1、Designation: E 721 07Standard Guide forDetermining Neutron Energy Spectra from Neutron Sensorsfor Radiation-Hardness Testing of Electronics1This standard is issued under the fixed designation E 721; the number immediately following the designation indicates the year oforiginal adoption or, in the ca
2、se of revision, 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.This standard has been approved for use by agencies of the Department of Defense.1. Scope1.1 This gui
3、de covers procedures for determining theenergy-differential fluence spectra of neutrons used inradiation-hardness testing of electronic semiconductor devices.The types of neutron sources specifically covered by this guideare fission or degraded energy fission sources used in either asteady-state or
4、pulse mode.1.2 This guide provides guidance and criteria that can beapplied during the process of choosing the spectrum adjust-ment methodology that is best suited to the available data andrelevant for the environment being investigated.1.3 This guide is to be used in conjunction with Guide E 720to
5、characterize neutron spectra and is used in conjunction withPractice E 722 to characterize damage-related parameters nor-mally associated with radiation-hardness testing of electronic-semiconductor devices.NOTE 1Although Guide E 720 only discusses activation foil sensors,any energy-dependent neutron
6、-responding sensor for which a responsefunction is known may be used (1).2NOTE 2For terminology used in this guide, see Terminology E 170.1.4 The values stated in SI units are to be regarded as thestandard.1.5 This standard does not purport to address all of thesafety concerns, if any, associated wi
7、th its use. It is theresponsibility of the user of this standard to establish appro-priate safety and health practices and determine the applica-bility of regulatory limitations prior to use.2. Referenced Documents2.1 ASTM Standards:3E 170 Terminology Relating to Radiation Measurementsand DosimetryE
8、 261 Practice for Determining Neutron Fluence, FluenceRate, and Spectra by Radioactivation TechniquesE 262 Test Method for Determining Thermal Neutron Re-action and Fluence Rates by Radioactivation TechniquesE 263 Test Method for Measuring Fast-Neutron ReactionRates by Radioactivation of IronE 264 T
9、est Method for Measuring Fast-Neutron ReactionRates by Radioactivation of NickelE 265 Test Method for Measuring Reaction Rates andFast-Neutron Fluences by Radioactivation of Sulfur-32E 266 Test Method for Measuring Fast-Neutron ReactionRates by Radioactivation of AluminumE 393 Test Method for Measur
10、ing Reaction Rates byAnaly-sis of Barium-140 From Fission DosimetersE 523 Test Method for Measuring Fast-Neutron ReactionRates by Radioactivation of CopperE 526 Test Method for Measuring Fast-Neutron ReactionRates by Radioactivation of TitaniumE 704 Test Method for Measuring Reaction Rates by Ra-dio
11、activation of Uranium-238E 705 Test Method for Measuring Reaction Rates by Ra-dioactivation of Neptunium-237E 720 Guide for Selection and Use of Neutron Sensors forDetermining Neutron Spectra Employed in Radiation-Hardness Testing of ElectronicsE 722 Practice for Characterizing Neutron Energy Fluenc
12、eSpectra in Terms of an Equivalent Monoenergetic NeutronFluence for Radiation-Hardness Testing of ElectronicsE 844 Guide for Sensor Set Design and Irradiation forReactor Surveillance, E 706(IIC)E 944 Guide for Application of Neutron Spectrum Adjust-ment Methods in Reactor Surveillance, E 706 (IIA)E
13、1018 Guide for Application of ASTM Evaluated CrossSection Data File, Matrix E 706 (IIB)E 1297 Test Method for Measuring Fast-Neutron ReactionRates by Radioactivation of NiobiumE 1855 Test Method for Use of 2N2222A Silicon BipolarTransistors as Neutron Spectrum Sensors and Displace-ment Damage Monito
14、rs1This guide is under the jurisdiction of ASTM Committee E10 on NuclearTechnology and Applications and is the direct responsibility of SubcommitteeE10.07 on Radiation Dosimetry for Radiation Effects on Materials and Devices.Current edition approved Feb. 1, 2007. Published March 2007. Originallyappr
15、oved in 1980. Last previous edition approved in 2001 as E 721 01.2The boldface numbers in parentheses refer to the list of references at the end ofthis guide.3For referenced ASTM standards, visit the ASTM website, www.astm.org, orcontact ASTM Customer Service at serviceastm.org. For Annual Book of A
16、STMStandards volume information, refer to the standards Document Summary page onthe ASTM website.1Copyright ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959, United States.3. Terminology3.1 Definitions: The following list defines some of thespecial terms used
17、in this guide:3.1.1 effectthe characteristic which changes in the sensorwhen it is subjected to the neutron irradiation. The effect maybe the reactions in an activation foil.3.1.2 responsethe magnitude of the effect. It can be themeasured value or that calculated by integrating the responsefunction
18、over the neutron fluence spectrum. The response is anintegral parameter. Mathematically, the response, R =(iRi,where Riis the response in each differential energy region at Eiof width DEi.3.1.3 response functionthe set of values of Riin eachdifferential energy region divided by the neutron fluence i
19、n thatdifferential energy region, that is, the set fi= Ri/(F(Ei)DEi).3.1.4 sensoran object or material (sensitive to neutrons)the response of which is used to help define the neutronenvironment. A sensor may be an activation foil.3.1.5 spectrum adjustmentthe process of changing theshape and magnitud
20、e of the neutron energy spectrum so thatquantities integrated over the spectrum agree more closely withtheir measured values. Other physical constraints on thespectrum may be applied.3.1.6 trial functiona neutron spectrum which, when inte-grated over sensor response functions, yields calculated re-s
21、ponses that can be compared to the corresponding measuredresponses.3.1.7 prior spectruman estimate of the neutron spectrumobtained by transport calculation or otherwise and used asinput to a least-squares adjustment.3.2 Abbreviations:3.2.1 DUTdevice under test.3.2.2 ENDFevaluated nuclear data file.3
22、.2.3 NNDCNational Nuclear Data Center (atBrookhaven National Laboratory).3.2.4 RSICCRadiation Safety Information ComputationCenter (at Oak Ridge National Laboratory).3.2.5 TREEtransient radiation effects on electronics.4. Significance and Use4.1 It is important to know the energy spectrum of thepart
23、icular neutron source employed in radiation-hardness test-ing of electronic devices in order to relate radiation effects withdevice performance degradation.4.2 This guide describes the factors which must be consid-ered when the spectrum adjustment methodology is chosen andimplemented. Although the s
24、election of sensors (foils) and thedetermination of responses (activities) is discussed in GuideE 720, the experiment should not be divorced from theanalysis. In fact, it is advantageous for the analyst conductingthe spectrum determination to be closely involved with thedesign of the experiment to e
25、nsure that the data obtained willprovide the most accurate spectrum possible. These datainclude the following : (1) measured responses such as theactivities of foils exposed in the environment and their uncer-tainties, (2) response functions such as reaction cross sectionsalong with appropriate corr
26、elations and uncertainties, (3) thegeometry and materials in the test environment, and (4) a trialfunction or prior spectrum and its uncertainties obtained froma transport calculation or from previous experience.5. Spectrum Determination With Neutron Sensors5.1 Experiment Design:5.1.1 The primary ob
27、jective of the spectrum characteriza-tion experiment should be the acquisition of a set of responsevalues (activities) from effects (reactions) with well-characterized response functions (cross sections) with re-sponses which adequately define (as a set) the fluence values atenergies to which the de
28、vice to be tested is sensitive. Forsilicon devices in fission-driven environments the significantneutron energy range is usually from 10 keV to 15 MeV. Listsof suitable reactions along with approximate sensitivity rangesare included in Guide E 720. Sensor set design is alsodiscussed in Guide E 844.
29、The foil set may include the use ofresponses with sensitivities outside the energy ranges neededfor the DUT to aid in interpolation to other regions of thespectrum. For example, knowledge of the spectrum below 10keV helps in the determination of the spectrum above thatenergy.5.1.2 An example of the
30、difficulty encountered in ensuringresponse coverage (over the energy range of interest) is thefollowing: If fission foils cannot be used in an experimentbecause of licensing problems, cost, or radiological handlingdifficulties (especially with235U,237Np or239Pu), a large gapmay be left in the foil s
31、et response between 100 keV and 2MeVa region important for silicon and gallium arsenidedamage (see Figs. A1.1 and A2.3 of Practice E 722). In thiscase two options are available. First, seek other sensors to fillthe gap (such as silicon devices sensitive to displacementeffects (see Test Method E 1855
32、),93Nb(n,n8)93mNb (see TestMethod E 1297)or103Rh(n,n8)103mRh. Second, devote thenecessary resources to determine a trial function that is close tothe real spectrum. In the latter case it may be necessary to carryout transport calculations to generate a prior spectrum whichincorporates the use of unc
33、ertainty and covariance information.5.1.3 Other considerations that affect the process of plan-ning an experiment are the following:5.1.3.1 Are the fluence levels low and of long duration sothat only long half-life reactions are useful? This circumstancecan severely reduce the response coverage of t
34、he foil set.5.1.3.2 Are high gamma-ray backgrounds present which canaffect the sensors (or affect the devices to be tested)?5.1.3.3 Can the sensors be placed so as to ensure equalexposure? This may require mounting the sensors on a rotatingfixture in steady-state irradiations or performing multiplei
35、rradiations with monitor foils to normalize the fluence be-tween runs.5.1.3.4 Does the DUT perturb the neutron spectrum?5.1.3.5 Can the fluence and spectrum seen in the DUT testlater be directly scaled to that determined in the spectrumcharacterization experiment (by monitors placed with thetested d
36、evice)?5.1.3.6 Can the spectrum shape and intensity be character-ized by integral parameters that permit simple intercomparisonof device responses in different environments? Silicon is asemiconductor material whose displacement damage functionE721072is well established. This makes spectrum parameter
37、ization fordamage predictions feasible for silicon.5.1.3.7 What region of the spectrum contributes to theresponse of the DUT? In other words, is the spectrum welldetermined in all energy regions that affect device perfor-mance?5.1.3.8 How is the counting system set up for the determi-nation of the a
38、ctivities? For example, are there enough countersavailable to handle up to 25 reactions from a single exposure?(This may require as many as six counters.) Or can theavailable system only handle a few reactions before theactivities have decayed below detectable limits?5.1.4 Once the experimental oppo
39、rtunities and constraintshave been addressed and the experiment designed to gather themost useful data, a spectrum adjustment methodology must bechosen.5.2 Spectrum Adjustment Methodology:5.2.1 After the basic measured responses, response func-tions, and trial spectrum information have been assemble
40、d,apply a suitable spectrum adjustment procedure to reach a“solution” that is as compatible as possible with that informa-tion. It must also meet other constraints such as, the fluencespectrum must be positive and defined for all energies. Thesolution is the energy-dependent spectrum function, F(E),
41、which approximately satisfies the series of Fredholm equationsof the first kind represented by Eq 1 as follows:Rj5*0sjE!FE! dE 1#j#n (1)where:Rj= measured response of sensor j,sj(E) = neutron response function at energy E for sensorj,F(E) = incident neutron fluence versus energy, andn = number of se
42、nsors which yield n equations.One important characteristic of this set of equations is thatwith a finite number of sensors, j, which yield n equations,there is no unique solution. With certain restrictions, however,the range of physically reasonable solutions can be limited toan acceptable degree.NO
43、TE 3Guides E 720 and E 844 provide general guidance on obtain-ing a suitable set of responses (activities) when foil monitors are used.Practice E 261 and Test Method E 262 provide more information on thedata analysis that generally is part of an experiment with activationmonitors. Specific instructi
44、ons for some individual monitors can be foundin Test Methods E 263 (iron), E 264 (nickel), E 265 (sulfur-32), E 266(aluminum), E 393 (barium-140 from fission foils), E 523 (copper), E 526(titanium), E 704 (uranium-238), E 705 (neptunium-237), E 1297 (nio-bium).5.2.2 Neutron spectra generated from se
45、nsor response datamay be obtained with either of two types of spectrum adjust-ment codes. One type is the iterative method, an example ofwhich is SAND II (2). The second is least squares minimiza-tion used by codes such as LSL-M2 (3). If used properly andwith sufficient, high-quality data, the two m
46、ethods will usuallyyield nearly the same values (610 to 15 %) for the primaryintegral parameters discussed in E 722.NOTE 4Another class of codes often referred to as Maximum Entropymay also prove useful for this type of analysis. These have historically notbeen used to estimate spectra for radiation
47、-damage purposes.5.2.3 Appendix X1 and Appendix X2 discuss in some detailthe implementation and the advantages and disadvantages ofthe two approaches as represented by SAND II and LSL-M2.5.3 Iterative Code Characteristics:5.3.1 The “iterative” codes use a trial function supplied bythe analyst and in
48、tegrate it over the response functions of thesensors exposed in the unknown environment to predict a set ofcalculated responses for comparison with the measured values.The calculated responses are obtained from Eq 1. The codeobtains the response functions from a library. See GuideE 1018 for the reco
49、mmendations in the selection of dosimetry-quality cross sections. Available dosimetry-quality cross sec-tion libraries include: the International Reactor Dosimetry File(IRDF-2002) cross section library (4), release 6 of the ENDF/B-VI (5, 6) cross section library and the SNLRML package (7)which is available through RSICC.5.3.2 The code compares the measured and calculatedresponses for each effect and invokes an algorithm designed toalter the trial function so as to reduce the deviations betweenthe measured and calculated responses. The process is repeatedwith code-
