authors: laurentiu aioanei dumitru dobrea karlsruhe november 21, 2012 inr pitesti

Authors: Laurentiu Authors: Laurentiu Aioanei Aioanei Dumitru Dumitru DobreaDobrea

Karlsruhe November 21, 2012Karlsruhe November 21, 2012

Authors: Laurentiu Authors: Laurentiu Aioanei Aioanei Dumitru Dumitru DobreaDobrea

Karlsruhe November 21, 2012Karlsruhe November 21, 2012

INR Pitesti INR Pitesti

C&I Introduction

In the present project stage locations for core instrumentation are not comprised. Thus instrumentation specifications will emphasize possible locations with suitable instrumentation, and, where needed, support equipment.

Where more detailed neutronic analysis is not available, qualitative or semi-quantitative evaluations are presented in order to emphasize alternatives with more detailed specifications.

In next project stages, detailed neutronic design could help selecting an alternative. The technical and design requirements subsection addresses also next project stages.

Accuracy, rangeability (accuracy over a wide range), response time, and reliability criteria allowed selecting measurement methods and instrumentation.

The core instrumentation is represented by neutron flux measurement instrumentation.

INR INR Pitesti Pitesti


2 General Description

The main functional requirements based on DEL006 (2011, Task 4.2) and DEL014 (2011, Task 4.4) are summarized as below: 1. To monitor low-level neutron flux and its rate of change during the

reactor start-up, and to provide trip signal for the protection system in case of high flux and high change rate (or small period) regard low-flux conditions.

2. To measure the neutron flux and its change rate, after automatic control becomes effective, at any representative position (in-core or ex-core), and to provide a trip signal in case of high flux, high rate of flux change (or log rate) low-power, and power reactor operations

3. To measure in-core flux at a representative set of locations in order to provide core power distribution information.

4. To monitor reactor flux at shutdown. The automatic control is performed by moving the control rods.


Reactor Operation Steps The approach to criticality is first phase (start-up, during core reloading, after long

shutdowns (cold start-up) and short shutdown (hot start-up)) Attaining the criticality and raising power until the automatic control becomes

effective An estimation of the spontaneous fissions source for a fresh FA is 4·106 n/s that could

produce a flux of the order of 1 n/(cm2·s) near the reactor vessel (more accurate calculations should confirm such values).

If the computed flux value is less than 0.1 n/(cm2·s), a neutron source is necessary during first approach to criticality placed:

in a special location in the core within a dummy location removed during last steps or after reaching criticality

Approach to criticality is also met: during core reloading after long shutdowns (cold start-up )

needs a special instrumentation including a high-sensitivity neutron detector before automatic control becomes effective, the high flux trip signal setpoints are

established based upon the current flux value) after a short shutdown (hot start-up).

During start-up a low signal value will indicate that the equipment (detector or the electronics) failed to follow the neutron flux; hence the channel will be tripped

Raising power until the automatic control becomes effective The automatic control is performed by moving the control rods. The above requirements are satisfied by ex-core, near-core and in-core detectors and

the associated electronics.


Signal Modes In pulse mode:

The signal should range up to 105 cps, to avoid pulse pile-up and count loss, To cover the last decades a fast pulse amplifier is required. The lower limit in pulse mode is usually 0.1 cps to avoid long time intervals to achieve

good statistics. In current mode:

a range of at least 3 flux decades should be covered, with linearity of signal vs. flux over that range ensured.

The upper limit of signal range should be greater than the signal value corresponding to the flux value at nominal power in the detector location.

Currents less than 1 nA could be measured by pico-ampermeters.

During the start-up an in-core or near-core high-sensitivity detector may be withdrawn several times from flux in order to decrease the output signal (pulse mode).

Without withdrawal, ex-core detectors should be used to continue flux monitoring, and the in-core detectors should be removed from the core (sensitive FC might be damaged by a high flux or a background signal might build up at low fluxes).

High sensitivity fission chambers can cover the entire flux range from start-up to full power by working in pulse, fluctuations (Campbelling) and current regimes.


Technical And Design RequirementsEx-core Detectors:

Boron coated ion chambers work only in current regime may cover maximum 6 decades of flux if shielded over the last two decades in-core detectors at cold start-up

high sensitivity fission chambers If is desensitized by104 vs. flux cover a flux range (at their location) between

0.1 and 1011 n.cm2.s-1

6 decades in pulse regimes 5 decades in fluctuation regimes 4 decades in current regime 1-2 decades overlap between regimes CFUC06 form Ref. [1] the sensitivities estimated for ALFRED core spectra

are around 100 times smaller than for thermal neutrons.


Ex-core Detectors (cont) The detectors may be placed:

out-of-vessel, needing locations in shielding (like LMFRs), FC was placed in a channel located in the lateral shielding another was placed in a lateral channel, near an intermediate heat exchanger

(PHENIX) Other FCs were located in a movable device in the central channel of the SPX core

behind and around the intermediate heat exchangers of that reactor [4]

between inner vessel and reactor vessel using: dedicated assemblies ensuring protection of detector cables for shocks and undesired movements (axial displacement of detector/cables

may be allowed).

out of core FAs, inside inner vessel, and will be called near-core detectors

dummy bundle assemblies, with dummy bundles replaced by detector, cables, and detector/cable supports/enclosures, could be used to insert detector assemblies


Ex-core Detectors (cont) We may assume that at an ex-vessel location distort its time-dependent response during transients In order to obtain required sensitivities (signal to average core neutron flux) the detector could be raised at various

elevations It is possible that fuel assemblies distort the dynamic response at high elevations so can use detectors with lower

sensitivity, placed at smaller elevations above the core. Placing ex-core detectors as far as possible of the core allows the detector signal to reflect the dynamic behavior

of the core as a whole

Possible ex-core detector placement and configuration:

a)-vertical section b)-horizontal section


Ex-core Detectors (cont) Protection against corrosion of in-vessel detectors, cabling and

abrasion in lead should be ensured by:

corrosion/abrasion resistant sheathing coating the outer surfaces of detectors and cables corrosion/abrasion resistant enclosures.

Ex-vessel detectors are not affected by corrosion. Cable lining, avoiding impact on fuel handling, connection boxes, and

penetrations should be considered into detailed design regarding installation of in-vessel detectors.

Cable insulation should withstand high temperature, thermal stress, and long irradiation effects.

electronics ensuring signal processing in pulse, fluctuations, and current regimes, switching automatically between regimes, with adequate hysteresis.

Electronic has low voltage power supply , preamplifiers, high voltage power supply for the detector, charge sensitive preamplifiers for FCs, pulse amplifier/SCA, counter/timer, log N module and log N-rate module


In-core and CLOSE-TO-CORE detectors

7 Detector Assembly

Safety Rods

In-core Detector (3 elevations)

Fuel Assembly

Control Rods

Dumy Element

Close-to-core Detector (top to bottom)

Possible in-core/close-to-core detector configuration

Measurements of flux or fission power distributions are significant at nominal power and over two decades below nominal power

Local power effects are not inherent for the ALFRED core, because the core is strongly coupled (Xenon effects are not important in fast spectra)

Distortions of the power distribution could appear incidentally, as emphasized in [3], due to control rod inadvertent movements, fuel assembly loaded in an incorrect position, and fuel assembly with incorrect composition


In-core and close-to-core detectors cont.

Generally, in-core neutron detectors could be used : For safety and control around nominal power the prompt/delayed gammas from fission dominate the delayed gammas of

the previously accumulated fission products. The gammas from fission will follow the reactor power dynamics, thus

linearity of the detector response vs. reactor power is not affected by the external gamma field

But gamma spatial distribution differs from the neutron spatial distribution so that can affect linearity at low power and hot start-up

prompt response of the detectors is not a strict requirement but enough fast for detecting signal variation of a rod drop.

Power computation based on in-core detector signal is based upon detector calibration at steady state by comparison with the power obtained from thermal balance around full power by weighting sums of their signals

This power value could be used directly by automatic reactor regulating systems. A weighted sum of the power computed as above, the power computed based on core flow and outlet-inlet temperature difference, and the power computed based on SG thermal balance could be used by the power computation unit of the overall control.


In-core and close-to-core detectors cont.

the influence of gamma field on in-core signal should be minimized. Such detectors could be miniature fission chambers or delayed response SPNDs (Ag, V, Rh, [1], [2]). For flux map one could use Rh or Hf SPNDs, as flux maps are performed at steady-state.

Miniature fission chambers are also prompt detectors. Prompt SPNDs (Co, Pt) are affected by gamma field, while miniature fission chambers working in pulse or fluctuation (Campbelling) modes could easily reject gamma contribution.

Prompt SPND detectors sensitivities per unit length are lower than delayed SPNDs sensitivities. For SPNDs having usual dimensions the signal in ALFRED will be less than 100 nA [2].

The SPND detector sensitivity could be increased using longer ones (coiling SPND detector, one may obtain 100 times increase of the sensitivity). ([1] around 1A).

However, for ALFRED coiling requires small coil radii, which may affect sheath and insulation.

Increasing detector length will decrease spatial resolution.


in-core and close-to-core detectors cont.

around nominal power the prompt gamma contribution dominates, still being proportional to fission power.

over at least a decade around the nominal power, a local change of fission rate will be sensed by prompt SPNDs better than ex-core detectors that respond mostly to global changes of the fission rate.

Prompt SPNDs could be calibrated vs. reactor thermal power at power levels where thermal balance provides accurate values, and used for thermal power measurements.

the relative sensitivities of the miniature FCs will depend on location due to neutron spectrum variations over the core (10-16 A/(n.cm2.s-1) in current mode, and 10-3 cps/(n.cm2.s-1) in pulse mode, 7 mm diameter, and 150 mm length, with 600 oC for ALFRED 1mA)

The relative sensitivities should be corrected with relative spectrum averaged cross-sections computed at the FCs positions by the flux distribution monitoring software routines.


In-core and CLOSE-TO-CORE detectors

Corrections will have got SPNDs alsoSPNDs advantages:

Not being externally powered (HV), small dependence on insulation resistance,simple, robust, and have negligible leakage current,could function at high temperatures, like thermocouples.

Protection against corrosion and abrasion in lead should be ensured for detectors, cabling, connection boxes, and penetrations should be considered into detailed design regarding installation of in-core/close-to-core detectors.

In-core or close-to core FCs shall be provided with the electronics ensuring signal processing current regimes.

The SPNDs electronics consists of a current (eventually low current) measurement line, and a current-to-voltage converter ensuring gain adjustment for calibration.


OPTIONS

for the neutron instrumentation:A - In-core detectors + ex-core detectorsB - Close-to-core detectors + ex-core detectorsC - Ex-core detectors only

The in-core or close-to-core detectors could be: IC-FC – low sensitivity FCs for neutron flux distributions and

local fission power measurements ICD-SPD – delayed SPNDs for neutron flux distributions ICP-SPD – prompt SPNDs for local fission power

measurementsThe ex-core detectors could be high-sensitivity FCs

placed: EC-NC - near-coreEC-IV – between inner vessel and reactor vesselEC-EV – in special locations outside reactor vessel


OPTIONS cont.

Regarding technical and design solutions: the most challenging options are those combinations involving in-

core detectors, like A+IC-FC+EC-NC. Options involving SPNDs are less challenging, due to SPND robustness, and lack of power supply

less challenging options are those involving close-to-core detectors, like B+IC-FC+EC-NC. For those options SPNDs will have low sensitivities

the least challenging options are those involving ex-core detectors only. Among them C+EC-IV seems most feasible, although most power fast reactors use option C+EC-EV.

One may take into account that detailed outlet temperature measurements produce also local power information

Computed spectral corrections applied to the sensitivity of FC will depend on burn-up

For ex-core detectors, which see the whole reactor, spectral influences on the signal are less important


OPTIONS cont.

The differences between FA and dummy spectra in low/intermediate energy range (up 0.01 MeV) and high energy range are comparable.

The spectrum outside the inner vessel is significant in low/intermediate energy range, while decreasing at higher energies.

Thus, apart from increased sensitivity outside inner core, the detector signal will be more sensitive to total flux variations.

One may conclude that in-core measurements have inherent imprecision that may be considered when choosing an option for neutron instrumentation.


OPTIONS cont.

0.0E+00

5.0E-02

1.0E-01

1.5E-01

2.0E-01

2.5E-01

1.E-07 1.E-06 1.E-05 1.E-04 1.E-03 1.E-02 1.E-01 1.E+00 1.E+01 1.E+02

Energy (MeV)

Sp

ec

tru

m (

gro

up

flu

x p

er

un

it l

eth

arg

y) FAs

DummyOutside inner vessel

Volume averaged normalized spectra in FA, dummy, and outside inner vessel regions


SAFETY FUNCTIONS

Ex-core or near-core detectors and in-core or close-to-core detectors (if they are involved in safety functions provide trip signals leading to activation of safety rod systems, shut-down systems)

they have to be triplicated, each with its electronic channel in order to ensure 2/3 trip logic (redundancy, independence).

Since there are two different shut-down systems, two triplets shall be used, in different places (separation).

In order to ensure diversity, a minimal requirement is to provide detectors/electronics from different suppliers.

Ex-core or near-core detectors and their electronics associated with the reactor automatic control should be separated from the detectors/electronics of the protection system

For redundancy at least two independent control channels are required. A loss-of-regulation signal should be provided to limitation and protection systems.


QUALIFICATION REQUIREMENTS

Neutron flux monitoring shall continue past DBE. As ex-core detectors and the associated electronics belong to the protection system, they should be qualified.

If dummy bundle assemblies are used for detectors and cables, their seismic qualification will cover the seismic qualification requirements for the detector assembly. For out of inner vessel placement, detector assembly and housing (if needed) require seismic qualification.

Qualification procedure shall confirm that the equipments perform their function correctly. These equipments require quality assurance and reliability data.

Qualification shall be performed following IEC 60780


PRIMARY COOLANT C&I DEVICES

As for core C&I, in the present project stage locations for coolant instrumentation are not comprised

Where more detailed thermal-hydraulics analysis is not available, qualitative or semi-quantitative evaluations are presented in order do emphasize alternatives with more detailed specifications

As for core C&I, accuracy, rangeability (accuracy over a wide range), response time, and reliability criteria allowed selecting measurement methods and instrumentation


Lead temperature measurements

According to DEL014 2011 of Task 4.4 [3] the following temperature values should be measured: 1. Lead temperature at each assembly outlet, and lead

temperature at core inlet (for loss of coolant flow, DBC4 fuel assembly partial blockage, and calculation of actual reactor power)

2. Local lead temperature measurements at significant locations in the reactor pool, including SG outlets

As stated in [3], the reactor should be tripped in the following situations: TT1-the temperature at some points in the reactor pool decrease

below safety values (risk of lead freezing) TT2-high lead temperature in the reactor pool (SG system line

break) TT3-high temperature difference through the core (high core

power level) TT4-excessive increase of the lead core outlet temperature TT5-high rate of local temperature decrease in the reactor pool

General description


General description cont.

places where freezing could be favored: colds spots around SGs outputs are the most significant locations Simulations emphasizing flow stratification in the pool could be

used to choose measurement locations The average coolant temperature in the reactor pool should be

calculated by weighting the measured local temperatures (req. 2 above) and core inlet temperature (req. 1 above) through “importance” coefficients, which might depend on the thermal capacity of the coolant volume represented by the local measurement and on the importance of the measuring point for plant safety [3].

The average coolant temperature is the measured parameter for trip conditions TT2 and TT5.

The core outlet temperatures will be regulated by reactor control rods [3].

The pool temperatures will be usually regulated by SG feedwater flow [3].


Technical and design requirements

the sensors, for measure T for power calculation, shall be placed as close as possible to fuel channels inlet and outlet (Ref. [1] thermowell, mechanical damage, corrosion and minimized dimensions) Wet thermowells preserve response times Dry thermowells increase response times

Simulations of DEC scenarios for ELSY (DEC-10-031) show that the outlet temperature could reach values around 750 oC so: upper limit of 800 oC could be stated. For the pool temperatures an upper limit of 750 oC could be stated.

Both for core leading dynamics transients (e.g. UTOP), or for SG/secondary leading dynamics transients (e.g. ULOHS), the heat transport processes involve large time constants (an upper limit of 1.5s for the sensor response time could be imposed)

For T measurements the accuracy should be better than 1%.


Technical and design requirements cont. Temperature sensors:

K-type Chromel-Allumel thermocouples more frequently used in nuclear environments.

J-type thermocouples (ref [1]) fast response (few milliseconds) in range 0-750 if it grounded a few ms delay response time (including possible filtering could rise time ) Accuracy

Temperature fluctuations measurements could be used to detect the onset of fuel assembly blockage

An increase of the temperature power spectrum density is more sensitive to flow blockage than flow measurements

For T measurements a differential scheme of thermocouples would be more accurate.

Core outlet temperatures for each FA could be measured by placing thermocouples at the “nose” of the fuel assembly

The setpoint for low temperature in pool (LOW TEMP) should be accompanied by a much lower LOW-LOW TEMP setpoint indicating a faulty sensor or transmitter

Excessive increase of the core lead outlet temperature (trip criterion TT4) could be produced by a flow blockage in a fuel assembly


Technical and design requirements cont. Metal-sheathed thermocouples are available with:

outer coatings for use in corrosive conditions [1]. they could be located inside a thermowell that protects it from

mechanical damage and corrosion, ensuring correct measurement and instrument reliability

Thermowells allow removal of sensors for calibration, replacement, or repairs [1].

Cable lining, avoiding impact on fuel handling, connection boxes, and penetrations should be considered into detailed design regarding installation of temperature measurement devices.



Safety functions In-pool temperature sensors, cables and the associated electronics function in

DBC1, DBC2, DBC3, and DBC4 conditions of EUR criteria. Core inlet temperature sensors, cables and the associated electronics function

in DBC1 and DBC3 conditions of EUR criteria. Core outlet temperature sensors, cables and the associated electronics function

in DBC1, DBC3, and DBC4 conditions of EUR criteria. They perform protection functions F1A and F1B functions (IEC 61226 categorization).

Trip criteria TT1, and TT5, may impose trips based of a local temperature value below or above the setpoint, because the temperature variation could be local

Grouping 3 sensors per region, 2/3 trip logic will be allowed (for trip signal). Since core outlet temperature sensors provide trip signals leading to activation

of safety rod systems, they have to be triplicated, each with its electronic channel in order to ensure 2/3 trip logic (redundancy, independence)

sensors from of different type (e. g. K and J), or different suppliers should be provided for different shut-down systems


QUALIFICATION REQUIREMENTS

As temperature sensors and the associated electronics belong to the protection system, they should be qualified.

If thermowells are used, they will require seismic qualification.

The equipment securing them in the vessel should be also qualified.

Qualification shall be performed following IEC 60780.


Pressure and lead flow measurements

Pressure measurements inside the cover gas are required by DEL014 2011 of Task 4.4 [3] for cover gas line break detection, and for differential pressure transducers associated with several flow measurement techniques.

The pressure gauge should resist high temperatures, corrosion and abrasion. Flow measurements could be used to assess flow blockage, complementary to

temperature measurements close to fuel assembly outlet. Bulk flow-rates in pool-type fast reactors are difficult to be measured because

flow to a component cannot be inferred from the sensors placed at any point [7]

In existing fast reactors several flow-meter types were used [7]: P flow-meters (Venturi, Sharp-edge orifice), magnetic flow-meters (permanent magnet, DC electromagnetic, and AC electromagnetic), ultrasonic flow-meters.

Venturi, permanent magnet and AC electromagnetic flow-meters were used in EBR-II


TECHNICAL AND DESIGN REQUIREMENTS Pump failure detection and DBC4 pump shaft break/seizure Low ratio of primary coolant flow to neutron flux (protection against low

flow characteristic to ULOF events). flow-meter could be inserted in a bypass of the main flow As suggested in [4] measurements of the Venturi flow-meter could be used

for in-situ recalibration by compensating the signal drifts in time of the electromagnetic flow-meters.

Another method of recalibration may consist in comparing the SG power at steady-state around nominal power with power computed using measured bulk flow and T measurements between core outlet and inlet.

Cable lining, avoiding impact on fuel handling, connection boxes, and penetrations should be considered into detailed design regarding installation of pressure and flow measurement devices.


Venturi flow-meters could be challenging for primary circuit design. Installing them on pump outlet pipes would cause worse accuracy and possible pressure field and flow perturbations.

Installing them on bypasses would increase the chance of oxide depositions.

AC electromagnetic flow-meters and ultrasonic flow-meters are less challenging. By performing their on-line calibration good performances could be achieved


Safety functions

Coolant bulk flow sensors, associated transmitters and transducers, cables and associated electronics function in DBC1, DBC2, and DBC4 conditions of EUR criteria. They perform protection functions F1A and F1B of IEC 61226 categorization

Since coolant bulk flow sensors, associated transmitters and transducers, cables and associated electronics provide trip signals leading to activation of safety rod systems, they have to be triplicated in order to ensure 2/3 trip logic (redundancy, independence).

Flow reduction below a limit value could trigger the passive shutdown system, hence only a single triplet would be necessary for the active shut-down systems.

One could use pump speed measurements, Venturi and AC electromagnetic flow-meters, or pump speed measurements, AC electromagnetic flow-meters and ultrasound flow-meters on each pump circuit to ensure triplication.

Pressure sensors in cover gas and the associated electronics work in DBC4 conditions performing F1A and F1B protection functions


Qualification requirements

As bulk flow sensors, associated transmitters and transducers, cables and associated electronics belong to the protection system, and flow measurements shall continue past DBE, they should be qualified.

Qualification procedure shall confirm that equipments perform their function correctly. These equipments require quality assurance and reliability data.



OXYGEN concentration

The oxygen control and coolant impurity removal system is a non-safety related system.

It maintains the required oxygen concentration in lead to protect structural steels from corrosion/erosion through protective oxide films on the steel and removes mechanical impurities and slag sediments from the lead coolant [8], [9]

The oxygen content should be maintained between a lower limit to allow corrosion protection by the self-healing oxide, and an upper limit to avoid the precipitation of lead oxide [9], [10].

For structural steels, the lower limit is defined by decomposition potential of the less stable oxide (usually, magnetite - Fe3O4).

The main oxygen control methods are [10]: continuous control of gas phase This method does not require

oxygen sensors. discontinuous control methods used when depletion or excess of

oxygen are detected by the oxygen sensors


OXYGEN concentration cont

Oxygen injection (bubbling) into coolant increases exchange area as compared to mass exchange at gas-liquid interface

Solid phase oxygen control is usually performed through mass exchange with compacted balls of solid oxidizers (like PbO) into a heated cavity through which molten lead passes continuously or periodically to avoid the oxidation in excess and the solid oxide formation within the system.

Oxygen concentration in molten lead is measured by sensors of lead’s oxidation potential based on measurement of the electromotive force (emf) arising in a cell formed by the coolant, solid electrolyte with ionic (oxygen-based) conduction and reference electrode with predetermined constant oxygen content (Bi/Bi2O3, In/In2O3 or Pt/Air)

Pt/Air electrode has safety issues, since air flow has to be maintained Cracking of the solid electrolyte Yttrium Stabilized Zirconia when using Bi/Bi2O3

as the reference electrode has been reported


Technical and design requirements

Following the recommendation of [10] for ELSY, control of H2O, H2 and O2 in the gas phase allows stabilizing the oxygen content in ALFRED’s pool by adjusting flow rates and H2O content to compensate oxygen consumption in the reactor.

Oxygen sensors placed at different places in pool should determine the actual oxygen content and adapt the gas phase if required [10].

Alternative concepts with oxygen supply via PbO pebbles have to be considered for a reactor as well [10].

The combination of the continuous gas phase process with the discontinuous PbO solid pebble method is recommended in [10].

Concentration of dissolved oxygen of about 10-5 – 10-7 % wt. is favourable for the creation and maintenance of the oxide layer on the surface of structural steels, while concentrations lower than 10-8 % wt. cause dissolution of the oxide layer, and concentrations higher than 10-5 % wt. support the creation of the solid phase of PbO [10].


Reactor Plate

BubblerSolid

oxidizer

OS

OS

OS

Filter

Pump

Pb level

OS

Simplified diagram of a system based on bubbling and solid oxidizers


Technical and design requirements cont.

We add the following requirements for oxygen concentration measurements [9]:

- specific to the dissolved oxygen, but not sensitive to bounded oxygen - rapid and continuous measurement, that is able to be implemented directly on-line

in the system, provided the leak tightness of the seal in between the liquid metal and the ceramic

- wide concentration range covered by one single sensor - very low detection limit - wide operating temperature range - no disturbance on the measured system

OS with Yttria Stabilized Zirconia as solid (ceramic) electrolyte with Bi/Bi2O3 as reference electrode could withstand the above requirements

Its main specifications could be stated for ALFRED also: - Range of TDA (thermodynamic activity ) oxygen measurement: 10-6 – 1 % wt. (10-6 %

wt. is the proposed limit for ELSY [13]) - Working temperature range: 350 – 650 °C - Working pressure: 0 – 1 MPa - Rate of temperature change: up to 100 °C/s - Life time: more than 10000 hours


Lead level in pool (for main vessel break) The radar method is recommended in [1] for monitoring the lead

level as providing reliable non-contact measurement, touch free indication

These devices have a range up to 40 m and it can measure up to 450 ºC or higher with additional cooling devices based on air or nitrogen gas. Radar sensors have accuracy about 2 mm.

Technical and design requirements The level sensors should have the range greater than 10 m, and

accuracy less than 5 mm. They have to withstand temperatures grater than 400 ºC.

The sensor could be installed outside the vessel, providing a non-metallic window through the reactor plate for radar pulses [1].

Level sensors will be steel rods hanging from the top cover. When the lead reaches a level rod, it will close a circuit between the level sensor and the vessel wall [1].

By closing the circuit of a rod whose elevation above the normal level equals the trip setpoint a relay contact of the shut-down system will open.


Safety functions and qualification

Level sensors and associated electronics function in DBC1 and DEC conditions of EUR criteria. They perform protection functions F1A and F1B of IEC 61226 categorization.

Since level sensors and associated electronics provide trip signals leading to activation of safety rod systems, they have to be triplicated in order to ensure 2/3 trip logic (redundancy, independence).

Using the discontinuous method only for protective functions (pre-alarm and trip), both continuous and the discontinuous techniques could be implemented, ensuring two triplets for both shut-down systems.

These equipments require quality assurance and reliability data.



Steam concentration in cover gas

Steam condensers comprised in the gas space monitoring system could be used to indicate a steam generator leakage on the secondary circuit, helping detect leakage initiation and follow its progression [11].

Continuous or periodic chromatographic analysis of protective gas serves for routine indirect detection of steam presence in lead [11].

Steam concentration in cover gas measurement devices work in DBC1 and DBC3 conditions of EUR criteria.

They perform protection functions F1A and F1B of IEC 61226 categorization, thus 2/3 trip logic, independence, diversity, separation and qualification are required


LOOSE PARTS MONITORING

Loose parts inside the coolant could be detected using acoustic emission sensors, placed on vessel and pipes where loose part are likely to travel and impact.

Acoustic monitors are used also in other reactors as backup for temperature measurements detecting onset of fuel assembly blockage.


IN-SERVICE INSPECTION

For in-service inspection (ISI) ultrasonic visualization techniques can be used [4].

For ISI of the reactor vessel and its internals a camera formed by several ultrasonic transducers will be deployed in the reactor pool during shutdowns (to avoid damaging neutron flux level).

For ISI of the SG tubes a small diameter probe is required [4].

Due to the hot and radioactive environment multiplexing electronics cannot be integrated in the head of the camera [4].

For ISI of the main vessel ultrasonic sensors operating in temperatures around 350°C are required.


[1] F. Rivero, State of the art Instrumentation and Control Survey, DEL006/2011

[2] L. Vermeeren, Task 4.2 Presentation, May 2011 [3 ] A. Campedrer, Normal, transient and accidental operational

modes: control and protection functions identification, DEL014/2011

[4] J.P. Trapp S. Haan. L. Martin J.L. Perrin M. Tixier, High Temperature Fission Chambers: State-of-the-Art

[5] D. Gugiu, Private communication [6] P.Swaminathan, Computer based on-line monitoring system for

Fast Breeder Test Reactor, IAEA, OECD/NEANSC, JAERI specialist meeting on in-core instrumentation, 14-17 October 96, Japan.

References


[7] John I. Sacket, Measurement of Thermal-Hydraulic Parameters in Liquid- Metal-Cooled Fast-Breeder Reactors, International Center for Heat and Mass Transfer Symposium: Measurement Techniques in Power Engineering August 29 - September 3, 1983, Beograd, Yugoslavia

[8] A. Bolind, Control of the Oxygen Content of the Cover Gas in a Molten Lead-Bismuth Euthectic System, Master Thesis, Urbana, Illinois

[9]***, “Handbook on Lead-bismuth Eutectic Alloy and Lead Properties,

Materials Compatibility, Thermal-hydraulics and Technologies”, 2007 Edition

[10] Vl. Sulc, UJV, Assessment of the Lead technology and Development

Needs, ELSY European Lead-cooled System, DEL-09-035

[11] V.V.Orlov, A.I.Filin, V.S.Smirnov, A.G.Sila-Novitsky, V.S.Tsykunov, V.N.Leonov, V.P.Smirnov, A.V.Lopatkin, I.Kh.Ganev, S.N.Bosin, V.A.Abramob, Z.I.Emelyantseva, G.A.Khacharesov, V.A.Kogut, Naturally-safe Lead-coolded Fast Reactor for Large-scale Nuclear Power, Moscow, 2001

References cont.


[12] ***, Proceedings of the Third Conference Heavy Liquid Metal Coolants in Nuclear Technology, Obninsk 2008

[13] P. Schuurmans, et. al., Future RD needs for ELSY, ELSY European Lead-cooled System, DEL-10-001

[14] C. Schroer, J. Konys, Physical Chemistry of Corrosion and Oxygen Control in Liquid Lead and Lead- Bismuth Eutectic, FZKA 7364, Dec. 2007

References cont.


1. By adapting the MVB model using an enthalpy averaging procedure to satisfy the imposed value of SG tube length, the results simulations of the non-linear dynamics of coupled SG and core do not deviate significantly from those reported in ref. [1], as regarding final steady-states. As regarding shape, including peak heights, comparisons could be performed based a common set of heat transport core-SG and pump. Most important differences at final steady-state could be observed for coolant flow rate variation. But the final values are close to initial values and slight differences in thermophysical data or correlations for water and lead could explain the deviations.

2. The linearized model results fit well the non-linear results for 1%, or 1 oC input variations for all transients, except turbine admission coefficient variations, where several percent deviations persist at very small input variation. This issue may be further investigated.

3. Restricting the range of heat transport time constants and working with actualized values of reactivity coefficients (for ALFRED) would be desired.

4. Conclusions

authors: laurentiu aioanei dumitru dobrea karlsruhe november 21, 2012 inr pitesti

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