Two LWR experiments have taken place using the above irradiation facility; the first contained UN uranium nitride and the second contained UO 2 fuel pellets inside SiC cladding. The facilities contain nine fuel pins—each comprising 10 fuel pellets—arranged as three fuel rods Figure 15 [ 11 ]. Design calculations indicate that a 1. Eventually, the screen was fabricated with a thickness of 0. Regarding the development of irradiation capabilities under fast neutron spectrum conditions at ATR Figure 9 , a boosted fast flux loop BFFL concept Figure 16 has been proposed.
The desired fast to thermal neutron flux ratio is over In the framework of the GTL Project Conceptual Design, several configurations of an experiment facility that could replicate a fast flux test environment have been studied. The BFFL combines boosters for neutron flux enhancing and thermal neutron absorber for neutron filtering and for reinforcing the increase of fast to thermal flux ratio.
The absorbing material is a Hf-Al composite. This material retains the high thermal conductivity of Al and has the thermal neutron absorption properties of Hf. With this approach, the produced heat can be removed by conduction and can be transferred from the experiment to pressurized water cooling channels [ 13 , 19 ]. In U-Si plates of the required fuel loading, meat thickness and curvature are prototypes [ 13 ]. The design shown in Figure 16 provided a fast flux of approximately 1.
Helium being inert, single phase, and without reactivity effects, it was chosen as the gas coolant medium. The greater the presence of Hf in Al, the greater the removed fraction of thermal neutrons Figure 17 would be.
The heating rates in the Hf-Al as a function of the Hf concentration in the absorber are shown in Figure It was indicated that with a 6. The loop can house 1—4 fuel elements or absorbent elements and can be inserted and removed while the reactor is operating. The aim of the screen was a higher than 0. Hf depletion is low, making its replacement during the irradiation unnecessary. Boron and its compounds find extensive application in nuclear reactors.
The extremely high thermal cross-section of 10 B together with its low abundance cause quick depletion of the boron screen. The depletion can be delayed by 10 B enrichment. In contrast with other absorbers, that is, cadmium, Boron has a significant neutron absorption in the epithermal energy range. Boron is usually combined as a carbide or oxide.
However, an Al-B alloy has been successfully used as neutron screen and another is under development. A factor that should be taken into account in a boron neutron screen design is the swelling that can be caused due to helium generation after boron irradiation with fast neutrons. The ITV is loaded in the central flux trap where the highest neutron flux occurs. The inert inner region of the ITV is filled with Al Figure 20 , in order to avoid water that would increase neutrons thermalization.
He or N 2 gases have been selected for the specimens cooling. The screen should be replaced after a certain irradiation time. In Table 2 the neutron flux values for a case with the Al-B alloy filter 4.
As can be seen, the presence of the Al-B screen increases the fast over thermal flux ratio about twice. An Al-B screen has been proposed to be loaded in a test facility in the ATR Figure 8 , in order to harden the spectrum, providing an acceptable environment for V alloy testing The two basic design goals of the test design assembly were the achievement of 10 displacements per atom dpa per year in vanadium while limiting the 51 V transmutation to less than 0.
The ITV is installed in the central flux trap, where the highest value of the neutron flux occurs. The results are presented in Table 3. Figure 21 shows the impact of an Al-B filter installation on the neutron.
As the enrichment increases, the hardening of the neutron spectrum is more intense. In order to hold the averaged ratio over 80, the filter needs to be replaced after EFPDs 4 typical operation cycles. The B 4 C filter was inserted between the walls of an Al irradiation tube. The area outside the irradiation tube was surrounded by Al displacers. The dimensions of the irradiation tube are shown in Table 4. Two changing parameters were investigated, that is, the thickness of the filter and the enrichment of 10 B.
In total, eight cases were studied and compared with the reference case, in which the B 4 C filter was replaced by Al. The energy spectrum was divided in five groups with their upper boundaries given in Table 5.
The influence of the thickness and enrichment variations on the energy spectrum and on neutron flux is shown in Figures 23 and 24 , respectively.
The following results were derived by this study. Although the results of the study have shown the great thermal neutron absorption capability of B 4 C, its utilization was not considered in practice because of its reactivity effect and its low heat conductivity [ 17 ]. Europium is a material with extremely high thermal neutron cross-section. The use of europium is limited since it is rare and hence expensive.
In nuclear reactors, europium is utilized in control rods. In this section, only one neutron screen with europium is presented, that is currently under development in HFIR. Cases of europium neutron screens already used in reactor facilities have not been found in the open literature, possibly due to the reasons described above. The screen has been planned to be installed at the reactor flux trap FT southeastern core part where the fast neutron flux exceeds 1. A tri-pin assembly design Figure 25 , occupying seven target locations, was selected for the application.
Calculations of performance characteristics such as linear heat generation rate, neutron flux magnitude, fast-to-thermal flux ratio, and displacements per atom dpa were performed in HFIR using the MCNP code [ 24 ]. The results are provided in Table 6.
Neutron screen performances can easily and inexpensively meet demands for a fast neutron flux facility, which arise due to the absence of adequate number of fast experimental reactors. In Section 2 , neutron screens which use a solid shielding material were presented.
The screens that were reported are either already successfully implemented and used or are still under development or study. The purpose of their use is the creation of an environment with a neutron energy spectrum free as much as possible from thermal and epithermal neutrons. Four solid materials were presented, that is, boron, cadmium, hafnium and europium. In principle the material selection depends on the reactor which will host the screen and on the required conditions that should be achieved.
Europium has not met great industrial development due to its excessively high cost compared to boron, cadmium, and hafnium which are widespread in nuclear reactors. From the safety point of view, europium oxide reacts readily with water and boron gets highly corrosive at high temperatures.
On the contrary, hafnium and cadmium are both corrosion-resistant metals. Regarding their mechanical properties, cadmium has the disadvantages of low melting and low boiling points. Irradiating boron with fast neutrons induces the generation of helium, causing swelling. Hafnium exhibits the best mechanical properties. The size of the screen depends on the materials depletion and on the experiment requirements.
The fact that europium and hafnium activation products have high-cross sections, delays their depletion with time, extending thus the neutron screen operational time. On the contrary, boron and cadmium have only one isotope with high cross-section, and in low abundance, so that both deplete fast with time. This is achieved either by increasing the volume of the material or through enrichment of the absorber i.
On the other hand, hafnium has the lowest cross-section of all four materials. Therefore, in order to fulfill the same requirements, larger volume of hafnium must be used, compared to the necessary volumes of other screening materials.
In general, few mm of absorber thickness are sufficient. From all of the four materials, boron has the highest cross-section in the epithermal energy region and its behavior is regular.
The epithermal neutron capture capability of the absorbers results to a neutron spectrum which better represents that of a fast reactor. A combination of absorbers can be utilized in order to create a more efficient neutron screen. In this chapter, neutron screens which use a gaseous or a liquid thermal neutrons absorbing material are presented. The purpose of fluid screens is to generate power transients, that is, power increase or decrease in a few seconds on the volume of a fuel sample.
The fluctuation is accomplished by pressure variation i. The performance of variable screens is necessary in order to examine the fuels behavior in case of exposure to sudden power variation conditions. Three different types of variable neutron screens are reported in this chapter, that is, gaseous 3 He, liquid H 3 BO 3 , and gaseous BF 3.
Parameters that can enhance the transient amplitude comprise the core location where the screen is adapted, the distance between the screen and the sample, the absorber type that is used in the screen, the absorber content in the screen, and the operational phase of the reactor. In addition, large amplitude transients can be achieved by combining the screen modification and the reactor power changes, for example, scram.
The fluid screens are classified by material and are divided into subclasses depending on whether they have already been used or they are under development or study. Most of power transients with a 3 He neutron screen have nowadays been abandoned for safety reasons aroused due to production of Tritium with 3 He activation.
The first experiment with a variable 3 He neutron screen was made in The screen surrounded a Na gas loop, in which the sample was inserted Figure Although the He gaseous screen performance in most reactors has been abandoned, ramp test facilities with 3 He screen are still in operation in the Halden reactor.
More precisely, in-pile loops with gaseous 3 He are utilized for studying the fuel rods performance under power transient conditions. The experiment is surrounded by Zircaloy, in order to be isolated from the reactor environment. In Figure 29 four typical power ramp tests are shown. Boron isotope 10 B has an extremely high thermal cross-section.
For power transient experiments boron compounds are used. As stated in Section 2. BF 3 utilization has been abandoned due to its corrosive and poisonous nature and its relatively low absorption cross-section, which requires screen utilization at high pressure.
The capsule carrier is placed in an aluminum filled space and its horizontal cross-section is described by nine concentric rings surrounding the The high thermal neutron absorption of 10 B can cause a power reduction, although the BF 3 gas was not sufficient for large power ramps in the in-pile experiments performed.
Stainless steel was used as a construction material and BF 3 gas was inserted into a special annular space surrounding the fuel. Due to the low cost of enriched B, the possibility of using enriched BF 3 —instead of natural—was considered.
With the same amount of enriched BF 3 less space would be necessary and more neutron absorption could be achieved, without pressure changes. Moreover, safety would also be increased. BF 3 is no longer in use at the HFR, because of its corrosiveness, poisonous nature, and its relatively low absorption cross section which requires screen utilization at high pressure.
The facility was placed outside the core in the pool side facility PSF. The region around the sodium containment was filled with BF 3 gas, acting as neutron shield. The variation of BF 3 concentration, the displacement of the facility to and from the core , or the combination of both allowed a power transient factor of 2 to 4.
B pressure increase varied slightly the power reduction and neutron absorption. CEN for fuel power transient experiments Figure The modification of boron concentration allows the tuning of the basic linear power.
The H 3 BO 3 concentration can vary between zero i. VANESSA not only provides the variable thermal neutron absorption but, via the thermal balance method, also serves for fuel rod power determination. Through the volume of H 3 BO 3 solution generated during the operation, limited number of transients during a cycle takes place.
The results showed that with natural boron a fuel power transient factor of about 2. The cooling of the screen is performed by the primary BR2 cooling system water. Three solutions for the plug material were studied; RODEO filled with water gives the highest transient ratio [ 27 ]. The results are shown in the Table 7. The power increase in the H3 and H4 channels Figure 1 was calculated.
For intermediate B concentrations, the influence of the RODEO rotation on the neighboring fuel elements will therefore be extremely small. RODEO rotation can be achieved on a time scale of the order of a few seconds.
In Section 3 , neutron screens which use a gaseous or a liquid thermal absorbing material were presented. Two different materials are typically utilized, helium and boron, that is, gaseous 3 He, gaseous BF 3 , and liquid H 3 BO 3.
BF 3 and H 3 BO 3 screens were developed with the purpose to replace the widely used 3 He screen for safety reasons. However, BF 3 performance also causes safety concerns: through neutron capture, 3 He generates tritium and BF 3 can potentially form F 2 which is a poisonous and corrosive gas. Compared to the previous materials, H 3 BO 3 can be considered a safe candidate for performing power transients, since it is not corrosive and does not produce active by-products.
For this reason its replacement with boron compounds having much lower cross-sections cannot provide the same power transients. This can be compensated either by pressure increment or by boron enrichment. Fluid neutron screens can provide successful power transients. Neutron screen technology has been developed in order to fulfill several application requirements. The key idea of the neutron screens is the utilization of some materials capability to absorb neutrons at specific energy range.
In this report only screens which utilize thermal and epithermal absorbers were presented. First, the capability to simulate fast nuclear reactor conditions in a specific area by adapting a neutron screen is reviewed through the presentation of relevant studies and applications. As arises from the available literature, several neutron spectra have been simulated in different reactors. Four solid materials were presented, that is, boron, cadmium, hafnium, and europium.
Their performance characteristics differ in terms of their mechanical properties, compatibility, depletion, and so forth. However, in most cases presented in this chapter, the utilized material could be replaced by one of the rest. The material selection is determined by factors such as the reactor conditions desired to be simulated, the reactor type in which the screen will be inserted and its operational conditions , the reactor coolant in order to prevent safety issues arising from possible interaction with coolant , and the available space for the screen.
Second, the capability of the existing reactors to irradiate samples under power transients by exploiting the performance of neutron screens is examined based on reported experience. More specifically, the utilization of neutron screens with fluid materials is reviewed. The method is based on the gas pressure or the liquid concentration variation so that power transients can be initiated.
Two different materials have been reported, that is, helium and boron, namely, gaseous 3 He, gaseous BF 3 , and H 3 BO 3. The last two neutron screens have been developed in order to replace the unsafe utilization of 3 He, because of the tritium generation problem. BF 3 utilization has been limited and eventually abandoned because of its corrosive and poisonous nature.
Moreover, its performance could not reach power transient ranges equivalent to those obtained using 3 He. Likewise, by using H 3 BO 3 the achieved power transients appear much lower, so that the screen is combined with a displacement system RODEO , thus providing very fast power transients.
The main conclusion is that neutron screens are worth studying and developing since they can successfully contribute to the creation of desirable special irradiation conditions, which cannot be achieved during normal reactor operation due to technological or economic reasons.
The authors declare that there is no conflict of interests regarding the publication of this paper. Investing in knowledge society through the European Social Fund. Chrysanthopoulou et al. This is an open access article distributed under the Creative Commons Attribution License , which permits unrestricted use, distribution, and reproduction in any medium, provided the original work is properly cited.
Article of the Year Award: Outstanding research contributions of , as selected by our Chief Editors. Read the winning articles. Journal overview. Special Issues. Chrysanthopoulou , 1,2 P. Savva, 1 M. Varvayanni, 1 and N. Academic Editor: Alejandro Clausse. Received 17 Oct Revised 22 Apr Accepted 23 Apr Published 11 Aug Abstract The presence of fast neutron spectra in new reactors is expected to induce a strong impact on the contained materials, including structural materials, nuclear fuels, neutron reflecting materials, and tritium breeding materials.
Introduction A neutron screen is a device, a model, a system, or a technology that provides the capability of locally tailoring the neutron spectrum. Solid Neutron Screens In this chapter, neutron screens which use a solid material are presented; four solid, thermal neutrons absorbing materials are reported, that is, cadmium, hafnium, boron, and europium.
Cadmium Neutron Screens Cadmium is a material widely used for thermal neutron filtering, because of its excellent thermal neutron capture capability. Figure 1. Horizontal cross-section of BR2 with a typical loading [ 2 ].
Figure 2. Neutron spectra in BR2 [ 2 ]. Figure 3. Figure 4. Figure 5. Figure 6. Neutron flux inside the HICU experiment computed for four conditions: with and without neutron screen, with natural Li 7. Table 1. Effect of Cd filter thickness on the neutron flux [ 6 ]. Figure 7. Effect of a 0. Figure 8.
ATR core cross-section—4 flux traps, 5 in-pile tubes, and 68 in reflector—[ 10 ]. Figure 9. Figure The influence of the thickness of the Hf shield on the power density in the two fuel pins [ 3 ].
The spectrum in the molybdenum shroud surrounding the fuel [ 3 ]. This means that if there is some sort of power failure or loss of signal the control rods are immediately released and fall into the reactor core because of gravity. This dropping motion can also be induced manually if the machinery holding the rods up fails in some way.
When the control rods are dropped into the reactor, it is a process known as scramming. Fossil Fuels. Nuclear Fuels. Acid Rain. Climate Change. Climate Feedback. Ocean Acidification. Rising Sea Level.
Figure 2. A schematic showing how reactor power output changes with how much the control rods shown in green are inserted. On the left, the control rods are inserted more than usual, reducing the power output of the reactor. On the right, the control rods are inserted less than usual, increasing the power output. July 7, Reactor Vessel Head [Online]. Nuclear Energy in the 21st Century , 2nd Ed. Burlington, MA, U.
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