[en] In this text we give the dosimetry principle on irradiated materials such baffle screw, pressure vessel and control element cans. This measure, made by gammametry, is based on the steel activation and comparison with calculated measures by Actige code. 4 figs., 6 refs

[en] The activation detector technique is very widely used in dosimetry of materials irradiations. The ⁹³Nb(n,n')sup(93m)Nb reaction is the most suitable. In this paper, the causes of this selection are summarized: high melting point and good compatibility of the metal with usual environments, long half-life and mainly low threshold cross-section comparable with radiation damage cross-sections. The selected solutions for its practical utilization are: irradiation of rolled metallic high purity niobium, (bare or under various jackets); measurement of the X ray Ksub(α) Nb activity with a Si-Li detector (cooled preamp.) and analysis of the X ray spectrum

[en] In the framework of a BWR fuel study, an experimental program consisting of the irradiation of poisoned fuel pins was carried out in the reflector of a swimming-pool type reactor (SILOE). Different types of fuels (uranium oxide 2.5% enriched or uranium-plutonium oxide, with 1 and 2% of gadolinium oxide), placed in two devices (with NaK or boiling water), were irradiated under specified conditions. For some irradiations, a continuous measurement of total power vs. time was performed, until the complete burn-up of the poison. For the others, the aim was to produce fuel samples of different burn-up rates of the poison, for further examinations. In the paper, dosimetry of these irradiations is described, which is based on the continuous signals observed from ''collections'' (self powered neutron detectors), located outside the device

[en] An experimental program has been undertaken to confirm divers fast neutron spectrum calculations. The verification of calculational methods consists to compare measurements in a highly heterogeneous media to two-dimensional calculations effectuated on approach geometries. The measurements have been carried out in two testing facilities located in the centre of the core and near the reflector of the MELUSINE pool-type reactor. Counting techniques used for the determination of the reaction rate by beta-gamma activity measurements and activation foils are described. The following reactions have been selected: ¹⁹⁷Au(n,γ) ¹⁹⁸Au, ¹¹⁵In(n,n') ¹¹⁵Insub(m), ⁴⁷Ti(n,p) ⁴⁷Sc, ⁵⁸Ni(n,p) ⁵⁸Fe(n,p) ⁵⁴Mn, Ti(n,x) ⁴⁶Sc, ⁵⁶Fe(n,p) ⁵⁶Mn, ⁶³Cu(n,α) ⁶⁰Co, ²⁷Al(n,α)²⁴Na, ⁹²Nb(n,2n) ⁹²Nbsub(m), ⁵⁸Ni(n,2n) ⁵⁷Ni. A spectrum form has been elaborated from such activities by using the unfolding code SAND-II. On other part, spectrum determinations are performed with transport computer codes ANISN and DOT-III. The spectrum obtained from Unfolding and Transport Codes have been compared. Good agreement between the measured (foil activation) and calculated (SAND-II) activities were found with all detectors. Reasonable agreement between transport (DOT-III) and iterative (SAND-II) solutions are observed in the two test cases. This analysis permits to retain a spectrum determination procedure in the actual case of a composite testing facility

[en] Experimental neutron characterization (or dosimetry) of surveillance capsule irradiation is a main step in the surveillance program of irradiation effects on pressurized water reactor (PWR) vessels. Recently in France, results were obtained with 50 capsules removed from 29 900-MW (electric) reactors and analyzed using a standard method implementing, among others, the TRIPOLI Code to determine the perturbations caused by structural parts (stiffeners) and the neutron spectra in the capsules with their variance-covariance matrices. The method is briefly explained, emphasizing some choices that were made concerning the uncertainty evaluations. Then the results of neutron fluxes and fluences are discussed, and the mean measured value for each reactor series is compared with the computed one. In France, PWRs in operation are highly standardized. There are only 2 models of 900-MW reactors, called CPO and CPY. The 50 capsules examined were taken from these 2 series of reactors and can be arranged into 6 classes only. Some classes have as many as 18 elements, as, for example the U capsule class, irradiated at a 20 degree position in the CPY series. In each class, the observed standard deviation and estimated uncertainty are discussed. Because some classes have a strong statistical weight, the averaged results are highly significant. These reactors are constructed and operated very close to a standard model in power, flux, and dosimetry measurements. Therefore, results are weakly scattered. They confirm the calculations and indicate that the uncertainty the authors find for capsule fluence is reasonable. The best is only 6.1% (2σ). Knowledge of the vessel fluence (and of its uncertainty, which is being evaluated) is an essential factor for plant life anticipation

[en] In order to use the ⁹³Nb(n, n')⁹³Nb reaction for the determination of fast neutron fluences in nuclear reactors the Xsub(k)-ray decay of three different samples of ⁹³Nbsub(m) was followed for several years by long runs of continous observations. The 16.6 keV X-ray was chosen to determine the half-time of ⁹³Nbsub(m). The ⁹³Nbsub(m) sample was prepared by irradiation of a niobium sample of high purity (less than 5 ppm Ta) in a SILOE reactor. The effect of the two probable impurities (⁹⁴Nb, T=2x10⁴ years; ¹⁸²Ta, T=115 days) was minimized. The detection efficiency was verified by alternate ²⁴¹Am source countings. The measurements resulted in a weighted average value of T=16.4+-0.4 years, which is greater than any other published one. (T.F.)

[en] Starting with the design of the CAP (Prototype Advanced NSSS), a programme for pressure vessel monitoring has been prepared, including dosimetry. The dosimetry programme encompasses activation dosimeters (Cu, Nb, Co) and fission dosimeters (²³⁷Np, ²³⁸U) installed either inside the pressure vessel with the monitoring test-samples, or in a counting tube outside the pressure vessel. In the first place, a description of the method for neutronic calculation is given; such calculations use the codes ANISN and MERCURE 4 allowing assessment of the neutron spectrum seen by the detectors and the related reaction coefficient. This is followed by a description of the instrumentation. The initial dosimetry results available after the initial operating cycles concur with calculations

[en] Activation detectors technique is very widely used in dosimetry of testing materials irradiations. In this scope, one of them, the ⁹³Nb(n,n')sup(93m)Nb reaction is the most suitable. In this paper, the motives of this choice are summarized: high melting point and good compatibility of the metal with usual environments, long half-life and mainly low threshold cross-section, which is comparable with radiation damage cross-sections. In order to use it in a common and practical way, the selected solutions are: the material is rolled metallic high purity niobium, wich can be irradiated bare or under various jackets; the X ray Ksub(α) Nb activity is measured with a Si-Li detector by the analyse of the complex X rays spectrum. Corrections are weak

[en] An experimental program has been undertaken to confirm fast neutron spectrum calculations. The verification of calculational methods consists of comparing the measurements in a highly heterogenous media to two dimensional calculations effectuated on approach geometries. The measurements have been carried out in two testing rigs located in the center of the core and near the reflector of the Melusine Pool-type reactor. Counting techniques used for the determination of the reaction rate by beta-gamma activity measurements and activation foils are described. The following reactions have been selected: ¹⁹⁷Au(n,γ)¹⁹⁸Au, ¹¹⁵In(n,n')/sup 115m/In, ⁴⁷Ti(n,p)⁴⁷Sc, ⁵⁸Ni(n,p)⁵⁸Co, ⁵⁴Fe(n,p)⁵⁴Mn, ⁴⁶Ti(n,x)⁴⁶Sc, ⁵⁶Fe(n,p)⁵⁶Mn, ⁶³Cu(n,α)⁶⁰Co, ²⁷Al(n,α)²⁴Na, ⁹²Nb(n,2n)/sup 92m/Nb, ⁵⁸Ni(n,2n)⁵⁷Ni. A spectrum form has been elaborated from such activities by using the unfolding code SAND-II. On the other part, spectrum determinations are performed with Transport codes ANISN and DOT-III. The spectrum obtained from Unfolding and Transport Codes has been compared. Good agreement between the measured (foil activation) and calculated (SAND-II) activities was found with all detectors. Reasonable (SAND-II) solutions are observed in the two test cases

[en] The surveillance dosimetry program of Electricite de France's reactors pressure vessel built by Framatome consists of neutronic computation by means of ANISN-DOT procedure on the one hand and of two experimental parts on the other hand. For the first one, a light instrumentation with activation wires was put outside the pressure vessel (PV) of one power plant, during the first 18 months cycle at nominal power. This instrumentation is described: it gave the possibility to do measurements along two vertical lines and an horizontal diameter under the vessel. Experimental and computed results are compared respectively for thermal, epithermal and fast neutrons. For the second experimental part, the first surveillance capsules have been extracted from six power plants and the dosimeters have been measured. The difficulties encountered during some steps of the process are described. The reproducibility of the different results is excellent. Their consistency and agreement with the calculations are discussed