1、Designation: E 1006 02Standard Practice forAnalysis and Interpretation of Physics Dosimetry Resultsfor Test Reactors, E 706(II)1This standard is issued under the fixed designation E 1006; the number immediately following the designation indicates the year oforiginal adoption or, in the case of revis
2、ion, the year 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 the methodology summarized inAnnex A1 to be used in the analysis and interpreta
3、tion ofphysics-dosimetry results from test reactors.1.2 This practice relies on, and ties together, the applicationof several supporting ASTM standard practices, guides, andmethods.1.3 Support subject areas that are discussed include reactorphysics calculations, dosimeter selection and analysis, exp
4、o-sure units, and neutron spectrum adjustment methods.1.4 This practice is directed towards the development andapplication of physics-dosimetry-metallurgical data obtainedfrom test reactor irradiation experiments that are performed insupport of the operation, licensing, and regulation of LWRnuclear
5、power plants. It specifically addresses the physics-dosimetry aspects of the problem. Procedures related to theanalysis, interpretation, and application of both test and powerreactor physics-dosimetry-metallurgy results are addressed inPractices E 185, E 560, E 853, and E 1035, Guides E 900,E 2005E
6、2006and Test Method E 646.1.5 This standard may involve hazardous materials, opera-tions, and equipment. This standard does not purport toaddress all of the safety concerns, if any, associated with itsuse. It is the responsibility of the user of this standard toestablish appropriate safety and healt
7、h practices and deter-mine the applicability of regulatory limitations prior to use.2. Referenced Documents2.1 ASTM Standards:E 185 Practice for Conducting Surveillance Tests for LightWater-Cooled Nuclear Power Reactor Vessels, E 706(IF)2,3E 482 Guide for Application of Neutron Transport Methodsfor
8、Reactor Vessel Surveillance, E 706 (IID)2,3E 560 Practice for Extrapolating Reactor Vessel Surveil-lance Dosimetry Results, E 706 (IC)2,3E 646 Test Method for Tensile Strain-Hardening Exponents(n-Values) of Metallic Sheet Materials4E 693 Practice for Characterizing Neutron Exposures inIron and Low A
9、lloy Steels in Terms of Displacements PerAtom (DPA), E 706 (ID)2,3E 706 Master Matrix for Light-Water Reactor PressureVessel Surveillance Standards, E 706 (O)3E 844 Guide for Sensor Set Design and Irradiation forReactor Surveillance, E 706 (IIC)2,3E 853 Practice for Analysis and Interpretation of Li
10、ght-Water Reactor Surveillance Results, E 706 (IA)2,3E 854 Test Method for Application and Analysis of SolidState Track Recorder (SSTR) Monitors for Reactor Sur-veillance, E 706 (IIIB)2,3E 900 Guide for Predicting Neutron Radiation Damage toReactor Vessel Materials, E 706 (IIF)2,3E 910 Specification
11、 for Application and Analysis of HeliumAccumulation Fluence Monitors for Reactor Vessel Sur-veillance, E 706 (IIIC)2,3E 944 Guide for Application of Neutron Spectrum Adjust-ment Methods in Reactor Surveillance, E 706 (IIA)2,3E 1005 Test Method for Application and Analysis of Radio-metric Monitors fo
12、r Reactor Vessel Surveillance, E 706(IIIA)2,3E 1018 Guide for Application of ASTM Evaluated CrossSection Data File, E 706 (IIB)2,3E 1035 Practice for Determining Radiation Exposures forNuclear Reactor Vessel Support Structures2E 2005 Guide for the Benchmark Testing of Reactor Do-simetry in Standard
13、and Reference Neutron Field, E 706(IIE-I)2,3E 2006 Guide for the Benchmark Testing of LWR Calcula-tions, E 706 (IIE-2)2,32.2 Nuclear Regulatory Documents:Code of Federal Regulations, “Fracture Toughness Require-ments,” Chapter 10, Part 50, Appendix G5Code of Federal Regulations, “Reactor Vessel Mate
14、rialsSurveillance Program Requirements,” Chapter 10, Part50, Appendix H51This 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 June 10, 2002. P
15、ublished September 2002. Originallypublished as E 1006 84. Last previous edition E 1006 96.2The reference in parentheses refers to Section 5 as well as to Figs. 1 and 2 ofMatrix E 706.3Annual Book of ASTM Standards, Vol 12.02.4Annual Book of ASTM Standards, Vol 03.01.5Available from Superintendent o
16、f Documents, U.S. Government PrintingOffice, Washington, DC 20402.1Copyright ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959, United States.Regulatory Guide 1.99, Rev 2, “Effects of Residual Ele-ments on Predicted Radiation Damage to Reactor VesselMaterials,”
17、 U.S. Nuclear Regulatory Commission, April197753. Significance and Use3.1 The mechanical properties of steels and other metals arealtered by exposure to neutron radiation. These propertychanges are assumed to be a function of chemical composition,metallurgical condition, temperature, fluence (perhap
18、s alsofluence rate), and neutron spectrum. The influence of thesevariables is not completely understood. The functional depen-dency between property changes and neutron radiation issummarized in the form of damage exposure parameters thatare weighted integrals over the neutron fluence spectrum.3.2 T
19、he evaluation of neutron radiation effects on pressurevessel steels and the determination of safety limits require theknowlege of uncertainties in the prediction of radiation expo-sure parameters (for example, dpa (Practice E 693), neutronfluence greater than 1.0 MeV, neutron fluence greater than 0.
20、1MeV, thermal neutron fluence, etc.). This practice describesrecommended procedures and data for determining theseexposure parameters (and the associated uncertainties) for testreactor experiments.3.3 The nuclear industry draws much of its informationfrom databases that come from test reactor experi
21、ments.Therefore, it is essential that reliable databases are obtainedfrom test reactors to assess safety issues in Light Water Reactor(LWR) nuclear power plants.4. Establishment of the Physics-Dosimetry Program4.1 Reactor Physics Computational Mode:4.1.1 IntroductionThis section provides a reference
22、 set ofprocedures for performing reactor physics calculations inexperimental test reactors. Although it is recognized thatvariations in methods will occur at various facilities, thepresent benchmarked calculational sequence has been usedsuccessfully in several studies (1-4)6and provides proceduresfo
23、r performing physics calculations in test reactors. Emphasisin these guidelines is placed on use of deterministic methods,but a short discussion of Monte Carlo techniques is alsoincluded.4.2 Determination of Core Fission Source DistributionThe total fission source distribution, in source neutrons pe
24、r unitvolume per unit time, defined as:Sx, y, z! 50*nE!(fx, y, z, E!fx, y, z, E!dE (1)where:n(E) = number of neutrons per fission,(f= macroscopic fission cross section, andf = fluence rate.is determined from a k-eigenvalue calculation of the reactorcore, with the neutron fluence rate normalized to g
25、ive thecorrect measured power output from the reactor, for example:P 5 *E*Vk(fx, y, z, E!fx, y, z, E!dxdydzdE (2)where:k = effective energy yield per fission, andP = experimentally determined thermal power with theintegral calculated over all energies E and the corevolume v.4.2.1 An accurate value f
26、or the reactor power, P, is impera-tive for absolute comparison with experimental data.4.2.2 If the axial core configuration is nonuniform, as mightresult from a partially inserted control rod, or from burnupeffects, then a three-dimensional k calculation is required. Thismay be calculated with a di
27、ffusion theory code such asVENTURE (5) or PDQ7 (6) using a few energy groups ( 1.0 and 0.1MeV, dpa, etc.) with assigned uncertainties.4.7.3 It is recommended to perform at least one dummyexperiment for each series of associated metallurgical experi-ments. The advantage of the dummy experiment is tha
28、t itallows greater latitude in the placement of dosimeters and thechoice of irradiation time. Thus, a larger variety of dosimetrysensors may be used providing a more detailed determinationof the fluence spectrum. However, in-situ dosimeters must alsobe placed in the metallurgical experiments to dete
29、rmine di-rectly the fluence exposure to the metallurgical specimen.4.7.4 Dosimeters used in both the dummy and metallurgicalexperiments are typically passive radiometric (foil) dosimeters.Other types of dosimeters (for example, solid state trackrecorders (SSTR), helium accumulation fluence monitors(
30、HAFM), and damage monitors (DM) should be added when-ever appropriate. Situations may arise for longer irradiationswhere some radiometric dosimeters will be ineffective due toshort half-life of the reaction product (see 4.7.5). There are twotypes of dosimeter sets that shall be used concurrently in
31、eachexperiment.4.7.4.1 Multiple Foil (MF) DosimetersThe MFs contain avariety of sensor materials appropriately encapsulated and areprimarily used to determine the energy dependence of theneutron spectra.4.7.4.2 Gradient Wires (GW)The GWs are dosimeters,generally in the form of wires that cover, in a
32、ll directions to thelargest extent possible, the dummy or metallurgical experimentin order to determine the spatial distribution of the neutronfluence. Typically, the54Fe(n, p) reaction (together withthe58Fe(n, g) reaction) is chosen for GW, but other reactionsand more than one material may be used
33、as appropriate.4.7.5 Dosimetry sensors shall be chosen whose reactioncross sections match as closely as possible the responsefunctions of the exposure parameters. The237Np(n, f ) and93Nb(n, n8) reactions are best suited for the determination ofdpa. The115In(n, n8) and103Rh(n, n8) reactions have thre
34、sholdsnear 1.0 MeV and are therefore well suited for the determina-tion of f 1.0 MeV. However, these two sensors can be usedonly in dummy experiments owing to the short half-life of theproduct isotopes. Two other important reactions are238U(n, f )and54Fe(n, p), but with responses above ;1 MeV and ;2
35、MeV, respectively. The addition of the HAFM reactions S(n,He), Ca(n, He), and N(n, He) could prove beneficial. Althoughexperimental testing is still required, the available cross-section data for the latter three reactions indicate some lowenergy sensitivity. In addition, the reaction product, He, i
36、sstable, thus eliminating half-life corrections.4.7.6 The other dosimetry sensors selected shall have re-sponse functions and threshold that are as diverse as possible incovering the neutron energy range of interest up to about 20MeV. It has been reported that using least squares adjustmenttechnique
37、s, exposure parameter values can be obtained atdosimeter locations with estimated uncertainties in the range of5to15%(1s) by using all three of the237Np(n, f ),238U(n, f ),and54Fe(n, p) reactions; in the range of 10 to 20 % (1s)byusing the latter two reactions; and only in the range of 20 to30 % (1s
38、)ifthe54Fe(n, p) reaction alone were to be used; seeRefs (14, 19,20). It is recommended to use at least six differentreactions for each MF set. Suitable sensors with associatedthresholds and other pertinent information are discussed inGuide E 844, Specification E 910, and Test Methods E 1005and E 85
39、4. See also Refs (14, 19, 20,21,22,23,24) for typicalMF sets and adjustment code results.4.8 Estimation of Neutron Exposure Parameters:4.8.1 Reports on the results of metallurgical irradiationexperiments shall contain the estimates for the uncertainties inthe determination of neutron exposure parame
40、ter values in theform of variances (or standard deviations) and covariances (orcorrelations). These data are necessary to perform reliable testsof damage models and to ensure consistency in data bankscomprising large numbers of metallurgical experiments fromtest reactors. An excellent discussion of
41、the uncertainties inneutron transport calculations of neutron exposure parameterscan be found in Refs (25) and (26).E10060244.8.2 Credible uncertainty data are very difficult to obtainfrom calculated spectra alone (see 4.6). The combination ofcalculations and appropriate dosimetry measurements bymea
42、ns of a least squares adjustment method greatly improvesthe values and reliability of uncertainty data as discussed in4.7.5 (see Guides E 482, E 944, E 1018, E 2005, and E 2006).4.8.3 The application of a least squares adjustment methodserves threefold purposes each of which is equally important:4.8
43、.3.1 Determination of the best (maximum likelihood orminimum variance) estimate for the damage exposure param-eter values.4.8.3.2 Determination of uncertainty bounds for these pa-rameters.4.8.3.3 Test for consistency for all input data.4.8.4 Each of the determinations and tests in 4.8.3.1-4.8.3.3sha
44、ll be performed and reported as recommended in GuideE 944. State-of-the-art information on the development, test-ing, and application of adjustment methods is provided in Refs( 18-26).5. Documentation5.1 The documentation of test reactor physics-dosimetryresults shall include the following items:5.1
45、.1 A complete spatial map of the exposure parametervalues dpa, f 1.0 MeV, f 0.1 MeV (and others, if needed)including a scheme to interpolate between spatial mesh points.5.1.2 Uncertainties of the exposure parameter values asexplained in 4.8. (These uncertainties are expected to be in therange of 5 t
46、o 15 %, 1s standard deviation, if appropriatedosimetry measurements have been performed. An explanationshall be provided if these values are exceeded in eitherdirection).5.1.3 Description of the methodology used including proce-dures for assigning input uncertainties.5.2 The following information sh
47、all also be available in theform of an appendix for possible use in later reviews. At a veryminimum, it shall be kept in archives if it is not included in themain report.5.2.1 The documentation of all dosimeter sensor QA results,as-built dosimeters, dosimetry, capsules, irradiation test rig,and the
48、replacement of dosimetry and metallurgy; including x,y, z, or r, u, z coordinates for each dosimetry sensor andmetallurgy specimen.5.2.2 The documentation of the test reactor components,as-built core region and test region dimensions, materials, andirradiation history.5.2.3 Nuclear data and constant
49、s used, raw measurementdata, derived dosimetry sensor reactions and reaction rates, andauxiliary computations with intermediate results and verifica-tion procedures.6. Keywords6.1 discrete ordinates; dosimetry; Monte Carlo; neutronexposure parameters; radiation transport; test reactorANNEX(Mandatory Information)A1. METHODOLOGY FOR THE ANALYSIS AND INTERPRETATION OF PHYSICS-DOSIMETRY RESULTS FROM TESTREACTORSA1.1 Establish a physics-dosimetry program in parallelwith material irradiation experiments which are designed tocorrelate damage in test specimens with neutron expos
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