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
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employees, makes any warranty, express or implied, or assumes any legal liability or responsi-
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ence herein to any specific commercial product, process, or service by trade name, trademark,
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mendation, or favoring by the United States Government or any agency thereof. The views
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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.
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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).