Thermal-hydraulic analysis

Thermal-Hydraulic Analysis must be extended to at least 24 hours into the accident. This study used calculations for feed-and-bleed operation with a charging pump, and with gravity teed from the refueling water storage tank (RWST),... 

Chu, K. J., and A. E. Dukler, 1974, Statistical Characteristics of Thin Wavy Films, AIChE J. 20 695. (3) Chu, P. T., H. Chelemer, and L. E. Hochreiter, 1973, THINC IV An Improved Program for Thermal Hydraulic Analysis of Rod Bundle Cores, WCAP-7965, Westinghouse Electric Corporation, Pittsburgh, PA. (App.)... 

Rowe, D. S, 1973, COBRA IIIC A Digital Computer Program for Steady State and Transient Thermal Hydraulic Analysis of Rod Bundle Nuclear Fuel Elements, BNWL-1695, Battelle Northwest Laboratory, Richland, WA. (3)... 

This implies that the LMTD or MTD as computed in equations 20 through 26 may not be a representative temperature difference between the two heat-transferring fluids for all tubes. The effective LMTD or MTD would be smaller than the value calculated, and consequently would require additional heat-transfer area. The tme value of the effective MTD may be determined by two- or three-dimensional thermal—hydraulic analysis of the tube bundle. Baffle—Tube Support Plate Area. The portion of a heat-transfer tube that passes through the flow baffle—tube support plates is usually considered inactive from a heat-transfer standpoint. However, this inactive area must be included in the determination of the total length of the heat-transfer tube. 

The thermal-hydraulic analysis and thermal stress analysis of plate-type SO3 decomposer had... 

Subchannel analysis is commonly used for thermal hydraulic analysis of single fuel subassemblies. Bulk average values characterize the fluid dynamics and thermal coolant conditions in each subchannel. Thermal and hydraulic interactions between subchannels are taken into account. 

Plant safety assessment projects improve the abilities of designers, operators, and regulators to evaluate the safety of their plants through the use of internationally accepted methodologies and computer analysis codes. An example of a key accomplishment in this area is training technical personnel to use the COBRA-SFS computer code to perform thermal-hydraulic analysis (shown in Fig. 4). 

Design of main system arrangement, of key components such as a steam reformer and a steam generator and of control method, especially start-up procedure and development of computer code for transient thermal-hydraulics analysis have been conducted. The steam generator is designed to have a safety function as an interface. 

Structural mechanics analysis indicates that the acceptable temperature difference between the primary temperatures from IHX outlets is about 40 K. Thermal hydraulics analysis indicates that with one module isolated in one loop, the temperature difference is 10 K at 90 % power. With 2 modules isolated, this temperature difference is 24 K at 67 % power. To achieve this, it is necessary to reduce the core flow (to 67 %) in proportion to the power as well as the sodium flow in the unaffected loop (50 %) as in the affected loop (50 %) The... 

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. 

Studies on development, validation and application of thermal-hydraulics analysis... 

The assessment should confirm that the framework for making these expert judgements is sound and the basis for the judgement is stated and shown to be valid as far as possible. This should take account of the thermal-hydraulic analysis that has been carried out, analyses for other similar plants and applicable research data. The quantification of the containment event trees should take account of the interdependences between the various phenomena that are being modelled. 

W. J. METE VIA, A. HUSAIN, and T. R. HENCEy, Relap-4 Thermal/Hydraulic Analysis of the Maine Yankee Spent Fuel Pit, Conf. Reactor Operating Experience, Trans. Am.Nucl. Soc., 21, SuppL 1, 18 (1975). 

K. V. MOORE and H. RETTIG, Relap-4,-A Computer Program for Transient Thermal Hydraulic Analysis, ... 

The canister temperatures and pressures shown by thermal hydraulic analysis are demonstrated to be maintained within their design basis limits. 

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. 

RETRAN (REactor TRansient ANalysis) is a best-estimate transient thermal-hydraulic analysis computer program (sponsored by EPRI) designed to provide analysis capabilities for BWR and PWR transients, small-break LOCAs, balance-of-plant modeling, and anticipated transients without scram (ATWS). 

Owing to the comprehensiveness of specific technical information such as respective FSAR chapters, EOF background documents, supporting thermal-hydraulic analysis, etc., it may be necessary to study some documentation in advance. The content and scope of such information, which will be made available to the review team, needs to be agreed upon beforehand and should reflect the intended scope of the review. 

SENICKI, J.J., WADE, D C., Thermal hydraulic analysis of the encapsulated nuclear heat source, ICONE-9 (Proc. 9 Int. Conf on Nuclear Engineering, Nice, France, April 8-12, 2001), ICONE9-113, ASME. 

FIG. XXIX-7. Simplified model of CHTR loop used in thermal-hydraulic analysis. 

The design analysis has been carried out using well-established nuclear computer codes. The validation and verification of those codes is described in topical reports. A summary of the major codes used in the thermal-hydraulic analysis is presented below. 

A thermal-hydraulic analysis is performed to determine the plant response for a Design Basis SGTR, the integrated primary-to-secondary break flow, and the mass releases from the ruptured and intact steam generators to the condenser and to the atmosphere. This information is then used to calculate the radioactivity release to the environment and the resulting radiological consequences. 

Over the next several years, the PRA was detailed to the point it included detailed fault trees of the mechanical, electrical, and instrumentation and control systems and the scope was expanded to include shutdown, fire, flood events, and large release frequency and off-site dose quantifications. Core damage frequency PRA was supported by extensive plant thermal-hydraulic analysis to justify success criteria. Extensive testing and thermal-hydrauUc analysis, to support containment integrity during core melt sequences, underpirmed the large release PRA. 

Velusamy, K., et. al. Thermal-hydraulic Analysis in the Design of PFBR Inner Vessel, Proceedings of the Fourth International Conference on Liquid Metal Engineering and Technology - Avignon, France, 1988. 

RECTOR, D.R., CUTA, I.M., LOMBARDO, N.J., MICHENER, T.E., WHEELER, C.L., OBRA-SFS A Thermal-Hydraulic Analysis Code, Rep. PNL-6049, Battelle Pacific Northwest Laboratory, Richland, WA (1986). 

Reactor physics and thermal hydraulics analysis methods that have been developed for earlier HTGR designs will need to be updated and qualified for apphcation to the GT-MHR... 

Reactor physics and thermal hydraulics analysis methods and codes... 

The thermal hydraulic analysis showed that the cladding temperature for both types of fuel elements are almost the same, e.g. maximum cladding temperature on the internal surface of the tube is 425 K and 426 K for LEU and HEU respectively. However, the maximum heat flux on the external wall of the fuel tube is 2,61 MW/m and 2,04 MW/m for the LEU and HEU fuel respectively. Moreover, the data acquired from the performed measurements point out that the coefficient of hydraulic resistance for LEU fuel elements exceeds by around 30% the resistance coefficient for the HEU fuel (Dorosz et al. 2009). 

COBRAIIIc is a widely used code for the thermal-hydraulic analysis of nuclear reactor cores. Its approach is based on subchannel analyses where conservation equations are axially solved and coupled through mixing coefficients (Todreas Kazimi 2001). 

In accordance with the requirement of Slovak Nuclear Regulatory Body in BNS 1.4.2 and general international requirements (SSG-3 and SSG-4), the PSA model shall be a comprehensive description of the NPP. In order to have useful information from PSA, a realistic model of the plant is necessary. To account for all relevant dependencies, all necessary support systems are included in the model such as HVAC, electrical supply and heat sink are included in the PSA model. The success criteria for front-end safety systems are derived based on detailed thermal hydraulic analysis specific for the plant and the fuel under analysis, both for core damage and for radiological releases. 

---