On the Thermal-Hydraulic Essentials of the Holtec Inherently Safe Modular Underground Reactor (HI-SMUR) System

2011 ◽  
Author(s):  
Kris Singh ◽  
Indresh Rampall ◽  
Joseph Rajkumar
Keyword(s):  
Reactor Core ◽  
Reactor Vessel ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Control Rod ◽  
Hydraulic Conditions ◽  
Deep Underground ◽  
The Stability ◽  
Core Cooling

HI-SMUR 140 is a small (145 MWe) modular pressurized water reactor designed to harness fission energy without the use of a recirculation pump. HI-SMUR’s core resides deep underground in a thick-walled reactor vessel (RV) enclosed by a stainless steel-lined reinforced concrete “Reactor Well”. The HI-SMUR Nuclear Steam Supply System (NSSS) is a conjugated pressure vessel assemblage wherein the steam generator(s) are integrally joined to the RV, i.e., without any interconnecting piping, and the pressurizer is integral to the reactor vessel. There are no penetrations in the RV for over a height of 120 feet above the reactor core, which precludes the scenario of loss of coolant to the core from a postulated pipe break event. The rejection of decay heat from the reactor in the wake of a scram is engineered to occur without the aid of on-site or off-site power, making the HI-SMUR NSSS demonstrably capable of withstanding a cataclysmic environmental phenomenon of Fukushima’s intensity without loss of core cooling or without precipitating any damage to the plant or the surrounding environment. The system design places a premium on accessibility and maintainability of vital equipment such as the steam generators, RV internals, and the control rod drive assemblies. This paper is the first in a series of papers planned to explain and quantify the performance and safety aspects of HI-SMUR 140. In this paper, the thermal-hydraulic characteristics of the HI-SMUR NSSS are explored using classical hydraulic correlations which have served as the tool for the scoping parametric study of the system. In particular, the stability of the system under varying power output conditions and the long-term reliability of the fuel under the most adverse thermal/hydraulic conditions are presented.

Coordinated Control of a Small Pressurized Water Reactor

2018 ◽  
Author(s):  
Peiwei Sun ◽  
Chong Wang
Keyword(s):  
Control System ◽  
Reactor Core ◽  
Reactor Power ◽  
Pressurized Water Reactors ◽  
Primary System ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Control Rod ◽  
Average Temperature ◽  
Water Reactors

Small Pressurized Water Reactors (SPWR) are different from those of the commercial large Pressurized Water Reactors (PWRs). There are no hot legs and cold legs between the reactor core and the steam generators like in the PWR. The coolant inventory is in a large amount. The inertia of the coolant is large and it takes a long time for the primary system to respond to disturbances. Once-through steam generator is adopted and its water inventory is small. It is very sensitive to disturbances. These unique characteristics challenge the control system design of an SPWR. Relap5 is used to model an SPWR. In the reactor power control system, both the reactor power and the coolant average temperature are regulated by the control rod reactivity. In the feedwater flow control system, the coordination between the reactor and the turbine is considered and coolant average temperature is adopted as one measurable disturbance to balance them. The coolant pressure is adjusted based on the heaters and spray in the pressurizer. The water level in the pressurizer is controlled by the charging flow. Transient simulations are carried out to evaluate the control system performance. When the reactor is perturbed, the reactor can be stabilized under the control system.

Probabilistic Model for Debris-Induced Loss of Long Term Core Cooling

2008 ◽  
Author(s):  
Heather L. Detar ◽  
Daniel T. McLaughlin ◽  
Robert J. Lutz
Keyword(s):  
Safety Issue ◽  
Management Strategies ◽  
Plant Design ◽  
Pressurized Water Reactor ◽  
Original Design ◽  
Pressurized Water ◽  
Design Basis ◽  
Design Basis Accident ◽  
Core Cooling

Generic Safety Issue (GSI) 191 deals with the potential for generation and transport of debris following a design basis accident that is in excess of quantities assumed in the original design basis and licensing of Pressurized Water Reactor (PWR) plants. In addition to physical modifications to the sump screens to comply with the Generic Letter requirements, some plants have also changed Emergency Operating Procedures (EOPs) to include contingency actions to prevent debris-induced loss of long term core cooling. ASME Probabilistic Risk Assessment (PRA) standard RA-Sb-2005 requires that the plant PRA be maintained and updated to reflect the current plant design and operation. Development of PRA models to quantify the potential for debris-induced loss of long term core cooling supports the PRA updated to reflect the as-built as-operated plant. The PWR Owners Group (PWROG) has undertaken a program to develop a generic PRA model for this issue. The generic PRA model was developed to address the overall plant risk, including new physical and procedural modifications. The model also addresses applications, such as Maintenance Rule screening and the assessment of the risk significance of deviations from the licensing basis analyses. The new PRA model probabilistically treats several facets of the potential for debris-induced challenges to long term core cooling; including debris generation and transport as a function of Reactor Coolant System (RCS) break size and location. The PRA model will permit plant operators to easily incorporate the potential for inadequate core cooling during emergency core cooling recirculation from the containment sump into their PRA Level 1 and Level 2 models. The methodology is based on realistically modeling the conditions that may lead to a debris-induced loss of long term core cooling. The PWROG model also includes consideration of water management strategies being implemented by several PWR plant operators.

Experimental Study of Debris-Loading Head Loss Test for PWR Fuel Assembly After LOCA

2017 ◽  
Author(s):  
Da Wang ◽  
Fenglei Niu ◽  
Weiqian Zhuo ◽  
Jingwen Ren ◽  
Zhangpeng Guo ◽  
...  
Keyword(s):  
Fuel Assembly ◽  
Safety Issue ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Pressure Drops ◽  
Loss Of Coolant Accident ◽  
Sensitivity Studies ◽  
Pass Through ◽  
Core Cooling

A loss of coolant accident (LOCA) in a PWR (pressurized water reactor) would generate debris from thermal insulation and other materials in the vicinity of the break. It is postulated that debris can transported to the containment sump strainer. Some of the debris may pass through the strainer and could challenge the long-term core cooling capability of the plant. To address this safety issue, the downstream effect tests for the PWR were performed. Sensitivity studies on pressure drops through LOCA-generated debris deposited on a fuel assembly were performed to evaluate the effects of debris type and flow rate. Fibrous debris is the most crucial material in terms of causing pressure drops.

Lower Plenum Voiding

1982 ◽  
Vol 104 (3) ◽  
pp. 479-486 ◽  
Author(s):  
D. Bharathan ◽  
G. B. Wallis ◽  
H. J. Richter
Keyword(s):  
Reactor Core ◽  
Reactor Vessel ◽  
Pressurized Water Reactor ◽  
Coolant Circuit ◽  
Two Phase ◽  
Primary Coolant ◽  
Pressurized Water ◽  
The Core ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant

One of the phenomena involved in a loss-of-coolant accident in a pressurized water reactor may be lower plenum voiding. This might occur during the blowdown phase after a cold-leg break in the primary coolant circuit. Steam generated in the reactor core may flow out of the bottom of the reactor core, turn in the lower plenum of the vessel, in a direction countercurrent to the emergency core coolant flow, and escape via the break. If its velocity is high enough, this steam may sweep water from the bottom (lower plenum) of the reactor vessel. Emergency coolant added to the vessel may also be carried out by the escaping steam and thus the reflooding of the core would be delayed. This paper describes a study of two-phase hydrodynamics associated with lower plenum voiding. Several geometrical configurations were tested at three different scales, using air to simulate the steam. Comparisons were made with data obtained by other researchers.

Post Small-Break LOCA Long Term Core Cooling Performance in ACME Test Facility

2016 ◽  
Author(s):  
Zhanfei Qi ◽  
Sheng Zhu
Keyword(s):  
Nuclear Power ◽  
Cooling System ◽  
Safety Margin ◽  
Test Facility ◽  
Passive Safety ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Safety Features ◽  
Core Cooling

CAP1400 Pressurized Water Reactor is developed by China’s State Nuclear Power Technology Corporation (SNPTC) based on the passive safety concept and advanced system design. The Advanced Core-cooling Mechanism Experiment (ACME) integral effect test facility, which was constructed at Tsinghua University, represents a 1/3-scale height of CAP1400 RCS and passive safety features. It is designed to simulate the performance of CAP1400 passive core cooling system in the small break loss of coolant accidents (SBLOCA) for design certification, safety review and safety analysis code development. The Long Term Core Cooling (LTCC) post-LOCA could be simulated by ACME as well. A series of test cases with various break sizes and locations with post-LOCA LTCC period were conducted in ACME facility. This paper describes the post-LOCA LTCC test conducted in ACME test facility. The LTCC phenomena in different cases are very similar. It’s found that the interval that switching from IRWST injection to sump recirculation has the least safety margin. However, it’s shown that the post-LOCA LTCC in ACME could be well maintained by passive core cooling system according to the test results even though the recirculation water level in ACME IRWST-2 is lower than the containment recircualtion level in CAP1400 conservatively.

CFD Analysis of the Coolant Mixing within the Upper Plenum of a Pressurized Water Reactor (PWR)

2013 ◽  
Vol 444-445 ◽  
pp. 411-415 ◽  
Author(s):  
Fu Cheng Zhang ◽  
Shen Gen Tan ◽  
Xun Hao Zheng ◽  
Jun Chen
Keyword(s):  
Temperature Distribution ◽  
Fluid Dynamic ◽  
Characteristic Temperature ◽  
Reactor Core ◽  
Flow Distribution ◽  
Sensitivity Study ◽  
Cfd Model ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor

In this study, a Computational Fluid Dynamic (CFD) model is established to obtain the 3-D flow characteristic, temperature distribution of the pressurized water reactor (PWR) upper plenum and hot-legs. In the CFD model, the flow domain includes the upper plenum, the 61 control rod guide tubes, the 40 support columns, the three hot-legs. The inlet boundary located at the exit of the reactor core and the outlet boundary is set at the hot-leg pipes several meters away from upper plenum. The temperature and flow distribution at the inlet boundary are given by sub-channel codes. The computational mesh used in the present work is polyhedron element and a mesh sensitivity study is performed. The RANS equations for incompressible flow is solved with a Realizable k-ε turbulence model using the commercial CFD code STAR-CCM+. The analysis results show that the flow field of the upper plenum is very complex and the temperature distribution at inlet boundary have significant impact to the coolant mixing in the upper plenum as well as the hot-legs. The detailed coolant mixing patterns are important references to design the reactor core fuel management and the internal structure in upper plenum.

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