ASTM E1035-2002 Standard Practice for Determining NeutronExposures for Nuclear Reactor Vessel Support Structures《核反应堆容器支承结构的中子辐照量测定标准实施规程》.pdf

《ASTM E1035-2002 Standard Practice for Determining NeutronExposures for Nuclear Reactor Vessel Support Structures《核反应堆容器支承结构的中子辐照量测定标准实施规程》.pdf》由会员分享，可在线阅读，更多相关《ASTM E1035-2002 Standard Practice for Determining NeutronExposures for Nuclear Reactor Vessel Support Structures《核反应堆容器支承结构的中子辐照量测定标准实施规程》.pdf（3页珍藏版）》请在麦多课文档分享上搜索。

1、Designation: E 1035 02Standard Practice forDetermining NeutronExposures for Nuclear ReactorVessel Support Structures1This standard is issued under the fixed designation E 1035; the number immediately following the designation indicates the year oforiginal adoption or, in the case of revision, the ye

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

3、 innuclear reactor vessel support structures located in the vicinityof the active core. This practice includes guidelines for:1.1.1 Selecting appropriate dosimetric sensor sets and theirproper installation in reactor cavities.1.1.2 Making appropriate neutronics calculations to predictneutron radiati

4、on exposures.1.2 This practice is applicable to all pressurized waterreactors whose vessel supports will experience a lifetimeneutron fluence (E 1 MeV) that exceeds 1 3 1017neutrons/cm2or 3.0 3 104dpa.2(See Terminology E 170.)1.3 Exposure of vessel support structures by gamma radia-tion is not inclu

5、ded in the scope of this practice, but see thebrief discussion of this issue in 3.2.1.4 This standard does not purport to address all of thesafety concerns, if any, associated with its use. It is theresponsibility of the user of this standard to establish appro-priate safety and health practices and

6、 determine the applica-bility of regulatory limitations prior to use.2. Referenced Documents2.1 ASTM Standards:E 170 Terminology Relating to Radiation Measurementsand Dosimetry3E 482 Guide for Application of Neutron Transport Methodsfor Reactor Vessel Surveillance, E 706 (IID)3,4E 693 Practice for C

7、haracterizing Neutron Exposures inIron and Low Alloy Steels in Terms of Displacements PerAtom (DPA), E 706 (ID)3,4E 844 Guide for Sensor Set Design and Irradiation forReactor Surveillance, E 706 (IIC)3,4E 854 Test Method for Application and Analysis of SolidState Track Recorder (SSTR) Monitors for R

8、eactor Sur-veillance, E 706 (IIIB)3,4E 910 Test Method for Application and Analysis of HeliumAccumulation Fluence Monitors for Reactor Vessel Sur-veillance, E 706 (IIIC)3,4E 944 Guide for Application of Neutron Spectrum Adjust-ment Methods in Reactor Surveillance, E 706 (IIA)3,4E 1005 Test Method fo

9、r Application and Analysis of Radio-metric Monitors for Reactor Vessel Surveillance, E 706(IIIA)3,4E 1018 Guide for Application of ASTM Evaluated CrossSection Data File, E 706 (IIB)3,42.2 ASME Standard:Boiler and Pressure Vessel Code, Section III52.3 Nuclear Regulatory Documents:Code of Federal Regu

10、lations, “Fracture Toughness Require-ments,” Chapter 10, Part 50, Appendix G6Code of Federal Regulations, “Reactor Vessel MaterialsSurveillance Program Requirements,” Chapter 10, Part50, Appendix H6Regulatory Guide 1.99, Rev. 1, “Effects of Residual Ele-ments on Predicted Radiation Damage on Reactor

11、 VesselMaterials,” U. S. Nuclear Regulatory Commission, April197763. Significance and Use3.1 Prediction of neutron radiation effects to pressure vesselsteels has long been a part of the design and operation of lightwater reactor power plants. Both the federal regulatory agen-cies (see 2.2) and natio

12、nal standards groups (see 2.1) havepromulgated regulations and standards to ensure safe operationof these vessels. Recently, it has become apparent that thesupport structures for pressurized water reactor vessels mayalso be subject to similar neutron radiation effects (1, 2, 3, 4,5).7The objective o

13、f this practice is to provide guidelines fordetermining the neutron radiation exposures experienced byindividual vessel supports.3.2 It is known that high energy photons can also produce1This practice is under the jurisdiction of ASTM Committee E10 on NuclearTechnology and Applications and is the di

14、rect responsibility of SubcommitteeE10.05 on Nuclear Radiation Metrology.Current edition approved June 10, 2002. Published September 2002. Originallypublished as E 103585. Last previous edition E 103595(1996).2Based on data from Table 5 of Master Matrix E 706 and Reference 5.3Annual Book of ASTM Sta

15、ndards, Vol 12.02.4The reference in parentheses refers to Section 5 of Matrix E 706.5Available from American Society of Mechanical Engineers, 345 E. 47th St.,New York, NY 10017.6Available from Superintendent of Documents, U.S. Government PrintingOffice, Washington, DC 20402.7The boldface numbers in

16、parentheses refer to a list of references at the end ofthis practice.1Copyright ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959, United States.displacement damage effects that may be similar to thoseproduced by neutrons. These effects are known to be much les

17、sat the belt line of a light water reactor pressure vessel thanthose induced by neutrons. The same has not been proven forall locations within vessel support structures. Therefore, it maybe prudent to apply coupled neutron-photon transport methodsand photon induced displacement cross sections to det

18、erminewhether gamma-induced dpa exceeds the screening level of 3.03 10-4, used in this practice for neutron exposures. See 1.2.4. Irradiation Requirements4.1 Location of Neutron DosimetersNeutron dosimetersshall be located along the support structure in the region wherethe maximum dpa or fluence (E

19、1 MeV) is expected to occur,based on neutronics calculations outlined in Section 5. Caremust be taken to ensure that reactor cavity structures notmodeled in the neutronics calculation offer no additionalshielding to the dosimeters. The neutron dosimeters will beanalyzed to obtain a map of the neutro

20、n fields within the actuallocation of the support structures.4.2 Neutron Dosimeters:4.2.1 Information regarding the selection of appropriatesensor sets for support structure application may be found inGuide E 844, Test Method E 1005, and Test Methods E 854and E 910.4.2.2 In particular, Test Method E

21、 910 also provides guid-ance for the additional possibility that operating plants may useexisting copper bearing instruments and cables within thereactor cavity as a priori passive dosimeter candidate.5. Determination of Neutron Exposure Parameter Values5.1 Neutronics CalculationsAll neutronics calc

22、ulationsfor (a) the analysis of integral dosimetry data, and (b) theprediction of irradiation damage exposure parameter valuesshall follow Guide E 482, subject to these additional consid-erations that may be encountered in reactor cavities:5.1.1 If the vessel supports do not lie within the coresacti

23、ve height, then an asymmetric quadrature set must bechosen for discrete ordinates calculations that will accuratelyreproduce the neutron transport in the direction of the supports.Care must be exercised in constructing the quadrature set toensure that “ray streaming” effects in the cavity air gap do

24、 notdistort the calculation of the neutron transport.5.1.2 If the support system is so large or geometricallycomplex that it perturbs the general neutron field in the cavity,the analysis method of choice may be that of a coupled discreteordinates/Monte Carlo calculation. The normal coupling forthis

25、type of problem would be to perform the two dimensionaldiscrete ordinates analysis only within the vessel. The neutroncurrents generated by this analysis would be used to create theappropriate cumulative distribution functions in the finalMonte Carlo analysis. For details of such analyses see Refs (

26、6),(7), and (8). In this instance, the above caveats still hold forthe discrete ordinates calculation, but in addition, thevariance of the Monte Carlo results must now be includedwith the overall assessment of the variance of the dosimetrydata.5.2 Determination of Damage Exposure Values andUncertain

27、tiesAdjustment procedures outlined in GuideE 944 and Guide E 1018 shall be performed to obtain damageexposure values dpa and fluence (E 1 Mev) using the integraldata from the neutron dosimeters and the calculation in 5.1.The cross sections for dpa are found in Practice E 693. Dpashall be determined

28、for this application rather than just fluence(E 1 MeV) because Ref (5) notes an increase in the ratio ofdpa to fluence (E 1 MeV) by a factor of two in going fromthe surveillance capsule position inside the reactor vessel to aposition out in the reactor cavity.REFERENCES(1) Docket 50338-207, North An

29、na Power Station, Units 1 and 2,Summary of Meeting Held on September 19, 1975 on Dynamic Effectsof LOCAs, Sept. 22, 1975.(2) Sprague, J. A., and Hawthorne, J. R., “Radiation Effects to ReactorVessel Supports,” U. S. Naval Research Laboratory Report NRC-03-79-148 for the U. S. Nuclear Regulatory Comm

30、ission, Oct. 22, 1979.(3) Unresolved Safety Issues Summary, NUREG-0606, Vol 4, No. 4, TaskA-11: Reactor Vessel Materials Toughness, November, 1982.(4) Asymmetric Blowdown Loads on PWR Primary Systems, NUREG-0609, U.S. Nuclear Regulatory Commission, 1981.(5) Hopkins, W. C., “Suggested Approach for Fr

31、acture-Safe PRV SupportDesign in Neutron Environments,” Transactions of the AmericanNuclear Society, Vol 30, 1978, pp. 187188.(6) Cain, V. R., “The Use of Monte Carlo with Albedos to Predict NeutronStreaming in PWR Containment Buildings,” Transactions of theAmerican Nuclear Society, Vol 23, 1976, p.

32、 618.(7) Straker, E. A., Stevens, P. N., Irving, D. C. and Cain, V. R., “TheMORSE CodeA Multigroup Neutron and Gamma-Ray MontreCarlo Transport Code,” ORNL-4585, September 1970.(8) Emmett, M. B., Burgart, C. E., and Hoffman, T. J., “DOMINO: AGeneral Purpose Code for Coupling Discrete Ordinates and Mo

33、nteCarlo Radiation Transport Calculations,” ORNL-4853, July 1973.ASTM International takes no position respecting the validity of any patent rights asserted in connection with any item mentionedin this standard. Users of this standard are expressly advised that determination of the validity of any su

34、ch 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 five years andif not revised, either reapproved or withdrawn. Your comments are invit

35、ed 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 committee, which you may attend. If you feel that your comments have not received a

36、 fair hearing you shouldmake your views known to the ASTM Committee on Standards, at the address shown below.E1035022This standard is copyrighted by ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 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).E1035023