1、Designation: E1035 18Standard 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, the yea
2、r 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 i
3、nnuclear 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 radiation
4、 exposures.1.2 The values stated in SI units are to be regarded asstandard; units that are not SI can be found in TerminologyE170 and are to be regarded as standard. Any values inparentheses are for information only.1.3 This practice is applicable to all pressurized waterreactors whose vessel suppor
5、ts will experience a lifetimeneutron fluence (E 1 MeV) that exceeds 1 1017neutrons/cm2or exceeds 3.0 104dpa.(1)2(See Terminology E170.)1.4 Exposure of vessel support structures by gamma radia-tion is not included in the scope of this practice, but see thebrief discussion of this issue in 3.2.1.5 Thi
6、s 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, health, and environmental practices and deter-mine the applicability of regulatory limitations prior to use.(For e
7、xample (2)1.6 This international standard was developed in accor-dance with internationally recognized principles on standard-ization established in the Decision on Principles for theDevelopment of International Standards, Guides and Recom-mendations issued by the World Trade Organization TechnicalB
8、arriers to Trade (TBT) Committee.2. Referenced Documents2.1 ASTM Standards:3E170 Terminology Relating to Radiation Measurements andDosimetryE482 Guide for Application of Neutron Transport Methodsfor Reactor Vessel SurveillanceE693 Practice for Characterizing Neutron Exposures in Ironand Low Alloy St
9、eels in Terms of Displacements PerAtom (DPA)E844 Guide for Sensor Set Design and Irradiation forReactor SurveillanceE854 Test Method for Application and Analysis of SolidState Track Recorder (SSTR) Monitors for Reactor Sur-veillanceE910 Test Method for Application and Analysis of HeliumAccumulation
10、Fluence Monitors for Reactor Vessel Sur-veillanceE944 Guide for Application of Neutron Spectrum Adjust-ment Methods in Reactor SurveillanceE1005 Test Method for Application and Analysis of Radio-metric Monitors for Reactor Vessel SurveillanceE1018 Guide for Application of ASTM Evaluated CrossSection
11、 Data FileE2956 Guide for Monitoring the Neutron Exposure of LWRReactor Pressure Vessels2.2 ASME Standard:Boiler and Pressure Vessel Code, Section III42.3 Nuclear Regulatory Documents:Code of Federal Regulations, “Fracture ToughnessRequirements,” Chapter 10, Part 50, Appendix G5Code of Federal Regul
12、ations, “Reactor Vessel MaterialsSurveillance Program Requirements,” Chapter 10, Part50, Appendix H5Regulatory Guide 1.99, Rev.2,“Effects of Residual Ele-ments on Predicted Radiation Damage on Reactor Vessel1This practice is under the jurisdiction of ASTM Committee E10 on NuclearTechnology and Appli
13、cationsand is the direct responsibility of SubcommitteeE10.05 on Nuclear Radiation Metrology.Current edition approved June 1, 2018. Published July 2018. Originally approvedin 1985. Last previous edition approved in 2013 as E103513. DOI: 10.1520/E1035-18.2The boldface numbers in parentheses refer to
14、a list of references at the end ofthis practice.3For referenced ASTM standards, visit the ASTM website, www.astm.org, orcontact ASTM Customer Service at serviceastm.org. For Annual Book of ASTMStandards volume information, refer to the standards Document Summary page onthe ASTM website.4Available fr
15、om 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.Copyright ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959. United StatesThis inte
16、rnational standard was developed in accordance with internationally recognized principles on standardization established in the Decision on Principles for theDevelopment of International Standards, Guides and Recommendations issued by the World Trade Organization Technical Barriers to Trade (TBT) Co
17、mmittee.1Materials,” U. S. Nuclear Regulatory Commission, May198853. 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 2.3) and natio
18、nal standards groups (see 2.1 and 2.2)have promulgated regulations and standards to ensure safeoperation of these vessels. The support structures for pressur-ized water reactor vessels may also be subject to similarneutron radiation effects (1, 3-6).6The objective of thispractice is to provide guide
19、lines for determining the neutronradiation exposures experienced by individual vessel supports.3.2 It is known that high energy photons can also producedisplacement damage effects that may be similar to thoseproduced by neutrons. These effects are known to be much lessat the belt line of a light wat
20、er 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 determinewhether gamma-induced dpa
21、exceeds the screening level of 3.010-4, used in this practice for neutron exposures. (See 1.3).4. Irradiation Requirements4.1 Location of Neutron DosimetersNeutron dosimetersshall be located along the support structure in the region wherethe maximum dpa or fluence (E 1 MeV) is expected to occur,base
22、d 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 neutron fields within the actuallocati
23、on of the support structures. Considerations discussed inGuide E2956 (especially with regard to ex-vessel surveillanceprograms) apply.4.2 Neutron Dosimeters:4.2.1 Information regarding the selection of appropriatesensor sets for support structure application may be found inGuide E844, Test Method E1
24、005, and Test Methods E854 andE910.4.2.2 In particular, Test Method E910 also provides guid-ance for the additional possibility that operating plants may useexisting copper bearing instruments and cables within thereactor cavity as a retrospective passive dosimeter candidate.5. Determination of Neut
25、ron 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 E482, subject to these additional consider-ations that may be encountered in rea
26、ctor 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 qua
27、drature 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 Monte
28、 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 distributi
29、on functionsin the final Monte Carlo analysis, or to develop bias (weight-ing) factors for use in a complete Monte Carlo model. Fordetails of analyses in which discrete ordinates and Monte Carlomethods were coupled see Refs (7-12). Reference (13) pro-vides a review of the available combined or hybri
30、d discreteordinates/Monte Carlo calculations. For hybrid calculations,the above caveats still hold for the discrete ordinatescalculation, but in addition, the variance of the Monte Carloresults must now be included with the overall assessment of thevariance of the dosimetry data.5.2 Determination of
31、 Damage Exposure Values andUncertaintiesAdjustment procedures outlined in Guide E944and Guide E1018 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 Pract
32、ice E693. Dpashall be determined for this application rather than just fluence(E 1 MeV) because Ref (1) 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.6. Keyw
33、ords6.1 dosimetry; dpa; hybrid transport methods; neutron ex-posure; neutron fluence; radiometric monitor; reactor vesselsupports; surveillance6The boldface numbers in parentheses refer to a list of references at the end ofthis standard.E1035 182REFERENCES(1) Hopkins, W. C.,“Suggested Approach for F
34、racture-Safe PRV SupportDesign in Neutron Environments,” Transactions of the AmericanNuclear Society, Vol 30, 1978, pp. 187188.(2) Resolution of Generic Safety Issues, NUREG-0933, Rev. 3, Sec. 3,Issue 15, Radiation Effects on Reactor Vessel Supports.(3) Docket 50338-207, North Anna Power Station, Un
35、its 1 and 2,Summary of Meeting Held on September 19, 1975 on Dynamic Effectsof LOCAs, Sept. 22, 1975.(4) 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, 197
36、9.(5) Unresolved Safety Issues Summary, NUREG-0606, Vol 4, No. 4, TaskA-11: Reactor Vessel Materials Toughness, November, 1982.(6) Asymmetric Blowdown Loads on PWR Primary Systems, NUREG-0609, U.S. Nuclear Regulatory Commission, 1981.(7) Cain,V. R., “The Use of Monte Carlo withAlbedos to Predict Neu
37、tronStreaming in PWR Containment Buildings,” Transactions of theAmerican Nuclear Society, Vol 23, 1976, p. 618.(8) Straker, E. A., Stevens, P. N., Irving, D. C. and Cain, V. R., “TheMORSE CodeAMultigroup Neutron and Gamma-Ray Monte CarloTransport Code,” ORNL-4585, September 1970.(9) Emmett, M. B., B
38、urgart, C. E., and Hoffman, T. J., “DOMINO: AGeneral Purpose Code for Coupling Discrete Ordinates and MonteCarlo Radiation Transport Calculations,” ORNL-4853, July 1973.(10) Chen, J. and Fero, A. H. “Krko Neutron Streaming Analysis andMeasurement for Concrete Missile Shield Removal,” in PHYSOR2010 A
39、dvances in Reactor Physics to Power the NuclearRenaissance, May 9-14, 2010.(11) Chen, J. and Fero, A. H. “Krko Radiation Streaming Evaluation forGamma-ray Doses in Containment,” Transactions of the AmericanNuclear Society, Vol 107, 2011, pp. 957-959.(12) Kulesza, J. A., “A New Method for Coupling 2D
40、 and 3D Determin-istic and Stochastic Radiation Transport Calculations,” MastersThesis, University of Tennessee, 2011,http:/trace.tennessee.edu/utk_gradthes/992/.(13) Wagner, J. C., Peplow, D. E., Mosher, S. W., and Evans, T. M.,“Review of Hybrid (Deterministic/Monte Carlo) Radiation TransportMethod
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