Power Uprate Research Articles

Heavy metal loading effects on special nuclear material and waste of an uprated Indonesian Reaktor Daya Eksperimental

This technical policy study investigated the effect of heavy metal loading (HML) quantity in the pebble fuel on the special nuclear material (SNM) and waste quantities in an uprated Indonesian pebble bed reactor design, the Reaktor Daya Eksperimental (RDE). A set of Monte Carlo simulations, performed using OpenMC, was deployed to simulate the pebble fuel depletion using an infinite lattice reactor calculation. This study considers HML, fuel residence time, leftover 235U, total Pu, medium- and long-lived wastes, number of depleted pebbles, and volume of waste at different HML values, at attainable discharged fuel burnup, at a target burnup of 80 GWd/MTU and at different reactor powers (10 and 40 MWt). Due to an over-moderation effect, a higher HML quantity shortened the attainable discharged fuel burnup level, which also translated to a lower fuel utilization. An 80% higher HML quantity resulted in about a 25% lower burnup level. Assuming a five-pass refueling scheme, this resulted in approximately three times more leftover 235U, 2.4 times more Pu, 1.4 times more medium-lived waste, and 1.5 times more long-lived waste collections per year. The study showed that a power uprating by a factor of four increased the SNM and waste quantities by four times.

Scale effects on core design, fuel costs, and spent fuel volume of pressurized water reactors

The desire to improve the economic competitiveness and deployment pace of nuclear energy through modularization, manufacturing, and series production had led to the development of smaller size reactors. As the standard 17x17 fuel technology is mainly maintained in the pressurized water reactors (PWRs) category, this translates into a lower number of fuel assemblies in the core and sometimes a reduced fuel height. To assess the impact of such scale change in core design on fuel cycle cost and spent fuel volume, a scoping analysis tool is developed based on infinite lattice calculations, leakage, fuel management reduced models, and levelized unit cost of electricity (LCOE) estimate. As such, cost dynamics driven by fuel specific power, burnup, core leakage, feed, cycle length, fuel assembly height as well as uranium market data are captured with consistent set of assumptions and analysis methods. A selection of 5 reactor designs representative of leading PWR developers is assessed and compared. Pursuing higher specific powers and optimal burnups are highlighted as the main fuel cost reduction drivers, nevertheless, practical limitations and opportunities must be evaluated to establish the feasibility of such enhanced fuel operation. In consequence, a detailed core design is performed using SIMULATE3 code for 5 PWR variations including natural and forced coolant circulation modes, two reactor scales, power densities of 73, 112, and 123 kW/l and higher discharge burnups. Design and optimization are performed at the lattice level, for the reflector, and at the core loading level. Satisfactory steady-state operation including power distribution, coolant operating limits, and reactivity requirements are analyzed and reported in this paper. The fuel economics of the detailed core designs confirm the scoping analysis findings. Despite the unlocked power uprates in small PWRs, the achievable burnup for a given fuel specific power requires more enrichment and shorter fuel height results in higher fabrication costs per mass of fuel, which makes scaling down core size a more expensive endeavor on the fuel cycle front. Spent fuel volumes are reported for the PWRs designed in this paper. These volumes are driven by the core average discharge burnup regardless of the scale in consideration. Additional cost and core performance aspects related to heavy reflector gains, fuel-reflector substitution, and disposal cost policy in the U.S. are examined.

Dynamic systems modeling of the spallation neutron source cryogenic moderator system to optimize transient control and prepare for power upgrades

Through support of the US Department of Energy’s Office of Basic Energy Sciences, Oak Ridge National Laboratory has begun applying machine learning methods to improve accelerator and target performance of the Spallation Neutron Source (SNS). These methods are being applied to the control optimization and power upgrade of the SNS Cryogenic Moderator System (CMS). A numerical model of the CMS has been developed to study these optimizations and system modifications using EcosimPro. This paper compares steady-state and transient numerical results with experimental data. Control optimization studies focused on dampening mass flow, temperature, and pressure fluctuations during sudden losses of accelerator beam power. This analysis was conducted by adjusting five proportional-integral-derivative controllers connected to four flow control valves and one heater. Future efforts include power uprate studies focused on increasing the CMS cooling capacity. The current accelerator power is 1.4 MW; the first target station is being upgraded to 2.0 MW as part of the Proton Power Upgrade effort. The CMS cooling capacity is sufficient for 2.0 MW operation.

Flux Flattening Large Heavy Water Power Reactors with Accelerator-Driven Photoneutron Source

Flux flattening and power uprating of large heavy water power reactors (HWRs) are demonstrated as an application of an accelerator-driven photoneutron source (ADS) in the ADS-CANDU concept where an array of electron linear accelerators is configured around the periphery of a subcritical CANDU-6 core. The localized ADS generated through (e−,γ,n) reactions in the HWR lattice perturbs the reactor power distribution by increasing the power of low-power bundles and depressing the power at the core center relative to the fundamental mode power distribution. Gross power uprating is feasible when the system is near critical, but the ADS array consumes tens of megawatts electric exceeding the power gained by a factor of more than 2 for the conservative ADS performance specifications assumed in the analysis. Several important challenges of fixed-source Monte Carlo simulations of near-critical multiplying media are investigated including severe load imbalance issues with distributed-memory parallel computing architecture and correlated local tallies in nonanalog (implicit absorption) Monte Carlo radiation transport. All subcritical fixed-source simulations in the study readily exceed the default random number stride used in most production Monte Carlo codes, and the stride exceedance causes both bias in local tally results (bundle powers) and spatial autocorrelation of these errors/biases in the large core. A legacy stride exceedance is critically reviewed, and the conclusions and subsequent interpretations of those conclusions are rejected. Several classes of radiation transport Monte Carlo problems are likely to be susceptible to stride exceedance, and this issue needs to be promptly addressed by the Monte Carlo analyst and code developer communities.

Condenser Retrofit in Leibstadt Nuclear Power Plant (BWR) - Far Beyond a Standard Modular Solution

Abstract The Leibstadt nuclear power plant (KKL) in Switzerland was originally commissioned in 1984 with a nominal power output of 1000 MWel. Since then, there have been several uprates of the boiling water reactor (BWR) which have increased the power to 1275 MWel. After the nearly four decades of operation, the original condenser was impacted by advanced erosion corrosion and by progressive performance deterioration. In addition, the thermal power uprates required load reduction during hot summer periods to avoid high condenser back pressure. Consequently, KKL asked GE to carry out a comprehensive retrofit study to identify potential improvements on both lifetime and performance. These included: robust material selection, optimized extraction of non-condensable gases by means of a bespoke air cooler section of the condenser and applying GE Steam Power's latest innovative tube pattern design which is aligned to BWR requirements. However, site constraints prevented a typical modular condenser retrofit. Hence, working in a close cooperation between KKL and GE Steam Power, a customized retrofit concept was developed which used semi-prefabricated skeletal modules and an on-site rebuild of the condenser. In total, approximately 475 tons of old condenser material were dismantled, 24 new semi-prefabricated skeletal modules were installed, a condenser with 51'600 tubes was rebuilt on-site and 8 waterboxes were replaced with an improved design made of stainless steel. More than 350 professionals contributed to the successful retrofit with an overall site effort of 270'000 working hours - increasing the power output to 1285 MWel.

Water Relief Through Safety or Relief Valves is not Just a Leak

Abstract Two Nuclear Safety Advisory Letters (NSALs), advise operators of PWRs that they may assume, in accident analyses, that certain safety and relief valves can open, relieve water, and then reseat properly. The NSALs claim water that exits the reactor coolant system (RCS) can be made up by water that is delivered into the RCS by the emergency core coolant system (ECCS). For example, during an inadvertent operation of the ECCS (IOECCS) event, The NSALs claim, “since the cause of the water relief is the ECCS flow, the magnitude of the leak will be less than or equivalent to that of the ECCS (i.e., operation of the ECCS maintains RCS inventory during the postulated event and establishes the magnitude of the subject leak).” A comparison of ECCS flow to water relief, at relevant RCS pressures, indicates that water relief, during an IOECCS, cannot be dismissed as a leak. Critical flow calculations indicate that ECCS flow cannot replace RCS inventory that is relieved, as water, through the pressurizer relief and safety valves when RCS pressures are near nominal operating levels. At much lower pressures, the ECCS could offset the water relief. However, by then the IOECCS will have been either resolved, or will have developed into a loss of coolant accident (LOCA). The NSALs were published in 1993 and 2007. NRC did not question the NSALs’ advice until 2015, when it was found in a licensee’s application for a power uprating. The licensee might have submitted a false statement, in support of its application.

Erosion and Corrosion of Pipelines and Equipment on the Power Uprate of Power Units at Nuclear Power Plants (NPP)

Erosion and Corrosion of Pipelines and Equipment on the Power Uprate of Power Units at Nuclear Power Plants (NPP)

Design and Neutronic, Thermal-Hydraulic Analysis of DSCF Assembly for a SMR and Investigation of the Effect on the Thermal Power Uprate

Small modular reactors (SMRs) can be a significant option for developing countries with low energy demand. Due to the lack of sufficient experience in the field of SMRs, extensive research should be done on SMRs to improve the performance of these systems. Using dual surface-cooled fuel (DSCF) is one of the methods that can increase the performance of SMRs. In this study, for the first time the core of a NuScale reactor (as a SMR) is designed based on DSCF without any change in core dimensions by analyzing neutronic, thermal-hydraulic, and natural circulation parameters. In addition, according to the departure from nucleate boiling ratio, the uprate of the thermal power in a reactor using DSCF is investigated. For this purpose, typical solid fuels as well as DSCFs under clean-cold and full-power conditions are primarily modeled for the four different lattices that maintain the same assembly dimensions, mass, and enrichment fuels as the original fuel assembly. The effective multiplication factor, and power peaking factor, as important neutronic parameters, are calculated. Then the departure from nucleate boiling, pressure drop, velocity, and temperature distribution calculations, as important thermal-hydraulic and natural circulation parameters, are accomplished via a computational fluid dynamics code. The best core configuration of DSCF for the NuScale core is determined based on comparing the neutronic, thermal-hydraulic, and natural circulation parameters of various lattices and typical solid fuels. Regarding the final result, a DSCF assembly configuration, called a 12 × 12 assembly, is suggested.

Capital cost estimation for advanced nuclear power plants

The first-of-a-kind (FOAK) nuclear plants built in the last 20 years are 2X over budget and schedule in the US and some European countries. One of the nuclear industry's proposed remedies is the small modular reactor (SMR). SMR designs leverage five factors to be more economically competitive than large reactors: 1) multiple units; 2) increased factory production and learning; 3) reduced construction schedules; 4) plant design simplification and 5) unit timing. There are currently no bottom-up studies that quantitatively account for these factors and compare different near-term light water reactor SMRs with Gen III + large plants. This work presents a nuclear plant cost estimating methodology using a detailed bottom-up approach for over 200 structures, systems, and components. The results compare relative costs for two large pressurized water reactors, one with active safety and one with passive safety, to two SMR designs, one with multiple reactor power modules and one with a single reactor module. Passive safety systems showed noticeable savings at both the large and small-scale reactors. The power uprating of an SMR by 20% resulted in ∼15% savings in the overnight unit capital cost. Overall, if built by an inexperienced vendor and work force, the two SMRs' overnight costs were higher than large reactors, since significant on-site labor still remains while losing economy of scale. However, the single-unit SMR had significantly less total person-hours of onsite labor, and if built by an experienced workforce, it could avoid cost-overrun risks associated with megaprojects.

Advanced gas-cooled reactors technology for enabling molten-salt reactors design – Optimisation of a new system

Molten salts present significant advantages as coolants over traditional light water or gas. Salt-cooled nuclear reactor can potentially i allow an increase in core power density and simplify the reactor safety case. Traditionally, molten salt reactors assume molten salt being both the fuel and the coolant. This, however, creates major challenges for the design and operation due to requirements for chemistry control and corrosion behaviour of the core structural materials. The Fluoride salt-cooled High-temperature Reactors (FHR) on the other hand, have conventional solid fuel geometry surrounded by moderator and coolant. Furthermore, some FHR variations share many common characteristics with the Advanced Gas-cooled Reactors (AGR). Thus, using the knowledge accumulated during many decades of successful AGR operation can speed up the FHR development and deployment. However, replacing carbon-dioxide coolant by molten salt significantly changes the reactor performance. Hence, a new core design is needed. In this work, a Multi-Objective Particle Swarm Optimisation is used to identify the most favourable configurations for a new system layout. Initially, the optimisation targeted the beginning of cycle parameters such as criticality and Coolant Temperature Coefficient (CTC). The present work is the next step in this analysis. Each configuration is examined with respect to its thermal-hydraulic performance to assess the power uprate potential which is limited by multiple temperature constraints (e.g. fuel centreline and cladding temperatures as well as the coolant freezing/boiling). The estimated maximum power was then used in the fuel burnup calculations, from which a discharge burnup and cycle average CTC were obtained. As a result of the optimisation process, several families of possible solutions were identified, which form an optimal Pareto front. The most attractive configurations in terms of achievable power density, however, were not necessarily on the Pareto front. Most of the identified design options had only a small amount of graphite moderator or no graphite at all, relying entirely on moderation in the salt. Increasing the graphite volume was consistently found to worsen most of the optimisation parameters. The newly identified design options have the potential to achieve power density which is higher than that of a typical AGR by up to a factor of five, while maintaining negative CTC through the burnup cycle.

A Review of Prognostics and Health Management Applications in Nuclear Power Plants

The US operating fleet of light water reactors (LWRs) is currently undergoing life extensions from the original 40- year license to 60 years of operation. In the US, 74 reactors have been approved for the first round license extension, and 19 additional applications are currently under review. Safe and economic operation of these plants beyond 60 years is now being considered in anticipation of a second round of license extensions to 80 years of operation. Greater situational awareness of key systems, structures, and components (SSCs) can provide the technical basis for extending the life of SSCs beyond the original design life and supports improvements in both safety and economics by supporting optimized maintenance planning and power uprates. These issues are not specific to the aging LWRs; future reactors (including Generation III+ LWRs, advanced reactors, small modular reactors, and fast reactors) can benefit from the same situational awareness. In fact, many small modular reactor (SMR) and advanced reactor designs have increased operating cycles (typically four years up to forty years), which reduce the opportunities for inspection and maintenance at frequent, scheduled outages. Understanding of the current condition of key equipment and the expected evolution of degradation during the next operating cycle allows for targeted inspection and maintenance activities. This article reviews the state of the art and the state of practice of prognostics and health management (PHM) for nuclear power systems. Key research needs and technical gaps are highlighted that must be addressed in order to fully realize the benefits of PHM in nuclear facilities.

Development of the NuScale Power Module in the INL Modelica Ecosystem

This paper provides a comprehensive overview of the development of a NuScale power module in the Modelica process model ecosystem at Idaho National Laboratory (INL) as part of the U.S. Department of Energy’s Office of Nuclear Energy Integrated Energy Systems (IES) program. Model development has led to the creation of a dynamic NuScale module in the Modelica language that operates under natural circulation and is consistent with design parameters set forth in the design certification documentation for the Nuclear Regulatory Commission (NRC). Controllers have been developed that are consistent with controllers mentioned by NuScale design certification documents. While characteristic time constants may differ due to controller gains, the actual control actions of the system are consistent with NuScale prescribed actions. In addition to natural circulation and controllers, the NuScale power module includes hot channel calculations, reactivity control, and a pressurizer. Hot channel calculations include bulk fluid temperature, outer clad temperature, fuel centerline temperature, and departure from nuclear boiling ratios. Reactivity control incorporates moderator and fuel temperature feedback mechanisms, boron concentration, and control bank positions. A pressurizer with sprays and heaters was developed to maintain primary system pressure within bounds. Further, the model has been developed to allow the user to input beginning-, middle-, or end-of-cycle reactivity conditions. Three simulations were run that demonstrate the capability of the model controllers to meet target parameters while preserving system values within reasonable bounds. The first simulation is an initialization study at 160 MW(thermal)/50 MW(electric) that compares the model steady-state values with the design certification values. This simulation demonstrates the capability of the model to operate in natural circulation mode and closely matches operating values set forth in the NRC design certification documents. The second simulation is a power uprate from 160 MW(thermal)/50 MW(electric) to 200 MW(thermal)/60 MW(electric), and the final simulation is of a typical summer day where the turbine demand oscillates throughout the day. These illustrate the capabilities of the system to operate under natural circulation while maintaining system parameters within controller bounds and to match turbine output with turbine demand. This work adds a systems-level model of the NuScale power module with associated control systems, reactivity coefficients, control banks, coolant through fuel temperature profiles, and associated mass flow rates throughout the system that have been added to the Modelica ecosystem. This model will be the basis of work that is utilized in the Joint Use Modular Plant program at INL, which proposes to utilize a single power module to support research, development, and demonstration activities, with a specific aim to demonstrate various IES configurations. As part of the ecosystem, the NuScale module can be quickly integrated with existing Modelica modules (e.g., thermal energy storage, reverse osmosis, high-temperature steam electrolysis, gas turbines) to model energy grids in different geographical locations.

Application of Expanded Event Trees combined with uncertainty analysis methodologies

During the last decades, the safety analysis of nuclear power plants, have been shifting from conservative models and hypotheses to Best Estimate. Within this trend, the different safety analyses can be categorized depending on their associated conservatisms. The approach that includes Expanded Event Trees together with Best Estimate model and conditions plus uncertainty avoiding conservatisms in system availability can be referred as Extended BEPU (EBEPU). In this sense, an Extended BEPU safety analysis relies on Expanded Event Trees and Best Estimate models to study an accidental sequence.In the present paper, an EBEPU study is performed, using a loss of coolant accident in a pressurized water reactor modeled with the TRACE code. A 5% core power uprate is used as an example of safety margin EBEPU analysis. The paper presents parametric and non-parametric uncertainty approaches. Observing the results from all event tree sequences, it is seen that the branches that contribute the most to the Core Damage frequency are successful branches, with all safety systems available, and some Core Damage branches have a possibility of success.

Uncertainty quantification analysis and statistical estimation for LBLOCA in a PWR using Monte-Carlo and alternative methods

The direct Monte-Carlo method has recently emerged as an alternative uncertainty propagation and quantification method to calculate the 95-percentile peak cladding temperature with 95% confidence level (PCT95/95) or the conditional exceedance probability (CEP) as the computational capability improves. However, it requires a large sample size, and it is still being debated that how many samples are required to obtain the result with low uncertainty and high convergence. In this study, assuming the 10% power uprate of APR-1400, the uncertainty quantification analysis of LBLOCA was conducted by using the direct Monte-Carlo method. Based on the PCT data calculated from different sample sizes (100, 200, 500, 1000, 2000, 5000) and different sampling methods (simple random sampling and Latin hypercube sampling), the PCT95, the CEP, and their 95% confidence intervals were estimated. Then, they were compared with results of Wilks’ method and with those of Johnson’s normal distribution transformation method, respectively.

Optimized 18-months low-leakage core loadings for uprated VVER-1000

Abstract The VVER-1000 life extension and power uprating are important areas of research and development activities of JSC “Rosenergoatom”. Recent research has shown that the core baffle resource is one of the major limiting factors for the life extension of the Balakovo NPP unit 3. The core loading optimization is necessary to decrease the core baffle radiation dose for NPP life extension at uprated power. The paper contains the results of the optimization of 18-month low-leakage core loadings for the uprated to 104% Balakovo NPP unit 3. Moreover, the results of the radiation dose calculations at the reactor pressure vessel and core baffle are presented in the work. It is shown that the optimization of the core loadings reduces peaks of radiation doses and flattens its distributions along the core baffle, as well as gives the opportunity to extend the NPP life up to 60 years at the uprated power.

Research on Optimization of Parameter Uncertainty Applied in Small Range Power Uprating for Nuclear Power Plants

The background and the method for the power uprating in nuclear power plants (NPP) in China were described, and the specifications on NPP power uprating in related laws and standards were analyzed and summarized. The classification of NPP power uprating was overviewed, and the specific requirements and main methods for NPP small range uprating were analyzed. Based on the analysis of NPP power and the calculation method for the uncertainty, the main parameters having effects on the calculation error for the thermal power of the core were specified, which were the main feedwater flow and the main feedwater temperature. The methods to apply the optimization of parameter uncertainty in the power uprating was analyzed and summarized, which was to add new differential pressure transmitter or ultrasonic flowmeter and temperature transmitter to optimize the parameter uncertainty. The economy for NPP small range power uprating was summarized.

Utilization of 3D fuel modeling capability of BISON to derive new insights in performance of advanced PWR fuel concepts

The Internally and eXternally cooled Annular Fuel (IXAF) and Lightbridge's metallic Helical Cruciform Fuel (HCF) are two innovative fuels, which could allow significant power uprate (∼20%) in current Pressurized Water Reactors (PWRs) while maintaining or improving safety margins. Both fuels would substantially benefit from 3D modeling to properly assess their performance in a PWR core and to further their technology readiness level. For IXAF, the possible misalignment of the annular pellets creates a mismatch in inner and outer gap size, which will result in an azimuthally asymmetric temperature distribution and different heat split between the inner and outer channel. For HCF, the fuel deformation requires 3D consideration due to its inherit helical geometry. In this work, the first attempt of building thermo-mechanical capability for analysis of these fuels was made in BISON fuel performance code. The 3D fuel behavior under normal operation along with loss of coolant accident (LOCA) and reactivity initiated accident (RIA) were simulated. The preliminary key insights derived from the 3D fuel performance modeling were the potential excessive ballooning of IXAF during LOCA and excessive swelling of HCF during RIA. However, overall both fuels did show promising and improved performance at 20% uprated conditions compared to traditional solid fuel geometry and material.

Power uprate operation at Chinshan Nuclear Power Station

This process of requesting operation at thermal power levels above the current licensed power level for nuclear power station is referred to as a Power Uprate (PU). There are many nuclear power stations around the world that have implemented power uprate operation for improving their efficiency and economic performance. The Chinshan Nuclear Power Station (CNPS) performed the 1.66% and 2% power uprate for MURPU and SPU, respectively. After successfully achieving the operation of power uprates, the total increase of electrical output for CNPS has amounted to 35.79 MWe. In addition, it also reaches the reduction of CO2 emission about 0.68 million tons per year compare to natural gas combined cycle (NGCC) power plant. Furthermore, the net increase in yearly revenue was compared to the cost of implementing the MURPU and SPU, and the payback time was less than one year. In addition to the calculation of thermal power uncertainty, the power ascension test results, major operating parameters after power uprates, cost benefit analysis, and calculation of CO2 reduction are also presented in this paper. Moreover, the key operating parameters of the turbine cycle measured after power uprate have established a reference base for monitoring station operating performance and provide useful information to turbine cycle design for nuclear power plant.

ANALYSIS OF CS-137 TO CS-134 ACTIVITY RATIO FOR FAILED FUEL EXPOSURE ESTIMATION

The Cs-134 to Cs-137 activity ratio of the Cs-134 and Cs-137 fission products released from failed fuel rods into primary coolant is very useful to identify the exposure along with the fuel batch of the failed fuel. The calculated and measured Cs-137 to Cs-134 radioactivity ratios of failed BWR and PWR fuels are compared and analyzed for better understanding of their relationship. Moreover, the impacts of power uprate and fuel reload outage on calculated Cs-137 to Cs-134 activity ratios are studied and the physics behind the impacts are provided.

Study of the impact of NPP rated thermal power uprate on process behavior at different transient conditions

Today, objective preconditions have been formed to find the ways on how to increase cost-effectiveness of NPPs operation, while providing the required safety level. One of such ways to increase thermal nominal power of power unit. The paper provides for the results of reactor behavior analysis at increased thermal power above nominal received using a one-dimensional system computer code RELAP5/MOD3.2 and relevant model of VVER-1000 (V-320) power unit. Calculation analyses are performed for quasi-static reactor operating conditions and transients using realistic approach in terms of initial performance parameters of reactor installation. In researches, representative initial events for transients have been selected according to the principle described further. For an abnormal operation, an event has been selected based on its high frequency and consequences, which require decreasing reactor power down to 50 % of nominal thermal power. For emergency conditions an event has been selected which is caused by external extreme impacts typical for Ukrainian NPP sites resulting in the worst consequences. Thus, the transients are represented by events associated with failure of a single turbine-driven feed water pump and total station blackout unit. To analyze emergency conditions caused by long-term blackout, they were additionally accompanied by a leakage through reactor coolant pump seals. Given that increase of steam flow in a turbine at increased thermal power above nominal requires additional studies on residual service life assessment of its critical components, a 3-D model of high-pressure rotor of a full speed turbine is proposed for further studies. Based on the calculations a comparative analysis of major parameters of the reactor at rated and increased thermal power is performed with assessment of significant factors to be considered in further studies on increase of installed thermal output of NPP unit.