ML20215M410
| ML20215M410 | |
| Person / Time | |
|---|---|
| Site: | 07000824 |
| Issue date: | 04/30/1987 |
| From: | BABCOCK & WILCOX CO. |
| To: | |
| Shared Package | |
| ML20215M379 | List: |
| References | |
| NUDOCS 8705130222 | |
| Download: ML20215M410 (177) | |
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TABLE OF CONTENTS b
Section Page 1.0 STANDARD CONDITIONS AND SPECIAL AUTHORIZATIONS 1-1 1.1 NAME 1-1 1.2 LOCATION 1-1 1.3 LICENSE NUMBER AND PERIOD 1-1 1.4 POSSESSION LIMITS 1-2 1.5 LOCATION OF POSSESSION AND USE.
1-3 1.6 DEFINITIONS 1-3 1.7 AUTHORIZED ACTIVITIES 1-5 i
1.8 EXEMPTIONS AND SPECIAL AUTHORIZATIONS 1-5 List of Figures Figure Page I:
1-1 SITE LOCATION WITHIN VIRGINIA 1-6 l
l 1-2 SITE, FIVE MILE RADIUS.
1-7 l
1-3 SITE, BUILDINGS 1-8 l
1 l
l l
l License No SNM 778 Docket No.70-824 Date April,1:' /
O 4
I~I Amendment No.
Revision No.
Pap Babcock &Wilcox a McDermott company
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PART I LICENSE CONDITIONS 1.0 STANDARD CONDITIONS AND SPECIAL AUTHORIZATIONS 1.1 NAME Name - McDermott International. Inc.
Babcock & Wilcox Naval Nuclear Fuel Division NNFD Research Laboratory McDermott International Inc. is incorporated under the laws of the Repubite of Panama.
Principle Office - 1010 Common Street, New Orleans, Louisiana.
1.2 LOCATION Address - Babcock & Wilcox NNFO Research Laboratory l
P. O. Box 11165 Lynchburg, Virginia 24506-1165 The NNFD Research Laboratory (site) is located in Campbell County, Virginia, near the James River, approximately four miles East of the city of Lynchburg.
Figure 1-1 shows the location of the site with l
respect to the Commonwealth of Virginia. Figure 1-2 shows the location of the site with respect to a five mile radius. Figure 1-3 l
shows the location of buildings and facility locations where licensed materialt are handled and stored.
1.3 LICENSE NUMBER AND PERIOD License Number - SNM-778 License No SNM 778 Docket No.70-824 Date April,1987 Amendment No.
O Revision No.
4 Page 1-1 Babcock &Wilcox a McDermott company
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Docket Number 70-824 Period of Time - It is requested that this license be renewed for a period of 10 years.
1.4 POSSESSION LIMITS Ma terial Physical Form Enrichment Amount
- 1. Uranium enriched Encapsulated or
> 20 %
3.5 Kg con-in U-235 trradiated tained U-235
- 2. Uranium enriched Unencapsulated
> 20 %
0.53 Kg con-in U-235 and unirradiated tained U-235
- 3. Uranium enriched Encapsulated or 5 % to <20%
1.2 Kg con-in U-235 irradiated tained U-235
- 4. Uranium enriched Unencapsula ted 5 % to <20%
0.5 Kg con-in U-235 and unirradiated tained U-235
- 5. Uranium enriched Encapsulated or
.711 % to <5%
55 Kg con-in U-235 irradiated tained U-235
/~N C')
- 6. Uranium enriched Unencapsulated
.711 % to <5%
11 Kg con-in U-235 and unirradiated tained U-235
- 7. Plutonium Unencapsula ted 0.05 Kg and unirradiated
- 8. Source Material Any 6000 Kg
- 9. Fission Products Irradiated Fuel Quanti ty
& Transuranium contained in Elements 4 f rradiated fuel as-semblies.
- 10. Fission Products Irradiated fuel 5,000,000 C1.
& Transuranium Elements License No SNM 778 Docket No.70-824 Date April,1987 Amendment No.
O Revision No.
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- 11. Any byproduct Irradiated 50,000 Cf.
material structural materials a componen ts
- 12. Byproduct Any 3,000 Ci each material with total not to at. nos. 3 exceed thru 83 1,000,000 C1.
- 13. Transuranium Any 20 milli-elements Curies each
- 14. Cf-252 Sealed Sources 4 milligrams
- 15. Am-241 Sealed Sources 30 C1
- 16. H-3 Sealed Sources 100 C1
- 17. H-3 0xide 3 Ci
- 18. H-3 Ni plated Sc 3 Ci tritide foil 1.5 LOCATION OF POSSESSION AND USE 1.5.1 Licensed material shall be possessed and used at the NNFO Research Laboratory (site).
1.5.2 Byproduct material in the form of sealed sources with activities of up to 500 mil 11 Curies may be possessed and used in locations other than the site for performing instrument calibration, elec-l tronic noise analysis, shielding studies, or similar operations.
1.5.3 A restricted zone shall be established on the area north of the Chessie System main line right-of-way with a fence line based on radiation levels not exceeding an exposure dose rate of 500 milli-rems / year.
1.6 DEFINITIONS 1.6.1 Site means NNFO Research Laboratory.
l License No SNM 778 Docket No.70-824 Date Aprfl.1987 Amendment No.
O Revision No.
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Babcock &Wilcox a McDermett company i
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1.6.2 SRC means Safety Review Committee.
1.6.3 SNM means Special Nuclear Material.
1.6.4 Licensed Material means source, byproduct, or SNM received, possessed, used or transferred under a general or specific license issued by the Nuclear Regulatory Commission.
1.6.5 Research and Development (R&D) means (1) theoretical analysis, exploration, or experinentation; or (2) the extension of investigative findings and theories of a scientific or technical nature into practical application for experimental and demon-stration purposes, including the experimental production and testing of models, devices, equipment, materials and processes. The administration of licensed material, internally or externally, to human beings is not included in this definition.
1.6.6 Safety Audit Subcommittee (SAS) means the subcommittee established under the SRC to perform audit functions.
1.6.7 Manager, Employee, Community, and Regulatory Relations (Manager, EC&RR) means the position with primary responsibility for the l
safety of operations at the site.
1.6.8 Authorized User means a person who may work with licensed material l (O unsupervised and may supervise others, not so designated, in the V
handling of licensed material.
1.6.9 Calibration means a comparison of a measurement standard of known accuracy that is traceable to the NBS with another standard or l
instrument to detect, correlate or adjust any variation in the accuracy of the item being compared, within the specified range and accuracy of the item. Calibration also includes standardization.
1.6.10 Standardization means, the act of using standards which are traceable to the NBS, a nationally accepted measurement system, l
or natural phenomena to set up an instrument.
Standardiza tion l
must be performed before and af ter use.
1.6.11 Unit means (1) a separate laboratory, room, or work area; (2) a transfer cart where SNM is separated from adjacent units by at least 8-inches edge-to-edge and 24-inches center-to-center. More than one unit may be on a cart provided the preceding edge-to-edge l
l License No SNM 778 Docket No.70-824 Date April, 1987 Amendment No.
O Revision No.
4 Page 1-4 l
Babcock & Wilcox a McDermott company
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and center-to-center values are maintained, and (3) a processing bench, glove box, furnace, fume hood, or other similar process equipment or container separated from adjacent units by at least 8-inches edge-to-edge and 24-inches center-to-center, k
1.6.12 Standing RWP's are Radiation Work Permits issued for a term not to exceed 6-months, authorizing entry into High Radiation Areas and Airborne Radioactivity Areas to perform routine work.
1.7 AUTHORIZED ACTIVITIES 1.7.1 Licensed material shall be used in the performance of Research and Development (e.g., hot cell examination of irradiated and radio-active components including irradiated fuel; analytical activities for other companies or 84W divisions including laboratory analysis, preparation of and testing of materials and equipment; pr(paration and modification of radiation sources; and preparation and decon-tamination of reactor-related hardware for inspecting, evaluating, and measuring reactor components).
1.7.2 The site may transport and possess licensed material in private l
carriage between NRC Ifcensed facilities within the United States pursuant to the regulations in 10 CFR 71 and 49 CFR.
1.8 EXEMPTIONS AND SPECIAL AUTHORIZATIONS 1.8.1 The uranium bloassay program sampling frequency shall comply with Tables 2 and 3 of Regulatory Guide 8.11 dated June,1974, except as follows:
1.8.1.1 When a worker is absent from the sita during a period when the l
bioassay counting service is on site, a special counting shall not be required for those workers for routine exposure control moni-toring. The maximum amount of time between in vivo counts shall not exceed 12-months.
License No SNM 778 Docket No.70-824 Date Apr11,1987 Amendment No.
O Revision No.
4 Page 1-5 Babcock &Wilcox A McDermott company
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Babcock &Wilcox a WDermott company
l i p TABLE OF CONTENTS v
Sectior Page 2.0 GENERAL ORGANIZATIONAL AND ADMINISTRATIVE REQUIREMENTS 2-1 2.1 POLICY 2-1 2.2 ORGANIZATION RESPONS!BILITIES AND AUTHORITIES 2-1 l
2.3 SAFETY REVIEW COMMITTEE 2-3 l
l 2.4 APPROVAL AUTHORITY FOR PERSONNEL SELECTION 2-4 2.5 PERSONNEL EDUCATION AND EXPERIENCE REQUIREMENTS 2-5 2.6 TRAINING 2-6 2.7 OPERATING PROCEDURES 2-7 1
1.
2.8 INTERNAL AUDITS AND INSPECTIONS 2-9 l
2.8.1 Nuclear Criticality Safety 2-9 AV 2.8.2 Health Physics 2-9 l
2.8.3 General Safety and Compliance 2-9 2.9 INVESTIGATIONS AND REPORTING OF 0FF-NORMAL OCCURRENCES 2-10 2.9.1 License Administrator 2-10 2.9.2 Supervisor, Health and Safety 2-10 t
2.9.3 Facility Supervisor 2-11 l
2.10 RECORDS 2-11 2.10.1 Health and Safety Group 2-11 i
2.10.2 Nuclear Criticality Safety Officer 2-12 l
2.10.3 License Administrator 2-12 l
Licenes No SNM 778 Docket No. 70 824 Date Aprfl.1987 O
4 2-1 Amendment No.
Revision No.
Pg i O Babcock &Wilcox l
a McDermott company o
TA8LE OF CONTENTS (Continued)
O 2.10.4 Emergency Records 2-12 List of Figures Figure Page 2-1 SITE ORGANIZATION 2-13 I
O Lloones No SNM 778 Docket No. 70 824 Date Apri1,1987 O
4 2-1i Amendment No.
Movision No.
Pm Babcock &Wilcox a McDermott compaiy
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t 2.0 GENERAL ORGANIZATIONAL AND ADMINISTRATIVE REQUIREMENTS l
2.1 POLICY l
It shall be the policy of the site to maintain radiation exposures to I
employees and the general public as low as is reasonably achievable.
The facility procedures to ensure the safe handling of Itcensed j
material are the Area Operating Procedures.
l 2.2 ORGANIZATION RESPONS!BILITIES AND AUTHORITIES 2.2.1 Manager, ECARR - The Manager, EC4RR is ultimately responsible for l
all safety at the site.
2.2.2 Facility Supervisor - The Facility Supervisor is responsible to the Manager, ECARR for the safe conduct of all operations at the site l
and for ensuring that all applicable operations are conducted in compliance with the license and applicable regulations.
To fulfill these responsibilities the Faciltty Super-visor shall have the authority to stop any operation that he feels is unsafe or in violation of Itcense. The Facilit Area Operating Procedures and RWP'y Supervisor shall review all new i
s, and changes there to, for O
license and regulatory compliance and for facility safety; and he i
U shall have approval authority for them.
He shall submf t Items for review to the SRC.
He shall have approval authortty for Area Super-
- visors, l
2.2.3 Area Supervisors - Area Supervisors are recommended by their l
division management and their appointment shall be jointly approved t
by the Supervisor, Health and Safety, and the Facility Supervisor.
l They shall functionally report to the facility Supervisor.
They l
shall be responsible for the safety and compliance of all operations in their assigned areas. They shall be responsible for the maintaining the exposures of personnel assigned to their area below 300 millf rom /weeki 1250 millf rems / quarter.
They shall have approval authority for Radiation Work Permits that apply to their I
assigned areas.
They shall keep the Facility Supervisor advised of a11 plans for projects and work to be carried out in their areas.
2.2.4 Manager, Safety and Licensing - The Manager Safety and Licensing l
reports to the Manager, EC&RR. The Supervisor, Health and Safety.
l Lloones No SNM.778 Docket No. 70 824 Date Apri1,1987 Amendment No.
O Reviolon No.
4 Pese 2-1 Babcock & Wilcox A McDermott comp.iny
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the Accountability Specialist, and the License Administrator report to this manager.
2.2.5 Supervisor Health and Safety - The Supervisor Health and Safety is responsible for providing adequate facilities, procedures, and properly trained personnel to implement the Health Physics and Industrial Safety Programs. He is responsible for health physics and industrial safety activities. The Supervisor Health and Safety reports to the Manager, Safety and Licensing.
The Super-l visor Health and Safety has the authority to stop any operation that he belfoves is contrary to accepted safety practicos or license requirements. He shall review all new Area Operating Procedures, Rtdiation Work Permits and changes thereto, for the radiation safety aspects of the procedure RWP, or change, and he shall have approval authority for them.
He shall conduct training programs for new employees and Authorized Users of Radioactive Ma terial.
He shall be responsible for the shipment of licensed ma terial. The Supervisor Health and Safety shall be a member of the Safety Review Committee but shall not be a member of the Safoty y
Audit Subcommittee. He shall have approval authority for Area Supervisors.
2.2.6 Senior Health Physics Engincor - The Senior Health Physics Engincor shall administer activities of the Health Physics Staf f.
He shall 77 report to the Supervisor Health and Safety.
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C' 2.2.7 Industrial Safety Officer - The Industrial Safety Officer shall administer the industrial safety program.
Ho shall report tJ the Supervisor Health and Safoty.
2.2.8 Nuclear Criticality Safety Of ficer - The Nuclear Criticality Safety l
Of ficer shall be responsible for ensuring that no operation at the sito results in the inadvertent assembly of a critical mss.
He l
shall review all new Area Operating Procedures and changes thorcto, for nuclear criticality safety and shall have approval authority for them. He shall conduct training programs in criticality safety and perform criticality safety calculations.
Ho shall report to the Manager Safety and Licensing.
2.2.9 License Administrator - The License Administrator shall be responsible for administering the license.
Ho is the primary liaison with the NRC and other federal, stato, and local agenclos in matters that pertain to nuclear activities, ho shall be the coordinator of the SR0 and the Safoty Audit Subcommittoo and shall represent managemont on both.
Ho shall maintain the permanent License No SNM 778 Docket No. 70 824 Date April, 1987 Amendment No.
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(V) records of the SRC and shall be responsible for assuring that appropriate action is taken to correct SAS audit findings that are i
approved by the Manager ECARR. He shall report to the Manager.
l Se fety and Licensing.
2.2.10 Accountabi1ity Specf alist - The Accountabi11ty Specf alist shall be responsible for the maintenance and retention of SNM accountability records.
The Accountabf11ty Specialist shall report to the Manager, Safety and Licensing.
i 2.3 SAFETY REVIEW COMMITTEE 2.3.1 Function 2.3.1.1 The SRC shall review and approve all new Area Operating Pro-cedures, and shall concur wtth all changes made to them in the time interval since their last regular meeting.
2.3.1.2 The SRC shall review and approve new projects and mejor changes to existing projects that utf1f ze Itcensed materials.
2.3.1.3 The SRC shall review the annual report prepared by the Supervisor, Health and Safety.
2.3.1.4 Tho SRC shall provide general consulting services in the field of
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radiation protection and the safe handling of Itcensed material.
2.3.1.5 The SRC sh411 review all SAS audtt findings, all overexposures and unusual occurrences which must be reported to the NRC.
These reviews shall be conducted during the next regularly scheduled meeting following the event and the results of the review shall be documented in the minutes.
2.3.1.6 The SRC Coordinator shall be responsible for resolving comments and recomendations made by the SRC.
2.3.2 Frequency of Meetings 2.3.2.1 The SRC shall meet at least four times annually for the purposes of conducting its business as specif fed in Sectfon 2.3.1.
2.3.3 Safety Audit Subcommittee 2.3.3.1 The SAS shall perform audits for the Safoty RoyIow Commf ttee.
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2.3.3.2 The SAS shall audit facilities, procedures, records, and opera-tions for compliance.1th written requirements and the exercise of acceptable safety practices.
2.3.3.3 The SAS shall perform at least three audits annually, distributed over a 12-month period. Audits shall be nido in accordance with written guidance to assure all aspect of 2.3.3.2 are audited.
2.3.3.4 SAS membership shall be appointed by the Manager, EC&RR.
l 2.3.4 Reporting 2.3.4.1 The SRC shall report to the Manager, EC8RR.
2.3.4.2 The SAS shall report to the Chairman, SRC.
2.3.5 Recordkeeping 2.3.5.1 Minutes of the SRC proceedings shall be prepared by the Chairman, SRC.
2.3.5.2 SRC Minutes sh.111 be forwarded to the Manager, EC8RR by the Chair-ntn, SRC.
2.3.5.3 The perninent records of the SRC shall be kept by the SRC
(~')
Coordina tor.
2.3.5.4 SAS audit reports shall be prepared by the Chairman, SAS.
2.3.5.5 SAS audit reports shall be forwarded to the Chairnin, SRC by the Chairnin, SAS.
2.3.S.6 SAS audit reports shall be forwarded to the Manager, ECARR by the l
Chairain, SRC with comments, as he deems appropriate.
2.4 APPROVAL AUTHORITY FOR PERSONNEL SELECTION 2.4.1 The Kinager, ECARR shall approve the personnel selected for safety-related positions specified in Section 2.2 of this Part and shall appoint the members of the Safety Review Committee in writing. The NNFD shall appoint the Kinager, ECARR.
Liceme No SNM.770 Docket No. 70 824 Date April,1987 Amendment No.
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p 2.5 PERSONNEL EDUCATION AND EXPERIENCE REQUIREMENTS 2.5.1 Manager, EC&RR - The Manager, EC&RR shall be appointed in ac-cordance with Company policy.
2.5.2 Facility Supervisor - The Facility Supervisor shall have a degree in his related work and three years experience in the use and handling of Itcensed material, or five years experience in the use and handling of licensed material. He must demonstrate to management proficiency in the application of good principles of radiation protection, industrial safety, and nuclear safety as related to the activities at the site.
I 2.5.3 Area Supervisors - Area Supervisors shall be Authorized Users and shall have demonstrated sufficient knowledge and experience in the equipment and techniques employed in projects performed in their assigned areas to ensure that all operations are conducted safely and in full compliance with applicable license conditions and area operating procedures.
2.5.4 Manager, Safety and Licensing - The Manager, Safety and Licensing shall have a 85 degree in a technical field and five years experience in the nuclear field.
2.5.5 Supervisor, Health and Safety - The Supervisor, Health and Safety shall have a BS degree in a technical field and professional Oi experience in assignments involving radiation protection at the supervisory level. He must have four years experience and demonstrate proficiency in the application of radiation safety principles and be knowledgeable in fields related to radiation protection.
2.5.6 Senior Health Physics Engineer - The Senior Health Physics Engineer l
shall have a BS degree which shall include at least 20 quarter hours health physics related course work, and two years of radi-ation control related experience or an MS degree and one year of radiation protection experience.
2.5.7 Industrial Safety Officer - The Industrial Safety Officer shall have at least one year's experience in radiation and industrial sa fe ty. He shall be familiar with the codes and requirements of the Occupational Health and Safety Act of 1970 and the National Fire Protection Association.
2.5.8 Nuclear Criticallty Safety Of ficor - The Nuclear Criticallty Safety Officer shall have a BS degree in science or engineering.
He must Lloonee No SNM 778 Docket No.70-824 Date April, 1987 Amendment No.
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t have either two years experience with nuclear criticality safety
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calculations similar to those associated with site-activities or he must have one year's experience with nuclear criticality safety i
i calculations similar to those associated with site activities if he has at least an additional two years' experience in nuclear reactor physics calculaticns.
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I 1
2.5.9 Accountability Specialist - The Accountability Specialist shall-have at least a high school education and three years' experience in the use of licensed material. He must demonstrate to Company management his knowledge of the principles necessary for the l
accountability and safegtarding of special nuclear materials.
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2.5.10 License Administrator - The License Administrator shall have a 85 degree in science or engineering and three years experience in i
nuclear technology or an AS degree in science or nuclear technology j
with five years experience in cuclear technology.
2.5.11 Safety Review Committee - The SRC membership, as a body, shall have l
expertise in chemistry, nuclear physics, haalth physics, and the safe handling of radioactive material. The SRC membership shall l
l have a general understanding of nuclear criticality safety as it pertains to site operation.
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2.6 TRAINING i
2.6.1 Program 1 - This course is presented to site workers and non-site l
workers who will be granted access to the res.'ricted area but who l
will not be granted unescorted access to the controlled areas. The course provides an introduction to radiation ano radioactivity i
(understandable to a non-technical person) and a thorough coverage j
of safety rules and procedures, including the site amargency pro-cedures. The Supervisor, Health and Safety may modify the course I
content for those individuals knowledgeable in the basics of radi-l ation and radioactivity. However, safety rules, procedures, and
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emergency procedures that apply at the site shall be covered.
i 2.6.2 Program 2 - This course is presented to site workers and non-site workers who will be granted unescorted access to the restricted area and controlled areas but who will not be permitted to work i
with radioactive materials without supervision.
The Supervisor.
l Health and Safety may modify the course content for those indi-l viduals knowledgeable in the basics of radiation and radioactivity.
j However, safety rules, procedures, and emergency procedures that i
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i Licenes No SNM 778 Docket No. 70 824 Date April 1987 Amendment No.
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apply at the site shall be covered. The effectiveness of the training shall be determined by a written examination.
2.6.3 Program 3 - This course shall be presented to site workers and non-site workers who will be granted unescorted access to the restricted area and controlled areas and will be permitted to work with radio-active materials and supervise such work. This course shall meet the requirements for designating a worker as an Authorized User.
The Supervisor, Health and Safety may modify the course content for those individuals knowledgeable in the basics of radiation and radioactivity.
However, safety rules, procedures and emergency procedures that apply at the site shall be covered. The effective-ness of the training shall be determioned by a written examination.
2.6.4 Retraining - Persons who are designated as Authorized Users shall be retrained annually. Satisfactory completion of the retraining shall be determined by passing a written examination.
2.6.5 Respiratory Protection Training - Training in respiratory pro-tection techniques and equipment shall be required of all workers before the use of such equipment will be permitted. Satisfactory completion of this training shall be determined by passing a written examination.
2.6.6 Respiratory Protection Retraining - Retraining in respiratory pro-tection shall be performed at two year intervals.
Satisfactory s
y'j completion of this retraining shall be determined by passing a written examination.
2.6.7 The training specified in Section 2.6 shall be administered by the Supervisor Health and Safety, or his designated and qualified al terna te.
2.6.8 Nuclear Criticality Safety Training - Nuclear Criticality Safety training provided as a part of the programs specified in Sections 2.6.2, 2.6.3 and 2.6.4, shall be performed by the Nuclear Criti-cality Safety Officer or his designated alternate. The designated alternate must meet the same minimum qualifications as those speci-fled for the Nuclear Criticality Safety Officer (2.5.8).
l 2.7 OPERATING PROCEDURES 2.7.1 Area Operating Procedures April, 1987 License No SNM 778 Docket No.70-824 Date Amendment No.
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2.7.1.1 All operations with licensed material shall be conducted in s
accordance with Area Operating Procedures or Radiation Work Permits (see 3.1.1).
2.7.1.2 Area Operating Procedures (AOP) - Area Operating Procedures shall be established for all routine operations in which SNM, source and byproduct materials are stored or handled. A0P's shall include those nuclear criticality and radiation safety controls and limits that apply to the operation.
Each A0P shall be approved by the Nuclear Criticality Safety Officer or his designated alternate, the Supervisor, Health and Safety or his designated alternate, the Facility Supervisor or his designated alternate, and the Safety Review Committee.
2.7.1.3 A0P's may be revised with the approval of the Nuclear Criticality Safety Officer or his designated alternate, the Supervisor, Health and Safety or his designated alternate, and the Facility Supervisor or his designated alternate. The revised procedure may be used with these approvals until the next scheduled regular meeting of the Safety Review Committee when the revision must be approved by the SRC.
2.7.1.4 A0P's shall be available in each operations area where they apply and shall be followed by site personnel.
O 2.7.1.5 Distribution of new and revised procedures shall be made in
/b accordance with a document control system which assures that the procedure manuals contain only the most. current revision of the procedures.
2.7.1.6 A0P manuals shall be reviewed annually by the Facility Supervisor to assure that the manuals contain the most current revision of the procedures.
2.7.2 Technical Procedures 2.7.2.1 Technical procedures shall be established, reviewed, approved, and followed for Health and Safety or Nuclear Criticality Safety.
They shall be reviewed and approved by the Senior Health Physics Engineer l
or the Nuclear Criticality Safety Officer, respectively, or their l
designated alternate. The designated alternate for a Senior Health l
Physics Engineer must meet the minimum qualifications specified in Sections 2.5.6.
The designated alternate for the Nuclear Criti-l cality Safety Officer must meet the same minimum qualifications specified in Section 2.5.8.
Approval signatures shall appear on the procedure.
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License No SNM 778 Docket No.70-824 Date April, 1987 i
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2.8 1NTERNAL AUDITS AND INSPECTIONS 2.8.1 Nuclear Criticality Safety 2.8.1.1 The Nuclear Criticality Safety Officer or his designated alternate shall conduct internal audits for the purpose of evaluating the nuclear criticality safety aspects of operations. This audit shall be conducted in accordance with written audit guidance.
This audit shall be conducted once each calendar quarter. A i
report of his findings shall be made to the Manager, EC&RR within I
two weeks of completing the audit. The audit reports shall be forwarded to the Facility Supervisor and the License Administrator.
The License Administrator shall be responsible for assuring that the appropriate corrective actions are taken to address the audit findings.
2.8.1.2 The Facility Supervisor shall perform an inspection weekly for compliance with the nuclear criticality safety aspects of the opera tions. Findings resulting from these inspections shall be reported to the Nuclear Criticality Safety Officer.
[
2.8.2 Health Physics 2.8.2.1 The Supervisor, Health and Safety or his designated alternate shall conduct internal audits for the purpose of evaluating the n('}
health physics aspects of operations. This audit shall be conducted in accordance with written audit guidance. This audit shall be conducted once each month. A report of his findings shall be made to the Manager, EC&RR within two weeks of completing J
the audit. The audit reports shall be forwarded to the Manager, EC&RR and the License Administrator.
The License Administrator shall be responsible for assuring the appropriate corrective actions are taken to address the audit findings.
2.8.3 General Safety and Compliance 2.8.3.1 The SAS performs audits of general safety and compliance. These audits shall be conducted three times annually. The audits shall be distributed over a 12-month period. The SAS shall include an audit of the Health and Safety Group at least once annually. This annual audit shall be performed by a qualified individual who is independent of the Health and Safety Group.
Other areas shall be audited for compliance with written requirements and the exercise of acceptable safety practices. Audits shall be made in accordance with written guidance to assure all aspects of Section 2.3.3.2 are License No SNM-778 Docket No.70-824 Date April, 1987 Amendment No.
O Revision No.
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audi ted. The Chairman, SAS shall file a report of the audit findings with the Chairman, SRC, with a copy to the License Admin-istrator and the Facility Supervisor and members of the SRC. The Chairman, SRC shall forward the report to the Manager, EC&RR with l
comments, as he deems appropriate. The License Administrator shall be responsible for assuring that the appropriate corrective actions are taken to address the audit findings.
2.9 INVESTIGATIONS AND REPORTING OF 0FF-NORMAL OCCURRENCES 2.9.1 License Administrator The License Administrator shall investigate and report, when required, the following types of off-normal occurrences:
2.9.1.1 Excessive levels of radiation from or contamination on packages upon receipt.
2.9.1.2 Thef ts, attempted thef ts, or losses of licensed material, other than normal operating losses.
2.9.1.3 Incidents as specified in 10 CFR 20.403 2.9.1.4 Overexposure of individuals and excessive levels and concentra-O tions of radioactivity.
V 2.9.1.5 Failures to comply and defects pursuant to 10 CFR 21.
2.9.1.6 Changes to security, safeguards, or emergency plans made without prior NRC approval, when prior approval is required.
2.9.1.7 Failures to comply with license requirements.
2.9.1.8 Unapproved storage or use of licensed material.
2.9.2 Supervisor, Health and Safety The Supervisor, Health and Safety shall perform investigations and issue reports of the following:
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2.9.2.1 Higher than expected personnel exposures.
2.9.2.2 Higher than expected concentration of airborne activity in the i
facili ty.
License No SNM-778 Docket No.70-824 Date April, 1987 l
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2.9.2.3 Unauthorized entry into a High Radiation or Airborne Radioactive Material area.
2.9.2.4 Failure of equipment or instrumentation to meet Health and Safety requirements.
2.9.3 Facility Supervisor The Facility Supervisor shall perform investigations of the following:
2.9.3.1 Any violation of nuclear criticality safety criteria.
2.9.3.2 Any violation of Area Operating Procedures or RWP's.
2.10 RECORDS The following positions or organizations shall be responsible For maintaining the indicated records, for the period specified.
Records may be kept in original form, microfilm or in computer storage. The symbol (*) indicates that the record will be retained until the NRC authorizes its disposition.
2.10.1 Health and Safety Group O
Health and Safety Supervisor audits 2 years Shipping and receiving RM forms 5 year!.
Waste disposal records
(*)
Personnel dosimetry records
(*)
Results of Bioassays and Whole Body Counting
(*)
Releases to the environment
(*)
Radiation survey data 2 yea, s Contamination survey data 2 years Radiation Work Permits (completed) 5 years Radiation detection instrument calibration 2 years Leak tests of sealed sources 2 years Personnel training
(*)
Personnel retraining
(*)
Airborne radioactivity sampling data
(*)
NRC-4 forms
(*)
NRC-5 forms
(*)
License No SNM-778 Docket No.70-824 Date April, 1987 Amendment No.
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2.10.2 Nuclear Criticality Safety Officer l
Nuclear criticality safety 6 months af ter l
evaluations and calculations termination of the approved
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process.
Nuclear Criticality Safety Officer 2 years Audit Reports.
2.10.3 License Administrator Safety Review Committee Minutes
(*)
Safety Audit Subcomittee Audit Reports 2 years Investigation reports of off-normal occurrences 2 years 2.10.4 Emergency Records Records pertaining to emergency response and preparedness shall be retained in accordance with Radiological Contingency Plan, Section 8.0.
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AU License No SNM-778 Docket No.70-824 Date April, 1987 Amendment No.
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FIGURE 2-1 SITE ORGANIZATION 1
1 NNFD RESEARCH LABORATORY ORGANIZATION EMPLOYEE, COMMUNITY, & REGULATORY RELATIONS MANAGER FACILITY SAFETY REVIEW SUPERVISOR COMMITTEE SAFETY & LICENSING MANAGER 1
es b
NUCLEAR LICENSE HEALTH & SAFETY CRITICALITY ACCOUNTABILITY ADMINISTRATOR SAFETY SPECIALIST OFFICER SUPERVISOR 1
4 i
SENIOR INDUSTRIAL HEALTH PHYSICS SAFETY ENGINEER OFFICER HEALTH PHYSICS GROUP I
I AREA AREA SUPERVISOR SUPERVISOR License No SNM-778 Docket No.70-824 Date April, 1987 Amendment No.
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4 Page 2-13 Babcock &Wilcox a McDermott company
3.0 LRADIATION PROTECTION L
3.1 SPECIALA_0M)NISTRATIVEREQUIREMENTS 3.1.1 Radiation Work Permits (RWP) 3.1.1.1 RWP's shall be issued whenever the activity is not covered by an 3
Area Operating Procedure and workers are likely to be exposed to
. levels of radiation or concentrations of radioactive material in i
excess of those specified in 10 CFR 20.1018 '20.103.
]
3.1.1.2 RWP's shall be approved by the Area Supervisor, Health Physics" Supervisor, and the Facility Supervisor.
In the absence of any of the above persons, a designated and qualified alternate may approve RWP's.
3.1.1.3 The RWP shall specify the radiological protection requirements-for the operation and specify levels of worker exposure above which a-documented ALARA evaluation shall be performed. RWP's that re-quire a documented ALARA evaluation must, in addition to 3.1.1.2, j-be' approved by the Manager, EC&RR.
i-
-3.1.1.4 RWP's shall-be approved at a meeting of all the signators of the i
form.
O
-3.1.1.5 The RWP form shall provide space for entering the estimated U
exposures to the whole body, extremities, and for the job. These are used to identify the areas of exposure concern and do not j
constitute' an exposure goal or limit.
3.1.1.6 The RWP form shall provide space for the Area Supervisor to sign or initial, attesting that the _ workers have been instructed in the requirements of,the RWP.
3.1.1.7 The term of a RWP shall not exceed 30-days, except that Standing RWP's shall have a term not to exceed 6-months.
3.1.2 ALARA Policy
,s-7 The management of the site is committed to a policy of maintaining exposures as low as is reasonably achievable.
~.1.2.1 Site workers shall be introduced to this policy during their 3
i j1 initial training and shall be reinforced during the annual re-1-
training of Authorized Users.
t License No SNM-778 Docket No. 70 824 Date April, 1987 Amendment No.
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Babcock &Wilcox a McDermott company
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h,/8 3.1.2.2 The ALARA policy shall be implemented through the Area Operating Procedures and Radiation Work Permits.
3.1.2.3 The ALARA policy shall be enforced by the Facility Supervisor and the Supervisor, Health and Safety in the exercise of their review and approval authority, their authority to terminate operations, and audits.
3.1.2.4 The SRC shall evaluate ALARA performance in exercising their review authority over procedures and proposed new projects and their review of the annual report from the Supervisor, Health and Safety.
3.1.3 Off-Site Possession Off-site possession and use of licensed material shall be the re-sponsibility of and under the control of site workers specifically I
approved by the Safety Review Committee.
3.2 TECHNICAL REQUIREMENTS 3.2.1 Access Control 3.2.1.1 High Radiation Areas - Entry into a High Radiation Area or an Airborne Radioactivity Area shall be controlled by an RWP.
3.2.1.2 Contamination Areas - Areas which are determined by the Health and Safety Group to present a risk of spreading radioactive contamination into non-contaminated areas shall be clearly marked at each entrance. Step-off pads shall be provided.
Personnel survey instrumentation shall be provided at the step-off pad. The minimum protective clothing required in Contamination Areas shall be shoe covers and lab coats. Exiting such areas shall require personnel to remove their protective clothing and survey them-selves with the instrumentation provided. Persons found to be contaminated above background levels must receive the approval of the Supervisor, Health and Safety, prior to leaving the
'l contaminated area.
l 3.2.2
' Ventilation Requirements 3.2.2.1 Air flows within Building B shall be in the direction of highest potential for airborne radioactive material. Air flow directions shall be checked monthly.
i License No SNM-778 Docket No.70-824 Date April,1987 l
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b 3.2.2.2 Potentially contaminated exhaust air from hood, hot cells, and U
glove boxes shall be discharged through the fifty meter high stack, except as noted in 3.2.2.7.
3.2.2.3 The exhaust stack shall be sampled isokinetically.
3.2.2.4 The stack sampling and monitoring system shall operate continu-ously except for periods when repair or calibration is required.
3.2.2.5 The following table presents the release limits and action levels associated with the exhaust stack. The Health and Safety Group shall be responsible for responding to releases in excess of these action levels. An operation that results in action levels being exceeded for 4-consecutive time periods, shall be shutdown until the cause is corrected.
STACK RELEASE LIMITS AND ACTION LEVELS Release Product Release Limit Action Level Beta Particulate 2 mci /yr 200 uCi/ week Alpha Particulate (long lived) 20 uCi/yr 1 uCi/2 weeks Kr-85 2500 C1/yr 70 C1/ week H-3 130 Ci/yr 3 Ci/ week I-131 6 mci /yr or 300 uCi/ week 200 uCi/ week 3.2.2.6 Exhaust systems that cannot be practicably discharged through the 50-meter stack, and where there exists a reasonable probability that the discharges to the atmosphere could exceed 10% of the applicable MPC for an unrestricted area, shall be monitored for gaseous and particulate activity in the effluent.
3.2.2.7 Exhaust air from areas in which there is no airborne radioactive material may be exhausted directly to the roof either with or without continuous sampling, if approved by the Safety Review Committee.
License No SNM 778 Docket No.70-824 Date April, 1987 Amendment No.
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O 3.2.2.8 Areas equipped with an air monitor may be exhausted to the roof V
through HEPA filters if the concentration of airborne radioactive material is below the appropriate MPC for an unrestricted area, if approved by the Safety Review Committee.
3.2.2.9 All hoods used for the handling of licensed material shall exhaust through one HEPA filter, except for hoods that are specifically designed and installed for use with perchloric acid.
3.2.2.10 Fume hoods utilized for the handling of unirradiated Pu shall be provided with two HEPA filters in series.
3.2.2.11 Hot cells shall be provided with two stages of HEPA filters.
3.2.2.12 Final HEPA filters which service facilities where licensed material is handled shall be tested, using the cold DOP test, annually or after a final HEPA filter is changed, whichever comes sooner.
3.2.2.13 The acceptance criteria for the testing of final HEPA filters (3.2.2.12) shall be 99.95% of all particles having a light-
- cattering mean diameter of approximately 0.7 micrometers.
3.2.3 Instrumentation Q
3.2.3.1 Portable Survey Instruments LJ 3.2.3.1.1 Portable instruments - A relatively large and diverse inventory of portable survey instruments is maintained. These instruments vary in range and sensitivity. The below listing is a representative sampling of the instruments on hand:
Instrument Sensitivity Characteristics Type Radiation Ionization 0 - 20K R/hr 6.5 Kev - 1.2Mev Beta & Gamma Chamber Geiger 0 - IK R/hr 23 Kev - 1.2Mev Beta & Gamma Counter Proportional 25 - 500K cpm Alpha & Beta Counter (gas flow)
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Scintillation 0 - 50K cpm Alpha Detector Geiger 0 - 50K' cpm
>40 Kev Beta Counter Scintillation 0 - 5 mR/hr Gamma Detector Neutron Dose 0 - SK mR/hr 25 Kev - 3MeV Neutron 3.2.3.1.2 Portable survey instruments shall be calibrated semiannually.
3.2.3.2 Air Monitors 3.2.3.2.1 Nuclear Measurements Corp. (NMC) Model AM-2A - This instruynt utilizes a gas flow proportional detector with a 1.0 mg/cm thick end window. These instruments are operated as alpha or beta-gamma monitors. Tgey utilize a fixed filter with a nominal air flow of 2.5 to 3 f t / min. The alarm setting is set at less than 40 MPC hours above normal background including Radon and Thoron daughters.
3.2.3.2.2 Eberline Model AIM-3S - These monitors are used for alpha monitoring only. They are typically located in areas where Pu D
or U is being processed. They use a ZnS(Ag) scintillation r with a fixed filter. The monitor air flow is nominally detecp/hr.
20 ft The alarm is set at less than 40 MPC hours above the normal background for Radon and Thoron daughters.
3.2.3.3 Air Samplers 3.2.3.3.1 Mine Safety Appliance (MSA) Model G - These personal air samplers utilize a Millipore field sample cassette. The nominal air flow rate is 2 liters / min.
Samples are collected for count-ing on a low background counter with sufficient sensitivity to detect 25% of the applicable MPC for 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> sampling intervals.
3.2.3.3.2 Fixed samplers are located at work stations where the concentra-tion of airborne radioactive material potentially exceeds 25% of the applicable MPC.
3.2.3.4 Criticality Monitors l
License No SNM 778 Docket No.70-824 Date April, 1987 Amendment No.
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(O) 3.2.3.4.1 Nuclear Measurements Corp. (NMC) Model GA-2TO and GA-2A - These monitors are designed as criticality alarm systems. Detection is by a Na! (TI) detector operated in the constant current mode.
Response is logarithmic r.nd non-saturating. Emergency power is provided. The nominal alarm setpoint is 20 mR/hr. Failure alarm function is provided.
Criticality monitors shall be calibrated semiannually.
3.2.3.5 Counting Equipment 3.2.3.5.1 Sharp Low Beta - Air samples and effluent samples may be counted on this instrument. This instrument utilizes a 4.5-inch and a 2.5-inch very thin end window proportional detector. Back-grounds and counter response are tested weekly and the instru-ment is calibrated annually.
3.2.3.5.2 Beckman Wide Beta - Air samples and effluent samples may be counted on this instrument.
It utilizes two 2.5-inch very thin end window proportional detectors. Backgrounds and counter response are tested weekly and the system is calibrated annually. The manual detector is used infrequently and it is tested when used.
3.2.4 Internal and External Exposure f;
3.2.4.1 Ventilation
%J 3.2.4.1.1 The minimum air velocity across the opening of fume hoods that are used to handle licensed material shall be at least 100 fpm.
Hood face velocities shall be measured monthly. Those hoods that do not meet the minimum requirement shall be placed out of service.
3.2.4.1.2 The maximum differential pressure across HEPA filters shall be limited to 4-inches of water, except the hot cell filters which shall be limited to 5-inches of water. HEPA filters shall be changed to prevent exceeding these limits. The differential pressure across HEPA filters shall be checked weekly.
3.2.4.1.3 The minimum differential pressure across the hot cell face shall be 0.25-inches of water. The differential pressure across the hot cell face shall be checked weekly. An additional hot cell fan will be automatically or manually started when the differential pressure reaches 0.25-inches of water.
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License No SNM 778 Docket No.70-824 Date April, 1987 i
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g 3.2.4.2. Air Sampling and Analysis 3.2.4.2.1 Continuous air sampling shall be performed in all areas where, in the judgment of the Supervisor, Health and Safety, there exists the potential for exposing personnel to concentrations of airborne radioactive materials in excess of 10% of the appli-cable MPC.
3.2.4.2.2 Air sampling filters shall be changed daily in areas where l
sample evaluations indicate concentrations of airborne radio-active materials in excess of 10% of the applicable MPC.
3.2.4.2.3 Air sampling filters shall be changed weekly in areas where l
sample evaluations indicate concentrations of airborne radio-active materials less than or equal to 10% of the applicable MPC.
3.2.4.2.4 Preliminary evaluation of air sample filters shall be performed within the next working day following their removal. Final evaluation of air sampling filters shall be performed within 7 working days following their removal.
3.2.4.2.5 An investigation by a Senior Health Physics Engineer shall be l
performed into the cause of unexpected air sampling results that indicate airborne activity at levels between 10% and 25% of the appitcable MPC. The Senior Health Physics Engineer shall assign l
responsibility for completion of any actions that may be indi-s cated by the investigation.
3.2.4.2.6 An investigation by the Supervisor, Health and Safety shall be performed into the cause of unexpected air sampling results that indicate airborne activity at levels exceeding 25% of the appli-cable MPC. The Supervisor, Health and Safety shall be resporsi-ble for specifying corrective actions and assuring that the specified actions are taken.
3.2.4.2.7 If fixed air samplers are used to determine concentrations of airborne radioactivity in the worker's breathing zone, the representativeness of the samplers shall be determined at least once every 12-montns.
3.2.4.3 Bioassay 3.2.4.3.1 Uranium Bioassay Program License No SNM 778 Docket No.70-824 Date April, IS87 Amendment No.
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- = - -
- 1. The uranium bioassay program sampling frequency shall comply with Regulatory Guide 8.11, June,1974, except as specified in section 1.8 of this applicatiori.
- 2. All workers who routinely work in uranium handling areas shall be subject to the uranium bioassay program. The follow-ing are the action criteria for the routine uranium bioassay program:
Action Analysis Level Action to be T5 ken
- a. Urinalysis
< 9 ug/l None
- b. Urinal,ysis 9-16 ug/l
- 1. Determine if area surveys support the analysis re sul ts.
- 2. If #1 is positive, investi-gate and correct as needed.
4
- 3. Make sure individual is in-vivo counted during the next time that the body counting service is at the B&W site.
i
- c. Urinalysis
> 16 ug/l
- 1. Restrict the worker from further exposure. Resample the individual within 5 i
working days.
- 2. Determine if area surveys support the analysis resul ts.
- 3. If #2 is positive, i
investigate the cause and correct as needed.
I
- 4. If exposure is confirmed by l
- 2, investigate to determine j
how exposure was incurred l
and correct it.
If the ex-posure exceeds 50% of the maximum permissible annual License No SNM 778 Docket No.70-824
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I dose, the worker shall be restricted from further exposure until the Super-visor, Health and Safety authorizes the 11f ting of the restriction.
- d. In-vivo
< 30 ug None U-235
- e. In vivo 30-120 ug
- 1. Determine if area surveys U-235
' support the analysis resul ts.
- 2. If #1 is positive, investi-gate and correct as needed.
- f. In vivo
> 120 ug
- 1. Resample the individual U-235 within 10 working days.
- 2. Determine if area surveys support the analysis resul ts.
j' 3.- If #2 is positive, investi-gate the cause and correct i
as needed.
I
- 4. If exposure is confirmed by
- 1, investigate to determine i
how exposure was incurred i
and correct it.
If the ex-l posure exceeds 120 ug,- the worker shall be-restricted 4'
from further exposure until the Supervisor, Health and I
Safety authorizes the lif ting of this restriction.
I 3.2.4.3.2 Plutonium Bioassay Program i
- 1. All workers who routinely work in Plutonium handling areas l
[
shall be subject to the Plutonium bioassay program. - The minimum frequency for urine sampling shall be six months. The j
minimum frequency for in vivo counting shall be annual. Ad-License No SNM 778 Docket No.70-824 Date April, 1987 Amendment No.
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ditional bioassays shall be performed when, in the judgment of the Supervisor, Health and Safety, conditions during a job and/or other data (air samples, floor smears or clothing con-tamination) indicate an internal exposure may have occurred.
- 2. The following are the action criteria for the routine Plu-tonium bioassay program:
Action Analysis Level Action to be Taken
- a. Urinalysis
< 0.2 dpm/L None
- b. Urinalysis
> 0.2 1pm/L
- 1. Resample the individual within 5 working days.
- 2. The Supervisor, Health and Safety shall consider the need for worker restriction to prevent further exposure until the diagnostic evalu-ation is complete. Only the Supervisor, Health and Safety may lift any work restriction g
once it is imposed.
- 3. Determine if area surveys support the analysis resul ts.
- 4. If #3 is positive, investi-l gate the cause and correct.
- 5. If the exrosure is confirmed by #1 investigate to de-termine how exposure was incurred and correct it.
If the exposure exceeds 50% of the maximum permissible annual dose, the worker shall be restricted from further exposure until the Supervisor, Health and Safety authorizes the lifting of this restriction.
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- c. In vivo
< 1.6E-8 Ci None Pu-239
- d. In vivo
> 1.6E-8 C1
- 1. Restrict the worker from Fu-239 further exposure.
- 2. Resample the individual within 10 working days.
- 3. Determine if area surveys support the analysis resul ts.
- 4. If #3 is positive, investi-gate the cause and correct as needed.
S. If exposure is confirmed by
- 2, the Supervisor, Health and Safety shall determine the organ dose.
If the confirmed exposure exceeds 50% of the maximum per-missible annual dose, the worker shall be restricted from further exposures until the Supervisor, Health and Safety authorizes the 4
lif ting of this restriction.
- 6. The restriction in #1 may be lifted by the Supervisor, i
Health and Safety if the results of the analysis per-formed under #2 fails to confirm the analysis.
3.2.4.3.3 Fission Product Bioassay Program
- 1. The fission product bioassay program sampling frequency shall comply with Regulatory Guide 8.26, September,1980.
- 2. Additional bioassays shall be performed when in the opinion of the Supervisor, Health and Safety, conditions during the job were such that significant internal exposure may have License No SNM 778 Docket No.70-824 Date April, 1987 Amendment No.
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occurred. The following are action criteria for additional bioassays.
Action
~
Analysis Level Action to be Taken In vivo
>10% MP08 Remeasure subject to determine effective half life of the contami-nant and plot decay curves.
Follow-up program will continue until the contamination present is
<5% MP0B or the effective half life has been determined.
Estimation >10% MP08 Submit in vitro sample for analysis from nasal within 5 working days, smears or air sample In vitro
>5% MP0B Resample excreta to confirm presence of contamination and to establish rate of elimination.
Perform isotopic analysis if >10%
^
of MP08 is a possibility.
In vitro
>10% MP08 In vivo measurement to be made as soon as practicable.
- 3. The Supervisor, Health and Safety, shall be responsible for evaluations to determine the location and amount of depo-sition; to provide data necessary for estimating internal dose rates, retention functions, and dose commitments; and to determine if work restrictions or referrals for therapeutic treatment are required for any case where a result indicating a greater than 10%/MP08 deposition of a radionuclide is verified.
3.2.4.4 Protective Clothing j
l 3.2.4.4.1 The use of protective clothing shall be specified in Area Operating Procedures and Radiation Work Permits.
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3.2.4.4.2 Protective clothing may also be specified by the Health and V
Safety Group.
In the event of conflicts between the Area Operating Procedure, Radiation Work Permit, and the Health and Safety Group, the decision of the latter shall prevail.
3.2.4.5 Respiratory Protection 3.2.4.5.1 The Respiratory Protection Program shall be conducted in ac-cordance with 10 CFR 20.103, and shall be a responsibility of the Health and Safety Group.
3.2.4.5.2 The Respiratory Protection Program shall be implemented througn written and approved procedures.
3.2.4.6 Surface Contamination Monitoring 3.2.4.6.1 The Health and Safety Group shall perform smear surveys in the below listed areas at the indicated minimum frequencies:
Action Level Area Frequency (dpm/100cm2)
<---------------------------ALPHA------------------------->
Unirradiated, unencapsulated Weekly 5000 b(7 fuel handling areas Building B Counting Lab.
Monthly 200 Hot Cell Oper. Area Monthly 200 Scanning Electron Monthly 200 Microscopy Lab.
Exit portals from Biweekly 200 controlled areas
<-----------------------BETA + GAMMA---------------------->
Building B Counting Lab.
Montaly 2000 Scanning Electron Monthly 2000 Microscopy Lab.
Hot Cell Operations Area Bimonthly 2000 License No SNM-778 Docket No.70-824 Date April, 1987 l
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w nQ Cask Handling Area Bimonthly 22000 Radiochemistry Lab.
Bimonthly 22000 Exit portals from Biweekly 2000 controlled areas 3.2.4.6.2 Large area smears are used to survey many square meters of surface area. To determine if these smears indicate that an action level has been exceeded, the assumed area covered shall not exceed 1-square meter.
3.2.4.6.3 Daily surveys shall be performed in the cafeteria, snack bars, and vending machine areas.
If contamination is detected in any of these areas, corrective action shall be taken at once.
3.2.4.7 Decon tamination 3.2.4.7.1 The Health and Safety Group shall determine and direct the action to be taken to protect personnel and reduce the levels of contamination below those specified in Section 3.2.4.6.
3.2.4.7.2 Decontamination to reduce levels of contamination shall begin within 24-hours of the discovering survey.
If the survey is made just prior to the beginning of a holiday or weekend, the f.)
contamination shall be marked and labeled, and decontamination v
shall commence during the first regular workday af ter the survey.
3.2.4.7.3 Fixed contamination that, in the opinion of the Supervisor, Health and Safety, does not substantially contribute to a worker's exposure, shall be posted and its location and radiation level recorded and its removal shall be scheduled as soon as practicable.
3.2.4.7.4 Fixed contamination that, in the opinion of the Supervisor, Health and Safety, may substantially contribute to workers' exposure shall be posted and removed as soon as practicable.
3.2.4.8 Emergency Evacuation 3.2.4.8.1 Refer to Radiological Contingency Plan, required by Order dated l
February 11, 1981, as amended.
l 3.2.4.9 Personnel Monitoring l
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lq 3.2.4.9.1 Site Workers (Restricted Area) shall be issued a film badge, a g
3.2.4.9.2 Site Workers (Controlled Areas) shall be issued a film badge, a SRD, and a TLD.
3.2.4.9.3 Visitors shall be issued a TLD.
3.2.4.9.4 Non-Site Workers (Restricted Area) shall be issued a film badge, a SRD, and a TLD.
3.2.4.9.5 Non-Site Workers (Controlled Areas) shall be issued a film badge, a SRD, and a TLD.
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TABLE OF CONTENTS d.
I Section Page 4.0 NUCLEAR CRITICALITY SAFETY 4-1 4.1 SPECIAL ADMINISTRATIVE REQUIREMENTS 4-1 l
4.2 TECHNICAL REQUIREMENTS 4-2 4.2.1 Nuclear Isolation 4-2 l
4.2.2 Building A 4-2 4.2.3 Building B 4-3 4.2.4 Building C 4-10 4.2.5 Outside Stcrage 4-10 4.2.6 Dry Waste 4-10 O
License No SNM-778 Docket No.70-824 Date4pril, 1987 O
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4.0 NUCLEAR CRITICALITY SAFETY 4.1 SPECIAL ADMINISTRATIVE REQUIREMENTS 4.1.1 Double Contingency Policy - The Double Contingency Policy as de-fined in the American National Standard ANSI /ANS-8.1-1983 shall be followed in establishing the basis for nuclear criticality safety of all operations.
4.1.2 Structural Integrity - Where structural integrity is necessary to provide assurance for nuclear criticality safety, the design and construction of those structures will be evaluated with due regard to load capacity and foreseeable abnormal loads, accidents, and deterioration. The Manager, Facilities shall be responsible for determining the structural integrity of equipment when it is necessary to provide assurance of nuclear criticality safety.
4.1.3 Nuclear Criticality Safety Evaluation - All modifications or additions or both to any operation, system or equipment must be approved by the Facility Supervisor.
It is the responsibility of the Facility Supervisor, in consultation with the Nuclear Criti-cality Safety Officer, to determine'whether or not a nuclear criti-cality safety evaluation is required for the proposed modification n
or addition. The Nuclear Criticality Safety Officer or a person j
C) designated by him shall provide any required evaluations, including i
calculational support. Nuclear criticality safety evaluations shall be reviewed by a second individual, either the Nuclear Criti-cality Safety Officer or by a per::on with the same minimum qualifi-cations required for the Nuclear Criticality Safety Officer.
4.1.4 Posting - Each unit shall be posted with the limits of SNM per-mitted in the unit. The Facility Supervisor shall be responsible for approving the posting of nuclear criticality safety limits.
4.1.5 Labeling - Each container containing greater than 0.5 grams of SNM shall be labeled to show the amount of element, the percent enrichment, when applicable, and the amount of fissile isotope.
This condition does not apply to irradiated SNM.
4.1.6 Compliance - Compliance with the nuclear criticality safety require-ments shall be in accordance with written area operating procedures, reviewed and approved by the Facility Supervisor, the Supervisor, Health and Safety, the Nuclear Criticality Safety Officer, and the l
SRC. Area operating procedures shall include all the controls and License No SNM-778 Docket No.70-824 Date April, 1987 Amendment No.
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limits significant to the nuclear criticality safety of the oper-ation. In addition, the Nuclear Criticality Safety Officer shall l
perform a quarterly audit for compliance with nuclear criticality safety requirements and verifies that process conditions have not been altered that may affect nuclear criticality safety. The re-sults of the audit shall be documented and submitted to the Manager, EC&RR.
4.2 TECHNICAL REQUIREMENTS 4.2.1 Nuclear Isolation - When nuclear isolation is required (the potential neutronic interaction between units is negligible) the unit or units isolated shall be separated from all other SNM by one of the following or equivalent conditions:
1.
Twelve inches of water.
3 2.
Twelve inches of concrete with density of at least 140 lb/f t when the unit (s) being nuclearly isolated are one of the units permitted by this license, (i.e., a mass limit specified in Section 4.2.2.2 or an authorized PWR or BWR fuel assembly or portion thereof, pursuant to Section 4.2.3.6.1) provided tha t the unit or units cannot be representable as a slab which interacts with other SNM primarily through its major face.
b 3.
The edge-to-edge separation of 12-feet, or the greatest distance across an orthographic projection of either accumulation on a plane perpendicular to a line joining their centers, whichever is larger.
4.2.2 Building A 4.2.2.1 General - Building A shall be limited to 40 units as defined in section 1.6 of this Part. Each unit shall be separated from adjacent units by at least 8-inches edge-to-edge and 24-inches cen ter-to-cen ter.
4.2.2.2 Unit Limits - Each unit shall be limited to one of the following:
4.2.2.2.1 Mass limits for mixtures of plutonium and U-235 I
License No SNM 778 Docket No.70-824 Date April, 1987 Amendment No.
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O C/
Pu (wt%)
Limit (total grams fissile) 0 350 1 to 20 313 20 to 40 283 40 to 60 258 60 to 80 237 80 to 100 220 4.2.2.2.2 Mass Limits for Low Enriched Uranium - 850 grams of 'F235 as contained in uranium enriched in the isotope U-235 tc and including 4 wt%.
1.
Whenever uranium enriched to 4 wt% U-235 is being processed under the 850 gram limit, no unit shall be permitted to have fissile material at an enrichment greater than 4 wt% within that laboratory, room, or work area.
- 2. Whenever an 850 gram, enriched controlled unit is in use in the building, the Facility Supervisor must approve all transfers involving materials with enrichments greater than 4 wt% within the building.
4.2.3 Building B 4.2.3.1 General - Building B shall be limited to 40 units, excluding the CI hot cells, underwater storage, and the examination of power reactor fuel assemblies. Each unit shall be separated from adjacent units by at least 8-inches edge-to-edge and 24-inches center-to-center.
4.2.3.2 Unit Limits - Each unit shall be limited to the values specified in Section 4.2.2.2 of this Part.
4.2.3.3 Hot Cell - The hot cells, except for examination of power reactor fuel assemblies, shall be limited to the following units:
1.
Three units in hot cell no.1, separated by at least 12-inches edge-to-edge.
2.
One unit in each of the other hot cells.
4.2.3.4 Underwater Storage (Transfer Canal & Storage Pool) - SNM in storage under water in the Transfer Canal & Storage Pool shall be in racks or containers limited to the values specified in 4.2.2.2, excluding License No SNM 778 Docket No.70-824 Date April, 1987 Amendment No.
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,,C) power reactor fuel assemblies, and separated by 12-inches edge-to-edge.
4.2.3.5 Storage Tubes - SNM in storage tubes shall be limited to the values specified in 4.2.2.2 for each tube. Storage tubes shall be spaced a minimum of 17-inches center-to-center, are approximately 5-inches in diameter, and totally immersed in concrete.
4.2.3.6 Power Reactor Fuel Assemblies - Examination of unirradiated and irradiated power reactor fuel assemblies, including both non-destructive and destructive testing is carried out in Building B subject to existing nuclear criticality safety ifmits and controls except as modified by the following conditions.
4.2.3.6.1 Fuel assemblies to be studied are identified as:
1.
Each assembly shall be of the enriched uranium oxide PWR type with a 15 X 15, or 17 X 17 square pin lattice not greater than 8.6-inches on a side (further identified as a Babcock & Wilcox Mark B or Mark C canless assembly).
2.
The maximum initial enrichment in an unirradiated assembly shall not exceed 4.05 wt%.
3.
Damaged fuel assemblies may be examined in air. Fuel I3 assemblies which have been damaged can be examined in water O
if they maintain their 8.6-inches on a side dimensions.
4.2.3.6.1.1 Other PWR or BWR fuel assemblies which do not meet the above may be studied, provided:
1.
The unirradiated, fully reflected fuel assembly (fueled with U0 only) with all control rods removed is shown by p
an appropriate nuclear safety evaluation to be subcritical by at least 5 % (K-eff <0.95).
2.
The fuel assembly is shown by an appropriate nuclear safety evaluation to be subcritical by at least 5 % (K-eff
<0.95) under specific conditions of disassembly.
3.
In 4.2.3.6.1.1 Items 1 and 2 above, the primary source for validation data shall be:
o DP-1014, " Critical and Safe Masses and Dimensions of Lattices of U and U0 Rods in Water," by H. K. Clark of g
License No SNM 778 Docket No.70-824 Date April, 1987 Amendment No.
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pQ Savannah River Laboratories, or; o Actual critical experiments having approximately the same enrichment and metal-to-water ratio as is present in the fuel assembly to be studied, or; o Criticality data supplied by the reactor designer for the actual fuel assemblies.
An appropriate cross-section set will be selected and used to calculate the validation data from one of the above described sources. Any bias between calculational results with that cross-section set and validation data will be included in the results of the safety evaluation called for in #1 and #2 above. A description of and the results of the validation and the nuclear safety analysis will be assembled into a report which shall contain the items specified in Section 4.3.6 of ANSI / ANS-8.1-1983, " Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors."
4.2.3.6.1.2 BWR fuel assemblies may be received and studied provided:
1.
They are evaluated pursuant to Section 4.2.3.6.1.1 of this Part, or
[
2.
The BWR fuel assemblies have a maximum initial unirradi-ated enrichment of 4.05 wt% U-235 and have a cross sectional area not exceeding that of a 22.5 cm (8.85 in.)
diameter cylinder.
4.?.3.6.2 Receipt and' Storage 4.2.3.6.2.1 Unirradiated Fuel Assemblies - Unirradiated fuel assemblies will be received at a maximum of two at a time in a shipping container licensed for two assemblies, or one assembly in a shipping container licensed for one assembly.
Unirradiated fuel assemblies may be stored in air in the Crane & Cask Handling Area, the Assembly & Machine Shop Area, or the i
Development Test Area subject to the following conditions:
1.
Assemblies may be stored in their shipping container as received.
2.
Assemblies may be stored no less than 21-inches apart cen ter-to-cen ter.
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/**'s Q-3.
Assemblies may be stored under water in the hot cell pool, mockup pool, or development test area pool in racks con-structed to maintain a 1-foot minimum surface-to-surface separation between assemblies and any other SNM.
4.
No more than four unirradiated assemblies may be kept at the site at one time.
4.2.3.6.2.2 Irradiated Fuel Assemblies - Irradiated fuel assemblies shall be received one at a time in a licensed single assembly shipping container or two at a time in a shipping container licensed for two assemblies. Fuel assemblies that have been irradiated will be stored in the hot cell pool which is limited to the following conditions:
1.
A maximum of four assemblies or portions thereof may be in the pool at a time.
2.
The assemblies shall be stored in a storage rack which is so constructed as to maintain a 1-foot minimum surface-to-surface separation between the stored assemblies and any other fissile material which might be in the pool.
3.
Only one assembly may be in a designated work area of the pool at any one time. There shall be at least 1-foot minimum surface-to-surface separation between the assembly
\\
in the work area and any other fissile material.
4.
Fuel rods which have been removed from an assembly shall be stored in a storage rack providing space in each position for a maximum of 75 rods. All positions shall be spaced from any other fissile material by a minimum surface-to-surface separation of 1-foot.
5.
Partially dismantled assemblies will be stored in the assembly storage rack.
6.
Each position in the assembly storage rack and in the fuel l
rod storage rack must limit contaired fuel to a square not to exceed the dimensions of a fresh fuel assembly or to a cross sectional area not exceeding that of a 22.5 cm (8.85 in.) diameter cylinder.
4.2.3.6.3 Work Area of Pool Under Hot Cell No.1 - The work area position
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under Cell No.1 is used to load and unload irradiated fuel assemblies into and out of shipping casks an! to dismantle. both irradiated and unirradiated fuel assemblies. The following conditions govern operations in this work area:
1.
Only two assemblies at a time shall be permitted outside of
.their shipping container provided they are separated by a minimum surface-to-surface separation of 1-foot.
2.
An associated storage position shall be permitted fo fuel rods or components which have been removed from de assemblies.
3.
The assemblies and associated rod storage positions shall be separated from each other and any other fissile material by a minimum surface-to-surface separation of 1-foot.
4.
Fissile material and fuel rods or components in the associ-ated storage positions shall each be restricted to a square not exceeding the dimensions of a fresh fuel assembly or to a cross sectional area not exceeding that of a 22.5 cm (8.85 in.) otameter cylinder.
5.
Only one fuel rod at a time shall be removed from or inserted into the assembly or the rod storage position. A (v-)
Maximum of 75 rods shall be permitted in the rod storage posi tion. -
6.
A fuel assembly may be completely dismantled by withdrawing one fuel rod at a time from the assembly; during all stages of dismantlement,- the partially dismantled assembly shall be maintained within the confines of a square not exceeding the dimensions of a fresh fuel assembly or to a cross sectional area not exceeding that of a 22.5 cm (8.85 in.) diameter cylinder.
7.
An assembly and its associated rod storage position may be withdrawn from the pool into the cell.
Free water drainage from both the assembly and rod storage position as well as l
1-foot separation from other fissile materials and each other shall be assured.
4.2.3.6.4 Assembly and Machine Shop and Development Test Area -
The work areas on the first floor of Building B may be used to l
License No SNM-778 Docket No.70-824 Date April, 1987 l
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disassemble unirradiated fuel assemblies for testing. The v
following conditions govern operations in the work area:
1.
Only one assembly at a time shall be permitted to be dismantled.
2.
An associated storage position will be permitted for fuel rods which have been removed from the assembly, and shall be spaced and stored as stated in items 3 and 4 (4.2.3.6.4) below.
3.
The assembly and associated rod storage position shall be separated from each other and any other fissile material by l
a minimum surface-to-surface separation of 21-inches.
4.
The associated rod storage position shall be no larger in any dimension than the fuel assembly. There shall be one such storage position for each assembly to be dismantled.
Rods may be stored or handled in a slab up to 4-inches thick provided the slab is separated from other fissile material by a minimum of 12-feet.
5.
Only one fuel rod at a time may be removed from or inserted into the assembly or any rod storage position.
Only one rod may be in transit to any one location at a time.
6.
The fuel assembly may be completely disassembled by with-drawing one fuel rod at a time from the assembly; during all stages of disassembly, the partially disassembled assembly shall be maintained within the confines of the assembly whether damaged or undamaged.
7.
Fuel rods may be removed one at a time from this area as required. These rods shall be subject to all fuel handling requirements pertinent to the area they are in.
8.
Assemblies may be handled and dismantled under water in these areas (mock-up pool and development test area pool) subject to the same requirements of the hot cell pool.
4.2.3.6.5 Hot Cell Operations - Fuel rods removed from irradiated assemblies may be examined including destructive examination in the hot cells. Operations in the hot cells shall be governed according to the following conditions:
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1.
An assembly and its associated rod storage position may be withdrawn from the pool into Hot Cell No.1 pursuant to Item No. 7 of Section 4.2.3.6.3 of this Part.
2.
Two units in Hot Cell No. I may have a total of 64 fuel rods each, stored, provided that rods shall be confined within a cross sectional area not exceeding that of a 22.5 cm (8.85 in.) diameter cylinder, drainage of any free water within the unit shall be assured and the units must be maintained 1-foot from each other and any other SNM in the cell.
3.
In addition to the two units of stored rods, another unit limited to the values in 4.2.2.2.1 may be present in Hot Cell No. 1.
In this unit under mass control, rods may be destructively examined.
4.2.3.6.6 Fuel Rod Dismantlement - Fuel rods from unirradiated assemblies can be dismantled in any area where the license permits handling of unirradiated fuel. The following conditions must also be met in areas to dismantle fuel rods:
1.
The area shall be mass limited to 350 grams of.U-235. This area must be separated from the assembly and slab storage area by minimum of 12-feet.
2.
Dismantlement must be completed under approved procedures.
4.2.3.6.7 Shipment and Disposal - Af ter examination, assemblies, partially dismantled assemblies, fuel rods, and scrap generated during destructive examination shall be disposed of according to the following conditions.
1.
Fuel rods, including fuel rod segments may be placed in any available hole in a fuel assembly, including the instrument and control rod guide tube positions, i.e., 225 and 285 fuel rods in Mark B and Mark C assemblies, respectively. Fuel rod segments shall have their ends sealed, and shall be encapsulated in steel tubing with ends sealed, prior to insertion into an available hole in a fuel assembly.
2.
Unir.adiated assemblies may be reassembled (one rod at a time) for later use.
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3.
Assemblies, including partially dismantled assemblies, shall be loaded into shipping casks approved for such assemblies for shipment.
4.
Scrap, including rod segments, shall be disposed of according to present LRC procedures and limits.
4.2.4 Building C 4.2.4.1 General - Building C is limited to 90 units. Each unit shall be separated from adjacent units by at least 8-inches edge-to edge and 36-inches center-to-center.
4.2.4.2 Unit Limits - Each unit shall be limited to the values specified in section 4.2.2.2 of this Part.
4.2.5 Outside Storage 4.2.5.1 General - Outside storage consists of underground storage, ship-ments, and the fenced storage area located adjacent to Building J.
4.2.5.2 Underground Storage - Radioactive materials stored in underground storage tubes shall be limited to one SNM unit per tube, with values as specified in section 4.2.2.2 of this Part. Tubes shall be spaced 20-inches center-to-center and are 5-inches in. diameter,
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and totally immersed in concrete.
t 4.2.5.3 Shipments - Each shipment of fissile material being stored out-side must conform with all license requirements for the type of shipping container. Additionally, each shipment must be nuclearly isolated from all other SNM.
4.2.6 Dry Waste - Dry waste is accumulated in 55-gallon drums, or other suitable containers, with a maximum of 45 grams of SNM per con-tainer. These containers may be located throughout the labora-tories as required to collect contaminated laboratory waste. Filled containers are transferred, to the radioactive waste storage building after scanning. Dry waste containing 0.5 grams of SNM or less per container may be stored in a fenced, locked and paved outside storage area adjacent to Building J.
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5.0 ENVIRONMENTAL PROTECTION 5.1 EFFLUENT CONTROL SYSTEM 5.1.1 Responsibility - The Supervisor, Health and Safety is responsible for the Effluent Control System.
5.1.2 Solid Waste - Solid radioactive waste, including solidified liquid wastes, shall be sent off site to a licensed disposal facility.
5.1.3 Liquid Waste - Low-level, liquid radioactive waste is discharged from process areas to the Liquid Waste Disposal Facility. This facility is comprised of several tanks where the waste is accumu-lated for eventual transfer to the B & W Naval Nuclear Fuel Division (NNFD) for ultimate release to the James River.
5.1.3.1 Prior to release to NNFO the contents of a tank shall be mixed and sampled.
5.1.3.2 The contents of liquid waste tanks shall not be released to the NNFD unless the concentration of radioactivity is less than 25% of the MPC values of Table I, Col. 2, of 10 CFR 20, Appendix B.
The limiting values-in water shall be determined in accordance with
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the note at the end of Appendix B,10 CFR 20.
5.1.3.3 Process liquid wastes may be collected and stored indoors. These wastes may be solidified and handled as dry waste.
5.1.3.4 Storage tanks in the Liquid Waste Disposal Facility shall be inspected visually upon each tank voiding or annually, whichever is sooner, to assure that there is no unsafe accumulation of Special Nuclear Material. Storage tanks that have not been used during a year will not be inspected.
5.1.3.5 Samples of liquid waste are grab sampled. A small portion of the sample is pipetted into a planchet and brought to dryness. This planchet is counted on a low background counter, either Low Beta or Wide Beta, and the waste activity concentration is calculated.
The samples are counted for gross alpha and beta. Waste tanks that may receive Sr-90 waste will have their samples analyzed for Sr-90.
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5.1.3.6 Waste tanks that indicate concentrations of activity greater than those specified in section 5.1.3.2 shall be appropriately diluted prior to release.
5.1.3.7 The NNFD must approve the release of liquid waste to their waste treatment facility prior to the release.
5.1.3.8 The 10,000 sq. ft. Storm Drain Collection Pond shall be grab sampled quarterly. The sample shall be analyzed for gross alpha and gross beta.
5.1.4 Gaseous Effluent - Discharge air from process areas is released to the general environment through the 50-meter high stack. The discharge rate of the stack is approximately 20,000 cubic feet per minute. The annual discharge volume is 1.1E10 cubic feet. Planned discharges to the air shall be in compliance with 40 CFR 61. The annual exposure resulting from these planned discharges shall not exceed 25 millirem whole body and 75 millirem to any organ.
5.1.4.1 Action levels - The action levels for releases from the stack are specified in section 3.2.2.5 of this Part I.
5.1.4.2 Analyses - The fixed filter of the stack particulate monitor shall be counted on the Low Beta or Wide Beta counting system, after an fm appropriate decay period. -The results shall be recorded and
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maintained on file.
5.1.4.3 Sampling - The stack shall be sampled isokinetically on a continu-ous basis.
5.1.4.4 Monitoring - The stack sample shall be passed through a monitoring system that consists of the following:
1.
Particulate Monitor - The stack particulate monitor consists of an alpha and beta channel, with a dual channel ratio de-tector. This monitor uses a fixed filter and a nominal sampling flow rate of 2 - 3 cubic jeet per minute.
The de-tector is a thin window (1.0 mg/cm ) gas flow proportional detector. Alpha and Beta-gamma radiations are monitored through two single channel analyzers and log rate meters. The ratio between these two channels is also displayed as a log-rhythmic ratio. This system effectively compensates for the presence of Radon and Thoron daughters and increases the sen-sitivity of the system.
Alpha and Beta-gamma sensitivities License No SNM 778 Docket No.70-824 Date April, 1987 Amendment No.
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V are himilar for both channels. Alarm settings, based on the ratio system, are sufficient to alarm at or below short term stack concentrations that are specified in section 3.2.2.5 of.
this Part I and that which would result in concentrations in unrestricted areas exceeding 10 times the applicable limits given in 10 CFR 20 for the nuclides in use at the site.
l 2.
Gas Monitor - The stack gas monitoring system consists of a 2
shielded chamber with one or two GM tube detectors (30 mg/cm stainless steel). The Beta-gamma count rate is directly proportional to the stack concentration and system sensitivity is approximately 3E-9 uCi/ml per CPM for Kr-85. A con-ventional alarming and recording log ratemeter is used to monitor the gas channel. The alarm level is set to activate below the level representing 70 Curies / week of Kr-85.
5.2 ENVIRONMENTAL MONITORING 5.2.1 The environment surrounding the site and the Mount Athos plant site is sampled periodically to determine whether the radiation and radioactive material levels in the area surrounding the site have changed as a result of the operations at this location.
5.2.2 The-following types of samples shall be -taken at the below indicated frequencies:
g 1.
Site boundary air sample - monthly.
2.
Grab sample of the James River above and below the point of discharge - monthly.
3.
Continuous sampling of rain water.
4.
Grab sample of river silt - quarterly.
5.
Direct radiation survey shall be made of the water channel passing through the railroad right-of-way - annually.
6.
Direct radiation survey shall be made at the east end of the canal - annually.
I 7.
Vegetation sample - semiannually.
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Direct radiation monitoring at the site boundary -
continuously.
9.
Accumulated water from the soil retention basin shall be sampled and if its activity exceeds 10% of the concentration specified in Appendix B, Table 2, column 2, of 10 CFR 20, the collected water shall be disposed of through the liquid waste disposal system - annually.
5.2.3 The evaluation of environmental sampling shall be performed by either the site personnel or a qualified outside concern.
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6.0 SPECIAL PROCESS COMMITMENTS The site is engaged in research and development and for this reason I
there are a large number of small special processes that are special only because they are outside of the few repetitive operations.
Ex-
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amples of operations performed at the site are given in section 1.7.1.
The site relies on established administrative controls to determine what proposals fall outside of the bounds of work that is authorized by the license, in which case amendments are applied for. Those proposals that are authorized by the license but are significantly different from previously reviewed proposals shall be brought before the Safety Review Committee for review and approval.
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License No SNM-778 Docket No.70-824 Date April, 1937 Amendment No.
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- [']j 7.0 DECOMMISSIONING PLAN u.
The site is committed to decommissioning the facilities which have l
been used for the use and storage of licensed material at the end of their useful life. At the time of this application for renewal of License No. SNM-778, two programs are underway to decommission Buildings A and C.
It is presently estimated that ther,e two facili-ties will be decontaminated and ready for release for unrestricted use by March, 1987.
7.1 PLANNING CONSIDERATIONS 7.1.1 The history of the facility shall be determined to facilitate the identification of services, equipment, and areas that should be included in the survey plan.
7.1.2 The decontamination of facilities and equipment must meet the levels of contamination specified in Table 1, Annex C to License SNM-778, Guidelines for Decontamination of Facilities and Equipment Prior to Release for Unrestricted Use or Termination of Licenses for Byproduct, Source, or Special Nuclear Material, dated November, 1976.
In addition, a reasonable effort will be made to further reduce contamination levels to those which are as low as reasonably achievable.
7.1.3 No covering will be applied to remaining surfaces until it has been determined that contamination levels are below those of Table 1, Annex C to License SNM-778, Guidelines for Decontamination of Facilities and Equipment Price to Release for Unrestricted Use or Termination of Licenses for Byproduct Source, or Special Nuclear Material, dated November,1976, and until it has been determined that a reasonable effort has been made to further reduce contamination below those specified above.
7.1.4 The radioactive contamination of interior surfaces of pipes, ductwork, and other conduits will be determined by taking measurements at all traps and other appropriate access points, provided contamination at these locations is likely to be representative of interior conditions.
If such access locations are not likely to be representative, or if interior surfaces are inaccessible, the interior surfaces will be assumed to be contaminated in excess of levels specified in Table 1, Annex C to License SNM-778, Guidelines for Decontamination of Facilities and License No SNM-778 Docket No.70-824 Date April, 1987 Amendment No.
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8.0 RADIOLOGICAL CONTINGENCY PLAN
'v The site shall maintain and execute the response measures of the l
Radiological Contingency Plan submitted to the NRC on November 15, 1983, in accordance with provisions of the February 11, 1981 order.
The site shall maintain implementing procedures for the Radiological l
Contingency Plan as necessary to implement the Plan. The site shall l
make no change in the Radiological Contingency Plan that would decrease the response effectiveness of the Plan without NRC approval.
The site may make changes to the Radiological Contingency Plan without l
prior NRC approval if the changes do r.ot decrease the response effec-tiveness of the Plan. The site shall maintain records of changes that l
were made to the Plan without prior approval for a period of two years-from the date of change and shall furnish the Chief, Uranium Fuel Licensing Branch, Division of Fuel Cycle and Material Safety, NMSS, USNRC, Washington, DC 20555 and Region II, a report containing a des-cription of each change within six months after the change is made.
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U,G TABLE OF CONTENTS Section Page 9.0 OVERVIEW 0F OPERATION.
9-1 9.1 CORPORATE INFORMATION.
9-1 9.2 FINANCIAL QUALIFICATIONS.
9-3 9.3
SUMMARY
OF OPERATING OBJECTIVE AND PROCESS.
9-3 9.4 SITE DESCRIPTION 9-3 9.5 LOCATION OF SITE BUILDINGS 9-4 9.6 LICENSE HISTORY.
9-4 9.7 CHANGES IN PROCEDURES, FACILITIES, AND EQUIPMENT.
9-5 List of Figures i
u Figure Page 9-1 SITE LOCATION IN VIRGINIA 9-7 l
9-2 FIVE MILE RADIUS OF SITE.
9-8 l
9-3 SITE BUILDING LAYOUT 9-9 l
9-4 FACILITY WORK ORDER FORM.
9-10 License No SNM 778 Docket No. 70 824 DateApril, 1987 Amendment No.
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PART II SAFETY DEMONSTRATION 9.0 OVERVIEW 0F OPERATION 9.1 CORPORATE INFORMATION
9.1.1 LICENSEE
MCDERMOTT INTERNATIONAL, INC.
BABC0CK & WILCOX NAVAL NUCLEAR FUEL DIVISION NNFD RESEARCH LABORATORY
9.1.2 Address
Babcock & Wilcox Naval Nuclear Fuel Division NNFD Research Laboratory P. O. Box 11165 Lynchburg, Virginia 24506-1165 9.1.3 Principal Offices:
McDermott International, Inc.
Babcock & Wilcox 1010 Common Street P. O. Box 60035 New Orleans, Louisiana 70160 9.1.4 Principal Officers:
J. E. Cunningham Chairman of the Board &
Chief Executive Officer 1010 Common Street P. O. Box 60035 New Orleans, Louisiana 70160 U. S. Citizen License No SNM 778 Docket No.70-824 Date April, 1987 Amendment No.
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Robert E. Howson
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President & Chief Operating Officer 1010 Common Street P. O. Box 60035 New Orleans, Louisiana 70160 U. S. Citizen John A. Lynott Executive Vice President Chief Financial Officer 1010 Common Street P. O. Box 60035 New Orleans, Louisiana 70160 U. S. Citizen Robert E. Howson President and Chief Operating Officer Babcock & Wilcox 1010 Common Street P. O. Box 60035 New Orleans, Louisiana 70160 U. S. Citizen 9.1.5 State of Incorporation:
Deltware 9.1.6 Alien or Foreign Control:
Babcock & Wilcox is incorporated under the laws of the State of Delaware.
In 1978 McDermott Inc., a Delaware corporation, acquired Babcock & Wilcox.
In 1983 McDermott International, Inc., incorpo-rated under the laws of the Republic of Panama, became the parent company of the McDermott group of companies. This reorganization was reviewed by the NRC prior to its implementation.
In a letter dated December 17, 1982, to Mr. J. H. MacMillan, William Dirks summarized the Commission's finding that the change was not inimical to the common defense and security or the health and safety of the public. Based on this conclusion, the change was approved under Section 104(b) of the Atomic Energy Act as it related to the operation of a critical experiment reactor owned by Babcock & Wilcox at the site.
Such approval was not required under j
the Act for material licensees.
The organization of McDermott International, Inc. has not changed since that action.
License No SNM 778 Docket No. 70 824 Date April, 1987 Amendment No.
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/^s ij 9.2 FINANCIAL QUALIFICATIONS 9.2.1 The financial qualifications of the Corporation to continue opera-tions at the site and to perform the necessary decommissioning at l
the end of plant life is demonstrated in the latest (1984) copy of the Corporation's Annual Report, which is enclosed with this appli-ca tion.
9.3
SUMMARY
OF OPERATING OBJECTIVE AND PROCESS 9.3.1 Research and development activities, utilizing licensed material, are conducted at the site in support of the operating divisions of l
Babcock & Wilcox and for other companies and government organt-za tions. The broad range of projects that have been conducted pursuant to the license cannot be described in terms of through-put or any single process.
Radioactive materials are handled and stored, principally in Building 8.
That building houses the Hot Cells, Radiochemistry Laboratory, Scanning Electron Microscopy Laboratory, Metallurgy Laboratories, Analytical Chemistry Labora-tory, Fatigue and Fracture Laboratory, Failure Analysis Laboratory, Crane and Cask Handling Area, a Hot Machine Shop, the Counting Room, and a Health Physics Laboratory.
Licensed material in the form of liquid waste is collected in tanks O
that are located in the Liquid Waste Disposal Facility. Solid V
radioactive waste is stored in Building J, the Annex to Building J, and the storage area adjacent to Building J.
9.4 ' SITE DESCRIPTION The site is located on the James River about four miles east of Lynch-l burg, Virginia. The site, which comprises 525 acres, lies within Campbell County and borders on Amherst County. The site occupies about 13.6 acres of the site. The location of the site within the l
Commonwealth of Virginia is shown in Figure 9-1.
The irregularly shaped property is bounded on three sides by a large loop of the James River and on the fourth side by State Route 726, which closely follows the base of Mount Athos.
This mountain rises License No SNM 778 Docket No. 70 824 Date April, 1987 Amendment No.
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rapidly from about 500 feet MSL to 900 feet MSL, making it the dominant feature of the surrounding landscape. The Babcock & Wilcox property consists of large sections of relatively flat floodplain along the James River lying at about 470 feet MSL. The interior of the property is largely composed of rolling hills, one of which rises to almost 700 feet MSL. The property boundary and topography within about two miles of the site are shown in Figure 9-2.
l The land in the immediate vicinity of the plant is sparsely inha-bi ted. The severe topography makes it unsuitable for commercial farming. The Lynchburg Foundry, a producer of linht metal castings, occupies a parcel of land which abuts the south bour.dary of the Babcock & Wilcox property. The Foundry is approximately.5 miles from the site.
The site is sr.eviced by a spur of the Chessie System Railroad which l
runs through the Babcock & Wilcox property. The property is also conveniently located for truck and automobile access. About three miles from the site, State Route 726 connects with U.S. Highway 460, a l
major link between Roanoke and Richmond. The site is located about l
100 feet above the James River and for that reason no dams on the river would threaten the site should they fail.
I 9.5 LOCATION OF SITE BUILDINGS V
9.5.1 Figure 9-3 shows the layout of buildings at the tite. All buildings l
are of masonry construction.
9.6 LICENSE HISTORY 9.6.1 License SNM-778 was issued on September 16, 1966.
Since that time the following renewals and amendments have been approved:
February 15, 1974
.First renewal July 21, 1980 Second renewal August 28, 1981 Amendment No.1, approved a change in the organization.
February 25, 1982 Amendment No. 2, approsed the Radiological Contingency Plan License No SNM 778 Docket No.70-824 Date April, 1987 Amendment No.
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4 Page 9-4 v
Babcock & Wilcox a McDermott company
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/~N FIGURE 9-1 k ',
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3e IIO* l1 2
D A
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5 E SITE 4
13 6*
7, O.
50 10 0 qQ SCALE IN MILES
- 1. ARLINGTON (174,2543 A, JAMES RtVER
- 2. ALIXANDRI A (110,9383 8.COWPA5fbRE R1VIR 1 FREDERICK 58URGild,45(8 C. JACK 50N RIVER
- 4. MWPORINEWS (141TD
- 0. AMHERST COUNTY 426,072)
- 5. ROANOKE(92.1158 E. BEDFORD COUNTY 126 T288 L MARilN5VIL11(19,65h F. CAMP 8tLL COUNTY 143,3191
- 7. DANVILLE146,191)
G. APPOMATT0X COUNTY 19,184)
L LYNCit9URG (54,0th
- 9. STAUNTON(25,504) 1(L WAYNES80R0(16,70D (NUMBERS IN PARINih15(5 (l ARE 1970 CENSUS DATA)
LL CHARLOTT15VittiI48401
- 12. RtCHMOND1249.62D 11 NORFolx (307.951)
License No SNM 778 Docket No.70-824 Date April, 1987 Amendment No.
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License No SNM-778 Docket No.70-824 Date April, 1987 Amendment No.
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Babcock &Wilcox a McDermott company
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License No SNM 778 Docket No. 70 824 Date April, 1987 Amendment No.
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Page 9-9 Babcock &Wilcox a McDermott company
_ _ _ _ _ _. _ _ _. _ _ _ _ _ _... ~... _ _ _ _ _
O TABLE OF CONTENTS V
Section Page
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10.0 FACILITY DESCRIPTION 10-1 10.1 PLANT LAYOUT.
10-1 10.2 UTILITIES INCLUDING EMERGENCY POWER 10-1 10.2.1 Potable Water 10-1 10.2.2 Process Water 10-1 10.2.3 Ga s.
10-1 10-1 10.2.4 Fuel Oil 10.2.5 Electricity 10-1 10.3 HEATING, VENTILATION, AND AIR CONDITIONING.
10.2 10.3.1 Heating 10-2 10.3.2 Ventilation 10-2 10.4 WASTE HANDLING 10-4 10.4.1 Liquid Wastes 10-4 10.4.1 Solid Wastes.
10-5 10.5 FIRE PROTECTION.
10-6 10.5.1 Codes and Standards 10-6 10.5.2 Insurance Inspection Reports 10-8 l
M.S.3 Fire Protection Equipment 10-8 l
l 10.5.4 Combustible Waste Storage 10-9 License No SNM 773 Docket No.70-824 Date April,1987 4
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10.0 FACILITY DESCRIPTION V
10.1 PLANT LAYOUT Figures 10-1 through 10-3 show the layout of the site buildings.
l 10.2 UTILITIES INCLUDING EMERGENCY POWER 10.2.1 Potable Water Potable water is provided to the site by the NNFD.
It is pumped l
from wells.
It is stored and treated at the NNFD and is gravity fed to the site.
l 10.2.2 Process Water Process water is provided to the site by the NNFD. The source of I
process water is the James River.
It is pumped from the river, filtered by the NNFD and is gravity fed to the site. There is a l
storage capacity of 6,000,000 gallcns on site.
Process water is also used for fire fighting.
10.2.3 Gas Natural gas is supplied at the site via pipeline which enters the l
g B & W property on the western side of the site.
Natural gas is V;
used for space heating, fuel for emergency engines, and laboratory uses. This system is provided with a backup source of propane gas which is stored on site.
10.2.4 Fuel Oil Fuel oil is available for space heating to provide a backup source in the event of curtailed availability of natural gas.
Fuel oil is purchased locally and stored on the site site. There is a storage l
capacity of approximately 24,000 gallons.
10.2.5 Electricity Electricity is furnished to the site from a substation located on l
the west side of the site. This source provides the normal source of power for the stack fans, hot cell fans, criticality monitors, emergency evacuation alarm, and lighting. When normal site power is lost, emergency sources are provided for these loads in the following manner.
License No SNM 778 Docket No.70-824 Date April, 1987 Amendment No.
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OU of being locked and provided with side panels which permit the roof to fit flush with the top of the block walls. Containers are loaded into the Annex from the top. A curbing will be placed on the approach side of the addition to prevent a loading vehicle from accidentally contacting the wall. Two individuals are involved in loading containers into this facility to prevent a container from striking the walls. This facility provides storage of waste that is contaminated with irradiated fuel and is t
being stored on site until it is accepted by the 00E under the Nuclear Waste Policy Act of 1982. The maximum qantity of SNM per container shall be limited to 45 grams.
10.4.2.5 The Outside Waste Storage Area is located adjacent to Building J.
This area is fenced, locked and paved. Waste stored in this area is limited to that contained in closed metal containers. Each container is limited to not more than a Type A quantity (10 CFR 71.4) or 0.5 grams of fissile material or both. Pu shall not be stored in this area. Containment of tared waste is assured by a quarterly visual inspection by the Sups" visor, Health and Safety.
10.4.2.6 The High Level W1ste Storage Tubes are located adjacent to the south side of the Liquid Waste Disposal Facility. These tubes are constructed of two sections of iron pipe, immersed in concrete, and below ground level. The upper section of pipe p
(approximately 42-inches long) is 6-inches-in diameter. The Q
lower section (approximately 80-inches long) is welded to the upper section and is 5-inches in diameter.
Each tube is fitted with a concrete-filled iron plug.
These tubes are locked and under the direct control of the Health and Safety Group. Waste stored in these tubes is limited to that which is produced in the Hot Cells and must be in closed metal containers. The quantity of fissile material permitted in each tube is limited to one unit.
10.5 FIRE PROTECTION 10.5.1 Codes and Standards - The development and building construction program of the site has taken place over the period 1955 to the l
present. For the three main buildings under consideration in this renewal request, the design and construction efforts took place from 1955 to 1969. There have been a number of alterations and use changes over the past ten years, but generally these changes have not significantly altered the structural characteristics of the buildings.
License No SNM 778 Docket No. 70 824 Date April, 1987 Amendment No.
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All three buildings were built as staged or "added-on" phased construction. Building A was built in four distinct phases,
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Building B was built in two stages, and Building C was constructed in five phases. Building A was designed in-house by B&W engi-neering personnel. Building B was designed by Wiley & Wilson, a l
Lynchburg consulting engineering firm.
Building C was also l
primarily designed by Wiley & Wilson, with some design by B&W.
l The physical layout of all three buildings is highly functional, i.e., based on the specialized requirements of research work related to the nuclear industry.
For the most part, the building structure envelopes are quite conventional in nature, both from a b
design and construction materials standpoint.
With the exception of highly specialized portions of these buildings, such as the hot cells, engineering design of the buildings would be considered as state-of-the-art fnr light industrial / heavy commercial class buildings (for each of the design and construction time periods involved).
The overall quality of the building construction is well above average. Aside from some roof leakage problems and minor 1
settlement cracking in some of the masonry construction, the per-
)
formance of the building structures and envelopes has been good.
There have been no repairs related to significant structural defects in any of the three buildings. As would be anticipated for O
a complex of this type and importance, maintenance of the buildings has been excellent and contributes to the overall good condition of such a facility.
During the period of design and construction for Buildings A, B, and C, it should be noted that there was very little in the way of code requirements or guidance for construction of such a facility.
Virginia did not adopt a state-wide building code until September 1, 1973.
Up until that time, various localities in the state had adopted their own local building codes; the Southern Building Code being the one generally used. Many counties, however, had no code at all; Campbell County, in which the site is located, had no l
building code during this time period. The only state-wide code directly applicable to building construction prior to 1973, was the Virginia Fire Safety Regulations, enacted in 1949.
The lack of a state-wide building code should not be taken as implication that the design and construction of the site facilities l
were accomplished on an inferior basis. On the contrary, where good accepted engineering and construction practices, coupled with License No SNM 778 Docket No.70-824 Date April, 1987 Amendment No.
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V stringent requirements from insuring companies such as Factory Mutual, are used as the main criteria for such facilities, the resulting structures usually far exceed the minimum requirements of various building codes.
Such is typically the case for all three buildings under consideration, when examined from the load capacity standpoint as specified in the present Virginia building code, B0CA (Building Officials and Code Administrators) 1978, Seventh Edition.
Conventional construction materials are used throughout all three buildings.
Structural steel yield strength varies from 33,000 PSI i
(ASTM A7 steel) for the 1955 construction to 36,000 PSI ( ASTM A36 steel) for the 1969 construction. Concrete strengths vary from 3000 PSI for conventional cast-in-place concrete construction to 5000 PSI for the precast prestressed concrete elements found in Building 8. Concrete reinforcing steel is typically ASTM A615, Grade 40 (40,000 PSI yield strength). Working stress design was used as the basis for concrete and steel design for all structures on site. Applicable design criteria used for the facility includes l
the standards of the American Concrete Institute ( ACI), the Pre-stressed Concrete Institute (PCI), and the Americ.an Institute of Steel Construction ( AISC). These various standar c's serve as both design and code basis for the respective types or construction, both at the time of original design as well as the present.
q 10.5.2 Insurance Inspection Reports - The site is inspected twice annually l
Q by the Arkwright-Boston Insurance Company on behalf of the Mutual Atomic Energy Reinsurance Pool (MAERP).
The inspection reports If st the following items in each report; housekeeping, mainte-nance & repair, Supervision fire equipment, watchmen, radioisotope handling, areas sprinklered, water supply, all valves found open, criticality control, and until the decommissioning of the last reactor, nuclear reactor operation.
These reports have con-sistently found that the site meets the requirements in each category for a " satisfactory" rating. On a few occasions there have been recommendations that the site add fire protection equipment when the use of an area has been changeJ. Each such recommendation has been addressed at the site in a manner that has been found acceptable to the inspectors upon their reinspection.
The reports on which the above statement is based are dated from 1977 through 1985.
10.5.3 Fire protection equipment is installed in response to recommen-dations made by the Industrial Safety Officer, the Corporate Fire Protection Engineer, or the insurance underwriters.
Installed systems are approved and inspected by Factory Mutual Engineering License No SNM 778 Docket No.70-824 Date April, 1987 Amendment Ho-0 Revision No.
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Ob Association. Routine inspection and maintenance is described below:
EQUIPMENT MAINTENANCE RESPONSIBILITY REFERENCE Portable fire Insp./ test Industrial Safety NFPA 10 extinguishers FM 4-5 Fire hoses Insp./ test Industrial Safety NFPA 10 Sprinklers Test Plant Engineering NFPA 13 FM 4-5 Fire suppres.
Inspection Plant Engineering NFPA 12-A systems (Halon)
FM 4-8N Mfg.
Smoke det.
Test Plant Engineering Mfg.
Heat det.
Test Plant Engineering Mfg.
Housekeeping Inspection Health and Safety Emer, equip.
Inspection Health and Safety Mfg.
10.5.4 Combustible Waste Storage - Combustibles are not routinely stored at the site except; when work requiring such materials is in l
progress, in containers for shipping and receiving, or in sprinklered areas. Combustible wastes are discarded in metal containers and disposed of by an off site disposal firm.
Contami-nated combustible waste is discarded in metal containers and shipped to an off-site licensed disposal facility.
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TABLE OF CONTENTS Section Page 11.0 ORGANIZATION AND PERSONNEL 11-1 11.1 SITE LINE ORGANIZATION 11-1 11.1.1 Manager, Employee, Conununity, and Regulatory Relations.
11-1 11.1.2 Facility Supervisor 11-1 11.1.3 Area Supervisors 11-1 11.1.4 Manager, Safety and Licensing 11-1 11.1.5 Supervisor, Health and Safety 11-1 11.1.6 Senior Health Physics Engineer.
11-2 11.1.7 Industrial Safety Officer 11-2 11.1.8 Accountability Specialist 11-3 11.1.9 License Administrator.
11-3 11.1.10 Nuclear Criticality Safety Officer 11-4 11.1.11 Facility Supervisor 11-4 11.1.12 Safety Review Committee 11-5 11.2 EDUCATION AND EXPERIENCE OF KEY PERSONNEL 11-6 l
11.2.1 Safety and Licensing Manager 11-6 11.2.2 Health and Safety Supervisor 11-7 l
11.2.3 Senior Health Physics Engineer.
11-9 f
11.2.4 Industrial Safety Officer 11-10 11.2.5 Accountability Specialist 11-11 License No SNM 778 Docket No.70-824 Date April 1987 O
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Section Page 11.2.6 License Administrator.
11-13 Facility Supervisor 11-13 11.2.7 Nuclear Criticality Safety Officer 11-14 11.3 PROCEDURES 11-16 11.3.1 Area Operating Procedures (A0P) 11-16 11.3.2 Technical Procedures 11-17 I
11.4 TRAINING 11-17 11.4.1 General Radiation Protection Training 11-17 11.4.2 Program 1.
11-18 11.4.3 Program 2.
11-18 11.4.4 Program 3.
11-18 11.4.5 Respiratory Protection Training 11-18 11.5 FACILITY CHANGE.
11-21 List of Figures i
Figure Page 11-1 LINE ORGANIZATION 11-21 11-2 FACILITY WORK ORDER FORM.
11-22 License No SNM 778 Docket No. 70 824 Date April 1987 O
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11.0 ORGANIZATION AND PERSONNEL 11.1 SITE ORGANIZATION 11.1.1 The Manager, Employee, Community, and Regulatory Relations (Manager, EC&RR) - The Manager, EC&RR is responsible for all site operations.
The Manager, Safety and Licensing, Manager, Facilities, and the Facility Supervisor report to him.
11.1.2 Facility Supervisor - Research and development work at the site will be performed by personnel who do not report to the Manager, EC&RR. Therefore, the positions of Facility Supervisor and Area l
Supervisor have been established to control the workers and their activities.
The Facility Supervisor shall report to the Manager, EC&RR. He shall be responsible for the safety of all operations performed pursuant to License SNM-778. He shall utilize the expertise of the Supervisor, Health and Safety, the Accountability Specialist, Nuclear Criticality Safety Officer, and the Industrial Safety Officer to ensure the safety of operations.
11.1.3 Area Supervisors - Area Supervisors are selected by their Division Management and shall be jointly approved by the Facility Supervisor (p
and the Supervisor, Health and Safety. Area Supervisors function-
'j ally report to the Facility Supervisor and are responsible for the safe performance of all activities in their assigned area and that all activities within their assigned areas are performed in full compliance with the license.
11.1.4 Manager, Safety and Licensing - The Manager of Safety and Licensing is appointed by and reports to the Manager, EC&RR.
He is respon-l sible for the proper management of the materials accounting function, licensing function, nuclear criticality safety function, and the Health and Safety Group. He manages the allotment of funds and other resources and assures the proper assignment of personnel priorities. The Supervisor, Health and Safety, Accountability Specialist, Nuclear Criticality Safety Officer, and License Admin-l istrator, report to him.
11.1.5 Supervisor, Health and Safety - The Supervisor of Health and Safety is appointed by the Manager, EC&RR and reports to the Manager, Safety and Licensing. The Supervisor directs the overall operation of the Health and Safety Group and the Industrial Safety Officer.
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'N He also serves on the Safety Review Committee. He has the authority (h
to stop any operation that he believes is contrary to accepted safety practices, or license requirements. The Supervisor has over-all responsibility for the shipment and receipt of licensed material and exercises signature authority on all Area Operating Procedures.
He performs audf ts of the site for compliance with Health and Safety rules. The Seniar Health Physics Engineer and Industrial Safety Officer report to him.
11.1.6 Senior Health Physics Engineer - A Senior Health Physics Engineer l
reports to the Supervisor, Health and Safety.
He administers the activities of the Health Physics staff, which include:
1.
Performing area surveys 2.
Administering the air sampling program 3.
Administering the respiratory protection program 4.
Administering the bioassay program 5.
Leak testing radioactive sources 6.
Supervising shipping and receiving of licensed material 7.
Supervising and coordinating the waste disposal program b,)
'V 8.
Assisting in personnel, equipment, and facility decontamination 9.
Conducting radiation safety training m
10.
Providing expertise in all aspects of radiation protection 11.
Generating, maintaining and distributing records and reports that are required by NRC regulations or Health Physics procedur.ss 12.
Providing expertise in health physics to the Facility 4
Supervisor.
11.1.7 Industrial Safety Officar - The Industrial Safety Officer reports to the Supervisor, Health.and Safety. His responsibilities include the following:
1.
Administering the' industrial safety program k:
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2.
Reviewing proposed facility changes to ensure fire safety V
3.
Providing expertise in fire prevention to the Facility Supervisor and the Safety Review Committee 4.
Performing tests, maintenance, and inspection of fire protection, control, and extinguishing equipment 5.
Providing training for the site Fire and Rescue Team and off l
site support agencies 6.
Inspecting all areas of the site periodically to ensure:
a.
Proper storage and use of flaunable solvents b.
Proper placement of fire extinguishing equipment c.
Elimination of fire hazards d.
Reduction, to the extent practicable, of the accumulation of flammable materials e.
Proper use and maintenance of electrical equipment.
7.
Working with Area Supervisors to formulate safety rules and elimination of hazards 8.
Investigation of all personnel injuries 9.
Keeping management informed concerning industrial safety activities
- 10. Conducting industrial safety training.
11.1.8 Accountability Specialist - The Accountability Specialist reports to the Manager, Safety and Licensing. He is responsible for the maintenance and retention of SNM accountability records. He prepares and transmits the reports required by regulation to inform regulatory agencies of SNM transactions.
11.1.9 License Administrator - The License Administrator reports to the Manager, Safety and Licensing. The License Administrator is responsible for administering the license.
He is the primary liaison between the site and the NRC and other federal, state, and local agencies regarding nuclear matters. He is the coordinator of License No SNM-778 Docket No.70-824 Date April, 1987 Amondment No.
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the Safety Review Committee and Chairman of the Safety Audit Sub-committee and represents site management on both. The License Administrator is responsible for ensuring that corrective action is taken in response to audit findings as they per tain to licensed activities.
11.1.10 Nuclear Criticality Safety Officer - The Nuclear Criticality Safety Officer is appointed by and reports to the Manager, Safety and Licensing. The Nuclear Criticality Safety Officer is responsible for ensuring that no operation at the site can lead to the inad-vertent assembly of a critical mass. To help assure this, he has signature authority for all new Area Operating Procedures and changes to these procedures, he observes operations, institutes educational programs if and when he deems them necessary, and carries out confirming nuclear criticality safety calculations.
The Nuclear Criticality Safety Officer will inspect all rite oper-l ations where special nuclear material is being processed, quarterly.
Other areas may be inspected less frequently, but all licensed fa-cilities will be inspected at least twice a year. He will consider area operations when scheduling these inspections and will, if necessary, schedule his inspection at more frequent intervals.
His consideration should include inspection of new operations, an audit of nuclear safety records, a check for area posting, a review of current practices and a review of corrective actions recommended g
- during previous-audits and-the status of the recommended actions.
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He shall submit a report of his finding to the Manager, EC&RR, with
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a copy to the License Administrator. Prior to the submission of the report, he will discuss its contents with the Facility Supervisor.
The following information is to be included:
1.
Areas visited 2.
Operations observed 3.
Unsafe practices and situations noted 4.
Nuclear safety activity of the quarter 5.
Recommendations.
11.1.11 Facility Supervisor - The Facility Supervisor is appointed by and reports to the Manager, EC&RR.
He shall be responsible to the Manager, EC&RR for the safe conduct of all operations at the site and for ensuring that these operations are conducted in accordance with all license conditions. The Facility Supervisor shall review License No SNM-778 Docket No.70-824 Date April, 1987 Amendment No.
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i and have approval authority for Area Operating Procedures. He shall have authority to terminate any operation that he deems contrary to license conditions, Area Operating Procedures, or general safety conditions. The Facility Supervisor shall become familiar with all license conditions and procedures concerned with radiation safety, nuclear safety, industrial safety, and nuclear materials safeguards.
He may consult with the following personnel to ensure compliance with all safety regulations and principles:
Supervisor, Health and Safety Nuclear Safety Oft'icer Industrial Safety Officer Accountability Specialist 11.1.12 Safety Review Committee - The Safety Review Committee (SF.C) shall be comprised of at least five technically trained and experienced members appointed by the Manager, EC&RR. One member shall be se-1ected by tne Manager, EC&RR to be the SRC Chairman. The Chairman shall preside at the meetings and keep the minutes. The Manager, EC&RR shall appoint an ATternate Chairman who shall act for the Chairman during absences. One member shall be appointed by the Manager, EC&RR to be the SRC Coordinator. The Coordinator shall g
represent site management on the SRC, set the meeting agenda, and l
5 maintains the permanent files of the Committee.
The SRC membership shall have expertise in chemistry, nuclear physics, health physics, and the safe handling of radioactive material. The SRC membership shall have a general understanding of nuclear criticality safety as it pertains to site operations.
j Consultants with special expertise are available to the Committee when needed.
The SRC shall meet at least four times a year. A quorum shall consist of a simple majority of the membership including the Chairman. The SRC shall review and approve all Area Operating Procedures.
It shall review and approve new projects that utilize licensed material that are significantly different from previously reviewed and approved projects.
The SRC shall review the annual report issued by the Supervisor, Health and Safety which summarizes site workers' exposures, environmental releases, and a summary of
[
the ALARA program accomplishments.
The SRC Chairman shall forward the Committee minutes to the Manager, EC&RR, with a copy to the SRC
{
__Coordina tor.
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The Manager, EC&RR shall appoint the members of the Safety Audit Sub-committee (SAS). The SAS shall be comprised of at least two individuals, one of whom shall be designated as Chairman and he shall report to the Chairman, SRC. The SAS shall audit site oper-l ations at least three times annually, with successive audits separated by at least two months.
Additional audits may be performed at any time. The SAS Chairman shall develop the audit report and submit it to the SRC Chairman. The SRC Chairman shall submit the audit report to the Manager, EC&RR with appropriate l
comments, with a copy to the License Administrator.
11.2 EDUCATION AND EXPERIENCE OF KEY PERSONNEL 11.2.1 Safety and Licensing Manager - Richard L. Bennett Education:
B.Ch.E. - Chemical Engineering, University of Delaware,1958 Experience:
(1985-Present)
Babcock & Wilcox, Manager, Safety and Licensing, Lynchburg Research Center, Lynchburg, Virginia.
See Section 11.2.1 (1982-1985)
Babcock & Wilcox, Manager, Building C Decommissioning, Lynchburg Research Center, Lynchburg, Virginia He was responsible for decontaminating facilities that were used for preparation of experimental quantities of nuclear fuels containing plutonium.
(1973-1982)
Babcock & Wilcox, Supervisor, Process Technology Group, Lynchburg Research Center, Lynchburg, Virginia This group was responsible for long-range studies, design assistance, start-up assistance, and preparation of environmental reports and safety analyses related to nuclear fuel conversion.
Some of the specific projects performed by the group were prepa-ration of the designs for a low-enriched nuclear fuel conversion plant, preparation of a conceptual design for a spiked nuclear fuel License No SNM 778 Docket No.70-824 Date April, 1987 Amendment No.
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fabrication plant, process engineering assistance to nuclear fuel conversion plants, development of a halide volatility scrap recovery process, development of alternative effluent treatment systems for various nuclear fuel conversion processes, and evaluation of fabrication methods for advanced fuels.
'(1971-1973)
Babcock & Wilcox, Senior Research Engineer, Lynchburg Research Center, Lynchburg, Virginia He was responsible for the conceptual design of a facility to treat the effluent from a nuclear fuel plant and developing and evaluating processes for recovering byproducts from B&W wastes.
(1959-1971)
American Cyanamid Company, Process Engineer, Piney River, Virginia He has had broad experience in chemical engineering. This includes research and development, designing equipment and processes, testing and operating new equipment, pilot plant operation, process engineering, and economic evaluation. He has specific knowledge in pigment manufacture, effluent treatment, and byproduct recovery.
Professional Affiliations:
American Institute of Chemical Engineers- (Member)
American Nuclear Society (Member) 11.2.2 Supervisor, Health and Safety - Gary S. Hoovler Education:
B.S. - Nuclear Engineering, University of Virginia M.S. - Nuclear Engineering, University of Virginia
- 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> toward Master of Engineering Administration, George Washington University
- Respiratory Protection for Nuclear Industry Experience:
(1986-Present) Babcock & Wilcox, Supervisor, Health and Safety, Lynchburg Research Center, Lynchburg, Virginia Mr. Hoovler is responsible for the Health Physics and Industrial Safety functions at the Lynchburg Research Center; reporting to the Manager, Sa Sty and Licensing.
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ob Mr. Hoovler is responsible for assuring that all radioactive ma-terials at the LRC are properly handled, labeled, and stored. He is responsible for the proper packaging and shipping of radioactive materials, and for radioactive waste disposal.
He is responsible for the proper use, storage and disposal of other hazardous materi-als at the LRC.
Mr. Hoovler is responsible for establishing, maintaining and admin-istering training programs in health physics and industrial safety for employees. He is responsible for reviewing all Area Operating Procedures, Radiation Work Permits, and all Technical Procedures that apply to the Health Physics and Industrial Safety operations.
He is a member of the Safety Review Committee.
1984-1986)
Babcock & Wilcox, Project Manager, Decommissioning, Lynchburg Research Center, Lynchburg, Virginia Mr. Hoovler was the Project Manager for the Decommissioning Program reporting the the Director, LRC.
He developed methods and directed activities for decontamination and survey of Building A and the Critical Experiment Facility, which together compromise an area of 14,000 square feet. He was also responsible for the development of methods and directing the m
activities for the decontamination and survey of the 20,000' square g
foot Plutonium Development Laboratory. The objective of both pro-jects was for their release for unrestricted use and designation as non-use areas for NRC-licensed materials.
(1981-1983)
Babcock & Wilcox, Supervisor, Radiation Experiments Group, Lynchburg Research Center, Lynchburg, Virginia Mr. Hoovler was Supervisor of the Radiation Experiments Group in the Nuclear Physics Section.
He helped to develop the soil assay method used in the plutonium decontamination project, worked on the EPRI Radiation Control Program, and performed experimental work developing radiation gauges for several company applications.
(1976-1981)
Babcock & Wilcox, Research Experiments Group, l
Lynchburg Research Center, Lynchburg, Virginia l
Mr. Hoovier worked in the Reactor Experiments Group of the Nuclear Physics Section. He joined B&W in 1976 as a research eqgineer.
In 1973, af ter receiving his Senior Reactor Operator's Licensc, he was l
l i
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appointed as the CX-10 Operations Supervisor and promoted to Senior Research Engineer in'1980.
He worked on the Department of Energy's Spent Fuel Storage critical experiment program and served on ANS Working Group 15.1.
He worked on the neutron spectrum unfolding code SAND II, eddy current analysis, and on experiments to demonstrate the use of beryllium-gold as a passive technique for measuring fission product distribution in spent fuel.
11.2.3 Senior Health Physics Engineer - Steven W. Schilthelm-
~
Education:
B.S. - Nuclear Engineering, University of Wisconsin, Madison,1983 M.S. - Health Physics, University of Wisconsin, Madison,1985
- Domestic & International Shipping of Radioactive Material.
Experience:
(1985-Present) Babcock & Wilcox, Senior Health Physicist, Lynchburg Research Center, Lynchburg, Virginia
- Mr. Schilthelm is responsible-for administering-the Health Physics p]
Program at the Lynchburg Research Center. His duties include external and internal exposure control, shipping and receiving of radioactive material, maintaining the respiratory protection pro-gram, preparation and presentation of radiological safety training courses, maintaining the support for licensed activities.
Mr. Schilthelm is the Emergency Radiological Safety Officer and is the designated alttrnate for the position of Supervisor, Health and Safe ty.
(1984-1985) Research Specialist, Synchrotron Radiation Center, University of Wisconsin, Madison, Wisconsin Mr. Schilthelm was responsible for Radiation Surveys and subsequent shielding calculations and design at the 800 Mev electron acceler-ator/ storage ring. He co-authored a shielding upgrade proposal that was presented to the National Science Foundation, and he pro-vided the experimental basis for the proposal. Mr. Schilthelm presented a paper at the 1985 Health Physics Society meeting, en-titled " Radiation Survey Measurements at the Aladdin Synchrotron Light Source."
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Professional Affiliation:
American Nuclear Society (Member)
~ Health Physics Society (Member) 11.2.4 Industrial Safety Officer - Reginald R. Spradlin Education: - Graduate, Appomattox County High School
- Certified Instructor Trainer, Basic Cardiac Life Support, American Heart Association
- Certified Instructor, First Aid & Advanced First Aid, American Red Cross
- Training in the following areas:
Industrial Safety Fire Fighting Rescue Extrication Fire Protection Fire Extinguishing Equipment and Materials Arson Investigation.
Experience:
(1972-Present) Babcock & Wilcox, Industrial Safety Officer,
/
Lynchburg Research Center, Lynchburg,-Virginia O]
Mr. Spradlin is the LRC's Industrial Safety Officer. As such he is responsible for compliance with the regulations of the Occupational Health and Safety Administration. He advises the LRC on the standards and requirements of the National Fire Protection Associ-ation and performs reviews of equipment and systems for compliance with NFPA standards. He performs inspections of facilities and equipment for fire protection purposes. He reviews facility changes and modifications to ensure fire safety. Mr. Spradlin performs-tests, maintenance, and inspection of fire protection, control and extinguishing equipment. He is responsible for investi-gating all accidents, and keeping his management informed of safety activities. He performs fire and rescue training for the members of the LRC's Fire and Rescue Team, and serves as the Captain of the team.
He is a certified Shock Trauma Technician, an Emergency Medical Technician, and certified instructor in CPR and Standard and Advanced First Aid.
(1971-1972) Babcock & Wilcox, Accountability Technician, Lynchburg Research Center, Lynchburg, Virginia Ucense No SNM-778 Docket No.70-824 Date April, 1987 Amendment No.
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r Mr. Sprad1'in served as the Accountability Technician.
In this i
capacity he was responsible for the recordkeeping. system for SNM accountability _ in the Plutonium Development Laboratory.. He recorded all transfers of SNM, performed inventories, and updated -
the unit log records.
(1969-1971) Babcock & Wilcox, Health Physics Technician, Lynchburg Research Center, Lynchburg,- Virginia Mr. Spradlin was a health physics technician in the Plutonium Development Laboratory. He was responsible for performing contamination surveys of the facility, assisting in the monitoring of bagging operations, and supervising decontamination. He implemented the surveillance program for airborne radioactive material. He performed maintenance, testing, and calibration of alpha particle survey instrumentation and counting equipment. _He implemented the respiratory protection program in that laboratory.
(1967-1969) Babcock'a Wilcox, Plant Engineering Technician, Lynchburg Research Center, Lynchburg - Virginia As a plant engineering technician, Fr. Spradlin performed-installation,- modification, and repair of facilities, equipment, and experimental apparatus at the LRC. He performed these duties on electrical, mechanical and plumbing systems.
(1952-1967) Mead Corporation, Maintenance Superintendent,
' Mead Paper Company, Lynchburg, Virginia Mr. Spradlin served in several capacities during this period,'
including: finishing operation, paper machine operation, Millwright, Maintenance Foreman, Maintenance Superintendent, Safety Inspector and Accident Investigator.
Professional Affiliations:
L Cor. cord. Rescue Squad - Founding President American Heart Association -' Cardiac Care Committee t-l 11.2.5 Accountability -Specialist
.Kenneth D. Long Education:
l Graduate - White Sulphur Springs High School,1958 l
Certificate - Bookkeeping, Central Virginia Community College,1983 License No SNM-778 Docket No.70-824 Date April, 1987 Amendment No.
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Experience:
(1974-Present)
Babcock & Wilcox,- Accountability Specialist Lynchburg Research Center, -Lynchburg, Virginia Mr. Long, as'the Accountability Specialist, is responsible to the Manager of Safety and Licensing for the accurate accounting of all Special Nuclear, Source, and Byproduct material at the LRC. He is responsible -for recording all transfers of SNM that are made within the LRC and for preparing the reports and records of off site transfers.- He prepares all NRC/ DOE 741 Transaction Forms.
He. i s responsible for the timely completion of inventories of licensed material.
He initiates the paper work required for all shipments of licensed material.
In addition to.his normal duties he is a Document Custodian.
In this capacity, he is responsible for the safe storage of all classified DOE and' D0D documents at the LRC. He is also an authorized classiffer and an authorized courier of classified f
ma terial.
.(1970-1974) Babcock & Wilcox, Shipping & Receiving Clerk Lynchburg Research Center, Lynchburg, Virginia i
O.
-- Mr. -Long was responsible for-the shipment and' receipt of all materials at the LRC. This assignment included the processing of all the necessary forms and documents used for shipping and receiving. licensed materials as well as the many items that are required for operation of a research and development laboratory.
(1967-1970) Babcock & Wilcox,. Technician Lynchburg Research Center,- Lynchburg, Virginia Mr. Long was a technician in the Plutonium Development Laboratory during this period.. He performed chemical operations utilizing i.
uranium and plutonium materials and was responsible for the accountability of SNM materials into and out of his area.
Professional Affiliations:
t Institute of Nuclear Materials Management (Senior Member)
Nuclear Materials Control Committee, B&W (Secretary)
American Nuclear Society, Virginia Chapter (Member) c l
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11.2.6 License Administrator - Arne F. Olsen Facility Supervisor
- Arne F. Olsen Education:
AAS - Nuclear Technology, Central Virginia Community College,1978 Experience:
(1972-Present) Babcock & Wilcox, Senior License Administrator and Facility Supervisor, Lynchburg Research Center, Lynchburg, Virginia Mr. Olsen is responsible for preparing, amending, and administering the licenses that the LRC possesses with the NRC and the Common-wealth of Virginia. He acts as the primary liaison between the LRC and the NRC and other federal, state, and local agencies regarding nuclear matters. He coordinates the visits made by the NRC's Office of Inspection and Enforcement, and coordinates the LRC's compliance with NRC and state regulations and the licenses. He is the coordinator of the Safety Review Committee and is Chairman of the Safety Audit Subcommittee, and represents LRC management on both. Mr. Olsen is the Facility Supervisor and as such is responsible to the Manager, Lynchburg Technical Operations for the safety of all operations-at-the LRC.
U Mr. Olsen is the Alternate LRC Security Officer, Alternate Emergency Officer and an internal auditor.
(1968-1972) Babcock & Wilcox, Health Physics Technologist, Lynchburg Research Center, Lynchburg, Virginia In this capacity, Mr. Olsen was responsible to the site Health Physicist (Supervisor, Health and Safety) for the implementation of the Health Physics Program in the Plutonium Development Laboratory.
This responsibility included the implementation of the smearing, survey, air sampling, environmental sampling, and waste disposal l
programs.
(1964-1968)
Babcock & Wilcox, Technician and Shift Leader, Babcock & Wilcox Test Reactor, Lynchburg Research Center, Lynchburg, Virginia Mr. Olsen possessed a Senior Reactor Operator's License for the BAWTR.
He was in charge on one of four shif ts of reactor operators l
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V charged with the proper operation and maintenance of the BAWTR. He supervised the loading and unloading of fuel and experiments in the reactor and kept all required records of operations and maintenance performed on his shift.
(1960-1964)
U. S. Navy, Reactor Plant Electrical Supervisor, USS Enterprise CVA(N)-65 Mr. Olsen was an Electrician, First Class and was responsible for the proper operation and maintenance of all electrical equipment serving one of the reactor plants aboard the Enterprise.
Professional Affiliation:
Health Physics Society (Member)
Site Environmental Committee, B&W (Member) 11.2.7 Nuclear Criticality Safety Officer - Francis M. Alcorn l
Education:
B.S.
- Nuclear Engineering, North Carolina State College,1957 M.B.A - Business Administration, Lynchburg College,1974
- Graduate study in Nuclear Engineering, University of Virginia U
Experience:
(1971-Present)
Babcock & Wilcox, Supervisor, Nuclear Criticality Safety Group, Lynchburg Research Center, Lynchburg, Virginia This group is the Company's central organization which provides guidance, develops and validates the analytical methods needed for criticality evaluations, does criticality calculations, performs nuclear safety audits, and gives assistance to the various divisions of the Company and the Company's customers in matters related to nuclear criticality safety.
In addition to his responsibility as supervisor of this group, he is the Nuclear Safety Officer for the Lynchburg Research Center.
(1969-1971) Babcock & Wilcox, Criticality Specialist, Nuclear Safety Engineer, Lynchburg Research Center, Lynchburg, Virginia l.
License No SNM 778 Docket No.70-824 Date April,1987 Amendment No.
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Transferred to thelLRC as Nuclear Criticality Safety Specialist for Babcock & Wilcox's Naval Nuclear Fuel Plant, Commercial Nuclear Fuel Plant, and the LRC. He was appointed Nuclear Safety Officer for the LRC.
(1964-1969)' Babcock & Wilcox, Power Generation Division',
Lynchburg, Virginia L
Mr. Alcorn was a physicist in the PWR Development Section and was responsible for determining the most economical method for utilizing plutonium as a recycle fuel in 3&W's pressurized water reactor concepts.
In addition, he was Nuclear Criticality Safety Advisor' to the Company's Naval Nuclear Fuel Division.
'(1961-1964) Babcock & Wilcox, Nuclear Power Generation Division Lynchburg, Virginia i
He has been concerned with core neutron physics analysis and design of the Consolidated Edison Reactor, the Liquid Metal Fuel Reactor, the Babcock & Wilcox Test Reactor, the Advanced Test i
Reactor, the Heavy Water-Organic Cooled Reactor Concept, and Babcock & Wilcox Pressurized Water Reactor Concepts.
He developed methods for and performed calculations for criticality, fuel depletion, nuclear safety coefficients, power profiles, p
nuclear fuel costs and critical experiment analysis. -He has also
- Q worked in the areas of kinetic safety analysis.
(1957-1960) Babcock & Wilcox, Atomic Energy Division Lynchburg, Virginia He functioned as a nuclear engineer doing both core neutron
- hysics and shielding calculations.
(1960-1961) General Nuclear Engineering Corporation, Staff Physicist.
i Mr. Alcorn engaged in core neutron physics design and analysis of the Boiling Nuclear Superheat Reactor. He also wrote physics articles for Power Reactor Technology which were published by i
GNEC for the AEC.
Professional Affiliations:
Sigma Pi Sigma (Member)
Tau Beta Pi (Member) l License No SNM-778 Docket No.70-824 Date April, 1987 i
Amendment No.
O Revision No.
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Babcock &Wilcox j
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p American Nuclear Society - Past Chairman of ANS Nuclear v
Criticality Safety Division
- Member Standards Subcommittee ANS-8.
11.3 PROCEDURES 11.3.1 Area Operating Procedures (A0P) - All operations with licensed material shall be conducted in accordance with Area Operating Precedures or a Radiation Work Permit. Area Operating Procedures are prepared by any technically competent person. The proposed procedure is delivered to the Facility Supervisor who ensures that the procedure is in the proper format. The Facility Supervisor routes the procedure to the Nuclear Criticality Safety Officer who reviews it to assure that any nuclear criticality safety issues are properly addressed.
If the Nuclear Criticality Safety Officer has additions or corrections, he notes them on the procedure and forwards it to the Supervisor, Health and Safety.
If the Nuclear Criticality Safety Officer approves it, he signs the procedure in the space provided and forwards it to the Supervisor, Health and Safety. The Supervisor, Health and Safety reviews it for proper radiological and industrial safety content.
If he has additions or corrections, he notes them on the procedure and forwards it to the Facility Supervisor.
If the Supervisor, Health and Safety approves the procedure, he signs the procedure in the space provided and forwards it to the-Facility Supervisor. The Facility Supervisor p) reviews it for general safety and determines its impact on other i
i work and facilities. The Facility Supervisor is responsible for l
resolving all additions or changes recommended by the previous re-f viewers. When the procedure is approved by the three reviewers, the Fac.flity Supervisor forwards it to the Safety Review Committee.
The Safety Review Committee (SRC) may approve the procedure as written, approve the procedure conditionally with specific changes to be made prior to issuance or the SRC can disapprove it. The SRC coordinator signs for the SRC when approval is voted. The procedure l
may be implemented subsequent to SRC approval.
Revisions to A0P's will follow this same approval route, except that the revised procedure may be implemented after receiving the ap-l preval signatures of the Nuclear Criticality Safety Officer, Super-visor, Health and Safety, and the Fa'cility Supervisor. The revised procedure will be placed on the agenda for the next regularly scheduled meeting of the SRC. AOP manuals shall be placed in areas where the procedures apply.
License No SNM-778 Docket No.70-824 Date April, 1987 Amendment No.
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11.3.2 Technical Procedures - Technical procedures provide detailed techni-cal standards and instructions for performing specific tasks.
Pur-suant to this license application, they are not intended for use by operations personnel and are not distributed in the same manner as A0P's. Neither are they necessarily approved by the Safety Review Committee.
Technical procedures for the Health and Safety Group and the Nuclear Criticality Safety Group are reviewed and approved by the Senior Health Physics Engineer and the Nuclear Criticality Safety Officer, respectively, or by their designated alternates. The distribution list for each procedure is specified in the procedure.
11.4 TRAINING 11.4.1 General Radiation Protection Training The site provides two training programs covering the nature, use l
and control of radiation, and radioactivity. These courses are presented to ensure that all site personnel receive training l
appropriate to their activitics and to fulfill obligations under the N'lC license to provide such training.
n The courses consist of a serias of lectures. intended to present the Q
proper background and technical base to allow workers to understand the principles of radiation safety. The Supervisor, Health and Safety administers the course and, in general, teaches each course.
Where practical, basic general procedures and federal regulations are included and discussed. Training aids, such as motion pictures and self-study materials, are used as appropriate.
Program 1 is intended for site workers and non-site workers who will be authorized access to the restricted area.
Program 2 is intended for site and non-site workers who may enter the restricted and controlled areas but who will not be permitted to work with licensed material without supervision.
Program 3 is intended for authorized users (those who will be authorized to work with licensed l
material and to supervise such work).
l i
Training in area operating procedures a'id special area procedures is the responsibility of the Area Supervisor. This training should be accompanied with appropriate formal and on-the-job training as the job requirements dictate.
I License No SNM-778 Docket No.70-824 Date April, 1987 l
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11.4.2-Program 1 This course is presented to site workers and non-site workers who will be granted access to the restricted area but who will not be granted unescorted access to the controlled areas. The course pro-vides an introduction to radiation and radioactivity (understand-able to a non-technical person) and a thorough coverage of safety rules and procedures, including the site emergency procedures.
Subjects include the types of radiation, ALARA, radiation effects on humans, decontamination procedures, radiation exposure to females, warning signs, basic health physics rules, a history of radiation protection, worker's rights and responsibilities, and health physics terms.
11.4.3 Program 2 This course is presented to site workers and non-site workers who will be granted unescorted access to the restricted area and con-trolled areas but who will not be permitted to work with radioactive materials without supervision.
This course is intended to provide the workers with a knowledge of the hazards of working in radiation i
and controlled areas and ways to minimize their dose.
Subjects include types of radiation, radiation exposure limits, ALARA, per-l sonnel dosimetry and its use, dose calculation, biological effects, radiation exposure to females, radiation protection measures, warn-i bsT ing signs and labels, radiation work permits, emergency procedures, j
V rights and responsibilities of workers, and health physics terms.
11.4.4 Program 3 This course is presented to site workers and non-site workers who l
will be granted unescorted access to the restricted area and con-e trolled areas and will be permitted to work with radioactive materials and supervise such work. This course is intended for l
neeeting the requirements for designation of a worker as an author-ized user.
Subjects include fundamentals of radiation, external and internal radiation protection, biological effects, radiation detection, instrumentation, contamination control, license require-ments, site organization, rights and responsibilities under 10 CFR 19, ALARA, dose calculation, personnel dosimetry requirements and use, posting and labeling, and health physics terms.
11.4.5 Respiratory Protection Training Training in respiratory protection techniques will be required of License No SNM-778 Docket No.70-824 Date April, 1987 Amendment No.
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all workers before the use of such equipment will be allowed. This training will be carried out by a qualified individual, as defined in NUREG-0041 (Section 12.1), who will document that such training as been completed. Those persons who direct the work of workers I.
i using respiratory protection will be included in the training courses. Biennial retraining will be sched.iled, at the discretion of the qualified individual, to ensure that a high degree of pro-ficiency in the use of respiratory protective devices is maintained.
Training in respiratory protection shall include the following subjects:
a.
Discussion of the airborne contaminants present in the work environment including their physical properties, physiological actions, toxicity, means of detection, and maximum permissible concentrations (MPC's).
b.
Discussion of the importance of selecting the proper respirator based on the hazard and the dangers of using respirators for a purpose other than that intended.
c.
Discussion of the construction, operating principles, and limitations of the available respirators.
d.
Discussion of the use of engineering controls as a-substitute
,<m) for respiratory protection and the need to make every reason-able effort to reduce or eliminate the need for respiratory protection, e.
Instruction in methods to be used to determine that the respirator is in proper working order, f.
Instruction in fitting the respirator properly, field testing for proper fit, and factors that may influence a proper fit.
g.
Instructions in the proper use and maintenance of the respirator, h.
Discussion of the uses of various cartridges and canisters available for air-purifying respirators.
- i. Review of radiation and contamination hazards, including a review of other protective equipment that may be used with respirators.
License No SNM 778 Docket No. 70 824 Date April, 1987 Amendment No.
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Instruction in emergency actions to be taken in the event of respirator malfunction.
k.
Classroom instruction to recognize and cope witn emergency situations while working with a respirator.
1.
Any additional training as needed for special use.
m.
The wearer must pass a written examination on the material presented on respiratory protection.
11.5 FACILITY CHANGE Changes and modifications to buildings, exhaust ventilation systems, gas supply systems, emergency electrical systems, etc. are requested on Form LRC-229, " Facilities Work Order Form" (Figure 9-4).
All work orders are forwarded to the maintenance supervisor.
The Plant Engineering Supervisor determines if the request involves a facility change.
If a facility change is involved, the work order is forwarded to the Facility Supervisor.
It is the Facility Super-visor's responsibility to determine that all safety and licensing considerations have been addressed and if the request must be approved by the Safety Review Committee. Space is provided on the form for the approval-signatures of the Supervisor, Health and-e(m)
Safety, the Industrial Safety Officer, and the Facility Supervisor.
v Completed forms are kept on file by the maintenance supervisor and are audited once a month by the Health Physics Group.
License No SNM-778 Docket No.70-824 Date April, 1987 Amendment No.
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3 FIGURE 11-1
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NNFD RESEARCH LABORATORY ORGANILATION EMPLOYEE, COMMUNITY, & REGULATORY RELATIONS J. A. EANES MANAGER FACILITY SAFETY REVIEW SUPERVISOR A. F. OLSEN COMMITTEE CAFETY & LICENSING R. L. BENNETT MANAGER
'fN NUCLEAR i
LICENSE HEALTH & SAFETY CRITICALITY ACCOUNTABILITY vV ADMINISTRATOR SAFETY SPECIALIST G. S. HOOVLER OFFICER A. F. OLSEN SUPERVISOR F. M. ALCORN K. D. LONG SENIOR INDUSTRIAL HEALTH PHYSICS SAFETY ENGINEER OFFICER S. W. SCHILTHELM R. R. SPRADLIN HEALTH PHYSICS GROUP l
i AREA AREA SUPERVISOR SUPERVISOR l
License No SNM-778 Docket No.70-824 Date April, 1987 l
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FIGURE 11-2 LRC229 FACluTIES WORK ORDER FORM l
l T0 Plant Engineering Date From:
Section:
Signed:
Section Mgr.:
Date:
0o1 Required:
Charge No.:
(, Labor)
(Materia.)
DESCRIPTION OF WORK TO BE DONE l
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$1GN ATURE REQUIRED Industrial Safety Officer H eal,th Physic s:
Facility Supervisori l,
seece setow TMs Line For Plent Engineerlag use only l
Order Received Date Signedi Planned Starting Dater Planned Completion Date:
Order Completedi Work Order Number Datei sig n atur ei l
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License No SNM-778 Docket No.70-824 Date Aprile 1987 Amendment No.
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N TABLE OF CONTENTS Section Page 12.0 RADIATION PROTECTION 12-1 12.1 PROGRAM 12-1 12.2 POSTING AND LABELING 12-1 12.2.1 Radioactive Materials Area 12-1 12.2.2 Contamination Area 12-1 l
12.2.3 Radiation Area 12-2 12.2.4 High Radiation Area 12-2 12.2.5 Airborne Radioactivity Area 12-2 12.3 EXTERNAL RADIATION - PERSONNEL MONITORING 12-3 12.3.1 Administrative Exposure Control 12-3 12.3.2 Personnel Monitoring for Site and Non-site Workers 12-3 l
12.3.3 Personnel Monitoring and Escort Requirements 12-3
-l 12.3.4 Monitoring Devices 12-3 l
12.4 DIRECT RADIATION SURVEYS 12-5 l
12.5 REPORTS AND RECORDS 12-6 l
12.6 INSTRUMENTS 12-6 l
12.6.1 Types 12-6 l
12.6.2 Calibration 12-8 l
12.7 PROTECTIVE CLOTHING 12-8 l
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12.7.1 Clothing 12-8 l
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,O TABLE OF CONTENTS (Continued)
Section Page 12.7.2 Emergency Clothing 12-9 l
12.8 ADMINISTRATIVE CONTROL LEVELS 12-9 l
12.8.1 Internal Occupational Exposure 12-9 l
12.8.2 External Occupational Exposure 12-14 l
12.8.3 Airborne Activity 12-14 l
12.8.4 Liquid Activity 12-16 l
12.8.5 Surface Contamination 12-16 l
12.9 RESPIRATORY PROTECTION 12-19 l
12.10 OCCUPATIONAL EXPOSURE ANALYSIS 12-20 l
12.10.1 External Exposure 12-20 l
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12.10.2 Internal Exposure 12-25 l
12.11 MEASURES TAKEN TO IMPLEMENT ALARA 12-34 l
12.12 BI0 ASSAY PROGRAM 12-35 l
12.13 AIR SAMPLING AND MONITORING 12-36
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12.13.1 Air Sampling Program 12-36 l
12.13.2 Air Monitoring Program 12-37 l
12.14 SURFACE CONTAMINATION 12-38 l
12.14.1 Smear Surveying 12-38 l
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12.14.2 Direct Radiation Surveys 12-41 l
12.14.3 Personnel Contamination Surveys 12-43 l
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List of Tables Table Page 12-1 PORTABLE RADIATION PROTECTION INSTRUMENTATION 12-7 l
12-2 STATIONARY RADIATION PROTECTION INSTRUMENTATION
~ 12-7 l
12-3 PLUT0NIUM BI0 ASSAY ACTION CRITERIA 12-9 l
12 PLUT0NIUM BI0 ASSAY ACTION CRITERIA 12-10 l
12-5 URANIUM BI0 ASSAY ACTION CRITERIA 12-11 l
12-6 FISSION PRODUCT. ACTION CRITERIA 12-13 l
12-7 STACK RELEASE ACTION LEVELS 12-16 l
12-8 SMEAR SURVEYS IN WORK AREAS 12-17 l
12-9 ACTION LEVELS FOR LARGE AREA SMEARS 12-18 l
12-10 MAXIMUM PERMISSIBLE CONTAMINATION FOR SKIN SURFACES 12-19 l
12-11 1984 EXPOSURES BY RANGE 12-21 l
nh 12-12 1985 EXPOSURES BY RANGE 12-22 l
12-13 RADIATION EXPOSURE.
12-23 l
12-14
. EXPOSURE BY GROUP (PERSON' REMS) 12-24 l
12-15 NUMBER OF URINE BI0 ASSAY SAMPLES 12-25 l
12-16 1983 AIR ACTIVITY 12-26 l
12-17 1984 AIR ACTIVITY 12-28 l
12-19 WHOLE BODY COUNTS 1983 12-35 l
12-20 Am - Pu LUNG COUNTING 1983 12-36
_L 12-21 URANIUM LUNG COUNTING 1983
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Table Page 12-21 WHOLE BODY COUNTS 1984 12-33 l
12-22 ACTION LEVELS FOR LARGE AREA SMEARS 12-39 l
12-23 SMEAR SURVEY FREQUENCIES AND ACTION LEVELS 12-40 l
12-24 CONTAMINATION ACTION LEVELS 12-42 l
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O 12.0 RADIATION PROTECTION 12.1 PROGRAM The radiation protection program at the site is implemented to protect l
employees and the general public from the harmful effects of radi-ation and radioactive material, to comply with NRC regulations, and to maintain personnel exposures as far below the limits established by the NRC as is reasonably achievable.
Implementation of the program requires the active participation of all I-personnel who work with licensed material or in areas were licensed material is handled. To support the worker, the site has established I
the Health and Safety organization and vested it with the authority and resources necessary to meet the program goals.
12.2 POSTING AND LABELING Many areas in the site are required to be posted to indicate the l
hazard present.
This posting is required by the federal regulations and is a fundamental part of an effective radiation protection
(
program. Posting of areas makes the workers aware of the potential Q]
hazards in the area and assists workers in keeping their exposures ALARA. Permanent postings are the responsibility of the Health and Safety Group. Temporary postings are the responsibility of Authorized Users. This section discusses the posted areas at the site.
Persons j
not directly familiar with conditions existing in a posted area shall contact the area supervisor prior to entering and shall enter only l
under his direction.
12.2.1 Radioactive Materials Area - Any area where radioactive materials are stored, handled, or processed in amounts exceeding 10 times the quantities specified in 10 CFR 20, Appendix C is designated a radioactive materials area.
Each area is clearly marked at every normal entry with a sign bearing the radiation caution symbol and RADI0 ACTIVE MATERIAL (S). Monitoring CAUTION the words equipment and protective clothing required for use in the area will be specified by the Health and Safety Group.
12.2.2 Contamination Area - This is any area in which loose contamination is present in quantities in excess of those specified in Table 12-24 or an area designated by the Health and Safety Group as one License No SNM 778 Docket No.70-824 Date April, 1987 Amendment No.
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in wnich there is a risk of contamination. Each contamination area is clearly marked at every normal entry. Work in these areas may require a Radiation Work Permit.
Protective clothing, respiratory protection, and personnel monitoring devices required for entry into these areas must be specified by the Health and Safety Group.
Entry into the area without the prescribed equipment is prohibited.-
When exiting a contamination area, workers must remove the protec-l tive clothing and monitor himself in accordance with established procedures.
12.2.3 Radiation Area - A Radiation Area is an area in which an individual could receive a radiation exposure to a major portion of the body greater than 5 mrem in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or 100 mrem in 5 consecutive days.
Each radiation area is clearly marked at every normal entry with a CAUTION sign bearing the radiation caution symbol and the words
- RADIATION AREA. Work in these areas may require a Radiation Work Permit. Personnel monitoring devices and protective clothing to be worn in the area will be specified by the Health and Safety Group.
12.2.4 High Radiation Area - Any area in which an individual may receive an exposure to a major portion of the body greater than 100 mrem in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> is a High Radiation Area.
High radiation areas are desig-nated by a sign at each normal entrance bearing the radiation caution symbol and the words - CAUTION - HIGH RADIATION AREA.
A Entry into high radiation areas is limited to qualified persons, or V
under the direct supervision of a qualified person and, working under an approved radiation work permit. Protective clothing, protective equipment, and personnel monitoring devices appropriate for the area will be specified by the Health and Safety Group and must be worn. When protective clothing is required, each person must remove the protective clothing and monitor himself in accordance with established procedures, when exiting the area.
12.2.5 Airborne Radioactivity Area - This is an area in which airborne radioactivity concentrations could exceed the maximum permissible concentration limits given in 10 CFR 20, Appendix B or in which the concentration of airborne radioactivity averaged over the number of hours individuals are in the area could exceed 25% of the limits given in 10 CFR 20, Appendix B.
Each area is clearly designated by a sign at each normal entrance bearing the radiation caution symbol CAUTION - AIRBORNE RADI0 ACTIVITY AREA. Entry is and the words limited to those qualified persons classified as radiation workers, working under an approved radiation work permit. No entry is permitted until an appropriate area survey has been made and a member of the Health and Safety Group is present.
Protective License No SNM-778 Docket No.70-824 Date April, 1987 Amendment No.
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pj clothing, protective equipment, and personnel monitoring devices to be worn in the area will be specified by the Health and Safety Group and must be worn. When exiting these areas, each person must remove the protective clothing and monitor himself in accordance with established procedures.
12.3 EXTERNAL RADIATION - PERSONNEL MONITORING 12.3.1 Administrative Exposure Control - Limits for external radiation exposure are set forth in 10 CFR 20.101 and these general limits are used at the site. The applicable exposure limits to be used i
for operations at the site are:
I 1.
Whole body - 300 mrem / week (with long-term exposure controlled within the 1.25 Rem / quarter limit by the worker's immediate supervisor) 2.
Skin of the whole body - 1.5 Rem / week 3.
Hands and forearms, feet and ankles - 3.0 Rem / week.
The Manager, Safety and Licensing has the authority to approve whole body exposures up to, but not exceeding, 3.0 Rem / calendar p
quarter.
In emergencies, the Emergency Officer-is authorized to Q
allow personnel exposures to the whole body of up to 3.0 Rem / calendar quarter. Higher exposures may be authorized by the Emergency Officer in accordance with the Radiological Contingency Plan.
12.3.2 Personnel Monitoring for Site and Non-site Workers - All site and non-site workers will be issued a film badge, a SRD, and a TLD.
This dosimetry will be worn by the workers when they are in the restricted area. When the workers leaves the restricted area they will place their dosimetry on a rack provided for this purpose.
12.3.3 Visitor Monitoring and Escort Requirements - Visitors to the re-stricted area will be issued a TLD. This dosimetry will be worn by the visitor when they are in the restricted area and will be sur-rendered to the receptionist when they depart the site. Visitors must be escorted by a site worker when in the restricted area.
12.3.4 Monitoring Devices The primary device used for monitoring exposure on site is the film l
License No SNM-778 Docket No.70-824 Date April, 1987 Amendment No.
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d badge. The exposure measured by this badge (reported in units of dose equivalent) becomes a part of the workers permanent exposure I
record. Films are changed monthly and are mailed off-site for evalua tion.
In some cases, a Health Physics Engineer may choose to base the monthly exposure of an employee on the monthly thermo-luminescent dosimeter (TLD).
This determination shall be recorded in.the employees exposure record.
' In general, the worker should wear the dosimeters on the portion of l
the whole body expected to receive the highest dose (with the exception of extremity dosimetry issued in special cases). The film badge and/or nwnthly TLD badge should always be worn in the proper orientation to ensure that exposure to non-penetrating radi-ation (e.g., beta radiation) is recorded.
For cases in which the exposure may vary significantly.within a small area, several. badges may be worn to ensure that the maximum whole body dose is measured.
In this context, whole body includes the head, lens of the eyes, the gonads, the upper legs above the knees, and the upper arms above the elbows.
12.3.4.1 Pocket Dosimeters - These dosimeters are small, air-filled ionization chanters used to provide a check of the daily exposure of workers and to ensure that the administrative limit for weekly l
exposure is not exceeded. Indirect dosimeters are capable of m
measuring ~ external exposure to gamma radiation-in-the range 0 to-
[V) 200 mR (other ranges are also available). These dosimeters are read, recorded, and rezerced daily. Daily readings are used also as an indication of the need to evaluate the primary dosimeter before the normal exchange period.
Some workers may be issued self-reading pocket dosimeters (SRD).
l These dosimeters do not require reading and recharging on a daily frequency and the worker may evaluate his accumulated exposure I
without the need for a special reading device. Workers are en-l couraged to read their self-reading dosimeters at least on a daily basis. These dosimeters are capable of measuring external exposure to gamma radiation in the range 0 to 200 mR, but other ranges are available.
12.3.4.2 Film Badges - These dosimeters are the primary monitoring device used on site, i.e., the film badge results are entered in the l
errployee's permanent exposure record.
Film badges monitor external exposure to beta and gamma radiation typically in the range 15 mRems to 500 Rems. For situations in which neutron exposure is probable, film packets sensitive to neutrons also are used.
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Films in use on site are changed monthly and mailed to an off-site dosimetry service for processing (reading, recording, and reporting).
12.3.4.3 Thermoluminescent Dosimeters (TLD) - TLD's are small, solid-state dosimeters capable of measuring external exposure from beta and gamma radiation in the range 10 mRems to 10,000 Rem. The monthly TLD's are used to duplicate the readings of the film badge.
These badges are also changed monthly and mailed off-site for processing.
At the discretion of a Senior Health Physics Engineer, persons l
handling radioactive materials may be issued extremity dosi-meters. These dosimeters are small TLD chips attached to a ring and are to be worn on the fingers. TLD " finger rings" are capable of measuring external exposure to beta and gamma radi-ation in the range 10 mRems to 10,000 Rems. These dosimeters are evaluated on a frequency established by the Health and Safety Group.
12.4 DIRECT RADIATION SURVEYS Surveys of the direct radiation exposure in areas on site are to be p
performed on a frequency established by the Health and Safety Groupr.
gj In general, these surveys require the selection of the appropriate portable survey instruments based upon the anticipated radiation levels, the types of radiation expected, e.nd the nature or type of su *vey to be performed.
Survey maps of the areas to be surveyed may be used to record the measured ambient radiation levels and/or, in some cases, to designate specific areas in which the exposure rates should be measured. The survey should also include a visual exami-nation of the area for any unusual conditions or work habits which could affect the exposures received by personnel working in these areas.
Items of this nature should be reported immediately to a Health Physics Engineer or corrected immediately, if practical.
Results of these surveys should be reviewed by a Health Physics l
Engineer to ensure that the proper posting requirements are in effect i
for the area and to ensure that appropriate actions are taken to keep all exposures ALARA.
License No SNM-778 Docket No.70-824 Date April, 1987 Amendment No.
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12.5 REPORTS AND RECORDS The following records will be maintained by the Health and Safety Group for the periods indicated.
Health and Safety Supervisor audits 2 years Shipping and receiving RM forms 5 years Waste disposal records
(*)
Personnel dosimetry records
(*)
Results of Bioassays and Whole Body Counting
(*)
Releases to the environment
(*)
Radiation survey data 2 years Contamination survey data 2 years Radiation Work Permits (completed) 5 years Radiation detection instrument calibration 2 years Leak tests of sealed sources 2 years Worker training
(*)
Worker. retraining
(*)
Airborne radioactivity sampling data
(*)
NRC-4 forms
(*)
NRC-5 forms
(*)
- - indicates that the record will be retained until the NRC authorizes its disposition.
(
12.6 INSTRUMENTS 12.6.1 Types - The commitment of site management to an effective radiation l
, protection program includes the obligation to provide the adequate equipment and supplies for such a program.
It is the responsi-bility of the Manager, Safety and Licensing and the Supervisor of Health and Safety to ensure the appropriate radiation protection instrumentation is available for use on site.
In addition, the l
Health and Safety Group has the responsibility to ensure that this instrumentation is used properly, and is calibrated, maintained, and repaired as necessary. Minimum instrumentation requirements for maintaining an effective radiation protection program are listed in Tables 12-1 and 12-2. Other specialized instrumentation may not be included in this list.
However, the exclusion of these instruments does not imply that their availability does not enhance the effectiveness of the radiation protection program.
License No SNM 778 Docket No.70-824 Date April, 1987 Amendment No.
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TABLE 12-1 PORTABLE RADIATION PROTECTION INSTRUMENTATION Radiation Instrument Sensi tivity Range Window Thickness Low-range GM Be ta, Gamma Bkgd. to 30 mg/sq. cm.
mR/hr Intermediate Be ta, Gamma
- mR/hr to Beta:
1 mg/sq. cm.
range ion R/hr chamber High range Gamma up to 500 R/hr
>100 mg/sq. cm.
fon chamber Proportional Alpha, Beta Bkgd. to 1 mg/sq. cm.
counters 500,000 cpm Proportional Neutron, fast Bkgd. to N/A counters and thermal 5000 mrem /hr A
Portable air Air particulate N/A N/A
()
samplers collection only TABLE 12-2 STATIONARY RADIATION PROTECTION INSTRUMENTATION Radiation Instrument Sensitivi ty Range Window Thickness Laboratory Alpha, Beta Bkgd. to
<1 mg/sq. cm.
proportional 100,000 cpm counter Air particulate Alpha, Be ta Bkgd. up
<1 mg/sq. cm, moni tors License No SNM-778 Docket No.70-824 Date April, 1987 l
Amendment No.
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O Stack particulate Alpha, Beta Bkgd. to
<1 mg/sq. cm.
monitor 1,000,000 cpm Stack gas Beta, gamma Bkgd. to 30 mg/sq. cm.
monitor 100,000 cpm 12.6.2 Calibration - Portable survey instruments shall be calibrated twice annually using approved procedures and sources traceable to the National Bureau of Standards.
In addition, frequent operational checks will be performed on survey instruments while in use. For example, Geiger-Mueller survey instruments always indicate the presence of radiation above the ambient background. This provides an indication that the instrument is functioning.
Portable alpha survey instruments are equipped with check sources which can be used to ensure that the instruments are operating correctly.
Portable ionization chamber survey instruments are not equipped with an internal check source and the user must make sure these instruments are functioning before making a radiation survey.
Fixed and stationary radiation monitoring equipment is calibrated on either a semi-annual or annual basis depending on the applicable manufacturer's recommendations and established health physics procedures. Operational checks are performed routinely by the p
Health Physics technicians on the laboratory counting equipment and (j -
" friskers" located at exits from selected areas on site.
I 12.7 PROTECTIVE CLOTHING 12.7.1 Clothing - The following is a list of protective clothing that is available for use by personnel during normal and maintenance condi-tions:
1.
Laboratory coats 2.
Coveralls 3.
Shoe covers, treated fabric (reusable) 4.
Shoe covers, plastic 5.
Pants, plastic l
6.
Coats, plastic l
7.
Hoods, fabric (reusable) 8.
Shields, spatter 9.
Glasses, plastic 10.
Glasses, glass
- 11. Gloves, plastic l
License No SNM 778 Docket No.70-824 Date April, 1987 l
Amendment No.
0 Revision No.
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- 12. Gloves, surgeons 13.
Gloves, heat resistant 4
14.
Coats, heat reflective 15.
Hard-ha ts.
1 12.7.2 Emergency Clothing - In the event of an accident that requires special clothing or personnel protective equipment, the Fire and Rescue Team is provided with the following:
1.
Hard-hats, heat resistant with face shields 2.
Coats, flame resistant 3.
Boots, high top rubber with steel toe shields 4.
Gloves, chemical resistant 12.8 ADMINISTRATIVE CONTROL LEVELS 12.8.1 Internal Occupational Exposure 12.8.1.1 Plutonium bioassay action criteria.
TABLE 12-3 PLUT0NIUM BI0 ASSAY ACTION CRITERIA OJ Bioassay Technique Action Level Action To Be Taken Urinalysis
< 0.2 dpm/L None
> 0.2 dpm/L 1.
Resample the individual within 5 working days.
2.
The Supervisor, Health and Safety shall con-sider the need for worker restriction to prevent further exposure until the diagnostic evaluation is complete.
Only the Supervisor, Health and Safety may lift any work restriction once it is imposed.
License No SNM 778 Docket No.70-824 Date April, 1987 Amendment No.
0 Revision No.
4 Page 12-9 Babcock &Wilcox a McDermott company
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\\,_/ -
3.
If #1 is positive, in-vestigate the cause and correct.
4.
If the exposure is con-firmed by #1, investi-gate to determine how exposure was incurred and correct it.
If the exposure exceeds 50% of the maximum permissible annual dose, the worker shall be restricted from further exposure until the Supervisor, Health and Safety authorizes the lif ting of their restriction.
TABLE 12-4 P'.UT0NIUM BI0 ASSAY ACTION CRITERIA Bioassay Technique Action Level Action To Be Taken In-vivo
< 1.6E-8 Ci None Pu-239
> 1.6E-8 C1 1.
Restrict worker from Fu-239 further exposure.
2.
Resample the individual within 10 working days.
3.
Determine if area surveys support the l
analysis results.
l 4.
If area surveys confirm result, investigate the License No SNM 778 Docket No.70-824 Date April,1987 i
Amendment No.
O Revision No.
4 Page 12-10 O(""N Babcock &Wilcox a McDermon company
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'v_ I cause and take correc-tive actions.
5.
If the resample results do not confirm the exposure, the Super-visor, Health and Safety may lift the work restrictions.
6.
If resample results con-firm the exposure, the Supervisor, Health and Safety shall determine the organ dose.
7.
If.the exposure has exceeded 50% of the maximum permissible l
annual dose, the worker shall remain on a work restriction until the Supervisor, Health and Safety authorizes the es removal of the re-
)
striction.
(
12.8.1.2 Uranium bioassay action criteria.
TABLE 12-5 URANIUM B10 ASSAY ACTION CRITERIA Bioassay Technique Action Level Action To Be Taken
- a. Urinalysis
< 9 ug/L None
- b. Urinalysis 9-16 ug/L
- 1. Determine if area surveys support the analysis re-sults.
License No SNM-778 Docket No.70-824 Date April, 1987 Amendment No.
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- 2. If #1 is positive, in-vestigate and correct as needed.
- 3. Make sure individual is in-vivo counted during the next time that the counting service is at the B&W site.
- c. Urinalysis
> 16 ug/L
- 1. Restrict the worker from further exposure.
Resample the individual within 5 working days.
- 2. Determine if area surveys support the analysis resul ts.
- 3. If #2 is positive, in-vestigate the cause and correct as needed.
- 4. If exposure is confirmed by #2, investigate to y
determine how exposure was incurred and correct it.
If the exposure ex-ceeds 50% of the maximum permissible annual dose, the worker shall be re-stricted from further exposure until the Super-
~
visor, Health and Safety authorizes the lif ting of this restriction.
- d. In-vivo
< 30 ug
- 1. None U-235
- e. In-vivo 30-120 ug
- 1. Determine if area surveys support the analysis re-sults.
License No SNM 778 Docket No.70-824 Date April, 1987 Amendment No.
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- 2. If #1 is positive, in-vestigate and correct as needed.
- f. In-vivo
> 120 ug
- 1. Resample the individual U-235 within 10 working days.
- 2. Determine if area surveys support the analysis re-suits.
- 3. If #2 is positive, in-vestigate the cause and correct as needed.
- 4. If exposure is confirmed by #1, investigate to de-termine how exposure was incurred and correct it.
If the exposure exceeds 120 ug, the worker shall be restricted from further exposure until the Supervisor, Health and Safety authorizes the lifting of this restric-tion.
12.8.1.3 Beta-gamma activity - Workers who work in areas where beta-gamma internal exposure is likely (Hot Cells, Radiochemistry, Health Physics) shall be in-vivo counted at approximately annual in terval s.
TABLE 12-6 FISSION PRODUCT ACTION CRITERIA i
Analysis Action Level Action to be Taken In-vivo
>10% MP0B Remeasure subject to determine effective half life of the contami-nant and plot decay curves.
License No SNM 778 Docket No.70-824 Date April, 1987 Amendment No.
O Revision No.
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-.___,,..,.-,..-_,.,,,.,n.,,,,
,,-.,-.,_ _-- -,_.,.,_,- _,- n.
p
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Followup program will continue until the contamination present is
<5% MP08 or the effective half life has been determined.
Estimation
>10% MP08 Submit in vitro sample for analysis from nasal within 5 working days.
smears or air sample In-vitro
>5% MP0B Resample excreta to confirm presence of contamination and to establish rate of elimination.
Perform isotopic analysis if >10%
MP08 is a possibility.
In-vitro
>10% MP08 In vivo measurement to be made as soon as practicable.
The Supervisor, Health and Safety shall be responsible for evalu-ations to determine the location and amount of deposition; to provide data necessary for estimating internal dose rates, retention functions, and dose commitments; and to determine whether work restrictions or referrals-for-therapeutic treatment b) are required for any case where a result indicating a greater V
than 10% MP08 deposition of a radionuclide is verified.
12.8.2 External Occupational Exposure - Personnel monitors (film badges, dosimeters, or other suitable devices) are provided to measure the radiation exposure of visitors and workers.
Personnel dosimeters l
issued pursuant to 10 CFR 20.202 shall be read on a monthly basis.
The Area Supervisors are responsible for keeping exposures below I
300 millirem per week and 1250 millirem per quarter. The Super-visor, Health and Safety may approve weekly exposures above 300 millirem, but the quarterly limit of 1250 millirem shall not be exceeded without the approval of the Manager, EC&RR.
If a worker has received the quarterly limit and the Manager, EC&RR has not authorized exceeding the limit, the worker shall be restricted to prevent further exposure for the remainder of the quarter.
12.8.3 Airborne Activity 12.8.3.1 Air Monitoring Program - Air monitoring in operating areas of the License No SNM 778 Docket No. 70 824 Date April, 1987 Amendment No.
O Revision No.
4 Page 12-14 Babcock &Wilcox a McDermott company
b' site is accomplished with continuous monitors in predetermined, fixed locations. A monitor is placed in each radioactive materials handling area in which there is a potential for the release of airborne radioactivity. Locations are selected based upon the ability of the monitor to provide a reasonable evalu-ation of the airborne activity in a particular area and to provide adequate. warnings to those in the area of changing condi-tions. The determinations are made by the Health and Safety Group based upon the operations in the area, the potential for release, the quantity and chemical form of the material.
Alarms are set in accordance with a particular operation, the material being handled, and the potential for release. Actual alarm points are set as low as possible commensurate with the ambient radiation levels in the area. Personnel are instructed through procedures and training to evacuate, up wind, if an air monitor alarms and to notify the Health and Safety Group.
Re-entry is controlled by the Health and Safety Group.
12.8.3.2 Effluent Monitors - Potentially contaminated air from chemical hoods, hot cells, and glove boxes is discharged ultimately through the 50-meter stack.
Generally, exhaust air containing beta-gamma activity is passed through a single-stage HEPA filter which is sufficient to remove airborne particulates. Air from more hazardous operations, e.g.,- from glove boxes,-is' passed
(
through a two-stage HEPA filter.
V Discharges through the stack are monitored with a sampling head located in the stack about 25 feet above the base. Air removed by the sampler passes through a fixed filter, into the chamber of the gas monitor, and is returned to the stack. The fixed filter
~
is monitored continuously for alpha and beta activity by a gas-flow proportional counter. The second monitor, the gas monitor operates continuously utilizing a halogen-quenched GM tube. The stack monitor flow rate is maintained at a minimum of 2 cfm.
Both monitors are equipped with adjustable alarms. The set points for these alarms are determined by the Health and Safety Group. The alarms are connected to an alarm panel located in the Health Physics Area in Building 8.
Alarms of the system are responded to by the Health and Safety Group. The alarm condition is first verified by the Health and Safety Group.
If the alarm is actual, the exhaust fan is secured, operations personnel are advised to stop all operations with radioactive material, the cause is investigated by the Health and Safety Group, corrected by operations personnel, and the fan restarted.
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TABLE 12-7 STACK RELEASE ACTION LEVELS Release Product Action Levels Beta Particulate 200 uC1/ week Alpha Particulate 1 uC1/2 weeks (long lived)
Kr-85 70 Ci/ week H-3 3 Ci/ week I-131 200 uCi/ week 12.8.4 Liquid Activity - Liquids containing radioactive material are dis-charged from the area where they are generated, to the Liquid Waste Disposal Facility. This facility is comprised of a series of tanks. All radioactive liquid waste is held in this facility for sampling prior to release.
If the concentration of radioactivity exceeds 25% of the MPC values listed in Table I, Col. 2, of-10 CFR p
20, Appendix B, the waste must be diluted to levels that meet this specification. Liquid waste is discharged to the liquid waste processing system at the NNFD. The NNFD must be notified and approve of each discharge from the site prior to discharge.
No l
alarms are associated with this system because its operation is under the positive control of the Health and Safety Group.
12.8.5 Surface Contamination 12.8.5.1 Work Areas - The Health and Safety Group performs smear surveys in the work areas listed in Table 12-8.
The frequencies specified in Table 12-8 are minimum frequencies. More frequent surveys are performed based on the level of work performed in the specified areas. Action is taken to protect personnel and reduce the levels of contamination below those specified. The Health and Safety Group will supervise and direct the protection and decontamination activities. Decontamination to reduce levels of contamination will commence within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of discovery. The Supervisor, Health and Safety shall evaluate and approve any delays on decontamination work that are longer than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
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TABLE 12-8 SMEAR SURVEYS IN WORK AREAS Action Level Area Frequency *
(dpm/100 cm2)
<---------------------------ALPHA------------------------->
Unirradiated, unencapsulated Weekly 5000 fuel handling areas Building B Counting Lab.
Monthly 200 Hot Cell Oper. Area Monthly 200 Scanning Electron Monthly 200 Microscopy Lab.
Exit portals from Biweekly 200 controlled areas
<-------------------BETA + GAMMA-------------------------->
Building B Counting Lab.-
Monthly 2000 Scanning Electron Monthly 2000 Microscopy Lab.
Hot Cell Operations Area Bimonthly 2000 Cask Handling Area Bimonthly 22000 Radiochemistry Lab.
Bimonthly 22000 Exit portals from Biweekly 2000 controlled areas
- Minimum frequency specified. More frequent surveys are performed, based on work loads.
Large area smears are used to survey many square meters of surface area.
Action levels for large area smears are given below.
License No SNM 778 Docket No.70-824 Date April, 1987 i
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f TABLE 12-9 ACTION LEVELS FOR LARGE AREA SMEARS
- 1.. Routine Large Area Smears (1000-5000 dpm) - Repeat' the large
-area smear.
If results show levels.of contamination above 1000 dpm, take smears. in. smaller areas to locate the source.
Decontaminate all areas in which the smear results indicate contamination above 1000 dpe/100 square feet.
2.-
Routine Large Area Smears (5000-10,000 dpm)
. Repeat the
- large area smear.
If results show levels of contamination above 5000 dpe, isolate the contaminated area. Take smears in smaller areas to locate the source. Decontaminate all areas in which the smear results show contamination in excess of 1000 dpm/100 square feet.
3.
Routine Large Area Smears (>10,000 dpm) - Isolate the con-taminated area. Survey all personnel in the contaminated area. Take smaller smears in the area to locate the source.
Decontaminate all areas in which the, smear results show con-tamination in excess of 1000 dpm/100 square feet. Survey all persons leaving the building.
12.8.5.2 Personnel Contamination Surveys - Personnel are required to monitor themselves for activity present on their hands, shoes, clothing and person before exiting a contamination area. Con-taminat. ion monitors (friskers) are located at all normal exits from contamination areas for this purpose. The detector should be held as close to the ' surface of the item being monitored as possible, without touching the item, and the probe should be moved at a slow speed over the surface. Allowable levels of contamination on skin surfaces and on items of clothing are given in Tables 12-10. Any contamination in excess of these Ifmits should be reported'immediately to the Hec 1th and Safety Group.
The Health and Safety Group will supervise the decontamination and determine if clothing must be discarded. The approval of the Supervisor, Health and Safety shall be required to allow any individual +.o leave a contaminated area who is contaminated above background radiation levels.
LloenseNo SNM 778 Docket No. 70 824 Date April, 1987 Amendment No.
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k TABLE 12-10 MAXIMUM PERMISSIBLE CONTAMINATION FOR PROTECTIVE CLOTHING 2
(dpm/100 sq. cm )
Item Alpha Beta + Gamma Clothing 2,200 22,000 Shoes 22,000 220,000 12.8.5.3 Release of Equipment or Packages - Packages and equipment are surveyed by the Health and Safety Croup. The Health and Safety Group has the authority to prohibit the release of items that are found to exceed the limits specified in Annex C to License SNM-778 " Guidelines for Decontamination of Facilities c,nd Equip-ment Prior to Release for Unrestricted Use of Termination of Licenses for Byproduct, Source, or Special Nuclear Material, dated July,1982."
j t]
12.9 RESPIRATORY PROTECTION U
The primary objective of a respiratory protection program is to limit the inhalation of airborne radioactive materials and other hazardous ma terials. This objective is normally accomplished through the use of engineering controls, including process, containment, and venti-lation equipment. When engineering controls a.e not feasible or cannot be applied, respiratory protection must be used. The Health and Safety Group is responsible for the implementation of the respiratory protection program. The program is based on the guidance contained in 10 CFR 20, Regulatory Guide 8.15, " Acceptable Programs for Respiratory Protection," and NUREG-0041, " Manual of Respiratory Protection Against Airborne Radioactive Materials."
The respiratory protection program will include the following:
1.
Air sampling and other surveys sufficient to identify the hazard, to evaluate individual exposures, and to permit proper selection of respiratory protection equipment.
License No SNM 778 Docket No.70-824 Date April, 1987 Ametuiment No.
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2.
Written procedures to ensure proper selection, supervision, and training of personnel using such protective equipment.
3.
Written procedures to ensure the adequate individual fitting of respirators, as well as procedures to ensure the testing of respiratory protective equipment for operability insnediately prior to each use..
4.
Written procedures for nnintenance to ensure full effectiveness of respiratory protective equipment, including procedures for cleaning and disinfecting, decontaminating, inspecting, repair-ing, and storing.
5.
Written operational and administrative procedures for the control, issuance, proper use, and return of respiratory pro-tective equipment, including provisions for planned limitations on duration of respirator use for any individual as necessi-tated by operational conditions.
6.
Bioassays and other surveys, as appropriate, to evaluate individual exposures and to assess the protection actually provided.
7.
Records sufficient to permit periodic evaluation of the adequacy p
of the respiratory protection program. -
8.
Determination prior to assignment of any individual to tasks requiring the use of respirators that such an individual is physically able to perform the work and use the respiratory protective equipment. A physician is to determine what health and physical conditions are' pertinent. The medical status of each respirator user is to be reviewed at 12-month intervals.
l Other details of an effective respiratory protection program can be found in the above mentioned documents and the health physics procedures.
12.10 OCCUPATIONAL EXPOSURE ANALYSIS 12.10.1 External Exposure - The external radiation exposure received by workers is presented in Tables 12-11 through 12-14. Tables 12-11 and 12-12 show the exposures by ranges and the number of workers in each range for calendar years 1984 and 1985 respectively.
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~.-(d -
N TABLE ~12-11 1984 EXPOSURES BY RANGE Annual Whole Body Dose Number of Individuals Ranges (Rems)
In Each Range No Measurable Exposure 87 Measurable Exposure <0.100 77 0.100 to 0.250 33 0.250 to 0.500 12 0.500 to 0.750 6
0.750 to 1.000 1
1.000 to 2.000 3
2.000 to 3.000 1
3.000 to 4.000 0
4.000 to 5.000 0
5.000 to 6.000 0
6.000 to 7.000 0
7.000 to 8.000 0
8.000 to 9.000 0
9.000 to 10.000 0
10.000 to 11.000 0
11.000 to 12.000 0
>12.000 0
220 License No SNM 778 Docket No. 70 824 Date April 1987 Amendment No.
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TABLE 12-12 l
1985 EXPOSURES 8Y RANGE Annual Whole 8ody Dose NL.mber of Individuals Ranges (Rems)
In Each Range No Measurable Exposure 185 Measurable Exposure <0.100 83 O.100 to 0.250 21 0.250 to 0.500 17 0.500 to 0.750 5
0.750 to 1.000 3
1.000 to 2.000 2
2.000 to 3.000 2
3.000 to 4.000 0
4.000 to 5.000 0
5.000 to 6.000 0
6.000 to 7.000 0
7.000 to 8.000 0
8.000 to 9.000 0
9.000 to 10.000 0
10.000 to 11.000 0
11.000 to 12.000 0
>12.000 0
318 i
License No SNM 778 Docket No. 70 824 Date April,1987 Amendment No.
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U' Table 12-13 presents the exposures received by workers for calendar years 1981 through 1984. The row entitled "Off Site" gives the exposures received by workers at other licensed I
facilities.
TABLE 12-13
]
RADIATION EXPOSURE 1984 1983 1982 1981 Total Person Rems 23.5 18.4 19.4 26.3 Off Site 3.5 2.0 2.5 3.0 LRC 20.0 16.4 16.9 23.3 Average Exposure 0.09 0.088
.105
.137 Number of Workers 220 208 184 192 Highest Exposure 2.25 2.04 1.9 1.7 License No SNM 778 Docket No.70-824 Date April, 1987 Amendment No.
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The exposure received by workers is categorized by group in Table 12-14 for exposures received for calendar years 1983 and 1984.
TABLE 12-14 EXPOSURE BY GROUP (PERSON REMS)
Group 1984 1983 Plant Engineering 5.10 2.25 Project Services 0.07 0.05 Health & Safety 1.85 2.16 Nuclear Materials 11.40 9.60 Chemical & Nuclear Engineering 1.30 1.53 Nondestructive Methods 0.58 0.17 Process Control 0.00 0.49 Systems Design & Engineering 2.12 2.09 O
Calendar year 1984 brought increased activity in our hot cell facility. This typically results in increased exposures to personnel in the Nuclear Materials, Plant Engineering, and Health and Safety Groups. Table.12-14 reflects this in all categories.
Table 12-14 also reflects this increase in two of the three affected groups. Only Health and Safety saw a reduction in the group's exposure. The amount of exposure received from off-site work reversed a three year period of decreases. Table 12-14 I
reflects this in the increase in the Systems Design & Engineering Group's exposure.
The increases noted in Tables 12-13 and 12-14 do not indicate a decrease in the vigilance given by site management to personnel exposures nor do they suggest a decreased ALARA emphasis.
Exposure history at the site shows wide variances because of the I
variety of work that is performed here. Clear trends have not been evident.
If the amount of hot cell work is considered and the fact that objects received for examination exhibit higher levels of radioactivity, the effectiveness of the ALARA program License No SNM.778 Docket No.70-824 Date April, 1987 Amendment No.
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can be appreciated. The preliminary exposure information required on the Radiation Work Permit form was increased in early 1985.
This has resulted in many improvements in the manner that cell entries are made.
12.10.2 Internal Exposure - The bioassay sampling, lung counting, and air sampling programs show that the worker is exposed to extremely low levels of respirable activity.-
12.10.2.1 Bioassay Results - Urine bioassay samples are taken primarily of workers who perform work with unclad uranium and those involved in any work with plutonium. Table 12-15 below presents the l
number of urine bioassay samples taken during 1983 and 1984.
TABLE 12-15 l
NUMBER OF URINE BI0 ASSAY SAMPLES 1983 1984 Month U
Pu U
Pu 20 4
January 5
13 6
February 15 12 March April 19 19 18 9
13 5
May June 16 16 17 8
July 11 8
15 6
August 10 8
15 7
September 11 14 14 7
October 11 9
16 5
November 3
1 December 5
5 14 6
In 1983, all samples for uranium were less than 5 micrograms /
I liter (lower limit of detection), except on four occasions when the analysis indicated the presence of uranium but none met the resample limit of 20 micrograms / liter. All plutonium analyses 1
were below the minimum sensitivity which varied from 0.00 + 0.1 to 0.3 + 0.4 dpm per sample.
l License No SNM.778 Docket No. 70 824 Date April, 1987 Amendment No.
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i Babcock &Wilcox j
a McDermctt company w
-,-e- - - - -. - - -,.
-._.7 maq.
.,,.,y.-p-.
py,+,,g--,m.
w-.,ry,,ep--
-w
--w.w
-4.,-
_-.--n._,%-.
-y
.1 Di)
In 1984, all samples for uraniuni were less than 5 micrograms /
l liter (lower limit of detection), except on one occasion 27 l
micrograms / liter was reported. A resample showed that the level had returned below the lower limit of detection. All plutonium l
samples indicated 0.0 + (0.01 to 0.6) dpm per sample.
12.10.2.2 Air Sampling Results - The air sampling program is the first line of defense for all operations of this type, but the bio-assay progrrm, along with lung counts, is the final step in the estimation of exposure that may occur.
12.10.2.2.1 Table 12-16 presents a summary of the air sampling program for l
calendar year 1983, for fixed air samplers.
TABLE 12-16 l
1983 AIR ACTIVITY (VALUES IN pCi/ml)
Approximate Maximum Labs Average Concentra tion MPC
-15
~14
-10 15 3x10 1.2x10 1x10
-15
~I4
-10
\\v]
/
16 3x10 1.5x10 1x10
-15
~I3
~II 17 8.7x10 1.8x10 4x10
-15
-10 19 7x10 8.7x10"I4 1x10 27 2.4x10"I6 5.7x10 1x10
-15
-10
-15
-15
-10 44*
2x10 6.5x10 1x10
-12
-10
~9 Cask Handling Area 1.9x10 1.27x10 9x10
-15
-13
~II 6.7x10 4.5x10 4x10 Hot Cell 1x10~I4 1.25x10"I3 9x10~9
-16
-15
~II 5x10 1.2x10 4x10 Recirculated Air "C" 1.5x10'I4 3.5x10 9x10
-13
~9 4x10 1.93x10 4x10'II
-15
-14 Waste Storage Area 1.5x10-14 2.6x10"I4 9x10~9 7x10~IO 1.7x10 4x10"II
-15 4
License No SNM 778 Docket No.70-824 Date April,1987 Amendment No.
O Revision No.
4 Page 12-26 Babcock &Wilcox a McDermott company
a, s
by 4
- _ Laundry 3x10"I4 1.5x10 9x10
~13
-9
-15
-14
~11 3x10 2.5x10 4x10
-14
-12
-8 Radio Chem Lab 7x10 2.3x10 9x10
-15.
~14
~II 1.5x10 1.5x10 4x10
- Discontinued in S<spt.
l 12.10.2.2.2 On 338 occasions in 1983, breathing zone air samples were taken to rsasure the airborne activity to which workers were-exposed.
In no case was anyone exposed to greater than 2 MPC of airbcene activity in any one week.
In most cases, respira-tory protection was used and exposure levels were at least a factor of 1,000 below the limits.
There are three major operations which require respiratory protection, and several minor ones.
1.
Entries into the isolation area behind the hot cell. A supplied air. respiratory system was installed in January, 1980, in the hot cell area which has a protection factor of at least 1,000. This-system incorporates-a -double bibb i
O hood which has reduced airborne activity to which a worker is exposed to below detectable levels.
2.
Operations outside of the isolation area in the cask handling area using the 3M hood and the supplied air i
respiratory system. This system incorporates the 3M hard i
hat which is NIOSH approved with a protection factor of l
1,000. Breathing zone samples are taken outside of the hood each time this system is used.
I 3.
Operations in Building C may involve bagging operations with plutonium glove boxes. All operations of this type r
require respiratory protection. When it is used, a breathing zone sample is taken. Normally, the powered respirator with 1,000 protection factor is used; however, i
the full face masks with a protection factor of 50 may be j
used.
l.
4.
Other minor operations requiring respiratory protection l
are: changing HEPA filters, repair work on NPD site I
License No SNM 778 Dooket No.70-824 Date April,1987 t
i i
Amendment No.
O Revision No.
4 Pese 12-27
!O i
Babcock &WHcox i
a McDermott company i
f
e4 support equipment, and any other operations where Health and Safety believes that there is a potential of airborne activity.
5.
It should to be noted that a major operation is occurring in the decommissioning of Building C that is requiring the use of respiratory protection for industrial safety reasons, not for protection from radioactive materials. A number of operations are very dusty (paint chipping, concrete destruction, etc.). A NIOSH approved full flow hard hat system is used. With no protection factor, no one in Building C has been exposed in excess of 2 MPC hr in one week.
In most cases, radioactivity above back-ground is undetectable.
12.10.2.2.3 Table 12-17 presents a summary of the air sampling program for l
calendar year 1984, for fixed air samplers.
TABLE 12-17 1984 AIR ACTIVITY (VALUES IN 9C1/ml)
~
Approxima te Maximum Labs Average Concentra tion MPC 15*
SE-15 3.9E-13 4E-11 19 7E-15 1E-13 1E-10 27**
2.4E-14 7.5E-15 IE-10 Soil Processing ***
1E-15 7.4C-15 4E-11 Cask Handling Area SE-13 1.2E-11 9E-9 SE-15 SE-13 4E-11 License No SNM 778 Docket No.70-824 Date April,1987 Amendment No.
O Revision No.
4 Pass 12-28 Babcock &Wilcox a McDermott company
)
'V SE-16 1.5E-14 4E-11 Recirculated Air 1.5E-14 1.1E-13 9E-9 Building C 1.5E-15 3.3E-13 4E-11 Waste Storage 1.5E-14 2.9E-14 9E-9 7E-16 3.3E-15 4E-11 Laundry 3E-14 6.3E-14 9E-9 2E-15 3.3E-15 4E-11 Radio Chem 3E-14 1.0E-12 9E-9 l
1.5E-15 2.4E-15 4E-11
- Discontinued November 1984
- Discontinued June 1984
- Begun May 1984 12.10.2.2.4 On 278 occasions in 1984, breathing zone air samples were taken to measure the airborne activity to which workers were exposed.
In no case was anyone exposed to greater than 3 MPC hour of airborne activity in any one week.
In most cases.
/
respiratory protection was used and exposure levels were at L
least a factor of 1000 below the Ilmits.
There are three major operations which require respiratory protection, and several minor ones.
I 1.
Entries into the isolation area behind the hot cell. A supplied air respiratory system was installed in January, 1980 in the hot cell area which has a protection factor of at least 1000. This system incorporates a double bibb hood which has reduced airborne activity to which a worker is exposed to below measurable levels.
l 2.
Operations outside of the isolation area in the cask I
handling area use the 3M hood and the supplied air i
respiratory system. This system incorporated the 3M hard hat which is NIOS't approved with a protection factor of 1000.
Breathing zone samples are taken outside of the hood each time this system is used.
License No SNM 778 Docket No. 70 824 Date Apr11,1987 l
l Amendment No. 'o Hewision No.
4 Page 12-29 O
Babcock &Wilcox i
l A McDermott company e
n 3.
Operations in Building C may involve bagging operations h
with plutonium glove boxes. All operations of this type require respiratory protection. When it is used, a breathing zone sample is taken. Normally, a 20T air line respirator with a 1000 protection factor is used; however, the full face mask with a protection factor of 50 may be used.
4.
Other minor operations requiring respiratory protection are: changing of HEPA filters, repair work on NPD site support equipment, and any other operation where Health and Safety believes that there is a potential of airborne activity.
5.
It should to be noted that a major operation is occurring in the decommissioning of Building C that is requiring the use of respiratory protection for industrial safety reasons, not for protection from radioactive materials. A number of operations are very dusty (paint chipping, con-crete destruction, etc.).
A NIOSH approved full flow hard hat system is used. With no protection factor, no one in Building C has been exposed in excess of 2 MPC hours in one week.
In most cases, radioactivity above background is undetectable.
12.10.2.3 In-vivo Results (1983) - Whole body counting was performed by O
Helgeson Scientific Services, Inc. on 32 workers during 1983.
l Three had detectable activities, no other workers indicated detectable activity. The results of the three workers with detectable activity is presented in Table 12-18.
TABLE 12-18 I
WHOLE BODY COUNTS - 1983 (ALL VALUES IN NAN 0 CURIES)
Worker Isotope MPB8 1
2 3
Cs-137 3E4 8+2 4+2 Mn-54 3.6E3 5+2 4+1 Co-60 1.1E3 3+1 7+1 License No SNM 778 Docket No. 70 824 Date April,1987 Amendment No.
O Revision No.
4 Page 12-30 O
v Babcock &Wilcox a McDermott company
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In-vivo counting was performed on seven workers during 1983, for I
plutonium and Americium-241.
These results are summarized in Table 12-19.
I I
TABLE 12-19 Am - Pu LUNG COUNTING 1983 (ALL VALUES IN NAN 0 CURIES)
Worker P_u Am u
1 0
0.00+0.10 2
0 0.00+0.11 3
0 0.13+0.13 4
0 0.00+0.14 5
0 0.00+0.15 6
0 0.00+0.19 7
0 0.00+0.16
~
O v
License No SNM-778 Docket No.70-824 Date April,1987 Amendment No.
O Revision No.
4 Page 12-31 Babcock &Wilcox a McDermott company
(~x In-vivo lung counting was perfornutd on nine workers in 1983, for I
uranium. The results are listed in Table 12-20.
Four of the I
nine indicated positive results. However, these results were not confirmed, in followup urinalyses.
I TABLE 12-20 URANIUM LUNG COUNTING
- 1983, (ALL VALUES IN MICROGRAMS)
I Worker U-235 1
0+30 2
0+43 3
0139 4
42+37 5
0+41 6
38+33 p
7 76+45 d
8 0_+39 9
49+44 License No SNM-778 Docket No. 70 824 Date April,1987 Amendment No.
0 Revision No.
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12.10.2.4 In-vivo results (1984) - Whole body counting was performed by Helgeson Scientific Services, Inc. on 99 workers during 1984.
I Twelve had positive results but these were very low levels. A summary is presented in Table 12-21.
l I
TABLE 12-21 WHOLE BODY COUNTS 1984 (EXPOSURE VALUES IN NAN 0 CURIES)
Number of Maximum Isotope Workers Observed MPBB l
Cs-134 1
3.0 2E4 Cs-137 7
9.0 3E4 Co-60 4
4.0 1.1E3 In-vivo lung counting was performed on 14 workers during 1984 for l
O Plutonium-239 and Americium-241.
No plutonium was reported.
The presence of Americium-241 was indicated for 5 workers with the i
highest quantity being 0.26 NanoCuries (+0.14) for one person.
In-vivo lung counting was performed on 20 workers during 1984 for l
Uranium-235.
In 5 instances, the results were positive with the highest result being 48 micrograms (+37) for one person.
l l
l l
l License No SNM-778 Docket No.70-824 Date April,1987 i
Amendment No.
O Revision No.
4 Page 12-33 Babcock &Wilcox a McDermott company
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o 12.11 MEASURES TAKEN TO IMPLEMENT ALARA 12.11.1 Irradiated metal specimens had been stored on.the roof of the hot cells in an open top cave. This configuration caused this roof to be _ designated as a.high radiation area.
Several small totally enclosed caves have been constructed for the storage of these specimens which has eliminated the high radiation area on the cell roof, thus reducing exposures received by workers who periodically l
enter the area for maintenance on the HEPA filters and to cali-brate an area monitor.
It also eliminated the radiation area on 1-the roof of Building B which no longer contributes to the exposure j
of workers who maintain the building ventilation system.
h
Cleaning of the hot cells contributed significantly to exposure 12.11.2 doses of workers. This cleaning operation, which is performed at three or four year intervals, requires the set-up table in the L
cell to be dismantled.
In 1985, this operation was performed remotely with a modified saw so that workers did not enter the
{
cell.for this high exposure work.
_12.11.3 Trash removal from the hot cell. during cell cleaning operations E
was significant in the past.
During the cleaning operation in 1985, trash was remotely loaded into special_ metal drum liners that were designated to fit into 30-ga11on drums and to be handled.
with long poles. -This process modification reduced personnel exposures for this part of the operation considerably.
12.11.4 The site has purchased a TLD reader which provides immediate in-l-
formation on worker exposure. This system is not intended to replace the normal contract service for dose measurement but i
rather to provide prompt indication of unexpected exposures for non-routine operations. The system makes possible the estimation of exposures to hard to measure areas of the body such as the soles of feet, hands and fingers.
4 12.11.5 A supplied air respiratory system has been installed to support j
hot cell work, principally during hot cell entries. This system L
provides a greater protection factor for workers in addition to providing greater worker comfort while performing the strenuous work.-
12.11.6 The Radiation Work Permit (RWP) approval process has been revised.
Previously, the worker or. his supervisor completed the RWP form and carried it to those personnel who were required to sign it.
1 License No SNM-778 Docket No.70-824 Date April,1987 Amendment No.
O Revision No.
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} _l This method has been changed such that the workers, Area Super-l visors and signators of the RWP gather at a
- meeting wnere the proposed work scope.and methods are discussed in detail. All facets of work are agreed to before any authorization signatures are placed on the RWP.-_This new approval process requires more time being. spent for the planning stage of a -task but considerable exposure savings have resulted.
12.12 BI0 ASSAY PROGRAM Those workers routinely working in contamination or airborne radio-1 activity areas will be scheduled for participation in the bioassay
. program. The Health and Safety Group will' select those workers to l-be sampled in the program. This selection will be based on the probability of exposure, the worker's work habits, the type of work I
in the area air sample data, previous bioassay data, etc. _ Routine bioassay may consist of check or whole-body counting (in-vivo bioassay) or excretion analysis (in-vitro bioassay).
In-vivo bioassay is performed routinely by a bioassay service which comes
[
' on-site for the _ evaluations.
In-vitro bioassay is performed by. a commercial laboratory located off-site.
Bioassay action criteria for plutonium are outlined in Table 12-3 &.
~
'12-4.
In general', no. action is required if the excretion result (i.e., urinalysis) is less than 0.2 dpm/ liter or the in-vivo measurement of material in the lung is less than 16 nanoCuries. All compounds of plutonium are considered to be either class W or Y.
This classification refers to the most recent evaluation of the ICRP for internal dose calculations. Class W compounds are moderately soluble and clear from the pulmonary region of the lung with half-times in the range 10 to.100 days. Class Y compounds are essentially insoluble and are ~ considered to clear from the pulmonary region with _ half times of.>100 days. No compounds of plutonium are considered by the ICRP to be readily soluble (i.e., class D compounds which clear from the lungs in <10 days).
The bioassay program for uranium generally follows that outlined in
' Regulatory Guide 8.11, " Application of Bioassay For Uranium," June l.
1974. There are two exceptions to this general guidance:
('
1.
Workers off-site during the regular visit of the bioassay I
service will not be scheduled for a special, make-up count, if the count was scheduled only for routine exposure control mo7iT-toring.
License No SNM-778 Docket No.70-824 Date April,1987 Amendment No.
O Revision No.
4 Page 12-35 I
Babcock &Wilcox a McDermott company
... -.. _ _ _ _. ~. _. _ _ _ _ _ _. _ _ _.. _ _ _ _ _ _ _. _ _ _ _ _ _ _ _ _ _. _ _ _. _., - _ -. _,, _ _ - -
(j 2.~
Bioassays of workers working in areas in which both plutonium I
and uranium may be airborne shall be evaluated for both plutonium and uranium. The Supervisor, Health and Safety may decide to analyze for only one of these elements, if it can be demonstrated that the analysis for a single element is a more sensitive indicator of an uptake.
Bioassay action criteria for uranium are outlined in Table 12-5 &
12-6.
Workers working primarily with beta and gamma emitting radionuclides I
will also be included in the in-vivo bioassay analysis program. Any worker suspected of an exposure greater than 40 MPC-hours will be I
scheduled for a bioassay evaluation as soon as practicable after the exposure.
Bioassay action criteria for beta-gamma are outlined in Table 12-7.
12.13 AIR SAMPLING AND MONITORING
- The presence of airborne radioactive materials in the working areas i
is determined through the combined use of air samplers and monitors.
These. programs are discussed below:
1 12.13.1 Air' Sampling Program s
l l
The air sampling program can be divided into two categories; fixed i
and portable.
Selection of the sampling category and the frequency of sampling is left to the discretion of the Supervisor, Health and Safety.
12.13.1.1 Fixed Air Samplers - Air samples are obtained at designated points through the use of a central vacuum system.
Sampling points are located as close as possible to a permanent operator station to permit continuous sampling of the air near the worker's breathing zone. These samples are usually collected weekly.
However, the frequency may vary as the situation dictates.
Normally, these are evaluated within two weeks, af ter allowing the appropriate decay period for the radon daughter products.
However, based on the particular operation, etc., a Health Physics Engineer may determine that it is necessary to evaluate the samples without allowing for the decay period.
In these 1.icense No SNM-778 Docket No.70-824 Date April,1987 Amendment No.
O Revision No.
4 Page 12-36 AV Babcock &Wilcox a McDermott company i
d cases, an applicable radon decay correction factor must be applied to the results.
12.13.1.2 Portable Samplers - Air samples in the approximate breathing zone of a worker may be obtained through the use of a lapel sampler. The lapel sampler consists of a small sampling head attached to tre worker's lapel (or collar) connected through a small flexible tube to a small air-pump worn at the waist. The flow rates through these samplers are quite low when compared to the fixed system. However, since the. sampler is located near the nose and nouth and moves with the worker as he moves about the area, it provides a reasonable estimate of the concentration of airborne radioactivity in the breathing zone of the worker.
Air samples obtained with these samplers are evaluated on a low background, proportional counting system.
Factors are applied to the counting results to account for background activity and detector. efficiency. All results are reported in units of activity / unit volume of air sampled.
12.13.2 Air Monitoring Program Air monitoring in operating areas is accomplished with continuous monitors in predetermined, fixed locations.
Normally, a monitor is placed in each radioactive materials handling area in which-
-(]
there is a potential for the release of airborne radioactivity.
Locations are selected based upon the ability of the monitor to provide a reasonable evaluation of the airborne activity in a particular area and to provide adequate warnings to those in the area of changing conditions. These determinations are made by the Health and Safety Group based upon the operations in the area, the potential for release, and the quantity and chemical form of the ma terial.
Alarms are set in accordance with the particular operation, the material being handled, and the potential for release.
Actual alarm points are set as low as possible commensurate with the ambient radiation levels in the area, i
License No SNM-778 Docket No.70-824 Date April,1987 Amendment No.
O Revision No.
4 Page 12-37 Babcock &Wilcox a McDermott company
N N
12.14 SURFACE CONTAMINATION
]
12.14.1 Smear Surveying Smear surveys are performed in all areas specified in the license and which, in the judgment of the Supervisor, Health and Safety, have a potential for surface contamination. The frequency of these surveys will be based upon the potential for contamination in the area, previous experience with contamination in the area, and the need to keep the area free from contamination. Typical areas and survey schedules are listed in Table 12-9, however, both the areas included and the frequencies of surveys are subject to change based upon the current research activities. The frequency of. smear surveys in areas not included in the table are generally specified in the procedure covering the particular area.
12.14.1.1 Smear Samples - Smear samples are obtained with small, absorbent filter papers. The smear paper is moved across an area of approximately 100 sq. cm. using about 5 pounds of pressure. The smear may be counted with a portable gas-flow proportional counter capable of detecting alpha or beta radiation. Normally, smear samples are evaluated in a stationary counter located in the Health Physics Laboratory. Appropriate conversion factors are applied to the net counts to express the smear results in units of disintegrations per minute.
(
12.14.1.2 Large Area Smears - Large area smears are obtained using the dust mop technique in areas around the site, the hot cell w
operations area, the change room and main hallways in Building B.
These smears are intended to indicate the general contami-nation environment in an area and may lead to a more extensive survey, if unexpected contamination is indicated. Normally, large area smears are evaluated with a hand-held, portable survey instrument (e.g., a gas-flow proportional counter such as the PAC 4G). Actions to be taken in response to the results of large area smears are outlined in Table 12-22.
l 12.14.1.3 Action Levels - Included in Table 12-24 are the appropriate I
action levels to be used in designated areas. Decontamination shall be initiated in areas in which the removable surface contamination levels exceed these action levels. The Health and Safety Group shall determine and direct the actions to be taken to protect workers working in these areas and to reduce contami-l nation levels as far below those listed in Table 12-1 as is possible. Normally, decontamination of an identified area shall begin within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of the discovery.
License No SNM-778 Docket No.70-824 Date April,1987 2
Amendment No.
O Revision No.
4 Page 12-38 Babcock &Wilcox a McDermott company
=-
In some. cases, for example, if the contamination is discovered just prior to a weekend or a regularly scheduled holiday, the contaminated area may be marked.and posted appropriately. _Such a determination'shall be made by the Health and Safety Group based upon the severity and extent o." the contamination and the potential for further contamination of equipment and/or i -
personnel during the interval. Decontamination of the area shall begin on the first regular work-day after discovery.-
TABLE'12-22 l
ACTION LEVELS-FOR LARGE AREA SMEARS 1.
Routine Large Area Smears (1000 - 5000 dpm)
Repeat the large area smear.
If results show levels of contamination above 1000 dpm, take smears in smaller areas to locate the source.
Decontaminate all areas in which the smear results indicate contamination above 1000 dpm per 100 sq. ft.
2.
Routine Large Area Smears (5000 - 10,000 dpm)
Repeat the large area smear.
If results show levels of contamination above 5000 dpm, isolate the contaminated area.
Take smears in smaller areas to locate the source.
Decontaminate all areas in which the smear results show contamination _ in excess of -1000 dpm per 100 sq. ft.
3.
Routine Large Area Smears (>10,000 dpm)
Isolate the contaminated area..
Survey all personnel in the contaminated area.
Take smaller smears in the area to locate the source.
t License No SNM-778 Docket No.70-824 Date April,1987 Amendment No.
O Revision No.
4 Page 12-39 I
Babcock &Wilcox a McDermott company 1
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1
/]
Decontaminate all areas in which the smear results show V
contamination in excess of 1000 dpm per 100 sq. ft.
Survey all persons leaving the building.
NOTE:
Routine large area smears are normally taken in the early after-noon to facilitate clean-up of areas found to be contaminated before the end of the normal work-day.
TABLE 12-23' SMEAR SURVEY FREQUENCIES AND ACTION LEVELS Alpha Radiation Smear Survey Action Level Area Frequency (dpm/100 sq. cm.)
Unirradiated, unencapsulated weekly 5,000 fuel handling areas f%g Building B counting laboratory monthly 200 Hot cell operations area monthly 200 Scanning electron microscopy monthly 200 laboratory Exit portals from controlled twice weekly 200 Beta Radiation Smear Survey Action Level Area Frequency (dpm/100 sq. cm.)
Building B Counting Laboratory monthly 2,000 Scanning Electron Microscopy monthly 2,000 Laboratory l
License No. SNM-778 Docket No.70-824 Date April,1987 Amendment No.
O Hevision No.
4 Page 12-40 l
Babcock &Wilcox a McDermott company I
1
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x : t 1 h . c':- ~ F 6%_ 9 + s, License No SNM 778 Docket No. 70 824 Date April,1987 0 4 13-4 l Amendment No. Revision No. Page Babcock &Wilcox a McDermott company N TABLE OF CONTENTS Section Page 14.0 NUCLEAR CRITICALITY SAFETY 14-1 14.1 ADMINISTRATIVE AND. TECHNICAL PROCEDURES. 14-1 14.2 PREFERRED APPROACH TO DESIGN 14-2 14.3 BASIC ASSUMPTIONS 14-2 14.3.1 Nuclear Isolation 14-2 l 14.3.2 Building A 14-3 14.3.3 Building B 14-8 l 14.3.4 Building C 14-16 14.3.5 Outside Storage. 14-17 l 14.3.6 Dry Wasta. 14-17 l 14.4 ANALYTICAL METHODS AND YALIDATION REFERENCES 14-17 l 14.5 OATA SOURCES. 14-18 l 14.6 FIXED POISONS 14-19 l 14.7 STRUCTURAL INTEGRITY 14-19 l 14.8 SPECIAL CONTROLS 14-19 l l l i License No SNM 778 Docket No.70-874 Date April,1987 l l 0 4 14-1 Amendment No. Revision No. Page O Babcock &Wilcox a McDermott compary 1 ~ d - (] List of Tables G Table Page 14-1 K,ff TID-7016 14-4 14-2 K,f f FOR PRESENT LIMIT 14-5 14-3 K,ff FOR 6x6x6 ARRAY OF 850 g U-23S UNITS ON 30-INCH CENTERS. 14-6 l 14-4 K,ff FOR ARRAYS OF 850 GRAM U-235 UNITS ON 24 AND 36-INCH CENTERS 14-7 l 14-5 COMPARISON OF THE MARK B AND MARK C FUEL ASSEMBLIES. 14-10 l 14-6 REACTIVITY FOR MARK B AND MARK C FUEL ASSEMBLIES UNDER DISMANTLEMENT 14-13 14-7 K-EFFECTIVE OF INDIVIDUAL MARK B AND MARK C FUEL ASSEMBLIES. 14-14 l 14-8 REACTIVITY-FOR AN INFINITE BY 14-UNIT ARRAY Os GF FUEL ASSEMBLIES. 14-15 l List of Figures Figure Page 14-1 FUEL R0D REMOYAL SCHEMATIC - MARK B 14-20 l 14-2 FUEL R00 REMOVAL SCHEMATIC - MARK C 14-21 l l l License No SNM-778 Docket No.70-824 Date April,1987 O 4 14"II Amendment No. Revision No. Pe Babcock &Wilcox a McDermott company 1 () 14.0 NUCLEAR CRITICALITY SAFETY Y 14.1 ADMINISTRATIVE AND TECHNICAL PROCEDURES The ultimate responsibility for nuclear criticality safety rests with the Manager, EC&RR. However, first-line responsibility is with the l Facility Supervisor supported by the Nuclear Criticality Safety l Officer. I The Nuclear Criticality Safety Officer is generally responsible for establishing nuclear safety limits and nuclear safety considerations in operating procedures, processes, and the like. His duties are shown more specifically in the following statement. I The position of Nuclear Criticality Safety Officer has been estab-I lished at the site. It will be this officer's responsibility to I ensure, as far as possible, that no operations on site can lead to the inadvertent assembly of a critical mass. To this end, he will review all new procedures which involve the handling of special nuclear materials as well as changes in old procedures, observe oper-ations, inaugurate educational programs if and when he deems them necessary, and carry out confirming criticality calculations. This appointment does not in any way relieve the Facility Supervisor of his responsibilities for ensuring the safety of operations, nor n) will it eliminate the necessity for the reviews by the Safety Review ( Committee required by the license. Once a quarter the Nuclear Criticality Safety Officer or qualified I person designated by him will inspect all site operations where I special nuclear materials are being processed. Other areas shall be inspected less frequently; however, all areas shall be inspected at least once a year. He shall consider area operations when schedul-ing these inspections and shall, if necessary, schedule his inspection at more frequent intervals. His consideration should include inspection of new facilities, inspection of hazardous non-routine operations, an audit of nuclear criticality safety records, a check for area posting and a review of current practices. A written report is to be filed with the Panager, EC&RR quarterly l with a copy to the License Administrator. Prior to submission of the report, he shall discuss any findings with the Facility Supervisor. The report shall be brief, concerning itself with inspections mide during the quarter and with the nuclear criticality safety activity of the quarter. License No SNM 778 C wket No.70-824 Date April,1987 Amendment No. O Revision No. 4 Page 14-1 V Babcock &Wilcox a McDermott company El The following information is to be included: . N' o Areas visited o Operations observed o Unsafe practices or situations noted o Nuclear safety activity of the quarter (brief summary) o Reconsnendations o Resolution of previous recommendations. 14.2 PREFERRED APPROACH TO DESIGN Research and Development activities are performed at the site. While I the use of safe geometry is the preferred approach in a production facility, it is not appropriate nor practical at a research labora-I tory. Since most projects require only small amounts of SNM on labo-ratory benches and in hoods, the preferred approach at the is through safe masses ir simple arrays; the lattice density model or arrays found in TID-7016, Rev.1 is the adopted model. The one exception to use of safe masses is when examining and testing reactor fuel assem-blies. The approach to such uses is to accept only a-limited number ,s (] of fuel assemblies and then to maintain the fuel of an assembly within the dimensional envelope of the original assembly's dimensions. Where this is not possible, the fuel of an assembly is handled within the dimensions of safe geometry or as a safe mass. 14.3 BASIC ASSUMPTIONS This section describes basic assumptions and evaluations that have been made to demonstrate nuclear criticality safety for the speci-fications of Section 4.2 (Technical' Requirements for Nuclear Criti-cality Safety). 14.3.1 Nuclear Isolation - Special nuclear material is isolated from all other special nuclear material for nuclear criticality safety pur-poses if any of the three conditions (or equivalent) listed in 4.2.1 are met. These three isolation criteria are accepted industrial practice for maintaining nuclear criticality safety. It is recognized that 12 inches of high density concrete may not be License No SNM 778 Docket No.70-824 Date April,1987 l l Amendment No. O Revision No. 4 Page 14-2 i O V t i Babcock &Wilcox a McDermott company I /7 adequate as isolation between two large parallel slabs of SNM; this (,/ does not describe any SNM configuration and will not be permitted without additional evaluation and NRC approval. 14.3.2 Building A 14.3.2.1 General - From Figure 22. TID-7016, Revision 1, 74 units is read as the maximum allowable number of units in a cubic array on 24 inch centers (with at least 8 inches edge-to-edge between units), assuming full reflection on the array. The number of allowable units has been reduced from 74 to 40 units to permit use of the 850 grams of U-235 at low enrichment as a unit. The subdivisions defining a unit are for clarification of the general definition of a unit as any physically identifiable accumulation of SNM. The terminology of TID-7016, Revision 1, applies. 14.3.2.2 Mass Limits 1. The mass limits for plutonium, U-233, and U-235 are based on the recommended limits in Table I, TTO-7016 (Rev.1). The values for Pu-235-U mixtures in 4.2.2.2.1 were derived to satisfy the following relationship: grams Pu fissile # grams U-235,1 220 350 2. The values for U-233 - Pu and U-233 - U-235 mixtures were found by taking the lowest limit of any isotope in the m1 xture. 3. From DP-1014, uranium metal-water lattices which have the minimum U-235 mass at critical are 2.36 kg for 3.0 wt% U-235 and 1.47 kg for 5.0 wt% U-235. A conservative interpolation between these two points gives 1.9 kg at 4.0 wtt U-235; 45% of this is 850 g U-235. The present array control is based on the lattice density model using Figure 22 and Table IV (modified) in TID-7016 (Rev. 1). Our calculations demon-strate that the 850 gram unit is an allowable unit if the number of units permitted in TID-7016 (Rev.1) is set at 40. All computer calculations were made using either the NULIF code for fully reflected spheres or with the Monte Carlo code KENO. Four series of computer calculations were made. Tables 14-1 and 14-3 summarize the results. License No SNM 778 Docket No.70-824 Date April,1987 Amendment No. O Revision No. 4 Page 14-3 1 1 Babcock &Wilcox l a McDermott company f p. 1 14-1 Table for determination of K for the mass limits listed in TID-7016 fof fattice density f model at the upper H/X limit (made with NULIF). 14-2 Table for determination of K Vs H/X for 349 grams of U-235 contained in ibfly enriched uranium metal (made with NULIF). The K values for the lattice density limits ranged from 0.800 to 0.8N and are tabulated in the following table. TABLE 14-1 g7f TID-7016 Sphere
- Mass, Radius, g
g Kg U-235 cm H/U-235 H/ Total U eff = 10.0 6.828 2.0 1.87 1.86 0.800 9.0 7.182 3.0 2.81 1.84 0.804 O' 7.3 7.562 5.0 4.68 1.82 0.803 5.2 8.178 10.0 9.36 1.81 0.815 3.6 8.922 20.0 18.71 1.84 0.854 License No SNM-778 Docket No. 70 824 Date April,1987 Amendment No. O Rwision No. 4 Page 14-4 , O Babcock &Wilcox a McDermott company ( TheK[f350) T H/X are given in the following table. values for the present limit (349 was used instead of 'v TABLE 14-2 (ff FOR PRESENT LIMIT Sphere Mass
- Radius, g U-235 cm H/U-235 H/ Total U K.
Eeff 349 13.32 736.8 689 1.49 0.780 349 9.82 293.8 275 1.76 0.753 349 7.79 146.2 137 1.86 0.671 By analogy, values for Pu would be similar. O License No SNM-778 Docket No.70-824 Date April,1987 Amendment No. O Revision No. 4 Page 14-5 Babcock &Wilcox a McDermott company (~ The effect of interspersed water moderation in a concrete V} reflected finite array of 850 gram U-235 units is shown in Table 14-3. These data show that maximum array multiplication occurs with almost no interspersed water. TABLE 14-3 (ff FOR 6x6x6 ARRAY OF 850 g U-235 UNITS ON 30-INCH CENTERS ( y = 18.14 cm, = 0.85 9 U/cc) Volume g Fraction H2O eff + 2o 1.00 0.833 + 0.010 0.15 0.826 + 0.010 0.10 0.851 + 0.010 0.07 0.868 + 0.009 0.05 0.907 + 0.011 0.03 0.924 + 0.009 p () 0.02 0.931 + 0.009 0.01 0.937 + 0.009 0.001 0.930 + 0.009 License No SNM.778 Docket No. 70 824 Date April,1987 Amendment No. O Revision No. 4 Page 14-6 Babcock &Wilcox a McDermott company ) (V The effect of varying the number of 850 U-235 units in a concrete reflected array while maintaining a constant center-to-center spacing with void between them is shown in Table 14-4. Interpo-l lating between the heterogeneous values for the 24 inch center-to-center system predicts a K 2 for 40 units of 0.936 + 0.12 whereas the 36 inch spacTkh + system has about 512 units fiir the same K,ff. TABLE 14-4 K FOR ARRAYS OF 850 GRAM U-235 UNITS ON gf 24 AND 36 INCH CENTERS Array Number Center-to-Center g Size of Units eff f_2o Spacing, in. 4x3x3 36 0.910 + 0.010 24 4x3x3 36 0.929 + 0.013* 24 4x4x3 48 0.932 + 0.010 24 4x4x3 48 0.947 + 0.011* 24 Q 4x4x4 64 0.949 + 0.011 24 4x3x3 36 0.807 + 0.011 36 4x4x4 64 0.845 + 0.011 36 5x5x5 125 0.860 + 0.010 36 6x6x6 216 0.881 + 0.009 36 7x7x7 343 0.912 + 0.009 36 8x8x8 512 0.938 + 0.008 36 9x9x9 729 0.949 + 0.009 36
- Assumes heterogeneous UO -water mixture.
2 License No SNM 778 Docket No.70-824 Date April, 1987 Amendment No. O Revision No. 4 Page 14-7 O Babcock &Wilcox a McDermott company N [) From these data it is concluded that for 850 grams, U-235 per unit V an array of 24-inch centers should be safe for 40 units or less and on 36-inch centers, an array would be safe with 90 units or less. A slight' increase in the array multiplication, on the order of 1%, may occur for low levels of interspersed water modera tion. However, the safety of these arrays would still be maintained.,- To avoid confusion and possible mistakes, additional procedural controis are applied when low-enrichment limits are used. These preclude enrichment combinations of below and above 4.0 wt% U-235. (These are not necessarily unsafe - no calculations were made and no such combinations are desired.) 4. The unit and its limit (laboratory, furnace, transfer cert, etc.) are established by the Facility Supervisor, who author-izes posting the limit showing the maximum quancity of plutonium, U-233, and U-235 allowed. The fissile material content of the material transferred to or from a unit is established from process records, analyses, or previous analytical data. Only authorized users of SNM may transfer SNM between units and must do so only according to approved procedures. A board, sign, or other acceptable device is used to record the new balance and compares to balance with p the unit limit. 'd 14.3.3 Building B 14.3.3.1 General - The demonstration for units and the array is identical to that of Building A (14.3.2.1 & 14.3.2.2). 14.3.3.2 Hot Cell - The demonstration for the units and array is identical to that of Building A. The individual hot cells are isolated from all other arrays by a minimum of 2 feet of high density concrete. 14.3.3.3 Underwater Storage - Transfer Canal - Underwater aluminum or stainless steel storage racks are constructed to ensure 12-inch edge-to-edge spacing of each unit. Units are limited to those in 4.2.2.2.1 & 4.2.2.2.2 excluding PWR fuel assemblies and, since they are separated by 12 inches of water, units are considered i sola ted. Therefore, any number of these units may be used. i License No SNM.778 Docket No.70-824 Date April,1987 Amendment No. Revision No. O 4 Page 14-8 L) ( i Babcock &Wilcox ( a McDermott company l /^' Racks and fixtures are constructed with sufffcient integrity and \\, strength to withstand reasonable structural c'eformity, thereby providing the spacing previously outlined. Supervisory approval is required for removing or inserting any subcritical unit out of or into its storage rack. There is no credible way in which water can be lost from the storage pool and transfer canal. However, assuming loss of water, stored units would drain and be unmoderated and sub-cri tical. 14.3.3.4 Underground Storage Tubes - Underground storage tubes are 5 inches in diameter, approximately 10 feet long, and on 17 inch centers (minimum) in a straight line. Material stored is first placed in a storage can with an inside diameter of 4-1/2 inches. Maximum units demonstrated safe in Section 14.3.3.2 are stored one per tube.. These are nuclearly isolated frcm each other by 12 inches of concrete (minimum). The average edge-to-edge separation approximates 13 inches of concrete. 14.3.3.5 Power Reactor Fuel. Assemblies I i 14.3.3.5.1 General - The site will receive and examine PWR fuel assemblies for both nondestructive and destructive examination. Irradi-ated assemblies will have been subjected to a reactor environ-ment. From a nuclear criticality safety viewpoint,-these assem-(m) blies are in their most reactive state when fresh or unirradi-a ted. Therefore, nuclear safety is demonstra ted by appropriate evaluation of the unirradiated assembly. The current plans call for examination of B&W-manufactured fuel assemblies from B&W power reactors. The current models of interest are designated as the Mark B and Mark C canless assembly. The Mark 8 assembly is described in the SNM license for B&W's Commercial Nuclear Fuel Plant (SNM License No.,1168, Docket 70-1201). In 7.10 of of the unrodded and Section III in SNM License 1168, the K*N shown to be 0.92 at fully moderated and reflected assembly maximum enrichment. Maximum enrichment is defined as 4.0 per-cent nominal which could go to 4.05 percent in manufacturing. b Table 14-5 shows a comparison of the Mark C and Mark B assem-blies. The K of the Mark C asserb h under the same condi-tions listed $bve has a value of M. The reactivity as well as the spectral and physics Eidct f these assemblies are essentf ally the same. All o' t c na ' ear safety calculations shown in this section were ma a wi. Cie Mark B assembly model (except Tables 14-5 and 14-6). Results were obtained for a License No SNM 776 Docket No.70-824 Date April, 1987 l Amendment No. O Revision No. 4 Page 14-9 l O l a 1 Babcock &Wilcox a McDermott company ,w-- ('] fully reflected infinite array 12 inch edge-to-edge of maxi-V mumly enriched assemblies that were fully moderated, i.e., under water. The Mark B and C assemblies are to be disas-sembled in air only in an unirradiated state. The similarity in nuclear characteristics and the large decrease in reactivity in air-moderated assemblies ensure nuclear safety during the disassembly operations. Conditions given in 4.2.3.6.1 are sufficient to ensure that these two assembly types are indeed those to be examined. A damaged assembly which is restrained to 8.6 inches on a side will be no more reactive in air or water even if part of the fuel is missing; this will be demon-strated in Section 14.3.3.5.3. TABLE 14-5 COMPARISON OF THE MARK B AND MARK C FUEL ASSEMBLIES Mark B Mark C Fuel assembly array 15 x 15 17 x 17 Fuel assembly dimensions, in. 8.45 x 8.45 8.536 x 8.536 Control rod tubes per assembly 16 24 b Instrument tube per assembly 1 1 Fuel rods per assembly 208 264 Fuel rod pitch, in. 0.568 0.501 Fuel active height, in. 144 143 Pellet OD, in. 0.370 0.324 Theoretical density, % 92.5 94.0 Enrichment, % 4.0 4.0 Fuel rod clad ID, in. 0.377 0.332 Fuel rod clad OD, in. 0.430 0.379 Fuel rod clad material Zr-4 Zr-4 V IY in fuel rod cell 1.65 1.68 water fuel 1 I License No SNM-778 Docket No.70-824 Date April,1987 Amendment No. 0 Revision No. 4 Page 14-10 1 0 1 v l Babcock &Wilcox a McDermott company m. =. (n -) V IY in assembly with water fuel water completely filling control rod and instrument cells 1.90 1.98 K,ff of one assembly in H O 0.92 0.92 2 Since The Babcock & Wilcox Company is continuing to improve its assemblies and will supply reload fuel to reactors initially fueled by other reactor manufacturers, the site may destruc-l tively examine other types of assemblies. The conditions given in 4.2.3.6.1.1 for additional evaluation are adequate to ensure nuclear safety for different assemblies. Acceptance of BWR fuel assemblies for study is acceptable if the assemblies have a maximum enrichment of 4.05 wt% U-235 and have a cross sectional area not exceeding that of a 22.5 cm diameter cylinder. By reference to DP-1014 this is 90% of the minimum critical cylinder diameter for an infinitely long, water reflected, optimally moderated cylinder with four wt% enriched heterogeneous UO. This value is further supported by Figure 2(page10)inANS$/ANS 8.1-1983 and Figure 2.15 (page
- 44) in TID-7016, Rev. 2.
14.3.3.5.2 Receipt and Storage A. Unirradiated Assemblies - Unirradiated fuel assemblies may be stored in their shipping containers since their nuclear safety has been proven prior to their licensing. Assemblies that are unirradiated may also be stored in air if the distance between assemblies is no less than 21 by 38 inches. (Refer to SNM-1168, Docket 70-1201, Section 3, page 173, dated 2/27/81). This distance assures criti-cality safety for less than 100 assemblies of either the Mark B and/or Mark C assembly types. This ensures the safety of the maximum of four assemblies stored on site. Unirradiated assemblies may also be stored under water (hot cell pool, mock-up pool, or development test area pool). Assemblies stored in air will be stored either: l i 1. Horizontally - on the floor or on tables constructed l with sufficient integrity and strength to withstand reasonable structural deformity thus assuring the above i mentioned spacing. i l License No SNM-778 Docket No. 70 824 Date April, 1987 Amendment No. O Revision No. 4 Page 14-11 Nq l l Babcock &Wilcox a McDermott company l (O) sufficient integrity and strength to withstand 2. Vertically - in. racks and fixtures constructed with reasonable structural deformity and assuring the above mentioned spacing. Supervisory approval is required to move any other fissile material into the area where the assemblies are stored. No more than four unirradiated assemblies may be stored at once. The limit of four assemblies is an arbitrary limit which the site imposes upon itself and j does not affect nuclear safety. Partially disassembled unirradiated Mark B or Mark C assemblies may also be stored in air. This is safe due to the lower moderation characteristics of air compared to water. Air moderated values of K*ff will be less th.in those shown in Table 14-6. Fuel rods from unirradiated, disassembled Mark B or Mark C assemblies will be stored in air in slabs not to exceed 4 ' inches in height (see Section 14.3.3.5.4, statement 2). B. Irradiated Fuel Assemblies - Assemblies which have been irradiated may also be stored in their shipping containers or in the hot cell pool. Storage in the hot cell pool is limited to four irradiated assemblies. --The-limit of four ( assemblies in the pool is an arbitrary limit which the site l b imposes upon itself and does not affect nuclear safety since each fuel assembly or rod storage position is neutronically isolated from any other fissile material by a minimum of 1 foot of water. Racks and fixtures in the pool are constructed with sufficient integrity and strength to withstand reasonable structural deformity, thereby providing the spacing previously outlined. The racks are also constructed to preclude inadvertently placing other fissile material closer than the 1-foot minimum spacing. Supervisor approval is required for removing or inserting fissile material into or out of any of the racks or fixtures. Storage of Mark B and j Mark C fuel rods and partially dismantled assemblies into storage racks which restrain the size of each position to a square not exceeding the dimensions of a fresh fuel assembly, i.e., 8.6 inches, is safe based upon the analysis demonstrating safety of as assembly during dismantlement. Fuel rods may also be stored in an ever safe cross sectional l License No SNM 778 Docket No.70-824 Date Apr11,1987 L l Amendment No. O Revision No. 4 Page 14-12 0 v I Babcock &'Wilcox l a McDermott company t N (~] area fixture, i.e., a cross sectional area not exceeding (j that of a 22.5 cm diameter cylinder. 14.3.3.5.3 Work Area Of Pool Under Hot Cell No.1 - This area will be used to dismantle irradiated and unirradiated assemblies. Nuclear criticality safety for Mark B and Mark C assemblies under varying stages of dismantlement has been demonstrated via use of NULIF and PDQ-07 physics computer codes. Reactivity was calculated by PDQ (coefficients having been generated by NULIF) for a fully reflected and flooded, unrodded fresh assembly and for the same assembly under five conditions of dismantlement. The cases run with the number of rods ~ removed in each case and the resulting K,ff is given in Table 14-6. TABLE 14-6 REACTIVITY FOR MARK B AND MARK C FUEL ASSEMBLIES UNDER DISMANTLEMENT
- II
- /3 No. of Removed (j Case No. Rods Mark B Mark C 1 0 0.891 0.921 2 4 0.894 0.920 3 12 0.897 0.919 4 24 0.898 0.922 5 36 0.896 0.917 6 8 0.888 0.918 l l i I License No SNM 778 Decket No.70-824 Date April, 1987 Amendment No. O Revision No. 4 Page 14-13 O v Babcock &Wilcox a McDermott company l D] Reactivity was also calculated by KENO-IV using the 123-group / XSDRN cross section set for a fresh assembly fully submerged in water under conditions of 24 rods removed and with all instru-ment and control rod guide tube positions loaded with fuel rods. The cases run with the number of rods removed or added and the resulting K-effectives are given in Table 14-7. TABLE 14-7 K-EFFECTIVE OF INDIVIDUAL MARK B AND MARK C FUEL ASSEMBLIES Change in No. K-effective + 2 o Of Fuel Rods From Normal Mark B Mark C 0 .895 +.010 .900 +.013 -24* .900 7.015 .906 T.014 +17 .890 T.015 +25 .876 +.016 [ \\
- Same configurations as case 4 in Table 14-5.
Cases 2 through 5 represent removal of rods " uniformly" through the assembly while Case 6 represents the removal of eight rods clustered about the center. Rods removed are shown sche-matica11y in Figures 14-1 and 14-2 for Mark 8 and C assemblies, respectively. The calculations reported above demonstrate the nuclear safety of an assembly under various conditions of dis-r assembly and reloading. If any of the fuel rods inserted into the fuel assembly are further encased in metal tubing, the assembly would still be safe due to the tubing displacing l moderator with absorber. A grouping of 75 fuel rods confined within a 8.6-inch square merely describes a dismantled assemb1 and is also safe. Fuel rods inserted into instrument and control rod guide tubes shall be held in place with a flat metal plate which shall be bolted to the top of the assembly. i l License No SNM-778 Docket No.70-824 Date April,1987 Amendment No. O Revision No. 4 Page 14-14 V l Babcock & Wilcox a McDermott company (V 3 The safety of withdrawing an assembly and its associated rod storage position partially into the cell is demonstrated safe by comparison to a series of KENO runs made for pool storage at a reactor site. To demonstrate the safety of flooding a reactor site storage pool filled with fresh Mark B fuel assemblies, an array of fuel assemblies 14 units wide, infinitely long, and reflected on the sides and bottom by concrete was calculated by KENO. Each assembly was spaced 1 foot from the other on the concrete reflector, as appropriate. Four cases at different degrees of pool flooding were evaluated and are described in Table 14-8. TABLE 14-8 REACTIVITY FOR AN INFINITE BY 14-UNIT ARRAY OF FUEL ASSEMBLIES Calculated K Water Height eff Fully Flooded 0.951 + 0.006 3/4 0.946 7 0.007 ((7 1/2 0.928 7 0.007 ) 0 (dry) 0.506}0.004 The confidence levels quoted above are one standard deviation. K for the fully flooded condition is higher than that calcu-1$Nd by PDQ because of simplifications made in running the cases. The series of runs were to demonstrate safety of a partially flooded pool, a much more restrictive condition than partial withdrawal into one cell. The similarity in the Mark B and Mark C nuclear characteristics and the simplifying assump-tions assure these calculations are also valid for the Mark C assembly type. 14.3.3.5.4 Assembly and Machine Shop and Development Test Areas - As-semblies of either Mark 8 or C disassembled in air are far less l reactive than the cases listed in Table 14-6. Either assembly type may be disassembled in air. A safe reactivity level (u0.95) is assured provided the handling in Section 4.2.3.6.4 is followed. The conditions stated in Section 4.2.3.6.4 are based on KENO calculations that show: License No SNM-778 Docket No. 70 824 Date April, 1987 i Amendment No. O Revision No. 4 Page 14-15 O Babcock &Wilcox a McDermott company I O 1. Two assemblies in air 21 inches or more apart are nuclearly 'V safe. 2. Fuel pins at a maximum enrichment when optimum 1y moderated are fully reflected in an infinite slab have a K,ff = 0.95 .if the slab is no more than 4 inches thick. 3. Fuel rods in any configuration or number, up to the number in the assembly, when limited to the confines of the assembly size are no more reactive than the intact assembly (Ref. Table 14-6). 14.3.3.5.5 Hot Cell Operations - Work within the hot cell will, by and large, follow existing centrols. Three units in addition to an assembly and its associated rod storage position are permitted within Cell No. 1. Two of the three units are restricted to rods confined within an ever safe cross sectional area, i.e., a cross sectional area not exceeding that of a 22.5 cm diameter cylinder; in addition these two units must be free draining of any water. The third unit of Cell No.1 under mass control is permitted. All other Hot Cells are limited to one unit eacn. 14.3.3.5.6 Fuel Rod Dismantlement - Fuel rods of either assembly type may be dismantled. Dismantlement can be performed in any area which present licensing conditions permit fuel handling. In m ~ addition, mass control must-be-limited to 350 grams of U-235, ) proper spacing must be maintained, and approved procedures must be followed. 14.3.3.5.7 Shipment and Disposal - The conditions of 4.2.3.6.7 are consis-tent with the above demonstratic and/or current ifmits. 14.3.4 Building C The demonstration for units and the array is similar to that of Building A (14.3.2.1 and 14.3.2.2). The values of all units in Building C are less than or equal to the value of the maximum storage unit defined in Table IV TID-7016, Revision 1 (as amended), or they have been evaluated above in Section 14.3.2.2. The allowable number of units on 36-inch centers is 90 units with r at least 8 inches edge-to-edge between units. The allowable number i of units according to Figure 22 of TID-7016, Revisic.1 1, is about i 190. The number of units has been reduced to 90 to permit the low enriched units. Administrative procedures for posting and con-l l l License No SNM 778 Docket No. 70 824 Date Apr11, 1987 Amendment No. O Revision No. 4 Page 14-16 ! O l v l Babcock &Wilcox a McDermott company (A) 14.3.2.2.4. trolling transfers of SNM to and from units are those described in 14.3.5 Outside Storage 14.3.5.1 General - The underground storage and shipments are nuclearly isolated by distance or matter. 14.3.5.2 Underground Storage - The underground storage tubes are 5 inches in diameter, approximately 20 feet long and 20-inch centers. Maximum units demonstrated safe in Section 14.3.2.2 are stored, one per tube. These are neutronically isolated from each other by 15 inches of concrete. 14.3.6 Dry Waste Nuclear criticality safety of dry waste is ensured by maintaining the concentration of SNM to a value much less than an ever safe concentration. Forty-five grams of SNM in a 55-gallon drum yields a concentration of less than 0.25 g/ liter. These low concentra-tions are guaranteed by the nature of the material being stored which is contaminated laboratory waste. The nature of the waste as borne out by more than 20 years of experience will maintain an approximate uniform dispersion within the container. Dry waste containers are stored in the radioactive waste building af ter gamma scanning to ensure that the maximum SNM is not exceeded.-- There is ( therefore no requirement in the number or arrangement of containers within the radioactive waste building. One dimensional transport u calculation shows that, at a U-235 concentration of 0.25 g/ liter with optimum water moderation, a fully concrete reflected sphere 5 having the same volume as 8 x 10 55-gallon drums has a neutron multiplication of < 0.95. Therefore, the 45 grams of U-235 per drum limit is safe in that the maximum number of drums on site can-5 not credibly exceed 8 x 10 14.4 ANALYTICAL METHODS AND VALIDATION REFERENCES Nuclear criticality safety computer calculations presented in this chapter have used the computer codes NULIF, PDQ-07 and/or KEN 0. The physics codes NULIF and PDQ-07 are not only routinely used in nuclear criticality safety to evaluate highly moderated low-enriched systems but also are the standard codes used by the reactor design group of the Babcock & Wilcox Company (both codes have been certified by the Company's Quality Assurance Program for reactor calculations). The Monte Carlo code KENO is state-of-the-art in industry for nuclear License No SNM 778 Docket No.70-824 Date April,1987 Amendment No. O Revision No. 4 Page 14-17 Babcock &Wilcox a McDermott company n V(3 - cribed in Appendix A,Section III, pages 3 through 12 of SNM License criticality safety evaluations. These three computer codes are des-No.1168 (Docket 70-1201); validation for these codes are given in Appendix A,Section III, of the same document on pages 19 throJgh 21. Future calculations for nuclear criticality safety will make use of these codes and the Nuclear Criticality Safety Codes in SCALE 3 (NITAWL-S, XSDRNPM-S, KENO-IVS and KENO-Va). SCALE 3 is described in NUREG/CR-0200. Before use of the SCALE 3 package, the proper wording of the various codes will be assured and appropriate benchmarking activity will be carried out. 14.5 DATA SOURCES Data and Guidance for Nuclear Criticality Safety is taken from one or more of the sources specified below. 1. Calculations using methods described in Section 14.4. 2. " Nuclear Safety Guide, TID-7016 Revision 2," NUREG/CR-0095 (ORNL/NUREG/CS0-6), (June, 1978.) 3. " Nuclear Safety Guide, TID-7016, Revision 1," (1961). TID-7016, Revision 1 is used only for application of the lettices density method which is Table IV and Figure 22 on page 26. Table IV has been modified according to -information published in the Federal Register, March 5,1963 on page 2130. L 4. American National Standard for Nuclear Criticality Safety in Operations with Fissionable Materials Oxide Reactors, ANSI /ANS-8.1-1983). 5. H. K. Clark, " Critical and Safe Masses and Dimensions of Lattices of U and UO Rods in Water" DP-1014, Savannah River Laboratory 2 (1966). 6. H. C. Paxton, " Criticality Control in Operations with Fissile Material," LA-3366(Rev), Los Alamos Scientific Laboratory, (1972). 7. R. D. Carter, et al, " Criticality Handbook," ARH-600 Revised last l November 6,1973, Atlantic Richfield Hanford Company. l License No SNM-778 Docket No.70-824 Date April, 1987 Amendment No. O Revision No. 4 Page 14-18 Babcock &Wilcox a McDermott company l 14.6 FIXED POISONS The site does not now use Fixed Poisons to maintain nuclear criti-I cality safety. 14.7 STRUCTURAL INTEGRITY Where structural integrity is necessary to provide assurance for nuclear criticality safety in any operation, the design and con-struction of those structures will be evaluated with due regard to load capacity and foreseeable abnormal loads, accidents and deteri-ora tion. This engineering activity is the responsibility of the Manager, Facilities with review and approval by a qualified person. l 14.8 SPECIAL CONTROLS There are no special controls for nuclear criticality safety at the si te. i i f% (. J License No SNM 778 Docket No. 70 824 Date April,1987 Amendment No. O Revision No. 4 Page 14-19 O V Babcock &Wilcox a McDermott company ~' p ) FIGURE 14-1 tv FUEL ROD REMOVAL SCHEMATIC - MARK B Case 2 Case 3 4 rods removed 12 rods removed g x _c g'..x ._.c X x o o x o o O X O O o i f Case 4 Case 5 24 rods removed 36 rods removed G X- -D<---s O X -~ - --X - --s X X X i X O O X O O O i x o x x o x o o X X X X X l\\ e 4 \\ Location of removed fuel rods h case 6 i 8 refs removed l Instrument Tube (1 per assembly) OM l -' -t XX O O caaerat nad caida tuba 1 (16 per assembly) l e o i Rod I 1 License No SNM 778 Docket No.70-824 Date April, 1987 Amendment No. O Revision No. 4 Page 14-20 O v l Babcock &Wilcox a McDermott company (" FIGURE 14-2 FUEL ROD REMOVAL SCHEMATIC - MARK C Instrumentation Tube (1) / / D D / A B / B A O G .G B I / B Control Rod D C C / D Cg de Tube (24) G G G / G e / D C C' Fy C' C D g s_g,_ g _g g D C C' k C' C D e e e e e D C i C D B I B e e e I A B B A D g D I Q MARK C FUEL ASSEMBLY (17 x 17) Fuel Rods Removed for Reactivity Under Dismantlement (Section 4.4.7.3) No. of Rods Pins Shown Removed By Letter 4 A 12 A, B 24 A, B, C, C ' 36 A, B, C, C', D 8 C ', E License No SNM 778 Docket No. 70824 Date April, 1987 Amendment No. O Revision No. 4 Page 14-21 Babcock &Wilcox a McDermott company b (,/> 15.0 PROCESS DESCRIPTION AND SAFETY ANALYSES The operations and projects at the site do not lend themselves to l flow sheets. There is no operation presently in progress that has a regular measured feed material or a regular measured product output. O c i License No SNM 778 Docket No.70-824 Date April,1987 Amendment No. O Revision No. 3 Page 15-1 Babcock & Wilcox a McDermott company