Thermal-hydraulic and neutronic analyses of the submersion-subcritical, safe space (S) reactor

The solid-core Submersion-Subcritical Safe Space (S⁴) reactor is cooled with He-28% Xe gas (molecular weight of 40 g/mole) and nominally generates 471 kWth for at least 7 years. To avoid single point failures in reactor cooling and energy conversion, the S⁴ reactor core is divided into three hydraulically independent, but neutronically and thermally coupled sectors. Each sector feeds a separate Closed Brayton Cycle (CBC) power conversion loop with separate heat rejection radiator panels. Detailed thermal-hydraulic analyses of the S⁴ reactor core are performed to ensure that the maximum fuel temperature during nominal operation stays below 1300 K. In addition, a neutronics analysis performed using MCNP 5 confirms that the S⁴ reactor satisfies the design reactivity requirements. These are at least $ 4 of cold clean excess reactivity, at least $ 2.25 of shutdown margin, and at least $ 1 subcritical in the worst-case of submersion and flooding, following a launch abort accident. Mass estimates of the S⁴ reactor design that meets both the thermal and the reactivity requirements are provided.     

The modular high temperature gas-cooled reactor (MHTGR) in the U.S.

The MHTGR is an advanced nuclear reactor concept being developed in the USA, under a cooperative program involving the U.S. Government, the nuclear industry, and the utilities. As its objective, this program is developing a safe, reliable, and economic nuclear power option for the USA, and the other nations of the world to consider in meeting their individual nationalistic electrical generation or process heat needs by the turn of the century. The design is based on a concept of modularization that can meet the various power needs by combining any number of 350 MW(t) reactor modules in parallel with a selected number of turbine plants in a variety of arrangements. Basic HTGR features of ceramic fuel, helium coolant, and graphite are sized and configured to provide a low power density core with passive safety features such that no operator action or external source of power is needed for the plant to meet 10CFR100 or Protective Action Guidelines limits at the 425 m site boundary. This precludes the necessity to plan for the evacuation or sheltering of the public during any licensing basis event. The safe behavior of the reactor plant is not dependent upon operator action and it is insensitive to operator error. The Conceptual Design is presently being vigorously reviewed by the U.S. Nuclear Regulatory Commission (NRC). A safety evaluation report and a licensability statement are scheduled for issuance by the NRC in January 1988.     

Safe analysis for cooling pipes for prestressed concrete reactor vessels

A step-by-step analysis is given for the direct computation of two-dimensional heat flow to and safe pitching of the cooling pipes. Two models of the cooling system have been selected and calculations have been carried out for an existing vessel. On one model this analysis is compared with the three-dimensional finite element analysis for obtaining insulation conductances for various cooling pipe pitches.     

Compact ignition test reactor (CITR)

Design considerations have been developed for a compact ignition test reactor (CITR). The objectives of this tokamak device are to achieve ignition, to study the characteristics of plasmas that are self-heated by alpha particles, and to investigate burn control. To achieve a compact design, the toroidal field magnet consists of copper-stainless steel plates to accommodate relatively high stresses; it is inertially cooled by liquid nitrogen. No neutron shielding is provided between the plasma and the toroidal field magnet. The flat-top of the toroidal field magnet is 10 s. Strong auxiliary heating is employed. In one design option, adiabatic compression in major radius is employed to reduce the neutral beam energy required for adequate penetration; thiscompression boosted design option has a horizontally elongated vacuum chamber; illustrative parameters are a compressed plasma witha=0.50 m, R=1.35 m,B _T =9.1 T, and a neutral beam power of 15 MW of 160 keVD ⁰ beams. A design option has also been developed for alarge bore device, which utilizes a circular vacuum chamber. Thelarge bore design provides increased margin and flexibility; both direct heating with RF or neutral beam injection and compression boosted startup are possible. The large bore design also facilitates the investigation of high-Q driven operation. Illustrative plasma parameters for full use of the large bore area=0.85 m,R=1.90 m, andB _T =7.5 T.     

Temperature and burnup reactivities and operational lifetime for the submersion-subcritical, safe space (S4) reactor

The He–Xe gas-cooled, S4 reactor has a sectored, Mo–14%Re solid core for avoidance of single point failures in reactor cooling and Closed Brayton Cycle (CBC) energy conversion. The reactor core is loaded with UN fuel and each of its three sectors is thermal-hydraulically coupled to a separate CBC loop and radiator panels. The solid core minimizes voids, and the BeO reflectors are designed to easily disassemble upon impact, ensuring that the bare S4 reactor is sufficiently subcriticial when submerged in wet sand or seawater and flooded with seawater, following a launch abort accident. Spectral shift absorber (SSA) additives in the core and thin SSA coatings on the outer surface of the core can also be used to ensure subcriticality in such an accident. This paper investigates the effects of various SSAs (Re, Ir, Eu-151, B-10 and Gd-155) on the temperature and burnup reactivity coefficients and the operating lifetime of the S4 reactor at a steady thermal power of 550 kW. The calculations of the burnup, reactivity feedback coefficient used a mixture of the top 10 light and top 10 heavy fission products plus Sm-149 and are performed for isothermal reactor core and reflector temperatures of 1200 and 900 K. In this fast spectrum space reactor, SSAs markedly increase fuel enrichment and decrease the burnup reactivity coefficient, but only slightly decrease the temperature, reactivity feedback coefficient. With no SSAs, the UN fuel enrichment is lowest (58.5 wt.%), the temperature and burnup reactivity coefficients are the highest (−0.2709 ¢/K and −1.3470 $/at.%), and the estimated operating lifetime is the shortest (7.6 years). The temperature and burnup reactivity coefficients decrease to −0.2649 ¢/K and −1.0230 $/at.%, and the operating lifetime increases to 8.3 years when rhenium additives are used. With europium-151 and gadolinium-155 additions, fuel enrichment (91.5 and 94 wt.%) and operating lifetime (9.9 and 9.8 years) are the highest and both the temperature reactivity feedback coefficient (−0.2382 and −0.2447 ¢/K) and the burnup reactivity coefficient (−0.9073 and −0.8502 $/at.%) are the lowest.     

Assessment of reactor vessel integrity (ARVI)

The assessment of reactor vessel integrity (ARVI) project involved a total of nine organizations from Europe and USA. The work consisted of experiments and analysis development. The modeling activities in the area of structural analyses were focused on the support of EC-FOREVER experiments as well as on the exploitation of the data obtained from those experiments for modeling of creep deformation and the validation of the industry structural codes. Work was also performed for extension of melt natural convection analyses to consideration of stratification, and mixing (in the CFD codes). Other modeling activities were for (1) gap cooling CHF and (2) developing simple models for system code. Finally, the methodology and data was applied for the design of IVMR severe accident management scheme for VVER-440/213 plants.The work was broken up into five packages. They were divided into tasks, which were performed by different partners. The major experimental project continued was EC-FOREVER in which data was obtained on in-vessel melt pool coolability. In previous EC-FOREVER experiments data was obtained on melt pool natural convection and lower head creep failure and rupture. Those results obtained were related to the following issues: (1) multiaxial creep laws for different vessel steels, (2) effects of penetrations, and (3) mode and location of lower head failure. The two EC-FOREVER tests reported here are related to (a) the effectiveness of gap cooling and (b) water ingression for in vessel melt coolability. Two other experimental projects were also conducted. One was the COPO experiments, which was concerned with the effects of stratification and metal layer on the thermal loads on the lower head wall during melt pool convection. The second experimental project was conducted at ULPU facility, which provided data and correlations of CHF due to the external cooling of the lower head.     

Integrated modular water reactor (IMR) design

Integrated modular water reactor (IMR) has been developed as one of the advanced small-scale light water reactors, with a thermal output of 1000 MW. The IMR adopts natural circulation and self-pressurization in the primary cooling system, and a reactor vessel built-in steam generators. The core design has been performed using the current light water reactor technology. Thermal-hydraulic sensitivity analyses have been done from the viewpoint of the departure from nucleate boiling (DNB) limitation. The IMR core, with 97 21×21-type fuel assemblies and natural circulation in the primary coolant system, shows a good nuclear and thermal-hydraulic behavior and good allowable margins for the DNB phenomenon. The reactivity change with burnup is about 1%Δk by using burnable absorbers, and only 12 rod-cluster-controls are used through the operating cycle. The 20 m-height reactor vessel encloses steam generators in vapor and liquid portions. Plant dynamic analyses have been also performed in order to evaluate the IMR behavior from the viewpoints of plant operation and control. This study shows that the IMR will operate with enough margins for the core safety and will be stably controlled for load demand changes expected during normal operations.     

Space reactor power systems with no single point failures

Nuclear reactor power systems could revolutionize space exploration and support human outpost on the moon and Mars. This paper reviews various energy conversion technologies for use in space reactor power systems and provides estimates of the system's net efficiency and specific power, and the specific area of the radiator. The suitable combinations of the energy conversion technologies and the nuclear reactors, classified based on the coolant type and cooling method, for best system performance and highest specific power, are also discussed. In addition, a number of power system concepts with both static and dynamic energy conversion, but with no single point failures in reactor cooling, energy conversion and heat rejection, and for nominal electrical powers up to 110 kW_e, are presented. The first two power systems employ reactors cooled with lithium and sodium heat pipes, SiGe thermoelectric (TE) and alkali-metal thermal-to-electric conversion (AMTEC), and potassium heat pipes radiators. The reactors heat pipes operate at a fraction of the prevailing capillary or sonic limit, and in the case of a multiple heat pipes failure, those in the adjacent modules remove the additional heat load, thus maintaining the reactor adequately cooled and the power system operating at a reduced power. The third power system employs SiGe TE converters and a liquid metal cooled reactor with a divided core into six sectors that are neurotically and thermally coupled, but hydraulically decoupled. Each sector has a separate energy conversion loop, a heat rejection loop, and a rubidium heat pipes radiator panel. When a core sector experiences a loss-of-coolant, the fission power of the reactor is reduced, and that generated in the sector in question is removed by the circulating coolant in the adjacent sectors. The fourth power system employs a gas cooled reactor with a core divided into three identical sectors, and each sector is coupled to a separate Closed Brayton Cycle (CBC) loop with He-Xe binary mixture (40 g/mol) working fluid, a secondary loop with circulating liquid Nak-78, and two water heat pipes radiator panels.     

Core Safety Characteristics of High Flux Safe Reactor (HFSR)

Abstract

Transient analyses are performed for graphite moderated helium-cooled high flux reactor to obtain the high flux safe reactor design. In order to promote the safety of the high flux reactor, the present design adopts the pebble bed reactor and its fuel technology. In the transient analyses, among the postulated off-normal events and accidents, the reactivity accident followed by a loss of helium forced circulation with system depressurization is found to be the severest potential event which may threaten the reactor safety from fission products release point of view. Several neutronic and thermal-hydraulic design parameters are indicated and exploited to promote the reactor safety. Neutronic and core thermal-hydralic models are proposed and used to simulate the reactor responses to the off-normal events and accidents. As the results of the transient analyses and accident simulation, safe and optimal design parameters are obtained which provide high thermal neutron flux with a desirable spectrum and large usable volume constrained by safety limitations.     

SSTAR: The US lead-cooled fast reactor (LFR)

It is widely recognized that the developing world is the next area for major energy demand growth, including demand for new and advanced nuclear energy systems. With limited existing industrial and grid infrastructures, there will be an important need for future nuclear energy systems that can provide small or moderate increments of electric power (10-700 MWe) on small or immature grids in developing nations. Most recently, the global nuclear energy partnership (GNEP) has identified, as one of its key objectives, the development and demonstration of concepts for small and medium-sized reactors (SMRs) that can be globally deployed while assuring a high level of proliferation resistance. Lead-cooled systems offer several key advantages in meeting these goals. The small lead-cooled fast reactor concept known as the small secure transportable autonomous reactor (SSTAR) has been under ongoing development as part of the US advanced nuclear energy systems programs. It is a system designed to provide energy security to developing nations while incorporating features to achieve nonproliferation goals, anticipating GNEP objectives. This paper presents the motivation for development of internationally deployable nuclear energy systems as well as a summary of one such system, SSTAR, which is the US Generation IV lead-cooled fast reactor system.     

The prestressed cast iron reactor pressure vessel (PCIPV)

The concept of the prestressed cast iron reactor pressure vessel (PCIPV) emerges from the utilization of cast iron in the design of radiation and thermal shields. The principles of construction are explained using a model which is at present being assembled. Salient differences between the proposed vessel concept and a prestressed concrete reactor pressure vessel (PCPV) are discussed.     

Safety characteristics of the multipurpose fast reactor (MPFR)

The safety characteristics of a long-life multipurpose nuclear reactor (MPFR) with self-sustained liquid metallic fuel and lead coolant, which is proposed to meet the requirements for the energy production in the future, were investigated. The application of liquid plutonium–uranium metallic alloys used as a nuclear fuel demonstrated high potential to reach excellent reactor shutdown characteristics against anticipated transients without scram such as unprotected loss-of-flow and unprotected transient overpower. The calculations indicated that the thermal expansion of liquid fuels would cause the negative reactivity insertion that would be larger in magnitude than any other thermally induced reactivity changes. This created the reactivity balance for the passive shutdown and power stabilization capabilities of the MPFR core. It was found that MPFR satisfies such design characteristics to be a potential candidate providing the replacement of fossil fuels by alternative energy sources in the next century.     

Noise measurements on the KEMA suspension test reactor (KSTR) and the Dodewaard boiling water reactor

Noise measurements have been performed on the neutron flux and the fuel concentration of the KEMA Suspension Test Reactor. The behaviour of the autopower spectral density function of the neutron flux is given with respect to reactor power and fuel burn up.

To investigate the concentration fluctuations the load of the suspension pump has been analysed.

Apart from the measurements on the Suspension Test Reactor, noise analysis has been performed on the temperature on the outside of the feed water inlet nozzle and on its thermal sleeve of the Dodewaard Boiling Water Reactor.

Temperature distributions indicate the leakage of feed water through the annulus, formed by the nozzle and the thermal sleeve. Amplitude distribution measurements show the magnitude of the circumferential temperature fluctuations on the thermal sleeve. From cross correlation measurements of the temperature fluctuations in nozzle and thermal sleeve the flow pattern in this region has been determined. From the results, temperature fluctuations at the inner radius and the bore of the feed water inlet nozzle can be derived.     

Static design of a liquid-salt-cooled pebble bed reactor (LSPBR)

A renewed interest has been raised for liquid-salt-cooled nuclear reactors. The excellent heat transfer properties of liquid-salt coolants provide several benefits, like lower fuel temperatures, higher average coolant temperature, increased core power density and better decay heat removal, and thus higher achievable core power. In order to benefit from the on-line refueling capability of a pebble bed reactor, the liquid salt pebble bed reactor (LSPBR) is proposed. This is a high temperature pebble bed reactor with a fuel design similar to existing HTRs, but using a liquid-salt as coolant. In this paper, the selection criteria for the liquid-salt coolant are described. Based on its neutronic properties, LiF–BeF₂ (flibe) was selected for the LSPBR. Two designs of the LSPBR were considered: a cylindrical core and an annular core with a graphite inner reflector. Coupled neutronic thermal-hydraulic calculations were performed to obtain the steady state power distribution and the corresponding fuel temperature distribution. Calculations were performed to investigate the decay heat removal capability in a protected loss-of-forced cooling accident. The maximum allowable power that can be produced with the LSPBR is hereby determined.     

Objectives in developing the European pressurized water reactor (EPR)

In 1991, major German utilities and Electricité de France agreed to develop, together with Siemens and Framatome, the nuclear island for the next generation of nuclear power plants. This nuclear island design is based on German and French experience in the construction and operation of pressurized water reactors. The major step in the evolutionary European pressurized water reactor (EPR) design is the systematic inclusion of events beyond classic design events. The mitigation of core melt accidents by special means, such as reinforcement of the containment function, primary loop depressurization, hydrogen reduction and ultimate heat removal, is part of the design, in addition to a number of features which increase the reliability of plant operation and accident control. Nevertheless, utilities and designers are aware of the economic challenge facing the EPR by the need to compete with other nuclear power plant designs and fossil plants for electricity generation.