h1

% IMPORTANT: The following is UTF-8 encoded.  This means that in the presence
% of non-ASCII characters, it will not work with BibTeX 0.99 or older.
% Instead, you should use an up-to-date BibTeX implementation like “bibtex8” or
% “biber”.

@PHDTHESIS{Chan:783468,
      author       = {Chan, Hsiao-Wei},
      othercontributors = {Allelein, Hans-Josef and Cheng, Xu and Lauster, Michael},
      title        = {{D}ose rate in a {PWR}-containment in consequence of a core
                      melt accident},
      school       = {Rheinisch-Westfälische Technische Hochschule Aachen},
      type         = {Dissertation},
      address      = {Aachen},
      reportid     = {RWTH-2020-02522},
      pages        = {1 Online-Ressource (XX, 220 Seiten) : Illustrationen,
                      Diagramme},
      year         = {2019},
      note         = {Veröffentlicht auf dem Publikationsserver der RWTH Aachen
                      University; Dissertation, Rheinisch-Westfälische Technische
                      Hochschule Aachen, 2019},
      abstract     = {Nowadays, the available radioactive source term analyses in
                      a nuclear power plant are done by so called lumped parameter
                      codes. These analyses are performed by coupling a
                      thermohydraulic calculation and an aerosol transport and
                      deposition calculation. When applying the result of such an
                      analysis in a code simulating nuclear reaction, it can be
                      used to evaluate the dose rate in the containment of a
                      nuclear power plant. To prove the feasibility and the
                      practicality of the mentioned procedure, an evaluation chain
                      containing two calculation parts has been established. The
                      first part of the evaluation chain is related to the
                      isotopic source term analyses, which utilizes the codes
                      ATHLET-CD and COCOSYS in a coupled version to obtain source
                      term results on isotope basis. An SB LOCA scenario of a 50
                      cm2 leak at the hot-leg of the reactor cooling system surge
                      line loop with limited emergency measurement leading to core
                      degradation without reactor pressure vessel failure is
                      assumed. Thermohydraulic responses and fission product
                      transport and distribution in both RCS and containment are
                      analyzed. The second part of the evaluation chain is
                      performed with Monaco. Monaco is a time-independent,
                      fixed-source, multi-group Monte Carlo particle transport
                      code for shielding application. For the radiological
                      assessment, two source locations are chosen to perform
                      particle transport analyses in three different containment
                      geometry models. The three containment geometry models are
                      built based on one another. The first model represents a
                      strongly simplified containment and reactor building and the
                      last model has the highest geometrical resolution. The
                      quality of the applied evaluation chain is impressive. This
                      is proven by comparing the results of exemplary calculations
                      with the measurements in the containment of the
                      Three-Mile-Island Unit 2. They are in very good agreement.},
      cin          = {419810 / 413110},
      ddc          = {620},
      cid          = {$I:(DE-82)419810_20140620$ / $I:(DE-82)413110_20140620$},
      typ          = {PUB:(DE-HGF)11},
      doi          = {10.18154/RWTH-2020-02522},
      url          = {https://publications.rwth-aachen.de/record/783468},
}