, which would develop an understanding of nuclear and radiation safety. Such experience shall be of a nature which demonstrates to the Plant Manager sufficient judgment and capability to establish and maintain an effective nuclear criticality and radiation safety program for the activities authorized by license. 2.2.3 Health Physicist The Health-Safety Section shall include a person who shall act as the plant Health Physicist. This person shall have a Bachelor's Degree in Science or Engineering. A minimum of 2 years experience in applied health physics is required along with sufficient formal 1 training that provides an understanding of the health physics and nuclear safety hazards involved at the CNFP. 2.2.4 Health-Safety Foreman The Health-Safety Foreman shall have a high school education and three years experience in radiation safety which would develop an understanding of nuclear and radiation safety. PAGE: 2-3 DATE: 6-22-90 REV.: 0 SUPERSEDES: PAGE: DATE: REV.:

B&W FUEL COMPANY, COMMERCIAL NUCLEAR FUEL PLANT USNRC LICENSE SNM-1168, DOCKET 70-1201 CHAPTER 2.0 ORGANIZATION AND ADMINISTRATIVE PART I 2.2 Personnel Education and Responsibilities and Authority 2.2.5 Health-Safety Monitors Health-Safety Monitors set up and conduct routine monitoring, sample collection and analytical tests in the plant to determine whether the amount of radioactivity is within acceptable limits and assists in verifying the radiological and industrial safety of employees. The Health-Safety Monitors shall have, as a minimum, a high school diploma or GED equivalent with six months of H y experience as a radiation monitor. They may fulfill the experience requirements on the job as a Health-Safety gg, Monitor trainee. 2.2.6 Nuclear Criticality Safety Specialist The Nuclear Criticality Safety Specialist is a separate component within the corporate structure and thus is organizationally independent of the CNFP, with no interest in plant operations, other than the nuclear criticality safety aspects. The Nuclear Criticality Safety Specialist is responelble for evaluating the basic nuclear criticality safety limitations upon which plant safety was originally assessed, potential changes, validity of assumption, and accuracy of results. The minimum qualifications for the Nuclear Criticality Safety Specialist shall be a Bachelor's Degree in Science or Engineering and a minimum of two years experience in nuclear reactor physics and one year experience in-nuclear criticality analysis or two years experience performing nuclear criticality safety analyses. 2.3 Safety Review Board The Safety Review Board reviews the following as a minimum on a quarterly basis o New or revised facilities o Analysis of equipment and processes involving hazardous materials o Maintenance of fire safety 24 PAGE: 2-4 DATE: 6-22-90 REV.: 0 SUPERSEDES: PAGE: DATE: REV.:

B C FUEL COMPANY, COMMERCIAL NUCLEAR FUEL PLANT USNRC LICENSE SNM-ll68, DOCKET 70-1201 ORGANIZATION AND ADMINISTRATIVE PART I CHAPTER 2.0 2.2 Personnel Education and Experience Requirements o The continuing effectiveness of established controls and safeguards o Maintenance of ALARA criteria (review of quarterly air sample averages, review of surface contamination surveys) o Safety-related audit and inspection findings 2.3 Safety Review Board o Other items (such as abnormal occurrences) that Safety Review Board members wish to discuss. The Safety Review Board Chairman shall have a Bachelor's Degree in Science or Engineering and a minimum of five years MT; experience in responsible positions which would develop an gy understanding of nuclear and radiation safety. The Safety Review Board Chairman shall be directly responsible to the Plant Manager for the proper conduct of the Safety Review Board. The Plant Manager shall be kept informed in writing of Safety Review Board action. The permanent membership of the Board shall consist of representatives from production management, Quality and Safety and others as deemed g, necessary by the chairman. Technical representatives of outside consulting organizations shall be included as necessary. Board meetings may be convened at the discretion of the Safety Review Board Chairman, but shall be held at least quarterly. The Safety Review Board Chairman shall decide whether or not the necessary disciplines are present during a board meeting to evaluate the item (s) under consideration. There shall be a minimum of 4 Safety Review Board members present during a board meeting. The Safety Review Board Chairman reviews all requests for changes in process and equipment which involve hazardous materials and determines if Board review is necessary. In the case of minor changes where existing safety practice remains the same, the Safety Review Board Chairman may determine that Board review is not necessary. Safety Review Board members shall be kept appraised of actions taken by the Safety Review Board Chairman on such minor changes. PAGE: 2-5 DATE: 6-22-90 REV.: 0 SUPERSEDES: PAGE: DATE: REV.:

t DRAET. r B&W FUEL COMPANY, COMMERCIAL NUCLEAR FUEL PLANT USNRC LICENSE SNM-1168, DOCKET 70-1201 ORGANIZATION AND ADMINISTRATIVE CHAPTER 2.0 PART I 2.3 Safety Review Board Where other than minor changes are involved, the Safety Review Board review and approval process shall be conducted in accord with procedures are approved by the Plant Manager. Records of Safety Review Board proceedings, including supporting calculations and approvals, shall be retained for 2 years minimum after the completion or termination of the subject activity. An annual ALARA report shall be prepared under the direction l of the Manager, Quality and Safety. The report shall be submitted to the Safety Review Board in which they will review h; to determine: 1) if there are any upward trends developing -in personnel exposures (internal and external) for identifiable categories of workers, types of operations, or 7d effluent releases; 2) if exposures and releases might be lowered in accordance with the ALARA concept; and 3) if equipment for effluent and exposure controls is being properly used, maintained, and inspected. A copy of the report shall be sent to the Plant Manager along with the results of the review and recommendations. l At least every two years, the Safety Review Board shall pr I evaluate the the effectiveness of the radiation / nuclear safety training program. \\9 2.4 Approval Authority for Personnel Selection i Personnel selection for those CNFP staff level positions shall be approved by the Plant Manager. 2.5 Training i Initial indoctrination of employees to nuclear and radiological safety shall be the responsibility of Health-Safety and shall conform with 10 CFR 19. Initial indoctrination training shall, aa a minimum, include the following topics: i o license conditions L o federal regulations o operating procedures o radiation safety o nuclear safety o emergency procedure yr o chemical and fire safety PAGE: 2-6 DATE: 6-22-90 REV.: 0 SUPERSEDES: PAGE: DATE: REV.: 1

L l B C FUEL COMPANY, COMMERCIAL NUCLEAR FUEL PLANT DIDkb1 USNRC LICENSE SNM-1168, DOCKET 70-1201 PART I CHAPTER 2.0 ORGANIZATION AND A'sMINISTRATIVE 2.5 Training The extent and depth of the training, relative to the detailed aspects of the health physics and nuclear safety programs, is dependent on the employee's job assignment and potential exposure to radioactive materials as determined by Health-Safety. The initial indoctrination training shall be reinforced (as appropriate to the individual's job assignment) by the employee's immediate supervisor or his designee with respect to individual unit safety requirements, location of emergency exits, contamination control techniques, specific local controls, and operating procedures, prior to the employee being released to operate independently. The employee's immediate supervisor shall complete a new employee training verification form prior to allowing the employee to operate independently. A continuing safety training program shall be conducted by Health-Safety to the extent necessary to assure the maintenance of acceptable safety practices. Such training may be conducted on an individual or group basis. The content of retraining programs may be varied by Health-Safety but will include radiological and nuclear safety as a minimum. Emphasis is placed on new or revised safety criteria or areas in need of reinforcement. A formal retraining of radiation workers shall be conducted at least annually. Documentation of formal training and retraining shall be maintained by y Health-Safety and retained for at least two years, b The Manager, Quality and Safety shall be responsible to assure l that personnel assigned to Health-Safety are properly trained. The extent and depth of the training is based on the l specific job assignment involved. Health-Safety monitoring personnel shall receive a combination of formal and "on-the-job" training such that they can successfully demonstrate their proficiency in basic nuclear and radiation physics monitoring and control techniques and regulatory requirements before being allowed to function without direct oversight. i PAGE: 2-7 DATE: 6-22-90 REV.: 0 SUPERSEDES: PAGE: DATE: REV.:

r-i DRAET. B&W FUEL COMPANY, COMMERCIAL NUCLEAR FUEL PLANT USNRC LICENSE SNM-ll68, DOCKET 70-1201 ORGANIZATION AND ADMINISTRATIVE CHAPTEK 2.0 PART I i 2.6 Operating Procedures Written procedures for the conduct of specific operations including maintenance and development of work within the plant are prepared by the functional component responsible for that activity and shall be reviewed and approved by appropriate production management and Manager, Quality and Safety. g Operating procedures which involve SNM shall be reviewed at least every two years by the appropriate production manager l}I; and Manager, Quality and Safety. Applicable procedures shall be available in the work area and adherence to procedure shall byi be required of all personnel. Procedures for operations where nuclear and radiological safety are involved shall include lh) l specific reference to applicable safety requirements. ( Procedure and format shall be such that operations are clearly detailed and specific directions are provided for operation l under both normal and abnormal conditions. Deviation from written procedures for the handling of radioactive materials shall be approved by the Manager, Quality and Safety, or his qualified designee. Procedural control of activities at the CNFP are categorized as follows: l-Health-Safety Procedures developed by Health-Safety specify the method by which safety related functions are to be accomplished. The procedures shall encompass all health physics activities required by the license. Such procedures may be for internal Health-Safety use or may be intended for general distribution to affected l individuals within other components. As a minimum, Health-Safety procedures shall be approved in writing by the Manager, Quality and Safety as well as approved by affected members of plant management. t SNM Accountability Nuclear Materials Control procedures provide techniques for the accountability and measurement of SNM. As a minimum, such procedures shall be approved in writing by the Manager, Quality and Safety and the Manager, Production and Materials Control. 1 PAGE: 2-8 DATE: 6-22-90 REV.: 0 SUPERSEDES: PAGE: DATE: REV.:

B&W FUEL COMPANY, COMMERCIAL NUCLEAR FUEL PLANT OS USNRC LICENSE SNM-ll68, DOCKET 70-1201 PART I CHAPTER 2.0 ORGANIZATION AND ADMINISTRATIVE I i 2.6 Operating Procedures Other Plant Groups l Procedures from other plant groups (i.e., Manufacturing, Quality Assurance) where nuclear or radiological safety, license conditions, or regulatory requirements are involved require prior approval by the Manager, Quality L and Safety as well as approval by affected members of plant management. New operations and major operational changes shall require the written recommendation of the Safety Review Board Chairman l prior to inaplementation. Revised procedures shall be subject to approval in the same l manner as new procedures. Health-Safety procedures shall be l reviewed at least annually for technical correctness and i applicability. The Manager of Quality & Safety shall use his l discretion to assure that the appropriate personnel of Section i 2.2 performs the procedure review. l Procedure distribution and control shall be in accord with procedures approved by plant management. 2.7 Audits and Inspections An internal Health-Safety inspection program shall ha maintained to provide assurance that plant activities are conducted safely and in accord with license specifications. The Manager, Quality and Safety shall be responsible to assure that the inspection program is conducted effectively. The internal Health-Safety inspection program at the CNFP is structured as follows: Monthly Safety Inspections Health-Safety personnel shall conduct, at least monthly, a formal inspection of plant status relative to safety g related functions to include fire safety, except during plant shutdown of a week or longer. Inspection results j{o shall be documented, reported to plant management and supervision as appropriate, and will be maintained on file by Health-Safety for at least 2 years, i PAGE: 2-9 DATE: 6-22-90 REV.: 0 SUPERSEDES: PAGE: DATE: REV.:

n 1 BC FUEL COMPANY, COMMERCIAL NUCLEAR FUEL PLANT DRN USNRC LICENSE SNM-ll68, DOCKET 70-1201 PART I CHAPTER 2.0 ORGANIZATION AND ADMINISTRATIVE l 2.7 Audits and Inspections The monthly safety inspections shall be conducted by personnel technically qualified to perform this function and in the application of license specifications. Informal Daily Inspections Health-Safety personnel shall, as part of their routine duties, conduct informal daily inspections of plant activities. These inspections are not formally documented unless adverse findings are identified. Other Inspections ventilation, containment, and air cleaning equipment shall be routinely inspected at least annually by Health-Safety personnel to assure continued effectiveness and compliance with license specifications. 1 l-1 Independent Audits l l Independent auditors shall conduct, as a minimum, semi-annually nuclear safety, fire safety and health A physics inspections at the CNFP. These aud'its shall be conducted in accordance with written instructions or 3Io l procedures. The audit scope shall consist of physical inspections and records reviews for the industrial, r I nuclear, and radiological safety elements of plant activities including - effectiveness of procedural controls impacting on operational safety parameters. - audit of operating records, where such records provide a means of verifying procedural compliance with safety specifications. - review and evaluation of contamination survey data. 1 - ascertaining the overall performance of the plant functions in providing adequate controls, surveillance, and follow-up to assure safety and license compliance. 1 l l l l PAGE: 2-10 DATE: 6-22-90 REV.: O SUPERSEDES: PAGE: DATE: REV.:

B&W FUEL COMPANY, COMMERCIAL NUCLEAR FUEL PLANT ()Rbg"' USNRC LICENSE SNM-1168, DOCKET 70-1201 PART I CHAPTER 3.0 RADIATION PROTECTION 3.1 Administrative Requirements 3.1.1 General The business of the CNFP includes radiation hazards from alpha contamination (fuel fabrication) and from beta / gamma contamination (field service work for commercial power utilities). In order to properly discuss the controls used in both facets of our operations, we have split chapter 3.0 to address the alpha contamination controls (Section 3.2) separate from the beta / gamma contamination controls (Section 3.3). The demonstration given in Chapter 12 is also structured j accordingly. l 3.1.2 ALARA It is the policy of the CNFP to keep occupational radiation exposures and radioactive contamination in effluents as low as is reasonably achievable. The responsibility for implementation at the ALARA policy is designated to the Health-Safety Section 3.1.3 Radiation Work Permit (RWP) Procedures Operations which are not covered by an operating procedure and are judged by cognizant supervision as j being likely to exceed the concentration limits specified in Table I of Appendix B to 10 CFR Part 20, shall be covered by a Radiation Work Permit (RWP). The RWP shall be reviewed for industrial safety and U approved by the Health Physicist. An RWP previously approved may be reissued and approved by the 2() Health-Safety Foreman. It shall specify the safety requirements, protective clothing and equipment, and monitoring requircments necessary to assure the operation is conducted in a safe manner. Active RWPs shall be reviewed at least monthly to ensure the operation is conducted in a safe manner and the RWP is still-applicable, l 3.1.4 Written Procedures All license activities related to radiation protection shall be conducted in accordance with approved written procedures. Approval, scope, format and distribution requirements are described in Section 2.6. PAGE: 3-1 DATE: 6-22-90 REV.: O SUPERSEDES: ) PAGE: DATE: REV.: l

B&W FUEL COMPANY, COMMERCIAL NUCLEAR FUEL PLANT N USNRC LICENSE SNM-1168, DOCKET 70-1201 PART I CHAPTER 3.0 RADIATION PROTECTION 3.1 Administrative Requirements 3.1.5 Postings Local safety rules approved by Health-Safety providing personnel and supervision with specific directions essential to assuring radiation safety shall be posted in areas where appropriate. 3.1.6 Personnel Monitoring Systems Personnel monitoring devices are selected by Health-Safety on the basis of radiation type and l sensitivity requirements, and may include TLD, or pocket chambers. Neutron dosimeters shall be used by operators and other persons as deemed necessary by Health-Safety when a potential for measurable neutron exposure exists. Rapid dosimetric evaluation of neutron exposure in the event of accidental criticality is provided by indium j foil issued to personnel as part of their standard dosimetric package. Dosimeters are issued to all 1 designated plant employees and selected visitors based on the requirements of 10 CFR 20, with routine monthly or quarterly exchanges. Use of special dosimeters may be required by Health-Safety where unusual exposure levels may be encountered, as in source manipulation. An indication of 100 mR/or more on a pocket chamber requires the expeditious evaluation of the employee's TLD or film. Extremity exposure is monitored using TLD's as determined necessary by the Health Physicist, or Health-Safety Foreman. TLD dosimeters are processed by a vendor with monthly or quarterly printout forwarded i to CNFP. More rapid notification is available by telephone if needed. Radiation exposure and absorbed dose to employees are derived based on dosimetric results and in accord with 10 CFR 20. Health-Safety shall review and maintain the dose records and prepare such reports as are required by regulations. PAGE: 3-2 DATE: 6-22-90 REV. 0 SUPERSEDES: PAGE: DATE: REV.:

B&W FUEL COMPANY, COMMERCIAL NUCLEAR FUEL' PLANT USNRC LICENSEz SNM-1168, DOCKET 70-1201 O CHAPTER'3.0 PART I RADIATION PROTECTION

3.2 Technical

Requirements - Alpha Contamination following major maintenance. The minimum acceptable system efficiency shall be 99.9%. The ventilation system shall incorporate the following requirements. Air recirculated back into the controlled' area is sampled on a continuous basis to verify filter effectiveneas. Air will not be recirculated if levels are above 25% MPC of 10 CFR 20 Appendix B. At least one filter housing or bank in each system shall be equipped with a device for monitoring differential pressure. Differential pressure shall. be checked weekly and filters replaced when damage is evident, or when the L differential pressure exceeds 4 inches of water. Gaseous effluents shall be representative 1y sampled for gross alpha on a continuous basis, and when the facilities are in. operational status, the samples shall be collected daily and counted after allowing for decay of radon and its daughters. 3.2.2.2 Uncontrolled Area Air Effluent Limits l r Compliance.with the following-release limits shall be maintained-in order to assure that airborne releases to uncontrolled areas are maintained as low as reasonably achievable. ( I PAGE: 3-5 DATE: 6-22-90 REV.: 0 SUPERSEDES: PAGE: DATE: REV.:

i B&W FUEL COMPANY, COMMERCIAL NUCLEAR FUEL PLANT DRAfl USNRC LICENSE SNM-1168,_ DOCKET 70-1201 CHAPTER 3.0 RADIATION PROTECTION PART'I 3.3 Technical' Requirements - Beta / Gamma Contamination b. Contaminated equipment that exceeds the Radiation i control Zone limits in Table 1 shall be-I decontaminated prior to maintenance or gp calibration. Clean-up operations shall commence no later than the start of the next work shift. 74, c. During periods of maintenance on the contaminated equipment, radiological surveys of the RCZ shall be performed at least daily or more frequently if considered necessary by Health-Safety. The levels may not exceed Table 1. When operations are-l complete, Item 3 of Table 1 must be below 5,000 i dpm/100-cm2. d. The change room shall be surveyed each operational fg day and shall not exceed the limits established in Table 2. 25 j e. During periods of long term inactivity, weekly ) surveys of the RCZ for external radiation and a removable contamination shall be performed, except during CNFP plant shutdown of a week or i longer, but not to exceed 3 weeks. f. Instrumentation shall be calibrated as indicated in Chapter 3.2.4.4. - i g. Gaseous effluents to unrestricted areas shall be controlled to the limits specified in 10 CFR 20, 1 Air samples shall be collected daily when'the I facilities are in an operational status. HEPA-filtered containment. facilities shall be used j 1 in accord with good Health Physics practice as required by Health-Safety to control airborne j adioactivity. HEPA filters shall be changed when the differential pressure exceeds 4" of gater. Differential pressure shall be recorded weekly when containment facilities are in use. J.o.4 Personnel Exposure Control a. All personnel entering the RCZ shall wear as a minimum protective lab coats and shoe covers. All personnel handling contaminated equipment shall also wear protective gloves. Upon exiting the RCZ, personnel shall survey to assure they l meet unrestricted area release limits given in Table 2 of this part. PAGE: 3-15 DATE: 6-22-90 REV.: 0 l SUPERSEDES: i PAGE: DATE: REV.:

B&W FUEL COMPANY, COMMERCIAL NUCLEAR FUEL PLANT gd USNRC LICENSE SNM-1168, DOCKET 70-1201 PART I CHAPTER 3.0' RADIATION PROTECTION i 3.3 Technical Requirements - Beta / Gamma Contamination Radiation workers shall receive initial indoctrination trajning before they are allowed to work in the RCZ. This training, as well as the annual retraining, is as described in Section 2.5. b. Radiation Areas and High Radiation Areas will be established as appropriate and maintained as described in 10 CFR 20.202 and 10 CFR 20.203. c. Respi?*. tory protection shall be used in unusual-cond1Laons where airborne concentrations of radioactive natorial may cause personnel to receive exposures in excess of permissible levels. . Typical unusual conditions may be certain maintenance operations, short-term operations, etc. Procedures for maintenance of the respiratory protection system and training of personnel using respiratory protection shall be under the cognizance of Health-Safety. The respiratory protection program shall be conducted in accord with 10 CPR 20.103. d. Personnel and/or fixed air samplers shall be used to determine exposure to airborne radioactivity. D' Fixed air samples shall be used to sample air continuously. Analyses for airborne concentrations 73g3 of radioactivity shall be conducted after each operational shift. Health-Safety shall specify ' frequency of personnel sampling based on operational experience and Health Physics judgement to keep exposures within 10 CFR 20.103(a) limits. e. Prior to Health-Safety approval of an individual to work in the RCZ, the individual's occupational radiation exposure history (both internal and external) shall be verified to ensure regulatory limits are not exceeded. l I PAGE: 3-16 DATE: 6-22-90 REV.: 0 SUPERSEDES: PAGE: DATE: REV.:

B&W FUEL COMPANY, COMMERCIAL NUCLEAR FUEL PLANT C$lbp-{'

USNRC LICENSE SNM-ll68, DOCKET 70-1201 CHAPTER 3.0 JPART I. RADIATION PROTECTION c 1 II TABLE 1: RADIATION CONTROL ZONE (RCZ) ~ LIMITS Beta-Gamma n Item Radiation Intensity Removable Contamination

l. Protective. Cloth-ing Available for Use

.5 mR/hr. at 1 cm 22,000 DPM/100 cm (2)

2. Equipment 2

Maintenance 50 mR/hr. at 1 ft. 500,000 DPM/100 cm

3.-Floors, Walls, and other 2

Surfaces 1.0 mR/hr. at 1 ft. 20,000 DPM/100 cm III Alpha contamination will be controlled as indicated in Section 3.2.6. (2)As measured by direct survey. PAGE: 3-19 DATE: 6-22-90 REV.: 0 SUPERSEDES: $PAGE: DATE: REV.: g p .1.

p B&W. FUEL COMPANY, COMMERCIAL NUCLEAR FUEL PLANT USNRC; LICENSE SNM-1168, DOCKET-70-1201 RADIATION PROTECTION CHAPTER 3.0-PART I nn /JT ( TABLE 2!: UNRESTRICTED AREA RELEASE LIMITS Beta-Gamma Item Radiation intensity Removable Contamination l

1. Skin, Hair,.and Personal (2) clothing Min. Detect. Level Min. Detect. Level 2

2. Equipment 0.2 mR/hr. at 1 cm' 1,000 DPM/100 cm

3. Floors,. Walls,

-Change Room and '2 M Other. Surfaces-0.2 mR/hr. at 1 cm 1,000 DPM/100 cm 25-III Alpha contamination will be controlled as indicated in Section.3.2.6. (2) As measured by direct survey. 4 PAGE: 3-20 DATE: 6-22-90 REV.: 0 SUPERSEDES: PAGE: DATE: REV.:

B&W FUEL COMPANY, COMMERCIAL NUCLEAR FUEL PLANT ggpy}

USNRC' LICENSE SNM-1168, DOCKET 70-1201 NUCLEAR CRITICALITY SAFETY CHAPTER 4.0 PART I 4.1 Administrative conditions 4.1.1 Design Philosophies i

= The double contingency principle as defined in the American Nuclear Standard ANSI /ANS-8.1 shall be followed-4t I in establishing nuclear criticality safety for all equipment, systems and operations. Process designs shall 2f incorporate sufficient factors of safety to require at least-two unlikely, independent, and concurrent changes in process' conditions before a criticality accident is possible. _ Where possible and practicable, reliance will be placed on equipment design in which dimensions (i.e., favorable geometry) are limited rather than on administrative controls. Where structural integrity is necessary to provide assurance for safety, the design and construction of the equipment will be made with due regard to abnormal loads, accidents and_ deterioration. 4.1.2 Criticality Safety Analyses With respect to the overall plant nuclear criticality safety, the Manager, Quality and Safety is responsible for controlling all modifications and/or additions to any-operation, system or equipment. Nuclear criticality safety evaluations are performed by qualified nuclear criticality safety specialists.- These specialists.must have a B.S. Degree in Science or Engineering and either a minimum of two years experience performing nuclear criticality safety analyses or a minimum of two years experience in reactor physics and one year experience performing nuclear criticality safety; analyses. Individuals not satisfying the above requirements may perform safety evaluations provided the evaluations are approved in writing by a qualified nuclear criticality safety specialist. All nuclear criticality safety evaluations shall be independently reviewed by an individual meeting the i qualifications of nuclear criticality safety specialist as defined above with two years of experience as a EI Nuclear Criticality Specialist. All evaluations shall l4 - include an appropriate statement of this review. Both the analyzing and reviewing nuclear criticality safety specialist are independent of CNFP manufacturing operations. A library of validated computer codes and cross sections shall be maintained and utilized for performing nuclear criticality safety evaluations. PAGE: 4-1 DATE: 6-22-90 REV.: 0 SUPERSEDES: PAGE: DATE: REV.:

o }',' 4< Ao .B&W FUEL COMPANY, COMMERCIAL NUCLEAR FUEL PLANT 'DRN1 M USNRC LICENSE SNM-1168, DOCKET 70-1201 NUCLEAR CRITICALITY SAFETY CHAPTER 4.0 PART I -r +p W 4.1 Admi~nistrative Controls 4.1.2 Criticality Safety Analyses M Clarifications and conformance of-limits established by license or by the nuclear criticality safety evaluation Zl3 are made by qualified personnel in the Health-Safety ks Section addressed in 2.2. Similarly, determinations of i safe spacing utilizing the surface density model (Section 15.2.5) may be made by-Quality and Safety personnel provided the fraction. critical and reflected critical surface density for each unit has already been determined e in a previous nuclear criticality safety evaluation. All surface density calculations shall include an appropriate statement of review performed by the Safety Review Board 7 Chairman. 4.1.3 Approvals and Documentation Modifications of product or process are authorized by the Safety Review Board. Operation of the Board and procedures for modification are described below: [ 4.1.3.1 Authorized Modifications Authorized modifications shall be limited to the following areas: i: a. Modification of product' specifications -(except enrichment) as described in 4.2.1 and therefore in the safety criteria for ..l fuel assembly ~ processing, storage and packaging. t" b. Changes in the SNM handling and fuel rod q; processing areas' limited to: Relocation or expansion of fines and rs ti scrap storage areas

y Relocation, arrangement, redesign, or hi addition'of equipment i

n N '" I l PAGE: 4-2 DATE: 6-22-90 REV.: 0 pg SUPERSEDES: PAGE: DATE: REV.: o + l-

I i _BsW FUEL COMJANY, COMMERCIAL NUCLEAR FUEL PLANT -USNRC LICENSE SNM-ll68, DOCKET 70-1201 ~ NUCLEAR CRITICALITY-SAFETY PART I CHAPTER 4.0 4.1-Administrative Controls 4.1.3.3 Review and Approval Process 1 a. The proposal shall be reviewed by the Safety Review Board Chairman or his qualified designee for content, it completeness, and conformance with previously evaluated conditions. If gj3 ' required, the proposal is forwarded for formal nuclear criticality safety evaluation as described in 4.1.2. b. The criticality safety analysis is performed in accord with 4.1.3.2 and' results are forwarded in writing to the Safety Review Board Chairman or his qualified designee. All criticality safety evaluations are reviewed as specified in Part 4.1.2 of this section.. c. The results and recommendations obtained-from the criticality safety evaluations shall be reviewed by the Safety Review Board and/or the Safety Review Board Chairman and the permissibility of the change determined and documented. d. Appropriate personnel are notified in writing of the result of the review. e. The proposed change shall not be implemented until a pre-operational audit -(Ref. 4.1.6) has been satisfactorily ft completed. The individual.who performed '9 2 either the nuclear criticality evaluation or the independent review shall participate in the pre-operational inspection. f. Correspondence, calculations, and other j material shall be maintained on file for a minimum of 2 years or six months following termination of the operation whichever is longer. l PAGE: 4-4 DATE: 6-22-90 REV.: 0 SUPERSEDES: PAGE: DATE: REV.: J

C&W FUEL COMPANY,. COMMERCIAL NUCLEAR FUEL PLANT g(( aft.- USNRC LICENSE SNM-1168,. DOCKET 70-1201 CHAPTER-4.0~-- ~ NUCLEAR CRITICALITY SAFETY PART I 4.2 Technical Criteria TABLE 4.1 U-235 Enrichment Maximum Slab Thickness (percent)- (inches) 1 2.5 5.4 .> 2.5 1 3.0 4.8 > 3.0 1 3.5 4.4 > 3.5 1 4.1 4.0 The vault is made up of several storage cubicles..Each cubicle is separated from each other by.at least 8 inches of concrete plus a neutron poison. This neutron poison shall be equivalent to 35 wt % B C in a.168" thick 3 aluminum plate with an averall minimum density of 2.46 gm/cc. Each cubicle may contain a maximum of two tiers of storage shelves.- Each storage tier may contain up to five - 18" wide shelves for safe geometry slab storage. Vertical spacing between shelves shall be at least 16 inches. -A neutron poison as described above shall be placed on each shelf level except for the bottom. Horizontal spacing between tiers within a cubicle shall be at least 36 inches. Other nuclear safety conditions on pellet vault storage are ae follows: Shelves with multiple enrichments are limited to a maximum slab thickness of 4 inches. The_ vault is separated from other_SNM l storage and processing by eight inch thick concrete walls. l PAGE: 4-12 DATE: 6-22-90 REV.: 0 l SUPERSEDES: L PAGE: DATE: REV.: 1

30. .4 B&W FUEL COMPANY, COMMERCIAL NUCLEAR FUEL PLANT. USNRC LICENSE SNM-1168, DOCKET 70-1201 t, PART:I NUCLEAR CRITICALITY SAFETY CHAPTER 4.0' 4.2 Technical Criteria e 4.2.4.2 Pellet Vault Storage s Material storage and handling in the-vault is controlled based on criteria specified 1 4 by Health-Safety personnel. Enrichment determination and selection of proper slab I/. thickness shall be made utilizing appropriate records or test data. Reject SNM, collected fines, retainer-samples, etc. may be stored in the vault provided the material does not exceed-4.0" slab. Transport of SNM in the pellet vault area shall be with carts applying the same nuclear safety criteria as described for that area. L 4.2.4.3 Fuel Rod Loading The fuel rod loading area is located' adjacent to the pellet vault and may involve handling and storage of unclad pellets in areas other than the vault. Nuclear safety in this area is maintained as safe geometry slabs as described. in 4.2.3.2. Other nuclear safety conditions associated with 4 g~ the fuel rod loading' area are: - All accumulations of SNM are_placed within a 6 ft. vertical space. This six foot distance describes the limit of vertical: displacement-in areas where surface density criteria is applied for nuclear safety. - No SNM accumulation shall be positioned above or below any other SNM accumulation. - SNM in the rod loading area shall:be arranged such that the average SNM slab thickness (presuming all the SNM was uniformly smeared over its assigned area) does not exceed 2.3 inches. L l PAGE: 4-13 DATE: 6-22-90 REV.: 0 SUPERSEDES: -PAGE: DATE: REV.: p L jo

0 B&W FUEL COMPANY, COMMERCIAL NUCLEAR FUEL PLANT l USNRC LICENSE SNM-1168, DOCKET 70-1201 i CHAPTER ~- 4 '. 0 NUCLEAR CRITICALITY SAFETY PART I i j~ 4.2 Technical Criteria i-- 4.2.4.6 Fuel Assembly Storage and Packaging l Fuel assembly dust wrappers, if used, shall be arranged to permit drainage of water from within. Moderation, such as polyethylene, etc., shall not be permitted within the i L . assemblies. No stream sources or sprinklers shall be located near the-strorage array.

• k l

. Restriction that-prohibits any significant quantity of moderating material such as paper, plastic, oil, and etc. within the fuel assembly shall be prominently posted in the boundries of a the assembly storage array. The following nuclear safety restrictions shall 'W ~ l be imposed on fire fighting within the fuel assembly storage arrays 1 - The use.of hoselines shall be prohibited L unless authorization has been received from a management representative of the plant emergency. response-organization. - Simultaneous application of more than one l hoseline is not authorized. - Area postings shall be maintained at the array perimeter specifying limitations on fire fighting techniques. 4.2.4.7 Fuel Assembly Shipping Containers Fuel assembly packaging and unpackaging operations involving licensed shipping containers shall be performed within the L following limitations: - Fuel assembly packaging shall be in accord i with the requirements of the container l certificate. - Fuel assemblies in adjacent containers shall l have a minimum 18" edge-to-edge separation distance. l l l I PAGE: 4-17 DATE: 6-22-90 REV.: 0 L SUPERSEDES: PAGE:- DATE: REV.:

I B&W FUEL COMPANY, COMMERCIAL NUCLEAR FUEL PLANT USNRC LICENSE SNM-1168, DOCKET 70-1201 ENVIRONMENTAL PROTECTION ()Rb -3' PART I CHAPTER 5.0 ' 5.1 Effluent Control Systems A' report providing the effluent results in accordance with M section 5 of Regulatory Guide 4.16 dated December 1985 shall be j5\\ submitted to the NRC semiannually. 5.1.1 Gaseous Effluent Control Gaseous effluents to uncontrolled areas are restricted and monitored as described in Chapter 3. 5.1.2 Liquid Effluent Control Potentially contaminated liquid effluents are normally processed through an evaporator and released as gaseous effluent through the HEPA filtration system. The release limits are the same as those used for the gaseous effluents to uncontrolled areas as described in Chapter 3. Utilizing this process for the management of liquid U effluents, the discharge from the retention tanks shall not exceed 2.5% MPC for uranium. An investigation-f$2 shall be conducted for levels exceeding 2.5% MPC. Circumstances may require that we dispose of liquid effluents through-our retention tank system..If this route is used, the liquid effluent will be analyzed and evaluated for compliance with 10 CFR 20, Appendix B limits prior to release to unrestricted areas as indicated in the following table. LIQUID EFFLUENTS % 10 CFR 20 App. B. "MPC" Action 20% No action required. 21 - 75% Individual releases authorized by Manager, Quality and Safety or his alternate. > 75% Discharge prohibited. Effluent routed for further treatment or disposal. PAGE: 5-1 DATE: 6-22-90 REV.: 0 SUPERSEDES: PAGE: DATE: REV.:

) i B&W FUEL COMPANY, COMMERCIAL NUCLEAR FUEL PLANT C)P AE3-USNRC LICENSE SNM-ll68, DOCKET 70-1201 ENVIRONMENTAL PROTECTION PART I - CHAPTER 5.0 52 Environmental Monitoring An environmental monitoring program shall be conducted to assess the effective control of airborne and liquid effluent it releases to unrestricted areas. Environmental monitoring shall be in accordance with approved proceduros which require the 32E generated data to be evaluated against internal action levels.' The program as shown on Figure 5.1 and Table 5.1, provides coverage to areas immediately currounding the plant,and includes locations in the upwind and upstream directions for-background comparisons. The satuple' results are maintained on

file, i

i i s PAGE: 5-2 DATE: 6-22-90 REV.: 0 SUPERSEDES: PAGE: DATE: REV.: I:

B&W FUEL COMPANY,-COMMERCIAL NUCLEAR FUEL PLANT gh USNRC LICENSE SNM-ll68,~ DOCKET 70-1201 PART 1 CHAPTER 6.0 SPECIAL PROCESSES 6.1 Nonexempt Sealed Source Control 6.1.1 Use of nonexempt sources-for training and instrument calibration shall be limited to, or under the direct control of, the Health-Safety Section. 6.1.2 Sources utilized as a functional component of devices designated for manufacturing and quality control purposes shall be operated only by approved personnel who have been instructed in safe practice by Health-Safety.. Health-Safety shall provide appropriate monitoring support during maintenance-or other operations that may entail increased exposure levels. A register of approved operators shall lx3 maintained in the Health-Safety Office. 6.1.3 Maximum whole body exposure rates in any constantly occupied area in'the vicinity of operating manufacturing or quality control units utilizing by-product material sources shall not exceed 2 mrem /hr.- 6.1.4 In addition to dosimetric devices routinely worn by designated CNFP employees, appropriate self-reading dosimeters shall be utilized by personnel involved in source manipulation in cases where the exposure rate may exceed 2 mrem /hr. 6.1.5 Each sealed source shall be tested for leakage at intervals not.to exceed six (6) months. In.the absence of a certificate from a transferor indicating that a test has been made within six (6) months prior to the transfer, the sealed source shall not be put into use until tested. 6.1.5.1 The test shall be capable of detecting the presence of~0.005 microcurie of contamination on the test sample. The test sample shall be taken from the source or from appropriate accessible surfaces of the device in which the sealed source is permanently or semi-permanently mounted or stored. Records of leak test results shall be kept in units of microcuries and maintained for inspection by the Commission. PAGE: 6-1 DATE: 6-22-90 REV.: 0 SUPERSEDES: PAGE: DATE: REV.:

h J ..c. ' B&W FUEL COMPANY, COMMERCIAL NUCLEAR' FUEL PLANT () USNRC LICENSE.SNM-ll68, DOCKET 70-1201 DECOMMISSIONING PLAN PART,1 -- CHAPTER 7.0 TABLE 7.2 CNFP Plant Area Summary Building / Area Contamination Contamination Equipment Area Potentia) Level Involved Involved ft2 Main Plant' Bldg. South Bay -SERF 1 YES MID-HIGH YES 5700 Pellet-Vault YES LOW YES 1200 Rod Loading YES MID-HIGH YES 1200 Change Rooms YES LOW YES 700 i Rod Assbly & Fab NO 19000 Assbly Storg & Ship NO 8640 Machine Shop NO 7200 Ship & Receiving / Grid NO 10000 Offices NO 16000 Ancillary Area & Structures LSA Building YES LOW NO 300-Retention Tank & Line YES LOW YES 300 SERF 2 YES LOW YES 750 Garage / Maintenance NO 1000 Wet Weather Stream YES LOW NO-24000 UF6 Cyl. Storage YES LOW YES 10800 l l, l R 1 i l I l l l h-l- i l y PAGE: 7-8 DATE: 6-22-90 REV.: 0 l SUPERSEDES: h PAGE: DATE: REV.:

9 ph1 B&W FUEL COMPANY, COMMERCIAL NUCLEAR FUEL PLANT USNRC-LICENSE SNM-1168, DOCKET 70-1201 FACILITY DESCRIPTION l CHAPTER 10.0 PART II' 10.1 Plant Layout and Operations j Figure 10.1 is'a layout of the CNFP illustrating the various production areas on site. The plant's primary function is the manufacture of nuclear fuel assemblies for use in commercial power reactors. These operations may be subdivided into three production phases: -(Unclad SNM Handling, Fuel Rod Processing and Inspection, and Fuel Bundle Assembly). The numbers in parenthesis are taken from Figure 10.1. The CNFP also supports the Field Operations Department for the refurbishemnt of contaminated equipment. The majority of the operations take place in the south bay of the facility known as the Service Equipment Refurbishemnt Facility (SERF). 10.1 Unclad SNM Handling Unclad SNM receiving, storage, and rod loading are located at the south end of the plant, as shown in Figure 10.1. The area includes pellet receiving (#1), the pellet vault (#4), and the pellet loading room (#6). Other than the laboratory (#10), this is the only part of the process in which unclad special a nuclear material (SNM) is handled. The entire pellet-vault / rod loading area is separated from the remainder of the plant by means of concrete block and metal walls. A slight negative pressure is maintained in this area with respect to the rest of the plant to. prevent contamination spread. 10.1.2. Fuel Rod Processing and Inspection Loaded fuel. rods are processed and stored in the central portion of the plant (#7). Processing includes end cap welding, quality control inspection, cleaning (#14), helium leak testing, and accumulation of rods into groups of the number required for a fuel-assembly. Rods are then stored until needed for assembly production. Individual unclad fuel pellets are processed in the laboratory (#10) which is located in this portion of the plant. 10.1.3 Fuel Bundle Assembly Fuel rods are assembled into their final configuration (#11), checked for quality, and shipped to the customer from the north end of the plant (#12). Ancillary production activities conducted within the plant consist of non-nuclear component fabrication which may be characterized as light machining and fabrication. Examples of this type of activity, PAGE: 10-1 DATE: 6-22-90 REV.: 0

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B&W FUEL COMPANY, COMMERCIAL NUCLEAR FUEL PLANT D USNRC' LICENSE SNM-1168, DOCKET 70-1201 CHAPTER 10.0. FACILITY DESCRIPTION PART II 10.4 Radioactive Waste Handling l 10.4.1 Liquid Wastes Potentially contaminated liquids generated at the CNFP are controlled by means of a dedicated evaporation system. The liquid effluent is collected and allowed to evaporate (with heat if necessary) into the existing airborne effluent control system where it is HEPA filtered prior to release. The HEPA system and 10 CFR 20 airborne effluent release limits used are as described in 8.1.1. Vessels used to collect / evaporate the liquid effluent shall be inspected monthly for sludge accumulation. Any dried sludge or other solids collected from the holding / evaporation vessels will be disposed of as LSA waste. As a backup to the evaporation system, we will. maintain a liquid retention tank system that will' collect the contaminated liquid if necessary. _The accumulated' liquids in these retention tanks would be sampled, radiometrically analyzed, and treated as necessary,_ prior to release. The retention tank system incorporates capacity alarms, and air agitation capability. Analytical sensitivity is 1% of the applicable 10 CFR 20, Appendix B, Table II limit. The sampling program is under the control of Health-Safety and no releases are made without the prior approval:of Health-Safety.- The retention tanks are housed in the Rad Waste Retention Buildings shown in Figure 10.1. Figure 10.3 is a schematic of our contaminated liquid waste system. 10.4.2 Solid Wastes Uncontrolled disposal of solid wastes or equipment is authorized'when contamination levels do not exceed the 2 'levles defined in section 1.7.4 and under the concept of ALARA. Establishment of the above contamination limits to permit disposal in accord with routine industrial PAGE: 10-9 DATE: 6-22-90 REV.: 0 SUPERSEDES: PAGE: DATE: REV.:

1-1B&W FUEL COMPANY, COMMERCIAL NUCLEAR FUEL PLANT DRbg' USNRC' LICENSE SNM-ll68, DOCKET 70-1201 FACILITY DESCRIPTION CHAPTER 10.0 PART II 4 10.4 Radioactive Waste Handling 10.4.2 Solid Wastes practice does not present a hazard to the general public..The limits are generally accepted within the nuclear industry, as not-presenting any significant radiological or nuclear safety hazards, ik Routine monitoring programs are. conducted by Health-Safety to assure that material, contaminated in excess of specification limits, is not released for s uncontrolled disposal and to detect and alleviate increasing contamination trends. Non-contaminated solid wastes are disposed of through s a contract hauler. Contaminated solid wastes consist primarily of low specific activity material and are disposed of by a licensed contractor by land burial on an NRC or state licensed site.- LSA wastes are packaged in appropriate containers as required by 10 CFR and 49 CFR. SNM content for each package is estimated using gamma scan or by gross alpha count. 10.5 Fire Protection 10.5.1 General-All CNFP. buildings are of steel and/or masonry construction and the roofs of all main buildings are Class I construction. Class I construction requires that the vapor barrier be non-combustible. Plant operations are typical of metal working type facilities; therefore, very few Class A type combustibles are present.. Accumulations of combustible materials within the CNFP shall be limited to the greatest extent practicable, consistent with operational requirements. Supervision is responsible PAGE: 10-10 DATE: 6-22-90 REV.: 0 SUPERSEDES: PAGE: DATE: REV.:

B&W FUEL COMPANY, COMMERCIAL NUCLEAR FUEL PLANT gg[Af3 'USNRC LICENSE SNM-1168, DOCKET 70-1201 CHAPTER 11.0 PART II-ORGANIZATION AND PERSONNEL 11.1 Organizational Responsibilities Figure 11.1 illustrates the departmental and managerial organization at the CNFP. The key organization responsible for maintaining the health and safety aspects at the CNFP is the Health-Safety Section. The Health-Safety Section is a part of the Quality I and Safety Group. The Health-Safety Section is headed by the Health Physicist who reports directly to the Manager, Quality and Safety. The Manager, Quality and Safety reports directly to the Plant Manager, 11.2 Key Personnel Function 11.2.1 Overall Program Management Responsibility for planning, coordinating, administering and managing the health and safety aspects of the CNFP is vested in the Manager, Quality and Safety. This position is organizationally parallel to other member of the Plant Manager's staff such as the Managers of Manufacturing Engineering and Fuel Operations. 11.2.2 The Health-Safety Section Health-Safety personnel are responsible for the general surveillance of all plant safety related functions. Specifically, these ft.netions are described as.follows: Maintaining appropriate control of hazardous material, shipments, and receipts. Supervising _and coordinating the hazardous waste disposal program. Assisting in personnel and equipment decontamination. Distribution and processing of personnel monitoring equipment. Maintaining individual exposure records. Orienting and training CNFP personnel in radiological and nuclear safety. PAGE: 11-1 DATE: 6-22-90 REV.: 0 SUPERSEDES: PAGE: DATE: REV.: i

o $l \\ 4 M B&W FUEL COMPANY, COMMERCIAL NUCLEAR FUEL PLANT C T -USNRC LICENSE SNM-1168, DOCKET 70-1201 ORGANIZATION AND PERSONNEL CRAPTER 11.0-PART II 11.2 Key Personnel Functions 11.2.5 Health-Safety Monitor The Health-Safety Monitor is responsible for conducting routine monitoring, sample collection and analytical tests to determine radiation and contamination levels. L 11.3 Resumes Since it is also the responsibility.of the entire plant management to assure safe operations and regulatory compliance, we are including resumes from other managerial I organizations within the CNFP as well as those within the Quality and Safety Group. These are as follows: Name Title l l: l R. A. Alto Plant Manager, Commercial Nuclear Fuel Plant D. V. Ferree Manager, Fuel Operations B. W..Pugh Manager, Production & Materials Control l E. J. Coppola Manager, Quality & Safety K. S. Lester Manager, Health Physics & Licensing G. B. Lindsey Health-Safety Foreman L C. W. Speight Manager, Facilities and Services R. W. Penoza Manager, Field Operations 'F. M. Alcorn Manager, Nuclear Criticality Safety Engineering-J. M. Harwell Nuclear Criticality Specialist Engineer PAGE: 11-3 DATE: 6-22.90 REV.: 0 L SUPERSEDES: 'PAGE: DATE: REV.: lL

B&W FUEL COMPANY, COMMERCIAL. NUCLEAR FUEL PLANT USNRC LICENSE SNM-1168, DOCKET 70-1201 DSb CRAPTER 11.0 ORGANIZATION AND PERSONNEL PART II NAME: Ernest J. Coppola TITLE: Manager, Quality and Safety CITIZEN OF UNITED STATES I EDUCATION: 1964 U. S. Naval Academy - B.S. Naval Science 1969 MIT - Naval Engineer 1969 MIT - MS Nuclear Engineering EXPERIENCE: US NAVY 1964-1966 Two years sea duty on' Navy Destroyer Escorts. Variety of shipboard assignments, including direct control of the ship during officer of the Deck watches. 1969-1973 Four years as a Navy Ship Superintendent at a U.S. Navy nuclear shipyard, responsible for waterfront management of construction, overhaul, repair, and refueling of nuclear and non-nuclear submarines and surface ships. Babcock & Wilcox 1973 Senior Engineer, Reactor Operations. Shift Test Coordinator for Oconee 1 startup physics testing. 1973-1974 Supervisory Engineer. Directed the j planning and execution of startup physics testing on five L&W reactors, 0-2, 0-3, TMI-1, ANO, and Rancho SECO. 1974-1976 Manager, Plant Performance Services. Section Manager, responsible for all physics-related services, including startup physics testing, instructing customer personnel in nuclear physics and and reactor (core) operation, test and operating procedures, and field assignments for core physics personnel. PAGE: 11-8 DATE: 6-22-90 REV.: 0 SUPERSEDES: PAGE: DATE: REV.:

-B&W FUEL COMPANY,. COMMERCIAL NUCLEAR FUEL. PLANT {)RbE1 USNRC LICENSE SNM-1168, DOCKET 70-1201 CHAPTER 11.0-PART II ORGANIZATION'AND PERSONNEL i NAME: Ernest J. Coppola EXPERIENCE: 1976-1983 Fuel Project Manager. Responsible for the technical.and commercial performance of assigned fuel contracts. Represent B&W to'the customer and the customer to B&W. Assignments included Oconee 1,2,&. 3; Rancho Seco,: Davis Besse, WPPS 1 & 4, and Bellefonte-1 & 2. 1983-1989 Senior (Fuel) Project Manager - TMI 1 restart, GE Fuel Powder / Pellet Project, McGuire/ Catawba: Fuel Project. 1989-Manager, Quality-and Safety. Manage Quality Assurance and Health Safety Departments. Responsible for functions of inspection, data _ evaluation, radiation-and industrial safety. Serves as the chairman of the Safety Review Board. PAGE: 11-9 DATE: 6-22-90 REV.: 0 SUPERSEDES: PAGE: DATE: REV.:

s QRN1 B&W FUEL COMPANY, COMMERCIAL NUCLEAR FUEL PLANT USNRC-LICENSE SNM-ll68, DOCKET 70-1201-CHAPTER 11.0-ORGANIZATION AND PERSONNEL ,PART.II NAME: J. Wayne Harwell TITLE: Principal Engineer, Nuclear Criticality Safety

CITIZEN OF-UNITED STATES EDUCATION:

B.S. Nuclear Engineering, 1963 Mississippi State University M.S. Nuclear Engineering, 1968 -Mississippi State University EXPERIENCE: 1963-1964 Ingalls Shipbuilding Corporation - Engineer, Shielding Structure Unit Performed nuclear shield design modifications and project management related duties for shielded structures on nuclear submarines during construction. 1964-1968 Mississippi State University - Instructor & Graduate Assistant Engineering Graphics Department. Graduate assistant.and instructor teaching freshmen engineering drawing classes. Attended graduate school in nuclear engineering. 1968-1976 Babcock & Wilcox, Nuclear Power Division, Senior Engineer, Nuclear Development Work related to self-powered neutron detectors signal-to-power conversion, core physics analytical modeling and benchmarking, core model analyses, core and fuel. assembly design optimization and reactor vessel fluence analysis. 1976-1976 Southern Company Services, Senior Core Analysis Engineer. Developed core physics models for the Farley PWR cores including generation of cross section tablesets and geometries for PDQ07 using EPRI ARMP code package. PAGE: 11-22 DATE: 6-22-90 REV.: 0 SUPERSEDES: PAGE:' DATE: REV.:

B&W~ FUEL COMPANY;: COMMERCIAL NUCLEAR FUEL. PLANT {)$N.- USNRC LICENSE SNM-1~168,-DOCKET 70-1201 ' CHAPTER.11.0 ORGANIZATION AND PERSONNEL ~ PART II '. 1 NAMES. J. Wayne Harwell' EXPERIENCE: '1976-1988 Babcock & Wilcox, Fuel Management Analysis. Responsible for fuel cycle design and fuel management analyses for Connecticut Yankee and B&W design 177 fuel assembly reactor cores using the-PDQO7 computer code. Work includes cross section tableset generation and fitting strategy development, advanced fuel and reactivity control concept n development, new fuel management' concepts and use of transport codes for-analytical model development. 1988-Babcock & Wilcox Company - Principal Engineer Nuclear Criticality Safety _ Engineering. Performs nuclear criticality safety evaluations using the SCALE computer code package that utilizes the Monte-Carlo computer codes (KENO-4-and KENO-5) and transport computer code (XSDRN). Responsible for methods development-along with computer codes benchmarking, verification, and. validations for the codes used in nuclear criticality calculations. 9 4 i' l PAGE: 11-23 DATE: 6-22-90 REV.: 0 SUPERSEDES: PAGE: DATE: REV.:

M -B&W FUEL COMPANY, COMMERCIAL NUCLEAR FUEL PLANT USNRC LICENSE SNM-1168, DOCKET 70-1201-PART II CHAPTER 12.0-RADIATION PROTECTION 12.1 General -Operations involving potential exposure to, radioactive t materials will be performed in a manner that assures the radiation safety of employees and the general public. This policy is implemented by maintaining a staff of qualified M personnel'and appropriate equipment, procedures, and records. Operations will be conducted in accordance with 9 applicable Federal, State, and Local requirements. Exposures to radioactive materials, or other hazards, will be maintained as low as reasonably achievable (ALARA). Health-Safety has the authority to stop hazardous or potentially hazardous operations until correction or resolution by_ plant management is obtained. Program effectiveness will not be reduced as.a' result of changes instituted by plant management. Section'12.'15 is a discussion of the measures used in conjunction with the field services operations (by-product material contamination).- Typical sources of information and . guidance for the CNFP radiation protection program:are: - USNRC Regulatory Guide series (as appropriate) - International Commission on Radiation Protection publications (as appropriate, --Handbook of Industrial Loss Prevention,_ Factory Mutual Engineering Corporation - Radiological Health Handbook U.S. Department of Health, Education, and Welfaro, Public Health Service. 12.2 Posting and Labeling L 12.2.1 ' General y l We request a continued exemption from the labeling L and posting requirements of 10 CFR 20.203 (e) (1) L and 20.203 (f) (1) because of the-nature of.our operation. To meet the intent'of these 9t j regulations, we post designated entrances with R signs that incorporate the radiation symbol (10 2%f CFR 20.203 (a) (1) ) and the following warning: CAUTION: RADIOACTIVE-MATERIALS l ANY AREA OR CONTAINER WITHIN THIS PLANT MAY CONTAIN RADIOACTIVE MATERIALS PAGE: 12-1 DATE: 6-22-90 REV.: 0 i. L SUPERSEDES: PAGE: DATE: REV.: L

Jh. I + Attachment III OTHER CHANGES INCORPORATED INTO THE LICENSE RENEWAL APPLICATION 1) The exemption for Ken Shy, section 2.2.3, page-2-3 is no l longer applicablo, therefore, it has been removed. 2) The section regarding personnel monitoring systems, section 3.2.1,'page 3-2, was addressed under the technical requirementa for alpha contamination. It has been relocated to a more appropriate location, section 3.1.6, page 3-2. Also sections (6.1.4 & 3.1.6) that stated that all personnel-l would wear TLDs was changed to designated employees. 3) The minimum acceptable criteria for the ventilation ayatem'a efficiency waa' inadvertently left out. It has been added to section 3.2.2.1, page 3-5. 4) The acceptable. contamination on the floor in Radiation i. Control Zone during operations was increased to 20,000 l dpm/100 cm2 provided it la decontaminated upon completion of i the work.- Section-3.3, page 3-15 and Table 1, page 3-19 i l reflects this change. 5) Section 4.2, page 4-12 was changed to read " Shelves with ~ multiple enrichmenta are limited to a maximum alab thickness of'4 inches."lin place of "Each shelf la limited to a single alab thickness." i E Section 4.2.4.2, page 4-13 was changed to read "... does not l exceed 4.0" slab" instead of "... la maintained as a 4.0" slab. 6)- Table 7.2 was updated to include the SERF building as part H L of the decommissioning plan. l r 7) Section 10.1, page_10-1 was updated to include field. operations as part of the CNFP operations. 8) Chapter 11 was updated to include the CNFP current j management and organization. I l

ATTACHl1EllT IV .MPC-HRS. DPM1 Rance 1988 1989 1 0-5 87.6 84.2 5-10

8. 0

7. 9 11 - 15

4. 4

2. 6 l

16 - 20

0. 0

2. 6 21 - 25

0. 0

1. 8 26 - 30

0. 0

0. 9 I

i AVERAGE MILLIREM EXPOSURE PER QUARTER i 1988 1989 F O - 50 98.7 96.2 51 100

1. 3

3. 3 101 150

0. 0

0. 5

] 151 - 200

0. 0

0. 0 201 - 250

0. 0

0. 0 1

251 - 275

0. 0

0. 0 276 - 300

0. 0

0. 0 301 - 350

0. 0

0. 0 351 - 375

0. 0

0. 0 376 - 400

0. 0

0. 0 401 - 450

0. 0

0. 0 451 - 500

0. 0

0. 0 501 - 550

0. 0

0. 0 551 - 575

0. 0

0. 0 576 - 600

0. 0

0. 0 Maximum exposure for 1988 - 358 millirem (Joe Chambers)

Maximum exposure for 1989 532 millirem (Joe Chambers) ......}}