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safetyanalysese 4 NPPS in Ro
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5/21/2018 Safety Analysese for NPPs in Romania
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NATIONAL COMMISSION FOR
NUCLEAR ACTIVITIES CONTROL
Regional Workshop on Application of Best Estimate Plus Uncertainty (BEPU)Analyses, 10-14 March 2008, Budapest, Hungary
CENTER OF TECHNOLOGY AND
ENGINEERING FOR NUCLEAR PROJECTS
ROMANIA 1
Safety Analyses Performed for Nuclear Power Plants
in Romania
Elena DINCA, National Commission for Nuclear Activities Control
Daniel BOGDAN - (CNCAN), Bucharest, Romania
Virgil IONESCU Center of Engineering and Technology for Nuclear
Projects (CITON), Bucharest-Magurele, Romania
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10-14 March 2008, Budapest, HungaryROMANIA 2
Content
1. Introduction
2. Conservative and Best Estimate Safety Analysis
Methods for CANDU-6 Design NPPs
3. CNCAN policy for NPPs licensing
4. Nuclear safety analyses performed for Cernavoda
NPPs
5. Specific safety issues to CANDU NPPs
6. Conclusions
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Regional Workshop on Application of Best Estimate Plus Uncertainty (BEPU) Analyses,10-14 March 2008, Budapest, Hungary
ROMANIA 3
Romaniainterested in CANDU 6 project (700 MWe)
CANDUCANadian Deuterium Uranium Cernavoda NPPs: Unit 1 in operation (since 1996) Cernavoda NPP Unit 2 at 100%FP (September 2007) Cernavoda NPP Units 3, 4, 5 under preservation (feasibility study)
Nuclear Safety Analyses, performed for: NPPs licensing NPPs ageing assessment Support for a continuous nuclear safety improvement, by:
Plant modifications
New design Support for plant lifetime extension, etc. Analysis methodology models and input data
Are in a continuous improving process. The conservative approach isgoing to be replaced by a more realistic best estimate plusuncertainty analysis.
1. Introduction
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ROMANIA 4
Deterministic safety analyses may be performed: at the limit of operating envelope (LOE)or
a best estimate and uncertainty (BEAU)methodology may be used.
As a rule, uncertainties are not included in the LOE analysis
CANDU reactors were licensed using deterministic conservative
safety analysis, which evaluates consequences from postulatedinitiating events and sequences of events (LOE method).
The essential elements of the LOE analysis are as follows:
Analysis input parameters:key and nonkey operating/design parameters
Modeling parameters
Plant operating state
Deterministic assumptions
Computer models
2. Conservative and Best Estimate Safety Analysis
Methods for CANDU-6 Design NPPs
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Regional Workshop on Application of Best Estimate Plus Uncertainty (BEPU) Analyses,10-14 March 2008, Budapest, Hungary
ROMANIA 5
The CANDU design featureshistorically have influenced formation ofthe Canadian regulatory philosophy, as well as of the safety analysismethods and acceptance criteria.
According to RD-310 Canadian regulatory document: The safetyanalysis shall build in a degree of conservatism to off-set anyuncertainties associated with both NPP initial and boundary conditionsand modeling of nuclear power plant performance in the analyzedevent. This conservatism shall depend on event class, and shall becommensurate with the analysis objectives.
Historically, the reasons for excluding random modeling uncertaintieswere as follows:
large originally predicted margins
belief that conservatism achieved by assuming the limiting values of
operational parameters and imposing of certain deterministic assumptionsmore than adequately covers modeling uncertainties
lack of well defined modeling uncertainties
The impact of modeling uncertainties is usually investigated byperforming sensitivity studies.
CANDU-6 Safety Philosophy
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ROMANIA
Deterministic Nuclear Safety Analysis
SIMPLIFIED DIAGRAM
Deterministic
Safety Analysis
Support AnalysisDesign Analysis Licensing AnalysisPerformance Analysis Uncertainty Analysis
Normal
Operation
Abnormal
Operation
Accident
Condition
Radiological
Analysis
Containment
Analysis
Thermal-Hydraulic
Analysis
Fuel and Fuel
Channel Analysis
Neutron
Analysis
Structural Analysis
Common Cause Event Analysis
Nuclear Safety
Assessment
LOCA events NON-LOCA events
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ROMANIA 7
From a safety analysis perspective, the CANDU reactor has some
distinctive features and characteristics:
The natural uranium fuel resides in a matrix of individual horizontal
fuel channels within short fuel bundles and is irradiated to relatively
low burnup
The primary circuit is relatively complicated, with heavy water at
high pressure and high temperature (100 bar, 312C)
The moderator system is separated from the coolant, contains a
significant amount of heavy water (~260 Mg)at low pressure and
low temperature (1.02 bar, 65 C) with a cooling capacity of about
5% FP
Re-fuelling is performed at power
The reactor has a positive core void
reactivity coefficient.
CANDU 6 - Main Characteristics
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ROMANIA 8
CANDU-6 NPP scheme
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ROMANIA 9
The operating power reactors in Canada have been licensed according
to requirements as those presented in the C-6 regulatory guideRequirements for the Safety Analysis of CANDU Nuclear Power Plants(C-6, 1980, issued by AECB).
According to C-6 and RD-310 regulatory documents of CNSC, theevents in CANDU NPPs could be classified as: Anticipated Operational Occurrences (AOOs) include all events with
frequencies of occurrence equal to or greater than 10-2 per reactor year; Design Basis Accidents (DBAs) include events with frequencies of
occurrence equal to or greater than 10-5 per reactor year but less than 10-2per reactor year; and
Beyond Design Basis Accidents (BDBAs) include events with frequencies ofoccurrence less than 10-5 per reactor year.
Accidents are also categorized into 5 classes which reflect the frequencyof the accident. For example:
Class 1 category: highest frequency; high number of occurrences perreactor year (1 per 100 years < 1/ f < 1 per year)
Class 5 category: lowest frequency; low number of occurrences per reactoryear (1 per 100000 years > 1/f)
CANDU 6Events Classification (1)
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ROMANIA 10
The evaluation methodology and acceptance criteria are
different for different event categories. Briefly, the evaluationmethods are as follow:
Category Aevents are deterministically analyzed. Conservativeassumptions are used for initial plant conditions and mitigatingsystems availability, which impose the most stringent conditions onsafety system design. The category A events are called Design
Basis Events; their analysis is the subject of Chapter 15 of a NPPPreliminary/Final Safety Report.
Category B events are analyzed probabilistically. Realisticassumptions are used to provide information to operators on themost probable plant response in case of the analyzed event.
Category C and Devents are qualitatively assessed. Category Ccomprises external hazard events and Category D comprisesevents with a very small occurrence frequency (e. g. steamgenerator support or shell failure).
CANDU 6Events Classification (2)
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According to C-6 Canadian consultative document, the analysisfor AOOs and DBAs shall demonstrate that:
Radiological doses to members of the public do not exceed theestablished limits (acceptance criteria level 0); and
The derived acceptance criteria, established for accidents are met
Safety analysis for AOOs and DBAs shall demonstrate thecorrect application of safety principles.
Analysis for BDBAs shall be performed as part of the safetyassessment to demonstrate that:
The nuclear power plant as designed can meet the established
objectives of safe operating envelope; and The accident management program and design provisions put in
place to handle the accident management needs, are effective.
CANDU 6 Safety Analysis Methodology
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According to C-6 consultative document:
Mathematical models and calculational methods(including inputdata) recommended for CANDU NPPs represent a conservativeprediction for each of the safety analysis requirements;
Allow for bias in calculational methods at high confidence limits (95percent). In higher event classes, assess sensitivity analysis,degraded mitigating system functional capability, a second diversemitigating system actuation parameter, and worse plant states assurrogates for calculational tolerances of higher confidence.
All physical phenomena should be accounted for, and simplificationsshould be appropriate.
When choosing conservative assumptions and error tolerances,
identify and account for the presence of each effect separate(thresholds, timing, competing effects, different failure mechanism,different reactions, different transport processes, structural integrity).
CANDU 6 Safety Analysis Methodology (2)
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CANDU 6 Conservative Assumptions for
Deterministic Safety Analysis
The deterministic approach uses several generic assumptionswhich are applied in assessing the consequences resulting fromthe postulated accidents. These include the following:
Reactor trip occurs at the second trip signal on the less effectiveshutdown system;
Intervention by the operator is not credited during the first 15
minutes following the the clear and unambiguous indication thatan initiating event has occurred and that operator action isrequired initiating event;
Mitigating action by process system response is not credited;
Each special safety system is assumed to be in its minimum
allowable performances configuration.
Computer codes and models are in a continuous improvingprocess and allow a better simulation of CANDU specificphenomena
D t i i ti f t l i d i
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- local moderatorbehaviour
ContainmentPRESCON2
GOTHIC
Thermal hydraulics
CATHENA
Reactor Physics
RFSP
Fuel Channel
CATHENA
Fuel
ELESTRES
ELOCA
Atmospheric
DispersionADDAM
- header boundary
conditions
- power transients
- power transient
- thermal hydraulic boundary
conditions
- fuel/sheath temperatures
- metal/water reaction- fission product inventory
distribution
- fuel failure
- fission product release
- pressure tube strain
- post-contact pressure tube/
calandria tube behaviour
- high building pressure trip
- ECC conditioning signal
- activity release
- weather scenario
- release height/location
- power transient
Public
Dose PEAR
Moderator
MODTURC_CLAS
- coolant characteristics
Deterministic safety analysis process used inpresent for CANDU-6 NPPs
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CATHENA Computer Code Characteristics (1)
CATHENA- The main computer code used in present for
deterministic thermalhydraulic safety analysis CATHENA- CanadianAlgorithm for THErmalhydraulic NetworkAnalysis
One-dimensional, two-fluid system thermalhydraulics code
Developed by AECL primarily for analysis of postulated LOCAevents in CANDU reactors and then developed
Non-equilibrium model (2-velocities, 2-temperatures, 2-pressuresplus noncondensables)
CATHENA interfaces to other codes:
Fuel Behaviour: CATHENA / ELOCA
Reactor Physics: CATHENA / RFSP Containment Thermalhydraulic Behavior:
CATHENA/PRESCON2
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CATHENAs heat transfer model:
Multiple surfaces perthermalhydraulic node
Radial and circumferentialconduction modeled
Models heat transfer within
bundles subjected to stratifiedflow
Radiation heat transfer calculated
Built-in temperature dependentmaterial property tables
Models deformed geometry andpressure/ calandria tube contact
CATHENA Computer Code Characteristics (2)
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CATHENA Validation
CATHENA has been validated, in a formal process, for safety and
licensing analysis of CANDU Reactors, including by the experimentalloop simulation RD-14
Validation has proceeded on a phenomenon-by-phenomenon basis
Standardized and documented models of facilities used where they exist
Default code settings used throughout unless otherwise specified and
justified
Data selected in validation process includes numerical tests, separateeffects, component and integral tests, as well as transients in CANDUplants
Sensitivity analysis conducted to identify impact on simulations of
experimental errors used as boundary conditions (e.g., power) andnodalization
Uncertainty analysis conducted to identify impact on code results (e.g.,uncertainty in heat transfer correlations)
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Development of Best Estimate Analysis Methods for
CANDU Project (1)
It has been recognized, by both the industry and the Canadianregulator, that BE+UA methods have reached sufficient maturity
to allow more accurate and realistic modeling of accident
transients, thus presenting an opportunity to better quantify
safety margins.
It is expected that in many cases a BE+UA analysis will be ableto show larger margins than it was possible to demonstrate using
the conservative approach.
If a BEAU-type of analysis methodology is used, the acceptance
criteria should be met at a certain level of probability andconfidence limit commensurate with the risk posed by the
postulated event.
D l f B E i A l i M h d f
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The industry has started several projects aimed at thedevelopment and application of BE+UA methods
Results of the BE+UA analysis are expected to play animportant role in decisions related to:
avoid of economic penalties,
relaxation of overly restrictive operational practices,
dealing with plant ageing effects,
resolution of outstanding safety analysis issues
The industry has requested the Canadian regulatory body toevaluate the admissibility of such methods for licensingpurposes
There is confidence that the best estimate methods will find wideuse in the licensing process in Canada in the next future and inthe other countries which operate CANDU NPPs, includingRomania.
Development of Best Estimate Analysis Methods for
CANDU Project (2)
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3. CNCAN Policy for NPPs Licensing
General licensing conditions of a Romanian NPP are provided in Law
No. 111/1996 on Safe Deployment of Nuclear Activities withsubsequent modifications and completions
Specific requirements for Romanian NPPs licensing related to safety
analyses are provided in CNCAN Norms. CNCAN elaborated specific
regulations for special safety systems, Periodic Safety Review, and
Probabilistic Safety Assessment reports content. CNCAN uses the following international standards in the process of
nuclear regulation in Romania: IAEASafety Standards and Guides;
AECL (Atomic Energy Canada Limited) Standards and
Guides
Regulatory documents developed by Canadian Nuclear
Safety Commission (CNSC) and US NRC;
Applicable Standards and Codes (CSA, ANSI, ASME, IEC,
IEEE, etc.);
Doc ments containing req irements sed b
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Documents containing requirements used by
CNCAN in the licensing process
Law
Regulations, Standards and
Codes
Regulatory letters
+licensees documents previously approved by CNCAN
(such as Safety Analysis Reports, Management Manuals,
etc.)
(
CNCAN dispositions and actions stated in the inspection reports
+licensees procedures previously approved by CNCAN (such as reference
documents, station instructions, etc.)
Regulatory requirements,
criteria and conditions
Regulations,
Standards and Codes
Law
Regulatory letters+
licensees documents
previously approved by CNCAN
CNCAN dispositions and actions
stated in the inspection reports +
licensees procedures previously approved
by CNCAN
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CNCAN Approach for Safety Analysis Methods (1)
Romanian regulatory systeminfluenced by the Canadian
regulatory philosophy but much more prescriptive. The current Canadian regulatory regime is based on the principle that
the licensee has primary responsibility for safety and that detailedregulatory prescription is unnecessary and detrimental to the licenseecarrying out that responsibility.
It is a firmly established principle in the Romanian regulatory practice torequire that adequate safety margins be maintained and demonstratedby the safety analysis.
The analysis must show that the facility meets all specified criteria withsufficient margins to cover any uncertainties in the methods of analysis.
In Romania, as in Canada, two types of acceptance criteria are used in
safety analyses:
radiological dose limits
derived acceptance criteria.
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Deterministic safety analyses for Design Basis Accidents are
provided in Chapter 15 of FSAR for NPP licensing The requirements for the Design Basis Accidents for Cernavoda
NPP Unit 1 &2 were reviewed before the commissioning of theplant started, and during the commissioning phases. The reviewwas based on:
the feedback from other projects, updated the systematic review using PSA level 1 analyses in parallel with
the licensing process
the use of external independent expertise for those topics forwhich independent review of the evaluations was needed.
It was reviewed the trip coverageas defined in the FSAR,which was of concern mainly for partial and low power states.
current status of the research was considered, like theexperiments on molten fuel -moderator interaction, review of theanalyses for flow blockage etc..
CNCAN Approach for Safety Analysis Methods (2)
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Aplication of CANDU Defense in Depth concept
Operating
Limit
Trip imit
Safety imit
Operating Margin
Safety Margin
Operating
Domain
Design Center
Normal Operation
Design Basis Accidents
Severe Accidents
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Accident Analysis Acceptance Criteria
Acceptance criteria level 0Releases limiting individual doses
REQUIREMENT EVENT CLASS
1 2 3 4 5
(i) Effective dose mSv
(ii) Eye mSv
(iii) Skin (average on 1 cm2)mSv
(iv) Liquid effluent emissions during 30 days
are less than derived annual dose limits for normal operation
0,5
5
20
+
5
50
200
+
30
300
1200
N
100
1000
4000
N
250
1500
5000
N
Legend: + the limiting condition will be satisfied for the worst failure sequence of the
specified class of events
N the limiting condition is not necessary
Acceptance criteria level 1Criteria derived from associated licensing requirements
Acceptance criteria level 2Criteria derived from analysis modelling assumptions
EVENT CLASS Probability1/reactor * year
1
2
3
4
5
>10-2
10-2 10-3
10-3 10-4
10-4 10-5
10-5 10-7
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Requirements for the licensee
The Licensee has to take actions to develop and maintain thecapability for full independent from the original designercalculations.
The Licensee will address the issue of the systematical review
of safety within the periodical safety review process, in thelong term research and development program, in cooperationwith COG.
The performance of all the calculations for the FSAR and
support documents, are to be performed based on codeswhich are in V&V (verification and validation ) and/or alreadyvalidated. The data are checked against the site specific data.
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4. Nuclear Safety Analyses Performed for Cernavoda NPPs
For Cernavoda NPP Unit 1, the process systems failures analysed in
Chapter 15of FSAR, include: loss of reactor regulation;
LOCA events (large LOCA and small LOCA);
pressure tube rupture;
channel flow blockage;
end-fitting failure; fuelling machine events;
pipe breaks in HT auxiliary systems;
loss of off-site power;
seizure of a primary heat transport system main pump;
pressurisation events - primary side;
depressurisation events - primary side;
feedwater line breaks;
steam main breaks;
steam generator tube
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Accident Analysis Technology for Cernavoda NPP Unit 1
ACCIDENT ANALYSIS APPROACH:
Conservative approach = Conservative codes + Conservative assumptions
DESIGN BASIS
ACCIDENT
Single Failure
F/H
PHTS
LOCA
-S-LOCA
- L-LOCA
- PT rupture
- EF rupture
- Flow blockage
- Feeder break
- Single SG tub rupture
- Multiple SG tub rupt.
- HTS auxiliary failures
LOF
-Loss of normal power
- Pump seizure
LOR
-Loss of P&IC
- Loss of P&IC compon
- Loss of react. control
S/G + FW
F/W break
-Up. Check valve
-Dw. Check valve
LOF
Loss of normal
power
LOR
-Loss of BPC
- Loss of BLCLOCA
On reactor
ESCS
LOI
Pipe
breaks
LOF
Loss of
normal
power
LOHS
Loss of
RCW,
RSW
MMS
LOI
Pipe
breaks
LOF
Loss of
normal
power
LOHS
Loss of
RCW,
RSW
LOCA
Off reactor
S/G + MSS
MSLB
Inside R/B
MSLB
Outside R/B
Dual Failure =
Single Failure
+ECCS impairment
Containment impairment
Nuclear safety analyses performed for
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Nuclear safety analyses performed for
Cernavoda NPPs (2)
The safety analyses for Cernavoda NPP Unit 2were based onthe guidelines provided in the document Requirements for theSafety Analysis of CANDU Nuclear Power Plants (C-6, 1980,issued by AECB).
For Cernavoda NPP Unit 2, the analyses provided in the
Chapter 15 of the FSAR were grouped in sections dedicated to:
Heat transport system LOCA events
Heat transport system non-LOCA
Steam and feedwater circuit events
Moderator events
Shield cooling events
Nuclear safety analyses performed for
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Examples of safety analysis requirements introduced by C-6consultative document that differ from previous practices aregiven as follows:
a requirement for a systematic review for the identification ofpostulated initiating events;
five event classes, replacing the two concepts of single and dualfailures;
correlation of event classes with probability of occurrence andallowable release limit;
more explicit consideration of combinations of postulated initiating
events with failures of mitigating systems (not just the dual failuresconcept).
Nuclear safety analyses performed for
Cernavoda NPPs (3)
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Licensing requirements for Unit 2
Cernavoda NPP embedded in the basic safety design the coverage of
a broad area of BDBA events (Category A of events, analyzed in
Chapter 15 of PSAR/FSAR)
The results of the review lead to a situation when some events in the
category of BDBA should be included in the DBA category for
Cernavoda unit 1. For Unit 1, the modifications are to be addressed in
the Periodical Safety Review process starting from 2008.
These results are applicable for Unit 2. The review of the licensing
requirements for unit 2 led to postulating more events from the BDBA
category, some of them involving the re-qualification (EQ for instance)
by comparison with Unit 1.
There are going on evaluations of the Cernavoda safety, including theDBA definition.
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Category A Events (1)
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Category A Events (2)
N l f t l f d f
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A Safety Analysis Strategic Programme(SASP)was developed by
Cernavoda NPP Unit 1 and approved by CNCAN The main objective of SASP is to get a better definition of the plant
safe envelope
SASPintended also to create and develop a group that will beable to perform and re-evaluate the safety analyses results.
SAPS purpose was to update, based on plant specific models andstate of the art computer codes, the entire set of accident analysesincluded in the Cernavoda Unit 1 Safety Analyses Report.
In the framework of SASP, Cernavoda NPP started to perform anew set of analyses for Unit 1, in accordance with the currentregulatory requirements and standards.
Cernavoda NPP is a member of COG (CANDU Owners Group) and
uses the last versions of CANDU specific computer codes and shares
with COG members a common data base, SEREX, containing
operating experience events.
Nuclear safety analyses performed for
Cernavoda NPPs (4)
Nuclear safety analyses performed for
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The generic severe accident analyses and severe accidentmanagement guidelines (SAMGs) developed by the CANDUOwners Group for CANDU 6 are going to be used byCernavoda NPP in the elaboration of plant specific analysesand SAMGs.
At the moment, specific information reports and procedures areprepared at the plant for establishing the framework for thisprogramme, including the allocation of resources and thenecessity of training, the activities that need to be performedwith external support, etc.
The specific computer code MAAP4-CANDU will be used forintegrated analysis of severe accident sequences.
Nuclear safety analyses performed for
Cernavoda NPPs (5)
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Computer code and plant model val idat ion
Recent experimental data in reactor physics area identified several
shortcomings of the major analysis tools in cell codes such as
POWDERPUFS-V. The most important shortcomings found are:
inaccurate predictions of key parameters for accident conditions,
lack of proper validation data for important phenomena and
range of conditions, and
significant gap between the state of knowledge reflected in the
licensees computer codes and the current state of knowledge in
this area.
5. Specific safety issues to CANDU NPPs
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Compl iance w ith Bundle and Channel Power Lim i ts
The limiting values for bundle and channel powers are specified inthe Operating License for each station.
Licensees ensure compliance with these limits by following operatingprocedures, which are based on analyses.
However, current validation of the channel and bundle poweranalyses method is such that the errors associated with theircalculations are not well-defined.
If larger allowances for uncertainties were needed, channel or bundlepower may become more limiting than bulk power.
(source: Generic Safety Issues For Nuclear Power Plants WithPressurized Heavy Water Reactors And Measures For TheirResolution, IAEA, Vienna, 2007, IAEA-TECDOC-1554 chapter AA2).
Specific safety issues to CANDU NPPs (2)
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Operat ion w ith a Flux Ti l t :
The adequacy of Regional Overpower Protection (ROP) or NeutronOverpower Protection (NOP) trips for reactor operation with a fluxtilt is demonstrated by analyses, which take into account differentplant states for which continued operation is permitted.
ROP/NOP system design is based on information derived fromsimulations of certain reference and perturbed flux shapes in thereactor core.
Trip setpoints are established from these simulations to prevent anychannel reaching its critical power limit in case of a bulk loss ofregulation.
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Operat ion w ith a Flux Ti l t :
In order to demonstrate the adequacy of ROP/NOP trip setpoints,licensees should:
determine the maximum tilt permitted by the current operatingprocedures for prolonged operation with a flux tilt, prior to any
operator action, generate a steady state flux distribution, corresponding to the
maximum tilt permitted by the current operating procedures, anddesign-basis and abnormal perturbation flux shapes,corresponding to this steady state shape,
assess simulation ratios (ratios of changes in fluxes andchannel powers due to perturbations) for the above flux shapes,and assess the ROP/NOP trip coverage by determining whetherthe ratios are invariant within any postulated error allowance.
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PHT Pump Operation Under Two -Phase Flow Cond i t ions:
The operation of Primary Heat Transport (PHT) pumps under lowsuction pressure and significant void can be detrimental to theintegrity of the PHT system piping due to the generation of largeamplitude pressure pulsations and excessive pump set vibration.
In the past, the PHT piping fatigue analysis was done using a limitingforcing function (harmonic excitation) obtained from laboratory tests offull-scale PHT pumps.
Given the underlying assumptions, especially the amplitude andfrequency of excitation, this approach was very sensitive tointerpretation of the test data and their application to the PHT system.
Consequently, the assessment of the piping fatigue life may not havebeen conservative.
Specific safety issues to CANDU NPPs (5)
6 C l i
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The safety analyses support provided for the licensing of the two units ofCernavoda NPPwas based mainly on the plant designer and theanalyses of similar CANDU NPPs.
However, over the years the license holder has developed its owncapability of performing accident analyses, using state-of-the-artcomputer codes and appropriate methodologies, according to the bestpractice of CANDU NPPs .
A Safety Analysis Strategic Programme was developed by CernavodaNPP Unit 1 and approved by CNCAN. Up to date a main part of theanalyses from Chapter 15 of FSAR has been performed and a strongteam of Safety Analysis is working at Cernavoda NPP Unit 1.
Best estimate computer codes with conservative assumptions are used
for deterministic safety analyses. As an active member of COG (CANDU Owners Group), Cernavoda NPP
uses the last versions of CANDU specific computer codes and safetyanalysis methodologies and has access to a common data base.
6. Conclusions
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References (excerpt)
Law No. 111/1996 on Safe Deployment of Nuclear Activities
Safety Analysis of CANDU NPP- Regulatory Guide, AECB 1999
C-006, Requirements for the Safety Analysis of CANDU NPPRegulatory
Guide, AECB 1980
C-006 (Rev. 1) (E) Safety Analysis of CANDU Nuclear Power Plants, AECB,
1999
Guidance for Accident Analysis of Commercial NPPSafety Report, IAEA
1999 Regulatory Document RD310, Safety Analysis for Nuclear Power Plants,
2008
The Technology of CANDU Loss of Coolant AnalysisTTR 276, AECL 1991
CSA N286.7Quality Assurance of Analytical, Scientific and Design
Computer Programs for NPP, rev. 4, 1998
Manual on Quality Assurance for Computer Software Related to the Safety ofNPPTRS 282, IAEA 1988
Generic Safety Issues For Nuclear Power Plants With Pressurized Heavy
Water Reactors And Measures For Their Resolution, IAEA, Vienna, 2007,
IAEA-TECDOC-1554