Diversified Reactor Level Measurement System in the Boiling Water Reactor of Olkiluoto

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Energy (2017) 000–000 139–147 EnergyProcedia Procedia127 00 (2017) www.elsevier.com/locate/procedia

International Youth Nuclear Congress 2016, IYNC2016, 24-30 July 2016, Hangzhou, China International Youth Nuclear Congress 2016, IYNC2016, 24-30 July 2016, Hangzhou, China

Diversified Reactor Level Measurement System in the Boiling Diversified Reactor Level Measurement System in the Boiling Water Reactor The 15th International Symposiumof onOlkiluoto District Heating and Cooling Water Reactor of Olkiluoto Pitkänen Janne a,a,the * Assessing the feasibility of using Pitkänen Janne * heat demand-outdoor Teollisuuden Voima safety engineering: Olkiluoto, Finland, temperature function forOyj,aNuclear long-term district heat26100 demand forecast Teollisuuden Voima Oyj, Nuclear safety engineering: Olkiluoto, Finland, 26100 a a

I. Andrića,b,c*, A. Pinaa, P. Ferrãoa, J. Fournierb., B. Lacarrièrec, O. Le Correc

Abstract a Abstract IN+ Center for Innovation, Technology and Policy Research - Instituto Superior Técnico, Av. Rovisco Pais 1, 1049-001 Lisbon, Portugal b Veolia Recherche & reactor Innovation, 291level Avenue Dreyfous Daniel, 78520 FranceWater Reactor (BWR) of This study investigated the feasibility of a new water measurement system forLimay, the Boiling c Département Systèmes Énergétiques et Environnement IMT Atlantique, 4 rue Alfred Kastler, 44300 Nantes,Reactor France (BWR) This study investigated the feasibility of a new reactor water level measurement system for the Boiling Water of the Olkiluoto Nuclear Power Plant (NPP). The operation of this new system is based on a different physical principle than the the Olkiluoto Nuclear Power Plant (NPP). The operation of this new system is based on a different physical principle than the existing ones. It is capable of indicating low or high water levels in the downcomer and of initiating the related safety functions. existing ones. It is capable indicating low or high levels in the downcomer and of relatedtosafety functions. This diversified system is of designed to operate underwater Design Extension Cases (DEC), in initiating which in the addition an Anticipated This diversified system (AOO) is designed to operate under Design Extension (DEC), inlevel which in addition system, to an Anticipated Operational Occurrence a Common Cause Failure (CCF) occurs inCases the preliminary measurement thus when Abstract Occurrence (AOO) a Common Cause Failure (CCF) occurs in the preliminary level measurement system, thus when Operational a conventional system cannot execute its safety function. The target of this study was to analyze the most relevant AOOs, where aDiversified conventional system cannot execute safety function. of this studytowas to analyzestate the1.most relevant AOOs, where Level Measurement Systemits(DLMS) is neededThe for target bringing the NPP a controlled District heating networks are System commonly addressed in the as one most effective Diversified Level Measurement (DLMS) is needed forliterature bringing the NPPofto the a controlled state1. solutions for decreasing the greenhouse gas emissions from the building sector. These systems require high investments which are returned through the heat © 2016 The Authors. Published by Elsevier Ltd. ©sales. 2017 Due The Authors. Published by Elsevier Ltd. to the changed conditions renovation policies, heat demand in the future could decrease, © 2016 The Authors. Publishedclimate by Elsevier Ltd. and building Peer-review under responsibility of the Peer-review under responsibility ofperiod. the organizing organizing committee committee of of IYNC2016. IYNC2016 prolonging the investment return Peer-review under responsibility of the organizing committee of IYNC2016. The main scope of this paper is to assess the feasibility of using the heat demand – outdoor temperature function for heat demand Keywords: Diversified reactor level measurement system; thermal hydraulics; nuclear safety; design extension; common cause failure forecast.Diversified The district of level Alvalade, locatedsystem; in Lisbon used as design a caseextension; study. The district is failure consisted of 665 Keywords: reactor measurement thermal(Portugal), hydraulics; was nuclear safety; common cause buildings that vary in both construction period and typology. Three weather scenarios (low, medium, high) and three district renovation scenarios were developed (shallow, intermediate, deep). To estimate the error, obtained heat demand values were compared with results from a dynamic heat demand model, previously developed and validated by the authors. The results showed that when only weather change is considered, the margin of error could be acceptable for some applications (the error in annual demand was lower than 20% for all weather scenarios considered). However, after introducing renovation scenarios, the error value increased up to 59.5% (depending on the weather and renovation scenarios combination considered). The value of slope coefficient increased on average within the range of 3.8% up to 8% per decade, that corresponds to the decrease in the number of heating hours of 22-139h during the heating season (depending on the combination of weather and renovation scenarios considered). On the other hand, function intercept increased for 7.8-12.7% per decade (depending on the * Corresponding author. Tel.: +358-44-5511123 coupled scenarios). The values suggested could be used to modify the function parameters for the scenarios considered, and * E-mail Corresponding Tel.: +358-44-5511123 address:author. [email protected] improve the accuracy of heat demand estimations. E-mail address: [email protected] In controlled state reactor has been shut down and removal of its decay heat has been secured . ©1 2017 The Authors. Published by Elsevier Ltd. In controlled state reactor has been shut down and removal of its decay heat has been secured . Peer-review under responsibility of the Scientific Committee of The 15th International Symposium on District Heating and Cooling. 1

1876-6102 ©Heat 2016demand; The Authors. Published bychange Elsevier Ltd. Keywords: Forecast; Climate 1876-6102 2016responsibility The Authors. of Published by Elsevier Ltd. of IYNC2016. Peer-review©under the organizing committee Peer-review under responsibility of the organizing committee of IYNC2016.

1876-6102 © 2017 The Authors. Published by Elsevier Ltd. Peer-review under responsibility of the Scientific Committee of The 15th International Symposium on District Heating and Cooling.

1876-6102 © 2017 The Authors. Published by Elsevier Ltd. Peer-review under responsibility of the organizing committee of IYNC2016 10.1016/j.egypro.2017.08.100

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1. Introduction The Olkiluoto 1 and 2 Nuclear Power Plant (OL1/OL2 NPP) in Finland, consists of two identical boiling water reactors (BWR) with nominal thermal power of 2500 MW. The operation started in OL1 in 1978 and in OL2 in 1980, respectively. Both units are operated by Teollisuuden Voima Oyj (TVO), Finland, and were designed and constructed by Asea Atom AB, Sweden. The Reactor Water Level Measurement Systems of OL1/OL2 uses currently differential pressure sensors, which is the most common principle for BWRs in the world. For the safety of a BWR it is essential that the water level is kept continuously above the top of the active fuel, which ensures sufficient cooling of the reactor core. Another constraint is that the water level should remain below the elevation of the main steam line nozzles to prevent any liquid phase water from entering the steam lines. Otherwise, the integrity of the steam line isolation valves, and even that of the turbine, could be jeopardized. Indication of either low (L) or high (H) water level actuates corresponding safety functions; auxiliary feed water flow is started by a low level indication, and water injection into the reactor is stopped on a high level indication, respectively. From the safety point of view, the current level measurement systems in OL1/OL2 units rely on redundancy principle; the system is based on four identical and independent trains in which 2/4-indication of either low or high water level will lead to actuation of safety functions. Applying the diversity principle will improve the reliability of the level measurement even further. This report introduces a planned design of TVO for a diversified reactor level measurement system (DLMS) based on a different physical principle than the present system. The current system is based on an indirect measurement by measuring the pressure difference between the reactor steam dome and downcomer at a certain elevation. The new design will include level switches determined by the water level in the reactor. Validation of the new system is based on both experimental results and thermal hydraulic modeling. Two codes which were utilized in validation calculations were APROS and TRACE. Nomenclature 321-RHRS 326 AOO BE BWR CCF DBC DEC DLMS H2 HP ECCS L2 OL1/OL2 RHRS RT RPV STUK TVO

Residual Heat Removal System at the suction of pipeline 321 Cooling System of Reactor Pressure Vessel Head Anticipated Operational Occurrences Best Estimate Boiling Water Reactor Common Cause Failure Design Basis Conditions Design Extension Cases Diversified Level Measurement System High level (2) High Pressure Emergency Core Coolant System Low level (2) Olkiluoto 1 and 2 nuclear power plant Residual Heat Removal System Reactor Trip Reactor Pressure Vessel Säteilyturvakeskus (Finnish Radiation and Nuclear Safety Authority) Teollisuuden Voima



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2. Design And Purpose of the Systems 2.1. Design of the Current System Current level measurement systems (fine level and coarse level measurements) consist of impulse pipes, a reference pipe, a reference chamber, pressure and temperature measurement equipments, and a density compensation unit. Impulse pipes are connected to the Reactor Pressure Vessel (RPV), one below (minus side) and one above (plus side) the actual water level. The other end of impulse pipe (plus side) is connected to a reference chamber, which in turn is connected to a vertical reference pipe. Water level is proportional to the pressure difference between the impulse pipe (minus side) and reference pipe at certain level below the minus side RPV connection. Pressure drops in these pipes below the minus side connection due to the elevation cancel each other. A scheme of the current level measurement system is shown in Fig 1.

Fig 1. A scheme of the current level measurement system in OL1/OL2 (For simplicity this figure does not show two-phase mixture. Red colors illustrates gas phase and blue color liquid phase)

When the reactor is at full power or at hot standby state, the reactor coolant is under saturated conditions. The saturated steam flows through the reference pipe in plus side (1) to the uninsulated reference chamber, and condenses there. Excess water drains back to the RPV. The impulse pipe (2) is hence continuously filled with water under such conditions. Because the water inside the system is at lower temperature than in the RPV, the density compensation is needed for calculating the actual water level inside the RPV. There are not horizontal parts in the impulse pipes, which causes non-condensable gases to flow back to the RPV.

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2.2. Safety Functions of the Current System Current systems are designed to operate under all the Design Basis Conditions (DBC) of the plant. The level measurements provide information necessary to the operators and to certain automatic functions, to bring the nuclear facility in a controlled state after a postulated accidents. During these events it actuates the corresponding safety features on low or high reactor water levels. The level measurement systems are hence classified to the second highest safety class (which is the highest functional safety class in Finland) according to Finnish Nuclear Safety Authority (STUK) [1]. 2.3. Safety Functions of the New Design The DLMS is designed to operate under so called Design Extension Cases (DEC), which are postulated to occur with a frequency lower than once in 100,000 years. DEC-events are not included in the original design basis of OL1/OL2. In 2013 the Ministry of Employment and the Economy in Finland defined the first category of DECevents (DEC-A) as a Common Cause Failure (CCF) in a system that executes a safety function during an Anticipated Operational Occurrences (AOO) or class 1 postulated accident [2]. An AOO is expected to occur with a frequency more than 10-2 i.e. once in a life time of a single unit, and a class 1 postulated accident with a frequency between 10-2 and 10-3 i.e. once in a lifetime of multiple units, respectively. The DLMS system will be designed to accomplish the diversity principle and to ensure the feasibility of bringing the facility into a controlled state in case of a failure in preliminary reactor water level measurement system. Thus it will be classified to a lower safety class than the preliminary reactor water level measurement systems [1]. Below is a list of events that should be initiated by low (or high) water level, regardless of function of the preliminary reactor water level measurement. Table 1. Initiated safety functions of the DLMS system Trip Signal

Initiated events

L2 (low level (2) indication)

Reactor Trip (RT) Starting injection of High Pressure Emergency Core Coolant System (HP ECCS) / Auxiliary Feed Water System

H2 (high level (2) indication)

Reactor Trip (RT) Closure of Feed Water Isolation Valves (FWIVs) and Main Steam Isolation Valves (MSIVs)

2.4. First Application of a Diversified System The first plan was to install a DLMS to a pipeline of the Residual Heat Removal System (321 - RHRS). The suction of 321 pipeline connects to the RPV at an elevation below the normal reactor water level. The upper and lower ends of the DLMS pipes are connected to different elevations in 321 line as shown in Fig 2a. A float level switch / float house (Fig 2b) is located at a certain elevation in the DLMS, which corresponds to a desired low level indication (L2). The switch gives a low level indication if the water level decreases under the limit.

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a

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b

Fig 2. (a) Configuration of the first design of a DLMS; (b) A schematic picture of a float level switch

One can calculate the water level in a diversified system indirectly according to the pressure difference between the inlet and outlet of the system. The pressure difference is expressed as follows:

p  p fric  ploc  pacc  pE

(1)

Frictional pressure losses Δpfric are caused due to the wall friction, and the local pressure losses Δploc are due to the area changes close to the level switch. Acceleration pressure losses Δpacc indicate the pressure losses due to changes in the fluid momentum. Elevation pressure drop Δp E can be expressed by a liquid level h and the total height H between the inlet and outlet of the system, as following:

pE   L gh  G g ( H  h)

(2)

The sub cooling margin of water in RPV, close to the system 321 connections, is almost zero. Furthermore, there would have been continuous water flow through the DLMS pipelines due to a constant mass flow rate of the system 321, when the reactor is under operation. The level switches were shown to give false trip signals caused by water flashing inside the impulse pipes, especially during start-ups and shutdowns of the reactor. The design was not approved and thus no further investigations were made with this configuration. 2.5. Second Application of a Diversified System A new plan was created, which specify to connect the bottom end of the impulse pipes to the system 321 line and the top end to the RPV head cooling system (326). System 326 is used only in reactor shut down states, and it is connected to the steam dome of the RPV. During full power operation system 326 is filled with saturated steam. The configuration of second version of the system is shown in Fig 3.

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Fig 3. Second version of a DLMS design

The second version consists of two separate level indication units: high level indication H2 (upper level switches), and low level indication L2 (lower level switches). In this case there is not continuous flow rate of reactor water trough the DLMS system. Only a small liquid flow rate through the float level switches will occur due to the steam condensing inside the uninsulated impulse pipes. The sub cooling margin of water can be therefore higher compared to the original design, and hence the flashing effect will be avoided. In steady state conditions frictional, local and acceleration pressure losses can be ignored, and the total pressure loss can be described again by the equation 2. The density of water inside an insulated pipe is higher compared to the density of saturated water inside the RPV at operating pressure. Therefore an elevation of water level in the DLMS system is lower than the corresponding water level in RPV. 3. Feasibility studies The DLMS system was validated by the results of the feasibility studies and the experimental tests. 3.1. Analyzed cases The main target of the feasibility studies is to ensure that the level switches will give actuation signals on both high and low water levels in RPV in adequate timing even if the normal level measurement system is out of the service. In addition the DLMS system should not actuate any false trip signals. The most relevant AOO events (with CCF in the primary reactor water level measurement system) for the DLMS in OL1/OL2 are:  Turbine trip with dump blocking  Fault in reactor water level control (max set value)  Total loss of feed water In the licensing analyses these cases were analyzed by Westinghouse Sweden AB using the BISON code with minor conservatives in the input data and calculation methodology [3]. These analyses were recalculated by TVO with best estimate assumptions with the system code APROS [4]. Furthermore, Vattenfall Sweden used the results of the APROS calculation as boundary conditions same conditions for a detailed TRACE model of the DLMS [5]. The TRACE calculations were used to verify that the DMSL design fulfills the design criteria regarding its functional behavior. In addition an inadvertent failure of a 321-pump was investigated to ensure that the DLMS will not initiate any false signals in case of flow rate ceasing in the water inlet of the DLMS.



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3.2. Results The thermal hydraulic modeling with APROS gave good comparison results against licensing analyses of OL1/OL2 in all analyzed cases. Figure 4 shows an example of the results in case total loss of feed water, which was a bounding case for low level (L2D) indication. a

b

Fig 4. Example of results in total loss of feed water: neutron power (APRM), feed water flow rate, and reactor water levels in preliminary level measurement systems; (a) conservative case (APROS); (b) best estimate case (APROS)

In the example above the water level decreases after the feed water flow ceases. This leads also to lower subcooling of the core inlet, as less cold feed water is mixed with hot recirculated water from the steam separators and steam dryer. In addition recirculation pump runback starts after loss of feed water. These events cause an increase in core void fraction, which in turn cause reactor power and pressure to decrease. After recirculation pumps have reached their minimum speed, the reactor power increases slightly until reactor scram is initiated by low level (L2) indication. One can see that the sequence of events is slower in best estimate case than in conservative case (Fig. 4a). This is due to a slightly different assumptions in reactor initial state, in delay times of automation signals, and in driving times of valves etc. The best estimate results were further used in detailed analyses of the DLMS system. Figure 5 shows an example of the behavior of void fraction in a DLMS water connection pipe (321-pipe), and water level in the DLMS system during total loss of feed water -transient.

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a

b

Fig 5. (a) Void fractions in the connection points of the DLMS system and system 321 after total loss of feed water accident (APROS BE calculations) [4]; (b) Level in the DLMS system after total loss of feed water accident (APROS and TRACE calculations) [5]

Comparison between figures 5a and 5b shows that during the transient, after the void fraction in the DMLS inlet connection increases, the level inside the DLMS system increases momentarily. The maximum level is still not enough to initiate false H2D signal. The water inside the DLMS remains in liquid phase, and the level continues decreasing after the short increase. The DLMS indicates L2 level few seconds before the void fraction in 321 pipeline increases to 1. One can notice that figure 5 shows only effect of inlet void fraction to level of DLMS. In addition effects of pressures, temperatures, densities, and mass flow rates were studied as key parameters for every selected case. Figure 5b also shows that the absolute level inside the DLMS differs from level measurement from the RPV. This is mainly caused due to the different densities of the liquid water inside RPV and DLMS. According to the results shown in Fig 5 the timing of the diversified low level trip signals (L2D) are in consistent with the L2 signals of current level measurement system. The other analyzed AOO events gave similar results, and the results from inadvertent failure in a 321-pump showed that the DLMS does not initiate any false signals during that transient.



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3.3. Future Works Few low pressure experimental tests were made in Lappeenranta University of Technology (LUT) with a test rig to prove the operation of level switches with decreasing and increasing water level inside the RPV. Also the temperature profiles inside the DLMS were investigated in these tests. The results of these experimental tests will be confirmed with CFD simulations. These will also provide further information of the best location of temperature measurements in the test facility. This information could be used in further experiments with high pressure tests. An effect of non-condensable gases inside the DLMS pipelines and float houses will be investigated with separate studies. 4. Conclusion The RPV level measurement is one of the most important safety related measurements of the BWRs. According to the key results of references [4] and [5] the DLMS system located between RHRS (321) and RPV head cooling system (326) is feasible to indicate L2 and H2 levels of the OL1/OL2 RPV in case of CCF in preliminary water level measurement system. The results are valid for the most relevant AOO events where indication of high/low level is needed. The DLMS is totally independent of the preliminary water level measurement and it hence fulfils the diversity principle. The DLMS lowers the risk of missing or wrong high/low level indications of the RPV and therefore increases the safety of the OL1/OL2 plant. References [1] STUK, "YVL B.3, Classification of systems, structures and components of a nuclear facility", Stuklex, Helsinki, (2013) [2] Ministry of Employment and the Economy, Finland, "Government Degree on the Safety of Nuclear Power Plants 717/2013", Finlex, Helsinki, (2013) [3] M. Lemmetty, L. Saarelainen, "FSAR/9.3 – Anticipated Operational Transients", TVO, Olkiluoto, (2014) [4] J. Pitkänen, "Boundary Conditions for Diversified Reactor Pressure Vessel Level Measurement System - Apros Calculations", TVO. document number: 161052, (2015) [5] J. Bertilsson, D. Palko, L. Facciolo, F. Lillienwall, "QP.50018.004-68236796 Feasibility Report", Vattenfall, (2015)