Offsite Power Research Articles

Assessment of the grid-related loss of offsite power to the nuclear power plants in the presence of wind farms

Grid stability affects the progression of loss of offsite power scenarios to the nuclear power plants, as the most dominant core-damage-inducing initiating event. The ever-increasing tendency toward integration of large-scale wind farms into the transmission systems drastically challenges their dynamic behavior.In this study, we proposed a dynamic-probabilistic framework to evaluate the effect of wind penetration scenarios on the frequency of occurrence of LOOP to the NPPs. We developed 30 initial grid states, including 3 wind penetration percentages, 5 operational loading levels, in both optimal and normal power flows.We estimated 150 GR-LOOP frequencies for 5 NPP locations upon 3Ph-Shc faults on the transmission lines. Extending from 1.3 × 10−7 to 9.6 × 10−5 (year)−1 for the safest to the riskiest situations, the mean GR-LOOP frequencies have been assessed as 1.74 × 10−5, 2.96 × 10−5, and 2.77 × 10−5 (year)−1, for grids with 0%, 10%, and, 20% penetration percentage of installed wind generation, respectively.Our results indicate the undeniable influence of variations in the operational loadings and the wind penetration percentage on the GR-LOOP frequency, in addition to the NPPs’ locations. Such intensification effect was more pronounced in instantaneous frequencies of unique states than in the averaged ones. Furthermore, we observed more outstanding growth in the corresponding values of optimal cases. Our findings suggested that the reliability of the offsite grid to the NPPs be considered as a decision-making criterion, in addition to the existing security and economic issues in the grids with wind penetration.

Study and assessment of the newly proposed passive management of MSLB + SBO accidents for the VVER 1000/V446 NPP using the secondary side Deaerator tank

To manage accidents in actual conditions, we need to evaluate different accident scenarios on nuclear power plants. Removing the residual heat from the reactor core after the loss of off-site power (LOOP) accident and Station black-out (SBO) accident is one of the biggest challenges in such an incident. In this paper, LOOP and SBO Accidents with the main steam line break (MSLB) accident of No. 2 (inside the Containment) have been simulated by the thermal-hydraulic RELAP5 code on the VVER-1000/v446 reactor. A total of 5 different scenarios have been simulated, assuming the most difficult conservative conditions, and the results reveal the critical role of diesel generators in supplying emergency feed water pumps (EFWP), as well as the part of operator performance in securing a nuclear power plant in the condition of an accident. In the case of using a series of operator operations and using the water resources of the Deaerator tank on the secondary side with the Feed and Bleed system on the primary side, in The most brutal conservative mode (MSLB + SBO accident) by injecting the water into the steam generator-4, the temperature and pressure of the coolant in the primary side are reduced and by absorbing more heat from the reactor, core melting is delayed about 25,000 s, and the safety of Bushehr nuclear power plant (BNPP) against this accident will be improved.

A Pragmatic Approach to Modeling Combinations of Plant Operational States in Multi-Unit Nuclear Power Plant Probabilistic Safety Assessment

One of the technical challenges in multi-unit probabilistic safety assessment (MUPSA) is dealing with numerous combinations of plant operational states (POSs) for each nuclear power plant unit. Since the number of possible combinations of POSs increases exponentially with the number of units, it is impractical to develop separate MUPSA models and assess the risk for each POS combination. This paper proposes a pragmatic approach to modeling combinations of POSs for each reactor unit in MUPSA involving up to 10 reactor units. This approach does not focus on selecting representative POS combinations but rather on screening out non-risk-significant accident sequences in a stepwise manner according to the model quantification results. The effectiveness of the approach is demonstrated by application to cases with four different numbers of units. As a result, in the 2-, 4-, and 6-unit cases, the site and multi-unit core damage frequency (CDF) due to a multi-unit loss of offsite power initiating event are successfully calculated without screening out any accident sequences for each unit. In the 10-unit case, the quantification fails without screening, but it succeeds after reducing the model size by about 43% via the exclusion of the accident sequences in each integrated single-unit model with a CDF contribution of less than 0.1%. The results show that the minimal cut sets obtained in each case cover many POS combinations and that most non-risk-significant POS combinations can be truncated by the cutoff value of 1E-14/yr. In addition, comparing the quantification results according to the stepwise screening criteria shows that the proposed approach can effectively reduce the computational burden in MUPSA without losing much accuracy or realism.

Estimating the adverse effects of inter-unit radioactive release on operator actions at a multi-unit site

Radioactive materials released from a nuclear power plant can degrade the performance of plant operators by disturbing their physical actions, complicating the mitigation procedures, or heightening their psychological stress. This study provides an approach to quantitatively estimate the inter-unit radioactive influence at a multi-unit site by using a conventional probabilistic safety analysis (PSA) model. Three steps are suggested as follows: (1) select the risk-significant core damage sequences of the unit(s) that release radioactive materials, (2) select the risk-significant human failure events (HFEs) at the unit(s) that are affected by the radioactive releases, and (3) develop an integrated PSA model with a fault tree linking approach. The effects of release timing and operator action timing are also considered, called timing effects. To demonstrate the applicability of the approach, the risk of five identical units at a multi-unit site was quantitatively evaluated with a loss of offsite power initiating event as case studies. The multi-unit core damage frequency increased by a maximum factor of 5.7 without consideration of timing effects, while it increased up to 19 percent with consideration of timing effects.

Evaluation of inter-unit dependency effect on site core damage frequency: Internal and seismic event

Since the Fukushima Daiichi accident occurred, the probabilistic safety assessment on multi-unit sites has been studied so far. Several studies addressed to reflect the factor that can affect an adjacent unit, called inter-unit dependency, in the assessment and various research has been made to investigate or evaluate the dependencies. In this study, researches on inter-unit dependency are implemented to internal and seismic level 1 site risk assessment in the purpose of the evaluation on their effect on the probabilistic safety assessment. The assessments are conducted on the hypothetical site in Korea with 4 units of APR-1400. To reflect the redundancy changes in safety system by operation state, the approach is suggested based on the study of dependent accidents. The result showed that the earthquake and loss of offsite power would induce most of the multi-unit core damage. The inter-unit dependency models related to the electric system were sensitive to the result and the synergies between the models were found. The seismic correlation made site core damage frequency decrease. The expectation is proposed as alternative metric for the risk of site and the necessity of site risk assessment is pointed out in this study.

On the Estimation and Reduction of the Frequency of the Loss of Offsite Power Event

Loss of offsite power (LOOP) contributes significantly to the estimated core damage frequency (CDF) for the majority of nuclear power plants (NPPs). Based on experience from some near-miss events around the world and especially in Japan from Fukushima Daiichi accident and several other NPPs it was proven the how important power supply is for nuclear safety. Therefore comprehensive description of the LOOP initiating event is given in this paper. Specific experience of the NPP Krško (NEK) related operation is also presented with comparison to the U.S. NPPs’ approaches and experience. Statistical data are given to explain main characteristics (frequency and duration) of the LOOP event. Important historical trends are also identified. LOOP event frequency is often calculated using generic data because of statistical data base. Thereat, many important factors that influence LOOP are neglected, such as the specific power network configuration, load profiles, climate conditions, and ageing of the equipment. Several different approaches that take these factors into consideration are therefore presented and discussed. Familiarity with the actual state of the NPP, power network and weather conditions helps in the proper planning process of the test and maintenance (T&M) activities and reducing of the risk. This is significant for better online risk estimate. Due to the Fukushima accident, it is necessary to review the safety of the NPPs from new different angles. This paper also shortly reviews planned modifications in NEK that will increase the electric power supply availability (i.e., reduce LOOP event frequency. Other, worldwide solutions that can help to avoid or mitigate LOOP effects are briefly presented.

Liquid-salt-cooled Reactor Start-up with Natural Circulation

The Liquid-Salt-Cooled Very High-Temperature Reactor (LS-VHTR) was modeled using the neutronics analysis code SCALE6.0 and the thermal-hydraulics and kinetics modeling code RELAP5-3D with objective to devise, analyze, and evaluate the feasibility and stability of a start-up procedure for this reactor using natural circulation of the coolant and under the Loss Of Offsite Power (LOOP) conditions. This Generation IV reactor design has been initially developed by Oak Ridge National Laboratory and studied by researchers worldwide for almost a decade. While neutronics and thermal-hydraulics analyses have been previously performed to show the performance of the reactor during normal operation and for shutdown scenarios, no study has heretofore been published to examine the active or passive start-up of the reactor. The fuel temperature (Doppler) and coolant density coefficient of reactivity of the LSVHTR were examined using the CSAS6 module of the SCALE6.0 code. Negative Doppler and coolant density feedback coefficients were calculated. Two initial RELAP5 simulations were performed to obtain the steady-state conditions of the model and to predict the changes of the thermal-hydraulic parameters during the shutdown of the reactor. Next, a series of step reactivity additions to the core were simulated to determine how much reactivity can be inserted without jeopardizing the safety and stability of the core. Finally, a start-up procedure was developed, and the restart of the reactor with natural convection of the coolant was simulated. The results of the simulations demonstrated the potential for natural circulation start-up of the LS-VHTR.

RELAP5/MOD3.3 Analyses of Core Heatup Prevention Strategy during Extended Station Blackout in PWR

The accident at the Fukushima Dai-ichi nuclear power plant demonstrated the vulnerability of the plants on the loss of electrical power for several days, so called extended station blackout (SBO). A set of measures have been proposed and implemented in response of the accident at the Fukushima Dai-ichi nuclear power plant. The purpose of the study was to investigate the application of the deterministic safety analysis for core heatup prevention strategy of the extended SBO in pressurized water reactor, lasting 72 h. The prevention strategy selected was water injection into steam generators using turbine driven auxiliary feedwater pump (TD-AFW) or portable water injection pump. Method for assessment of the necessary pump injection flowrate is developed and presented. The necessary injection flowrate to the steam generators is determined from the calculated cumulative water mass injected by the turbine driven auxiliary feedwater pump in the analysed scenarios, when desired normal level is maintained automatically. The developed method allows assessment of the necessary injection flowrates of pump, TD-AFW or portable, for different plant configurations and number of flowrate changes. The RELAP5/MOD3.3 Patch04 computer code and input model of a two-loop pressurized water reactor is used for analyses, assuming different injection start times, flowrates and reactor coolant system losses. Three different reactor coolant system (RCS) coolant loss pathways, with corresponding leakage rate, can be expected in the pressurized water reactor (PWR) during the extended SBO: normal system leakage, reactor coolant pump seal leakage, and RCS coolant loss through letdown relief valve unless automatically isolated or until isolation is procedurally directed. Depressurization of RCS was also considered. In total, six types of RCS coolant loss scenarios were considered. Two cases were defined regarding the operation of the emergency diesel generators. Different delays of the pump injection start following the station blackout were assumed and analysed. For each scenario, two kinds of SBO calculations for two-loop PWR were performed, base and verification. Base calculations were needed to determine necessary minimum flowrate for steam generators feeding in such a way that they are not overfilled or emptied. Namely, it was assumed that instrumentation is not available during extended SBO. The verification calculations have been then performed to verify if the determined minimum flowrates are sufficient to prevent the core heatup. The calculated results show effectiveness of the proposed extended SBO prevention strategy provided that the water injection is available in the first two hours after SBO occurring at full power. If diesel generator is running after loss of offsite power for some time, e.g. one hour, the available time for steam generator water injection is significantly longer. The obtained results demonstrate the need for assessment of the pump injection flowrates before the utilization of the pump for mitigation of the event. The applicability of the developed method for assessment of the required pump injection flowrate has been validated on PWR.

Investigation of Loss of Feedwater (LOFW) Accident in the APR-1400 Using Fault Tree Analysis

Nuclear power plants play a significant role in the contribution of electricity generation on a global scale. Various reactor designs have advantages over others in different aspects. APR-1400 is a pressurized water reactor that is deemed safe due to the redundancy and independence of the multiple safety systems. Probabilistic safety assessment (PSA) is well known for its effectiveness in the representation of risk and safety analysis of the systems in a nuclear power plant. It provides different scenarios of system failure and accident progression via fault tree analysis. A loss of feedwater (LOFW) accident may occur due to numerous reasons such as spurious closure of valves, component failure of heaters, pumps, tanks, or a loss of offsite power (LOOP) event. In the present research, a methodology has been developed that aims to investigate different factors contributing to the loss of feedwater. This research also aims to analyze LOFW accidents by developing fault tree models for the main feedwater system of the APR-1400 to identify the basic events, which may lead to a loss of feedwater accidents. The results of the top event probabilities, risk decrease factor (RDF), risk increase factor (RIF), minimal cut sets (MCS), basic event probabilities, and sensitivity analysis were compared with the WASH-1400 database. It has been found that the control valve (V04) and main feedwater isolation valve (V05) have more contribution to the LOFW accident. The common cause failure (CCF) analysis has been carried out, and it was found that the flow toward the check valve and steam generator are most critical for CCF.

The Contribution of Small Modular Reactors to the Resilience of Power Supply

In recent years, there has been a growing interest in the design, development and commercialization of nuclear power Small Modular Reactors (SMRs). Actual SMR designs cover the full spectrum of nuclear reactor technologies, including water-, gas-, liquid-metal-, and molten-salt-cooled. Despite physical and technological differences, SMRs share some relevant design features, such as small size, modularity, inherent and passive safety systems. These features are expected to enhance availability, recoverability, promptness and robustness, thereby contributing to the resilience of power supply. Thanks to the peculiar design features of SMRs, they are likely to satisfy a number of Functional Requirements (FRs) for this objective, namely: (i) low vulnerability to external hazards; (ii) natural circulation of primary coolant; (iii) prompt, unlimited and independent core cooling under shutdown conditions; (iv) shutdown avoidance in response to variations of the offsite power supply quality and electrical load; (v) island mode operation; (vi) robust load-following; (vii) independent, self-cranking start. These make advanced Nuclear Power Plants (aNPPs) comprised of SMRs perfect candidates to withstand a broader range of natural disruptions and to recover faster from them, compared to conventional Nuclear Power Plants (cNPPs), thus rendering them a major potential asset for guaranteeing resilience and security of power supply. The review focuses on Natural Technological (NaTech) events that impact a typical Integrated Energy System (IESs) within which SMRs are embedded: IESs are, indeed, being developed to integrate different power generation plants with gas facilities, through gas and electricity infrastructures, because they are expected to bring increased security and resilience of power supply, as shown in the qualitative case study presented.

Loss of offsite power (LOOP) accident analysis by integration of deterministic and probabilistic approaches in Bushehr-1 VVER-1000/V446 nuclear power plant

Loss of offsite power (LOOP) accident analysis by integration of deterministic and probabilistic approaches in Bushehr-1 VVER-1000/V446 nuclear power plant

PCTRAN enhancement for large break loss of coolant accident concurrent with loss of offsite power in VVER-1000 simulation

Abstract PCTRAN capability to simulate a large break loss of coolant accident concurrent with the loss of offsite power in Bushehr Nuclear Power Plant is enhanced and investigated. Following the correction of the accident scenario for Bushehr nuclear power plant in PCTRAN, simulation results are compared with the final safety assessment report of that plant. As a result, the primary loop thermal hydraulics parameters including pressure, total flow rates, leakage flow rates and reactor power are in a good agreement with the reference data. Hot and cold leg temperature variations have the same trends as reference data but have a maximum of 80 8C disagreement at the transient initiation. The reason for this disagreement is explained and its adjustment is discussed. Improvements of PCTRAN simulator are mainly due to enhancing user control for atmospheric steam dump valve, containment pressure and emergency core cooling systems which are thoroughly described in this paper.

Enhanced Seismic Protection System for an Emergency Diesel Generator Unit

Nuclear power plants are required to maintain operation after an earthquake, leading to a safe shutdown if necessary. In the case of a loss of offsite power, the onsite emergency diesel generator is critical to ensure procedural operations of the nuclear power plant. As a means to reduce the overall seismic risk, a three-dimensional seismic protection system is proposed to enhance the seismic performance of the emergency diesel generator. The proposed seismic isolation system decouples the horizontal and vertical components of shaking and considers available hardware to achieve an effective isolation solution over the range of excitation frequencies considered. Numerical analysis of the proposed system demonstrates a reduction in seismic demands on the emergency diesel generator and provides a higher safety margin than conventional base installation procedures. Umbilical lines that cross the isolation plane are considered and impose additional constraints on the displacement capacity of the isolation system. However, increasing the displacement capacity of these components can significantly increase the safety margin against failure. The seismic protection system can be customized depending on the seismic hazard and application to different seismic regions.

Dynamic Reliability Evaluation of Diesel Generator System of One Chinese 1000MWe NPP Considering Temporal Failure Effects

Loss of power supply from the diesel generator system (DGS) after loss of offsite power (LOOP) will pose great threat to the safety of GEN-II pressurized water reactors (PWR). Therefore, it is very desirable to evaluate the DGS’s reliability. The traditional analyzing tools are limited to static approaches neglecting the dynamic sequence failure behaviors, such as reliability block diagram (RBD), static fault tree (SFT). Static reliability modeling techniques are not capable of capturing the dynamic sequence-dependent failure behaviors typically existing in NPP safety systems such as DGS, and thus often overestimate the unreliability of systems. In this paper, motivated to study the effects of sequence failure behaviors, dynamic fault tree (DFT) is applied to evaluate the reliability of the DGS of one Chinese 1000MWe Nuclear Power Plant (NPP), and an integrated two-phased Markov Chain model is also developed, which can be considered as a contribution of this article. Comparative study of DGS reliability between DFT and SFT is carried out. The results indicate that compared with the result derived from the DFT model, the unreliability of DGS calculated by SFT is greatly overestimated by about one to two orders of magnitude. Therefore, DFT has a potential to improve the economy of NPP by relaxing the overestimated unreliability of nuclear power systems.

Analysis of ESFR Decay Heat Removal Systems in Protected Station Blackout

Abstract As part of the European sodium fast reactor (ESFR)-safety measures assessment and research tools (SMART) project, safety assessments are being conducted on the updated ESFR design. An important part of the study is the evaluation of the reactor's behavior within hypothetical accidental conditions to assess and ensure that the accident would not lead to unexpected and disastrous events. In this paper, the analyzed accidental scenario is the so-called protected station blackout (PSBO), where the offsite power is lost for the power plant, simulated by using the TRACE and simulator-sodium fast reactor (SIM-SFR) system codes. The assessment started from the simulation of the reactor behavior without the decay heat removal systems (DHRSs). Following this, calculations of multiple DHRS arrangements have been performed to evaluate the individual and combined efficiency of the systems. Where it was possible, the results from the two system codes have been compared to show the consistency of the separate calculations. Through this study, the design of the DHRSs proposed at the beginning of the project has been investigated, and certain recommendations have been made for further improvement of the DHRS systems performance.

Simulation of loss of off-site power accident of CN HCCB TBS

Helium Cooled Ceramic Breeder Test Blanket System (HCCB TBS) is proposed by China, which will be tested in ITER facility to validate the tritium breeding blanket technology. Test Blanket Module (TBM) is the core component associated with its helium cooling system (HCS) and other ancillary systems. The accident analysis is necessary during the preliminary design process, and the results obtained provide safety feedback to the following design work. Loss of Off-site Power (LOSP) is one of the design basis accidents (DBA), and is simulated using the best estimated code RELAP5 MOD3.4. For the safety simulation, besides the thermal conductive and convective model, the radiation model is adopted as well. The influence of the natural circulation of the HCS primary loop and the necessity of the Class III power supply (safety class) are discussed. The simulation results show that the peak temperature of the structure material exceeds the design temperature (550 °C) for short duration, which has almost no impact on the structure. In the meantime, the duration of the temperature over 550 °C drops sharply while the natural circulation and the Class III Power are considered. The results verify the integrity and the safety of the HCS primary loop during the LOSP accident.

Modeling of flow coastdown transient of a reactor coolant system under different pump failures

The flow coastdown transient calculation is important for the safety analysis of a nuclear reactor. An accurate transient analysis during coastdown is necessary and important for the design and manufacture of a reactor coolant pump, since only the reliable work of reactor coolant pumps can guarantee the safety of a nuclear power plant during this important period.In the present work, mathematical models are solved for the non-dimensional transient flow rates in four loops and a reactor core under different power failures of reactor coolant pumps. Python 3.6, high level programming language, is used to solve the mathematical equations of core thermal-hydraulics for the flow coastdown transient. A standard 4-Loop Westinghouse design pressurized water reactor is used in the analysis. Three cases are studied; failure of one of the four pumps, sequential failure of the four pumps and simultaneous failure of the four pumps due to loss of off-site power. The important parameters for the safety analysis of the reactor are flow coastdown time, flow rates, and coolant temperature. The model results for a single pump failure are compared with the published experimental data during flow coastdown of Qinshan reactor coolant pump. The comparison results show acceptable agreements. Sensitivity analyses for two different parameters during flow coastdown transient are investigated. These two parameters are the fluid inertia ratio, α and core time constant, τ. Effects of each of the parameters on the flow coastdown transient, are given separately.

Investigating the influence of the variation of the load characteristic on the occurrence frequency of Grid-related Loss of Offsite Power using a probabilistic-deterministic methodology

Generic frequencies are used for Grid-related Loss of Offsite Power (GR-LOOP) in the Probabilistic Safety Analysis (PSA) of Nuclear Power Plants (NPPs). When historical databases are used, the influence of variation in the grid conditions over time and the impact of the NPP location are not considered in the estimation of GR-LOOP frequency. Therefore, relying purely on historical data induces uncertainty in total Core Damage Frequency (CDF).This paper aims at assessing the influence of the dynamic load behavior on the GR-LOOP frequency, for different locations. For that purpose, sensitivity cases are defined. The occurrence frequency of GR-LOOP following the occurrence of a three-phase-short-circuit fault on a transmission line is evaluated via a probabilistic-deterministic methodology for all cases.The minimum and maximum estimated values of the GR-LOOP frequency, 2.98E-04/year and 9.95E-03/year respectively, indicate the dependence on the variation in either the dynamic load behaviors or the NPP locations.The results reveal that GR-LOOP frequency varies with load model in the initial grid. In the presented probabilistic-deterministic methodology it is beneficial to consider the influence of timely changes of the grid conditions on the GR-LOOP occurrence frequency for different NPPs’ locations, especially where relevant or sufficient generic databases are not available.

Assessment of PSA Level 2 for core catcher design using a combined probabilistic and deterministic method for a case study: IR-360 NPP

Mitigating of molten corium in severe accidents is a vital concern in safety of Nuclear Power.Plants (NPPs). As one of the main features, core catcher is provisioned in the design of NPPs. In order to appropriately design an ex-vessel core catcher, numerous core damage (CD) sequences with vessel breach are required to be studied and analyzed, which requires massive efforts. Therefore, this study intends to develop a comprehensive hybrid method, to extract a logical number of CD sequences with bounding thermal hydraulic parameter for core catcher design. Such method interprets the result of Level 2 Probabilistic Safety Analyss (PSA), from both the probabilistic and deterministic perspectives. For this analysis, high pressure, high-temperature massive melt volumes are considered as core catcher bounding parameters. Moreover, a multi-level qualitative and quantitative screenings is accomplished on the list of Plant Damage States (PDSs), to screen the candidate CD sequence(s) with vessel breach. The thermal hydraulic parameters of the screened CD sequences are subsequently studied using MELCOR code, to conclude with the final list of worst sequences, regarding the core catcher design. This study is applied for IR-360 NPP as a case study. In the worst case scenario, the core catcher shall handle a CD sequence with pressure about pressurizer safety valve set point, temperature of 2500 K and corium of 82 tones. According to the results, among the entire CD sequences, a Large Break Loss of Coolant Accident (LB-LOCA) and Loss of Offsite Power (LOOP), induced CD sequence are identified as the most severe sequences for the core catcher design.

Net zero energy buildings in New Zealand: Challenges and potentials reviewed against legislative, climatic, technological, and economic factors

A net zero energy building (NZEB) is a highly energy efficient building that can generate energy through renewable sources that is sufficient to meet its own needs. While NZEBs are being realized in many countries, its emergence in New Zealand remains slow. This study investigates several factors that may cause this issue, namely: legislative, geographical and climatic, technological, and economic factors. The investigation is carried out by secondary data collection and a systematic review of the literature. For the legislative factor, different government statutes, policies, and programs are critically reviewed. For the geographical and climatic factor, the conditions of New Zealand are compared with those of Italy and Japan. For the energy efficiency technological factor, the existing top-rated Green Buildings in the country are analyzed, and the most recurring energy efficient features are discussed. For the renewable energy generation and distribution technological factor, various potential options for on-site and off-site building power sources are discussed, as well as energy storage systems and load matching and grid interaction. For the economic factor, the cost and payoffs of NZEBs are investigated, including the different investment schemes available for potential NZEB developers and how constructing NZEBs can still be feasible post-COVID-19. The study found that New Zealand's climate is suitable for NZEBs and the required technology and economic resources are available. However, more focused and intentional legislations and policies are required to both encourage and incentivize the wider adoption of the NZEB concept in the country.