ML20137Y800
| ML20137Y800 | |
| Person / Time | |
|---|---|
| Site: | 07000824 |
| Issue date: | 10/31/1985 |
| From: | BABCOCK & WILCOX CO. |
| To: | |
| Shared Package | |
| ML20137Y745 | List: |
| References | |
| BAW-381, BAW-381-R, BAW-381-R00, NUDOCS 8512110151 | |
| Download: ML20137Y800 (195) | |
Text
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I BAW-381 s
October, 1985 1
DEMONSTRATION AND CONDITIONS FOR LICENSE SNM-778 S
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l BABC0CK & WILC0X Research and Development Division Lynchburg Research Center P. O. Box 11165 Lynchburg, VA 24506-1165 l
l License No SNM-778 Docket No.70-824 Date October, 1985
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l su2ggg $b$r sabcocka wiscox DR a McDermott company
r8b TABLE OF CONTENTS Section Page i
1.0 STANDARD CONDITIONS AND SPECIAL AUTHORIZATIONS.
1-1 2.0 GENERAL ORGANIZATIONAL AND ADMINISTRATIVE REQUIREMENTS 2-1 3.0 RADIATION PROTECTION 3-1 4.0 NUCLEAR CRITICALITY SAFETY 4-1 5.0 ENVIRONMENTAL PROTECTION 5-1 6.0 SPECIAL PROCESS COMMITMENTS 6-1 7.0 DECOMMISSIONING PLAN 7-1 8.0 RADIOLOGICAL CONTINGENCY PLAN 8-1 9.0 OVERVIEW 0F OPERATION 9-1 10.0 FACILITY DESCRIPTION 10-1 11.0 ORGANIZATION AND PERSONNEL 11-1 y
12.0 RADIATION PROTECTION 12-1 13.0 ENVIRONMENTAL SAFETY 13-1 14.0 NUCLEAR CRITICALITY SAFETY 14-1 15.0 PROCESS DESCRIPTION AND SAFETY ANALYSES 15-1 16.0 ACCIDENT ANALYSES 16-1 License No SNM-778 Docket No.70-824 Date October,1985 O
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TABLE OF (,0NTENTS
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Section Page 1.0 STANDARAD CONDITIONS AND SPECIAL AUTHORIZATIONS 1-1 1.1 NAME.
1-1 1.2 LOCATION 1-1 1.3 LICENSE NUMBER AND PERIOD 1-1 1.4 POSSESSION LIMITS 1-2 1.5 LOCATION OF POSSESSION AND USE.
1-3
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DEFINITIONS 1-3 1.7 AUTHORIZED ACTIVITIES.
1-4 1.8 EXEMPTIONS AND SPECIAL AUTHORIZATIONS 1-5
~s License No SNM-778 Docket No.70-824 Date October,1985 O
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OO PART I LICENSE CONDITIONS 1.0 STANDARD CONDITIONS AND SPECIAL AUTHORIZATIONS 1.1 NAME Name - McDermott International, Inc.
Babcock & Wilcox Research and Development Division Lynchburg Research Center McDermott International Inc. is incorporated under the laws of the Republic of Panama.
Principle Office - 1010 Common Street, New Orleans, Louisiana.
1.2 LOCATION
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j Address - Babcock & Wilcox Lynchburg Research Center P. O. Box 11165 Lynchburg, Virginia 24506-1165 The Lynchburg Research Center is located in Campbell County, Virginia, near the James River, approximately four miles East of the city of Lynchburg.
1.3 LICENSE NUMBER AND PERIOD License Number - SNM-778 Docket Number 70-824 Period of Time - It is requested that this license be renewed for a period of 10 years, i
l License No SNM-778 Docket No.70-824 Date October,1985
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/^N 1.4 POSSESSION LIMITS Material Physical Form Enrichment Anount
- 1. Uranium enriched Encapsulated or
> 20 %
3.5 Kg con-in U-235 irradiated tained U-235
- 2. Uranium enriched Unencapsulated
> 20 %
0.27 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-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 Unencapsulated 0.31 Kg and unirradiated
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- 8. Source Material Any 6000 Kg
- 9. Fission Products Irradiated Fuel Quantity
& Transuranium contained in Elements 4 irradiated fuel as-semblies.
- 10. Fission products Irradiated fuel 5,000,000 Ci.
- 11. Any byproduct Irradiated 50,000 Ci.
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.
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- 13. Transuraniun Any 20 milli-elements curries each
- 14. Cf-252 Sealed Sources 4 milligrams
- 15. Am-241 Sealed Sources 30 Ci
- 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 millicuries may be possessed and used in locations other than the Lynchburg Research Center for performing instrument calibration, electronic noise analysis, shielding studies, or p
similar operations.
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1.6 DEFINITIONS 1.6.1 LRC means Lynchburg Research Center.
1.6.2 SRC means Safety Review Committee.
1.6.3 SNM means Special Nuclear Material.
1.6.4 Licensed Material means source, byproduct, or SNM received, possessed, used or transferred under a general or specific license issued by the Nuclear Regulatory Commission.
1.6.5 Research and Development (R&D) means (1) theoretical analysis, exploration, or experimentation; or (2) the extension of investigative findings and theories of a 3cientific or technical nature into practical application for exparimental and demon-stration purposes, including the experimental production and License No SNM-778 Docket No.70-824 Date October,1985 O
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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 Director means the Director, Lynchburg Research Center.
1.6.8 Qualified Person means a person who is assigned by his supervisor to work in an area where licensed material is handled and who is familiar with the hazards in the area.
A qualified person may also be referred to as a Category A Person.
1.6.9 Authorized User means a person who may work with licensed material unsupervised and may supervise others, not so designated, in the handling of licensed material.
1.6.19 Calibration means a comparison of a measurement standard of known accuracy with another standard or instrument to detect, correlate or adjust any variation in the accuracy of the item being com-pared.
Calibration also includes standardization.
1.6.11 Standardization means, the act of using standards which are p
traceable to the NBS, a nationally accepted measurement system, or natural phenomena to set up an instrument.
Standardization
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must be performed before and after use.
1.6.12 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 units by at least 8-inches edge-to-edge and 24-inches center-to-center.
1.7 AUTHORIZED ACTIVITIES 1.7.1 Licensed material shall be used in the performance of Research and Development.
1.7.2 The LRC may deliver licensed material to a carrier for transport in accordance with the regulations in 10 CFR 71 and 49 CFR.
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1.7.3 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.
1.7.4 The LRC may dispose of licensed material pursuant to the regula-tions in 10 CFR 20 and 10 CFR 61.
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.
- 1. 8.1'. P.
Bioassay samples (urinalyses and in-vivo lung counting) shall be analyzed for plutonium and uranium if the sanple involves an employee working in an area where both Pu and U are present in air except when the Supervisor, Health and Safety can demon-strate that the analysis for Pu is more sensitive, then he can authorize the analysis for Pu only.
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(7 TABLE OF CONTENTS Section Page 1
2.0 GENERAL ORGANIZATIONAL AND ADMINISTRATIVE REQUIREMENTS.
2-1 2.1 POLICY.
2-1 2.2 ORGANIZATION RESPONSIBILTIES AND AUTHORITIES 2-1 2.3 SAFETY REVIEW COMMITTEE 2-3 2.4 APPROVAL AUTHORITY FOR PERSONNEL SELECTION.
2-4 2.5 PERSONNEL EDUCATION AND EXPERIENCE REQUIREMENTS 2-4 2.6, TRAINING 2-5 2.7 OPERATING PROCEDURES 2-6 2.8 INTERNAL AUDITS AND INSPECTIONS 2-7 p
2.8.1 Nuclear Criticality Safety 2-7
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2.8.2 Health Physics 2-7 2.8.3 General Safety and Compliance 2-7 2.9 INVESTIGATIONS AND REPORTING 0F 0FF-NORMAL OCCURRENCES.
2-8
- 7. 9.1 License Ad'ministrator.
2-8 2.9.2 Supervisor, Health and Safety 2-9 2.9.3 Facility Supervisor 2-9 2.10 RECORDS 2-9 2.10.1 Health and Safety Group 2-9 2.10.2 Nuclear Safety Officer 2-10 2.10.3 License Administrator.
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G 2.0 GENERAL ORGANIZATIONAL AND ADMINISTRATIVE REQUIREMENTS 2.1 POLICY It shall be the policy of the LRC to maintain radiation exposures to employees and the general public as low as is reasonably achievable.
The facility procedures to ensure the safe handling of licensed material are the Area Operating Procedures.
2.2 ORGANIZATION RESPONSIBILITIES AND AUTHORITIES.
2.2.1 Director - The Director is ultimately responsible for all safety at the LRC.
2.2.2 Laboratory Managers - The Laboratory Managers are responsible for the safety of personnel in their laboratories. The Laboratory Managers report to the Director.
2.2.3 Section Managers - Section Managers are responsible for the safe performance of projects under their purview. To this end, they
(~N are responsible for ensuring that personnel in their sections follow all applicable Area Operating Procedures. The Section
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Managers report to the Laboratory Managers.
2.2.4 Facility Supervisor - The Facility Supervisor is responsible to the Director for the safe conduct of all operations at the LRC and for ensuring that all applicable operations are conducted in compliance with the license and applicable regulations. To fulfill these responsibilities the Facility Supervisor shall have the authority to stop any operation that he feels is unsafe or in violation of license. The Facility Supervisor has approval authority for all Area Operating Procedures and Radiation Work Permits.
He shall submit items for review to the SRC.
2.2.5 Manager, Safety and Licensing - The Manager, Safety and Licensing reports to the Director. The Supervisor, Health and Safety, the Accountability Specialist, and the License Administrator report to this manager.
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2.2.6 Supervisor, Health and Safety - The Supervisor, Health and Safety is responsible for providing adequate facilities, procedures, and properly trained personnel to implement the Health Physics Plan and industrial safety program. He is responsible for health physics and industrial safety activities. The Supervisor, Health j
and Safety reports to the Manager, Safety and Licensing, and has direct access to the Director in matters pertaining to Health and Safety. The Supervisor, Health and Safety has approval authority for all Area Operating Procedures and Radiation Work Permits. He shall conduct training programs for new employees and Authorized Users of Radioactive Material.
He shall be responsible for the shipment of licensed material.
He shall be a member of the SRC.
2.2.7 Health Physics Engineer - A Health Physics Engineer shall adminis-ter activities of the Health Physics Staff. He shall report to the Supervisor, Health and Safety.
2.2.8 Industri11 Safety Officer - The Industrial Safety Officer shall administer the industrial safety progran. He shall report to the Supervisor, Health and Safety.
2.2.9 Nuclear Safety Officer - The Nuclear Safety Officer shall be responsible for ensuring that no operation at the LRC results in the inadvertent assembly of a critical mass. He shall have
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approval authority for all Area Operating Procedures. He shall (3
conduct training programs in criticality safety and perform
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criticality safety calculations. He shall report to the Director.
2.2.10 License Administrator - The License Administrator shall be responsible for administering the license. He is the primary liaison with the NRC and other federal, state, and local agencies in matters that pertain to nuclear activities.
He shall be the coordinator of the SRC and the Safety Audit Subcommittee and shall represent management on both.
He shall maintain the permanent records of the SRC and shall be responsible for assuring that appropriate action is taken to correct SAS audit findings that are approved by the Director.
He shall report to the Manager, Safety and Licensing.
2.2.11 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.
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2.3 SAFETY REVIEW COMMITTEE 2.3.1 Function 2.3.1.1 The SRC shall review and approve all Area Operating Procedures.
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 licensed material.
2.3.1.5 The SRC may approve an operation utilizing licensed material without written procedures, after reviewing the proposed operation.
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.
(n) 2.3.3 Safety Audit Subcommittee LJ 2.3.3.1 The SAS shall perform audits of the LRC for the Safety Review Committee.
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. At least one calendar month shall separate succeeding audits.
2.3.3.4 SAS membership shall be appointed by the Director.
2.3.4 Reporting 2.3.4.1 The SRC shall report to the Director.
2.3.4.2 The SAS shall report to the Chairman, SRC.
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2.3.5 Recordkeeping 2.3.5.1 Minutes of the SRC proceedings shall be prepared by the Chairman, SRC.
2.3.5.2 SRC Minutes shall be forwarded to the Director 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 Director by the Chairman, SRC with comments, as he deems appropriate.
2.4 APPROVAL AUTHORITY FOR PERSONNEL SELECTION 2.4.1 The Director shall approve the personnel selected for safety-related positions specified in Section 2.2 of this Part. The Director is appointed by the R & D Division Vice President.
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2.5 PERSONNEL EDUCATION AND EXPERIENCE REQUIREMENTS 2.5.1 Director - The Director shall be appointed in accordance with Company policy.
2.5.2 Laboratory Managers - The Laboratory Managers are appointed by the Director 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.
2.5.4 Facility Supervisor - The Facility Supervisor shall have a degree in his related work and three years experience, 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.
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2.5.5 Manager, Safety and Licensing - The Manager, Safety and Licensing shall have a BS degree in a technical field and five years er.perience 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 five years experience and demonstrate proficiency in the application of radiation safety principles and be knowledgeable in fields related to radiation protection.
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 the National Fire Protection Association.
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 O
experience with nuclear criticality safety calculations similar to
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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 nust demonstrate to Company management his knowledge of the principles necessary for the accountability and safeguarding of special nuclear materials.
2.6 TRAINING 2.6.1 Progran 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 coverage of safety rules and procedures including emergency procedures.
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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. Pirts 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 protection 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.
2.7 OPERATING PROCEDURES 2.7.1 Area Operating Procedures (A0P) - Area Operating Procedures are n
facility procedures for the safe handling of licensed naterial.
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A0P's contain the provisions to assure the safety of the operation with regard to Health Physics and Nuclear Criticality Safety.
Each A0P shall be approved by the Nuclear Safety Officer or his desig-nated alternate, the Supervisor, Health and Safety or his desig-nated alternate, the Facility Supervisor or his designated alternate, and the Safety Review Committee.
2.7.2 A0P's may be revised 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.
2.7.3 A0P's shall be available in each operations area where they apply and shall be followed by operations personnel.
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2.7.4 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.
2.7.5 A0P manuals shall be reviewed annually by the Facility Supervisor to assure that the manuals contain the most current revision of the procedures.
2.8 INTERNAL. AUDITS AND INSPECTIONS 2.8.1 Nuclear Criticality Safety 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 Director within two weeks of completing the audit. The audit reports shall be forwarded to the Facility Supervisor and the License Adminis-trator.
The License Administrator shall be responsible for assuring that the appropriate corrective actions are taken to address the audit findings, i
V 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 evaluating the health physics aspects of operations. This audit shall be conducted in accordance with written audit guidance.
This audit shall be conducted once each month. A report of his findings shall be made to the Director within two weeks of com-pleting the audit. The audit reports shall be forwarded to the Director and the License Administrator. The License Administrator shall be responsible for assuring the appropriate corrective actions are taken to address the audit findings.
2.8.3 General Safety and Compliance License No SNM 778 Docket No.70-824 Date October,1985 O
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2.8.3.1 The SAS performs audits of general safety and compliance at the
'y LRC.
These audits shall be conducted three times annually with at least one calendar month separating succeeding audits. The SAS shall include an audit of the Health and Safety Group at least once annually. Other areas of LRC operations shall be audited for compliance with written requirements and the exercise of acceptable safety practices. The Chairman, SAS shall file a report of the audit findings with the Chairman, SRC and with a copy to the License Administrator and the Facility Supervisor.
The Chairman, SRC shall forward the report to the Director 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:
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 materf al, other than normal operating losses.
v 2.9.1.3 Incidents as specified in 10 CFR 20.403 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.
2.9.1.8 Unapproved storage or use of licensed material.
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2.9.2 Supervisor, Health and Safety The Supervisor, Health and Safety shall perform investigations and issue reports of the followiag:
2.9.2.1 Higher than expected personnel exposures.
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 Supervisor The Facility Supervisor shall perform investigations of the following:
2.9.3.1 Any violation of nuclear criticality safety criteria.
2.9.3.2 Any violation of Area Operating Procedures or RWP's.
O 2.10 RECORDS The following positions or organizations snall be responsible for maintaining the indicated records, for the period specified.
Records may be kept in original form, nicrofilm 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
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Personnel dosimetry records
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Results of Bioassays and Whole Body Counting
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Releases to the environment
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Radiation survey data 2 years Contanination survey data 2 years Radiation Work Permits (completed) 5 years Radiation detection instrument calibration 2 years Leak tests of sealed sources 2 years Employee training
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Employee retraining
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Airborne radioactivity sampling data
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NRC-4 forms
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NRC-5 forms
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2.10.2 Nuclear Safety Officer Nuclear criticality safety calculations 6 months after termination of the approved process.
2.10.3 License Administrator Safety Review Committee Minutes
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Safety Audit Subcomnittee Audit Reports 2 years Investigation reports of off-normal occurrences 2 years O
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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 3.2.1 Access Control 3-2 3.2 2 Ventilation Requirements.
3-2 3
3.2.3 Instrumentation.
3-4 3.2.4 Internal and External Exposure.
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i b) 3.0 RADIATION PROTECTION 3.1 SPECIAL ADMINISTRATIVE REQUIREMENTS 3.1.1 Radiation Work Permits (RWP) 3.1.1.1 RWP's shall be issued whenever the activity is not covered by an l
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.
3.1.1.2 RWP's shall be approved by the Work Area Supervisor, Employee's Supervisor, Health Physics Supervisor, and the Facility Supervisor.
3.1.1.3 The RWP form shall specify levels of personnel exposure above which a documented ALARA evaluation shall be required. RWP's that require a documented ALARA evaluation must, in addition to 3.1.1.2, be approved by the Director.
3.1.1.4 RWP's shall be approved at a meeting of all the signators of the l
form.
3.1.1.5 The RWP form shall provide space for entering the estimated A
exposures to the whole body, extremities, and for the job. These l V 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 sign or initial, attesting that the workers have been instructed in the requirements of the RWP.
3.1.2-ALARA Policy The management of the LRC is committed to a policy of maintaining exposures as low as is reasonably achievable.
3.1.2.1 Employees shall be introduced to this policy during their initial training and shall be reinforced during the annual retraining of Authorized Users.
3.1.2.2 The ALARA policy shall be applied to Area Operating Procedures and RWP's during the approval process.
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3.1.2.3 The ALARA policy shall be implemented through the Area Operating Procedures and Radiation Work Permits.
3.1.2.4 The ALARA policy shall be enforced by the Facility Supervisor and the Supervisor, Health and Safety in the exercise of their review and approval authority, their authority to terminate operations, and audits.
3.1.2.5 The SRC shall evaluate ALARA performance in exercising their review authority over procedures and proposed new projects and their review of the annual report from the Supervisor, Health and Safety.
3.2 TECHNICAL REQUIREMENTS 3.2.1 Access Control 3.2.1 1 High Radiation Areas - High radiation areas shall be posted and access controlled in accordance with 10 CFR 20.203(c).
In addition, entry into a high radiation area shall be controlled pursuant to a RWP.
3.2.1.2 Radiation Areas - Radiation areas shall be posted and access controlled in accordance with 10 CFR 20.203(b).
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3.2.1.3 Airborne Radioactivity Areas - Airborne radioactivity areas shall be posted and access controlled in accordance with 10 CFR 20.203(d). Entry into an airborne radioactivity area shall be controlled pursuant to a hWP.
3.2.1.4 Contamination Areas - Areas which are detemined by the Health and Safety Group to present a risk of spreading radioactive contanination into non-contaminated areas shall be clearly marked at each entrance.
Stop-off pads shall be provided.
Personnel survey instrumentation shall be provided at the step-off pad. The Health and Safety Group shall specify the protective clothing needed to enter such areas. Exiting such areas shall require personnel to remove their protective clothing and survey them-selves with the instrumentation provided.
3.2.2 Ventilation Requirements License No SNM 778 Docket No. 70 824 Date October,1985 Amendment No.
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3.2.9.1 Potentially contaminated exhaust air from hood, hot cells, and
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glove boxes shall be discharged through the fifty meter high stack, except as noted in 3.2.2.6.
3.2.2.2 The exhaust stack shall be sampled isokinetically.
3.2.2.3 The minimum air flow rate in the stack sampling system shall be 2 cfm.
3.2.2.4 The stack sampling and monitoring system shall operate continu-ously except for periods when repair or calibration is required.
3.2.2.5 The following table presents the release limits and action levels associated with the exhaust stack. The Health and Safety Group shall be responsible for responding to releases in excess of these action levels. An operation that results in action levels being exceeded for 4-consecutive time periods, shall be shutdown until the cause is corrected.
STACK RELEASE LIMITS AND ACTION LEVELS Release Product Release Limit Action Level
/~'T Beta Particulate 2 mci /yr 200 uCi/ week b
Alpha Particulate (long lived) 20 uCi/yr 1 uCi/2 weeks Kr-85 2500 C1/yr 70 Ci/ week H-3 130 C1/yr 3 C1/ week 1-131 6 nCi/yr or 303 uC1/ week 200 uCi/ week 3.2.2.6 Exhaust systems that cannot be practicably discharged through the 50-meter stack, and where there exists a reasonable probability that the discharges to the atmcaphere could exceed 10% of the applicable MPC for an unrestricted area, shall be monitored by taking a continuous particulate sample.
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3.2.2.7 Exhaust air from areas in which there is no airborne radioactive material may be exhausted directly to the roof either with or without continuous sampling, if approved by the Safety Review Committee.
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 l
approved by the Safety Review Committee.
1 3.2.2.9 All hoods used for the handling of licensed material shall exhaust through at least one prefilter and one HEPA filter, except for hoods that are specifically designed ard installed for use with perchloric acid.
3.2.2.10 Fume hoods utilized for the handling of unirradiated Pu shall be I
provided with two HEPA filters in series.
3.2.9,.11 Hot cells shall be provided with two stages of HEPA filters.
l 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, i
3.2.2.13 The acceptance criteria for the testing of final HEPA filters V)
(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 3.2.3.1 Portable instruments - The LRC maintains a relatively large and diverse inventory of portable survey instruments. These instruments vary in range, sensitivity, and manufacturer. The below listing is a representative sampling of the instruments on hand:
Sensitivity Operating Calibration Instrument Range Characteristics Freq.
Method Eberline R0-3C 0 - SK mR/hr 6.5 Kev - 1.2Mev 6 mo.
Source Eberline E-530
.01 - 200 mR/hr 23 Kev - 1.2Mev 6 Mo.
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Eberline PAC-4G 25-500K cpn Alpha 6 Mo.
Source Eberline R0-3A 0 - 50K mR/hr 6.5 Kev - 1.2Mev 6 fio.
Source Eberline RM-3A 0 - 50K cpm Alpha 6 Mo.
Source Eber11ne PNR-4 0 - SK mR/hr
.025Mev - 4Mev 6 Mo.
Source Eberline Teletector 0 - IK R/hr
.05Mev - 3Mev 6 Mo.
Source Eberline E-520 0 - 2K mR/hr 23 Kev - 1.2Mev 6 Mo.
Source Eberline PRM-7 0 - SK uR/hr 6 fio.
Source 3.2.3.2 Air Monitors 3.2.3.2.1 Nuclear Measurements Corp. (NMC) Model AM-2A - This instrument utilizes a gas flow proportional detector with a 1.0 ng/cm2 (q
thick end window. These instruments are operated as alpha or Cj beta-gamma monitors. They utilize a fixed filter with a nominal 3
air flow of 2.5 to 3 ft / min.
The alarm setting is set at less than 40 MPC hours above normal background including Radon and Thoron daughters.
3.2.3.2.2 Eberline Model AIM-3S - Toese monitors are used for alpha nonitoring only. They are typically located in areas where Pu or U is being processed.
They use a ZnS( Ag) scintillation detector with a fixed filter. The monitor air flow is nominally 20 ftJ/hr.
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.
Sanples 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.
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3.2.3.'3.2 ' Fixed sublers areElocated at work stations where the concentra-
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tion of strunrne radioactive material potentially exceeds 25% of the applicablo MPC.
These samplers which may be used to determin concentrations in the workers breathing zone shall be evaluated foi Yepresentativeness at lent once every 12 months or when a' licensed process or equipment r.hange is made that could disturb' the air. flow pattern.
3.2.3.4 Criticality Monitors 3.2.3.4.1 Nuclear Measurements Corp. (NMC) Nodel GA-2TO and GA-2A - These monitore are designed as criticality alarm systems. Detection is by 2 -Na! (Tl) detector operated in the constant current mode.
Response,is logrithmic and non-saturating.
Emergency power is provided. ' The nominal alarm setpoint is 20 mR/hr. Failure alarm function -is provided.
3}2.3.5 Counting Equipment 3.2.$.5.1 Sharp Low Beta' - Air sanples and effluent samples may be counted
~on this instrument. This instrument utilizes a 4.5-inch and a s
2.5-inch very thin end ' window proportional detector.
Back-
' grounds and counter response are tested weekly and the instrunent is calibrated annually.
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3.2.3.5.2 Beckman Wide Beta - Air samples and effluent samples may be C
counted on this instrument.
It utilizes two 2.5-inch very thin end window proportional detectors. Backgrounds and counter i
response are tested weekly and the system is calibrated annually. The manual detector is.used infrequently and it is t.s!,ted when used.
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' ' 3.2.4 Internal and External Exposure 3.2.4.1 Ventilation 3.2.4.1.1 The minimod air velocity across the opening of fune hoods that are used to handle licensed material shall be at least 100 fpm.
Hood face velor,ities shall be measured monthly.
Those hoods that do not meet the minimum requirement shall be blaced out of service.
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3.2.4.1.2 The maximum differential pressure across HEPA filters is limited to 4-inches of water.
HEPA filters shall be changed to prevent exceeding this limit.
3.2.4.1.3 The minimum differential pressure across the hot cell face shall be 0.25-inches of water.
An additional hot cell fan will be automatically or manually started when the differential pressure reaches 0.25-inches of water.
3.2.4.1.4 The minimun air flow rate through any opened door to a hot cell shall be 100 fpm. An additional hot cell fan will be automati-cally 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 Air Samples shall be taken in all areas where operations could cause personnel to be exposed to airborne radioactive materials.
3.2.4.2.2 Any area in which the concentration of airborne radioactive material potentially exceeds 25% of the applicable MPC shall be continuously monitored for as long as the process that caused the airborne activity is in progress, 3.2.4.2.3 Permanently mounted air sampling equipment used to determine n
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concentrations in the worker's breathing zone shall be evaluated
'v' for representativeness at least once every 12 months. This evaluation shall also be performed when changes are made in a process or equipment that could effect the sample's representativeness.
3.2.4.2.4 An evaluation of the representativeness of air sampling equip-ment shall be performed at the start up of a process that has been shutdown for more that 6 months.
3.2.4.2.5 Permanently mounted air samplers used to determine concentra-tions in a worker's breathing zone shall be changed as follows:
- 1. Process areas during normal operations - once/ shift.
License No SNM 778 Docket No.70-824 Date October,1985 O
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- 2. All areas during periods when normal operations are shut m
down - 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.
- 3. Sanples that are less than 10% of the MPC - weekly.
- 4. Sanples that exceed 25% of the MPC during one sample change period - once/ shift until the situation is corrected and at least one month's samples are less than 10% of the MPC.
3.2.4.2.6 Permanently mounted air samplers shall have a minimum sample flow rate of 8 LOM except when a personnel air sampler is used as a temporary replacement, in which case the minimum flow rate shall be 1.8 LPM.
3.2.4.3 Bioassay 3.2.4.3.1 Uranium Bioassay Program I
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- 1. The uranium bioassay program sampling frequency shall comply with Regulatory Guide 8.11, June,1974, except as specified in section 1.8 of this' application.
- 2. The action levels for the results of urinalysis sampling shall be 55 dpm/ sample for >4% enriched U-235 and 20 ug/
7x sample for <4% enriched U-235.
If these levels are exceeded,
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additional samples shall be taken.
If the additional samples exceed the limits, the worker shall be removed from further uranium work and the Supervisor, Health,and Safety shall estimate the worker's exposure utilizing additional urine samples, fecal samples, in vivo counting or other means at his disposal. Any worker whose estimated body burden is greater than 50% shall be invivo counted as soon as practicable.- Any workar whose estimated body burden is between 10 and 50% will be invivo counted during the next time that the body counting service is at the B & W site.
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s (g) 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 minimum frequency for urine sampling shall be six months.
The minimun frequency for in vivo counting shall be annual.
- 2. The following are the action criteria for W and Y compounds exposure for the plutonium bioassay program:
Action Analysis Level Action to be Taken Urinalysis < 0.2 dpn/L None 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, investigate the cause and correct.
- 4. If the exposure is confirmed
n by #1 investigate to determine how exposure was incurred and correct it.
If the exposure exceeds 50% of the maximum permissible annual dose, the worker shall be restricted from further exposure until the Supervisor, Health and Safety authorizes the lifting of this restriction.
Urinalysis > 4 dpn/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 />.
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- 3. Initiate an investigation.
- 4. The Supervisor, Health and Safety only, may lift the work restriction.
In vivo
< 1.6E-8 Ci None In vivo
> 1.6E-8 C1
- 1. Restrict the worker from further ex,posure.
- 2. Resample the individual wi',nin 10 working days.
- 3. Determine if area surveys support the analysis results.
- 4. If #3 is positive, investigate the cause and correct as needed.
- 5. If exposure is confirmed by #2, the Supervisor, Health and Safety shall determine the organ dose.
If the confirmed
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exposure exceeds 50% of the
\\- I maximum permissible annual dose, the worker shall be restricted from further exposures until the Supervisor, Health and Safety authorizes the lifting of this re-striction.
- 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.
License No SNM-778 Docket No.70-824 Date October, 1985 es 0
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V 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 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 ibnitoring 3.2.4.6.1 The Health and Safety Group shall perform smear surveys in the below listed areas at the indicated frequencies:
Action Level Area Frequency (dpm/100cm2) p'
'N. s
<---------------------------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 License No SNM-778 Docket No.70-824 Date October,1985 O
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<---------------------------BETA-------------------------->
Building A Labs.
Monthly 2000 Building B Counting Lab.
Monthly 2000 Scanning Electron Monthly 2000 Microscopy Lab.
Hot Cell Operations Araa Bimonthly 2000 Cask Handling Area Bimonthly 22000 Radiochemistry Lab.
Bimonthly 22000 Exit portals from Biweekly 2000 controlled areas 3.2.4.6.2 Large area smears are used to survey many square meters of surface area. To determine if these smears indicate that an action level has been exceeded, the assumed area covered shall not exceed 1-square meter.
3.2.4.7 Decontamination n
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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 within 24-hours of discovery.
If discovery is made just prior to the beginning of a holiday or weekend, the contanina-tion shall be marked and labeled, and decontamination shall commence during the first regular workday after discovery.
3.2.4.7.3 Fixed contanination 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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3.2.4.7.4 Fixed contamination that, in the opinion of the Supervisor, Health and Safety, may substantially contribute to workers' exposure shall be posted and removed as soon as practicable.
3.2.4.8 Emergency Evacuation 3.2.4.8.1 Emergency evacuation drills shall be conducted semiannually.
3.2.4.8.2 The emergency evacuation alarm will sound automatically when any tw6 criticality monitors reach their alarm set point.
3.2.4.8.3 An emergency evacuation may be announced by the Receptionist on the Public Address System.
3.2.4.8.4 All employees and the Emergency Control Organization shall respond to energency evacuations in accordance with the Energency Procedures.
3.2.4.9 Personnel Monitoring 3.2.4.9.1 All employees of the LRC shall be issued a TLD monitor. This monitor has a range of from 10 nRem to 10,000 Rem.
This dosimetry shall be attached to the employee's identification badge.
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3.2.4.9.2 Employees whose annual exposure, as projected by the Supervisor, Health and Safety, will exceed 100 milliren (Radiation Workers) shall be issued a film budge and two indirect reading pocket dosimeters or one self-reading dosimeter and one TLD. This dosimetry shall be worn when the individual is in a radiation, high radiation or airborne activity area.
3.2.4.9.3 Visitors shall wear one TLD which shall be changed weekly.
This monitor is inserted in the visitor identification badge and shall be worn at all times while on site.
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.
License No SNM-778 Docket No.70-824 Date October,1985 Amendment No.
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IO 3.2.4.9.5 B & W employees from facilities other than the LRC will not be g
issued LRC dosimetry if they wear the dosimetry from their own facility.
If they do not have their own dosimetry, they shall 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 dosimeter and a film badge.
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O' TABLE OF CONTENTS Section Page 4.0 NUCLEAR CRITICALITY SAFETY 4-1 4.1 S.ECIAL ADMINISTRATIVE REQUIREMENTS 4-1 4.2 TECHNICAL REQUIREMENTS 4-2 4.2.2 Building A 4-2 4.2.3 Building B 4-3 4.2.4 Building C 4-9 4.2,5 Outside Storage.
4-9 4.2.6 Dry Waste.
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4.0 NUCLEAR CRITICALITY SAFETY 4.1 SPECIAL ADMINISTRATIVE REQUIREMENTS 4.1.1 Double Contingency Policy - The Double Contingency Policy as de-fined in the American National Standard ANSI /ANS-8.1-1983 shall be followed in establishing the basis for nuclear criticality safety of all operations.
4.1.2 Structural Integrity - Where structural integrity is necessary to provide assurance for nuclear criticality safety, the design and construction of those structures will be evaluated with due regard to load capacity and foreseeable abnormal loads, accidents, and deterioration.
4.1.3 Nuclear Criticality Safety Evaluation - All moditications or additions or both to any operation, system or equipment must be approved by the Facility Supervisor.
It is the responsibility of the Facility Supervisor, in consultation with the Nuclear Safety Officer, to determine whether or not a nuclear criticality safety evaluation is required for the proposed modification or addition.
The Nuclear Safety Officer or a person designated by him shall provide any required evaluations, including calculational support.
Nuclear safety evaluations shall be reviewed by a second indi-(,)
vidual, either the Nuclear Safety Officer or by a person with the
'v' same minimum qualifications required for the Nuclear Safety Officer.
4.1.4 Posting - Each unit shall be posted with the limits of SNM permitted in the unit.
4.1.5 Labeling - Each container containing greater than 0.5 grams of SNM shall be labeled to show the amount of element, the percent enrichment, when applicable, and the amount of fissile isotope.
This condition does not apply to irradiated SNM.
4.1.6 Compliance - Compliance with the nuclear criticality safety requirements shall be in accordance with written area operating procedures, reviewed and approved by the Facility Supervisor, the Supervisor, Health and Safety, the Nuclear Safety Officer, and the SRC.
In addition, the Nuclear Safety Officer will perform a quarterly audit of compliance with nuclear criticality safety requirements and document same in writing to the Director.
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4.2 TECHNICAL REQUIREMENTS 4.2.1 Nuclear Isolation - When nuclear isolation is required (the potential neutronic interaction between units is negligible) the unit or units isolated shall be separated from all other SNM by one of the following or equivalent conditions:
1.
Twelve inches of water.
2.
Twelve inches of concrete with density of at least 140 lb/ft3 provided that the isolated unit or units cannot be representable as a slab which interacts with the 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 center-to-center.
U' 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 1 to 20 313 20 to 40 283 40 to 60 258 60 to 80 237 80 to 100 220 4.2.2.2.2 Mass Limits for Low Enriched Uranium - 850 grams of U-235 as contained in uranium enriched in the isotope U-235 to and including 4 wt%.
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Whenever uranium enriched to 4 wt% U-235 is being processed under the 850 gram limit, no unit shall be permitted to have fissile material at an enrichment greater than 4 wt% within that laboratory, room, or work area.
- 2. Whenever an 850 gram, enriched controlled unit is in use in the build 1.1g, the Facility Supervisor must approve all transfers involving materials with enrichments greater than 4 wt% within the building.
4.2.3 Building B 4.2.3.1 General - Building B shall be limited to 40 units, excluding the hot cells, underwater storage, and the examination of power reactor fuel assemblies. Each unit shall be separated from adjacent units by at least 8-inches edge-to-edge and 24-inches center-to-center.
4.2.3.2 Unit Limits. Each unit shall be limited to the values specified in Section 4.2.2.2 of this Part.
4.2.3.3 Hot Cell - The hot cells, except for examination of power reactor fuel assemblies, shall be liinited to the following units:
1.
Three units in hot cell no.1, separated by at least 12-inches edge-to-edge.
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v 2.
One unit in each of the other hot cells.
4.2.3.4 Underwater Storage (Transfer Canal & Storage Pool) - SNM in storage under water in the Transfer Canal & Storage Pool shall be in racks or containers limited to the values specified in 4.2.2.2, excluding power reactor fuel assemblies, and separated by 12-inches edge-to-edge.
4.2.3.5 Storage Tubes - SNM in storage tubes shall be limited to the values specified in 4.2.2.2 for each tube. Storage tubes shall be spaced a minimum of 17-inches center-to-center, are approximately 5-inches in diameter, and totally immersed in concrete.
4.2.3.6 Power Reactor Fuel Assemblies - Examination of unirradiated and irradiated power reactor fuel assemblies, including both non-destructive and destructive testing is carried out in Building B subject to existing nuclear criticality safety limits and controls except as modified by the following conditions.
License No SNM-778 Docket No.70-824 Date October,1985 Amendment No.
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4.2.3.6.1 Fuel assemblies to be studied are identified as:
1.
Each assembly shall be of the enriched uranium oxide PWR type with a 15 X 15, or 17 X 17 square pin lattice not greater than 8.6-inches on a side (further identified as a Babcock & Wilcox Mark B or Mark C canless assembly).
2.
The maximum initial enrichment in an unirradiated assembly shall not exceed 4.05 wt%.
3.
Damaged fuel assemblies may be examined in air. Fuel assemblies which have been damaged can be examined in water if they maintain their 8.6-inches on a side dimensions.
4.2.3.6.1.1 Other PWR or BWR fuel assemblies which do not meet the above may be studied, provided:
1.
The unirradiated, fully reflected fuel assembly (fueled with U02 only) with all control rods removed is shown by an appropriate nuclear safety evaluation to be subcritical by at least 5 % (K-eff <0.95).
2.
The fuel assembly is shown by an appropriate nuclear safety evaluation to be subcritical by at least 5 % (K-eff em
<0.95) under specific conditions of disassembly.
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3.
The nuclear safety evaluation called for in #1 and #2 above shall be made pursuant to Section 4.1.3 of this Part.
The review shall be for descriptive and conceptual accuracy of the assembly and conditions of disassembly, proper use of appropriately benchmarked calculative nethods, and reasonableness of results and absence of noncenservative assumptions in the evaluation.
4.2.3.6.1.2 BWP fuel assemblies may be received and studied provided:
1.
T'iey are evaluated pursuant to Section 4.2.3.6.1.1 of this Part, or 2.
The BWR fuel assemblies have a maximun initial unirradi-ated enrichment of 4.05 Nt% U-235 and have a cross sectional area not exceeding that of a 22.5 cm (8.85 in.)
diameter cylinder.
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4.2.3.6.2 Receipt and Storage 4.2.3.6.2.1 Unirradiated Fuel Assemblies - Unirradiated fuel assemblies will be received at a maximum of two at a time in a 0 ' ppis.g container licensed for two assemblies, or one assembly la a shipping container licensed for one assembly. Unirradiated fuel assemblies may be stored in air in the Crane & Cask Handling Area, the Assembly & Machine Shop Area, or the.
Development Test Area subject to the following conditions:
1.
Assemblies may be stored in their shipping container as received.
2.
Assemblies may be stored no less than 21-inches apart center-to-center.
3.
Assemblies may be stored under water in the hot cell pool, mockup pool, or development test area pool in racks con-structed to maintain a 1-foot minimum surface-to-surface separation between assemblies and any other SNM.
4.
No more than four unirradiated assemblies may be kept at the LRC at one time.
4.2.3.6.2.2 Irradiated Fuel Assemblies - Irradiated fuel assemblies shall
- g) be received one at a tiine in a licensed single assembly
'v' shipping container or two at a time in a shipping container licensed for two assemblies. Fuel assemblies that have been irradiated will be stored in the hot cell pool which is limited to the following conditions:
1.
A maximum of four assemblies or portions thereof may be in the pool at a time.
2.
The assemblies shall be stored in a storage rack which is so constructed as to maintain a 1-foot minimum surface-to-surface separation between the stored assemblies and any other fissile material which might be in the pool.
3.
Only one assembly may be in a designated work area of the pool at any one time. There shall be at least 1-foot minimum surface-to-surface separation britween the assembly in the work area and any other fissile raaterial.
License No SNM-778 Docket No.70-824 Date October,1985 0
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Fuel rods which have been removed from an assembly shall be stored in a storage rack providing space in each position for a naximun of 75 rods. All positions shall be spaced from any other fissile material by a minimum surface-to-surface separation of 1-foot.
5.
Partially dismantled assemblies will be stored in the assembly storage rack.
6.
Each position in the assembly storage rack and in the fuel rod storage rack must limit contained fuel to a square not to exceed the dimensions of a fresh fuel assembly or to a cross sectional area not exceeding that of a 22.5 cm (8.85 in.) diameter cylinder.
4.2.3.6.3 Work Area of Pool Under Hot Cell No.1 - The work area position under Cell No.1 is used to load and unload irradiated fuel assemblies into and out of shipping casks and to dismantle both irradiated and unirradiated fuel assemblies.
The following conditions govern operations in this work area:
1.
Only two assemblies at a time shall be permitted outside of their shipping container provided they are separated by a minimum surface-to-surface separation of 1-foot.
(V) 2.
An associated storage position shall be parmitted for fuel rods or components which have been removed from the assemblies.
3.
The assemblies and associated rod storage positicas shall be separated from each other and any other fissile material by a ninimum surface-to-surface separation of 1-foot.
4.
Fissile material and fuel rods or components in the associ-ated storage positions shall each be restricted to a square not exceeding the dimensions of a fresh fuel assembly or to a cross sectional area not exceeding that of a 22.5 cm (8.85 in.) diameter cylinder.
5.
Only one fuel rod at a time shall be removed from or inserted into the assembly or the rod storage position. A Maximum of 75 rods shall be permitted in the rod storage position.
License No SNM-778 Docket No.70-824 Date October,1985
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6.
A fuel assembly may be completely dismantled by withdrawing one fuel rod at a time from the assembly; during all stages of dismantlenent, the partially dismantled assembly shall be maintained within the confines of a square not exceeding the dimensions of a fresh fuel assembly or to a cross sectional area not exceeding that of a 22.5 cm (8.85 in.) diameter cylinder.
7.
An assembly and its associated rod storage position may be withdrawn from the pool into the cell. Free water drainage from both the assembly and rod storage position as well as 1-foot separation from other fissile materials and each other shall be assured.
4.2.3.6.4 Assembly and Machine Shop and Development Test Area -
The work areas on the first floor of Building B may be used to disassemble unirradiated fuel assemblies for testing. The following conditions govern operations in the work area:
1.
Only one assembly at a time shall be permitted to be dismantled.
2.
An associated storage position will be permitted for fuel rods which have been removed from the assembly, and shall be
.f 3 spaced and stored as stated in items 3 and 4 (4.2.3.6.4)
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below.
3.
The assembly and associated rod storage position shall be separated from each other and any other-fissile material by a minimum surface-to-surface separation of 21-inches.
4.
The associated rod storage position shall be no larger in any dimension than the fuel assembly.
There shall be one such storage position for each assembly to be dismantled.
Rods may be stored or handled in a slab up to 4-inches thick provided the slab is separated from other fissile material by a minimun of 12-feet.
5.
Only one fuel rod at a time may be removed from or inserted into the assembly or any rod storage position. Only one rod may be in transit to any one location at a time.
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The fuel assembly may be completely disassembled by with-
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drawing one fuel rod at a time from the assembly; during all stages of disassembly, the partially disassembled assembly shall be maintained within the confines of the assembly whether damaged or undamaged.
7.
Fuel rods may be removed one at a time from this area as required. These rods shall be subject to all fuel handling requirements pertinent to the area they are in.
8.
Assemblies may be handled and dismantled under water in these areas (mock-up pool and development test area pool) subject to the same requirements of the hot cell pool.
4.2.3.6.5 Hot Cell Operations - Fuel rods removed from irradiated assemblies may be examined including destructive examination in the hot cells. Operations in the hot cells shall be governed according to the following conditions:
1.
An assembly and its associated rod storage position may be withdrawn from the pool into Hot Cell No.1 pursuant to Item No. 7 of Section 4.2.3.6.3 of this Part.
2.
Two units in Hot Cell No.1 may have a total of 64 fuel rods each, stored, provided that rods shall be confined within a n
('~)
cross sectional area not exceeding that of a 22.5 cm (8.85 in.) diameter cylinder, drainage of any free water within the unit shall be assured and the units must be maintained 1-foot from each other and any other SNM in the cell.
3.
In addition to the two units of stored rods, another unit limited to the values in 4.2.2.2.1 may be present in Hot Cell No.1.
In this unit under mass control, rods may be destructively examined.
4.2.3.6.6 Fuel Rod Dismantlement - Fuel rods from unirradiated assemblies can be dismantled in any area where the license permits handling of unirradiated fuel. The following conditions must also be met in areas to dismantle fuel rods:
1.
The area shall be mass limited to 350 grams of U-235. This area must be separated from the assembly and slab storage area by minimum of 12-feet.
2.
Dismantlement must be completed under approved procedures.
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4.2.3.6.7 Shipment and Disposal - After examination, assemblies, partially dismantled assemblies, fuel rods, and scrap generated during destructive examination shall be disposed of according to the following conditions.
1.
Fuel rods, including fuel rod segments may be placed in any available hole in a fuel assembly, including the instrument and control rod guide tube positions, i.e., 225 and 285 fuel rods in Mark B and Mark C assemblies, respectively.
Fuel rod segments shall have their ends sealed, and shall be encapsulated in steel tubing with ends sealed, prior to insertion into an available hole in a fuel assembly.
2.
Unirradiated assemblies may be reassembled (one rod at a time) for later use.
3.
Assemblies, including partially dismantled assemblies, shall be loaded into shipping casks approved for such assemblies for shipment.
4.
Scrap, including rod segments, shall be disposed of according to present LRC procedures and limits.
4.2.4 Building C 7_
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4.2.4.1 General - Building C is limited to 90 units. Each unit shall be
~
separated from adjacent units by at least 8-inches edge-to edge and 36-inches center-to-center.
4.2.4.2 Unit Limits - Each unit shall be limited to the values specified in section 4.2.2.2 of this Part.
4.2.5 Outside Storage 4.2.5.1 General - Outside storage consists of underground storage, shipments, and the fenced storage area located adjacent to Building J.
4.2.5.2 Underground Storage - Radioactive materials stored in underground storage tubes shall be limited to one SNM unit per tube, with values as specified in section 4.2.2.2 of this Part. Tubes shall be spaced 20-inches center-to-center and are 5-inches in diameter, and totally immersed in concrete.
License No SNM 778 Docket No.70-824 Date October,1985 m
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b(m 4.2.5.3 Shipments - Each shipment of fissile material being stored out-side must conform with all license requirements for the type of shipping container. Additionally, each shipment must be nuclearly isolated from all other SNM.
4.2.6 Dry Waste - Dry waste is accumulated in 55-gallon drums, or other suitable containers, with a maximum of 45 grams of SNM per con-tainer.
These containers may be located throughout the laboratories as required to collect contaminated labcratory waste.
Filled containers are transferred, to the radioactive waste storage building after scanning. Dry waste containing 0.5 grams of SNM or less per container may be stored in a fenced, locked and paved outside storage area adjacent to Building J.
g License No SNM 778 Docket No.70-824 Date October, 1985 Q
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TABLE OF CONTENTS Section Page 5.0 ENVIRONMENTAL PROTECTION 5-1 5.1 EFFLUENT CONTROL SYSTEM 5-1 5.2 ENVIRONMENTAL MONITORING 5-3 License No SNM 778 Docket No.70-824 Date October,1985 O
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V 5.0 ENVIRONMENTAL PROTECTION 5.1 EFFLUENT CONTROL SYSTEM 5.1.1 Responsibility - The Supervisor, Health and Safety is responsible for the Effluent Control System.
5.1.2 Solid Waste - Solid radioactive waste, including solidified liquid wastes, shall be sent off site to a licensed disposal facility.
5.1.3 Liquid Waste - Low-level, liquid ractioactive waste is discharged from process areas to the Liquid Waste Disposal Facility. This facility is comprised of several tanks where the waste is accumu-lated for eventual transfer to the B & W Naval Nuclear Fuel Division (NNFD) for ultimate release to the Janes River.
5.1.3.1 Prior to release to NNFD the contents of a tank shall be mixed and sampl ed.
5.1.3.2 The contents of liquid waste tanks shall not be released to the NNFD unless the concentration of radioactivity is less than 25% of the MPC values of Table I, Col. 2, of 10 CFR 20, Appendix B.
The limiting values in water shall be determined in accordance with O
the note at the end of Appendix B, 10 CFR 20.
5.1.3.3 Process liquid wastes may be collected and stored indoors. These wastes may be solidified and handled as dry waste.
5.1.3.4 Storage tanks in the Liquid Waste Disposal Facility shall be inspected visually upon each tank voiding or annually, whichever is sooner, to assure that there is no unsafe accumulation of Special Nuclear Material.
Storage tanks that have not been used during a year will not be inspected.
5.1.3.5 Samples of liquid waste are grab sampled. A small partion of the sanple is pipetted into a planchet and brought to dryness. This planchet is counted on a low background counter, either low Beta or Wide Beta, and the waste activity concentration is calculated.
The samples are counted for gross alpha and beta. Waste tanks that may receive Sr-90 waste will have their samples analyzed for Sr-90.
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~x V) l 5.1.3.6 Waste tanks that indicate concentrations of activity greater than those specified in section 5.1.3.2 shall be appropriately diluted prior to release.
5.1.3.7 The NNFD must approve the release of liquid waste to their waste treatment facility prior to the release.
5.1.3.8 The 10,000 sq. ft. Storm Drain Collection Pond shall be grab sampled quarterly. The sample shall be analyzed for gross alpha and gross beta.
5.1.4 Gaseous Effluent - Discharge air from process areas is released to the general environment through the 50-meter high stack. The discharge rate of the stack is approximately 20,000 cubic feet per minute.
The annual discharge volume is 1.1E10 cubic feet.
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 appropriate decay period. The results shall be recorded and maintained on file.
- O 5.1.4.3 Sanpling - The stack shall be sampled isokinetically on a continu-V ous basis.
5.1.4.4 Monitoring - The stack sample shall be passed through a monitoring system that consists of the following:
1.
Particulate Monitor - The stack particulate monitor is a Nuclear Measurements Corporation (NMC) Model AM-22R with dual channel ratio detector. This monitor uses a fixed filter and a nominal sampling flow rate of 2 - 3 cubic feet per minute.
The detector is a thin window (1.0 mg/cm2) gas flow pro-portional 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 logrithmic ratio. This system effectively compensates for the presence of Radon and Thoron daughters and increases the sensitivity of the system. Alpha and Beta-gamma License No SNM-778 Docket No.70-824 Date October,1985 O
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sensitivities 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 concen-trations in unrestricted araas exceeding 10 times the appli-cable 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 shielded chamber with one or two GM tube detectors (30 mg/cm2 stainless steel). The Beta-gamma count rate is directly proportional to the stack concentration and systen sensitivity is approximately 3E-9 pCi/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.
gh 5.2.2 The following types of samples shall be taken at the below indicated frequencies:
1.
One continuous on site background air sample.
2.
Grab sanples from the James River below the point of discharge
- monthly.
3.
Continuous sampling of rain water.
4.
Grab sample of river silt and plant life - annually.
5.
A direct radiation survey shall be made of the water channel passing through the railroad right-of-way - annually 6.
A direct radiation survey shall be made at the east end of the canal - annually.
License No SNM 778 Docket No.70-824 Date October,1985
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7.
Accumulated water from the soil retention basin shall be sampled and if its activity exceeds 10% of the concentration specified in Appendix B, Table 2, column 2, of 10 CFR 20, the collected water shall be disposed of through the liquid waste disposal system - annually.
5.2.3 The evaluation of environmental sampling shall be performed by either the LRC or a qualified outside concern.
O License No SNM 778 Docket No. 70 824 Date October,1985 C)
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O TABLE OF CONTENTS Section Page 6.0 SPECIAL PROCESS COMMITMENTS 6-1 O
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6.0 SPECIAL PROCESS COMMITMENTS The Lynchburg Research Center is engaged in research and development and for this reason there are a large number of small special processes that are special-only because they are outside of the few repetitive operations. The LRC relies on established administrative controls to determine what proposals fall outside of the bounds of work that is authorized by the license, in which case amendments are applied for. Those proposals that are authorized by the license but are significantly different from previously reviewed proposals shall be brought before the Safety Review Committee for review and approval.
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O TABLE OF CONTENTS Section Page 7.0 DECOMMISSIONING PLAN 7-1 7.1 PLANNING CONSIDERATIONS 7-1 7.2 COST AND FINANCIAL ARRANGEMENTS 7-2 (DD License No SNM 778 Docket No.70-824 Date October,1985
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i 7.0 DECOMMISSIONING PLAN V
The Lynchburg Research Center is connitted to deconlissioning the facilities which have been used material et the end of their uss, for the use and storage of licensed ful, 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,1986.
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 raust meet the levels of contanination s~pecified-ia Table 1, Regulatory Guide 1.86, dated June, 1974.
In additice., a reasonable effort will be made to further reduce contanination levels to those'which are as low as reasce.aolf achievable.
7.1.3 No covering will be applied to remaining surfaces until it has been determined that contemination levels are below those of Table 1, C_)T Reguldtory Guide 1.86, dated June, 1974,.and unt.il it has been determined that a reasonable effort has been made to further reduce-s contamination below those specified above.
7.1.4 The radioactive contanination of interior surfaces of pipes, ductwork, and other conduits will,be determined by taking neasurements at all traps and othcr 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 contaninated in excess of levels specified in Table 1, Regula-tory Guide 1.86, dated June, 1974.
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7.1.5 A radiological survey plan shall be designed and implemented to assess the extent of decontamination necessary. The survey plan will include the taking of smear samples of roofs, ceilings, walls, floors, and equipment.
It shall also include the taking of core samples of concrete floors, walls, ceilings and pools to a depth of one-eighth of one inch. Core samples shall also be taken of soil beneath floors and pools and in the vicinity of underground drainlines. Records of the survey and sample results shall be prepared and maintained.
7.1.6 Equipment will be disposed of by burial or sale. The decontami-nation of equipment scheduled for burial will be sufficient to meet the requirements of transportation and those of the receiving facility. The decontamination of equipment scheduled for sale to a licensed facility will be sufficient to meet the requirements of transportation and those of the receiving facility. Equipment scheduled for receipt by a nonlicensee shall be decontaninated to levels specified in Table 1, Regulatory Guide 1.86, dated June, 1974.
7.1.7 At the completion of decommissioning activities, a report of the final survey shall be prepared and submitted to the NRC with a request that a confiratory inspection be made and that the facility be released for unrestricted use.
'd 7.2 COST AND FINANCIAL ARRANGEMENTS 7.2.1 Financial Arrangements - Financial arrangements to cover the cost of decommissioning are set forth in the letter dated March 13, 1978 to Mr. John B. Martin, Assistant Director, Fuel Cycle Safety and Licensing, USNRC, from Mr. George P. Zipf, Chairman and President, Babcock & Wilcox Company.
7.2.2 Cost Estinate Guidelines - The following are the general guidelines upon which the estimates are based.
7.2.2.1 No facility will be razed.
7.2.2.2 No "mothballing" of any facility or perpetual surveillance is contempl ated.
All facilitis will be released for unrestricted use upon approval by NRC.
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7.2.2.3
" Reasonable" efforts to decontaminate below levels spec'fied in Table 1, Regulatory Guide 1.86, dated June 1974, shall be required only to the extent that the benefits derived clearly justify the additional cost expenditures.
'7.2.2.4 There will be no requirement to prepare or furnish prior to, concurrent with, or subsequent to decommissioning, any environ-mental data or impact reports.
l,7.2.2.5 Physical security pursuant to 10 CFR 73 will not be required at the time the Decommissioning Plan is implemented.
7.2.2.6 All estimates are given in 1985 dollars.
No factor for escalation l
has been included.
l 7.2.3 Technical Guidelines - The following are the technical guidelines upon.which the estimated costs are based.
7.2.3.1 Surface removal will be required for decont$nination of the, interior surface of hot cells, transfer canal and storage pooM !
liquid waste retention basins SK-1, SK-2, 10K-1, 10K-2 and the Building A basin, primary equipment cell pump pit, and the Building B Containe.ent Pool.
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7.2.3.2 All plutonium contaminated glove boxes will be disposed of, at-burial facilities in Richland, Washington.
7.2.3.3 No decontamination of cold areas d 'l be required.
7.2.3.4 Removal of ventilation ductMt bec ad the first stage of HEPA filtration will not be reuji A.
7.2.3.5 Removal of piping in the liquid waste ' system will be required.
7.2.3.6 Disassembly and off site burial of liquid waste tanks 2K-1, 2K-2, 2K-3, 13K, 4K-1, 4K-2, 300-1 and 300-2 will be required.
7.2.3.7 Disassembly and off site burial of the waste storage tubes will be required.
l 7.2.3.8 l Health and Safety labor will be approximately 20 percent of' the l
(total labor cost.
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7.2.3.9 Rental of special shipping containers will be required.
7.2.4 Cost Estimate 7.2.4.1 Building A
$500,000 Building B
$8,500,000 Building C
$4,500,000 Building J
$500,000 Liquid Waste Disposal Facility
$1,000,000 TOTAL
$15,000,000 O
License No SNM-778 Docket No.70-824 Date October,1985 O
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O TABLE OF CONTENTS Section Page 8.0 RADIOLOGICAL CONTINGENCY PLAN 8-1 B
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8.0 RADIOLOGICAL CONTINGENCY PLAN Reference Amendment No. 2, License SNN-778, Docket 70-824.
E License No SNM 778 Docket No.70-824 Date0ctober, 1985 0
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TABLE OF CONTENTS Section Page 9.0 OVERVIEW 0F OPERATION 9-1 9.1 CORPORATE INFORMATION 9-1 9.2 FINANCIAL QUALIFICATIONS 9-3 9.3
SUMMARY
OF OPERATING OBJECTIVE AND PROCESS 9-3 9.4 SITE DESCRIPTION.
9-3 9.5 LOCATION 0F SITE BUILDINGS 9-4 9.6 LICENSE HISTORY 9-4 9.7 CHANGES IN PROCEDURES, FACILITIES, AND EQUIPMENT 9-5 O
List of Figures b
Figure Page 9-1 LRC LOCATION IN VIRGINIA 9-7 9-2 FIVE MILE RADIUS OF LRC 9-8 9-3 LRC BUILDING LAYOUT.
9-9 9-4 FACILITY WORK ORDER FORM 9-10 License No SNM 778 Docket No.70-824 Date October, 1985 0
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PART II SAFETY DEMONSTRATION 9.0 OVERVIEW 0F OPERATION 9.1 CORPORATE INFORMATION
9.1.1 Licensee
McDernott International, Inc.
Babcock & Wilcox Research & Development Division Lynchburg Research Center
9.1.5 Address
Babcock & Wilcox Research & Development Division Lynchburg Research Center P. O. Box 11165 IV)
Lynchburg, Virginia 24506-1165 9.1.3 Principal Offices:
McDermott International, Inc.
Babcock & Wilcox 1010 Common Street P. O. Box 60035 New Orleans, Louisiana 70160 9.1.4 Principal Officers:
J. E. Cunningham Chairman of the Board &
Chief Executive Officer 1010 Common Street P. O. Box 60035 New Orleans, Louisiana 70160 U. S. Citizen License No SNM 778 Docket No.
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Robert E. Howson President & Chief Operating Officer 1010 Common Street P. O. Box 60035 New Orleans, Louisiana 70160 U. S. Citizen John A. Lynott Executive Vice President Chief Financial Officer 1010 Common Street P. O. Box 60035 New Orleans, Louisiana 70160 U. S. Citizen Walter M. Vannoy President and Chief Operating Officer Babcock & Wilcox 1010 Common Street P. O. Box 60035 New Orleans, Louisiana 70160 U. S. Citizen 9.1.5 State of Incorporation:
U Delaware 9.1.6 Alien or Foreign Control:
Babcock & Wilcox is incorporated under the laws of the State of Delaware.
In 1978 McDermott Inc., a Delaware corporation, acquired Babcock & Wilcox.
In 1983 McDernott International, Inc., incorpo-rated under the laws of the Republic of Panana, became the parent company of the McDermott group of companies. This reorganization was reviewed by the NRC prior to its implementation.
In a letter dated December 17, 1982, to Mr. J. H. MacMillan, Willian Dirks summarized the Commission's finding that the change was not inimical to the common defense and security or the health and safety of the public. Based on this conclusion, the change was approved under Section 104(b) of the Atomic Energy Act as it related to the operation of a critical experiment reactor owned by Babcock & Wilcox at the Lynchburg Research Center.
Such approval was not required under the Act for material licensees. The organization of McDermott International, Inc. has not changed since that action.
October, 1985 License No SNM-778 Docket No.70-824 Date 0
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(3 C) 9.2 FINANCIAL QUALIFICATIONS 9.2.1 The financial qualifications of the Corporation to continue operations at the Lynchburg Research Center and to perform the necessary decommissioning at the end of plant life is demon-strated in the latest (1984) copy of the Corporation's Annual Report, which is enclosed with this application.
9.3
SUMMARY
OF OPERATING OBJECTIVE AND PROCESS 9.3.1 Research and development activities, utilizing licensed material, are conducted at the Lynchburg Research Center in support of the operating divisions of Babcock & Wilcox and for other companies and government organizations. The broad range of projects that have been conducted pursuant to the license cannot be described in terms of through-put or any single process. Radioactive materials are handled and stored, principally in Building B.
That building houses the Hot Cells, Radiochemistry Laboratory, Scanning Electron Microscopy Laboratory, Metallurgy Laboratories, Analytical Chemistry Laboratory, Fatigue and Fracture Laboratory, Failure Analysis Laboratory, Crane and Cask Handling Area, a Hot Machine Shop, the Counting Room, and a Health Physics Laboratory, n
Licensed material in the form of liquid waste is collected in tanks
(
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that are located in the Liquid Waste Disposal Facility.
Solid radioactive waste is stored in Building J, the Annex to Building J, and the storage area adjacent to Building J.
9.4 SITE DESCRIPTION The Lynchburg Research Center (LRC) is located on the James River about four miles east of Lynchburg, Virginia. The site, which com-prises 525 acres, lies within Campbell County and borders on kiherst County. The LRC occupies about 13.6 acres of the site. The location of the site within the Commonwealth of Virginia is shown in Figure 9-1.
The irregularly shaped property is bounded on three sides by a large loop of the James River and on the fourth side by State Route 726, which closely follows the base of Mount Athos.
This mountain rises License No SNM-778 Docket No.70-824 Date October,1985 0
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rapidly from about 500 feet MSL to 900 feet MSL, making it the dominant feature of the surrounding landscape.
The Babcock & Wilcox property consists of large sections of relatively flat floodplain along the James River lying at about 470 feet MSL.
The interior of the property is largely composed of rolling hills, one of which rises to almost 700 feet MSL. The property boundry and topography within about two miles of the LRC are shown in Figure 9-2.
The land in the immediate vicinity of the plant is sparsely inha-bited.
The severe topography makes it vnsuitable for commercial farming.
The Lynchburg Foundry, a producer of light metal castings, occupies a parcel of land which abuts the south boundary of the Babcock & Wilcox property. The Foundry is approximately.5 miles from the LRC.
The site is serviced by a spur of the Chessie System Railroad which runs through the Babcock & Wilcox property. The property is also conveniently located for truck and automobile access. About three
, miles from the LRC, State Route 725 connects with U.S. Highway 460, a najor link between Roanoke and Richmond. The LRC is located about 100 feet above the James River and for that reason no dams on the river would threaten the LRC should they fail.
9.5 LOCATION OF SITE BUILDINGS
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9.5.1 Figure 9-3 shows the layout of buildings at the LRC. All buildings are of masonary construction.
9.6 LICENSE HISTORY 9.6.1 License SNM-778 was issued on September 16, 1966.
Since that time the following renewals and amendments have been approved:
February 15, 1974 First renewal July 21, 1980 Second renewal August 28, 1981 Amendment No. 1, approved a change in the organization.
February 25, 1982 Amendment No. 2, approved the Radiological Contingency Plan License No SNM-778 Docket No.70-824 Date October,1985 f)
Amendment No.
Revision No.
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Babcock &Wilcox a McDermott company
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May 11, 1982 Amendment No. 3, approved the reduction in the
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possession limit for unirradiated Pu to 0.9 kilograns May 14, 1982 Amendment No. 4, added four license conditions requiring surveys of B & W and railroad property and established a restricted zone.
August 11, 1983 Amendment No. 5, approved the reduction of the possession limits of SNM to below ore effective kilogram.
February 28, 1984 Amendment No. 6, approved a revision to the Radiological Contingency Plan.
November 6, 1984 Amendment No. 7, approved a revision to the definition of the Restricted Area.
July 25, 1985 Amendment No. 8, approved the extension of the expiration date of the license to January 1, 1986.
October 15, 1985 Amendment No. 9, approved a change in the organization.
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9.7 CHANGES IN PROCEDURES, FACILITIES, AND EQUIPMENT 9.7.1 Changes in Procedures - Changes, revisions, or additions to Area Operating P'rocedures are submitted to the Facility Supervisor.
The Facility Supervisor is responsible for assuring that changes are reviewed as required on the procedure form. Area Operating Procedures (A0P) must be reviewed and approved by the Nuclear Safety Officer, Supervisor, Health and Safety, Facility Supervisor and the Safety Review Committee.
Changes and revisions to A0P's may be implemented after the change has been approved by the Nuclear Safety Officer, Supervisor, Health and Safety, and the Facility Supervisor. The Facility Supervisor is responsible for assuring that the change is placed on the agenda of the Safety Review Committee at their next regularly scheduled meeting.
9.7.2 Changes in Facilities - Changes and modifications to buildings, exhaust ventilation systems, emergency electrical systems, and License No SNM 778 Docket No.
70 824 Date October,1985 0
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other facilities and services that may affect the safe handling of licensed material are requested on the Facilities Work Order Form (Figure 9-4). These forms are first subnitted to the Plant Engi-neering Supervisor.
He shall determine if the requested change involves a " Facility Change."
If a facility change is involved, the work order is forwarded to the Facility Supervisor.
It is the Facility Supervisor's responsibility to assure that all safety and licensing considerations have been addressed.
He shall assure that the change is approved by the Industrial Safety Officer and the Supervisor, Health and Safety. He shall consult with the License Administrator to determine if the change requires an amendment to the license.
Completed forms are kept on file by the Plant Engineering Super-visor and are audited nonthly by the Health and Safety Group.
9.7.3 Changes in Equipment - Changes or modifications in equipment which is used to handle licensed material are requested on the Facility Work Order Form. This form is submitted to the Plant Engineering Supervisor who determines that the change or modification involves a piece of equipment involved in licensed work.
If he determines that the change does involve such equipment, he forwards the form to the Facility Supervisor. The Facility Supervisor exercises the same responsibilities as those described in Section 9.7.2, above.
v Licensa No SNM 778 Docket No.70-824 Date October,1985 O
Amendment No.
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Babcock &Wilcox a McDermott company
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FIGURE 9-4 LRC139 FACIUTIES WORK ORDER FORM M
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Revision No.
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Babcock &Wilcox a McDermott company
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TABLE OF CONTENTS Section Page 10.0 FACILITY DESCRIPTION 10-1 10.1 PLANT LAYOUT.
10-1 10.2 UTILITIES INCLUDING EMERGENCY POWER 10-1 10.2.1 Potable Water 10-1 10.2.2 Process Water 10-1 10.2.3 Gas.
10 10.2.4 Fuel Oil
- -1 10.2.5 Electricity 10-1 10.3 HEATING, VENTILATION, AND AIR CONDITIONING.
10-2 10.3.1 Heating 10-2 10.3.2 Ventilation 10-2 10.4 WASTE HANDLING 10-4 10.4.1 Liquid Wastes 10-4 10.4.2 Solid Wastes.
10-5 10.5 FIRE PROTECTION.
10-6 10.5.1 Codes and Standards 10-6 10.5.2 Insurance Inspection Reports 10-8 10.5.3 Fire Protection Equipment 10-9 10.5.4 Combustible Waste Storage 10-9 License No SNM 778 Docket No. 70 824 Date October,1985 0
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Babcock &Wilcox a McDermott company
O List of Figures Figure Page 10-1 BUILDING A 10-10 10-2 BUILDING B 10-11 10-3 BUILDING C 10-12 10-4 BUILDING B VENTILATION SYSTEM.
10-13 10-5 LIQUID WASTE DISPOSAL SYSTEM 10-14
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i License No SNM 778 Docket No.70-824 Date October,1985
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10.0 FACILITY DESCRIPTION v
10.1 PLANT LAYOUT Figures 10-1 through 10-3 show the layout of the LRC buildings.
10.2 UTILITIES INCLUDING EMERGENCY POWER 10.2.1 Potable Water Potable water is provided to the LRC by the NNFD.
It is pumped from wells.
It is stored and treated at the NNFD and is gravity fed to the LRC.
10.2.2 Process Water Process water is provided to the LRC by the NNFD. The source of process water is the James River.
It is pumped from the river, filttred by the NNFD and is gravity fed to the LRC.
There is a storage capacity of 6,000,000 gallons on site.
Process water is also used for fire fighting.
10.2.3 Gas
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Natural gas is supplied at the LRC via pipeline which enters the A_
B & W property on the western side of the site. Natural gas is s
used for space heating, fuel for emergency engines, and laboratory uses.
This system is provided with a backup source of propane gas which is stored on site.
10.2.4 Fuel Oil Fuel oil is available for space heating to provide a backup source in the event of curtailed availability of natural gas. Fuel oil is purchased locally and stored on the LRC site. There is a storage capacity of approximately 24,000 gallons.
10.2.5 Electricity Electricity is furnished to the LRC frcm a substation located on the west side of the site.
This source provides the normal source of power for the stack fans, hot cell fans, criticality nonitors, emergency evacuation alarm, and lighting. When nornal site power is lost, energency sources are provided for these loads in the following nanner.
License No SNM 778 Docket No.70-824 Date0ctober,1985 p
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Babcock &Wilcox a NkDemon compaiy
10.2.5.1 The Onan emergency generator will start automatically on loss of site power and provides emergency power to one Hot Cell exhaust fan, the stack monitoring system, and emergency lighting. This engine / generator is fueled by natural gas and is provided with a source of propane gas as a backup source of fuel.
The Onan system is load tested weekly.
10.2.5.2 An Invertastat provides an uninterruptable source of electrical power for the criticality monitors, emergency evacuation alarm, and the alarm panel in Building B.
The invertastat operates, upon loss of normal site power, from a battery, with sufficient capacity to provide its loads with power for up to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
The invertastat and battery are load tested weekly.
10.2.5.3 The stack fans are provided with backup engines which start automatically upon loss of normal site power. One fan supplies off gas for Building B and the other provides off gas for i
Building C.
They are coupled to the fan shafts by a centrifugal clutch, such that when the engine speed equals the fan speed, the clutch engages. Both engines are fueled with natural gas and this is backed up by a source of propane gas. Both emergency fan i
engines are load tested weekly.
10.3 HEATING, VENTILATION, AND AIR CONDITIONING 10.3.1 Heating - Space heating in Buildings A, B, and C is provided by gas fired boilers.
The room air in Buildings A and B is partially recirculated. Room air in Building C is recirculated.
10.3.2 Ventilation 10.3.2.1 Building A - Building A.is presently undergoing decommissioning and very little licensed material remains in the building. The ventilation consists of the normal room air heating and air conditioning. No special ventilation is required and none is provided.
i License No SNM 778 Docket No. 70 824 Date October,1985 O
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10.3.2.2 Building B - The overall ventilation system for Building B (Figure L
10-4) has been designed so that the flow of air is toward the areas of highest potential airborne contamination; the lowest pressure being within the hot cell. All processing of licensed material in a form that could result in harmful airborne contamination or could generally contaminate personnel, equipment, or buildings, is carried out in the hot cells, glove boxes, and fume hoods. Fume hoods and glove boxes for use with licensed material are provided with a HEPA filter, except for those specified for use with perchloric acid, in which case they exhaust directly to the atmosphere.
Fume hoods and glove boxes used in the handling of unirradiated Pu nust exhaust through two HEPA filters connected in series.
Final HEPA filters are 00P tested annually or after the final filter is changed.
Final HEPA filters are provided with differential pressure indication and filters are changed when this differential pressure reaches 4-inches of water.
Special ventilation systems in Building B are described below:
10.3,2.2.1 Hot Cells - The hot cells are provided with off-gas that is passed through two stages of HEPA filters in series. The ducting between the hot cells and the HEPA filters is con-structed of steel.
There is a fire damper positioned upstream c,f the filters that is operated by a fuseable link that will close the damper in the event of fire and protect the filter.
The cells are provided with two off-gas fans that operate in G)
"and/or" medes. This provision permits increased off-gas when
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the cells are opened or maintenance of the off-gas system when one fan requires maintenance or fails.
The off-gas exhausts to the intake of the Building B stack fan. The off-gas system provides sufficient capacity to maintain an air flow of 100 linear feet per minute through an opening. The minimum differential pressure across the hot cell face is 0.25 inches of water.
The system was designed so that either the roof slab or personnel door can be opened and a differential pressure of 0.6 inch of water would be maintained across the cell face. The opening of either of these major openings is' controlled administratively so that both are not opened at the same time.
The volume of air removed is sufficient to remove the heat generated by lighting plus 7.5 kW from other sources and to maintain ambient' temperatures below 120 degrees F during normal operations.
License No SNM 778 Docket No.70-824 Date October,1985 0
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10.3.2.2.2 Radiochemistry Laboratory - The Radiochemistry Laboratory off-gas is provided by fume hoods. The hoods are generally provided with a single stage of HEPA filters.
Hoods that are used to handle unirradiated Pu will be provided with two stages of HEPA filters in series. The minimum air flow rate through the hood opening is maintained at 100 linear feet per minute or greater.
Each final filter is provided with differential pressure indication. The filter will be changed when the differential pressure reaches 4 inches of water. The Radiochemistry Laboratory is provided with heating or air conditioning from the Building B room air handling system.
10.4 WASTE HANDLING 10.4.1 Liquid Wastes 10.4.1.1 Liquid wastes that are potentially contaminated are piped to the Liquid Waste Disposal Facility (LWDF). A piping diagram of Building B is presented in Figure 10-5.
Wastes generated in the North-east end of Building B, which includes the Hot Cell Area, Radiochemistry Laboratory, Failure Analysis Laboratory, the Primary Equipment Cell (PEC), and the Containment, collect in a large tank located in the PEC and are pumped to the LWDF. The remainder of the building is drained by gravity to the LWDF.
The LWDF is a below grade tank farm. A schematic diagram of the tank arrangements, piping and pumps comprising the LWDF is shown l
in Figure 10-6.
Potentially contaminated waste water is piped from material handling areas to specific tanks so that the type of activity in any given tank can be anticipated. Each tank is provided with piping for thorough mixing by air sparge. The drain lines from each tank are located in the bottm of the tank to permit complete emptying. Exceptions to this are the two 4000 gallon waste tanks which have floating drains. The rectangular tanks shown in Figure 10-6 are constructed of concrete with concrete tops. The interiors are treated with a water-proofing material. Access to these tanks is gained through manholes. The tanks indicated by circles are constructed of steel. The 300 gallon tanks are stainless steel and are not provided with inspection ports. The 2000-gallon tanks are constructed of carbon steel, treated with an epoxy lining and are equipped with manholes for inspection.
10.4.1.2 Liquids with high concentrations of radioactive materials are solidified and disposed of as solid wastes.
License No SNM 778 Docket No. 70 824 Date October, 1985 O
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(b' 10.4.2 Solid Wastes 10.4.2.1 Solid wastes are generated in Building A and C as a result of the ducommissioning efforts in progress in those buildings and in Building B as a result of normal operations associated with the Hot Cells, Radiochemistry Laboratory, Failure Analysis Labora-tory, Fatigue and Fracture Laboratory, and support facilities.
S.olid waste that is generated in the Cask Hanoling Area ano the Radiochemistry Laboratory is compacted. These wastes are generally low level byproduct materials. High level wastes generated in the Hot Cells are not compacted.
10.4.2.2 Solid radioactive waste is stored, awaiting shipment for offsite disposal, in Building J, the Building J Annex, the high level waste storage tubes, and the outside storage area. Waste is stored in closed containers suitable for offsite shipment.
In Building J these containers may be any approved by 00T or NRC.
Containers that are stored outdoors must be constructed of metal.
Any container may be stored autdoors for short periods of time incidental to transportation. Waste stored in the high level storage tubes must be packaged in containers used in the renoval of waste from the hot cells.
10.4.2.3 Building J provides 1400 square feet of storage space.
The building is equipped with a smoke detector, air sampler, and a p
criticality monitor. The interior of the building is divided V
into three areas that are partitioned by concrete block walls.
This permits storage of high, intermediate, and low level waste in the same facility in a manner that results in minimum exposure to personnel maneuvering waste within the building. The maximum quantity of SNM permitted in Building J is 45 grams per container.
10.4.2.4 The Building J Annex is constructed of unmortared concrete block.
Three walls consist of four courses of block which provide about a four-foot thick shield wall.
The wall that is adjacent to Building J consists of three courses of block.
The Annex is provided with exhaust ventilation through ducting which connects Building J with the Annex thereby permitting the smoke detector and air sampler in Building J to serve both. The Annex is provided with a metal roof which is hinged to Building J, capable of being locked and provided with side panels which permit the License No SNM 778 Docket No.70-824 Date October,1085 0
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Babcock &Wilcox a McDermott company
i pd root.to fit flush with the top of the block walls. Containers are kaded into the Annex from the top. A curbing will be placed on tne. approach side of the addition to prevent a loading vehicle from accidently contacting the wall. Two individuals are involved in loading containers into this facility to prevent a container from striking the walls. This facility provides storage of waste that is contaminated with irradiated fuel and is being stored on site until it is accepted by the DOE under the Nuclear Waste Policy Act of 1982. The maximun quantity of SNM per container shall be limited to 45 grams.
10.4.2.5 The Outside Waste Storage Area is located adjacent to Building J.
This area is fenced, locked and paved. Waste stored in this area is limited to that contained in closed metal containers.
Each container is limited to not more than a Type A quantity (10 CFR i
71.4) or 0.5 grams of fissile material or both.
Pu shall not be stored in this area.
10.4.2.6 The High Level Waste Storage Tubes are located adjacent to the j
south side of the Liquid Waste Disposal Facility. These tubes are constructed of two sections of iron pipe, immersed in concrete, and below ground level. The upper section of pipe t
(approximately 42-inches long) is 6-inches in diameter. The lower section (approximately 80-inches long) is welded to the
,q upper section and is 5-inches in diameter.
Each tube is fitted Q
with a concrete-filled iron plug. These tubes are lock'ed and under the direct control of the Health and Safety Group. Waste stored in these tubes is limited to that which is produced in the Hot Cells and must be in closed metal containers. The quantity of fissile material permitted in each tube is limited to one j
unit.
1 i
10.5 FIRE PROTECTION 10.5.1 Codes and Standards - The development and building construction I
program of the Lynchburg Research Center complex has taken place over the period 1955 to the present.
For the three main buildings i
i under consideration in this renewal request, the design and con-struction efforts took place from 1955 to 1969.
There have been a number of alterations and use changes over the past ten years, but generally these changes have not significantly altered the structural characteristics of the buildings, i
License No SNM 778 Docket No.70-824 Date October,1985 O
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,GV All three buildings were built as staged or "added-on" phased construction.
Building A was built in four distinct phases, Building B was built in two stages, and Building C was constructed in five phases.
Building A was designed in-house by B&W engi-neering personnel. Building B was designed by Wiley & Wilson, a Lynchburg consulting engineering firm. Building C was also primarily designed by Wiley & Wilson, with some design by B&W.
The physical layout of all three buildings is highly functional, i.e., based on the specialized requirements of research work related to the nuclear industry.
For the most part, the building structure envelopes are quite conventional in nature, both from a design and construction materials standpoint. With the exception of highly specialized portions of these buildings, such as the hot cells, engineering design of the buildings would be considered as state-of-the-art for light industrial / heavy commercial class buildings (for each of the design and construction tme periods involved).
The overall quality of the building construction is well above average. Aside from some roof leakage problens and ninor settlement cracking in some of the rtasonry construction, the per-formance of the building structures and envelopes has been good.
There have been no repairs related to significant structural o
defects in any of the three buildings. As would be anticipated for
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a complex of this type and importance, maintenance of the buildings has been excellent and contributes to the overall good condition of auch a facility.
Durir.w the period of design and construction for Buildings A, B, and C, it should be noted that there was very little in the way of code requirements or guidance for construction of such a facility.
Virginia did not adopt a state-wide building code until September 1, 1973.
Up until that time, various localities in the state had adopted their own local building codes; the Southern Building Code being the one generally used. Many counties, however, had no code at all; Campbell County, in which the LRC complex is located, had no building code during this time period. The only state-wide code directly applicable to building construction prior to 1973, was the Virginia Fire Safety Regulations, enacted in 1949.
License No SNM 778 Docket No. 70 824 Date October, 1985 p) 0 0
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The lack of a state-wide building code should not be taken as implication that the design and construction of the LRC facility was accomplished on an inferior basis. On the contrary, where good accepted engineering and construction practices, coupled with stringent requirements from insuring companies such as Factory Mutual, are used as the main criteria for such facilities, the resulting structures usually far exceed the minimun requirements of various building codes. Such is typically the case for all three buildings under consideration, when examined from the load capacity standpoint as specified in the present Virginia building code, B0CA (Building Officials and Code Administrators) 1978, Seventh Edition.
Conventional construction materials are used throughout all three buildings.
Structural steel yield strength varies from 33,000 PSI
( ASTM A7 steel) for the 1955 construction to 36,000 PSI ( ASTM A36 steel) for the 1969 construction. Concrete strengths vary fron 3000 PSI for conventional cast-in-place concrete construction to 5000 PSI for the precast prestressed concrete elements found in Building B.
Concrete reinforcing steel is typically ASTM A615, Grade 40 (40,000 PSI yield strength). Working stress design was used as the basis for concrete and steel design for all structures at LRC.
Applicable design criteria used for the facility includes the standards of the American Concrete Institute ( ACI), the Pre-stressed Concrete Institute (PCI), and the American Institute of n
Steel Construction (AISC). These various standards serve as both
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design and code basis for the respective types of construction, i
s both at the time of original design as well as the present.
10.5.2 Insurance Inspection Reports - The LRC is inspected twice annually by the Arkwright-Boston Insurance Company on behalf of the Mutual Atomic Energy Reinsurance Pool (MAERP). The inspection reports list the following items in each report; housekeeping, mainte-
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nance A renair, Supervision fire equipment, watchmen, radioisotope handling, dreas sprinklered, water supply, all valves found open, criticality control, and until the decommissioning of the last reactor, nuclear reactor operation.
These reports have con-sistently found that the LRC meets the requirements in each category for a " satisfactory" rating. On a few occasions there have been reconnendations that the LRC add fire protection equipment when the use of an area has been changed.
Each such recommendation has been addressed by the LRC in a manner that has been found acceptable to the inspectors upon their reinspection.
The reports on which the above statement is based are dated from 1977 through 1985.
License No SNM 778 Docket No.70-824 Date October, 1985 0
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10.5.3 Fire protection equipment is installed at the LRC in response to reconnendations made by the Industrial Safety Officer, the Corporate Fire Protection Engineer, or the insurance underwriters.
Installed systems are' approved and inspected by Factory Mutual Engineering Association. Routine inspection and maintenance is described below:
EQUIPMENT MAINTENANCE RESPONSIBILITY REFERENCE Portable fire Insp./ test Industrial Safety NFPA 10 extinguishers FM 4-5 Fire hoses Insp./ test Industrial Safety NFPA 10 Sprinklers Test Diant Engineering NFPA 13 FM 4-5 Fire suppres.
Inspection Plant Engineering NFPA 12-A systems (Halon)
FM 4-8N Mfg.
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Inspection Health and Safety Mfg.
10.5.4 Combustible Waste Storage - Combustibles are not routinely stored at the LRC except; when work requiring such materials is in progress, in containers for shipping and receiving, or in sprinklered areas. Combustible wastes are discarded in metal containers and disposed of by an off site disposal firm. Contani-nated combustible waste is discarded in metal containers and shipped to an off-site licensed disposal facility.
License No SNM 778 Docket No. 70 824 Date October,1985 m
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Amendment No.
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O TABLE OF CONTENTS Section Page 11.0 ORGANIZATION AND PERSONNEL 11-1 11.1 LRC LINE ORGANIZATION.
11-1 11.2 LRC SAFETY ORGANIZATION 11-1 11.2.1 Manager, Safety and Licensing 11-1 11.2.2 Supervisor, Health and Safety 11-1 11.2.3 Health Physics Engineer 11-2 11.2.4 Industrial Safety Officer 11-2 11.2.5 Accountability Specialist 11-3 11.2.6 License Administrator.
11-3 11.2.7 Nuclear Safety Officer 11-4
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11.2.8 Facility Supervisor 11-4 11.2.9 Safety Review Committee 11-5 11.3 EDUCATION AND EXPERIENCE OF KEY PERSONNEL 11-6 11.3.1 Safety and Licensing Manager 11-6 11.3.2 Health and Safety Supervisor 11-7 11.3.3 Health Physics Engineer 11-9 11.3.4 In'dustrial Safety Officer 11-10 11.3.5 Accountability Specialist 11-12 11.3.6 License Administrator.
11-13 Facility Supervisor 11-13 License No SNM-778 Docket No.70-824 Date October, 1985 O
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TABLE OF CONTENTS (Cont'd)
Section Page 11.3.7 Nuclear Safety Of ficer 11-14 11.4 OPERATING PROCEDURES 11-16 11.4.1 Area Operating Procedures (A0P) 11-16 11.4.2 Availability.
11-17 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 y
11.6 FACILITY CHANGE.
11-21 List of Figures Figure Page 11-1 LRC LINE ORGANIZATION.
11-22 11-2 LRC SAFETY ORGANIZATION 11-23 11-3 FACILITY WORK ORDER FORM.
11-24 License No SNM-778 Docket No.70-824 Date October, 1985 l
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11.1 LRC LINE ORGANIZATION The Director is responsible for all LRC operations. Two Laboratory Managers report to him. The Laboratory Managers are responsible for operations that fall within the areas of expertise encompassed by the sections under their control.
Section Managers report to the Labora-tory Managers and are responsible for the safe performance of projects under their purview.
Research and development performed at the Lynchburg Research Center may result in projects being performed by sections of different laboratories in the same building.
For this reason, the Director has established the position of Facility Supervisor.
He advises the Laboratory and Section Managers in the safety aspects and the control of licensed material and coordinates the safety progran within their areas.
The Facility Supervisor utilizes the expertise
' of the Supervisor, Health and Safety, the Nuclear Safety Officer, the Accountability Specialist, and the Industrial Safety Officer to ensure the safety of operations performed at the LRC.
Figure 11-1 shows the LRC Line Organization and Figure 11-2 shows the safety organization.
11.2 LRC SAFETY ORGANIZATION 11.2.1 Manager, Safety and Licensing - The Manager of Safety and Licensing is appointed by and reports to the Director. He is responsible for the proper management of the materials accounting function, licensing function, and the Health and Safety Group. He manages the allotment of funds and other resourses and assures the proper assignment of personnel priorities.
The Supervisor, Health and Safety, Accountability Specialist and License Administrator, report to him.
11.2.2 Supervisor, Health and Safety - The Supervisor of Health and Safety is appointed by the Director and reports to the Manager, Safety and Licensing, but nay have direct access to the LRC Director in matters concerning Health and Safety. The Supervisor directs the overall operation of the Health and Safety Group and the Industrial Safety Officer. He also serves on the Safety Review Committee. He has the authority to stop any operation that he believes is contrary to accepted safety practices, or license requirements.
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The Supervisor has overall responsibility for the shipment and receipt of licensed naterial and exercises signature authority on all Area Operating Procedures.
He performs audits of the LRC for compliance with Health and Safety rules. The Health Physics Engineer and Industrial Safety Officer report to him.
11.2.3 Health Physics Engineer - A Health Physics Engineer reports to the Supervisor, Health and Safety. He administers the activities of the Health Physics staff, which include:
1.
Perforfling area surveys 2.
Administering the air sampling program 3.
Administering the respiratory protection progran 4
Administering the bioassay program 5.
Leak testing radioactive sources 6.
Supervising shipping and receiving of licensed material 7.
Supervising and coordinating the waste disposal program 8.
Assisting in personnel, equipment, and facility decontamination U
9.
Conducting radiation safety training 10.
Providing expertise in all aspects of radiation protection
- 11. Generating, maintaining and distributing records and reports that are required by NRC regulations or Health Physics procedures 12.
Providing expertise in health physics to the Facility Supervisor.
11.2.4 Industrial Safety Officer - The Industrial Safety Officer reports to the Supervisor, Health and Safety. His responsibilities include the following:
1.
Administering the industrial safety progran 2.
Reviewing proposed facility changes to ensure fire safety License No SNM-778 Docket No.70-824 Date October,1985 O
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Providing expertise in fire prevention to the Facility Supervisor and the Safety Review Committee 4.
Performing tests, maintenance, and inspection of fire protection, control, and extinguishing equipment 5.
Providing training for the LRC Fire and Rescue Team and off site support agencies 6.
Inspecting all areas of the LRC periodically to ensure:
a.
Proper storage and use of flammable solvents b.
Proper placement of fire extinguishing equipment c.
Elimination of fire hazards d.
Reduction, to the extent practicable, of the accumulation of flanmable materials e.
Proper use and maintenance of electrical equipment.
7.
Working with supervisors to formulate safety rules and elimi-nation of hazards n
(
)
8.
Investigation of all personnel injuries 9.
Keeping management informed concerning industrial safety activities
- 10. Conducting industrial safety training.
11.2.5 Accountability Specialist - The Accountability Specialist reports to the Manager, Safety and Licensing. He is responsible for the maintenance and retention of SNM accountability records. He prepares and transmits the reports required by regulation to inforn regulatory agencies of SNM transactions.
11.2.6 License Administrator - The License Administrator reports to the Manager, Safety and Licensing.
The License Administrator is responsible for administering the license.
He is the primary liaison between the LRC and the NRC and other federal, state, and License No SNM-778 Docket No.70-824 Date October,1985 (G)
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local agencies regarding nuclear natters. He is the coordinator of the Safety Review Committee and Chairman of the Safety Audit Sub-committee and represents LRC management on both. The License Administrator is responsible for ensuring that corrective action is taken in response to audit findings as they pertain to licensed activities.
11.2.7 Nuclear Safety Officer - The Nuclear Safety Officer is appointed by and reports to the Director. The Nuclear Safety Officer is re-sponsible for ensuring that no operation at the LRC can lead to the inadvertent assembly of a critical mass. To help assure this, he has signature authority for all new Area Operating Procedures and changes to these procedures, he observes operations, institutes educational programs if and when he deems them necessary, and carries out confirming nuclear criticality safety calculations.
The Nuclear Safety Officer will inspect all LRC operations where special nuclear material is being processed, quarterly. Other areas may be inspected less frequently, but all licensed facilities will be inspected at least twice a year. He will consider area operations when scheduling these inspections and will, if neces-sary, schedule his inspection at more frequent intervals. His consideration should include inspection of new operations, an audit of nuclear safety records, a check for area posting and a review of current practices. He shall submit a report of his finding to the V)
Director, with a copy to the License Administrator.
Prior to the
(
submission of the report, he will discuss its contents with the Facility Supervisor. The following information is to be included:
1.
Areas visited 2.
Operations observed 3.
Unsafe practices and situations noted 4.
Nuclear safety activity of the quarter 5.
Recommendations.
11.2.8 facility Supervisor - The Facility Supervisor is appointed by and reports to the Director. He shall be responsible to the Director for the safe conduct of all operations at the LRC and for ensuring that these operations are conducted in accordance with all license conditions. The Facility Supervisor shall review and have approval authority for Area Operating Procedures. He shall have authority to terminate any operation that he deems contrary to license con-ditions, Area Operating Procedures, or general safety conditions.
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V The Facility Supervisor shall become familiar with all license conditions and procedures concerned with radiation safety, nuclear safety, industrial safety, and nuclear materials safeguards. He nay consult with the following personnel to ensure compliance with all safety regulations and principles:
Supervisor, Health and Safety Nuclear Safety Officer Industrial Safety Officer Accountability Specialist 11.2.9 Safety Review Committee - The Safety Review Committee (SRC) shall be comprised of at least five technically trained and experienced members appointed by the Director.
One nenber shall be selected by the Director to be the SRC Chairman. The Chairman shall preside at the neetings and keep the minutes. The Director shall appoint an Alternate Chairman who shall act for the Chaiman during absences.
One member shall be appointed by the Director to be the SRC Coordinator. The Coordinator shall represent LRC management on the SRC, set the meeting agenda, and maintains the permanent files of the Committee.
cx The SRC membership shall have expertise in chemistry, nuclear "1
)
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 operations.
Consultants with special expertise are available to the Committee when needed.
The SRC shall meet at least four times a year. A quorum shall consist of a simple majority of the membership including the Chairman. The SRC shall review and approve all Area Operating Procedures.
It shall review and approve new projects that utilize licensed material that are significantly different from previously reviewed and approved projects. The SRC shall review the annual report issued by the Supervisor, Health and Safety which summarizes LRC personnel exposures, environmental releases, and a summary of the ALARA program accomplishments.
The SRC Chairman shall forward the Committee minutes to the Director, with a copy to the SRC Coordinator.
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The Director shall appoint the menbers of the Safety Audit Sub-committee (SAS). The SAS shall be comprised of at least two individuals, one of whom shall be designated as Chairman and he shall report to the Chairman, SRC. The SAS shall audit operations at the LRC at least three times annually, with successive audits separated by at least two months. Additional audits may be performed at any time. The SAS Chairman shall develop the audit report and submit it to the SRC Chairman. The SRC Chairman shall subnit the audit report to the Director with appropriate comments, with a copy to the License Administrator.
11.3 EDUCATION AND EXPERIENCE OF KEY PERSONNEL 11.3.1 Safety and Licensing Manager - Richard L. Bennett Education:
B.Ch.E. - Chemical Engineering, University of Delaware,1958 Experience:
(1985-Present)
Babcock & Wilcox, Manager, Safety and Licensing, Lynchburg Research Center, Lynchburg, Virginia.
)
See Section 11.2.1 v
(1982-1985)
Babcock & Wilcox, Manager, Building C Decommissioning, 1,ynchburg Research Center, Lynchburg, Virginia He was responsible for decontaminating facilities that were used for preparation of experimental quantities of nuclear fuels containing plutonium.
(1973-1982)
Babcock & Wilcox, Supervisor, Process Technology Group, Lynchburg Research Center, Lynchburg, Virginia This group was responsible for long-range studies, design assistance, start-up assistance, and preparation of environmental reports and safety analyses related to nuclear fuel conversion.
Some of the specific projects performed by the group were prepa-ration of the designs for a low-enriched nuclear fuel conversion plant, preparation of a conceptual design for a spiked nuclear fuel License No SNM-778 Docket No.70-824 Date October,1985
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I' i fabrication plant, process engineering assistance to nuclear fuel
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conversion plants, development of a halide volatility scrap recovery process, development of alternative effluent treatment systems for various nuclear fuel conversion processes, and evaluation of fabrication methods for advanced fuels.
(1971-1973)
Babcock & Wilcox, Senior Research Engineer, Lynchburg Research Center, Lynchburg, Virginia He was responsible for the conceptual design of a facility to treat the effluent from a nuclear fuel plant and developing and evaluating processes for recovering byproducts from B&W wastes.
(1959-1971)
Anerican Cyanamid Company, Process Engineer, Piney River, Virginia He has had broad experience in chemical engineering. This includes research and development, designing equipment and processes, testing and operating new equipment, pilot plant operation, process engineering, and economic evaluation.
He has specific knowledge in pigment manufacture, effluent treatment, and byproduct recovery.
Professional Affiliations:
American Institute of Chemical Engineers (Member)
(~~)
American Nuclear Society (Member)
V 11.3.2 Health and Safety Supervisor - John W. Cure, III Education:
B. S. - Electrical Engineering, Virginia Military Institute,1952 M. S. - Physics, Vanderbilt University, 1956
- AEC-Sponsored Radiological Physics Fellowship, 1952-54 (in conjunction with Vanderbilt University and Oak Ridge National Laboratory)
- Nuclear Safety Training Course, Oak Ridge,1957
- Certified Health Physicist, American Board of Health Physics, 1961
- In-Place filter Testing, Harvard,1976 License No SNM-778 Docket No.70-824 Date October,1985 1
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Experience:
(1956-Present)
Babcock & Wilcox, Supervisor, Health and Safety, Lynchburg Research Center, Lynchburg, Virginia Mr. Cure established the Health Physics Program at B&W's Critical Experiment Laboratory, which is now Building A, at the Lynchburg Research Center. This program was expanded as the nuclear activities at the Center grew.
In 1972 this program provided health physics coverage for four critical experiment reactors, a one-megawatt pool-type research reactor, a six-megawatt test reactor, hot cells, a radiochemistry laboratory, and uranium, thorium, and plutonium fuel laboratories. The Health and Safety Group, in addition to providing operational surveillance, is responsible for the solid waste disposal program, shipping and receipt of radioactive materials, liquid waste disposal progran, and environmental surveillance. This group is responsible for administering the bioassay program, implementing the respiratory protection progran and naintaining the personnel exposure records systen. At the present time the Health and Safety Group consists of three health physics engineers, three technicians, and the survey monitors associated with two decommissioning projects.
Mr. Cure also provides calculational support for the Health and
(^3 Safety Group.
Mr. Cure is also responsible for Industrial Safety which includes compliance with the regulations of the Occupational Health and Safety Administration, administering the fire prevention and fire protection programs, and enforcing the safety standards of the Company's insurance underwriters.
Mr. Cure serves on the LRC's Safety Review Committee and the Safety Committee at the Alliance Research Center.
He has served on the Radiation Safety Committee at the Virginia Polytechnic Institute and State University.
In 1973, he was a member of the Oak Ridge National Laboratories Applied Health Physics Advisory Committee.
He has taught formal courses in health physics and radiation safety in B&W educational programs such as that provided for the N.S. Savannah deck officers (first and second crews), CAMEN research reactor operators, and operator training for Arkansas Power and Light Company, Sacramento Municipal Utility District, Florida Power Company, Metropolitan Edison Power Company and Toledo Edison Power Company.
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(1954-1956)
U. S. Air Force, Kirtland Air Force Base, New Mexico Mr. Cure was a nuclear research officer. He participated as a health physicist during operation " Tea Pot."
(1953-1954) Junior Health Physicist, Oak Ridge National Laboratory, Junior Health Physicist, Oak Ridge, Tennessee Mr. Cure was employed as a junior health physicist at the Oak Ridge National Laboratory and received experience in the design and testing of new instruments and also in field work.
Professional Affiliations:
American Nuclear Society (Member)
Health Physics Society (Member)
Subcomittee on Internal Dosimetry Working Committee - 1972-75 Virginia Health Physics Society - President, 1975-76, 1981-82
- Councilman, 1976-77 American Industrial Hygiene Association Virginia Manufacturers Association, Water and Air Control Comm.
Virginia Safety Association - Board of Directors O
11.3.3 Health Physics Engineer - W. Scott Pennington G
Education:
B.S. - Biology with an option in Health Physics, Virginia Polytechnic Institute and State University, 1978 Experience:
(1979-Present)
Babcock & Wilcox, Senior Health Physicist, Lynchburg Research Center, Lynchburg, Virginia Mr. Pennington ad.ninisters the LRC's health physics program.
The program includes the measurement and control of the external exposure, internal exposure, environmental sampling, the respiratory protection program, the solid waste disposal program, the liquid waste disposal program, and providing radiation and contamination surveillance for the Center's decommissioning projects. He provides expertise in radiation safety to project engineers. He implements the bioassay (in vivo and in vitro)
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program. He has approval authority for radiation work permits, facility work orders, and area operating procedures, in the absence of the ~ Supervisor, Health and Safety.
(1978-1979)
Southwest Research Institute, Environmental Science Technician, Environnental Science Department, Houston, Texas Mr. Pennington was involved in a study for the Bureau of Land Management of off-shore oil platforms in the Gulf of Mexico, and studies for the Houston Lighting and Power Company on thermal pollution and commercial fish survivability. ~
Professional Affiliation:
Health Physics Society (Member)
Virginia Health Physics Society (Member)
American Nuclear Society (Member)
B 11.3.4 Industrial Safety Officer - Reginald P,. Spradlin Education: - Graduate, Appomattox County High School
- Certified Instructor Trainer, Basic Cardiac Life Support, American Heart Association
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- Certified Instructor, First Aid & Advanced First Aid,
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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 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 License No SNM 778 Docket No.70-824 Date October,1985 f'j Amendment No.
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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 certifled Shock Trauma Technicico, an Energency 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)
Babcou & Wilcox, Health Physics Technician, Lynchburg Research Center, Lynchburg, Virginia Mr. Spradlin was a health physics technician in the Plutonium Development Laboratory. He was responsible for performing contamination surveys of the facility, assisting in the nonitoring of bagging operations, and supervising decontamination. He implemented the surveillanace program for airborne radioactive material.
He performed maintenance, testing, and calibration of alpha particle survey 10strumentation and counting equipment.
He implemented the respiratory protection program in that laboratory.
(1967-1969)
Babcock & Wilcox, Plant Engineering Technician, Lynchburg Research Center, Lynchburg, Virginia As a plant engineering technician, Mr. Spradlin performed installation, modification, and repair of facilities, equir-ant, and experimental apparatus at the LRC. He performtd these cuties on electrical, mechanical and plumbing systems.
(1952-1967)
Mead Corporation, thintenance Superintendent, Mead Paper Company, Lynchburg, Virginia
\\
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Mr. Spradlin served in several capacities during this period, including:
finishing operation, paper machine operation,
- lillwright, Maintenance Foreman, Maintenance Superintendent, Safety Inspector and Accident Investigator.
Professional Affiliations:
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 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
-s 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 material.
He initiates the paper work required for all shipments of licensed material.
In addition to his normal duties he is a Document Custodian.
In this capacity, he is responsible for the safe storage of all classified DOE and 000 documents at the LRC. He is also an authorized classifier and an authorized courier of classified naterial.
(1970-1974)
Babcock & Wilcox, Shipping & Receiving Clerk Lynchburg Research Center, Lynchburg, Virginia License No SNM-778 Docket No.70-824 Date October,1985 O
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Mr. Long was responsible for the shipment and receipt of all V
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 5 9 materials into and out of his area.
Professional Affiliations:
Institute of Nuclear Materials Management (Senior Menber)
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 Education:
O (V
AAS - Nuclear Technology, Central Virginia Community College,1978 Experience:
(1972-Present)
Babcock & Wilcox, Senior License Administrator and Facility Supervisor, Lynchburg Research Center, Lynchburg, Virginia Mr. Olsen is responsible for preparing, amending, and administering the licenses that the LRC possesses with the NRC and the Common-wealth of Virginia.
He acts as the primary liaison between the LRC and the NRC and other federal, state, and local agencies regarding nuclear matters. He coordinates the visits made by the NRC's Office of Inspection and Enforcement, and coordinates the LRC's compliance with NRC and state regulations and the licenses. He is the coordinator of the Safety Review Committee and is Chairman of the Safety Audit Subcommittee, and represents LRC management on both. Mr. Olsen is the Facility Supervisor and as such is responsible to the Director for the safety of all operations at the LRC.
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Mr. Olsen is the Alternate LRC Security Officer, Alternate
'd 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 Shift Leader, Babcock & Wilcox Test Reactor, Lynchburg Research Center, Lynchburg, Virginia Mr. Olsen possessed a Senior Reactor Operator's License for the BAWTR.
He was in charge on one of four shifts of reactor operators charged with the proper operation and maintenance of the BAWTR.
He supervised the loading and unloading of fuel and experiments in the reactor and kept all required records of operations and maintenance performed on his shift.
(1960-1964)
U. S. Navy, Reactor Plant Electrical Supervisor, O(V 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:
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 License No SNM-778 Docket No.70-824 Date October,1985 O
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Babcock &Wilcox a McDermott company
7Q 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 Transferred to the LRC as Nuclear Criticality Safety Specialist for Babcock & Wilcox's Naval Nuclear Fuel Plant, Commercial Nuclear Fuel Plant, and the LRC. He was appointed Nuclear Safety Officer for the LRC.
(1964-1969)
Babcock & Wilcox, Power Generation Division, g)
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 He has been concerned with core neutron physics analysis and design of the Consolidated Edison Reactor, the Liquid Metal Fuel Reacter, the Babcock & Wilcox Test Reactor, the Advanced Test 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.
License No SNM-778 Docket No.70-824 Date October,1985 Amendment No.
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(1957-1960)
Babcock & Wilcox, Atomic Energy Division
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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:
Sigma Pi Sigma (Member)
Tau Beta Pi (Member)
American Nuclear Society - Past Chairman of ANS Nuclear Criticality Safety Division
- Member Standards Subconmittee ANS-8.
11.4 OPERATING PROCEDURES
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11.4.1 Area Operating Procedures ( A0P) - Area Operating Procedures are V
prepared by any technically competent person. The proposed pro-cedure 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 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.HAS).
If the NSO approves it, he signs the procedure in the space provided and forwards it to the 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 ti:e 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 License No SNM-778 Dochet No.70-824 Date October,1985
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Review Committee. The Safety Review Co mittee (SRC) may approve the procedure as written, 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.
11.4.2 Availability A0P's are entered in 3-ring binder manuals.
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-(nV) 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 proper background and technical base to allow workers to understand the principles of radiation safety. The Health and Safety Group administers the course and, in general, teaches each course. Where practical, basic general procedures and federal regulations are included and discussed. Training aids, such as motion pictures and self-study materials, are used as appropriate.
Program 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 License No SNM-778 Docket No.70-824 Date October,1985 O
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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 fornal and on-the-job training as 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 r,
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.
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 License No SNM-778 Docket No.70-824 Date October,1986 i]
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The importance of maintaining exposures as low as is s
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.
2.
Health Physics Instruments a.
Personnel monitoring devices b.
Cutie pie and Geiger-Mueller counter c.
Alpha survey meter d.
Air i.onitors 4
e.
Criticality alarm system f.
Emergency equipment g.
Instructions in field use of instruments.
V)
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3.
Regulations and Procedures a.
Code of Federal Regulations (including 10 CFR 19) b.
License requirements 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.
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V 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 doct. ment that such training as been completed. Those persons who direct the work cf 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 devices is maintained.
Training in respiratory protection shall include the following subjects:
a.
Discussion of the airborne contaminants present in the work environment including their physical properties, physiological actions, toxicity, neans of detection, and naximum 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, pQ c,
Discussion of the construction, op erating principles, and limitations of the available respi ators, d.
Discussion of the use of engineering controls as a substitute for respiratory pentection and the need to make every reason-able effort to reduce or aliminate the need for respiratory protection.
e.
Instruction in methods to be used to determine that the respirator is in proper working order, f.
Instruction in fitting the respirator properly, field testing for proper fit, and factors that may influence a proper fit.
9 Instructions in the proper use and maintenance of the respirator.
h.
Discussion of the uses of various cartridges and canisters available for air-purifying respirators.
License No SNM 778 Docket No.70-824 Date October,1985 Amendment No.
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- i. Review of radiation and contanination 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 presented on respiratory protection.
11.6 FACILITY CHANGE
, Changes and nodifications to buildings, exhaust ventilation systems, gas supply systems, emergency electrical systems, etc. are requested on Form LRC-229, " Facilities Work Order Form" (Figure 9-4). All work orders are forwarded to the maintenance supervisor. The Plant Engineering Supervisor determines if the request involves a facility change.
If a facility change is involved, the work order is forwarded to the Facility Supervisor.
It is the Facility Super-
[,^;
visor's responsibility to determine that all safety and licensing K./
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.
Completed forms are kept on file by the maintenance supervisor and are audited once a month by the Health Physics Group.
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.u FIGURE 11-1
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7 LYNCH 8dRG RE'!%1CH CENTER g
T. C, Engelder s
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SECRETART q
J. G. Billino les t
QUALITY ASSURMCE DECOMISSIONING Lynchburg Research Center s
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G. 5. Hoovier I
G. W. Roberts Manager Manager l
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f FACILITIES P(RSONNEL s PURCHA5!NG ACCOUNTING AND T
3AFETY' AfD LICEN51%
ADMINISTRATIVE SEAf!CES C. E. Bell J. R. Parsell M. R. Keith R'. t.. Dennett J. P. Doc.i' Manager Manager Manager Manager Manager 4s I
r SYSTEMS PEVELOPMENT LABORATORY
. SENIOR MIENTIST ikTERIALS ENGINEERING LABORATORY A. E. Wehrmeister C. 5. Caldwell
- Lynchburg Operations Manager P. 5. Ayres Manager 4
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Licer;se No SNM 778 Docket No.70-824 Date October,1985 Amendment No.'
Revision No.
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I FIGURE 11-2 x_/
DIRECTOR T. C. Engelder SAFETY REVIEW C0lHITTEE I
i FACILITY SUPERVISOR SAFETY & LICENSING l
NUCLEAR SAFETY OFFICER A. F. Olsen R. L. Bennett F. H. Alcorn Manager.
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HEALTH & SAFETY J. W. Cure LICENSE ADMINISTRATOR Supervisor ACCOUNTABILITY SPECIALIST A. F. Olsen K. D. Long I
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INDUSTRIAL SAFETY IN R OFFICER W. S. Pennington S. W. Schilthelm R. R. Spradlin License No SNM-778 Docket No.70-824 Date October,1985 0
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FIGURE 11-3 J'~ _...
FACIUTIES WORK ORDER FORM T0e Plent Engineering Date 1
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From
. Section:
Signed:
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.a Section Mgr.:
Dates M'
Date Required Charge No.:
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(Labor)
(Material)
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~h DESCRIPTION OF WORK TO BE DONE s
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,6 SIGN ATURE REQUIREDe industrial Safety Officers a
hy Health Physicer F acility Supervisori _
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Order Received Date Signed l
Planned Stor6ng Date Planned Completion Date Order Completed:
Work Order Number Date Signatween n
Licensa No SNM-778 Docket No.70-824 Date October, 1985 0
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%J TABLE OF CONTENTS Section Page 12.0 RADIATION PROTECTION.
12-1 12.1 PROGRAM 12-1 12.2 POSTING AND LARELING.
12-1 12.2.1 Radioactive Materials Area.
12-1 12.2.2 Contanination Area 12-2 12.2.3 Radiation Area.
12-2 12.2.4 High Radiation Area 12-2 12.2.5 Airborne Radioactivity Area 12-2 12.3 EXTERNAL RADIATION - PERSONNEL MONITORING.
12-3 12.3.1 Administrative Exposure Control 12-3 s
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12.3.2 Personnel Monitoring for LRC Employees.
12-3 12.3.3 Personnel Monitoring and Escort Requirements for Non-LRC Employees.
12-5 12.3.4 Monitoring Devices 12-7 12.4 DIRECT RADIATION SURVEYS 12-8 12.5 REPORTS AND RECORDS 12-9 12.6 INSTRUMENTS.
12-9 12.6.1 Types.
12-10 12.6.2 Calibration.
12-11 12.7 PROTECTIVE CLOTHING 12-12 License No SNM 778 Docket No.70-824 Date October, 1985 0
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TABLE OF CONTENTS (Cont'd)
Section Page 12.7.1 Clothing.
12-12 12.7.2 Emergency C1othing 12-12
- 17. 8 ADMINISTRATIVE CONTROL LEVELS.
12-12 12.8.1 Internal Occupational Exposure 12-12
- 19. 8.7.
External Occupational Exposure 12-19 12.8.3 Airborne Activity.
12-19 12.8.4 Liquid Activity 12-20 12.8.5 Surface Contanination 12-21 12.8.6 Chnpliance with 40 CFR 190.
12-25 12.9 RESPIRATORY PROTECTION 12-25 n
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- 12. 10 0CCUPATIONAL EXPOSURE ANALYSIS 12-26 12.10.1 External Exposure Analysis.
12-26 12.10.2 Internal Exposure.
12-28 12.11 MEASURES TAKEN TO IMPLEMENT ALARA 12-37
- 17. 11.1 Management Connitment-12-37 12.11.2 Training.
12-37 12.11.3 ALARA Connittee 12-41 12.11.4 Management Supervision 12-42 1?. 11.5 Authority 12-42 License No SNM-778 Docket No.70-824 Date October,1985 0
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TABLE OF CONTENTS (Cont'd)
Section Page 12.12 BI0 ASSAY PROGRAM 12-42 12.13 AIR SAMPLING AND MONITORING 12-43 12.13.1 Air Sanpling Program.
12-43 12.13.2 Air Monitoring Program 12-44 12.14 SURFACE CONTAMINATION 12-45 12.14.1 Smear Surveying 12-45 12 14.2 Direct Radiation Surveys 12-48 12.14.3 Personnel Contanination Surveys 12-50
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List of Tables i
L/I Table Page 12-1 PORTABLE RADIATION PROTECTION INSTRUMENTATION 12-10 17.-2 STATIONARY RADIATION PROTECTION INSTRUMENTATION.
12-11 12-3 PLtlTONIUM 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 tlRANIUM BI0 ASSAY ACTION CRITERIA.
12-17 12-7 BETA-GAMMA ACTION CRITERIA.
12-18 12-8 STACK RELEASE ACTION LEVELS 12-20 License No SNM 778 Docket No.70-824 Date October, 1985 0
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List of Tables (Cont'd)
Table Page 12-9 SMEAR SURVEYS IN WORK AREAS 12-21 12-10 ACTION LEVELS FOR LARGE AREA SMEARS.
12-22 12-11 MAXIMUM PERMISSIRLE CONTAMINATION FOR SKIN SURFACES 12-23 12-12 MAXIMUM PERMISSIBLE CONTAMINATION OF CLOTHING 12-24 17.-13 LRC RADIATION EXPOSURE 12-27 12-14 EXPOSURE BY GROUP (PERSON REMS) 12-27 12-15 NUMBER OF URINE RI0 ASSAY SAMPLES.
12-29 12-16 1983 AIR ACTIVITY.
12-30 12-17 1984 AIR ACTIVITY.
12-32 12-18 WHOLE BODY COUNTS 1983 12-34
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12-19 Am - Pu LUNG COUNTING 1983 12-35 12-20 URANIUM LUNG COUNTING 1983 12-36 12-21 WHOLE B0DY COUNTS 1984 12-36 12-22 ACTION LEVELS FOR LARGE AREA SMEARS.
12-46 12-23 SMEAR SURVEY FREQUENCIES AND ACTION LEVELS 12-47 12-24 CONTAMINATION ACTION LEVELS 12-49 License No SNM.778 Docket No.70-824 Date October, 1985 0
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12.0 RADIATION PROTECTION 12.1 PROGRAM The radiation protection progran at the LRC is implemented to protect employees and the general public from the harmful effects of radi-ation and radioactive material, to comply with NRC regulations, and to maintain personnel exposures as far below the limits established by the NRC as is reasonably achievable.
Implementation of the program requires the active participation of all employees who work with licensed material or in areas were licensed material is handled. To support the worker, the LRC has established the Health and Safety organization and vested it with the authority and resources necessary to meet the program goals.
12.2. POSTING AND LABELING Many areas in the LRC are required to be posted to indicate the hazard present. This posting is required by the federal regu-lations and is a fundamental part of an effective radiation pro-tection program. Posting of areas makes the workers aware of the potential hazards in the area and assists workers in keeping their (o")
exposures ALARA.
Permanent postings are the responsibility of the Health and Safety Group. Temporary postings are the responsibility of Authorized Users. This section discusses the posted areas at the LRC. Persons not directly familiar with conditions existing in a posted area shall contact a qualified person prior to entering and shall enter only under his direction.
12.2.1 Radioactive Materials Area - Any area where radioactive materials are stored, handled, or processed in amounts exceeding 10 times the quantities specified in 10 CFR 20, Appendix C is designated a radioactive materials area. Each area is clearly marked at every normal entry with a sign bearing the radiation caution synbol and the words CAUTION - RADI0 ACTIVE MATERIAL (S). Monitoring equipment and protective clothing required for use in the area will be specified by the Health and Safety Group.
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- 17. 2.7.
Contanination Area - This is any area in which loose contamination is present in quantities in excess of those specified in Table 12-24 or an area designated by the Health and Safety Group as one in which there is a risk of contamination. Each contamination area is clearly marked at every normal entry. Work in these areas may require a Radiation Work Permit. Protective clothing, respiratory protection, and personnel monitoring devices required for entry into these areas must be specified by the Health and Safety Group.
Entry into the area without the prescribed equipment is prohibited.
When exiting a contamination area, the employee must remove the protective clothing and nonitor himself in accordance with established procedures.
12.2.3 Radiation Area - A Radiation Area is an area in which an individual could receive a radiation exposure to a major portion of the body greater than 5 mrem in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or 100 nRen in 5 consecutive days.
Each radiation area is clearly marked at every nomal entry with a sign bearing the radiation caution symbol and the words - CAUTION
- RADIATION AREA. Work in these areas may require a Radiation Work Pemit. Personnel monitoring devices and protective clothing to be worn in the area will be specified by the Health and Safety Group.
12.2.4 High Radiation Area - Any area in which an individual may receive an exposure to a major portion of the body greater than 100 mrem in n
1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> is a High Radiation Area. High radiation areas are desig-()
nated by a sign at each normal entrance bearing the radiation caution symbol and the words - CAUTION - HIGH RADIATION AREA.
Entry into high radiation areas is limited to qualified persons, or under the direct supervision of a qualified person and, working under an approved radiation work pemit.
Protective clothing, protective equipment, and personnel monitoring devices appropriate for the area will be specified by the Health and Safety Group and must be worn. When protective clothing is required, each person must remove the protective clothing and nonitor himself in accordance with established procedures, when exiting the area.
12.2.5 Airborne Radioactivity Area - This is an area in which airborne radioactivity concentrations could exceed the maximun pemissible concentration limits given in 10 CFR 20, Appendix B or in which the concentration of airborne radioactivity averaged over the number of hours individuals are in the area could exceed 25% of the limits given in 10 CFR 20, Appendix B.
Each area is clearly designated by a sign at each normal entrance bearing the radiation caution symbol License No SNM-778 Docket No.70-824 Date October,1985 O
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and the words CAUTION - AIRBORNE RADI0 ACTIVITY AREA. Entry is limited to those qualified persons classified as radiation workers, working under an approved radiation work permit. No entry is permitted until an appropriate area survey has been made and a member of the Health and Safety Group is present.
Protective clothing, protective equipment, and personnel monitoring devices to be worn in the area will be specified by the Health and Safety Group and must be worn. When exiting these areas, each person must remove the protective clothing and monitor himself in accordance with established procedures.
12.3 EXTERNAL RADIATION - PERSONNEL MONITORING 12.3.1 Administrative Exposure Control - Limits for external radiation exposure are set forth in 10 CFR 20.102 and these general limits are used at the LRC. The applicable exposure limits to be used for operations at the LRC are:
1.
Whole body - 300 mrem / week (with long-term exposure controlled within the 1.25 Rem / quarter limit by the worker's immediate supervisor) 2.
Skin of the whole body - 1.5 Rem / week
[)J 3.
Hands and forearms, feet and ankles - 3.0 Rem / week.
8_
The Director, LRC has the authority to approve whole body exposures up to, but not exceeding, 3.0 Rem / calendar quarter.
In energencies, the Emergency Officer is authorized to allow personnel exposures to the whole body of up to 3.0 Ren/ calendar quarter.
Higher exposures may be authorized by the Emergency Officer in accordance with the Radiological Contingency Plan.
12.3.2 Personnel Monitoring for LRC Employees - All LRC employees will be monitored for radiation exposure while on site. This monitoring will be accomplished in two ways:
1.
All employees will be issued a thermoluminescent dosimeter (called an Annual TLD).
2.
Employees will be classified as radiation or non-radiation workers.
In addition to the Annual TLD, those employees License No SNM-778 Docket No.70-824 Date October, 1985 0
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ij classified as radiation workers will be issued a film badge and two indirect-reading pocket dosimeters. These monitoring devices will be worn at all times while the employee is on the LRC site. As a substitute for the indirect-reading dosimeters, radiation workers may be issued a self-reading pocket dosimeter (SRD) and another TLD badge (called a monthly TLD).
All new employees, and all individuals to be monitored as empl,oyees, must be admitted initially using Form LRC-129 and must receive the proper indoctrination.
12.3.2.1 Annual TLD - Each LRC employee will be issued an Annual TLD.
.This dosimeter will be taped to the employee's identification (ID) badge. The employee is required to have the ID badge and dosimeter on his person at all times he is on the LRC site. The ID badge and dosimeter may be worn in an exposed location on the body or it may be carried in the employees wallet. Secretaries, and other female employees, who do not normally carry a wallet separate from their purse are pemitted to leave this dosimeter in their purse at their desk, if they wish. These employees have normal access to areas at the LRC. However, entry into a radi-ation area is not pemitted.
Employees wearing only an Annual TLD are permitted to visit or work in radiation and radioactive materials areas provided their (m) estimated whole body dose does not exceed 100 mrem for the year.
e Entry into a high radiation area or an airborne radioactivity area will not be permitted.
12.3.2.2 Radiation Workers - An' employee whose occupational whole body dose is projected to exceed 100 mrem / year is defined as a radi-ation worker. The Health and Safety Group.is responsible for making the determination of the employees projected exposure.
This evaluation will include consideration of the job assignment, past exposure history, and other factors relevant to the determi-nation.
Radiation workers shall wear a film badge and two indirect-reading pocket dosimeters. These personnel monitoring devices shall be carried at all times while on the Mt. Athos site. When not in use, the film badge and pocket dosimeters are required to be left in the film badge rack at the LRC. The Annual TLD must be carried on the person but it is' not required that this dosimeter be worn in close proximity to the other monitoring devices.
License No SNM 778 Docket No. 70-824-Date October,1985 O
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Employees visiting or working at another facility, for periods of
'~'
30 days or less, will wear the film badge on an exposed portion of the body. This requirement applies to visits or assignments to other Company facilities on the Mt. Athos site.
If the visit or assignment should be for more than 30 days or will extend beyond the monthly badge change-out period (the first of each month), the employee must contact the Health and Safety Group for additional instructions concerning film badge exchange.
12.3.3 Personnel Monitoring and Escort Requirements for Non-LRC Emloyees Visitors and non-LRC employees are classified into three categories
.for purposes of personnel monitoring and dose control. These categories are as follows:
1.
Non-employee Visitor A visitor to the LRC who will remain on site for a short period of time.
2.
Employee Visitor An employee from another facility on the Mt. Athos site wearing his own personnel monitoring devices.
(m) 3.
Tour Groups x.;
Reserved for individual tours and/or large tour groups on site for a short period.
A visitor may not receive a radiation exposure exceeding 10 mrem in one week unless he is working under a raciation work permit.
Visitors shall have access to non-radiation areas on site while under the direct supervision of an LRC employee. Entry into radi-ation areas on site shall be made only after a stay-time calcu-lation has been nade by the escort to ensure that the above exposure limit will not be exceeded.
If such a calculation shows that the 10 mRen/ week limit will be exceeded, the visitor must enter the area under an approved radiation work permit.
Certain non-LRC employees may be badged and monitored as employees.
All decisions as to proper classification and monitoring of non-LRC License No SNM-778 Docket No.70-824 Date October, 1985 O
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employees will be made by a Health Physics Engineer.
In general, non-LRC employees will require an escort and will be badged and nonitored to comply with one of the following categories.
12.3.3.1 Non-employee Visitors - Persons visiting the LRC for business purposes are classified as non-employee visitors. These visitors are required to sign in with the Receptionist who will issue each visitor a personnel monitoring device. The visitor is required to return the monitoring device and sign out with the Receptionist each time he leaves the site.
A non-employee visitor, on-site for more than one day, will be issued the same personnel monitor each day.
In some cases, a visitor on-site frequently (e.g., a construction worker) may be assigned a position in the badge rad to store his personnel monitor when not in use. This assignment is made at the discretion of the Health and Safety Group.
It does not relieve the visitor from the responsibility of signing in and out with the Receptionist nor of obtaining and returning the personnel monitoring devices.
12.3.3.2 Employee Visitors - A Company employee from another facility on the Mt. Athos site who is wearing his own personnel monitoring devices, is required to sign in with the Receptionist. The
(~N Receptionist will issue the employee a visitor's personnel
)
monitor. The employee is responsible for returning the visitor's i
personnel monitor and signing out upon leaving the LRC.
Employee visitors coming frequently to the LRC nay be assigned a position in the badge rack. A Health Physics Engineer, at his discretion, nay also waive the sign-in requirements.
In this case, the employee visitor must receive the new employee in-doctrination required for all LRC employees.
12.3.3.3 Tour Groups - Individuals touring the Research Center must sign-in with the Receptionist.
If the group is large, a roster of the members of the group may be substituted for the require-ment that each individual sign-in. The Receptionist will list the time the group signed in and the tour guides. Several members of the group will be issued personnel monitoring devices to provide a representative sampling of the group exposure. At
_least one person in each subgroup will receive a personnel rionitor.
Each subgroup must have a tour guide and it is the License No SNM-778 Docket No.70-824 Date October,1985 O
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I guide's responsibility to keep the subgroup under constant super-vision. At the end of the tour, all personnel monitoring devices must be returned to the Receptionist and the group will be signed out.
Members of a tour group required to make side-trips must be reclassified as a non-employee visitor (with the exception of trips to restroom facilities, etc.).
12.3.4 Monitoring Devices The primary device used for monitoring exposure at the LRC is the film badge. The exposure measured by this badge (reported in units of dose equivalent) becomes a part of the employees permanent exposure record. Films are changed monthly and are mailed of f-site for evaluation.
In some cases, a Health Physics Engineer may choose to base the monthly exposure of an employee on the monthly thermoluminescent dosimeter (TLD). This determination shall be recorded in the employees exposure record.
In general, the employee should wear the dosimeters on the portion of the whole body expected to receive the highest dose (with the exception of extremity dosimetry issued in special cases). The film badge and/or monthly TLD badge should always be worn in the proper orientation to ensure that exposure to non-penetating radi-(n) ation (e.g., beta radiation) is recorded. For cases in which the
'u' exposure may vary significantly within a small area, several badges may be worn to ensure that the maximum whole body dose is measured.
In this context, whole body includes the head, lens of the eyes, the gonads, the upper legs above the knees, and the upper arms above the elbows.
12.3.4.1 Pocket Dosimeters - These dosimeters are small, air-filled ionization chambers used to provide a check of the daily exposure of employees and to ensure that the administrative limit for weekly exposure is not exceeded.
Indirect dosimeters are capable of measuring external exposure to gamma radiation in the range 0 to 200 mR (other ranges are also available). These dosimeters are read, recorded, and rezerced daily. Daily readings are used also as an indication of the need to evaluate the primary dosimeter before the normal exchange period.
License No SNM-778 Docket No.70-824 Date October,1985
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oj Sone employees may be issued self-reading pocket dosimeters
( SRO). These dosimeters do not require reading and recharging on a daily frequency and the employee may evaluate his accumulated exoosure without the need for a special reading device.
Employees are encouraged to read their self-reading dosimeters at least on a daily basis. These dosimeters are capable of measuring external exposure to gamma radiation in the range 0 to 200 nR, but other ranges are available.
12.3.4.2 Film Badges - These dosimeters are the primary monitoring device used at the LRC, i.e., the film badge results are entered in the employee's permanent exposure record.
Film badges monitor external exposure to beta and gamma radiation typically in the range 15 mRems to 500 Rems. For situations in which neutron exposure is probable, film packets sensitive to neutrons also are used.
Films in use at the LRC are changed monthly and mailed to an off-site dosimetry service for processing (reading, recording, and reporting).
12.3.4.3 Thermoluminescent 00simeters (TLD) - TLD's are small, solid-state dosimeters capable of measuring external exposure from beta and gamma radiation in the range 10 mRems to 10,000 Rem. The monthly TLD's are used to duplicate the readings of the film badge.
(n' ')
These badges are also changed monthly and mailed off-site for processing. The Annual TLD, issued to all employees, also measures beta and ganma exposure and has the same range. The badges are evaluated at least annually by the off-site dosimetry service.
At the discretion of a Health Physics Engineer, persons handling radioactive materials may be issued extremity dosimeters. These dosimeters are small TLD cnips attached to a ring and are to be worn on the fingers. TLD " finger rings" are capable of neasuring external exposure to beta and gamma radiation in the range 10 mRens to 10,000 Rems. These dosimeters are evaluated on a frequency established by the Health and Safety Group.
12.4 DIRECT RADIATION SURVEYS Surveys of the direct radiation exposure in areas of the LRC are to be performed on a frequency established by the Health and Safety Group.
In general, these surveys require the selection of the appropriate portable survey instruments based upon the anticipated License No SNM-778 Docket No.70-824 Date October, 1985 O
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V radiation levels, the types of radiation expected, and the nature or type of survey to be perfomed.
Survey maps of the areas to be surveyed may be used to record the measured ambient radiation levels and/or, in some cases, to designate specific areas in which the exposure rates should be measured.
The survey should also include a visual exami-nation of the area for any unusual conditions or work habits which could affect the exposures received by personnel working in these areas.
Items of this nature should be reported immediately to a Health Physics Engineer or corrected immediately, if practical.
Results of these surveys should be reviewed by a Health Physics 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.
12.5 REPORTS AND RECORDS 4
The following records will be maintained by the Health and Safety Group for the periods indicated.
Health and Safety Supervisor audits 2 years Shipping and receiving RM foms 5 years Waste disposal records
(*)
(g Personnel dosimetry records
(*)
)
Results of Bioassays and Whole Body Counting
(*)
Releases to the environment
(*)
Radiation survey data 2 years Contamination survey data 2 years Radiation Work Permits (completed) 5 years Radiation detection instrument calibration 2 years Leak tests of sealed sources 2 years Employee training
(*)
Employee retraining
(*)
Airborne radioactivity sampling data
(*)
NRC-4 forms
(*)
NRC-5 forms
(*)
- - indicates that the record will be retained until the NRC authorizes its disposition.
12.6 INSTRUMENTS License No SNM 778 Docket No.70-824 Date October,1985 O
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12.6.1 Types - The connitment of LRC management to an effective radiation protection progran includes the obligation to provide the adequate equipment and supplies for such a program.
It is the responsi-bility of the Manager, Safety and Licensing and the Supervisor of Health and Safety to ensure the appropriate radiation protection instrumentation is available for use at the LRC.
In addition, the Health and Safety Group has the responsibility to ensure that this instrumentation is used properly, and is calibrated, maintained, and repaired as necessary. Minimum instrumentation requirements for maintaining an effective radiation protection program are listed in Tables 12-1 and 12-2.
Other specialized instrumentation in use at the LRC may not be included in this list. However, the exclusion of these instruments does not imply that their availa-bility does not enhance the effectiveness of the radiation protection program.
TABLE 12-1 PORTABLE RADIATION PROTECTION INSTRUMENTATION Radiation Instrument Sensitivity Range Window Thickness
(]
Low-range GM Beta, Gamma Bkgd. to 30 mg/sq. cm.
(,/
mR/hr Internediate Beta, Gamma nR/hr to Beta:
1 mg/sq. cm.
range ion R/hr chamber High range Ganma up to 500 R/hr
>100 mg/sq. cm.
ion chanber Proportional Alpha, Beta Bkgd. to 1 mg/sq. cm.
counters 500,000 cpm Proportional Neutron, fast Bkgd. to N/A counters and thermal 5000 nRen/hr Portable air Air particulate N/A N/A samplers collection only License No SNM 778 Docket No.70-824 Date October, 1985 0
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TABLE 12-2 STATIONARY RADIATION PROTECTION INSTRUMENTATION Radiation Instrument Sensitivity Range Window Thickness Laboratory Alpha, Beta Rigd. to
<1 mg/sq. cm.
proportional 100,000 cpm counter Air particulate Alpha, Beta Rkgd. up
<1 ng/sq. cm.
nonitors Stack particulate Alpha, Beta Bkgd to
<1 ng/sq. cm.
nonitor 1,000,000 cpn Stack gas Beta, ganna Bkgd. to 30 mg/sq. cm.
nonitor 100,000 cpn 1?. 6.2 Calibration - Portable survey instruments shall be calibrated twice annually using approved procedures and sources traceable to the
(
)
National Bureau of Standards.
In addition, frequent operational k/
checks will be oerforned on survey instruments while in use.
For example, Geiger-Mueller survey instruments always indicate the presence of radiation above the ambient background. This provides an indication that the instrument is functioning.
Portable alpha survey instruments are equipped with check sources which can be used to ensure that the instruments are operating correctly.
Portable ionization chamber survey instruments are not equipped with an internal check source and the user must make sure these instruments are functioning before making a radiation survey.
Fixed and stationary radiation monitoring equipment is calibrated on either a semi-annual or a'inual basis depending on the applicable manufacturer's recommendations and established health physics procedures. Operational checks are performed routinely by the Health Physics technicians on the laboratory counting equipment and
" friskers" located at exits from selected areas in the LRC.
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A 12.7 PROTECTIVE CLOTHING
%/
12.7.1 Clothing - The following is a list of protective clothing that is available for use by personnel during normal and maintenance condi-tions:
1.
Laboratory coats 2.
Coverall s 3.
Shoe covers, treated fabric (reuseable) 4.
Shoe covers, plastic 5.
Pants, plastic 6.
Coats, plastic 7.
Hoods, fabric (reuseable) 8.
Shields, spatter 9.
Glasses, plastic
- 10. Glasses, glass
- 11. Gloves, plastic
- 12. Gloves, surgeons
- 13. Gloves, t. eat resistant
- 14. Coats, heat reflective
- 15. Ha rd-hats.
12.7.2 Emergency Clothing - In the event of an accident that requires special clothing or personnel protective equipment, the Fire and
(T Rescue Team is provided with the following:
\\~
1.
Hard-hats, heat resistant with face shields 2.
Coats, flame resistant 3.
Boots, high top rubber with steel toe shields 4
Gloves, chemical resistant 12.8 ADMINISTRATIVE CONTROL LEVELS 12.8.1 Internal Occupational Exposure 12.8.1.1 Plutonium bioassay action criteria.
License No SNM 778 Docket No.
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TABLE 12-3 PLUT0NIUM BI0 ASSAY ACTION CRITERIA (CLASS W AND Y COMP 0UNDS ONLY)
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-
,_s
(
)
firm exposure, determine
's /
if exposure has exceeded 50% of the naximun permissible annual dose.
5.
If the exposure has exceeded 50% of the naximun permissible annual dose, the worker shall be restricted fron further exposure until the Supervisor, Health and Safety authorizes the removal of this restriction.
> 4 dpm/L 1.
Restrict the individual i
from any further work with plutonium.
License No SNM-778 Docket No.70-824 Date October,1985 0
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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 (CLASS W AND Y COMP 0UNDS ONLY)
O Bioassay b
Technique Action Level Action To Be Taken In-vivo
< 1.6 E-8 Ci None
> 1.6 E-8 Ci 1.
Restrict worker from (16 nanocuries) further exposure.
2.
Resample the individual within 10 working days.
3.
Determine if area surveys support the analysis results.
4.
If area surveys confirm result, investigate the cause and take correc-tive actions.
License No SNM-778 Docket No.70-824 Date October, 1985 O
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A) 5.
If the resample results do not confirm the exposure, the Super-visor, Health and Safety may lift the work restrictions.
6.
If resanple results con-firm the exposure, the Supervisor, Health and Safety shall determine the organ dose.
7.
If the exposure has exceeded 50% of the naximun permissible annual dose, the worker shall remain on a work restriction until the Supervisor, Health and Safety authorizes the renoval of the re-striction.
I';
12.8.1.2 Uranium bioassay action criteria.
k.J TABLE 12-5 URANIUM BI0 ASSAY ACTION CRITERIA (URANIUM ENRICHED IN U-235 <4%)
Bioassay Technique Action Level Action To Be Taken Urinalysis
< 20 ug/L No action
> 20 ug/L 1.
Resample within 5
~
working days.
2.
Perforn an area survey.
License No SNM-778 Docket No.
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If the area survey (#2) confirms the exposure, investigate and correct the cause.
4.
If the resample (#1) results confirm a dispo-sition >50% of a body burden, the worker will be restricted from further exposure and will be in-vivo counted as soon as practicable.
1 5.
If the resample (#1) results indicate >10%
but <50% of a body burden, the worker will be in-vivo counted during the next scheduled period when the service is on site, 7
6.
If the in-vivo counting
(#4) indicates >50% of a (O
body burden, the worker will be restricted from further exposure until the Supervisor, Health and Safety removes the restriction.
12.8.1.3 Uranium bioassay action criteria.
License No SNM-778 Docket No.70-824 Date October,1985 O
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TABLE 12-6 (m)
~'
URANIUM BI0 ASSAY ACTION CRITERIA (URANIUM ENRICHED IN U-235 >4%)
Bioassay Technique Action Level Action To Re Taken Urinalysis
< 55 dpr1/L No action
> 55 dpm/t 1.
Resample within 5 working days.
2.
Survey the work area.
3.
If #2 confirms an exposure, investigate and correct the cause.
4 If the resar ple results
(#1) confirm a depo-sition >50?,of a body burden, the worker will be restricted from
/7 further exposure and
(.)
will be in-vivo counted as soon as practicable.
5.
If the resample results
(#1) indicate >10% but
<50% of a body burden, the worker will be in-vivo counted during the next scheduled period when the service is on site.
6.
If the in-vivo counting results (#4) indicated
>50% of a body burden, the worker will be restricted from further exposure until the Supervisor, Health and Safety removes the restriction.
License No SNM 778 Docket No. 70 824 Date October,1985 p
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A fV) 12.8.1.4 Beta-ganna activity - Workers who work in areas where beta-gamna internal exposure is likely (Hot Cells, Radiochemistry, Health Physics) shall be in-vivo counted at approximately annual intervals.
TABLE 12-7 BETA-GAMMA ACTION CRITERIA Method Action Level Action To Be Taken Estimation
> 10% of a la. Submit in-vitro samples from air Body Burden for analysis with 5
- sampling, working days.
nose smears, in-vivo counting.
> 40% of a 2a. Submit in-vitro samples Body Burden for analysis within 5 working days, (n) restricted from further 2b. Worker will be exposure.
.In-vitro
> 50% of a 3a. Worker will be Fody Burden restricted from further exposure and will be in-vivo counted as soon as practicable.
In-vivo
> 70% of a 4a. The worker shall be Body Burden restricted from further exposure until the Supervisor, Health and Safety authorizes the removal of this restriction.
License No SNM 778 Docket No. 70 824 Date October,1985 O
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AU 12.8.2 External Occupational Exposure - Personnel nonitors (film badges, dosimeters, or other suitable devices) are provided to measure the radiation exposure of visitors and employees. Personnel dosineters issued pursuant to 10 CFR 20,202 shall be read on a monthly basis.
The employee's line supervisor is responsible for keeping exposures below 300 millirem per week and 1250 millirem 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 Director.
If an employee has received the quarterly limit and the Director has not authorized exceeding the limit, the employee's work shall be restricted to prevent further exposure for the remainder of the quarter.
12.8.3 Airborne Activity 12.8.3.1 Air Monitoring Program - Air monitoring in operating areas of the LRC is accomplished with continuous nonitors in predetermined,-
fixed locations. A monitor is placed in each radioactive materials handling area in which there is a potential for the release of airborne radioactivity. Locations are selected based upon the ability of the monitor to provide a reasonable evalu-ation of the airborne activity in a particular area and to provide adequate warnings to those in the area of changing condi-tions. The determinations are made by the Health and Safety p ),
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 anbient 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 through the 50-meter stack. Generally, exhaust air containing beta-gama 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.
License No SNM 778 Docket No.70-824 Date October,1985 O
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i i s' Discharges through the stack are monitored with a sanpling 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 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 12-8 STACK RELEASE ACTION LEVELS p()
Release Product Action Levels Beta Particulate 200 uCi/ week Alpha Particulate 1 uCi/2 weeks (long lived)
Kr-85 70 C1/ week H-3 3 Ci/ week I-131 200 uCi/ week 12.8.4 Liquid Activity - Liquids containing radioactive material are dis-charged from the area where they are generated, to the Liquid Waste Disposal Facility.
This facility is comprised of a series of tanks. All radioact:ye liquid waste is held in this facility for sampling prior to release.
If the concentration of radioactivi;y License No SNM-778 Docket No.70-824 Date October, 1985 O
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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 systen 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 12-9.
Action is taken to protect personnel and reduce the levels of contanination below those specified. The Health and Safety Group will supervise and direct the protection and decontamination activities. Decontani-nation to reduce levels of contanination will connence within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of discovery. However, if the contamination is discovered just prior to the beginning of a holiday or weekend, the contamination will be narked and labeled, and decontanination will begin during the first regular work day af ter discovery.
TABLE 12-9 p.,
SMEAR SURVEYS IN WORK AREAS s
LJ 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 License No SNM 778 Docket No.70-824 Date October, 1985 O
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si y
,\\-
}
23 lly (L')
<---------------------------BETA-------------------------->
s E
~2 v'
T 6
Building A' Labs.
Monthly 2000 N,
%'y Bull' ding B Counting Lab.
Monthly 2000~
Scanning Electron Monthly s 200'O
- y Mg'roscopyLab.
3 g'
s i
~
Hut Cell Operations AOa Bimonthly 2000 Cisk H5npling g' a Bimonthly 22000' y
i
? ' Bimonthly 2200'0 W
Radlochemistry'(:abp. 2 3
st
-t l
~.b Exit portals f&pJ '",
Biweekly 2000 q
r4ntrolled areas s
- 2. g. xV G
q 4
7 Large area smears are usd to survey many square neters of surface Action levels for large q ea smears are given below.
i area.
4 \\!
q a,
- g. TABLE 12-10
- s
- \\g, ACTIONL!gtSFORLARGEAREASMEARS w
i
\\
1.
RoutIneLarge5Ar\\ea Smears (1000-5000 dpm) - Repeat the 1arge
(-q\\
area smear. 'If results show levels of contamination above 1000 dpm, take smears in smaller areas to locate the source.
t N.t Decontaminate all areas in which.the smear results indicate
{
yntatinationabovq1,000dpm/100squarefeet, k
l'arge(areasmear.Large Ar'ea Smars (5000-10,000 dpm)L-Re Ihutin Q.s,, g) 2.
4 If results show levels of contamination i
s, above 5000 dpm, isolate the contaminated erea. Take smears 5,
in smiler areas to locate the source..9econtaminate all fib areaddn which tM smearresults show contmination in excess y
of 1000 dpm/,1 3,quare feet, Q
1 hY.
.'[,
,Q y I\\>
\\
\\
s License No, SNM 778 Docket No.70-824 Date October,1985 3
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],
i
i i
()
3.
Routine Large Area Smears (>10,000 dpm) - Isolate the con-taminated area.
Survey all personnel in the contaminated area. Take smaller snears 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-tanination monitors (friskers) are located at all normal exits from contamination areas for this purpose. The detector should be held as close to the surface of the item being monitored as possible, without touching the item, and the probe should be noved at a slow speed over the surface. Allowable levels of contamination on skin surfaces and on items of clothing are given in Tables 12-11 & 12-12. Any contamination in excess of these linits should be reported immediately to the Health and Safety Group. The Health and Safety Group will supervise the decontani-nation and determine if clothing must be discarded.
TA3LE 12-11 v)
MAXIMUM PERMISSIBLE CONTAMINATION FOR SKIN SURFACES Fixed Alpha Fixed Beta-Gamma Smearable Surface dpm/100 sq. cm.
dpn/100 sq. cm.*
(Alpha, Beta-gamna)
Body 220 2200 None Detectable Hands 220 2200 None Detectable License No SNM 778 Docket No.70-824 Date October,1985 O
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pU TABLE 12-12 MAXIMUM PERMISSIBLE CONTAMINATION OF CLOTHING (dpm/100 sq. cm)
Smearable Item Fixed Alpha-Fixed Beta-Gamma
- Alpha, Beta-Gamma Shoes:
Contaminated Zone I
Inside 2,200 22,000 220 2,200 Outside 22,000 220,000 2,200 22,000 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 V
12.8.5.3 Release of Equipment or Packages - Packages and equipment are 1
surveyed by the Health and Safety Group. The Heat th and Safety 4
Group has the authority to prohibit the release of itens that are found to exceed the below release limits.
1.
Fixe'd Aipha - 2200 dpm/100 cm2 2.
Fixed Beta-gamma - 22,000 dpn/100 cm2 i
3.
Smearable Alpha 30 dpm/100 cn2 4.
Smearable Beta-gamma 2200 dpm/100 cn2 License No SNM-778 Docket No.70-824 Date October,1985 O
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12.8.6 Compliance with 40 CFR 190 - Compliance with 40 CFR 190 has been demonstrated by calculation.
Two release pathways were considered; release through the 50-meter stack and release of liquids. The stack release assumed all alpha activity was plutonium. The lung dose to a nenber of the general public at the maximum point was calculated to be 8.6E-6 millirems /yr. The exposure calculation for release of liquid to the James River was based on the assumption that exposure was achieved by ingestion of fish caught and eaten at a rate of 10 kilograms each year. The exposure was calculated to be 5.3E-7 millirem / year.
12.9 RESPIRATORY PROTECTION The primary objective of a respiratory protection progran is to limit the inhalation of airborne radioactive materials and other hazardous materials. 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 progran 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 (s'
Materi al s."
j tj 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.
3.
Written procedures to ensure the adequate individual fitting of respirators, as well as procedures to ensure the testing of respiratory protective equipment for operability 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.
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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 physically able to perforn 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 0CCUPATIONAL EXPOSURE ANALYSIS 12.10.1 External Exposure - The external radiation exposure received by LRC employees is presented in Table 12-13.
The years 1981 through 1984 are presented. The row entitled "Off Site" gives the exposures received by LRC employees at other licensed facilities.
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TABLE 12-13 LRC RADIATION EXPOSURE 1984 1983 1982 1981 Total Person Rems 23.5 18.4 19.4 26.3 Off Site 3.5 2.0 2.5 3.0 LRC 20.0 16.4 16.9 23.3 Average Exposure 0.09 0.088
.105
.137 Number of Workers 220
?.08 184 192 Highest Exposure 2.25 2.04 1.9 1.7 The exposure received by LRC employees is categorized by group in Table 12-14 for exposures receivea for calendar years 1983 and 1984.
TABLE 12-14 EXPOSURE BY GROUP (PERSON REMS)
Group 1984 1983 Plant Engineering 5.10 2.25 Project Services 0.07 0.05 Health & Safety 1.85 2.16 Nuclear Materials 11.40 9.60 Chemical & Nuclear Engineering 1.30 1.53 Nondestructive f%thods 0.58 0.17 Process Control 0.00 0.49 Systems Design & Engineering 2.12 2.09 License No SNM-778 Docket No.70-824 Date October,1985 O
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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 12-13 reflects this in all categories.
Table 12-14 also reflects this increase in two of the three affected groups. Only Health and Safety saw a reduction in the group's exposure. The amount of exposure received from off-site work reversed a three year period of decreases. Table 12-14 reflects this in the increase in the Systems Design & Engineering Group's exposure.
The increases noted in the two tables 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 perfomed 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 radio-activity, the effectiveness of the ALARA progran can be appreci-ated. The preliminary exposure information required on the Radiation Work Permit form was increased in early 1985. This has resulted in many improvements in the manner that cell entries are made.
12.10.2 Internal Exposure - The bioassay sampling, lung counting, and air g) sampling programs show that the worker is exposed to extremely low tL' levels of respirable activity.
12.10.2.1 Bioassay Results - Urine bioassay samples are taken primarily of workers who perforn work with unclad uranium and those involved in any work with plutonium. Table 12-15 below presents the number of urine bioassay samples taken during 1983 and 1984.
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TABLE 12-15 j
i NUMBER OF URINE BI0 ASSAY SAMPLES 1983 1984 Month U
Pu U
Pu 20 4
January 5
13 6
February March 15 12 April 19 19 18 9
13 5
May June 16 16 17 8
July 11 8
15 6
August 10 8
15 7
September 11 14 14 7
October 11 9
16 5
November 3
1 December 5
5 14 6
In 1983, all samples for uranium were less than 5 ygrams/ liter (lower limit of detection), except on four occasions when the I,l\\
analysis indicated the presence of uranium but none met the C
resample limit of 20 p grams / liter. All plutonium analyses were below the minimum sensitivity which varied from 0.00 + 0.1 to 0.310.4 dpm per sample.
In 1984, all samples for uranium were less than 5 pgrans/ liter (lower limit of detection), except on one occasion 27 pg/ liter was reported. A resample showed that the level had returned below the lower limit of detection. All plutonium samples indicated 0.01 (0.01 to 0.6).
12.10.2.2 Air Sampling Results - The air sampling program is the first line of defense for all operations of this type, but the bio-assay program, along with lung counts, is the final step in the estimation of exposure that may occur.
12.10.2.2.1 Table 12-16 presents a summary of the air sampling program at the LRC for calendar year 1983, for fixed air samplers.
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TABLE 12-16 1983 AIR ACTIVITY (VALUES IN pCi/ml)
Approximate Maximun Labs Average Concentration MPC 15 3x10-15 1.2x10-14 1x10-10 16 3x10-15 1.5x10-14 1x10-10 17 8.7x10-15 1.8x10-13 4xio-11 19 7x10-15 8.7x10-14 1x10-10 27 2.4x10-15 5.7x10-15 1x10-10 44*
2x10-15 6.5x10-15 1x10-10 Cask Handling Area 1.9x10-12 1.27x10-10 9x10-9 6.7x10-15 4.5x10-13 4xin-11 Hot Cell 1x10-14 1.25x10-13 9x10-9 5x10-16 1.2x10-15 4x10-11
(
Recirculated Air "C"
1.5x10-14 3.5x10-13 9x10-9 Y
4x10-15 1.93x10-14 4x10-ll Waste Storage Area 1.5x10-14 2.6x10-14 9x10-9 7x10-16 1.7x10-15 4x10-11 Laundry 3x10-14 1.5x10-13 gxio-9 3x10-15 2.5x10-14 4x10-ll Radio Chem Lab 7x10-14 2.3x10-12 9x10-9 1.5x10-15 1.5x10-14 4x10-Il
- Discontinued in Sept.
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12.10.2.2.2 On 338 occasions in 1983, breathing zone air sanples were taken to neasure 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 (s) with plutonium glove boxes. All operations of this type N/
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 nay 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 activity.
5.
It should to be noted that a major operation is occurring in the decommissioning of Building C that is requiring the use of respiratory protection for industrial safety reasons, not for protection from radioactive materials. A number of operations are very dusty (paint chip,ing, License No SNM-778 Docket No.70-824 Date October,1985 0
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concrete destruction, etc.). A NIOSH approved full flow hard hat systen is used. With no protection factor, no one in Building C has been exposed in excess of 2 MPC hr in one week.
In most cases, radioactivity above back-ground is undetectable.
12.10.2.2.3 Table 12-17 presents a summary of the air sampling progran at the LRC for calendar year 1984, for fixed air samplers.
TABLE 12-17 1984 AIR ACTIVITY (VALUES IN pCi/ml)
Approximate Maximun Labs Average Concentration MPC 15*
(
,)
17*
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 i
Waste Storage 1.5E-14 2.9E-14 9E-9 1
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Laundry 3E-14 6.3E-14 9E-9 2E-15 3.3E-15 4E-11 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 measure the airborne activity to which workers were exposed.
In no case was anyone exposed to greater than 3 MPC hour of airborne activity in any one week.
In most cases, respiratory protection was used and exposure levels were at least a factor of 1000 below the limits.
3 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
( w/
i 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.
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 3f1 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 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.
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4 Other minor operations requiring respiratory protection are: changing of HEPA filters, repair work on NPD site support equipment, and any other operation where Health and Safety believes that there is a potential of airborne activity.
5.
It should to be noted that a major operation is occurring in the decommissioning of Building C that is requiring the use of respiratory protection for industrial safety reasons, not for protection from radioactive materials. A number of operations are very dusty (paint chipping, concrete destruction, etc.). A NIOSH approved full flow hard hat system is used. With no protection factor, no one in Building C has been exposed in excess of 2 MPC hours in one week.
In nost cases, radioactivity above background is undetectable.
12.10.2.3 In-vivo Results (1983) - Whole body counting was perforned 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 12-18.
m TABLE 12-18 l
I
'0 WHOLE BODY COUNTS 1983 (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 sunnarized in Table 12-19.
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TABLE 12-19 Am - Pu LUNG COUNTING 1983 (ALL VALUES IN NAN 0 CURIES)
Employee P_u Am u
1 0
0.0010.10 2
0 0.0010.11 3
0 0.1310.13 4
0 0.0010.14 5
0 0.0010.15 6
0 0.0010.19 7
0 0.0010.16 In-vivo lung counting was performed on nine employees in 1983, for uranium. The results are listed in Table 12-20.
Four of the nine indicated positive results. However, these results were not confirmed in followup urinalyses.
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i TABLE 12-20 v
URANIUM LUNG COUNTING 1983 (ALL VALUES IN MICROGRAMS)
Employee U-235 1
0130 2
0143 3
0139 4
42137 5
0141 6
38133 7
76145 8
0139 9
49144 x
12.10.2.4 In-vivo results (1984) - Whole body counting was performed by
{d Helgeson Scientific Services, Inc. on 99 employees during 1984.
Twelve had positive results but these were very low levels. A summary is presented in Table 12-21.
TABLE 12-21 WHOLE BODY COUNTS 1984 (EXPOSURE VALUES IN NAN 0 CURIES)
Number of Maximum Isotope Employees Observed MPBB Cs-134 1
3.0 2E4 Cs-137 7
9.0 3E4 Co-60 4
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-( V In-vivo lung counting was performed on 14 employees during 1984
)
for Plutonium-239 and Americiun-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 micrograns (137) for one person.
12.11 MEASURES TAKEN TO IMPLEMENT ALARA 12.11.1 Management Connitment - LRC management has made a commitment to maintaining exposures to radiation and radioactive material as low as reasonably achievable.
12.11.1.1 The commitment is reinforced in training and retraining sessions presented to Authorized Users and Radiation Workers.
It also appears as a policy statement in several procedures.
12.11.1.2 Exposure histories of employees are reviewed annually in a summary report prepared by the Supervisor, Health and Safety, reviewed by the Safety Review Committee and forwarded to the Di rector.
s f(")
12.11.1.3 Each RWP requires an estimate of the exposure dose that will be received by workers performing the described task. This estimate is used by the RWP reviewers in evaluating the protection requirements specified on the form.
12.11.1.4 The RWP form requires the preparation of an ALARA study for tasks where the estimated exposure approaches LRC self imposed limits.
12.11.2 Training - The following training prograns are presented to LRC employees.
12.11.2.1 General Radiation Protection Training The LRC provides two training programs covering the nature, use and control of radiation and radioactivity. These courses are presented to ensure that all LRC personnel receive training appropriate to their activities and to fulfill obligations under the NRC license to provide such training.
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The courses consist of a series of lectures intended to present the proper background and technical base to allow workers to understand the principles of radiation safety. The Health and Safety Group is responsible for the course content and, in general, teaches each course. Where practical, basic general procedures and federal regulations are included and discussed.
Training aids, such as motion pictures and self-study materials, are used as appropriate.
Program 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 radio-active 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 as an Authorized User.
Workers who are exposed to ionizing radiation are classified as radiation workers and will receive training commensurate with their exposure as required by Tilte 10, Code of Federal Regu-lations, Part 19 (10 CFR 19). This training will include n
(
)
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 the job requirements dictate.
12.11.2.2 Program I This course is available to new office employees and is presented to employees within 30 days of report;'q to work at the LRC.
It provides an introduction to radiation and radio-activity (understandable to the employee with no technical education or experience) and a thorough coverage of safety rules and procedures, including the site energency procedures.
Subjects include types of radiation, radiation effects on humans, permissible levels, basic health physics rules, a history of radiation protection, and personal hygiene.
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12.11.2.3 Program II New laboratory employees who work with radioactive materials are required to complete thic course and pass a written test.
Subjects include the following:
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
,o females and the embryo / fetus 1.
Radiation exposure compared to other hazards in the work place.
2.
Health Physics Instruments a.
Personnel monitoring devices b.
Cutie pie and Geiger-Mueller counter c.
Alpha survey meter d.
Air monitors e.
Criticality alarm system f.
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Instructions in field use of instruments.
3.
Regulations and Procedures a.
Code of Federal Regulations (including 10 CFR 19) b.
License requirements 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. How-ever, such personnel must pass the written exanination required for Program II.
12.11.2.4 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
[mU)
Health Physics individual, as defined in NUREG-0041 (Section 12.1), who will document that such training has 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 devices is maintained.
Training in respiratory protection shall include the following subjects:
1.
Discussion of the airborne contaminants present in the work environment including their physical properties, physio-logical actions, toxicity, means of detection, and maximum permissible concentrations (MPC's).
2.
Discussion of the importance of selecting the proper respirator based on the hazard and the dangers of using respirators for purposes other than that intended.
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3.
Discussion of the construction, operating principles, and limitations of the available respirators.
4.
Discussion of the use of engineering controls as a substi-tute for respiratory protection and the need to make every reasonable effort to reduce or eliminate the need for respiratory protection.
5.
Instruction in methods to be used to determine that the respirator is in proper working order.
6.
Instruction in fitting the respirator properly, field testing for proper fit, and factors that may influence a proper fit.
7.
Instructions in the proper use and maintenance of the respirator.
8.
Discussion of the uses of various cartridges and canisters available for air-purifying respirators.
9.
Review of radiation and contamination hazards, including a review of other protective equipment that may be used with respirators.
f]
10.
Instruction in emergency actions to be taken in the event V
of respirator malfunction.
- 11. Classroon instruction to recognize and cope with emergency situations while working with a respirator.
- 12. Any additional training as needed for special use.
13.
The wearer must pass a written examination on the material presented on respiratory protection.
12.11.3 ALARA Committee - The Safety Review Committee serves as the ALARA Comittee at the LRC.
This committee has the final signature authority for all Area Operating Procedures, it reviews and approves new projects, and reviews the annual summary of personnel exposures and environmental releases.
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12.11.4 Management Supervision - The Supervisor, Health and Safety is i
responsible for implementing the respiratory protection program.
The supervisor meets the qualifications as the " Radiation Pro-tection Supervisor" or " Radiation Safety Officer" as specified in ANSI /ANS-3.1-1981, proposed revision 2 to Regulatory Guide 1.8 and proposed Division 8 Regulatory Guide Task OP 722-4.
12.11.5 Authority - The authority to implement the ALARA policy at the LRC is vested in the Authorized Users.
Specific authority is vested in the Supervisor, Health and Safety, the Facility Supervisor, and the Safety Review Committee. The Supervisor, Health and Safety and the Facility Supervisor or their designated alternates have the authority to terminate any work that, in their judgment, does not conform with the LRC's ALARA policy. These two positions have signature authority for all Area Operating Procedures and Radi-ation Work Permits. The Safety Review Committee exercises its authority through its approval authority when reviewing Area Oper-ating Procedures and Projects.
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 (7
employees to be sampled in the program. This selection will be V
based on the probability of exposure, the employee's work habits, the type of work in the area, air sample data, previous bioassay data, etc. Routine bioassay may consist of check or whole-body counting (in-vivo bioassay) or excretion analysis (in-vitro bioassay).
In-vivo bioassay is performed routinely by a bioassay service which comes on-site for the evaluations.
In-vitro bioassay is performed by a commercial laboratory located off-site.
Bioassay action criteria for plutonium are outlined in Table 12-3 &
12-4.
In general, no action is required if the excretion result (i.e., urinalysis) is less than 0.2 dpn/ liter or the in-vivo measurement of material in the lung is less than 16 nanoCuries. All compounds of plutonium are considered to be either class W or Y.
This classification refers to the most recent evaluation of the ICRP for internal dose calculations. Class W compounds are moderately soluble and clear from the pulmonary region of the lung with half-times in the range 10 to 100 days. Class Y compounds are essentially insoluble and are considered to clear from the puimonary License No SNM-778 Docket No.70-824 Date October, 1985 0
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No compounds of plutontun are
~
considered by the ICRP to be readily soluble (i.e., class D compounds which clear from the lungs in <10 days).
The bioassay program for uranium generally follows that outlined in Regulatory Guide 8.11, " Application of Bioassay For Uranium," June 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 monf-toring.
2.
Bioassays of employees working in areas in which both plutonf un and uranium may be airborne shall be evaluated for both plutonium and uranium. The Supervisor, Health and Safety may decide to analyze for only one of these elements, if it can be demonstrated that the analysis for a single element is a more sensitive indicator of an uptake.
Bioassay action criteria for uranium are outlined in Table 12-5 &
12-6.
Employees working primarily with beta and gamma emitting radio-nuclides will also be included in the in-vivo bioassay analysis
[mV}
progran. Any employee suspected of an exposure greater than 40 MPC-hours will be scheduled for a bioassay evaluation as soon as practicable after the exposure. Bioassay action criteria for beta-gamma are outlined in table 12-7.
12.13 AIR SAMPLING AND MONITORING The presence of airborne radioactive materials in the working areas 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 left to the discretion of the Supervisor, Health and Safety.
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12.13.1.1 Fixed Air Samplers - Air samples are obtained at designated points through the use of a central vacuum system. Sampling points are located as close as possible to a permanent operator station to permit continuous sampling of the air near the worker's breathing zone. These samples are usually collected weekly. However, the frequency may vary as the situation dictates.
Normally, these are evaluated within two weeks, 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 for the decay period.
In these cases, an applicable radon decay correction factor must be applied to the results.
12.13.1.2 portable Sanplers - 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 systen. However, since the sampler is located near 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.
(O Air samples obtained with these samplers are evaluated on a low background, proportional counting system.
Factors are applied to the counting results to account for background activity and detector efficiency. All results are reported in units of activity / unit volume of air sampled.
12.13.2 Air Monitoring Program Air monitoring in operating areas of the LRC is accomplished with continuous nonitors 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-activity. Locations are selected based upon the ability of the monitor to provide a reasonable evaluation of the airborne activity in a particular area and to provide adequate warnings to those in the area of changing conditions. These determinations are made by the Health and Safety Group based upon the operations in the area, the potential for release, and the quantity and chemical form of the material.
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Alarns are set in accordance with the particular operation, the naterial 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.
12.14 SURFACE CONTAMINATION 12.14.1 Snear Surveying Snear 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 contanination.
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 Snear Samples - Smear samples are obtained with small, absorbent filter papers. The snear paper is moved across an area of
,o approximately 100 sq. cm. using about 5 pounds of pressure. The
(l snear may be counted with a portable gas-flow proportional counter capable of detecting alpha or beta radiation.
- Normally, smear samples are evaluated in a stationary counter located in the Health Physics Laboratory. Appropriate conversion factors are applied to the net counts to express the smear results in units of disintegrations per minute.
12.14.1.2 Large Area Snears - Large area snears 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 contani-nation environment in an area and may lead to a more extensive survey, if unexpected contamination is indicated.
Normally,
large area smears are evaluated with a hand-held, portable survey instrument (e.g., a gas-flow proportional counter such as the PAC 4G). Actions to be taken in response to the results of large area smears are outlined in Table 12-22.
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12.14.1.3 Action Levels - Included in Table 12-23 are the appropriate action levels to be used in designated areas of the LRC.
Decon-tanination 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 contaninated area may be marked and posted appropriately.
Such a determination shall be nade 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 after discovery.
TABLE 12-22 ACTION LEVELS FOR LARGE AREA SMEARS b,
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 dpn per 100 sq. ft.
2.
Routine Large Area Smears (5000 - 10,000 dpm)
Repeat the large area smear.
If results show levels of contamination above 5000 dpm, isolate the contaminated area.
Take smears in smaller areas to locate the source.
Decontaminate all areas in which the snear results show contamination in excess of 1000 dpm per 100 sq. ft.
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3.
Routine Large Area Smears (>10,600.dpm)
Isolate the contaninated 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 contamination in excess of 1000 dpm per 100 sq. ft.
Survey all persons leaving the building.
NOTE:
Routine large area smears are normally taken in the early after-noon to facilitate clean-up of areas found to be contaminated before the end of the normal work-day.
TABLE 12-23 rw SMEAR SURVEY FREQUENCIES AND ACTION LEVELS t
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V Alpha Radiation Smear Survey Action Level 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 October,1985 rw 0
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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 8
Radiochemistry Laboratory twice monthly 22,000 Exit Portals From Controlled twice monthly 2,000 Areas g
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12.14.2 Direct Radiation Surveys 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 inicude 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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C 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 12-24.
TABLE 12-24 CONTAMINATION ACTION LEVELS Transferable Surface Type of Fixed Contamination
. Area Radiation Surface Reading (dpm/100 sq. cm.)
Uncontrolled Al pha 300 dpm/100 sq. cm.
30 Beta-Gamma 0.1 mrad /h 220 77 Contanination*
Al pha 3000 dpm/100 sq. cm.
2,200' Beta-Gamma 1.0 mrad /h**
22,000
- 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 nomally radiation areas.
This level of contanination will not cause a sig-nificant iacrease in radiation exposure.
NOTE:
This table provides limits above which decontamination must be initiated.
These action levels pertain to areas nomally accessible to personnel perfoming nomal 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.70-824 Date October,1985 O
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17.14.3 ' Personnel Contamination Surveys ti.
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- Personnel are required to monitor themselves for activity present h'-
on their. hands, shoes, clothing, and person before exiting a con-
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't tamnation area. Contamination monitors (friskers) are located at S
[ all exits"from contamination areas for this purpose. The detector T
j' (probe) shored be held as close to the surface of the item being monitored 45' possible (without touch';.ng 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 arf given in Tables 12-11 and 12-12. Any contamination in excess s
of these limits should be reported immediately'to the Health and Safety Group.
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O TABLE OF CONTENTS Section Page 13.0 ENVIRONMENTAL SAFETY.
13-1 13.1 ENVIR0' MENTAL MONITORING 13-1 13.2 EFFLUENT AIR MONITORING.
13-1 13.3 LIQUID EFFLUENT MONITORING.
13-2 4
License No SNM 778 Docket No.70-824 Date October, 1985
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13.0, ENVIRONMENTAL SAFETY t
13.h ENVIRONMENTAL > MONITORING r
Envi'ronmental sampling of the area surrounding the LRC is performed x
on a regular basf s to evaluate changes in the levels of radioactivity in air, water, and vegetation. The minimun environmental program consists of the following.
e one continuous on-site background air sample e monthly water samples f-om the James River' collected above and below the liquid discharge point s
e con,tinuous sampling of rain water on-site e quarterly sanples of river silt and near-river vegetation.
I 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 commercisl laboratory.
Records of the results of these analyses will be mai.ntained by the Health and Safety Group.
V 13.2 EFFLUENT AIR MONITORING Potentially contaminated ' air from chemical hoods, hot cells, and glove boxes is di'scharged ultimately through the 50-meter stack.
Generally, exhaust air containing beta-ganma 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 thro. ugh a two-stage HEPA filter.
Discharge through the stack is accomplished with a large blower, powered normally by a large electric motor operated on off-site power. Emergency power is suppl,ied by an internal combustion engine coupled to the bl~ower shaft through a centrifugal clutch. On loss of off-site power, the e'ngine starts automatically and takes over the load upon reaching the proper speed.
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Discharges through the stack are monitored with a sampling head
\\d 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 nonitor, 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 nonitor 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.
Prior approval of the Safety Review Committee is required for both of the above situ-
' ations, i
13.3 LIQUID EFFLUENT MONITORING All potentially radioa:tive liquids are collected in tanks located in f]
the Liquid Waste Disposal Facility. The contents of each tank are
\\d mixed, samples are obtained, 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 evaluation 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.
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U('T TABLE OF CONTENTS Section Page 14.0 NUCLEAR CRITICALITY SAFETY.
14-1 14.1 ADMINISTRATIVE AND TECHNICAL PROCEDURES 14-1 14.2 PREFERRED APPROACH TO DESIGN 14-2 14.3 BASIC ASSUMPTIONS.
14-2 14.3.1 Nuclear Isolation.
14-2 14.3.2 Building A 14-3 14.3.3 Building B 14-7 14.3.4 Building C 14-16 14.3.5 Outside Storage 14-16 14.3.6 Dry Waste 14-16 14.4 ANALYTICAL METHODS AND VALIDATION REFERENCES.
14-17 14.5 DATA SOURCES 14-17 14.6 FIXED POIS0NS 14-18 14.7 STRUCTURAL INTEGRITY.
14-18 14.8 SPECIAL CONTROLS 14-18 License No SNM-778 Docket No.70-824 Date October, 1985 b
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/^N tVl List of Tables Table Page 14-1 Keff TID-7016 14-4 14-2 Keff FOR PRESENT LIMIT 14-5 14-3 Keff FOR 6x6x6 ARRAY OF 850 g U-235 UNITS ON 30-INCH CENTERS.
14-5 14-4 Keff FOR ARRAYS OF 850 GRAM U-235 UNITS ON 24 AND 36 INCH CENTERS 14-6 14-5 COMPARIS0N OF THE MARK B AND MARK C FUEL ASSEMBLIES.
14-9 14 fi REACTIVITY FOR MARK B AND MARK C FUEL ASSEMBLIES UNDER DISMANTLEMENT 14-13 14-7 K-EFFECTIVE OF INDIVIDUAL MARK B AND MARK C FUEL ASSEMBLIES.
14-13 14-8 REACTIVITY FOR AN INFINITE BY 14-UNIT ARRAY V) 0F FUEL ASSEMBLIES.
14-14 List of Figures Fi gure Page 14-1 FUEL R0D REMOVAL SCHEMATIC - MARK B.
14-19 14-2 FUEL R0D REMOVAL SCHEMATIC - MARK C.
14-20 License No SNM-778 Docket No.70-824 Date October, 1985 0
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14.0 NUCLEAR CRITICALITY SAFETY 14.1 ADMINISTRATIVE AND TECHNICAL PROCEDURES The ultimate responsibility for nuclear safety rests with the Di rector.
However, first-line responsibility is with the Facility Supervisor supported by the Nuclear Safety Officer.
The Nuclear Safety Officer is generally responsible for establishing nuclear safety limits and nuclear safety considerations in operating procedures, processes, and the like. His duties are shown more spe-cifically in the following statement.
The position of Nuclear Safety Officer has been established at the Lynchburg Research Center.
It will be this officer's responsibility to ensure, as far as possible, that no operations in the Lynchburg Research Center can lead to the inadvertent assembly of a critical
. mass. To this end, he will review all new procedures which involve the handling of special nuclear materials as well as changes in old procedures, observe operations, inaugurate educational programs if and when he deems them necessary, and carry out confirming criti-cality calculations.
p This appointment does not in any way relieve the Facility Supervisor
(~j of his responsibilities for ensuring the safety of operations, nor will it eliminate the necessity for the reviews by the Safety Review Committee required by the license.
Once a quarter the Nuclear Safety Officer or qualified person designated by him will inspect all LRC operations where special nuclear materials are being processed. Other areas shall be inspected less frequently; however, all areas shall be inspected at least once a year. He shall consider area operations when schedul-ing these inspections and shall, if necessary, schedule his inspection at more frequent intervals. His consideration should include inspection of new facilities, inspection of hazardous non-routine operations, an audit of nuclear safety records, a check for area posting and a review of current practices.
A written report is to be filed with the Director quarterly with a copy to the LRC License Administrator. Prior to submission of the report, he shall discuss any findings with the Facility Supervisor.
The report shall be brief, concerning itself with inspections made during the quarter and with the nuclear safety activity of the quarter.
License No SNM-778 Docket No.70-824 Date October, 1985
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The following information is to be included:
a e Areas visited e Operations observed e Unsafe practices or situations noted e Nuclear safety activity of the quarter (brief summary) e Recommendations e Resolution of previous recommendations.
14.2 PREFERRED APPROACH TO DESIGN The LRC is a research and development organization. While the use of
, safe geometry is the preferred approach in a production facility, it is not appropriate or practical at a research laboratory.
Since most projects require only small amounts of SNM on laboratory benches and in hoods, the preferred approach at the LRC is through safe masses in simple arrays; the lattice density model or arrays found in TID-7016, Rev.1 is the adopted nodel.
The one exception to use of safe masses is when examining and testing reactor fuel assemblies. The LRC's bsj approach to such uses is to accept only a limited number of fuel V
assemblies and then to maintain the fuel of an assembly within the dimensional envelope of the original assembly's dimensions. Where this is not possible, the fuel of an assembly is handled within the dimensions of safe geometry or as a safe mass.
14.3 BASIC ASSUMPTIONS This section describes basic assumptions and evaluations that have been nade to demonstrate nuclear criticality safety for the speci-fications of Section 4.2 (Technical Requirements for Nuclear Criti-cality Safety).
14.3.1 Nuclear Isolation - Special nuclear material at the LRC is isolated from all other special nuclear material for nuclear criticality safety purposes if any of the three conditions (or equivalent) listed in 4.2.1 are met. These three isolation criteria are accepted industrial practice for maintaining nuclear criticality l
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It is recognized that 12 inches of high density concrete may not be adequate as isolation between two large parallel. slabs of SNM; this does not describe any SNM configuration at the LRC and will not be permitted without additional evaluation and NRC approval.
14.3.2 Building A 14.3.2.1 General - From Figure 22, TID-7016, Revision 1, 74 units is read as the maximum allowable number of units in a cubic array on 24 inch centers (with at least 8 inches edge-to-edge between units),
assuming full reflection on the array. The nunber of allowable units has been reduced from 74 to 40 units to permit use of the 850 grams of U-235 at low enrichment as a unit. The subdivisions defining a unit are for clarification of the general definition of a unit as any physically identifiable accumulation of SNM.
The terninology of TID-7016, Revision 1, applies.
14.3,2.2 Mass Limits 1.
The mass limits for plutonium, U-233, and U-235 are based on the recommended limits in Table I, TID-7016 (Rev.1). The values for Pu-235-U mixtures in 4.2.2.2.1 were derived to satisfy the following relationship:
(~'N grams Pu fissile grams U-235 220 350
-<1 2.
The values for U-233 - Pu and U-233 - U-235 nixtures were found by taking the lowest limit of any isotope in the mixture.
3.
From DP-1014, uranium metal-water lattices which have the minimum U-235 mass at critical are 2.36 kg for 3.0 wt% U-235 and 1.47 kg for 5.0 wt% U-235.
A conservative interpolation between these two points gives 1.9 kg at 4.0 wt% U-235; 45%
of this is 850 g U-235.
The present array control is based on the lattice density model using Figure 22 and Table IV (modified) in TID-7016 (Rev.1). Our calculations demon-strate that the 850 gram unit is an allowable unit if the number of units permitted in TID-7016 (Rev. 1) is set at 40.
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All computer calculations were made using either the NULIF code for fully reflected spheres or with the Monte Carlo code KEN 0.
Four series of computer calculations were made. Tables 14-1 and 14-3 summarize the results.
14-1 Table for determination of Keff for the mass limits listed in TID-7016 for lattice density model at the upper H/X limit (made with NULIF).
14-2 Table for determination of Keff Vs H/X for 349 grams of U-235 contained in fully enriched uranium metal (made with NULIF).
The Keff values for the lattice density limits ranged from 0.800 to 0.854 and are tabulated in the following table.
TABLE 14-1 Keff TID-7016 p.
Sphere (j
- Mass, Radius, g
g eff Kg U-235 cm H/U-235 H/ Total U
=
10.0 6.828 2.0 1.87 1.86 Q.800 9.0 7.182 3.0 2.81 1.84 0.804 7.3 7.562 5.0 4.68 1.82 0.803 5.2 8.178 10.0 9.36 1,81 0.815 3.6 8.922 20.0 18.71 1.84 0.854 350) eff values for the present limit (349 was used instead of The K Vs H/X are given in the following table.
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TABLE 14-2 Keff FOR PRESENT LIMIT Sphere Mass
- Radius, K
Keff g U-235 cm H/U-235 H/ Total U
=
349 13.32 736.8 689 1.49 0.780 349 9.82 293.8 275 1.76 0.753 349 7.79 146.2 137 1.86 0.671 By anology, values for Pu would be similar.
The effect of interspersed water noderation in a concrete reflected finite array of 850 gram U-235 units is shown in Table 14-3.
These data show that maximum array multiplication occurs with almost no interspersed water,
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TABLE 14-3 Keff FOR 6x6x6 ARRAY OF 850 g U-235 UNITS ON 30-INCH CENTERS
( y = 18.14 cm, p = 0.85 g U/cc)
Vol ume K
Fraction H2O eff 2o 1.00 0.833 1 0.010 0.15 0.826 1 0.010 0.10 0.851 1 0.010 License No SNM-778 Docket No.70-824 Date October, 1985 O
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0.07 0.868 1 0.009 0.05 0.907 1 0.011 0.03 0.924 1 0.009 0.02 0.931 1 0.009 0.01 0.937 1 0.009 0.001 0.930 1 0.009 The effect of varying the number of 850 U-235 units in a concrete reflected array while maintaining a constant center-to-center spacing with void between them is shown in Table 14-3 (b).
In-terpolating between the heterogeneous values for the 24 inch center-to-center systen predicts a Keff 1 2 o for 40 units of 0."36 1 0.12 whereas the 36 inch spacing systen has about 512 units for the same Keff.
TABLE 14-4 Keff FOR ARRAYS OF 850 GRAM U-235 UNITS ON
's 24 AND 36 INCH CENTERS (V
Array Number Center-to-Center K
Size of Units eff 2a Spacing, in.
4x3x3 36 0.910 1 0.010 24 4x3x3 36 0.929 1 0.013*
24 4x4x3 48 0.932 1 0.010 24 4x4x3 48 0.947 1 0.011*
24 4x4x4 64 0.949 1 0.011 24 4x3x3 36 0.807 1 0.011 36 4x4x4 64 0.845 1 0.011 36 5x5x5 125 0.860 1 0.010 36 License No SNM-778 Docket No.70-824 Date October,1985 O
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6x6x6 216 0.881 + 0.009 36 7x7x7 343 0.912 1 0.009 36 8x8x8 512 0.938 1 0.008 36 9x9x9 729 0.949 + 0.009 36
- Assumes heterogeneous U0 -water mixture.
2 From these data it is concluded that for 850 grams U-235 per unit an array of 24-inch centers should be safe for 40 units or less and on 36-inch centers, an array would be safe with 90 units or less.
A slight increase in the array multiplication, on the order of 1%, may occur for low levels of interspersed water moderation.
However, the safety of these arrays would still be maintained.
4 To avoid confusion and possible nistakes, additional procedural controls are applied when low-enrichment limits are used.
These preclude enrichment combinations of below and above 4.0 wt%
(These are not necessarily unsafe - no calculations were (3
made and no such combinations are desired.)
4.
The unit and its limit (laboratory, furnace, transfer cart, etc.) are established by the Facility Supervisor, who author-izes posting the limit showing the maximun quantity of plutonium, U-233, and U-235 allowed. The fissile material content of the material transferred to or from a unit is established from process records, analyses, or previous analytical data.
Only authorized users of SNM may transfer SNM between units and must do so only according to approved procedures.
A board, sign, or other acceptable device is used to record the new balance and compares to balance with the unit limit.
14.3.3 Building B 14.3.3.1 General - The demonstration for units and the array is identical to that of Building A (14.3.2.1 & 14.3.2.2).
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14.3.3.2 Hot Cell - The demonstration for the units and array is identical
'd to that of Building A.
The individual hot cells are isolated from all other arrays by a ninirium of 2 feet of high density concrete.
14.3.3.3 Underwater Storage - Transfer Canal - Underwater aluminun or stainless steel storage racks are constructed to ensure 12-inch edge-to-edge spacing of each unit.
Units are limited to those in 4.2.2.2.1 & 4.2.2.2.2 excluding PWR fuel assemblies and, since they are separated by 12 inches of water, units are considered isolated. Therefore, any number of these units may be used at the LRC.
Racks and fixtures are constructed with sufficient integrity and strength to withstand reasonable structural deformity, thereby providing the spacing previously outlined.
Supervisory approval is required for removing or inserting any subcritical unit out of or into its storage rack.
There is no credible way in which water can be lost from the storage pool and transfer canal.
However, assuming loss of water, stored units would drain and be unmoderated and sub-critical.
14.3.3.4 Underground Storage Tubes - Underground storage tubes are 5 O
inches in diameter, approximately 10 feet long, and on 17 inch V
centers (minimum) in a straight line. Material stored is first placed in a storage can with an inside diameter of 4-1/2 inches.
Maximun units demonstrated safe in Section 14.3.3.2 are stored one per tube. These are nuclearly isolated from each other by 12 inches of concrete (mininun). The average edge-to-edge separation approximates 13 inches of concrete.
14.3.3.5 Power Reactor Fuel Assemblies 14.3.3.5.1 General - The LRC will receive and examine PWR fuel assemblies for both nondestructive and destructive examination.
Irradiated assemblies will have been subjected to a reactor environment. From a nuclear criticality safety viewpoint, these assemblies are in their most reactive state when fresh or unirradiated. Therefore, nuclear safety is demonstrated by appropriate evaluation of the unirradiated assembly. The LRC's current plans call for examination of B&W-manufactured fuel assemblies from B&W power reactors. The current models of License No SNM-778 Docket No.70-824 Date October,1985 O
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interest are designated as the Mark B and Mark C canless assembly. The Mark B assembly is described in the SNM license for B&W's Commercial Nuclear Fuel Plant (SNM License No.1168, Docket 70-1201).
In 7.10 of Section III in SNM License 1168, the Keff of the unrodded and fully moderated and reflected assembly is shown to be 0.92 at maximum enrichment. Maximun enrichment is defined as 4.0 percent nominal which could go to 4.05 percent in manufacturing. Table 14-5 shows a comparison of the Mark C and Mark 3 assemblies. The Keff of the Mark C assembly under the same conditions listed above has a value of 0.92.
The reactivity as well as the spectral and physics kinetics of these assemblies are essentially the same. All of the nuclear safety calculations shown in this section were made with the Mark B assembly model (except Tables 14-5 and 14-6).
Results were obtained for a fully reflected infinite array 12 inch edge-to-edge of maximumly enriched assemblies that were fully moderated, i.e., under water.
The Mark R and C assemblies are to be disassembled in air only in an unirradi-ated state. The similarity in nuclear characteristics and the large decrease in reactivity in air-moderated assemblies ensure nuclear safety during the disassembly operations.
Conditions given in 4.2.3.6.1 are sufficient to ensure that these two assembly types are indeed those to be examined. A damaged assembly which is restrained to 8.6 inches on a side will be no more reactive in air or water even if part of the fuel is (m'v) missing; this will be demonstrated in Section 14.3.3.5.3.
TABLE 14-5 COMPARISON OF THE MARK B AND MARK C FUEL ASSEMBLIES Mark B Mark C Fuel assembly array 15 x 15 17 x 17 Fuel assembly dimensions, in.
8.45 x 8.45 8.536 x 8.536 Control rod tubes per assembly 16 24 Instrument tube per assembly 1
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Fuel rod pitch, in.
0.568 0.501 Fuel active height, in.
144 143 Pellet OD, in.
0.370 0.324 Theoretical density, %
92.5 94.0 Enrichment, %
4.0 4.0 Fuel rod clad ID, in.
0.377 0.332 Fuel rod clad OD, in.
0.430 0.379 Fuel rod clad material Zr-4 Zr-4 Vwater/V uel in fuel rod cell 1.65 1.68 f
Vwater/V uel in assembly with f
water completely filling control rod and instrument cells 1.90 1.98 Keff of one assembly in H 0 0.92 0.92 2
Since The Babcock & Wilcox Company is continuing to improve its n
assemblies and will supply reload fuel to reactors initially
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fueled by other reactor manufacturers, the LRC may destruc-tively examine other types of assemblies. The conditions given in 4.2.3.6.1.1 for additional evaluation are adequate to ensure nuclear safety for different assemblies.
Acceptance of BWR fuel assemblies at the LRC for study is acceptable if the assemblies have a maximum enrichnent of 4.05 wt% U-235 and have a cross sectional area not exceeding that of a 22.5 cm diameter cylinder.
By reference to DP-1014 this is 90% of the ninimun critical cylinder diameter for an infinitely long, water reflected, optimally moderated cylinder with four wt% enriched hetrogeneous U0.
This value is further supported 2
by Figure 2 (page 10) in ANSI /ANS 8.1-1983 and Figure 2.15 (page 44) in TID-7016, Rev. 2.
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14.3.3.5.2 Receipt and Storage A. - Unirradiated Assemblies - Unirradiated fuel assemblies may be stored in their shipping containers since their nuclear safety has been proven prior to their licensing.
Assemblies that are unirradiated may also be stored in air if the distance between assemblies is no less than 21 by 38 inches.
(Refer to SNM-1168, Docket 70-1201, Section 3, page 173, dated 2/27/81). This distance assures criti-cality safety for less than 100 assemblies of either the Mark B and/or Mark C assembly types. This ensures the safety of the maximua of four assemblies stored on site.
Unirradiated assemblies may also be stored under water (hot cell pool, mock-up pool, or developnent test area pool).
Assemblies stored in air will be stored either:
1.
Horizontally - on the floor or on tables constructed with sufficient integrity and strength to withstand reasonable structural deformity as assuring the above mentioned spacing.
2.
Vertically - in racks and fixtures constructed with f)-
sufficient integrity and strength to withstand reasonable structural deformity and assuring the above V
mentioned spacing.
Supervisory approval is required to move any other fissile material into the area where the assemblies are stored.
No more than four unirradiated assemblies may be stored at the LRC at once. The limit of four assemblies is an arbitrary limit which the LRC imposes upon itself and does not affect nuclear safety.
Partially disassembled unirradiated Mark B or Mark C assemblies may also be stored in air. This is safe due to the lower moderation characteristics of air compared to water. Air moderated values of Keff will be less than those shown in Table 14-6.
Fuel rods from unirradiated, disassembled Mark B or Mark C assemblies will be stored in air in slabs not to exceed 4 inches in height (see Section 14.3.3.5.4, statement 2).
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Irradiated Fuel Assemblies - Assemblies which have been
'V irradiated may aiso be stored in their shipping containers or in the hot cell pool.
Storage in the hot cell pool is limited to foui irradiated assemblies.
The limit of four assemblies in tne pool is an arbitrary limit which the LRC imposes upon itself and does not affect nuclear safety since each fuel assembly or rod storage position is neutronically isolated from any other fissile material by a minimun of 1 foot of water.
Racks and fixtures in the pool are constructed with sufficient integrity and strength to withstand reasonable structural defornity, thereby providing the spacing previously outlined. The racks are also constructed to preclude inadvertently placing other fissile material closer than the 1-foot ninimun spacing.
Supervisor approval is required for removing or inserting fissile material into or out of any of the racks or fixtures.
Storage of Mark B and Mark C fuel rods and partially dismantled assemblies into storage racks which restrain the size of each position to a square not exceeding the dimensions of a fresh fuel assembly, i.e., 8.6 inches, is safe based upon the analysis demonstrating safety of as assembly during dismantlement.
Fuel rods may also be stored in an ever safe cross sectional area fixture, i.e., a cross sectional area not exceeding
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that of a 22.5 cn diameter cylinder.
U 14.3.3.5.3 Work Area Of Pool Under Hot Cell No.1 - This area will be used to dismantle irradiated and unirradiated assemblies.
Nuclear criticality safety for Mark B and Mark C assemblies under varying stages of dismantlement has been demonstrated via use of NULIF and PDQ-07 physics computer codes.
Reactivity was calculated by PDQ (coefficients having been generated by NULIF) for a fully reflected and flooded, unrodded fresh assembly and for the same assembly under five conditions of dismantlement. The cases run with the number of rods renoved in each case and the resulting Keff is given in Table 14-6.
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REACTIVITY FOR MARK B AND f1 ARK C FUEL ASSEMBLIES UNDER DISMANTLEMENT Calculated Keff No. of Removed Case No.
Rods Mark B Mark C 1
0 0.891 0.921 2
4 0.894 0.920 3
12 0.897 0.919 4
24 0.898 0.922 5
36 0.896 0.917 6
8 0.888 0.918 Reactivity was also calculated by KENO-IV using the 123-group XSDRN cross section set for a fresh assembly fully submerged in water under conditions of 24 rods removed and with all instru-nent and control rod guide tube positions loaded with fuel rods. The cases run with the number of rods removed or added and the resulting K-effectives are given in Table 14-7.
A TABLE 14-7 K-EFFECTIVE OF INDIVIOVAL MARK B AND MARK C FUEL ASSEMBLIES Change in No.
K-effective + 2o Of Fuel Rods From Normal Mark B Mark C 0
.895 +.010
.900 +.013
-24*
.900 +.015
.906 +.014
+17
.890 T.015
+25 876 +.016
- Sane configurations as case 4 in Table.14-5.
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the assembly while Case 6 represents the removal of eight rods
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Cases 2 through 5 represent removal of rods "unifornly" through clustered about the center. Rods removed are shown sche-natically in Figures 14-1 and 14-2 for Mark B and C assemblies, respectively. The calculations reported above demonstrate the nuclear safety of an assembly under various conditions of dis-assembly and reloading.
If any of the fuel rods inserted into the fuel assembly are further encased in metal tubing, the assembly would still be safe due to the tubing displacing noderator with absorber. A grouping of 75 fuel rods confined within a 8.6-inch square merely describes a dismantled assembly and is also safe.
Fuel rods inserted into instrument and control rod guide tubes shall be held in place with a flat metal plate which shall be bolted to the top of the assembly.
The safety of withdrawing an assembly and its associated rod storage position partially into the cell is demonstrated safe by comparison to a series of KENO runs made for pool storage at a reactor site.
To demonstrate the safety of flooding a reactor site storage pool filled with fresh Mark B fuel assemblies, an array of fuel assemblies 14 units wide, infinitely long, and reflected on the sides and bottom by concrete was calculated by KEN 0.
Each assembly was spaced 1 foot from the other on the concrete reflector, as appropriate.
Four cases at different degrees of pool flooding were evaluated
,s V) and are described in Table 14-8.
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TABLE 14-8 REACTIVITY FOR AN INFINITE BY 14-UNIT ARRAY OF FUEL ASSEMBLIES Calculated K Water Height eff Fully Flooded 0.951 + 0.006 3/4 0.946 + 0.007 1/2 0.928 T 0.007 0 (dry) 0.506}0.004 License No SNM-778 Docket No. 70 824 Date October,1985 O
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The confidence levels quoted above are one standard deviation.
Keff for the fully flooded condition is higher than that calcu-lated by PDQ because of simplifications made in running the cases. The series of runs were to demonstrate safety of a partially flooded pool, a much more restrictive condition than partial withdrawal into one cell. The similarity in the Mark B and Mark C nuclear characteristics and the simplifying assump-tions assure these calculations are also valid for the Mark C assembly type.
14.3.3.5.4 Assembly and Machine Shop and Development Test Areas - As-semblies of either Mark B or C disassembled in air are far less reactive than the cases listed in Table 14-6.
Either assembly type may be disassembled in air. A safe reactivity level (u0.95) is assured provided the handling in Section 4.2.3.6.4 is followed. The conditions stated in Section 4.2.3.6.4 are based on KEN 0 calculations that show:
1.
Two assemblies in air 21 inches or more apart are nuclearly safe.
2.
Fuel pins at a maximum enrichment when optimanly moderated are fully reflected in an infinite slab have a Keff = 0.95 if the slab is no more than 4 inches thick.
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3.
Fuel rods in any configuration or number, up to the number V
in the assembly, when limited to the confines of the assembly size are no more reactive than the intact assembly (Ref. Table 14-6).
14.3.3.5.5 Hot Cell Operations - Work within the hot cell will, by and large, follow existing controls. Three units in addition to an assembly and its associated rod storage position are permitted within Cell No.1.
Two of the three units are restricted to rods confined within an ever safe cross sectional area, i.e., a cross sectional area not exceeding that of a 22.5 cm diameter cylinder; in addition these two units must be free draining of any water.
The third unit of Cell No.1 under mass control is permitted.
All other Hot Cells are limited to one unit each.
14.3.3.5.6 Fuel Rod Dismantlement - Fuel rods of either assembly type may be dismantled.
Dismantlement can be performed in any area which present licensing conditions permit fuel handling.
In addition, mass control must be limited to 350 grams of U-235, proper spacing must be maintained, and approved procedares must be followed.
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14.3.3.5.7 Shipment and Disposal - The conditions of 4.2.3.6.7 are consis-x_
tent with the above demonstration and/or current limits.
14.3.4 Building C The demonstration for units and the array is similar to that of Building A (14.3.2.1 and 14.3.2.2). The values of all units in Building C are less than or equal to the value of the maximum storage unit defined in Table IV, TID-7016, Revision 1 (as amended), or they have been evaluated above in Section 14.3.2.2.
The allowable number of units on 36-inch centers is 90 units with at least 8 inches edge-to-edge between units.
The allowable number of units according to Figure 22 of TID-7016, Revision 1, is about 190.
The number of units has been reduced to 90 to permit the low enriched units. Administrative procedures for posting and con-trolling transfers of SNM to and from units are those described in 14.3.2.2.4.
14.3.,5 Outside Storage 14.3.5.1 General - The underground storage and shipments are nuclearly isolated by distance or matter.
14.3.5.2 Underground Storage - The underground storage tubes are 5 inches in diameter, approximately 20 feet long and 20-inch centers.
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Maximum units demonstrated safe in Section 14.3.2.2 are stored, i j one per tube.
These are neutronically isolated from each other by 15 inches of concrete.
14.3.6 Dry Waste Nuclear criticality safety of dry waste is ensured by maintaining the concentration of SNM to a value much less than an ever safe concentration. Forty-five grams of SNM in a 55-gallon drum yields a concentration of less than 0.25 g/ liter. These low concentra-tions are guaranteed by the nature of the material being stored which is contaminated laboratory waste.
The nature of the waste as borne out by more than 20 years of experience will maintain an approximate uniform dispersion within the container. Dry waste containers are stored in the radioactive waste building after gamma scanning to ensure that the maximum SNM is not exceeded. There is therefore no requirement in the number or arrangement of containers within the radioactive waste building.
One dimensional transport calculation shows that, at a U-235 concentration of 0.25 g/ liter License No SNM 778 Docket No.70-824 Date October,1985 0
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h with optimun water moderation, a fully concrete reflected sphere having the same volume as 8 x 105 55-gallon drums has a neutron V
multiplication of < 0.95.
Therefore, the 45 grams of U-235 per drum limit is safe'~in that the maximun number of drums on site can-5 not credibly exceed 8 x 10,
14.4 ANALYTICAL METHODS AND VALIDATION REFERENCES Nuclear criticality safety computer calculatior s presented in this chapter have used the computer codes NULIF, PDQ-07 and/or KEN 0.
The physics codes NULIF and PDQ-07 are not only routinely used in nuclear criticality safety to evaluate highly moderated low-enriched systems but also are the standard codes used by the reactor design group of the Babcock & Wilcox Company (both codes have been certified by the Company's Quality Assurance Program for reactor calculations). The Monte Carlo code KEN 0 is state-of-the-art in industry for nuclear criticality safety evaluations. These three computer codes are des-
, cribed in Appendix A,Section III, pages 3 through 12 of SNM License No. 1168 (Docket 70-1201); validation for these codes are given in Appendix A,Section III, of the same document on pages 19 through 21.
Future calculations for nuclear criticality safety at the LRC will make use of these codes and the Nuclear Criticality Safety Codes in SCALE 3 (NITAWL-S, XSDRNPM-S, KENO-IVS and KENO-Va).
SCALE 3 is described in NUREG/CR-0200.
Before use of the SCALE 3 package, the proper wording of the various codes will be assured and appropriate v'
benchmarking activity will be carried out.
14.5 DATA SOURCES Data and Guidance for Nuclear Criticality Safety at the LRC is taken from one or more of the sources specified below.
1.
Calculations using nethods described in Section 14.4.
2.
" Nuclear Safety Guide, TID-7016, Revision 2,"
NUREG/CR-0095 (0RNL/NUREG/CS0-6),(June,1978.)
3.
" Nuclear Safety Guide, TID-7016, Revision 1," (1961). TID-7016, Revision 1 is used only for application of the lattices density method which is Table IV and Figure 22 on page 26. Table IV has been modified according to information published in the Federal Register, March 5,1963 on page 2130.
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4.
American National Standard for Nuclear Criticality Safety in Operations with Fissionable Materials 0xide Reactors, ANSI /ANS-8.1-1983).
5.
H. K. Clark, " Critical and Safe Masses and Dimensions of Lattices of U and U02 Rods in Water" DP-1014, Savannah River Laboratory (1966).
6.
H. C. Paxton, " Criticality Control in Operations with Fissile Material;" LA-3366(Rev), Los Alamos Scientific Laboratory, (1972).
7.
R. D. Carter, et al, " Criticality Handbook," ARH-600 Revised last November 6,1973, Atlantic Richfield Hanford Company.
14.6 FIXE 0 POIS0NS
. The LRC does not now use Fixed Poisons to maintain nuclear criti-cality safety.
14.7 STRUCTURAL INTEGRITY Where structural integrity is necessary to provide assurance for g;
nuclear criticality safety in any operation, the design and con-('-
struction of those structures will be evaluated with due regard to load capacity and foreseeable abnormal loads, accidents and deteri-oration. This engineering activity is the responsibility of the LRC's Facilities Department with review and approval by a qualified person.
14.8 SPECIAL CONTROLS There are no special controls for nuclear criticality safety at the LRC.
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1 (mv) e FIGURE 14-1 FUEL R00 REMOVAL SCHEMATIC - MARK B Case 2 C+.e 3 4 rods removed 12 tot a removed Q
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FIGURE 14-2 FUEL R00 REMOVAL SCHEMATIC - MARK C Instrumentation Tube (1)
/
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D D
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MARK C FUEL ASSEMBLY (17 x 17)
Fuel Rods Removed for Reactivity Under Dismantlement (Section 4.4.7.3)
No. of Rods Pins Shown Removed By Letter 4
A 12 A, B 24 A, B, C, C' 36 A, B, C, C',
D 8
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I TABLE OF CONTENTS Section Page 15.0 PROCESS DESCRTPTION AND SAFETY ANALYSES.
15-1 O
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7) 15.0 PROCESS DESCRIPTI0tl AND SAFETY ANALYSES The operations and projects at the LRC do not lend themselves to flow sheets.
There is no operation presently in progress that has a regular measured feed material or a regular measured product output.
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(G 16.0 ACCIDENT ANALYSES The analyses of accidents may be found in the Environmental Report supporting this application and in the Radiological Contingency Plan for the Lynchburg Reseach Center, August, 1981, as revised.
O License No SNM 778 Docket No.70-824 Date October,1985
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