Reactor core thermal hydraulics

The computer codes which are used to carry out the anticipated operational occurrences and DBA analysis should be properly verified and validated. This includes the codes used to predict the behaviour of the reactor core, thermal-hydraulic codes and the radiological release and consequence codes. In addition, the analysts and users of the codes should be suitably qualified, experienced and trained. 

Thermal hydraulic design. The preliminary core thermal hydraulic design reported here needs to be continued using multidimensional methods to establish the heat removal capability and coolant pump sizing for the target power level and for assessing the reactor s performance in anticipated transients. 

M. Richards, A. Shenoy, Y. Kiso, N. Tsuji, N. Kodochigov, and S. Shepelev Thermal Hydraulic Design of a Modular Helium Reactor Core Operating at 1 000°C Coolant Outlet Temperature, Proceedings of the 6 International Conference on Nuclear Thermal Hydraulics, Operations and Safety (NUTHOS-6), October 4-8, 2004, Nara, Japan, Atomic Energy Society of Japan, Tokyo, Japan (2004). 

Kelly JM, Stewart CW, Cuta JM (1992) VIPRE-02 - A two-fluid thermal-hydraulics code for reactor core and vessel analysis Mathematical modeling and solution methods. Nucear Technology 100 246-259... 

The steady-state performance of the reactor core is discussed in the following subsections on nuclear, thermal/hydraulics, structural, and fuel performance. 

The CAREM reactor under development in Argentina is a 100 MW(th) (about 27 MW(e)) design based on natural circulation. It has an integrated primary circuit comprising the core, steam generators, control rods with their drive mechanisms and the entire primary coolant. Several experiments examining core neutronics and thermal hydraulics have been conducted in test facilities. The construction of a prototype is planned. 

In 1993 the chief reactor designer had calculated the neutron-physical and thermal hydraulic characteristics of reactor core for the RCCA drop time up to 10 sec, keeping in mind the case when all RCCA are inserted except the most effective RCCA. It was shown that requirements of regulatory safety documents were met. However, the results obtained were not used to cancel operational restrictions according to a conservative principle. 

Computer code CALPER — a thermal hydraulic subchannel analysis code for the assessment of coolant local conditions in the fuel assemblies and in the core of PWRAVWER-type nuclear reactors... 

In the period of 1998-99, two sets of experiments focused on problems of rapid decrease of concentration of boric acid in reactor coolant at nuclear reactor core inlet were performed at the University of Maryland, US, under the auspices of OECD. The situation, when there is an inadvertent supply of boron-deficient water into the reactor vessel, could lead to a rapid (very probably local) increase of reactor core power in reactor, operated at nominal power, or to a start of fission reaction in shut-down reactor (secondary criticality). In the above mentioned experiments the transport of boron-deficient coolant through reactor downcomer and lower plenum was simulated by flow of cold water into a model of reactor vessel. These experiments were selected as the International Standard Problem ISP-43 and organisations, involved in thermal — hydraulic calculations of nuclear reactors, were invited to participate in their computer simulation. Altogether 10 groups took part in this problem with various CFD codes. The participants obtained only data on geometry of the experimental facility, and initial and boundary conditions. 

The database information consists of the thermal-hydraulic/radiological variables of the core, reactor coolant system, containment system, and safety systems which are generated from integrated severe accident analysis code, such as MAAP or MELCOR program. 

The reactor coolant pumps provide sufficient forced circulation flow through the Reactor Coolant System to assure adequate heat removal from the reactor core during power operation. A low limit on reactor coolant pump flow rate (i.e., design flow) is established to assure that Specified Acceptable Fuel Design Limits (SAFDLs) are not exceeded. Design flow is derived on the basis of the thermal-hydraulic considerations presented in Section 5.2. 

The reactor coolant pumps are sized to deliver flow that equals or exceeds the design flow rate utilized in the thermal hydraulic analysis of the Reactor Coolant System. Analysis of steady-state and anticipated transients is performed assuming the minimum design flow rate. Tests are performed to evaluate reactor coolant pump performance during the post-core load hot functional testing to verify adequate flow. 

Thermal-hydraulic codes which model the behaviour of the reactor core and the associated coolant systems during normal operation and following accidents,... 

These design objectives were carried over to the work on the power reactor PIUS, basically a pressurized water reactor (PWR) in which the primary system has been rearranged in order to accomplish an efficient protection of the reactor core and the nuclear fuel by means of thermal-hydraulic characteristics, in combination with inherent and passive features, without reliance on operator intervention or proper functioning of any mechanical or electrical equipment. Together with wide operating margins, this should make the plant design and its function, in normal operation as well as in transient and accident situations, much more easily understood and with less requirements on the capabilities and qualifications of the operators. 

The thermal-hydraulic analysis is performed using special codes that receive the fission power spatial deposition from the reactor analysis pin power deconstruction and perform a heat transfer and fluid flow calculation to calculate the expected temperature profile throughout the core. Because neutron cross sections depend on temperature, this calculation must be iteratively linked into the previous six steps. In many cases, this linkage results in an expansion of the parametric expansions of the neutron cross sections (described as part of Step 6) to be a function of temperature and total assembly fission power. 

Though the fission in the reactor core is analyzed by the nuclear consideration, the heat generated by the reactor through fission and its use in the generation of power for a given reactor core is largely limited by thermal processes and material properties rather than by nuclear considerations. The safety analysis of the reactor during normal and abnormal operational conditions involves detailed thermal-hydraulics analysis. 

Natural circulation systems may undergo thermal-hydraulic instabilities under low-power and low-pressure conditions, which occur during start-up. The void reactivity feedback and void fraction fluctuations in the reactor core would create power oscillations during start-up. Three kinds of thermal-hydraulic instabilities may occur during start-up in natural circulation BWRs, which are as follows (1) geysering induced by condensation, (2) natural circulation instability induced by hydrostatic head fluctuation in steam separators, and (3) density wave instabilities. 

Such approaches, which aim at the establishment of a single quantity to represent the fraction of the core inventory that would be released in the event of a severe reactor accident, have now been widely abandoned. As has been summarily described by Malinauskas and Kress (1991), currently fission product release is recognized to be both scenario- and plant-specific, and it has become necessary to replace simple release fractions by computer codes that couple release and transport with thermal-hydraulic aspects. In particular it has been realized that the timing of fission product release can be of great importance and that the releases... 

Such interactions can only occur, however, when the volatile fission products and the primary aerosols appear simultaneously in the primary system, in spite of the large differences in their volatilization behavior. As was discussed above, uniform thermal-hydraulic conditions do not prevail within the reactor core during a severe accident (for example, the peripheral fuel rods may fail relatively late in the accident sequence, at a point when a large part of the central rods may already be molten) and it can be assumed that the broad time-envelope of significant release of structural aerosols will encompass the release of the volatile fission products. However, as was mentioned in Section 7.3.1.2., the amount of primary aerosols formed and the timing of their formation depend highly on the specific accident sequence this is particularly true for the control rod materials. 

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