radiochem nucl power reactor

8/8/2019 Radiochem Nucl Power Reactor

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Karl-Heinz Neeb

Tile Hadiochernistry ofNuclear Power Plantswith Light Water Reactors

With a preface written by Gunter Marx

DE

G Walter de Gruyter . Berlin New York 1997


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Part BRadiochemistry during normal operationof the plant

3. Radionuclides in the reactor coreThe reactor pressure vesselencloses the region where the largest fraction by far ofthe total radionuclide inventory of a nuclear power plant is concentrated. All theradionucl ides present in the primary system and in the other systems of the planthave been produced here by reactions of neutrons with different substances. Forthis reason, the reactor core isthe starting point for all further considerations withregard to radiochemistry in a nuclear power plant.

Uranium is a natural ly occurring radioactive element and, therefore, freshnuclear fuel also contains radionuclides. Plutonium and reprocessed uranium, onthe other hand, are accompanied by isotopes which were generated during thepreceding exposure of nuclear fuel in a reactor. Al l the radionucl ides present infresh nuclear fuel are of minor importance in reactor operation; nonetheless, theyhave to be t aken into considera ti on in fuel fabri ca tion in order to pro tect theemployees from undue radiation exposure.In the course of the operation of a nuclear power plant , a very large radionuelide inventory generated by nuclear fission and neutron capture is accumulated inthe fuel. Since this inventory normally is hermetically confined in the gas-t ightsealed fuel rod claddings, these radionuclides do not appear to a significant extentin other areas of the reactor plant during normal operation; in this phase, thei ronly sign of existence isthe radiation they emit, which has to be shielded by appropriate means. Only in the subsequent stage of reprocessing of the spent fuel is thechemical behavior of the fission products and actinides of vital interest. Nevertheless,there are various reasons for the efforts undertaken in nuclear reactor technology to acquire detailed knowledge of the chemical state and the behavior of theradionuclides in the irradiated fuel. One reason is that the chemical compositionof the fuel is significantly altered by the accumulation of fission products belongingto chemical elements originally not present in the fuel; moreover, some of thefission products arc suspected of being potential ini tiators of fuel rod claddingdeterioration (e. g. initiation of stress corrosion cracking) when they are depositedon the inner sur face of the Zirca loy cladding. In the event of fuel rod failuresduring reactor operation, radionucl ides are released from the fuel to the reactorprimary coolant, there leading to contamination; knowledge of the mechanismsand the extent of their t ransport from the defective fuel rod to the coolant is anessential prerequisite for dealing with the consequences of such failures. Even if


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60 Radiochemistry during normal operation of the plant Radionuclides in the reactor core 61these are comparatively rare events (when compared to the total number of fuelrods present in the reactor core), they cannot be ruled out completely. Finally, inthe highly improbable case of a severe reactor acc iden t it has to be assumed thatconsiderable fractions of the fission product inventory will be re leased f rom thecore; the extent as wel las the kinet ics of this release depend highly on the chemicalstate of the radionuclides in the irradiated fuel.In addition to the radionuclides produced in the nuclear fuel, others are generated in the core structural mater ia ls such as the fuel rod cladding, spacers, fuelassembly end pieces, control rods, and core support materials. Here, the mainproduction mechanism is neutron activation. A fraction of the radionuc1ides generated in t hi s manner is also re le as ed to the primary coolant during steady-stateoperation of the plant or to the primary circuit in the course of a severe reactoraccident. Therefore, the following discussion will also cover these radionuc1ides.

3.1 Radionuclides in fresh nuclear fuelsUranium as the main starting material fOT nuclear energy production is a naturallyradioactive element, composed of the three long- lived isotopes 234U, 235U and238U. In undisturbed natural deposits, these isotopes appear in a s ecul ar d ec ayequilibrium with their daughter products of the 4n +2 and 4n+3 series, which arepresented in a simplified version in Figs. 3.1. and 3.2.

In the course of the fabrication of nuclear fuel from naturally occurring uranium, these equilibria are disturbed; in the f ir st s teps of ore processing, the bulkof the uranium decay products is separated from the e lement to be used furtheron. The following preparation steps, chemical separations as well as the conversionto UF6 and i ts volat il izat ion in the gas dif fusion or gas centrifuge processes for235U enrichment, give rise to an additional purification, so that at the end of theseprocesses a uranium compound is obtained which is vir tual ly f ree of decay products . However , every delay in these processes, such as the t ime intervals betweenisotope enrichment and pellet fabrication as well as that between the fabrication ofthe pellets and their encapsulation into the fuel rods, leads to a new growth ofdecay products. As can be seen from Figs. 3.1. and 3.2., in the time per iods involved, ranging from a few weeks to seve ral months, this growth concerns exclusively the short-lived protactinium and thor ium i sotopes, which reach their decayequilibria with the starting nuclides within a few days or months, respectively, afterthe final chemical separation. The buildup of additional 234U, the growth of itsdaughter nuclide 230Th as well as of 23 lPa (in the 4n+3 series) and, subsequently,of their decay products proceeds very slowly. Therefore, their activity concentrat ions in the final nuclear fuel remain negligibly small.

The isotope enrichment process does not only le ad to an i nc re as e in the 235Uconcentration in the fuel, but also in a more than proportional increase in the 234Uconcentration, as can be seen from Table 3.1., where the mass concentrations ofthe actinide isotopes in different types of fresh nuclear fuel are compared. As a

r-- 234U 2 ~ U ~ranium 2.44-105a 4 4 7 ~ O 9 ' 1t \234PaProtactinium 1.1SmI6.75h

. '-' ~ _ . - _ - . P _ ~ _ '\ i230Th 234ThThorium 7.7104a 24.1da ~

Actiniumr---" f---.--- _ _ ' w " ' _ ~ I - --

226RaRadium 1.600.103aa1-------Francium i

/---------- ji

I 222RnRadon 3.824dI aIAstatine

1---- I210po 214Po 218pOPolonium 138.38d 1.64-10-45 3.05mII ex, a , aII I' r- !10Bi 214Bi IiBismuth 5.0d 19.8m Ie ~ I----/-. - 1


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62 Radiochemistry during normal operation of the plant Radionuclides in the reactor core 63

235UUranium 7.04-108aa231PaProtactinium 3.25-104aa ,

'\.227Th I 234ThThorium 25.6


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64 Radiochemistry during normal operation of the plant RadionucJides in the reactor core 65starting with mixed-oxide fuel processing. Among the a emitters, 238pU is the mos tintense source. The decay products of the a-emitting plutonium isotopes as well asof 241Am are long-lived nuclides, so that their contribution and that of their daughter products to the total activity of the materials used for fuel fabrication is onlyof minor significance. The only exception is 236pu, which is mainly generated via237Np by (n,2n) reaction and p'" decay of the resulting 236Np; this radionuclidedecays to 232U, an undesirable nuclide in the use of reprocessed uranium for thefabrication of new fuel (see below).Due to the high a activity of the plutonium isotope mixture originating fromhigh-burnup LWR fuel, all the steps in mixed-oxide fuel fabrication have to becarried out under the most stringent safety precautions, 1. e. inside gas-tight boxes.In modern mixed-oxide processing, the plants are highly automated and remotecontrolled equipment is used to protect the employees against r radiation.

Up to now, reprocessed uranium has only been used for the fabrication of alimited number of test fuel assemblies. For this type of fuel, the uranium fractionfrom the spent fuel reprocessing process is again subjected to an isotope enrichmentprocedure to obtain a 235U content which is sufficiently high for reactor operation(3.8% 23SU in the example shown in Tables 3.1 and 3.2.). Besides the naturallyoccurring isotopes 234U, 235U and 238U, this material contains 236U, mainly generated by neutron capture in 235U, and 232U as a decay product of 236PU. The 236Ucontent of reprocessed uranium strongly depends on the preceding fuel history(fuel burnup, initial fuel enrichment); it ranges between 1 and 3%, per weight. 236Uis a strong neutron absorber. which reduces the efficiency of the fuel; i t does not,however, cause any problems in fuel manufacturing. 232U, on the o th er hand , as amember of the 4n series, leads via comparatively short-lived decay products (seeFig. 3.3.) to several intense y emitters, in particular 208TI which shows y energiesof 2.62 and 0.58 MeV. Due to the short halflives of the intermediate nuclides, thebuildup of the v-emitting radionuclides proceeds comparatively quickly; therefore,fabrication of reprocessed uranium fuel requires shielding of the process lines inorder to keep radiation exposure of the employees well below the regulatory limits.The 232U content in the fuel depends both on the fuel bumup and on the decaytime (halflife of 236pU 2.85 years). For a fuel initially enriched in 235U to 2.4%, the232U proportion will more than double when the bumup increases from 15 to23MWd/kg U. The cooling period after irradiation is the second main factor; i tbecomes particularly significant when associated with high burnup.Likewise, the enhanced a activity of reprocessed uranium (greater by a factorof 4 to IOthan that of natural uranium) has to be considered in fuel manufacturing.The resulting complications can, in principle, be reduced by optimized fabricationlogistics, so that the time elapse between isotope enrichment and insertion of thefuel into the reactor would amount to less than half a year.According to the generally accepted specifications, the maximum allowablecontent of residual fission products in the reprocessed uranium amounts to about10% of the natural uranium ~ activity; as an upper limit for residual y-emittingfission products 1.1 . 105MeV'Bq/d-kgf.J is usually specified. In most cases, theresidual fission product activity in reprocessed uranium is below 1000Bq/g. Duringthe enrichment process of the reprocessed uranium the fission products behave in

232UUranium 71.7a(J

Protactinium~

228ThThorium 1.913aaActinium

224RaRadium 3.64d(J-_. ..

Francium

Radon j 220Rn55.65(J

Astatine-- I- .212PO 216PoPolonium 46s13-10 7s 0.155

_...._ ' ' " - ~ - - - - - ~ ctIBismuth 212Bi ISO.55mjr,(J.

"-

Lead 208Pb 212Pb IStable 10,64h i, p-Ihallium 208TI3.054m I~

Figure 3.3. 232U decay chain (4n series)

a different manner. 99Tc is the most important fission product with regard to massconcentrations; during fluorination of the uranium fraction, technetium is converted into TcF6 which has a vapor pressure almost identical to that of UF6. In


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66 Radiochemistry during normal operation of the plant Radionuc1ides in the reactor core 67

In general , the independent yield is highest for the first members of an isobaricchain and decreases significantly in the development of the chain; the final mem-

235U fission their average number v amounts to 2.4, in the thermal 239pU fissionthe corresponding number is 2.9. TIle prompt neutrons show a continuous energyspectrum in the range f rom vir tual ly zero to more than 10MeV wit h a peak valueat about 0.7 MeV and a median energy of about 1.7MeV. Simultaneously, prompty quants are emitted in the fission reaction, showing an average energy of about1MeV.

The energy released in the thermal 235U fission amounts to about 200 MeV,consisting of the following individual contributions

Since the neutrinos escape quantitatively f rom the reactor, the fission energy whichis available for production of utilizable energy amounts to 189MeV pe r fissionreaction.

In the the rmal neutron-induced f ission react ion, the heavy nuclides preferentially disintegrate asymmetrically into a lighter and a heavier fragment with a statistical distribution of the fragment masses. This leads in the 235U fission to maximaof formation at mass numbers 95 and 138, with each showing fission yields (isobaric chain yield) of about 7%. As can be seen from the well-known fission productmass distribution curves, between the two maxima a broad and deep minimumexists with fission yields on the order of 10--2%. In the 239pufission the maxima offormation are shifted towards the mass numbers 99 and 140, while the basic shapesof the two distribution curves are quite similar. The mass distribution curve of the235U fission products shows a fine structure near mass number 133and, to a somewhat l es se r ext en t, also in the r eg ion of mass number 95; the 239pU fission yieldcurve shows a similar fine st ructure. The r easons for the fine structure are thepreferential formation of fission fragments with the magic neutron numbers 50 and82 in the nucleus (i. e. saturated neutron shells) , as well as the emission of promptneutrons from fragments which, in addition to having a stable configuration, alsohave weakly bound neutrons in the nucleus.

The mass distributions in the fission reaction are expressed by the fission yields.Here, three different yields have to be distinguished:the independent yield (or fragment yield) describing the formation of one individual fission product nuclide by the primary fission process;the cumulative yield, which means the total yield of one individual fission product nuclide as the sum of its independent yield and of its formation by decayof i ts chain precursors;the cha in yield (or i soba ri c yield), which is the cumulative yield of a de caychain.

the isotope separation process, TcF6 preferentially accompanies the lighter uraniumfraction, thus leading to a technetium content in the enriched uranium on the orderof a few ppm (specification limit 5 ppm). Ruthenium f luor ide also shows a vaporpressure similar to that of UF6, so that virtually no decontamination effect isachieved by the volatilization process. For the other fission products, as well as forthe daughter products of 232U, the fluorides are non-volatile and are collected inthe residues of the volatilization process.

The transuranium elements such as neptunium, plutonium or americium formhexafluoride compounds at their highest valency state with physical propertiesclose to those of UF6; but these compounds are not stable when the fluorine partialpressure decreases. Under such conditions they arc converted into a lower valencystate and remain as a solid product. This means that the main fraction of theseimpurities can be col lected as ash or dust ; if there isstill a small proportion remaining in the liquid UF 6 , it can be removed by a special fi lter before the container isfilled. The specified upper limit for residual transuranium act iv ity in the reprocessed uranium amounts to 2.5 Bq/gU, with 238pU, 239pU and 237Np as the guideisotopes.

The chemical process leading from the uranyl nitrate solution to the UF6 to bedelivered to the enrichment process resul ts in a decontamination factor for alphaactivity of about 100; beta and gamma activities are reduced to a lower proportion,i. e. by a factor of 2 to 4 (Beck, 1985). From the onset of spent fuel reprocessing,enriched natural uranium has also been contaminated in some instances with tracesof 236U, 99Tc,and plutonium isotopes.

3.2 Radionuclides in irradiated nuclear fuels3.2.1 Fission product generation and fuel structureHeavy nuclides are unstable; they can be disintegrated not only by a or decayor , in some cases, by spontaneous fission, bu t also by neutron-induced fission. Thefission cross sections, however, of the various nuclides show great differences.Amongthe naturally occurring fissile nuclides 235U is the only onethat can be usedin thermal reactors; in addition, the artificially produced nuclides 233U, 239pU and241pU show fission cross sections and halflives which make them appropriate foruse as nuclear fuels. In the currently operating light water reactors, which areexclusively based on uranium as the starting element, only the fissile nuclides 23SU,239pU and 241pUare of real interest.

Neutron-induced nuclear fission can be descr ibed by the general equation (according to Lieser, 1991)

A + n B + D + v . n' + ~ Upon neutron capture in the target nucleus A, a highly excited compound nucleusis generated which decays within about 10.-13 s, forming the two fragments BandD. About 10.. 17 s later , the prompt fission neutrons n' are emitted; in the thermal

Kinetic energy fission fragmentsKinetic energy prompt neutronsEnergy prompt y emissionEnergy fission product decayEnergy fission product y decayEnergy neutrinos

167MeV5 MeV7 MeV5 MeV5MeVI I MeV


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68 Radiochemistry during normal operation of the plant Radionuclides in the reactor core 69bers of an isobaric decay chain are predominantly produced by the ~ decay oftheir precursors.The two fragments formed in the fission process exhibit a total kinetic energyof 167MeV, which is distributed between the two fragments corresponding to thereciprocal rat io of the ir masses. This kinet ic energy effects a movement of thefragments in opposite directions in the fuel matrix; after a recoil length of about10urn in this matrix, they reach their rest position normally inside a D02 crystalliteafter about 10- 9 seconds. Some 95% of the kinetic energy of the fission fragmentis transformed into electronic stopping power, with a minor portion causing latticedefects, e. g. through displacement cascades. Usually, the spur of a fission fragmentis cons idered to be a cylindr ical tube hav ing a d iamete r of 10 nm and a length ofabout 6urn. Thus, the high-speed movement of the fragments through the D02lattice results in a short-term intense ionization of the fragments (average atomiccharge number +20) as well as of the lat tice a toms; near the s ta rt ing point of thespur, very high local temperatures on the order of 3000C are reached within ad iamete r of about 10 nm. The knock ing out of atoms from their regular latticepositions results in the formation of lattice defects. In addition, other direct effectsof the fission fragments are to be observed, with the most important ones beingradiation-induced creep and fission-induced densification of the VOl matrix, aswell as redissolution of small fission gas bubbles.The primary fission fragments are unstable due to an excess of neutrons in thenucleus compared to the neutron-to-proton ratios in stable medium-weight atoms.In order, therefore, to reach a stable nucleus configuration, neutrons have to beconverted into protons by emission of ~ particles. Because of the negligibly smallmass of the emitted electron, the mass of the remaining nucleus is virtually unchanged, but its electrical charge increases by one unit; this means that the transformations of the nuclei mainly occur in an isobaric chain until the stable finalp roduct of the chain has been formed. In most cases, the ~ decay reactions donot lead directly to the ground state of the daughter nucleus but to an intermediateexcited state; thus the - decay is usually accompanied by associated y emission.Examples of such isobar ic decay cha ins are shown in Fig. 3.4. At the most, theisobaric chains have about 7 members, with the average being 3 to 5.In general, the halflives of the individual chain members increase from the primary product to the final p roduct of the cha in ( there are, however , numerousexceptions from this rule). In chains with odd mass numbers there is a continuousincrease in hal fli fe; in chains with even mass numbers , on the o ther hand, themembers show shorter and longer halflivcs alternately, but with an increasing tendency. In many cases there are branched ~ decays, leading to the formation ofisomeric states which subsequently decay to the ground state by internal transitionwith y or Xvray emission and a shorter or longer halflife.However, there are deviations from this isobaric chain behavior, the most im

portant ones will be shortly mentioned.In some cases the fission product nucleus decays by neutron emission. Examplesof such delayed neutron emitters are the iodine isotopes 137, 138, 139with halflivesof 24, 6,and 2.7 seconds, respectively, and the bromine isotopes 87, 88, 89, 90withhalflives of 55, 16, 4.5, and 1.6 seconds, respectively. As a consequence of the

Figure 3.4. Fission product decay chains (schematic)

136Xestable


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Investigations performed by Bleier et al. (1984) yielded 3H concentrations in thefuel which were systematica lly higher than those calcula ted using the KORIGENcode. The reasons for these differences are no t exactly known; beyond possibleuncertainties in the code basicdata, contributions by other light element impuritiessuch as boron or l ithium in the fuel or 3He in the fuel rod fill gas cannot be ruledout. The fiss ion 3Hfragment shows a normal-distribut ion energy spectrum with amost probable energy of about 7MeV, which resul ts in a recoil l ength of about10 urn in the VO z lattice before the fragment reaches its rest position (Ray, 1968).Ternary f issions resul ting in the formation of three fragments with approximately equal masses are very seldom, with a probability of about 10-6(10.

During its stay in the operating reactor core, the nuclear fuel changes its properties, in particular its chemical composition, its radionuclide inventory, and itspellet structure. Some important aspects of U02 behavior during reactor operationhave been summarized in the review paper of Assmann and Stehle (1982). In Tables3.3. and 3.4., the element concentrations of the fission products generated in theirradiated fuel are given for different burnup values, both for an enriched uraniumfuel and a mixed-oxide fuel. These data were calculated using the KORIGEN code;the ir accuracy isassumed to be better than 10%. The by far greatest contributionsto the e lement concentra t ions are del ivered by the stable fina l products or verylong-lived intermediate nuclides of the isobaric chains; those isotopes which aremainly responsible for the radioactivity inventory of the fuel generally representonly a small f ract ion of the total e lement mass. In total , the mass concentra tionof fission products in the fuel increases nearly linearily with increasingfuel burnup,reaching about 5.4% at a burnup of 52MWd/kg HM (heavy metal) in an enricheduranium fuel and about 5.2% in a mixed-oxide fuel at the same burnup; each ofthe burnup steps chosen in Tables 3.3. and 3.4. corresponds approximately to oneadditional operating cycle of the fuel. Because of their greater number of stableisotopes, the elements with even proton numbers are general ly present in higherconcentrations than those with odd proton numbers. The fission product elementwith the highest mass concentra tion in the fuel is xenon, fol lowed by the sum ofthe rare earth elements, the sum of the light platinum elements, and zirconium;however, as concerns chemistry in the fuel, these elements are of minor relevancedue to t he ir rather inert character . Molybdenum, on the other hand, which isalso present in the fuel in compara tive ly high concentra tions , is of great significance in fuel chemistry; its ability to change between different valency states effectsa protection of the existing chemical states of various other fission product elements, in particular that of iodine, against the influence of slightly hyperstoichiometric fuel composition. Furthermore, the cesium-to-iodine atomic r at io in thefuel is of particular interest as regards the release and transport behavior of iodineduring a reactor accident; this aspect wil l be treated in more detai l in Chapters 6and 7.

neutron emission, the relevant nucleus leaves its original isobaric chain and joinsthe one with a mas s number one unit lower . The delayed neutrons account forabout 0.75% of the total neutron production in the fission reaction; their energiesare in the range 250 to 620 keV. About two-thirds of the delayed neutrons originatefrom the lighter fission product nuclides, the remainder from the heavier products.Delayed neutrons play an important role in reactor core react ivity control ; inheavy-water reactors with separate fuel channels they are taken advantage of forquasi-continuous detection of failed fuel assemblies.Fission product nuclides may also be transformed by neutron capture; the probability of an n, r reaction increases with increasing halflife and with increasingneutron absorption cross section of the relevant nuclide. Therefore, in some casesthe n,1 react ion competes with the ~ _ decay and is the limiting parameter for theconcentration of the relevant radionuclide in the fuel. The most important exampleof such a neutron capture conversion is 135Xe with a physical halflife of 9.17 hoursand a reactor neutron absorption cross section ( j of 6 .25' 10- 18 cm-; at a neutronflux, ftl, of relevant energy dis tr ibut ion in the fuel at ful l-power operation of theplant of 5 . 1013 cm-2S- I, its effective halflife is reduced according to

A.crr = A + c . (Iv decay constant)from the 9.17 h mentioned above to about 0.58 h. At a constant supply rate fromits precursor 1351, the 135Xeconcentration in the fuel iscomparatively low at reactorfull-power operation; following a significant reduction in reactor load, e. g. in thecourse of a shutdown, the 13SXe concentration in the fuel increases strongly becauseof the continuing production by 1351 decay (half li fe 6.59h) and lower or a lmostzero consumption by neutron capture. In practice, reactor operation is affected bythis enhancement in 135Xe concentrat ion in the fuel , leading to an increase in theparasitic neutron absorption cross section (xenon poisoning) and, possibly, to problems in reactor startup at low excess reactivity of the fuel , c. g. towards the end ofa fuel cycle.

Another secondary effect is the production of radionuclides which themselvesare not fission products but which are generated by neutron capture in long-livedor stable fission product nuclides. Examples of the products of such reactions are134CS and J36CS, which are separated from the true members of the isobaric chainsby the stable nuclides 134Xe and 136Xe, and which are formed by neutron capture inthe fission products 133CS and 135CS, respectively. Because of the two-fold neutroninduced nuclear reaction which is necessary for their production, their concentration in the irradiated fuel depends approximately on the square of the local neutrontluence.Symmetric fission, in which the 235U or 239pU nuclei disintegrate into two products of equal mass number, has a low probability. For this reason, their productsdo not playa significant role in the radionuclide composition of irradiated fuel.In 0.2 to 0.3% of all fissions a thi rd , light f ragment bes ide the two mediumweight products isgenerated. The light product of ternary fissions with the greatestsignificance in reactor radiochemistry is tritium 3H, the yields of which are approximately

in thermal 235U fissionin thermal 239pU fissionin thermal 241pU fission9 10- 3%2.10-- 2%3 . 10- 2% .


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72 Radiochemistry during normal operation of the plant Radionuclides in the reactor core 73Table 3.3. Fission product element concentrations(glkg HM) in irradiated LWR Table 3.4. Fission product element concentrations (g/kg HM) in irradiated LWRuranium fuel (initial enrichment 4.0% nsU) mixed-oxide fuel (initial Pu content 4.0% Pun)(By courtesy of Siemens/KWU) (By courtesy of Siemens/KWU)Element Fuel burnup (MWd/kg HM) Element Fuel burnup (MWd/kg HM)

13.0 26.0 39.0 52.0 65.0 13.0 26.0 39.0 52.0 65.0

Bromine 0.0093 0.018 0.026 0.034 0.041 Bromine 0.0078 0.016 0.023 0.030 0.038Krypton 0.16 0.31 0.43 0.54 0.64 Krypton 0.087 0.17 0.26 0.34 0.43Rubidium 0.16 0.29 OA1 0.51 0.60 Rubidium 0.076 0.15 0.23 0.30 0.38Strontium 0.47 0.82 1.11 1.36 1.57 Strontium 0.22 0.41 0.60 0.78 0.96Yttrium 0.24 0.42 0.58 0.71 0.82 Yttrium 0.11 0.21 0.31 0.40 0.50Zirconium 1.56 2.97 4.27 5.48 6.62 Zirconium 1.07 2.10 3.13 4.16 5.19Niobium 0.045 0.044 0.042 0.040 0.038 Niobium 0.036 0.037 0.037 0.036 0.036Molybdenum 1.23 2.57 3.89 5.18 6.46 Molybdenum 1.16 2.42 3.67 4.92 6.17Technetium 0.33 0.64 0.91 1.14 1.33 Technetium 0.32 0.63 0.91 1.15 1.37Ruthenium 0.84 1.76 2.76 3.85 5.00 Ruthenium 1.22 2.32 3.40 4.49 5.61Rhodium 0.17 0.35 0.50 0.60 0.66 Rhodium 0.30 0.59 0.82 0.98 1.07Palladium 0.23 0.68 1.34 2.18 3.18 Palladium 0.70 1.55 2.49 3.52 4.62Silver 0.015 0.042 0.073 0.11 0.14 Silver 0.069 0.13 0.17 0.21 0.23Cadmium 0.011 0.037 0.080 0.15 0.23 Cadmium 0.040 0.098 0.17 0.27 0.39Indium 0.0007 0.0013 0.0016 0.0017 0.0018 Indium 0.0016 0.0026 0.0030 0.0031 0.0030Tin 0.0]4 0.032 0.054 0.079 O.ll Tin 0.028 0.056 0.085 0.11 0.14Antimony 0.0058 0.013 0.020 0.027 0.034 Antimony 0.012 0.023 0.031 0.039 0.045Tellurium 0.16 0.34 0.53 0.74 0.96 Tellurium 0.24 0.47 0.70 0.93 J.I6Iodine 0.080 0.17 0.27 0.37 0.47 Iodine 0.14 0.27 0.39 0.51 0.63Xenon 2.02 4.07 6.16 8.28 10.4 Xenon 1.97 3.94 5.92 7.92 9.94Cesium 1.14 2.27 3.34 4.36 5.32 Cesium 1.25 2.49 3.68 4.80 5.85Barium 0.56 1.10 1.66 2.26 2.89 Barium 0.48 0.96 1.48 2.04 2.64Lanthanum 0.51 0.99 1.45 1.90 2.32 Lanthanum 0.44 0.88 1.31 1.74 2.16Cerium 1.30 2.34 3.28 4.19 5.07 Cerium 1.09 2.02 2.90 3.76 4.61Praseodymium 0.43 0.87 1.30 1.71 2.ll Praseodymium 0.38 0.80 1.21 1.61 2.00Neodymium 1.38 2.89 4.42 5.93 7.41 Neodymium l.15 2.45 3.80 5.17 6.55Promethium 0.13 0.18 0.19 0.19 0.17 Promethium 0.12 0.19 0.21 0.21 0.20Samarium 0.23 0.51 0.81 1.10 1.36 Samarium 0.27 0.56 0.86 1.16 1.43Europium 0.036 0.10 0.19 0.27 0.34 Europium 0.056 0.14 0.24 0.35 0.45Gadolinium 0.0094 0.037 0.10 0.22 OAO Gadolinium 0.025 0.063 0.12 0.21 0.35Totals l3.5 26.9 40.3 53.6 66.8 Totals 13.1 26.2 39.2 52.2 65.3

When comparing the data fo r the two dif fe rent t ypes of fuel, uranium fuel an d less than linear buildup. The main reasons for these deviations are the increasingmixed-oxide fuel, o ne c an see the main differences in the region of the lighter contribution of plutonium fissions having other fission yields and the consumptionplatinum elements ruthenium, rhodium and palladium where the higher fission of stable primary end products of isobaric chains by neutron capture; the extent ofyields in the 239pu fission lead to higher concentrations of these elements in the the latter effec t i s a l so influenced by the shift in the neutron energy spectrum withirradiated mixed-oxide fuel. increasing fuel burnup.

There are, however, some devia tions from the linear relationship between the The activity concentrations of selected fission product radionuclides in a ura-element concentrations of fission products, on the one hand, and fuel burnup on nium standard fuel are shown in Table 3.5. From these data, i t becomes evidentthe other, part ly r es ul ti ng in a greater than proport ional buildup, and partly a that during reactor operation the overwhelming fraction of radioactivity is caused


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74 Radiochemistry during normal operation of the plant Radionuc1idcs in the reactor core 75

by the great number of short-lived radionuclides (the very short-lived ones a re no t Table 3.5. (contd.)given in the Table). These radionuclides are already in their activity saturation state Nuclide Ha1f1ife Fuel burn up (MWdlkg HM)after a short period of reactor ope ra ti on and, t hu s, show a radioactivity which isvirtually constant over time; for this reason, the total activity concentration i n the 13.0 26.0 39.0 52.0fuel is no t sensitive to the fuel burnup, in contrast to the activity concentration of I05Rh 35.5 h 2.5 E+4 3.2 E+4 3.8 E+4 4.4 E+4the long-lived fission products. l06Rh 30 s 7.2 E+3 1.5 E+4 2.2 E+4 2.8 E+AWith regard to the radionuclide composition of irradiated fuel, there are also I09Pd 13.4 h 4.8 E+3 7.9 E+3 I . IE+4 1.4 E+4deviations f rom the simple relationship between fission product activity concentra- llOmAg 249.9 d 9.8 4.9 E+ l 1.2 E+2 2.2 E+2tions of longer-lived nuclides and fuel burnup. Similarly to the buildup of the mass IllAg 7.45 d 1.0 E+3 1.5E+ 3 1.8 E+ 3 2.2 E+3concentrations, these deviations are due to the increas ing contribution of pluto- 125Sb 2.77 a l. l E+2 2.4 E+2 3.7 E+2 5.0 E+2nium fissions to radionuclide production as well as to consumption of long-lived 11'>Te 69.6 m 9.4 E+3 1.1 E+4 1.2 + 4 1.3 E+4radionuclides by neutron capture; in extreme cases, such as with the short-lived 13ITe 25.0 m 3.0 E+4 3,1 E+4 3.2 E+4 3.2 E+4I 32Te 76.3 h 4.9 E+4 5.1 E+4 5.1 E+4 5.3 E+4

133Te 12,5 m 4.2 E+4 4.1 E+4 4.1 E+4 4.1 E+4Table 3.5. Fission product activity concentrations (GBqJkg HM; end of irradiation) 134Te 41.8 m 6.5 + 4 5.9 +4 5.4 E+4 5.0 E+4U02 fuel, initial enrichment 4.0% 2 ~ 5 U 1291 1.57 10' a 3.8 -4 8.4 -4 1.3E-3 1.9 -3(By courtesy of SiemenslKWU) 1311 8.02d 3.4 + 4 3.5 E+4 3.6 E+4 3.7 E+41321 2.3 h 5.0 E+4 5.1 + 4 5.3 + 4 5.4 E+4Nuclide Halflife Fuel burnup (MWd/kg HM) 133J 20.8 h 7.4 + 4 7.3 E+4 7.3 + 4 7.3 E+4

1341 52.0 m 8.1 E+4 8.0 E+4 7.8 E+4 7.7 E+413.0 26.0 39.0 52.0 1351 6.61 h 6.9 E+4 6.8 E+4 6.8 E+4 6.7 +4J ~ 3 X e 5.25d 7.4 E+4 7.3 E+4 7.3 + 4 7.3 E+43H 12.3 a 6.6 1.4E+l*) 2,1 E+ l 2.8 E+1 mXe 9.10 h 2.1 E+4 1.9 E+4 1.7 E+4 1.6 E+479Se 6.4' 104 a 5.7 E-3 1.1 E-2 1.6-2 2.2 E-2 13SmXe 15.3 m 1.4 + 4 1.5 +4 1.5 E+4 1.6 +4s2Br 35.3 h 4.5 E+ 1 9.0 E+ I ].4 +2 2.] + 2 13RXe 14.1 m 6.4 E+4 6.1 + 4 5.8 E+4 5.6 E+483Br 2.40 h 5.0 + 3 4.5 E+3 4.1 +3 3.8 +3 \34CS 2.06 a 9.3 E+2 3.4 + 3 7.0 E+3 1.2 E+485Kr 10.76 a 1.7 E+2 3.1 + 2 4.2 +2 5.1 E+2 136CS 13.2 d 7.3 + 2 1.4 E+3 2.2 E+3 3.1 E+38s"'Kr 4.48 h 1.2 E+4 1.0 E+4 9.0 E+3 8.0 E+3 137CS 30.17 a 1.6 E+3 3.1 E+3 4.7 E+3 6.1 E+3&7Kr 76.3 m 2.2 E+4 1.9 E+4 1.7 E+4 1.4E+4 138CS 32.2 m 7.0 E+4 6.7 E+4 6.4 E+4 6.3 E+488Kr 2.84 h 3.2 E+4 2.7 E+4 2.4 E+4 2.0 E+4 139Ba 83.06 m 6.8 E+4 6.6 E+4 6.4 E+4 6.3 E+488Rb 17.8m 3.2 E+4 2.8 E+4 2.4 E+4 2.1 + 4 140Ba 12.75d 6.7 E+4 6.5 E+4 6.3 E+4 6.1 E+489Rb 15.2 m 4.3 E+4 3.6 E+4 3.1 E+4 2.7 E+4 14La 40.27 h 6.8 E+4 6.6 E+4 6.5 E+4 6.5 E+4&9S r 50.5d 4.4 E+4 3.8 E+4 3.3 E+4 2.8 E+4 141La 3.93 h 6.2 E+ 4 6.0 E+4 5.9 E+4 5.8 E+490Sr 28.5 a 1.3 E+3 2.4 E+3 3.3 E+3 4.1 E+3 141Ce 32.5d 6.3 E+4 6.1 E+4 6.0 E+4 5.9 E+491S r 9.5 h 5.3 +4 4.6 E+4 4.1 E+4 3.6 +4 144Ce 284.8 d 3.3 E+4 4.4 E+4 4.6 E+4 4.5 E+490y 64.1 h ].4+3 2.5 E+3 3.5 E+3 4.3 E+3 143Pr 13.57d 5.9 E+ 4 5.6 E+4 5.4 E+4 5.1 + 4

91 y 58.5 d 5.5 E+4 4.8 + 4 4.3 E+4 3.7 + 4 147Nd 10.98d 2.4 E+4 2.4 E+4 2.4 E+4 2.3 E+493Zr 1.5. 106 a 2.9 E-2 5.6 E-2 7.9 E-2 1.0 E-1 6.3 E+6 6.2 E+6 6.1 E+6 6.1 E+695Zr 64.0 d 6.5 + 4 6.3 E+4 6.0 E+4 5.7 + 4 Totals- -7Zr 16.8 h 6.3 E+4 6.2 E+4 6.] E+4 6.0 E+4 *) read as 1.4 . I019sNb 35.0 d 6.5 E+4 6.3 E+4 6.0 E+4 5.7 + 497Nb 74m 6.4 E+4 6.3 E+4 6.1 E+4 6.1 + 4 krypton and rubidium isotopes, the activity c o ~ c e n t r a t i o n ? some r a d i o n ~ c l i d e s9Mo 66.0h 6.7 E+4 6.6 E+4 6.6 E+4 6.6 E+4IOIMo 14.6m 5.8 E+4 6.0 E+4 6.1 E+4 6.2 E+ 4 decreases steadily with increasing fuel burnup. Since the a c t l v l t ~ ' c ~ n c e n t r a t l o n s of99']c 2.1 . 105a 2.1 E-I 4.0 E-I 5.7 E-I 7.1 E-1 the fission products in irradiated mixed-oxide fuels show no baSICdifferences, Table99mTc 6.0 h 5.8 E+4 5.8 + 4 5.8 E+4 5.8 E+4 3.5. is limited to uranium fuel only.I03Ru 39.4 d 4.7 + 4 5.4 E+4 6.0 E+4 6.5 + 4 In addition to the fission products, actinide nuclides are generated in the nuclearlO5Ru 4.4 h 2.6 E+4 3.4 E+4 4.0 E+4 4.7 + 4 fuel during irradiation. The main starting reaction for thebui ldup of the transura-106Ru 368 d 6.4 + 3 ].4 E+4 2.0 E+4 2.6 E+4 nium nuclides is neutron capture in the 238U nucleus leading to short-lived 239U


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76 Radiochemistry during normal operation of the plant Radionuclides in the reactor core 77which decays by ~ . emission to 239Np; this nuclide , in turn, decays to 239PU. Ascan be seen from Fig. 3.5., where a highly s implif ied version of the buildup anddecay react ions is presented, a large number of different nuclides is formed. Theprinc ipal react ions for buildup of the transpJutonium elements from the lightertransuranium elements neptunium and plutonium are again neutron capture andsubsequent p- decay leading to the curium isotopes; the comparatively short halflives of these isotopes and their large fiss ion cross sec tions are the main reasonsthat the yield of higher act inides is negligibly small . In most cases , the l ightertransuranium nuclides are a emitters with rather long halflives; some of them showalso disintegration by spontaneous fission.In Tables 3.6. and 3.7., the actinide element concentrations are given as a function of fuel burnup, both for uranium and for mixed-oxide fuel. In both types offuel, besides uranium, plutonium is the most abundant element, amounting in theuranium fuel at a burnup of 52MWd/kg HM to about one third of i ts concentrat ion in a mixed -ox ide fuel at the s ame burnup. Whereas in the mixed-oxide fuelthe plutonium content decreases s teadily with increas ing burnup, in the uraniumfuel the initially steep increase declines at burnup values beyond about 40MWd/kgHM, reaching an almost constant value with continued irradia tion. The higheractinides americium and curium show a steady growth with increasing burnup inboth types of fue l; americium, curium and the heavier e lements are present in themixed-oxide fuel in s igni fi can tly h ighe r concent ra tions than in uranium fuel,whereas neptunium shows higher concentrations in the high-burnup uranium fuel.In Tables 3.8. and 3.9. the activity concentrations of actinide nuclides both inuranium and in mixed-oxide fuel are shown. The 239pu activity in the uranium fuelincreases during the firs t fue l cycle and then remains a lmost constant with only aslight further increase, while the 241pU activity increases steadily with increasingburnup. Tnthe mixed-oxide fuel the 239pU concentration decreases steadily with241PU remaining virtually constant over the whole i rr ad ia tion per iod under consideration. During reactor operation, the fissile nuclides 239pU and 241PU producedare partly consumed again by in-situ fission; in a uranium fue l in the burnup rangefrom 40 to 60MWd/kg U, about 30 to 400- '71:10Nt--

,.-...S:lEQ..c:'02cG"'0(V(:5:.0. .5-r:Q'0::l>=o:95(.)'-o!ci5.r;


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78 Radiochemist ry during normal operat ion of the plant Radionuclides in the reactor core 79Table 3.6. Actinide element concentrations (g/kgHM) in irradiated LWR uranium fuel Table 3.8. Actinide activity concentrations(GBq/kg HM ; end of irradiation)(initial enrichment 4.0%235U) U02 fuel, initial enrichment 4.0%235U(By courtesy of SiemenslKWU) (Bycourtesy of Siemens/KWU)Element Charge Fuel burnup (MWd/kg HM) Nuclide Halflife Decay Fuel burnup (MWdlkg HM)(g/kg HM) mode13.0 26.0 39.0 52.0 65.0 Charge 13.0 26.0 39.0 52.0

Uranium 1.00E+3 9.82 E+2 9.65E+2 9.49 E+2 9.34 E+2 9.19 E+2232U 70.0 a 0- 0 4.2 E-5 1.9E-4 5.3 E-4 !.I E- 3

Neptunium 0 1.73 E-1 3.8 E-I 6.3 -1 8.6 E-1 1.04 233U 1.59 105 a 0- 0 2.4 E- 6 3.4 - 6 3.5 E-6 3.] E-6Plutonium 0 5.01 8.0 1.01 + 1 1.17E+l 1.28E+ I 234U 2.45 105 a 0- 7.2 E-2 6.1 E- 2 5.1 E--2 4.1 E- 2 3.3 E- 2Americium 0 5.3 -3 4.9 E-2 1.6 E-1 3.3 E-1 5.5 -1 235U 7.04 lOS a a 3.2 E-3 2.1 E- 3 1.4 E- 3 8.3 E- 4 4.5 E- 4Curium 0 5.3 E-4 9.8 E-3 5.1 -2 1.6E-1 3.7 E-l 236U 2.34 ' 107 a 0- 0 5.8 E- 3 9.6 --3 1.2 E-2 1.3 E- 2Berkelium 0 1.0 E-I I 7.7 -10 1.6 E-8 1.5 E-7 237U 6.75d W 0 1.6 E+4 2.6 E+4 3.4 +4 4.0 E+4Californium 0 4.1 E -12 4.5 E-IO 1.2 E-8 1.4E-7 238U 4.47 J09 a (J. 1.2 E- 2 1.2 E- 2 1.2 E- 2 1.2 E- 2 1.1 E-2Einsteinium 0 1.6 E-16 4.9 E-14 2.7 E-12 5.0 E-ll 239U 23.5 m ~ 0 5.9 E+5 6.6 E+5 7.6 E+5 8.5 E+5236Np 1.15 105 a .. 0 6.8 -2 2.2 E-1 4.6 E- { 7.3 E-I237Np 2.14 J06 a (J. 0 2.7 E- 3 8.0 E-3 1.4 -2 2.0 - 2238Np 2.12 d ... 0 2.1 E+3 7.1 E+3 1.4 E+4 2.3 E+4gamma scan onecan observe additional activity minima at the pellet interfaces due 239Np 2.36 d p- O 5.9 E+5 6.6 E-+5 7.5 E+5 8.5 E+5to the lower fuel densi ty in these regions, which is caused by the pellet dishings; at 236pU 2.85 a 0- 0 2.6 E-3 1.5 E-2 3.8 E- 2 7.1 E- 2high burnup levels, the dishings have largely disappeared as a consequence of fuel 23SpU 87.7a a 0 6.1 3.7 E+I 1.1 E+2 2.1 E+2swelling. In such fuel rods the pellet interfaces are no longer (or only very weakly) 239pU 2.41. 104 a 0- 0 9.0 1 .2 E+ 1 1.3 E+ 1 1.3 E+ 1indicated by gamma scan minima. The gamma scans demonstrate that during 140pU 6.55 103 a 0- 0 6.2 1.4 E+ 1 1.9 E+ I 2.3 E+I241PU 14.4a W 0 1.2 E+3 3.8 E+3 5.8 E+3 6.9 E+3steady-state operation at linear heat ratings of the fuel rods typical for LWRs there 242pU 3.76' 105 a (J. 0 3.7 E- 3 2.8 E- 2 7.7 E- 2 1.4 E-Ii s no measurable axial fission product migration. In BWR fuel rods, the axial gross 244pu 8.26' !Oi a a 0 8.9 E-11 1.8 E- 9 9.8 E- 9 3.0 E- 8gamma distribution is asymmetric, which demonstrates the impact of the control 241Am 432.6 a 0- 0 4.7 E-I 2.9 6.1 8.2rod positions. However, as can be concluded from the distributions of long-lived 242"'Am 141 a IT 0 1.7 E- 3 1.6 E-I 3.7 - 1 5.1 E-]and short-lived fission products, this effect results from the fission rate distribution 243Am 7.37. l O a a 0 1.1 E-2 1.9 E-I 8.0 E-1 1.9in the f ina l fue l cyc le rather than from burnup distribution. By an appropriate 241Cm 162.8d a 0 5.1 E+I 6.5 E+2 2.0 E+3 3.5 E+3operational mode of the reactor core, an evenly distributed burnup can be finally 24JCm 28.5 a a 0 6.6 E- 3 1.9 E-l 9.7 E-I 2.4obtained. 244Cm 18.1 a 0- 0 3.[ E-1 [,3 E+ I 9.7 E+I 3.7 E+2245Cm 8.50 ' 103 a ().. 0 9.8 E- 6 7.7 E- 4 8.3 E- 3 4.0 E-2Table 3.7. Actinide element concentrations (g/kg HM) in irradiatedLWR mixed-oxide fuel(initial Pu content 4.0% Pun)(By courtesy of Siemens/KWU)Element Charge Fuel burnup (MWdJkgHM)(g1kg HM) 13.0 26.0 39.0 52.0 65.0Uranium 9.37 E+2 9.29 E+2 9.22 E+2 9.14 E+2 9.05 E+2 8.94 E+2Neptunium 0 9.35 E-2 1.45 E-I 2.03 E-l 2.74 E-l 3.46 E-IPlutonium 6.32E +1 5.61 E+I 4.93E+ 1 4.29 E+I 3.75 E+ 1 3.32 + IAmericium 0 1.08 1.95 2.64 3.21 3.67Curium 0 1.2 E-l 4.1 -} 8.6 E-1 1.51 2.37Berkelium 0 4.7 E-12 4.6 E-IO 7.0 E-9 5.7 E-8 3.2 E-7Californium 0 1.7E-lO 3.7 E-9 3.9 E-8 2.7 E-7Einsteinium 0 1.9 E-15 i.: E-13 2.7 E-]2 3.5 E-11

An important feature in characterizing irradiated nuclear fuels is the analysis ofthe distribution of fission products and activation products in the irradiated fuelrods, as well as i n i nd iv idua l fuel pel le ts . For this reason , some of the analyticalt echn ique s devel oped t o t hi s end wi ll be shortly described, each of which meetsspecific requirements.Very often, determination of the f ission gas fraction released during operationf rom the fuel pel le t to the rod free volume is required. To this end, the fuel rod ispunctured inside a ho t cell using a special experimental device; the gas inventoryof the rod, mainly consisting of helium, is then col lec ted in an evacuated volumeand measured by volumetric methods. Analysis of the individual gas compositionis usually performed by mass spectrometry.Axial distribution of fission products in a fuel rod usua lly i s determined bygamma scanning; in thi s technique , the fuel rod is passed before the slit of a colli-


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80 Radiochemistry during normal operation of the plant Radionuclides in the reactor core 81Table 3.9. Actinide activity concentrations (GBq/kg HM;end of irradiation)mixed-oxide fuel, initial Pu content 4.0% PUfj"s(By courtesy of Siemens/KWU)Nuclide Fuel burnup (MWdJkg HM)

Charge 13.0 26.0 39.0 52.0232U 0 2.0 -5 8.0 -5 1.9 E-4 3.6 E-42 ~ 3 U 0 5.3 -7 9.3 E- 7 1.2 E-6 1.5 -6234U 1.1 -2 1.3 E-2 1.4 E--2 1.5 -2 1.5 -22J5U 5.3 E-4 4.4 -4 3.5 E-4 2.7 E- 4 1.9 -4236U 0 6.3 -4 1.2 E- 3 1.6 E- 3 1.9 -3237U 0 4.1 +3 5.5 E+3 6.9 E+ 3 9.1 +323SlJ 1.2 E-2 1.2 -2 1.1 E- 2 1.1 -2 1.1 E- 2239U 0 4.7 E+5 5.0 E+5 5.4 E+5 6.4 E+5236Np 0 2.7 -2 6.3 --2 1.1 E-I 1.9 - 1237Np 0 1.0 -3 2.3 -3 3.7 E-3 5.2 E- 3238Np 0 5.3 +2 1.3 E+ 3 2.4 +3 4.1 E+3239Np 0 4.6 E+5 5.0 + 5 5.4 E+5 6.4 E+5236pu 0 3.5 E- 3 8.4 E - 3 1.5 E-2 2.5 E ~ 238pU 1.2 E+ 3 1.1 +3 9.7 E+2 9.3 +2 9.2 E+2239pu 7.7 +1 5.9 E+ l 4.3 E+ l 3.2 +1 2.5 +1240pU 1.3 +2 1.4 E+2 1.3 E+2 1.2 + 2 1.0 + 2241PU 2.4 + 4 2.9 E+4 3.0 E+4 3.0 E+4 2.7 + 4242pU 7.6 -1 7.4 E-I 7.7 E-I 8,1 E-I 8.5 E-I244PU 0 4.6 E- 8 9.3 -8 1.5 E- 7 2.3 -7241Am 0 3.4 E+ I 6.0 E+ l 7.1 +1 7.0 +1242mAm 0 1.1 2.8 4,1 4.324,'Am 0 6.0 1.1 E+ I 1.5 E+ I 2.0 E+1242Cm 0 3.5 E+3 1.0 E+ 4 1.6 E+4 2.1 +4243Cm 0 6.5 E-) 3.5 8.5 1.5 E+I244Cm 0 2.6 E+2 9.7 E+2 2.1 E+3 3.9 E+324SCm 0 3.3 E- 3 2.0 E-2 5.8 E-2 1.2 s- :

mated gamma detector, either in a hot cell or in the spent fuel pool of the reactorplant (provided the required devices can be installed there). By using appropriatedetectors, gross gamma scans (representing the total gamma activity) can be obtained as well as nuclide-specific scans (for example, l:nCs distribution). The axialresolution of this technique depends widely on the collimator used; with a largeopening of the col limato r, the relat ive burnup of whole fuel assembl ies can bedetermined. When absolute values of the local fission product concentrations or ofthe fuel assembly burnup are required, a careful calibration of the detection systemhas to be performed.For the determination of the fission product distribution over the cross sectionof a fuel pellet, autoradiography isthe simplest method with regard to instrumentalneeds; this technique, however, requires extensive experience in order to obtainoptimum results at the very high radioactivity level of the materials. The distribu-

T I i l d i ' ' ' : : m f . t ' J ~ m B 3 r r r itlr::jH;j'iUlr i ;J M 1 mf JJ l i f ' l Tflrl Hm: iJ Ii jill jjj! !!:t;!i mm!li :jti"v. '!1 H" ' Hi:" ., -t rt .. . :.1.11 " ~ I I - i II - 'J 'III I < I' - ": n II ,. "i '. : ,. , " I "I' Of{,t':\i' ' ' I t I t . l i ' ~ W l--!iP. 1J F i. " "; , ~ Tti:H' , -'1 . .c' Ii! It I . I Ii J.:,'IH .1' I ; qiffii"i.1li!i!l: i; Iii' H:; Ji.. t ". tL , , -W- " f14" 'J


14/45

82 Radiochemistry during normal operation of the plant Radionuciides in the reactor core 83tion of p-- 1Y emitters can be imaged by special X-ray emulsions, while a emitterscan be local ized even in the presence of a large excess of p- ' f nuclides by usingspecial polymere foils, with the defects caused by a damaging being developedby caustic etching. A further advantage of this technique is that it can be easi lyaccornodated to different activity levelsof the specimen by selection of an appropriate exposure time. Usually, local activity resolutions on the order of a few micrometers can be achieved. Primarily, the results obtained are of a qualitative nature; by optical density measurement after an appropriate calibration of the relationship between radioactivity and density, semi-quantitative results are also possible.The distribution of y-ernitting fission products across a fuel pellet section canalso be determined by micro-gamma scanning. With a carefully designed system, alocal resolution on the order of 250 to 500Jim can be achieved (ManzeI et al. ,1984).A global analysis of all radionuclides, as wellas of non-radioactive constituents,can be performed by taking microsamples from the fuel pellet, e.g. by mechanicaldrilling or by an ultrasonic technique. By using sophisticated sampling techniques,a local resolution on the order of 300urn is possible. In order to determine thefission gas contents, the microsamples are dissolved under vacuum in de-aeratednitric acid; the released gases are collected and analyzed by mass spectrometry.Other constituents are determined by an appropriate separation and measurementmethod; as an example, determination of ! 291 in such microsamples by neutronactivation analysis is described in Section 3.2.3.3.Determination of the concentrations of the different isotopes of uranium, plutonium and the transplutonium clements plays an important role in the characterization of irradiated nuclear fuels. Radiochemical analysis procedures with subsequentradiation measurement generally cannot deliver the high accuracy required of theresults. For this reason, the standard procedure applied for this task is based onan anion exchange separation of uranium and plutonium from the fission productsand from each other in nitric acid solution, after addition of 233U and 242pU spikesto the solution to monitor the separation yields. In the isolated uranium and plutonium fractions the ratios of the concentrations of the individual isotopes to theconcentration of the relevant spike nuclide are determined by high-resolution massspectrometry using thermionic excitation. The high sensitivity of mass spectrometryrenders possible the isotope analysis in microgram to milligram samples of irradiated fuel with satisfying reliability and accuracy. Because of its comparatively lowmass concentration, 238pU determination by a spectrometry in an aliquot of theisolated plutonium solution usually offers higher accuracy than mass spectrometry.In a second aliquot of the same sample, the concentrations of the burnup monitor 146Nd and of the gadolinium isotopes in irradiated U02 - Gd 20 3 fuel can alsobe determined. In this procedure, the rare earth elements are first separated fromuranium, plutonium and the fission products by anion exchange from hydrochloricacid solut ion; by addit ion of known amounts of J50Nd and 16Gd spikes to thesamplesolution, correction for lossesof both rare earth elements during the separation process can be made. In the second step, the neodymium and gadol in iumfractions are isolated from the other rare earth elements by cation exchange using

Figure3.7. Enhanced product ion of transuranium nuclides at the fuel pel let rim (autoradiographic image)(By courtesy of SiemenslKWU)a-hydroxy isobutyric acid as an eluant; finally, the ratios of the concentrations ofthe isotopes to be determined to that of the added spike isotopes are measured bymass spectrometry.In recent years, instrumental analysis techniques have gained more and moreimportance in the determination of fission product distribution in the fuel pellet,X-ray microanalysis and secondary ion microanalysis with local resolution ofabout 2urn, Auger electron microanalysis with a local resolution down to 0.] urn,as well as secondary ion mass spectrometry (e. g. Zwicky et al., 1989) are capableof furnishing informations on almost all chemical elements. By appropriate shielding and encapsulation of the excitation system, highly radioactive and also highlya-active samples can be analyzed without difficulties. These techniques furnish abetter local resolution than micro-gamma scanning and microsampling; however.in most cases they are only able to measure the element composition, and do notdistinguish between different isotopes. Often, the instrumental microanalyticaltechniques (such as electron probe microanalysis) give the concentrations of thefission products dissolved in the U02 lattice and of those trapped in small bubbleswithin the grains. Little, if any, of the fission gases and volatile fission productscontained on the grain boundaries and in intergranular bubbles contribute to themeasured X-ray intensities because in most cases the analyses are made away fromthe grain boundaries.The radial distribution of the fission products over the cross section of the fuelpellet is primarily governed by the profile of the thermal neutron flux. However,there are two effects leading to variations in the distribution of some of the fissionproducts. The first one of these effects is the preferential formation of plutoniumin the outermost pel let zones due to epi thermal neutron capture in the 238U nucleus.This effect ismore pronounced in high-burnup fuel than in fuel with a lowerburnup. Fig. 3.7. shows an alpha autoradiographic image of a fuel pel let crosssection where the enhanced a activity in the outer ring can be seen; it must nonetheless be pointed out that the greatest fraction of this a act iv ity is not due to the


15/45


Figure 3.8. Uranium and plutonium concentrations in the surface region of a high-burnupoxide fuel pellet(Kleykarnp, 1990 a)

- ~ ' - - - - - - - I . . ' - ".. --

os I. '. -r-

----- ----- r/t 0 .---.--.-

10I i 't2 r Xenon

1 0 ~ < gf ' '. ,S , ~ 2 o1.5 I

; L 910.m10 . --------- ,u Neodymium A.A~ 2 "$ OS f A .."$.I

"f " Z " , " N ~OS i A ., " 'A~ 2 o ,o 02 0[. 06 08 10

Figure 3.9. Fission product distribution as a function of the relative fuel pellet radius in aLWR high-burnup oxide fuel(Kleykamp, 1990a)

as a consequence of isotope burnup. Gadolinium consists of several naturally occurring isotopes with markedly different nuclear properties. During fuel operation,the strongly neutron-absorbing isotopes 157Gd (natural isotopic abundancy 15.7%,thermal neutron absorption cross section 2 .54 ' 10- J9 cm-) and 155Gd (20.6%,6.1 . W- 20cm 2) are preferentially consumed, thus r esul ti ng in an inc rease in therelative concentrations of the weakly neutron-absorbing isotopes 158Gd (24.7%,2.5 . 1O-24cm2) and 156Gd (20.6%, 1.5 . IO-24cm2). The nuclear reactions create apronounced profile of the ind iv idua l gadol in ium i so topes across the spent fuelpellet. In the outer pellet zone, 155Gd and 157Gd decrease to nearly zero concentration after one fuel cycle of irradiation, whereas 156Gd and 158Gd show a cor re sponding increase; 154Gd and '6oGd distributions are virtually no t affected by theneutron irradiation.

The second effect leading to an inhomogeneous fission product distribution, inparticular in the radial direction, is migration in the thermal gradient. This effectmainly affects the gaseous and the volatile fission products; its extent depends onseveral parameters suc h as the linea r heat rating of the relevant fuel rod and,consequently, the temperatures in the pellets during reactor operation, as well as

1.0

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:$. '0

! :t ; ..;.3o 200 [.00 600 800 1000.---- Jl.m - . - - . - -.------------

plutonium isotopes but to the higher actinide Isotopes generated, in particular tothe curium isotopes.Detailed studies of the preferential plutonium production in the outer pelletzones were also per formed using X-ray microanalysis techniques. As a typical example, the investigation of a high-burnup PWR fuel rod (initial enrichment 3.2%235D, burnup 55.9 MWd/kg D, time-averaged linear heat rating 210W/cm) reportedby Kleykamp (1990 a) shall be mentioned here. In this material , the average PU02concentration (which represents the difference between total plutonium productionand plutonium consumption during fuel operation) amounted to 1.38%; in theouter 100 flJI1 zone this value increased steeply to about 3.8% in the surface region,whi le in the center of the pellet it only amounted to 1.2% PU02 (see Fig. 3.8.) . Thisincrease in concentration in the pel let r im zone applies to all plutonium isotopes.The enhanced plutonium concentration in the outer zone of the pellet means higherfission density, resulting in a corresponding increase in the concentrations of thefission products from about a 6% average value to about 14(10 in the rim zone (seeFig. 3.9.). The almost identical distribution of neodymium, zirconium and cesiumindicates that cesium migration in the thermal gradient contributes only insignificant ly to the higher level near the pellet surface. Xenon distribution seems to showan opposite behavior to that of the other fission products, since the analyticalresults indicate that the fuel grains in the outer region are depleted in xenon. However, th is region is character ized by a high gas bubble density. It was shown , e. g.by Manzel et al. (1984) and by Manzel and Eberle (1991), that most of the xenonis confined in these bubbles and has not been released to the fuel rod free volume.

The radial distribution of gadolinium, which is added to both PWR and BWRfuels as a burnable poison (see Section 1.1.2.), is also altered during fuel operation


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86 Radiochemistry during normal operation of the plant Radionuclides in the reactor core 87the stoichiometry of the fuel . Migra tion in the thermal gradien t is a secondaryeffect and will be treated in more detail in Section 3.2.3.As was described in Section 1.1.2., fresh mixed-oxide fuels show an inhomogeneous structure with U/Pu oxide master-mix agglomerates embedded in the V0 2matrix. In the course of reactor operation under steady-state conditions, only littleU-Pu interdi ffus ion is observed up to burnup values of about 40 MWd/kg HM(linear heat ratings around 240W!em). However, as a consequence of power tran

sients which reach values near 400W/cm for about 50 hours (corresponding to apellet central temperature on the order of 2000 C), the mixed-oxide agglomeratesin the central pellet region are completely dissolved (Goll et al., 1993).Due to theirhigher concentration of fissile nuclides, these agglomerates show a fission densityand a resulting burnup which is considerably higher than that in the bulk of thefuel pellet. In a mixed-oxide fuel with an integral burnup of 38.8 MWdlkg HM,local burnup values in the mixed-oxide phases between 130 and 200MWd/kg HMhave been measured using an electron probe microanalyzer (Walker ct al., (991).As a consequence of fission recoil, the fission fragments travel in the fuel overa distance ofabout )0urn, i. e.a distancecomparable to the sizeof the D02 crystallite dimensions. Because of the isotropic distribution of fragment movement, thiseffect does not lead to variations in the homogeneous distr ibution of the fissionproducts in D02 fuel. In mixed-oxide fuels, however, the recoil effect results in

a measurable depletion of the fission product concentrations in the master-mixagglomerates and a corresponding increase in the surrounding U02 matrix. Measurements using microanalytical techniques (Walker et al., 1991) have determinedexcess xenon and cesium concent ra tions in these matrix zones of about 35%. avalue which agrees satisfactorily with recoil calculations (the principles of calculation of recoil release from samples showing dimensions comparable to the fissionfragment recoil length were presented e. g. by Wise, 1985). Likewise, fission fragment recoil is the main reason for the buildup of the radionuc1ide inventory in thefuel pellet - cladding gap (see Section 3.2.4.).While the bulk of the fission products is more or less homogeneously distributedin the fuel matrix in an a tomic-dispersed s ta te , new fission product phases areformed with increasing burnup (i. e. increasing fission product concentration), obviously resulting from an exceeding of the solubility limits of the relevant clementsin the oxide matrix. Two such types of phases can be distinguished, a ceramic oneand a metallic one. According to Kleykamp et al. (1985), the newly formed ceramicphase crystallizes in the cubic perovskite lattice and consists of oxides of the elements uranium, plutonium, barium, strontium, cesium, zirconium, molybdenumand the rare earth elements (RE), with a general composition (Bal-x-ySrxCsy)(U,PU,RE,Zr,Mo)03, as has been derived from numerous microanalytical investigations. As for the mechanism of formation, it is assumed that the elements strontium, yttrium, zirconium, which in lower concentrations are soluble in the V02lattice, begin to form a particular phase after having reached their solubility limits:rare ear th elements appear in this new phase to a s igni fican t extent only at veryhigh burnup values of the fuel.The metallic s phase is a crystalline compound with a hexagonal lattice consisting of the elements molybdenum (24-43 weight%), technetium (8-16


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Figure 3.10. Crack formation in irradiated fuel pellets; longitudinal and cross-sectional view(By courtesy of Sicmens/KWU)

During reactor operation, the structure of the fuel is significantly altered due toneu tron i rrad ia tion and thermal load. The most obv ious evidence of this is theformation of an irregular network of cracks in the pellets in the radial as wellas theaxial direction (see Fig. 3.10.) evenat low burnup levels. These cracks are formed asa consequence of the steep temperature gradient between the hot pellet center andthe cooled cladding. One of the effects of the cracks is a significant increase in thereal surface area of the pellet, which facilitates the release of gaseous fission products from the pellet matrix to the gap. In statistical evaluations of a great numberof metallographic sections of spent fuel rods which were operated under standard LWR conditions, the specific surface area of the fuel was found to be210-230mm2/ g UOl (compared to about 50 mm2/g in fresh fuel pellets), with nosignificant dependence on fuel burnup; in BWR fuels, the specific surface area is

about 180mm2/ g UOl, i. e. somewhat lower than in PWR fuels. In general, withincreasing heat ratings the average fuel surface area increases, reaching about260 rnm2/ g U02 at 350 W/crn. Mixed-oxide fuels and U02-Gd203 fuels show anapproximately identical crack development, leading to similar average specific surface areas as in a s tandard UO l fuel.A high-burnup PWR fuel which was operated at an essentially constant load ischaracterized by a thin outer zone in which the original grain structure has disappeared and has been replaced by a fine-crystall ine structure in which small rodshaped precipitates are embedded, as well as by a multitude of gas bubbles havingdiameters between 0.1 and 1.5urn. In the pores of this rim zone, precipitated particles can be observed using high-resolution electron microscopy. Next to the rimzone, the pellet shows an annular zone with a grain structure quite similar to theoriginal state , which contains a few gas bubbles and precipitated phases in thegrains. Another characteristic feature of such a high-burnup fuel is the development of one or more annular zones which are connected with the local release offission gases. Apparently, these annuli indicate areas of beginning release of fissiongases; the conditions necessary for this transformation such as temperature, fissionrate and fission gas concentration are reached repeatedly during the residence timeof the fuel inside the reactor core. It isassumed that the outer annulus was generated in the course of the final fuel cycle. In the inner annulus, the number of gasbubbles inside the grains decreases strongly, whereas the size and number of thebubbles at the grain boundaries increase markedly. In the pellet center, only isolated gas bubbles are observed inside the grains while large bubbles are present atthe grain boundaries and in the channels connecting the grain surfaces, which aretypical for the hot central region.During the first irradiation phase, fuel porosity decreases due to the dissolutionof the fine pores; with increasing burnup, the porosity increases again due to theprecipitation of fission gas bubbles, and the pore size distribution shifts to somewhat larger pore diameters. The porosity of mixed-oxide fuel is characterized bythe large bubbles in the plutonium-containing particles mentioned above, which inthe pellet center develop large bubbles at the grain boundaries, partly generatingconnected channels. At high burnup, the fraction of open porosity originatingfromfuel fabrication is largely covered by channel formation between the fission gasbubbles. The fraction of open porosity as related to total porosity fluctuates between I and 30%.The size of the oxide grains also changes during reactor exposure of the fuel,mainly caused by the prevailing temperature which shows a radial parabolic distribution from 800-1200 C in the pellet center to about 450C at the pellet r im (seeFig. 1.9.) . In the outermost rim zone with a width of 50 to 130urn (depending onthe burnup), f ine particles with a grain size of less than I urn were formed. Anintermediate zone shows grain sizes quite similar to the original shape, whereas inthe di rect ion towards the pellet cen te r the grain sizes moderately increase to8-12 urn. The fuel structuremay also be changed by power transients, inparticularat high burnup levelsand a high final heat rating of the transient. For a U02 fuelirradiated to 45MWd/kgU i t was observed after a transient reaching 410W/crnthat the central zone of the pellet was characterized by elongated grains and by aporosity resembling to some extent that known from columnar grain growth. Fur-

Fig.3.lObig.3.lOa


18/45


19/45

92 Radiochemistry during normal operation of the plant Radionuclidcs in the reactor core 93be applied to determine the relat ive contribution of the final irradiation period tothe total burnup). Second, the nuclear data of the nuclide to be measured, such ascumulative fission yield, and, for radioactive burnup monitors, the halflife anddecay mode, must be known with sufficient accuracy; further, the neutron capturecross section of the monitor nuclide should be small in order to minimize errorsdue to a neutron-induced burnup of the nuclide. Third, the monitor nuclide shouldnot show any mig ra ti on t endency in the fuel during operation to avo id errorscaused by inhomogeneous distribution of the relevant nuclide in the fuel, resultingpotentially in non-representative sampling. When determination of the integral fuelburnup is required, the fission yields of the monitor nuclide in 235U fi ssion and in239pu fission should be as similar as possible; on the other hand, when differentiation between the two types of f issions is required, these yields should show pronounced differences. Finally, the analytical procedure necessary for determinationshould be comparatively simple to facilitate the work.It is quite hard to identify a fission product nuclide that fulfils all these requirements ideally. For LWR fuels showing comparatively low operating temperatures,137CS has often been applied; its gamma ray at 0 .662 MeV is easy to measure, evenwithout chemical separation of the nuclide from the other radionuclides, its halflifeof 30 years is long enough even for very long operation and decay periods and itsdecay scheme is known with a high deg ree of accuracy. However, in high-burnupfuels migration of this radionuc1ide in the fuel cannot be ruled out completely.Because of the lack of other suitable radioactive fission products, attention hasbeen f ocused on the use of stable fission product nuclides. Extensive research hasshown that !46Nd is an optimum burnup monitor; i t does not show any migrationtendency in the fuel, the fission yield is known with high accuracy, and mass spectrometry determination is not interfered with by other long-lived or stable fissionproducts present in the fuel in significant concentrations. Usually, analysis is performed by separation of the nuclide from the solution of the fuel sample by l iquidchromatography using a-hydroxy isobutyricacid as an eluant; the elution behaviorof the nuclide from the chromatographic column can be monitored by measuringthe y emission of i ts radioactive isotope 150Nd . The neodymium yield in the courseof the chemical separation procedure can be monitored by addition of a 142Ndspike to the initial solution. since this isotope is not produced in nuclear fission.Using nanogram amounts of the element in mass spect rometry, the concentrationof the i sotope in the fuel can be determinedwith an accuracy of 0.1 to 0.5% (Greenet al., 1989).

The only disadvantage in the application of th is nuclide as a burnup monitor isthe fact that non-destructive determination of the fuel burnup is not possible; therefore, to obtain a quick survey of the burnup state of whole fuel rods or fuel assemblies, 137CS has to be measu red. Depending on the question posed, either burnupdistribution in the fuel rod can be determined by gamma scanning or an averagevalue can be measured using an appropriate collimator designwith a comparativelylarge opening. When the numbe r of f issions which occurred in the final stage ofoperation is of interest, a short-lived radionuclide such as 140Ba/140La can be measured in the same manner. These measurements can be car ri ed ou t e it he r in a hotcell or in the spent fuel pool, provided that measures have been taken to install the

collimator - detector unit. When absolute burnup values are required, calibrationof the y det ec to r uni t can be done either by calculations taking into account thespecific parameters of the measuring sys tem or by subsequent destructive burnupanalysis of a sample of the relevant fuel rod.

For safeguards fuel characterization within the framework of the Non-Proliferat ion Treaty, non-destructive techniques for burnup determination are required,even for fuels with comparatively long decay t imes. Attempts have been made toselect pairs of y-emitting fission products whose rat io can be used not only toverify the fuel burnup, but also to determine the Pu : U ratio in the fuel as well asthe coo li ng time. It has been r epor ted that the r at io s of the radionuclides154Eu: J37C S and 134CS: 137CS fulfil these requirements (e. g . Berndt , 1988). Theactivity concentrations of t hese r ad ionucl ides in the fuel can be det ermined by asingle measurement using a collimated high-resolution y spectrometer. However,the correlations between the activity concentrations in the fuel and the requireddata have to be established separately for fuels with dif ferent ini tial 235U enrichment values; likewise, the neutron energy spectrum is of importance. Taking intoaccount these parameters, the resul ts obta ined are of satisfactory reliability forverification, in particular for fuels with long cool ing per iods (up to 20 years).

U si ng a combination of active and passive neutron interrogation, burnup ofPWR a nd B WR fuel assemblies can be determined with an accuracy of 1.2MWdlkg U and 2 MWd/kg U, respectively (Wiirz, 1991). Measurements of a greatnumber of fuel assemblies performed in the spent fuel poo ls of LWR plants haveshown that the accuracy of the resul ts is not affected by the particular data of thefuel assembly and that i tholds as long as 244Cm is the predominant neutron emitterin the fuel.

3.2.3 Chemical state and behavior of the fission products inthe fuel3.2.3.1 General aspectsAswas discussed in Section 3.2.1., immediately after the fission reaction the newlyformed fission products t ravel at very high speed through the UO l lattice. Since inthis phase the fragments are highly ionized atoms (average atomic charge number+20), a well-defined chemical state cannot be attributed to t hem. A ft er b ei ngstopped and reaching i ts rest posit ion, such an atom re-arranges its electron shellso that the final s ta te is compatible with the condi ti ons p reva il ing in the fuel.Identification of the chemical state of fission products in irradiated nuclear fuel isa complex task, because of the great number of influencing parameters and of thelarge differences in the concentration of fission products from that of the matrixsubstance. Due to these factors, experimental investigations are as difficult as aretheoretical calculations. The following sections, therefore, are far f rom giving acomplete picture; mainly those topics will be addressed which are of importancefor understanding the succeeding chapters.


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94 Radiochemistry during normal operation of the plant Radionuclides in the reactor core 95An understanding of the principles of fission product chemistry requires a t f irst

a description of the chemical system nuclear fuel. In a pure uranium fuel thi ssys tem in i ts original state, i. e. prior to irradiation, consists of sintercd pellets ofstoichiometric U02.lXIO containing only ncglibly small amounts of impurities. Thegap between the pellets and the Zircaloy cladding is fil led wit h hel ium at a startingpressure of about 2MPa (at ambient temperature), showing very low concentrations of gaseous impurities. The cladding i s s ea led gas -t ight ; as a consequence aclosed chemical system exists, except in the very seldom case of a cladding failure(sec Section 4.3.2.). Nonetheless, this closed system iscomplicated by the fact thatZircaloy is able to trap oxygen irreversibly by oxide formation, affecting therebythe oxygen partial pressure inside the system. During irradiation, the fuel composition is changed by the buildup of the fission products belonging to different chemical clements; thus, a gross composition of about (Uo.9sFPo.o6TUo.ol)Oz (where FPmeans fission product and TU transuranium element) can be assumed to ex is t ata burnup of 34MWdJkg HM, when one ignores the chemically inert noble gasesas well as the formation of metallic and oxide precipitates within the fuel g ra insand at the grain boundaries.

Fresh mixed-oxide fuels consist of a U02 matrix wit h a content of 3 to 4% fissileplutonium, with the plutonium concentrated in the master-mixture grains. Duringirradiation, fission product and transuranium elements are built up in the samemanner as in uranium fue l; the newly formed transuranium elements ar c homogeneously distributed in the fuel matrix with the exception of the preferential pluton ium breed ing i n the outermost zone of the pellet . The plutonium constituent hasli tt le effect on the chemical conditions in the fuel; therefore, both types of fuels arequite s imi lar wi th regard to the chemistry environment of the fission productsunder the operating conditions of light water reactors.

U02 (as well as (U,Pu)Oz for thermal reactors) form a c ryst al l at ti ce of thecalcium fluorite type showing a lat ti ce constant of 0.5468 nm (5.468 A) at stoichiometric composition. As can be seen from Fig. 3.12., the structure consists of a cubic

Figure3.12. lJ02 crystal lattice structure (scheme)o U atom; .0 atom; Vacancy

face-centered lattice (Fvtype lattice)of uranium atoms incorporating a s imple cubiclattice of oxygen atoms (P-type lattice), the edge length of which is half of that ofthe uranium lattice. 'The zero position of the oxygen lat t ice i sshi f ted byone quarterof the volume diagonal of the uranium latt ice. Every second cube of oxygen atomsi s occupied by an uranium atom in a volume-centering position, whereas the corresponding position i n t he neighbouring ce ll i s vacant. Such a vacancy can be occupied in a hypcrstoichiornetricU02+x by excess oxygen atoms up to the compositionU40 g (corresponding to U02.25); within this range, the lattice constant decreasessteadily with increasing hyperstoichiometry, whereas the lat t ice type remains unchanged.

After being stopped, the fission fragments reach their rest positions usually inside a U02 crystallite. There, their posi t ion might be at one of three differentlocations:

at a regular lattice position, either at one which has been made available by thefission of the relevant uranium (or plutonium) atom, or at the lattice vacancyposition;at an interstitial position o r a t a defec t position within the U02 lattice;in precipitations inside the crystall i tes or a t the crystal l it e grain boundaries.

At l ea st in the first two above-mentioned cases, the fission products are present inthe fuel in an atomic-dispersed state which docs not seem to allow the identificationof a definite chemical compound. Usually t hi s is the situation at low fuel burnupwith correspondingly low fission product concentrations in the fuel matrix. Thechemical state of such finely dispersed fission product atoms cannot be describedby the properties of chemical compounds; ra ther it has to be characterized byatomic properties, the most important of which, under the given circumstances,are the position in the crystal l at t ice and the valency state.

The number of available latt ice positions, i . c. th at o f fissioned uranium (orplutonium) atoms plus that of the origina l vacancies , i s sufficiently large to incorporate all the fission product atoms generated up to very high burnup values. Afirst indication of the fac t that the original vacancies are also occupied by fissionproduct atoms was obtained by early X-ray analytical observations showing thatthe lat ticeconstant decreases with increasing burnup (Schmitz et al., 1971; Benedictet a l. , 1972) . The lattice contraction of the CaF2- type structure was a lso reportedby a number of other investigators, as mentioned in the review paper of Kleykamp(1985). The reason for this lat ti ce contraction i s the enhanced mutual attractiondue to additional lattice atoms, comparable to the situation in hyperstoichiornetricU02+ x On the other hand, the proportion of the fission products incorporated atintersti t ial si tes in the U02 lattice can hardly be quant if i ed; i t can be assumed that,due to t he high densi ty of fission product fragments and fast neutrons in the material, numerous l at t ice defec ts are formed which are only incompletely annealedat the operational temperatures. In any case, an important precondition for theincorporation of foreign atoms into a crys tal l at ti ce , either at regular lat t ice positions or a t inte rs ti t ia l posi t ions, i s the compatibility of their radii with the lat t icedimensions.


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96 Radiochemistry during normal operation of the plant Radionuclidcs in the reactor core 97

Figure 3.I3. Typical properties of fission product elements in oxide fuels(Kleykarnp, 1985; with kind permission of Elsevier Science - NL. Sara Burgerhartstraat 25,1055KV Amsterdam, The Netherlands)

u+ v 2 FP+"V+ (2 V - V) e:where V means the valency state (oxidation state) of the fissile element (e. g. uranium) in the fuel and V the average valency state of the fission product mixturerepresented by the chemistry symbol FP. V can be calculated according to

V = 0.5' LYz' Y'I.where Yz s tands for the fiss ion yield and Yz for the valency state of the chemicalatom with the pro ton number z. I t has to be pointed out that the definition of thevalency state does not necessarily mean the presence of a particular kind or structure of a chemical species. In o ther words , the symbol Zr( IV) only means thatzirconium is present in a tetravalent state in the relevant medium, not necessarilyas a zirconium dioxide compound.The general validity of the electro-neutrality condition requires that in a closedsystem no net generation or net consumption of free electrons will take place. Inthe presence of ionizing radiation, free electrons are produced whose concentrationis compensated for by unusual atomic valency states; with respect to equilibriumconsiderations, therefore, this effectcan be ignored. As long as the FP in the aboveformula stands for the most easily oxidizable or reducible species in the system,the equat ion ( 2 V -V ) = 0 is valid. A single crystallite of the V02 material can

ED Volatile fission productso Metal lic precipitates- (alloys)Ceramic p recipi ta tes {oxides IOxides dissolved in the fuel

81 lIB IVB VB VIB VIIB Ne

18 n s l AI ArCo INi ICu IZn IGa IGe IAs JSe ! j r ~ ~ ~ :VllI

Osl Ir 1Pt IAulHg ITlIPb IBi IPolAt IRnRu IRh IPd IAg ICd IIn ISn ISb m ; . , ~ : m ~ : : ~ ~ ~ ~

A"'!o( TIA HALi I BeNalMgI l llA IVA VA VIA VlIA

For general considerations concerning the chemical state of the fission productsin the fuel mat ri x (in pa rt ic ul ar of the a tomic dispersed ones, but to a c er tai nextent also of those contained in precipitations), one may establish the followingpostulates:The chemical state is characterized by the existing valency state of the relevantfission product atom; chemical compounds with a well-defined chemical composition and an own habitus cannot be defined, with the exception of fuel inclusions.Fo r thermodynamic reasons, the existing valency state of the fission productatom must be compatible with the prevailing conditions. This is a consequenceof the postulate mentioned earlier that the progress of chemical reactions iscontrolled by the ambient conditions.Chemical compounds which are thermodynamically stable as isolated, purecompounds under the relevant conditions are no t necessarily formed in the fuelmatrix. There are problems due to kinetic interferences, to the very high U0 2excess, and to the unfavorable mass ratios of some of the fission product elements under consideration for the formation of compounds that are possible inprinciple.The impact of the high-energy radiation (fast fission fragments and neutrons,as wel l as a, and y radiation) may lead to the formation of unusual valencystates, bu t only for a very limi ted volume and over a very short t ime. Globa lchanges in the chemical principles are not assumed to be caused by radiationeffects.The princ ip le of e lect ro -neu tral ity has to be obeyed, i. e. the sum of all thepositive and negative charges contained in a chemical system (e. g. in a crystall ite) has to bezero.

One can expect, therefore, that the fission product atoms in the fuel are in a quasiequilibrium state with regard to their environment.With regard to their behav io r in the fuel matrix, the f iss ion products can bedivided very generally into four groups (see Fig. 3.13. , according to Kleykamp,1985):Fission product noble gases and other volatile fission products.Fission products dissolved as oxides in the fuel matrix; employing the atomisticaspects discussed above, the term "oxide" should be replaced by "atoms in theiradequate oxidized valency state".Fission products forming metaJlic precipitates.Fission products forming oxide precipitates.

There are transitions between the different groups which are partly caused by theoxygen potential of the fuel, while the formation and composition of precipitates,in particular, mainly depends on the concentrations of the individual fission product elements in the fueLAccording to the early considerations of Robinson (1958), nuclear fission maybe understood as a nuclear chemistry reaction


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98 Radiochemistry during normal operat ion of the plant Radionuclides in the reactor core 99Table 3.10. Most probable valency states of fission products in irradiated U02 fuel(Robinson, 1958; Copyright 1958by the American Nuclear Society, La Grange Park, Illinois)

be understood as a cl osed system in the sense of the above-mentioned electroneutrality requirement. Since the recoil length of the fission f ragments is on thesame order as the dimensions of the fuel gra ins (about 10urn), it can be a ssumedthat a considerable fraction of them will leave the original crystallite. But becauseof the isotropic distribution of the movement of the f ragments in the fuel, anequivalent amount of fission fragments generated in adjacent crystallites can beassumed to enter the given crystallite; therefore, this recoil effect is not expected tochange the distribution of the fission product elements in a single crystallite locatedin the volume of the pellet to an exten t that would violate the electro-neutralitycondition.Using simple electrochemical series, the most probable valency states of fissionproducts in U02 fuel presented in Table 3.10. can be def ined; the set ting of anelement in brackets means that i t is able to exis t in different valency sta tes. Withincreasing burnup, the e lectrochemical s ituat ion in the fuel is assumed to changedue to changing concentration ratios of the fission products. These variations arecompensated for by changes in the valency states of different fission products (e. g.iodine, ruthenium, molybdenum): thus, redox reactions at the D(IY) matrix atomsare not to be expected.These electrochemical considerations indicate that the prevailing redox potentialof the fuel is of high significance for the chemical state of the fission products.Since determination of this property is an important prerequisite, a short surveyof the methods for its determination shall be given. In non-irradiated uranium andmixed-oxide fuels, the redox potential is clearly controlled by the oxygen-to-metalratio of the mater ia l and can be determined either by direct measurement of themetallic constituent, by determination of the non-s toichiometric fract ion, or bydetermination of the redox property by equilibration with CO-C02 or H20 -H2gas mixtures. In irradiated fuels, the possibilities of determination arc more limited,not only becausc of the high radioactivity of the material, but mainly because of thecomplex multi-component chemical system of the fuel. Since no definite oxygen-tometal ratio can be defined, the oxygen potential of the mater ia l as the most essential propertymust bedirectly measured. This can be done by measuring the electromotoric force of a solid-state galvanic cell, consisting of an yttria-doped thoriacrucible which contains the fuel specimen to be measured; in most cases, Fe/FeO

Figure3.14. Relativepartial molar Gibbs free energy of oxygen of the fission product oxidesand OfU02(according10 Assmann and Stehle, 1982)

UO,.Ooo(-1

BaD

2000

1500 DC

OK1000

500

Cs20

o

oi '"800

1200

L\Go 2

kJ/mol0 2

is used as a reference system. This technique, which has to be car ri ed ou t at anelevated temperature (about 1000 K) has been employed by various investiga tors ;a miniature ce ll for the determination of radial oxygen profiles of irradiated fastbreeder oxide fuel has been descr ibed by Matzke et a1. (988). Another possiblemethod for determining the stoichiometry of irradiated fuel is measurement of thelattice parameters of the material by X-ray diffractornetry. However, in doing so ithas to be taken into account that the lattice parameters of the fuel oxides are notonly influenced by the stoichiometry of the fuel but, simultaneously, by its fissionproduct content. For this reason, analysis is usual ly performed in several steps,namely lattice parameter measurement of the fuel sample after irradia tion, thenequilibration treatment in a CO-C02 atmosphere to r each a ilG(02) level corresponding to the stoichiometriccomposition, and finally a second lattice parametermeasurement of the equilibrated specimen. The difference between the two measurements is directly proportional to the change in oxide composition. Finally, theoxygen potential of the irradia ted fuel can be derived from the Mo/Mo02 distribution in metal lic or oxide precipi ta tions which can be measured, e. g., by electronmicroprobe analysis.A first impression of the chemical state of the fission products in the mat rixU0 2 can be obtained by thermodynamic considerations, despite the limitationsdiscussed below. In Fig. 3.14. a simplified presentation of the partial free enthalpies

Elements(Se), (Te)(Br), (I)Kr , XeRb, Cs, (Ag)Sr, Ba, (Cd), (Pd)Y, Rare earths, (Rh), (As), (Sb), (In)Zr, (Mo). (Nb), (Te), (Ru), (Ge), (Sn)

Valency state-2-Io+1+2+3+4


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lOO Radiochemistry during normal operation of the plant Radionuclides in the reactor core IOJof formation of fission product oxides and of U0 2 is shown as a function oftemperature. Since the oxygen partial pressures of the most stable oxides of Ba,Ce, Zr, Sr, Pr, La, Nb and Yare significantly lower than that of stoichiometricU02, itcan beasswned that in the fuel the valency states Ba(II), Ce(lV) or Ce(lU),Zr(IV), Sr(Il), Pr(I1I), La(III), Nb(III) and Y(IlI) are stable, which is in generalagreement with the data given i

radiochem nucl power reactor

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