1、Designation: E482 111E482 16Standard Guide forApplication of Neutron Transport Methods for ReactorVessel Surveillance, E706 (IID)Surveillance1This standard is issued under the fixed designation E482; the number immediately following the designation indicates the year oforiginal adoption or, in the c
2、ase of revision, 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 NOTEMissing references to Practice E693 were added to 3.2.1.6, 3.2.1.7 and 3.4.8.3 editorially in N
3、ovember 2012.1. Scope1.1 Need for Neutronics CalculationsAn accurate calculation of the neutron fluence and fluence rate at several locations isessential for the analysis of integral dosimetry measurements and for predicting irradiation damage exposure parameter values inthe pressure vessel. Exposur
4、e parameter values may be obtained directly from calculations or indirectly from calculations that areadjusted with dosimetry measurements; Guide E944 and Practice E853 define appropriate computational procedures.1.2 MethodologyNeutronics calculations for application to reactor vessel surveillance e
5、ncompass three essential areas: (1)validation of methods by comparison of calculations with dosimetry measurements in a benchmark experiment, (2) determinationof the neutron source distribution in the reactor core, and (3) calculation of neutron fluence rate at the surveillance position andin the pr
6、essure vessel.1.3 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibilityof the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatoryrequirements prior to use.2.
7、 Referenced Documents2.1 ASTM Standards:2E693 Practice for Characterizing Neutron Exposures in Iron and Low Alloy Steels in Terms of Displacements Per Atom (DPA),E 706(ID)E706 Master Matrix for Light-Water Reactor Pressure Vessel Surveillance Standards, E 706(0) (Withdrawn 2011)3E844 Guide for Senso
8、r Set Design and Irradiation for Reactor Surveillance, E 706 (IIC)E853 Practice for Analysis and Interpretation of Light-Water Reactor Surveillance ResultsE944 Guide for Application of Neutron Spectrum Adjustment Methods in Reactor Surveillance, E 706 (IIA)E1018 Guide for Application of ASTM Evaluat
9、ed Cross Section Data File, Matrix E706 (IIB)E2006 Guide for Benchmark Testing of Light Water Reactor Calculations2.2 Nuclear Regulatory Documents:4NUREG/CR-1861 LWR Pressure Vessel Surveillance Dosimetry Improvement Program: PCA Experiments and Blind TestNUREG/CR-3318 LWR Pressure Vessel Surveillan
10、ce Dosimetry Improvement Program: PCA Experiments, Blind Test, andPhysics-Dosimetry Support for the PSF ExperimentsNUREG/CR-3319 LWR Pressure Vessel Surveillance Dosimetry Improvement Program: LWR Power Reactor SurveillancePhysics-Dosimetry Data Base CompendiumNUREG/CR-5049 Pressure Vessel Fluence A
11、nalysis and Neutron Dosimetry3. Significance and Use3.1 General:1 This guide is under the jurisdiction ofASTM Committee E10 on NuclearTechnology andApplications and is the direct responsibility of Subcommittee E10.05 on NuclearRadiation Metrology.Current edition approved June 1, 2011July 1, 2016. Pu
12、blished June 2011August 2016. Originally approved in 1976. Last previous edition approved in 20072011 asE482 07E482 111. DOI: 10.1520/E0482-11E01.10.1520/E0482-16.2 For referencedASTM standards, visit theASTM website, www.astm.org, or contactASTM Customer Service at serviceastm.org. For Annual Book
13、of ASTM Standardsvolume information, refer to the standards Document Summary page on the ASTM website.3 The last approved version of this historical standard is referenced on www.astm.org.4 Available from Superintendent of Documents, U.S. Government Printing Office, Washington, DC 20402.This documen
14、t is not an ASTM standard and is intended only to provide the user of an ASTM standard an indication of what changes have been made to the previous version. Becauseit may not be technically possible to adequately depict all changes accurately, ASTM recommends that users consult prior editions as app
15、ropriate. In all cases only the current versionof the standard as published by ASTM is to be considered the official document.Copyright ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959. United States13.1.1 The methodology recommended in this guide specifies cr
16、iteria for validating computational methods and outlinesprocedures applicable to pressure vessel related neutronics calculations for test and power reactors. The material presented hereinis useful for validating computational methodology and for performing neutronics calculations that accompany reac
17、tor vesselsurveillance dosimetry measurements (see Master Matrix E706 and Practice E853). Briefly, the overall methodology involves: (1)methods-validation calculations based on at least one well-documented benchmark problem, and (2) neutronics calculations for thefacility of interest. The neutronics
18、 calculations of the facility of interest and of the benchmark problem should be performedconsistently, with important modeling parameters kept the same or as similar as is feasible. In particular, the same energy groupstructure and common broad-group microscopic cross sections should be used for bo
19、th problems. Further, the benchmark problemshould be characteristically similar to the facility of interest. For example, a power reactor benchmark should be utilized for powerreactor calculations. The neutronics calculations involve two tasks: (1) determination of the neutron source distribution in
20、 thereactor core by utilizing diffusion theory (or transport theory) calculations in conjunction with reactor power distributionmeasurements, and (2) performance of a fixed fission rate neutron source (fixed-source) transport theory calculation to determinethe neutron fluence rate distribution in th
21、e reactor core, through the internals and in the pressure vessel. Some neutronics modelingdetails for the benchmark, test reactor, or the power reactor calculation will differ; therefore, the procedures described herein aregeneral and apply to each case. (See NUREG/CR5049, NUREG/CR1861, NUREG/CR3318
22、 and NUREG/CR3319.)3.1.2 It is expected that transport calculations will be performed whenever pressure vessel surveillance dosimetry data becomeavailable and that quantitative comparisons will be performed as prescribed by 3.2.2. All dosimetry data accumulated that areapplicable to a particular fa
23、cility should be included in the comparisons.3.2 ValidationPrior to performing transport calculations for a particular facility, the computational methods must be validatedby comparing results with measurements made on a benchmark experiment. Criteria for establishing a benchmark experiment forthe p
24、urpose of validating neutronics methodology should include those set forth in Guides E944 and E2006 as well as thoseprescribed in 3.2.1. A discussion of the limiting accuracy of benchmark validation discrete ordinate radiation transport proceduresfor the LWR surveillance program is given in Referenc
25、eRef (1). Reference (2) provides details on the benchmark validation fora Monte Carlo radiation transport code.3.2.1 Requirements for BenchmarksIn order for a particular experiment to qualify as a calculational benchmark, the followingcriteria are recommended:3.2.1.1 Sufficient information must be a
26、vailable to accurately determine the neutron source distribution in the reactor core,3.2.1.2 Measurements must be reported in at least two ex-core locations, well separated by steel or coolant,3.2.1.3 Uncertainty estimates should be reported for dosimetry measurements and calculated fluences includi
27、ng calculatedexposure parameters and calculated dosimetry activities,3.2.1.4 Quantitative criteria, consistent with those specified in the methods validation 3.2.2, must be published anddemonstrated to be achievable,3.2.1.5 Differences between measurements and calculations should be consistent with
28、the uncertainty estimates in 3.2.1.3,3.2.1.6 Results for exposure parameter values of neutron fluence greater than 1 MeV and 0.1 MeV (E 1 MeV and 0.1 MeV)and of displacements per atom (dpa) in iron should be reported consistent with Practices E693 and E853.3.2.1.7 Reaction rates (preferably establis
29、hed relative to neutron fluence standards) must be reported for 237Np(n,f) or238U(n,f), and 58Ni(n,p) or 54Fe(n,p); additional reactions that aid in spectral characterization, such as provided by Cu, Ti, andCo-A1, should also be included in the benchmark measurements. The 237Np(n,f) reaction is part
30、icularly important because it issensitive to the same neutron energy region as the iron dpa. Practices E693 and E853 and Guides E844 and E944 discuss thiscriterion.3.2.2 Methodology ValidationIt is essential that the neutronics methodology employed for predicting neutron fluence in areactor pressure
31、 vessel be validated by accurately predicting appropriate benchmark dosimetry results. In addition, the followingdocumentation should be submitted: (1) convergence study results, and (2) estimates of variances and covariances for fluence ratesand reaction rates arising from uncertainties in both the
32、 source and geometric modeling. For Monte Carlo calculations, theconvergence study results should also include (3) an analysis of the figure-of-merit (FOM) as a function of particles history, andif applicable, (4) the description of the technique utilized to generate the weight window parameters.3.2
33、2.1 For example, model specifications for discrete-ordinates method on which convergence studies should be performedinclude: (1) neutron cross-sections or energy group structure, (2) spatial mesh, and (3) angular quadrature. One-dimensionalcalculations may be performed to check the adequacy of grou
34、p structure and spatial mesh. Two-dimensional calculations shouldbe employed to check the adequacy of the angular quadrature. A P3 cross section expansion is recommended along with a S8minimum quadrature.3.2.2.2 Uncertainties that are propagated from known uncertainties in nuclear data need to be ad
35、dressed in the analysis. Theuncertainty analysis for discrete ordinates codes may be performed with sensitivity analysis as discussed in References (3, 4). InMonte Carlo analysis the uncertainties can be treated by a perturbation analysis as discussed in Reference (5). Appropriatecomputer programs a
36、nd covariance data are available and sensitivity data may be obtained as an intermediate step in determininguncertainty estimates.55 Much of the nuclear covariance and sensitivity data have been incorporated into a benchmark database employed with the LEPRICON Code system. See Ref (6).E482 1623.2.2.
37、3 Effects of known uncertainties in geometry and source distribution should be evaluated based on the following test cases:(1) reference calculation with a time-averaged source distribution and with best estimates of the core, and pressure vessel locations,(2) reference case geometry with maximum an
38、d minimum expected deviations in the source distribution, and (3) reference casesource distribution with maximum expected spatial perturbations of the core, pressure vessel, and other pertinent locations.3.2.2.4 Measured and calculated integral parameters should be compared for all test cases. It is
39、 expected that larger uncertaintiesare associated with geometry and neutron source specifications than with parameters included in the convergence study. Problemsassociated with space, energy, and angle discretizations can be identified and corrected. Uncertainties associated with geometryspecificat
40、ions are inherent in the structure tolerances. Calculations based on the expected extremes provide a measure of thesensitivity of integral parameters to the selected variables. Variations in the proposed convergence and uncertainty evaluations areappropriate when the above procedures are inconsisten
41、t with the methodology to be validated.As-built data could be used to reducethe uncertainty in geometrical dimensions.3.2.2.5 In order to illustrate quantitative criteria based on measurements and calculations that should be satisfied, let denotea set of logarithms of calculation (Ci) to measurement
42、 (Ei) ratios. Specifically,5$qi:qi 5wilnCi/Ei!,i 51N% (1)where qi and N are defined implicitly and the wi are weighting factors. Because some reactions provide a greater response overa spectral region of concern than other reactions, weighting factors may be utilized when their selection method is w
43、ell documentedand adequately defended, such as through a least squares adjustment method as detailed in Guide E944. In the absence of the useof a least squares adjustment methodology, the mean of the set q is given byq 5 1N (i51Nqi (2)and the best estimate of the variance, S2, isS 2 5 1N 21 (i51N q
44、2 qi! 2 (3)3.2.2.6 The neutronics methodology is validated, if (in addition to qualitative model evaluation) all of the following criteria aresatisfied:(1) The bias, |q|, is less than 1,(2) The standard deviation, S, is less than 2,(3) All absolute values of the natural logarithmic of the C/E ratios
45、 (|q|, i = 1 . N) are less than 3, and(4) 1, 2, and 3 are defined by the benchmark measurement documentation and demonstrated to be attainable for all itemswith which calculations are compared.3.2.2.7 Note that a nonzero log-mean of the Ci/Ei ratios indicates that a bias exists. Possible sources of
46、a bias are: (1) sourcenormalization, (2) neutronics data, (3) transverse leakage corrections (if applicable), (4) geometric modeling, and (5 ) mathematicalapproximations. Reaction rates, equivalent fission fluence rates, or exposure parameter values for example, (E 1 MeV) anddpa may be used for vali
47、dating the computational methodology if appropriate criteria (that is, as established by 3.2.2.5 and 3.2.2.6)are documented for the benchmark of interest.Accuracy requirements for reactor vessel surveillance specific benchmark validationprocedures are discussed in Guide E2006. The validation testing
48、 for the generic discrete ordinates and Monte Carlo transportmethods is discussed in References (1, 2).3.2.2.8 One acceptable procedure for performing these comparisons is: (1) obtain group fluence rates at dosimeter locationsfrom neutronics calculations, (2) collapse the Guide E1018 recommended dos
49、imetry cross section data to a multigroup setconsistent with the neutron energy group fluence rates or obtain a fine group spectrum (consistent with the dosimetry cross sectiondata) from the calculated group fluence rates, (3) fold the energy group fluence rates with the appropriate cross sections, and (4)compare the calculated and experimental data according to the specified quantitative criteria.3.3 Determination of the Fixed Fission SourceThe power distribution in a typical power reactor undergoes significant changeduring the life of the rea
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