1、Designation: E 1006 08Standard 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 () 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 interpretat
3、ion 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, expo
4、-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 p
5、ower 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 2005, E
6、 2006 and 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 hea
7、lth practices and deter-mine the applicability of regulatory limitations prior to use.2. Referenced Documents2.1 ASTM Standards:2E 185 Practice for Design of Surveillance Programs forLight-Water Moderated Nuclear Power Reactor VesselsE 482 Guide for Application of Neutron Transport Methodsfor Reacto
8、r Vessel Surveillance, E706 (IID)E 560 Practice for Extrapolating Reactor Vessel Surveil-lance Dosimetry Results, E 706(IC)E 646 Test Method for Tensile Strain-Hardening Exponents(n -Values) of Metallic Sheet MaterialsE 693 Practice for Characterizing Neutron Exposures inIron and Low Alloy Steels in
9、 Terms of Displacements PerAtom (DPA), E 706(ID)E 706 Master Matrix for Light-Water Reactor PressureVessel Surveillance Standards, E 706(0)E 844 Guide for Sensor Set Design and Irradiation forReactor Surveillance, E 706(IIC)E 853 Practice for Analysis and Interpretation of Light-Water Reactor Survei
10、llance Results, E706(IA)E 854 Test Method for Application and Analysis of SolidState Track Recorder (SSTR) Monitors for Reactor Sur-veillance, E706(IIIB)E 900 Guide for Predicting Radiation-Induced TransitionTemperature Shift in Reactor Vessel Materials, E706 (IIF)E 910 Test Method for Application a
11、nd 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 for Application and Analysis of Radio-metric Monitors for Reactor Vessel Surveill
12、ance, E706(IIIA)E 1018 Guide for Application of ASTM Evaluated CrossSection Data File, Matrix E 706 (IIB)E 1035 Practice for Determining Neutron Exposures forNuclear Reactor Vessel Support StructuresE 2005 Guide for Benchmark Testing of Reactor Dosimetryin Standard and Reference Neutron FieldsE 2006
13、 Guide for Benchmark Testing of Light Water Reac-tor Calculations2.2 Nuclear Regulatory Documents:Code of Federal Regulations, “Fracture Toughness Require-ments,” Chapter 10, Part 50, Appendix G31This practice is under the jurisdiction of ASTM Committee E10 on NuclearTechnology and Applications and
14、is the direct responsibility of SubcommitteeE10.05 on Nuclear Radiation Metrology.Current edition approved Nov. 1, 2008. Published December 2008. Originallyapproved in 1984. Last previous edition approved in 2002 as E 1006 02.2The reference in parentheses refers to Section 5 as well as to Figs. 1 an
15、d 2 ofMatrix E 706.3Available from Superintendent of 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.Code of Federal Regulations, “Reactor Vessel MaterialsSurveillance P
16、rogram Requirements,” Chapter 10, Part50, Appendix H3Regulatory Guide 1.99, Rev 2, “Effects of Residual Ele-ments on Predicted Radiation Damage to Reactor VesselMaterials,” U.S. Nuclear Regulatory Commission, April197733. Significance and Use3.1 The mechanical properties of steels and other metals a
17、realtered by exposure to neutron radiation. These propertychanges are assumed to be a function of chemical composition,metallurgical condition, temperature, fluence (perhaps alsofluence rate), and neutron spectrum. The influence of thesevariables is not completely understood. The functional depen-de
18、ncy between property changes and neutron radiation issummarized in the form of damage exposure parameters thatare weighted integrals over the neutron fluence spectrum.3.2 The evaluation of neutron radiation effects on pressurevessel steels and the determination of safety limits require theknowlege o
19、f 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 0.1MeV, thermal neutron fluence, etc.). This practice describesrecommended procedures and data for determining theseexposure param
20、eters (and the associated uncertainties) for testreactor experiments.3.3 The nuclear industry draws much of its informationfrom databases that come from test reactor experiments.Therefore, it is essential that reliable databases are obtainedfrom test reactors to assess safety issues in Light Water R
21、eactor(LWR) nuclear power plants.4. Establishment of the Physics-Dosimetry Program4.1 Reactor Physics Computational Mode:4.1.1 IntroductionThis section provides a reference set ofprocedures for performing reactor physics calculations inexperimental test reactors. Although it is recognized thatvariat
22、ions in methods will occur at various facilities, thepresent benchmarked calculational sequence has been usedsuccessfully in several studies (1-4)4and provides proceduresfor performing physics calculations in test reactors. The MonteCarlo technique is used with about the same frequency asdiscrete or
23、dinates techniques in test and research reactordosimetry. The method is used more frequently in test/researchreactors, as compared to power reactors, because of the veryheterogeneous geometry often encountered in test/researchreactors. Very complex geometries can be handled in 3D spaceusing the Mont
24、e Carlo approach.4.2 Determination of Core Fission Source DistributionThe total fission source distribution, in source neutrons per unitvolume per unit time, defined as:Sx, y, z! 5*0nE!(fx, y, z, E!fx, y, z, E!dE (1)where:n(E) = number of neutrons per fission,(f= macroscopic fission cross section, a
25、ndf = fluence rate.is determined from a k-eigenvalue calculation of the reactorcore, with the neutron fluence rate normalized to give 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 = expe
26、rimentally determined thermal power with theintegral calculated over all energies E and the corevolume V.4.2.1 An accurate value for 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 partiall
27、y inserted control rod, or from burnupeffects, then a three-dimensional k calculation is required. Withthe computing capability of today, multigroup discrete ordi-nates or Monte Carlo is used almost exclusively to model thecore (that is, not few group diffusion theory). This is particu-larly importa
28、nt where there are special purpose loops in thecore or at a reflector/core boundary where the spectrum of theflux changes very rapidly. In these cases, the few groupdiffusion models are typically not adequate.4.2.3 Whenever the axial shape of the neutron fluence rate isseparable from the shape in th
29、e other variables, then a fullthree-dimensional calculation is not required. In many experi-mental reactors, the axial dependence of the fluence rate is wellapproximated by a cosine shifted slightly from the midplane. Inthis case only a two-dimensional calculation (with a bucklingapproximation for a
30、xial leakage) is needed. In this case it ispossible to use two-dimensional transport theory.4.2.4 For reactor cores that generate a non-negligibleamount of thermal power, the shape of the fission source maychange with time due to burnup and changes in control rodpositions. In this case, the source s
31、hould be averaged over thetime period during which the experiment was performed.4.2.5 If a few-group set is used to model the fission sourcedistribution, it is recommended that a fine-group cross-sectionlibrary of approximately 100 groups with at least 10 thermalgroups be used to generate the few-gr
32、oup set. Resonanceshielding of the fine-group cross sections can be done with anyof the methods acceptable for LWR analysis (5) (shieldingfactor, Nordheim, integral transport theory, etc.). The fine-group cross-section library shall be collapsed with weightingspectra obtained from cell calculations
33、for each type of unit cellfound in the core. If experiments are located near control rodsor reflectors, then a separate calculation shall be performed foradjacent cells to account for the influence of these regions onthe thermal spectrum in the experiment.4.3 Transport Calculations-Discrete Ordinate
34、s Method:4.3.1 Transport calculations for test reactors may be per-formed by discrete ordinates or Monte Carlo methods, or by acombination of the two. The use of Monte Carlo codes isdescribed in 4.5. If discrete ordinates methods are used, it isrecommended that a multi-dimensional (2D or 3D) discret
35、eordinates code such as DORT/TORT (6) or DANTSYS (7) be4The boldface numbers in parentheses refer to the list of references appended tothis practice.E1006082used for the transport theory calculations of both in-core andex-core dosimeters. At least an S8 order quadrature with a P3cross section expans
36、ion should be used. The space-dependentfission source from the core calculation is input as a volumetricdistributed source with a fission spectrum energy distribution.It is recommended that the ENDF/B-VI representation (8) ofthe235U thermal fission spectrum (MAT 9228, MF 5, MT 18),which is based on
37、the Madland-Nix formalism (9) be used torepresent the fission neutron energy distribution. The latestapplicable ENDF/B cross section data files shall be used (8,10).If a three-dimensional discrete ordinates transport code is notused, it is recommended that the three-dimensional fluence ratedistribut
38、ion be synthesized from two two-dimensional calcula-tions. A simple synthesis procedure that has been found toproduce accurate results in benchmark dosimetry calculationsis given in (2,3).4.3.2 This synthesis procedure has been used successfully ina number of experiments in which the ex-core configu
39、ration isuniform axially along the full core height. For these types ofproblems, the three-dimensional synthesized fluence rates givedosimeter reactions that agree to within 10 % of the measuredvalues, even off the core midplane. However, for experimentsthat contain short (relative to the core heigh
40、t) attenuatingbodies, neutron streaming may occur around the edges of thebody, and this effect is not well-predicted with the synthesisprocedure. A “leakage iteration” procedure has been developedfor such problems (11), but since most experiments do notexperience this difficulty, it will not be disc
41、ussed in thispractice.4.4 Calculation of Bias Factors:4.4.1 In order to reduce the number of mesh intervals in thetwo-dimensional discrete ordinates calculations, it is oftennecessary to smear some detailed structure into a homogeneousmixture or completely ignore it. The experimental data com-puted
42、with the homogeneous two-dimensional model can becorrected for the effects of local heterogeneities with biasfactors. An example in which bias factors may be useful is incorrecting for fluence rate perturbations caused by the experi-ment itself. This factor has been observed to be as high as 1.3for
43、a 1-in2. container in an ex-core location. For in-coreexperiments the effects of heterogeneities within the experi-mental assembly should be examined.4.4.2 Bias factors can be obtained with detailed one-dimensional (usually cylindrical) discrete ordinates calcula-tions (12) in the vicinity of the de
44、sired data. Two cellcalculations are usually done: one in which the experiment ismodeled with as much detail as possible, and the other inwhich it is smeared in the same manner as in the two-dimensional calculation. In both the heterogeneous and homo-geneous cases, the experiment zone should be surr
45、ounded by ahomogenized zone corresponding to the same material whichsurrounds the experiment in the two-dimensional model. Thisregion should be several mean free paths thick. It is recom-mended that the discrete ordinates calculations be performed asboundary source problems with an isotropic fluence
46、 rateboundary condition which is equal to the corresponding scalarfluence rate from the two-dimensional calculation. Group-dependent bias factors for the experiment zone are defined asthe ratio of the group fluence rates for the heterogeneous andhomogeneous geometries. These bias factors should mult
47、iplythe multigroup fluence rates for the experiment zone in thetwo-dimensional calculation.4.5 Transport CalculationsMonte Carlo Method:4.5.1 While this practice permits the use of a discrete-ordinates technique for test reactor analysis (4.3), the alterna-tive Monte Carlo technique may be preferred
48、 in many situa-tions. This approach has the inherent advantage, over thedeterministic method described in 4.3, of being able to treatthree-dimensional aspects as well as geometrical complexity inexplicit detail. Equally, there are drawbacks, and these have ledto the limited use of Monte Carlo analys
49、is in test reactordosimetry in the U.S., although it has been used effectively inthe United Kingdom (13). Three Monte Carlo codes used forreactor analysis are MCNP (14). MCBEND (15) and TRIPOLI(16).4.5.2 The Monte Carlo technique may be employed for theproduction of detailed core power distributions (for example,“eigenvalue” calculations). A dosimetry analysis, using theMonte Carlo method, however, may be initiated by way of adiffusion theory core-source calculation as described in 4.2.4.5.3 A relevant restriction of Monte Carlo lies in thedifficulty of calculating reaction
