Preliminary Safety Analysis Report for the General Atomic Gas-Turbine Modular Helium Reactor
Ed Blandford
Ali Moheet
Jeff Seifried
Evan Thomas
NE 167/267 Final Report
May 14th, 2007
1.1Introduction
The purpose of the report is to demonstrate that the Gas-Turbine Modular Helium Reactor (GT-MHR) meets the goals laid out by the Staff as described in the Policy Statement on Regulation of Advanced Nuclear Power Plants and can be licensed as designed. The development of an actual Preliminary Safety Analysis Report (PSAR) requires several years and hundreds of experienced engineers coupled with a rigorous R&D program. This report represents a culmination of associated GT-MHR-related research publications and a preliminary conceptual design report issued by General Atomics (1999) which are appropriately cited.
The NRC has over 30 years of experience with licensing and regulating light water reactors (LWR). The original Rasmussen reactor safety study (WASH-1400), followed up by the NUREG-1150 report, really provided the foundation for assessing the associated safety risks of the current fleet of LWRs and contributed greatly to the development of Probabilistic Risk Assessment (PRA) methods. The NUREG-1150 report represented one element of the NRCs effort to close the book on severe accident issues associated with the set of currently operating U.S.nuclear power plants and provided the results of the estimated plant risks for five commercial nuclear power plants of different design. This work, coupled with a successful operating history, has led to a familiarity in licensing and regulating which is evident by a host of recent LWR plant uprates.
With the increasing demand of emission-free power generation, a nuclear power renaissance is becoming more realistic. Innovative reactor designs are being pushed by both industry and academia. The latest report issued by the DOE has indicated a desire to demonstrate large-scale hydrogen production using a nuclear plant. The project, Next Generation Nuclear Plant (NGNP) demonstration, calls for the use of a high-temperature gas reactor to produce hydrogen using high temperature process heat or electricity. The GT-MHR is a prime candidate for the NGNP nuclear plant and presents several licensing issues to the NRC. Advanced reactors, such as the GT-MHR, are unique to the current fleet of LWRs and proposed Gen III+ designs and create several challenges to the regulator.
1.2Overall Licensing Approach of GT-MHR
Developing a licensing strategy for the GT-MHR requires an extensive understanding of the NRCs current stance on advanced reactors. The Advanced Reactor Policy Statement was issued some years ago and the NRC, along with NEI, has done extensive work since then to help advanced plant designers with developing licensing strategies. The purpose of this section is to discuss these latest developments and how they may impact the licensing of the GT-MHR.
The current NRC approach for licensing advanced reactors consists of a four part process (Figure 1-1) which will results in an overall technology-neutral regulatory structure with technology-specific regulatory guidance. The NRC is required to take this approach due to the extreme diversity of the advanced reactors proposed. For example, fast sodium cooled reactors have unique neutronic characteristics such as a positive void coefficient that gas-cooled reactors and current LWRs are not concerned with. Therefore the proposed framework uses the reactor Safety Goal Policy quantified health objectives (QHO) in the Commission’s Reactor Safety Goal Policy to ensure that design, construction, and operations are consistent with the performance goals for all proposed reactor types.
Figure 1-1Framework for Regulatory Structure for New Plant Licensing
In addition to meeting the QHO objectives, the Policy Statement on Regulation of Advanced Nuclear PowerPlants also mandated thatadvanced reactors will make larger safety margins. The NRC has developed generic frequency-consequence curves that are consistent with the overall safety goal objective and are applicable to all reactor concepts. The approach utilized by the staff combines both probabilistic risk criteria and design-basis criteria. The risk criteria portion deals with preventing accidents and ultimately the development of mitigation criteria while the probabilistic criteria are used to select appropriate design basis accidents (DBA) and the overall safety classification of the reactors systems, structures and components (SSC). Design basis criteria are used todetermine fixed acceptance criteria for events that are used for comparison to siting requirements.A frequency-consequence curve (Figure 2-2) was developed by the NRC to determine an acceptable region for advanced reactors based on offsite dose guidelines laid out in 10CFR100 and 10CFR50.34
Figure 1-2 Frequency-consequence curve for public health and safety
In Chapter 3, a set of design basis accidents are considered and shown to fall within the acceptable region as defined by Figure 2-2. The curve as a whole is meant to provide guidance on the frequency and consequence of accidents and to be reasonably consistent with the QHOs of the Commission’s Safety Goal Policy Statement. The QHOs limit the total risk of all accidents to the “average” individual within specified distances of the exclusion area boundary.
2.1 Plant Description
Each GT-MHR plant consists of four reactor modules. The primary components for each module are contained within a steel vessel system, which includes a reactor vessel and a power conversion vessel, connected by a cross vessel. The vessel system is located inside an underground concrete silo 25.9m in diameter by 42.7m deep, which serves as the containment structure. The reactor vessel is made of high strength 9Cr-1Mo-V alloy steel and is approximately 8.4m in diameter and about 31.2m high. It contains the reactor core, the reactor internals, control rod drives, refueling access penetrations, and the shutdown cooling system. The reactor vessel is surrounded by a Reactor Cavity Cooling System
Figure 2-1 GT-MHR module arrangement
which provides totally passive safety related decay heat removal by natural draft air circulation. The shutdown cooling system located at the bottom of the reactor vessel provides forced helium circulators for decay heat removal for refueling and maintenance activities (General Atomic).
Power conversion vessel is also made of modified 9Cr-1Mo-V alloy steel and is approximately 8.5m flange outside diameter and about 35.4m high. This vessel houses the turbo machine, a plate-fin recuperator, and a helical tube water-cooled intercooler and precooler. The turbomachine includes a generator, a turbine, and 2 compressor sections all mounted on a single shaft supported by magnetic bearings.
2.2Module Description
The standard reactor module, which is the basic building block of the reference GT-MHR, consists of a reactor core and power conversion equipment.
Figure 2-2 GT-MHR simplified schematic flow diagram
The reactor core and power conversion equipment are housed in separate welded steel vessels that are connected by a cross vessel. The same helium that flows through the reactor is the working fluid in the power conversion portion of the module (Figure 2.2).
The single standard reactor module, which is the building block of the MHR, contains the nuclear heat source and all the power conversion equipment required to generate electricity within the primary pressure boundary. This equipment includes the turbo-compressor-generator set, plate-fin recuperator modules, precooler, intercooler and the interconnecting flow ducting (General Atomic).
2.3Plant Systems
The gas turbine plant includes the following key systems:
- Reactor System, which includes the reactor core, core supports, internal structures, reactivity control assemblies, and hot duct.
- Vessel System, which includes the reactor vessel, power conversion vessel, cross vessel, vessel supports, and lateral restrains.
- Power Conversion System, which includes the turbomachine, recuperator modules, precooler, intercooler, internal supports, shrouds, and seals. This system also includes the equipment and handling casks necessary for the removal and replacement of PCS components.
- Shutdown Cooling System, an independent forced convection cooling system for backup decay heat removal, which includes the shutdown circulator, shutdown heat exchanger, and shutdown cooling control.
- Reactor Cavity Cooling System, a safety-related passive air cooling system for backup decay heat removal, which includes structures for inlet/outlet of atmospheric air, a set of cooling panels surrounding the reactor vessel, and the hot/cold duct work for transporting the air.
- Fuel Handling System, which handles fuel and reflector elements, and transports them between the receiving facility, the reactor core, and the fuel packaging and shipping facility.
Figure 2-3 Helium flow path in power conversion module.
- Helium Services System, which includes the helium purification system and the helium transfer and storage system.
- Reactor Protection System, which performs automatic safety-related plant protection functions.
- Investment Protection Systems, which performs automatic non safety intersystem investment-related protection functions.
- Plant Control, Data and Instrumentation System, which monitors plant parameters, automatically regulates plant conditions, provides information to the operator, and accepts and executes manual control commands from the operator.
2.4Vessel System
The principal functions of the Vessel System (VS) are to contain the primary coolant inventory and to maintain primary coolant boundary integrity. In addition, the VS provides structural support and alignment for the Reactor System components and Shutdown Cooling System components that are housed within the reactor vessel and all power Conversion System components that are housed within the power conversion vessel.
The radionuclide control function of the VS are to transfer decay heat from the reactor core to the reactor cavity cooling system (RCCS) during conduction cooldown events, to maintain the geometry of the reactor core with respect to the neutron control assemblies (NCAs) to control heat generation, and to prevent air ingress and consequent core oxidation (General Atomics).
The Vessel System is located below grade, enclosed and supported in a reinforced concrete silo. The reactor vessel and power conversion vessel are places side-by-side with the power conversion vessel at a lower elevation than the reactor vessel. This arrangement provides for thermal isolation and protection of the power conversion components from the high temperature core during conduction cooldown events.
2.5Shutdown Cooling System
A Shutdown Cooling System (SCS) provides reactor cooling when the Power Conversion System is non-operational. The SCS consist of the shutdown circular and shutoff valve, the shutdown heat exchanger, and shutdown cooling control. Also included as part of the SCS are the shutdown circular and shutdown heat exchanger service equipment.
The SCS consist of a single loop with shutdown heat exchanger in series with the shutdown circular and shutdown loop shutoff valve assembly, all located at the bottom of the reactor vessel. Hot helium from the core outlet plenum flows through multiple parallel openings (pips) in the center of the core support structure and into the shutdown heat exchanger. Once cooled, the helium continues downward through the shutdown loop shutoff valve to the shutdown circulator where it is compressed and discharged into the reactor vessel bottom heat cavity. The loop is completed as the helium flows down through the reactor core. Heat is rejected from the shutdown cooling water to the atmosphere through the air cooled heat exchanger (General Atomics).
2.6Reactor Cavity Cooling System
The Reactor Cavity Cooling System (RCCS) performs 2 safety functions. It provides a passive means of transporting core residual heat from the reactor cavity when neither the Power Conversion System nor the Shutdown Cooling System is available, thereby preventing the reactor vessel from exceeding design temperature limits. It also protects the concrete walls of the reactor cavity from exceeding design temperature limits for all modes of operation. The RCCS removes heat by conduction through the graphite reflector and by radiation and natural convection from the uninsulated vessel. The system, which receives the heat transferred from the vessel, includes a cooling panel placed around the reactor vessel. Heat is removed from the reactor cavity by natural circulation of outside air through the cooling panel.
The natural draft air cooling concept is shown in Figure 2.4. The design has no pumps, circulators, valves, or any other active components. The surface of the cooling panel serves to separate the outside atmosphere from the reactor cavity atmosphere. This minimizes the site boundary dose due to release of air activated in the cavity. The system has multiple inlet/outlet ports and interconnected parallel flow paths to ensure continued cooling in the event of blockage of any single duct or opening.
The system is required to operate continuously in all modes of plant operation to support normal operations, and, if forced cooling is lost, it functions to remove decay heat to ensure investment and safety protection. Since the RCCS is relied upon to meet 10CFR100 requirements, the system is classified as “safety-related” (General Atomics).
Figure 2-4 Shutdown cooling water flow system.
2.7Safety Features
Health and safety of workers and of the public is a fundamental consideration in GT-MHR plant design. A defense in depth approach to safety was used in the design of the GT-MHR. Implementation of defense-in-depth results in the provision of multiple barriers to the release of fission products and systems which limit the challenges to and protect those barriers. Furthermore, these systems are capable of functioning despite credible failures, by being redundant, independent, and divers.
The fundamental, inherent characteristics of the GT-MHR are listed below. These characteristics tend to dominate the safety characteristics of the plant as a whole and serve to prevent and mitigate accidents.
Coated Fuel Particles; Coated Fuel Particles can withstand extremely high temperature without losing their ability to retain radio nuclides. Core temperature can remain at 1600C for several hundred hours without losing particle coating integrity. For design basis events, peak expected fuel temperatures do not exceed 1460oC.
Graphite Moderator; Graphite can withstand even higher temperatures than the fuel and without structural damage, which complements the fuel’s high temperature capability. The graphite also holds up certain fission products, further reducing potential radioactivity releases.Massive graphite structures in the core provide extremely large heat capacity. Even under extreme conditions, reactor heat up is slow, so that days are available for the operators to respond to an unusual event, such as loss of all AC powers.
Helium Reactor Coolant;Helium is chemically inert and neutronically transparent, meaning it will not aggravate an accident by participating in any chemical or nuclear reaction. Helium will not change phase in the reactor; therefore, it is impossible to have problems of 2 phase flow within the reactor, such as steam bubbles which affect reactivity and temperature control. Pump cavitation can not occur. The use of helium minimizes the problems of primary system corrosion and greatly reduces the resultant buildup of radioactive by-products associated with water-cooled reactors.
Negative Temperature Coefficient of Reactivity;The GT-MHR reactor core is designed to have a negative temperature coefficient of reactivity. This characteristic means that as the reactor gets hotter, the change in temperature alone tends to reduce reactor power. For all credible reactivity addition events, the negative temperature coefficient is sufficient to control reactor power (General Atomics).
3.1Accidents Scenarios
In accordance with guidance laid out by the NRCs Technology Neutral Framework, three classifications of events have been defined as a function of the event frequency:
- Frequent Events (Anticipated Operational Occurrences)
- Infrequent Events (Design Basis Accidents)
- Rare (Beyond Design Basis Accidents)
Associated does releases for each event category are defined by the NRC based on 10CFR100 and 10CFR50.34 criteria (see Figure 2-2). DBA offsite dose guideline is 25 Rem as defined in 10 CFR 50.34. All postulated events for the GT-MHR are expected to fall within the acceptable region. Potential pathways for radionuclide release are shown in Figure 3-1.
The accident classifications used are consistent with what is defined in the GT-MHR design conceptual report and the classification levels described above. Work performed by Oak Ridge National Laboratory (ORNL) analyzed the fuel response under various accident conditions. The main concern under accident conditions is whether the fuel temperature exceeds the failure limit of 1600°C and a code developed by ORNL was used to calculate these values over the evolution of an accident. The Graphite Reactor Severe Accident Code (GRSAC) was developed to study a wide spectrum of core transient and heatup accident scenarios for both the PBMR and the GT-MHR design. A detailed 3-D thermal-hydraulics model was implemented and models were used to characterize the SCS and RCCS.
Figure 3-1 Radionuclide Containment System
3.2Safety-related Systems, Structures and Components (SSC)
The safety-related Systems, Structures and Components identified by the GA conceptual design report include: