In-Core Thermocouples at Different Elevations and Radial Positions in Reactor Core, 56174-56182 [2013-22234]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 78, Number 177 (Thursday, September 12, 2013)]
[Proposed Rules]
[Pages 56174-56182]
From the Federal Register Online via the Government Printing 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 https://www.regulations.gov and search on Docket ID NRC-2012-0056. Address 
questions about NRC dockets to Carol Gallagher, telephone: 301-492-
3668; email: Carol.Gallagher@nrc.gov.
     The NRC's Agencywide Documents Access and Management 
System (ADAMS): You may access publicly available documents online in 
the NRC Library at https://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 PDR.Resource@nrc.gov. 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: Tara.Inverso@nrc.gov.

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 https://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 https://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 
---------------------------------------------------------------------------
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 https://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 https://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            https://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.     https://
                              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.       https://
                              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             https://
                              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