ASTM C1769-2015 Standard Practice for Analysis of Spent Nuclear Fuel to Determine Selected Isotopes and Estimate Fuel Burnup《为测定选定同位素以及评估燃料燃耗的废核燃料分析的标准实施规程》.pdf

《ASTM C1769-2015 Standard Practice for Analysis of Spent Nuclear Fuel to Determine Selected Isotopes and Estimate Fuel Burnup《为测定选定同位素以及评估燃料燃耗的废核燃料分析的标准实施规程》.pdf》由会员分享，可在线阅读，更多相关《ASTM C1769-2015 Standard Practice for Analysis of Spent Nuclear Fuel to Determine Selected Isotopes and Estimate Fuel Burnup《为测定选定同位素以及评估燃料燃耗的废核燃料分析的标准实施规程》.pdf（7页珍藏版）》请在麦多课文档分享上搜索。

1、Designation: C1769 15Standard Practice forAnalysis of Spent Nuclear Fuel to Determine SelectedIsotopes and Estimate Fuel Burnup1This standard is issued under the fixed designation C1769; the number immediately following the designation indicates the year oforiginal adoption or, in the case of revisi

2、on, 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 A sample of spent nuclear fuel is analyzed to determinethe quantity and atomic ratios of uranium and p

3、lutoniumisotopes, neodymium isotopes, and selected gamma-emittingnuclides (137Cs,134Cs,154Eu,106Ru, and241Am). Fuel burnupis calculated from the148Nd-to-fuel ratio as described in thismethod, which uses an effective148Nd fission yield calculatedfrom the fission yields of148Nd for each of the fission

4、ingisotopes weighted according to their contribution to fission asobtained from this method. The burnup value determined inthis way requires that values be assumed for certain reactor-dependent properties called for in the calculations (1, 2).21.2 Error associated with the calculated burnup values i

5、sdiscussed in the context of contributions from random andpotential systematic error sources associated with the measure-ments and from uncertainty in the assumed reactor-dependentvariables. Uncertainties from the needed assumptions areshown to be larger than uncertainties from the isotopicmeasureme

6、nts, with the largest effect arising from the value ofthe fast fission factor. Using this factor will provide the mostconsistent burnup value between calculated changes in heavyelement isotopic composition.1.3 This standard practice contains explanatory notes thatare not part of the mandatory portio

7、n of the standard.1.4 The values stated in SI units are to be regarded as thestandard. Mathematical equivalents are given in parentheses.1.5 This standard does not purport to address all of thesafety concerns, if any, associated with its use. It is theresponsibility of the user of this standard to e

8、stablish appro-priate safety and health practices and determine the applica-bility of regulatory limitations prior to use.2. Referenced Documents2.1 ASTM Standards:3C1625 Test Method for Uranium and Plutonium Concentra-tions and Isotopic Abundances by Thermal IonizationMass SpectrometryC859 Terminol

9、ogy Relating to Nuclear MaterialsD1193 Specification for Reagent WaterE244 Test Method forAtom Percent Fission in Uranium andPlutonium Fuel (Mass Spectrometric Method) (With-drawn 2001)43. Terminology3.1 DefinitionsFor definitions of other standard terms inthis practice, refer to Terminology C859.3.

10、2 Definitions of Terms Specific to This Standard:3.2.1 gigawatt days per metric tonthe gigawatt days ofheat produced per metric ton of uranium plus plutoniuminitially present in a nuclear fuel.3.2.2 heavy element atom percent fissionthe number offissions per 100 uranium plus plutonium atoms initiall

11、y presentin a nuclear fuel.3.3 Symbols: Symbols used in the procedural equations aredefined as follows:3.3.1 F5,F9,F1,F8heavy element atom percent fissionfrom fission235U,239Pu,241Pu, and238U.3.3.2 FTtotal heavy element atom percent fission.3.3.3 F80,N50heavy element atom percent238U and235U,in the

12、pre-irradiated fuel.3.3.4 R580,R680,R650atoms ratios of235Uto238U,236Uto238U, and236Uto235U in the pre-irradiated fuel.3.3.5 R58 ,R68 ,R65 atom ratios of235Uto238U,236Uto238U, and236Uto235U in the final irradiated sample.3.3.6 R98 ,R08 ,R18 atom ratios of239Pu,240Pu,241Pu,242Pu and to238U in the fin

13、al irradiated sample.1This practice is under the jurisdiction of ASTM Committee C26 on NuclearFuel Cycle and is the direct responsibility of Subcommittee C26.05 on Methods ofTest.Current edition approved June 1, 2015. Published July 2015. DOI: 10.1520/C1769-15.2The boldface numbers in parentheses re

14、fer to a list of references at the end ofthis standard.3For referenced ASTM standards, visit the ASTM website, www.astm.org, orcontact ASTM Customer Service at serviceastm.org. For Annual Book of ASTMStandards volume information, refer to the standards Document Summary page onthe ASTM website.4The l

15、ast approved version of this historical standard is referenced onwww.astm.org.Copyright ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA 19428-2959. United States13.3.7 R18 atom ratio of241Pu to238U in the final irradiatedsample corrected for neutron capture, fission, an

16、d decay duringand after irradiation.3.3.8 82.67 6 0.30 neutrons per fission of238U (3).3.3.9 52.426 6 0.006 neutrons per fission of235U (4).3.3.10 9/5ratio of number of neutrons per fission of239Pu to235U = 1.192 6 0.005 (4).3.3.11 1/5ratio of number of neutrons per fission of241Pu to235U = 1.237 6

17、0.017 (4).3.3.12 telapsed time from the end of irradiation to mea-surement.3.3.13 tirradiation time, s.3.3.14 1decay constant of 153 10-9s-1.3.3.15 cratio of the238U fission rate ot the fission ratefrom all other sources expressed as equivalent235U fission rate.3.3.16 fast fission factor (defined in

18、 Ref (5) which is1.00 for fully enriched reactors. Typically, ranges from 1.03to 1.07 for low enrichment systems.3.3.17 a5effective ratio of235U(n, ) capture-to-fissioncross sections obtained from reactor designer, experiment, ormachine calculation. If not otherwise available, it may beestimated fro

19、m Fig. 1 for well-moderated thermal reactors.3.3.18 a9effective ratio of239Pu (n, ) capture-to-fissioncross sections obtained from reactor designer, experiment, ormachine calculation. If not otherwise available, it may beestimated from Fig. 2 for well-moderated thermal reactors.3.3.19 a1effective ra

20、tio of241Pu (n, ) capture-to-fissioncross sections = 0.40 6 0.15 for thermal reactors Ref (6). Itsneutron spectrum dependence has not been measured.3.3.20 a8effective ratio of238U(n, ) capture-to-fissioncross sections averaged over a fission spectrum = 0.58 6 0.45(3).3.3.21 repithermal index which i

21、s a measure of theproportion of epithermal neutrons in a reactor spectrum. In Ref(7), r is defined and related mathematically to the cadmiumratio. Note that for r = 0 the spectrum is pure Maxwellian.3.3.22 neutron flux, neutrons/cm2-s.3.3.23 1, 5, 6total neutron absorption cross sections of241Pu,235

22、U, and236U. For boiling water reactors, typical coreaverage values are 188 10-23cm2,64.610-23cm2,and510-23cm2, respectively. For pressurized water reactors, typicalcore average values are 155 10-23,55.610-23cm2, and 8.410-23cm2, respectively.3.3.24 Ptotal239Pu neutron captures per initial238U atom.4

23、. Summary of Practice4.1 Atomic ratios of the isotopes234U,235U,236U, to238Uand240Pu,241Pu, and242Pu to239Pu are measured by massspectrometry in accordance with Test Method C1625 or asimilar methodology. The atom percent fission attributed tofission of235U,238U,239Pu, and241Pu are separately calcula

24、tedand then summed to obtain the total heavy element atompercent fission (6, 8).4.2 Fission product neodymium (Nd) is chemically sepa-rated from irradiated fuel and determined by isotopic dilutionmass spectrometry. Enriched150Nd is selected as the neo-dymium isotope diluent and the mass-142 position

25、 is used tomonitor for natural neodymium contamination. The two rareearths immediately adjacent to neodymium do not interfere.Interference from other rare earths, such as natural or fissionproduct142Ce or natural148Sm and150Sm is avoided byremoving them in the chemical purification (9, 10).4.3 After

26、 addition of a blended150Nd,233U, and242Pu spiketo the sample, the neodymium, uranium, and plutonium frac-tions are separated from each other by ion exchange. Eachfraction is further purified for isotope dilution mass spectrom-etry analysis. Two alternative separation procedures are pro-vided.4.4 Th

27、e gross alpha beta, and gamma decontaminationfactors are in excess of 103and are normally limited to thatvalue by traces of242Cm,147Pm, and241Am, respecitvely (andFIG. 1 Calculated Dependence of a5on Neutron Temperature andEpithermal Index, r, for Well-Moderated Thermal ReactorsFIG. 2 Calculated Dep

28、endence of a9on Neutron Temperature andEpithermal Index, r, for Well-Moderated Thermal ReactorsC1769 152sometimes106Ru), and are insignificant to the analysis. The 70ng148Nd minimum sample size recommended in the procedureis large enough to exceed by 100-fold a typical naturalneodymium blank of 0.70

29、 6 0.7 ng148Nd (for which acorrection is made) without exceeding radiation dose rates of20 Sv/h (20 mrem/h) at 1 m for 60-day cooled fuel to 20 Sv/h (2 mrem/h) at 1 m for 1-year cooled fuel. Beta dose ratesare an order of magnitude greater, and may be significantlyshielded with a 12.7 mm (12-in.) th

30、ick plastic sheet. By use ofsuch simple local shielding dilute solutions of irradiatednuclear fuel dissolver solutions can be analyzed for burnupwithout an elaborate shielded analytical facility. The decon-taminated neodymium fraction is mounted on a rhenium (Re)filament for isotope dilution mass sp

31、ectrometry analysis.Samples from 20 ng to 20 g run well in the mass spectrometerwith both NdO+and Nd+ion beams present.5. Significance and Use5.1 This standard practice defines a measure of heavyelement atom percent fission from which the output of heatduring irradiation can be estimated.5.2 This st

32、andard practice is restricted in use to sampleswhere accurate pre-irradiation U and Pu isotopic analysis isavailable. This data should be available from the fuel manu-facture.5.3 The contribution of238U fast fission is not subject tomeasurement from isotopic analysis. For reactors in which themajori

33、ty of fissions are caused by thermal neutrons, thecontribution may be estimated from the fast fission factors, ,found in each reactor design document.5.4 In post-irradiation isotopic analysis, take extreme careto avoid environmental uranium contamination of the sample.This is simplified by using sam

34、ple sizes in which the amount ofeach uranium isotope is more than 1000 times the levelsobserved in a blank carried through the complete chemistry andmass spectrometry procedure employed.5.5 Take care to make sure that both the pre-irradiation andthe post-irradiation samples analyzed are representati

35、ve. In thepre-irradiation fuel, the235U and236U atom ratio content mayvary from lot to lot.236U is not found in naturally uranium inmeasurable quantity (2 ppm of a u basis) but forms duringirradiation and increases with each successive pass through thefuel cycle. In the post-irradiation examination

36、of a large fuelelement, the atom percent fission normally varies radially andaxially. Radial and axial profiles of atom percent fission can bedetermined by analyzing samples obtained from along theradius or axis of the fuel element. An average value of atompercent fission can be obtained by totally

37、dissolving the fuel tobe averaged, and then mixing and analyzing an aliquot of theresultant solution.5.6 The burnup of an irradiated nuclear fuel can be deter-mined from the amount of a fission product formed duringirradiation. Among the fission products,148Nd has the follow-ing properties to recomm

38、end it as an ideal burnup indicator: (1)It is not volatile. (2) It does not migrate in solid fuels belowtheir recrystallization temperature. (3) It has no volatile pre-cursors. (4) It is nonradioactive and requires no decay correc-tions. (5) It has a low destruction cross section. (6) Formationof148

39、Nd from adjacent mass chains can be corrected for. (7)Ithas adequate emission characteristics for mass analysis. (8) Itsfission yield is nearly equivalent for235U and239Pu. (9) Itsfission yield is essentially independent of neutron energy (11).(10) It has a shielded isotope,142Nd, which can be used

40、forcorrecting natural neodymium contamination. (11)Itisanatypical constituent of unirradiated fuel.6. Apparatus6.1 Dissolution bomb.56.2 Oven-convection.7. Reagents and Materials7.1 Purity of ReagentsReagent grade chemicals shall beused in all tests. Unless otherwise indicated, it is intended thatal

41、l reagents conform to the specifications of the Committee onAnalytical Reagents of the American Chemical Society wheresuch specifications are available. Other grades may be used,provided it is first ascertained that the reagent is of sufficientlyhigh purity to permit its use without lessening the ac

42、curacy ofthe determination.7.2 Purity of WaterUnless otherwise indicated, referencesto water shall be understood to mean reagent grade as definedby Type I of Specification D1193 or water exceeding thesespecifications.7.3 Hydrochloric Acid (sp gr 1.19)-concentrated HCl.7.4 Nitric Acid (sp gr 1.42)-co

43、ncentrated (HNO3).7.5 Hydrobromic Acid (sp gr 1.18)-concentrated (HBr).7.6 Perchloric Acid (sp gr 1.67)-concentrated.7.7 Nitric Acid (2 M)-Add 126 mL concentrated nitric acidto a volume of water and dilute, with water, to a final volumeto 1000 mL.7.8 Nitric Acid (0.3 M)-Add 19 mL concentrated nitric

44、 acidto a volume of water and dilute, with water, to a final volumeto 1000 mL.7.9 Nitric Acid (0.1 M)-Add 6 mL concentrated nitric acidto a volume of water and dilute, with water, to a final volumeto 1000 mL.7.10 Methanol.7.11 Nitric Acid/methanol (1.56 M HNO3/80 % CH3OH)Add 99 mLconcentrated nitric

45、 acid to 50 mLof water then 800mL of methanol and dilute to a final volume of 1000 mLvolume with water.7.12 Dilute nitric acid/methanol (0.156 M HNO3/80 %CH3OH) Take a 100 mL aliquot of the solution made in 7.11and dilute with 800 mL of methanol and dilute to a finalvolume of 1000 mL.7.13 Nitric Aci

46、d/Methanol (0.10 M HNO3/80 % CH3OH)Add 6.3 mL concentrated nitric acid, 800 mL methanol anddilute, with water, to a final volume to 1000 mL volume withwater.5Parr dissolution bomb Model 4749 has shown to be adequate.C1769 1537.14 Hydrofluoric Acid (HF, sp gr 1.18) concentrated.7.15 Hydrochloric/Hydr

47、ofluoric Acid (0.05 M HCl/0.001HF) Add 4.2 mL concentrated HCl to 50 mL of water;separately add 0.04 mL concentrated HF of 50 mL of water.Combine the two diluted acids and dilute to a final volume of1000 mL.7.16 Ion Exchange ResinsAnion, quatinary amine (100-200 and 200-400 mesh).7.17 Exchange resin

48、6, transuranic, octylphenyl-N,N-di-isobutyl carbamoylphosphine oxide dissolved in tri-n-butylphosphate (TBP) and placed on a solid support. WarningHydrofluoric acid is a highly corrosive acid that can severelyburn skin, eyes, and mucous membranes. Hydrofluoric aciddiffers from other acids because th

49、e fluoride ion readilypenetrates the skin causing destruction of deep tissue layers.Unlike other acids that are rapidly neutralized hydrofluoric acidreactions with tissue may continue for days if left untreated.Familiarization and compliance with the Safety Data Sheet isessential.8. Operational Procedure8.1 Inside Hot Cell:8.1.1 Measure the mass of a portion of the pulverized fuelinto the polytetrafluoroethylene (PTFE) cup of a ParrTMbomb.Test samples are typically 50 g originating from a knownlocation