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Journal ArticleDOI

Post-implementation review of inadequate core cooling instrumentation

01 Feb 1989-IEEE Transactions on Nuclear Science (IEEE)-Vol. 36, Iss: 1, pp 1248-1250

AbstractInstrumentation needs for the detection of inadequate core cooling are discussed, and a review of reactor coolant inventory tracking systems (RCITS) is presented. RCITS installation experience is considered, and attention is given to operating experience, accuracy, and procedures. It is noted that the addition of inadequate core cooling accident monitoring instrumentation to all PWRs (pressurized water reactors) is nearing completion after more than eight years. The response of and acceptance by the utilities have been largely positive, with only a few holdouts. It is concluded that the systems developed, while not providing the precision that might be desired, are sufficient to provide vital information for accident avoidance and recovery. In addition, unexpected benefits for normal operation have resulted, and operator acceptance and confidence have been good. >

Summary (3 min read)

Instrumentation Needs for Detection of Inadequate Core Cooling

  • Studies of the Three Mile Island (TMI) accident identified the need for additional instrumentation to detect inadequate core cooling (ICC) in nuclear power plants.
  • Industry studies by plant owners and reactor vendors 1 supported the conclusion that improvements were needed to help operators diagnose the approach to or existence of ICC and to provide more complete information for operator control of safety injection flow to minimize the consequences of such an accident.
  • In 1980, the U.S. Nuclear Regulatory Commission (NRC) required further studies by the industry 2 and described ICC instrumentation design requirements that included human factors and environmental considerations.

Review of RCITS

  • The NRC, with assistance from Oak Ridge National Laboratory, reviewed generic RCITSs proposed by reactor vendors, instrument manufacturers, and individual utilities.
  • Both of these systems were tested extensively under simulated accident conditions, including extremes of temperature and twophase flow, and the results have been reviewed.
  • In addition to the generic reviews, the plant-specific installations at 66 PWRs were reviewed for conformance to NUREG-0737 requirements.
  • Of these, 48 plants use one of the generic vessel water level measurement systems, 18 plants have unique designs, and 2 plants (3 units) use a gamma-thermometer (GT) level measurement scheme very similar in principle to the HJTC system.
  • The remainder of the unique designs use dp measurement schemes with various configuration differences.

ICC Instrument Performance

  • Subcooling margin monitoring provides an early indication of potential voiding but does not of itself provide any additional information about the possible approach of ICC.
  • This situation has occurred in steam generator tube rupture events and could happen in any overcooling event that results in loss of pressurizer water level.
  • Some loss-of-fluid tests (LOFT) 7 suggest that during reflood or coolant injection, CETs may be subcooled while the core remains voided.
  • Some of those problems and the current status of industry efforts to achieve a reliable and unambiguous indication of water level were discussed recently in another paper.
  • The possible effects of level measurement uncertainty for a typical small-break transient are shown in Fig. 1 .

Generic RCITS

  • The Westinghouse Reactor Vessel Level Instrumentation System uses redundant sets of 3 dp cells to measure pressure drop from the bottom to the top of the reactor vessel and from the hot legs to the top of the vessel.
  • A narrow-range transducer is calibrated to indicate full-scale with the static head across the vessel and pumps off, and the output is conditioned to display the equivalent collapsed liquid level in the vessel.
  • The head-tohot-leg measurement is used for head venting operations during long-term recovery.
  • For plants with upper head injection, the measurement from the bottom to the top of the vessel is omitted.
  • The Combustion Engineering HJTC system measures reactor coolant liquid inventory using discrete HJTC sensors located at different levels within a separator tube that extends from the reactor vessel head to the top of the core.

RCITS Installation Experience

  • Fifty operating power plant units have installed one of the two generic RCITS.
  • The operating experience obtained so far has been generally satisfactory, although a number of early failures and problems occurred in both system types, some of which necessitated field design changes.
  • Eleven units have chosen to design their own dp measurement systems or use other vendors.
  • B&W type reactors, because of their geometry, were required by NRC to have level measurement from head to hot leg and additional measurement of hot-leg (candy cane) level.
  • Thirteen additional units included ICC instrumentation in their initial licensing submittals and were reviewed separately within NRC.

Operating Experience

  • All but a few of the PWRs have completed installation of the ICC instrumentation systems and have acquired operating experience with them over 1 to 3 fuel cycles.
  • While ICC instrumentation was intended primarily for accident monitoring, expanded uses have been found to improve routine operation as operators develop confidence in the indications.
  • The SMM and CET system requirements were rather explicit in NUREG-0737.
  • Conformance to these requirements has been good, and both performance and acceptance have also been good.
  • As a result, there has been much more variation in design, performance, conformance, and acceptance.

Accuracy and Procedures

  • NRC requirements have not included an absolute accuracy specification for water level asasurements.
  • Analyses based on SfiLOCA scenarios suggest that ±6% is acceptable for procedural action requirements dealing with ICC.
  • Typically the operator is not instructed to take action on the basis of level measurement system indication alone.
  • Procedures were developed for some units with "conservative" decision points which, coupled with appropriate operator training, purportedly would not mislead operators into initiating inappropriate action.
  • This procedure may cause problems in the case of events that call for a strategy of keeping the core covered while maintaining the water level below the vessel nozzles to reduce the flow of water out of the nozzles where a break may exist in the hot leg.

Human Factors

  • Most utilities have done a good job of integrating ICC information displays into already full control rooms.
  • An appropriate choice taken by many is to incorporate the primary ICC display in the safety parameter display system (SPDS) or other advanced information display systems.

Conclusions

  • The addition of inadequate core cooling accident monitoring instrumentation to all PWRs is nearing completion after more than eight years.
  • Indirect costs for items such as procedures and maintenance may add another $2H to $411.
  • Cost effectiveness cannot be measured directly, but must be judged subjectively in relation to the socioeconomic impact of another accident similar to the one at TMI.
  • The response and acceptance of the utilities has been largely positive, with only a few holdouts.
  • The systems developed, while not providing the precision that might be desired, are sufficient to provide vital information for accident avoidance and recovery.

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POST-IMPLEMENTATION REVIEW OF INADEQUATE
CORE COOLING INSTRUMENTATION
J. L. Anderson
. R. L. Anderson
E. W. Hagen CONF-881103—14
T. C. Morelock
Oak Ridge National Laboratory* DE89 002347
Oak Ridge, Tennessee
Tai L. Huang
Laurence E. Phillips
U.S.
Nuclear Regulatory Commission
Washington, D.C. 20555
Paper presented to the
IEEE 1SSS Nuclear Science Symposium
Orlando,
Florida
November 7-11, 1988
DISCLAIMER
This report was prepared as an account of work sponsored by an agency of the United States
Government. Neither the United States Government nor any agency thereof, nor any of their
employees, makes any warranty, express or implied, or assumes any legal liability or responsi-
bility for the accuracy, completeness, or usefulness of any information, apparatus, product, or
process disclosed, or represents that its use would not infringe privately owned rights. Refer-
ence herein to any specific commercial product, process, or service by trade name, trademark,
manufacturer, or otherwise does not necessarily constitute or imply its endorsement, recom-
mendation, or favoring by the United States Government or any agency thereof. The views
and opinions of authors expressed herein do not necessarily state or reflect those of the
United States Government or any agency thereof.
Operated by Martin Marietta Energy Systems, Inc., for the
U.S.
Department of Energy under Contract No. DE-AC05-840R21400.
MASTER
DISTRIBUTION
OF
THIS OOCUMENT
JS U.\.

POST-IMPLEMENTATION REVIEW OF INADEQUATE
CORE COOLING INSTRUMENTATION
J. L. Anderson Tai L. Huang
R. L. Anderson Laurence E. Phillips
E.
W. Hagen U.S. Nuclear Regulatory Commission
T. C. Morelock Washington, D.C. 20555
Oak Ridge National Laboratory*
P.O. Box 2008
Oak Ridge, TN 37831-6008
Instrumentation Needs for Detection of
Inadequate Core Cooling
Studies of the Three Mile Island (TMI) accident
identified the need for additional instrumentation to
detect inadequate core cooling (ICC) in nuclear power
plants.
Industry studies by plant owners and reactor
vendors
1
supported the conclusion that improvements
were needed to help operators diagnose the approach to
or existence of ICC and to provide more complete
information for operator control of safety injection
flow to minimize the consequences of such an accident.
In 1980, the U.S. Nuclear Regulatory Commission (NRC)
required further studies by the industry
2
and
described ICC instrumentation design requirements that
included human factors and environmental
considerations.
3
On December 10, 1982, NRC issued to
Babcock & Wilcox (B&W) licensees' orders for
Modification of License and transmitted to all
pressurized water reactor (PWR) licensees Generic
Letter 82-28 to inform them of the revised NRC
requirements. The instrumentation requirements for
detection of ICC include upgraded subcooling margin
monitors
(SMMs),
upgraded core exit thermocouples
(CETs),
and installation of a reactor coolant
inventory tracking system
(RCITS).
NRC Regulatory
Guide 1.97, which covers accident monitoring
instrumentation,* was revised (Rev. 3) to be
consistent with the requirements of item II.F.2 of
NUREG-0737.
3
Following are some of the more
significant requirements specified in that item.
Operated by Martin Marietta Energy Systems,
Inc.,
for the U.S. Department of Energy under Contract
No.
DE-AC05-84OR21400.

1. Instrumentation should provide an unambiguous
indication of the approach to and existence of
ICC.
2.
Reactor water level measurement is to be
considered.
3. The system must indicate the existence of ICC
caused by various phenomena (e.g., high void
fraction pumped flow and stagnant
boil-off).
4.
The presence of an unrelated phenomenon must not
cause the system to erroneously indicate ICC.
5. Advance warning of the approach of ICC must be
given.
6. Instrumentation must conform to Appendix B (Class
IE) of NUREG-0737.
7.
Alarms and displays should be selected based on a
human factors analysis.
8. Instrumentation indications must be integrated
into emergency procedures and operator training
programs.
Review of RCITS
The NRC, with assistance from Oak Ridge National
Laboratory, reviewed generic RCITSs proposed by
reactor vendors, instrument manufacturers, and
individual utilities. Two water level measurement
methods were developed by vendors and approved
generically by NRC for application in nuclear power
plants:
the Westinghouse differential pressure (dp)
system and the Combustion Engineering heated junction
thermocouple (HJTC) system. Both of these systems
were tested extensively under simulated accident
conditions,
including extremes of temperature and two-
phase flow, and the results have been reviewed.
5
*
6
Both systems were shown to have adequate response to
the accidents simulated. In addition to the generic
reviews,
the plant-specific installations at 66 PWRs
were reviewed for conformance to NUREG-0737
requirements. Of these, 48 plants use one of the
generic vessel water level measurement systems, 18
plants have unique designs, and 2 plants (3 units) use
a gamma-thermometer (GT) level measurement scheme very
similar in principle to the HJTC system. The
remainder of the unique designs use dp measurement
schemes with various configuration differences.

ICC Instrument Performance
Subcooling margin monitoring provides an early
indication of potential voiding but does not of itself
provide any additional information about the possible
approach of ICC. Sometimes the reactor vessel head
water level monitor provides an initial indication of
voiding at the same time that the hot-leg resistance
temperature detectors
(RTDs),
or even the
CETs,
indicate that subcooling still exists. This situation
has occurred in steam generator tube rupture events
and could happen in any overcooling event that results
in loss of pressurizer water level. The upper head
can be the region of highest temperature, thus acting
as the system pressurizer. In addition, a small-break
loss-of-cooling accident (SBLOCA) with a leak in the
upper head of the reactor vessel can result in upper
head voiding. Small line breaks or faulty valves can
lead to such accidents.
CETs provide perhaps the most reliable indication
of the existence of ICC during stagnant boil-off when
superheated steam conditions are present. However,
some loss-of-fluid tests
(LOFT)
7
suggest that during
reflood or coolant injection, CETs may be subcooled
while the core remains voided. There is some
indication from LOFT that CETs can be cooled by water
falling back from the steam generator. A serious
condition can occur if the loss of pumps in a highly
voided situation permits the water level to collapse
below the bottom of the core. The core can then heat
up without detection by the CETs because there is no
coolant in the core to
boil,
and thus no superheated
steam is generated to flow past and heat the
CETs.
A
diverse measurement system that includes measurements
of other parameters, such as coolant inventory or
water level, is needed to complement GET measurements
under abnormal conditions. When the primary coolant
pumps are running, the voids tend to be distributed
throughout the system, and the resulting "froth" can
provide adequate cooling when high void fractions
exist.
However, if pumping is continued, high void
fraction mixtures (> -25%) are likely to cause pump
damage.
The problems of ambiguities in water level
measurement and the difficulties of achieving adequate
accuracy with water level or inventory tracking
systems are considerable. Some of those problems and
the current status of industry efforts to achieve a
reliable and unambiguous indication of water level
were discussed recently in another paper.
8
The
possible effects of level measurement uncertainty for
a typical small-break transient are shown in Fig. 1.

Generic RCITS
The Westinghouse Reactor Vessel Level
Instrumentation System (RVLIS) uses redundant sets of
3 dp cells to measure pressure drop from the bottom to
the top of the reactor vessel and from the hot legs to
the top of the vessel. A wide-range transducer
includes the pump dynamic head and is used to infer
void fraction with the pumps running. A narrow-range
transducer is calibrated to indicate full-scale with
the static head across the vessel and pumps off, and
the output is conditioned to display the equivalent
collapsed liquid level in the vessel. The head-to-
hot-leg measurement is used for head venting
operations during long-term recovery. For plants with
upper head injection, the measurement from the bottom
to the top of the vessel is omitted. The Westinghouse
RVLIS dp system uses hydraulic isolators and sealed
lines compensated for temperature and density effects
inside containment, with the transmitters located
outside containment to achieve an accuracy of better
than ±6% (- ±2.5 ft) under degraded environmental
conditions in containment.
The Combustion Engineering HJTC system measures
reactor coolant liquid inventory using discrete HJTC
sensors located at different levels within a separator
tube that extends from the reactor vessel head to the
top of the core. The separator tube allows any steam-
water mixture to collapse and hence provide a steam-
water interface at the collapsed liquid level. Heated
thermocouple sensors at discrete axial levels within
the tube indicate the presence of water at the
measurement level. The HJTC system claims very high
accuracy (within a few
inches),
but resolution is
limited to 2 to A ft because of the spacing between
the limited number of discrete measurements points
(typically 8). Additional uncertainty may be
introduced during rapid level change caused by sensor
response time delay (15 to 60 s).

Citations
More filters

Journal ArticleDOI
Abstract: Since the Three Mile Island accident, an important focus of pressurized water reactor (PWR) transient analyses has been a small-break loss-of-coolant accident (SBLOCA). In 2002, the discovery of th...

12 citations


Proceedings ArticleDOI
Abstract: Presented are experiment results of the Large Scale Test Facility (LSTF) conducted at the Japan Atomic Energy Agency (JAEA) with a focus on core exit thermocouple (CET) performance to detect core overheat during a vessel top break loss-of-coolant accident (LOCA) simulation experiment. The CET temperatures are used to start accident management (AM) action to quickly depressurize steam generator (SG) secondary sides in case of core temperature excursion. Test 6-1 is the first test of the OECD/NEA ROSA Project started in 2005, simulating withdraw of a control rod drive mechanism penetration nozzle at the vessel top head. The break size is equivalent to 1.9% cold leg break. The AM action was initiated when CET temperature rose up to 623K. There was no reflux water fallback onto the CETs during the core heat-up period. The core overheat, however, was detected with a time delay of about 230s. In addition, a large temperature discrepancy was observed between the CETs and the hottest core region. This paper clarifies the reasons of time delay and temperature discrepancy between the CETs and heated core during boil-off including three-dimensional steam flows in the core and core exit. The paper discusses applicability of the LSTF CET performance to pressurized water reactor (PWR) conditions and a possibility of alternative indicators for earlier AM action than in Test 6-1 is studied by using symptom-based plant parameters such as a reactor vessel water level detection.Copyright © 2008 by ASME

4 citations


Journal ArticleDOI
Abstract: Presented are results of an application of a new primary coolant inventory tracking method which was invented for a Westinghouse-type pressurized water reactor, to various kinds of small break loss-of-coolant accident experiments conducted at the Large-Scale Test Facility in Japan Atomic Energy Research Institute. The uniqueness of this method is that it can track the primary coolant depletion prior to the initiation of inadequate core cooling. The primary coolant inventory is tracked by measuring the water level in the vertical region of each primary loop including the steam generator outlet plenum and by using a simple correlation between the level and coolant inventory. The principal level measuring range corresponds to the primary coolant volumes of approximately 30 to 60% of the initial volume. A limitation of the reactor vessel level indication system is also shown in comparison with this method.

3 citations


Journal ArticleDOI
Abstract: Presented are experimental results on the general performances of core exit thermocouples (CETs) to detect core overheat for accident management (AM) action by using the Large-Scale Test Facility (LSTF) of the ROSA Program of the Japan Atomic Energy Agency. The LSTF is a full-height, full-pressure, and 1/48-volumetric-scaled model of a 4-loop pressurized water reactor (PWR). This study was motivated by a significant delay in the time and temperature rise of the CETs from core heat-up during a vessel top head small break loss-of-coolant accident (SBLOCA) test. A certain delay in time and temperature rise of the CETs was also observed in various SBLOCA and abnormal transient tests. Such CET performances are derived from thirteen LSTF tests as follows: (1) general CET performances are obtained in the form of equations including cases under limited influences of water fall-back from hot legs, (2) the major reason for the delay is the interaction of three-dimensional steam flows with low-temperature structures...

3 citations


Proceedings ArticleDOI
30 Jul 2012
Abstract: In the transients produced during Small Break Loss-Of-Coolant Accidents (SBLOCA), the maximum Peak Cladding Temperature (PCT) in the core could suffer rapid excursions which might strongly affect the core integrity. Most Pressurized Water Reactors (PWR) have Core Exit Thermocouples (CETs) to detect core overheating by considering that superheated steam flows in the upward direction when core uncovery occurs during SBLOCAs. Operators may start Accident Management (AM) actions to mitigate such accident conditions when the CET temperature exceeds a certain value. However, in a Vessel Upper Head SBLOCA, a significant delay in time and temperature rise of CETs from core heat-up can be produced.This work is developed in the frame of OECD/NEA ROSA Project Test 6-1 (SB-PV-9 in JAEA) handled in the Large Scale Test Facility (LSTF) of the Japan Atomic Energy Agency (JAEA). Test 6-1 simulated a PWR pressure vessel Upper-Head SBLOCA with a break size equivalent to 1.9% of the cold leg break under the assumption of total failure of High Pressure Injection System (HPIS).The paper shows several analyses about the geometry variables (size, location, flow paths and Upper Head nodalization) which can influence on the pressure vessel Upper Head SBLOCA model performed using the thermal-hydraulic code TRACE5.© 2012 ASME

3 citations


References
More filters

01 Jan 1986
Abstract: Instrumentation needs for detection of inadequate core cooling. Studies of the Three Mile Island accident identified the need for additional instrumentation to detect inadequate core cooling (ICC) in nuclear power plants. Industry studies by plant owners and reactor vendors supported the conclusion that improvements were needed to help operators diagnose the approach to or existence of ICC as well as to provide more complete information for operator control of safety injection flow to minimize the consequences of such an accident. In 1980, the US Nuclear Regulatory Commission (NRC) required further studies by the industry and described ICC instrumentation design requirements that included human factors and environmental considerations. On December 10, 1982, NRC issued to Babcock and Wilcox (B and W) licensees orders for Modification of License and transmitted to pressurized water reactor licensees Generic Letter 82-28 to inform them of the revised NRC requirements. The instrumentation requirements include upgraded subcooling margin monitors (SMM), upgraded core exit thermocouples (CET), and installation of a reactor coolant inventory tracking system. NRC Regulatory Guide 1.97, which covers accident monitoring instrumentation, was revised (Rev. 3) to be consistent with the requirements of item II.F.2 of NUREG-0737.

4 citations


"Post-implementation review of inade..." refers background in this paper

  • ...Some of those problems and the current status of industry efforts to achieve a reliable and unambiguous indication of water level were discussed recently in another paper.(8) The possible effects of level measurement uncertainty for a typical small-break transient are shown in Fig....

    [...]


Book ChapterDOI
01 Jan 1984
Abstract: Historically, light water cooled reactors (LWRs) have heen designed with reliance on automatic protection systems to assure that the reactor is shut down and adequate coolant is provided to the core to prevent severe core damage under all design basis accident conditions. A combination of high pressure and low pressure safety injection systems are designed to replenish coolant lost from the reactor coolant system (RCS) for the entire spectrum of potential breaks in the reactor coolant pressure boundary. Design philosophy emphasized that protection systems should be redundant and highly reliable and should require minimal interaction with the plant operators. The latter consideration was based on a dominant concern for the large break loss of coolant accident in which the sequence of events is too fast to permit reliable diagnosis and response by the operator. Consequently, design attention given to accident monitoring instrumentation was generally limited to those instruments required to help the operator assess the effectiveness of long term core cooling in the recirculation mode after termination of an accident transient.

3 citations