US20090122948A1 - Method of Determining at Least One Technological Uncertainty Factor for Nuclear Fuel Elements, and Corresponding Methods of Designing, Fabricating, and Inspecting Nuclear Fuel Elements - Google Patents

Method of Determining at Least One Technological Uncertainty Factor for Nuclear Fuel Elements, and Corresponding Methods of Designing, Fabricating, and Inspecting Nuclear Fuel Elements Download PDF

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US20090122948A1
US20090122948A1 US11/991,712 US99171206A US2009122948A1 US 20090122948 A1 US20090122948 A1 US 20090122948A1 US 99171206 A US99171206 A US 99171206A US 2009122948 A1 US2009122948 A1 US 2009122948A1
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fabrication
individual
variation
nuclear fuel
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Pascal Charmensat
Michel Pasquet
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Areva NP SAS
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    • G—PHYSICS
    • G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
    • G21C—NUCLEAR REACTORS
    • G21C17/00—Monitoring; Testing ; Maintaining
    • G21C17/06—Devices or arrangements for monitoring or testing fuel or fuel elements outside the reactor core, e.g. for burn-up, for contamination
    • G—PHYSICS
    • G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
    • G21C—NUCLEAR REACTORS
    • G21C3/00—Reactor fuel elements and their assemblies; Selection of substances for use as reactor fuel elements
    • G—PHYSICS
    • G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
    • G21D—NUCLEAR POWER PLANT
    • G21D3/00—Control of nuclear power plant
    • G—PHYSICS
    • G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
    • G21D—NUCLEAR POWER PLANT
    • G21D3/00—Control of nuclear power plant
    • G21D3/008—Man-machine interface, e.g. control room layout
    • Y—GENERAL TAGGING OF NEW TECHNOLOGICAL DEVELOPMENTS; GENERAL TAGGING OF CROSS-SECTIONAL TECHNOLOGIES SPANNING OVER SEVERAL SECTIONS OF THE IPC; TECHNICAL SUBJECTS COVERED BY FORMER USPC CROSS-REFERENCE ART COLLECTIONS [XRACs] AND DIGESTS
    • Y02—TECHNOLOGIES OR APPLICATIONS FOR MITIGATION OR ADAPTATION AGAINST CLIMATE CHANGE
    • Y02E—REDUCTION OF GREENHOUSE GAS [GHG] EMISSIONS, RELATED TO ENERGY GENERATION, TRANSMISSION OR DISTRIBUTION
    • Y02E30/00—Energy generation of nuclear origin
    • Y—GENERAL TAGGING OF NEW TECHNOLOGICAL DEVELOPMENTS; GENERAL TAGGING OF CROSS-SECTIONAL TECHNOLOGIES SPANNING OVER SEVERAL SECTIONS OF THE IPC; TECHNICAL SUBJECTS COVERED BY FORMER USPC CROSS-REFERENCE ART COLLECTIONS [XRACs] AND DIGESTS
    • Y02—TECHNOLOGIES OR APPLICATIONS FOR MITIGATION OR ADAPTATION AGAINST CLIMATE CHANGE
    • Y02E—REDUCTION OF GREENHOUSE GAS [GHG] EMISSIONS, RELATED TO ENERGY GENERATION, TRANSMISSION OR DISTRIBUTION
    • Y02E30/00—Energy generation of nuclear origin
    • Y02E30/30—Nuclear fission reactors

Definitions

  • the present invention relates to a method of determining technological uncertainty factors.
  • the invention applies by way of example, but not exclusively, to the nuclear fuel used in light water reactors, such as pressurized water reactors.
  • the nuclear fuel used in such reactors is conditioned in the form of pellets.
  • pellets are placed in cladding to form nuclear fuel rods that are grouped together within assemblies. Such assemblies are for loading in the cores of nuclear reactors.
  • the factor F Q E has been calculated fabrication parameter by fabrication parameter, taking into consideration the permitted value that maximizes linear power density, and even though certain compensation phenomena exist. Furthermore, the uncertainty in that calculation is such that it does not enable different variations to be compared.
  • That method makes it possible to quantify the influence of each fabrication parameter on the core state of a nuclear reactor. It was developed by analyzing variations in fabrication parameters in a fuel based on a mixture oxide fuel (MOX).
  • MOX mixture oxide fuel
  • the invention provides a method of determining at least one technological uncertainty factor for nuclear fuel elements as a function of variations in fabrication parameters of the elements about nominal values, the method comprising a step of making use, for at least one fabrication parameter, of collective variation in said parameter about the nominal value within a batch of fabricated elements.
  • the method may include one or more of the following characteristics taken in isolation or in any technically feasible combination:
  • T i and TL i respectively designate individual and collective variations in a fabrication parameter Figure, where ⁇ i and ⁇ i respectively designate the microscopic and macroscopic sensitivity coefficients for the fabrication parameter, and where ⁇ i designates the mean for the fabrication parameter F i ;
  • TL i designates the collective variation in a fabrication parameter F i
  • ⁇ i and ⁇ i designate respectively the microscopic and the macroscopic sensitivity coefficients in the fabrication parameter
  • ⁇ i designates the mean for the fabrication parameter F i ;
  • the invention also provides a method of designing a nuclear fuel element, the method including a step of using a method of determining at least one technological uncertainty factor as defined above.
  • the method comprises the steps of:
  • the invention also provides a method of fabricating nuclear fuel elements designed by a method as defined above.
  • the invention also provides a method of inspecting fabricated nuclear fuel elements, the method comprising the steps of:
  • FIG. 1 is a diagrammatic side view of a nuclear fuel assembly for a pressurized water reactor
  • FIG. 2 is a diagrammatic longitudinal section of a rod of the FIG. 1 assembly
  • FIG. 3 is a diagrammatic fragmentary view on a larger scale showing the shape of a pellet of the FIG. 2 rod;
  • FIG. 4 is a diagrammatic plan view showing the pattern used for calculating microscopic sensitivity coefficients for implementing a determination method of the invention.
  • FIG. 1 is a diagram showing a nuclear fuel assembly 1 for a pressurized water reactor. Water thus performs therein both a refrigeration function and a moderation function, i.e. it slows down neutrons produced by the nuclear fuel.
  • the assembly 1 extends vertically and in rectilinear manner along a longitudinal direction A.
  • the assembly 1 mainly comprises nuclear fuel rods 3 and a structure or skeleton 5 for supporting the rods 3 .
  • the support skeleton 5 comprises:
  • the nozzles 7 and 9 are connected to the longitudinal ends of the guide tubes 11 .
  • the rods 3 extend vertically between the nozzles 7 and 9 .
  • the rods 3 are disposed at the nodes of a substantially regular square-mesh array in which they are held by the grids 13 . Some of the nodes of the array are occupied by the guide tubes 11 , and possibly by an instrumentation tube.
  • each rod 3 comprises outer cladding 17 closed by a bottom plug 19 and a top plug 21 , and containing the nuclear fuel.
  • a helical retention spring 25 is placed in the cladding 17 between the top pellet 23 and the top plug 21 .
  • each pellet 23 is substantially cylindrical in shape having diameter D and height H.
  • the pellet 23 presents chamfers 26 between its end faces and its side face. These chamfers are of height h and of inside diameter dc.
  • a recess 27 in the form of a spherical cap is formed in each end face, substantially in the center thereof. It is of diameter d and of depth p.
  • the height/diameter ratio H/D may be arbitrary, for example about 1.6, but it could equally well be smaller, e.g. about 0.5.
  • Diametral clearance j e.g. lying in the range 100 micrometers ( ⁇ m) to 300 ⁇ m, is provided between the pellets 23 and the cladding 17 .
  • the fuel of the pellets 23 is constituted by uranium oxide (natural UO 2 ) that is enriched in isotope 235 .
  • the fuel may equally well be based on enriched retreated uranium, a mixture of uranium and plutonium oxides, and/or it may contain neutron poisons based on rare earth (gadolinium, erbium).
  • a pellet 23 is defined by a series of characteristics that are common to all types of pellet, i.e. whether based on enriched natural uranium (ENU), enriched reprocessed uranium (ERU), mixed oxide (MOX), or containing neutron poisons such as gadolinium oxide.
  • EEU enriched natural uranium
  • ERU enriched reprocessed uranium
  • MOX mixed oxide
  • neutron poisons such as gadolinium oxide.
  • the geometrical density Dg of the pellet is defined as being the ratio of its oxide mass in ceramic form to its theoretical oxide mass.
  • the gadolinium oxide content t which is defined as the ratio of Gd 3 O 3 to the total mass of UO 2 plus Gd 2 O 3 :
  • the isotope composition is generally the same as that of natural gadolinium.
  • the fuel manufacturer thus does not induce any variability in this composition CI and the isotope composition of gadolinium is therefore not taken into account in the example described below.
  • the isotopes 234 U, 235 U, and 236 U are characterized by their own concentrations also referred to as enrichment. These fabrication parameters are defined as follows:
  • e 4 mass ( 234 ⁇ U ) mass ( 234 ⁇ U + 235 ⁇ U + 236 ⁇ U + 238 ⁇ U )
  • e 5 mass ⁇ ( 235 ⁇ U ) mass ( 234 ⁇ U + 235 ⁇ U + 236 ⁇ U + 238 ⁇ U )
  • e 6 mass ⁇ ( 236 ⁇ U ) mass ( 234 ⁇ U + 235 ⁇ U + 236 ⁇ U + 238 ⁇ U )
  • the enrichment es is then defined by:
  • the isotope composition CI is thus defined by:
  • the nuclear fuel is modeled by using nominal values F i0 for the fabrication parameters F i .
  • the hot point is the rod of a core that has the greatest power. Below, it is sometimes referred to as the hot rod or located as being the power peak.
  • the first is the technological uncertainty factor for linear power density at the hot point F Q E .
  • P max maximum value of the linear power density at the hot point taking account of variations associated with fabrication
  • P nom linear power density calculated at the hot point with the nominal fabrication values F i0 .
  • the linear power density depends on pellet-to-pellet variations in the following fabrication parameters F i :
  • the second factor is the technological uncertainty factor of the hot channel F ⁇ H E1 .
  • ⁇ H max maximum value for enthalpy rise in the hot channel due to fabrication variations
  • ⁇ H nom calculated enthalpy rise in the hot channel calculated using nominal fabrication values F i0 .
  • the hot channel is the channel, i.e. the gap between adjacent fuel rods, in which enthalpy variation is the greatest. It is generally situated beside the hot rod.
  • variable P is distributed with a normal distribution
  • variable ⁇ P/P 0 is likewise distributed with a normal distribution that is characterized by an average of zero and a standard deviation of ⁇ . 95% of the values of ⁇ P/P 0 are less than 1.645 ⁇ with a confidence index of 95%, considering unilateral dispersion.
  • the technological uncertainty factors of the pertinent magnitude are determined from formulae for the sensitivity of the pertinent magnitude (maximum linear power density or maximum enthalpy rise) to the fabrication parameters F i of the nuclear fuel under consideration.
  • ⁇ P/P 0 is the relative variation in the magnitude of the hot point
  • ⁇ F i /F i0 designates relative variations in each fabrication parameter F i
  • ⁇ i designates the power sensitivity coefficients.
  • the value of the first technological uncertainty factor F Q E is such that 95% of the local variations in linear power are less than:
  • the value F ⁇ H E1 is such that 95% of the enthalpy rise variations are less than:
  • sensitivity formula Two types of sensitivity formula are calculated depending on whether it is considered that the deviation of the fabrication parameter applies to the hot rod, with the other rods of the assembly to which it belongs being unaffected (microscopic sensitivity formula), or to all of the assembly to which it belongs (macroscopic sensitivity formula):
  • the coefficients ⁇ i and ⁇ i can be calculated for different burnups of the fuel, as described below.
  • ⁇ i can be measured or approximated as described below.
  • an estimate with a confidence level of 95% for the maximum value of its effect in terms of overall power is given by a quadratic accumulation of the estimates, with 95% confidence level, for the maximum values of the mean and local power effects.
  • the fabrication tolerance T i is then defined in such a manner that 95% of the pellets lie in the range F i0 ⁇ T i about the nominal value F i0 . If F i is distributed with a normal distribution of mean ⁇ i and a standard deviation of ⁇ i , then:
  • This calculation is identical to that which consists in calculating an individual technological factor F Q E (i) for each fabrication parameter F i and then in accumulating them in order to obtain the overall F Q E .
  • F ⁇ ⁇ ⁇ H E ⁇ ⁇ 1 1 + ( 1 - F ⁇ ⁇ ⁇ H E ⁇ ⁇ 1 ⁇ micro ) 2 + ( 1 - F ⁇ ⁇ ⁇ H E ⁇ ⁇ 1 ⁇ macro ) 2
  • F ⁇ ⁇ ⁇ H E ⁇ ⁇ 1 ⁇ micro 1 + 1.645 1.96 ⁇ ⁇ i ⁇ ⁇ i 2 ⁇ TL i 2 ⁇ i 2
  • F ⁇ ⁇ ⁇ H E ⁇ ⁇ 1 ⁇ macro 1 + 1.645 1.96 ⁇ ⁇ i ⁇ ⁇ i 2 ⁇ TL i 2 ⁇ i 2
  • Sensitivity coefficients ⁇ i and ⁇ i are obtained by calculations performed on a pattern or “cluster” (in the mathematical sense) for the microscopic aspect with the help of the generalized conventional theory of perturbations and by calculations modeling the entire core for the macroscopic aspect, as described below.
  • the method is based on a first-order Taylor development of the powers P of the rods (or point powers) as a function of the fabrication parameters F i .
  • the underlying assumption is the possibility of linearizing around the nominal values F i0 and of considering that the impact on the final result (the point power P) of the variation of a plurality of input variables (the fabrication parameters F i ) is equal to the sum of the impacts of each of the variables taken separately.
  • ⁇ i ⁇ P P o ⁇ ⁇ micro ⁇ F i F io ⁇ ⁇ micro
  • Sensitivity calculations are performed in a pattern or “cluster” in an infinite medium with the help of a calculation code.
  • ⁇ ⁇ ⁇ ⁇ ⁇ U U + ⁇ ⁇
  • an assembly in an infinite medium corresponds to a single assembly with zero flow conditions.
  • the zero flow condition is equivalent to mirror symmetry: the assembly is repeated to infinity.
  • a rod situated in a corner of the assembly under consideration were it to be disturbed, would be surrounded by three disturbed rods in the virtual assemblies, even though the model would require those three rods to be remain nominal.
  • the model is redefined by moving the axes of symmetry further away, i.e. by using a pattern or “cluster”, while nevertheless retaining a reasonable dimension therefor.
  • the selected pattern thus comprises a complete assembly 1 surrounded by four assembly halves and four assembly quarters. This pattern is shown in FIG. 4 .
  • the central assembly 1 is the assembly containing the hot rod.
  • composition of the assemblies 1 of the pattern is representative of the fuel management under study, e.g. with a central MOX assembly surrounded by half and quarter ENU assemblies when studying a MOX fuel.
  • the method of calculating the sensitivity coefficients is based on separating neutron calculations into two portions: calculating neutron constants (macroscopic cross-sections) by transport calculation, and calculating the set of parameters for the core in diffusion. This split comes from approximating the transport equation as a diffusion equation.
  • the method is thus functionally split into two or three portions.
  • the fabrication parameters are varied without passing via the isotope concentrations. This produces directly the sensitivity coefficients relating the relative power variation of the hot point ⁇ P/P 0 to the relative variations in the fabrication parameters ⁇ F i /F i0 .
  • the isotope concentrations are varied, initially to obtain sensitivity coefficients relating the relative power variation of the ⁇ P/P 0 of the hot rod to relative variations ⁇ n j /n j0 in isotope concentration.
  • These are deduced from the fabrication parameter variations ⁇ F i /F i0 (except for the pellet diameter D) using a simple analytic formula given below, and thereafter the sensitivity coefficients obtained that relate ⁇ P/P 0 to each ⁇ F i /F i0 for F i ⁇ [F v , D g , CI, t, e 4 , e 5 , e 6 , e s ]
  • This functional split into three portions is used for the isotope composition CI and for the uranium support enrichment es for MOX.
  • the portion ( 1 b ) thus makes use of an analytic formulation given below.
  • the portions ( 1 a ) and ( 2 b ) calculate the macroscopic cross-sections by a transport code.
  • the portions ( 2 a ) to ( 3 b ) use the generalized conventional perturbation theory (GCPT), as explained below.
  • GCPT generalized conventional perturbation theory
  • the methodology that is developed is applied at the beginning of the lifetime of the fuel. It is repeated at operating intervals that are fixed in megawatt days per metric tonne (MWd/t).
  • This portion relates to all of the fabrication parameters with the exception of pellet diameter D.
  • the concentrations of plutonium or of gadolinium isotopes (or of the uranium support) in MOX or gadolinium-containing ENU fuels are obtained using the following analytical formula:
  • n j F v ⁇ M v ⁇ t ⁇ C j m j ⁇ R oxy ⁇ N ⁇ C th ⁇ ( T )
  • the mass per unit volume M v is equal to the product of the geometrical density D g multiplied by the theoretical density ⁇ th of the fuel.
  • C th (T) and R oxy are a priori functions of the operating point of the reactor under study and of the fabrication parameters F i for F i ⁇ [F v , D g , CI, t, es], but the variations therein can readily be ignored compared with the variations ⁇ F i /F i0 .
  • the relative isotope concentration variations ⁇ n j /n j0 are thus obtained from the fabrication parameter variations ⁇ F i /F i0 with F i ⁇ [F v , D g , CI, t, es].
  • n j F v ⁇ M v ⁇ C j m j ⁇ R oxy ⁇ N ⁇ C th ⁇ ( T )
  • relative isotope concentration variations ⁇ n j /n j0 are thus obtained from fabrication parameters ⁇ F i /F i0 with F i ⁇ [F v , D g , e 4 , e 5 , e 6 ].
  • the macroscopic cross-section variations are determined with the help of a transport code by calculating the impact solely:
  • the direct portion is determined by the product of the power cross-section variation multiplied by flux:
  • the spectral portion is conventionally obtained by the product of the important function r with perturbation multiplied by flux:
  • a and F are the absorption and diffusion production operators.
  • ⁇ ⁇ ⁇ P P 0 ⁇ ⁇ f , ⁇ ⁇ rod ⁇ ⁇ f , ⁇ ⁇ rod - ⁇ ⁇ f , ⁇ ⁇ pattern ⁇ ⁇ f , ⁇ ⁇ pattern - ⁇ ⁇ , ( ⁇ ⁇ ⁇ A - ⁇ ⁇ ⁇ F k ) ⁇ ⁇ ⁇ pattern
  • isotope concentrations n j for F i ⁇ [F v , D g , CI, t, e 4 , e 5 , e 6 , es].
  • the sensitivity coefficients ⁇ j are then calculated relating the relative variations of isotope concentrations to the relative variation in the power of the hot point:
  • the sensitivity coefficient ⁇ D is calculated that directly relates relative variation in pellet diameter to relative variation in the power of the hot point:
  • the coefficients may be determined for a plurality of degrees of burnup up to about 50,000 MWd/t or 60,000 MWd/t.
  • the method implemented is based on a first-order Taylor development of the point powers P of the rods, as a function of fabrication parameters F i .
  • ⁇ i ⁇ P P 0 ⁇ ⁇ macro ⁇ F i F i ⁇ ⁇ 0 ⁇ ⁇ macro
  • the particular core assembly to be perturbed is chosen as a function of the reference fuel management under consideration.
  • the reference management used is the management planning that was used as a basis for developing the fuel assembly under study. In such management, it is the position of the assembly 1 containing the hot rod that is studied.
  • the method of calculating the sensitivity coefficients ⁇ i is similar to that described for the microscopic aspect, with the GCPT calculation over a pattern being replaced by model calculations over the entire core using a calculation sequence that includes a transport code.
  • the portion ( 1 b ) relies on an analytic formulation. It is identical to that used for the microscopic aspect.
  • the portions ( 1 a ) and ( 2 b ) calculate macroscopic cross-sections by the transport code.
  • the portions ( 2 a ) and ( 3 b ) are performed by direct calculations modeling the entire core.
  • the coefficient relating to F v is taken as being equal to that calculated for D g . No calculation is thus performed for F v .
  • a first direct calculation modeling the entire core from the reference library gives the power of the reference hot point P 0 ref (sometimes written simply as P 0 ).
  • ⁇ ⁇ ⁇ P P 0 ⁇ j ⁇ ⁇ j ⁇ ⁇ ⁇ ⁇ n j n j ⁇ ⁇ 0 + ⁇ i ⁇ ⁇ i ⁇ ⁇ ⁇ ⁇ F i F i ⁇ ⁇ 0
  • the coefficients may be determined for a plurality of degrees of burnup, generally over one irradiation cycle.
  • F Q E 1 + 1.645 1.96 ⁇ ⁇ i ⁇ [ D , Fv , Dg , es , t ] ⁇ ⁇ ( ⁇ i 2 ⁇ T i 2 ⁇ i 2 + ⁇ i 2 ⁇ TL i 2 ⁇ i 2 ) ( 3 )
  • F ⁇ ⁇ ⁇ H E ⁇ ⁇ 1 1 + 1.645 1.96 ⁇ ⁇ i ⁇ [ D , Fv , Dg , es , t ] ⁇ ⁇ ( ⁇ i 2 ⁇ TL i 2 ⁇ i 2 + ⁇ i 2 ⁇ TL i 2 ⁇ i 2 ) ( 4 )
  • F Q E 1 + 1.645 1.96 ⁇ ⁇ i ⁇ [ D , Fv , Dg , CI ] ⁇ ⁇ ( ⁇ i 2 ⁇ T i 2 ⁇ i 2 + ⁇ i 2 ⁇ TL i 2 ⁇ i 2 ) ( 5 )
  • F ⁇ ⁇ ⁇ H E ⁇ ⁇ 1 1 + 1.645 1.96 ⁇ ⁇ i ⁇ [ D , Fv , Dg , CI ] ⁇ ⁇ ( ⁇ i 2 ⁇ TL i 2 ⁇ i 2 + ⁇ i 2 ⁇ TL i 2 ⁇ i 2 ) ( 6 )
  • This method of determining technological uncertainty factors can be used, for example:
  • the values for F Q E and F ⁇ H E1 are known and depend on the reactor in which the nuclear fuel that is to be fabricated is going to be loaded. As a function of the type of fuel to be fabricated, values are also known for the pertinent sensitivity coefficients ⁇ i and ⁇ i .
  • the nominal values F i0 are used as means ⁇ i for the parameters F i . It is then possible, e.g. iteratively, to determine the individual and collective fabrication tolerances T i and TL i for the pertinent fabrication parameters F i .
  • T i corresponds to the individual tolerance that needs to be complied with by each pellet 23 , i.e. the real values as measured for F i must lie within the range F i0 ⁇ T i for 95% of the pellets.
  • TL i corresponds to the tolerance that the mean of the measured real values in a batch coming from the same fabrication run must comply with, i.e. for 95% of the batches, the mean of the real values of F i must lie within the range F i0 ⁇ TL i .
  • T i is thus referred to as the individual or microscopic tolerance, and TL i is referred to as the collective or macroscopic tolerance.
  • the manufacturer In order to increase the freedom of the pellet manufacturer, it is also possible to supply the manufacturer with means for calculating technological uncertainty factors locally on the basis of the individual and collective tolerances. Under such circumstances, the manufacturer is provided with no more than the limit values to be complied with for the technological uncertainty factors, and it is the manufacturer who determines the corresponding values for T i and TL i .
  • the pellet manufacturer gives the real measured values for the parameters F i of the corresponding batch of pellets. These measurements can be performed only on a statistically representative sample of the fabricated batch.
  • the fabricated pellets can be used for obtaining technological uncertainty factor values that remain below the limit values for the reactor under consideration.
  • the method of determining the above-mentioned uncertainty factors is found to be more accurate than that used in the state of the art.
  • the determining factor turns out to be taking account of macroscopic variation.
  • Pellet manufacturers thus gain in flexibility, thereby making it possible to reduce the cost of producing nuclear fuels, for given safety levels.
  • the invention can be implemented using calculations other than those in the example described above. Nevertheless, account should always be taken of a macroscopic aspect for at least one fabrication parameter, as in the example described above for F ⁇ H E1 or of a microscopic aspect and a macrocosmic aspect in combination, as in the above-described example for F Q E .
  • the invention can also be applied to fabricating fuels for other light water reactors, e.g. boiling water reactors.
  • the invention can be applied to nuclear fuels for other types of reactor, such as high temperature reactors (HTR) for example.
  • HTR high temperature reactors
  • the nuclear fuel elements taken into consideration are therefore not necessarily pellets, but they could be spheres.

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US11/991,712 2005-09-09 2006-08-21 Method of Determining at Least One Technological Uncertainty Factor for Nuclear Fuel Elements, and Corresponding Methods of Designing, Fabricating, and Inspecting Nuclear Fuel Elements Abandoned US20090122948A1 (en)

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FR0509228A FR2890767B1 (fr) 2005-09-09 2005-09-09 Procede de determination d'au moins un facteur d'incertitude technologique d'elements de combustible nucleaire, procede de conception, procede de fabrication et procede de controle d'elements de combustible nucleaire correspondants.
FR0509228 2005-09-09
PCT/FR2006/001965 WO2007028870A1 (fr) 2005-09-09 2006-08-21 Procede de determination d'au moins un facteur d'incertitude technologique d'elements de combustible nucleaire, procede de conception, procede de fabrication et procede de controle d'elements de combustible nucleaire correspondants

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US20110176650A1 (en) * 2008-09-30 2011-07-21 Areva Np Nuclear reactor green and sintered fuel pellets, corresponding fuel rod and fuel assembly
US20200025956A1 (en) * 2016-08-14 2020-01-23 Nuclear Research Center Negev Neutron detector and method for its preparation

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RU2647486C1 (ru) * 2017-03-22 2018-03-16 Федеральное государственное унитарное предприятие "Научно-исследовательский институт Научно-производственное объединение "ЛУЧ" (ФГУП "НИИ НПО "ЛУЧ") Способ испытания высокотемпературных тепловыделяющих элементов
KR102102975B1 (ko) * 2018-01-24 2020-05-29 한전원자력연료 주식회사 우수한 충격 저항성을 갖는 핵연료 소결체
RU2682238C1 (ru) * 2018-05-07 2019-03-18 Федеральное государственное унитарное предприятие "Научно-исследовательский институт Научно-производственное объединение "ЛУЧ" (ФГУП "НИИ НПО "ЛУЧ") Способ реакторных испытаний высокотемпературных вентилируемых тепловыделяющих элементов
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JP5379476B2 (ja) 2013-12-25
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FR2890767A1 (fr) 2007-03-16
ZA200802030B (en) 2009-08-26
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US8687760B2 (en) 2014-04-01

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