ASTM E1035-2013 red 5000 Standard Practice for Determining Neutron Exposures for Nuclear Reactor Vessel Support Structures《测定核反应堆容器支座结构中子辐照的标准实施规程》.pdf

《ASTM E1035-2013 red 5000 Standard Practice for Determining Neutron Exposures for Nuclear Reactor Vessel Support Structures《测定核反应堆容器支座结构中子辐照的标准实施规程》.pdf》由会员分享，可在线阅读，更多相关《ASTM E1035-2013 red 5000 Standard Practice for Determining Neutron Exposures for Nuclear Reactor Vessel Support Structures《测定核反应堆容器支座结构中子辐照的标准实施规程》.pdf（4页珍藏版）》请在麦多课文档分享上搜索。

1、Designation: E1035 08 E1035 13Standard Practice forDetermining Neutron Exposures for Nuclear ReactorVessel Support Structures1This standard is issued under the fixed designation E1035; the number immediately following the designation indicates the year oforiginal adoption or, in the case of revision

2、, the year of last revision. A number in parentheses indicates the year of last reapproval. Asuperscript epsilon () indicates an editorial change since the last revision or reapproval.1. Scope1.1 This practice covers procedures for monitoring the neutron radiation exposures experienced by ferritic m

3、aterials in nuclearreactor vessel support structures located in the vicinity of the active core. This practice includes guidelines for:1.1.1 Selecting appropriate dosimetric sensor sets and their proper installation in reactor cavities.1.1.2 Making appropriate neutronics calculations to predict neut

4、ron radiation exposures.1.2 This practice is applicable to all pressurized water reactors whose vessel supports will experience a lifetime neutron fluence(E 1 MeV) that exceeds 1 1017 neutrons/cm2 or 3.0 104 dpa.2 (See Terminology E170.)1.3 Exposure of vessel support structures by gamma radiation is

5、 not included in the scope of this practice, but see the briefdiscussion of this issue in 3.2.1.4 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibilityof the user of this standard to establish appropriate safety and health pra

6、ctices and determine the applicability of regulatorylimitations prior to use.2. Referenced Documents2.1 ASTM Standards:3E170 Terminology Relating to Radiation Measurements and DosimetryE482 Guide for Application of Neutron Transport Methods for Reactor Vessel Surveillance, E706 (IID)E693 Practice fo

7、r Characterizing Neutron Exposures in Iron and Low Alloy Steels in Terms of Displacements Per Atom (DPA),E 706(ID)E844 Guide for Sensor Set Design and Irradiation for Reactor Surveillance, E 706 (IIC)E854 Test Method for Application and Analysis of Solid State Track Recorder (SSTR) Monitors for Reac

8、tor Surveillance,E706(IIIB)E910 Test Method for Application and Analysis of Helium Accumulation Fluence Monitors for Reactor Vessel Surveillance,E706 (IIIC)E944 Guide for Application of Neutron Spectrum Adjustment Methods in Reactor Surveillance, E 706 (IIA)E1005 Test Method for Application and Anal

9、ysis of Radiometric Monitors for Reactor Vessel Surveillance, E 706 (IIIA)E1018 Guide for Application of ASTM Evaluated Cross Section Data File, Matrix E706 (IIB)2.2 ASME Standard:Boiler and Pressure Vessel Code, Section III42.3 Nuclear Regulatory Documents:Code of Federal Regulations, “Fracture Tou

10、ghness Requirements,” Chapter 10, Part 50, Appendix G5Code of Federal Regulations, “Reactor Vessel Materials Surveillance Program Requirements,” Chapter 10, Part 50, AppendixH51 This practice is under the jurisdiction of ASTM Committee E10 on Nuclear Technology and Applicationsand is the direct resp

11、onsibility of Subcommittee E10.05 onNuclear Radiation Metrology.Current edition approved Nov. 1, 2008Jan. 1, 2013. Published December 2008January 2013. Originally approved in 1985. Last previous edition approved in 20022008as E103502.08. DOI: 10.1520/E1035-08.10.1520/E1035-13.2 Based on data from Ta

12、ble 5 of Master Matrix E706 and Reference 5.3 For referenced ASTM standards, visit the ASTM website, www.astm.org, or contact ASTM Customer Service at serviceastm.org. For Annual Book of ASTM Standardsvolume information, refer to the standards Document Summary page on the ASTM website.4 Available fr

13、om American Society of Mechanical Engineers, 345 E. 47th St., New York, NY 10017.5 Available from Superintendent of Documents, U.S. Government Printing Office, Washington, DC 20402.This document is not an ASTM standard and is intended only to provide the user of an ASTM standard an indication of wha

14、t changes have been made to the previous version. Becauseit may not be technically possible to adequately depict all changes accurately, ASTM recommends that users consult prior editions as appropriate. In all cases only the current versionof the standard as published by ASTM is to be considered the

15、 official document.Copyright ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959. United States1Regulatory Guide 1.99, Rev. 1, “Effects of Residual Elements on Predicted Radiation Damage on Reactor Vessel Materials,” U.S. Nuclear Regulatory Commission, April 1977

16、53. Significance and Use3.1 Prediction of neutron radiation effects to pressure vessel steels has long been a part of the design and operation of light waterreactor power plants. Both the federal regulatory agencies (see 2.22.3) and national standards groups (see 2.1 and 2.2) havepromulgated regulat

17、ions and standards to ensure safe operation of these vessels. The support structures for pressurized waterreactor vessels may also be subject to similar neutron radiation effects (1, 2, 3, 4, 5).6 The objective of this practice is to provideguidelines for determining the neutron radiation exposures

18、experienced by individual vessel supports.3.2 It is known that high energy photons can also produce displacement damage effects that may be similar to those producedby neutrons. These effects are known to be much less at the belt line of a light water reactor pressure vessel than those inducedby neu

19、trons. The same has not been proven for all locations within vessel support structures. Therefore, it may be prudent to applycoupled neutron-photon transport methods and photon induced displacement cross sections to determine whether gamma-induceddpa exceeds the screening level of 3.0 10-4, used in

20、this practice for neutron exposures. See(See 1.2.).4. Irradiation Requirements4.1 Location of Neutron DosimetersNeutron dosimeters shall be located along the support structure in the region where themaximum dpa or fluence (E 1 MeV) is expected to occur, based on neutronics calculations outlined in S

21、ection 5. Care must betaken to ensure that reactor cavity structures not modeled in the neutronics calculation offer no additional shielding to thedosimeters. The neutron dosimeters will be analyzed to obtain a map of the neutron fields within the actual location of the supportstructures.4.2 Neutron

22、 Dosimeters:4.2.1 Information regarding the selection of appropriate sensor sets for support structure application may be found in GuideE844, Test Method E1005, and Test Methods E854 and E910.4.2.2 In particular, Test Method E910 also provides guidance for the additional possibility that operating p

23、lants may use existingcopper bearing instruments and cables within the reactor cavity as a priori passive dosimeter candidate.5. Determination of Neutron Exposure Parameter Values5.1 Neutronics CalculationsAll neutronics calculations for (a) the analysis of integral dosimetry data, and (b) the predi

24、ctionof irradiation damage exposure parameter values shall follow Guide E482, subject to these additional considerations that may beencountered in reactor cavities:5.1.1 If the vessel supports do not lie within the cores active height, then an asymmetric quadrature set must be chosen fordiscrete ord

25、inates calculations that will accurately reproduce the neutron transport in the direction of the supports. Care must beexercised in constructing the quadrature set to ensure that “ray streaming” effects in the cavity air gap do not distort the calculationof the neutron transport.5.1.2 If the support

26、 system is so large or geometrically complex that it perturbs the general neutron field in the cavity, theanalysis method of choice may be that of a Monte Carlo calculation or a combined discrete ordinates/Monte Carlo calculation. Thecombined calculation involves a two or three dimensional discrete

27、ordinates analysis only within the vessel. The neutron currentsor fluences generated by this analysis may be used to create the appropriate source distribution functions in the final Monte Carloanalysis, or to develop bias (weighing) factors for use in a complete Monte Carlo model. For details of an

28、alyses in which discreteordinates and Monte Carlo methods were coupled see Refs (6),(7), and (8). InReference this(9) instance, the provides a review ofthe available combined or hybrid discrete ordinates/Monte Carlo calculations. For hybrid calculations, the above caveats still holdfor the discrete

29、ordinates calculation, but in addition, the variance of the Monte Carlo results must now be included with the overallassessment of the variance of the dosimetry data.5.2 Determination of Damage Exposure Values and UncertaintiesAdjustment procedures outlined in Guide E944 and GuideE1018 shall be perf

30、ormed to obtain damage exposure values dpa and fluence (E 1 Mev)MeV) using the integral data from theneutron dosimeters and the calculation in 5.1. The cross sections for dpa are found in Practice E693. Dpa shall be determined forthis application rather than just fluence (E 1 MeV) because Ref (5) no

31、tes an increase in the ratio of dpa to fluence (E 1 MeV)by a factor of two in going from the surveillance capsule position inside the reactor vessel to a position out in the reactor cavity.6 The boldface numbers in parentheses refer to a list of references at the end of this practice.E1035 132REFERE

32、NCES(1) Docket 50338-207, North Anna Power Station, Units 1 and 2, Summary of Meeting Held on September 19, 1975 on Dynamic Effects of LOCAs, Sept.22, 1975.(2) Sprague, J. A., and Hawthorne, J. R., “Radiation Effects to Reactor Vessel Supports,” U. S. Naval Research Laboratory Report NRC-03-79-148 f

33、orthe U. S. Nuclear Regulatory Commission, Oct. 22, 1979.(3) Unresolved Safety Issues Summary, NUREG-0606, Vol 4, No. 4, Task A-11: Reactor Vessel Materials Toughness, November, 1982.(4) Asymmetric Blowdown Loads on PWR Primary Systems, NUREG-0609, U.S. Nuclear Regulatory Commission, 1981.(5) Hopkin

34、s, W. C., “Suggested Approach for Fracture-Safe PRV Support Design in Neutron Environments,” Transactions of the American NuclearSociety, Vol 30, 1978, pp. 187188.(6) Cain, V. R., “The Use of Monte Carlo with Albedos to Predict Neutron Streaming in PWR Containment Buildings,” Transactions of the Ame

35、ricanNuclear Society, Vol 23, 1976, p. 618.(7) Straker, E. A., Stevens, P. N., Irving, D. C. and Cain, V. R., “The MORSE CodeA Multigroup Neutron and Gamma-Ray Montre Carlo TransportCode,” ORNL-4585, September 1970.(8) Emmett, M. B., Burgart, C. E., and Hoffman, T. J., “DOMINO: A General Purpose Cod

36、e for Coupling Discrete Ordinates and Monte Carlo RadiationTransport Calculations,” ORNL-4853, July 1973.(9) Wagner, J. C., Peplow, D. E., Mosher, S. W., and Evans, T. M., “Review of Hybrid (Deterministic/Monte Carlo) Radiation Transport Methods, Codes,and Applications at Oak Ridge National Laborato

37、ry,” In Progress in Nuclear Science and Technology, Vol 2, Toshikazu Takeda, Ed., Atomic EnergySociety of Japan, October 2011, pp. 808-814.E1035 133ASTM International takes no position respecting the validity of any patent rights asserted in connection with any item mentionedin this standard. Users

38、of this standard are expressly advised that determination of the validity of any such patent rights, and the riskof infringement of such rights, are entirely their own responsibility.This standard is subject to revision at any time by the responsible technical committee and must be reviewed every fi

39、ve years andif not revised, either reapproved or withdrawn. Your comments are invited either for revision of this standard or for additional standardsand should be addressed to ASTM International Headquarters. Your comments will receive careful consideration at a meeting of theresponsible technical

40、committee, which you may attend. If you feel that your comments have not received a fair hearing you shouldmake your views known to the ASTM Committee on Standards, at the address shown below.This standard is copyrighted by ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, P

41、A 19428-2959,United States. Individual reprints (single or multiple copies) of this standard may be obtained by contacting ASTM at the aboveaddress or at 610-832-9585 (phone), 610-832-9555 (fax), or serviceastm.org (e-mail); or through the ASTM website(www.astm.org). Permission rights to photocopy the standard may also be secured from the ASTM website (www.astm.org/COPYRIGHT/).E1035 134