ML20210D345
| ML20210D345 | |
| Person / Time | |
|---|---|
| Site: | 07000824 |
| Issue date: | 12/31/1986 |
| From: | BABCOCK & WILCOX CO. |
| To: | |
| Shared Package | |
| ML20210D305 | List: |
| References | |
| NUDOCS 8702100067 | |
| Download: ML20210D345 (104) | |
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LYNCHBURG RESEARCH CENTER SNM-778 LICENSE RENEWAL APPLICATION REVISION 3, DECEMBER, 1986 REMOVE INSERT PAGE REV DATE PAGE REV DATE 1-1 2
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TABLE OF CONTENTS v
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 P,ERIOD 1-1 1.4 POSSESSIONUMbS.
1-2 1.5 LOCATION OF POSSESSION AND USE 1-3 1.6 DEFINITIONS.
1-3 1.7 AUTHORIZED ACTIVITIES 1-5 1.8 EXEMPTIONS AND SPECIAL AUTHORIZATIONS 1-5 List of Figures 1-1 LRC LOCATION WITHIN VIRGINIA 1-6 l
1-2 LRC, FIVE MILE RADIUS 1-7 i
1-3 LRC, BUILDINGS.
1-8 I
License No SNM-778 Docket No.70-824 DateDecember,1986 Amendment No.
O Revision No.
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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 Material j Physical Form Enrichment Amount
- 1. Uranium enriched Encapsulated or
> 20 %
3.5 Kg con-in U-235 irradiated 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 Unencapsulated 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-
_s in U-235 irradiated tained U-235
- 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 l
and unirradiated
- 8. Source Material Any 6000 Kg
- 9. Fission Products Irradiated Fuel Quantity
& Transuranium contained in Elements 4 irradiated fuel as-semblies.
- 10. Fission Products Irradi.ated fuel 5,000,000 Cf.
& Transuranium Elements License No SNM-778 Docket No.70-824 Date December,1980 Amendment No.
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- 11. Any byproduct Irradiated 50,000 Cf.
material structural materials &
components
- 12. Byproduct Any 3,000 Ci each material with total not to at. nos. 3 exceed thru 83 1,000,000 Ci.
- 13. Transuranium Any 20 milli-elements Curies each l
- 14. Cf-252 Sealed Sources 4 milligrams
- 15. Am-241 Sealed Sources 30 C1
- 16. H-3 Sealed Sources 100 Ci
- 17. H-3 0xide 3 C1
- 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 Lynchburg Research Center.
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 Lynchburg Research Center for performing instrument calibration, electronic noise analysis, shielding studies, or similar operations.
1.6 DEFINITIONS 1.6.1 LRC means Lynchburg Research Center.
1.6.2 SRC means Safety Review Committee.
License No SNM 778 Docket No.70-824 Date December,1986 Amendment No.
O Revision No.
3 Page 1-3 Babcock &Wilcox a McDermott company
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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 experimentation; or (2) the extension of investigative fincjings and theories of a scientific or technical nature into practical application for experimental and demon-stratton 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, Lynchburg Technical Operations means the Manager of Technical Operations for the Lynchburg Research Center.
1.6.8 Authorized User means a person who may work with licensed material unsupervised and may supervise others, not so designated, in the Cl handling of licensed material.
l 1.6.9 Calibration means a comparison of a measurement standard of known accuracy that is traceable to the NBS with another standard or 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, or natural phenomena to set up an instrument.
Standardization 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 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 unib by at least 8-inches edge-to-edge and 24-inches center-to-center.
License No SNM-778 Docket No.70-824 Date December,198h Amendment No.
O Re ision No.
3 Page 1-4 Babcock &Wilcox a McDermott company
pV 1.6.13 Standing RWP's are Radiation Work Permits issued for a term of more than 30-days, authorizing entry into High Radiation Areas and Air-borne 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 B&W divisions including laboratory analysis, preparation of and testing of materials and equipment; preparation 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 LRC may transport and possess licensed material in private carriage between NRC licensed 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 bioassay 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 an employee is absent from the LRC during a period when the bioassay counting service is on site, a special counting shall not be required for those employees for routine exposure control monitoring. The maximum amount of time between in vivo counts shall not exceed 12-months.
License No SNM 778 Docket No. 70 824 Date December,1980 Ammdment No.
O Revision No.
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TABLE OF CONTENTS, Section
- Page 2.0 GENERAL ORGANIZATIONAL AND ADMINISTRATIVE REQUIREMENTS 2-1 2.1 POLICY 2-1 2.2 ORGANIZATION RESPONSIBILITIES AND AUTHORITIES 2-1 2.3 SAFETY REVIEW COMM,ITTEE 2-1 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.-
2.8 INTERNAL AUDITS AND INSPECTIONS 2-8 2.8.1 Nuclear Criticality Safety i2-9 2.8.2 Health Physics 2-9 _
g 2.8.3 General Safety and Compliance 2-9 2.9 INVESTIGATIONS AND REPORTING 0F 0FF-NORMAL OCCURRENCES 2-10 2.9.1 License Administrator 2 _0 2.9.2 Supervisor, Health and Safety 2-10 2.9.3 Facility Supervisor 2-11 2.10 RECORDS 2-11 2.10.1 Health and Safety Group 2-11 e 2.10.2 Nuclear Safety Officer 2-12 2.10.3 License Administrator 2-12 2.10.4 Emergency Records 2-12 License No SNM 778 Docket No. 70424 DateDecember, 1986 O
Amendment No.
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List of Figures Figure Page 2-1 LRC LINE ORGANIZATION 2-13 2-2 LRC SAFETY ORGANIZATION 2-14 4
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License No SNM-778 Docket No.70-824 DateDecember, 1986 2-11
, Amendment No.
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approved by the Manager, Lynchburg Technical Operations. He shall report to the Manager, Safety and Licensing.
'?.2.11 Accountability Specialist - The Accountability Specialist shall be responsible for the maintenance and retention of SNM accountability records. The Accountability Specialist shall report to the Manager, Safety and Licensing.
2.3 SAFETY REVIEW COMMITT'EE i
2.3.1 Function
,2.3.1.1 The SRC shall review and approve all new Area Operating Pro-cedures, and shall concur with 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 major changes to existing projects that utilize licensed materials.
2.3.1.3 The SRC shall review the annual report prepared by the Supervisor, Health and Safety.
2.3.1.4 The SRC shall provide the LRC with general consulting services in the field of radiation protection and the safe handling of s
licensed material.
2.3.1.5 The SRC shall review all SAS audit 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 documetned in the minutes.
2.3.1.6 The SRC Coordinator shall be responsible for resolving comments and recommendations 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 specified in Section 2.3.1.
2.3.3 Safety Audit Subcommittee License No SNM-778 Docket No.70-824 Date December,1980 Amendment No.
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2.3.3.1 The SAS shall perform audits of the LRC for the Safety Review Commi ttee.
e 2.3.3.2 The SAS shall audit facilities, procedures, records, and operations at the LRC for compliance with 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 made 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, Lynchburg Technical Operations.
2.3.4 Reporting 2.3.4.1 The SRC shall report to the Manager, Lynchburg Technical Operations.
2.3.4.2 The SAS shall report to the Chairman, SRC.
j 2.3.5 Recordkeeping O
2.3.5.1 Minutes of the SRC proceedings shall be prepared by the Chairman, l
V SRC.
I 2.3.5.2 SRC Minutes shall be forwarded to the Manager, Lynchburg Technical Operations by the Chairman, SRC.
2.3.5.3 The permanent records of the SRC shall be kept by the SRC Coordinator.
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 Chairman, SRC by the Chairman, SAS.
2.3.5.6 SAS audit reports shall be forwarded to the Manager, Lynchburg l
Technical Operations by the Chairman, SRC with comments, as he deems appropriate.
2.4 APPROVAL AUTHORITY FOR PERSONNEL SELECTION License No SNM-778 Docket No.70-824 Date December,1986 O
Amendment No.
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2.4.1 The Manager, Lynchburg Technical Operations 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 R&D Division shall appoint the Manager, Lynchburg Technical Operations.
2.5 PERSONNEL EDUCATION AND EXPERIENCE REQUIREMENTS 2.5.1 Manager, Lynchburg Technical Operations - The Manager, Lynchburg Technical Operations shall be appointed in accordance with Company policy.
2.5.2 Laboratory Managers - The Laboratory Managers are appointed by the Manager, Lynchburg Technical Operations in accordance with Company policy.
2.5.3 Section Managers - The Section Managers shall have a BS degree and three years post graduate work or equivalent experience in the pertinent technical field. Managers of sections handling licensed materials shall have demonstrated knowledge of the application of radiation and nuclear criticality safety requirements relative to their projects.
l l D) 2.5.4 Facility Supervisor - The Facility Supervisor shall have a degree in his related work and three years experience in the use and handling of licensed 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 LRC.
2.5.5 Manager, Safety and Licensing - The Manager, Safety and Licensing shall have a BS degree in a technical field and five years I
experience in the nuclear field.
2.5.6 Supervisor, Health and Safety - The Supervisor, Health and Safety shall have a BS degree in a technical field and professional 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.
License No SNM-778 Docket No.70-824 Date December,1986 Amendment No.
O Revision No.
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2.5.7 Health Physics Engineer - A Health Physics Engineer shall have a BS degree which shall include at least 20 quarter hours health physics related course work or the equivalent in work experience.
2.5.8 Industrial Safety Officer - The Industrial Safety Officer shall have at least one year's experience in radiation and industrial safety. He shall be familiar with the codes and requirements of the Occupational Health and Safety Act of 1970 and tne National Fire Protection Association.
V 2.5.9 Nuclear Safety Officer - The Nuclear Safety Officer shall have a BS degree in science or engineering. He must have either two years experience with nuclear criticality safety calculations similar to those associated with LRC activities or he must have one year's experience with nuclear criticality safety calculations similar to those associated with LRC activities if he has at least an additional two years' experience in nuclear reactor physics calculations.
2.5.10 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
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2.5.11 License Administrator - The License Administrator shall have a BS degree in science or engineering and three years experience in nuclear technology or an AS degree in science or nuclear technology with five years experience in nuclear technology.
2.5.12 Safety Review Committee - The SRC membership, as a body, 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 LRC operation.
2.6 TRAINING 2.6.1 Program I - Each new employee shall receive training within thirty days of reporting to work. This training, denoted as Program I, provides an introduction to radioactivity and a thorough coverane of safety rules and procedures including emergency procedures.
License No SNM-778 Docket No.70-824 Date December,1980 p
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2.6.2 Program II - New laboratory employees who will be working with licensed material shall be required to complete Program II training. Completion of this program requires passing a written examination. Parts of Program II may be waived by the Supervisor, Health and Safety for technical and scientific personnel already knowledgeable and experienced in working in radiation areas and with licensed material. However, such personnel must pass the written examination required for Program II. Persons who complete this course may be designated as an Authorized User.
2.6.3 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.
2.6.4 Respiratory Protection Training - Training in respiratory pro-tection techniques and equipment shall be required of all employees before the use of such equipment will be permitted. Satisfactory completion of this training shall be determined by passing a written examination.
2.6.5 Respiratory Protection Retraining - Retraining in respiratory pro-tection shall be performed at two year intervals.
Satisfactory j
completion of this retraining shall be determined by passing a l
written examination.
v 2.6.6 The training specified in Section 2.6 shall be administered by the Supervisor, Health and Safety, or his designated and qualified al terna te.
2.6.7 Nuclear Criticality Safety Training - Nuclear Criticality Safety training provided as a part of the programs specified in Sections 2.6.2 and 2.6.3 shall be performed by the Nuclear Safety Officer or his designated alternate. The designated alternate must meet the same minimum qualifications as those specified for the Nuclear Safety Officer (2.5.9).
l 2.7 OPERATING PROCEDURES l
2.7.1 Area Operating Procedures 2.7.1.1 All operations with licensed material shall be conducted in accordance with Area Operating Procedures or Radiation Work Permits (see 3.1.1).
l License No SNM-778 Docket No.70-824 Date December,1986 l p Amendment No.
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2.7.1.2 Area Operating Procedures (A0P) - 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 i
that apply to the operation. Each A0P shall be approved by the 1
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Nuclear Safety Officer or his designated alternate, the Super-l visor, Health and Safety or his designated alternate, the Facility Supervisor or his designated alternate, and the Safety Review l
Commi ttee.
2.7.1.3 A0P's may be ievised with the approval of the Nuclear 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 operations personnel.
2.7.1.5 Distribution of new and revised procedures shall be made in accordance with a document control system which assures that the procedure manuals contain only the most current revision of the procedures.
I 2.7.1.6 AOP 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 developed for Health and Safety or Nuclear Criticality Safety shall be reviewed and approved by a Health l
Physics Engineer or the Nuclear Safety Officer, respectively, or l
their designated alternate. The designated alternate for a Health Physics Engineer must meet the minimum qualifications specified in Sections 2.5.7.
The designated alternate for the Nuclear Safety Officer must meet the same minimum qualifications specified in Section 2.5.9.
Approval signatures shall appear on the procedure, 2.8 INTERNAL AUDITS AND INSPECTIONS License No SNM-778 Docket No.70-824 Date December,1986 Amendment No.
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1 2.8.1 Nuclear Criticality Safety 4
2.8.1.1 The Nuclear Safety Officer or his designated alternate shall conduct internal audits of the LRC 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 report of his findings shall be made to the Manager, Lynchburg Technical Operations within two weeks of completing the audit.
The audit reports shall be forwarded to the Facility Supervisor 1
and the Licdnse Administrator. The License Administrator shall be responsible for assuring that the appropriate corrective actions are taken to address the audit findings.
4 2.8.1.2 The Facility Supervisor shall perform an inspection weekly for compliance with the nuclear criticality safety aspects of the operations. Findings resulting from these inspections shall be reported to the Nuclear Safety Officer.
2.8.2 Health Physics 2.8.2.1 The Supervisor, Health and Safety or his designated alternate shall conduct internal audits of the LRC for the purpose of n
evaluating the health physics aspects of operations. This audit shall be conducted in accordance with written audit guidance. This v
audit shall be conducted once each month. A report of his I
findings shall be made to the Manager, Lynchburg Technical i
Operations within two weeks of completing the audit. The audit l
reports shall be forwarded to the Manager, Lynchburg Technical Operations and the License Administrator. The License Adminis-i l
trator shall be responsible for assuring the appropriate corrective actions are taken to address the audit findings, i
2.8.3 General Safety and Compliance 2.8.3.1 The SAS performs audits of general safety and compliance at the LRC.
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 of LRC operations shall be audited for compliance with written requirements and the exercise of acceptable safety practices. Audits shall be made in accordance with written License No SNM-778 Docket No.70-824 Date December, 198n Amendment No.
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guidance to assure all aspects of Section 2.3.3.2 tre audited.
The Chairman, SAS shall file a report of the audit findings with the Chairman, SRC, with a copy to the License Admin,strator and the Facility Supervisor and members of the SRC. The Chairman, SRC shall forward the report to the Manager, Lynchburg Technical Operations with 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:
4 2.9.1.1 Excessive levels of radiation from or contamination on packages upon receipt.
2.9.1.2 Thefts, attempted thefts, or losses of licensed material, other than normal operating losses.
l 2.9.1.3 Incidents as specified in 10 CFR 20.403 l
2.9.1.4 Overexposure of individuals and excessive levels and concentra-tions of radioactivity.
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.
l l
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:
1
[
2.9.2.1 Higher than expected personnel exposures, l
l License No SNM 778 Docket No.70-824 Date December,1986 Amendment No.
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bi 2.9.2.2 Higher than expected concentration of airborne activity in the facility.
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 Superviso'r i
The Facility Supervisor shall perform investigations of the following:
2.9.3.1 Any violation of nuclear criticality safety criteria.
i 2.9.3.2 Any violation of Area Operating Procedures or RWP's.
2.10 RECORDS i
The following positions or organizations shall be responsible for maintaining the indicated records, for the period specified.
p) y 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 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 l
Radiation Work Permits (completed) 5 years l
Radiation detection instrument calibration 2 years l
Leak tests of sealed sources 2 years l
Employee training
(*)
l Employee retraining
(*)
Airborne radioactivity sampling data
(*)
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s NRC-4 forms
(*)
NRC-5 forms
(*)
2.10.2 Nuclear Safety Officer Nuclear safety evaluations 6 months after and calculations termination of the approved j
process.
Nuclear Safety Officer 2 years Audit Reports.
2.10.3 License Administrator Safety Review Committee Minutes
(*)
Safety Audit Subcommittee 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 n
retained in accordance with Lynchburg Research Center Radiological n,y Contingency Plan, Section 8.0.
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FIGURE 2-1 LRC LINE ORGANIZATION RESEARCH AND DEVELOPMENT
.4 DIVISION VICE PRESIDENT p LYNCHSURG RESEARCH CENTER SPECIAL PROJECTS /
LYNCH 8URG TECHNICAL OPERATIONS DECOMMISSIONING MANAGER
- MANAGER DECOMMISSIONING
~~~~~~~~~
PURCHASING PURCHASING MANAGER R&DO LRC MANAGER MANAGER SAFETY f
AND l
ACCOUNTING UCENSING CONTROLLER AND R&DD ADMINISTRATIVE MANAGER SERVICES LRC
}
MANAGER SYSTEMS FACILITIES DEVELOPMENT AN FACILITIES LABORATORY ASSURANCE LRC R&OD MANAGER MANAGER MANAGER QUALITY QUALITY MATERIALS i
ASSURANCE ASSURANCE ENGINEERING l
R&DD LRC LABORATORY MANAGER ADMINISTRATOR MANAGER SCIENTIST
---INDICATES FUNCTIONAL REPORTING License No SNM-778 Docket No. 70 824 Date December,1986 Amendment No.
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4 FIGURE 2-2 LRC SAFETY ORGANIZATION s
I LYMCHOURG RESEARCH CENTER LYNCHOURG TECHNICAL OPERATIONS MANAGER i
r SAFETY REVIEW
. COMMITTEE I
i i
f FACILITY SUPERVISOR SAFETY AND UCENSING l
NUCLEAR SAFETY l
OFFICER MANAGER I
I I
i i
I LICENSING HEALTH AND SAFETY ACCOUNTABILITY l
ADMINISTRATOR SUPERVISOR SPECIAUST I
l HEALTH PHYSICS INOUSTRIAL STAFF SAFETY
---INDICATES FUNCTIONAL REPORTING l'
License No SNM 778 Docket No.70-824 Date December, 1986 Amendment No. O Revision No.
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1 1
k TABLE OF CONTENTS Section Page 3.0 RADIATION PROTECTION.
3-1 3.1 SPECIAL ADMINISTRATIVE REQUIREMENTS.
3-1 3.1.1 Radiation Work Permits (RWP) 3-1 3.1.2 ALARA Policy 3-1 3.2 TECHNICAL REQUIREMENTS 3-2 2
l 3.2.1 Access Control.
3-2 3.2.2 Yentilation Requirements 3-2 3.2.3 Instrumentation 3-4
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3.2.4 Internal and External Exposure 3-6 O
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License No SNM 778 Docket No.70-824 Date0ecember, 1986 i
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G) 3.0 RADIATION PROTECTION 3.1 SPECIAL ADMINISTRATIVE REQUIREMENTS 3.1.1 Radiation Work Per. nits (RWP) 3.1.1.1 RWP's shall be issued whenever the activity is not covered by an Area Operating Procedure and personnel are likely to be exposed to levels of radiation or concentrations of radioactive material in excess of those specified in 10 CFR 20.101 & 20.103.
l 3.1.1.2 RWP's shall be approved by the Work Area Supervisor, Employee's Supervisor, Health Physics Supervisor, and the Facility Supervisor.
In the absence of any of the above persons, a designated and quali-fled alternate may approve RWP's.
3.1.1.3 The RWP shall specify the radiological protection requirements for the operation and specify levels of personnel 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, be approved by the Manager, Lynchburg Technical Operations.
3.1.1.4 RWP's shall be approved at a meeting of all the signators of the form.
3.1.1.5 The RWP form shall provide space for entering the estimated exposures to the whole body, extremities, and for the job. These are used to identify the areas of exposure concern and do not constitute an exposure goal or limit.
3.1.1.6 The RWP form shall provide space for the workers' supervisor to w
sign or initial, attesting that the workers have been instructed in the requirements of the RWP.
3.1.1.7 The term of an RWP shall not exceed 30-days except for Standing RWP's.
3.1.2 ALARA Policy The management of the LRC is committed to a policy of maintaining exposures as low as is reasonably achievable.
l 3.1.2.1 Employees shall be introduced to this policy during their initial training and shall be reinforced during the annual retraining of i
Authorized Users, i
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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.
3.2.2.2 Potentially contaminated exhaust air from hood, hot cells, and 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.
l The stack samplinh and monitoring system shall operate continu-3.2.2.4 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 V
Beta Particulate 2 mci /yr 200 uCi/ week Alpha Particulate (long lived) 20 uCi/yr 1 uCi/2 weeks Kr-85 2500 Ci/yr 70 Ci/ week H-3 130 Cf/yr 3 Ci/ week I-131 6 mci /yr or 300 uCi/ week 200 uCf/ 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 License No SNM 778 Docket No.70-824 Date December,1986 Amendment No.
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without continuous sampling, if approved by the Safety Review Committee.
3.2.2.8 Areas equipped with an air monitor may be exhausted to the roof 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 00P 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 O'
(3.2.2.12) shall be 99.95% of all particles having a light-scattering mean diameter of approximately 0.7 micrometers.
3.2.3 Instrumentation l
3.2.3.1 Portable Survey Instruments 3.2.3.1.1 Portable instruments - The LRC maintains a relatively large and diverse inventory of portable survey instruments. These instru-ments vary in range and sensitivity. The below listing is a representative sampling of the instruments on hand:
l Instrument Sensitivity Characteristics Type Radiation l
Ionization 0 - 20K R/hr 6.5 Kev - 1.2Mov Beta & Gamma Chamber Geiger 0 - IK R/hr 23 Kev - 1.2Mev Beta & Gamma Counter i
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Proportional 25 - 500K cpm Alpha & Beta Counter (gas flow)
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 Key - 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 instrunynt 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.
Tpy utilize a fixed filter with a nominal air flow of 2.5 to 3 ft / min.
The alarm setting is set at less k-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 or U is being processed. They use a ZnS(Ag) scintillation or with a fixed filter. The monitor air flow is nominally detec}/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-License No SNM 778 Docket No.70-824 Date December,1986 Amendment No.
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tion of airborne radioactive material potentially exceeds 25% of the applicable MPC.
3.2.3.4 Criticality Monitors 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 NaI (T1) detector operated in the constant current mode.
Response is logarithmic and 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 g
g 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 3.2.4.1 Ventila tion l
3.2.4.1.1 The minimum air velocity across the opening of fume hoods that are used to handle ifcensed material shall be at least 100 fpm.
Hood face velocities shall be measured monthly. Those hoods that do not meet tha minimum requirement shall be placed out of service.
3.2.4.1.2 The maximum differer'ial 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.
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V 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.
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 judgmentjof 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 weekly in areas where 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 daily in areas where sample evaluations indicate concentrations of airborne radio-active materials less than or equal to 10% of the applicable MPC.
l 3.2.4.2.4 Air sampling filters shall be evaluated within two weeks of the i
filter removal.
3.2.4.2.5 An investigation by a Health Physics Engineer shall be performed into the cause of unexpected air sampling results that indicate airborne activity at levels between 10% and 25% of the appli-cable MPC. The Health Physics Engineer shall assign responsi-bility for completion of any actions that may be indicated by j
the investigation.
3.2.4.2.6 An investigation by the Supervisor, Health and Safety shall be l
performed into the cause of unexpected air sampling results that l
indicate airborne activity at levels exceeding 25% of the appli-cable MPC.
The Supervisor, Health and Safety shall be responsi-ble for specifying corrective actions and assuring that the specified actions are taken.
3.2.4.3 Bioassay License No SNM 778 Docket No.70-824 Date December,1986 Amendment No.
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L.J 3.2.4.3.1 Uranium Bioassay Program
- 1. The uranium bioassay program sampling frequency shall comply with Regulatory Guide 8.11, June,1974, except as specified in section 1.8 of tris application.
- 2. The following are the action criteria for the routine uranium bioassay program:
Action Analysis 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 resul ts.
- 2. If #1 is positive, investi-gate and correct as needed.
- 3. Make sure individual is in-vivo counted during the next pi t
time that the body 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, investigate the cause and
{
correct as needed.
- 4. If exposure is confirmed by
- 2, investigate to determine how exposure was incurred and correct it.
If the ex-posure exceeds 50% of the License No SNM 778 Docket No.70-824 Date December,198 i O
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maximum permissible annual dose, the worker shall be restricted from further exposure until the Super-visor, Health and Safety authorizes the lifting 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.
- 3. If #2 is positive, investi-gate the cause and correct as needed.
- 4. If exposure is confirmed by
- 1, investigate to determine how exposure was incurred and correct it.
If the ex-posure exceeds 120 ug, the worker shall be restricted from further exposure until the Supervisor, Health and Safety authorizes the lifting of this restriction.
3.2.4.3.2 Plutonium Bioassay Program
- 1. All personnel who routinely work in Plutonium handling areas shall be subject to the Plutonium bioassay program. The License No SNM 778 Docket No.70-824 Date December, 198 i Amendment No.
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minimum frequency for urine sampling shall be six months. The minimum frequency for in vivo counting shall be annual. Ad-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 dpm/L
- 1. Resample the individual within 5 working days.
- 2. Determine if area surveys support the analysis results.
- 3. If #2 is positive, investi-gate the cause and correct.
V
- 4. If the exposure 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.
- c. Urinalysis
> 4 dpm/L
- 1. Restrict the individual from further Pu work.
- 2. Resample the individual with 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
License No SNM 778 Docket No.70-824 Date December,1986 Amendment No.
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- 3. Initiate an investigation.
- 4. The Supervisor, Health and Safety only, may lif t the work restriction.
- d. In vivo
< 16E-8 Ci None I
- e. In vivo.
> 16E-8 Ci
- 1. Restrict the worker from l
further exposure.
- 2. Resample the individual within 10 working days.
- 3. Determine if area surveys support the analysis results.
- 4. If #3 is positive, investi-gate the cause and correct as needed.
- 5. If exposure is confirmed by O
- 2, the Supervisor, Health O
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 lifting of this restriction.
- 6. The restriction in #1 may be lifted by the Supervisor, Health and Safety if the results of the analysis per-formed under #2 fails to confirm the analysis.
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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 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 If fe of the contami-nant and plot decay curves.
Follow-up program will continue until the contamination present is
<51 MP08 or the effective half life has been determined.
O-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 License No SNM 778 Docket No.70-824 Date December,1986 Amendment No. O Revision No.
3 Page 3-12 Babcock &Wilcox a McDermott company
O treatment are required for any case where a result indicating a greater than 105/MP08 deposition of a radionuclide is veri fied.
3.2.4.4 Protective Clothing 3.2.4.4.1 The use of protective clothing shall be specified in Area Operating Procedures and Radiation Work Permits.
3.2.4.4.2 Protective clothing may also be specified by the Health and 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 through 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 fuel handling areas Building B Counting Lab.
Monthly 200 I
Hot Cell Oper. Area Monthly 200 Scanning Electron Monthly 200 Microscopy Lab.
Exit portals from Biweekly 200 controlled areas License No SNM 778 Docket No. 70 824 Date December,1986 Amendment No.
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<-----------------------BETA + GAMMA---------------------->
l Building B Counting Lab.
Monthly 2000 Scanning Electron Monthly 2000 Microscopy Lab.
Hot Cell Operations Area Bimonthly 2000 CaskHandlingArea Bimonthly 22000 Radiochemistry Lab.
Bimonthly 22000 Exit portals from Btweekly 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 exccaded, 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 Decontamination 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 discovery.
If discovery is made just prior I
to the beginning of a holiday or weekend, the contamination shall be marked and labeled, and decontamination shall commence during the first regular workday after discovery.
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.
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e 3.2.4.7.4 Fixed contamination that, in the opinion of the Supervisor
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Health and Safety, may substantially contribute to workers exposure shall be posted and removed as soon as practicable.
y 3.2.4.8 Emergency Evacuation
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3.2.4.8.1 Refer to Radiological Contingency Plan, required by Order dated February 11, 1981, as amended.
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3.2.4.9 Personnel Monitoring 3.2.4.9.1 All employees of the LRC shall be issued a TLD monitor. This e
monitor has a range of from 10 mrem to 10,000 Rem.
This a
dosimetry shall be attached to the employee's identification
~"^
badge.
3.2.4.9.2 Employees whose annual exposure, as projected by the Supervisor, Health and Safety, will exceed 100 mil 11 Rem (Radiation Workers) shall be issued a film badge and two pocket dosimeters or one self-reading dosimeter and one TLD. This dosimetry shall be worn when the individual is in the restricted area.
Pocket and self-reading dosimeters are read and recorded daily. Film badges and TLD's shall be sent for processing monthly.
Ex-posures indicated by the film badge and TLD shall be ovaluated O
monthly. The Supervisor, Health and Safety shall notify the radiation worker's supervisor when the total weekly exposure to the worker exceeds 100 mrem, as indicated by the pocket dosi-meter, or self-reading dosimeter.
l 3.2.4.9.3 Visitors shall wear one TLD.
3.2.4.9.4 Visitors in large tour groups shall be issued one TLD dosimeter.
for each 10 persons. At least one monitored visitor shall be in each subgroup.
3.2.4.9.5 B & W employees from facilities other than the LRC will not be issued LRC dosimetry if they wear the dosimetry from their own facili ty.
If they do not have their own dosimetry, they shall l
be monitored the same as other visitors.
3.2.4.9.6 Visitors who perform special work at the LRC may be badged the same as Radiation Workers after appropriate training.
3.2.4.9.7 Delivery truck drivers shall be issued one self-reading j
dosimeter and a film badge.
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. TABLE OF CONTENTS
, e' Section Page g
A, 4.0 NUCLEAR CRITICALITY SAFETY.
4-1 4.1
/SPECIAL ADMINISTRATIVE. REQUIREMENTS.
4-1
[
4.2 ' ' TECHNICAL REQUIREMENTS 4-2 4.2.2 8uilding A 4-2 4.2.3 Building a~.
4-3 4.2.4 [Efilding C 4-10 4.2.5 Outside Storage 4-10 4.2.6 Dry Waste 4-10 "O-ig
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4 SRC. Area operating procedures shall include all the controls and limits significant to the nuclear criticality safety of the oper-ation. In addition, the Nuclear Safety Officer shall perform a quarterly audit for compliance with nuclear criticality safety requirements and veriffes that process conditions have not been altered that may affect nuclear criticality safety. The results of the audit shall be documented and submitted to the Manager, Lynch-burg Technical Operations.
4.2 TECHNICALREQUIREMENT$
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/ft 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 that O
the unit or units cannot be representable as a slab which U
interacts with other SNM primarily through its major face.
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 Pu (wt%)
Limit (total grams fissile) 0 350 License No SNM-778 Docket No.70-824 Date December,1986 Amendment No.
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p 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.
j 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
/~'N appropriate decay period. The results shall be recorded and V
maintained on file.
5.1.4.3 Sampling - The stack shall be sampled isokinetically on a continu-cus 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 {eet 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-rithmic 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 l
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are similar 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 LRC.
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 3
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 LRC 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.
l 5.2.2 The following types of samples shall be taken at the below indicated frequencies:
1.
Site boundary air sample - monthly.
I 2.
Grab sample of the James River above and below the point of discharge - monthly.
3.
Continuous sampling of rain water.
I 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.
7.
Vegetation sample - semiannually.
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8.
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 i
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 LRC or,a qualified outside concern.
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TABLE OF CONTENTS Section Page l
7.0 DECOMMISSIONING PLAN 7-1 7.1 PLANNING CONSIDERATIONS 7-1 7.2 COST AND FINANCIAL ARRANGEMENTS 7-2 O
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S 7.0 DECOMMISSIONING PLAN V
The Lynchburg Research Center is committed to decommissioning the facilities which have 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 these two facilities will be decontaminated and ready for release for unrestricted use by March,1987.
l 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.
i p 7.1.3 No covering will be applied to remaining surfaces until it has been V
determined that contamination levels are below those of 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, 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, l
ductwork, and other conduits will be determined by taking l
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 l
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TABLE OF CONTENTS (Continued)
O-Section Page 11.3.7 Nuclear Safety Officer 11-14 11.4 OPERATING PROCEDURES.
11-16 11.4.1 Area Operating Procedures (A0P) 11-16 11.4.2 Availability 11-17 l
11.5 TRAINING.
11-17 11.5.1 General Radiation Protection Training 11-17 11.5.2 Program I 11-18 11.5.3 Program II 11-18 11.5.4 Respiratory Protection Training 11-20 11.6 FACILITY CHANGE 11-21 i
List of Figures Figure Page l
11-1 LRC LINE ORGANIZATION 11-23 11-2 LRC SAFETY ORGANIZATION.
11-24 11-3 FACILITY WORK ORDER FORM 11-25 i
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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 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 f
(
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 l
pigment manufacture, effluent treatment, and byproduct recovery.
Professional Affiliations:
American Institute of Chemical Engineers (Member)
American Nuclear Society (Member) 11.3.2 Supervisor, Health and Safety - Gary S. Hoovler l
Education:
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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 Univeristy
- Respiratory Protection for Nuclear Industry Experience:
(1986-Present) Babcock & Wilcox, Supervisor, Health and Safety, Lynchburg Research Center, Lynchburg, Virginia Mr. Hoovler is rebponsible for the Health Physics and Industrial Safety functions at the Lynchburg Research Center; reporting to the Manager, Safety and Licensing.
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 wste 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 s
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 activities for the decontamination and survey of the 20,000 square 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.
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g (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, Lynchburg Research Center, Lynchburg, Virginia Mr. Hoovler worked in the Reactor Experiments Group of the Nuclear Physics Section. He joined B&W in 1976 as a research engineer.
In 1978, after receiving his Senior Reactor Operator's License, he was 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.
l He worked on the neutron spectrum unfolding code SAND II, eddy current analysis, and on experiments to demonstrate the use of l p beryllium-gold as a passive techniqud for measuring fission product l
d distribution in spent fuel.
i l
11.3.3 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, l
Lynchburg Research Center, Lynchburg, Virginia Mr. Schilthelm is responsible for administering the Health Physics 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-l License No SNM-778 Docket No.70-824 Date December, 198 i l
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O 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 alternate for the position of Supervisor, Health and Safety.
(1984-1985) Research Specialist, Synchrotron Radiation Center, University of Wisconsin, Madison, Wisconsin Mr. Schilthelm wasfresponsible 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."
Professional Affiliation:
American Nuclear Society (Member)
Health Physics Society (Member) 11.3.4 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 License No SNM-778 Docket No.70-824 Date December,1986 Amendment No.
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p 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 Mr. Spradlin served as the Accountability Technician.
In this 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 l
l 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 surveillanace program for airborne radioactive material.
He performed maintenance, testing, and calibration of l
alpha particle survey instrumentation and counting equipment. He l
implemented-the respiratory protection program in that laboratory.
l (1967-1969) Babcock & Wilcox, Plan Engineering Technician, Lynchburg Research Center, Lynchburg, Virginia As a plant engineering technician, Mr. Spradlin performed l
installation, modification, and repair of facilities, equipment, l
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O and experimental apparatus at the LRC.
He performed these duties b
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, M111 wright, Maintenance Foreman, Maintenance Superintendent, Safety Inspector and Accident Investigator.
Professional, Affiiiations:
Concord Rescue Squad - Founding President American Heart Association - Cardiac Care Committee 11.3.5 Accountability Specialist - Kenneth D. Long Education:
Graduate - White Sulphur Springs High School,1958 Certificate - Bookkeeping, Central Virginia Community College,1983 Experience:
(1974-Present)
Babcock & Wilcox, Accountability Specialist gg 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/D0E 741 Transaction Forms. He is responsible for the timely completion of inventories of licensed ma terial.
He initiates the paper work required for all shipments l
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 D00 documents at the LRC. He is also an authorized classifier and an authorized courier of classified ma terial.
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p (1970-1974) Babcock & Wilcox, Shipping & Receiving Clerk Lynchburg Research Center, Lynchburg, Virginia 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 uranium and plutonium materials and was responsible for the accountability of SNM materials into and out of his area.
Professional Affiliations:
Institute of Nuclear Materials Management (Senior Member)
Nuclear Materials Control Committee, B&W (Secretary)
American Nuclear Society, Virginia Chapter (Member) 11.3.6 License Administrator - Arne F. Olsen Facility Supervisor
- Arne F. Olsen 7
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Education:
1 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 l
and the NRC and other federal, state, and local agencies regarding l
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 License No SNM-778 Docket No.70-824 Date December, 198 i Amendment No.
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\\s 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.
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 programs.
(1964-1968)
Babcock & Wilcox, Technician and Shif t 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 o
charged with the proper operation and maintenance of the BAWTR. He C
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) 11.3.7 Nuclear Safety Officer - Francis M. Alcorn Education:
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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 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 p
Transferred to the LRC as Nuclear Criticality Safety Specialist Q
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 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 B&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 l
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 License No SNM-778 Docket No.70-824 Date December, 198 i 1
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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, nuclear fuel costs and critical experiment analysis. He has also 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 physics and shielding calculations.
(1960-1961) General Nuclear Engineering Corporation, Staff Physicist 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 GNEC for the AEC.
Professional Affiliations:
i Sigma Pi Sigma (Member)
Tau Beta Pi (Member)
(sk ')
American Nuclear Society - Past Chairman of ANS Nuclear i
i Criticality Safety Division
- Member Standards Subcommittee ANS-8.
l 11.4 OPERATING PROCEDURES I
11.4.1 Area Operating Procedures (A0P) - All operations with licensed material shall be conducted in accordance with Area Operating Procedures 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 Safety Officer who reviews it I
to assure that any nuclear criticality safety issues are properly addressed.
If the Nuclear Safety Officer (NS0) has additions or corrections, he notes them on the procedure and forwards it to the Supervisor, Health and Safety (S.H&S).
If the NS0 approves it, he signs the procedure in the space provided and forwards it to the License No SNM-778 Docket No.70-824 Date December, 198 i Amendment No.
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gQ S.H&S. The S.H&S 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 S.H&S approves the procedure, he signs the procedure in the space provided and forwards it to the Facility Supervisor. The Facility Supervisor reviews it for general safety and determines its impact on other work and facilities. The Facility Supervisor is responsible for resolving all additions or changes recommended by the previous reviewers. When the procedure is approved by the three reviewers, the Facility Supervisor forwards it to the Safety Review Committee., The Safety Review Committee (SRC) may approve the procedure as Eritten, approve the procedure conditionally with specific changes to be made prior to issuance or the SRC can dis-approve it. The SRC coordinator signs for the SRC when approval is voted. The procedure 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 approval signatures of the NSO, S.H&S and the Facility Supervisor.
The revised procedure will be placed on the agenda for the next regularly scheduled meeting of the SRC.
t A0P's are entered in 3-ring binder manuals.
11.4.2 Availability Manuals are issued to individual workers and placed in areas where the procedures apply.
11.5 TRAINING 11.5.1 General Radiation Protection Training The LRC provides two training programs covering the nature, use and control of radiation, and radioactivity. These courses are pre-sented to ensure that all LRC personnel receive training appropri-ate to their activities and to fulfill obligations under the NRC license to provide such training.
The courses consist of a series of lectures intended to present the l
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 License No SNM-778 Docket No.70-824 Date December, 198 3 l
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qV are included and discussed. Training aids, such as motion pictures and self-study materials, are used as appropriate.
Program I is intended for new employees who will be scheduled for such training within 30 days of reporting to work at the LRC.
Program II is intended for personnel working with radioactive materials.
Personnel selected by their section manager to be an Authorized User (Section 1.6) of radioactive materials (i.e.,
employees who may handle licensed material unsupervised, health physics technicians, etc.) will be scheduled for these courses.
Personnel will not.be permitted to work unsupervised with licensed material until they are trained in radiation protection and criticality safety and designated an Authorized User. Retraining of Authorized Users of radioactive materials is performed annually.
Workers who are exposed to ionizing radiation are classified as radiation workers and will receive training commensurate with their exposure as required by Title 10, Code of Federal Regulations, Part 19 (10 CFR 19). This training will include Program II as necessary.
Training in area operating procedures and special area procedures is the responsibility of the line supervisor.
This training should be accompanied with appropriate formal and on-the-job training as q
the job requirements dictate.
11.5.2 Program I This course is available to new office employees and is presented to employees within 30 days of reporting to work at the LRC.
It provides an introduction to radiation and radioactivity (under-standable to the employee with no technical education or experi-ence) and a thorough coverage of safety rules and procedures, including the site emergency procedures. Subjects include types of radiation, radiation effects on humans, permissible levels, basic health physics rules, a history of radiation protection, and personal hygiene.
11.5.3 Program II New laboratory employees who work with radioactive materials are required to complete this course and pass a written test. Subjects include the following:
1 1
License No SNM-778 Docket No.70-824 Date December,1986 Amendment No.
O Revision No.
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V 1.
Radioactivity a.
Types of radiation b.
Radioactive decay c.
Radiation dose and dose rates d.
Exposure control methods - time, distance, shielding e.
External and internal exposure hazards f.
Respiratory protection g.
The importance of maintaining exposures as low as is reasonably achievable (ALARA) h.
Risks from radiation exposure including exposure of females and the embryo / fetus 1.
Radiation exposure compared to other hazards in the work place.
l 2.
Health Physics Instruments l
a.
Personnel monitoring devices b.
Cutie pie and Geiger-Mueller counter c.
Alpha survey meter d.
Air monitors e.
Criticality alarm system f.
Emergency equipment g.
Instructions in field use of instruments.
3.
Regulations and Procedures a.
Code of Federal Regulations (including 10 CFR 19) b.
License requirements l
License No SNM-778 Docket No.70-824 Date December,1986 Amendment No.
O Revision No.
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d c.
Shipment of radioactive materials d.
Waste disposal e.
Internal procedures.
Parts of Program II may be waived as appropriate for technical and scientific personnel already knowledgeable and experienced in working in radiation areas and with licensed material. However, such personnel must pass the written examination required for Program II.
j 11.5.4 Respiratory Protection Training Training in respiratory protection techniques will be required of all employees 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 employees using respiratory protection will be included in the training courses.
Periodic retraining will be scheduled, at the discretion of the qualified individual, to ensure that a high degree of employee proficiency in the use of respiratory protective I
devices is maintained.
l OV 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 I
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 for respiratory protection and the need to make every reason-able effort to reduce or eliminate the need for respiratory protection.
License No SNM-778 Docket No.70-824 Date December,1986 Amendment No.
O Revision No.
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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 respira tor.
h.
Discussion of the uses of various cartridges and canisters available for air-purifying respirators.
1.
Review of radiation and contamination hazards, including a review of other protective equipment that may be used with respirators.
j.
Instruction in emergency actions to be taken in the event of respirator malfunction.
k.
Classroom instruction to recognize and cope with 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 l
3 l
presented on respiratory protection.
{
11.6 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 l
Engineering Supervisor determines if the request involves a facility l
change.
If a facility change is involved, the work order is forwarded to the Facility Supervisor.
It is the Facility Super-i 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 Safety, the Industrial Safety Officer, and the Facility Supervisor.
l l
License No SNM-778 Docket No.70-824 Date December, 1986 Amendment No. O Revision No.
3 Page 11-21 i
Babcock &Wilcox l
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i O
V Completed forms are kept on file by the maintenance supervisor and are audited once a month by the Health Physics Group.
l O License No SNM-778 Docket No.70-824 Date December,1986 Amendment No.
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FIGURE 11-1 a
RESEARCH AND DEVELOPMENT DIVISION E. A. WOMACK VICE PRESIDENT LYNCH 8URG RESEARCH CENTER SPECIAL PROJECTS /
LYNCHSURG TECHNICAL OPERATIONS DECOMMISSIONING P. S. AYRES T. C. ENGELDER
, MANAGER "
MANAGER DECOMMISSIONING
~~~~~~~~~
PURCHASING PURCHASING MANAGER R&DD LRC J. R. PARSELL MANAGER MANAGER SAFETY AND ANNTWG UCENSWG CONTROLLER ANO R. L. BENNETT R&DD ADMINISTRATIVE MAllAGER l
SERVICES LRC J. P. DORAN W,ANAGER
\\
I FACILITIES DEVELOPMENT MO FACILITIES LABORATORY A. E. WEHRMEISTER LRC UR CE R&DD C. E. BELL MANAGER MANAGER MANAGER I
l
(
QUALITY OUALITY MATERIALS ASSURANCE ASSURANCE ENGINEERING R&DO LRC LABORATORY J. E. KRAMER P. S. AYRES MANAGER ADMINISTRATOR
- MANAGER, l
l SCIENTIST C. S. CALDWELL l
l
= =-INDICATES FUNCTIONAL REPORTING License No SNM-778 Docket No.70-824 Date December,1986 Ameruiment No.
O Revision No.
3 Page 11-23 O
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/~3 FIGURE 11-2 V
LYNCH 8URG RESEARCH CENTER LYNCH 8URG TECHNICAL OPERATIONS P. S. AYRES MANAGER SAFETY Review
. COMMITTEE 4
I I
FACILITY SUPERVISOR SAFETY AND LICENSING l
NUCLEAR SAFETY R. L. BENNETT l
OniCER A. F. OLSEN:
1 F. M. ALCORN I
I I
I O
I I
LICENSING HEALTH AND SAFETY ACCOUNTABILITY ADMINISTRATOR SUPERVISOR SPECIALIST A. F. OLSEN G. S. H00VLER K. D. LONG HEALTH PHYSICS INDUSTRIAL STAFF SAFETY S. W. SCHILTHELM R. R. SPRADLIN
. INOICATES FUNCTIONAL REPORTING License No SNM.778 Docket No.70-824 Date December,1986 Amendment No.
O Revision No.
3 Page 11-24 G
Babcock &Wilcox a McDermott company
A TABLE OF CONTENTS (Continued) b Section Page 12.7.1 Clothing 12-12 12.7.2 Emergency Clothing 12-12 12.8 ADMINISTRATIVE CONTROL LEVELS 12-12 12.8.1 Internal Occupational Exposure 12-12 12.8.2 External Occupatio'al Exposure 12-18 n
12.8.3 Airborne Activity 12-19 12.8.4 Liquid Activity 12-20 12.8.5 Surface Contamination 12-20 12.9 RESPIRATORY PROTECTION 12-24 12.10 OCCUPATIONAL EXPOSURE ANALYSIS 12-25 12.10.1 External. Exposure 12-25 A
(]
12.10.2 Internal Exposure 12-30 12.11 MEASURES TAKEN TO IMPLEMENT ALARA 12-38 12.12 BI0 ASSAY PROGRAM 12-39 i
12.13 AIR SAMPLING AND MONITORING 12-40 l
12.13.1 Air Sampling Program 12-40 l
12.13.2 Air Monitoring Program 12-41 12.14 SURFACE CONTAMINATION 12-42 12.14.1 Smear Surveying 12-42 12.14.2 Direct Radiation Surveys 12-45 12.14.3 Personnel Contamination Surveys 12-47 1
i License No SNM-778 Docket No.70-824 DateDecember, 1986 l
I Amendment No.
O Revision No.
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suV List of Tables Table Page 12-1 PORTABLE RADIATION PROTECTION INSTRUMENTATION 12-10 12-2 STATIONARY RADIATION PROTECTION INSTRUMENTATION 12-11' 12-3 PLUT0NIUM BI0 ASSAY ACTION CRITERIA 12-13 12-4 PLUT0NIUM BI0 ASSAY ~ ACTION CRITERIA 12-14 12-5 URANIUM BI0 ASSAY ACTION CRITERIA 12-15 12-6 FISSION PRODUCT ACTION CRITERIA 12-17 12-7 STACK RELEASE ACTION LEVELS 12-20 N'
12-8 SMEAR SURVEYS IN WORK AREAS 12-21
~
12-9 ACTION LEVELS FOR LARGE AREA SMEARS 12-22 12-10 MAXIMUM PERMISSIBLE CONTAMINATION FOR SKIN SURFACES 12-23 12-11 MAXIMUM PERMISSIBLE CONTAMINATION OF CLOTHING 12-23 12-12 1984 LRC EXPOSURES BY RANGE 12-26 12-13 1985 LRC EXPOSURES BY RANGE 12-27 12-14 LRC RADIATION EXPOSURE 12-28 12-15 EXPOSURE BY GROUP (PERSON REMS) 12-29 12-16 NUMBER OF URINE BI0 ASSAY SAMPLES 12-30 12-17 1983 AIR ACTIVITY 12-31 12-18 1984 AIR ACTIVITY 12-33
~
License No SNM-778 Docket No.70-824 DateDecember, 1986 Amendment No.
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12-111 Revision No.
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x s.
/~'h
[
List of Tables (Continued)
V i
Table Page
, 12-19 WHOLE BODY COUNTS 1983 12-35
~
12-20 Am Pu LUNG COUNTING 1983 12-36 12-21 URA!!IUM LUNG COUNTING 1983 12-37
'12-22 WHOLE BODY COUNTS.1984 12-37 12-23 ACTION LEVELS FOR LARGE AREA SMEARS 12-43 12-24 SMEAP. SURVEY FREQUENCIES AND ACTION LEVELS 12-44 12-25 CONTAMINATION ACTION LEVELS 12-46 l
l
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License No SNM-778 Docket No.70-824 DateDecember, 1986 O
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. _ _... _. _ _. _ _ _....., _ ~ _
(V TABLE 12-3
.5 PLUT0NIUM BI0 ASSAY ACTION CRITERIA l
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.
Determine if area surveys support the analysis results.
3.
If area surveys confirm result, investigate the cause and take correc-tive ~ action.
4.
If resample results con-firm exposure, determine if exposure has exceeded O
50% of the maximum permissible annual dose.
5.
If the exposure has exceeded 50% of the maximum permissible annual dose, the worker shall be restricted from further exposure until i
the Supervisor, Health and Safety authorizes the removal of this restriction.
> 4 dpnf t 1.
Restrict the individual l
from any further work i
with plutonium.
rt License No SNM-778 Docket No.70-824 Date December,1986 Amendment No.
O Revision No.
3 Page 12-13 Babcock &Wilcox a McDermott company
V(O 2.
Resample the individual within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
3.
Investigate the exposure and take corrective action as needed.
4.
Evaluate for possible referral to a competent physician.
5.
Remove work restriction only with the approval of the Supervisor, Health and Safety.
TABLE 12-4 PLUT0NIUM BI0 ASSAY ACTION CRITERIA I
Bioassay Technique Action Level Action To Be Taken In-vivo
< 16E-8 Ci None
-> 16E-8 Ci 1.
Restrict worker from further exposure.
2.
Resample the individual within 10 working days.
3.
Determine if area surveys support the analysis results.
4.
If area surveys confirm l
result, investigate the cause and take correc-tive actions.
l License No SNM-778 Docket No. 70 824 Date December,1986 l
Amendment No.
O Revision No.
3 Page 12-14 Babcock &Wilcox a McDermott company
.+
p 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 annual dose, the worker shall remain on a work restriction until the Supervisor, Health and Safety authorizes the removal of the re-striction.
,1O l
12.8.1.2 Uranium bioassay action criteria.
TABLE 12-5 URANIUM BI0 ASSAY ACTION CRITERIA 1
Bioassay Technique Action Level Action To Be Taken
- a. Urinalysis
< 9 ug/L None l
- b. Urinalysis 9-16 ug/L
- 1. Determine if area surveys l
support the analysis re-sults.
License No SNM-778 Docket No.70-824 Date December,1986 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.
l by #2, investigate to determine how exposure was incurred and correct i t.
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 lifting of this restriction.
- d. Invivo
< 30 ug
- 1. None U-235
- e. Invivo 30-120 ug
- 1. Determine if area surveys support the analysis re-sults.
License No SNM-778 Docket No.70-824 Date December,1986 Amendment No.
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(VD
- 2. If #1 is positive, in-vestigate and correct as needed.
- f. Invivo
> 120 ug
- 1. Resample the individual U-235 within 10 working days.
- 2. Determine if area surveys support the analysis re-p sults.
- 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 i
be restricted from further exposure until l
r the Supervisor, Health I (
and Safety authorizes the lifting of this restric-tion.
12.8.1.3 Beta-gamma activity - Workers who work in areas where beta-gamma j
internal exposure is likely (Hot Cells, Radiochemistry, Health Physics) shall be in-vivo counted at approximately annual interval s, l
1 l
TABLE 12-6 i
FISSION PRODUCT ACTION CRITERIA Analysis Action Level Action to be Taken 1
In-vivo
>10% MP08 Remeasure subject to determine effective M1f life of the contami-nant and plot decay curves.
License No SNM-778 Docket No.70-824 Date December,1986 Amendment No.
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\\.J Followup program will continue until tl.a 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 55%MP08 Resample excreta to confirm presence of contamination and to establish rate of elimination.
Perform isotopic analysis if >10%
MP08 is a possibility.
In-vi tro
>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 i
provide data necessary for estimating internal dose rates, retention functions, and dose commitments; and to determine l ]s whether 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.
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 employees.
Personnel dosimeters issued pursuant to 10 CFR 20.202 shall be read on a monthly basis.
l The employee's line supervisor is responsible for keeping exposures below 300 millirem per week and 1250 millfrem per quarter. The Supervisor, 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, Lynchburg Technical Operations.
If an employee has received the quarterly ifmit and I
the Manager, Lynchburg Technical Operations has not authorized exceeding the limit, the employee's work shall be restricted to prevent further exposure for the remainder of the quarter.
License No SNM-778 Docket No.70-824 Date December,1986 Amendment No.
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(3 G) 12.8.3 Airborne Activity 12.8.3.1 Air Monitoring Program - Air monitoring in operating areas of the LRC 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 adequatejwarnings 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.
Reentry 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 O
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.
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 B.
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 License No SNM 778 Docket No.70-824 Date December,1986 Amendment No.
O Revision No.
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p 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.
TABLE 12-7 STACK RELEASE ACTION LEVELS i
Release Product Action Levels Beta Particulate 200 uCi/ week Alpha Particulate 1 uCi/2 weeks (long lived)
Kr-85 70 Ci/ week H-3 3 Ci/ week l
I-131 200 uCi/ week l
l Ob 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 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 LRC prior to discharge.
No 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 l
surveys are performed based on the level of work performed in the l
l License No SNM 778 Docket No.70-824 Date December,1986 Amendment No.
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V 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 />.
j 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 Building A Labs.
Monthly 200 Hot Cell Oper. Area Monthly 200 Scanning Electron Monthly 200 Microscopy Lab.
Exit portals from Biweekly 200 controlled areas l
<-------------------BETA + GAMMA-------------------------->
i Building A Labs.
Monthly 2000 Building B Counting Lab.
Monthly 2000 Scanning Electron Monthly 2000 Microscopy Lab.
Hot Cell Operations Area Bimonthly 2000 License No SNM-778 Docket No.70-824 Date December,1986 Amendment No.
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Cask Handling Area Bimonthly 22000 Radiochemistry Lab.
Bimonthly 22000 Exit portals from Biweekly 2000 coitrolled areas
- Minimum frequency specified. More frequent surveys are performed, based on work loads.
Large area smears are usd to survey many square meters of surface area. Action levels for large area smears are given below.
TABLE 12-9 ACTION LEVELS FOR LARGE AREA SMEARS 1.
Routine Large Area Smears (1000-5000 dpm) - Repeat the large area smear.
If results rhow levels of contamination above 1000 dpm, take smears in 2ma11er areas to locate the source.
Decontaminate all areas in which the smear results indicate p
contamination above 1000 dpm/100 square feet.
C 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/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-tamination monitors (friskers) are located at all normal exits License No SNM-778 Docket No.70-824 Date December,1986 Amendment No.
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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 & 12-11. Any contamination in excess of these limits should be reported immediately to the Health and Safety Group. The Health and Safety Group will supervise the decontami-nation and determine if clothing must be discarded. The approval of the Health and Safety Group shall be required to allow any individual to leave a contaminated area who is contaminated above background radiation levels.
TABLE 12-10 MAXIMUM PERMISSIBLE CONTAMINATION FOR SKIN SURFACES Fixed Alpha Fixed Beta-Gamma Smearable Surface dpm/100 sq. cm.
dpm/100 sq. cm.
( Alpha, Beta-gamma )
Body 220 2200 None Detectable Hands 220 2200 None Detectable TABLE 12-11 MAXIMUM PERMISSIBLE CONTAMINATION OF CLOTHING (dpm/100 sq. cm)
Smearable Item Fixed Alpha Fixed Beta-Gamma Alpha, Beta-Gamma Shoes:
Contaminated Zone Inside 2,200 22,000 220 2,200 Outside 22,000 220,000 2,200 22,000 License No SNM 778 Docket No.70-824 Date December,1986 Amendment No.
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)
D' Personal Inside 2,200 2,200 220 2,200 Outside 22,000 22,000 220 2,200 Clothing:
Contaminated Zone 2,200 2,200 Not Detectable Personal 2,200 2,200 Not Detectable i
12.8.5.3 Release of Equipment or Packages - Packages and equipment are surveyed by the Health and Safety Group. 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 and Equip-ment Prior to Release for Unrestricted Use of Termination of Licenses for Byproduct, Source, or Special Nuclear Material, dated November,1976."
12.9 RESPIRATORY PROTECTION The primary objective of a respiratory protection program is to limit p
the inhalation of airborne radioactive materials and other hazardous t d ma terials.
This objective is normally accomplished through the use of engineering controls, including process, containment, and venti-lation equipment. When engineering controls are 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 at the LRC. 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 Ma terial s. "
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.
2.
Written procedures to ensure proper selection, supervision, and training of personnel using such protective equipment.
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pi 3.
Written procedures to ensure the adequate individual fitting of g'd respirators, as well as procedures to ensure the testing of respiratory protective equipment for operability immediately prior to each use.
4.
Written procedures for maintenance 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 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 Q
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 least annually.
Other details of an effective respiratory protection program can be found in the above mentioned documents and the LRC health physics procedures.
12.10 OCCUPATIONAL EXPOSURE ANALYSIS 12.10.1 External Exposure - The external radiation exposure received by LRC employees is presented in Tables 12-12 through 12-15. Tables 12-12 and 12-13 show the exposures by ranges and the number of employees in each range for calendar years 1984 and 1985 respectively.
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TABLE 12-12 1984 LRC 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 O'.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 December,1986 Amendment No.
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C; V
TABLE 12-13 1985 LRC EXPOSURES BY RANGES Annual Whole Body Dose Number of Individuals Ranges (Rems)
In Each Range No Measurable Exposure 185 Measurable Exposure <0.100 83 0.100 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 Table 12-14 presents the exposures received by LRC employees for calendar years 1981 through 1984. The row entitled "Off Site" gives the exposures received by LRC employees at other licensed facilities.
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O TA8LE 12-14 LRC RADIATION EXPOSURE 1984 1983 1982 1981 Total Person Rems 23.5 18.4 19.4 26.3 0ff 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 The exposure received by LRC employees is categorized by group in
'O Table 12-15 for exposures received for calendar years 1983 and 1984.
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(9 TABLE 12-15 G'
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 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 12-15 reflects this in all categories.
Table 12-15 also reflects this increase in two of the three C) affected groups. Only Health and Safety saw a reduction in the b
group's exposure. The amount of exposure received from off-site work reversed a three year period of decreases. Table 12-15 reflects this in the increase in the Systems Design & Engineering Group's exposure.
The increases noted in Tables 12-14 and 12-15 do not indicate a decrease in the vigilance given by LRC management to personnel exposures nor do they suggest a decreased ALARA emphasis.
Exposure history at the LRC shows wide variances because of the 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 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.
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(3 V) 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 12-16 below presents the number of urine bioassay samples taken during 1983 and 1984.
TABLE 12-16 j
NUMBER OF URINE BI0 ASSAY SAMPLES 1983 1984 Month U
Pu U
Pu January 5
20 4
February 13 6
March 15 12 April 19 19 18 9
13 5
May June 16 16 17 8
July 11 8
15 6
pd 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 grams / liter (lowe.- ifmit of detection), except on four occasions when the analysis indicated the presence of uranium but none met the resample limit of 20 grams / liter. All plutonium analyses were below the minimum sensitivity which varied from 0.00 + 0.1 to 0.3 + 0.4 dpm per sample.
In 1984, all samples for uranium were less than 5 grams / liter (lower limit of detection), except on one occasion 27 g/ liter was reported. A resample showed that the level had returned below the lower limit of detection. All plutonium samples indicated 0.0 + (0.01 to 0.6).
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12.10.2.2 Air Sampling Results - The air sampling program is the first m
[V) line of defense for all operations of this type, but the bio-assay program, along with lung counts, is the final step in the estimation of exposure that may occur.
12.10.2.2.1 Table 12-17 presents a summary of the air sampling program at the LRC for calendar year 1983, for fixed air samplers.
TABLE 12-17 1983 AIR ACTIVITY j
(VALUES IN Ci/ml)
Approximate Maximum Labs Average Concentra tion MPC
-15
-14
-10 15 3x10 1.2x10 1x10
-15
~14
-10 16 3x10 1.5x10 1x10
-15
-13
~11 17 8.7x10 1.8x10 4x10
-15
~14
-10 19 7x10 8.7x10 1x10
-15
-15
-10 27 2.4x10 5.7x10 1x10
-15
-15
-10 44*
2x10 6.5x10 1x10
-12
-10
-9 P
Cask Handling Area 1.9x10 1.27x10 9x10 k
6.7x10 4.5x10 4x10
-15
-13
-11
-14
-13
~9 Hot Cell 1x10 1.25x10 9x10
-16
-15
~11 5x10 1.2x10 4x10
-14
-13
-9 Recirculated Air "C" 1.5x10 3.5x10 9x10
-15
~I4
-11 4x10 1.93x10 4x10
~14
~14
-9 Waste Storage Area 1.5x10 2.6x10 9x10
-16
-15
-II 7x10 1.7x10 4x10
~14
-13
~9 Laundry 3x10 1.5x10 9x10
-15
~14
~II 3x10 2.5x10 4x10
-14
-12
-9 Radio Chem Lab 7x10 2.3x10 9x10
-15
~14
-II 1.5x10 1.5x10 4x10
- Discont..iued in Sept.
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[
12.10.2.2.2 On 338 occasions in 1983, breathing zone air samples were U
taken to measure the airborne activity to which workers were exposed.
In no case was anyone exposed to greater than 2 MPC of airborne 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 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 respiratory system. This system incorporates the 3M hard hat which is NIOSH approved with a protection factor of 1,000. Breathing zone samples are taken outside of the hood each time this system is used.
3.
Operations in Building C may involve bagging operations p/
with plutonium glove boxes. All operations of this type U
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, the full face masks with a protection factor of 50 may be used.
4.
Other minor operations requiring respiratory protection are: changing HEPA filters, repair work on NPD site support equipment, and any other operations where Health and Safety believes that there is a potential of airborne activi ty.
5.
It should to be noted that a major operat;:.n 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 License No SNM 778 Docket No.70-824 Date December 1986 Amendment No.
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one in Building C has been exposed in excess of 2 MPC hr 6
in one week.
In most cases, radioactivity above back-ground is undetectable.
12.10.2.2.3 Table 12-18 presents a summary of the air sampling program at the LRC for calendar year 1984, for fixed air samplers.
TABLE 12-18 1984 AIR ACTIVITY (VALUES IN yC1/ml)
Approximate Maximum Labs Average Concentration MPC 15*
SE-15 3.9E-13 4E-11 19 7E-15 1E-13 1E-10 27**
2.4E-14 7.5E-15 1E-10 Soil Processing ***
1E-15 7.4E-15 4E-11 Cask Handling Area SE-13 1.2E-11 9E-9 SE-15 SE-13 4E-11 Hot Cell 8E-15 6.7E-13 9E-9 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 License No SNM 778 Docket No. 70 824 Date December,1986 Amendment No.
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Od Radio Chem 3E-14 1.0E-12 9E-9 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 measyre 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 least a factor of 1000 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 1000. This system incorporates a double bibb hood which has reduced airborne activity to which a worker is exposed to below measurable levels.
O'v 2.
Operations outside of the isolation area in the cask handling area use the 3M hood and the supplied air respiratory system. This system incorporated the 3M hard hat which is NIOSH approved with a protection factor of 1000. Breathing zone samples are taken outside of the hood each time this system is used.
3.
Operations in Building C may involve bagging operations with plutonium glove boxes. All operations of this type t
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 License No SNM 778 Docket No. 70 824 Date December,1986 Amendment No.
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I 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 trom 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 Bujlding 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 Helgeson Scientific Services, Inc. on 32 employees during 1983.
Three had detectable activities, no other workers indicated detectable activity. The results of the three employees with detectable activity is presented in Table 12-19.
4 TABLE 12-19 WHOLE BODY COUNTS 1983 Ob (ALL VALUES IN NAN 0 CURIES)
Employee Isotope MPBB 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 In-vivo counting was performed on seven employees during 1983, for plutonium and Americium-241. These results are summarized in Table 12-20.
License No SNM 778 Docket No.70-824 Date December,1986 Amendment No.
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(
TABLE 12-20 Am - Pu LUNG COUNTING 1983 (ALL VALUES IN NAN 0 CURIES)
Employee Pu Am
'1 0
0.0010.10 2
0 0.0010.11 3
0 0.1310.13 4
0 0.0010.14 5
0 0.0010.1s 6
0 0.0010.19 7
0 0.0010.16 In-vivo lung counting was performed on nine employees in 1983, O
for uranium. The results are listed in Table 12-21.
Four of the nine indicated positive results. However, these results were not confirmed in followup urinalyses.
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License No SNM 778 Docket No.70-824 Date December,1986 Amendment No. O Revision No.
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_ _ _. ~.
TABLE 12-21 URANIUM LUNG COUNTING 1983 (ALL VALUES IN MICROGRAMS)
Employee U-235 1
0130 2
0143 j
3 0139 4
42137 5
0141 6
38133 7
76145 8
0139 9
49144 12.10.2.4 In-vivo results (1984) - Whole body counting was performed by Helgeson Scientific Services, Inc. on 99 employees during 1984.
i Twelve had positive results but these were very low levels. A summary is presented in Table 12-22.
TABLE 12-22 WHOLE BODY COUNTS 1984 (EXPOSURE VALUES IN NAN 0 CURIES)
Number of Maximum Isotope Employees Observed MPBB i
Cs-134 1
3.0 2E4 Cs-137 7
9.0 3E4 Co-60 4
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.. _ - - - ~..,. - _ -.. - - -, _ -
4 OCl In-vivo lung counting was performed on 14 employees during 1984 for Plutonium-239 and Americium-241. No plutonium was reported.
The presence of Americium-241 was indicated for 5 employees with the highest quantity being 0.26 Nanocuries (10.14) for one person.
In-vivo lung counting was performed on 20 employees during 1984 for Uranium-235.
In 5 instances, the results were positive with...
the highest result being 48 micrograms (137) for one person.
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 personnel who periodi-cally enter the area for maintenance on the HEPA filters and to calibrate an area monitor.
It also eliminated the radiation area on the roof of Building B which no longer contributes to the exposure of workers who maintain the building venti 11ation system.
12.11.2 Cleaning of the hot cells contributed significantly to exposure l
doses of workers. This cleaning operation, which is performed at O
three or four year intervals, requires the set-up table in the V
cell to be dismantled.
In 1985, this operation was performed remotely with a modified saw so that personnel did not enter the cell for this high exposure work.
12.11.3 Trash removal from the hot cell during cell cleaning operations 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-gallon drums and to be handled with long poles. This process modification reduced personnel j
exposures for this part of the operation considerably.
12.11.4 The LRC has purchased a TLD reader which provides immediate information on worker exposure. This system is not intended to l
replace the normal contract service for dose measurement but rather to provide prompt indication of unexpected exposures for non-routine operations. The system makes possible the estimation of exposures too hard to measure areas of the body such as the i
soles of feet, hands and fingers.
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/3
\\
)
12.11.5 A supplied air respiratory system has been installed to support hot cell work, principally during hot cell entries. This system provides a greater protection factor for workers in addition to providing greater worker comfort while performing '.he 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.
This method has b9en changed such that the workers, supervisors and signators of the RWP gather at a meeting where 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 consider.ible exposure savings have resulted.
[
12.12 BI0 ASSAY PROGRAM Those employees routinely working in contamination or airborne radioactivity areas will be scheduled for participation in the bio-assay program. The Health and Safety Group will select those t
employees to be sampled in the program. This selection will be based on the probability of exposure, the employee's work habits, A
the type of work in the area, air sample data, previous bioassay da ta, e tc.
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 classifi cation 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 License No SNM 778 Docket No. 70 824 Date December.1986 Amendment No.
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t
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Lj considered by the ICRP to be readily soluble (i.e., class D s
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 1974. There are two exceptions to this general guidance:
- 1.
Employees off-site during the regular visit of the bioassay service will not be scheduled for a special, make-up count, if the count was scheduled only for routine exposure control moliT-toring.
2.
Bioassays of employees working in areas in which both plutonium
"~
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 elenents, 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.
Employees working primarily with beta and gamma emitting radio-nuclides will also be included in the in-vivo bioassay analysis O
program. Any employee suspected of an exposure greater than 40 MPC-hours will be scheduled for a bioassay evaluation as soon as practicable af ter 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 of the LRC is determined through the combined use of air samplers and monitors. These programs are discussed below:
12.13.1 Air Sampling Program The air sampling program can be divided into two categories; fixed and portable.
Selection of the sampling category and the frequency of sampling is lef t to the discretion of the Supervisor, Health and Safety.
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O 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 worxer's breathing zone. These samples are usually collected weekly. However, the frequency may vary as the situation dicta tes.
Normally, these are evaluated within two weeks, after 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 fcr the decay period.
In these 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 the 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 m
the nose and mouth 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 l
Air monitoring in operating areas of the LRC is accomplished with l
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 radio-activi ty. Locations are selected based upon the ability of the monitor to provide a reasonable evalu'ttion 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 License No SNM-778 Docket No.70-824 Date December,1986 Amendment No.
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in the area, the potential for release, and the quantity and chemical form of the material.
Alarms are set in accordance with the particular operation, the material being handled, and the potential for release. Actual l
alarm points are set as low as possible commensurate with the ambient radiation levels in the area.
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 at the LRC. 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 ::amples are obtained with small, absorbent filter papers. The smear paper is moved across an area of approximately 100 sq. cm. using about S 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 l
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 operations area, the change room and main hallways in Building B.
These smears are intended to indicate the general contami-t nation environment in an area and may lead to a more extensive survey, if unexpected contamination is indicated. Normally, large area smears are eveluated 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 12-23.
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7 12.14.1.3 Action Levels - Included in Table 12-25 are the appropriate action levels to be used in designated areas of the LRC. Decon-tamination 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 LRC personnel working in these areas and to reduce contamination 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.
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 of the contamination and the potential for further contamination of equipment and/or personnel during the interval.
Decontamination of the area shall begin on the first regular work-day af ter discovery.
TABLE 12-23 l
m 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) l 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.
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m 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.
Decontaminafe all areas in which the smear results show contamination in excess of 1000 dpm per 100 sq. ft.
t 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 12-24 SMEAR SURVEY FREQUENCIES AND ACTION LEVELS l
Alpha Radiation Smear Survey l
l Action Level l
Area Frequency (dpm/100 sq. cm.)
Unirradiated, unencapsulated weekly 5,000 fuel handling areas Building B counting laboratory monthly 200 Building A laboratories monthly 200 Hot cell operations area monthly 200 Scanning electron microscopy monthly 200 laboratory License No SNM-778 Docket No.70-824 Date December,1986 Amendment No.
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3(Q Exit portals from controlled twice weekly 200 Beta Radiation Smear Survey Action Level Area Frequency (dpm/100 sq. cm.)
Building A Laboratories monthly 2,000 Building B Counting Laboratory monthly 2,000 Scanning Electron Microscopy monthly 2,000 Laboratory Hot Cell Operations Area twice monthly 2,000 Cask Handling Area twice monthly 22,000 Radiochemistry Laboratory twice monthly 22,000 Exit Portals From Controlled twice month 1y 2,000 Areas l
J 12.14.2 Direct Radiation Surveys l
Surveys of the direct radiation exposure in areas of the LRC are to be performed on a frequency established by a Health Physics engineer.
In general, these surveys require the selection of the appropriate portable survey instruments based upon the anticipated radiation levels, the types of radiation expected, and the nature or type of survey to be performed.
General 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 examination of the area for any unusual con-ditions or work habits which could affect the exposures received by personnel working in these areas.
Items of this nature should be reported immediately to the Supervisor, Health and Safety, or corrected immediately, if practical.
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(3 Results of these surveys should be reviewed by a Health Physics
(,)
Engineer to ensure that the proper posting requirements are in effect for the area and to ensure that appropriate actions are taken to keep all exposures ALARA.
Action levels for direct radiation surveys are presented in Table 12-25.
TABLE 12-25 CONTAMINATION ACTION LEVELS Transferable Surface Type of Fixed Contamina tion Area Radiation Surface Reading (dpm/100 sq. cm.)
Uncontrolled Alpha 300 dpm/100 sq. cm.
30 Beta-Gamma 0.1 mrad /h 220 Contamination
- Alpha 3000 dpm/100 sq. cm.
2,200 Beta-Gamma 1.0 mrad /h**
22,000 3(V
- The Supervisor, Health and Safety may raise these action levels.
Justification for this action must be documented and forwarded to the Safety Review Committee for their review and approval.
- This action limit applies to contamination areas which are normally radiation areas. This level of centamination will not cause a sig-nificant increase in radiation exposure.
NOTE:
This table provides limits above which decontamination must be initiated.
These action levels pertain to areas normally accessible to personnel performing normal work functions. The levels do not apply to areas requiring extraordinary precautions for entry, e.g.,
the Isolation Area, waste water tanks, etc.
In these cases, direct health physics coverage is the primary control mechanism.
License No SNM-778 Docket No.
70824 Date December,1986 Amendment No.
O Revision No.
3 Page 12-46
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Babcock &Wilcox a McDermott company 7
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V 12.14.3 Personnel Contamination Surveys 4
Personnel are required to monitor themselves for activity present on their hands, shoes, clothing, and person before exiting a con-tamination area. Contamination monitors (friskers) are located at all exits from contamination areas for this purpose. The detector (probe) 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 speed of about 0.5 inch /second. Allowable levels of contamination on skin surfaces and on items of clothing are given in Tables 12-11 and 12-12. Any contamination in excess of these limits should be reported immediately to the Health and Safety Group.
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License No SNM-778 Docket No. 70 824 Date December,1986 Amendment No.
O Revision No.
3 Page 12-47
!O 1
Babcock &Wilcox a McDermott company
TABLE OF CONTENTS Section Page 13.0 ENVIRONMENTAL SAFETY 13-1 13.1 ENVIRONMENTAL MONITORING 13-1 13.2 EFFLUENT AIR MONITORING 13-1 13.3 LIQUID EFFLUENT MONITORING 13-2 e
O License No SNM-778 Docket No.70-824 DateDecember, 1986 Amendment No.
O Revision No.
3 13-1 p,
Babcock &Wilcox a McDermott company
13.0 ENVIRONMENTAL SAFETY 13.1 ENVIRONMENTAL MONITORING Environmental sampling of the area surrounding the LRC is performed on a regular basis to evaluate changes in the levels of radioactivity in air, water, and vegetation. The minimum environmental program consists of the following.
o one continuous on-site site boundry air sample (Figure 13-1) l o monthly water samples from the James River collected above and below the liquid discharge point (Figure 13-2) o continuous sampling of rain water on-site (Figure 13-1) o quarterly samples of river silt and near-river vegetation (Figure 13-2).
Normally, LRC personnel are responsible for collecting the environ-mental samples. Analysis of these samples may be performed on-site or the samples may be analyzed by a commercial laboratory.
Environmental sampling data for the calendar years 1982,1983, and 1984
/m is given in "Lynchburg Research Center, Environmental Report, October, l
1985," Tables 2.2 through 2.6.
l 13.2 EFFLUENT AIR MONITORING Planned discharges of air to the environment shall be in compliance with the limits specified in 40 CFR 61.
Compliance with 40 CFR 61 demonstrated by calculation that the total annual release limits per-mitted by the license will not exceed the specified 25 millirem whole body and 75 millirem organ dose limits for persons located at the point of maximum ground level concentration.
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 routed through a two-stage HEPA filter.
License No SNM-778 Docket No. 70 824 Date0ecember, 1986 Amendment No.
O Revision No.
3 p,,,13-1 Babcock &Wilcox a McDermott company
O Discharge through the stack is accomplished with a large blower, powered normally by a large electric motor operated on off-site power.
Emergency pcNer is supplied by an internal combustion engine coupled to the blower shaft through a centrifugal clutch. On loss of off-site power, the engine starts automatically and takes over the load upon reaching the proper speed.
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 a'ctivity 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.
Set points for these alarms are determined by the Health and Safety Group. These alarms are connected to an alarm panel located in the Health Physics Laboratory in Building B.
Air from areas equipped with continuous air monitors (and which is below the applicable MPC for an unrestricted area) may be exhausted, through HEPA filters, directly to the roof of the building. Air from areas which have a low potential for airborne activity may be exhausted directly to the roof of the building.
qQ 13.3 LIQUID EFFLUENT MONITORING All potentially radioactive liquids are collected in tanks located in the Liquid Waste Disposal Facility. The contents of each tank are mixed, samples are obtt..ned, and are analyzed for radioactivity before the liquids are released to the waste treatment plant at the Naval Nuclear Fuel Division (NNFD).
Liquid waste tanks are sampled on a quarterly frequency, before release to the NNFD or at other times determined by the Health and Safety Group. Results of all analyses are reported in units of activity per unit volume and records of these evaluations are retained by the Health and Safety Group.
Water samples are also obtained on a quarterly basis from the retention basin located behind Building C and the holding pond located near Building J.
License No SNM-778 Docket No.70-824 DateDecember, 1986 Amendment No.
O Revision No.
3 p g,13-2 Babcock &Wilcox a McDermott company
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