Mixed-spectrum Reactor Research Articles

Response of unalloyed tungsten to mixed spectrum neutrons

Neutron irradiation in mixed spectrum reactors is an effective approach to investigate the synergistic effects of solid transmutations and atomic displacement damage in tungsten. In this article the findings and interpretations from a large-scale tungsten irradiation campaign conducted using the High Flux Isotope Reactor of Oak Ridge National Laboratory are summarized. The response of originally-unalloyed tungsten to mixed spectrum neutron irradiation is characterized by enormous hardening, degradation of modulus of toughness eventually leading to embrittlement, and decreased thermal conductivity. These property changes may be attributed primarily to production, segregation, and precipitation of rhenium and osmium and assisted by displacement damage. The stages describing the evolution of the microstructural development and property changes with the neutron dose and the accompanying accumulation of transmutation products are proposed.This manuscript has been co-authored by UT-Battelle, LLC under Contract No. DE-AC05-00OR22725 with the U.S. Department of Energy. The United States Government retains and the publisher, by accepting the article for publication, acknowledges that the United States Government retains a non-exclusive, paid-up, irrevocable, worldwide license to publish or reproduce the published form of this manuscript, or allow others to do so, for United States Government purposes. The Department of Energy will provide public access to these results of federally sponsored research in accordance with the DOE Public Access Plan (http://energy.gov/downloads/doe-public-access-plan).

Performance and safety evaluation of a mixed-spectrum reactor design

Mixed-spectrum reactors (MXR) have been investigated for a wide variety of applications. Typical MXR core designs are based on fast configurations with moderating material inserted within localized regions of the core. Limited analysis has been conducted on assessing important performance and safety aspects. Three main safety-related metrics are devised in this article to validate the feasibility of MXR designs. A long-lived MXR variant was taken as a case study for this analysis. Reactivity feedback mechanisms were evaluated, along with power peaking effects and fast flux damage. Neutron transport simulations found that distortions are manageable in all three areas, with notable improvements in fluence limits and Doppler broadening coefficients. Power peaking effects can be significantly dampened by slight addition of gadolinium in the fuel and by carefully selecting the reflector material.

Design of a Mixed-Spectrum Reactor With Improved Proliferation Resistance for Long-Lived Applications

Design of a Mixed-Spectrum Reactor With Improved Proliferation Resistance for Long-Lived Applications

Effect of neutron irradiation on rhenium cluster formation in tungsten and tungsten-rhenium alloys

To clarify the irradiation hardening behavior of neutron-irradiated tungsten(W), the formation behavior of rhenium(Re)-enriched cluster and W-Re precipitation with displacement damage in fission reactors environment was investigated by the atom probe tomography (APT) technique, and the results obtained by APT and microstructural observation by transmission electron microscopy (TEM) were compared. Neutron-irradiated pure W and W-10Re alloys were used for the analysis. These samples were irradiated to 0.96 dpa at 538 °C in a fast experimental reactor Joyo, and irradiated to 0.90 dpa at 500 °C in a mixed spectrum reactor HFIR. The chemical compositions of the matrix, Re clusters, and precipitates were determined by the AP analysis. The Joyo-irradiated samples contained lower levels of transmuted Re and Os, while the HFIR-irradiated samples contained higher levels of Re and Os. The Re-enriched clusters demonstrated characteristic spatial distribution related to defect cluster distribution in each irradiated specimen.

Fission products measured from highly-enriched uranium irradiated under 10B4C in a research reactor

Prior work has demonstrated the use of a natural B4C capsule for spectral-tailoring in a mixed spectrum reactor as an alternate and complementary method to critical assemblies for performing nuclear data measurements at near 235U fission-energy neutron spectrum. Previous fission product measurements showed that the neutron spectrum achievable with natural B4C was not as hard as what can be achieved with critical assemblies. New measurements performed with the Washington State University TRIGA reactor using a B4C capsule 96 % enriched in 10B resulted in a neutron spectrum very similar to a critical assembly and a pure 235U fission spectrum. Fission product yields measured following an irradiation of a sample with this new method and subsequent radiochemical separations are presented here.

Design and testing of a 10B4C capsule for spectral-tailoring in mixed-spectrum reactors

A boron carbide capsule highly enriched in 10B has been designed and used for spectral-tailoring experiments at the TRIGA reactor at Washington State University. New experiments show that enriching the boron to 96% B-10 results in additional absorption of neutrons in the resonance region thereby producing a neutron spectrum that is much closer to a pure 235U fission spectrum than measured previously with a natural boron carbide capsule. A cadmium outer cover was used to reduce thermal alpha heating. The neutron spectrum calculated with MCNP was found to be in very good agreement with measured activation rates from neutron fluence monitors.

Supercritical Water-Cooled Reactors

Supercritical water-cooled reactor (SCWR) is the watercooled reactor using supercritical pressure water as coolant [1]. It is considered as one of the promising Generation IV reactors, due to its advantages of plant simplification and high thermal efficiency [2, 3]. Several design concepts of SCWRs have been proposed: (a) supercritical water-cooled thermal neutron reactor; (b) supercritical water-cooled fast neutron reactor; (c) supercritical water-cooled mixed neutron spectrum reactor; (d) supercritical water-cooled pebble bed reactor; (e) supercritical heavy-water-cooled reactor. The detailed design parameters of some typical SCWR concepts around the world are summarized in Table 1 [4]. Recently, the use of thorium in the SCWRs has been investigated [5, 6]. The advantages of the SCWRs are shown as follows [1].

Fission yield measurements from highly enriched uranium irradiated inside a boron carbide capsule

A boron carbide capsule was previously designed and tested by Pacific Northwest National Laboratory (PNNL) and Washington State University (WSU) for spectral-tailoring in mixed spectrum reactors. The presented work used this B4C capsule to create a fission product sample from the irradiation of highly enriched uranium (HEU) with a fast fission neutron spectrum. An HEU foil was irradiated inside of the capsule in WSU’s 1 MW TRIGA reactor at full power for 200 min to produce 5.8 × 1013 fissions. After 3 days of cooling, the sample was shipped to PNNL for radiochemical separations and analysis by gamma and beta spectroscopy. Fission yields for products were calculated from the radiometric measurements and compared to measurements from thermal neutron induced fission (analyzed in parallel with the non-thermal sample at PNNL) and published evaluated fast-pooled and thermal nuclear data. Reactor dosimetry measurements were also completed to fully characterize the neutron spectrum and total fluence of the irradiation.

Design and Testing of a Boron Carbide Capsule for Spectral Tailoring in Mixed-Spectrum Reactors

Abstract A boron carbide capsule has been designed and used for spectral-tailoring experiments at the TRIGA reactor at Washington State University. Irradiations were conducted in pulsed mode and in continuous operation for up to 4 h. A cadmium cover was used to reduce thermal heating. The neutron spectrum calculated with the Monte Carlo N-particle transport code was found to be in good agreement with reactor dosimetry measurements using the STAY’SL computer code. The neutron spectrum resembles that of a fast reactor. The design of a capsule using boron carbide fully enriched in 10B shows that it is possible to produce a neutron spectrum similar to that of 235U fission.

Irradiation creep of SA 304L and CW 316 stainless steels: Mechanical behaviour and microstructural aspects. Part I: Experimental results

Solution annealed 304L (SA 304L) and cold work 316 (CW 316) austenitic stainless steel irradiation creep behaviour have been studied thoroughly. Irradiations were carried out in fast breeder reactors BOR-60 (at 330 °C, up to 120 dpa) and EBR-II (at 375 °C, up to 10.5 dpa), and in the OSIRIS mixed spectrum reactor (at 330 °C, up to 9.8 dpa). After an incubation threshold, the irradiation creep of the austenitic stainless steels is linear in stress and in dose. Creep appears to be athermal in this temperature range. A significant difference in the behaviour is measured between the creep of SA 304L and CW 316. In order to study the anisotropy of loop population, which would be the signature of a possible stress induced preferential absorption (SIPA) mechanism for irradiation creep, special attention was given to the measurement of anisotropy of loop distribution between the four families. The anisotropy induced by an applied stress has been shown to be in the range of the statistical scatter in the situation where no stress is applied. TEM microstructural analyses performed on this sample show slight difference between the microstructure of specimens deformed under irradiation and the microstructure of specimens irradiated without stress under the same irradiation conditions.

Microstructural Evolution of Neutron Irradiated Stainless Steels

Materials of the core internals of Pressurized Water Reactors (austenitic stainless steels) are submitted to neutron irradiation. Some baffle/former bolt cracking were detected by non-destructive examinations and removed for metallurgical examination. The most likely cause to cracking was deduced to be Irradiation Assisted Stress Corrosion Cracking.To understand the ageing mechanisms associated to irradiation and propose life predictions of the component, irradiation damage is investigated experimentally by Transmission Electron Microscopy for two representative austenitic stainless steels, SA 304L and CW 316. The microstructure of these austenitic stainless steels constitutive of internals structures of PWR is characterized by TEM and compared to the microstructure obtained after irradiation for high doses up to 40 dpa at 320°C in experimental OSIRIS (mixed flux spectrum) and BOR-60 (fast breeders) reactors and at 300°C in experimental SM-2 (mixed flux spectrum) reactor. Only minor differences were detected.

Void swelling of AISI 321 analog stainless steel irradiated at low dpa rates in the BN-350 reactor

In several recently published studies conducted on a Soviet analog of AISI 321 stainless steel irradiated in either fast reactors or light water reactors, it was shown that the void swelling phenomenon extended to temperatures as low as ∼300 °C or less, when produced by neutron irradiation at dpa rates in the range 10 −7–10 −8 dpa/s. Other studies yielded similar results for AISI 316 and the Russian analog of AISI 316. In the current study a blanket assembly duct from BN-350, constructed from the Soviet analog of AISI 321, also exhibits swelling at dpa rates on the order of 10 −8 dpa/s, with voids seen as low as 281 °C and only 0.65 dpa. It appears that low-temperature swelling occurs at low dpa rates in 300 series stainless steels in general, and also occurs during irradiations conducted in either fast or in mixed spectrum reactors as shown in other studies.

Mechanical properties of neutron-irradiated nickel-containing martensitic steels: I. Experimental study

Tensile and Charpy specimens of 9Cr–1MoVNb (modified 9Cr–1Mo) and 12Cr–1MoVW (Sandvik HT9) steels and these steels doped with 2% Ni were irradiated at 300 and 400 °C in the High Flux Isotope Reactor (HFIR) up to ≈12 dpa and at 393 °C in the Fast Flux Test Facility (FFTF) to ≈15 dpa. In HFIR, a mixed-spectrum reactor, ( n, α) reactions of thermal neutrons with 58Ni produce helium in the steels. Little helium is produced during irradiation in FFTF. After HFIR irradiation, the yield stress of all steels increased, with the largest increases occurring for nickel-doped steels. The ductile–brittle transition temperature (DBTT) increased up to two times and 1.7 times more in steels with 2% Ni than in those without the nickel addition after HFIR irradiation at 300 and 400 °C, respectively. Much smaller differences occurred between these steels after irradiation in FFTF. The DBTT increases for steels with 2% Ni after HFIR irradiation were 2–4 times greater than after FFTF irradiation. Results indicated there was hardening due to helium in addition to hardening by displacement damage and irradiation-induced precipitation.

Towards reduced activation structural materials data for fusion DEMO reactors

The development of first wall, blanket and divertor materials that are capable of withstanding for many years high neutron and heat fluxes, is a critical path to fusion power. Therefore, the timely availability of a sound materials database has become an indispensable element in international fusion road maps. In order to provide materials design data for the short-term needs of ITER test blanket modules (TBMs) and for a DEMOnstration fusion reactor, a wealth of R&D results on the European reduced activation ferritic–martensitic steel EUROFER, and on oxide dispersion strengthened variants are being characterized, mainly in the temperature window 250–650°C. The characterization includes irradiations up to 15 dpa in the mixed spectrum reactor HFR and up to 75 dpa in the fast breeder reactor BOR60. Industrial EUROFER-batches of 3.5 and 7.5 tons have been produced with a variety of semi-finished, quality-assured product forms. To increase the thermal efficiency of blankets, high temperature resistant SiCf/SiC channel inserts for liquid metal coolant tubes are also developed. Regarding radiation damage resistance, broad based reactor irradiation programmes deal with several ranges from ≤ 5 dpa (ITER TBMs) up to 75 dpa (DEMO). For the European divertor designers, a materials database is currently being set up for pure W and its alloys, and related reactor irradiations are foreseen with temperatures from 650°C to 1000°C.

Characterization of 08Cr16Ni11Mo3 stainless steel irradiated in the BN-350 reactor

In several recently published studies conducted on a Soviet analog of AISI 321 stainless steel irradiated in either fast reactors or light water reactors, it was shown that the void swelling phenomenon extended to temperatures as low as ∼300° C, when produced by neutron irradiation at dpa rates in the range 10 −7 to 10 −8 dpa/s. Other studies yielded similar results for AISI 316. In the current study a blanket duct assembly from BN-350, constructed from the Soviet analog of AISI 316, also exhibits swelling at dpa rates on the order of 10 −8 dpa/s, with voids seen as low as 281 °C and only 1.3 dpa. It appears that low-temperature swelling at low dpa rates occurs in 300 series stainless steels in general, and during irradiations conducted in either fast or mixed spectrum reactors.

Pros and cons of nickel- and boron-doping to study helium effects in ferritic/martensitic steels

In the absence of a 14 MeV neutron source, the effect of helium on structural materials for fusion must be simulated using fission reactors. Helium effects in ferritic/martensitic steels have been studied by adding nickel and boron and irradiating in a mixed-spectrum reactor. Although the nickel- and boron-doping techniques have limitations and difficulties to estimate helium effects on the ferritic/martensitic steels, past irradiation experiments using these techniques have demonstrated similar effects on the swelling and Charpy impact properties that are indicative of a helium effect. Although both techniques have disadvantages, it should be possible to plan experiments using the nickel- and boron-doping techniques to develop an understanding of the effects of helium on mechanical properties.

Calculation and measurement of helium generation and solid transmutants in Cu–Zn–Ni alloys

A method recently proposed by Garner and Greenwood [1]would allow the separation of the simultaneous and possibly synergistic effects of solid and gaseous transmutation in Cu–Zn–Ni alloys. Copper transmutes primarily to Ni and Zn, leading to an accelerated production of helium from the ingrowth of zinc in mixed-spectrum reactors. Adding Zn to Cu alloys eliminates the time required for this ingrowth and leads to accelerated helium rates which are comparable to the well-known effect in Ni. Using a suitable Cu–Zn–Ni alloy matrix in a comparative irradiation of thermal neutron shielded and unshielded specimens, the separate influences of the solid and gaseous transmutants should be distinguishable. To assess the potential of this proposal, it is necessary to show that we can accurately predict the isotopic and elemental evolution of these alloys. In the current study, several different Cu–Zn–Ni alloys, irradiated in the HFIR JP-23 experiment, were characterized by measuring of the helium content as well as the solid transmutants. The measured values are in good agreement with calculations based on neutron dosimetry, therefore allowing the successful design of the transmutation experiment proposed by Garner and Greenwood.

Irradiation resistance of DS copper/stainless steel joints fabricated by friction welding methods

Copper alloys tubes are assumed to be used for the heat sink system of the ITER first wall and divertor. To join copper tubes with type 316LN stainless steel tubes (for the collector system) various methods are provided, the most promising being the friction welding. Specimens of joints and the base materials were irradiated in the SM-2 mixed spectrum reactor to a dose of 0.2 dpa at T irr=150°C and 300°C. Specimens of GlidCop Al25/316LN friction welded joints have a relatively high level of strength properties and a rather low plasticity after irradiation.

Effect of neutron dose and spectra, He/dpa ratio and Ni and Zn accumulation on irradiation damage of pure copper and PH and DS copper alloys

The present analysis is based mainly on the results of recently undertaken investigations of radiation resistance of copper alloys irradiated in the SM-2 and BOR-60 reactors, as well as on the results of the just accomplished collaborative spectrally tailored experiment in the SM-2 reactor. Low-temperature radiation embrittlement at T irr∼100°C presents considerable problems for dispersion strengthened (DS) and precipitation-hardened (PH) copper alloys, as their total elongation at T test= T irr=100°C drops to ∼1% and uniform elongation to ∼0.1%. At irradiation temperatures above 300°C, the most significant mechanical property change for PH copper alloys is the softening effect due to radiation enhanced precipitate coarsening and dislocation recovery and recrystallization processes. The softening effect is sensitive to the neutron spectrum, and the threshold temperature of its manifestation in a mixed-spectrum reactor is apparently ∼100°C less than in a fast neutron reactor. The results of copper alloys irradiation in mixed-spectrum reactors serve as the basis for an analysis of the neutron spectrum effect and the problem of fission–fusion correlation. At irradiation temperatures greater than 350°C, helium embrittlement associated with the high helium accumulation rate for copper alloys exposed to fusion neutrons will be one of the main problems limiting the lifetime. At increased irradiation temperatures, creep will present a special problem. It is apparent that the DS copper alloys have some advantage in their radiation resistance compared to PH alloys such as CuCrZr, this being attributable to the high-temperature stability of their hardening structure.

The effect of rhenium on the radiation damage resistivity of Mo–Re alloys

The results of investigations of alloys in the Mo–Re system are presented with Re varying from 0.5 to 47%. Samples were irradiated in the SM-2 reactor up to doses of 1–5.5×10 25 n/m 2 ( E>0.1 MeV) at 100°C, 210°C, 260°C and 330°C and in the BOR-60 reactor to doses of 0.8–2.0×10 26 n/m 2 ( E>0.1 MeV) at 480°C, 550°C, 760°C and 800°C. Irradiation to high doses in the SM-2 and BOR-60 reactors results in pronounced embrittlement of Mo–Re alloys in the testing temperature range of 300–600°C. The study of the electrical conductivity of the alloys has shown that irradiation in SM-2 results in a linear increase of the electrical resistivity with increasing Re concentration. Based on the analysis of available data the tentative conclusion has been made, that during irradiation in a mixed-spectrum reactor the resonance reaction proceeds on Re giving rise to the accumulation of transmutants (Os) reducing the electrical conductivity of the alloys.