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AbstractThe extension of the operating domain of PreussenElektra’s Konvoi-type pressurized water reactors (PWRs) beyond the natural end of cycle is known as stretch-out operation. A range of possibilities exists to increase nuclear fuel utilization to continue operation after the boron concentration reaches its dilution limit. The most basic option is to continue operation with constant average moderator temperature, which results in a relatively fast decrease in reactor power. From a fuel utilization point of view, this is the least optimal procedure. In PreussenElektra’s PWR fleet, an enhanced operation mode is adopted, leading to a comparatively modest decrease in reactor power and very high utilization of nuclear fuel. Initially, the stretch-out mode provided an option to gain flexibility regarding outage planning. More recently, the stretch-out method has served as a practical approach to optimizing electricity generation costs during the last cycles before the final shutdown as stipulated by law, as operators can extend the cycle length in a range of 30 to 60 days after the natural end of cycle. This paper describes the licensing rationale, the feasibility of this type of operation, and the operating requirements and experience. The system parameters affected by stretch-out operation are discussed. Adjustments of set points of thermal-hydraulic variables in the primary and secondary systems are explained. Licensing requirements for safe reactor operation in stretch-out mode are reviewed. Furthermore, aspects of neutronic and thermal-hydraulic core surveillance are included. After more than 35 years and counting, the methods of increasing fuel utilization are not new, and an evaluation of experience and effectiveness is in order.

Defence in depth is a time-tested strategy in some technical fields, such as cybersecurity or nuclear reactor technology. In view of potentially devasting consequences of exponential growth, the implementation of such a strategy is almost self-evident in this case. Less obvious is a formal explanation of why and when such a control strategy is necessary compared to traditional risk management approaches and which systems or organisations can benefit from it. Quantitative arguments are given based on the properties of heavy tail stochastic variables. Particularly, their single big-jump property provides the rationale to require a hierarchy of barriers for containment of a cascade of daughter events. The COVID-19 pandemic has vividly demonstrated how difficult it is to implement an appropriate control strategy under these circumstances. Concretely, it is shown that the spread of transport vectors in this case has heavy tail properties and consequently requires a defence in depth response.

ABSTRACTDefense in depth is a pillar of nuclear power plant design and it also plays an important factor for the reliability of fuel supply. While most safety systems of a classical light water reactor usually remain unchanged for many years of operation the first and second fission product barrier – the fuel and the cladding – are regularly exchanged by loading fresh fuel assemblies into the reactor core. We review our experience of achieving high reliability in fuel supply by means of real options. The required activities for option creation are explained in detail. Real option valuation is known to be more difficult than financial option valuation and we review the valuation metrics which we found most useful. Due to the incompleteness and illiquidity of the considered option market and due to the barriers of realizing arbitrage, lower and upper bounds are easier to determine than exact option values. Notwithstanding these constraints using concepts based on volatility and stochastic processes in our experience is good practice to avoid naked bets on future power plant availability and avoids the pitfalls of static forecasts.

Fuel assembly (FA) bow in pressurized water reactor (PWR) cores is considered to be a complex process with a large number of influencing mechanisms and several unknowns. Uncertainty and sensitivity analyses are a common way to assess the predictability of such complex phenomena. To perform such analyses, a structural model of a row of 15 FAs in the reactor core is implemented with the finite-element code ANSYS Mechanical APDL. The distribution of lateral hydraulic forces within the core row is estimated based on a two-dimensional Computational Fluid Dynamics model with porous media, assuming symmetric or asymmetric core inlet and outlet flow profiles. The influence of the creep rate on the bow amplitude is tested based on different creep models for guide tubes and fuel rods. Different FA initial states are considered: fresh FAs or FAs with higher burnup, which may be initially straight or exhibit an initial bow from previous cycles. The simulation results over one reactor cycle demonstrate that changes in the creep rate and the hydraulic conditions may have a considerable impact on the bow amplitudes and the bow patterns. A good knowledge of the specific creep behavior and the hydraulic conditions is therefore crucial for making reliable predictions.

ABSTRACTThis work shows the impact of potential displacements of the fuel assembly positions in the reactor core on the signal values of the ex-core instrumentation of a pressurized water reactor in order to understand in detail the impact on the calibration factor of ex-core detectors. This was done with a range of Monte Carlo calculations that simulated the detailed geometrical effect by stepwise changing of the positions of fuel assemblies for selected, conservative scenarios. First, criticality calculations were carried out for the chosen core configurations, and corresponding surface sources on the core barrel were determined. In these calculations, the distances were varied between the fuel assemblies which were in the line of sight of the ex-core instrumentation. A maximal change of the fluxes on the surface of the core barrel of 4%/mm could be calculated under conservative assumptions for the combination of displaced fuel assemblies. In addition, a dependence of this effect as a function of cycle burn-up was analyzed. In a second step, transport calculations for the ionization chambers were performed using the surface sources. An increase of the reaction rate at the chambers of up to 3%/mm has been calculated.

ABSTRACTThe aeroball measurement system (AMS) is an important in-core instrumentation in German pressurized water reactors. Therefore, it is essential to know the possible uncertainties of this system. One is the lack of knowledge of the positions of balls in the guide tubes. The position changes can be up to 7 mm. Since the neutron flux distribution is not constant across the guide tubes, different reaction rates can result from the displacements. Both fuel assembly and full core calculations were carried out with the Monte Carlo code MCNP5. Differences in the reaction rates of up to 2% could be determined. In most cases, differences are only up to 0.5%. The results were hardly influenced by burnup and boron concentration in the water moderator. For fuel assemblies containing gadolinium as a burnable poison, a more pronounced reduction could be observed in the direction towards the gadolinium fuel rods. Overall, it was found that the AMS measurement values are very robust with regard to possible variations of ball positions.

The safety behaviors of a nuclear power plant (NPP) after an external hazard-initiated event, as well as after a small break (SB) loss of coolant accident (LOCA), are already well known as part of the analyses made for standard license application. The coincidence of both events leads to a beyond-design basis consideration. Such a combination of both event categories is investigated by means of the thermohydraulic system code ATHLET. The scenario assumes an external event with a LOCA caused by induced vibrations on a small pipe attached to the primary circuit, although all pipes are designed to withstand the loads created by such an external event. Furthermore, in the context of both robustness and enveloping analyses, both a loss of offsite power (LOOP) and an unavailability of the emergency diesel power supply are postulated. The NPP in the scenario considered only has access to the passive accumulators and to systems supplied via the safeguard emergency diesel engines (second quartet of emergency diesel engines), which are housed in the bunkered emergency feed building. The dedicated type of external event itself is not in focus, but rather the thermohydraulic behavior of the NPP is considered. Apart from the model's assumptions, the accident sequence is explained in detail. The remaining systems for emergency core cooling are capable of handling the LOCA under such demanding boundary conditions. Long-term cooling can be ensured. Furthermore, heat removal out of the core is always sufficient. Eventually, all safety protection objectives have been complied for this beyond-design basis scenario.