A practical procedure of bilinear weighted core kinetics parameters computation for the purpose of experimental reactivity determination

A practical procedure for the computation of bilinear weighted core kinetics parameters of WWER-440 reactors, intended to be used in experimental reactivity determination by the inverse kinetics method is described. The results from its application are benchmarked against those obtained on the basis of a many-group fine-mesh two-dimensional solution of the core boundary problem. The sensitivity of core kinetics parameters and of the resultant experimental reactivity values to the variation of primary delayed neutron data is evaluated.     

A feasibility study of LEU enrichment uranium fuels for MNSR conversion using MCNP

A neutronics feasibility study has been performed to determine the enrichment that would be required to convert a commercial Miniature Neutron Source Reactor (MNSR) from HEU (90.2%) to LEU (<20%) fuel. Two LEU cores with uranium oxide fuel pins of different dimensions were studied. The one has the same dimensions as the current HEU fuel while the other has the dimensions as the special MNSR, the In-Hospital Neutron Irradiator (INHI), which is a variant of the MNSR. The LEU cores that were studied are of identical core configuration as the current HEU core, except for potential changes in the design of the fuel pins. The following reactor core physics parameters were computed for the two LEU fuel options; clean cold core excess reactivity (ρ_ex), control rod (CR) worth, shut down margin (SDM), neutron flux distributions in the irradiation channels and kinetics data (i.e. effective delayed neutron fraction, β_eff and prompt neutron lifetime, l_f). Results obtained are compared with current HEU core and indicate that it would be feasible to use any of the LEU options for the conversion of NIRR-1 in particular from HEU to LEU.     

Determination of neutron generation time in miniature neutron source reactor by measurement of neutronics transfer function

The prompt neutron generation time Λ and the total effective fraction of delayed neutrons (including the effect of photoneutrons) β have been experimentally determined for the miniature neutron source reactor (MNSR) of Syria. The neutron generation time was found by taking measurements of the reactor open-loop transfer function using newly devised reactivity-step- ejection method by the reactor pneumatic rabbit system. Small reactivity perturbations i.e. step changes of reactivity starting from steady state, were introduced into the reactor during operation at low power level i.e. zero-power. Relative neutron flux and reactivity versus time were obtained. Using transfer function analysis as well as least square fitting techniques and measuring the delayed neutrons fraction, the neutron generation time was determined to be 74.6±1.57 μs. Using the prompt jump approximation of neutron flux, the total effective fraction of delayed neutrons was measured and found to be 0.00783±0.00017. Measured values of Λ and β were found to be very consistent with calculated ones reported in the Safety Analysis Report.     

Behavior of mechanical properties of nickel-alloyed reactor pressure vessel steel under neutron irradiation and post-irradiation annealing

The effect of neutron irradiation and post-irradiation thermal annealing on tensile and impact properties of Cr–Ni–Mo steel used for WWER-1000 reactor pressure vessel (RPV) manufacturing was studied. A gap in yield stress and ultimate tensile stress fluence dependence at the fluence range of 0–3×10²³ neutrons m⁻² was observed while ductile-to-brittle transition temperature (DBTT) was continuously increasing with damage dose. The post-irradiation annealing recovery of tensile properties was found to be higher than the one of impact properties. Over-recovery of tensile properties due to 460 and 490°C post-irradiation annealings were observed. The annealing effectiveness of WWER-440 and WWER-1000 grades was compared. Nickel was supposed to affect both the radiation sensitivity and the post-irradiation residual DBTT shift of WWER-1000 type steel.     

New analysis of prompt supercritical process with temperature feedback

The prompt supercritical process of a nuclear reactor with temperature feedback and initial power while inserting large step reactivity (ρ₀ > β) is analyzed. The new analytic expressions of reactivity and output power as well as reactor's temperature increase during the course of the prompt supercritical are derived. The maximal power and the related reactivity and time are obtained. The effects of the inserted step reactivity and initial power on the prompt supercritical process are analyzed and discussed. It is found that for considering only the effect of prompt neutron the reactivity, power and temperature vary much more quickly and the power peak value and the increase values of temperature are larger than those for considering the effect of both prompt neutron and delayed neutron.     

Validation of Monte Carlo predictions of LWR-PROTEUS safety parameters using an improved whole-reactor model

The recent experimental programme conducted in the PROTEUS research reactor at the Paul Scherrer Institute (PSI) has concerned detailed investigations of advanced light water reactor (LWR) fuels. More than fifteen different configurations of the multi-zone critical facility have been studied, each of them requiring accurate estimation of operational safety parameters, in particular the critical driver loadings, shutdown rod worths and the effective delayed neutron fraction β_eff. The current paper presents a full-scale 3D Monte Carlo model for the facility, set up using the MCNPX code, which has been employed for calculation of the operational characteristics for seven different LWR-PROTEUS configurations. Thereby, a variety of nuclear data libraries (viz. ENDF/B6v2, ENDF/B6v8, JEF2.2, JEFF3.0, JEFF3.1, JENDL3.2, and JENDL3.3) have been used, and predictions of k_eff and shutdown rod worths compared with experimental values. Even though certain library-specific trends have been observed, the k_eff predictions are generally very satisfactory, viz. with discrepancies of <0.5% between calculation (C) and experiment (E). The results also confirm the consistent determination of reactivity variations, the C/E values for the shutdown (safety) rod worths being always within 5% of unity. In addition, the MCNP modelling of the multi-zone reactor has yielded interesting results for the delayed neutron fraction (β_eff) in the different configurations, a breakdown being made possible in each case in terms of delayed neutron group, fissioning nuclide, and reactor region.     

固定棒位法测量控制棒总价值

控制棒价值测量的准确度与效率对核电厂的安全性与经济性具有重要影响。在动态刻棒等反应性测量工作中,本底与中子源对探测器有显著影响,致使根据实测电流计算得到的反应性显著偏离真实值。基于点堆逆动态方程,通过对本底与中子源影响的分析,利用固定棒位状态下的测量数据计算反应性并得到控制棒总价值,给出了一种不受本底与中子源影响的简便的控制棒总价值测量计算方法,并在零功率实验装置上进行验证。结果表明,该方法可有效避免本底和中子源组件对反应性探测的影响,并简化了离线理论计算,其与周期法计算结果的相对偏差在1%以内。     

Consolidation of scientific and technological expertise to assess the reliability of reactor pressure vessel embrittlement prediction in particular for the arctic area plant (COBRA)

The evaluation and prognosis of reactor pressure vessel (RPV) material embrittlement in WWERs and the allowable period of their safe operation are performed on the basis of impact test results of irradiated surveillance specimens. The main problem concerns the irradiation conditions (irradiation temperature, neutron flux and neutron spectrum) of the surveillance specimens that have not been determined yet with the necessary accuracy. These conditions could differ from the actual RPV wall condition. In particular, the key issue is the possible difference between the irradiation temperature of the surveillance specimens and the actual RPV wall temperature. It is recognized that the direct measurement of the irradiation temperature by thermocouples during reactor operation is the only way to obtain reliable information. In addition, the neutron field's parameters in the surveillance specimens location have not been determined yet with the necessary accuracy. The use of state of the art dosimeters can provide high accuracy in the determination of the neutron exposure level.The COBRA project, which started in August 2000 and had a duration of 3 years, was designed to solve the above-mentioned problems. Surveillance capsules were manufactured which contained state of art dosimeters and temperature monitors (melting alloys). In addition, thermocouples were installed throughout the instrumentation channels of the vessel head to measure directly the irradiation temperature in the surveillance position during reactor operation. The selected reactor for the experiment was the Unit 3 of Kola NPP situated in the arctic area of Russia. Irradiation of capsules and online temperature measurements were performed during one fuel cycle. On the base of statistical processing of thermocouples readings, the temperature of irradiated surveillance specimens in WWER-440/213 reactor can be accepted as 269.5 ± 4 °C. Uncertainties were evaluated also with experimental work carried out in the WWRSZ research reactor and by finite element modelling of surveillance capsules. The results obtained show that there is not need to perform temperature correction when surveillance data of irradiated specimens are used for embrittlement assessment of WWER-440(213) reactor pressure vessels. Maximum neutron flux evaluated using detectors, which were placed in the Charpy specimen simulators, equals 2.7 × 10¹² cm⁻² s⁻¹ with E > 0.5 MeV. It is established that depending on the orientation of the capsules with respect to the core, the detectors of the standard surveillance capsules can give both overestimated and underestimated neutron flux values, as compared to the actual flux received by the surveillance specimens. The overestimation or underestimation can reach 10%.     

Void reactivity evaluation by modified conversion ratio measurements in LWR critical experiments

We have developed a void reactivity evaluation method by using modified conversion ratio measurements in a light water reactor (LWR) critical lattice. Assembly-wise void reactivity is evaluated from the “finite neutron multiplication factor”, k*, deduced from the modified conversion ratio of each fuel rod. The distributions of modified conversion ratio and k* on a reduced-moderation LWR lattice, for which the improvement of negative void reactivity is a serious issue, were measured. Measured values were analyzed with a continuous-energy Monte Carlo method. The measurements and analyses agreed within the measurement uncertainty.

The developed method is useful for validating the nuclear design methodology concerning void reactivity.     

Prediction of irradiation embrittlement of vanadium alloyed low nickel steel for future reactors

A comparison between pearlitic 2CrMoV steel (WWER-440) and 9% Cr based ferritic-martensitic steels (EUROFER 97 and LA12TaLC) is presented as regards irradiation induced ductile-brittle transition temperature shifts. For neutron doses of 1.5-2 dpa and irradiation temperatures around 300 °C the transition temperature shifts for WWER-440 steel and EUROFER 97 welds are comparable. In the temperature range 350-500 °C the radiation embrittlement levels of both steels are low. Moreover, post-irradiation annealing is proposed as a promising method to predict results of high temperature irradiation embrittlement from existing lower temperature irradiation embrittlement data.     

Prompt alpha and reactivity measurements on fast metal assemblies

This paper summarizes and reviews the methods of reactivity determination and measurement of the prompt-neutron decay, briefly describes the equipment requirements for such measurements for unmoderated metal assemblies, and presents experimental results to illustrate the methods. These assemblies include fast metal critical assemblies and fast pulsed reactors. The primary reactivity determination methods used have been: (1) stable reactor period measurements which are usually used near delayed criticality to obtain the reactivity or to calibrate the reactivity prior to burst initiation; (2) prompt reactor period measurements which are useful to determine the reactivity early in superprompt critical excursions; (3) inverse kinetics rod drop measurements which obtain the reactivity as a function of time after a rod or reactor component is removed from a delayed critical assembly, and (4) prompt neutron decay constant measurements from which the reactivity can be obtained if corrections are made for changes in the neutron lifetime. Inverse kinetics and decay constant measurements are usually used below delayed criticality, although decay constant measurements have been performed above delayed critical.The decay constant is usually obtained by the traditional pulsed-neutron method, using a pulsed neutron source such as a Cockcroft-Walton accelerator, or by the Rossi-α method. The recent use of ²⁵²Cf has resulted in some new techniques for determining the decay constant and reactivity; this method has some unique advantages over the traditional methods. The theory of these new measurements is reviewed, and some recent results are presented.     

Simulation of start-up measurements with the KIKO3D code and its application for determining the scram rod worth uncertainty in the VVER-440 reactor

The reasons for large discrepancies between the computed and measured values of the efficiency of control rods observed during start-up experiments on the Russian pressurized water type VVER reactors are discussed. The numerical simulation of the measurements including the prediction of the ex-core detector signals was used to resolve the discrepancies. The time and space dependent neutron flux in the core during these measurements have been calculated by the KIKO3D nodal kinetic code. For calculating the ionization chamber signals the Green function technique has been applied. The Green functions of ionization chambers have been evaluated via solving the neutron transport equation in the reflector regions with the MCNP Monte Carlo code. The detector signals have been calculated and compared with measured ones using the inverse point kinetics transformation. Large number of asymmetric rod drop measurements (with one rod stuck) and some differential rod worth measurements from the Zero Power Physics Tests were provided by the Paks NPP for validation. The experiments cover different fuels (without and with enrichment zoning) and loading patterns. The intermediate range ionization chambers have been used during the scram measurements. The newly developed method provides fairly sufficient match of measured and calculated results. The time behavior of the detector readings observed in the measurements are described by the code in a consistent manner.As a further application the uncertainty of scram rod worth of the KARATE-440 code system was determined by static calculations and subsequent simulation of rod drop with the KIKO3D code. The calculated results were compared to measurements carried out by the Paks NPP. The uncertainty of scram rod worth is established by statistical analysis.     

Modeling of WWER-440 fuel pin behavior at extended burn-up

Currently, there is an ongoing effort to increase fuel discharge burn-up of all LWRs fuel including WWERs as much as possible in order to decrease power production cost. Therefore, burn-up is expected to be increased from 60 to 70 MWd/kg U. The change in the fuel radial power distribution as a function of fuel burn-up can affect the radial fuel temperature distribution as well as the fuel microstructure in the fuel pellet rim. Both of these features, commonly termed the “rim effect.” High burn-up phenomena in WWER-440 UO₂ fuel pin, which are important for fission gas release (FGR) were modeled. The radial burn-up as a function of the pellet radius and enrichment has to be known to determine the local thermal conductivity.In this paper, the radial burn-up and fissile products distributions of WWER-440 UO₂ fuel pin were evaluated using MCNP4B and ORIGEN2 codes. The impact of the thermal conductivity on predicted FGR calculations is needed. For the analysis, a typical WWER-440 fuel pin and surrounding water moderator are considered in a hexagonal pin well. The thermal release and the athermal release from the pellet rim were modeled separately. The fraction of the rim structure and the excessive porosity in the rim structure in isothermal irradiation as a function of the fuel burn-up was predicted. A computer program; RIMSC-01, is developed to perform the required FGR calculations. Finally, the relevant phenomena and the corresponding models together with their validation are presented.     

A measurement of the Li(n,α) cross-section

The relative shape of the ⁶Li(n,α)/²³⁵U(n,f) cross-section ratio has been determined in the range of incident neutron energy from 2 to 800 keV. The measurements were made by the time-of-flight method using the Harwell 45 MeV linac to provide the pulsed source of neutrons. A thin ⁶Li-glass scintillator was used to register the (n,α) events, and the (n,f) events (in a metallic sample of ²³⁵U) were registered with fission neutron detectors. The shape of the ⁶Li(n,α) cross-section was obtained by combining the measured ⁶Li(n,α)/²³⁵U(tn,f) cross-section ratio with an evaluation of the ²³⁵U(n,f) cross-section. The cross-section so derived was placed on an absolute scale by normalization in the neutron energy interval 2–10 keV, where the ⁶Li(n,α) cross-section is accurately known. The cross-section at the peak of the prominent p-wave resonance near 240 keV is found to be 3.29 ± 0.12 b. The results are compared with other measurements and also with a recent theoretical calculation of the cross-section.     

Development of mathematical models and the dual-gauge principle for surface-type neutron moisture content gauges

Mathematical models are developed for the response of surface-type neutron moisture content gauges. Models are developed for both cadmium-covered and bare BF₃-filled proportional counter detectors so that the dual-gauge principle of measurement can be applied. This consists of the simultaneous solution of the two responses to minimize the effect of variations in sample composition.The response models are based on epithermal and thermal line and area flux models, where if flux as a function of radial distance r from the source is φ(r), then line flux is ∫φ(r) dr and area flux is ∫φ(r)r dr. The flux models are obtained from a solution of the three-group diffusion equation for a point source of fast neutrons at the surface of an infinite half-space and are expressed as inverse Hankel transforms. The models are checked by calculation and are verified with Monte Carlo and experimental results.To minimize the effects of variations in sample composition, represented by the weight fraction of thermal neutron absorber, two gauge response models are solved simultaneously for moisture and absorber weight fractions. This application of the dual-gauge principle is tested with experimentally obtained responses on samples of known density and composition. The results of this feasibility study are encouraging in terms of eventual commercial application of the response models and the dual-gauge principle.     

SCALE calculation for evaluation of spent fuel condition during long-term storage

Experiences with an advanced spent nuclear fuel management in Slovakia are presented in this paper. The evaluation and monitoring procedures are based on practices at the Slovak wet interim spent fuel storage facility in NPP Jaslovské Bohunice. Since 1999, leak testing of WWER-440 fuel assemblies were completed using a special leak tightness detection system developed by Framatome-anp, “Sipping in Pool”. This system utilized external heating for the precise defects determination.Optimal methods for spent fuel disposal and monitoring were designed. A new conservative factor for specifying of spent fuel leak tightness is introduced in the paper. Limit values of leak tightness were established from the combination of SCALE4.4a (ORIGEN-ARP) calculations and measurements from the “Sipping in Pool” system. These limit values are: limiting fuel cladding leak tightness coefficient for tight fuel assembly – k_FCT(T) = 3 × 10⁻¹⁰, limiting fuel cladding leak tightness coefficient for fuel assembly with leakage – k_FCT(L) = 8 × 10⁻⁷.     

Stress waves in elastic rods induced by radiation heating

Consider a stress-free, thin, elastic rod, freely suspended. At time t = 0, the rod is rapidly exposed to a high temperature tube field inside, uniform over the cross-section T(x, t), t 0. Such a case arises in an external target when bombarded with a fast extracted proton beam of high energy and high intensity. In contrast to the quasi-static solution σ_x ≡ 0, stress waves are created at both free end-faces, at time t = 0, which propagate into the rod. Closed-form solutions are presented for a fast rise of a homogeneous temperature, during rise time t₀ to the final value T₀. Maximum compressive and tensile stresses will occur which may easily lead to failure of the rod. In particular, for an extremely fast rise t₀ < t_m, t_m being the characteristic mechanical time, |σ_x|max = EαT₀, which is the absolute value of a compressive stress in a rod fixed at the ends. For t₀ > t_m, |σ_x|max = EαT₀t_m/t₀, i.e. the maximum stresses are proportional to the length of the rod and inversely proportional to the rise time.     

A modernized and versatile startup reactivity measuring system installed at NPP Paks and its application for subcritical systems

In 2004 the Hungarian Paks NPP completed a project for upgrading the reactivity measuring system applied during reactor startup experiments. Almost all components of the previous system were replaced, only ex-core ionisation chambers remained unaltered. New hardware and software components were introduced for neutron flux signal handling, for data acquisition, as well as for measurement evaluation and data presentation. High-precision picoamper meters were installed at each reactor unit, current signals are handled by a portable signal processing unit. The system applies an accurate on-line reactivity calculation algorithm based on the point-kinetic model with six delayed neutron groups. Detailed off-line evaluation and analysis of startup measurements can be performed on the portable unit, as well.The paper describes the architecture, data acquisition modules, services and man–machine interface of the new system. Functions and results are illustrated with measured data recorded during a startup of Unit 3. In 2003 and 2004 the RMR was installed and tested at all Paks NPP units successfully and now it is in regular use during unit startups.The second part of the paper illustrates an extension of the new system to perform reactivity measurements using the well-known Rossi-α and Feynman-α statistical methods. The modified system was needed to estimate the reactivity of a subcritical system formed by damaged fuel assemblies stored at the fuel service pit of Paks Unit 2. Theoretical background of the applied algorithms is outlined, then results of validation tests and on site measurements are treated. The measurements have shown that the subcriticality of the damaged fuel was sufficiently deep if the high boron concentration in the fuel service pit was maintained.     

Reactivity feedback coefficients of a material test research reactor fueled with high-density U3Si2 dispersion fuels

The reactivity feedback coefficients of a material test research reactor fueled with high-density U₃Si₂ dispersion fuels were calculated. For this purpose, the low-density LEU fuel of an MTR was replaced with high-density U₃Si₂ LEU fuels currently being developed under the RERTR program. Calculations were carried out to find the fuel temperature reactivity coefficient, moderator temperature reactivity coefficient and moderator density reactivity coefficient. Nuclear reactor analysis codes including WIMS-D4 and CITATION were employed to carry out these calculations. It is observed that the average values of fuel temperature reactivity feedback coefficient, moderator temperature reactivity coefficient and moderator density reactivity coefficient from 20 °C to 100 °C, at the beginning of life, followed the relationships (in units of Δk/k × 10⁻⁵ K⁻¹) −2.116 − 0.118 ρ_U, 0.713 − 37.309/ρ_U and −12.765 − 34.309/ρ_U, respectively for 4.0 ≤ ρ_U (g/cm³) ≤ 6.0.     

The Fort St. Vrain high temperature gas-cooled reactor: I. Low power physics tests

The core was loaded with fuel and reflector elements with subcriticality ensured by temporary neutron absorbers. Pulsed-neutron measurements initially showed the core to be more subcritical than calculated. Adjustment of the calculational model by including the correct reflector impurities led to good agreement between calculated and measured values of the multiplication constant and prompt neutron generation time. Axial fluxes showed consistent good agreement between calculated and measured values for a variety of configurations, but the ratios of average regional fluxes in the top half of the core to their respective regional averages in the bottom half differed from the calculated ratios by as much as 10%. The helium pressure coefficient was essentially zero. The calculated and measured isothermal temperature coefficients up to 300°F agreed within 3%. Control rod calibrations using the reactivity computer were satisfactory only for the central rod pair. Peripheral rod pairs were calibrated by comparison to the central rod pair.