# Deterministic Analysis of Natural Circulation Events at  the Ignalina NPP

Authors: Algirdas Kaliatka; Eugenijus Uspuras; Virginijus Vileiniskis
Journal: Science and Technology of Nuclear Installations
 (2008)

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## Abstract

Eight main circulation pumps (MCPs) are employed for the cooling of water forced circulation through the RBMK-1500 reactor at the Ignalina nuclear power plant (NPP). These pumps are joined into groups of four pumps each (three for normal operation and one on standby).
In the case of all MCPs trip, the reactor shutdown system is activated due to decrease of coolant flow rate. At the same time, after the pump trip, the coolant to the reactor fuel channels during the first few seconds is supplied by pump coastdown. Later, the reactor is cooled by natural circulation. The main question arises whether this coolant flow rate is sufficient to remove the decay heat from the reactor core.
This paper presents the investigation of all MCPs trip events at the Ignalina NPP by employing best estimate code RELAP5 and methodology of uncertainty and sensitivity analysis.


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## Body

## 1. Introduction

The Ignalina nuclear
power plant is a twin unit with two RBMK-1500, graphite moderated, boiling
water, and multichannel reactors. Unit 1 was permanently shutdown in 2004. A schematic representation of one main
circulation circuit (MCC) loop is given in Figure1. The MCC is divided into two halves: the
left and the right
loops. The MCPs (5) are joined in groups of four pumps each (three for normal
operation and one on standby). The MCPs feed a common pressure header (PH) (8) on each side of
the reactor. Each pressure header provides coolant to 20 group distribution
headers (GDH) (9), each of which in turn feeds from 38 to 43 fuel channels (11).
The coolant flow rate
through the individual
fuel channel (FC) is regulated by isolation and control valves (ICV) (10), which are mounted in the lower
water communication lines. Coolant passing through FCs is boiled and part of
the water is evaporated. Steam-water mixture through steam-water communication
lines (12) flows to drum separators (DS). The steam, which is separated in the DS, is supplied to turbines through steam lines  (13). A
detailed description of the Ignalina NPP can be found in [1].Figure 1
Schematic representation of one loop of the RBMK-1500 MCC: 1-DS, 2-downcomers, 3-MCPs suction header, 4-MCP suction piping, 5-MCPs, 6-MCP discharge piping, 7-bypass line, 8-MCPs pressure header, 9-GDHs, 10-lower water communication line, 11-FC, 12-steam-water communication line, 13-steam lines.At the Ignalina NPP type CVN-8, MCPs are employed for the forced
circulation of cooling water through the RBMK-1500 reactor. These pumps belong
to the “wet” stator pump group. The CVN-8 type is a centrifugal, vertical,
single-stage pump with a sealed shaft. To increase the rotary inertia in order
to prolong the rotation of the shaft in the event the electric motor fails, the
massive 0.2 m outside diameter and the 0.195 m thick steel flywheel are mounted on the motor shaft. In the all-pumps-trip case, the coolant, due to high inertia of pump
flywheel during the first 40–60 seconds to the
reactor fuel channels, is supplied by pumps coastdown. Later, the natural
circulation through the core is established. The
reactor shutdown system is activated due to the decrease of coolant flow rate.
All 211 rods of the control and
protection system (CPS) are inserted into the reactor core within approximately
8 seconds. The main question that arises is whether this coolant flow rate is
sufficient to remove the decay heat from the reactor  core.Best estimate model of
RELAP5 Ignalina NPP RBMK-1500 reactor cooling circuit was developed by
Lithuanian Energy Institute (LEI) for the investigation of MCP trip events.
This model includes forced circulation circuit, steam lines, and safety systems
necessary for transient and accident process analyses. Detailed description of
RELAP5 nodalization scheme can be found in [2]. The obtained results were compared with the measurements of Ignalina  NPP.Initial and boundary
conditions (coolant pressure, flow rate, feed water temperature, amount of
steam for in-house needs, reactor power, and flow energy loss in different MCC
components) and RELAP5 code models, assumptions, and correlations may impact
the uncertainty of calculation results. The GRS method SUSA 3.5 [3] was used for the
sensitivity and uncertainty analysis. The parameters, the initial values of which may have the
greatest impact on the simulation results, were used for the analysis on the
basis of earlier performed benchmarking calculations. The selection of these
parameters is described in [4, 5].


## 2. Best-Estimate Analysis of Loss-of-All-MCPs Event

Loss-of-all-MCPs event occurred at Ignalina NPP on March 26, 1986.
During this event, all the six operating MCPs at Ignalina Unit 1 were tripped
simultaneously. Before this event, the reactor operated at thermal power level
of 4650 mW. In response to multiple pump trips, an emergency protection signal
AZ-1 was generated and the reactor was shut down. The MCC flow decreased in
response to the MCPs coastdown. Long-term flow was due to natural circulation
in the MCC.RELAP5 analysis results were compared with the plant data. The
uncertainty and sensitivity analysis was performed using a two-sided tolerance
limit (with 0.95 of probability and 0.95 of confidence); 100 runs were performed.The agreement of
the calculation results, obtained using RELAP5 code with the real plant data,
was evaluated using the adequacy standard, presented in the Guideline for
performing code validation, and issued by DOE International Nuclear Safety
Center [6]. The agreement is judged to be excellent, when the code exhibits no
deficiencies in the modelling of a given behaviour; major and minor phenomena
and trends are predicted correctly; calculation results are judged to agree
closely with the real plant data. The agreement is judged to be reasonable when
the code exhibits minor deficiencies, although it provides an acceptable
prediction; all major trends and phenomena are correctly  predicted, but
differences between the calculation and data are greater than deemed acceptable
for excellent agreement. According to the standard, both excellent and
reasonable agreements of the calculation results and the real plant data are
considered as being acceptable.Data of the flow rate in twelve individual channels during natural circulation
regime were available and are presented in Figure2. For the comparison of
calculated and measured flow rates, calculated maximum flow rate through
maximum loaded FC and calculated minimum flow rate through minimum loaded FC
were used. These calculated flow rates represent the whole interval (from
minimal up to maximal) of flow rates through the core channels. The calculated
maximum flow rate through maximum loaded FC, the calculated minimum flow rate through minimum
loaded FC, and real plant data showed reasonable agreement. Coolant flow rate
through one MCC loop is presented in Figure 3. As it is seen from the figure,
measured values are available only in a limited range, and become equal to 0
after approximately 115 seconds from the beginning of the accident. The
comparison of measured coolant flow rate through the fuel channels and through
one MCC loop at the flow <7000 m3/h shows that the last measures
are unreliable. This is because the throttling devices, which are not designed
for the measurement of low coolant, are employed. The flow rates show a
coastdown associated with the loss of forced circulation by the MCPs. The
coastdown continues during the first 40 seconds from the beginning of the
transient. Later, a natural circulation of the coolant was established at a flow
rate equal to approximately 15% of the initial flow.Figure 2
Loss-of-all-MCPs event. Coolant flow rate through individual channels.Figure 3
Loss-of-all-MCPs event. Coolant flow rate through one MCC loop.After reactor shutdown, the reactor is reliably cooled by natural
circulation because heat flux decreases faster than coolant flow rate due to
pumps coastdown (see Figure4).Figure 4
Loss-of-all-MCPs event. Comparison of behaviour of coolant flow rate through one
MCC loop and heat flux in maximum loaded fuel channel.Figure5 shows pressure in the MCC. Pressure in the pressure header
decreases immediately after MCP trip because the MCP head is decreasing. During
the first seconds, before the reactor shutdown, the coolant flow rate decrease
through the core causes the short-term increase of steam generation. The
increase of steam generation causes the short-term pressure increase in the DS.
Steam generation in the core decreases and pressure in the MCC also decreases
after the reactor shutdown. The reloading process of turbines is starting
immediately after the reactor shutdown. When the turbine control valves are
closed, the pressure starts to increase again. Furthest pressure changes depend
on the amount of removed steam for in-house needs. As could be seen from the
presented comparison, pressure losses in different parts of the MCC are
predicted correctly using the developed model.Loss-of-all-MCPs event. Pressure in
the MCC.
(a)(b)(c)Uncertainty and
sensitivity analyses were performed for the following.(i)
Coolant
flow rate through one MCC loop—one of the
important technological parameters which allows identifying the existence of
natural circulation.(ii)
Calculated critical heat flux ratio (CHFR) from
the side of fuel assembly to coolant. CHFR is defined as a relationship
between the calculated critical and real heat transfer fluxes. If critical heat
flux ratio is greater than one, no critical heat flux will be observed in any
fuel channel segment, and drying of FC will not occur.
Parameters, which
may impact the calculation results, are presented in Table 1. Selected RELAP5 code parameters are
varied in the area where two-phase flow conditions might occur: in the vertical
section before the heated channels, in the heated channels, above the heated
channels, and in the steam water communications. The areas with single-phase
conditions are excluded due to the fact that these parameters do not have
impact on the results in such region. Additionally, one parameter, which might
impact the coolant flow regime in the reactor fuel channels, was selected—the 13th position of ICV, what
affects coolant flow rate through fuel channels.Table 1
Parameters which may impact the uncertainty
of calculation results in case of all MCPs trip.
No.ParameterValue of   theExplanationWidth of parameter distributionparameter inMean deviationProbabilityMin.Max.the basic caseand error[%]distributioncalculationInitial conditionsPar. 1Pressure in6.35⋅1067.035·1066.70·1061.675·105[5%]NormalMeasurementDS, PaerrorPar. 2Coolant flow rate79388262810081 [2%]Normalthrough singleMeasurementMCP,  m3/herrorPar. 3Feed water458.52467.78463.152.32 [1%]NormalMeasurementtemperature,KerrorPar. 4Coefficients0.981.021.00.01 [2%]Normalwhich defineMeasurementsteam flow rate toerrorthe in-houseneeds
headerPar. 5Reactor thermal4.510·1094.790·1094.650·1097.0·107[3%]NormalMeasurementpower,WerrorModelling parameters in
the corePar. 6Water packing0 (on)1 (off)1 (off)—NonparametricModelassumptionPar. 7Stratification0 (on)1 (off)0 (on)—NonparametricModelassumptionPar. 8PV term0 (off)10 (off)—NonparametricModelassumptionPar. 9CCFL0 (off)1 (on)0 (off)—NonparametricModelassumptionPar. 10Thermal front0 (off)1 (on)0 (off)—NonparametricModeltrackingassumptionPar. 11Mixture level0 (off)1 (on)1 (on)—NonparametricModeltrackingassumptionPar. 12Nonhomogeneous0 (nonhomogeneous1 (homogeneous0  (nonhomogeneous—NonparametricModelmodel)model)model)assumptionFlow energy lossPar. 13Coefficients0.91.11.00.05 [10%]NormalModelwhich define ICVpositionassumptionResults of the
sensitivity analysis are presented using plots with the parameters impact on
the results. Higher absolute value of impact on the results means higher
parameter influence on the result. The positive impact means that when the
higher value of the parameter is selected, the higher coolant flow rate is
obtained; the negative impact means that when the higher value of the parameter is selected,
the lower coolant flow rate in the affected MCC loop is obtained.The performed
analysis shows that the selection of homogeneous or nonhomogeneous models (see Figures6, 7, 9, and 10) has the largest impact on the calculation results. Homogeneous model
selection is a nonphysical conservative assumption and it is not recommended
for best estimate codes.Figure 6
Loss-of-all-MCPs event. The impact of parameters no.1–7 on the coolant flow
rate through one MCC loop. The numbering of parameters is the same as in Table1.Figure 7
Loss-of-all-MCPs event. The impact of parameters no.8–13 on the coolant flow
rate through one MCC loop. The numbering of parameters is the same as in Table1.The
parameter-dependent sensitivity analysis shows that initial plant conditions
(coolant flow rate, pressure in the DS, feed water temperature, and reactor
power) and ICV position have only insignificant influence on coolant flow rate
through the pumps (natural circulation regime) (see Figures6 and 7).Reactor core is reliably cooled because the
CHFR is greater than 1 in all 100 calculations (see Figure10).The
parameter-dependent sensitivity analysis (see Figures8 and 9) shows that reactor
thermal power (Par. 5) and pressure in DS (Par. 1) have the greatest impact on
the reactor cooling conditions if the selection of homogeneous or nonhomogeneous
models is not taken
into account.Figure 8
Loss-of-all-MCPs event. The impact of parameters no.1–7 on the CHFR. The
numbering of parameters is the same as in Table1.Figure 9
Loss-of-all-MCPs event. The impact of parameters no.8–13 on the CHFR. The
numbering of parameters is the same as in Table1.Loss-of-all-MCPs event. CHFR in case of all MCPs trip.
(a)(b)Thus, the performed uncertainty and sensitivity
analysis shows that in presented all MCPs trip case, reactor core is reliably
cooled due to natural circulation regime.


## 3. Best-Estimate Analysis of Sequential MCP Trip

A three-MCPs
sequential trip event occurred at the Ignalina NPP on August 23, 2000. Before
this event, the reactor operated at 2300 mW thermal power level. At11:17, due
to the short circuit into   the control cable, fire-prevention signal of Ignalina
NPP was activated by mistake. This caused the fire-prevention pump to provide a
foam mixture into the MCPs compartments of one MCC loop. The foam was found on
the cabinets of MCP electric motors control. The short circuit protections were
activated. At 11:23, the first MCP was switched off. As the core thermal power
was less than 2860 mW, AZ-4 signal was not generated. Still after three
minutes, the second MCP of the same MCC loop was switched off. According to two
MCPs trip in one loop of the MCC, AZ-1 signal was generated. According to this
signal, all CPS rods were inserted within 12–14 seconds, and
the reactor was shut down. Approximately 20 seconds after AZ-1 initiation, the
steam supply for turbine was suspended. At 11:29, the last operating MCP in the
affected loop of the MCC was switched off. In order to decrease flow rate
differences in both loops, operators stopped one pump in intact loop of the
MCC.Results of the analysis are presented in
Figures11–14. The
calculated and measured pressure in the DS, SH, and PH agrees well (see Figure 11). Switching off of one MCP leads to the increase of coolant flow rate
through other MCPs of the same loop due to the lower pressure drop in the
downstream system caused by cutting off the flow rate of a single pump (see Figure 12). The second MCP of the affected MCC loop was switched off after
approximately 180 seconds. Output of the only operating pump increases up
to 10500 m3/h. Approximately 400 seconds after the first
MCP was tripped, the last MCP was switched off. The calculated maximum-minimum values of
pressures and coolant flow rates are in reasonable agreement with real plant
data (Figures 11–13).Three MCPs sequential trip. Pressure in the MCC.
(a)(b)(c)Figure 12
Three MCPs sequential trip. MCPs throughput in the affected MCC loop.Figure 13
Three MCPs sequential trip. MCPs throughput in the intact MCC loop.Three MCPs sequential trip. Comparison of behaviour of coolant flow rate through
one MCC loop and heat flux in average loaded fuel channel.
(a)(b)Calculations showed that after MCP trip, the
coolant flow rate through it decreases smoothly due to high inertia of
flywheel, pump, and motor rotors. In approximately 60 seconds after the last
MCP trip, the coolant natural circulation starts in the affected loop of the
MCC (see Figure12). It is necessary to note that coolant flow rate through the
first two tripped pumps is also re-established in natural circulation node.
Unfortunately, due to insensibility of measuring devices to low coolant flow
rates at the Ignalina NPP, the coolant natural circulation was not identified.
Coolant flow rate through MCPs of intact loop is presented in Figure 13. In the
three-MCPs
sequential trip in one MCC loop event, the reactor core is also reliably cooled by natural circulation
because heat flux in the core decreases faster than coolant flow rate through
the MCPs (see Figure 14), and CHFR is
greater than 1 (see Figure 15).Three MCPs sequential trip. Critical heat flux ratio.
(a)(b)In this case,
only coolant flow rate through the affected MCC loop was selected for the
uncertainty and sensitivity analysis because CHFR in base case calculations (see
Figure15) is greater than in all MCPs trip case (see Figure 10).Parameters, which
may impact the calculation results, are presented in Table2. The analysis
presented in the previous chapter shows that the selection of the homogeneous
model has the biggest impact on the calculation results. However, the
homogeneous model is not proper for two-phase flow modelling; it presents too
conservative results. Thus, only the nonhomogeneous model was used in the
accident analysis described in this chapter.Table 2
Parameters which may impact the uncertainty of calculation results in
sequential MCPs trip case.
No.ParameterValue of   theExplanationWidth of parameter distributionparameter inMean deviationProbabilityMin.Max.the basic caseand error[%]distributioncalculationInitial conditionsPar. 1Pressure in6.517·1067.203·1066.86·1061.715·105[5%]NormalMeasurementDS, PaerrorPar. 2Coolant flow rate71547446730073 [2%]NormalMeasurementthrough singleerrorMCP, m3/hPar. 3Feed water458.52467.78463.152.32 [1%]NormalMeasurementtemperature,KerrorPar. 4Steam flow rate to81.3484.66830.83 [2%]NormalMeasurementthe in-houseerrorneeds header, kg/sPar. 5Reactor thermal2.247·1092.387·1092.317·1093.48·107[3%]NormalMeasurementpower,WerrorModelling parameters in
the corePar. 6Water packing0 (on)1 (off)1 (off)—NonparametricModelassumptionPar. 7Stratification0 (on)1 (off)0 (on)—NonparametricModelassumptionPar. 8PV term0 (off)10 (off)—NonparametricModelassumptionPar. 9CCFL0 (off)1 (on)0 (off)—NonparametricModelassumptionPar.10Thermal front0 (off)1 (on)0 (off)—NonparametricModeltrackingassumptionPar.11Mixture level0 (off)1 (on)1 (on)—NonparametricModeltrackingassumptionThe performed
analysis shows that the initial coolant flow rate through MCP (Par. 2) has the
largest positive impact on the calculation results,
whereas pressure in DS (Par. 1) has the largest
negative impact. Parameter number 11—mixture level
tracking model usage (see Figures16 and17)—has the
largest influence in the core. Mixture level tracking
model has negative
impact on the coolant flow rate in the affected MCC loop, that is, at the switching off of this model, calculated values of
flow rate will be lower.Figure 16
Three MCPs sequential trip. The impact of parameters No.1–No.6 on the coolant
flow rate through the affected MCC loop. The numbering of parameters is the same as in
Table2.Figure 17
Three MCPs sequential trip. The impact of parameters No.7–No.11 on the coolant
flow rate through the affected MCC loop. The numbering of parameters is the same as in
Table2.


## 4. Conclusions

The uncertainty and
sensitivity analysis was performed for the simultaneous trip of all MCPs and
sequential trip of three MCPs events.The performed
analysis of all MCPs trip shows that the selection ofthe homogeneous model has the biggest
impact on the calculated flow rate through one MCC loop and CHFR. Homogeneous
model selection is a nonphysical conservative assumption and it is not recommended
for best estimate codes. Calculation results also show that reactor
thermal power and pressure in DS have the biggest impact on the reactor cooling conditions.In the sequential
MCPs trip case, initial coolant flow rate through MCP, pressure in DS, and
selection of mixture level tracking model have the biggest impact on the
reactor cooling conditions. Switch off of the mixture level tracking model
decreases the calculated values of flow rate.Performed benchmark
analysis of MCPs trip events showed that the initial reactor thermal
power has insignificant influence on cooling conditions during natural
circulation regime. Even at an
initial maximal thermal power level (loss-of-all-MCPs case in Ignalina NPP), the reactor core is reliably cooled with 0.95 of probability and 0.95 confidence level.



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*Source: 672519-2008-02-20.xml*