ANSI ANS 15.11-2016 Radiation Protection at Research Reactor Facilities.pdf

1、An American National StandardPublished by the American Nuclear Society 555 N. Kensington AveLa Grange Park, IL 60526ANSI/ANS-15.11-2016Radiation Protection at Research Reactor FacilitiesANSI/ANS-15.11-2016ANSI/ANS-15.11-2016 American National Standard Radiation Protection at Research Reactor Facilit

2、ies Secretariat American Nuclear Society Prepared by the American Nuclear Society Standards Committee Working Group ANS-15.11 Published by the American Nuclear Society 555 North Kensington Avenue La Grange Park, Illinois 60526 USA Approved May 13, 2016 by the American National Standards Institute, I

3、nc. American National Standard ANSI/ANS-15.11-2016 American National Standard Designation of this document as an American National Standard attests that the principles of openness and due process have been followed in the approval procedure and that a consensus of those directly and materially affec

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13、n Nuclear Society (ANS) Standards Committee will provide responses to inquiries about requirements, recommendations, and/or permissive statements (i.e., “shall,” “should,” and “may,” respectively) in American National Standards that are developed and approved by ANS. Responses to inquiries will be p

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18、an National Standard ANSI/ANS-15.11-2016 American National Standard ANSI/ANS-15.11-2016 i Foreword (This foreword is not a part of American National Standard “Radiation Protection at ResearchReactor Facilities,” ANSI/ANS-15.11-2016.) In the fall of 1970, the American Nuclear Society Standards Commit

19、tee established ANS-15, Operation of Research Reactors, under the auspices of the N17 Consensus Committee, Research Reactors, Reactor Physics, Radiation Shielding, and Computational Methods, to provide needed standards for the operation, use, and regulation of research reactors. Since then, numerous

20、 standards have been developed, and several working groups have been established, among them ANS-15.11. In 1977, two standards dealing with radiation protection at research reactors were published: ANSI/ANS-15.11-1977 (N628), “Radiological Control at Research Reactor Facilities,” and ANSI/ANS-15.12-

21、1977 (N647), “Design Objectives for and Monitoring of Systems Controlling Research Reactor Effluents.” These two standards were revised later and combined into a single standard, ANSI/ANS-15.11-1987, “Radiation Protection at Research Reactor Facilities.” Major changes to the regulations on standards

22、 for radiation protection and to the recommendations of the National Council on Radiation Protection and Management and the International Commission on Radiological Protection, significantly impacting research reactors, were incorporated into the revision, ANSI/ANS-15.11-1993, “Radiation Protection

23、at Research Reactor Facilities.” The standard was revised in 2009 with minor changes. Work on this standard began in November of 2013 and culminated in March of 2016 with approval by the Research and Advanced Reactor Consensus Committee. The current revision addresses applicable changes and provides

24、 directions on implementation, including meeting the objectives and principles of as-low-as-is-reasonably-achievable (ALARA) levels of radiation. In preparing this standard, the intent has been to specify objectives that will achieve the following results: (1) Establish a comprehensive radiation pro

25、tection program that deals with all matters involving radiation and radioactive materials at research reactors; (2) Limit exposures and releases to ALARA levels without seriously restricting the operation of existing reactors, inhibiting growth and upgrade, or discouraging the development of new res

26、earch reactors; (3) Set a reasonably low activity level threshold, above which measurements will be required that will allow for the use of readily available instrumentation without resorting to extraordinary means. In the process of creating standards with respect to existing and varied practices i

27、n many operating facilities, it is important to consider the following: (1) It is not intended that the standard be used as a demand model for backfitting purposes; (2) The standard can be a significant aid for existing and new owners or operators; (3) The standard can be helpful for a facility unde

28、rgoing change or modification; (4) The standards considered use can assist in implementing regulatory requirements. Prior to using the standard, individual facilities ought to carefully examine their license, permit, or other requirements for limiting conditions that might not be compatible with the

29、 American National Standard ANSI/ANS-15.11-2016 ii standard or new regulatory requirements and that might require change, amendment, or special authorization. Care also ought to be exercised in using appropriate units as might be specified by authorities. The standard does not address certain condit

30、ions that do not occur or are known not to exist at research reactor facilities such as planned special exposures, facilities-specific public dose limits, and hot particle contamination. Individual facilities ought to address these issues, if needed, in their programs. The family of American Nationa

31、l Standards developed by ANS-15 for research reactors are the following: ANSI/ANS-15.1-2007 (R2013), “The Development of Technical Specificationsfor Research Reactors”; ANSI/ANS-15.2-1999 (R2009), “Quality Control for Plate-Type Uranium-Aluminum Fuel Elements”; ANSI/ANS-15.4-2016, “Selection and Tra

32、ining of Personnel for ResearchReactors”; ANSI/ANS-15.7-1977 (R1986), “Research Reactor Site Evaluation”(withdrawn); ANSI/ANS-15.8-1995 (R2013), “Quality Assurance Program Requirements forResearch Reactors”; ANSI/ANS-15.10-1994, “Decommissioning of Research Reactors” (withdrawn); ANSI/ANS-15.11-2016

33、, “Radiation Protection at Research Reactor Facilities”; ANSI/ANS-15.15-1978 (R1986), “Criteria for the Reactor Safety Systems ofResearch Reactors” (withdrawn); ANSI/ANS-15.16-2015, “Emergency Planning for Research Reactors”; ANSI/ANS-15.17-1981 (R2000), “Fire Protection Program Criteria for Researc

34、hReactors” (withdrawn); ANSI/ANS-15.19-1991, “Shipment and Receipt of Special Nuclear Material(SNM) by Research Reactor Facilities” (withdrawn); ANSI/ANS-15.21-2012, “Format and Content for Safety Analysis Reports forResearch Reactors.”This standard might reference documents and other standards that

35、 have been superseded or withdrawn at the time the standard is applied. A statement has been included in the “References” section that provides guidance on the use of references. This standard does not incorporate the concepts of generating risk-informed insights, performance-based requirements, or

36、a graded approach to quality assurance. The user is advised that one or more of these techniques could enhance the application of this standard. This standard was prepared by the ANS-15.11 Working Group of the American Nuclear Society. The following members contributed to this standard: S. R. Reese

37、(Chair), Oregon State University C. H. Bassett, U.S. Nuclear Regulatory Commission D. R. Brown, National Institute of Standards and Technology R. J. Dobey, University of Missouri W. D. Frey, University of California at Davis American National Standard ANSI/ANS-15.11-2016 iii The Operations of Resear

38、ch Reactors Subcommittee had the following membership at the time of its approval of this standard: D. S. OKelly (Chair), Idaho National Laboratory A. Adams, U.S. Nuclear Regulatory Commission M. J. Burger, Sandia National Laboratories D. J. Cronin, University of Florida at Gainesville L. P. Foyto,

39、University of Missouri S. Miller, Armed Forces Radiobiology Research Institute M. K. Morrison, Idaho National Laboratory T. J. Myers, Individual D. M. Pinkston, Oak Ridge National Laboratory S. R. Reese, Oregon State University T. R. Schmidt, Individual R. Strader, National Institute of Standards an

40、d Technology The Research and Advanced Reactor Consensus Committee had the following membership at the time of its approval of this standard: G. F. Flanagan (Chair), Oak Ridge National Laboratory B. B. Bevard (Vice Chair), Oak Ridge National Laboratory T. Newton (Vice Chair), National Institute of S

41、tandards and Technology A. Adams, U.S. Nuclear Regulatory Commission J. K. August, Southern Nuclear Operating Company E. D. Blandford, University of New Mexico R. E. Carter, Individual L. P. Foyto, University of Missouri A. Grenci, Chicago Bridge the word “should” is used to denote a recommendation;

42、 and the word “may” is used to denote permission, neither a requirement nor a recommendation. 2.2 List of acronyms (the) Act: Atomic Energy Act of 1954 (42 U.S.C. 2011 et seq.) ALARA: as low as is reasonably achievable ALI: annual limit on intake DAC: derived air concentration DAC-hour: derived air

43、concentration-hour ICRP: International Commission on Radiological Protection ICRP 60: ICRP Publication 60 ICRP 103: ICRP Publication 103 ICRU: International Commission on Radiation Units and Measurements LDE: lens dose equivalent MQA: measurement quality assurance NCRP: National Council on Radiation

44、 Protection and Measurements NCRP 91: NCRP Report No. 91 NRC: U.S. Nuclear Regulatory Commission NVLAP: National Voluntary Laboratory Accreditation Program TEDE: total effective dose equivalent TLD: thermoluminescent dosimeter American National Standard ANSI/ANS-15.11-2016 2 2.3 Definitions For radi

45、ation quantities, the definitions herein do not conflict with those of the International Commission on Radiation Units and Measurements (ICRU), and the ICRU definitions may be used where appropriate. A number of definitions not used in this document are nevertheless included for completeness and ref

46、erence. The definitions of shall, should, and may are listed under “shall.” Many of the definitions are based on those in 10 CFR 20 1.1)In no case has any wording from that source been deleted. A few have been amplified to make them more generally applicable to all research reactors. These are noted

47、 in the text. absorbed dose: The energy imparted by ionizing radiation per unit mass of irradiated material. The units of absorbed dose are the rad and the gray (Gy). accessible area: The area that can reasonably be occupied by a significant portion of an individuals body (also see “radiation area”

48、and “high radiation area”). activity: The rate of disintegration (transformation) or decay of radioactive material. The units of activity are the curie (Ci) and the becquerel (Bq). adult: An individual 18 or more years of age. airborne radioactive material: Radioactive material dispersed in the air

49、in the form of dusts, fumes, particulates mists, vapors, or gases; also commonly referred to as airborne radioactivity. airborne radioactivity area: A room, enclosure, or area in which airborne radioactive materials, composed wholly or partly of permitted or licensed material, exist in concentrations (1) in excess of the derived air concentrations (DACs) for controlling occupational exposures, e.g., those specified in 10 CFR 20, Sec. 20.1001 (“Purpose”) through Sec. 20.2401 (“Violations”) and Appendix B “Annual Limits on Intake (ALIs) and Derived Air Concentrations