Reactivity control options of space nuclear reactors

This paper compares two ex-core control options of the gas-cooled Submersion Subcritical Safe Space (S^4) reactor with a fast neutrons energy spectrum: (a) rotating BeO drums with 120° thin segments of enriched B₄C in the BeO radial reflector; and (b) sliding segments in the BeO radial reflector. Investigated are the effects on the beginning-of-life (BOL) excess reactivity, reactivity depletion rate and operation life, and the spatial neutron flux distributions and fission power profiles in the core. Also investigated is the effect of reducing the thickness of the enriched B₄C segments in the control drums on the BOL excess reactivity, when one or two of the 6 drums are stuck in the shutdown position. Reducing the thickness of the B₄C segments from 0.5 mm to 0.238 mm, with one drums stuck in the shutdown position, increases BOL cold and hot-clean excess reactivity from +$1.71 and +$0.47 to +$2.38 and +$0.89, respectively. These reactivity values are almost identical to those of the reactor with one of the six reflector segments stuck open in the shutdown position. Results also showed that the control options made little difference in the reactor performance. The power peaking in the reactor core with sliding reflector segments is slightly lower and the spatial power profiles are relatively flatter. The operation life of the reactor with a sliding reflector segments control, when operating at a nominal thermal power of 471 kW, is only 22 full power days longer than with rotating drums control.     

Submersion criticality safety of fast spectrum space reactors: Potential spectral shift absorbers

Compact, fast spectrum, nuclear reactors are being considered to support NASA's future space exploration sometime in the next decade. In order to secure launch approval, these reactors should remain sufficiently subcritical when submerged in seawater or wet sand and subsequently flooded, following a launch abort accident. In such an accident, the neutron spectrum in the reactor is thermalized, typically increasing reactivity, and potentially making the reactor supercritical. Incorporating “Spectral Shift Absorbers” (or SSAs), which have significantly higher absorption cross-sections for thermal versus fast neutrons, could offset the reactivity increase. It has always been the assertion that the worst-case submersion accident involves a fully flooded reactor; however, this work shows that, depending on the type and amount of SSA in the reactor, a submerged but unflooded reactor could be more reactive. A screening of the existing nuclear database for potential SSAs yielded 28 elements and nuclides, which are examined in detail as additives to a representative homogenous space reactor core by varying the SSA-to-U²³⁵ atom ratio. The effect of placing a thin coating of different SSA materials on the outside surface of the reactor core is also investigated. Nine SSAs (boron-10, cadmium, cadmium-113, samarium-149, europium-151, gadolinium, gadolinium-155, gadolinium-157, and iridium) are recommended for further consideration in actual space reactor designs.     

Thermal striping limits for components of sodium cooled fast spectrum reactors

With the objective of establishing thermal striping limits for future sodium cooled fast spectrum reactors (SFR), a fracture mechanics-based method employing ‘σ-d approach’ recommended in RCC-MR: Appendix A16 has been followed. Towards this, an idealized geometry, thermal fluctuations in the form of constant power spectral density and pessimistic material data were considered and temperature and thermal stresses are computed taking in to account frequency-dependent thermal attenuation on the structural wall. The effect of attenuation is found to be significant. The limits are derived at various potential locations in SFRs, which are also subjected to creep-fatigue damage due to major cycles caused by startup, shutdown, power failures and pump trips, etc. The maximum range of temperature fluctuations can be as high as 70 K where there is practically no accumulated creep-fatigue damage and 45 K is acceptable where the creep-fatigue is significant (0.9). These limits are found to be consistent with the broad limits extrapolated from the failure experiences of international SFRs and sodium facilities. Pool hydraulic computations carried out to identify and quantify the thermal striping zones confirmed that the proposed limits can be respected with good margins for SFRs.     

Electromagnetic flowmeters for fast reactors

This paper is in four parts. Section 1 explains the theory of the induced-voltage electromagnetic flowmeter and then considers various types which have been used. For the primary circuit of fast reactors both flow-through type and probe type have been proposed, although obtaining magnets which operate satisfactorily at high temperatures has been a problem. In the secondary circuit the high magnetic Reynolds numbers cause the field to be swept out of the magnet gap and this has led to the use of the long saddle-coil flowmeter.In Section 2 flux-distortion flowmeters are described. These have been proposed mainly for monitoring the primary circuit flow and again both flow-through and probe types have been tested. Sections 3 and 4 continue the discussion of the flux-distortion flowmeter by introducing two methods of analysing its performance. The first is a finite difference method which solves the non-linear problem by using a time marching method. It is shown that a linear approximation is adequate for the likely levels of flow encountered in the fast reactor and consequently two linearised solutions are used. The first method is a finite difference one and allows the instantaneous response of a step change in velocity to be observed as well as the effect of bubbles.In Part 4 the second linearized method uses current rings to divide up the conducting material. By considering the interaction of all the rings, it is possible to obtain the current distribution and hence the magnetic field. In conclusion it is suggested that further development would be useful of the devices which are most suited to the liquid metal fast breeder reactor.     

Testing of programs for substantiation of the nuclear safety of space reactors

Russian Scientific Center “Kurchatovskii Institut.” Translated from Atomnaya énergiya, Vol. 77, No. 5, pp. 326–329, November, 1994.     

A design study on MOX-fueled small fast reactors for standardization of a small fast nuclear reactor system

A way of development to standardize a small fast nuclear reactor system, which is considered one of the suitable concepts at next generation for satisfying such needs as generality, small dependence on natural resources, safety and non-proliferation, is proposed. This process consists of three steps : the first is to demonstrate the basic system within a short period based on current techniques, the second is to achieve greatly higher economy, and the final is to standardize the commercial system that can economically compete with or overcome current light water reactors. A technical investigation is conducted on the performance of a mixed-oxide (MOX)-fueled small fast reactor with a reflector-driven reactivity control system to satisfy the needs at the first step, considering plenty of accomplishments on the MOX fuel and its advantage for limiting the duration of development to the level required at the stage. The results obtained from a series of neutronic and thermal-hydraulic calculations show the feasibility of a small fast reactor that produces the electric power of about 50MW, achieves about two-year consecutive operation with high safety performance and is greatly flexible for updating the system. A mixed-nitride-fueled core is found to be promising past the first stage.     

Methods for determining operation life and reactivity depletion for space reactors with fast energy spectra

Space reactors with fast neutron energy spectrums are preferred for their compactness and high fission power density, but require a high fissile inventory. The operation life estimates of these reactors are important to mission planning. This paper examines a number of fuel depletion and neutronics code packages for determining the operation lives of two space reactors with hard fast neutron energy spectra. These are: the lithium-cooled, Sectored, Compact Reactor (SCoRe-S₁₁), and the submersion subcritical safe space (S^4) reactor, cooled with a He–Xe binary gas mixture (40 g/mol). This work investigated the code packages of Monteburns 2.0, MCNPX 2.6C and TRITON and validated their prediction with fuel depletion data for a PWR fuel bundle, with satisfactory results. The operation life predictions of the two space reactors using these code packages are compared with those calculated using a simplified method that couples MCNP5 to a burnup analysis model using the Simulink^® platform. This method considers only the 10 most probable low-Z and high-Z elements of the fission yield peaks plus ¹⁴⁹Sm, and neglects the depletion of fission products due to capture and radioactive decay. The simplified method requires significantly shorter running time and its predictions of the operation lives for the two space reactors are within 0.29–12.5% of those obtained using Monteburns 2.0 and MCNPX 2.6C code packages. This method, however, is not recommended for operation life predictions for space or commercial reactors with thermal neutron spectrums.     

Nuclear data requirements for fast reactors

Conclusions Summarizing, we emphasize again that precise knowledge of estimated microscopic neutron data is necessary in order to ensure the required accuracy in calculating a number of parameters which are important for the reactor operation and decisions on promising reactor concepts, for the technology of the external fuel cycle, including transportation, reprocessing of spent fuel, and the manufacture of new fuel elements, and for investigating the stability of fuel elements in relation to the neutron flux in the reactor. These requirements have not yet been satisfied, in spite of the progress made in the field of experiments and evaluations.It should also be emphasized that, for effective progress toward the target accuracy of reactor parameters, it is necessary to determine, the correlative properties of allowable errors, which, unfortunately, has not yet been done in the WRENDA international list of requirements.DeceasedTranslated from Atomnaya Énergiya, Vol. 57, No. 4, pp. 234–241, October, 1984.     

International cooperation on control for safe operation of nuclear research reactors

The purpose of this paper is to present some of the main reasons, objectives, planning, and recent advances of an SCK·CEN/ININ joint project, which deals with the design and application of modern/expert control and real-time simulation techniques for the safe operation of a TRIGA Mark III research nuclear reactor. This project has been proposed as the first of its kind under a general cooperation agreement between the Belgian Nuclear Research Centre (SCK·CEN) and the National Nuclear Research Institute (ININ) of Mexico.     

Optimal control algorithms for space-based nuclear power stations with thermal reactors

Krasnaya Zvezda Scientific Production Union. Translated from Atomnaya Énergiya, Vol. 71, No. 5, pp. 386–391, November, 1991.