Quantification of initiating event frequencies and component reliability data in level 1 probabilistic safety assessment at Puspati TRIGA research reactor M. Maskin a,b , A. Hassan c , F.C. Brayon d , P.T. Phongsakorn b , M.F. Zakaria b , Z. Ramli d , F. Mohamed a,⇑ a Faculty of Science and Technology, University Kebangsaan Malaysia, Selangor, Malaysia b Reactor Technology Center, Malaysian Nuclear Agency, Selangor, Malaysia c Faculty of Applied Sciences and Technology, Universiti Tun Hussein Onn Malaysia, Johor, Malaysia d Atomic Energy Licensing Board, Selangor, Malaysia article info Article history: Received 30 April 2018 Received in revised form 3 July 2018 Accepted 6 July 2018 Keywords: Reliability data Initiating event Probabilistic safety assessment Research reactor Malaysia abstract Prior to developing level-I probabilistic safety assessment in Puspati TRIGA Reactor (RTP), collecting and analyzing data is the cornerstone of the probabilistic safety assessment processes. This paper provides on how data quantification is made available for both initial initiating events frequency and components reliability. Taking into consideration on the availability of data, it was decided that generic data to be used in combination with plant-specific data. Methods for estimating the parameters used in quantifying is also presented in this paper. This includes, data treatment by applying Bayesian updates in forming a better refine posterior data. It was found out that, result with the presence of plant-specific data history tends to have high value, loss of offsite power (LOOP) with 1.27E-03. Meanwhile, for other IEs that used generic data, shown lower result with maximum value of 2.68E-06 (loss of flow accident at secondary pump: LOSC-P). As for components data, basic event, No signal from fission chamber 1 (fail to function) is the highest value with 4.07E-01, followed with Manual valve SV1A fail to close (2.75E-01). This pub- lication is perhaps one of the first publically available document providing information on generic and plant specific data for initiating event and components reliability parameters that were used in a Level-1 PSA study for a nuclear research reactor. In addition, this information can be used by PSA tech- nical staff in how to analyze and treat initial incident event data, including, type and what data need to be collected. Ó 2018 Elsevier Ltd. All rights reserved. 1. Introduction Probabilistic safety assessment (PSA) technique is increasingly being used in many countries operating research reactor as an assessment tool of risk associated with a broad range of potential accident scenarios, identify adverse effects of various risk contrib- utors (e.g. equipment failures, human errors, etc.), and determines the measures for further enhancement of plant safety by incorpo- rating the risk insights in the decision-making process (IAEA, 1987; IAEA, 2010; Brayon et al., 2014; IAEA, 2001). PSA is usually developed based on logical and systematic approach that makes use of realistic assessments of equipments and plant personnel performance as a basis for calculations of risk. This in principle has the potential to produce an understanding of the inherent risk of the operating plant over a much wider range of conditions than the traditional deterministic methods which gen- erally define what is assumed to be a bounding set of fault condi- tions (IAEA, 1200; Barón, 2010; IAEA, 1989). Furthermore, the adoption of conservative assumptions relating to plant and system performance is an accepted approach to address uncertainty when performing deterministic analyses. The combination of considering a limited number of faults and a conservative approach to the anal- ysis of each fault can produce inappropriate, or worse, misleading insights, and therefore decisions solely based on these determinis- tic types of analyses might not always be the most appropriate way for reducing plant risk (IAEA, 1200). In order to estimate the frequencies of occurrence for accident sequences analyzed in PSA, three basic input parameters must be quantified: (a) initiating events (IEs) frequency; (b) component https://doi.org/10.1016/j.anucene.2018.07.013 0306-4549/Ó 2018 Elsevier Ltd. All rights reserved. ⇑ Corresponding author. E-mail addresses: mazleha@nm.gov.my (M. Maskin), hassan@uthm.edu.my (A. Hassan), fedrick@aelb.gov.my (F.C. Brayon), phongsakorn@nm.gov.my (P.T. Phongsakorn), fazli@nm.gov.my (M.F. Zakaria), zulfadli@aelb.gov.my (Z. Ramli), faizalm@ukm.edu.my (F. Mohamed). Annals of Nuclear Energy 121 (2018) 22–28 Contents lists available at ScienceDirect Annals of Nuclear Energy journal homepage: www.elsevier.com/locate/anucene