1、Designation: C 1295 05Standard Test Method forGamma Energy Emission from Fission Products in UraniumHexafluoride and Uranyl Nitrate Solution1This standard is issued under the fixed designation C 1295; the number immediately following the designation indicates the year oforiginal adoption or, in the
2、case 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.1. Scope1.1 This test method covers the measurement of gammaenergy emitted from fission products in
3、 uranium hexafluoride(UF6) and uranyl nitrate solution. It is intended to provide amethod for demonstrating compliance with UF6specificationsC 787 and C 996 and uranyl nitrate specification C 788.1.2 The lower limit of detection is 5000 MeV Bq/kg(MeV/kg per second) of uranium and is the square root
4、of thesum of the squares of the individual reporting limits of thenuclides to be measured. The limit of detection was determinedon a pure, aged natural uranium (ANU) solution. The value isdependent upon detector efficiency and background.1.3 The nuclides to be measured are106Ru/106Rh,103Ru,137Cs,144
5、Ce,144Pr,141Ce,95Zr,95Nb, and125Sb. Other gammaenergy-emitting fission nuclides present in the spectrum atdetectable levels should be identified and quantified as requiredby the data quality objectives.1.4 This standard does not purport to address all of thesafety concerns, if any, associated with i
6、ts 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:2C 761 Test Methods for Chemical, Mass Spectrometric,Spectrochemical, N
7、uclear, and Radiochemical Analysis ofUranium HexafluorideC 787 Specification for Uranium Hexafluoride for Enrich-mentC 788 Specification for Nuclear-Grade Uranyl Nitrate So-lutionC 996 Specification for Uranium Hexafluoride Enriched toLess than 5 %235UD 3649 Practice for High Resolution Gamma-Ray Sp
8、ec-trometry of Water3. Summary of Test Method3.1 A solution of the uranium sample is counted on ahigh-resolution gamma-ray spectroscopy system. The resultingspectrum is analyzed to determine the identity and activity ofthe gamma-ray-emitting radioactive fission products.The num-ber of counts recorde
9、d from one or more of the peaks identifiedwith each nuclide is converted to disintegrations of that nuclideper second (Bq). The gamma-ray energy for a nuclide iscalculated by multiplying the number of disintegrations persecond of the nuclide by the mean gamma-ray energy emissionrate of the nuclide.
10、The calculated gamma-ray energy emissionrates for all observed nuclides are summed, then divided by themass of the uranium in the sample to calculate the overall rateof gamma energy production in units of million electron voltsper second per kilogram of uranium.4. Significance and Use4.1 The gamma-r
11、ay emitting fission products in UF6areidentified and quantified using a high-resolution gamma-rayenergy analysis system, which includes a high-resolutiongermanium detector. This test method shall be used to meet thehealth and safety specifications of C 787, C 788, and C 996regarding applicable fissi
12、on products in reprocessed uraniumsolutions.5. Apparatus5.1 High-Resolution Gamma-Ray Spectrometry System,asspecified in Practice D 3649.5.2 Sample Container with Fitted CapA leak-proof plas-tic container capable of holding the required sample volume.The dimensions must be consistent between contain
13、ers used forsamples and standard to keep the counting geometry constant.The greatest detection efficiency will be achieved with alow-height sample container with a diameter slightly smallerthan the detector being used.5.3 Sample Holder, shall be used to position the samplecontainer such that the det
14、ector view of the sample is repro-ducible. To minimize the effects of coincident summing, thesample holder shall provide a minimum separation of 5 mmbetween the sample container and the detector end cap.1This test method is under the jurisdiction ofASTM Committee C26 on NuclearFuel Cycle and is the
15、direct responsibility of Subcommittee C26.05 on Methods ofTest.Current edition approved July 1, 2005. Published August 2005. Originallyapproved in 1995. Last previous edition approved in 1998 as C 1295 98.2For referenced ASTM standards, visit the ASTM website, www.astm.org, orcontact ASTM Customer S
16、ervice at serviceastm.org. For Annual Book of ASTMStandards 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.6. Calibration and Standardization of Det
17、ector6.1 Prepare a mixed radionuclide calibration standard stocksolution covering the energy range of approximately 50 to2000 keV.6.1.1 Commercial calibration standards are available.6.2 Prepare a solution of ANU at 6.74 gU/100 g. Theuranium and its progenys relationship must not have beenaltered fo
18、r at least eight months.6.3 Transfer a known, suitable activity of the mixed nuclidecalibration standard stock solution (40 to 50 kBq) to acontainer identical to that used for the sample measurement.Add ANU solution to the mixed nuclide standard so that thefinal volume and uranium concentration matc
19、h those expectedin the sample measurement. Practice D 3649 provides infor-mation on calibration of detector energy, efficiency, resolution,and other parameters.6.4 The detector energy scale and efficiency are calibratedby placing the container with the mixed nuclide calibrationstandard in a sample h
20、older that provides a reproduciblegeometry relative to the detector. Collect a spectrum over aperiod up to 1 h that includes all the gamma photopeaks in theenergy range up to ;2000 keV.All counting conditions (exceptcount time) must be identical to those that will be used foranalysis of the actual s
21、ample.6.5 Determine the net counts under each peak of everynuclide in the mixed radionuclide standard, then divide by thecount time (live time) to determine the rate in counts persecond for each radionuclide. If a background count on thedetector shows any net peak area for the peaks of interest, the
22、semust be subtracted from the standard counts per second.6.6 Divide the observed count rate determined for eachgamma peak by the calculated emission rate of the gamma raythat produced the peak in the mixed calibration standard(gammas per second).6.6.1 Calculation of the gamma emission rate for each
23、peakfrom the mixed calibration standard must account for thefollowing:6.6.1.1 Activity of the nuclide that produces the peak in itsoriginal standard (disintegrations/second/unit volume). This istaken from the standard certificate of measurement suppliedwith the standard.6.6.1.2 Volume of each isotop
24、ic standard taken for themixed standard and the final volume of the mixed standard.6.6.1.3 Fraction of the volume of the mixed standard takenfor counting.6.6.1.4 Decay of the activity of each isotope in the standardbetween its date of standardization and the date of countingaccording to the equation
25、:Ai5 Ai0e2lit(1)where:Ai= activity of isotope i on the date of counting in Bq,Ai0= activity of isotope i on the date of standard charac-terization in Bq,li= decay constant of isotope i in units of inverse time(values for some isotopes of interest may be found incolumn 3 of Table 1), andt = elapsed t
26、ime between the calibration reference dateand the date of counting. Time units must be thesame as in the decay constant.6.6.1.5 The abundance of gamma rays of the energy ofinterest emitted by each disintegration (see Table 1).6.7 Plot a detector efficiency curve of counts/gamma versusgamma energy. M
27、ost multichannel analyzers and data reduc-tion programs are able to store individual values from thiscurve or the equation of the curve for later use.6.8 This efficiency calibration will remain valid providednone of the sample or instrument parameters are changed (forexample, volume of sample, conta
28、iner geometry, distance fromdetector, and detector) and instrument response to the controlstandard remains within the statistical limits established.7. Measurement of Control Standard Solution7.1 Measure the control standard solution prepared in 6.3with the geometry as used during detector efficienc
29、y calibra-tion. Ten measurements of the control standard solution aremade. The calculated data for the fission products is used toestablish precision and bias of the test method.7.1.1 Most multichannel analyzers have automatic routinesfor determining the net counts under single peaks and doublepeaks
30、 that are not resolved. If the available analyzer does nothave such capabilities, refer to Reilly3for single-peak analysismethods and 7.2.1 and 7.2.2 for double-peak problems that arelikely to be encountered.7.1.2 Peaks that are determined for this analysis are listed inTable 1,4along with the abund
31、ance factors, decay constants,and the mean gamma energy per disintegration for eachnuclide.7.2 Determination of the following peak areas may causeproblems during calibration or sample measurements.7.2.1 The peak produced by the 765.9-keV gamma rayof95Nb is not resolved from the peak produced by the7
32、66.4-keV gamma ray of234mPa, a daughter of238U. The3Reilly, T. D., and Parker, J. L., A Guide to Gamma Ray Assay for NuclearMaterials Accountability, LA-5794M, Los Alamos National Laboratory, 1975.4The information in Table 1 is from the Joint European File: 1 data file suppliedby the Nuclear Energy
33、Agency, Paris, France. The user may use other publisheddata.TABLE 1 Gamma-Ray-Emitting Fission Products Found in UF6NuclideHalf-LifeDecayConstant(lI)MeasurementPeaks,MeVAbundanceGamma/Disintegration(GI)Mean GammaEnergyDisintegration,MeVBq (EI)103Ru/103Rh 39.35d 0.01761/d 0.4971 0.889 0.4840.6103 0.0
34、56106Ru/106Rh 366.5d 0.001891/d 0.5119 0.207 0.2090.6222 0.0981141Ce 32.55d 0.02129/d 0.1454 0.484 0.0718144Ce/144Pr 284.5d 0.002436/d 0.1335 0.1110 0.0518137Cs/137Ba 30.17y 0.02297/y 0.6616 0.851 0.565595Nb 34.97d 0.01982/d 0.7658 1.000 0.76695Zr 63.98d 0.01083/d 0.7242 0.444 0.7370.7567 0.549125Sb
35、 2.71y 0.256/y 0.4279 0.294 0.4330.6008 0.178C1295052following procedure is suggested to determine the count rateof95Nb in the double peak.7.2.1.1 Perform a series of count measurements for periodsupto1hofasample ofANU under the same conditions as thecalibration standard or sample. The counting peri
36、od should beadjusted so that the counting errors are less than 1 % for theappropriate peaks of interest.7.2.1.2 For each measurement, determine the ratio of countsin the234mPa peaks at 766.4 and 1001 keV using the equation:RPa5 C766 total/ C1001(2)where:RPa= ratio of counts in the 766.4 and 1001-keV
37、peaks of234mPa,C766 total= total counts in the double peak near 766 keV,andC1001= counts in the 1001-keV peak of234mPa.7.2.1.3 Calculate the mean value for the ratio (RPa).7.2.1.4 Determine the95Nb counts at 765.9 keV by use ofthe equation:CNb5 C766 total2 C1001!RPa!# (3)where:CNb= counts in the pea
38、k near 766 keV resulting from765.9-keV gamma rays of95Nb.7.2.2 The peak produced by the 145.4-keV gamma rayof141Ce is not resolved from the peak produced by the143.8-keV gamma ray of235U. The following procedure issuggested to determine the count rate of141Ce in the doublepeak.7.2.2.1 Perform a seri
39、es of measurements of up to 1-hcounting time of a sample of ANU under the same conditionsas the calibration standard or sample.7.2.2.2 For each measurement, determine the ratio of countsin the235U peaks at 143.8 and 185.7 keV using the equation:RU5 C144 total/ C185.7(4)where:RU= ratio of counts in t
40、he 143.8 and 185.7-keVpeaks of235U,C144 total= total counts in the double peak near 144 keV,andC185.7= counts in the 185.7-keV peak of235U.7.2.2.3 Calculate the mean value for the ratio (RU).7.2.2.4 Determine the141Ce counts at 145.4 keV by use ofthe equation:CCe5 C144 total2 C185!RU!# (5)where:CCe=
41、 counts in the peak near 144 keV resulting from145.4-keV gamma rays of141Ce.8. Procedure8.1 Hydrolyze the UF6sample for counting as in TestMethod C 761 or prepare the uranyl nitrate solution sample.Ensure that sample preparation parameters (solution volume,uranium concentration, sample container, ge
42、ometry, and soforth) are the same as used during detector efficiency calibra-tion. Note the mass of uranium (W) taken in grams.8.2 Place the container and sample into the counter with thesame geometry as used during detector efficiency calibration.Count the sample for 60 min to collect a gamma spect
43、rum ofthe sample.8.3 Determine the net counts under one or more peaks foreach nuclide, then divide by the count time (live time) todetermine the count rate for each gamma peak in counts persecond. See 7.2.1 and 7.2.2 for methods to deal with unre-solved double peaks.9. Calculation9.1 Determine the g
44、amma energy release rate for eachnuclide according to the following equation:Fi51000W3CiEff 3 Gi3 Ei(6)where:Fi= rate of energy released in gamma radiation as a resultof fission nuclide i decay in MeV Bq/kg U (MeV/kgper second),Ci= count rate calculated in 8.3 for a single gamma-raypeak of nuclide i
45、 (counts per second),Eff = the detector efficiency (counts/gamma) determinedin Section 6 for the energy of the gamma-ray peakbeing analyzed,Gi= the gamma-ray production rate (gammas/disintegration) by nuclide i for the energy of gammaray being analyzed (from Table 1),Ei= mean gamma energy release pe
46、r disintegration ofnuclide i in MeV (from Table 1), andW = uranium sample weight, in grams.9.2 Determine the total fission product energy release rate,FTotal, by summing the contributions from all nuclides de-tected, as follows (expressed in units of MeV Bq/kg U(MeV/kg U per second).FTotal5 (Fi(7)10
47、. Precision and Bias10.1 Precision:10.1.1 Precision data was obtained from ten measurementsof a uranyl fluoride (UO2F2) solution prepared from ANUhexafluoride and spiked nuclides106Ru,134Cs,60Co, and137Csfrom an international traceable standard. The work was carriedout by one analyst over a period o
48、f weeks, and the data is inTable 2.10.2 Bias Estimate:10.2.1 No standard material is certified for fission productsin UF6or UNO2solution. The bias estimates were obtainedTABLE 2 Precision and Bias DataNuclidePreparedActivity Level,MeV Bq/kg UMeasuredActivity Level,MeV Bq/kg UStandard Deviation ofMea
49、sured Activity (1s),MeV Bq/kg U106Ru 1.7 3 1051.4 3 1051.10 3 103134Cs 1.0 3 1058.8 3 1041.40 3 10460Co 1.2 3 1051.2 3 1051.79 3 103137Cs 2.4 3 1042.4 3 1043.40 3 102C1295053from the same data used to calculate the precision. The data aresummarized in Table 2.10.2.2 The data gave a relative bias of 18 % for106Ruand 12 % for134Cs. This negative bias is probably because ofthe effects of coincidence summing and absorption.11. Keywords11.1 fission products; gamma energy; uranium hexafluoride;uranium nitrateASTM International takes no position respecting the validity of any
