The fundamental characteristics of groundwater colloids, such as composition, concentration, size, and stability, were analyzed using granitic groundwater samples taken from the KAERI Underground Research Tunnel (KURT) site by such analytical methods as inductively coupled plasma-mass spectrometry, field emission-transmission electron microscopy, a liquid chromatography-organic carbon detector, and dynamic light scattering technique. The results show that the KURT groundwater colloids are mainly composed of clay minerals, calcite, metal (Fe) oxide, and organic matter. The size and concentration of the groundwater colloids were 10–250 nm and 33–64 μg·L−1, respectively. These values are similar to those from other studies performed in granitic groundwater. The groundwater colloids were found to be moderately stable under the groundwater conditions of the KURT site. Consequently, the groundwater colloids in the fractured granite system of the KURT site can form stable radiocolloids and increase the mobility of radionuclides if they associate with radionuclides released from a radioactive waste repository. The results provide basic data for evaluating the effects of groundwater colloids on radionuclide migration in fractured granite rock, which is necessary for the safety assessment of a high-level radioactive waste repository.

The role of the women divers (Haenyo and their unique meeting place (Buleok)of peace to solve disputes is another feature of Jeju Peace Island that recommends it as a locus for peace action. Sammu speaks of neighbourly-ness, although the concept itself refers to three elements lacking on Jeju: The Thief, the Gate and the Beggar. Going deep into the peasant past, Jeju islanders held virtues ``diligence, thrift and interdependence”. This meant that no one was moved to become a thief, there was no need for a secure Gate and, so, the Beggar was not found either. Jeju Islander society and culture was without avarice and promoted values of independence, self-reliance and honor. Entry to a traditional Jeju compound was barred by a Jeongnang or log that showed that the homeowner was away and, so, no one crossed the threshold out of respect. All of this may sound rather negative and stern were it not for Samryeo, the “Three Treasures” in the developing heritage of Jeju peace culture, taken to be Nature (folklore, native industries), Crops of special use, such as marine products and, today, tourism; finished off with Generosity of the beauty of nature, including controls on the level of industry so as to preserve the natural blessings. Whilst in keeping with the three-part philosophy of ancient Jeju, Samryeo is a development from 1960 and more recent times, as Jeju and South Korea recovered from the Jeju Tragedy. Indeed, it is only in 2003(English edition in 2013) that the final definitive report on these disturbing events has been compiled and published. Like the Mangaian war ax, the Jeju Tragedy is a constant reminder of the need for peace and the avoidance of violence and war for harmony of civilization.”.

When we think about an idea of United Nations Tolerance Program UNTP, it is important for us to share memory with world citizens that we had 2314 veterans from 11 countries in the Busan U.N. Forces Cemetery since January of 1951. Recently, we had 13 veterans, who got buried coming from 4 Americans, 3 Hollanders, 2 Frenchmen, 2 Germans, 2 Englishmen, 1 Canadian and 1 headband after they died in their countries. (Joongang daily newspaper June 19, 2022). "I believe that creating a fifth office of the UNTP in Republic of Korea (Seoul, Busan, Kyunggi or Jeju Island), as a twin organization of the UNEP in Nairobi, Kenya, would provide a geopolitical balance to the international organization and a useful base for peace island studies in the conflict laden region of East Asia. Establishing this office in S. Korea(Seoul, Busan, Kyunggi or Jeju Island), with its unique geographical location as a hub for peace talks and environmental negotiations between Russia, North Korea, Japan, China, and Taiwan, would contribute to peace and development in the region." (Yates. 2022).

The effect of oxidation time on the characteristics and mechanical properties of spent nuclear fuel cladding was investigated using Raman spectroscopy, tube rupture test, and tensile test. As oxidation time increased, the Raman peak associated with the tetragonal zirconium oxide phase diminished and merged with the Raman peak associated with the monoclinic zirconium oxide phase near 333 cm−1. Additionally, the other tetragonal zirconium oxide phase peak at 380 cm−1 decreased after 100 d of oxidation, whereas the zirconium monoclinic oxide peak became the dominant peak. The oxidation time had no effect on the tube rupture pressure of the oxidized zirconium alloy tube. However, the yield and tensile stresses of the oxidized nuclear fuel cladding tube decreased after 100 d of oxidation. The results of the scanning electron microscopy and transmission electron microscopy were represented with the in-situ Raman analysis result for the oxide characteristics generated on the cladding of spent nuclear fuel.

We assume that, even though Jeju Island ‘Peace’ Cherry Trees (1913), happened at different time and places, it connects each other as (1919) and Busan UN Forces Cemetery (1951) as an important connected diplomatic event of Provisional Government and Republic of Korea positively. Especially, it will bring a significant positive impact to Asia Community if we organize 2023 Korean Week Event: The First Korea Congress (1919) and Busan UN Forces Cemetery (1951) and suggest the invitation issues of UN Asia Headquarters to the Republic of Korea (2023) into Busan metropolitan city(Seoul, Kyunggi, Jeju Special Self-Governing Province) of Republic of Korea. We also assume Korean’s tolerance philosophy will Koreans to unite each other together if they will succeed to share spirit of tolerance from Busan UN Forces Cemetery. We have 2314 graves from 11 countries at the graveyard for UN Forces in Busan, which was built as the UN Headquarters Cemetery in January of 1951 and used as memorial space for participation in the Korean War, who had involved the War beyond their sacrifices. Recently we had 13 veterans, who got buried coming from 4 Americans, thre Hollanders, 2 Frenchmen, 2 Germans, 2 Englishmen, 1 Canadian and 1 headband after they died in their countries. (Joongang daily newspaper June 19, 2022). In the end they become Asian spirit of Toleran명 at the Busan UN Forces Cemetery beyond borders.

Graphite is widely used as a reflector or moderator in nuclear reactors, and since it is exposed to high flux neutron irradiation, long-lived radioactive isotopes such as C-14 are formed. Therefore, quantitatively prediction of production amount is a very essential task for reliable radioactive waste management. In this study, considering nuclear reactions such as (􀝊, 􀟛) reaction by thermal neutrons and elastic scattering, various characteristics such as the rate of formation of C-14 and energy distribution of thermal neutrons according to the penetration depth from the graphite surface were numerically analyzed. The evaluation was carried out in consideration of the average energy of neutrons and reaction/collision cross-sectional area at given energy, and a comparative study was also performed when the thermal neutrons were in thermal equilibrium and when they were not. The numerically evaluated results were compared qualitatively with the experimental study, and methods to further increase the accuracy were also discussed.

Complexation of actinides and lanthanides with carboxylic organic ligands is known to facilitate migration of radionuclides from deep geological disposal systems of spent nuclear fuel. In order to examine the ligand-dependent structures of trivalent actinides and lanthanides, a series of Eu(III)-aliphatic dicarboxylate compounds, Eu2(oxalate)3(H2O)6, Eu2(malonate)3(H2O)6, and Eu2(succinate)3(H2O)2, were synthesized and characterized by using X-ray crystallography and time-resolved laser fluorescence spectroscopy. Powder X-ray diffraction results captured the transition of the coordination modes of aliphatic dicarboxylate ligands from side-on to end-on binding as the carbon chain length increases. This transition is illustrated in malonate bindings involving a combination of side-on and end-on modes. Strongly enhanced luminescence of these solid compounds, especially on the hypersensitive peak, indicates a low site symmetry of these solid compounds. Luminescence lifetimes of the compounds were measured to be increased, which is ascribed to the displacement of water molecules in the innersphere of Eu center upon bindings of the organic ligands. The numbers of remaining bound water molecules estimated from the increased luminescence lifetimes were in good agreement with crystal structures. The excitation-emission matrix spectra of these crystalline polymers suggest that oxalate ligands promote the sensitized luminescence of Eu(III), especially in the UV region. In the case of malonate and succinate ligands, charge transfer occurs in the opposite direction from Eu(III) to the ligands under UV excitation, resulting in weaker luminescence.

There are three questions arise in radioactivity measurements: (1) Dose the measured value originate from the radioactivity being present in the sample? (2) Is the measurement procedure suitable for the intended measurement purpose with respect to the requirements? (3) What is a range of values fairly sure our true value lies in with a specified probability? These three questions are answered by determining characteristic limits (decision threshold, detection limit and limits of the coverage interval), which are widely used as part of quality assurance in radioactivity measurements. In the past, numerous papers have focused on the questions in different ways, and have drawn a variety of conclusions about the meaning of the different characteristic limits using various terms and symbols. In recent years, substantial efforts were made in order to obtain a systematic and unified way to calculate and express these limits. As a result, the ISO 11929 Series which specify a procedure for calculation of the characteristic limits have been developed. This paper is focused on the calculations of characteristic limits for noble gas monitor (NGM204 monitor) that offers the continuous measurements of radioactive noble gases discharged from the stacks of the HANARO reactor facility. The calculations are based on the standard ISO 11929 as well as the traditional formulas provided by NUREG 1576, ANSI N42-18, ANSI N42-17 and DIN 25482. A comparison is made among the results obtained from the formulas given in each literature.

Safety evaluation of high-level radioactive waste disposal facilities including spent nuclear fuel is a very urgent and critical issue, and in order to do so, it is very important to develop a safety case that includes Feature, Event, Process (FEP) analysis, scenario development, and scenario uncertainty evaluation. In the case of Korea, the disposal of spent nuclear fuel is recognized as an unavoidable option, and in the end, Korea’s specific FEP (SFEP) development and safety evaluation according to the scenario should be conducted. Because each country’s situation and environment are different, it is necessary to develop an SFEP based on a generic FEP (International FEP). To this end, an understanding of IFEP is essential. In this study, about 1,000 major terms appearing in the OECD/NEA IFEP are classified to where each of them belongs among F, E, and P, and which FEP each word belongs to, and the correlation between the frequency of occurrence and each term is analyzed. This result will serve as a reference for the results of SFEP analysis such as POSIVA and SKB, which our research team will analyze later. In addition, each term belongs to which academic field, and the most appropriate translation for translating each term into Korean is also described.

FTIR (Fourier Transform Infrared) and Raman spectra of KJ-II bentonite provided by Clariant Korea were compared with those of MX-80 bentonite. The FTIR spectra were obtained using a Nicolet 5 FTIR spectrometer (Fisher Scientific) equipped with a diamond ATR (Attenuated Total Reflection) module. The spectra were collected for 64 scans with a resolution of 4 cm−1. Raman spectra were obtained using an optical microscope (Olympus, BX43) and a spectrometer (Andor, SR- 500). The laser beam was focused using an objective lens with a magnifying power of 50. The backscattered light from the sample was collected into an optical fiber with a core diameter of 0.4 mm. The Raman signals were recorded with CCDs (Andor, DV401A-BV for 532 nm laser wavelength and DV420A-OE for 638 and 785 nm laser wavelengths). Each pixel of CCD received the signal for 1 s and its 1000 times accumulated data were collected. The FTIR spectra of the two bentonite samples are very similar. The FTIR spectra of both bentonites showed absorption bands at 3623, 3399, 3231 cm−1 in the higher wavenumber region and at 1637, 1443, 1117, 997, 914, 887, 847, 797, 611, 515, 414 cm−1 in the lower wavenumber region. A sharp band at 3623 cm−1 and the strong band at 997 cm−1 correspond to the OH stretching of structural hydroxyl groups and the Si-O stretching vibration, respectively. In addition to these clear bands, several absorption bands observed in this experiment are well matched with the results reported in various literatures. Unlike the FTIR spectrum, it is not easy to observe the Raman bands of bentonite. The reason is that strong fluorescence interferes with the Raman spectrum. The two bentonite samples showed different fluorescence intensities. In the case of MX-80 bentonite, no clear Raman band was observed due to the influence of very strong fluorescence. KJ-II bentonite showed a relatively weak fluorescence intensity and Raman bands were partially visible at around 147, 260, 397, 709, and 1279 cm−1. In particular, the band at 1279 cm−1 is wide and sturdy. It was observed that the non-powder samples showed a better quality spectra. The Raman characteristics of KJ-II bentonite, which depend on the incident laser wavelength and the sample pretreatment, are discussed in detail.

A deep geological disposal system, which consists of the engineered and natural barrier components, is the most proven and widely adopted concept for a permanent disposal of the high level radioactive waste (HLW) thus far. The clay-based engineered barrier is designed to not only absorb mechanical stress caused by the geological activities, but also prevent inflow of groundwater to canister and outflow of radionuclides by providing abundant sorption sites. The principal mineralogical constituent of the clay material is montmorillonite, which is a 2:1 phyllosilicate having two tetrahedral sheets of SiO2 sandwiching an octahedral sheet of Al2O3. The stacking of SiO2 and Al2O3 sheets form the layered structures, and ion-exchange and water uptake reactions occur in the interlayer space. In order to reliably assess the radionuclide retention capacity of engineered barrier under wide geochemical conditions relevant to the geological disposal environments, sorption mechanisms between montmorillonite and radionuclides should be explicitly investigated in advance. Thus far, sorption behavior of mineral adsorbents with radionuclides has been quantified by the sorption-desorption distribution coefficient (Kd), which is simply defined as the ratio of radionuclide concentration in the solid phase to that in the equilibrium solution; the Kd value is conditional, and there have been scientific efforts to develop geochemically robust bases for parameterizing the sorption phenomena more reliably. In this framework, application of thermodynamic sorption model (TSM), which is theoretically based on the concept of widely accepted equilibrium models for aquatic chemistry, offers the potential to improve confidence in demonstration of radionuclide sorption reactions on the mineral adsorbents. Specifically, it is generally regarded in the TSM that coordination of radionuclides on montmorillonite takes place at the surficial aluminol and silanol groups while their ion-exchange reactions occur in the interlayer space also. The effects of electrical charge on the surface reactions are additionally corrected in accordance with the numerous theories of electrochemical interface. The present work provides an overview of the current status of application of TSM for quantifying sorption behaviors of radionuclides on montmorillonite and experimental results for physical separation and characterization of Ca-montmorillonite from the newly adopted reference bentonite (Bentonil- WRK) by means of XRD, BET, FTIR, CEC measurement, and acid-base titration. The determined mineralogical and chemical properties of the montmorillonite obtained will be used as input parameters for further sorption studies of radionuclides with the Bentonil-WRK montmorillonite.

Molten Salt Reactor (MSR) is one of the generation-IV advanced nuclear reactors in which hightemperature molten salt mixture is used as the primary coolant, or even the fuel itself unlike most nuclear reactors that adopt solid fuels. The MSR has received a great attention because of its excellent thermal efficiency, high power density, and structural simplicity. In particular, since the MSR uses molten salts with boiling points higher than the exit temperature of the reactor core, there is no severe accident such as a core melt-down which leads to a hydrogen explosion. In addition, it is possible to remove the residual heat through a completely passive way and when the fuel salt leaks to the outside, it solidifies at room-temperature without releasing radioactive fission products such as cesium, which make the MSR inherently safe. Both fluoride and chloride mixtures can be used as liquid fuel salts by adding actinide halides for MSRs. However, the MSRs using chloride-based salt fuels can be operated for a long time without adding nuclear fuel or online reprocessing because the actinide solubility in chloride salts is about six times higher than that in fluoride salts. Therefore, the chloride-based MSRs are more effective for the transmutation of long-lived radionuclides such as transuranic elements than the fluoride-based MSRs, which is beneficial to resolve the high radioactive spent nuclear fuel generated from light water reactors (LWRs). This paper examines liquid fuel fabrication using an improved U chlorination process for the chloride-based MSRs and presents the strategy for the management of gaseous fission products generated during the operation of MSR.

Nuclear operators must sort out their digital assets as Critical Digital Asset (CDA) and manage their vulnerabilities. Since vulnerabilities are continually found and can be abused anytime, and the number of digital assets in nuclear facilities is increasing, it is important to collect publically-known vulnerabilities in automated mechanism to reevaluate their risks. KINAC is now in progress of establishing an automated mechanism of collecting publically-known vulnerabilities for nuclear facilities. This paper will discuss about criteria of selecting database when establishing an automated mechanism of collecting publically-known vulnerabilities for nuclear facilities. When selecting sets of vulnerability database, the characteristic of sets of digital assets need to be managed, importance of each digital asset, and where and who will use the set of digital assets should be mainly considered. Most of safety-related CDAs are made and used in the United States, and safety-related CDAs are similar to Information and Communication Technology (ICT) facilities. Therefore, the main vulnerability database used in the United States should be included when collecting the database of vulnerabilities. Especially, US government actively provides vulnerabilities of digital assets, enacting vulnerability disclosure policy to make each organization report their own potential vulnerabilities. The main vulnerability database of the US is National Vulnerability Database (NVD) of NIST. It contains over 150,000 vulnerabilities on ICT and Industrial Control System (ICS). Nuclear Energy Institute (NEI) published “Cyber Security Vulnerability and Risk Management”, Addendum 5 to NEI 08-09, and informed that US-CERT, ICS-CERT, and NVD can be used as publically-known vulnerability database, and US National Regulatory Commission (NRC) endorsed the publication. In South Korea, KrCERT and National Cyber Threat Intelligence (NCTI) share publically-known vulnerabilities, however, the number of vulnerabilities are less than those of NVD, and most of the data are duplication of those of NVD. Moreover, certain portion of information are only opened to authorized organizations, so it is unable to access those databases. Therefore, considering the fact that most information of vulnerabilities of CDAs are included in NVD and ICS-CERT, vulnerability database should also contain information from NVD and ICS-CERT. Otherwise, the database must contain equivalent information compared to NVD and ICS-CERT. Furthermore, the methodology for collecting vulnerabilities of digital assets from other countries is also required to be studied in the future research.

Molten salt reactor (MSR) has a unique characteristic using liquid fuel and/or coolant salt among six type of GEN IV reactors. Liquid fuels and on-site processing are fundamentally different from a solid fuel reactor where separate facilities produce the fresh solid fuel and process the Spent Nuclear Fuel. Because the choice of fuel cycle affects the safeguards and non-proliferation characteristics of the reactor system, different MSR concepts may have different proliferation resistance and physical protection characteristics. For example, MSR design variants that use solid fuel but are cooled with liquid salts such as FHR are very close to the Very High Temperature Reactor design concept. The composition of various fuel salts is a representative factor that makes it difficult to generalize the PRPP evaluation principle of MSR. In addition, the flow of molten salts containing fissile materials is also complex depending on the design of the reactor. The path through which radioactive materials travel not only inside the reactor but also to nuclear fuel cycle facilities can act as a difficult factor in measuring nuclear materials. As a further complication, some of the plants include fuel salt drain tanks intended to provide decay heat removal while others are designed to provide decay heat removal while the salt is maintained within the reactor vessel. Some lessons learned from the prior molten salt breeder reactor program are reflected in all of the new designs. Interior reflectors/shielding are frequently employed to reduce the radiation damage to the reactor vessel, and fuel salt chemistry control is employed to substantially limit oxidizing the container alloy constituents. However, even with the vessel interior shielding, the containment environment around both solid and liquid fueled MSRs during operation is likely to have substantially higher dose rates than at LWRs due to the fission process and fission products in the case of circulating liquid fueled reactors, and the shortlived activation products of fluorine (16N, 20F, and 19O) in the case of FHRs. Consequentially due to insufficient shielding from the coolant and the vessel wall, MSR containments will be remote access only for liquid fueled systems and remote access only during operation for FHRs.

Liquid-fueled Molten Salt Reactors (MSRs) do not contain their fuel in assemblies. It is then not possible to perform traditional item counting and visual accountability of the salt fuel. These facilities are closer to bulk accounting facilities, such as reprocessing plants, and require inventory determinations based on measurements of the actinide content of salts. This can be problematic due to the difficulty of sampling and the destructive analysis of actinide-containing molten salts. Some problems arise from the unique combination of high temperature and high radiation environments present in molten salt fuels. Another challenge is the continuous change in the isotopic concentration of fuel salts due to burn-up, conversion, plating out, and online chemical processing. There is a potential for fuel stocks outside the reactor containment vessel in on-site salt processing. In terms of proliferation resistance of 233U-232Th fuel cycle, the nuclide 232U is an important nuclide in thorium fuel cycle from the standpoint of proliferation resistance, because its daughter Thallium (208Tl) is a strong gamma (2.6 MeV) emitter. The hard gamma ray is not only barrier from to nuclear material theft, but also an effective means of detecting lost fissile material. However, there is a theoretical weakness in obtaining pure 233U at the core of the initial two weeks with a concentration of 232Pu less than 1,000 ppm. Therefore, Pu separation process is one of the most sensitive parts in online reprocessing facility. The decision to use a fertile blanket should also be based on proliferation risk considerations in addition to operational parameters. MSRs can be designed without a separate fertile blanket, which should be considered. In the case of the MSFR, even if fertile blankets are used, the production of 232U is large enough to make difficult the utilization of blankets for proliferation purpose. For the liquid-fueled MSRs without fissile materials separations, many of the observations from the previous section apply, except salt processing is minimized. The reactors will still need some method of estimating total actinide content. These reactor designs reduce proliferation risk for the reactor by not separating any actinides during operation.

Since the commercial operation of Kori unit 1 in 1978, nuclear power has provided cheap, stable and clean electricity in South-Korea. For decades, the discussion about the spent fuel management has been dominated and the government is responsible for on-going research and development (R&D) related to long-term spent nuclear fuel management. The effective management of spent fuel should be applied from the early stage of the R&D process to licensing phases with the step-by-step evaluation system. As part of follow-up efforts after the Fukushima nuclear accident, the Nuclear Promotion Commission and Nuclear Safety Commission were divided in function as an independent agency for enhancing national nuclear safety and security, which aims to protect the public and environment from undue radiological hazard. The national spent fuel project must have a vibrant program for spent fuel management. Due to the nature of these projects, the establishment of a ‘conformity assessment’ system that collects the opinions from the licensing organisations on the results of research projects from the initial R&D stage should be applied advertently in order to efficiently conduct research projects and enhance public confidence. For the government-led project for spent nuclear fuel management, the adequacy and applicability of its technology R&D as well as its sustainability that includes financial, social and environmental performance measures should be evaluated in each stage. The institutionalisation arrangement, so called ‘conformity assessment system’ for the development of a national spent nuclear fuel management plan and related technology should be developed. This study aims to propose the basic principles for the introduction of the conformity assessment system: (1) national management responsibility, (2) spent fuel management project scope, (3) its management main principles, (4) project implementation system, (5) final management project scope and securing financial resources.

The nuclear facilities sites handling radioactive materials are contaminated due to accidents or its activities. Since the experimental reactor was operated in Idaho, USA in 1951, nuclear power and its related activities were carried out in many countries. Thus, the legacy sites in which radioactive materials have been deposited are a matter of public concern. As a result of the operation of facilities related to the use of nuclear energy, it is faced with the restoration of the environment containing radioactive wastes. In particular, the activities carried out in the early years caused its contamination of the sites and areas with significant releases of fission products and natural radioactive wastes adventitiously or intentionally due to the lack of its technology and the unclear radiation risks. In addition, nuclear weapons tests conducted in the 1950s and 1960s and the Chernobyl power plant accident in 1986 caused radioactive contamination in the Pacific Ocean and extensive territories in Europe. Accordingly, the IAEA discussed the restoration of the contaminated areas at the conference on the radiation legacy of the 20th Century in terms of environmental restoration (RADLEG-2000) held at Moscow, Russia in October 2000 and its cases from the former Soviet Union and Eastern Europe were reported. Also, the Nuclear Energy Agency (NEA) has discussed the potential issues on the legacy sites occurring radioactive materials and suggested that these sites should be managed in an open, transparent and consistent manner in order to build sustainable solutions with trust. However, there is currently no overall regulatory system for the management of legacy sites in Korea. The TRIGA Mark-II research reactor, which was commissioned in 1962, was decontaminated and dismantled in 1997. As such, it is necessary to consider the regulatory system related to the existing legacy site. Most of all, the legacy site management and regulatory standards of facilities that have operated in the past and are not currently operating or that deal with natural radioactive materials should be established. Also, the dismantling of NK’s nuclear facilities and management of related radioactive waste, particularly NK’s uranium mines and refining facilities particularly at Pyongsan and Bakcheon in NK are one of the key issue on the North Korea’s denuclearization. In this study, the international standards on the reuse of the legacy site after dismantling are analyzed, and its regulatory considerations used for domestic application are suggested.

희생은 아일랜드 민족주의의 중요한 가치이자 독립 투쟁의 수단이기도 했다. 희생은 예이츠 초기희곡 세 작품—『캐슬린백작부인』,『캐슬린 니 훌리한』,『왕의 문지방』—의 주제이다. 이 논문은 이 세 작품을 중심으로 각기 다른 형태의 희생 을 연구한다.『캐슬린백작부인􋺸에서 캐슬린은 가난한 농부들을 구하기 위해 영혼까지 팔고, 『캐슬린 니 훌리한』에서 젊은 농부인 마이클은 아일랜드의 상징인 노파의 부름 에 응해 신부와 가족을 뒤에 두고 노파를 위해 싸우러 떠난다. 『왕의 문지방』에서 국 민시인인 샤나한은 국가평의회 위원이라는 시인의 오래된 권리를 수호하기 위해 목숨 을 희생한다. 이 논문의 목적은 주인공들의 희생의 의미와 그 희생에 대한 예이츠의 태도를 분석하는 것이다.