ML20154G405
| ML20154G405 | |
| Person / Time | |
|---|---|
| Site: | 07001201 |
| Issue date: | 10/05/1998 |
| From: | Matheson J FRAMATOME COGEMA FUELS (FORMERLY B&W FUEL CO.) |
| To: | Lamastra M NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM), NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS) |
| References | |
| NUDOCS 9810130163 | |
| Download: ML20154G405 (100) | |
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1 October 5,1998 Director, Office of Nuclear Material Safety and Safeguards i
U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555 l
Reference:
Docket 70-1201, SNM 1163 j
Subject:
Organizational Changes at the Lynchburg Manufacturing Facility of Framatome Cogema Fuels
Dear Mr. Lamastra,
Framatome Cogema Fuels formally requests an amendment of SNM-1168, Docket 70-1201, to support organizational changes. These chang:s are driven by two different forces.
First, while on maternity leave, Ms. O.F. Elliott resigned from her position as Manager, Safety & Licensing. As previously submitted to your office, T. V. Allsep continues to serve as the Acting Manager, Safety & Licensing. As a result of this resignation, FCF has l
opted to reorganize the Quality, Safety, Health & Licensing Department. The Manager, Safety & Licensing is presently responsible for M!terial Control & Accountability, Radiation Protection, Industrial Safety, Training, Corporate Safety, and Licensing. The proposed organizatio's will distribute these responsibilities across four new positions. Hey are: Manager, Licensing and Quality Systems, to be held by C. Armontrout; Manager, Radiation Protection, to be held by T. Allsep; Manager, Uranium Utilization and Accountability, to be held by P. Burneson: and Program Manager, Compliance and SkillsTraining, to be held by J. Whitt. Resumes reflecting these appointments are attached.
Second, FCF has opted to change from one Production Manager to three Product Center Managers representing key product lines.
We believe this will further develop the ability of Production and Quality to support the goals of FCF, In summary, FCF requests an amendment to our license for operation based on a need for organizational changes. It is our opinion these changes will not reduce operational safety or regulatory compliance at the facility. It is the desire of FCF to implement these changes by October 15*,1998. Six copies of the affected license sections are attached. Should you have questions regarding this matter, you may contact me at 804-832-5202. Hank you for your time and effort.
Sincerely, ramatome Cyema Fuels J.. Matheson ice President, Operations l
cc Mr. Michael A lamasus l
Unite 8tmes Nuclear Reguimory Comminnen omse of Nuclear MmerWs Safety & safeguards Dmsian of Fuel Cycle Safety & Safeguards Ucensmg Branch.Ucensms eenon 2
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s Mail stop TEDl2 Two Wbte Fina North
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l l1355 Rockville Pike Rodville, MD 20852-2733 7
9810130163 981005 PDR ADOCK 07001201 PM S
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Framatome Cogema Fuels l
F R AM ATO M E P.O. Box 11646, Lynchburg, VA 24506-1646 TEcHHOLOGlE5 Telephone: 804 832 5000 Fax: 804 832 5167 I
4 Att: chm:nt # 1 Paga 1,of 8 a
Chao./Paae Para.
Chanae 2
2-1 2.1.1 Revised LMF management structure to show Manager, Licensing & Quality Systems, i
Manager, Radiation Protection, Program Manager, Compliance
& Skills Training and Manager, Uranium Utilization Accountability to report directly to Manager of Quality.
Changed Production managers to Product Centers Changed from " includes Manufacturing Engineering, Fuel Manufacturing, etc. to " includes Process Engineering and Fuel j
Manufacturing functions",
j 2
2-2 2.1.2 Reworded last sentence to read "The Managers of Assembly, Rods, & Components are Production Managers".
2 2-2 2.1.4 Deleted " maintain training programs".
2 2-3 2.1.4 Added " radiological" to last paragraph.
)
i Changed " Manager Safety &
Licensing" to " Manager, Radiation Protection".
l 2
2-3 2.2.3 Changed " Manager, Safety &
Licensing" to Manager, Licensing & Quality Systems".
t Reworded first sentence from
" understanding of nuclear and l
l l
[
_]
h Attachmtnt # 1 Paga 2 of 8 radiation safety" to
" understanding of industrial safety & licensing".
Reworded last sentence from
" maintain an effective nuclear criticality and radiation safety program for the activities authorized by the license" to effective training industrial safety and licensing programs".
2 2-3 2.2.4 Changed " Health Physicist" to 2-4
" Manager, Radiation -
Protection".
Deleted first sentence; changed "this person" to " Manager, Radiation Protection"; changed
" Science or Engineering" to
" science or engineering".
Changed "A" to "and"; changed "2 years" to "5 years".
Added last sentence stating "Such experience shall be..."
2 2-4 2.2.7 Changed " Manager, Safety &
Licensing" to " Manager, Radiation Protection".
2 2-5 2.2.8 Added new paragraph.
2 2-5 2.3 Eliminated " Quality Manager" 2-6 from third paragraph.
Changed " Manager, Safety &
Licensing" to Manager, l
Radiation Protection".
l Changed " Manager, Safety &
Licensing" to " Manager, Radiation Protection".
j
Att:chmtnt # 1 -
Pzga 3 of 8 i
2 2-6 2.3.1 Changed " Health Safety" to
" Health Safety & Licensing".
2 2-7 2.5 Changed " Health Safety" to "the Program Manager, Compliance & Skills Training"in the first paragraph.
2-7 Changed " Health Safety" to "
the Health Safety and the Licensing departments" in second paragraph.
Changed " Health Safety" to "the Program Manager, Compliance & Skills Training"in fourth paragraph (3 places).
Changed " Manager, Safety &
Licensing to " Manager, Radiation Protection".
2 2-8 2.6 Changed " Manger, Safety &
Licensing" to " Manager, Radiation Protection" (5 places).
Changed " Manager, Safety &
licensing" to " Manager, Uranium Utilization &
Accountability" (2"' bullet) 2 2-9 2.7 Changed " Manager, Safety &
Licensing" to " Manager, Radiation Protection".
I Changed " semi-annually" to l
" semi-annual" in bullet noting l
Independent Audits.
1 Deleted "LMF" from last l
paragraph.
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. _ _. _ _ _ _. _ _~
Attachmtnt #-1 t
b" Pega 4 of 8 l
3 3-1 3.1.3 Changed " Health Physicist" to
" Health Safety.
3 3-7.
3.2.3 Changed " Safety & Licensing organization" to " Health Safety".
i 4
4-1 4.1.2.
Changed from " Manager, l
Safety Licensing" to l
" Manager, Radiation Protection".
L 6-6-3 6.2 Changed " health Safety Section" to " Licensing Section".
l Changed " Manager, Safety &
Licensing" to " Manager, Licensing & Quality Systems".
6 6-3 6.2.3 Changed from " Health &
Safety" to " Manager, Licensing
& Quality Systems".
11 11-1 11.1 Changed " Safety & Licensing" organization to " Quality" organization.
Eliminated third sentence.
Changed " Manager, Safety &
Licensing" to " Manager, l
Radiation Protection (Health Safety Section)".
11 11-1 11.2.1 Changed " Manager, Safety &
Licensing" to " Manager, Radiation Protection".
l Changed " Managers of l
Manufacturing" to " Product l
Center Managers".
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.Atttchm::nt # 1 Paga 5 of 8 Deleted Manufacturing Engineering and Fuel from last sentence.
11
_11.1 11.2.2 Added " radiological" to first 11.2 sentence.
Deleted sixth bullet item.
Reworded paragraph to reflect current organization changes.
" Health Physicist and Regulatory Compliance Officer" replaced with " Manager, Radiation Protection".
11.
11-2 11.2.3 Changed " Health Physicist" to
" Manager, Radiation Protection". (3 places)
Deleted " entire" from first sentence.
Eliminated " industrial &
i chem cal" from first sentence.
Changed " Manager,.chety &
Licensing" f.o "Gusisty Manager".
11 11-2 11.2.4 Renumbered paragraph.
Paragraph 11.2.5 changed to paragraph 11.2.4 Changed " Health Physicist" to
" Manager, Radiation Protection".
Deleted "directly" from last sentence.
11 2 11.2.5 Added new paragraph.
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11-3 l
L N
t..
Atttchm:nt # 1 Paga 6 of 8 11 11-3
-11.2.6 Renumbered paragraph.
Paragraph 11. 4 is now paragraph 11.2.6 Changed " Manager, Safety &
Licensing" to " Manager, Licensing & Quality Systems".
11 11-3 11.2.7 Renumbered paragraph.
. Paragraph 11.2.6 is now paragraph 11.2.7.
Changed " Safety & Licensing" organization to " Radiation Protection" organization.
Changed from "2 Health Safety Technicians" to "3 Health
. Safety Technicians".
'11 '
11-3 11.2.8 Added new paragraph.
11 11-3 11.3 Deleted G. F. Elliott, D. L.
Gordon, G. B. Lindsey. Added T. V. Allsep, C. A. Armontrout, D. M. Minor, C. A. Conrad, and J. S. Whitt.
Changed title of T. S. Wilkerson from Manager, Fuel Manufacturing (Production Manager) to Manager Assembly.
11 11-9 11.3 Experience updated for T. S.
Wilkerson. Changed title to
" Manager, Assembly".
11-11 11.3 Resume added to reflect 11-12 organization change of 11-13 T. V. Allsep.
11-14 11.3 Resume added to reflect 11-15 organization change of
Attachm:nt # 1 Paga 7 of 8,
C. A. Armontrout.
11 11-16 11.3 Resume added to reflect 11-17 organization change of 11-18 D. M. Minor.
11-19 11 11-20 11.3 Resume added to reflect 11-21 organization change of 11-22' C. A. Conrad.
11 11-25 11.3 Resume added to reflect 11-26 organization change of J. S.
11-27 Whitt.
11 11-25 11.4 Changed " Manager, Safety &
Licensing" to " Manager, Radiation Protection". (2 places) 11 11-25 11.5 Changed " Health Safety" to
" Program Manager, Compliance
& Skills Training".
11 11-25 11.5.1 Changed " Health Safety" to
" Program Manager, Compliance
& Skills Manager".
11 11-26 11.5.1 Changed " Health Safety" to
" training".
11 11-26 11.5.2 Added " Program Manager, Compliance & Skills Training" to first sentence.
Changed " Health Safety" to L
" Health Safety & Program Manager, Compliance & Skills Training".
Changed " health safety record" to " training record".
t
. Attzchm:nt # 1
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Paga 8 of 8 11 11 11.5.3.1 Changed " Health Safety" to
" Health Safety or Program Manager, Compliance & Skills Training".
11 11-27 11.5.3.2 Added " Program Manager, Compliance & Skills Training" to second sentence.
11 11-26 11.6.1
- Changed " Manager Safety &
Licensing" to " Manager, Radiation Protection".
11 11-29 11.6.2.4 Added " Licensing Personnel".
14 14-1 14.1 Changed from " Manager, Safety & Licensing" to i
" Manager, Radiation Protection".
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FRAMA TOME COGEMA FUELS - L 5'NCHBURG MANUFACTURING FA CILITY USNRC LICENSE SNM-1168, DOCKET 70-1201 PARTI - CHAPTER 2 - ORGANIZATION AND ADMINISTRATIVE l
r 2.1 Oraanizational Responsibilities and Autfiority 2.1.1 Manaaement it is the responsibility of the Plant Manager and Quality 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 of program effectiveness j
prompt correction of nonconforming conditions The LMF management structure is as shown in Figure 2.1.
l FCF l
PRESIDENT l
l i
l QUALITY NUCLEAR LMF PLANT MANAGER CRITICAllTY MANAGER SPECIALIST j
MANAGER MANAGER
- PRODUCT
^ '^
l CENTERS
^
(PRODUCTION l
MANAGERS)
(LICENSING) l l
I i
MANAGER PROGRAM URANIUM
- MANAGER, 1
UTILIZATION &
COMPLIANCE &
OPERATIONAL ACCOUNTABILITY SKILLS SUPERVISORS TRAINING (e.g., includes Process Engineering & Fuel Manufacturing Functions)
Page: 2-1 September 9,1998 Revision: 10
FRAMA TOME COGEMA FUELS - LYNCHBURG MANUFACTURING FACillTY
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USNRC LICENSE SNM-1168, DOCKET 70-1201 l
PARTI - CHAPTER 2 - ORGANIZATION AND ADMINISTRATIVE
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2.1.2 Production Manaaers Production managers report directly to the plant manager with the exception of the Manager, of SERF-3 & 4. SERF-3 & 4's I
organization is discussed in section 2.2.7. Production managers are responsible for managing operational area supervisors and are responsible for production functions. The managers of Assembly, Rods and Components are Production Managers.
i 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 is 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 qualification of operational area supervision shallinclude:
1 (a)-
A high school education and 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 LMF operations.'
2.1.4 The Health-Safety Section The Health-Safety Section chall be responsible to interpret the license conditions, provide monitoring facilities, develop safe operation guidelines anu review and approve operating procedures to assure safe operation and license compliance. These responsibilities include maintenance of nuclear safety 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 September 9,1998 Revision: 10
L FRAMA TOME COGEMA FUELS - L YNCH2URG MANUFACTURING FACILITY USNRC LICENSE SNM-1168, DOCKET 70-1201 PARTI - CHAPTER 2 - ORGANIZATION AND ADMINISTRATIVE l
The Manager, Radiation Protection or their qualified designee shall be responsible to provide management with assurance of the effectiveness of the radiological safety program by maintaining an audit program that includes periodic inspection of controls and i
operations, reports to management, follow-up or nonconforming conditions and necessary documentation (see Audits, Section 2.7).
l 2.2 Personnel Education and Experience Reauirements 2.2.1 Plant Manaaer The Plant Manager shall have a Bachelor's Degree in Science or l
Engineering, a minimum of 10 years experience in the nuclear industry, and 5 years experience in management.
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2.2.2 Quality Manaaer The Quality Manager shall have a Bachelor's of Science, a minimum of 10 years experience in the nuclear industry, and 5 years experience in management.
l 2.2.3 Manaaer. Licensina & Quality Systems The Licensing & Quality Systems Manager shall have e Bachelor's Degree and a minimum of five years experience, which would develop an understanding of industrial safety and licensing. Such experience shall be of a nature which demonstrates to the Quality Manager sufficient judgment and capability to establish and maintain effective training, industrial safety and licensing programs.
2.2.4 Manaaer. Radiation Protection The Manager of Radiation Protection shall have a Bachelor's Degree in science or engineering and a minimum of 5 years experience in applied health physics is required along with sufficient formal training that provides an understanding of the health physics and nuclear safety hazards involved at the LMF. Such experience shall l
be of a nature which demonstrates to the Quality Manager sufficient
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judgement and capability to establish and maintain an effective i
Page: 2-3 September 9,1998 Revision: 10 i
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FRAMA TOME COGEMA FUELS - L YNCHBURG MANUFACTURING FACILITY USNRC LICENSE SNM-1168, DOCKET 70-1201 PARTI - CHAPTER 2 - ORGANIZA, TION AND ADMINISTRATIVE nuclear criticality and radiation safety program for the activities authorized by the license.
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 experience as a radiation monitor. They may fulfill the experience requirements on the job as a Health-Safety Monitor trainee.
2.2.6 Nuclear Criticality Safety Specialist The Nuclear Criticality Safety Specialist is organizationally independent of the LMF, with no interest in plant operations, other i
than the nuclear criticality safety aspects. The Nuclear Criticality Safety Specialist is responsible 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 Engineerlag 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 or a Master's Degree in Nuclear Engineering and one year experience performing nuclear criticality safety analyses.
2.2.7 SERF-3 & 4 Oraanization l
Regulatory compliance to include all facets of safety of the SERF-3 l
& 4 facilities will be the responsibility of the Quality Manager in l
which he may delegate the responsibility to the Manager, Radiation Protection. The safety organization described in Section 2.2.3 and 2.2.4 shall be responsible for implementation of the safety program.
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FRAMA TOME COGEMA FUELS - L YNCHBURG MANUFACTURING FA CILITY USNRC LICENSE SNMs1168, DOCKET 70-1201 PARTI - CHAPTER 2 - ORGANIZA TION AND ADMINISTRA TIVE 4
?
l The SERF-3 & 4 Manager will be responsible for all service and i
support activities and will report to the management of Framatome Technologies, Inc.
l 2.2.8 Procram Manaaer. Comoliance & Skills Trainino The Program Manager, Compliance and Skills Training, shall be responsible for the development and implementation of a training l
program to ensure personnel engaged in work activities at the facility l
are fully training and cognizant of their responsibilities in respect to i
radiological safety, criticality safety, industrial safety and regulatory compliance issues.
i 2.3 Safety Review Board l
The Safety Review Board reviews the following as a minimum on a quarterly basis:
l e
New or revised facilities e
Analysis of equipment and processes involving hazardous materials e
Maintenance of fire safety The continuing effectiveness of established controls and safeguards e
Maintenance of ALARA criteria (review of quarterly air sample, review e
of surface contamination survey anomalies etc.)
e Safety-related audit and inspection findings Other items (such as abnormal occurrences) that Safety Review Board e
members wish to discuss.
The Safety Review Board Chairman shall have a Bachelor's Degree in Science l
or Engineering and a minimum of five years experience in responsible positions which would develop an understanding of nuclear and radiation safety.
The Safety Review Board Chairman shall be directly responsible to the l
Quality Manager for the proper conduct of the Safety Review Board. The Plant Manager shall be kept informed in writing of Safety Review Board I
action. The permanent membership of the Board shall consist of representatives from production management (section 2.1.2), Manager, Radiation Protection and others as deemed necessary by the Chairman.
Technical representatives of outside consulting organizations shall be included as necessary.
1 Page: 2-5 September 9,1998 Revision: 10
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$rAMAT6ME COGEMA F5ELS - LYNCHBURG MNNUFACiUR hACILITY l
USNRC LICENSE SNM-1168, DOCKET 70-1201 l-PARTI - CHAPTER 2 - ORGANIZATION AND ADMINISTRATIVE l
i-I Board meetings may be convened at the discretion of the Safety Review l
Board Chairman, but shall be held at least quarterly. The Safety Review L
Board Chairman shall decide whether or not the necessary disciplines are
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- 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.
Records of Safety Review Board proceedings, including supporting calculations and approvals, shall be retained for 2 years minimum after the l
completion or termination of the subject activity'.
An annual ALARA report shall be prepared under the direction of the Manager, Radiation Protection. The report shall be submitted to the Safety Review Board in which they will review 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 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.
At least every two years, the Safety Review Board shall evaluate the effectiveness of the radiation / nuclear safety training program.
2.3.1 Pre-ooerational Evaluations New operations and facilities and major operational changes require the Health Safety and the Licensing departments to perform an evaluation prior to initial operation to ensure that adequate radiation, nuclear, fire, and chemical protection is established.
The Safety Review Board Chairman reviews all pre-operational l
evaluations which involve hazardous materials and determines if l
Board review is necessary.
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In the case of minor changes where existing safety practice remains i
the same, the Safety Review Board Chairman may determine that Board review is not necessary. Safety Review Board members shall l
be kept appraised of actions tr' a by the Safety Review Board I
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FRAMA TOME'C5GEMA FOELS - L YNCHBURG MANUFACTURIN FACILlTY
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USNRC LICENSE SNM-1168, DOCKET 70-1201
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PARTI - CHAPTER 2 - ORGANIZATION AND ADMINISTRATIVE Chairman on such minor changes. 'Where other than minor changes are involved, the Safety Review Board review and approval process l
shall be conducted in accord with procedures approved by the Plant Manager.
l 2.4 Acoroval Authority for Personnel Selection Personnel selection for the FCF staff level positions shall be approved by the President.
2.5 Trainino initial indoctrination of employees to nuclear and radiological safety shall be the responsibility of the Program Manager, Compliance & Skills Training and shall conform with 10 CFR 19. Initial indoctrination training shall, as a minimum, include the following topics:
e license conditions e
federal regulations e
operating procedures e
radiation safety e
nuclear safety -
e emergency procedure e
chemical and fire safety The extent and depth of the training, relative to the detailed aspects of the radiation, chemical, fire and nuclear safety programs, is dependent on the employee's job assignment and potential exposure to radioactive materials as determined by the Health Safety and the Licensing departments.
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 continaing safety training program shall be conducted by the Program Manager, Compliance & Skills Training to the extent necessary to assure the f
Page: 2-7 September 9,1998 Revision: 10 i
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FRAMA TOME COGEMA FUELS - L YNCHBURG MANUFACTURING FACILITY USNRC LlCENSE SNM-1168, DOCKET 70-1201 PARTI - CHAPTER 2 - ORGANIZATION AND ADMINISTRATIVE m'aintenance of acceptable safety practices. Such training may be conducted on an individual or group basis. The content of retraining programs may be varied by the Program Manager, Compliance & Skills Training but willinclude j
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 the Program Manager, Compliance & Skills Training and retained for at least two years.
The Manager, Radiation Protection shall be responsible to assure that personnel assigned tc Health-Safety are properly trained. The extent and depth of the training is based on the specific job assignment involved.
Health-Safety monitoring personnel shall receive a combination of formal and wn-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.
2.6 Operatino Procedures Written procedures for the conduct of specific operations including maintenance end development of work within the plant are prepared by the functional component responsible for that activity and shall be reviewed and approved by production management and Manager, Radiation Protection as appropriate. Operating procedures which involve SNM shall be reviewed at i
least every two years by the appropriate production manager and Manager, Radiation Protection. Applicable procedures shall be available in the work i
area and adherence to procedures is required of all personnel. Procedures for operations where nuclear and radiological safety are involved shall include specific reference to applicable safety requirements. Procedure and format shall be such that operations are clearly detailed and specific directions are provided for operation under both normal and abnormal conditions.
Administrative procedures shall supplement operating procedures to ensure proper proceduralimplementation. Procedural control of activities at the LMF are categorized as follows:
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 Page: 2-8 Septembar 9,1998 Revision: 10
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FRAMA TOME COGEMA FUELS - L YNCH2URG MANUFACTURING FACILITY
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USNRC LICENSE SNM-1168, DOCKET 70-1201 1
PARTI - CHAPTER 2 - ORGANIZATION AND ADMINISTRATIVE l
i procedures may be for internal Health-Safety use or may be intended for general distribution to affected individuals within other components. As a minimum, Health-Safety procedures shall be approved in writing by the Manager, Radiation Protection as well as i
approved by affected members of plant management.
SNM Accountability l
l Nuclear Materials Control procedures provide techniques for the l'
accountability and measurement of SNM. As a minimum, such procedures shall be approved in writing by the Manager, Uranium l
Utilization and Accountability (SNM Accountability and Inventory).
Other Plant Groups l
l Procedures from other plant groups (i.e., Manufacturing, Quality, etc.)
l where nuclear, chemical, fire or radiological safety, license conditions, or regulatory requirements are involved require prior approval by the Manager, Radiation Protection as well as approval by affected j
members of plant management.
New operations and major operational changes shall require the written recommendation of the Safety Review Board Chairman prior to implementation. If the change requires revisions to procedures or the local safety rules, these modifications shall be in place prior to implementation.
Revised procedures shall be subject to approval in the same manner as new procedures. Health-Safety procedures shall be reviewed at least annually for
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technical correctness and applicability. The Manager, Radiation Protection shall use discretion to assure that the appropriate personnel of Section 2.2 performs the procedure review.
l Procedure distribution and control shall be in accord with procedures i
approved by plant management.
I 2.7 Audits and Insoections An internal Health-Safety inspection program shall be maintained to provide j
assurance that plant activities are conducted safely and in accord with i
I.
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FRAMA TOME COGEMA FUELS - L YNCHBURG MANUFACTURING FACILITY USNRC LICENSE SNM-1168, DOCKET 70-1201 c
PARTI - CHAPTER 2 - ORGANIZATION AND ADMINISTRATIVE l
license specifications. The Manager, Radiation Protection shall be responsible to assure that the inspection program is conducted effectively.
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The internal Health-Safety inspection program at the LMF is structured as i
follows:
Monthly Safety Insoections Health-Safety personnel shall conduct, at least monthly, a formal j
inspection of plant status relative to safety related functions to include fire safety, except during plant shutdown of a week or longer.
l Inspection results shall be documented, reported to plant management l
- and supervision as appropriate, and will be maintained on file by Health-Safety for at least 2 years.
The monthly safety inspections shall be conducted by personnel technically qualified to perform this function and in the application of l
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.
Indeoendent Audits Independent auditors shall conduct, as a minimum, semi-annual nuclear safety, fire safaty and health physics inspections at the LMF.
These audits shall be conducted in accordance with written instructions or procedures. The audit scope shall consist of physical inspections and records reviews for the industrial, nuclear,.and radiological safety elements of plant activities including:
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September 9,1998 Revision: 10 l
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. FRAMA TOME COGEMA FUELS - L YNCHBURG MANUFACTURING FACIUTY USNRC UCENSE SNM-11.68, DOCKET 70-1201 i
PARTI - CHAPTER 2 - ORGANIZATION AND ADMINISTRA TIVE effectiveness of procedural controle impacting on operational safety parameters, t
- audit of operating records, where such records provide a means of verifying procedural compliance with safety specifications.
review and evaluation of contamination survey data.
- ascertaining the overall performance of the plant functions in providing adequate controls, surveillance, and follow-up to assure safety and license compliance.
Independent auditor's reports shall be submitted to the Quality Manager for his review. He will ensure that the proper management reviews the report. The audit report shall include any audit findings or i
recommendations. Actions taken as a result of audit findings shall be documented.
Qualifications of the independent auditors shall include competence in the areas of health physics or nuclear physics as appropriate at a level at least equivalent to Paragraph 2.2.3 or 2.2.5 respectively.
Designation of the independent auditors shall be the responsibility of the Quality Manager.
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2.8 Investiaations and Reportina Unusual events requiring reporting under NRC regulations shall be investigated as appropriate, with results reported to plant management and NRC. Events not otherwise requiring a report may be reported to NRC based on potential public or media involvement, etc., in order to keep NRC appraised of the situation.
4 2.9 Records Plant alterations or additions, abnormal occurrences, events associated with radioactive releases, criticality analyses, audits, inspections, instrument calibration, Al_ ARA findings, employee training and retaining, personnel exposures, routine radiation surveys, and environmental surveys shall be maintained on file for a minimum of 2 years or as otherwise required by
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federal regulation or other license condition, for review by LMF management and regulatory agencies.
Page: 2-11 September 9,1998 Revision: 10
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' FRAMATOME COGEMA FUELS - L YNCHBURG MANUFACTURING FACILITY USNRC LICENSE SNM-1168, DOCKET 70-1201 PARTI - CHAPTER 3 - RADIATION PROTECTION 3.1 Administrative Reauirements 3.1.1 General The business of the LMF includes radiation hazards from alpha contamination (fuel fabrication) and from beta-gamma contamination (field service work for commercial nuclear utilities).
While the levels of these respective hazards are different, the controls for each are similar.
3.1.2 ALARA lt is the policy of the LMF to maintain occupational exposures to radiation and radioactive contamination in effluents as low as reasonably achievable. The responsibility for implementat;on of the ALARA policy is designated to the Health-Safety section.
3.1.3 Radioloaical Work Permit (RWP)
Radiological Work Permits (RWP) shall be used to define the protective clothing and equipment required to perform work involving surface contamination. RWPs shall be used to control all work in a RCA and work involving surface contamination above clean area limits outside of a permanent RCA that is not addressed by an approved operating procedure. RWPs are reviewed for industrial safety and approved by Health Safety.
3.1.4 Written Procedures All licensed activities related to radiation protection shall be conducted in accordance with approved written procedures.
Approval, scope, and distribution requirements are described in Section 2.6.
3.1.5 Postinas Local safety rules approved by Health-Safety providing personnel and supervision with specific directions essential to ensuring radiation safety shall be posted in areas where appropriate. Other l
radiation safety postings or warnings as required by 10CFR20 shall be placed in areas as required.
Page: 3-1 September 9,1998 Revision: 6 I
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FRAMA TOME COGEM4 FUELS - L YNCH~URG MANUFACTORING FACILITY USNRC LICENSE SNM-1168, DOCKET 70-1201 PARTI - CHAPTER 3 - RADIATION PROTECTION 3.1 Administrative Reauirements 3.1.1 General The business of the LMF includes radiation hazards from alpha contamination (fuel fabrication) and from beta-gamma contamination (field service work for commercial nuclear utilities).
While the levels of these respective hazards are different, the controls for each are similar.
3.1.2 ALARA It is the policy of the LMF to maintain occupational exposures to radiation bud radioactive contamination in effluents as low as reasonably achievable. The responsibility for implementation of the ALARA policy is designated to the Health-Safety section.
3.1.3 Radioloaical Work Permit (RWP)
Radiological Work Permits (RWP) shall be used to define the protective clothing and equipment required to perform work involving surface contamination. RWPs shall be used to control all work in a RCA and work involving surface contamination above clean area limits outside of a permanent RCA that is not addressed by an approved operating procedure. RWPs are reviewed for industrial safety and approved by Health Safety.
3.1.4 Written Procedures All licensed activities related to radiation protection shall be conducted in accordance with approved written procedures.
Approval, scope, and distribution requirements are described in Section 2.6.
3.1.5 Postinas Local safety rules approved by Health-Safety providing personnel and supervision with specific directions essential to ensuring radiation safety shall be posted in areas where appropriate. Other radiation safety postings or warnings as required by 10CFR20 shall be placed in areas as required.
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USNRC LICENSE SNM-1168, DOCKET 70-1201 PARTI - CHAPTER 3 - RADIATION PROTECTION l
j 3.1.6 Radiation Control Radiation areas shall be posted and controlled according to 10CFR20 requirements. Personnel radiation exposures shall be monitored by Health-Safety using appropriate devices based on the type of radiation and sensitivity requirements, including thermoluminescient dosimeters (TLD) or self-reading pocket ion chambers (SRD). Dosimetry issue to plant personnel and visitors l
will be determined based on the monitoring requirements of 10CFR20. Accidental neutron radiation exposures will be monitored by indium foil issued to personnel as an integral part of the standard identification badge. Other dosimetry will be issued as necessary for unusual circumstances such as source manipulation or work involving highly transient exposure rates. Extremity exposure monitoring is accomplished using TLD badges as required by 10CFR20. TLDs are processed by a vendor at monthly or quarterly intervals. Immediate processing is available for rapid evaluation of exposures. Personnel monitoring reports shall be prepared as required by applicable regulations.
3.2 Technical Reauirements 3.2.1 Controlled Areas - Personnel Contamination Control Radiologically Controlled Areas (RCA) shall be established by Health-Safety as an area used to control work involving surface contamination above uncontrolled area limits. RCAs may contain a
" clean area" which is a potentially contaminated area and one or more Contaminated Areas (CA), which are areas known to be contaminated. Personnel must frisk prior to leaving a RCA or the immediate area adjacent to the RCA boundary. Upon leaving a CA, personnel must frisk prior to working in another area of the RCA.
Personnel working in a RCA shall be properly trained prior to work or be escorted by a qualified workerc Contaminated Areas shall be designed to include a step-off pad or an intermediate change room. The purpose of the step-off pad is to establish a designated area where personnel enter and exit the Contaminated Area. Prior to stepping on the Step-off pad, all l
protective equipment and anti-contamination clothing shall be removed. In RCAs which do not contain a clean area (i.e., the l
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FRAMATOME COGEMA FUELS - LYNCHBURG MANUFACTURING FACIUTY USNRC UCENSE SNM-1168, DOCKET 70-1201 PARTI - CHAPTER 3 - RADIATION PROTECTION entire RCA is a Contaminated Area), an intermediate change room may be provided in lieu of the step-off pad. Guidelines for entry
{
into intermediate change rooms shall be established for each area, Upon completion of radiological work involving surface contamination, personnel will perform a complete wholebody frisk.
Frisking instruments will be provided and maintained by Health-Safety and will be selected based on the type of radiation or contamination being monitored within the particular area or the scope of work being performed. Friskers shall include a visual and audible alarm. The alarm set point will be established as low as
)
possible taking into account the need to minimize the number of j
false alarms.
Personnel Decontamination Policy i
Health-Safety shall be notified if contamination above the frisker alarm set point is detected on personnel skin or clothing as they exit from a RCA and when initial decontamination efforts fail to reduce the contamination below the alarm set point. Health-Safety shall assist with further decontamination efforts as necessary to reduce the exposure to a level as low as reasonably achievable, consistent j
with good health physics practice, before releasing the employee.
3.2.2 Ventilation Svstem 3.2.2.1 Airborne Effluents to Uncontrolled Areas Airborne effluents to uncontrolled areas shall be l
controlled to the limits specified in 10CFR20 by means of heat resistant absolute type filters with rated collection efficiencies of 99.95% for 0.3 micron DOP particulate. Effluents shall pass through l
single stage HEPA filtration before release. Also, recirculated air shall be HEPA filtered prior to reentering controlled areas. Filtration efficiency shall be evaluated in accord with Regulatory Guide 3.2 (Efficiency Testina of Air-Cleanina Svstems Containina Devices for Removal of Particles,1/8/73) upon installation, and following major maintenance.
The minimum acceptable system efficiency shall be l
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FRAMXTOIWC COGEMA FUELS - L YNCHBURG MANUFACTURING FACILITY USNRC LICENSE SNM-1168, DOCKET 70-1201 PARTI - CHAPTER 3 - RADIATIO" PROTECTION N
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 effectiveness. Air will not be recirculated if levels are above 25% DAC of 10CFR20 Appendix B, Table 1.
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 differential pressure exceeds 4 inches of water.
Gaseous effluents shall be representatively sampled 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 Eftluent Limits Compliance with the following release limits shall be maintained in order to ensure that airborne releases to uncontrolled areas are maintained as low as reasonably achievable (ALARA).
If the gross alpha radioactivity (excluding radon and its daughters) in planned gaseous effluent discharges from fuel manufacturing operations exceeds 10 uCi per calendar quarter, a written report shall be submitted to NMSS and the Regional Office of the Commission within 30 days; identifying the cause for exceeding the limit, and the corrective actions to reduce release rates.
Page: 3-4 September 9,1998 Revision: 6
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PARTI - CHAPTER 3 - RADIATION PROTECTION l
l If the gross beta radioactivity in planned effluent discharges exceeds 25% (based on a quarterly average) of the corresponding DAC specified in i
Table 2 of Appendix B to'10CFR20, an investigation of the system will be conducted to determine the reasons for abnormallevels of release.
If the parameters important to a dose assessment change such that the parameters no. longer represent boundary conditions, a report shall be submitted within 30 days which describes the changes in parameters and includes an estimate of the resultant change in dose commitment (Ref.: Order to Modify License, January 28, 1980).
Personnel exposures monitored in uncontrolled areas will be assessed and evaluated according to section 3.2.3.
3.2.2.3 Ventilation Neaative Pressure A ventilation system will be provided for each Radiologically Controlled Area to maintain areas of higher contamination at a slight negative pressure to j
uncontrolled areas. For temporary RCAs, negative i
pressure does not have to be maintained if HP-review determines radiological conditions do not require negative pressure.
I 3.2.2.4 Hoods and Gloveboxs Hoods or equivalent airborne activity control devices are installed and utilized so as to maintain airborne contamination levels as low as reasonably achievable, consistent with operational requirements. The degree and type of containment l
required for individual operations shall be determined l
by Health-Safety based on the airborne radioactive particulate generetion potential including the:
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Revision: 6 l
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FRAMA TOME COGEMA FUELS - LYNCH:URG MANUFACTURING FA0lLITY USNRC LICENSE SNM-1168, DOCKET 70-1201 PARTI - CHAPTER 3 - RADIATION PROTECTION potential for alternation of physical form or characteristics of the material being processed.
possibility for resuspension based on the morphological characteristics of the material and the nature of the operation.
potential for development of abnormal or unusual conditions.
previous operational history and air sample documentation.
Hoods and gloveboxes shall be constructed of fire resistant materials. Windows and viewing ports, when included shall be fire resistant. Exhaust flow rates from gloveboxes shall be such that a minimum negative pressure of 0.25 inches of water is maintained when openings are closed. Each hood or air capture device shall be evaluated and assigned a specific flow criteria against which its performance will be measured in accordance with industrial standard " Industrial Ventilation: A Manual of Recommended Practice, American Conference of Governmental Industrial Hygienists" as applicable.
" Elephant trunk" drops shall be used as required for maintenance or non-routine activities requiring exhaust ventilation and where other, more permanent types of containment are not practicable.
Minimum flow rates on elephant trunk drops will be 1000 LFM.
Except as noted above, elephant trunks shall not be used as a routine control measure without the specific approval of Health-Safety.
Air velocity surveys shall be conducted using heated l
thermocouple anemometers or equivalent. Higher velocities may be required, based on Health-Safety l
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l FRAMA TOME COGEMA FUELS - L NCHBURG MANUFACTURING FACILIT USNRC LICENSE SNM-1168, DOCKET 7.0-1201 PARTI - CHAPTER 3 - RADIATION PROTECTION survey data. Glovebox negative pressures, air capture device face velocities, and elephant trunk air l
flows shall be monitored weekly, except during plant shutdown of a week or longer.
l Pressure differential shall be monitored by manometers or equivalent devices.
3.2.3 Work-Area (Controlled Areal Air Samolina To verify the effectiveness of the ventilation systems, contamination control enclosures, or air capture devices, air samples in work areas and at fixed locations will be collected and analyzed.
Work area air sampling will be performed in accordance with Regulatory Guide 8.25, " Air Sampling in the Workplace".
Where monitoring is required, breathing zone air sample results will be evaluated on a shiftly basis to ensure employee exposures are consistent with estimates for the work operation and within the scope of the company's ALARA goals and policy. Static or fixed air sample results may be used to support the breathing zone sample evaluation and determine the effectiveness of various air capture l
systems and contamination control enclosures. The evaluation period for fixed air samples will not exceed a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period during periods of plant operation.
Upon initiation of new operations, non-repetitive operations, or operations modified such that previous airborne contamination levels may be affected or are unknown, air sampling shall be used to verify that satisfactory control is established and maintained.
Records of air sample result investigations including corrective actions taken will be maintained by Health Safety.
3.2.4 Radioactivity Measurement Instruments 3.2.4.1 Portable Instrumentation l
Portable instrumentation used to evaluate genera!
area radioactivity shall be capable of detecting i
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,USNRC LICENSE SNM-1168, DOCKET 70-1201 PARTI - CHAPTER 3 - RADIATION PROTECTION l
contamination levels at or below license limits or action levels.
3.2.4.2 Effluent Measurements Analytical capability for evaluation of effluent samples shall be such that instrument sensitivity and sample preparation techniques allow detection of the effluent stream at or below the most restrictive action level or release limit.
3.2.4.3 Criticality Monitorina System A criticality monitoring system shall be maintained in j
compliance with the appropriate sections of 10CFR70. Response time for the systen shall be in accordance with Regulatory Guide 8.5. " Criticality i
and Other Interior Evacuation Signals" dated March 1981. The criticality monitoring system will be functionally tested at least quarterly and detector units calibrated annually.
3.2.4.41nstrument Calibration All instrumentation shall be calibrated prior to first use, following major maintenance, and at other times as deemed necessary, in any case, all instruments used for documented survey results or monitoring of personnel exposures shall be calibrated at six month intervals. Other instruments used for informational purposes (e.g., self-reading pocket dosimeters and frisking equipment used +o determine decontamination effectiveness within a known contaminated area) may be calibrated at annual frequencies. Laboratory counting instruments shall be calibration checked on a daily basis when in use.
Calibration records shall be maintained for a minimum of two years. Calibration shall be performed on-site by trained personnel or by another j
facility licensed to perform such work. The
[
cnticality monitoring system is calibrated as j
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FRAMA TOME COGEMA FUELS - I.YNCHBURG MANUFACTURIN FACKITY USNRC LICENSE SNM-1168, DOCKET 70-1201 o
PARTI - CHAPTER 3 - RADIATION PROTECTION indicated in section 3.2.4.3. All instruments shall be j
subject to a routine functional test for response to l
the appropriate type of radiation using a check.
l-3.2.5 Radioloolcal Surveillance and Monitorino To monitor the radiological conditions in work areas and ensure the l
continued effectiveness of the radiological control program, routine surveys will be conducted. Routine surveys shallinclude surface contamination and radiation levels in and adjacent to Radiologically l
Controlled Areas (RCA) and Contaminated Areas (CA). Surveys shall also be conducted in uncontrolled areas to monitor the spread l
of contamination beyond established RCA boundaries. Survey l
techniques and instrument selection will be based on the isotopes l
and type of radiation being monitored.
Radiation levels identified above ambient will be posted and controlled as required in 10CFR20. Contamination levels identified will be posted and controlled with regard to isotope and work i
operation as follows. For Uranium fuel fabrication operations, an item or area will be controlled as a Radiologically Controlled Area and Contaminated Area (RCA/CA) when the alpha contamination 2
2 levels exceed 200 DPM/100 cm smearable or 1000 DPM/100 cm l
fixed. For SERF operations (by-product materials), an item or area will be controlled as a RCA/CA when beta / gamma contamination l
2 levels exceed 1000 DPM/100 cm smearable or 0.1 mR/Hr fixed.
To the extent practicable, uranium fuel processing and by-product material handling will not be pcrformed simultaneously in the same RCA/CA. Alpha smearable contamination levels will be investigated 2
above 20 DPM/100cm smearable in by-product material work I
areas.
Contamination levels which exceed those stated above will be l
evaluated with regard to the type of work to be performed on the equipment or in the area. This information will be the basis for development of the Radiological Work Permit and/or design of l
engineering controls for the work activity.
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FRAMA TOME COGEMA FUELS - LYNCHBURG MANUFACTURING FACILITY
~ USNRC LICENSE SNM-1168, DOCKET 70-1201
. PARTI - CH, APTER 3 - RADIATION PROTECTION 3.2.6 Personnel 'Exoosure Monitorina Employee exposures shall be determined using a variety of methods based on the operations being performed in each area. Monitoring requirements will be established for uncontrolled work areas and be established by RWPs for entry into Radiologically Controlled Areas (RCAs).
External monitoring for all work areas shall be performed by thermoluminescent dosimeters (TLD) or similar devices. These devices include whole body and extremity dosimeters as required when employees are likely to exceed 10% of the applicable annual limit.
Internal monitoring for fuel fabrication operations shall be performed by measuring air concentrations with lapel air samples, and may be supplemented by routine or diagnostic bioassay measurements for the purpose of verifying the effectiveness of the airborne monitoring program or when a significant intake is suspected. Bioassay techniques will generally be limited to in-vitro urine sarnples or in-vivo lung count measurements. In general, bioassay measurements will not be used to determine exposures for routine operations.
j in mixed fusion product work areas (SERF RCAs), historical data indicates that personnel are not likely to receive in excess of 10%
of the annual Committed Effective Dose Equivalent (CEDE) limit.
For this reason, periodic air sampling in the area and in-vivo 2
bioassay will be used to confirm that breathing zone air sampling is not required. Workers will be monitored using in-vivo measurements prior to and following work periods at this site or upon termination of employment. This measurement will be performed to confirm that no significant intakes have occurred during work at this facility. For work where intakes are considered possible, specific monitoring requirements (using air monitoring or bioassay methods) will be established by the RWP for the specific task.
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USNRC LICENSE SNM-1168, DOCKET 70-1201 PARTI - CHAPTER 3 - RADIATION PROTECTION l
l 3.2.7 Use for Unrestricted Handlino (RFUH) l I
Equipment and areas previously contaminated above the limits l
specified in section 3.2.5 of this document shall be surveyed and l
released for unrestricted handling (RFUH) in accordance with
" Guidelines for Decontamination of Facilities and Equipment Prior to Release for Unrestricted Use or Termination of Licenses for l
Byproduct, Source, or Special Nuclear Material", USNRC, August 1987, Exhibit A to section 1 of this document.
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USNRC LICENSE SNM-1168, DOCKET 70-1201 PARTI - CHAPTER 4 - NUCLEAR CRITICALITY SAFETY 4.1 Administrative Conditions 4.1.1 Desian Philosophies The double contingency principle as defined in the American National Standard ANSI /ANS-8.1 shall be followed in establishing nuclear criticality safety for all equipment, systems and operations.
Process designs shallincorporate 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, Radiation Protection is responsible for controlling all l
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 qualifications of nuclear criticality safety specialist as defined above with two years of experience as a Nuclear Criticality Specialist or a PhD in Nuclear l
Engineering with one year experience in nuclear criticality analysis.
l All evaluations shallinclude an appropriate statement of this review.
Both the analyzing and reviewing nuclear criticality safety specialist i
are independent of LMF manufacturing operations. A library of Page: 4-1 September 16,1998 Revision: 10 i
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USNRC LICENSE SNM-1168, DOCKET 70-1201 l
PARTI - CHAPTER 4 - NUCLEAR CRITICALITY SAFETY J
l validated computer codes and cross sections shall be maintained l
and utilized for performing nuclear criticality safety evaluations.
1 Clarifications and conformance of limits established by license or by the nuclear criticality safety evaluation are made by qualified personnel in the Health-Safety Section addressed in 2.2.
4.1.3 Aporovals and Documentation I
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 i
Authorized modifications shall be limited to the following areas.
a.
Modification of product specifications (except enrichment) as described in 4.2.1 and therefore in the safety criteria for fuel assembly processing, storage and packaging.
b.
Changes in the SNM handling and fuel rod processing areas limited to:
Relocation or expansion of fines and scrap storage areas Relocation, arrangement, redesign, or addition of equipment Relocation or substitution of individual process units when such changes do not alter the associated fundamental nuclear safety specifications. In these cases, changes from slab control to volume control or substitution of different process steps shall be permissible as long as nuclear interaction and individual unit safety f
criteria are met.
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^ FRAMA TOME COGEMA FUELS - L YNCHBURG MANUFACTURING FACILITY
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USNRC LICENSE SNM-1168o DOCKET 70-1201 PARTI - CHAPTER 4 - NUCLEAR CRITICALITY SAFETY l
4.1.3.2 Basic Criteria Modification of product or process involving SNM may be made without license amendment provided such changes are demonstrated to satisfy the following criteria:
The license conditions specified in this part are not violated.
j Nuclear safety calculations (when needed) shall be performed by a method equal to or better than those originally utilized to demonstrate safety for the area and type of operation under consideration. Calcu-lations shall consider both individual unit and interaction safety.
Nuclear safety evaluation is not required where the anticipated changes does not affect existing nuclear safety control or bases. The Safety Review Board Chairman or his qualified designee determines when a change in product, process, or equipment requires a nuclear criticality evaluation or license amendment.
4.1.3.3 Review and Acoroval Process
)
The review and approval procedure for plant modifications shall be as follows:
a.
The proposal shall be reviewed by the Safety Review Board Chairman or his qualified designee for content, completeness, and conformance with previously evaluated conditions. If 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 l
the Safety Review Board Chairman or his qualified designee. All criticality safety evaluations are re-viewed as specified in Part 4.1.2 of this section.
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', "*FRAMATOME COGEMA FUELS:L5NCHiURG MAN 5F5CVORiNG F5CILITY USNRC LICENSE SNM-1168, DOCKET 70-1201 PARTI - CHAPTER 4 - NUCLEAR CRITICALITY SAFETY l
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.
l d.
Appropriate personnel are notified in writing of the result of the review.
e.
The proposed change shall not be implemented until a preoperational audit (Ref. 4.1.7) has been satisfac-torily completed. The individual who performed either the nuclear criticality evaluation or the independent review shall participate in the preoperational inspection.
f.
Correspondence, calculations, and other material shall be maintained on file for a minimum of 2 years or six months following termination of the operation whichever is longer.
4.1.4 Procedures I
Written procedures approved by plant management shall be utilized for all operations involving SNM. Criticality safety requirements shall be appropriately referenced in these procedures to assure continued safe handling of SNM. The process for approval of plant modifications as l
describad in 4.1.3.2 shall be covered by approved written procedure.
Procedure control and distribution is as described in Section 2.6.
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FRAMATOME CGGEMA FUELS -TYWCHRGRG MANUFACTURING FACILITY
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1 USNRC LICENSE SNM-1168, DOCKET 70-1201 PARTI - CHAPTER 4 - NUCLEAk CRITICALITY SAFETY i
4.1.5 Postinas l
Nuclear safety postings approved by Health-Safety shall be maintained i
specifying nuclear safety parameters that are subject to procedural controls.
l Area postings may be used for those parts of the process where nuclear safety is maintained by means of " area-wide" requirements.
j Operations for which specific (dedicated) controls are applicable shall be individually posted.
4.1.6 Dedicated Controls i
The LMF primary criticality control for SNM prior to fuel bundle assembly is the four inch slab criteria and moderation control. Upon completion of fuel rod fabrication, spacing and moderation control are j
relied upon. These controls are maintained by the following:
l 4.1.6.1 Moderation Control 1
All accumulations of liquids and hydrogenous materials are regulated through the local safety rules. The autho-rized volume per container is one gallon with a total volume not to exceed 20 gallons. This criteria is exempt when SNM materialis absent. For the fuel assembly ar-eas, water lines are baffled.
t 4.1.6.2 Pellet Receipt l
Pellets are received by the vendor in approved shipping containers which contain up to six cardboard boxes of fuel pellets. When they are removed from the shipping container, the boxes are stacked two high (8 inch stab) in-transit to the conveyor. This only occurs for a very short period of time and only one shipping container (up to six l
boxes of pellets) are transferred at any one given time.
l They are placed on the conveyor one cardboard box at a l
tirr,e which initiates the four inch slab. There is a l
physical barrier on the conveyor to prevent anything above four inches from entering the pellet loading room.
4.1.6.3 Pellet Vault Storage Page: 4-5 September 16,1998 Revision: 10 l
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PARTI - CHAPTER 4 - NUCLEAR CRITICAllTY SAFETY I
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The pellets remain in the box for storage purposes. The i
vault is constructed with a barrier to prevent double stacking of the boxes. Only if the boxes were manually lifted above this barrier could double stacking result and this would take great effort on one's part to do and could I
not be accomplished accidently.
4.1.6.4 Fuel Rod Loading j
For loading purposes, the pellets remain stacked on the corrugated trays in a four inch slab. All hoods, the pellet j
down draft table and carts are either twenty-eight inches in width or less or a red line indicates the allowable twenty-five inch width space. The four inch height limitation is also marked in the hoods.
4.1.6.5 Fuel Rod Fabrication All fuel rod channels are designed to support the four inch l
slab criteria The rods could not be stacked above the four inches in the channels without spillage.
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4.1.6.6 Fuel Assembly Room j
Fuel rods are stored in the four inch channels and the channels are stored on'a rack. Local safety rules specify that the channels cannot be stored above or below one
- another, j
L The fuel bundle tables are constructed out of granite block weighing approximately 8000 pounds and are L
spaced in accordance with the nuclear criticality evalua-tion and would be impossible to move.
I 4.1.6.7 Fuel Assembly Storage The assembly racks are constructed to meet the spacing criteria for 4.1% For enrichments greater than 4.1%,
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every other storage space in the rack will be used. A dedicated modified rack with 42 inch center to center spacing will be used for assemblies > 4.1 wtE i
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USNRC LICENSE SNM-1168, DOCKET 70-1201 o
PARTI - CHAPTER 4 - NUCLEAR CRITICALITY SAFETY
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i l-f 4.1.7 Preooerational Testina I
Specifications and design criteria for purchased or locally fabricated equipment where nuclear and radiological safety considerations are involved shall be approved by a knowledgeable representative of the Health-Safety Section addressed in 2.2. Before being released for l
production operation, new equipment shall be tested to assure that l
safety specifications are satisfied. Favorable geometry equipment i
shall be measured by a knowledgeable person to ascertain that it is of proper dimensions before it is put into service. Where operational safety is based wholly or in part on the use of electrical or mechanical interlocks, the proper functioning of interlocks shall be verified upon in-L stallation and on an annual basis thereafter. Routine plant inspections l
place added emphasis on new operations. No equipment is used after being removed from service until an equipment checkout for continued effectiveness of safety related parameters is performed.
4.2
' Technical Criteria -
I 4.2.1 Individual Units Unless otherwise provided for, the safety of individual units assumes l
all SNM is uraniurn oxide with a maximum 2asU enrichment of 5.1 i
wt.%. The material may be pellets and pellet chips / powder. The most I
reactive heterogeneous geometry is fuel pellets. Pellet diameters have
.been varied over the range from 0.2" to 0.5" and it has been found that the most reactive pellet with optimum moderation is about 0.30 inches. Fuel pellets can have any density up to and including the l
theoretical maximum for uranium oxide (10.96 g/cm ).
Certain plant areas utilize slab thickness as a means of nuclear safety.
The 4 inches thick slab is the maximum allowed for any operation in the plant except for the unloading of pellets where an 8 inch slab is allowed during the time that the fuel boxes are being transferred from the shipping container to the vault conveyor where the 4 inch slab limit starts. Boxes of fuel pellets come in shipping containers with fuel stacked up to 8 inches high. The size of the 4 inches thick slab is un-limited for enrichments below 4.1 wt% and is limited to a 25 inches wide and infinite length slab with 12" horizontal separation between slabs for enrichments above 4.1 wt% when fuel is in the unciad state.
l The 4" thick stao aize for clad fuel is unlimited for enrichments up to l
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l USNRC LICENSE SNM-1168, DOCKET 70-1201 l.
PARTI - CHAPTER 4 - NUCLEAR CRITICALITY SAFETY l
l 5.1 wt%. This provision for different nuclear safety limits for different enrichmeats implies knowledge of the enrichment level of the SNM in-volved. As a result, the enrichment of SNM at the LMF must be known at all times. Fuel boxes with enrichments greater than 4.1 wt% will be prominently identified. If the enrichment is unidentified, the SNM in question will be stored and handled as if it were the most reactive enrichment authorized (5.1 wt%) pending reestablishment of the identity. Enrichment of SNM is verified by shipper's documents and overchecked by Quality and nuclear material accountability re-quirements.
1 All operations, except for fuel assembly processing, assume, as an accident condition, the unrestricted presence of water moderation and reflection of variable density. All manufacturing steps, except for the final product (assembled fuel bundles) and where positive controls are present, are assumed to be optimally moderated as an accident condition. The potential presence of other moderators, such as polyethylene, is considered and controlled.
The safety of individual units is based either on a calculation of K,,, or on data from an approved handbook or document. Calculated safe units are those evaluated according to 4.1.3.3 and shown to have a K,,, not exceeding 0.87 under normal conditions and 0 95 under assumed accident conditions considering both statistical and methodology limits of error.
Calculations of K.,, assume the presence of any nearby reflector, such as a concrete floor.
For simple shapes, safe values are determined by application of the following limits:
Mass:
45% of minimum critical reflected mass Mass:
75% of minimum critical reflected mass when double batching is not credible Volume:
75% of minimum critical reflected volume l
Diameter:
90% of minimum critical reflected cylinder diameter Thickness:
90% of minimum critical reflected slab thickness l
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,USNRC LICENSE SNM-1168, DOCKET 70-1201 PARTI - CHAPTER 4' - NUCLEAR CRITICALITY SAFETY 4.2.2 Multiple Units and Arravs The safety of multiple units and arrays is determined by calculation of K,,, using an industry recognized validated and benchmarked code such as the SCALE computer code package. Calculated safe arrays are those evaluated according to 4.1.3.3 and shown to have a K,, not exceeding 0.87 under normal conditions and 0.95 under assumed accident conditions. The maximum allowed K,, for a postulated accident is equal to the calculated K,, + 2 sigma where sigma would apply if the calculation was made by a statistical code such as KENO.
The allowed maximum value of 0.95 including bias is considered to have an adequate safety margin based on our benchmark calculations which show that when compared to critical experiments, our bias has not exceeded 0.02 A K.,,in the non-conservative direction. In the event that any future benchmark comparisons show a non-conservative bias in excess of 0.02 A K.,,, then the acceptable K,,
would be correspondingly reduced. The accident condition assumes any credible interspersed moderation of varying density. The presence of any nearby concrete reflectors is accounted for in calculations of K.,,. The 4-inch thick, 25-inch wide finite slabs of fuel pellets in 4.2.1 with enrichments greater than 4.1 wt% must be spaced a minimum of 12 inches apart horizontally.
l 4.2.3 Technical Data and Calculational Methods 4.2.3.1 Isolated Units or Arrays Acceptable nuclear interaction between individual units and between arrays have been computed using the SCALE computer code package. The generalized relation for acceptable nuclear isolation is given by the following:
The greatest distance across an orthographic projection of either accumulation on a plane perpendicular to a line joining their centers, or a distance of 12 feet, whichever is greater.
L The Safety Review Board Chairman or his qualified l
designee, with the concurrence of a nuclear criticality safety specialist, shall determine which arrays are more reactive than those used as a basis for the evaluations in Page: 4-9 September 16,1998 Revision: 10
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I this license. In these cases, the interaction acceptance criteria shall be evaluated on an individual basis by a l
nuclear criticality safety specialist or it shall be as defined in the generalized criteria given above.
4.2.3.2 Safe Units l
The following individual units are defined for the LMF as safe, independent of the degree of water moderation or l
reflection. All are authorized for uranium in any form up to 5.1 wt.% ssU enrichment:
2 l
l a.
Safe volume - 12 liters maximum capacity b.
Geometrical safe containers - defined as a container with a dimension of 8 inches ID by 14.5 inches high, or other cylinders that may be demonstrated to be safe.
c.
Safe mass - 700 grams assU d.
Safe size slabs of 4 inches thickness were 2
determined as a function of asU enrichment (other than vault storage) for both unciad fuel pellets and clad fuel rods. Slabs are assumed to be infinite in length for all cases. Slabs are to be a minimum of 12" above any concrete floor. Concrete is the best reflector that is available at the LMF. Reflecting material better than concrete such as beryllium or lead are not available in any significant quantities at the LMF and any significant quantities of these materials will not be allowed in any of the fuel processing areas.
l l
i l
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23sU Maximum Slab Enrichment Width Wt.%
inches 1 4.1 Infinite (unciad)
> 4.1 15.1 25.0 (unciad) 1 5.1 Infinite (clad) 4.2.3.3 Calculational Methods Computer codes will be used to calculate K,, of individual units and of arrays. Computer codes and associated cross section sets are always benchmarked to demonstrate their validity for application at the LMF. The particular family of codes may change over time as improvements are realized in the calculational methods.
The current computing package in use is the system of computer codes for performing Standardized Computer Analyses for Licensing Evaluations (SCALE) which was developed by ORNL for the USNRC. The LMF's goalis to maintain its computing capability at state-of-the-art for nuclear criticality safety and to assure that all codes and cross sections are properly benchmarked and validated.
The SCALE package, Rev 5, has four basic cross section libraries relevant to LMF evaluations. The SCALE package also has a supplernentary set which is not used at the LMF.
1.
The Hansen-Roach 16-group library which includes the Knight modification, and additional nuclides based upon the 218-group ENDF/B-IV cross sections set as modified for compatibility for SCALE usage; 2.
The 27-group ENDF/B-IV broad group library collapsed from the 218-group ENDF/B-IV library; l
3.
The 44-group ENDF/B-V broad group library collapsed from the 238-group ENDF/B-V library; Page: 4-11 September 16,1998 Revision: 10
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PARTI - CHAPTER 4' - NUCLEAR CRITICAllTY SAFETY l
- 4.
The 238-group ENDF/B-V broad group library, including ENDF/B-VI cross sections from selected nuclides; FCF has made extensive validation calculations against l
critical experiments using the four cross-section sets and I
has determined that a 0.02 bias added to the calculated value is more than adequate to adjust all validation calculations to a predicted k,n of unity or greater except l.
for carbon moderated systems which do not apply at the LMF. The critical experiments used for validation have l-included: a) uranium enrichments ranging from 2% to l
near 100%, b) ~ optimum and non-optimum moderation, c) moderation (H O) densities' ranging from 100% down to 2
24%, and d) both clean and poisoned lattices.
l Generally the 44-group cross-section set is used for i
enrichments encountered at the LMF. This is done to avoid excessive conservatism and to provide a good l
treatment of the U resonance absorption which is 28 l
particularly important with low enriched material.
i 4.2.4 Special Controls i
l This section outlines those nuclear safety controls concerning plant specific equipment, arrays or areas at the LMF. Each area will be addressed separately. Alternative designs satisfying the provisions of 4.1.3.3 may be substituted for any area described in this section.
i
.4.2.4.1 Fuel Pellet Receivina and Temoorary Storaae A single shipment of SNM may be stored in the shipping containers in an array no more reactive than as received pending transfer to planned storage facilities. Nuclear l
interact _'on between the shipping containers and other-L SNM will be as specified in Section 4.2.3.1.
Shipping containers shall be examined for transit damage after receipt and Health-Safety notified if damage is found. Health-Safety shall examine the damaged con-i tainers for evidence of water entrainment or other l
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potential hazardous conditions. Damaged c'ontainers shall be opened one at a time and the contents placed in a 1
nuclear safe configuration (i.e., safe slab or 700 grams 23sU) in accordance with applicable storage license conditions. Damaged containers containing SNM shall not be returned to the shipper.
The use of conveyors to transport pellets shall be 1
maintained under the safe 4" thick X 25" wide infinite
]
length slab criteria outlined in Section 4.2.3.2.
Transport of SNM in this area (other than by conveyor) shall be with carts as follows:
No SNM accumu!ation shall be positioned above or r
below any other SNM accumulation; SNM on the cart shall be arranged such that the 1
SNM slab thickness is less than, or equal to,4 inches and less than 25 inches wide.
Other than in transit, a 12' separation distance must be maintained between the edge of the cart and the edge of any other SNM accumulation.
4.2.4.2 Pellet Vault Storaae l
Handling and storage of SNM with enrichments up to 5.1 wt% in the Pellet Vault Storage area will be as sate j
geometry slabs with a maximum thickness of 4" and a maximum width of 25". The slabs can be infinite in length.
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 aluminum plate with an overall 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 and 19 feet long shelves for safe geometry 4" thick slab storage. Vertical spacing between l
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l l
' shelves shall be at least 16 inches. A neutron poison as j
described above shall be placed on each shelf level except j
for the bottom. Horizontal spacing between tiers within a cubicle shall be at least 9 feet-11 inches.
Other nuclear safety controls on pellet vault storage are as follows:
Shelves with multiple enrichments are limited to a j
maximum slab thickness of 4 inches.
The vault is separated from other SNM storage and processing by eight inch thick concrete walls.
Material storage and handling in the vault is controlled based on criteria specified by Health-Safety personnel. Enrichment determination j
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 the 4.0" slab thickness.
Transport of SNM in the pellet vault area shall be with
]
carts applying the same nuclear safety criteria as described for that area.
4.2.4.3 Fuel Rod Loadina The fuel rod loading area is located adjacent to the pellet vault and may involve handling and storage of unclad pellets in a:eas other than the vault. Nuclear safety in this area is maintained as safe geometry slabs as de-scribed in 4.2.3.2.
j Other nuclear safety controls associated with the fuel rod l
loading area are:
)
No SNM accumulation shall be positioned above or below any other SNM accumulation.
Page: 4-14 September 16,1998 Revision: 10
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~FRAMATOME COGEMA FUELS - LYNCHL'URG MANUFACTURING FACILITY USNRC LICENSE SNM-1168, DOCKET 70-1201 l.
PARTI - CHAPTER 4 - NUCLEAR CRITICALITY SAFETY l
l l
SNM in the rod loading area shall be arranged such
)
that the SNM slab thickness is less than or equal to 4 inches and less that 25 inches wide. The slab is i
to be spaced 12" horizontally from any other SNM l
accumulation. The slab will also be spaced a l
minimum of 12" above any concrete floor.
j 4.2.4.4 Fuel Rod Processina and Storaae Fuel rod processing operations are those performed after the cladding is loaded with UO pellets. Typically, these 2
functions include welding, leak testing, cleaning, visual and dimensional inspection, etc. Certain operations may require that the rods be partially or wholly immersed in liquids. These liquids shall be of such a nature that the cladding would not be compromised.
Nuclear safety controls applied to the fuel rod processing area are as follows:
Individual accumulations shall be limited to a 4.0 inch thick slab unless specifically authorized by Section 4.1.3.
No SNM accumulation shall be positioned above or below any other SNM accumulation.
Transport of fuel rods from storage to the assembly area shall be in maximum of 4.0" slab.
4.2.4.5 Fuel Assembiv Processina Fuel assembly processing operations are those which consists of loading fuel rods into an assembly configuration, dismantling assemblies, and other fabrication steps such as alignment, welding, inspection operations, etc.
The actual assembly of rods into bundles (or removal l
therefrom) shall be performed in an area containing no i
sprinkler system or similar facility for general moderation.
(
Moderating materials in the assembly area shall be l
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USNRC LICENSE SNM-1168, DOCKET 70-1201 PARTI - CHAPTER 4 - NUCLEAR CRITICALITY S/ FETY limited to only those essentic! to the operations. Liquid moderators shall be cleaning agents used only for local wiping rather than immersion and shall be controlled by limiting the volume of individual containers to one gallon not to exceed more than a total of 20 gallons. Water or similar moderator piping that must be in the area shall be protected from accidental rupture by virtue of its location, or by protective guards, and shall have the joints baffled to prevent spraying in-process assemblies in the event of a break. The use of polyethylene or sirnilar moderators shall be limited to one sheet per fuel rod channel.
Nuclear safety controls applicable to the fuel assembly processing area are:
A minimum separation of 4 feet shall be maintained between fuel bundle assembly stations.
The assembly station equipment shall be such that liquids cannot be retained in and around the in-process operation.
Assembly processing operations may be performed without moderation control provided the fuel rods are restrained in the fuel assembly configuration.
Adjacent assemblies shall be separated by a center-to-center distance of 38 inches.
l jor the purposes of nuclear interaction control, the fuel assembly processing and fuel assembly storage arrays are considered as a single array. Adjacent arrays are considered to be no more reactive than an array of fuel assemblies loaded in adjacent containers as described in 4.2.4.6.
The following nuclear safety controls shall be imposed on fire fighting within the fuel assembly processing area:
The use of hoselines shall be prohibited unless authorization has been received from a i
management representative of the plant emergency response organization.
Page: 4-16 September 16,1998 Revision: 10
FRAMA TOME COGEMA FUELS - LYNCHBURG MANUFACTURING FACILITY USNRC LICENSE SNM-1168, DOCKET 70-)201 PARTI - CHAPTER 4 - NUCLEAR CRITICALITY SAFETY Simultaneous application of more than one hoseline is not authorized.
Area postings shall be maintained at the array perimeter specifying limitations on fire fighting techniques.
4.2.4.6 Fuel Assembiv Storaae and Packaaina Fuel assemblies may be stored in racks meeting the following spacing and packaging criteria:
Fuel assemblies stored in linear arrays shall have a minimum 21" center-to-center spacing for enrichments below 4.1 wt% and a minimum of 42" center-to-center spacing for enrichments above 4.1 wt% but less than or equal to 5.1 wt% 23sU.
A minimum of 38" center-to-center spacing shall be maintained between the nearest assemblies of adjacent planar or linear storage arrays.
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 assemblies. No stream sources or sprinklers shall be located near the storage array. A 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 boundaries of the assembly storage array.
The following nuclear safety controls shall be imposed on fire fighting within the fuel assembly storage array:
The use of hoselines shall be prohibited unless authorization has been received from a management representative of the plant emergency response organization.
Page: 4-17 September 16,1998 Revision: 10
FRAMA TOME COGEMA FUELS - LYNCHBURG MANUFACTURING FACILITY USNRC LICENSE SNM-1168, DOCKET 70-1201 PARTI - CHAPTER 4 - NUCLEAR CRITICALITY SAFETY Simultaneous application of more than one hoseline is not authorized.
Area postings sha l be maintained at the array perimeter specifying limitations on fire fighting techniques.
4.2.4.7 Fuel Assembiv Shionina Containers Fuel assembly packaging and unpackaging operations involving licensed shipping containers shall be performed within the following limitations:
Fuel assembly packaging shall be in accord with the requirements of the container certificate.
Fuel assemblies in adjacent containers shall have a minimum 18" edge-to-edge separation distance.
The edge-to-edge separation distance between fuel assemblies in the storage rack and assemblies in shipping containers shall be a minimum of 38 inches.
Fuel assemblies may be stored in licensed shipping containers after being prepared for shipment. During such storage, the following conditions shall be maintained:
Loaded shipping containers shall be handled using only authorized methods.
Loaded shipping containers shall not be stacked more than two (2) high.
Fuel assemblies stored in NRC licensed shipping containers are exempt from the criticality monitoring requirements of 10 CFR 70.24 when the following conditions are maintained:
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FRAMA TOME COGEMA FUELS - L YNCHBURG MANUFACTURING FACILITY USNRC LICENSE SNM-1168, DOCKET 70-1201 PARTI - CHAPTER 4 - NUCLEAR CRITICAllTY SAFETY I
I The containers and contents have beeri subject to j
the inspections and determinations required by 10 CFR 71, Subpart D.
The containers are sealed, properly identified, and meet shipment requirements.
i A minimum of 38" edge-to-edge separation is maintained between the loaded container array and storage rack.
l 4.2.4.8 Outside Storaae of Radioactive Material UF cylinders or other radioactive materials in shipping containers may be stored outside the LMF main building within the perimeter fence. The area used for such storage shall be prepared and maintained under weed control.
Areas used for UF cylinder and other radicactive material storage shall be monitored quarterly. No combustibles shall be stored in the area except for timbers / pallets necessary for structural support. Other specific conditions for the area are as follows:
UF; Cylinder Storaae I
The UF area will be kept segregated from other radioactive material storage areas for ease of handling.
The UF. storage area is enclosed by a barrier (i.e.,
rope, fence, etc.) when cylinders are not being loaded or unloaded.
UF cylinders shall be equipped with valve protectors during movement and storage.
Enriched UF, will be stored in a single planar array l
(natural or depleted U does not require geometric j
controls for criticelity safety).
Other Radioactive Material Storaae Page: 4-19 September 16,1998 Revision: 10 t
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\\
O l
l The material is properly packaged per NRC/ DOT i
requirements.
Predefined storage areas must be approved by the l
Safety Review Board.
l The loaded containers will be stored in an enclosure (i.e., building, trailer, sea container, other type structure, etc.) that will provide protection from the elements.
l Natural or depleted uranium does not require geometric controls for criticality safety.
4.2.4.9 Miscellaneous Criticality Controls Laboratory Ooerations Each laboratory shall be limited to 700 grams 23sU with enrichments up to 5.1 wt%. Inventory will maintained by l
logging materials into and out of the laboratory.
l l
Development Ooerations Development operations are of a nature such that specific l
geometric controls cannot be established in advance.
Development operations shall be limited to either the geometric controls imposed for the manned processing operations, or 350 grams assU limit per accumulation.
Shiocina Preparation for shipment and shipping shall be in accord with specific NRC and DOT approvals for the container and materials involved.
Solid Waste Solid wastes generated such as packing media, wiping cloths, etc., but not including reject SNM do not require special handling for nuclear safety.
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, USNRC LICENSE SNM-1168, DOCKET 70-1201 PARTI - CHAPTER 4 - NUCLEAR CRITICAllTY SAFETY Liould Waste Liquid wastes containing significant quantities of uranium shall be stored in a safe container limited to 700 grams of 23sU. The container (s) may be stored in the laboratory as a part of the laboratory inventory, as a 4.0" slab in the pellet vault, or in other approved locations.
Shiocina Container Storaae SNM may be stored in authorized shipping containers in an array that iF no more reactive than that specified for the same mater 6i under transport conditions.
Vacuum Cleaners Vacuum cleaners used for cleanup / decontamination purposes shall satisfy the safe geometric limits specified in 4.2.1 of this section.
4.2.4.10 Evaporator Sludoe The evaporator sumps shall be inspected monthly for sludge accumulation. If the accumulation exceeds one inch, it shall be removed.
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FRAMA TOME COGEMA FUELS - L YNCHBURG MANUFACTURING FA ILITY USNRCPCENSE SNM-1168, DOCKET 70-1201 PARTI - CHAPTER 6 - SPECIAL PROCESSES 6.1 Nonexemot 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 l
entail increased exposure levels. A register of approved operators shall be maintained in the Health-Safety Office.
6.1.3 Maximum whole body oxposure 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 LMc employees, appropriate self-reading dosimeters shall be utilized by personnelinvolved in source manipulation in cases where the exposure 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 transfer or 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 shal! ' i 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 micro curies and maintained for inspection by the Commission.
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6.1.5.2 If the test reveals the presence of 0.005 microcurie or more of removable contamination, the source Page: 6-1 Septemeber 16,1998 Revision: 5
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FRAMA TOME COGEMA FUELS - LYNCHBURG MANUFACTURING FACILITY USNRC LICENSE SNM-1168, DOCKET 70-1201 PARTI - CHAPTER 6 - SPECIAL PROCESSES shall be withdrawn from use and shall be decontaminated and repaired by a person appropriately licensed to make such repairs or j
disposed of in accordance with Commission regulations. Within five (5) days after determining that any source has leaked, a report shall be filed with the U.S. Nuclear Regulatory Commission describing the source, the test results, the extent of contamination, the apparent or suspected cause of source failure, and the corrective action taken. A copy of the report shall be sent to the Director the nearest NRC Regional Compliance Office listed in Appendix D of Title 10, Code of Federal Regulations, Part 20.
6.1.5.3 The periodic leak test required by this condition does not apply to sealed sources that are stored and not being used. The sources excerpted from this test i
shall be tested for leakage prior to any use or transfer to another person unless they have been leak tested within six (6) months prior to the date of use or transfer.
l 6.1.6 Adequate records shall be maintained to insure effective source documentation, including leak test results.
6.1.7 When not in use, sources shall be stored in approved secured containers in a manner selected to prevent unauthorized removal or use. Adequate posting of the source container and storage / operation area shall be maintained to insure compliance with appropriate regulations.
l l
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USNRC LICENSE SNM-1168, DOCKET 70-1201, PARTI - CHAPTER 6 - SPECIAL PROCESSES t -
O 6.2 Fire Protection The LMF fire protection program is administered by the Licensing Section j
with the Manager, Licensing & Quality Systems overseeing the program.
LMF utilizes internal Health Safety procedures to ensure adequate fire protection is established, inspected and maintained for new facilities, equipment and/or operations. Internal procedures also provide instruction for our emergency rescue team in the event of a minor or major fire. Our pre-fire plan provides a hazard analysis associated with separate plant l
areas, fire emergency plans, and training. The Plan shall be reviewed by all fire fighting support agencies and updated accordingly.
j l
Accumulations of combustible materials within the LMF shall be limited to the greatest extent practicable, consistent with operational requirements.
Supervision is responsible for assuring that areas under their cognizance are maintained in accord with good housekeeping and fire prevention practice.
6.2.1 Flammable liquids shall be stored in approved containers.
l 6.2.2 Fire extinguisher systems compatible with area nuclear safety l
requirements shall be installed or provided in accord with insurance i
and federal regulations. The above systems shall include portable extinguisher of a type (pressurized water, CO and dry chemical) 2 and size based on the potential hazard. Agents such as "Metl-X" are available in areas where metal fires may occur. Sprinkler systems and other water-type extinguishing systems are not installed in moderation controlled areas.
l 6.2.3 As part of its emergency program, the LMF organization shall j
include a " fire brigade" staffed by qualified personnel, familiar with basic fire fighting techniques, the equipment available for their immediate use onsite, and the nuclear safety and health physics considerations that are involved. Fire Brigade members shall be i
retrained at least annually. The Manager, Licensing & Quality Systems shall have oversight of the fire brigade.
I 6.2.4 Fire protection water shall be available at all times. There shall be a Iow pressure detection and alarm device and annual water i
flow / pressure tests shall be conducted routinely.
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USNRC LICENSE SNM-1168, DOCKET 70-1201 l-PARTI - CHAPTER 6 - SPECIAL PROCESSES l
l 6.2.5 Fire protection equipment shall be inspected routinely in accordance with NFPA standards to ensure reliability.
6.2.6 Areas under moderation controls shall be prominently posted and approved fire fighting techniques defined. Deviations from the posted techniques can be approved only by one of the following members of the LMF Emergency Team:
i Plant Manager, Emergency Officer, Health-Safety Officer During periods when the plant is not in operation, at least two tours per shift shall be made by a representative of the security force or a LMF employee.
6.2.7 An early warning fire detection / alarm system shall be installed in the S-1 storage facility. The system shall alarm at the continuously manned guard house. Further precautions, detailed in 10.5 of this document, will also be implemented.
6.2.8 A sprinkler system shall be installed and maintained in GERF-3 & 4 in accord with NFPA codes. Further precautions, ouuined in 10.5 of this document, will also be implemented.
6.3 Emeraency Utilities Backup battery power is provided for the criticality alarm, fire alarm, public address system and for emergency lighting. The nature of our operations is such that a loss of utilities simply results in a totally safe halt in operations.
6.4 Radioactive Wasto Control The " Guideline for Decontamination of Facilities and Equipment Prior to Release for Unrestricted Use or Termination of Licenses for By-Product, Source or Special Nuclear Material", USNRC August 1987, Exhibit A of Chapter 1 shall be used for disposal control of materials or equipment.
Additionally, these guidelines will be followed regarding radioactive waste control:
a.
Any item which cannot be thoroughly surveyed due to physical construction, painting, or other reason shall be assumed to be in Page 6-4 Septemeber 16,1998 Revision: 5 i
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PARTI - CHAPTER 6 - SPECIAL PROCESSES i
excess of the above limits and shall be disposed of in a controlled manner, unless indirect methods such as acid etching verify the absence of significant contamination.
l b.
When possible and practicable, reasonable attempts shall be made to decontaminate allitems with " detectable contamination" to a non-detectable level.
6.5 Chemical Safety l
6.5.1 Written procedures shall be used for the implementation of the i
criteria set forth in 29CFR and 40CFR for chemical safety. This is l
to include the implementation of a Hazardous Communications Program that defines the proper storage facility, equipment, and chemical handling procedures.
6.5.2 Storage of snemicals in tanks and/or buildings shall incorporate a pressure relief system. Storage shall be maintained an adequate distance from the main plant.
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11.1 Oraanizational Responsibilities The key organization responsible for maintaining the health and safety aspects at the LMF to include all SERF Facilities is the Health-Safety Section. The Health-Safety Section is a part of the Quality organization.
The Manager, Radiation Protection (Health Safety Section) reports directly to the Quality Manager.
11.2 Kev Personnel Function 11.2.1 Overall Proaram Maneaement Responsibility for planning, coordinating, administering and managing the health and safety aspects of the LMF is vested in the Manager, Radiation Protection. This position is organizationally parallel to members of the Plant Manager's staff such as the Product Center Managers.
11.2.2 The Health-Safety Section Health-Safety personnel are responsible for the general surveillance of all plant radiological safety related functions. Specifically, these functions 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.
Furnishing consulting services and advice on radiation protection to plant supervision and management.
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FRAMATOME COGEMA FUEL - LYNCH:URG MANUFACTURING FA ILITY USNRC LICENSE SNM-1168, DOCKET 70-1201 PARTll - CHAPTER 11 - ORGANIZATION AND PERSONNEL j
f 1
l Generating or acquinng, maintaining, and appropriately l
distributing all records and reports required by applicable j
regulations or procedures.
l Leak testing on sealed radioactive sources.
Developing and disseminating procedures related to radiation safety and reviewing procedures prepared by other operating sections for regulatory compliance and the adequacy of safety 4
l considerations.
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The key positions within the Health-Safety Section are the Manager l
of Radiation Protection and the Health-Safety Monitors.
11.2.3 Manaaer, Radiation Protection The Manager, Radiation Protection is responsible to provide l
management with assurance of the effectiveness of the health and l
safety program from a radiological and nuclear safety aspect. This l
position is responsible for evaluating the potential for exceeding l
authorized control limits and to recommend appropriate restrictions or corrective measures.
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The Manager, Radiation Protection is also responsible for supervising the implementation of the Health-Safety Program to assure that the requirements as defined by license and procedures are carried out. The Manager, Radiation Protection reports directly to the Quality Manager, 11.2.4 Health-Safety Monitors l
l The Health-Safety Monitors are responsible for conducting routine monitoring, sample collection and analytical tests to determine radiation and contamination levels. The Health-Safety Monitors report to the Manager, Radiation Protection.
11.2. 5 The Licensina Section l
The Manager, Licensing and Quality Systems reports directly to the Manager of Quality and is responsible for Licensing and to provide management with assurance of the effectiveness of the industrial Page: 11-2 September 10,1998 Revision: 11
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PARTll - CHAPTER 11 - ORGANIZA TION AND PERSONNEL e
l and chemhal safety programs. Licensing' personnel are responsible for the general surveillance of all plant industrial safety functions and maintaining alllicensing documents.
The key position within the Licensing Section is the Regulatory Compliance Officer.
11.2.6 Reaulatory Comoliance Officer The Regulatory Compliance Officer is responsible for implementing the occupational and industrial safety programs to include chemical and fire safety. The Regulatory Compliance Officer reports directly to the Manager, Licensing and Quality Systems.
11.2.7 SERF 3 & 4 Oraanization The SERF 3 & 4 facilities have a Site Manager to oversee the operations of both facilities. His staff includes 2 supervisors and maintenance technicians. The SERF 3 & 4 Manager reports to the Vice-President, FTl Chemistry & Environmental Services whom reports to the FTl President.
A health physicist and 3 health safety technicians from the LMF Radiation Protection organization are also permanently assigned to the SERF 3 & 4 facilities.
11.2.8 Proaram Manaaer. Comoliance and Skills Trainina The Program Manager, Compliance and Skills Training reports directly to the Manager of Quality. The Program Manager, Compliance and Skills Training is responsible for orientating and training LMF personnel in radiological, nuclear, and industrial safety.
11.3 Resumes Resumes from managerial, health-safety, and criticality organizations within the LMF are included. These are as follows:
Name Title l
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USNRC LICENSE SNM-1168, DOCKET 70-1201 1
.PARTll - CHAPTER 11 - ORGANIZATION AND PERSONNEL J.E. Niatheson, Jr. Vice-President, Operations (Plant Manager)
A.D. McKim Manager, Quality, Health / Safety & Licensing l
T.S. Wilkerson Manager, Assembly T. V. Allsep Manager, Radiation Protection (Health Safety)
C. A. Armontrout Manager, Licensing & Quality Systems (Licensing)
D. M. Minor Manager, Rods C. A. Conrad '
Manager, Components l-L.A. Hassler Senior Principal Engineer l
P.L. Holman Senior Principal Engineer
)
J. S. Whitt Program Manager, Compliance and Skills Training l
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USNRC LICENSE SNM-1168, DOCKET 70-1201 PARTll - CHAPTER 11 - ORGANIZATION AND PERSONNEL r
l NAME:
John E. Matheson, Jr.
TITLE:
- Vice-President, Operations i
(Plant Manager, Lynchburg Manufacturing Facility)
CITIZEN OF UNITED STATES i
EDUCATION: Lowell Technologies Institute BS in Physics - 1968 The Catholic University of America MS in Aerospace Engineering - 1973 The University of Virginia MS in Mechanical Engineering - 1983 1
Registered Professional Engineer in State of Virginia i
Va. License No. 0402 012314 EXPER/ENCE: 1997 - Present Framatome Cogema Fuels, Lynchburg Manufacturing Facility, Lynchburg, Va. - Vice-President, Operations (LMF Plant Manager). Overall responsibility for all plant operations.
1994-1997 Framatome Cogema Fuels, Lynchburg, Va. -
l Manager of Design & Development. Responsible for the technical content and design basis of all fuel assemblies and components. Managed the design, analysis, configuration management, and design release functions.
1990-1994 Framatome Cogema Fuels, Lynchburg, Va. -
Contract Manager. Managed the fuel contracts for fuel delivered to Duke Power reactors (7).
Responsible for defining contract needs and issuing release authorizations for manufacturing and l
design. Coordinated all Framatome Cogema Fuel activities with Duke Power Company including l
negotiating all contract changes.
1987 -1990 Framatome Cogema Fuels, Lynchburg, Va. - Project j
Manager, High Level Waste. Managed the fuel consolidation project. Designed, developed and i
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FRAMATOME COGEMA FUELS - LYNCH:URG MANUFACTURING FACILITY USNRC LICENSE SNM-1168, DOCKET 70-1201 PARTH - CHAPTER 11 - ORGANIZA TlqN AND PERSONNEL l
l tested machine to consolidate spent nuclear fuel l
(joint project with SGN in France).
l 1985 -1987 Framatome Cogema Fuels, Lynchburg, Va. -
l Supervisor, Fuel Rod Design and Analysis.
Managed the design and analysis function for fuel l
rods. Lead the development of a code to predict the consequence of pellet cladding interaction.
1 1978-1985 Framatome Cogema Fuels, Lynchburg, Va. -
Principal Engineer Fuel Design. Performed various tasks in the design of fuel assembly components, grids, structures, etc. Acted as project leader of several major development projects.
1968-1978 Naval Ordinance Station, Indian Head, Maryland.
Held various positions from an entry level engineer to unit manager for air launched missiles.
Performed various engineering duties in the design l
and analysis of tactical rocket motors, in the last position (1975 - 1978) was responsible for the basis design engineering function and logistics support of rocket motors used on all air launched missiles in the U.S. Navy.
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FRAMA TOME COGEMA FUELS - L YNCHBURG MANUFACTURING FACILITY USNRC LICENSE SNM-1168, DOCKET 70-1201 FART ll-CHAPTER 11 - ORGANIZA TION AND PERSONNEL NAME:
Alvin D. McKim TITLE:
Manager, Quality, Health / Safety & Licensing CITIZEN OF UNITED STATES EDUCATION: AS Tool Engineering Technology, ITT - 1967 AS Mechanical Engineering, University of Evansville - 1973 BSBA Indiana State University of Evansville - 1975 Graduate Engineering Courses at University of Evansville EXPERIENCE: 1997 - Present Framatome Cogema Fuels, Lynchburg Manufacturing Facility, Lynchburg, Va. - Manager, Quality, Health / Safety & Licensing. Responsible for quality, safety & licensing of plant operations including; industrial, radiological, and nuclear safety programs; federal, state, and local licensing.
1996 -1997 Framatome Technologies, Inc., Lynchburg, Va. -
Director, Corporate Procurement. Responsible for all procurement in support of FTG. This included items from engineered products to MRO to j
services.
1989-1996 Framatome Technologies, Inc., Lynchburg, Va. -
Product Line Manager, Valve Services. Profit and loss responsibility for product line that ranged from seven to fifteen million dollars revenue per year.
Product Line provided valve engineering and diagnostic services to the commercial nuclear industry. Responsibilities included sales, marketing, engineering, field services, and product development.
1982-1989 Framatome Technologies, Inc., Lynchburg, Va. -
Unit Manager, Materials and Structural Analysis.
Managed engineering unit responsible for ASME stress, seicmic, flow-induced vibration, fracture mechanics, chemical and materials analyses associated with major NSS components and systems. Analyses were in support of plant Page: 11-7 September 10,1998 Revision: 11
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l licensing issues and field repairs / modifications including leak-before-break; pressure-temperature limits; flow evaluations; life extension; SG plugging limits, BJ pump failure analysis; steam generator auxiliary feedwater header repair and modification; i
steam generator plug, sleeve and stabilizer design analyses, surge line stratification; and internals bolt replacement program. Analyses were performed in accordance with ASME Section XI, USAS B31.7 and USAS B31.1.
1979 -1982 Framatome Technologies, Inc., Lynchburg, Va. -
)
i Supervisor, Component Structural Unit. Technical Supervisor for stress analysis of steam generators, piping and reactor vessels. Successfully supported operating plant field fixes - make-up nozzle and auxiliary feedwater nozzle thermal sleeve failures.
This included both stress analysis of field repairs / modifications and licensing support for restart.
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FRAMATOME COGEMA FUELS - LYNCH:URG MANUFACTURING FACILITY USNRC LICENSE SNM-1168, DOCKET 70-1201 PART11 - CHAPTER 11 -' ORGANIZATION,AND PERSONNEL NAME:
T. Scott Wilkerson TITLE:
Manager, Assembly CITIZEN OF UNITED STATES EDUCATION: University of Virginia, Charlottesville, Va.
M.E. in Applied Mechanics - 1986 Virginia Polytechnic Institute and State University, Blacksburg, Va.
B.S. in Mechanical Engineering - 1978 EXPERIENCE:
1998 - Present Framatome Cogema Fuels, Lynchburg Manufacturing Facility, Lynchburg, Va. - Manager Assembly. Responsible for assembly product cente'r including fuel assembly, contro component assembly, and in-core detector fabrication and production maintenance. Responsibilities include manufacturing, quality, safety, training, equipment maintenance, process engineering, in-process inspections, planning and scheduling.
1997 -1998 Framatome Cogema Fuels, Lynchburg Manufacturing Facility, Lyrichburg, Va. - Manager, Fuel Manufacturing. Unit manager with overall responsible for fuel rod, control component, grid, fuel bundle, and incore detector fabrication.
1992-1997 Framatome Cogema Fuels, Lynchburg Manufacturing Facility, Lynchburg, Va. - Manager, Manufacturing Engineering. Unit manager with responsibility for manufacturing procedures and processes used for production of nuclear fuel components. Unit provides technical responsibility for equipment design, fabrication methods, process qualifications and evaluation of deviated components. Function includes liaison with vendors, fuel design engineering and customers.
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FRAMATOME COGEMA FUELS - LYNCHBURG MANUFACTURING FACILITY USHRC LICENSE SNM-1168, DOCKET 70-1201 pal lT11 -- CHAPTER 11 - ORGANIZATION AND PERSONNEL 1'990 -1992 B&W Fuel Company, Commercial Nuclear Fuel Plant, Lynchburg, Va. - Group Supervisor in charge of the Present Production Engineering Group of the Fuel Mechanical Engineering Unit. As such, responsible for the interface between the design of fuel assemblies and control components and the fabrication facility. All drawings, specifications, and associated documentation for component fabrication are released, maintained, and administered by this group.
1980 -1990 B&W Fuel Company, Lynchburg, Va. - Principal Engineer, entered as an Associate Engineer (Engineer 11). Work consists primarily of the mechanical (structural) design testing, analysis, and associated documentation of commercial nuclear fuel assembly structural components and related equipment. This work typically involves the use of open shop computer programming as well as large scale structural analysis computer codes such as j
l-ANSYS. Project administrative tasks such as l
preparing proposals, schedules, presentations, cost estimates, and technical reports are routinely done l
as necessary. Often this work is in close cooperation with the manufacturing plant; J
therefore, familiarity'with standard manufacturing i
practices is also maintained.
l 1978 -1980 Lynchburg Foundry, Lynchburg, Va. - Associate l
l Engineer. In this capacity responsible for the budget, schedule, design, installation, and startup of capital projects and for solving a variety of i
mechanical plant engineering problems. Most of
[
the work was in direct support of the manufacturing and quality control process of the plant. Lynchburg Foundry makes gray and ductile iron castings primarily for the automotive and heavy equipment industries.
PROFESSIONAL AFFILIATION:
ASME i
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FRAMA TOME COGEMA FUELS - LYNCHBURG MANUFACTURING FACILITY USNRC LICENSE SNM-1168, DOCKET 70-1201 PART11 - CHAPTER 11 - ORGANIZA TION AND PERSONNEL NAME:
Timothy V. Allsep TITLE:
Manager, Radiation Protection CITIZEN OF UNITED STATES EDUCATION: Bachelor of Science, Applied Science and Technology (Specialization in Radiation Protection); Thomas Edison State College - 1998 EXPERIENCE: 1998 - Present Framatome Cogema Fuels, Lynchburg Manufacturing Facility, Lynchburg, Va. - Manager, Radiation Protection. Responsible for the maintenance, development and implementation of the FCF radiation protection program which controls, monitors and ensures the radiological safety of the plant, personnel, public and the environment, in accord with the ALARA philosophy and applicable regulations.
1996 -1998 Framatome Cogema Fuels, Lynchburg Manufacturing Facility, Lynchburg, Va. - Senior Health Physicist. Responsible for the development andimplementation of radiologicalprogram controls as wellas overseeing the daily conduct of radiologically controlled activities.
1992 - 1996 B&WNuclear Technologies, Lynchburg, Va. -
Senior Field Service Engineer, ALARA and Training Department. Responsible for allaspects of the ALARA controls for routine and specialized reactor maintenance, including proposal development, exposure estimates, engineered controls, process development, process training and on-site controls.
The acquisition of specialized knowledge through this process often resulted in technical expert or lead engineer roles.
1989 - 1992 B& W Nuclear Technologies, Lynchburg, Va. -
Supervisor, Radiological and ALARA Engineering.
Responsible for overseeing the application of the Page: 11-11 September 10,1998 Revision: 11 l
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FRAMA TOME COGEMA FUELS - LYNCHBURG MANUFACTURING FACILITY USNRC LICENSE SNM-1168, DOCKET 70-1201 PART11 - CHAPTER 11 - ORGANIZA TION AN,D PERSONNEL l
l ALARA Philosophy to all aspects or routine and specialized outage maintenance.
1985 - 1989 South Carolina Electric and Gas, V.C. Summer Nuclear Station, Jenkinsville, S.C., Senior Health Physics Specialist. Servedin the following capacities:
i Health Physics Instructor Corporate Health Physics - Radioanalytical Services Lead Health Physics Specialist Health Physics Reactor Building Outage Coordinator Health Physics Auxiliary Building Coordinator Health Physics Steam Generator Coordinator Health Physics Shift Leader 1983 - 1985 ANSI 3.1 Senior Health Physics Techni:ian at various sites for RAD Services, Numanco, and the Institute for Resource Management. Provided Health Physics coverage andleadership during l
l outage and non-outage periods.
1979 - 1983 Duke Power Company, Oconee Nuclear Station, Seneca, S.C. ANSI 3.1 Senior Health Physics Technician. After completion of the DPCo Systems-Health Physics training regimen, assumed role as a Health Physics Countroom Technician, performing gamma spectroscopy, liquid scintillation and other sample analysis procedures. Qualified as a Shift Health Physics Technician with responsibilities ranging from effluent controls to radiological emergency response.
SUMMARY
Nearly nineteen years of nuclear power related experience focusing upon Health Physics and the ALARA philosophy. Areas of l
Responsibility have included, but are not limited to:
l Radiological Surveillance and Control l
Radioactive Material Transport Page: 11-12 September 10,1998 Revision: 11 l
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FRAMA TOME COGEMA FUELS - LYNCHBURG MANUFACTURING FACILITY USNRC LICENSE SNM-1168, DOCKET 70-1201 PAR 7;H - CHAPTER 11 - ORGANIZATION AND PERSONNEL Respiratory Protection Personnel and Environmental TLD Programs l
Procedure and Work Permit Development l
^ Instrumentation Decommissioning and Decontamination
?
Criticality Controls Training Development and Implementation Program Compliance Emergency Response Environmental Sampling Member of the Health Physics Society, American Nuclear Society and National Registry or Radiation Protection Technologies.
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FRAMA TOME COGEMA FUELS - L YNCHBURG MANUFACTURING FACILITY USNRC LICENSE SNM-1168, DOCKET 70-1201 l
PARTII - CHAPTER 11 - ORGANIZA TION AND PERSONNEL I
l i
i NAME:
Charles A. Armontrout TITLE:
Manager, Licensing & Quality Systems CITIZEN OF UNITED STATES EDUCATION: BS - Physics, Mathematics - Lynchburg College - 1963 MS - Engineering Administration - George Washington University -
1980 EXPERIENCE: 1998 - Present Framatome Cogema Fuels, Lynchburg Manufacturing Facility, Lynchburg, Va. - Manager, Licensing and Quality Systems. Responsible for the establishment, implementation and maintenance of Licensing and Quality Systems for assurance of compliance with applicable licensing, quality, regulatory, occupational, and industrial safety requirements including systems for verification of compliance.
1993 - 1998 Framatome Cogema Fuels, Lynchburg Manufacturing Facility, Lynchburg, Va. - Lead, Quality Audits & Programs. Responsible for the establishment, implementation, and maintenance of quality systems for assurance of compliance with applicable quality and regulatory requirements including systems for verification of compliance.
Responsibilities also included the performance of self-assessments andinternalaudits to verify compliance with licensing and regulatory requirements for the control of special nuclear material and radioactive material shipping containers.
1989 - 1993 B&W Fuel Company, Lynchburg, Va. - Manager Project Quality Assurance. Responsible for the supervision and direction of QA activities on B&W l
Fuel Company projects, both contract and R&D, to l
assure / verify compliance with the requirements of applicable codes, standards, regulations, contracts, i
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FRAMA TOME COGEMA FUELS - LYNCHBURG MANUFACTURING FACILITY USNRC LICENSE SNM-1168, DOCXET 70-1201 PARTll - CHAPTER 11 - ORGANIZATION AND PERSONNEL and BWFC QA programs. Projects included High Level Waste & Spent Fuel.
1980 - 1989 B& W Nuclear Power Division, Lynchburg, Va. -
Manager, QA Audits and Programs. Directed QA audit and QA program development activities for the Nuclear Power Division. Responsibilities includedimplementation of contractual QA requirements for nuclear components and systems, f
development of specific QA programs and procedures for a wide variety of nuclearpower programs and the planning and execution of internal and supplier audits.
1973 - 1989 B& W Nuclear Power Division - Lynchburg, Va. -
Served as QA Audit Group Leader, QA Engineering Group Leader and Senior QA Engineer.
1963 - 1973 Bell Aerospace, Buffalo, NY, and Rocketdyne, Canoga Park, CA - Served as Engineer and Senior Engineer for the design, test, and fabrication of Propulsion System Components for the Gemini, Apollo, and Minuteman Projects.
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FRAMATOME COGEMA FUELSN NCHBURG MANUFACTURING FACILITY LY USNRC LICENSE SNM-1168, DOCKET 70-1201 l
PARTIf - CHAPTER 11 - ORGANIZA TION AND PERSONNEL NAME:
D. Idlichael Minor i
TITLE:
Manager, Rods CITIZEN OF UNITED STATES EDUCATION: BS Mechanical Engineering, NC State University - 1974 EXPERIENCE: 1998 - Present Framatome Cogema Fuels, Lynchburg Manufacturing Facility, Lynchburg, Va. - Manager, Rods. Responsible for fuel rod, burnable poison rod, and guide tube fabrication. Responsibilities include financial and schedule performance of the center, product quality, safety, training, equipment maintenance, process engineering, in-process inspections, planning and scheduling.
1997 - 1998 Framatome Cogema Fuels, Lynchburg Manufacturing Facility, Lynchburg, Va. - Manager, Process Engineering. Responsible for leading the activities of a team of fourteen engineers and technicians. The primary responsibility of the team was to develop, qualify and implement processes and procedures for the manufacture of the plant's fabricated products in a strict quality environment.
The team also provided engineering support to the manufacturing operations, proposes and implemented process improvements to reduce costs and improve quality, executed manufacturing capital and R&D projects, and dispositioned deviated product.
l Led the team in establishing goals, strategies, and priorities. Oversaw the review of design documents to assure compatibility with manufacturing capabilities. Interfaced with other organizations to facilitate completion of projects.
Interfaced with customers and management on technicalissues. Recommended capital and R&D i
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PARTH - CHAPTER 11 - ORGANIZATION AND PERSONNEL projects. Coached the team in the implementation of team processes and evaluated team member performance. Developed and maintained cost center budget.
1994 - 1997 Framatome Cogema Fuels, Lynchburg Manufacturing Facility, Lynchburg, Va. -
Supervisor, Product Engineering Group.
Responsible for leading the activities of six engineers in a team environment. The team was responsible for the development of manufacturing processes for all of the plant's fabricated products.
Other responsibilities of the team included qualification of processes, preparing route cards and manufacturing procedures, and dispositioning deviated product.
1992 - 1994 B&WFuel Company, Lynchburg, Va. - Supervisor, Field Service. Supervised from four to five Project Engineers and technicians. The group was responsible for the design, maintenance, and upgrades of equipment used for remote inspection and repair of nuclear fuel assemblies. In addition, the group was responsible for project engineering in support of field campaigns, including preparation of cost estimates and tooling lists, customer interface, site procedures, and site interface requirements.
1988 - 1992 B&WFuel Company, Lynchburg, Va. - Engineer IV Field Service. Designed remotely-operated equipment for the repair and inspection of nuclear fuel assemblies. Responsibilities included proposal preparation, mechanical design, analysis, fabrication, testing, and project management.
The design work involved mechanical components, high-precision dimensional measurement techniques, eddy current methods, and electronic controls, in addition, led several site campaigns Page: 11-17 September 10,1998 Revision: 11
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FRAMA TOME COGEMA FUELS - LYNCHBURG MANUFACTURING FACILITY i
USNRC LICENSE SNM-1168, DOCKET 70-1201 i
PARTH - CHAPTER 11 - ORGANIZA TION ANDPERSONNEL i
involving fuel reconstitution, recaging, and inspections.
1983 - 1988 Tweco Products, Inc., Wichita, KS - Project Engineer. Responsible for the design of special-purpose production machinery, automated assembly machines, controls, and fixtures to support manufacturing processes and cost reduction projects. Work involved mechanical design, hydraulics, pneumatics, programmable controllers, sensors, and stress analysis.
1982 - 1983 Cessna Aircraft, Wichita, KS - Design Engineer.
Responsible for the design of airframe components for twin-engine commercial jet airplanes.
1981 - 1982 Newport News Industrial Corp., Newport News, Va. - Equipment Engineer. Work involved the design and analysis of components for low-level radioactive waste encapsulation systems.
Responsibilities included stress analysis of piping and pressure vessels, design of remotely-operated processing and packaging equipment, fluid flow-analysis, and preparation of specifications.
1978 - 1981 Babcock & Wilcox Company, Lynchburg, VA. -
Research Engineer. Primary responsibilities were design, development, and analysis of remotely-operated equipment. Work involved hot cell equipment, fuel assembly repair and inspection systems, and special equipment to support various i
R&D projects. Was involved in remote repair and inspection of nuclear fuel assemblies during several site campaigns.
1974 - 1978 Union Carbide Corporation, Oak Ridge, TN. -
Design' Engineer. Work involved the design of uranium enrichment equipment. Responsibilities included stress analysis of large pressure vessels, design of assembly and welding fixtures for large l
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l FRAMATOME COGEMA FUELS - LYNCH URG MANUFACTURING FACILITY USNRC LICENSE SNM-1168, DOCKETc70-1201 t'
- PART11
, CHAPTER 11 - ^ ORGANIZATION AND PERSONNEL components, and specifications for components and hardware, j
PROFESSIONAL AFFILIATIONS:
MemberASME Registered ProfessionalEngineer 1
- PATENTS:
- 5,570,400: "On Line Sipping Air Delivery System"
- 5,461,547: " Drive Tool for Upper End Fitting Locking Arrangement"
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l FRAMA TOME COGEMA FUELS - L YNCH:URG MANUFACTURING FACILITY USNRC LICENSE SNM-1168, DOCKET 70-1201 PARTll - CHAPTER 11 - ORGANIZATION AND PERSONNEL i.
1 NAME Christopher A. Conrad T/TLE:
Manager, Components CITIZEN OF UNITED STATES EDUCAT/0N: Masters of Engineering Administration, George Washington University, Washington, DC. - 1986 BS Metallurgical Engineering, University of Cincinnati, Cincinnati, OH. - 1982 EXPERIENCE:
1998 - Present Framatome Cogema Fuels, Lynchburg Manufacturing Facility, Lynchburg, VA. - Manager l
Components. Responsible for component fabrication including machine shop and grid fabrication. Responsibilities include financial and schedule performance of the center, product quality, safety, training, equipment maintenance, process engineering, in-process inspections, planning and scheduling.
1997 - 1998 Framatome Cogema Fuels, Lynchburg Manufacturing Facility, Lynchburg, VA. - Manager, Fuel Rod & Control Rod Manufacturing.
Responsible for all direct labor, planning and administration activities associated with the manufacture of nuclear fuel and control rod assemblies. (Approximate 18 l
operators / technicians) l 1993 - 1997 Framatome Cogema Fuels, Lynchburg Manufacturing Facility, Lynchburg, VA. -
Supervisor, Manufacturing Engineering.
Responsible for all process and procurement engineering functions, document control, research and development administration. (Approximate 8 l.
engineers / technicians).
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FRAMATOME COGEMA FUELS IL NCHBURG MANtJFACTURING FACILITY USNRC LICENSE SNM-1168, DOCKET 70-1201 o
PARTll - CHAPTER 11 - ORGANIZATION AND PERSONNEL 1992 - 1993 Babcock & Wilcox, Naval Nuclear Fuel Division, l
Lynchburg, VA. - Senior Engineer. Responsible for l
providing technical support in the manufacture of nuclear core components for naval and research reactor applications. Directly responsible for all heat treating, cleaning, pickling, and corrosion testing processes. Involvement with low and high enriched uranium fuel processing. Familiarity with processing of zirconium and aluminum metal alloys and powders. Exposure to all aspects of nuclear core manufacturing including machining, welding, cleaning, assembly, inspection, material testing, and regulatory compliance. Interface with customer, design agency, and regulatory auditors.
1991 - 1992 Babcock & Wilcox, Naval Nuclear Fuel Division, Lynchburg, VA. - Manager, Element Fabrication.
Supervised manufacturing areas involving a variety of basic metal fabrication operations (including heat treating, pickling, cleaning, rolling, straightening, punching, and upset resistance welding). Oversaw downsizing transition resulting in eventual lay-off of a third of area workforce.
1987 - 1991 Babcock & Wilcox, Naval Nuclear Fuel Division, Lynchburg, VA. - Manager, Manufacturing Engineering. Managed unit responsible for manufacturing processes and equipment j
throughout naval reactors facility. Primarily involving machining, GTA & EB welding, assembly, roll forging, heat treating, and chemical processes.
Responsible for group planning and budgeting, assignment of tasks, department R&D, work proposals, and customer interface. Established statistical process control methods in number of core manufacturing applications.
1982 - 1987 Babcock & Wilcox, Naval Nuclear Fuel Division, Lynchburg, VA. - Manufacturing Engineer.
l Cognizance for manufacturing processes involving l
rolling, heat treating, assembly, cleaning, and l
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FRAMA TOME CO:EMA FUELS - LYNCHBURG MANUFACTURING FACILITY USNRC LICENSE SNM-1168, DOCKET 20-1201 PARTll
, CHAPTER 11 - ORGANIZA TION AND PERSONNEL pickling of naval nuclear core components.
Responsible for equipment / operator qualifications, resolving quality deficiencies, compliance with contractual / regulatory requirements, process improvement, and facility upgrades.
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FRAMATOME CO EMA FUELS - LYNCHBURG MANUFACTURING FACILITY
. USNRC LICENSE SNM-1168, DOCKET 70-1201 PARTll - CHAPTER 11 - ORGANIZATION AND PERSONNEL NAME:
L. A. Hassler TITLE:
Senior Principal Engineer i
CITIZEN OF THE UNITED STATES i
EDUCAT/0N:
' University of Va., Charlottesville, Va.
Ph.D. in Nuclear Engineering - 1973 St. Louis University, St. Louis, Mo.
B.S. in Physics - 1965 i
EXPERIENCE: Framatome Cogema Fuels, Lynchburg, Va. - Senior Principal Engineer in the Fuel Engineering Section with 24 years experience in the areas of radiation transport, shielding, and criticality safety. In the area of criticality safety he has been involved in the following tasks:
Attending ORNL KENOlV training course and KENOVa training i
j course.
Analysis of the disrupted core at TMI-Il and analytical design of the TMI-Il defueling canisters using KENOlV program.
On loan for three months to the Navel Nuclear Fuel Division (NNFD) criticality safety group. Supported criticality safety analysis for storage at NNFD.
Provided criticality design analysis for the BR-100 spent fuel shipping container using KENOlV program.
Provided criticality analysis for the NNFD 5X22 shipping container SARP license submittal.
Provided criticality analysis for the new fuel storage racks for Wisconsin Public Services Corporation and for the spent fuel storage racks for Toledo Edison Corporation.
Provided QA review for the licensing criticality calculations for two types of BWFC new fuel shipping containers.
i Page: 11-23 September 10,1998 Revision: 11 1
FRAMA TOME COGEMA FUELS - LYNCH:URG MANUFACTURING FAClllTY
?
USNRC LICENSE SNM-1168, DOCKET 70-1201 PARTH - CHAPTER 11 - ORGANIZATION AND PERSONNEL NAME:
P. L. Holman TITLE:
Senior Principal Engineer CITIZEN OF THE UNITED STATES EDUCATION: University of Va., Charlottesville, Va.
M.S. in Nuclear Engineering - 1978 B.S. in Nuclear Engineering - 1975 EXPERIENCE: Framatome Cogemt. Fuels, Lynchburg, Va. - Senior Princip:d Engineer with 20 years experience in reactor physics, including the following areas:
Core Follow analysis, Standard Models Development, Fuel Cycle Division for Oconee Units I,11, Ill, ANO-1 Unit 1, and Crystal River, Licensing, Reload Licensing Analysis Task Engineer, Maneuvering analysis, and Criticality calculations using Monte Carlo methodology.
Recent criticality work involved licensing the ANFIBWFC Model 51032-2 new fuel shipping container and also included separate criticality analyses for new and spent fuel storage racks for NUSCo and the Wisconsin Public Services Corporation. Preceding that work was the criticality analysis for the Toledo Edison new and i
spent fuel storage racks which involved an enrichment limit increase j
utilizing burnup credit. Additional criticality work involved the analysis for the GPU TMI-2 defueling canisters and the DOE 100 ton rail-barge shipping cask.
TECHNICAL PAPERS AND PUBLICATIONS:
"TMI-2 Defueling Canisters Reactivity Analysis", ANS Transactions, 1986.
"Three-Dimensional Analyses of a Highly Heterogeneous PWR Using the NOODLE Code", ANS Transactions 1988.
PROFESSIONAL AFFILIATIONS:
American Nuclear Society, Virginia Chapter ANS l
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USNRC LICENSE SNM-1168, DOCiC~T 421U
- PART11 - CHAPTER 11 - ORGANIZATI@ MD PERSONNEL i
NAME:
Jeffrey S. Whitt TITLE:
Program Manager, Compliance & Skills Training I
CITIZEN OF THE UNITED STATES:
' EDUCATION:
Attending Liberty University towards BS degree in Business Administration Certificate in Machine Operations, Central Virginia Community j
College, Lynchburg - 1982 (Summa Cum Laude)
EXPERIENCE:
Current Framatome Cogema Fuels, Lynchburg Manufacturing Facility, Lynchburg, VA - Manager, Compliance & Skills Training (new assignment, Oct 1998) Responsible for the development and implementation of compliance and skills training programs, including the identification of needed training and training personnel, as well as the establishment of a centralized process for documentation and monitoring.
1997-1998 Framatome Cogema Fuels, Lynchburg Manufacturing Facility, Lynchburg, VA - Manager, Facilities. Responsible for the -
engineering,. implementation, and maintenance of plant i
systems, equipment, and tooling.
Supervise six direct reports. Responsible for a'ctivities in: Tooling & Equipment Design, Mechanical Maintenance, Electrical Maintenance, Facilities Services, and Security. Overall team size at ~32 j
employees. Report to the VP, Operations.
1995-1997 Framatome Cogema Fuels, Lynchburg Manufacturing.
Facility, Lynchburg, VA - Supervisor, Tooling & Equipment Engineering Group.
Responsible for the development, design, and implementation of manufacturing systems, tooling,. and equipment for.the manufacturing facility.
Provided engineering design support for Field Services projects.
1991-1995 B&W Fuel Company, Commercial Nuclear Fuel Plant, l
l-Page: 11-25 September 10,1998 Revision: 11
FRAMA TOME COGEMA FUELS - L YNCH8URG MANUFACTURING FACILITY USNRC LICENSE SNM-1168, DOCKET 70-1201 PARTll - CHAPTER 11 - ORGANIZA TION AND PERSONNEL L7nchburg, VA - Lead Designer, Tooling & Equipment Group.
Lead designer for manufacturing tooling and equipment development, design, and implementation.
1989-1991 B&W Fuel Company, Commercial Nuclear Fuel Plant, Lynchburg, VA - Project Engineer, Specialty Manufacturing.
Responsible for project engineering, process and weld development, tooling design, and overall coordination of specialty manufacturing projects.
1985-1989 B&W Fuel Company, Commercial Nuclear Fuel Plant, Lynchburg, VA - Technical Specialist /. Designer, Manufacturing Engineering.
Responsible for process engineering and tooling design in support of manufacturing operations. Also provided design support for Field Service activities.
1983-1985 Babcock & Wilcox, Commercial Nuclear Fuel Plant, Lynchburg, VA - Machinist, Manufacturing. Responsible CNC machine tool programming and component machining.
1979-1982 General Electric Co., Mobile Communications Division, Lynchburg, VA, - Mechanical Skills Apprentice and Production Assembler. Graduate of GE's three year Mechanical Skills training program. Certifications: Virginia State Certificate and Machinist Joumeyman Card, Certificate of Apprenticeship, GE Co.
i Summary:
Five + years experience in direct leadership of supervisors, engineers, designers, and other professionals for implementation of engineering projects for manufacturing and. service organizations.
Ten + years experience in the management and implementation of manufacturing and inspection related tooling and equipment projects (capital and R&D); ranging up to $1M in work scope. Effective and innovative design skills, with a strength in guiding a design j
to its simplest form where possible; holder of five patents with two pending.
Technical writer with two articles published. Director for the FramaTONES, an employee choral group, '96 & '97. Personal computer skills include experience in CAD / CAM, spreadsheet, word processing, database, E-mail, and project management type software. Experience in the preparation and negotiation of Page: 11-26 September 10,1998 Revision: 11
FRAMA TOME COGEMA FUELS - L YNCHBURG MANUFACTURING FA CILITY USNRC LICENSE SNM-1168, DOCKET 70-1201 PARTll - CHAPTER 11 - ORGANIZA TION AND PERSONNEL commercial agreements with equipment suppliers. Various domestic and foreign travel destinations have included West Germany (equipment and process development at Mulheim-Karleich), Taiwan ROC (Maanshan ECHO field inspections), France, and the United Kingdom (equipment procurement and technology exchange)
Page: 11-27 September 10,1998 Revision: 11
t FRAMA TOME COGEMA FUELS - L YNCHBURG MANUFACTURING FACillTY USNRC LICENSE SNM-1168, DOCKET 7.0-1201 PARTH - C,HAPTER 11 - ORGANIZA TION AND PERSONNEL 11.4 Ooeratina 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. Health Safety activities are controlled by detailed operating procedures developed by Health-Safety to assure standardization and accuracy.
All written procedures are reviewed and approved by appropriate representatives of plant management. If SNM or other radioactive materials are involved in an activity, approval by the Manager, Radiation Protection or his designee shall be required prior to implementation. Likewise all Health-Safety procedures are approved by the Manager, Radiation Protection as well as by affected members of plant management. Health-Safety procedures are reviewed periodically and updated accordingly.
Applicable procedures are made available in the work area and adherence to procedure is required of all personnel.
11.5 Trainina All personnel receive basic training in radiological, industrial, and nuclear safety upon being hired. This initial training is a cooperative effort involving Human Resources, Program Manager, Compliance &
Skills Training and the employee's supervisor and is designed to satisfy the requirements of 10 CFR 19.12.
Particular emphasis is placed on the nature of the materials handled, ALARA plant eefety program and rules,10 CFR 19 requirements, and the emergency evacuation system. Specific areas covered in the safety training program are as follows:
11.5.1 Initial Emolovee Trainina Employees are referred to the Program Manager, Compliance
& Skills Training by the Human Resources Department for initial training in safety. The entire plant safety program is reviewed in some detail with particular emphasis being placed on specific areas depending on the employee's job assignment. A brief discussion of, and familiarization with, the general principles of health physics and nuclear safety is Page: 11-28 September 10,1998 Revision: 11
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FRAMATOMEBOGERTATOELS - L YNCHBURG MANUFA CTURING FA CILITY USNRC LICENSE SNM-1168, DOCKET 70-1201
^*
PART11 - CHAPTER 11 - ORGANIZATION AND PERSONNEL i
included. The employee is informed of his rights and responsibilities under CFR 19, and OSHA.
Following the initial indoctrination depending on the employees work location, they shall receive additional safety training from their immediate supervisor regarding the nuclear and radiological safety requirements of their specific job assignment. Training sessions are documented and filed as part of the employee's training record.
11.5.2 Emolovee Retrainino Continuing training of a general nature is provided as necessary by Health-Safety, Program Manager, Compliance &
Skills Training and supervision. This training may be formalized (i.e., " classes") or informal and conducted as part of routine Health-Safety audits. Formalized retraining may be utilized to explain operational changes affecting safety, control of special problems such as increased airborne activity, or changes in license specifications. The responsibility for determining the necessity for retraining or special training rests with Health-Safety and Program Manager, Compliance & Skills Training based on plant conditions or the request of supervision.
Radiation workers are all retrained annually as a routine part of the safety training program. The retraining sessions are documented and kept as part of the employee's training record.
If workers have had similar radiation worker training, the initial training or retraining may be by-passed by successful completion of a written exam with a score of 75%.
11.5.3 Soecialized Trainino 11.5.3.1 Respiratorv Protection Training and retraining in the use of respiratory protection devices is provided by Health-Safety or Program Manager, Compliance & Skills Training as Page: 11-29 September 10,1998 Revision: 11 L
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~~
^
USNRC LICENSE SNM-1168, DOCKET 70-1201 PARTH - CHAPTER 11 - ORGANIZATION AND PERSONNEL required. Points relating to proper use are covered as the unit is issued and fitted by Health-Safety.
This approach provides continuing review of respiratory protection requirements. Should situations arise where frequent use of a respirator is necessary, frequent Health-Safety surveillance will assure continued proper application.
11.5.3.2 Emeraency Teams Trainina Specialized training for special and emergency response units such as the Fire Brigade, Radiation Monitoring Team and First Aid personnelis coordinated by Health-Safety. Fire Brigade training is conducted by representatives of Health-Safety or Program Manager, Compliance & Skills Training and/or local Fire Departments and covers the use of fire fighting equipment and agents available at LMF. Radiation Monitoring Team members receive periodic training from Health-Safety in emergency response techniques, instrument use and maintenance, health physics and nuclear safety fundamentals, respiratory protection and contamination control. Annual evacuation drills are generally utilized as a training period for the emergency teams. First aid training is given by a qualified instructor and is the standard Red Cross program or equivalent.
11.6 Chanaes in Procedures. Facilities, and Eauioment 11.6.1 Procedural Chanaes Procedural changes are initiated by the functional component responsible for that activity. Such procedural changes are reviewed and approved by plant management prior to implementation. If the activity involves SNM or other radioactive materials, the Manager, Radiation Protection must approve the procedural change prior to implementation.
11.6.2 Facilities and Eautoment Chanaes Page: 11-30 September 10,1998 Revision: 11
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Changes or modifications to facilities and equipment that have a potential impact on nuclear, radiological, industrial, or chemical safety must be reviewed and approved by the Safety Review Board and/or the Safety Review Board Chairman or qualified designee prior to initiation. The Safety Review Board is described in detail in Chapter 2.0 of Part 1.
11.6.2.1 Initiatina Chanaes The responsibility for initiating changes as described in 11.6.2 is usually given to the immediate operational supervisor or manager. The requested change is documented and submitted to the Safety Review Board Chairman for initial review.
11.6.2.2 Analysis of Chanaes The Safety Review Board Chairman determines what safety evaluations are needed. If the proposed modification changes the basis on which the nuclear criticality safety was originally assessed, a technical evaluation by the nuclear criticality safety group will be initiated. The organizational structure and minimum qualifications of the nuclear criticality safety group is as described in Chapter 4.0.
Radiation safety evaluations will be performed for new or revised operations to assure personnel protection is maintained. Chemical and industrial safety aspects of proposed modifications will also be evaluated for acceptability. These evaluations 4
are documented and retained as described in 11.6.2.6.
l 11.6.2.3 Manaaement Review
]
1 s
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(A&Mf2v0 MEN 0GEMA~ FUEL.$ LYNCilBURG MANOFACTOIliNG FACIOTY USNRC LICENSE SNM-1168, DOCKET 70-1201 PARTll - CHAPTER 11 - ORGANIZATION AND PERSONNEL As a minimum, the Safety Review Board Chairman or his qualified designee shall review all safety analyses performed for the plant modifications prior to implementation. The Safety Review Board Chairman will determine if Safety Review Board approval is required.
11.6.2.4 Acoroval and Verificat on of Chanaes i
Approval and release of plant modifications for routine use is dependent upon satisfactory completion of a pre-operational evaluation. This svaluation is a final verification that the proposed change has been installed consistent with the analyses performed under 11.6.2.2. This evaluation will consider nuclear, radiological industrial, and chemical safety as well as license compliance. This evaluation is performed by Health-Safety personnel and Licensing personnel and is approved by the Safety Review Board Chairman prior to implementation. Compliance of plant modifications is assued by our existing Health-Safety controls and audit programs with regard to contamination control, personnel exposures, nuclear safety, chemical and industrial hazards.
11.6.2.5 Records All analyses, evaluations, pre-operational evaluations and other pertinent documentation relating to plant modifications will be maintained on file for at least six months after termination of the operation.
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l FRAMAT ME COGEMA FUELS - LYNCHBURG MANUFACTURING FACl0TY USNilC LICENSE SNM-1168, DOCKET 70-1201 PARTll - CHAPTER 14 - NUCLEAR CRITICAllTY SAFETY r
14.1 Administrative and Technical Procedures The ultimate responsibility for nuclear criticality safety at the LMF rests with the LMF Plant Manager. The Plant Manager has assigned this responsibility to the Safety Review Board Chairman (Manager, Radiation Protection) who must approve all modifications or additions which involve hazardous as well as nuclear materials. The Safety Rewew Board Chairman will determine if Board review is necessary. Written procedures approved by plant management shall be utilized for all operations involving SNM.
Nuclear safety postings approved by Health-Safety shall be maintained j
l specifying nuclear safety parameters that are subject to procedural controls.
l The training program conducted at the LMF (Sections 2.5 and 11.5) provide additional assurance that the criticality safety requirements are adhered to.
Administrative controls for performing and approving criticality safety analyses are described in Section 4.1.
14.2 Preferred Acoroach to De,sian The double contingency principle as defined in the American National Standard ANSI /ANS-8.1 shall be followed in establishing nuclear criticality safety for all equipment, systems, and operations. 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 abnormalloads, accidents, and deteriorations.
14.3 Basic Assumotions The basic nuclear criticality safety limitations discussed in Chapters 4,14, and 15 were developed assuming UO with a maximum U-235 enrichment 2
of 5.1 weight percent and a maximum pellet diameter of 0.4 inches.
Uranium pellets processed at the LMF can however range in diameter from 0.3 to 0.6 inches, and may include UO powder and pellet chips. All data 2
given is valid for the most reactive heterogeneous geometry appropriate to the situation being considered. The U-235 enrichment may not exceed 5.10 wt. % with measurement uncertainties.
l 14.3.1 Enrichment Page: 14-1 September 16,1998 Revision: 5
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FRAMA TOME COGEMA FULLS - L YNCHBURG MANUFACTURING FACILITY USNRC LICENSE SNM-1168, DOCKET 70-1201 o
PARTll - CHAPTER 14 - NUCLEAR CRITICALITY SAFETY b
Evaluations and limits discussed in this chapter and in Chapters 4 and 15 were based on a maximum enrichment of 5.1 wt. % U-235.
With product and statistical variations, the enrichment cannot exceed 5.10 wt. % U-235.
Figures 14.1 through 14.4 illustrate basic critical parameters (mass, volume, cylinder diameter and slab thickness) versus enrichment for optimum moderated and reflected UO, rods in water as taken from 2
DP-1014. These graphs demonstrate the maximum sizes for individual units and only a small extrapolation is needed to obtain values up to 5.1 wt% U-235.
14.3.2 Claddina Material Where appropriate, zircaloy type cladding material was input to the nuclear safety calculations. Stainless steel cladding provides an additional conservatism due to its higher effective cross section for neutron capture. Other clad materials may be used provided there is not an adverse affect on nuclear safety specifications.
14.3.3 Calculated Safe Unit A calculated safe unit is defined in terms of nuclear safety as having.
a K,,,10.87 for normal operating conditions and a K,,10.95 under assumed accident conditions. The statistical and method-ological limits of error will be considered when determining K,,...
14.3.4 Calculated Safe Arrav A calculated safe array is defined for nuclear safety purposes as having a K,,,10.87 for normal operating conditions and a K,,, <
0.95 under assumed accident conditions. The K,, values for these calculated safe arrays will also consider statistical and methodologicallimits of error.
14.3.5 Safety Factors for Simole Shaoes Page: 14-2 September 16,1998 Revision: 5
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l USNRC LICENSE SNM-1168, DOCKET 70-1201 PARTll - CHAPTER 14 - NUCLEAR CRITICAllTY SAFETY l
l For simple shap'es, safe values are determined by application of the following limits:
Mass:
45% of minimum critical reflected mass Mass:
75% of minimum critical reflected mass when double batching is not credible Volume:
75% of minimum critical reflected volume Diameter:
90% of minimum critical reflected cylinder diameter Thickness:
90% of minimum critical reflected slab thickness it is recognized that interaction between simple shapes must also be accounted for.
14.3.6 Nuclear Interaction Acceptable nuclear interaction between arrays is determined using calculations or as stipulated in Section 4.2.3.1. The limits are determined for each process as it was evaluated and its limits were set.
The interaction restriction given in 4.2.3.1 is a generally accepted practice for assuring nuclear isolation of suberitical accumulations under all conditions if a calculated value is not available.
14.4 Fixed Poisons Fixed neutron poisons are utilized as part of the nuclear criticality controls in the Pellet Storage Vault. Specifications in the Pellet Storage Vault layout are given in Section 4.2.4.2. The criticality safety analyses performed specifically on the Pellet Storage Vault are given in Section 15.2.
14.5 Structural Intearity Poliev and Review Proaram LMF plant policy regarding engineering design review for structural integrity and safety margins is indicated in Sections 14.2 and 2.3.
14.6 Analvtical Methods and Their Validation Computer codes were used to calculate K.,, of individual units and of arrays.
l Computer codes and associated cross section sets are always l
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~ USNRC VCENSE SNM-1168, DOCKET 70-1201 PARTll - CHAPTER 14 - NUCLEAR CRITICAllTY SAFETY i
b;,nchmarked to demonstrate their validity for application at the LMF The particular family of codes may change over time as improvements are l
realized in the calculational methods. The LMF's goal is to maintain its computing capability at state-of-the-art for nuclear criticality safety and to assure that all codes and cross sections are properly benchmarked and validated.
14.6.1 Physics Computer Codes All the calculations in this evaluation were done using the CSAS2 module of the SCALE-3 computer code package (reference 1). The SCALE (Standardized _Qomputer Analyses for Licensing Evaluation) l code package was developed for the USNRC by the Oak Ridge l
National Laboratory. Codes within SCALE-3 used by the CSAS2 module for nuclear criticality safety analyses are the cross section processing codes NITAWL and BONAMI, the one-dimensional trans-port code XSDRNPM (when needed for cell averaging), and the three-dimensional Monte Carlo' code KENO-IV. The 123GROUPGMTH cross section library was used for all cases. Any additional nuclear criticality safety evaluation requiring the calculation of the effective neutron multiplication (K-eff) will be made using a validated and benchmarked version of an industry accepted computer code package such as the SCALE computer code package. Safe units are those with a true k,,, below 0.87
'under normal conditions and below 0.95 under accident conditions.
The true k,,, may differ from the k,,, value from a properly modeled KENO calculation because of the calculational bias and statistical uncertainty. The calculated upper limit of true k,,is the value from j
KENO plus two standard deviations (also from KENO) plus a 0.02 bias. It is this true k,,, which is compared with the 0.87 and 0.95 i
limits to assure that a unit is safe. The 0.02 bias has been deter-mined to be the maximum non-conservative delta K,,, of our SCALE l
calculations based upon benchmark calculations of critical l
expenments.
i L
The codes and cross sections of SCALE have been well validated for l
criticality safety analyses. A limited number of computer cases
~
were provided within the SCALE manual which represent the evaluation of actual critical experiments; these not only provide an 1 -
initial basis for validity but also provide a means to verify that the codes are performing as planned on the computer system being used, in addition to these cases, a number of independent studies 4.
~
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USNRC LICENSE SNM-1168, DOCKET 70-1201 PARTll - CHAPTER 14 - NUCLEAR CRITICALITY SAFETY i
have been performed which demonstrate the reliability of the codes and cross sections in SCALE when they are used to calculate various classes of critical systems. See paragraph 4.2.3.3 for description of types and range of variables used in validating the code package. In addition to the above, FCF's nuclear criticality safety group has an on-going validation program which documents the results from current calculational methods and compares them with benchmark critical experiments. These comparisons are the basis for the 0.02 bias which we currently add to all calculations.
Our Nuclear Criticality Safety Benchmark Notebook along with numerous internal memos document these comparisons.
14.7 Soecial Controls The fuel assembly processing area (i.e., assembly room) is that plant area where the loaded fuel rods are configured into a reactor type fuel assembly.
This area is under moderation control and as a result, the following conditions apply:
No sprinkler systems are permitted in the assembly room.
1 gallon volume restriction on containers in the area to preclude any l
meaningful moderation in the event of a spill.
Baffling / shielding of water piping in the area to prevent general area moderation in case of pipe rupture.
Simultaneous application of more than one fire hose is prohibited.
The fuel assembly storage area is not a moderation controlled area but the use of more than one fire hose in the area is still prohibited.
Also, the fuel assembly dust wrappers are arranged to permit free drainage of water from within.
14.8 Data Sources The following documents will be used as sources of applicable data:
1.
ANSI /ANS-8.1-1983, Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors.
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PARTll - CHAPTER 14 - NUCLEAR CRITICALITY SAFETY i
2.
TID-7016, Rev. 2 (NUREG/CR-0095), Nuclear Safety Guide.
3.
LA-10860-MS " Critical Dimensions of Systems Containing 23sU, 23sPu, and U - 1986 Revision.
2 4.
LA-3366 (Rev.), Criticality Controlin Operations with Fissile Material.
5.
ARH 600, Criticality Handbook, Volume 1,2, and 3.
6.
DP-1014, Critical and Safe Masses and Dimensions of Lattices of U and UO Rods in Water.
2 14.9 References 1.
NUREG/CR-0200, ORNL/NUREG/CSD-2, " SCALE: #, Modular Code System For Performing Standardized Computer Analyses For Licensing Evaluations," Prepared by the Staff of the Nuclear Engineering Application Department, Union Carbide Corporation.
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USNRG LICENSE SNM-1168, DOCKET 70-1201 PARTll - CHAPTER 14 - NUCLEAR CRITICALITY SAFETY j
l i
I FIGURE 14.1 CRITICAL VALUES FOR UO2-WATER MIXTURES (FULL WATER REFLECTION ASSUMED) s.co ;
4.50 -
i 4.00 J 2n 7
3.50 -
a r*
f/Wi 3.00 -
v
\\
2 2.50 -
N 2.00 -
4,30
\\
1.50 i
i i
4 i
1.50 2.50 3.50 4.50 5.50 WElGHT PERCENT U-235 0
1.84 KG O 4.10%
l l
l l
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PARTII - CHAPTER 14 - NUCLEAR CRITICAllTY SAFETY FIGURE 14.2 CRITICAL VALUES FOR UO2-WATER MIXTURES (FUU. WATER REFLECTION ASSUMED) 40.00 39.00 -
38.00 -
37.00 -
3s.00 -
35.00 ;
24.00 -
33.00 -
m N
32.00 -
31.00 -
^ 30.00 -
29.00 -
+.-K nd 28.00 -
27.00 -
4.10 26.00 -
25.00 -
24.00 -
23.00 -
22.00 -
21.00 -
20.00 i
i i
i i
i i
i -- i i
7--
i 2.40 2.80 3.20 3.60 4.00 4.40 4.80 5.20 5.60 WOGHT PERCENT U-235 0
26.5 UTERS O 4.10,*:
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USNRC LICENSE SNM-1168, DOCKET 70-1201 PARTH - CHAPTER 14 - NIJCLEAR CRITICALITY SAFETY FIGURE 14.3 CRITICAL VALUES FOR UO2-WATER MIXTURES (FULL WATER REFLEC110N ASSUMED) 40.00 39.00 -
38.00 -
37.00 -
36.00 -
35.00 -
34.00 -
3 33.00 -
3 32.00 -
31.00 q 30.00 ]
f
-p 29.00 -
g 28.00 27.00 -
26.00 -
0 25.00 -
24.00 1 23.00 -
22.00 M 21.00 '
20.00 1.50 2.50 3.50 4.50 5.50 WDGHT PERCENT U-235 0
24.9 CMS. O 4.10%
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^ i USNRC LfCENSE SNM-1168, DOCKET 70-1201 PARTH - CHAPTER 14 - NUCLEAR CRITICAllTY SAFETY a
FIGURE 14.4 CRITICAL VALUES FOR UO2-WATER MIXTURES (FULL WATER REFLEC110N ASSUMED) 20.00 19.00 -
18.00 -
17.00 -
m 2
S.
16.00 -
g b
4
)
3 15.00 -
^[
14.00 -
s 13.00 -
12.00 -
,io 11.00 -i i
10.00 i
e i
i i
i 1.50 2.50 3.50 4.50 5.50 WElGHT PERCENT U-235 0
11.5 CMS. O 4.10%
Page: 14-10 '
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