ML19340C975
| ML19340C975 | |
| Person / Time | |
|---|---|
| Issue date: | 10/24/1980 |
| From: | Andrew Hon, Hsu Y NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES) |
| To: | |
| References | |
| IAEA-CN-39-100, NUDOCS 8012180430 | |
| Download: ML19340C975 (22) | |
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INTERNATIONAL ATOMIC ENERGY AGENCY INTERNATIONAL CONFERENCE ON CURRENT NUCLEAR POWER PLANT SAFETY ISSUES
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, Stockholm, 20-24 October 1980 IAEA CN 39/ 100 i
Possible Ways To improve The Nuclear Power Plant i
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Instrumentation
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- n. f Office of Nuclear Regulatory Research U. S. Nuclear Regulatory Commission Washington, D. C. 20555 801e186W30 gg This is a preprint of a paper intended for presentation at a scientific meeting. Because of the provisional nature of its content and since changes of substance or detail may have to be made before publication, the preprint is made available on the understanding that it will not be cited in the literature or in any way be reproduced in its present form. The views expressed and the statements made remain the resporaibility of the named author (s); the views do not necessarily reflec+ those of the govem-ment of the designating Member State (s) or of the designating organization (s). In particular, neither the /AEA. Tor any other omanization or body sponsoring this meeting can be held responsible for any merenal o.~oroduced in this preprint.
INTERNATIONAL ATOMIC ENERGY AGENCY INTERNATIONAL CONFERENCE ON CURRENT NUCLEAR POWER PLANT SAFETY ISSUES qyp w
Stockholm, 20-24 October 1980 IAEA CN 39/ 100 Possible Ways to Improve the Nuclear Power Plant Instrumentation Y. Y. Hsu & A. L. M. Hon ABSTRACT This paper examines the nuclear power plant instrumentation adequacy from lessons learned from the Three Mile Island Accident and available advanced technology.
From these, it identifies two areas that should be improved - unambiguous indication and reli-able data collection.
It suggests that these two areas can be improved by direct measurement of key parameters, grouping of information, disturbance analysis, self-verification of serisors, the sensors ability to survive in hostile environment and to measure in extended range.
The paper also reports the on-going reseirs p;ogram at the l
0.S. Nuclear Regulatory Commission to address suggested items.
Some of them are in-vessel liquid level measurement and on-line reactor surveillance system, i
1.
INTRODUCTION l
l The Three-Mile Island (TMI) nuclear power plant accident was the most serious one in the comercial nuclear power history.
Many lessons were learned through intensive investigations; these lessons are significant not only to the United States, but the world as well to enhance the safety of nuclear power.
A number of special investigation groups were comissioned to independently study this accident.
Since many reports have been written about this accident, it will not be repeated in detail here [1],[2],[3],[4], except that one area identified as critical d needs improvement is power plant instrumentation. NRC/RSR vivision has the responsibility to provide technical support through confinning research and techriical evaluation to determine the adequacy of instruments.
This is a preprint of a paper intended for presentation at a scientific meeting. Because of the provisional nature of its conterat and smce changes of substance or detail may have to be made before publication, the preprint is rna available on the understanding that it will not be M*eo in the literature or in any way be reproduced in its present form. The views expressrA and the statements made remain the.esponsibility of the named author (s); the views do not necessarily refitet those of the govem-ment of the designating Member state (s) or of the designating organization (s). In particular, neither the IAEA nor any other organization or body sponsoring this meeting can be held responsible for any materis) reproduced in thos preprint.
The studies [1],[2],[3],[4] generally concluded that certain existing instrumentation areas needed improvement, specifically in the following areas [1],[7]:
a.
Direct and unambiguous measurement of parameters, such as the water level in the reactor vessel, and the relief valve position.
b.
Extended range measurement of important parameters such as the in-core thermocouples and radiation monitors to cover both the operational and accident conditions.
c.
Be able to survive in hostile high radiation and temper-ature environments, especially during and after an accident.
d.
Assure the reliability and accuracy of key instruments so that the operator will not discard them when abnormal readings are shown.
e.
11 formation should be displayed to the operator in a coaorehensible form.
This paper sugsests some ways to meet these needs and reports some of the relt ted research activities.
2.
ADVANCED TECHNOLOGY TRANSFER When we conside.' major power plant instrumentation improve-ment, it is worthwhile to see what advanced technologies we can utilize.
In the past 10 years, the USNRC and its predecessor, the Atomic Energy Commission (AEC), have been studying the phenomena and behavior of nuclear reactors under loss-of-coolant-accident (LOCA) accidents [5]. Most of these studies were performed at the national laboratories with large test facilities, such as LOFT, Semi-Scale, PowerBurst-Facility (PBF), Blow-Down Heat Transfer (BDHT) facility, the U.S. -Japan-Germany 3-D Upper Plenum Test Facilities, etc. These tests usually call for accurate instru-2 ments which can perform under severe simulated LOCA conditions.
Since most of the commercially available instruments cannot meet the need, the NRC-contracted laboratories have been developing their own and have gained a great deal of experience [6]. This technology together with the advancement at the aerospace and defense industries can be utilized by the nuclear industry to improve power plant instrumentation.
e l-3.
INSTRUMENTATION IMPROVEMENT IDENTIFIED AND THE USNRC RESEARCH In response to the need for power plant instrumentation research and development (R&D) as shown in the previous discussions, the authors identified the following specific areas:
A.
Unambiguous Indication 1)
Direct measurement of key parameters 2)
Integrated functional display 3)
Disturbance analysis.
B.
Reliable Data Collection 1)
Self-verification 2)
Survival in hostile environment 3)
Extended range sensors.
We will discuss each of these items, as well as the related research at the USNRC Division of Reactor Safety Research (RSR) to assess the adequacy of existing instruments and to improve if necessary. We also mention some related industrial work.
3.1 Unambiguous indication 3.1.1 Direct measurement of key parameters In order to increase the confidence of the operator on the instruments and to reduce the burden on the operator to infer the critical information from secondary measurements, the key par-ameters should be measured directly by reliable sensors.
Some of them revealed by TMI are in-vessel liquid level indication, valve position and leakage detectors, saturation condition of the coolant, on-line radiation sampling capability [1]. We will briefly discuss each of them.
3.1.1.1 In-vessel liquid level detection Currently, the liquid level in PWR vessels is not measured directly but inferred from the pressurizer and other instruments designed for different purposes. This was judged inadequate by almost all of the study groups and the experts who examined the TMI incident. The NRC has set new requirements on this. Meanwhile, the NRC/RSR is evaluating viable techniques, such as heated thermo-couples, ultrasonic ribbon, differential pressure, etc.
l 3.1.1.1.1 Heated thermocouples (HTC)
Two types of HTCs are being developed and tested for their co1lity to detect the inadequate coolir.1 condition (liquid level).
One type is the differential temperature HTC and the other one is controlled absolute temperature HTC.
The differential type by the Oak Ridge National Laboratory (ORNL) uses a pair of K-type thermocouples [8]. One of the pair is heated by a separate wire with electric current, while the unheated one measures essentially the fluid temperature as a reference (Figure 1). The net temperature difference AT of the two is monitored to compensate for fluid temperature variations.
The aT is expected to indicate the cooling capability of the surrounding fluid. This HTC was tested in steady-state high-temperature saturated steam / water conditions from 0.1 to 8.3 MPa in a pressurizer. The tests showed the aT for uncovered state is significantly greater than that for the covered state (Figure 2).
Then it was tested in simulated pumping two-phase steam / water conditione. Again, the results showed the HTC indicated the cooling c pacity of the flowing mixture.
(Figure 3)
Another type of HTC developed by Idaho National Engi."eering Labora-tory (INEL) uses a single thermocouple with a heater and a unique electronic controller (Figure 4). This HTC's controller controls the heater power to maintain a preselected HTC temperature when the HTC is in steam (dry) phase. This temperature is set to be higher than the maximum plant temperature. However, the con-troller also limits the current so that when the HTC is in liquid phase, the current will not be sufficient to maintain the HTC at the same temperature. Thus, the changing temperature of the thermocouple can indicate whether the HTC is in liquid or steam phase. This HTC was tested in an autoclave under PWR operation temperature and pressure. The results showed it clearly indicated the level change [9].
Both HTCs will be uvaluated under additional simulated LOCA conditions at ORNL and in systems effects transient tests at the Semi-Scale Facility of INEL.
3.1.1. 2 Ultransonic technique ORNL is developing an ultrasonic level probe using a steel ribbon as the wave guide [10].
In principle, when a torsional
~
e pulse travels down a flattened ribbon and reflected from the end, the propagation velocity v is a function of the density of the fluid surrounding the ribbon. Since the material properties of the ribbon are affected by temperature due to thermal expansion, it must be compensated for. Fortunately, the temperature can be measured by sending and receiving extensional wave on the same probe and measure the propagation velocity. ORNL divided the wave guide into a series of zones with notches, so that continuous density and temperature profiles along the wave guide can be measured. The true liquid and the froth level can be displayed by combining]the density and temperature profiles with electronics
[ Figure 5.
ORNL tested a prototype from room temperature to 500 F with excellent results. More refinement and testing are planned for 1981.
3.1.1. 3 Differential pressure technique The differential pressure technique (dp) has been used exten-sively for liquid level measurements. Westinghouse Electric Corp.
is currently proposing to measure the vessel level with pressure difference from the bottom of vessel to the hot leg and to the top of the vessel (Figure 6). The NRC/RSR is planning to test it at the Semi-Scale Facility at the end of 1980, to determine the effects of length of lines and the effect of pumps.
3.1.1.4 Valve position indication At the beginning of the TMI accident, from the indicator, the operator thought the power-operated relief valve (PORV) was closed l
I while it actually stuck open for two hours, 5ecause the indicator only detected whether the solenoid was energized instead of the valve position itself. Meanwhile, the criticht emergency feed-water valve was manually shut off but no automat;c varning was I
given to the operator. These were two of the key cor,tributors to the TMI-2 accident. Shortly after TMI, the NRC took 'he Lessant Learned Task Force [3] recommendation and required the power plants to install direct valve position indicators or leakage monitors on the PORVs. Several existing instruments are avtilable to detect leakage, including acoustic leakage detector. However,
the same requirement should be extended to all critical valves including manually operated ones.
3.1.2 Functional int gration of signals On the present control panels, functionally related informa-tion often is displayed in widely separated locations [9]. Such separation causes great inconvenience to the operators. Thus, there is a need to group all the information on a functional basis. For example, all the information related to coolant inven-tory should be in one place.
If the same information is related to several functional groups, it should be displayed separately.
One of such examples is the core-cooling leakage monitor used by Westinghouse (Figure 7) [11].
The measured information can be further combined into a more functional form. All the related measured information can be verified, compared for consistency, such as level, density, temperature, pressure, etc., then digested into various functional parameters such as inventory, void fraction, subcooling, etc. An example is shown on Figure 8.
Thus, there can be two levels of information: the undigested measured quantities and the digested, functional parameters.
After discussing this with several senior operators, it appears that the grouped but undigested information should be the primary information for display, with digested information available for display when called for. Such an arrangement has several advantages:
(1)
It enables the operator to make sound judgment.
I (2) The digested information is available for the operat:r to verify his judgment, thus forming a " reinforced learning process."
(3)
It is less boring to the operator.
(4) When pressed by tim 9, the operator can directly call for digested information.
NRC is planning to study the nntimal way of forming func-l tional groups.
l l
3.1.3 Disturbance (or noise) analysis for surveillance The noise of a signal, such as temperature, pressure, neutron, noise, etc., during a particular operating condition is characterized by a particular power spectrum pattern. The pattern deviates from a normal " baseline" pattern when the operating condition is off-normal. Thus, the power spectrum of signal noise can be categorized.
By continuously collecting and analyzing the data, a baseline
" standard" pattern can be established and continuously up-dhted.
Any deviation of the pattern from baseline can be monitored as surveillance against onset of off-normal conditions. NRC/RSR is supporting a research program at ORNL to establish such noise analysis technique [12]. Typical example of noise pattern for PWR is shown in Figure 9 The monitoring system of ORNL has been installed on a commercial power plant to collect data from operating reactor.
A parallel effort of developing the surveillance system using disturbance analysis is undertaken by the Electric Power Research Institute (EPRI).
3.2 Reliable data collection It is important for the operator to be assured that the measured information is accurate and reliable. To meet such goals, the instruments must satisfy the following requirements:
i That it can survive the harsh environment of a nuclear f
a.
reactor, both under normal operation and off-normal conditions.
b.
That it can give readings over the expected range, both normal operation and off-normal conditions.
That it can be verified and desirably, self-verified.
l c.
f 3.2.1 Survival under hostile environment After TMI, the environmental requirement limits will be raised so that radioactivity will become 10eR/hr in containment I
and temperature will become 700'F in vessel.
[13]
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The survivability of instruments under the extended range of a hostile environment is still being assessed.
Fortunately, through years of R/D effort to develop research instruments, NRC contractors, notably INEL and CRNL, have developed the capability to fabricate instruments which can survive high temperature (1000*F) and severe thermal shocks (450*F/sec).
For instance, insulation seal is one of the most critical parts of the instruments. ORNL and INEL developed " Cermet" [la] and "Durole" [15] seals, respec-tively. This technology can be transferred to power plant.
For radiation-hardened electronics, Sandia has developed considerable technology for various radioactive environments, as shown in Table I [18].
These technologies are available and can be utilized by the industry to enhance the survivability of instruments.
3.2.2 Extended range As mentioned previously, the expected range of reading is 7 rad /hr. This seems to be within the capability of extended to 10 instruments produced by vendors. NRC is currently assessing such capability.
For effluent measurements, the sensors for noble gas seem to be adequate.
But the iodine monitor is expected to cover 10 10
-10 1" uCi/cc in normal and 100 pCi/cc in post-accident condi-tions. Some development work needs to be done to accurately cover such wide dual ranges without saturating the sensor.
3.2.3 Self-verification At present, an instrument is nanually verified through com-parison for consistency among redundant instruments of the same kind. Sometimes, when measurements from different kinds of instru-ments can be compared, verification can also be done by consistency But it would be more desirable to provide instruments with tests.
selfiverification capabilities through periodic consistency testing, with help of a computer is one way. There are other ways, includf-a subjecting a sensor with a known substitute input, or examining the impulse response when a step function is applied.
For example,
o l
ore can give a thermocouple a step-function of juole-heating. By comparing the transient temperature rise and decay with a pretested
" standard," the thermocouple's operational condition can be verified
[17]. The concept of self-verification and self-calibration is still relatively new to power plant applications.
4.
SUMMARY
The instrumentation plays a very important role in the nuclear power plant safety. The TMI accident revealed the deficiencies of the existing instrumentation and the need for improvements, par-ticularly the areas of unambiguous indication to the operator based upon reliable measurements. Meanwhile, the advanced technology is available from many years of LOCA research experience at NRC and the rapid advancement in the aerospace and defense industries.
The authors recommended several specific instrumentation improvements to meet the need. They are direct measurement of key parameters for unambiguous indications, integrated functional display, disturbance analysis, and sensors' self-verification capability as well as their survivability and extended range in hostile environments. We also briefly reported the plant instru-mentation research progress at the USNRC/RSR, such as in-vessel l
l liquid level detection, on-line noise diagnostics, seal material l
for the sensors, etc. to meet the need.
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4 REFERENCES TMI-2 Lessons Learned Task Force Status Report and Short
[1]
Term Recommendations, U.S. Nuclear Regulatory Commission (1979).
[2] KEMENY, J. G., et al., Report of the President's Commission on the Accident at Three Mile Island (1979).
[3] R0GOVIN, M., et al., Three Mile Island - A Report to the Commissioners and to the Public, NUREG/CR-1250, U.S. Nuclear Regulatory Commission (1979).
[4] Analysis of Three Mile Island Unit 2 Accident, NSAC-1, Electric Power Research Institute (1979).
[5] TONG, L. S., "USNRC LOCA Research Program," Current Nuclear Power Plant Safety Issues (Proceedings, International Conference in Stockholm, 1980) IAEA-CN-39/99 (1980).
[6] Review Group Conference on Advanced Instrumentation for Reactor Safety Research (Proceedings of USNRC Review Group Conference, U.S.A.) NUREG/CP-0007, U.S. Nuclear Regulatory Commission (1979).
IEEE-NRC Working Conference on Advanced Electrotechnology
[7]
Applications to Nuclear Power Plants (Proceedings, U.S.A.,
1980) IEEE Cat. No. TH-0077-8, The Institute of Electrical and Electronics Engineers, Inc., New York (1980).
[8] TURNAGE, K. G., In-Vessel Liquid Level Probe for PWRS -
Thermal Cavices (Proceedings of USNRC Advanced Instru-mentation Review Group Conference, U.S.A.,1980) NUREG/CP-0015, U.S. Nuclear Regulatory Commission (1980).
ANDERSON, J. V., Heatad Thermocouple Liquid Level System
[9]
(Proceedings of USNRC Advanced Instrumentation Review
' Group Conference, U.S.A.,1980) NUREG/CP-0015, U.S. Nuclear Regulatory Commission (1980).
Y e
i
[10] MILLER, G. N., In-Vessel Liquid Level Probes for PWRS -
Ultrasonic Devices (Proceedings of USNRC Advanced Instrumentation Review Group Conference, U.S.A., 1980)
NUREG/CP-0015, U.S. Nuclear Regulatory Commission (1980).
[11] GOPAL, R., Functional Integration of Signals, Oral Presentation in the NRC/IEEE Working Conference on Advanced Electrotech-nology Applications to Nuclear Power Plants, Washington, D.C., January 1980.
[12] SIDES, W. H., Jr., PIETY, K. R., FRY, D., Automated Pattern Recognition System for Noise Analysis, Trans. American Nuclear Society 34 (1980) 720.
[13] Regulatory Guide 1.97, Proposed Rev. 2, " Instrumentation for Nuclear Power Plant to Assess Plant and Environs Conditions During to Follow An Accident," U.S. Nuclear Regulatory Commission (1979).
[14] MOREHEAD, A.
J., " Fabrication of Sensors for High Temperature Steam Instrumentation System, NUREG/CR-1359, U.S. Nuclear Regulatory Commission (1980).
[15] STANLEY, M. L., Meeting Commercial Reactor Instrumentation Needs (Proceedings of USNRC Advanced Instrumentation Review Group Conference, U.S.A., 1980).
[16] GOVER, J., High Temperature, Radiation Hardened Electronics for Application to Nuclear Power Plants, SAND 80-0362C, Sandia National Laboratories (1980).
[17] CARNELL, R. M., SHEPARD, R. L., Measurement of the Transient Response of Thermocouples and Resistance Thennometers Using an In-Situ Method," ORNL/TM-4573, Oak Ridge National Laboratory (1977).
,