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Publications (10 of 87) Show all publications
Wallenius, J. & Bortot, S. (2019). A new paradigm for breeding of nuclear fuel. Annals of Nuclear Energy, 133, 816-819
Open this publication in new window or tab >>A new paradigm for breeding of nuclear fuel
2019 (English)In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 133, p. 816-819Article in journal (Refereed) Published
Abstract [en]

Breeding of nuclear fuel from fertile nuclides may allow to extend known nuclear fuel resources from a century to thousands of years. In this article, we show that the requirement for a breeder reactor fuel to feature an effective neutron production per absorption larger than 2 (eta > 2) breaks down for fertile and fissionable nuclides meeting two criteria related to the relative magnitude of capture and fission cross sections. Moreover, we find that a breeding ratio larger than unity can be achieved for fuels consisting of a single nuclide, in spite of the this nuclide featuring eta < 2. In particular, neptunium is identified as a nuclear fuel that can sustain a reactivity increase over time up to a burn-up exceeding 240 GWd/ton in a fast neutron spectrum.

Place, publisher, year, edition, pages
PERGAMON-ELSEVIER SCIENCE LTD, 2019
Keywords
Breeding, Conversion ratio, Reactivity based criterion, New paradigm, Neptunium
National Category
Subatomic Physics Energy Systems
Identifiers
urn:nbn:se:kth:diva-260989 (URN)10.1016/j.anucene.2019.07.028 (DOI)000484649800078 ()2-s2.0-85069550862 (Scopus ID)
Note

QC 20191008

Available from: 2019-10-08 Created: 2019-10-08 Last updated: 2019-10-16Bibliographically approved
Wallenius, J. (2019). Maximum efficiency nuclear waste transmutation. Annals of Nuclear Energy, 125, 74-79
Open this publication in new window or tab >>Maximum efficiency nuclear waste transmutation
2019 (English)In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 125, p. 74-79Article in journal (Refereed) Published
Abstract [en]

Efficient recycle of high level waste in spent nuclear fuel decreases the time required to isolate residual waste by a factor of 100 and reduces the volume of the waste repository by a factor of 4-6. Technical approaches to accomplish this feat include fast neutron Generation IV reactors and accelerator driven systems. Here, I present a novel design of a very small, passively safe lead-cooled reactor with (Np,Am)N fuel, which is shown to achieve maximum possible efficiency in transmutation of long-lived high level waste, while producing a nuclear fuel that is difficult to use for weapons production. Using this reactor for waste management minimises costs for demonstrating closure of the nuclear fuel cycle.

Place, publisher, year, edition, pages
PERGAMON-ELSEVIER SCIENCE LTD, 2019
Keywords
Nuclear waste transmutation, Lead coolant, Nitride fuel
National Category
Environmental Management
Identifiers
urn:nbn:se:kth:diva-244496 (URN)10.1016/j.anucene.2018.10.034 (DOI)000457657400008 ()2-s2.0-85055642581 (Scopus ID)
Note

QC 20190328

Available from: 2019-03-28 Created: 2019-03-28 Last updated: 2019-05-14Bibliographically approved
Wallenius, J. & Bortot, S. (2018). A small lead-cooled reactor with improved Am-burning and non-proliferation characteristics. Annals of Nuclear Energy, 122, 193-200
Open this publication in new window or tab >>A small lead-cooled reactor with improved Am-burning and non-proliferation characteristics
2018 (English)In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 122, p. 193-200Article in journal (Refereed) Published
Abstract [en]

In this paper, a novel approach for transmutation of americium in fast reactors is presented. Using enriched uranium as fissile support, rather than plutonium, it is shown that a minor actinide burning rate of 25 kg/TWh(th) is possible to achieve in a passively safe, critical lead-cooled reactor. Moreover, the plutonium produced by transmutation of Am-241 features up to 38% (PU)-P-238, making it difficult to use for weapons production.

Place, publisher, year, edition, pages
PERGAMON-ELSEVIER SCIENCE LTD, 2018
Keywords
Small lead-cooled nuclear reactor, Minor actinide transmutation, Non-proliferation
National Category
Energy Engineering
Identifiers
urn:nbn:se:kth:diva-238109 (URN)10.1016/j.anucene.2018.08.043 (DOI)000447107700019 ()2-s2.0-85052860299 (Scopus ID)
Note

QC 20190110

Available from: 2019-01-10 Created: 2019-01-10 Last updated: 2019-05-13Bibliographically approved
Wallenius, J., Qvist, S., Mickus, I., Bortot, S., Szakalos, P. & Ejenstam, L. (2018). Design of SEALER, a very small lead-cooled reactor for commercial power production in off-grid applications. Nuclear Engineering and Design, 338, 23-33
Open this publication in new window or tab >>Design of SEALER, a very small lead-cooled reactor for commercial power production in off-grid applications
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2018 (English)In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 338, p. 23-33Article in journal (Refereed) Published
Abstract [en]

In this paper, the conceptual design of a small lead-cooled nuclear reactor intended to replace diesel-power in off-grid applications is presented. In a vessel of dimensions making it transportable by air, the targeted design performance is to produce 3 MW of electrical power for up to 30 years without reloading of fuel. Consequently, the inner vessel can be sealed, delaying malevolent access to the nuclear fuel and improving security. Alumina forming alloys are applied to ensure long term corrosion protection of fuel cladding tubes, steam generator tubes and primary vessel over the operational temperature regime. Moreover, decay heat can be removed in a completely passive manner by natural convection from the core to the primary coolant and by thermal radiation from the primary vessel to the environment. Finally, the source term is such that relocation of population residing beyond 1 km from the reactor will not be required even in the case of a complete core melt.

Place, publisher, year, edition, pages
Elsevier, 2018
Keywords
Small lead-cooled nuclear reactor, Off-grid power production, Safety informed design
National Category
Physical Sciences
Identifiers
urn:nbn:se:kth:diva-235861 (URN)10.1016/j.nucengdes.2018.07.031 (DOI)000445542500004 ()2-s2.0-85051384904 (Scopus ID)
Note

QC 20181022

Available from: 2018-10-22 Created: 2018-10-22 Last updated: 2018-10-22Bibliographically approved
Lambrinou, K., Lapauw, T., Jianu, A., Weisenburger, A., Ejenstam, J., Szakálos, P., . . . Vleugels, J. (2016). Corrosion-resistant ternary carbides for use in heavy liquid metal coolants. In: Ceramic Engineering and Science Proceedings: . Paper presented at Ceramic Materials for Energy Applications V - 39th International Conference on Advanced Ceramics and Composites, ICACC 2015, 25 January 2015 through 30 January 2015 (pp. 19-34). (7)
Open this publication in new window or tab >>Corrosion-resistant ternary carbides for use in heavy liquid metal coolants
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2016 (English)In: Ceramic Engineering and Science Proceedings, 2016, no 7, p. 19-34Conference paper, Published paper (Refereed)
Abstract [en]

A primary concern in the development of accelerator-driven systems (ADS) with liquid leadbismuth eutectic (LBE) spallation target and Gen-IV lead-cooled fast reactors (LFRs) is the compatibility of the candidate structural steels with the heavy liquid metal (HLM) coolant In the accelerator-driven system MYRRHA, the envisaged primary coolant is liquid LBE, a potentially corrosive environment for various nuclear grade steels. The inherent LBE corrosiveness is the driving force behind diverse research incentives aiming at the development of corrosion-resistant materials for specific applications. I3ue to their superb corrosion resistance in contact with liquid LBE, MAX phases are currently being assessed as candidate materials for the construction of pump impellers suitable for MYRRHA and Gen-IV LFRs. In the case of the MYRRHA nuclear system, the pump impeller will be called to operate reliably at ∼270°C in contact with moderately-oxygenated (concentration of dissolved oxygen: [O] ≥ 7×10-7 mass%), fast-flowing LBE (LBE flow velocity: v ≈ 10-20 m/s locally on the impeller surface). Selected MAX phases are currently being screened with respect to their capability of meeting the targeted material property requirements, especially the enhanced erosion resistance requested by this particular application. This work gives a state-of-the-art overview of the processing and characterisation of selected MAX phases that are screened as candidate structural materials for the MYRRHA pump impeller. All considered MAX phases were produced via a powder metallurgical route and their performance was assessed by various mechanical tests in air/vacuum and corrosion/erosion tests in liquid LBE.

National Category
Ceramics
Identifiers
urn:nbn:se:kth:diva-186751 (URN)2-s2.0-84959575951 (Scopus ID)9781119040439 (ISBN)
Conference
Ceramic Materials for Energy Applications V - 39th International Conference on Advanced Ceramics and Composites, ICACC 2015, 25 January 2015 through 30 January 2015
Note

QC 20160530

Available from: 2016-05-30 Created: 2016-05-13 Last updated: 2016-05-30Bibliographically approved
Johnson, K. D., Wallenius, J., Jolkkonen, M. & Claisse, A. (2016). Spark plasma sintering and porosity studies of uranium nitride. Journal of Nuclear Materials, 473, 13-17
Open this publication in new window or tab >>Spark plasma sintering and porosity studies of uranium nitride
2016 (English)In: Journal of Nuclear Materials, ISSN 0022-3115, E-ISSN 1873-4820, Vol. 473, p. 13-17Article in journal (Refereed) Published
Abstract [en]

In this study, a number of samples of UN sintered by the SPS method have been fabricated, and highly pure samples ranging in density from 68% to 99.8%TD-corresponding to an absolute density of 14.25 g/cm3 out of a theoretical density of 14.28 g/cm3-have been fabricated. By careful adjustment of the sintering parameters of temperature and applied pressure, the production of pellets of specific porosity may now be achieved between these ranges. The pore closure behaviour of the material has also been documented and compared to previous studies of similar materials, which demonstrates that full pore closure using these methods occurs near 97.5% of relative density.

Place, publisher, year, edition, pages
Elsevier, 2016
Keywords
Generation IV, Nuclear fuel, Pore closure, Sintering, SPS, Uranium nitride, Nitrides, Nuclear fuels, Porosity, Spark plasma sintering, Uranium, Uranium compounds, Number of samples, Relative density, Sintering parameters, Specific porosity, Theoretical density
National Category
Mineral and Mine Engineering Physical Sciences
Identifiers
urn:nbn:se:kth:diva-186981 (URN)10.1016/j.jnucmat.2016.01.037 (DOI)000373490700003 ()2-s2.0-84959376726 (Scopus ID)
Note

QC 20160518

Available from: 2016-05-18 Created: 2016-05-16 Last updated: 2017-11-30Bibliographically approved
Bortot, S., Suvdantsetseg, E. & Wallenius, J. (2015). BELLA: a multi-point dynamics code for safety-informed design of fast reactors. Annals of Nuclear Energy, 85, 228-235
Open this publication in new window or tab >>BELLA: a multi-point dynamics code for safety-informed design of fast reactors
2015 (English)In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 85, p. 228-235Article in journal (Refereed) Published
Abstract [en]

In this paper, the multi-point dynamics code BELLA and its benchmarking with respect to SAS4A/SASSYS-1 is described for a small fast reactor cooled with natural convection of lead (ELECTRA). It is shown that BELLA is capable of reproducing the magnitude of mass-flow, reactivity, power and temperature excursions during design extension conditions with an accuracy better than 10%. Hence, the BELLA code can be used for safety-informed design and stability analyses of fast reactor systems, permitting to isolate essential phenomena and trends of significance for their safety assessment. (C) 2015 Elsevier Ltd. All rights reserved.

Place, publisher, year, edition, pages
PERGAMON-ELSEVIER SCIENCE LTD, 2015
Keywords
Fast reactors, Dynamics, Lumped parameters
National Category
Atom and Molecular Physics and Optics
Identifiers
urn:nbn:se:kth:diva-174909 (URN)10.1016/j.anucene.2015.05.017 (DOI)000361413800023 ()2-s2.0-84931259717 (Scopus ID)
Note

QC 20151019

Available from: 2015-10-19 Created: 2015-10-09 Last updated: 2017-12-01Bibliographically approved
Hania, P. R., Klaassen, F. C., Wernli, B., Streit, M., Restani, R., Ingold, F., . . . Wallenius, J. (2015). Irradiation and post-irradiation examination of uranium-free nitride fuel. Journal of Nuclear Materials, 466, 597-605, Article ID 49308.
Open this publication in new window or tab >>Irradiation and post-irradiation examination of uranium-free nitride fuel
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2015 (English)In: Journal of Nuclear Materials, ISSN 0022-3115, E-ISSN 1873-4820, Vol. 466, p. 597-605, article id 49308Article in journal (Refereed) Published
Abstract [en]

Two identical Phénix-type 15-15Ti steel pinlets each containing a 70 mm Pu<inf>0.3</inf>Zr<inf>0.7</inf>N fuel stack in a 1-bar helium atmosphere have been irradiated in the HFR Petten at medium high linear power (46-47 kW/m at BOL) and an average cladding temperature of 505 °C. The pins were irradiated to a plutonium burn-up of 9.7% (88 MWd/kg<inf>HM</inf>) in 170 full power days. Both pins remained fully intact. Post-irradiation examination performed at NRG and PSI showed that the overall swelling rate of the fuel was 0.92 vol-%/%FIHMA. Fission gas release was 5-6%, while helium release was larger than 50%. No fuel restructuring was observed, and only mild cracking. EPMA measurements show a burn-up increase toward the pellet edge of up to 4 times. All investigated fission products except to some extent the noble metals were found to be evenly distributed over the matrix, indicating good solubility. Local formation of a secondary phase with high Pu content and hardly any Zr was observed. A general conclusion of this investigation is that ZrN is a suitable inert matrix for burning plutonium at high destruction rates.

Keywords
Fission products, Fuels, Helium, Plutonium, Zirconium, Cladding temperatures, Destruction rates, Fission gas release, Helium atmosphere, Helium release, Post irradiation examinations, Secondary phase, Swelling rates, Irradiation
National Category
Chemical Sciences
Identifiers
urn:nbn:se:kth:diva-175609 (URN)10.1016/j.jnucmat.2015.08.054 (DOI)000364883400072 ()2-s2.0-84941800290 (Scopus ID)
Note

QC 20151102

Available from: 2015-11-02 Created: 2015-10-19 Last updated: 2017-12-01Bibliographically approved
Suvdantsetseg, E. & Wallenius, J. (2014). An assessment of prompt neutron reproduction time in a reflector dominated fast critical system: ELECTRA. Annals of Nuclear Energy, 71, 159-165
Open this publication in new window or tab >>An assessment of prompt neutron reproduction time in a reflector dominated fast critical system: ELECTRA
2014 (English)In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 71, p. 159-165Article in journal (Refereed) Published
Abstract [en]

In this paper, an accurate method to evaluate the prompt neutron reproduction time for a reflector dominated fast critical reactor, ELECTRA, is discussed. To adequately handle the problem, explicit time dependent Monte Carlo calculations with MCNP, applying repeated time cut-off technique, is used and compared against the σ ∼ 1/v time dependent absorber method, applying artificial cross section data in the Monte Carlo code SERPENT. The results show that when a reflector plays a major role in criticality for fast neutron reactor, the two methods predict different physical parameters (Λ = 69 ± 2 ns and Λ = 83 ± 1 ns for time cut-off and the 1/v method respectively). The reason is explained by applying Avery-Cohn’s two-region prompt neutron model. 

Keywords
Prompt neutron reproduction time, Reflector dominated critical system
National Category
Other Physics Topics
Identifiers
urn:nbn:se:kth:diva-146988 (URN)10.1016/j.anucene.2014.04.001 (DOI)000337985500018 ()2-s2.0-84899442089 (Scopus ID)
Note

QC 20140625

Available from: 2014-06-19 Created: 2014-06-19 Last updated: 2017-12-05Bibliographically approved
Malkki, P., Jolkkonen, M., Hollmer, T. & Wallenius, J. (2014). Manufacture of fully dense uranium nitride pellets using hydride derived powders with spark plasma sintering. Journal of Nuclear Materials, 452(1-3), 548-551
Open this publication in new window or tab >>Manufacture of fully dense uranium nitride pellets using hydride derived powders with spark plasma sintering
2014 (English)In: Journal of Nuclear Materials, ISSN 0022-3115, E-ISSN 1873-4820, Vol. 452, no 1-3, p. 548-551Article in journal (Refereed) Published
Abstract [en]

Applying a combination of hydriding/nitriding of metallic uranium with the spark plasma sintering technique, we show that uranium nitride pellets with an average porosity of 0.2% may be manufactured. This is considerably smaller than the lowest porosity previously reported in the literature. The high density is attained by sintering at 1650 °C for only three minutes.

Keywords
Porosity, Spark plasma sintering, Uranium compounds, Hydriding, Metallic uranium, Spark plasma sintering techniques, Uranium nitride pellets
National Category
Physical Sciences
Identifiers
urn:nbn:se:kth:diva-161521 (URN)10.1016/j.jnucmat.2014.06.012 (DOI)000339657200078 ()2-s2.0-84903788189 (Scopus ID)
Funder
Swedish Research CouncilEU, FP7, Seventh Framework Programme
Note

QC 20150313

Available from: 2015-03-13 Created: 2015-03-12 Last updated: 2017-12-04Bibliographically approved
Organisations
Identifiers
ORCID iD: ORCID iD iconorcid.org/0000-0002-6082-8913

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