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JAEA Reports

None

; ; ; Iguchi, Yukihiro; ;

JNC-TN3410 2000-014, 43 Pages, 2000/09

JNC-TN3410-2000-014.pdf:2.37MB

None

JAEA Reports

ComparaUve analyses on nuclear charaderistics of water-cooled breeder cores

; Sato, Wakaei*;

JNC-TN9400 2000-037, 87 Pages, 2000/03

JNC-TN9400-2000-037.pdf:3.48MB

ln order to compare the nuclear characteristics of water-cooled bleeder cores with that of LMFBR, MOX fuel cell models are established for boiling and non-boiling LWR, non-boiling HWR and sodium-cooled reactor. Frst, the comarison is made between the heterogeneous cell calculation results by SRAC and those by SLAROM. The results show some differences as for neutron energy spectrum, one-grouped cross section and conversion ratio due to the different grouped cross section library (both are based on JENDL-3.2, though) used for each code, however, the difference is acceptably small for grasping the basic characteristics of the above-mentioned cores. Second, using the SLAROM code, main core parameters such as mean neutron energy, ratio of fast neutron and $$eta$$-value, are analyzed. The comparison between the cores show that softened neutron spectrum by the scattering effect of hydrogen or heavy hydrogen increase the contribution of nuclear reaction (especially for neutron capture reaction rather than fission reaction) in lower energy region comparing with LMFBR. ln order to overcome the effect, tighter lattice than LMFBR is necessary for water-cooled cores to realize the breeding of fissile nuclides. Third, effects of Pu isotopic composition on the breeding ratio are evaluated using SRAC burnup calculation. From the results, it is confirmed that degraded Pu (larger ratio of Pu-240) show the larger breeding ratio. At last, sensitivity analyses are made for k-effective and main reaction ratios. As for k-effective, using a temporary covariance data of JENDL-3.2, uncertainty resulting from the cross sections' error is analyzed for a boiling LWR and a sodium-cooled reactor. The boiling LWR core shows larger sensitivity in lower energy region than the sodium-cooled reactor (especially for the energy region lower than 1kev), And, 18-group analysis that is considered acceptably good for LMFBR analysis, should not be enough for accurate sensitivity estimation of ...

JAEA Reports

None

*; *; *; *

JNC-TJ3410 2000-021, 73 Pages, 2000/03

JNC-TJ3410-2000-021.pdf:52.78MB

no abstracts in English

JAEA Reports

None

*; *; *; *

JNC-TJ3410 2000-020, 80 Pages, 2000/03

JNC-TJ3410-2000-020.pdf:41.34MB

no abstracts in English

JAEA Reports

Use results of MOX fuel in ATR Fugen nuclear power station

Ijima, Takashi; ; Matsumoto, Mitsuo; *

JNC-TN3410 2000-002, 93 Pages, 2000/01

JNC-TN3410-2000-002.pdf:2.54MB

Fugen Nuclear Power Station ("Fugen") is a prototype Advanced Thermal Reactor (ATR), it has been demonstrated the plutonium utilization by loading many Mixed Oxide Fuels (MOX) since the reactor start up March 1979, and no fuel defect had been occurred, The MOX fuel assemblies has the high reliability and has been loaded more than 700 fuel assemblies. This is the largest in the world as a thermal neutron reactor. However, "Fugen" is planning to stop its operation in the year 2003, because the role of the Fugen almost finished. Therefore, we are going to summarize the ATR project including the Plutonium utilization experience. This paper is summarized as part of the experience.

JAEA Reports

Effective multiplication factor measurement by Feynman-$alpha method (3)

Mori, Tomoaki;

PNC-TN9410 98-056, 72 Pages, 1998/06

PNC-TN9410-98-056.pdf:1.98MB

The sub-criticality monitoring system has been developed for criticality safety control in nuclear fuel handling plants. In the past experiments performed with the Deuterium Critical Assembly(DCA), it was confirmed that the detection of sub-criticality was possible to k$$_{eff}$$=0.3. To investigate the applicability of the method to more generalized system, experiments were performed in the light-water-moderated system of the modified DCA core. From these experiments, it was confirmed that the prompt decay constant($$alpha$$), which was a index of the sub-criticality, was detected between k$$_{eff}$$=0.623 and k$$_{eff}$$ 0.870 and the difference of 0.05$$sim$$0.1$$Delta$$k could be distinguished. The $$alpha$$ values were numerically calculated with 2D transport code TWODANT and monte carlo code KENO V.a, and the results were compared with the measured values. The differences between calculated and measured values were proved to be less than 13%, which was sufficient accuracy in the sub-criticality monitoring system. It was confirmed that Feynman-$$alpha$$ method was applicable to sub-critical measurement of the light-water-moderated system.

JAEA Reports

None

Matsumoto, Mitsuo; ;

PNC-TN1410 98-005, 96 Pages, 1998/03

PNC-TN1410-98-005.pdf:2.17MB

no abstracts in English

JAEA Reports

None

Kishida, Masako*; *; *

PNC-TJ3678 98-001, 206 Pages, 1998/03

PNC-TJ3678-98-001.pdf:5.53MB

no abstracts in English

JAEA Reports

Advanced graphical user interface to the MAAP/Fugen simulator system

Lund

PNC-TN3410 98-002, 34 Pages, 1998/01

PNC-TN3410-98-002.pdf:6.11MB

A new and improved Graphical User Interface (GUI) to the Modular Accident Analysis Program for FUGEN (MAAP/FUGEN) has been developed and implemented at Fugen. The new user interface is a superset of the existing GUI to MAAP - the MAAP/FUGEN/GRAAPH - in the meaning that it contains all the features of the GRAAPH, but in addition offers a number of new features. The new interface, named MAAP-PICASSO is based on the Picasso-3 technology developed by Institutt for Energiteknikk/OECD Halden Reactor Project. The main difference between the MAAP-PICASSO and MAAP-FUGEN-GRAAPH GUIs is that the MAAP-PICASSO GUI is completely decoupled from the numerical simulator. This gives a far higher flexibility regarding enhancement of the GUI, connection to other, external software and user friendliness. It also includes techniques for presenting 2 byte character set - i.e. displaying text in Japanese characters. A special software has been developed for automatic extraction and reuse of the graphical plant information provided in MAAP/GRAPH into Picasso language. This software-has been demonstrated not only on the Fugen plant data, but also other Nuclear Power Plant picture definitions provided by Fauske Inc. The new GUI requires a minimal modification of the MAAP code itself However, these modification is only for parameter communication and is not intrusive to the numerical computations of MAAP itself. The GUI has been developed using modular and object-oriented programming techniques, which makes it relatively easy to extend and modify to fulfill present and future requirements from the users at Fugen, and makes it compatible with future versions of the MAAP code. MAAP-PICASSO is developed on and currently running only on HP UNIX workstations. However, a new NT-based version of Picasso-3 will be released from the Halden Project in February 1998. This will further enhance the applicability and usability of the MAAP-PICASSO GUI.

JAEA Reports

Process condition monitoring at FUGEN

Lund

PNC-TN3410 98-001, 14 Pages, 1998/01

PNC-TN3410-98-001.pdf:3.07MB

At the FUGEN Power Station a system for online monitoring of selected process component behavior, CONFU (CONdition monitoring at FUgen) has been implemented. This system is based on MOCOM (Model Based Condition Monitoring System), developed at IFE/OECD Halden Reactor Project. The system is currently monitoring the heat exchangers for the Reactor Auxiliary Cooling Water System. These heat exchangers has shown a slowly degrading performance over time due to fouling, i.e. accumulation of a heat resisting layer of organic material on the sea water side. This slow degradation, which is not detected by the conventional control and alarm systems, is not an operational, but rather a maintenance problem. CONFU is using dynamically updated mathematical models to compute the performance degradation of the heat exchangers, expressed in overall heat transfer, heat transfer coefficients or heat exchanger efficiency. The results of testing CONFU on real plant data identify the expected degradation trends. The data from CONFU can, in addition to give the plant operator a good impression of the component's operational state, be utilized by the maintenance planning personnel for determination of the most optimal maintenance schedule. Furthermore, the process models in CONFU have been used for simulation purposes.

JAEA Reports

None

PNC-TN1440 97-005, 76 Pages, 1997/09

PNC-TN1440-97-005.pdf:3.92MB

no abstracts in English

JAEA Reports

None

PNC-TN1410 97-034, 338 Pages, 1997/09

PNC-TN1410-97-034.pdf:6.65MB

no abstracts in English

JAEA Reports

None

Onuki, Norihiko; ; Shuji, Yoshiyuki; ; ; ;

PNC-TN8410 97-272, 134 Pages, 1997/08

PNC-TN8410-97-272.pdf:5.33MB

None

JAEA Reports

None

Yamaguchi, Takashi

PNC-TN1410 97-030, 107 Pages, 1997/08

PNC-TN1410-97-030.pdf:1.98MB

no abstracts in English

JAEA Reports

None

Yamaguchi, Takashi

PNC-TN1410 97-029, 65 Pages, 1997/08

PNC-TN1410-97-029.pdf:1.26MB

no abstracts in English

JAEA Reports

None

Yamaguchi, Takashi

PNC-TN1410 97-027, 12 Pages, 1997/08

PNC-TN1410-97-027.pdf:0.25MB

no abstracts in English

JAEA Reports

None

Yamaguchi, Takashi

PNC-TN1410 97-028, 14 Pages, 1997/07

PNC-TN1410-97-028.pdf:0.28MB

no abstracts in English

JAEA Reports

None

Yamaguchi, Takashi

PNC-TN1410 97-026, 16 Pages, 1997/07

PNC-TN1410-97-026.pdf:0.29MB

no abstracts in English

JAEA Reports

None

PNC-TN1410 97-014, 87 Pages, 1997/03

PNC-TN1410-97-014.pdf:2.92MB

no abstracts in English

51 (Records 1-20 displayed on this page)