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[en] Objective: • Verification of Technical Basis for addition of Fire Fighting Water (FFW) under Severe Accident Management Guidelines (SAMG) for PHWR. • Suitable model development to address possible phenomena. • Analysis of small PHWR (220 MWe) and large PHWR (540 MWe) with the developed model with and without SAMG action.
[en] Highlights: • The present study deals with the issue of loosening of the nut in the Grayloc joint due to flow induced vibration and fret in the feeder pipes in contact due to differential creep in the neighbouring channels. • Accelerated test has been done on the Grayloc joint on simulated flow induced vibration to study the effect of loosening of the nut. • In the present accelerated test, the component has not been led to failure (loosening) and an estimation of its service life has been approached based on the severity of test. • The inverse square law approach based on PSD comparison for severity of test have been used to correlate the actual operational hours and the Laboratory test hours to verify the loosening of the Grayloc nut for the present study. • By inverse power law approach, the minimum number of reactor-hours equivalent to 80 h of testing is 46,080 h (5.26 full power years). - Abstract: The present study outlines the accelerated testing procedure of a Grayloc joint assembly for possible loosening of its nut due to flow induced vibration. The concern of the Grayloc nut getting loosened in the absence of a lock nut due to flow induced vibration and the resulting fretting in the feeder pipes in contact due to differential creep in the neighbouring channels has been addressed here. The severity of the test was decided based on actual site measurement under different operating flow conditions and comparison of power spectral density (PSD). The laboratory test results were extrapolated for estimation of life of the component under operating condition using inverse power law approach. The uniqueness of the accelerated test is that the component under test has not been led to failure for assessing its operating life unlike conventional accelerated testing. From the tests and analysis, it was deduced that 80 h of accelerated laboratory testing was equivalent to 5.26 full power years (46,080 h) of the reactor operating life. The test duration was enough to obtain approval from the regulatory body to operate the reactor.
[en] The pressure tubes of pressurized heavy water reactor (PHWR) operate under high temperature high pressure aqueous environment and are subjected to fast neutron irradiation. These are rolled joined with end fittings at both ends and are exposed to coolant temperature of 300°C and internal pressure of about 10 MPa. While operating under the above mentioned conditions pressure tubes’ average service life is limited to 30 years because of degradation mechanisms such as creep and delayed hydride cracking near rolled joints. To increase the life of coolant channel, Long Life Coolant Channel (LLCC) design has been proposed. LLCC consists of a pressure tube and an insulator made of Yttria Stabilized Zirconia (YSZ). The insulator shields the pressure tube against high temperature coolant. The pressure tube is in direct contact with moderator. This design brings the operating temperature of pressure tube close to moderator’s temperature. Reduction in pressure tube’s temperature reduces the impact of creep and hydriding related degradations thus enhancing its life. Though the pressure tube service life, in LLCC design is enhanced under normal operating condition, it must also be able to effectively transfer the heat from fuel to moderator in accidental conditions. Hence, heat transfer capability of LLCC design under the postulated loss of coolant accident (LOCA) with the total loss of emergency core cooling system (LOECCS) has been assessed. A one dimensional finite difference code, including conduction, convection and radiation solution scheme, has been developed for the purpose. Heat transfer capability has been accessed for different values of thermal conductivity of YSZ Insulator. It is found that LLCC effectively transfers decay heat from fuel elements to moderator (author)
[en] Highlights: • Core Damage Frequency and Large Early Release Frequency. • Multi –Unit Risk Metrics. • Aggregation of CDF of NPP through Mean Values. • Aggregation of CDF of NPP as Random Variable. - Abstract: The nuclear generating sites around the world are mostly twin unit and multi-unit sites. The PSA risk metrics Core Damage Frequency (CDF) and Large Early Release Frequency (LERF) currently are based on per reactor reference. The models for level 1 and level 2 PSA have been developed based on single unit. The Fukushima accident has spawned the need to address the issue of site base risk metrics, Site Core Damage Frequency (SCDF) and Site Large Early Release Frequency (SLERF), on the site years rather than reactor years. It is required to develop a holistic framework for risk assessment of a site. In the context of current study, the holistic framework refers to integration of risk from all units, dependencies due to external events and operation time of individual units. There is currently no general consensus on how to arrive at site-specific risk metrics. Some documents provide suggestions for site CDF and site LERF. This paper proposes a new method of aggregation of risk metric from the consideration of operating time of individual units under certain assumptions with a purpose to provide a new conceptual aspect for multi-unit PSA. The result of a case study on hypothetical data shows that site level CDF is not sum of CDF of all units but around 18% higher than unit level CDF. When the CDF is considered to be a random variable then, the new methodology produces site CDF as 50% higher than single unit CDF. These two approaches have been detailed in the paper. For a general data set of CDF for individual units, site CDF would more than individual unit CDF however, it would not be multiples of a single unit value.
[en] Finite Element Modeling is one of the efficient analytical tools for analysis of complicated structures subjected to variety of loads. However the reliability of the analyses is always questionable due to idealizations and assumptions made in the design. The model can be more realistic if it is refined based on experimental support. This paper presents refinement of finite-element model of Koyna Dam-foot Power House (KDPH) building, which is structurally complicated and asymmetrical. The dynamic properties of the building have been identified experimentally through Ambient Vibration Tests (AVT). The building has also been elaborately modeled analytically. The finite-element model is further refined so as to minimize the differences between analytical and the measured natural frequency of the building. The final refined finite-element model of KDPH building is able to produce natural frequency in good agreement with the measured natural frequency of the building. (author)