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BEPU and Safety Margins in Nuclear Reactor Safety
Get PDFAbstract. Approaches like Best Estimate Plus Uncertainty (BEPU) and concepts like Safety Margins (SM) are
well established in Nuclear Reactor Safety (NRS). However continuous improvements in analytical techniques
and in the sophistication of hardware products do not necessarily correspond to new industrial applications
within Nuclear Power Plants (NPP) technology. The declining condition for nuclear technology also contributes
to the lag between developments and applications definitely causing NPP safety at a level below the achievable
level. The possibility to extend BEPU to all areas of the Final Safety Analysis Report (FSAR), so-called BEPUFSAR
is outlined in the paper. This should be combined with the Extension of the SM concept (E-SM). BEPUFSAR
techniques may be at the origin of E-SM which also will need specific monitoring hardware. All of this
may open new horizons for NRS and for acceptance of NPP by the public and the decisions makers. The paper
describes recent accomplishments in the areas of BEPU and E-SM
Conjugating ALARA, BEPU, Safety Margins and Independent Assessment in Nuclear Reactor Safety
Get PDFALARA (As-Low-As-Reasonably-Achievable) is an early principle in Nuclear Reactor Safety, NRS (Nuclear Reactor Safety): Designers and Operators must do their best to minimize doses to the humans. BEPU (Best Estimate Plus Uncertainty) is an approach in Accident Analysis, part of NRS: one may state that BEPU implies the best use of computational tools to determine the safety of nuclear installations. Then, ALARA may be seen at the origin of BEPU, or ALARA is at the origin of BEPU. Furthermore, BEPU (and BEPU elements like V & V, Scaling, procedures of code application and code coupling, etc.) can be extended to all analytical parts of the Final Safety Analysis Report (FSAR). This brings to BEPU-FSAR. Safety Margin (SM) is an established concept in NRS: a few dozen SM values must be calculated in current safety analyses and demonstrated to be acceptable. The SM concept can be extended to everything part of the design, the operation and the environment for a Nuclear Power Plant (NPP) Unit. Here the environment includes the personnel in charge of activities connected with the NPP. The Extended SM concept, E-SM, implies the formulation of some ten-thousands SM values, which shall correspond to a similar number of monitored variables. Reasons for E-SM are the examples in section 4.1. Independent Assessment (IA) is an early requirement in NRS: data ownership and system complexity prevented so far a comprehensive application of the requirement. IA analyses conflict with industry policies to keep proprietary data. IA based BEPU-FSAR analyses are essential to finalize the E-SM design.
In the paper we discuss that: a) ALARA is at the origin of BEPU; b) BEPU-FSAR analyses are the natural origin of E-SM values; c) The implementation of E-SM equals to introducing an additional physical barrier against the release of fission products
The need of adding a safety barrier to water cooled nuclear reactors
Get PDFThe present paper deals with the proposal of an additional safety barrier for the class of large (1000 MWe or more) Light Water Reactors (LWR) now in operation, in construction, or under design. Emphasis is given to the motivations or the needs for the barrier. Two main parts of the paper can be distinguished. The following topics are discussed in the former part: (a) the weakness of the barrier constituted by the current design of nuclear fuel; (b) the continuously increasing complexity of the system, with main reference to the Instrumentation and Control (I&C); (c) the role that the Large Break Loss of Coolant Accident (LBLOCA) had for arriving at the current layout of the Reactor Coolant System (RCS). Furthermore avoiding the severe accidents in 1979, 1987 and 2011, is at the basis of the proposal. In the latter part, the elements of the proposed technological safety barrier are discussed: the As-Low-As-Reasonably-Achievable (ALARA) principle, the Best Estimate Plus Uncertainty (BEPU) approach, the Extended Safety Margin Detection (E-SMD) hardware, the Emergency Rescue Team (ERT) strategy (or a virtual entity for the reactor) and the Independent Assessment (IA) concept. The additional safety barrier, although not demonstrated in the paper, is expected to reduce for a factor in the range 10-1000 the probability of core melt and to have a cost in the order of 1% the cost of a nuclear reactor unit
Physical Phenomena in Nuclear Thermal Hydraulics and Current Status
No full text116 nuclear Thermal-Hydraulic Phenomena T-HP are identified in the present paper,
following documents issued during the last three decades by the Committee on the Safety
of Nuclear Installations of Nuclear Energy Agency of the Organization for Economic
Cooperation and Development (OECD/NEA/CSNI) and by the International Atomic
Energy Agency (IAEA). The derived T-HP list includes consideration of experiments
performed in Separate Effect Test (SET) and Integral Effect Test (IET) facilities relevant to
reactor coolant system and containment of Water Cooled Nuclear Reactors (WCNR). We
consider a dozen WCNR types: Pressurized Water Reactors (PWR), Boiling Water Reactors
(BWR), Russian reactors (VVER-440, VVER-1000 and RBMK), pressure tube heavy water
reactors by Canada (CANDU) and India (PHWR) and so-called ‘advanced’ reactors (e.g.
AP-1000 and APR-1400 designed in US and Korea, respectively).
We envisage a variety of applications for the T-HP list. Four of the phenomena are helpful
to characterize the current state of art in nuclear thermal-hydraulics: Counter Current Flow
Limitation (CCFL), Critical Heat Flux (CHF), reflood and Two-Phase Critical Flow
(TPCF). Furthermore, the T-HP identification contributes to addressing the scaling issue,
performing uncertainty evaluations, developing constitutive equations and ‘special models’
in codes and prioritizing the research
Introduction of an additional safety barrier for nuclear power reactors
Get PDFIn the present paper a proposal is formulated to improve the safety of existing and future nuclear reactors. The idea is based upon the introduction of a new safety barrier against the release of radioactivity generated by the fission chain process. The proposal aims at fixing bases for possible strengthening of current Nuclear Reactor Safety by combining the logical frameworks associated with the concepts As-Low-As-Reasonably-Achievable (ALARA), Best-Estimate-Plus-Uncertainty (BEPU),Extended-Safety-Margin (E-SM), Independent-Assessment (IA) and Emergency-Rescue-Team (ERT). The expected impact of the new barrier upon selected nuclear accidents is outlined. The cost for the implementation of the additional barrier is expected to be affordable from a financial viewpoint and to contribute to restoring the public confidence towards nuclear technology
The technological challenge for current generation nuclear reactors
Get PDFThe present paper deals with the proposal of an additional safety barrier for the class of large (1000 MWe or more) Light Water Reactors (LWR) now in operation, in construction, or under design. Emphasis is given to the motivations or the needs for the barrier. Two main parts of the paper can be distinguished. The following topics are discussed in the former part (section 2): (a) the weakness of the barrier constituted by the current design of nuclear fuel; (b) the continuously increasing complexity of the system, with main reference to the Instrumentation and Control (I&C); (c) the role that the Large Break Loss of Coolant Accident (LBLOCA) had for arriving at the current layout of the Reactor Coolant System (RCS). Furthermore avoiding the severe accidents in 1979, 1987 and 2011, is at the basis of the proposal. In the latter part (sections 3 and 4), the elements of the proposed technological safety barrier are discussed: the As-Low-As-Reasonably-Achievable (ALARA) principle, the Best Estimate Plus Uncertainty (BEPU) approach, the Extended Safety Margin Detection (E-SMD) hardware, the Emergency Rescue Team (ERT) strategy (or a virtual entity for the reactor) and the Independent Assessment (IA) concept. The additional safety barrier, although not demonstrated in the paper, is expected to reduce for a factor in the range 10–1000 the probability of core melt and to have a cost in the order of 1% the cost of a nuclear reactor unit
NEW SAFETY BARRIER FOR CURRENT AND FUTURE NUCLEAR REACTORS
Get PDFA proposal is formulated in the present paper to improve the safety of existing and future
nuclear reactors. The proposal is based upon the introduction of new safety barrier against the
release of radioactivity generated by the fission chain process.
Basically, two motivations for the proposal arise:
a) in the last two or three decades experimental evidence demonstrated that the established
barrier constituted by the fuel pin clad is weak and easily trespassed by fission products during
various phases of the fuel cycle, with main regard (here) to the in-core irradiation;
b) the probability of core melt must (and can) be substantially lowered by adopting
outcomes from recent researches with main reference to the reached capabilities of
computational tools.
Furthermore, the correct interpretation of the words ‘feedback from lessons learned’
impose that the conditions which led to the occurred severe accidents are understood to be not
replicable in the future and, according to the statement of Australian and Chinese scientists,
“upgrading and strengthening a nuclear regulatory system is not optional but imperative to
prevent the next core meltdown”.
The proposal aims at fixing bases for possible strengthening of current Nuclear Reactor
Safety by combining the logical frameworks connected with the terms As-Low-As-Reasonably-
Achievable (ALARA), Best-Estimate-Plus-Uncertainty (BEPU), Extended-Safety-Margin (ESM)
and Independent-Assessment (IA). The cost for the implementation of the proposal is
expected to be affordable from a financial viewpoint and to contribute to restoring the public
confidence towards nuclear technology
Thermal-hydraulic phenomena for water cooled nuclear reactors
No full textNuclear Reactor Safety (NRS), Deterministic Safety Assessment (DSA) and Accident Analysis (AA) constitute the
general framework for the topic of the present paper. The class of Water-Cooled Nuclear Reactors (WCNRs) is
concerned. This includes most of the nuclear reactors in operation, under construction or in advanced design
stage. The required licensing process for those reactors, are further necessary elements to establish the context of
the performed activity.
Best Estimate (BE) system thermal-hydraulic codes are adopted to demonstrate the safety of WCNR based on
AA, namely focusing on the class of Design Basis Accidents (DBAs). On the one hand, the validation of BE codes
is a necessary step to prove their applicability to calculate accident scenarios. On the other hand the knowledge
of accident scenarios is a requirement for the design and the operation of WCNR. The validation of BE codes and
the knowledge of accident scenarios needs the identification and the characterization of Thermal-hydraulic
Phenomena (T-HP).
A list of 116 T-HP is derived in the present paper, based on the documents issued in the last three decades by
the Committee on the Safety of Nuclear Installations of Nuclear Energy Agency of The Organization for
Economic Cooperation and Development (OECD/NEA/CSNI) and by the International Atomic Energy Agency
(IAEA). The T-HP list includes the consideration of Separate Effect Tests (SET) and Integral Effect Tests (IET)
relevant in Reactor Coolant System (RCS) and Containment of WCNRs. A dozen WCNR types are considered and
include Pressurized Water Reactors (PWRs), Boiling Water Reactors (BWRs), Russian design reactor types (e.g.
VVER-440, VVER-1000 and RBMK), pressure tube heavy water reactor designs in Canada (CANDU) and in India
(PHWR) and so-called ‘advanced’ reactors (in the text of this paper, they are sometimes assigned as “New
Reactors”), which are also equipped with passive systems (for instance, AP-1000 and APR-1400 designed in US
and Korea, respectively).
Each T-HP can be characterized by one or more parameters or variables which are part of numerical models
and constitute calculational results from system codes. A cross link process can be established between T-HP,
parameters and DBA scenarios. The basis for the process and selected cross-link examples are provided and
discussed.
A variety of applications for the T-HP list is envisaged in nuclear thermal-hydraulics. Insights are given in the
paper in relation to the use of phenomena: a) to address the scaling issue; b) to distinguish between constitutive
equations part of the balance equations and ‘special models’ in BE system codes; c) to prioritize research in
nuclear reactor thermal-hydraulics
BUILDING-UP THE LIST OF 116 THERMAL-HYDRAULIC PHENOMENA
No full textFOREWORD & SCOPE
- ISSUED PAPER
-J NED 2018
- ORIGINATING DOCS/ BRIEF HISTORY
- 116 PHENOMENA (TH-P) LIST
- CROSS-LINKING Reactor Types, TH-P, Accident Scenarios and Parameters
- CONCLUSIONS & PROPOSED DEVELOPMENTS
- ACKNOWLEDGEMENT
Best Estimate Plus Uncertainty (BEPU) approach and safety margins in nuclear reactor safety
No full textApproaches like Best Estimate Plus Uncertainty (BEPU) and concepts like Safety Margins (SM) are well established in Nuclear Reactor Safety (NRS). However continuous improvements in analytical techniques and in the sophistication of hardware products do not necessarily correspond to new industrial applications within Nuclear Power Plants (NPP) technology.
The declining condition for nuclear technology also contributes to the lag between developments and applications definitely causing NPP safety at a level below the achievable level. The possibility to extend BEPU to all areas of the Final Safety Analysis Report (FSAR), so-called BEPU-FSAR is outlined in the paper. This should be combined with the Extension of the SM concept (E-SM). BEPU-FSAR techniques may be at the origin of E-SM which also will need specific monitoring hardware. All of this may open new horizons for NRS and for acceptance of NPP by the public and the decisions makers. The paper describes recent accomplishments in the areas of BEPU and E-SM
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