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In the reactivity measurements of the China Experimental Fast Reactor (CEFR) start-up tests, five independent control rods were moved to compensate for the reactivity insertion. Since the recorded control rod positions were converted to reactivity via S-curves (rod worth curves), any uncertainty in the S-curves can propagate to all reactivity measurements. In this work, we rigorously derived the uncertainty of the reactivity in terms of the statistical uncertainty of the S-curves with Monte Carlo (MC) simulations. Additionally, the average error of the control rod worth from the MC calculation was estimated from experiments and embedded into the uncertainty formulation. The formulation shows that the error of the reactivity is highly correlated to the changes in the control rod position during the reactivity measurements. McCARD MC simulations were then conducted for the CEFR start-up tests, and the calculated reactivity and uncertainty were compared with the measurements. The main error factor of each reactivity calculation was figured out by quantifying the uncertainty components. With the uncertainty formulation, the calculation results showed a better agreement compared with the measurements, as the relative errors were observed mostly within 2 σ of the uncertainty. [ABSTRACT FROM AUTHOR]

Abstract: One of the lead-alloy cooled fast reactors, PEACER (Proliferation-resistant, Environment-friendly, Accident-tolerant, Continuable-energy, and Economical Reactor), has been designed by a university research team in Korea. This reactor is a medium-size modular fast reactor of 850MWt (300MWe) rated power with lead–bismuth coolant loop. A square lattice fuel assembly with large pitch size might be favorable to ensure sufficient mass flow rate not only in normal operation but also in natural cooling under accidents. A metal fuel of (U,TRU)10%Zr is chosen in this study for the advantage of high thermal conductivity. Multinational energy complex with a closed fuel cycle facilities should provide proliferation resistance with an institutional barrier. Under the material control safeguards, all fuel isotopes and radioactive material should be confined in a complex but as low-level radioactive waste. In order to achieve this goal, efficient transmutation of long-lived minor actinides (LLMAs) and long-lived fission products (LLFPs) should be completed in an LFR core. An energy complex, PEACER-Park, consists of four PEACERs and two collocated pyro-processing plants. Only low-level waste from spent fuel reprocessing plant is designed to be transported out of PEACER-Park. Recovery factors in reprocessing plant, therefore, should be high enough to make low-level tail satisfy US NRC class C classification. However, an interim storage should be collocated in PEACER-Park for storing radioactive fission products having short half-life such as Cs-135 and Sr-90. Parametric studies have been done for a core design to maximize its transmutation capability within safety criteria. [■] When a pitch to diameter ratio (P/D ratio) was increased, the transmutation capability of each LLMA isotope changed in different ways. [■] When neutron leakage was increased by making core shape flat, the transmutation enhanced. [■] It was confirmed that longer cycle length was favorable to transmutation. [■] In the case of feed fuel composition and coolant operation temperature, there were no significant effects on transmutation performance. The final optimized core was designed as follows: a flat core shape with 50cm of active core height and 5m core diameter, pitch to diameter ratio of 2.2 which could be compromised between different transmutation tendencies of Np-237 and Am-241 and three different enrichment zones with 1-year cycle length. This design could achieve a sufficient transmutation capability with a support ratio of 2.085. Maximum pin power peaking was less than 1.5. Large negative radial expansion feedback coefficient and a sufficient shutdown margin could also be guaranteed. [Copyright &y& Elsevier]

Highlights • Assessment of detectors' locations and number of detectors in PGSFR. • Generation of spatial weighting function using forward fixed-source calculation in MCNPX. • Calculation of transient power during withdrawal of a single control assembly by using FREK. Abstract This paper presents the assessment of the ex-core detectors in the Prototype Gen-IV Sodium-cooled Fast Reactor (PGSFR) in response to the movement of single primary control assembly, to evaluate the detectors' locations and to determine a proper number of detectors. The detectors are located below the containment vessel and distributed in different ranges from the center and different azimuthal angles. Another position considered is in the radial direction of the core, at the sodium coolant gap between the outermost assembly and core shroud. To calculate the detection rate, the contribution of each fuel region to the detector and the core power distribution are required. Initially, the aforementioned contribution is generated by performing a continuous-energy forward fixed-source Monte Carlo calculation using MCNPX. The reactor kinetics code FREK simulates the power change due to an inadvertent withdrawal of a primary control assembly up to 10% from its critical position followed by full insertion of secondary control assemblies. The response from each detector location was evaluated, and consequently, the detectors located below the containment vessel, which can "see" all the fuel assemblies, responded better to the core power increase with a minor error of less than ∼0.6%. In addition, the impact of locating the detector below the containment vessel at different locations from the center or at different azimuthal angles was insignificant but several detectors should be distributed at even azimuthal angles for effective core monitoring and redundancy purposes. [ABSTRACT FROM AUTHOR]

For a metal-fueled SFR (Sodium-cooled Fast Reactor), the negative feedback of the reactivity due to the thermal expansion of fuel has been a principal safety mechanism to guarantee passive reactor shutdown after undercooling accidents. In this paper, an experimental model of the fuel axial expansion reactivity in a metal-fueled SFR was proposed and measured values using the BFS-109-2A critical assembly was compared with calculated values using as-built MCNP models based on the ENDF/B-VII.0 library. The energy-dependent axial expansion reactivity components of the BFS-109-2A model showed good agreement with those of the target core. The difference in the calculation/experiment ratio between the as-built MCNP models and measurements vary from −10.2% to 2.6% for the axial expansion reactivity, meanwhile the suggested axial expansion model showed a 25% discrepancy comparing to the target core. [ABSTRACT FROM AUTHOR]

• An experimental model of the radial expansion reactivity in a SFR was proposed. • The model reflects well the spectral component of the radial expansion reactivity. • Measured reactivities during experiment were compared with calculated values. • The difference in the calculation/experiment ratio vary from −4.3% to 9.2%. For a metal-fueled SFR (Sodium-cooled Fast Reactor), the negative reactivity feedback due to the thermal expansion of the support grid which is known as a radial expansion reactivity, provides a principal safety mechanism in most severe accidents. In this paper, an experimental model of the radial expansion reactivity in a metal-fueled SFR is proposed, and experimental models are established in a collaboration between Korea Atomic Energy Research Institute (KAERI) and Institute for Physics and Power Engineering (IPPE). Modeling of the radial expansion reactivity in a realistic experiment condition is very difficult due to physical limitation of experimental facility. Hence, a sparse perturbation method suggested by authors was employed for reasonable modeling of the radial expansion phenomenon. Measured values were compared with calculated ones using the MCNP transport code with the ENDF/B-VII.0 library. The energy-dependent radial expansion reactivity components of the planned experimental model showed good agreement with those of the target Prototype Gen IV SFR core. Errors of C/M (calculation/measurement ratios) varied from −4.3% to 9.2%. Considering the 1 σ uncertainty in the C/M, 5% in maximum, errors of calculation/measurement ratios were within 2 σ uncertainty range. [ABSTRACT FROM AUTHOR]