Comparison of calculated and measured radionuclide inventory of a Zircaloy-4 cladding tube plenum section

MRS Advances ◽  
2018 ◽  
Vol 3 (19) ◽  
pp. 1031-1037 ◽  
Author(s):  
Michel Herm ◽  
Ron Dagan ◽  
Ernesto González-Robles ◽  
Nikolaus Müller ◽  
Volker Metz
Keyword(s):  
Chemical Interaction ◽  
Fission Products ◽  
Minor Actinides ◽  
Final Disposal ◽  
Fuel Pellets ◽  
Activation Products ◽  
Radionuclide Inventory ◽  
Interim Storage ◽  
Modelling Methods ◽  
Good Agreement

ABSTRACTCladding tubes of water-cooled nuclear reactors are usually made of Zircaloy and are an important retaining element for radionuclides present in the fuel both during predisposal activities such as reloading of fuel assemblies from interim storage casks to final disposal casks and during final disposal in the case of canister breaching. However, cladding integrity is affected by various processes during reactor operation and beyond, e.g. fuel cladding chemical interaction and fission product precipitation onto the inner cladding surface. Using experimental and modelling methods, the radionuclide inventory of an irradiated Zircaloy-4 plenum section is analyzed. Quantities of 235/238U, 237Np, 238/239/240/241/242Pu, 241/243Am, 243/244Cm besides 14C, 55Fe, 125Sb, 154Eu, and 134/137Cs were (radio-)chemically determined in digested Zircaloy-4 subsamples. Measured inventories of activation products in the Zr-alloy are in good agreement with calculated values. However, amounts of actinides and fission products exceed the calculated inventory by factor ∼57 (minor actinides and non-volatile fission products) and ∼114 (137Cs). Excess of minor actinides and part of enhanced Cs inventory originate from fuel residues deposited on the inner cladding surface during fuel rod fabrication, whereas vast amount of cesium is volatilized from subjacent fuel pellets and transported to the plenum.

Zirconia Inert Matrix Fuel for Plutonium and Minor Actinides Management in Reactors and as an Ultimate Waste Form

2008 ◽  
Vol 1104 ◽  
Author(s):  
Claude Degueldre ◽  
Wolfgang Wiesenack
Keyword(s):  
Near Field ◽  
Fission Products ◽  
Minor Actinides ◽  
Spent Fuel ◽  
Geological Disposal ◽  
Inert Matrix Fuel ◽  
Inert Matrix ◽  
Fuel Pellets ◽  
Reducing Conditions ◽  
Deep Geological Disposal

AbstractA plutonia stabilised zirconia doped with yttria and erbia has been selected as inert matrix fuel (IMF) at PSI. The results of experimental irradiation tests on yttria-stabilised zirconia doped with plutonia and erbia pellets in the Halden research reactor as well as a study of zirconia solubility are presented. Zirconia must be stabilised by yttria to form a solid solution such as MAz(Y,Er)yPuxZr1-yO2-ζ where minor actinides (MA) oxides are also soluble. (Er,Y,Pu,Zr)O2-ζ (with Pu containing 5% Am) was successfully prepared at PSI and irradiated in the Halden reactor. Emphasis is given on the zirconia-IMF properties under in-pile irradiation, on the fuel material centre temperatures and on the fission gas release. The retention of fission products in zirconia may be stronger at similar temperature, compared to UO2. The outstanding behaviour of plutonia-zirconia inert matrix fuel is compared to the classical (U,Pu)O2 fuels. The properties of the spent fuel pellets are presented focusing on the once through strategy. For this strategy, low solubility of the inert matrix is required for geological disposal. This parameter was studied in detail for a range of solutions corresponding to groundwater under near field conditions. Under these conditions the IMF solubility is about 109 times smaller than glass, several orders of magnitude lower than UO2 in oxidising conditions (Yucca Mountain) and comparable in reducing conditions, which makes the zirconia material very attractive for deep geological disposal. The behaviour of plutonia-zirconia inert matrix fuel is discussed within a burn and bury strategy.

scholarly journals Radioactive Fission Waste from Molybdenum-99 Production and Proliferation Risks

2021 ◽  
Vol 927 (1) ◽  
pp. 012041
Author(s):  
Aisyah ◽  
Pungky Ayu Artiani ◽  
Jaka Rachmadetin
Keyword(s):  
Nuclear Proliferation ◽  
Fission Products ◽  
High Level Waste ◽  
Blending Method ◽  
Activation Products ◽  
Enriched Uranium ◽  
High Level ◽  
Solid Form ◽  
Interim Storage

Abstract Molybdenum-99 (99Mo) is a parent radioisotope of Technetium-99m (99mTc) widely used in nuclear diagnostics. The production of this radioisotope by PT. INUKI generated radioactive fission waste (RFW) that theoretically contains239Pu and235U, posing a nuclear proliferation risk. This paper discusses the determination of radionuclides inventory in the RFW and the proposed strategy for its management. The radionuclides inventory in the RFW was calculated using ORIGEN 2.1 code. The input parameters were obtained from one batch of 99Mo production using high enriched uranium in PT. INUKI. The result showed that the RFW contained activation products, actinides, and fission products, including239Pu and235U. This result was then used for consideration of the management of the RFW. The concentration of 235U was reduced by a down-blending method. The proposed strategy to further manage the down-blended RFW was converting it to U3O8 solid form, placed in a canister, and eventually stored in the interim storage for high-level waste located in The Radioactive Waste Technology Center.

Separation of Trivalent Minor Actinides from Fission Products Using Single R-BTP Column Extraction Chromatography

2011 ◽  
Vol 48 (6) ◽  
pp. 855-858 ◽  
Author(s):  
Tatsuro MATSUMURA ◽  
Kazumi MATSUMURA ◽  
Yasuji MORITA ◽  
Yoshikazu KOMA ◽  
Yuichi SANO ◽  
...  
Keyword(s):  
Extraction Chromatography ◽  
Fission Products ◽  
Minor Actinides ◽  
Column Extraction

scholarly journals Control Experiences Regarding Clearable Materials from Nuclear Power Plants and Nuclear Installations: Scaling Factors Determination and Measurements’ Acceptance Criteria Definition

Environments ◽  
2019 ◽  
Vol 6 (11) ◽  
pp. 120
Author(s):  
Luca Albertone ◽  
Massimo Altavilla ◽  
Manuela Marga ◽  
Laura Porzio ◽  
Giuseppe Tozzi ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Fission Products ◽  
Scaling Factors ◽  
Activation Products ◽  
Long Time ◽  
Nuclear Installation ◽  
Enriched Uranium

Arpa Piemonte has been carrying out, for a long time, controls on clearable materials from nuclear power plants to verify compliance with clearance levels set by ISIN (Ispettorato Nazionale per la Sicurezza Nucleare e la Radioprotezione - National Inspectorate for Nuclear Safety and Radiation Protection) in the technical prescriptions attached to the Ministerial Decree decommissioning authorization or into category A source authorization (higher level of associated risk, according to the categorization defined in the Italian Legislative Decree No. 230/95). After the experience undertaken at the “FN” (Fabbricazioni Nucleari) Bosco Marengo nuclear installation, some controls have been conducted at the Trino nuclear power plant “E. Fermi,” “LivaNova” nuclear installation based in Saluggia, and “EUREX” (Enriched Uranium Extraction) nuclear installation, also based in Saluggia, according to modalities that envisage, as a final control, the determination of γ-emitting radionuclides through in situ gamma spectrometry measurements. Clearance levels’ compliance verification should be performed for all radionuclides potentially present, including those that are not easily measurable (DTM, Difficult To Measure). It is therefore necessary to carry out upstream, based on a representative number of samples, those radionuclides’ determination in order to estimate scaling factors (SF), defined through the logarithmic average of the ratios between the i-th DTM radionuclide concentration and the related key nuclide. Specific radiochemistry is used for defining DTMs’ concentrations, such as Fe-55, Ni-59, Ni-63, Sr-90, Pu-238, and Pu-239/Pu-240. As a key nuclide, Co-60 was chosen for the activation products (Fe-55, Ni-59, Ni-63) and Cs-137 for fission products (Sr-90) and plutonium (Pu- 238, Pu-239/Pu-240, and Pu-241). The presence of very low radioactivity concentrations, often below the detection limits, can make it difficult to determine the related scaling factors. In this work, the results obtained and measurements’ acceptability criteria are presented, defined with ISIN, that can be used for confirming or excluding a radionuclide presence in the process of verifying clearance levels’ compliance. They are also exposed to evaluations regarding samples’ representativeness chosen for scaling factors’ assessment.

Establishing a Store Baseline During Interim Storage of Waste Packages and a Review of Potential Technologies for Baselining

2013 ◽  
Author(s):  
Jennifer McTeer ◽  
Jenny Morris ◽  
Stephen Wickham ◽  
Gary Bolton ◽  
James McKinney ◽  
...  
Keyword(s):  
Waste Management ◽  
Storage System ◽  
Storage Facility ◽  
Original Condition ◽  
Final Disposal ◽  
Waste Storage ◽  
Secure Environment ◽  
Definition Of ◽  
Inspection Data ◽  
Interim Storage

Interim storage is an essential component of the waste management lifecycle, providing a safe, secure environment for waste packages awaiting final disposal. In order to be able to monitor and detect change or degradation of the waste packages, storage building or equipment, it is necessary to know the original condition of these components (the “waste-storage system”). This paper presents an approach to establishing the baseline for a waste-storage system, and provides guidance on the selection and implementation of potential baselining technologies. The approach is made up of two sections; assessment of baselining needs and definition of baselining approach. During the assessment of baselining needs a review of available monitoring data and store/package records should be undertaken (if the store is operational). Evolutionary processes (affecting safety functions), and their corresponding indicators, that can be measured to provide a baseline for the waste-storage system should then be identified in order for the most suitable indicators to be selected for baselining. In defining the approach, identification of opportunities to collect data and constraints is undertaken before selecting the techniques for baselining and developing a baselining plan. Baselining data may be used to establish that the state of the packages is consistent with the waste acceptance criteria for the storage facility and to support the interpretation of monitoring and inspection data collected during store operations. Opportunities and constraints are identified for different store and package types. Technologies that could potentially be used to measure baseline indicators are also reviewed.

scholarly journals Optimized Model-Functions for CCD-Camera Based Astrometric Plate Measurements

1994 ◽  
Vol 161 ◽  
pp. 288-290
Author(s):  
L. Winter
Keyword(s):  
Ccd Camera ◽  
Detailed Understanding ◽  
New Models ◽  
First Results ◽  
Optimized Model ◽  
Measuring Machine ◽  
Imaging Properties ◽  
Modelling Methods ◽  
Good Agreement

While digitizing the first 1500 photographic plates (astrograph plates and prime focus plates of various telescopes) using our CCD-based measuring machine, we gained a detailed understanding of the properties of the digitized stellar images. These images are different from those obtained by digitizing photographic plates with microdensitometers and similar scanners. Therefore it was necessary to reject the modelling methods given in the literature and compose new models that describe the stellar images more accurately. First results of simulating the whole imaging process, starting with the atmosphere, aberrations of the telescope, etc., and ending by including the imaging-properties of our measuring machine, show good agreement with the measured images as well as the modelling functions.

A CALORIMETRIC DETERMINATION OF THE AVERAGE KINETIC ENERGY OF THE FRAGMENTS FROM U235 FISSION

1955 ◽  
Vol 33 (7) ◽  
pp. 357-363 ◽  
Author(s):  
R. B. Leachman ◽  
W. D. Schafer
Keyword(s):  
Kinetic Energy ◽  
Average Energy ◽  
Fission Products ◽  
Average Kinetic Energy ◽  
U235 Fission ◽  
Induced Fission ◽  
Differential Calorimeter ◽  
Calorimetric Determination ◽  
Good Agreement

The average heat of thermal-neutron induced fission of U235 has been measured by a differential calorimeter. The average energy per fission observed by the calorimeter was 170.1 ± 1.2 Mev. On the basis of the thicknesses of the calorimeter materials and the theoretical energy loss equation, the β energy per fission observed by the calorimeter is 3.0 ± 1 Mev. and, on the same basis, the γ and neutron energy observed is negligible. The resulting 167.1 ± 1.6 Mev. for the average kinetic energy of the fission products is shown to be in good agreement with less direct determinations of this quantity.

Glass as a Material for the Final Disposal of Fission Products

1978 ◽  
Vol 25 (3-4) ◽  
Author(s):  
N. JACQUET-FRANCILLON ◽  
R. BONNIAUD ◽  
C. SOMBRET
Keyword(s):  
Fission Products ◽  
Final Disposal
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