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1

Elektrownia jądrowa im. Alvina W. Vogtle’a – największa w USA

Nowicki Jacek

Postępy Techniki Jądrowej

|

2023

|

z. 4

34--38

PL

W artykule przedstawiono historię rozwoju Elektrowni Jądrowej Vogtle, która wkrótce stanie się największym obiektem energetyki jądrowej w USA. Dwa najnowsze bloki: 3 i 4 zbudowane są na bazie projektu Westinghouse AP1000 – w technologii modułowej z uwzględnieniem zasad bezpieczeństwa pasywnego, charakterystycznego dla generacji III+ jądrowych bloków energetycznych.

EN

The article presents the history of the development of the Vogtle Nuclear Power Plant, which will soon become the largest nuclear energy facility in the USA. The two newest units: 3 and 4 are built on the basis of the Westinghouse AP100 design - in modular technology taking into account the principles of passive safety, characteristic for Generation III+ nuclear power units.

2

Performance comparison of PWR/EPR and steam generator with and without economizer

Laskowski Rafał

Modern Engineering

|

2022

|

2

21--27

EN

The article presents a model of a steam generator based on mass and energy balance equations, heat transfer coefficients, criterion relations and Peclet’s law. The presented model was applied to a steam generator without and with an economizer for a PWR/EPR nuclear power plant. Based on the calculations, the steam pressure at the outlet of the steam generator without the economizer is 74.17 bar, and for the model with the economizer it is 77.2 bar. The pressure difference for these two variants, based on the calculations, was therefore 3 bar. Higher steam pressure at the outlet of the steam generator translates into a greater enthalpy drop in the turbine and greater power generated by the steam turbine and the efficiency of the entire nuclear power plant. Based on the economic calculations carried out for 60 years of operation of a nuclear power plant with four steam generators, the profit from the use of the economizer amounted to about PLN 0.6 billion.

3

Jak zrealizować Program Polskiej Energetyki Jądrowej (PPEJ)

Kulczyński Dariusz Witold

Postępy Techniki Jądrowej

|

2021

|

z. 4

7--14

PL

W artykule porównano wersje 2020 i 2014 Programu Polskiej Energetyki Jądrowej i przedstawiono krótką charakterystykę reaktora AP1000. Podano źródła informacji, które mogą być pomocne przy ocenie generalnego wykonawcy, poddostawców i przy formułowaniu kontraktu. Wyliczono potencjalne zagrożenia w okresie budowy, rozruchu i eksploatacji elektrowni jądrowej ilustrując je przykładami opartymi na doświadczeniach autora. Podkreślono znaczenie przejrzystości harmonogramów (PERT). Omówiono cechy dobrze zorganizowanej dokumentacji i szkolenia oraz przedstawiono przykłady właściwie i niewłaściwie realizowanych przedsięwzięć. Podkreślono znaczenie stabilności struktur organizacyjnych. Zasugerowano zorganizowane przekazywanie wiedzy opartej na doświadczeniu zawodowym przez starsze pokolenia polskich specjalistów.

EN

The article briefly discusses the differences between the Polish Nuclear Power Programme update of 2020 and its 2014 version. A short characteristic of the AP1000 reactor was provided. Sources of information helpful in the assessment of the main contractor and subcontractors were listed. The importance of precision in contract language was stressed. Potential threats to the successful implementation of nuclear power were enumerated and illustrated by the author’s own experience. The importance of easily readable work plans (PERT) was stressed. The author discussed some features of well-organized documentation and training. Examples of well-run and poorly run projects were offered. Warning against unnecessary reorganization of nuclear business was presented. The orderly transfer of knowledge of the Baby Boomer generation of Polish ex-pats employed in the nuclear industry was suggested.

4

Optimization of the loading pattern of the PWR core using genetic algorithms and multi-purpose fitness function

Kubiński Wojciech, Darnowski Piotr, Chęć Kamil

Nukleonika

|

2021

|

Vol. 66, No. 4

147--151

EN

The study demonstrates an application of genetic algorithms (GAs) in the optimization of the first core loading pattern. The Massachusetts Institute of Technology (MIT) BEAVRS pressurized water reactor (PWR) model was applied with PARCS nodal-diffusion core simulator coupled with GA numerical tool to perform pattern selection. In principle, GAs have been successfully used in many nuclear engineering problems such as core geometry optimization and fuel confi guration. In many cases, however, these analyses focused on optimizing only a single parameter, such as the effective neutron multiplication factor (keff), and often limited to the simplified core model. On the contrary, the GAs developed in this work are equipped with multiple-purpose fitness function (FF) and allow the optimization of more than one parameter at the same time, and these were applied to a realistic full-core problem. The main parameters of interest in this study were the total power peaking factor (PPF) and the length of the fuel cycle. The basic purpose of this study was to improve the economics by finding longer fuel cycle with more uniform power/flux distribution. Proper FFs were developed, tested, and implemented and their results were compared with the reference BEAVRS first fuel cycle. In the two analysed test scenarios, it was possible to extend the fi rst fuel cycle while maintaining lower or similar PPF, in comparison with the BEAVRS core, but for the price of increased initial reactivity.

5

Analysis of the BEAVRS PWR benchmark using SCALE and PARCS

Darnowski Piotr, Pawluczyk Michał

Nukleonika

|

2019

|

Vol. 64, No. 3

87--96

EN

This paper presents an analysis of the Benchmark for Evaluation And Validation of Reactor Simulations (BEAVRS) performed using SCALE 6.1.2 and PARCS 3.2 computer codes. The benchmark specifi cation contains a detailed design, operational data and measurements for a real 4-loop Westinghouse pressurized water reactor (PWR). The lattice physics simulations were prepared using TRITON depletion sequence and NEWT neutron transport solver (SCALE package). The 238-neutron group library based on evaluated nuclear data fi le – ENDF/B-VII nuclear data libraries was applied. A set of branch and burnup calculations was prepared, and group constants in the form of PMAXS fi les were generated with GenPMAXS. The full-core models were prepared using the PARCS nodal-diffusion core simulator. The PMAXS libraries were used with PARCS to investigate the core operation. The hot zero power measurement data, including control rod worths and critical boron concentrations, were compared using simulations, and satisfactory results were achieved. The fi rst fuel cycle was simulated, and acceptable agreement with boron letdown curve and measurements were obtained. Finally, conclusions and recommendations for future research were presented.

6

Modelling of the reactor cycle cost for thorium-fuelled PWR and environmental aspects of a nuclear fuel cycle

Oettingen Mikołaj

Geology, Geophysics and Environment

|

2019

|

Vol. 45, no. 3

207--217

EN

The paper presents the methodology applied to the cost modelling of the uranium-thorium nuclear reactor cycle for PWR reactors as well as brief introduction to the environmental impact of the nuclear fuel cycle. The reactor core contains seed uranium fuel and blanket thorium fuel. In such a cycle, energy is produced in the fission of 235U included in the fresh fuel and in the fission of 233U bread from the fertile 232Th. A modified methodology developed by the OECD Nuclear Energy Agency was used for the reactor cycle cost modelling. The method is based on the levelized lifetime cost methodology for a reactor cycle, which is directly related to the heavy metal mass balance. Contrary to the case of uranium-fuelled nuclear reactors, the cost modelling includes the additional cash flow for thorium fuel. The abundance of thorium in the Earth’s crust is about 3–5 times larger than that of uranium, which suggests its promising potential as a nuclear fuel. However, this needs to be proved economically.

7

Trzy scenariusze energetyki jądrowej w Polsce

Mikulski A.

Postępy Techniki Jądrowej

|

2018

|

z. 1

26--29

PL

W artykule przedstawiono trzy możliwe scenariusze rozwoju energetyki jądrowej w Polsce oparte o duże bloki energetyczne, reaktory wysokotemperaturowe i zintegrowane bloki wodno-ciśnieniowe. Pierwszy jest kontynuacją kierunku określonego w Programie Polskiej Energetyki Jądrowej w 2009 r. wraz z przedstawieniem, co zostało zrealizowane w tym czasie oraz wymogiem weryfikacji tego kierunku. Drugi stanowi nowy kierunek zastosowania reaktorów jądrowych w celach kogeneracyjnych, czyli równoczesnego wytwarzania ciepła technologicznego i energii elektrycznej w nowych i obiecujących konstrukcjach. Natomiast trzeci kierunek stanowią obiekty o zmniejszonej mocy wodno-ciśnieniowych z pomysłem integracji całego obiegu pierwotnego w jednym zbiorniku. Bloki te ze względu na zwiększone bezpieczeństwo eksploatacyjne nie wymagają rozległej strefy bezpieczeństwa i mogłyby być lokalizowane w istniejących elektrowniach w ramach ich modernizacji. Poza tym będą znacznie tańsze i budowane w znacznie krótszym czasie, a umieszczone po kilka w jednej lokalizacji mogą dostarczać moc porównywalną z dużymi blokami jądrowymi.

EN

The paper presents three potential scenarios for the development of nuclear power in Poland based on: high power PWRs, high temperature reactors and integrated pressurized water cooled reactors. The first scenario is the continuation of the Polish Nuclear Power Program of 2009, supplemented by a short description of what has been achieved already, subject to further confirmation. The second scenario is the new possibility of using nuclear reactors for the cogeneration of technological heat and electricity using new, promising constructions of high temperature gas cooled reactors (HTGR). The third scenario is a construction of lower-power PWRs with an integrated primary loop in the one vessel (iPWR). Due to increased operational safety, such blocks require smaller safety zones and therefore may localized within refurbished old conventional power plants. Furthermore such blocks are cheaper to produce, faster to construct and, with several located in the same place, can deliver a power comparable to that from existing high power nuclear blocks.

8

Simulations of the AP1000-based reactor core with SERPENT computer code

Darnowski P., Ignaczak P., Orębski P., Stępień M., Niewiński G.

Archive of Mechanical Engineering

|

2018

|

Vol. LXV, nr 3

295--325

EN

The paper presents the core design, model development and results of the neutron transport simulations of the large Pressurized Water Reactor based on the AP1000 design.The SERPENT 2.1.29 Monte Carlo reactor physics computer code with ENDF/BVII and JEFF3.1.1 nuclear data libraries was applied. The full-core 3D models were developed according to the available Design Control Documentation and the literature. Criticality simulations were performed for the core at the Beginning of Life state for Cold Shutdown, Hot Zero Power and Full Power conditions. Selected core parameters were investigated and compared with the design data: eﬀective multiplication factors, boron concentrations, control rod worth, reactivity coeﬃcients and radial power distributions. Acceptable agreement between design data and simulations was obtained, conﬁrming the validity of the model and applied methodology.

9

Analysis of possible application of high-temperature nuclear reactors to contemporary large-output steam power plants on ships

Kowalczyk T., Głuch J., Ziółkowski P.

Polish Maritime Research

|

2016

|

nr 2

32--41

EN

This paper is aimed at analysis of possible application of helium to cooling high-temperature nuclear reactor to be used for generating steam in contemporary ship steam-turbine power plants of a large output with taking into account in particular variable operational parameters. In the first part of the paper types of contemporary ship power plants are presented. Features of today applied PWR reactors and proposed HTR reactors are discussed. Next, issues of load variability of the ship nuclear power plants , features of the proposed thermal cycles and results of their thermodynamic calculations in variable operational conditions, are presented.

10

Remonty kapitalne w kanadyjskich elektrowniach jądrowych

Kulczyński D. W.

Postępy Techniki Jądrowej

|

2015

|

z. 4

33--37

PL

Tematem artykułu jest przedłużanie okresu eksploatacji starzejących się elektrowni jądrowych. Omówiono szczegółowo remonty kapitalne w kanadyjskich elektrowniach jądrowych typu CANDU (PHWR). Zasygnalizowano także zagadnienia związane z wydłużaniem okresu użytkowania reaktorów lekkowodnych, które były tematem dyskusji konferencji Nuclear Power Plant Life Management & Extension w Paryżu w 2015 r.

EN

The article tackles various aspects of life extension of aging nuclear power plants. It describes in detail mid-life refurbishment of Canadian nuclear power plants (CANDU - PHWR). Some aspects of light water reactors life extension discussed at the Nuclear Power Plant Life Management & Extension 2015 in Paris were also mentioned.

11

Eastern-design VVER nuclear reactors in European Union

Oettingen M., Stanisz P., Malicki M.

Logistyka

|

2015

|

nr 4

9537--9551, CD3

EN

In total, 131 nuclear reactors of various types operates in fourteen European Union member states providing about 830 TWh electricity. The main technology used in European Union is the Western-type Pressurized Water Reactor. However, the second most popular technology is the Eastern-type Vodo-Vodyanoi Energetichesky Reaktor-VVER. The five out of fourteen nuclear countries (Bulgaria, Czech Republic, Finland, Hungary and Slovakia) operate 18 VVER reactors with total electricity output of 80 TWh, which corresponds to about 10% share in nuclear electricity generation and to 2.5% share in total electricity generation in European Union. The paper presents the background of the VVER nuclear reactor technology and analysis of its operation history in European Union member states. The analysis was performed using advanced data banks i.e. The Power Reactor Information System and The Country Nuclear Power Profiles of International Atomic Energy Agency, The Information Library of World Nuclear Association and Eurostat of European Commission. The performed research is devoted to the analysis of the electricity supply and performance indicators of VVER nuclear reactors, which also plays an important role for the logistics and transportation in European Union.

PL

Obecnie w czternastu krajach Uni Europejskie(UE) pracuje 131 reaktorów jądrowych różnego typu zapewniających ok. 830 TWh energii elektrycznej. Głównym typem reaktora jadowego używanego w krajach UE jest reaktor wodny ciśnieniowy Typu Zachodniego. Jednak, drugim najbardziej powszechnym typem reaktora jest reaktor wodny ciśnieniowy Typu Wschodniego - VVER (Vodo-Vodyanoi Energetichesky Reaktor). Pięć z czternastu krajów posiadających energetykę jądrową w UE (Bułgaria, Czechy, Finlandia, Węgry oraz Słowacja) posiada 18 reaktorów typu VVER produkujących ok. 80 TWh energii elektrycznej, co odpowiada ok. 10% produkcji energii elektrycznej z procesów jądrowych, jak również 2.5% całkowitej produkcji energii elektrycznej w Uni Europejskiej. Artykuł przedstawia ogólną charakterystykę reaktorów typu VVER jak i historię ich implementacji w krajach Uni Europejskiej. Analiza została przeprowadzona przy użyciu zasobów dostępnych w międzynarodowych bazach danych m.in.: „The Power Reactor Information System” oraz „The Country Nuclear Power Profiles” Międzynarodowej Agencji Energii Atomowej, „The Information Library” Międzynarodowego Stowarzyszenia Jądrowego oraz „Eurostat” Komisji Europejskiej. Przedstawiona analiza skupia się na charakterystyce produkcji energii elektrycznej oraz współczynników wydajności reaktorów VVER, co ma również istotne znaczenie dla logistyki i transportu w Uni Europejskiej.

12

Thermal-hydraulic analysis of single and multiple steam generator tube ruptures in a typical 3-loop PWR

Koszela Z., Sokołowski Ł.

Journal of Power Technologies

|

2015

|

Vol. 95, nr 3

175--182

EN

The response of the full-scale three-loop Pressurized Water Reactor (PWR) RELAP5 computational model on Steam Generator Break Rupture (SGBR) was investigated in this paper. This model was analyzed in terms of its applicability and performance regarding the research task conducted by Warsaw University of Technology and the National Center for Research and Development inWarsaw, Poland. In the paper break sizes corresponding to one, three and six ruptured tubes (which conform to a Loss-of-Coolant event break size area of 0.02%, 0.054 and 0.11%) were studied at three different locations (at the top of the hot-leg side tubesheet, U-bend and at the top of the cold-leg side tubesheet). The reactor at issue was a three-loop PWR of Westinghouse design with thermal output of 2775 MWt.

13

Symulacje Monte Carlo fizyki rdzenia reaktora jądrowego typu PWR

Oettingen M.

Logistyka

|

2013

|

nr 4

PL

Tematem niniejszej pracy jest modelowanie Monte Carlo fizyki rdzenia rektora jądrowego na poziomie kasety paliwowej. Symulacje numeryczne transportu neutronów oraz zmiany składu paliwa na skutek transmutacji i rozpadów promieniotwórczych zostały przeprowadzone za pomocą kodu MCB (The Monte Carlo Continuous Energy Burnup Code). Model numeryczny opracowany w celu przeprowadzenia symulacji Monte Carlo został zbudowany na podstawie geometrii oraz składu materiałowego kasety paliwowej typu 17x17 używanej w rektorach wodnych ciśnieniowych PWR (Pressurized Water Reactor). Kaseta zawiera czyste paliwo uranowe jak i paliwo z dodatkiem wypalającej się trucizny – Gd2O3. Obecność wypalającej się trucizny istotnie wpływa na charakterystyki kasety paliwowej w polu neutronów, co zostało poddane analizie w zaprezentowanym artykule. Głównymi parametrami otrzymanymi w symulacji numerycznej są: reaktywność układu, wypalenie paliwa oraz ewolucje wybranych nuklidów, takich jak U235, Pu239, Pu241, Gd155 oraz Gd157. Wyniki symulacji numerycznej przeprowadzonej przy pomoc kodu MCB są zgodne z prawidłowościami fizyki rdzenia rektorów jądrowych typu PWR. Opracowana metodologia symulacji rdzenia reaktora jądrowego PWR na poziomie kasety paliwowej z dużą wiarygodnością odzwierciedla rzeczywiste zachowanie systemu.

EN

The Study focuses on the Monte Carlo modelling of the nuclear reactor core at the level of the fuel subassembly. The simulations of neutron transport and fuel depletion due to the nuclear transmutations and decays were performed using The Monte Carlo Continuous Energy Burnup Code – MCB. The numerical model developed for the Monte Carlo simulation was built using the engineering geometry and material composition of the 17x17 fuel subassembly for Pressurized Water Reactor (PWR). The 17x17 fuel subassembly contains pure uranium fuel as well as Gd2O3 bearing fuel. The Gd2O3 burnable poison significantly influences the characteristics of the fuel subassembly in the neutron field, which was shown in the paper. The main parameters obtained in the simulation are: system reactivity, fuel burnup and evolutions of the U235, Pu239, Pu241, Gd155 and Gd157. The results of the numerical simulation performed with the MCB code show good agreement with the theoretical predictions of the nuclear reactor physics. The developed methodology for the numerical simulation of the PWR core at the level of the fuel subassembly with high accuracy reflects the reality.

14

Validation of the heat and mass transfer models within a pressurized water reactor containment using the International Standard Problem No. 47 data

Bury T., Składzień J., Fic A.

Archives of Thermodynamics

|

2012

|

Vol. 33, no. 2

23-46

EN

A lumped parameter type code, called HEPCAL, has been worked out in the Institute of Thermal Technology of the Silesian University of Technology for simulations of a pressurized water reactor containment transient response to a loss-of-coolant accident. The HEPCAL code has been already verified and validated against available experimental data, which in fact have been taken from separate effect tests mainly. This work is devoted to validation of the latest version of the HEPCAL code against experimental data from more complex tests. These experiments have been performed on three different test rigs (called TOSQAN,MISTRA and ThAI) and a part of them became the basis of the International Standard Problem No. 47 (ISP-47) dedicated to containment thermal-hydraulics. Selected experiments realized within the framework of the ISP-47 project have been simulated using the HEPCAL-AD code. The obtained results allowed for drawing of some important conclusions concerning heat and mass transfer models (especially steam condensation), two-phase flow model and buoyancy effects.

15

Stress corrosion cracking on cold-worked austenic stainless steels in PWR environment

Tribouilloy L., Vaillant F., Olive J.-M., Puiggali M.

Advances in Materials Science

|

2007

|

Vol. 7, nr 1(11)

61-69

EN

Types 304L and 316L austenitic Stainless Steels (SS) are widely used in PWR environment. These past few years, a limited number of cases of intergranular stress corrosion cracking (IGSCC) have been detected in cold worked areas of non sensitized austenitic stainless steel components. A first study has been initiated at EDF to assess the conditions of the cracking. The main results include cold work thresholds of 240 HV 0,1 for initiation cracking, and of 310 HV 0,1 for crack propagation, and propose that a dynamic loading is necessary for SCC. The aim of the present paper is to provide a basis of a crack propaga-tion model by investigating the effcet of loading, material and cold-work. In order to try to approach a static loading, a trape-zoidal cyclic loading is applied on high cold-worked (by rolling or by tensile loading) materials. It is shown that, for the most severe loading, the rolling cold-worked (RCW) materials undergo TGSCC whereas IGSCC is observed after tensile cold-working (TCW). The ratio of loading R bas such a strong impact on the crack growth rate (CGR) that it modifies the mechanism of cracking. Moreover, we notice that CGR increases with the applied K max but this evolution depends on the R value. Therefore, [delta K] is chosen to represent the mechanical loading effects on CGRs. Finally, the CGR after a hold time of 1 hour is quite the same than for 3 hours. Additionally, to address the critical issue of the effecet of the crack tip strain rate on crack growth rate, Slow Strain Rate Tests (SSTR) are carried out on RCW specimens and provide a first relation which is not consistent to a pure anodic process. This study is going on TCW specimens.