• Nuclear Engineering and Technology
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• v.49 no.5
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• pp.968-978
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• 2017

Three small-break loss-of-coolant accident (SBLOCA) tests with safety injection pumps were carried out using the integral-effect test loop for SMART (System-integrated Modular Advanced ReacTor), i.e., the SMART-ITL facility. The types of break are a safety injection system line break, shutdown cooling system line break, and pressurizer safety valve line break. The thermal-hydraulic phenomena show a traditional behavior to decrease the temperature and pressure whereas the local phenomena are slightly different during the early stage of the transient after a break simulation. A safety injection using a high-pressure pump effectively cools down and recovers the inventory of a reactor coolant system. The global trends show reproducible results for an SBLOCA scenario with three different break locations. It was confirmed that the safety injection system is robustly safe enough to protect from a core uncovery.

• Nuclear Engineering and Technology
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• v.33 no.1
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• pp.12-24
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• 2001

The post-LOCA long term cooling (LTC) performance of the Korean Standard Nuclear Power Plant (KSNPP) is analyzed for both small break loss-of-coolant accidents (LOCA) and large break LOCA at cold leg. The RELAP5/MOD3.2.2 beta code is used to calculate the LTC sequences based on the LTC plan of the Korean Standard Nuclear Power Plants (KSNPP). A standard input model is developed such that LOCA and the followed LTC sequence can be calculated in a single run for both small break LOCA and large break LOCA. A spectrum of small break LOCA ranging from \ulcorner.02 to 0.5 k2 of break area and a double-ended guillotine break are analyzed. Through the code calculations, the thermal-hydraulic behavior and the boron behavior are evaluated and the effect of the important action including the safety injection tank (SIT isolation and the simultaneous injection in LTC procedure is investigated. As a result, it is found that the sufficient margin is available in avoiding the boron precipitation in the core. It is also found that a further specific condition for the SIT isolation action need to be setup and it is recommended that the early initiation of the simultaneous injection be taken for larger break LTC sequences.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.417-422
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• 1996

The KNGR has an advanced ECCS design feature which employs four mechanically-separated SI trains where each train consisting of one HPSI pump and one SIT injects ECC water directly into the reactor vessel downcomer annulus. To demonstrate that the KNGR ECCS design features meet the EPRI ALWR requirements of no core uncovery for a break of up to 6 inch diameter, small break LOCA cases with various break sizes were analyzed using a best-estimate analytical procedure. Two kinds of break locations are considered: cold leg and DVI line breaks. It was observed that the KNGR ECC design can tolerate a cold leg break of up to 10 inches with no core uncovery. However. since DVI line break with 6 inch diameter undergoes slight core uncovery. further investigation is required for KNGR SIS optimization.

• Nuclear Engineering and Technology
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• v.16 no.1
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• pp.11-17
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• 1984

Sample calculations were done for KORI-1 to develop a better understanding of what happens after very small LOCA ($\leq$0.05 ft$^2$). For a water-side break with the break size larger than 0.006 ft$^2$, fluid-loss through break exceeds the makeup. If the break size is larger than 0.008ft$^2$, decay heat can be completely removed through break. Based on these results, it was concluded that KORI-1 is fairly safe for the whole spectrum of sizes in very small LOCA. However, for the reactor with 900 MWe or 1200 MWe, a certain spectrum of sizes in very small LOCA should be carefully considered. In the accident sequence the transition from natural circulation to pool boiling or from pool boiling to natural circulation may be troublesome to the operator or in the safety analysis. Operator's intervention was discussed; primary pump shutoff, HPI pump shutoff, break isolation, and opening relief valve. It was proved that continuous operation of HPI pumps after shutdown will not threaten the integrity of the primary system.

• Nuclear Engineering and Technology
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• v.53 no.10
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• pp.3256-3263
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• 2021

The main steam line break accident is an essential initiating event of the pressurized water reactor. In present work, the fuzzy set theory and the signal-based fault detection method has been used to detect the occurrence and diagnosis of the location and break area for the small scale MSLB. The models are validated by the AP1000 accident simulator based on MAAP5. From the test results it can be seen that the proposed approach has a rapid and proper response on accident detection and location diagnosis. The method proposed to evaluate the break area shows good performances for small scale MSLB with the relative deviation within ±3%.

• Nuclear Engineering and Technology
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• v.54 no.1
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• pp.193-206
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• 2022

Several algorithms for nuclear power plant (NPP) break event detection, isolation, localization, and size estimation are proposed. A break event can be promptly detected and isolated after its occurrence by simultaneously monitoring changes in the sensing readings and by employing an interquartile range-based isolation scheme. By considering the multi-sensor data block of a break to be rank-one, it can be located as the position whose lead field vector is most orthogonal to the noise subspace of that data block using the Multiple Signal Classification (MUSIC) algorithm. Owing to the flexibility of deep neural networks in selecting the best regression model for the available data, we can estimate the break size using multiple-sensor recordings of the break regardless of the sensor types. The efficacy of the proposed algorithms was evaluated using the data generated by Maanshan NPP simulator. The experimental results demonstrated that the MUSIC method could distinguish two near breaks. However, if the two breaks were close and of small sizes, the MUSIC method might wrongly locate them. The break sizes estimated by the proposed deep learning model were close to their actual values, but relative errors of more than 8% were seen while estimating small breaks' sizes.

• Proceedings of the Korean Society of Tribologists and Lubrication Engineers Conference
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• 2000.06a
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• pp.159-165
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• 2000

In this paper the residual stresses on lubricated sliding surfaces were measured during break-in procedure and up to scuffing by the X-ray diffraction method. The cylinder-on-disk type tribometer was used with the line-contact geometry. Scuffing tests were done using a constant load. In the break-in procedure the loads were increased from very low values in several steps. It was found that the sliding surfaces with break-in represented relatively higher values of residual compressive stresses than those without break-in. The residual stresses below the surfaces showed the small amount of stress increases. The results of scuffing tests with and without break-in showed the same trends as break-in tests did. However, in case of tests with break-in procedure the stresses below the surfaces showed very large increases in the residual compressive stresses. From the tests of break-in and scuffing, it was found that the increases in scuffing lives were related with the increases of residual stresses on the lubricated sliding surfaces with break-in.

• Tribology and Lubricants
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• v.16 no.3
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• pp.182-187
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• 2000

In this paper the residual stresses on lubricated sliding surfaces were measured during break-in procedure and up to scuffing by the X-ray diffraction method. The cylinder-on-disk type tribometer was used with the line-contact geometry. Scuffing tests were done using a constant load. In the break-in procedure the loads were increased from very low values in several steps. It was found that the sliding surfaces with break-in represented relatively higher values of residual compressive stresser than those without break-in. The residual stresses below the surfaces showed the small amount of stress increases. The results of scuffing tests with and without break-in showed the same trends as break-in tests did. However, in case of tests with break-in procedure the stresses below the surfaces showed very large increases in the residual compressive stresses. From the tests of break-in and scuffing, it was found that the increases in scuffing lives were related with the increases of residual stresses on the lubricated sliding surfaces with break-in.

• Nuclear Engineering and Technology
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• v.39 no.4
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• pp.287-298
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• 2007

As part of the $AP1000^{TM}$ pressurized water reactor design certification program, a series of integral systems tests of the nuclear steam supply system was performed at the APEX-1000 test facility at Oregon State University. These tests provided data necessary to validate Westinghouse safety analysis computer codes for AP1000 applications. In addition, the tests provided the opportunity to investigate the thermal-hydraulic phenomena expected to be important in AP1000 small-break loss of coolant accidents (SBLOCAs). The APEX-1000 facility is a 1/4-scale pressure and 1/4-scale height simulation of the AP1000 nuclear steam supply system and passive safety features. A series of eleven tests was performed in the APEX-1000 facility as part of a U.S. Department of Energy contract. In all, four SBLOCA tests representing a spectrum of break sizes and locations were simulated along with tests to study specific phenomena of interest. The focus of this paper is the SBLOCA tests. The key thermal-hydraulic phenomena simulated in the APEX-1000 tests, and the performance and interactions of the passive safety-related systems that can be investigated through the APEX-1000 facility, are emphasized. The APEX-1000 tests demonstrate that the AP1000 passive safety-related systems successfully combine to provide a continuous removal of core decay heat and the reactor core remains covered with considerable margin for all small-break LOCA events.

• Nuclear Engineering and Technology
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• v.55 no.7
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• pp.2438-2446
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• 2023

To cool down a nuclear reactor core and prevent the fuel damage without a pump-driven active component during any anticipated accident, the passive emergency core cooling system (PECCS) was designed and adopted in an advanced light water reactor, i-POWER. In this study, for a validation of the cooling capability of PECCS, thermal-hydraulic integral effect tests were performed with the ATLAS facility by simulating intermediate and small break loss-of-coolant accidents (IBLOCA and SBLOCA). The test result showed that PECCS could effectively depressurize the reactor coolant system by supplying the safety injection water from the safety injection tanks (SITs). The result pointed out that the safety injection from IRWST should have been activated earlier to inhibit the excessive core heat-up. The sequence of the PECCS injection and the major thermal hydraulic transient during the SBLOCA transient was similar to the result of the IBLOCA test with the equivalent PECCS condition. The test data can be used to evaluate the capability of thermal hydraulic safety analysis codes in predicting IBLOCA and SBLOCA transients under an operation of passive safety system.