[Federal Register Volume 78, Number 177 (Thursday, September 12, 2013)]
[Proposed Rules]
[Pages 56174-56182]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2013-22234]
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Proposed Rules
Federal Register
________________________________________________________________________
This section of the FEDERAL REGISTER contains notices to the public of
the proposed issuance of rules and regulations. The purpose of these
notices is to give interested persons an opportunity to participate in
the rule making prior to the adoption of the final rules.
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Federal Register / Vol. 78, No. 177 / Thursday, September 12, 2013 /
Proposed Rules
[[Page 56174]]
NUCLEAR REGULATORY COMMISSION
10 CFR Parts 50 and 52
[Docket No. PRM-50-105; NRC-2012-0056]
In-Core Thermocouples at Different Elevations and Radial
Positions in Reactor Core
AGENCY: Nuclear Regulatory Commission.
ACTION: Petition for rulemaking; denial.
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SUMMARY: The U.S. Nuclear Regulatory Commission (NRC) is denying a
petition for rulemaking (PRM), PRM-50-105, submitted by Mark Leyse (the
petitioner) on February 28, 2012. The petitioner requested that the NRC
require all holders of operating licenses for nuclear power plants
(NPPs) to operate NPPs with in-core thermocouples at different
elevations and radial positions throughout the reactor core to enable
the operators to accurately measure a large range of in-core
temperatures in NPP steady-state and transient conditions. The NRC is
denying the PRM because: there are no protection or plant control
functions that utilize inputs from core exit thermocouples (CETs);
there is no operational necessity for more accurate measurement of
temperatures throughout the core; the petition provided inadequate
justification of why precise knowledge of core temperature at various
elevations and radial positions would enhance safety or change operator
action; and the NRC believes that, despite the known limitations of
CETs, CETs are sufficient to allow NPP operators to take timely and
effective action in the event of an accident.
DATES: The docket for the petition for rulemaking, PRM-50-105, is
closed on September 12, 2013.
ADDRESSES: Please refer to Docket ID NRC-2012-0056 when contacting the
NRC about the availability of information for this petition. You may
access information related to this petition by any of the following
methods:
Federal Rulemaking Web site: Go to http://www.regulations.gov and search on Docket ID NRC-2012-0056. Address
questions about NRC dockets to Carol Gallagher, telephone: 301-492-
3668; email: [email protected].
The NRC's Agencywide Documents Access and Management
System (ADAMS): You may access publicly available documents online in
the NRC Library at http://www.nrc.gov/reading-rm/adams.html. To begin
the search, select ``ADAMS Public Documents'' and then select ``Begin
Web-based ADAMS Search.'' For problems with ADAMS, please contact the
NRC's Public Document Room (PDR) reference staff at 1-800-397-4209,
301-415-4737, or by email to [email protected]. The ADAMS Accession
Number for each document referenced in this document (if that document
is available in ADAMS) is provided the first time that a document is
referenced. In addition, for the convenience of the reader, the ADAMS
Accession Numbers are provided in a table in Section V, ``Availability
of Documents,'' of this document.
The NRC's PDR: You may examine and purchase copies of
public documents at the NRC's PDR, O1-F21, One White Flint North, 11555
Rockville Pike, Rockville, Maryland 20852.
FOR FURTHER INFORMATION CONTACT: Tara Inverso, Office of Nuclear
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, telephone: 301-415-1024; email: [email protected].
SUPPLEMENTARY INFORMATION:
I. Background
II. NRC Technical Evaluation
III. Public Comments on the Petition
IV. Ongoing NRC Activities Related to Reactor and Containment
Instrumentation
V. Availability of Documents
VI. Determination of the Petition
I. Background
The NRC received a petition for rulemaking (ADAMS Accession No.
ML12065A215) on February 28, 2012, and assigned it Docket No. PRM-50-
105. The NRC published a notice of receipt and request for public
comment in the Federal Register (FR) on May 23, 2012 (77 FR 30435).
The petitioner requested that the NRC amend its regulations in Part
50 of Title 10 of the Code of Federal Regulations (10 CFR), ``Domestic
Licensing of Production and Utilization Facilities,'' to require all
holders of operating licenses for NPPs to operate NPPs with in-core
thermocouples at different elevations and radial positions throughout
the reactor core to enable NPP operators to accurately measure a large
range of in-core temperatures in NPP steady-state and transient
conditions. The petitioner asserted that, in the event of a severe
accident, in-core thermocouples would provide NPP operators with
crucial information to help operators manage the accident. In support
of the petition, the petitioner cited several reports and findings,
including the Report of the President's Commission on the Accident at
Three Mile Island [TMI]: ``The Need for Change: The Legacy of TMI,''
dated October 1979. The petitioner asserted that ``[i]n the last three
decades, NRC has not made a regulation requiring that NPPs operate with
in-core thermocouples at different elevations and radial positions
throughout the reactor core to enable NPP operators to accurately
measure a large range of in-core temperatures in NPP steady-state and
transient conditions, which would help fulfill the President's
Commission recommendations.'' The petitioner further stated that, if
another severe accident were to occur in the United States, NPP
operators would not know what the in-core temperatures would be during
the progression of the accident, and concluded that, in a severe
accident, core-exit thermocouples would be the primary tool used to
detect inadequate core cooling and core uncovery.
II. NRC Technical Evaluation
The petitioner requested that the NRC require in-core thermocouples
be installed in all NPPs; this would include both pressurized water
reactors (PWRs) and boiling water reactors (BWRs). However, BWRs do not
use CETs, and thermocouple response in BWR applications is not
currently known. Furthermore, the experiments referenced throughout the
PRM studied only PWRs. Because the issues and arguments raised in the
PRM do not apply to BWRs, and because the PRM does not list any
limitations on BWR
[[Page 56175]]
instrumentation, there is no basis provided to evaluate this PRM for
BWRs. Therefore, the NRC is evaluating this PRM as it pertains to PWRs
only.
During normal operation in a PWR, reactor coolant system (RCS) hot
leg and cold leg temperatures are the primary indications of core
condition. Measurements of RCS hot and cold leg temperatures from
safety-related instrumentation provide the necessary input to a plant's
reactor protection system. There are no reactor protection or plant
control functions that use inputs from the CETs. Additionally, the CETs
are not the only source of information relied on to initiate reactor
operator responses to accident conditions. The uses of CETs will be
described in more detail, as part of the NRC's evaluation of the issues
raised in the PRM with respect to the use of CETs.
PRM Issue 1: Core Exit Thermocouple Limitations
The petitioner stated that, ``in a severe accident, in many cases,
a predetermined core exit temperature measurement (e.g.,
1200[emsp14][deg]F) would be used to signal the time for NPP operators
to transition from EOPs [Emergency Operating Procedures] to
implementing SAMGs [Severe Accident Management Guidelines].'' However,
experimental data indicates that CET measurements have significant
limitations. A report \1\ prepared by the Organization for Economic
Cooperation and Development (OECD) Nuclear Energy Agency (NEA),
Committee on the Safety of Nuclear Installations, entitled, ``Core Exit
Temperature (CET) \2\ Effectiveness in Accident Management of Nuclear
Power Reactor,'' dated November 26, 2010, concluded:
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\1\ Available at http://www.oecd-nea.org/nsd/docs/2010/csni-r2010-9.pdf.
\2\ Note that the OECD report uses the acronym CET to refer to
core exit temperature, but the NRC uses the acronym CET to refer to
core exit thermocouples in this document.
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The use of CET measurements has limitations in detecting
inadequate core cooling and core uncovery,
The CET indication displays in all cases a significant
delay (up to several hundred [seconds]), and
The CET reading is always significantly lower (up to
several 100 [Kelvin]) than the actual maximum cladding temperature.
The petition asserted that the NRC and the nuclear industry have
ignored experimental data indicating that CET measurements have
significant limitations. The results of four tests performed in the
loss-of-fluid test (LOFT) facility show that: 1) There was a delay
between the core uncovery and the thermocouple response, and 2) the
measured core exit thermocouple response was several hundred Kelvin
lower than the maximum cladding temperatures in the core. The
petitioner cited NUREG/CR-3386, ``Detection of Inadequate Core Cooling
with Core Exit Thermocouples: LOFT PWR Experience,'' dated November
1983 (ADAMS Accession No. ML13032A566), which states: ``There may be
accident scenarios in which these [thermocouples] would not detect
inadequate core cooling that preceded core damage.''
The NRC reviewed PRM Issue 1 and acknowledges that the CET
limitations cited by the petitioner are extensively documented in test
reports from the identified experimental programs. However, while these
test programs were conducted at large-scale test facilities
appropriately scaled (using a power to volume relationship) to produce
thermal-hydraulic phenomena similar to phenomena that could occur in a
commercial PWR, the scaling distortions introduced by the facilities
and the effects of plant-specific CET installation methods preclude the
direct extrapolation of the test results to reactor scale. In fact, the
same OECD report referenced by the petitioner also states:
Qualitative application/extrapolation of the CET response to
reactor scale is possible. However, direct extrapolation in
quantitative terms to the reactor scale should be avoided in general
or done with special care due to limitations of the experimental
facilities in terms of geometrical details, unavoidable distortion
in the scaling of the overall geometry, and of the heat capacity of
structures.
The NRC views these results within the context of their
applicability to full-scale plants in order to use the data to assess
the capability of the computer models used to perform full-plant
simulations. The separate test facilities, such as LOFT and
Primarkreislauf Test Facility Project (PKL), are simulated using
computer models, and the results from the simulations are compared with
the corresponding data. Once sufficient agreement between the
simulation and the data is achieved, or consistent biases are
determined, a full-plant simulation can be performed and more
definitive, quantitative statements about CET performance can be made.
Therefore, these experimental results cannot be, and are not intended
to be, quantitatively extrapolated to full-scale plants, as suggested
in the petition.
During normal operation, RCS hot leg and cold leg temperatures are
the primary indications of core condition. Measurements of RCS hot and
cold leg temperatures from safety-related instrumentation provide the
necessary input to a plant's reactor protection system. There are no
reactor protection or plant control functions that use inputs from the
CETs.
During accident conditions, the most significant functions provided
by CETs are the determination of a trend in RCS sub-cooling and the
known correlation of the indicated temperature to general core
conditions for the purposes of identifying the onset of core damage
(i.e., a severe accident). For these purposes, the CETs provide the
indication necessary to make operational decisions with respect to core
damage and perform these essential functions within the expected useful
range. In the initial stages of an accident, CETs provide accurate
indication of core temperatures for the purposes of determining sub-
cooling margin when forced circulation has been lost and confirming
that the core remains covered. As an event progresses, CETs provide an
indication of initial stages of core damage and are generally used as
an entry condition and diagnostic tool during implementation of SAMGs.
Upon entry into the SAMGs, core exit temperature is used as one
indication in a diagnostic process to determine core damage; other
indications include: RCS level, RCS pressure, containment pressure,
containment hydrogen concentration, nuclear instrumentation, and
containment high range radiation monitors. As CET readings rise above
1200 [deg]F, it becomes likely that the temperature for some sections
of cladding will have exceeded 1800 [deg]F, and therefore it can be
assumed that core damage has commenced. With this determination,
actions to restore key safety functions will continue in order to
restore core cooling and to ensure that fission product barriers remain
intact. At no point, either during diagnosis or follow-on actions to
restore core cooling, is there an operational necessity for an exact
measurement of core temperatures at various locations throughout the
core. The petitioner did not provide explicit examples where knowing
more precise temperatures would result in more effective operator
action. Further, the NRC's evaluation of this petition and relevant
information did not reveal added insights on how knowing precise in-
core temperatures would result in more effective operator action in a
core damage sequence. The correlation between CET readings and fuel
cladding temperature, in conjunction with other indications, is
sufficient for determining the onset of
[[Page 56176]]
fuel damage and the need for operator action. Actions taken to restore
core cooling would not depend upon a precise measurement of in-core
temperature. As the accident progresses, core vessel breach
determination is primarily made by utilizing containment pressure and
containment radiation indications, and nuclear instrumentation. Core
exit thermocouple indications are not used for this determination.
After considering the functions and indications provided by CETs in
normal and accident conditions, the NRC determined that the CETs
provide adequate indications for their intended purpose.
PRM Issue 2: Nuclear Power Plant Operators' Use of In-Core
Thermocouples
The petition asserted that, in the event of a severe accident, in-
core thermocouples would enable NPP operators to accurately measure in-
core temperatures better than CETs, and would provide crucial
information to help operators manage the accident; one example is an
indication that it is time to transition from EOPs to implementing
SAMGs. Therefore, the petition requested that all holders of operating
licenses for NPPs operate NPPs with in-core thermocouples at different
elevations and radial positions throughout the reactor core to enable
NPP operators to accurately measure a large range of in-core
temperatures in NPP steady-state and transient conditions.
As previously stated BWRs do not use CETs, and thermocouple
response in BWR applications is not currently known. Furthermore, the
experiments referenced throughout the PRM studied only PWRs. Therefore,
the NRC is evaluating this PRM as it pertains to PWRs only. The NRC
further notes that, in BWRs, saturation conditions exist within the
reactor vessel and fuel temperatures are closely related to the
saturation pressure. Under accident conditions, reactor vessel water
level is the best indication of conditions relating to imminent core
damage and drywell radiation monitors are typically the primary method
for determining core damage and SAMG entry conditions. For BWRs, SAMG
entry conditions are also tied to parameters such as water level,
containment hydrogen concentration, and component failures. With regard
to PWRs, CETs are located at various radial positions. Therefore, the
intent of the petitioner's request to account for various radial
temperatures is addressed by the current design.
The petition does not specify any benefit the data from in-core
thermocouples could provide or how that benefit would be greater than
that provided by core exit thermocouples. As discussed earlier, the
limitations of CETs are already well understood and accounted for in
existing SAMGs. The benefit provided by CETs, even in recognition of
their limitations, is discussed in greater detail in the NRC response
to PRM Issue 1. Furthermore, the petitioner cited no actions that would
be driven by the additional information obtained from in-core
thermocouples.
It is also important to note that the same OECD document referenced
by the petitioner contains additional information that provides a
perspective that is different from that of the petitioner. For example,
from page 48 of the report:
The conduct of the experiment was rather complicated with
repeated openings of two blowdown lines. The timeline for the
experiment was thus not very representative of a real accident. . .
. Measured cladding temperatures exceeded 2100 K . . . The
temperatures were in excess of 2100K for several minutes and the
peak temperatures were probably several hundred degrees higher than
that. Material examinations showed material formations consistent
with temperatures in the range of 2800 K and in local areas over
3000 K.
``An Account of the OECD LOFT Project'' of this experiment (LP-FP-
2) \3\ additionally states on page 53:
\3\ Available at http://www.oecd-nea.org/nsd/reports/OECD_LOFT_final_report_T3907_May1990.pdf.
Thermocouples used in the CFM [Center Fuel Module] were
calibrated as high as 2100 K. However, many of the CFM temperature
measurements were affected by thermocouple cable shunting effects
[formation of a new thermocouple junction due to exposure to high
temperature] before the temperature at the thermocouple location
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reached 2100 K.
These statements indicate that in-core thermocouples may not be any
more accurate than, or as reliable as, the core exit thermocouples
currently used in PWRs, and that they may be subject to additional
limitations. It is impractical to mount thermocouples to the fuel
cladding surface or fuel spacers. Reactor vessel head modifications
would be necessary, as well as the addition of a significant amount of
instrumentation wiring and support structures. Furthermore, the
addition of in-core thermocouples and the associated supporting
components would likely result in significant adverse effects on fluid
flow in the core. For instance, fin effects would disturb temperature
profiles within the core, and could create calibration difficulties. In
addition, installing in-core thermocouples could increase loose parts
potential, independence and separation issues, and seismic
considerations.
While the previous discussion applies to fuel-cladding-surface-
mounted thermocouples, the NRC also considered the petitioner's request
as it may relate to a requirement to install thermocouples in bulk
coolant areas within the fuel matrix, such as within instrument tubes.
Extensive research has been performed to characterize the relationship
between liquid and vapor temperatures and heat transfer rates in the
dispersed flow regime expected within the core during severe accident
conditions. Significant temperature differences can exist between the
bulk coolant, which would contain droplets of liquid water at
saturation conditions, and the fuel cladding surface. R.S. Dougall and
W.M. Rohsenow, for instance, characterized surface temperatures that
exceeded saturation temperatures by 400 to 700 degrees Fahrenheit in
their experimental work.\4\ Subsequent work has validated Dougall's and
Rohsenow's findings. Because of the significant temperature differences
that can exist within the post-accident core region, thermocouples
located within the instrument tubes would provide information that
offers no greater benefit than that provided by the CETs.
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\4\ R.S. Dougall and W.M. Rohsenow, ``Film Boiling on the Inside
of Vertical Tubes with Upward Flow of the Fluid at Low Qualities,''
1963, available at http://hdl.handle.net/1721.1/62142.
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For these reasons, the NRC determined that, for operating PWRs, in-
core thermocouples are not necessary, nor would they help operators
manage an accident. In addition to these reasons, the NRC notes that
the installation and maintenance associated with in-core thermocouples
would result in higher doses to plant workers, with no added safety
benefit.
The petition requested that the requirement for in-core
thermocouples be applied to ``all holders of operating licenses for
[nuclear power plants].'' The NRC interprets this request as applying
to both current and future holders of operating licenses under 10 CFR
Part 50, as well as current and future holders of combined licenses
under 10 CFR Part 52. The NRC believes that this is a reasonable
interpretation, inasmuch as combined licenses under 10 CFR Part 52
combine the authority provided under a construction permit and an
operating license (albeit with
[[Page 56177]]
certain conditions and restrictions as set forth in 10 CFR Part 52,
Subpart C \5\) into one license. In addition, because the two existing
combined licenses reference the AP1000 design certification rule (10
CFR Part 52, Appendix D), which controls the design of the reactor
instrumentation, including the placement of thermocouples, the NRC
interprets the petition as a request to amend the AP1000 design
certification rule.
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\5\ The conditions and limitations of a combined license issued
under 10 CFR Part 52 are consistent with, and are intended to comply
with, the statutory requirements for combined licenses in Section
185b of the Atomic Energy Act of 1954, as amended.
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Because the core of the AP1000 design is similar to the PWRs
described throughout this document, the NRC's evaluation of, and
determination on, this PRM with respect to PWRs also applies to the
AP1000 design and no changes to the AP1000 design are necessary.
PRM Issue 3: Post-Three Mile Island Accident Actions
The petition included a citation from an October 1979
recommendation from the President's Commission on the Three Mile Island
Accident, which stated:
Equipment should be reviewed from the point of view of providing
information to operators to help them prevent accidents and to cope
with accidents when they occur. Included might be instruments that
can provide proper warning and diagnostic information; for example,
the measurement of the full range of temperatures within the reactor
vessel under normal and abnormal conditions.
The petitioner asserted that the NRC has not made a regulation
requiring NPPs to operate with in-core thermocouples at different
elevations and radial positions throughout the reactor core to enable
NPP operators to accurately measure a large range of in-core
temperatures in NPP steady-state and transient conditions, which the
petitioner avows would help fulfill the President's Commission's
recommendations. The petitioner further asserted that if another severe
accident were to occur in the United States, NPP operators would not
know what the in-core temperatures were during the progression of the
accident.
Following the accident at TMI, the NRC ordered a broad range of
safety enhancements at U.S. NPPs. These enhancements include sub-cooled
margin monitors, post-accident monitoring instrumentation systems
(including CET indications available to operators), and the reactor
vessel level monitoring system. These enhancements, combined with other
post-TMI requirements for enhanced EOPs and operator training, form
part of the Agency's response to the recommendation of the President's
Commission on the Three Mile Island Accident.
Regarding the President's Commission's example of ``measurement of
the full range of temperatures within the reactor vessel under normal
and abnormal conditions,'' evidence of the NRC's consideration of in-
core thermocouples may be found in NUREG-0737, ``Clarification of TMI
Action Plan Requirements'' (ADAMS Accession No. ML051400209), Section
II.F.2, ``Instrumentation for Detection of Inadequate Core Cooling
(ICC).'' Item (6) on page 3-114 under Clarifications states:
The indication must cover the full range from normal operation
to complete core uncovery. For example, water-level instrumentation
may be chosen to provide advanced warning of two-phase level drop to
the top of the core and could be supplemented by other indicators
such as incore and core-exit thermocouples provided that the
indicated temperatures can be correlated to provide indication of
the existence of ICC [inadequate core cooling] and to infer the
extent of core uncovery. Alternatively, full-range level
instrumentation to the bottom of the core may be employed in
conjunction with other diverse indicators such as core-exit
thermocouples to preclude misinterpretation due to any inherent
deficiencies or inaccuracies in the measurement system selected.
The alternative noted in this excerpt, to use full-range level
indication combined with core exit thermocouples, was ultimately the
preferred option. Part of the consideration to use the alternative may
be found in the NRC's stated position on ICC that requires unambiguous,
easy-to-interpret indication of ICC. The NRC chose to use process
variables that map directly to clear, easy-to-interpret emergency
operating procedures to elicit safe and consistent operator responses
to accident scenarios.
PRM Issue 4: Consideration of Experimental Data
The petitioner asserted that the NRC and Westinghouse do not
consider that experimental data at four facilities (LOFT, PKL, Rig of
Safety Assessment Large-Scale Test Facility (ROSA/LSTF), and OECD/NEA
computer codes validation project (PSB-VVER)) indicate that CET
measurements would not be an adequate indicator for when to transition
from EOPs to implementing SAMGs in a severe accident. The petition
listed 13 conclusions from the OECD report that are common to the
evaluation of the tests in all four facilities summarized by that
report:
``The use of CET measurements has limitations in detecting
inadequate core cooling and core uncovery;''
``The CET indication displays in all cases a significant
delay (up to several 100 [seconds]);''
``The CET reading is always significantly lower (up to
several 100 [Kelvin]) than the actual maximum cladding temperature;''
``CET performance strongly depends on the accident
scenarios and the flow conditions in the core;''
``The CET reading depends on water fall-back from the
upper plenum (due to: e.g., reflux condensing [steam generator] mode or
water injection) and radial core power profiles. During significant
water fall-back the heat-up of the CET sensor could even be
prevented;''
``The colder upper part of the core and the cold
structures above the core are contributing to the temperature
difference between the maximum temperature in the core and the CET
reading;''
``The steam velocity through the bundle is a significant
parameter affecting CET performance;''
``Low steam velocities during core boil-off are typical
for [small-break loss-of-coolant accident] transients and can advance
3D flow effects;''
``In the core as well as above (i.e., at the CET
measurement level) a radial temperature profile is always measured
(e.g., due to radial core power distribution and additional effects of
core barrel and heat losses);''
``Also at low pressure (i.e., shut down conditions)
pronounced delays and temperature differences are measured, which
become more important with faster core uncovery and colder upper
structures;''
``Despite the delay and the temperature difference the CET
reading in the center reflects the cooling conditions in the core;''
``Any kind of [accident management] procedures using the
CET indication should consider the time delay and the temperature
difference of the CET behavior;'' and
``In due time after adequate core cooling is re-
established in the core the CET corresponds to no more than the
saturation temperature.''
Finally, the petitioner continued to reference the OECD report,
stating that, during the LOFT LP-FP-2 experiment when maximum core
temperatures were measured to exceed 3300 [deg]F, CETs were
[[Page 56178]]
typically measured at 800 [deg]F (more than 2500 [deg]F lower than the
maximum core temperatures). He provided that ``during the rapid
oxidation phase the CET appeared essentially to be disconnected from
core temperatures.''
The NRC and the industry have long acknowledged the limitations of
CETs, but conclude that the use of CETs remains appropriate and would
help operators to manage an accident. This awareness is documented in
several reports, such as ``Limitations of Detecting Inadequate Core
Cooling'' (U.S. Department of Energy's Office of Scientific and
Technical Information ID 6797561) published in 1984 and WCAP-14696-A,
Revision 1, ``Westinghouse Owners Group Core Damage Assessment
Guidance,'' dated July 1996 (ADAMS Accession No. ML993490267). The
delayed indication would not necessarily be a concern during a severe
accident. First, the NPP staff relies on other indications to diagnose
conditions, such as the reactor vessel level instrumentation system,
hot-leg resistance temperature detectors, and containment hydrogen and
radiation monitors. Second, whereas the CET indication delay may be up
to a few minutes, post-accident operator actions are determined and
implemented on a scale that exceeds several minutes. On this time
scale, the noted time delay is acceptable.
The petition cited a number of conclusions about CET deficiencies
that were noted in the OECD report, and cited on page 8 of the PRM, but
the PRM did not specifically acknowledge the following statement from
page 129 of the OECD report: ``Despite the delay and the temperature
difference the CET reading in the center reflects the cooling
conditions in the core.'' It is the NRC's position that scaling
challenges, described earlier in this document, exist when
extrapolating the results to a full-scale NPP, and these challenges
tend to exacerbate the extent of the CET deficiencies cited in the
experimental results. Therefore, while the noted deficiencies should be
considered qualitatively, overall, in terms of plant applicability, the
CETs performed the intended function, as described in the NRC's
response to PRM Issue 2.
III. Public Comments on the Petition
The NRC received three public comment submissions on the PRM, one
each from the following: the Nuclear Energy Institute (NEI), Exelon
Generation Company, and the petitioner. In addition to those
submissions, the NRC received a late-filed comment submission from the
petitioner in response to the NEI comment submission. The late-filed
comment submission, submitted by the PRM-50-105 petitioner, contains
some reiteration of information and assertions in PRM-50-105. The NRC
is not addressing those portions of the late-filed comment response.
However, the late-filed comment submission also discussed matters
related to the use of in-core thermocouples in gamma thermometers, the
use of in-core thermocouples in the Economic Simplified Boiling Water
Reactor (ESBWR) design, and the radiation dose to workers due to in-
core thermocouples; these issues were not raised in the original PRM.
Therefore, the NRC is addressing these three new matters in this
comment response section.
The comments are grouped into four comment categories: General
Discussion of PRM-50-105, Comments on In-Core Thermocouples, Comments
Related to Westinghouse AP1000, and Comments on Experimental Data. A
comment identifier (e.g., NEI-1) follows each comment summary. The
comments and the associated NRC responses follow.
General Discussion of PRM-50-105
Comment: The NRC should not amend its regulations to require all
holders of operating licenses to operate nuclear power plants with in-
core thermocouples at different elevations and radial positions
throughout the reactor core. (NEI-1)
NRC Response: The NRC agrees with this comment. The NRC is denying
PRM-50-105 for the reasons set forth in this document.
Comments on In-Core Thermocouples
Comment: Use of in-core thermocouples would result in higher doses
to workers both to implement plant modifications and to maintain the
proposed system with minimum if any benefit to plant safety. (NEI-2)
NRC Response: The NRC agrees with the comment, but notes that the
comment did not provide any basis for this assertion.
Comment: In response to another commenter's statement that in-core
thermocouples would result in a higher radiation dose to workers both
to implement plant modifications and to maintain the proposed system
with minimum, if any, benefit to plant safety, one commenter provided
the following quote from General Electric Hitachi (GEH) Nuclear Energy:
``A [gamma thermometer] system has no moving parts, no under vessel
tubing, virtually no radiation dose to maintenance since it is a fixed
in-core probe, and is expected to be very reliable.'' \6\ The commenter
asserts that in-core thermocouples could be placed inside instrument
tubes, distributed through the reactor core, like gamma thermometers
are, and thus cause virtually no radiation dose to workers during
maintenance. (Leyse2-5)
---------------------------------------------------------------------------
\6\ GE Hitachi Nuclear Energy, ``Licensing Topical Report: Gamma
Thermometer System for [Local Power Range Monitor] LPRM Calibration
and Power Shape Monitoring,'' NEDO-33197-A, p. 1 (available at ADAMS
Accession No. ML102810320).
---------------------------------------------------------------------------
NRC Response: The NRC disagrees with the comment that in-core
thermocouples would cause virtually no radiation dose to workers during
maintenance. The NRC notes that the GEH report, referenced by the PRM
as support for the comment, applies only to a comparison of the current
BWR moveable and retractable probe (the TIP system) with the ESBWR
fixed incore gamma thermometers. It does not apply to the installation
of in-core thermocouples in currently operating reactors. The NRC
agrees that the use of fixed versus bottom entry retractable sensors
may reduce exposure for routine maintenance. The NRC continues to
believe that in-core thermocouples would result in a higher radiation
dose to workers while implementing the necessary plant modifications
for installation and to maintain the proposed system, particularly when
replacement of sensor strings due to long-term radiation exposure is
required. Also, except for existing tubing for bottom-entry removable
sensors, any existing instrument tubes are already occupied. It is
likely that new instrument tubes would need to be installed. Tubes
installed through the vessel head would also require provisions for
mechanical and electrical connections. These installation efforts,
whether the new tubing enters the core through the vessel head or
bottom, are likely to require significant worker exposure, and may also
raise concerns related to pressure boundary integrity.
Comment: In some designs, in-core thermocouples could be more
susceptible to failures and misdiagnosis than CETs because of proximity
to thermal and radiation sources. It is not feasible to attach
thermocouples directly to the fuel cladding. Thermocouples would need
to be located in existing instrument tubes (e.g., BWR Local Power Range
Monitor tubes) and would not be in direct contact with the reactor
coolant. Therefore, thermocouples would provide only indirect readings
of fuel temperature and would be subject to heat transfer delays/
response times. The time response and accuracy of the reading as it
relates to the reactor
[[Page 56179]]
coolant would be highly questionable. The presence of the fuel channel
on a BWR fuel assembly would further inhibit and interfere with the
readings of a thermocouple in an instrument tube. (NEI-3) (Exelon-2)
NRC Response: The NRC acknowledges that in-core thermocouples could
be more susceptible to failure and misdiagnosis in some designs.
However, as stated throughout this document, because CETs perform their
desired functions and because precise knowledge of in-core temperatures
would not change operator actions, further consideration of the
potential limitations of in-core thermocouples is not necessary.
Comment: In response to another commenter's assertion that in-core
thermocouples may be more susceptible to failures and misdiagnosis than
CETs, one commenter stated that in-core thermocouples have been tested
and used in nuclear reactors for decades as the primary component of
in-core gamma thermometers (devices that measure gamma flux in nuclear
reactors). Radcal gamma thermometers were installed in PWRs in the
1980s. Radcal thermometers are also installed in BWRs. General Electric
Hitachi Nuclear Energy has plans to use in-core thermocouples in gamma
thermometers in the ESBWR design. (Leyse2-1) (Leyse2-2) (Leyse2-4)
NRC Response: The NRC continues to believe that CETs are acceptable
for use in current applications. Where current nuclear power plants
have fixed in-core gamma thermometers, they are for power shape
monitoring and calibration, not for actual temperature measurements.
Further, the gamma thermometer GEH plans to install in the ESBWR is a
device for measuring the gamma flux for the purpose of calibration of
the local power range monitors and power shape monitoring; the gamma
thermometers are not for the purpose of measuring axial and radial core
temperature. The GEH gamma thermometers utilize a local differential
temperature directly within the sensor at the specific sensor location
to infer the gamma flux inside the reactor core rather than the actual
temperature measurements at that location. Actual temperature
measurements are not available outside the reactor core. For these
reasons, the information about the use of gamma thermometers at nuclear
power reactors and in the ESBWR design certification do not affect the
NRC's position that CETs are acceptable for use in current applications
to perform their specified function.
Comment: An Idaho National Laboratory (INL) report stated that INL
``developed and evaluated the performance of a high temperature
resistant thermocouple that contains doped molybdenum and a niobium
alloy. Data from high temperature (up to 1500 [deg]C), long duration
(up to 4000 hours) tests and on-going irradiations at INL's Advanced
Test Reactor demonstrate the superiority of these sensors to
commercially-available thermocouples. However, several options have
been identified that could further enhance their reliability, reduce
their production costs, and allow their use in a wider range of
operating conditions.'' \7\ (Leyse2-3)
---------------------------------------------------------------------------
\7\ Joshua Daw, et al., Idaho National Laboratory, ``High
Temperature Irradiation-Resistant Thermocouple Performance
Improvements,'' INL/CON-09-15267, Sixth American Nuclear Society
International Topical Meeting on Nuclear Plant Instrumentation,
Control, and Human-Machine Interface Technologies, April 2009, p. 1
(available at http://www.inl.gov/technicalpublications/documents/4235634.pdf).
---------------------------------------------------------------------------
NRC Response: The information in the comment is not relevant to the
PRM, and therefore does not change the NRC's position that CETs are
acceptable for use in performing their specified function, thereby
obviating the need to install in-core thermocouples. The NRC also notes
that the pre-publication INL report dated 2009 referenced by the
commenter described a research product that is not yet ready for
commercial use by the nuclear industry. The NRC does not believe that
the statements in the report that are referenced in the comment are
relevant to the acceptability of CETs in current applications.
Comment: The transition from EOPs to SAMGs based on existing plant
parameters is adequate. Pressurized Water Reactors already use CETs to
make the transition to SAMGs. The potential delay in the response of
indirectly reading in-core thermocouples could actually be longer than
the response of other plant parameters, including CETs, in identifying
potential severe accident conditions. (Exelon-3)
NRC Response: The NRC agrees that the current transition from EOPs
to SAMGs is adequate. The NRC notes that SAMGs are developed based on
the recognition that CETs could differ from actual core temperatures.
This concept is described in Section II, ``NRC Technical Evaluation,''
of this document.
Comment: During steady-state operations for both PWRs and BWRs, the
fuel cladding (surface) temperature is a function of coolant
Temperature--Enthalpy (T-H) properties. The coolant steady-state
properties (i.e., temperature) do not vary significantly axially or
radially during steady-state operation and therefore, in-core
thermocouples would not provide useful information. There are more
accurate means of measuring core conditions than in-core thermocouples
already in place. Adding in-core thermocouples would not improve the
ability or accuracy of measuring core conditions. (Exelon-1)
NRC Response: The NRC agrees with the comment. The PWR in-core
conditions, for example, are measured using hot and cold leg
temperatures, reactor coolant pressure, and neutron flux. These
parameters are then used as inputs to the reactor protection system to
ensure that the reactor shuts down if core operating conditions deviate
significantly from the expected normal operating conditions. The BWRs
are equipped with similar equipment intended for monitoring normal,
steady-state operation. The addition of in-core thermocouples, either
to measure fuel surface or reactor coolant temperatures, would add
little value to the information already available for monitoring normal
operation.
Comment: The petitioner asserted that, in the event of a severe
accident, in-core thermocouples would provide nuclear power plant
operators with ``crucial information to help operators manage the
accident.'' However, the petitioner provided no basis that actions
taken by operators would be more effective than actions based on
existing CETs. Operators are trained to recognize off-normal operating
conditions that have potential for resulting in core damage and to
maneuver the plant to a more conservative operating envelope (i.e.,
provide coolant to the reactor core). In a severe accident, operator
strategies control parameters across large regions of the core or
across the entire core. The additional information regarding local fuel
temperature provided by in-core thermocouples would not be crucial
relative to restoring coolant, nor would it change the steps and
actions available to operators to maintain or restore adequate core
cooling conditions. There is no evidence to show that temperatures
sensed at a single location could be used more effectively than actions
based on CET temperatures. (Exelon-4) (NEI-4) (NEI-6)
NRC Response: The NRC agrees with the comment. Precise measurement
of local fuel temperatures at distinct locations throughout the core
would not provide essential data for informing severe accident
management decisions, and the petitioner cited no actions that would be
driven by the additional information obtained from in-core
[[Page 56180]]
thermocouples. In the event of an extended loss of core cooling that
leads to core damage, the actions taken by the operators will be
focused on restoring core cooling, with or without the knowledge of
precise fuel temperatures in the core.
Comments Related to Westinghouse AP1000
Comment: One commenter provided several comments on the emergency
response guidelines for Westinghouse's AP1000 design:
Westinghouse maintains that core exit gas temperature
would reach 1200 [deg]F in Time Frame 1, but the LOFT LP-FP-2
experiments show that core exit temperatures were measured at around
800 [deg]F when in-core thermocouples measured fuel cladding
temperatures exceeded 3300 [deg]F. Thus, after the onset of the rapid
zirconium-steam reaction, core exit temperatures were measured at
around 800 [deg]F. (Leyse-4)
There are problems with Westinghouse's emergency response
guidelines for the AP1000. Plant operators are instructed to actuate
the AP1000 containment hydrogen igniters after the CET measurements
exceeded 1200 [deg]F, which would most likely be some time after a
meltdown had commenced. (Leyse-6)
There are problems with Westinghouse's plan to have plant
operators rely on CET measurements in the event of a severe accident,
because plant operators might reflood an overheated core without
realizing that the core was in fact overheated. Consider a scenario
where there were similar temperature differences between in-core and
core exit temperatures as were observed in LOFT LP-FP-2. If plant
operators were to reflood the core when core exit temperatures were
well below 1200 [deg]F, the core could already be overheated (i.e.,
fuel-cladding temperatures could be over 3300 [deg]F), nearing the
temperature where zirconium melts. In such a case there would also be
some liquefaction of core components because of eutectic reactions
(i.e., the eutectic reaction between zirconium and stainless steel)
taking place at temperatures as low as 2200 [deg]F. Unintentionally
reflooding an overheated core could be very dangerous. In a severe
accident, during the reflooding of an overheated reactor core up to 300
kilograms of hydrogen could be generated in one minute. (Leyse-7)
It is evident that with Westinghouse's plan to have plant
operators rely on CET measurements in the event of a severe accident,
operators could unintentionally reflood an overheated core, which would
rapidly generate additional hydrogen, at a rate as high as 5.0
kilograms per second, which could, in turn, compromise the containment
if the hydrogen were to detonate. (Leyse-8)
For severe accidents, Westinghouse's plan for AP1000 plant
operators to rely on core exit temperature measurements to monitor the
condition of the core and to wait for a core exit temperature
measurement of 1200 [deg]F to signal when to actuate the hydrogen
igniters and implement other procedures would be neither productive nor
safe. (Leyse-10)
NRC Response: The NRC disagrees with the comments that the
Westinghouse emergency response guidelines for the AP1000 design are
inadequate, based upon CET limitations. As discussed throughout this
document, the CET limitations noted in both this comment and the PRM
are acknowledged by the NRC and have been documented in industry
reports. The CETs, even with their known limitations, are sufficient to
provide the necessary information to nuclear power plant operators.
More precise knowledge of in-core temperatures would not change the
operational decisions necessary in the event of a severe accident.
Therefore, the NRC does not believe that the comment provided
information supporting the PRM's request that nuclear power plant
licensees be required by rule to install in-core thermocouples.
To the extent that the comments raise issues with respect to the
adequacy of the AP1000 design and hydrogen control, the NRC regards
this portion of the comment to be outside the scope of the issues
raised in this PRM. The NRC notes, however, that these AP1000 issues
were raised in a 10 CFR 2.206 petition on Vogtle, Units 3 and 4 (ADAMS
Accession No. ML12061A218), and resolved as part of the NRC's action on
the petition. The NRC's resolution of the 10 CFR 2.206 petition is
available at ADAMS under Accession No. ML13105A308.
Comments on Experimental Data
Comment: The commenter cited the OECD Nuclear Energy Agency report,
which states: ``During the rapid oxidation phase [core exit
temperatures] appeared essentially to be disconnected from core
temperatures.'' (Leyse-5)
NRC Response: The following sentence appears in the same section of
the OECD report referenced by the commenter: ``For core runaway
conditions with rapid fuel oxidation, LOFT results indicated that the
CETs essentially were disconnected from the core temperatures. This is
perhaps a lesser problem since such conditions cannot be well addressed
by accident management measures.'' Currently, CET indications are used
to help determine core uncovery and initiate appropriate actions during
that phase of an accident. In following phases, core temperatures do
not provide information that is used to initiate actions to mitigate an
accident.
Comment: Two of the main conclusions from data from experiments
simulating design basis accidents conducted at four different
facilities are that core exit temperature measurements display in all
cases a significant delay (up to several hundred seconds) and that core
exit temperature measurements are always significantly lower (up to
several hundred degrees Celsius) than the actual maximum cladding
temperature. (Leyse-9)
NRC Response: The NRC agrees with this comment. The NRC was
directly involved in most of the experimentation referenced by the
petitioner, and the NRC and other nuclear industry stakeholders have
been aware for several years of the CET limitations concluded from the
experiments and verified by independent analyses. Evidence of this can
be seen in WCAP-14696-A, Revision 1 (November 1999; ADAMS Accession No.
ML993490267), which states that ``Analyses performed for the WOG
[Westinghouse Owners Group] ERGs [Emergency Response Guidelines] for
indication of inadequate core cooling concluded that the temperature
indicated by the core exit thermocouples, especially during transient
heat up conditions, is always several hundred degrees lower than the
fuel rod cladding temperatures.'' The NRC notes that SAMGs are
developed based on the recognition that CETs could differ from actual
core temperatures. This concept is described in Section II, ``NRC
Technical Evaluation,'' of this document.
Miscellaneous Comments
Comment: An April 2012 Advisory Committee on Reactor Safeguards
(ACRS) report states that the NRC ``has recognized the need for
enhanced reactors . . . instrumentation and is in the process of adding
this to the implementation of the NTTF [Near-Term Task Force]
recommendations.'' And the NTTF report ``recommends strengthening and
integrating onsite emergency response capabilities such as EOPs and
SAMGs.'' The April 2012 ACRS report states that ``such integration
could focus on the need to clarify the transition points'' that would
occur in a NPP accident. In-core
[[Page 56181]]
thermocouples would fulfill the need for enhanced reactor
instrumentation. In-core thermocouples would provide NPP operators with
crucial information to help them track the progression of core damage
and manage an accident; for example, indicating the correct time to
transition from EOPs to implementing SAMGs. (Leyse-1)
NRC Response: The NRC disagrees with this conclusion. As stated
previously in this document, at no point, neither during diagnosis nor
follow-on actions to restore cooling, is there an operational necessity
for an exact measurement of core temperatures at various locations
throughout the core. However, as noted in Enclosure 3 to SECY-12-0095,
``Tier 3 Program Plans and 6-month Status Update in Response to Lessons
Learned from Japan's March 11, 2011, Great Tohoku Earthquake and
Subsequent Tsunami,'' dated July 13, 2012 (ADAMS Accession No.
ML12208A210), the NRC indicated that it added the ACRS recommendation
that ``Selected reactor and containment instrumentation should be
enhanced to withstand beyond-design-basis accident conditions'' to the
Tier 3 activities implementing a set of the NRC's Near-Term Task Force
(NTTF) recommendations (Recommendations for Enhancing Reactor Safety in
the 21st Century, dated July 12, 2011, ADAMS Accession No.
ML112510271). The scope of the Tier 3 long-term evaluation is much
broader than, and does not focus on, the use of thermocouples. Rather,
the Tier 3 evaluation will focus on the entire suite of instrumentation
available to operators during a beyond-design-basis accident.
Comment: BWRs need to operate with in-core thermocouples and noted
the following:
CETs are not installed in BWRs. In the event of a severe
accident, BWRs are supposed to detect inadequate core cooling and core
uncovery by measuring the water level in the reactor core. However,
``BWR high drywell temperature and low pressure accidents ([for
example,] LOCAs) can cause the water level to read erroneously high . .
. and BWR water level readings are unreliable after core damage.''
(Leyse-2a)
By the time BWR operators confirm an accelerated core melt
(by measuring increased reactor and containment pressure rates and/or
wetwell water temperature rises), the reactor core would already be
overheated and reflooding an overheated core could generate hydrogen,
at rates as high as 5.0 kg per second. (Leyse-2b)
In the event of a BWR severe accident, in-core
thermocouple measurements would be more accurate and immediate for
detecting inadequate core cooling and core uncovery than readings of
the reactor water level, reactor pressure, containment pressure, or
wetwell water temperature. (Leyse-3)
NRC Response: The NRC considers this comment to be outside the
scope of the matters raised in the PRM. As discussed at the beginning
of the NRC's technical evaluation of this PRM, and in ``PRM Issue 2:
Nuclear Power Plant Operators' Use of In-Core Thermocouples,'' the NRC
is evaluating the PRM as it pertains to PWRs only for the reasons
indicated in those sections. Furthermore, the section addressing PRM
Issue 2 describes some challenges with the use of in-core
thermocouples, both surface-mounted thermocouples and thermocouples in
bulk coolant areas. Those challenges would exist in BWR applications,
as well.
Comment: The proposed additional instrumentation is relevant only
to postulated core conditions where CETs indicate some small amount of
sub-cooling while in-core thermocouples indicate locally higher
temperatures with less sub-cooling. Where CET sub-cooling is minimal,
operators are trained to take actions to increase this margin. Existing
procedures and a predetermined CET value concurrently provide adequate
indication for plant operators to transition from EOPs to implementing
SAMGs. (NEI-5)
NRC Response: The NRC agrees with the comment. As stated in
response to comments Exelon-4/NEI-4/NEI-6 and Leyse-5, operator actions
are not focused on localized core conditions. Rather, actions are based
on bulk CET readings. These readings are established in consideration
of expected differences between local conditions and the associated CET
conditions.
IV. Ongoing NRC Activities Related to Reactor and Containment
Instrumentation
As noted in the ``Miscellaneous Comments'' subsection of Section
III of this document, the NRC has added the ACRS recommendation that
``Selected reactor and containment instrumentation should be enhanced
to withstand beyond-design-basis accident conditions'' to the Tier 3
activities implementing a set of the NRC's NTTF recommendations. The
scope of the Tier 3 long-term evaluation will focus on the entire suite
of instrumentation available to operators during a beyond-design-basis
accident. These activities will support decisions on whether there is a
need for subsequent regulatory action, including rulemaking, in that
area. If the NRC decides that rulemaking is necessary in the area of
reactor instrumentation, the public will have an opportunity to provide
comments as part of publication of a proposed rule in the Federal
Register.
V. Availability of Documents
The following table provides information on how to access the
documents referenced in this document. For more information on
accessing ADAMS, see the ADDRESSES section of this document.
------------------------------------------------------------------------
ADAMS accession
number/Federal
Date Document Register
citation/URL
------------------------------------------------------------------------
February 28, 2012.......... Incoming Petition (PRM-50- ML12065A215.
105) from Mr. Mark Leyse.
May 23, 2012............... Mr. Mark Leyse; Notice of 77 FR 30435.
Receipt of Petition for
Rulemaking.
November 26, 2010.......... Organisation de http://www.oecd-
Cooperation et de nea.org/nsd/
Developpement docs/2010/csni-
Economiques; ``Core Exit r2010-9.pdf.
Temperature (CET)
Effectiveness in
Accident Management of
Nuclear Power Reactor
(NEA/CSNI/R(2010)9)''.
1963....................... Dougall, R.S. and W.M. http://
Rohsenow, ``Film Boiling hdl.handle.net/
on the Inside of 1721.1/62142.
Vertical Tubes with
Upward Flow of the Fluid
at Low Qualities''.
January 1, 1974............ Adams, J.P. and G.E. http://
McCreery, ``Limitations www.osti.gov/
of Detecting Inadequate energycitations/
Core Cooling''. product.biblio.
jsp?osti--id=67
97561.
November 1999.............. WCAP-14696-A, Revision 1, ML993490267.
``Westinghouse Owners
Group Core Damage
Assessment Guidance''.
November 1980.............. NUREG-0737, ML051400209.
``Clarification of TMI
Action Plan
Requirements''.
July 13, 2012.............. Enclosure 3 to SECY-12- ML12208A210.
0095, ``Tier 3 Program
Plans and 6-month Status
Update in Response to
Lessons Learned from
Japan's March 11, 2011,
Great Tohoku Earthquake
and Subsequent Tsunami''.
[[Page 56182]]
October 2010............... Licensing Topical Report, ML102810320.
``Gamma Thermometer
System for LPRM
Calibration and Power
Shape Monitoring''.
April 2009................. Idaho National http://
Laboratory, ``High www.inl.gov/
Temperature Irradiation- technicalpublic
Resistant Thermocouple ations/
Performance documents/
Improvements''. 4235634.pdf.
February 28, 2012.......... 2.206 Petition on Vogtle, ML12061A218.
Units 3 and 4.
April 30, 2013............. Closure Letter to Mr. ML13105A308.
Mark Leyse re. 2.206
Petition on Vogtle,
Units 3 and 4.
July 12, 2011.............. Recommendations for ML112510271.
Enhancing Reactor Safety
in the 21st Century.
August 2, 2012............. Comment Submission (1) ML12216A082.
from Nuclear Energy
Institute.
August 6, 2012............. Comment Submission (2) ML12219A362.
from Mr. Mark Leyse.
August 7, 2012............. Comment Submission (3) ML12230A296.
from Exelon Generation.
August 22, 2012............ Comment Submission (4) ML12237A263.
from Mr. Mark Leyse.
------------------------------------------------------------------------
VI. Determination of the Petition
During normal operation in a PWR, RCS hot leg and cold leg
temperatures are the primary indications of core condition.
Measurements of RCS hot and cold leg temperatures from safety-related
instrumentation provide the necessary input to a plant's reactor
protection system. There are no reactor protection or plant control
functions that use inputs from the CETs. Additionally, the CETs are not
the only source of information relied on to initiate reactor operator
responses to accident conditions.
The NRC has determined that there is no operational necessity for
an exact measurement of core temperatures at various locations
throughout the core. The petitioner provided no justification why the
precise knowledge of core temperature would enhance safety or change
operator actions during normal or accident conditions. Furthermore,
there are no reactor protection or plant control functions that use
inputs from the CETs.
Contrary to the petition's assertion that an OECD report supports a
determination that CETs have limitations, the NRC notes that the same
OECD report stated that ``despite the delay and the difference in the
measured temperatures, the time evolution of the CET signal readings in
the center section seem to reflect the change of the cooling conditions
in the core and thus the tendency of the maximum cladding temperatures
quite well.'' The NRC acknowledges the limitations of CETs but
concludes that CETs are sufficiently accurate to support appropriate
operator action in a timely fashion during an accident. The NRC's
conclusion is consistent with the conclusions of various industry
organizations that the use of CETs is appropriate and safe.
For these reasons, the NRC declines to undertake rulemaking to
require installation and use of in-core thermocouples. Accordingly, the
NRC is denying PRM-50-105 in accordance with 10 CFR 2.803. The NRC's
decision to deny the PRM included consideration of public comments
received on the PRM.
Dated at Rockville, Maryland, this 6th day of September, 2013.
For the Nuclear Regulatory Commission.
Richard J. Laufer,
Acting Secretary of the Commission.
[FR Doc. 2013-22234 Filed 9-11-13; 8:45 am]
BILLING CODE 7590-01-P