Progress in understanding fission product remobilization and hydrogen risk in water-cooled reactors: OECD/NEA THAI-3 project

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Appendix J of 10 CFR 50, “Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors,” sets the testing requirements for preoperational and periodic verification of the leak-tight integrity of the primary reactor containment, including systems and components which penetrate containment of light water-cooled power reactors, and establishes the acceptance criteria for such tests. The purposes of the tests are to assure that leakage through the primary reactor containment and systems and components penetrating primary containment would not exceed allowable leakage rate values as specified in the plants’ technical specifications. The current Appendix J requirements provide two options for performing the tests. Option A (fully deterministic) requires that after the preoperational leakage rate tests, a set of three integrated leak rate tests (ILRT — termed as Type A tests) shall be performed, at approximately equal intervals during each 10-year service period. Option B does not provide a quantitative requirement for scheduling the periodic Type A tests. However, NEI 94-01, “Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J,” and NRC Regulatory Guide (RG). The containment is a vital engineering safety feature of a nuclear power plant. It encloses the entire reactor and reactor coolant system and serves as a final barrier against the release of radioactive fission products to the environment under various accident conditions. Containment design is based on pressure and temperature loadings associated with a loss-of-coolant-accident resulting from a double-ended rupture of the largest pipe in the reactor coolant system. Recently, it has been reported that in certain 1.163, “Performance-Based Containment Leak Test Program,” provide guidelines for determining the frequencies of preoperational and periodic leak rate tests using a performance based approach. The Option B requirements and NEI 94-01 guidance are based on NUREG-1493, “Performance Based Containment Leak-Test Program.” Based on the information provided by the operating data and their risk significance, the study calculated the change in risk (in person-rem) to public for the 15 alternatives considered in the study. However, recognizing the non-sensitivity of risk to change in Type A leak rate testing frequencies, as depicted in NUREG-1493, a number of licensees are proposing changes to their Type A test frequencies using NRC risk-informed guidance in RG 1.174, “An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis.” The paper discusses certain key deterministic aspects of the risk-informed decision for the plant-specific changes in the ILRT frequencies.

The mobility of corrosion products is usually a concern in any process involving fluid flow. In a chemical plant, the result may be fouling of heat exchangers. In a water-cooled nuclear power reactor, fouling of fuel element surfaces is one result; the growth of radiation fields from radioactive corrosion products is another. The corrosion products are released to the flowing coolant, chiefly by slowly-corroding piping surfaces outside the reactor core. The dissolved and suspended corrosion products are carried into the core where some of them deposit. After being irradiated by neutrons for a time, the mixture of radioactive and inactive corrosion products is released to the coolant and carried outside the core, where they deposit again. The radioactive corrosion products, such as Co-60, Co-58 and Fe-59, decay to more stable nuclides but emit gamma rays during decay. Consequently, gamma radiation fields gradually grow around the working areas of the primary heat transport circuit. These fields can make maintenance and operation expensive.

Accumulation of radioactive corrosion products on steel surfaces of VVER-type nuclear reactors. II. 60Co

In support of global efforts to strengthen the nuclear non-proliferation regime, the IVG.1M research water-cooled thermal reactor at the National Nuclear Center of the Republic of Kazakhstan was successfully converted to low-enriched uranium (LEU, 19.75% 235U) fuel in 2023. The reactor’s operability with innovative bimetallic, fiber-type, dual-blade LEU fuel rods was experimentally verified during power start-up experiments. The test program included investigations of power distribution in the core, evaluation of temperature, power, and hydrodynamic reactivity effects, and the measurement of fission product release to the coolant. The results were in good agreement with safety calculations, confirming that the enrichment reduction did not degrade reactor performance characteristics. It was shown that the power reactivity effect increased by more than 1.5 times at a power level of 9 MW. The measured temperature reactivity coefficient (≈0.021 βeff/°C) and the level of fission product release remained within acceptable and expected limits.

Transport characteristics of fission product surrogates in water-cooled small modular reactor containments

The irradiation performance of Zircaloy-2 clad thorium-uranium fuel elements under water-cooled power reactor conditions continues to be excellent. Fuel swelling continues to be no more than that required to accommodate the fission product atoms. Measured fuel swelling is less than one percent at 3.1 {times} 10{sup 20} fissions/cc (9,300 MWD/T). The addition of iron and aluminum to metallic uranium fuel elements has been shown to markedly decrease fuel swelling. The improved swelling performance is attributed to a decreased tendency to form mechanically induced porosity and is believed to result from the dispersed second phase particles inhibiting the movement of structural defects formed as a consequence of the anisotropic growth of alpha uranium.

Fission product release modelling for application of fuel-failure monitoring and detection - An overview

The SCDAP/RELAP5 code has been developed for best estimate transient simulation of light water reactor coolant systems during a severe accident. The code models the coupled behavior of the reactor coolant system, the core, fission products released during a severe accident transient as well as large and small break loss of coolant accidents, operational transients such as anticipated transient without SCRAM, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits as much of a particular system to be modeled as necessary. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater conditioning systems. This volume contains detailed descriptions of the severe accident models and correlations. It provides the user with the underlying assumptions and simplifications used to generate and implement the basic equations into the code, so an intelligent assessment of the applicability and accuracy of the resulting calculation can be made.