Method to assess the radionuclide inventory of irradiated graphite from UNGG reactors (Uranium Naturel Graphite Gaz)

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1 Method to assess the radionuclide inventory of irradiated graphite from UNGG reactors (Uranium Naturel Graphite Gaz) B. PONCET/EDF-CIDEN SCIENTIFIC CONFERENCE Uranium Graphite Reactors Decommissioning Lituania 14th /16th of July, 2014

2 A radionuclide inventory: what is it and what for? The radionuclide inventory of a nuclear waste is a quantitative description as precise as possible of all the radionuclides it contains It is a legal commitment: waste producers DO HAVE TOdeclare the radionuclide inventory to ANDRA (public body in charge of long term radioactive waste management in France) It is needed to anticipate an «a priori» classificationin order to develop the best suited management strategies(repository sanitary impact) It makes possible to optimize the choice of the best decommissioning options, deconstruction tools and radiological shield for ensuring staff safety It enables the identification of radionuclidesthat could lead to significant discharges in the environment: in order to adapt the confinement and purification systems in order to limit the impact on the environment

3 Conventional activation calculations do not fit at all for irradiated graphite Direct calculation activation of an assumed impurity content can not be used Some radionuclide precursors were not always measured at the time of the reactors building (ex: Cl) Radionuclide (and precursors) release happened during reactor operation (radiolytic corrosion of graphite - thermal release) (Clayton and all, 1997, «Behavior of chlorine in nuclear graphite», Nirex technical report T/REP/20121/P/04) For some precursors such as 14 N( 14 C precursor) it is strictly impossible to know their content in operating graphite: It is an inherent impurity of nuclear graphite manufacturing coming from the pitch tar and/or entrapped air Could have been introduced onto graphite during operating period (leaks) or maintenance period Part of the nitrogen sorbed onto the graphite surface was released during operation (radiolytic corrosion)

4 The example of nitrogen for carbon 14 Huge differences in nitrogen content in published data: 40 to 150 mg/kg of adsorbed nitrogen on non irradiated nuclear graphite (R.Takahashi and all, 1999, «Investigation of morphology and impurities of nuclear graphite», IAEA technical meeting Manchester) 10 to 50 mg/kg of nitrogen taken into account for some activation calculations (where does it come from?) (IF.White and all, 1984, «Assessment management modes for graphite from decommissioning, European Projet Technical report EUR 9232 EN -BJ.Marsden and all, 2002, «The chemical forms of C-14 within graphite», Nirex technical report SA/RJCB/RD /R01) 4 mg/kg of nitrogen measured on some samples (IAEA, 1998, «Radiological characterization of shut down reactors for decommissioning purposes», technical report series n 389) An assumed nitrogen content of 10 mg/kg leads to a carbon14 inventory greater than the measurements of at least one to two order of magnitude in the case of UNGG i-graphite 4 -

5 The identification calculation-measurement method Cross section library Cross section library Radioactive chain decay Radiochemical analysis (x, y, z) Tripoli Monte-carlo transport code 3 D map of neutron flux Darwin / Pepin Activation code Radionuclide inventory (x, y, z) Pile geometry Chemical composition (without impurities) Operating conditions Chemical composition including impurities Operating history Shutdown delay Impurity level adjustment C/M =1 1) Creation of a 3D map of the flux density (315 intervals for energy) 2) Calculation of the inventory (reactor operational history) from impurity levels adjusted (least square method) from the result of activation calculation with the available measurements of the corresponding radionuclide 3) An upper-threshold value of each radionuclide activity is determined by using the standard specific deviation (depending on the number of available measurement points) and the confidence interval (2,5% risk of under-assessing the result) with respect to cobalt 60 (reference γemitter used for controls on final packages by Andra) 5

6 The identification calculation-measurement method 1) For each radionuclide measurement reverse activation calculation (3D flux map) of the precursor (impurity) content 2) Through a set of measurements (around 30 for each radionuclide) a mean value for each impurity content in graphite is determined with its confidence interval (depending on the number of measurements) 3) A radionuclide inventory is then calculated (direct activation calculation) from this mean value for each radionuclide: it is reported with respect to the cobalt 60 ratio (RN/ 60 Co) 4) The radionuclide inventory considered for disposal option is the upper value of the confidence interval of the RN/ 60 Co ratio multiplied by the upper value of the confidence interval of 60 Co (the waste producer must ensure the upper threshold representativeness of its radionuclide inventory) Advantages of the method No assumed impurity content Can take into account radionuclide (or precursor) release phenomenon Based on a set of measurements (representativeness) not on a single measurement

7 The use of a set of measurements is the only solution Some radionuclide measurements and especially those at the trace level such as chlorine 36 show a large dispersion... Material heterogeneity Size (representativeness) of the sample used for radiochemical analysis Chlorine 36 activity (Bq/g) Chlorine 36 measurement in Bugey1 pile according to the sampling point position (altitude) ,0 12,0 14,0 16,0 18,0 20,0 22,0 Sampling point (m) 7

8 Some results Piles Bq 3 H 14 C 36 Cl 137 Cs Chinon A3 (2,530 t) Saint-Laurent A1 (2,570 t) Saint-Laurent A2 (2,440 t) Bugey 1 (2,060 t) Radionuclide inventory for some EDF reactors determined according to the identification calculationmeasurement method The higher the neutron density (Bugey1), the higher the radionuclide content... Average calculated values (in Bq/g) of chlorine36 for the graphite of some fuel channels in the Bugey 1 graphite pile (squares represent measurements -x-axis: height of the pile) A radionuclide geographical distribution linked to the neutron flux

9 Some results Release phenomenon are taken into account: example, the chlorine 35 explanatory content is much lower (tens of mg/10 3 kg) than some measurements available in non irradiated graphite (mg/kg) Fission product content can be explained by uranium impurity content (not difference between reactors where fuel failure happened and the others) UNGG reactor Graphite mass pile (t) Specific power (W th / g U) Chinon A3 2, Coke used for graphite manufacturing Lockport L (74 %) Lockport M (26 %) Mean 36 Cl mass activity (Bq/g) 2.1 Saint-Laurent A1 2, Lockport M 1.7 Saint-Laurent A2 2, Lima 18.2 Bugey 1 2, Lima 25.5 Without any arbitrary assumption, the influence of the nature of the coke used for graphite manufacturing is evidenced

10 Conclusion A direct activation calculation is not possible (impurity content not well determined or actually unknown - see nitrogen example) Calculations based on a set of measurements is needed (wide variability of radiochemical measurements for radionuclide at the trace level) The identification calculation-measurement method allows A radionuclide inventory linked with the main phenomenon at the origin of radioactivity (neutron flux). while taking into account secondary phenomenon (release during operation) Representativeness depending on the number of available measurements (confidence interval) Highlights some forgotten parameters such as the origin of the graphite (coke used for manufacturing) 10 -

11 Any questions?