ML20205M980

Aj Blotcky Reactor Facility Radiation Protection Program

ML20205M980
Person / Time
Site:	05000131
Issue date:	04/05/1999
From:	DEPT. OF VETERANS AFFAIRS MEDICAL CENTER, OMAHA
To:
Shared Package
ML20205M954	List: ML20205M951 ML20205M968 ML20205M980 ML20205M999
References
PROC-990405, NUDOCS 9904160115
Download: ML20205M980 (9)

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A.J. BLOTCKY REACTOR FACILITY RADIATION PROTECTION PROGRAM O

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1 TABLE OF CONTENTS gQ 1.0 Introduction. 3 (j 2.0 Management and Administration . 3 2.1 Radiation Units . 3 22 Radiation Limits. 4 3.0 Training.. 4 4.0 Surveillance . 5 4.1 Radioactive Materials Accountability. 5 4.2 Effluent Monitoring. 5 4.3 Contamination Surveys. 5 4.4 Environs Monitoring. 6 4.5 Personnel Exposure . 6 5.0 ALARA Program . 7 5.1 Policy and Objectives.. 7 5.2 Implementation . 7 5.3 Elements of the ALARA Review and Report. 8 6.0 References . 8 O

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1. Introduction This Radiation Protection Program has been prepared by personnel of the A.J. BLOTCKY TRIGA Mk i Nuclear Reactor Facility in response to the requirements of Title 10, Part 20.1101, Code of Federal Regulations (10CFR20). The goal of the Program is the limitation of radiation exposures and radioactivity releases to a level that is as low as reasonably achievable (ALARA) without seriously restricting operation of the Facility for purposes of education and research.

The Facility is operated under License R-57 (Docket 50-151) issued by the U.S. Nuclear Regulatory Commission (NRC). The Program is executed in coordination with the Radiation Safety Department, Omaha VA Medical Center. It has been reviewed and approved by the Reactor Safeguards Committee for the Reactor Facility. Certain aspects of the Program deal with radioactive materials regulated by the OVAMC Broad Scope Materials License #26-00138-10.

This program is a part of the Operations Manual (completion pending) for the Reactor Facility, although it is published separately. A closely related part of the Operations Manual, also published separately, is the Emergency Plan. The Radiation Protection Program is designed to m.2et requirements of 10CFR20. It has been developed fol lowing the guidance of the American National Standard Radiation Protection at Research Reactor Facilities (1 ) and Regulatory Guides issued by the NRC (2-7].

2. Management and Administration (N Preparation, audit, and review of the Radiation Protection Program is the responsibility of the Reactor Manager of the Nuclear Reactor Facility. The activities of the Reactor Manager and annual audits prepared by the Reactor Manager, are reviewed by the Reactor Safeguards Committee chaired by the Chief of Staff for the OVAMC Medical Center. Records required by the Radiation Protection Program as well as audit reports by the Facility Reactor Manager are examined by the Reactor Safeguards Committee during their annual audit. Training, surveillance and record keeping are the responsibility of the Reactor Manager. ALARA activities, for which record keeping is the particular responsibility of the Reactor Manager, are incumbent upon all radiation workers associated with the Reactor Facility.

Substantive changes in the Radiation Protection Program require approval of the Reactor Safeguards Committee. Editorial changes, or changes to appendices, may be made on the authority of the Facility Reactor Manager. Changes made to operating or emergency procedures apply automatically to the Radiation Protection Program and corresponding changes may be made in the Program without further consideration by the Reactor Safeguards Committee. As with procedures, the Reactor Manager may override elements of the Program on a temporary emergency basis so long as the emergency changes are brought promptly to the attention of the Safeguards Committee.

2.1 Radiation Units l

The traditional units of Curie, rad, rem, and roentgen are to be used in record keeping. SI units of becquerel, gray, and sievert may be used in calculations, dose assessments and reports, so long as final results, conclusions, etc., are given in traditional units as well.

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External exposure is to be recorded in terms of deep or shallow dose equivalent (index).

According to the ICRP (8], the deep dose equivalent (in rem units) is within 4 percent of the free field exposure rate (in roentgen units) for gamma rays with energies between 0.6 and 8.0 MeV. Therefore, survey or area monitoring instruments calibrated in roentgen units may be used for assessment of deep dose equivalent in routine surveillance.

The total effective dose equivalent (TEDE) is the sum of the deep dose equivalent for external

exposure and the committed effective dose equivalent for internal exposure. Internal exposure I associated with the Reactor Facility has never been a threat to workers or the public Should it be considered as a potentiality, in connection with planned special exposures or in the conduct of ALARA reviews, its evaluation should follow the guidance of 10CFR20, Regulatory Guides l

(3-7] or the ICRP [9-11].

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l 2.2 Radiation Limits l l

l Occupational dose limits (except for planned special exposures), given by 10.CFR20.1201 are as follows. The annual limit for adults, in summary, is the more limiting of the following:

. 5 rem total effective dose equivalent (TEDE), or

. 15 rem to the lens of the eye, or

. 50 rem shallow dose equivalent to the skin or to any extremity, or

. 50 rem combined deep dose equivalent and committed dose equivalent to any organ other than the lens of the eye.

Dose limits for individual members of the public, given by 10CFR20.1301, are in summary as follows:

. 0.1 rem total ef,ective dose equivalent (TEDE) in one year, and

. 0.002 rem TEDE in one hour. I

3. Training l

Implementation of training for radiation protection is the responsibility of the Reactor Manager. 1 Re-training for active researchers must be administered biennially except for Reactor Operators and Senior Reactor Operators taking part in the annual Reactor Facility Requalification Program. Internal exposure monitoring is required only for adults likely to receive in 1 year in excess of 10% of the applicable annual limits of intake for ingestion and inhalation or for minors or pregnant women likely to receive in excess of 0.05 rem committed effective dose in one year.

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4. Surveillance

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( Surveillance requirements related to radiation protection and imposed by the Reactor Safeguards Committee are as follows:

l Daily (operational) Environmental surveillance of the reactor area before and after each days operation Monthly Wipe test reactor laboratory Environmental surveillance (survey meter)

Quarterly Source inventory and leak test (OVAMC RSO)

Semi-annually Emergency equipment inventory 4.1 Radioactive Materials Accountability Materials accountability and release of radioactive materialis the responsibility of the Reactor Manager or his delegate. Proceduros are in place to assure that applicable regulations and procedures are in complicance. The responsibility for the facility nuclear fuelinventory is delegated to the Reactor Manager by the Hospital Director. A written statement of this responsiblity is updated every two years. The fuel inventory is calculated semi-annually and the fuel report is sent to the Department of Energy and the Nuclear Regulatory Commission.

4.2 Effluent Monitoring

\ Monitoring of liquid effluents from the Facility is performed to assure compliance with 10CFR20.2003.

4.3 Contamination Monitoring and Surveys Personnel shall monitor hands and feet for contamination when leaving known contaminated areas or restricted areas that are likely contaminated. If contamination is detected, then a check of exposed areas of the body and clothing should be made. Monitoring control points shall be established for this purpose. Materials, tools, and equipment shall be monitored for contamination before removal from contaminated areas or restricted areas likely to be contaminated 4.3.1 Limits for Removable and Fixed Contamination:

Radioactive waste to be disposed of in this manner shall be held for decay a minimum of 10 half-lives. Before disposal as ordinary trash, byproduct material shall be surveyed at the container surface with the appropriate meter set on its most sensitive scale and with no interposed shielding to determine that its radioactivity cannot be distinguished from background.

All radiation labels shall be removed or obliterated. A record of each disposal permitted under this condition shall be retained for 3 years. The record must include the date of disposal, the date of which the byproduct material was placed in storage, the radionuclides disposed, the survey instruments used, the background dose rate, the dose rate measured at the surface of each waste container, and the name of the individual who performed the disposal. Ohterwise, s the material will be deposed of in acoordance with 10 CFR 20 subpart K and other appliccable regulations.

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4.4 Environs Monitoring Environs monitoring is required to assure compliance with IOCFR20, Subpart F and with l Technical Specifications for the Facility operating license. Technical Specifications require the i following monitoring during reactor operations:

a Area radiation monitor at the top of the tank. Calibration of area radiation monitors is required annually.

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b. Continuous air monitor. Calibration of the continuous air monitor is required annually and is performed following the facility standard operating procedures.

4.5 Personnel Exposure Regulation 10CFR.20.1502 requires monitonng of workers likely to receive, in one year from sources external to the body, a dose in excess of 10 percent of the limits prescribed in ,

10CFR20.1201. The regulation also requires monitoring of any individuals entering a high or I very high radiation area within which an F dividual could receive a dose equivalent of 0.1 rem in .

one hour. According to Regulatory Guide 8.7 (2), if a prospective evaluation of likely doses j indicates that an individual is not likely to exceed 10 percent of any applicable limit, then there i are no requirements for recordkeeping or reporting. Likewise, Regulatory Guice 8.3413]

indicates that, if individual monitoring results serve as confirmatory measures, but monitoring is not required by 10CFR20.1502, then such results are not subject to the individual dese recordkeeping requirements of 10CFR20.2106(a) even though they may be used to satisfy

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3 10CFR20.1501 requirements. As a result, the following procedures are implemented:

a. When the reactor is in operation, no person may enter the reactor room without the permission of the Reactor Operator on duty at the Console,

b. Unescorted visitors will not enter the reactor room.

c. Except as indicated below, no person may enter the reactor room without gamma-ray personnel dosimetry.

a. Escorted individual visitors must have personnel dosimetry prior to reactor room entry.

e. Escorted visiting groups will be monitored by not less than two self-reading pocket dosimeters or two film or TLD badges for every fifteen members of the group.

5. ALARA Program 5.1 Policy and Objectives Management of the Reactor Facility is committed to keeping both occupational and public radiation exposure as low as reasonably achievable (ALARA). The specific goal of the ALARA program is to assure that actual exposures are no greater than 10 percent of the occupational limits and 50 percent of the public limits prescribed by IOCFR20, namely, ALARA goals of:

Workers:

=> 5 500 mrem annual TEDE

=> $ 5 rem annual dose equivalent to any organ except the lens of the eye

=> 51.5 rem annual dose equivalent to the lens of the eye

=> 5 5 rem annual dose equivalent to the skin

> 5 50 mrem dose equivalent to the fetus during pregnancy 6

Public:

( =. < 50 mrem annual TEDE 5.2 Implementation of the ALARA Program Planning and scheduling of operations and experiments, education and training, and facility design are the responsibilities of the Reactor Supervisor. Any action which might lead to as much as half the annual ALARA dose limit (Section 5.1) to any one individualin one calendar quarter requires a formal ALARA review and report. Any staff member or experimenter, or any member of the Reactor Safeguards Committee may call for an ALARA review of a proposed action. Under any of these circumstances, it is the responsibility of the Reactor Supervisor to conduct an ALARA review and report.

5.3 Elements of the ALARA Review and Report The following topics shall be considered, if applicable. The report shall include but not be limited j to discussion of how these topics affect personnel exposure and specific actions recommended, categorized by topic:

l l Features for Internal and External Radiation Control:

l . Shielding and construction materials

. Radioactive material storage and disposal

. Monitoring systems

. Fac!'ity layout

. Control of access to high and very high radiation areas t . Contamination Control

' . Ventilation and filtration

. Containme".of contamination

. Construcuan materials to facilitate decontamination

. Liquid effluents

. Effluent monitoring

. Operations and Operations Planning

. Assessment of potentialindividual and collective exposures

. Application of shielding, time, and distance for dose reduction

. Use of ventilation and decontamination to reduce collective dose

. Provision of special personnel training and practice

. Provision of special supervision and surveillance

. Provision of special clothing or other protective gear

. Provision for determination of possible methods of intake

. Provision for appropriate calculation of possible intake l

6.0 References l

l 1. American National Standard Radiation Protection at Research Facilities ANSI /ANS-15.11 (Final Draft), American Nuclear Society, La Grange Park, Illinois, October,.1992.

2. Instructions for Recording and Reporting Occupational Radiation Exposure Data Regulatory Guide 8.7 (Rev.1), U.S. Nuclear Regulatory Commission, Washington, D.C.,1992.

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3. Monitoring Criteria and Methods to Calculate Occupational Radiation Doses Regulatory Guide 8.34, U.S. Nuclear Regulatory Commission, Washington, D.C.,1992.

4. Air Sampling in the Workplace Regulatory Guide 8.25 (Rev.1), U S. Nuclear Regulatorv Commission. Washington. D.C.1992.

S. Planned Special Exposures Regulatory Guide 8.35, U.S. Nuclear Regulatory Commission, Washington. D.C. 1992.

6. Radiation Dose to the Embryo / Fetus Regulatory Guide 8.36, U.S. Nuclear Regulatory Commission, Washington, D.C.,1992.

7. Interpretation of Bioassay Measurements Draft Regulatory Guide 8.9 (DG 8009), U.S. Nuclear Regulatory Commission, Washington, D.C.,1992.

8. Data for Use in Protection Against External Radiation Publication 51, International Commission on Radiological Protection.1987.

9 Limits for intakes of Radionuclides by Workers Publication 30, International Commission on Radiological Protection,1979.

10. 1990 Recommendations of the international Commission on Radiological Protection Publication 60, International Commission on Radiological Protection,1991.

11. Limits for intakes of Radionuclides by Workers Based on 1990 Recommendations of the International Commission on Radiological Protection Publication 61, International Commission on Radiological Protection,1991.

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