1、Designation: E 1035 08Standard Practice forDetermining Neutron Exposures 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 y
2、ear 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 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:3E 170 Terminology Relating to Radiation Measurementsand DosimetryE 482 Guide for Application of Neutron Transport Methodsfor Reactor Vessel Surveillance, E706 (IID)E 693 Practice for Chara
7、cterizing Neutron Exposures inIron and Low Alloy Steels in Terms of Displacements PerAtom (DPA), E 706(ID)E 844 Guide for Sensor Set Design and Irradiation forReactor Surveillance, E 706(IIC)E 854 Test Method for Application and Analysis of SolidState Track Recorder (SSTR) Monitors for Reactor Sur-v
8、eillance, E706(IIIB)E 910 Test Method for Application and Analysis of HeliumAccumulation Fluence Monitors for Reactor Vessel Sur-veillance, E706 (IIIC)E 944 Guide for Application of Neutron Spectrum Adjust-ment Methods in Reactor Surveillance, E 706 (IIA)E 1005 Test Method forApplication andAnalysis
9、 of Radio-metric Monitors for Reactor Vessel Surveillance, E706(IIIA)E 1018 Guide for Application of ASTM Evaluated CrossSection Data File, Matrix E 706 (IIB)2.2 ASME Standard:Boiler and Pressure Vessel Code, Section III42.3 Nuclear Regulatory Documents:Code of Federal Regulations, “Fracture Toughne
10、ss Require-ments,” Chapter 10, Part 50, Appendix G5Code of Federal Regulations, “Reactor Vessel MaterialsSurveillance Program Requirements,” Chapter 10, Part50, Appendix H5Regulatory Guide 1.99, Rev. 1, “Effects of Residual Ele-ments on Predicted Radiation Damage on Reactor VesselMaterials,” U. S. N
11、uclear Regulatory Commission, April197753. 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 national standards groups (see
12、2.1) havepromulgated regulations and standards to ensure safe operationof these vessels. The support structures for pressurized waterreactor vessels may also be subject to similar neutron radiationeffects (1, 2, 3, 4, 5).6The objective of this practice is toprovide guidelines for determining the neu
13、tron radiation expo-sures experienced by individual vessel supports.1This practice is under the jurisdiction of ASTM Committee E10 on NuclearTechnology and Applications and is the direct responsibility of SubcommitteeE10.05 on Nuclear Radiation Metrology.Current edition approved Nov. 1, 2008. Publis
14、hed December 2008. Originallyapproved in 1985. Last previous edition approved in 2002 as E 103502.2Based on data from Table 5 of Master Matrix E 706 and Reference 5.3For referenced ASTM standards, visit the ASTM website, www.astm.org, orcontact ASTM Customer Service at serviceastm.org. For Annual Bo
15、ok of ASTMStandards volume information, refer to the standards Document Summary page onthe ASTM website.4Available from American Society of Mechanical Engineers, 345 E. 47th St.,New York, NY 10017.5Available from Superintendent of Documents, U.S. Government PrintingOffice, Washington, DC 20402.6The
16、boldface numbers in 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.3.2 It is known that high energy photons can also producedisplacement damage effects that may be
17、 similar to thoseproduced by neutrons. These effects are known to be much lessat 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-
18、photon transport methodsand photon induced displacement cross sections to determinewhether 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
19、along the support structure in the region wherethe maximum dpa or fluence (E 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 dos
20、imeters. The neutron dosimeters will beanalyzed to obtain a map of the neutron 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 Meth
21、od E 1005, and Test Methods E 854and E 910.4.2.2 In particular, Test Method E 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 Ne
22、utron Exposure Parameter Values5.1 Neutronics CalculationsAll neutronics calculationsfor (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
23、reactor cavities:5.1.1 If the vessel supports do not lie within the coresactive 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
24、quadrature set toensure that “ray streaming” effects in the cavity air gap do 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 Mo
25、nte Carlocalculation or a combined discrete ordinates/Monte Carlocalculation. The combined calculation involves a two or threedimensional discrete ordinates analysis only within the vessel.The neutron currents or fluences generated by this analysis maybe used to create the appropriate source distrib
26、ution functionsin the final Monte Carlo analysis, or to develop bias (weighing)factors for use in a complete Monte Carlo model. For details ofanalyses in which discrete ordinates and Monte Carlo methodswere coupled see Refs (6), (7), and (8). In this instance, theabove caveats still hold for the dis
27、crete ordinates calculation,but in addition, the variance of the Monte Carlo results mustnow be included with the overall assessment of the variance ofthe dosimetry data.5.2 Determination of Damage Exposure Values andUncertaintiesAdjustment procedures outlined in GuideE 944 and Guide E 1018 shall be
28、 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 for this application rather than just fluence(E 1 MeV) because Ref (5) not
29、es 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 Anna Power Station, Units 1 and 2,Summary of Meeting Held on September 19, 1
30、975 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 Commission, Oct. 22, 1979.(3) Unresolved Safety Issues Summary, NUREG-0606, Vo
31、l 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 Fracture-Safe PRV SupportDesign in Neutron Environments,” Transactions of th
32、e AmericanNuclear Society, Vol 30, 1978, pp. 187188.(6) Cain, V. R., “The Use of Monte Carlo withAlbedos to Predict NeutronStreaming in PWR Containment Buildings,” Transactions of theAmerican Nuclear Society, Vol 23, 1976, p. 618.(7) Straker, E. A., Stevens, P. N., Irving, D. C. and Cain, V. R., “Th
33、eMORSE 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 MonteCarlo Radiation Transport Calculations,” ORNL-4853, July 1973.E1035082AS
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37、 on Standards, at the address shown below.This 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).E1035083
