0
SPECIAL SECTION PAPERS

Uncertainty Analysis for Source Term Evaluation of High Temperature Gas-Cooled Reactor Under Accident Conditions—Identification of Influencing Factors in Loss-of-Forced Circulation Accidents

[+] Author and Article Information
Yuki Honda

Japan Atomic Energy Agency,
4002, Narita-cho, Oarai-machi,
Higashiibaraki-gun,
Ibaraki-ken 311-1393, Japan
e-mail: honda.yuki@jaea.go.jp

Hiroyuki Sato

Japan Atomic Energy Agency,
4002, Narita-cho, Oarai-machi,
Higashiibaraki-gun,
Ibaraki-ken 311-1393, Japan
e-mail: sato.hiroyuki09@jaea.go.jp

Shigeaki Nakagawa

Japan Atomic Energy Agency,
4002, Narita-cho, Oarai-machi,
Higashiibaraki-gun,
Ibaraki-ken 311-1393, Japan
e-mail: nakagawa.shigeaki@jaea.go.jp

Hirofumi Ohashi

Japan Atomic Energy Agency,
4002, Narita-cho, Oarai-machi,
Higashiibaraki-gun,
Ibaraki-ken 311-1393, Japan
e-mail: ohashi.hirofumi@jaea.go.jp

Manuscript received October 30, 2017; final manuscript received January 10, 2018; published online May 16, 2018. Assoc. Editor: Tomio Okawa.

ASME J of Nuclear Rad Sci 4(3), 031013 (May 16, 2018) (11 pages) Paper No: NERS-17-1253; doi: 10.1115/1.4039066 History: Received October 30, 2017; Revised January 10, 2018

One of the key elements in probabilistic risk assessment is the identification and characterization of uncertainties. This paper suggests a procedure to identify influencing factors for uncertainty in source term evaluation, which are important to risk of public dose. We propose the following six steps for the identification in a systematic manner in terms of completeness and transparency of the results using both a logic diagram based on basic equations and expert opinions: (1) identification of uncertainty factors based on engineering knowledge of accident scenario analysis; (2) derivation of factors at the level of physical phenomena and variable parameters by expansion of dynamic equation for the system and scenario to be investigated, (3) extraction of uncertainties in variable parameters; (4) selection of important factors based on sensitivity study results and engineering knowledge; (5) identification of important factors for uncertainty analysis using expert opinions; and (6) integration of selected factors in the aforementioned steps. The proposed approach is tested with a case study for a risk-dominant accident scenario in direct cycle high-temperature gas-cooled reactor (HTGR) plant. We use this approach for evaluating the fuel temperature in terms of reactor dynamics and thermal hydraulic characteristics during a depressurized loss-of-forced circulation (DLOFC) accident with the failure of mitigation systems such as control rod systems (CRS) in a representative HTGR plan. In total, six important factors and 16 influencing factors were successfully identified by the proposed method in the case study. The selected influencing factors can be used as input parameters in uncertainty propagation analysis.

FIGURES IN THIS ARTICLE
<>
Copyright © 2018 by ASME
Your Session has timed out. Please sign back in to continue.

References

Strydom, G. , 2011, “PEBBED Uncertainty and Sensitivity Analysis of the CRP-5 PBMR DLOFC Transient Benchmark With the SUSA Code,” Idaho National Laboratory, Idaho Falls, ID, Report No. INL/EXT-10-20531. https://inis.iaea.org/search/search.aspx?orig_q=RN:42048254
DOE, 1987, “Probabilistic Risk Assessment for the Standard Modular High Temperature Gas-Cooled Reactor,” Department of Energy, Washington, DC, Standard No. DOE-HTGR-86-011. https://www.nrc.gov/docs/ML1113/ML111310342.pdf
Kunitomi, K. , Katanishi, S. , and Takada, S. , 2004, “Japan's Future HTR–the GTHTR300,” Nucl. Eng. Des., 233 (1–3), pp. 309–327. [CrossRef]
AESJ, 2009, “A Standard for Procedures of Probabilistic Safety Assessment of Nuclear Power Plant During Power Operation (Level 2 PSA): 2008,” Atomic Energy Society of Japan, Tokyo, Japan, Standard No. AESJ-SC-S001.
The RELAP5-3D© code Development Team, 2005, “RELAP5-3D© Code Manual,” Idaho Natioanl Laboratory, Idaho Falls, ID, Report No. INEEL-EXT-98-00834. http://www4vip.inl.gov/relap5/r5manuals/ver_2_3/rv1.pdf
Sato, H. , Yan, X. , Tachibana, Y. , Kazuhiko, K. , and Kato, Y. , 2013, “Transient Analysis of Depressurized Loss-of-Forced-Circulation Accident Without Scram in High Temperature Gas-Cooled Reactor,” Nucl. Technol., 185(3), pp. 227–238. [CrossRef]
Sato, H. , Ohashi, H. , Nakagawa, S. , Tachibana, Y. , and Kunitomi, K. , 2014, “Validation and Application of Thermal Hydraulic System Code for Analysis of Helically Coiled Heat Exchanger in High-Temperature Environment,” J. Nucl. Sci. Technol., 51(11–12), pp. 1324–1335. [CrossRef]
Sato, H. , Nakagawa, S. , and Ohashi, H. , 2016, “Selection of Design Basis Event for Modular High Temperature Gas-Cooled Reactor,” Japan Atomic Energy Agency Institute, Ibaraki, Kapan, Report No. JAEA-Technology-2016-014. https://inis.iaea.org/search/search.aspx?orig_q=RN:48014742
James, J. D. , and Louis, J. H. , 1976, Nuclear Reactor Analysis, Wiley, Hoboken, NJ, p. 650.
Kunitomi, K. , Nakagawa, S. , and Itakura, H. , 1991, “Thermal Transient Analyses During a Depressurization Accident in the High Temperature Engineering Test Reactor (HTTR),” Japan Atomic Energy Research Institute, Ibaraki, Kapan, Report No. JAERI-M 91-163. https://inis.iaea.org/search/search.aspx?orig_q=RN:23028236
Saito, S. , Tanaka, T. , Sudo, Y. , Baba, O. , Shindo, M. , Shiozawa, S. , Mogi, H. , Okubo, M. , Ito, N. , Shindo, R. , Kobayashi, N. , Kurihara, R. , Hayashi, K. , Hada, K. , Kurata, Y. , Yamashita, K. , Kawasaki, K. , Iyoku, T. , Kunitomi, K. , Maruyama, S. , Ishihara, M. , Sawa, K. , Fujimoto, N. , Murata, I. , Nakagawa, S. , Tachibana, Y. , Nishihara, T. , Oshita, S. , Shinozaki, M. , Takeda, T. , Sakaba, S. , Saikusa, A. , Tazawa, Y. , Fukaya, Y. , Nagahori, H. , Kikuchi, T. , Kawaji, S. , Isozaki, M. , Matsuzaki, S. , Sakama, I. , Hara, K. , Ueda, N. , and Kokusen, S. , 1994, “Design of High Temperature Engineering Test Reactor (HTTR),” Japan Atomic Energy Research Institute, Ibaraki, Kapan, Report No. JAERI 1332. http://jolissrch-inter.tokai-sc.jaea.go.jp/search/servlet/search?2074267&language=1
U.S. NRC, 2008, “Next Generation Nuclear Plant Phenomena Identification and Ranking Table (PIRTs),” U.S. Nuclear Regulatory Commission Washington, DC, Report No. NUREG/CR-6944. https://www.nrc.gov/reading-rm/doc-collections/nuregs/contract/cr6944/

Figures

Grahic Jump Location
Fig. 1

Procedure for identification of influencing factors

Grahic Jump Location
Fig. 2

Detail flow for identification of influencing factors

Grahic Jump Location
Fig. 3

Classification of uncertainty

Grahic Jump Location
Fig. 4

Typical fuel temperature and reactor power behaviors in HTGR DLOFC accident with CRS failure

Grahic Jump Location
Fig. 5

Initiating event group in HTGR and uncertainty factors for uncertainty evaluation of source term

Grahic Jump Location
Fig. 6

Temperature dependency of release rate of Iodine-131 from failed coated fuel particle

Grahic Jump Location
Fig. 7

Schematic of heat transfer model

Grahic Jump Location
Fig. 8

Logical diagram (neutronics)

Grahic Jump Location
Fig. 9

Logical diagram (thermal hydraulics)

Grahic Jump Location
Fig. 10

Additional logical diagram

Grahic Jump Location
Fig. 11

Results of sensitivity analysis for factors at the level of physical phenomena: (a) Doppler reactivity, (b) moderator temp., reactivity (c) Xenon reactivity, (d) dynamic parameter (λ), (e) decay heat, (f) heat transfer (inner core), (g) heat conductivity (inner core), (h) radiation (outer core), and (i) heat capacity (inner core)

Tables

Errata

Some tools below are only available to our subscribers or users with an online account.

Related Content

Customize your page view by dragging and repositioning the boxes below.

Related Journal Articles
Related eBook Content
Topic Collections

Sorry! You do not have access to this content. For assistance or to subscribe, please contact us:

  • TELEPHONE: 1-800-843-2763 (Toll-free in the USA)
  • EMAIL: asmedigitalcollection@asme.org
Sign In