Korean Next Generation Reactor Research Articles

암반 지반의 재해도 스펙트럼에 기반한 토사지반 원전 부지의 등재해도 스펙트럼 평가 기법

암반지반에 주어진 등재해도 스펙트럼에 상응하는 원전부지 토사지반에서의 등재해도 스펙트럼을 도출하기 위한 확률론적 방법론을 제시하였다. 이를 위해 지진 운동 및 지반의 불확실성을 고려한 지반응답 해석을 통해 토사지반 지표에서의 지진동 증폭계수를 산정하였다. 증폭계수는 가장 상관관계가 높은 지반운동의 스펙트럴 가속도 규모와의 회귀분석을 통해 계산되었다. 이 방법론을 적용하여 국내 KNGR (Korean Next Generation Reactor) 및 APR1400 (Advanced Power Reactor 1400) 원전의 포괄부지 지반 중 B1, B4, C1 및 C3 지반을 대상으로 등재해도 스펙트럼을 도출하였다. 등재해도 스펙트럼을 통해 지진동 발생 빈도 별 위험 주파수 대역을 평가하고 분석하였다. 이 결과는 원전의 종합적 지진리스크 평가 결과를 보다 합리적으로 개선하는 데에 활용될 수 있으며, 향후 다양한 종류의 토사지반에 대한 등재해도 스펙트럼을 평가하는 데에 적용할 수 있을 것으로 기대된다. We propose a probabilistic method to evaluate the uniform hazard spectra (UHS) of the soil of nuclear power plant(NPP) sites corresponding to that of a bedrock site. To do this, amplification factors on the surface of soil sites were estimated through site response analysis while considering the uncertainty in the earthquake ground motion and soil deposit characteristics. The amplification factors were calculated by regression analysis with spectral acceleration because these two factors are mostly correlated. The proposed method was applied to the evaluation of UHS for the KNGR (Korean Next Generation Reactor) and the APR1400 (Advanced Power Reactor 1400) nuclear power plant sites of B1, B4, C1 and C3. The most dominant frequency range with respect to the annual frequency of earthquakes was evaluated from the UHS analysis. It can be expected that the proposed method will improve the results of integrated risk assessments of NPPs rationally. We expect also that the proposed method will be applied to the evaluation of the UHS and of many other kinds of soil sites.

Proposed Neural Network Approach for Monitoring Plant Status in Korean Next Generation Reactors

This paper reports the development work carried out in respect of a proposed application of Neural Network approach for the Korean Next generation Reactor (KNGR) now referred as APR-1400. The emphasis is on establishing the methodology and the approach to be adopted towards realizing this application in the next generation reactors. Keeping in view the advantages and limitation of Artificial Neural Network Approach, the role of ANN has been limited to plant status or to be more precise plant transient monitoring. The simulation work carried out so far and the results obtained shows that artificial neural network approach caters to the requirements of plant status monitoring and qualifies to be incorporated as a part of proposed operator support systems of the referenced nuclear power plant.

Regulatory Concerns on the In-Containment Water Storage System of the Korean Next Generation Reactor

The in-containment water storage system (IWSS) is a newly adopted system in the design of the Korean Next Generation Reactor (KNGR). It consists of the in-containment refueling water storage tank, holdup volume tank, and cavity flooding system (CFS). The IWSS has the function of steam condensation and heat sink for the steam release from the pressurizer and provides cooling water to the safety injection system and containment spray system in an accident condition and to the CFS in a severe accident condition. With the progress of the KNGR design, the Korea Institute of Nuclear Safety has been developing Safety and Regulatory Requirements and Guidances for safety review of the KNGR. In this paper, regarding the IWSS of the KNGR, the major contents of the General Safety Criteria, Specific Safety Requirements, Safety Regulatory Guides, and Safety Review Procedures were introduced, and the safety review items that have to be reviewed in-depth from the regulatory viewpoint were also identified.

General Safety Criteria for the Korean Next Generation Reactor

In Korea, an evolutionary reactor called the Korean Next Generation Reactor (KNGR) is being developed. Safety and Regulatory Requirements and Guidance (SRRG) are also being developed for the regulation of the KNGR. The hierarchy of the SRRG consists of five tiers: Safety Objectives, Safety Principles (SP), General Safety Criteria (GSC), Specific Safety Requirements (SSR), and Safety Regulatory Guides.The GSC set out general and comprehensive criteria used to determine that SP are complied with in the design and to provide fundamental bases for the development of SSR.In this paper, the development approach, structure, and contents of the GSC are introduced. Particularly, the major features adopted for the safety enhancement of the GSC are addressed regarding such issues as severe accidents, human factors, shutdown and low-power operations, reliability, periodic safety review, radiation protection, and environmental-effect design bases.The GSC have been reviewed by the experts of various fields. The finalized GSC were at the preannouncement stage of the legislation process in 2001 and are expected to contribute to ensuring the enhanced safety level of the KNGR.

Development of core seismic analysis models for KNGR fuel assemblies associated with 0.3 g seismic loads

In order to evaluate the structural integrity of fuel assemblies associated with 0.3 g seismic loads in the Korean Next Generation Reactor (KNGR), detailed fuel assembly model and core series models with seven and 17 assemblies have been developed using the super-element capability of the msc/nastran code. The detailed fuel assembly model has been verified by comparison with the analysis results and test results. Non-linear transient analysis has been performed on the core series models. The detailed fuel assembly model and core series models for the KNGR developed with the msc/nastran code have a good correlation with test results and in-reactor impact behavior of fuel assemblies under applied seismic loads. Since the time histories considered in the transient analysis were generated from a seismic analysis of the System 80+ reactor internals that is reference plant of the KNGR based on 0.3 g ground motion, it is necessary to study further the evaluation for the KNGR fuel assembly and the models using the site specific time histories. It is expected that the further evaluation and some modifications can be performed effectively with the aid of the models developed in this study.

Defense-in-depth and diversity evaluation to cope with design bases events concurrent with common mode failure in digital plant protection system for KNGR

The Korean Next Generation Reactor (KNGR) has been evolved to adopt an advanced design feature, a digital Plant Protection System (PPS) as an effort of enhancing reliability and safety of the plant. Although the digital PPS can be designed with high reliability, it is considered to be vulnerable to the Common Mode Failure (CMF) in the system software resulting in a total loss of the built-in hardware redundancy. Therefore, a comprehensive evaluation has been performed to demonstrate the intrinsic capability of the KNGR design in coping with the design basis events concurrent with CMF in the digital PPS. Instead of the conservative bounding analysis methodology, a best-estimate analysis methodology has been developed and utilized since the design basis events accompanied by CMF in the digital PPS are categorized as beyond design bases events. A variety of diverse means such as Alternate Protection System (APS), process control systems, and timely operator actions have been verified to be effective in mitigating the design basis events with CMF in the digital PPS.

A study on the connection effects of subsurface slabs to exterior walls in the KNGR seismic analysis

The soil-structure interaction (SSI) analyses are being performed for the Korean next generation reactor (KNGR) design, because the KNGR is developed as a standard nuclear power plant concept enveloping various soil conditions. The SASSI program, which adopts the flexible volume method is used for the SSI analyses. Since the impedances are computed based on the assumption of a rigid foundation in the KNGR design, the impedances are computed at a single point around the center of the basemat. The base model used in the SSI analyses assumed that the subsurface slabs are free and not connected to the exterior walls. This results in a more flexible structure than reality. This study is performed to address the concern of the potential dynamic effect of this modeling assumption on the overall structural response. The super-structure model is modified with the subsurface portions of the sticks being laterally rigid, so that the entire subsurface cavity together with the super-structure model acts laterally rigid. For the comparison, the in-structure response spectra using the rigid subsurface model are compared with the original spectra without considering the connection effects.

The OBE Elimination in Piping Design

The Operating Basis Earthquake (OBE) and Safe Shutdown Earthquake (SSE) have been considered in the design of nuclear power facilities as required by Appendix A to 10 CFR Part 100. However, it is believed that the elimination of the OBE from the design of nuclear facilities would be necessary for plant optimization since the OBE criterion is too rigid and has excessive conservatism. Studies indicate that alternative piping designs can exhibit reliability and safety levels equal to or greater than the current analysis methods. The alternative rules for the Earthquake Engineering Criteria have been issued by the Appendix S to 10 CFR 50. In the System 80 + Design, the USNRC reviewed the alternate analysis methods which were proposed to eliminate the OBE based on the EPRI-URD and concluded that those were acceptable as stated in the NUREG-1462. In the Korean Next Generation Reactor (KNGR) developed as an ALWR, a typical piping model was selected to include ASME Classes 1, 2 and 3 piping and was analyzed according to the current method as well as the alternate analysis method, specified in NUREG-1462, for comparison.

Mode K—A Core Control Logic for Enhanced Load-Follow Operations of a Pressurized Water Reactor

New core control logic known as Mode K has been developed to enhance the load-follow operation (LFO) capability of a pressurized water reactor. The Mode K reactor regulating system, which actuates control bank movements, consists of two closed control loops, one for the coolant average temperature control and the other for the axial power shape control. Via its peculiar logic for selecting the control banks to be driven, the Mode K controls the coolant average temperature and axial power shape simultaneously and automatically within their allowed operating limits. In this way, the Mode K significantly reduces the operator burden associated with conventional manual power shape control during LFOs. A simple and flexible soluble boron scenario complements the Mode K logic and contributes toward reducing operational burden by its simplicity. The Mode K logic has been implanted in the Korean Next-Generation Reactor, a 1300-MW(electric) class evolutionary nuclear power plant under development in Korea, and various kinds of LFOs including frequency control have been simulated using the Framatome engineering simulator SAPHIR. The simulation results show reasonable core control performance of the Mode K as well as proper behaviors of other major nuclear steam supply system components such as the pressurizer and steam generator.

New leak detection technique using ceramic humidity sensor for water reactors

Leak-before-break (LBB) approach has been considered for its application to the main steam line (MSL) of Korean Next Generation Reactor (KNGR) — an advanced pressurized water reactor under development. Unlike the primary water leakage, the MSL leak detection must be based on principles other than radioactivity measurements. Among potential options that are being considered as indicators of leakage, it is believed that humidity at the proximity of the piping system is an effective one. A ceramic-based humidity sensor was developed, which can be qualified for LBB applications. The ceramic material, sintered and annealed MgCr 2O 4–TiO 2, is shown to increase its electrical conductivity upon water vapor adsorption without any negative impact due to gamma radiation over the entire temperature range of interest. In the plant applications, the sensor array can be positioned in the annulus between the piping and surrounding insulation. By the analysis of humidity distribution in the annulus, a leak rate of 1 l h −1 can be detected within an hour when the distance between two adjacent sensors does not exceed 1 m. In order to minimize the number of signal wires, the use of AC impedance technique is shown to be advantageous. In this paper, the results of the development and the performance characterization of ceramic humidity sensor for the LBB application to the MSL of KNGR are summarized.

The development of passive design features for the Korean Next Generation Reactor

Four passive design features, such as a fluidic device, a passive secondary condensing system (PSCS), a passive cavity flooding system, and a passive hydrogen ignitor, are under development as part of the Korean Next Generation Reactor (KNGR), an advanced PWR, to increase its safety. The fluidic device, which is located at the discharge of the safety injection tank, is a system to inject the borated water into the reactor coolant system in a passively regulating way to elongate the allowable start-up time of emergency diesel generators and to enhance performance against the loss of coolant accidents. The PSCS, which supplements the auxiliary feedwater system, secures the heat removal through steam generators in the case of the loss of feedwater event. The PSCS takes inlet flow from the steam line and returns condensate into the feedwater line after condensation through condenser tubes. Fusible plugs have been adopted for passive cavity flooding. If the ambient temperature is high enough, the plugs between the in-containment refueling water storage tank (IRWST) and the reactor cavity melt so that the IRWST water starts to flow into the reactor cavity by gravity. Passive hydrogen ignitors (i.e. catalytic ignitors) have been adopted in addition to the active hydrogen ignitors to maintain containment hydrogen concentrations below a detectable limit of 10 volume percent. Applicability of these passive features on KNGR has been studied following the three-step approach of (1) preliminary analyses using computer codes and small-scale experimental facilities, (2) detailed analyses through large-scale tests, code running, and uncertainty validation, and (3) quality assurance by design verification and analyses on interfaces with other systems.

Man-Machine Interface Design for Korean Next Generation Reactor: A Human Factors Perspective

The concept of advanced man-machine interface (MMI) technology is employed to the design of the main control room (MCR) for Korean Next Generation Reactor (KNGR), an advanced light water nuclear power plant (NPP) currently under development. In the KNGR MCR, computerized workstations and digital operating systems provide the operator with electronically generated graphics and information for system status displays and plant controls. The introduction of digital technology to the development of advanced power plant control station has brought new issues and concerns associated with the KNGR MMI design, especially in human factors perspective. This paper presents some of human factors efforts in the development of KNGR MMIs that includes large screen display, workstation CRT, computerized procedure, and soft controller.

Evaluation of analytically scaled models of a pressurized water reactor using the RELAP5/MOD3 computer code

This paper performs analytical evaluations for the potential distortions caused by the scaled models using RELAP5/MOD3 computer codes. By use of scaling analysis, two scaled models with the same volumetric ratio are constructed for the Korean next generation reactor (KNGR), which is an advanced light water reactor. The scaling methodology adopted in this paper preserves the two-phase natural circulation similarities between prototype and scaled models. One scaled model is at full height with reduced flow area. The other model is at reduced height with reduced flow area. By using appropriate scale factors the RELAP5/MOD3 input models are developed. Then, the transient responses of the two ideal scaled models are simulated for small break loss of coolant accidents (SBLOCAs) by using the RELAP5/MOD3 computer code. The transient responses of the two scaled models are compared with those of the prototype. The results indicate that qualitative and quantitative similarities are well preserved for both models during SBLOCA with different break sizes.

Design options for the safety injection system of Korean next generation reactor

Loss of coolant accident (LOCA) analyses for various configurations of safety injection system (SIS) are performed to optimize the emergency core cooling system (ECCS) performance for the Korean next generation reactor (KNGR). The KNGR is an advanced light water reactor (ALWR) adopting the advanced design feature of a direct vessel injection (DVI) configuration and passive fluidic device in the discharge line of the safety injection tank (SIT). To determine the feasible SIS configuration and the optimum capacities of the SIT and high pressure safety injection pump (HPSIP), licensing design basis and best estimate LOCA analyses are performed for the limiting large break and small break spectrum, respectively. The analyses results show that the four-train DVI injection with the current system design is a more feasible configuration than the other ones considered and the adoption of a fluidic device SIT enhances the ECCS performance for large break LOCA. For small break LOCA, in the case of cold leg break, the DVI4 configuration is better than other configurations and also meets the EPRI ALWR requirement of no core uncovery for up to a 15.24 cm (6 in) diameter small break. However, in the case of DVI line break, slight core uncovery is predicted and also the system behavior is significantly affected by reactor vessel (RV) downcomer modeling. Therefore, the DVI4 configuration is more feasible for KNGR ECCS performance, but further investigations are required to resolve the ECCS bypass issues for large break LOCA and to develop a proper RV downcomer model for analysis of DVI line break in small break LOCA.

Development of an advanced human–machine interface for next generation nuclear power plants

An advanced human–machine interface (HMI) has been developed to enhance the safety and availability of a nuclear power plant (NPP) by improving operational reliability. The key elements of the proposed HMI are the large display panels which present synopsis of plant status and the compact, computer-based work stations for monitoring, control and protection functions. The work station consists of four consoles such as a dynamic alarm console (DAC), a system information console (SIC), a computerized operating-procedure console (COC), and a safety system information console (SSIC). The DAC provides clean alarm pictures, in which information overlapping is excluded and alarm impacts are discriminated, for quick situation awareness. The SIC supports a normal operation by offering all necessary system information and control functions over non-safety systems. In addition, it is closely linked to the other consoles in order to automatically display related system information according to situations of the DAC and the COC. The COC aids operators with proper operating procedures during normal plant startup and shutdown or after a plant trip, and it also reduces their physical/mental burden through soft automation. The SSIC continuously displays safety system status and enables operators to control safety systems. The proposed HMI has been evaluated using the checklists that are extracted from various human factors guidelines. From the evaluation results, it can be concluded that the HMI is so designed as to address the human factors issues reasonably. After sufficient validation, the concept and the design features of the proposed HMI will be reflected in the design of the main control room of the Korean Next Generation Reactor (KNGR).

Design options for safety depressurization system

The Korean Next Generation Reactor (KNGR) adopted an advanced design feature, a safety depressurization system (SDS) to rapidly depressurize the primary system in case of events beyond the design basis. Two design approaches are considered for the SDS design. The use of bleed valves similar to the ABB-CE System 80+ is design option 1, while in design option 2, the French Sebim valve is considered to provide the combined function of overpressure protection and rapid depressurization. In this paper, thermal hydraulic analysis using a best-estimate version of CEFLASH-4AS/REM is performed for a total loss of feedwater (TLOFW) event to investigate the feasibility of those two design options. For each design option, various feed and bleed (F and B) procedures are investigated for a TLOFW event. For design option 1, the required bleed capacity is determined from the CEFLASH-4AS/REM simulation according to the EPRI Advanced Light Water Reactor (ALWR) requirements. The analysis results demonstrate that the TLOFW event can be mitigated in a proper manner with a sufficient margin using design option 1. For design option 2, the operator action times for initiating the F and B are investigated by varying the number of Sebim valves and high pressure safety injection (HPSI) pumps. If the operator opens two out of the three Sebim valves in conjunction with the four HPSI pumps before a hot leg saturation condition, the decay heat removal and core inventory make-up function can be successfully accomplished. The results of the present investigation demonstrate that the two design options are both feasible.