Time-temperature history in an underground cavity following rupture of a pressurized water nuclear reactor coolant system

Pressurizer is one of the most important components in reactor coolant system of a nuclear power plant, which operates normally at pressure of 15.4 MPa and temperature of 345°C[1]. The main function of pressurizer is to regulate the pressure in the reactor coolant system by either cooling the steam or heating the saturated water in its upper zone. When the pressure in the reactor coolant system increases, it will distribute cold water to decrease its temperature and pressure through atomizing the reactor coolant with swirl spray nozzle in pressurizer. Swirl nozzle is the key part of pressurizer with swirl structure of full cone spray pattern, and the atomization performance include drop size, spray angle and distribution, also it is characterized by huge flow rate and low pressure drop, and its atomization performance decides the quality of pressure control of the reactor coolant system. To enhance the independent design level of both pressurizer and cooling system, it’s necessary to study the atomization performance of swirl nozzle for nuclear reactor pressurizer. Aimed at improving atomization performance of swirl spray nozzle, the structure design methodology of nuclear reactor pressurizer was studied systematically in three aspects including theory design, numerical simulation and test confirm in this thesis. Through designing the swirl nozzle structure according to similar design formula of spray nozzle in theory, especially studying the influence of different structures that mainly include internal swirl structure on internal flow field of swirl nozzles, the primary structure parameters of swirl nozzle were confirmed. Then, through numerical simulation of the internal flow field, flow rate and pressure drop, and swirl core structure of the swirl nozzle (by building physical model and mathematic model according to the spray nozzle structure), the atomization performance of the nozzle was analyzed. On this basis, the typical swirl nozzle was designed and tested, which included spray angle, flow rate as well as pressure drop tests, and spray drop tests, and the applicability of the computational fluid dynamics (CFD) method was verified when it was applied in swirl nozzle design. Finally, the design method of swirl nozzle with deep groove of swirl core for pressurizer was put forward. Through this studying of theoretical calculation, numerical simulating and test, the correlation between the structural parameters of swirl nozzle and atomization performance was achieved, meanwhile design, analysis and test methods of spray nozzle with low pressure drop and huge flow rate were established. It is helpful to realize the independent design of pressurizer’s swirl nozzle and even to put forward the design methodology of pressurizer’s swirl nozzle with our own characteristic.

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An improved core thermal-hydraulic model for coastdown transient in pressurized water reactor

Proceedings of the Institution of Mechanical Engineers Part A Journal of Power and Energy ◽

10.1177/0957650917734745 ◽

2017 ◽

Vol 232 (4) ◽

pp. 416-424

Author(s):

Idrees Ahmad ◽

Saad Arshad ◽

Sajjad Tahir ◽

Qaisar Nadeem ◽

Abdus Samee

Keyword(s):

Time Constant ◽

Thermal Model ◽

Nuclear Reactor ◽

Transient Behavior ◽

Nucleate Boiling ◽

Hydraulic Model ◽

Half Time ◽

Accurate Analysis ◽

Pressurized Water ◽

Reactor Coolant

The safety of the nuclear reactor revolves around the accurate analysis of rapid flow transient for the design and manufacturing of reactor coolant pumps. In this article, the coastdown transient initiated by the loss of offsite power is simulated. In this case, the pumps are operated by the inertia of the flywheel, therefore, the reliable operation of reactor coolant pumps is the key to the safety of the nuclear reactor. A new hydraulic, as well as the thermal model, is developed for simulating various core parameters during the coastdown period. The present hydraulic model accounts for both the pump half-time and the loop half-time, which is used to increase the accuracy of predicted results over a longer period of time. The results predicted by the hydraulic model are incorporated into a thermal model, which also includes the decay heat following the reactor shutdown. This new model depends upon the core time constant, loop time constant, pump half-time, and hydraulic constant coefficient. The predicted results of flow rate, pressure, temperature, and departure from nucleate boiling ratio are compared with the experimental data and have found good agreement between the two cases. Finally, the departure from nucleate boiling ratio shows that the transient behavior of the reactor is moving toward safety.

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Deposition of hematite particles on alumina seal faceplates of nuclear reactor coolant pumps: Laboratory experiments and industrial feedback

Hemijska industrija ◽

10.2298/hemind110922101l ◽

2012 ◽

Vol 66 (3) ◽

pp. 291-299 ◽

Cited By ~ 1

Author(s):

Grégory Lefèvre ◽

Ljiljana Zivkovic ◽

Anne Jaubertie

Keyword(s):

Nuclear Reactor ◽

Primary Circuit ◽

Particle Removal ◽

Mechanical Seals ◽

Solution Ph ◽

Pressurized Water ◽

Reactor Coolant ◽

Key Factor ◽

Water Reactors ◽

Chemical Conditions

In the primary circuit of pressurized water reactors (PWR), the dynamic sealing system in reactor coolant pumps is ensured by mechanical seals whose ceramic parts are in contact with the cooling solution. During the stretch-out phase in reactor operation, characterized by low boric acid concentration, the leak-off flow has been observed to abnormally evolve in industrial plants. The deposition of hematite particles, originating from corrosion, on alumina seals of coolant pumps is suspected to be the cause. As better understanding of the adhesion mechanism is the key factor in the prevention of fouling and particle removal, an experimental study was carried out using a laboratory set-up. With model materials, hematite and sintered alumina, the adhesion rate and surface potentials of the interacting solids were measured under different chemical conditions (solution pH and composition) in analogy with the PWR ones. The obtained results were in good agreement with the DLVO (Derjaguin-Landau-Verwey- Overbeek) theory and used as such to interpret this industrial phenomenon.

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Thermal Fatigue of Pressurized Water Reactor Coolant System Loop Drain Line Down Horizontal Elbow

Volume 2: Nuclear Policy; Nuclear Safety, Security, and Cyber Security; Operating Plant Experience; Probabilistic Risk Assessments; SMR and Advanced Reactors ◽

10.1115/icone2020-16948 ◽

2020 ◽

Author(s):

Yue Zou ◽

Brian Derreberry

Keyword(s):

Thermal Fatigue ◽

Operating Experience ◽

Fatigue Cracking ◽

Screening Methods ◽

Pressurized Water Reactor ◽

Pressurized Water ◽

Water Reactor ◽

Thermal Transients ◽

Reactor Coolant ◽

Coolant System

Abstract Thermal cycling induced fatigue is widely recognized as one of the major contributors to the damage of nuclear plant piping systems, especially at locations where turbulent mixing of flows with different temperature occurs. Thermal fatigue caused by swirl penetration interaction with normally stagnant water layers has been identified as a mechanism that can lead to cracking in dead-ended branch lines attached to pressurized water reactor (PWR) primary coolant system. EPRI has developed screening methods, derived from extensive testing and analysis, to determine which lines are potentially affected as well as evaluation methods to perform evaluations of this thermal fatigue mechanism for the U.S. PWR plants. However, recent industry operating experience (OE) indicate that the model used to predict thermal fatigue due to swirl penetration is not fully understood. There are limitations with the EPRI generic evaluation. In addition, cumulative effects from various thermal transients such as the reactor coolant system (RCS) sampling and excess letdown may also contribute to the failure of RCS branch lines. In this paper, we report direct OE from one of our PWR units where thermal fatigue cracking is observed at the RCS loop drain line close to the welded region of the elbow. A conservative analytical approach that takes into account the influence of thermal stratification, in accordance with ASME Section III Class 1 piping stress formula, is also proposed to evaluate the severity of fatigue damage to the RCS drain line, as a result of various transients. Finally, recommendations are made for future operation and inspection based on results of the evaluation.

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Aging Management of Reactor Coolant System Mechanical Components in Pressurized Water Reactors for License Renewal

Selected Topics on Aging Management, Reliability, Safety and License Renewal ◽

10.1115/pvp2002-1372 ◽

2002 ◽

Author(s):

M. Subudhi ◽

R. Morante ◽

A. D. Lee

Keyword(s):

Reactor Vessel ◽

Pressurized Water Reactors ◽

Pressurized Water ◽

Reactor Coolant ◽

Management Review ◽

Mechanical Components ◽

Water Reactors ◽

Effects Of Aging ◽

License Renewal ◽

Coolant System

The reactor coolant system (RCS) mechanical components in pressurized water reactors (PWRs) that require an aging management review for license renewal include the primary loop piping and associated connections to other support systems, reactor vessel, reactor vessel internals, pressurizer, steam generators, reactor coolant pumps, and all other inter-connected piping, pipe fittings, valves, and bolting. All major RCS components are located inside the reactor building. Based on the evaluation findings of recently submitted license renewal applications for pressurized water reactors, this paper presents the plant programs and/or activities proposed by the applicants to manage the effects of aging. These programs and/or activities provide reasonable assurance that the intended function(s) of these mechanical components will be maintained for the period of extended operation. The license renewal application includes identification of RCS subcomponents that are within the scope of license renewal and are vulnerable to age-related degradation when exposed to environmental and operational conditions, determination of the effects of aging on their intended safety functions, and implementation of the aging management programs and/or activities including both current and new programs. Industry-wide operating experience, including generic communication by the NRC, is part of the aging management review for the RCS components. This paper presents a number of generic issues, including the time-limited aging analyses, associated with RCS components that require further review by the staff.

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