1、ANSI/ANS-5.4-2011method for calculating thefractional release ofvolatile fission products from oxide fuelANSI/ANS-5.4-2011ANSI/ANS-5.4-2011American National StandardMethod for Calculating theFractional Release ofVolatile Fission Products from Oxide FuelSecretariatAmerican Nuclear SocietyPrepared by
2、theAmerican Nuclear SocietyStandards CommitteeWorking Group ANS-5.4Published by theAmerican Nuclear Society555 North Kensington AvenueLa Grange Park, Illinois 60526 USAApproved May 19, 2011by theAmerican National Standards Institute, Inc.AmericanNationalStandardDesignation of this document as an Ame
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14、is not intended for genericapplication.Responses to inquiries on standards are published in ANSs magazine, NuclearNews, and are available publicly on the ANS Web site or by contacting the ANSstandards administrator.InquiryFormatInquiry requests must include the following:1! the name, company name if
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16、e.Inquiries should be addressed toAmerican Nuclear SocietyATTN: Standards Administrator555 N. Kensington AvenueLa Grange Park, IL 60526or standardsans.orgForewordThis Foreword is not a part of American National Standard “Method for Calculating theFractional Release of Volatile Fission Products from
17、Oxide Fuel,” ANSI0ANS-5.4-2011.!This standard provides a methodology for determining the radioactive fissionproduct releases from UO2fuel for use in assessing radiological consequences ofpostulated accidents that do not involve significant abrupt power transientssuch as in a reactivity-initiated acc
18、ident RIA!. When used with isotopic yields,this method will give the so-called “gap activity,” which is the inventory ofvolatile fission products that could be available for release from the fuel rod ifthe cladding were breached. The standard as developed applies to steady-stateconditions and, there
19、fore, should not be applied to accidents where abrupt tem-perature increases are experienced resulting in what is sometimes referred to as“burst release.” The standard can be applied to Condition 1 and 2 transients thatdo not result in large fuel temperature increases, i.e., H11350300 K temperature
20、in-crease, where significant burst release is possible. The standard does not con-sider escape-rate coefficients or other transport of fission products from the fuelrod void space to the reactor coolant or other medium; the standard applies onlyto available fission products in the fuel rod void spac
21、e available for release iffailure is experienced. The standard assumes that no significant fuel oxidationwill be present during the accident because fuel oxidation can significantlyenhance the release. Therefore, the standard does not apply to accidents wheresignificant fuel oxidation is present.The
22、 volatile and gaseous fission products of primary significance are krypton,xenon, iodine, and cesium. The radioactive nuclide that contributes the largest toequivalent dose to individuals is generally131I for accidents during in-reactoroperation or shortly after reactor operation e.g., a fuel-handli
23、ng accident! be-cause of its dose to the thyroid. These radioactive gaseous and volatile fissionproducts can be divided into two categories: a! short-lived radioactive nuclideshalf-life ,1yr! and b! long-lived radioactive nuclides half-life .1yr!. Thisdivision is significant because the most importa
24、nt release mechanism involvesthermally activated migration diffusion! processes that proceed slowly such thatthe short-lived nuclides decay appreciably before they are released from thepellet. Consequently, release calculations for short-lived nuclides must includetheir decay rate, whereas for long-
25、lived nuclides, decay does not have to beconsidered. The revised new standard is only applicable to the short-lived nu-clides with half-lives ,1yr.The ANS-5.4 standard was first implemented in 1982 with Dr. Stan Turner aschairman of the ANS-5.4 Working Group. The original methodology was basedon the
26、 use of the Booth diffusion model A. H. Booth, CRDC-721 and DCI-17,Atomic Energy of Canada 1957!; A. H. Booth and G. T. Rymer, CRDC-720,Atomic Energy of Canada 1958!; S. D. Beck, BMI-1433, Battelle Columbus1960!#. The coefficients to the model were determined from the measured re-lease data of stabl
27、e nuclides of xenon and krypton. The coefficients were derivedusing the stable nuclide release data because very few release data were avail-able for the radioactive nuclides. Also, because of the lack of data from theradioactive131I, the diffusion of this nuclide was assumed to be a factor of 7high
28、er than for the xenon and krypton nuclides. The 1982 standard was notreaffirmed in 1992 because newer data were demonstrating that the assump-tions used in the earlier standard were overly conservative. Many of the originalworking group members were not available to revise the standard at that time;
29、therefore, a new ANS-5.4 Working Group was formed in 2000 to revise thestandard. In the past this standard had primarily been used to estimate the doseto operators in the control room due to a fuel-handling accident but could be usedifor any accidents where a large fuel temperature increase is not e
30、xperienced andfuel failure has been determined or assumed.In the last 25 yr, fuel experiments in test reactors have been performed that havemeasured the release fractions of the radioactive nuclides including131I! up to95 MWd0kg U peak pellet average!, allowing for a verification of the standard toh
31、igher burnups. These new data have demonstrated that the diffusion coeffi-cients for the xenon, krypton, and iodine nuclides are essentially the same, suchthat the factor of 7 higher diffusion coefficient for iodine assumed in the 1982standard was overly conservative. In addition, the use of the 198
32、2 methodologyalso overpredicted the release fractions for the xenon and krypton stable nu-clides when used in modern fuel performance codes that include a fuel thermalconductivity model that is a function of burnup. As a consequence, new empiricaldiffusion coefficients have been developed; however,
33、the methodology still uti-lizes the simple phenomenological diffusion model that is often referred to as theBooth diffusion model with some modifications such as allowance for bubbleinterlinkage on grain boundaries that is temperature and burnup dependent.The interlinkage enhances the release with i
34、ncreasing burnup and temperature.The empirical coefficients have been developed from the radioactive xenon andkrypton release data published in the last 20 yr. The new coefficients have beencompared to the131I release data and found to predict these release data welland, therefore, be applicable to
35、radioactive iodine release that dominates mostdose consequences for design-basis accidents.The 1982 ANS-5.4 methodology and diffusion coefficients were considered to beconservative for application to UO2fuel and bounding for light water reactorLWR! applications. The revised coefficients for this rev
36、ision to the standard areconsidered to provide more of a best-estimate prediction of release from UO2;arelatively small conservatism may exist but cannot be determined quantitativelybecause of the uncertainties in the current data. As a result the uncertainties ofthe predictions using the revised me
37、thodology are provided in the standard withrecommendations on their application to determining radiological releases.As noted earlier, the working group has chosen to retain the simple phenom-enological Booth diffusion-type model proposed in the 1982 standard and hasfitted the model coefficients emp
38、irically to selected data, whose characteristicswill be described later. The “idealized” Booth model describes diffusion offission product atoms in a spherical grain of fuel material. The governingdifferential equation is?C?tH11005 BH11002lCH11001Dr?2rC!?r2, 1!where:C H11005 nuclide concentration in
39、 the fuel atomcmH110023!;t H11005 time s!;B H11005 production rate atomcmH110023sH110021!;lH11005 decay constant sH110021!;r H11005 radius at a point within the sphere cm!;D H11005 diffusion coefficient cm2sH110021!.The latter term of Eq. 1! is the local mass flow of atoms at a radius ratomcmH110023
40、sH110021!.iiThe solution to Eq. 1! used in the original ANS-5.4 standard for the releasefraction is sometimes referred to as R0B, the release-to-birth ratio, given byR0B H11005 3H208751!mcoth!m!H110021mH20876, 2!where:R H11005 number of atoms released per unit time accounting for decay and tem-perat
41、ure! at a point in time; R H11005H110023D0a!?C0?r!rH11005aatomcmH110023sH110021!;B H11005 number of atoms produced per unit time at the same point in time;C H11005 nuclide concentration in fuel; C H11005 0 is assumed at the sphere surface,where r H11005 a,withr H11005 the radius at a point within th
42、e sphere and a H11005 theidealized sphere radius;mH11005la20D unitless!.Equation 2! is for equilibrium conditions; i.e., it assumes a period of constantpower operation over several half-lives of the nuclide for the R0B in questiongenerally three half-lives is sufficient!. If the temperature and powe
43、r operationchange within fewer than three half-lives, Eq. 2! provides a conservative pre-diction of R0B if the maximum power and temperature are used during the timeperiod in question and the power change is not large enough to elicit large burstreleases.This diffusion equation assumes that a net fl
44、ow of matter occurs because of theexistence of a concentration gradient within the sphere and that the flux ofatoms is proportional to that gradient. The production rate B and decay constantl in Eq. 1! are known for the nuclides of interest, but the effective diffusioncoefficients D and idealized sp
45、here radius a in Eq. 2! are unknown and must bedetermined from experimental data.The release fraction for the current standard uses the same equation but definesthe radius of the sphere a in terms of a surface areatovolume ratio S0V, whichforasphereisa H11005 30S0V!. Substituting 30S0V! in place of
46、coefficient a inEq. 2! results in the following relationship:R0B H11005 S0V1H20906laDH20875coth!mH110021mH20876, 3!where:mH110059lS0V!2aD.Note that a is a unitless term for precursor effects that were generally ignored inthe 1982 standard with the exception of the nuclides133Xe and135Xe. However,the
47、 current standard will include precursor effects for additional nuclides. Equa-tion 3! can be further simplified with only a small overprediction by ,5 H1100310H110023relative! for release values ,0.02 with the following relationship:R0B H11005 S0VH20906aDl. 4! iii The maximum R0B for131I the longes
48、t lived nuclide of interest! in a peak rod intodays LWRs is determined to be ,0.02 from a best-estimate prediction with thenew standard. Also, the error introduced in Eq. 4! is several orders of magnitudeless than the 2s uncertainty in the revised standard between a factor of 2.5 to5!. Therefore, Eq
49、. 4! provides a reasonably accurate prediction of release fornuclides considering the uncertainties in the analysis, and this is the form of themodel recommended for determining the radiological release of the volatile nuclides.The Booth diffusion model is an oversimplification of the physical processesinvolved in the release of fission products, and as such, it cannot correctlycalculate the release for all fuel rod accident scenarios. One particular accidentis the RIA that involves a very large increase in rod power
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