A MESOSCOPIC OXIDE FUEL CLUSTERING AND ITS GLOBAL PERFORMANCE

Transient fuel behavior in a Light Water-cooled Reactor core depends on nuclear properties (Doppler broadening, moderation ratio, and, sometimes, neutron gas temperature etc.) and on variations of thermal-physics parameters (temperature distributions, fuel elongation and moderator density). Usually, in a rough reactor analysis one ignores the very details of temperature distributions largely staying in a frame of so-called adiabatic assumptions (when temperature and density distribution are changing in sync keeping given spatial shapes). In majority of practical applications the radially distributed temperature fields are represented as monotonically smeared ones as if fissile and other materials are homogeneously mixed. Moreover, no one measurement technique allows counting precise correlation between reactivity feedback and in-pellet temperature and materials space-time distributions. However, if fuel is made of Mixed Oxide Plutonium-Uranium compound the behavior of Light Water Reactor would be impacted by an appearance of Pu-rich agglomerates that could be large enough to change physical processes. In such case the fuel reacts on power and temperature variations no more as a homogeneous but a heterogeneous media (on a mesoscopic scale, of course). It leads to changes in a fission product distributions, a fission gas release and, even, to an appearance of multiple components in a Fuel Temperature Coefficient and in a Power Reactivity feedback. These components would depend non-linearly on power, power rate and on some details of a heat transfer. This paper is the only first step of a broad research program where we are estimating the relevant phenomena just by an order of magnitude.

Coupling Dependence of Multiple Operating Parameters on Burnup Credit Calculations for BWR Spent Fuel Assemblies

This study aims to investigate the coupling dependence resulting from three and four operating parameters for burnup credit calculations in boiling water reactor (BWR) spent fuel assemblies. Four operating parameters are under investigation, including fuel temperature, axial burnup profile, axial moderator density profile and control blade usage. In this study, the effects of variation on the curve of effective multiplication factor (keff) versus burnup (B) resulting from one and multiple operating parameters were defined as the single and compound effects, respectively. Particularly, the compound effects adopt more practical operating parameters than single effects does and thus affect the precise assessment to some extent. In our previous study, the compound effects resulting from two operating parameters were investigated in depth. However, the influence of compound effects resulting from three and four operating parameters on burnup credit calculation is still unknown. Therefore, this constitutes the purpose of this study. All the calculations were performed using SCALE 6.1 computer code together with the ENDF/B-VII 238 energy group neutron data library. Two geometrical models were established to represent the typical GE14 10 × 10 BWR fuel assembly and the GBC-68 storage cask. The results revealed that the reactivity deviation (or changes of keff, Δk) resulting from the compound effects was not a summation of the Δk’s resulting from the associated single effects. Moreover, such Δk discrepancies increase as B increases. In this study, the curves of keff versus B due to single and compound effects were approximated by a second degree polynomial of B. A general formula was thus proposed to express these curves.

ANALYSIS OF REACTIVITY COEFFICIENT CHANGE DUE TO BURN UP IN AP1000 REACTOR CORE USING NODAL3

One of the important things in reactor safety is the value of inherent safety parameter namely reactivity coefficient. These inherent safety parameters are fuel and moderator temperature coefficients of reactivity. The objective of the study is to obtain the change of those reactivity coefficients as a function of fuel burn up during the cycle operation of AP 1000 reactor core. Fuel and moderator temperature coefficients of reactivity and in addition moderator density coefficient of reactivity were calculated using SRAC 2006 and NODAL3 computer codes. Cross section generation of all core material was done by SRAC 2006 Code. The calculation of core reactivity as a function of temperature and burn up were carried out using NODAL3 Code. The results show that all reactivity coefficients of AP 1000 reactor core are always negative during the operation cycles and the values are in a good agreement to the design. It can be concluded that the AP 1000 core has a good inherent safety of its fuelKeywords: reactivity coefficient, burn up, AP1000, NODAL3. ANALISIS PERUBAHAN KOEFISIEN REAKTIVITAS AKIBAT FRAKSI BAKAR TERAS REAKTOR AP1000 MENGGUNAKAN NODAL3. Salah satu hal yang sangat penting dalam analisis kecelakaan pada reactor daya adalah koefisien reaktivitas untuk mengontrol daya reaktor. Penelitian ini bertujuan menentukan koefisien reaktivitas akibat perubahan fraksi bakar pada reaktor AP1000. Koefisien reaktivitas yang akan dihitung adalah koefisien reaktivitas bahan bakar dan moderator yang sering disebut inherent factor. Selain itu juga akan dihitung koefisien konsentrasi boron dan kerapatan moderator. Semua koefisien reaktivitas ini dihitung saat terjadi perubahan fraksi bakar untuk mempertimbangkan produk fisi dan konsumsi bahan bakar. Perhitungan neutronik teras reactor disimulasi dengan menggunakan program SRAC2006 dan NODAL3. Perhitungan tampang lintang seluruh perangkat bahan bakar dan batang kendali reaktor AP1000 dilakukan dengan program SRAC2006. Perhitungan parameter neutronik sebagai fungsi temperature dan fraksi bakar dilakukan menggunakan program NODAL3. Perhitungan koefisien reaktivitas ditentukan berdasarkan perbedaan nilai reaktivitas. Hasil perhitungan menunjukkan bahwa koefisien reaktivitas teras reaktor AP 1000 selalu berharga negative untuk sepanjang siklus operasinya dan mendekati harga desain. Kesimpulan yang dapat ditarik adalah bahwa teras AP 10000 mempunyai keselamatan melekat yang baik.Kata kunci: koefisien reaktivitas, fraksi bakar, AP 1000, NODAL3.

Parametric Study of the Burnup Credit Criticality Analysis in BWR Fuels

The purpose of this study is to investigate the effects of operation parameters on BWR fuel reactivity and nuclide inventory. The three high-priority parameters that are suggested by NRC/ORNL [1], that is, axial burnup profile, axial moderator density profile, and control blade insertion during operation, were examined in order to evaluate their impact on BWR burnup credit. All of the calculations were performed using SCALE6.1 based on the ENDF/B-VII nuclear data library and involved 238 energy groups. The results indicate that the axial burnup profile is the most critical factor in reactivity when the burnup value is greater than 10 GWd/MTU. The insertion of control blades during operation is also more important for the reactivity of spent fuel as compared to the influence of the axial moderator density profile. At a burnup of 60 GWd/MTU, these three operation parameters (i.e., axial burnup profile, axial moderator density profile, and control blade insertion during operation) in turn lead to multiplication factor differences of 18850, 7600, and 1980 pcm, respectively, for set2 nuclides. The influence between different parameters was also investigated in this study. The results reveal that multiple operation parameters affect the reactivity reduction tendency. In addition to the three parameters investigated in this study, other important factors related to burnup credit analysis are also currently under investigation.

Study on Thermal Characteristics of Water-Rod in SCWR

Water-rod is introduced into Supercritical Water Cooled Reactor (SCWR) to guarantee the homogeneous distribution of core axial power, and there could be many types of flow in water-rod. Based on three types of water-rod and flow, which named S type, D1 type, and D2 type, choosing different values of structural parameters and material thermal conductivity, the distributions of moderator density and average density in each case are obtained through making program to associated-calculate the fuel rod, coolant and moderator. The results indicate that considering only moderator density, the D2 type double water-rod which has smaller thermal conductivity and bigger outer length of inner water-rod is the best choice; considering average density, both the S type single water-rod and the D2 type double water-rod which have smaller thermal conductivity and smaller outer length of inner water-rod is the best choice; the thermal performance of D1 type double water-rod is inferior because of its inherent defect, so it is not suitable for use.

Modeling the Spectral History in the Depletion of a PWR Core

Codes for reactor core calculations use few-group cross sections (XS) which depend on local burnup, given in terms of the energy produced per fuel mass (MWd/kgHM). However, a certain burnup value can be reached under different spectral conditions depending on moderator density and other local parameters. Neglecting these spectral effects, i.e. applying the summary-burnup value only, can cause considerable errors in the calculated power density. This paper describes a way to take into account spectral-history effects. It is shown that the respective XS correction linearly depends on the actual Pu-239 concentration. The applicability of the method was proved not only for usual uranium oxide fuel, but also for mixed uranium/plutonium oxide (MOX) and fuel assemblies with burnable absorber. The code DYN3D was extended by new subroutines which calculate the actual distribution of Pu-239 in the core and apply a spectral-history correction for the XS.

RCS Pressure Predictions Under ATWS for Maanshan With MUR

The reactor coolant system (RCS) pressure transients under ATWS for Maanshan PWR with MUR conditions were analyzed in this study. The main concern was to observe whether the peak RCS pressure still maintained within 3200 psig limit (defined by ASME Code Level C service limit criterion under 100% rated power). The Maanshan RETRAN model, which had been developed for PWR transients, was used for ATWS safety analysis and related sensitivity analysis under MUR conditions. In document SECY-83-293, all initializing events were classified as either turbine trip or non-turbine trip events and their ATWS risks were also evaluated according to these two events. Loss of condenser vacuum (LOCV) and Loss of normal feedwater (LONF) ATWS were identified as limiting transients of turbine trip and non-turbine trip events in this study. In order to analyze the RCS pressure variations, a moderator density coefficient (MDC) model was included to observe the influences of moderator temperature and system pressure to the reactor reactivity, and the possibility of the peak RCS pressure reaching its limit. In addition, this report also covers the following sensitivity studies, (1) different MTC settings, and (2) different initial conditions of steam generator outlet pressure. The peak RCS pressure was found to stay within the criterion of 3200 psig even at a conservative MTC setting of −4 pcm/°F. Furthermore, a lower steam generator outlet pressure also resulted in a lower peak RCS pressure.

Influence of Low Boron Core Design on PWR Transient Behavior

In conventional pressurized water reactor (PWR) designs, the concentration of boron in primary coolant is limited by the requirement of having a negative moderator density coefficient. As high boron concentrations have significant impact on reactivity feedback properties, design changes to reduce boron concentration in the reactor coolant are of general interest in view of improving PWR inherent safety. In the framework of an investigation into the feasibility of low boron design, a PWR core configuration based on fuel with higher gadolinium (Gd) content has been developed which permits to reduce the natural boron concentration at begin of cycle (BOC) by approx. 50% compared to current German PWR technology. For the assessment of the potential safety advantages, a Loss-of-Feedwater Anticipated Transient Without Scram (ATWS LOFW) has been simulated with the system code ATHLET for two PWR core designs: a low boron design and a standard core design. The most significant difference in the transient performance of both designs is the total primary fluid mass released through the pressurizer (PRZ) valves. It is reduced by a factor of four for the low boron reactor, indicating its improved density reactivity feedback.