[Federal Register Volume 81, Number 109 (Tuesday, June 7, 2016)]
[Notices]
[Pages 36610-36611]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2016-13387]
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NUCLEAR REGULATORY COMMISSION
[NRC-2016-0097]
Consequential SGTR Analysis for Westinghouse and Combustion
Engineering Plants With Thermally-Treated Alloy 600 and 690 Steam
Generator Tubes
AGENCY: Nuclear Regulatory Commission.
ACTION: Draft NUREG; request for comment.
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SUMMARY: The U.S. Nuclear Regulatory Commission (NRC) is issuing for
public comment a draft NUREG, NUREG-2195, ``Consequential SGTR Analysis
for Westinghouse and Combustion Engineering Plants with Thermally
Treated Alloy 600 and 690 Steam Generator Tubes.'' This report
summarizes severe accident-induced consequential steam generator tube
rupture (C-SGTR) analyses recently
[[Page 36611]]
performed by the NRC's Office of Nuclear Regulatory Research. The
analyses described in this report include risk assessment, thermal-
hydraulic analyses, and materials behavior analyses.
DATES: Submit comments by August 8, 2016. Comments received after this
date will be considered if it is practical to do so, but the Commission
is able to ensure consideration only for comments received before this
date.
ADDRESSES: You may submit comments by any of the following methods
(unless this document describes a different method for submitting
comments on a specific subject):
Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2016-0097. Address
questions about NRC dockets to Carol Gallagher; telephone: 301-415-
3463; email: [email protected]. For technical questions, contact
the individual listed in the FOR FURTHER INFORMATION CONTACT section of
this document.
Mail comments to: Cindy Bladey, Office of Administration,
Mail Stop: OWFN-12-H08, U.S. Nuclear Regulatory Commission, Washington,
DC 20555-0001.
For additional direction on accessing information and submitting
comments, see ``Obtaining Information and Submitting Comments'' in the
SUPPLEMENTARY INFORMATION section of this document.
FOR FURTHER INFORMATION CONTACT: Selim Sancaktar, Office of Nuclear
Regulatory Research; U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001; telephone: 301-415-2391; email: [email protected].
SUPPLEMENTARY INFORMATION:
I. Obtaining Information and Submitting Comments
A. Obtaining Information
Please refer to Docket ID NRC-2016-0097 when contacting the NRC
about the availability of information for this action. You may obtain
publicly-available information related to this action by any of the
following methods:
Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2016-0097.
NRC's Agencywide Documents Access and Management System
(ADAMS): You may obtain publicly-available documents online in the
ADAMS Public Documents collection 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].
Draft NUREG-2195 can be found in ADAMS under at Accession No.
ML16134A029.
NRC's PDR: You may examine and purchase copies of public
documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555
Rockville Pike, Rockville, Maryland 20852.
B. Submitting Comments
Please include Docket ID NRC-2016-0097 in your comment submission.
The NRC cautions you not to include identifying or contact
information that you do not want to be publicly disclosed in your
comment submission. The NRC posts all comment submissions at http://www.regulations.gov as well as entering the comment submissions into
ADAMS. The NRC does not routinely edit comment submissions to remove
identifying or contact information.
If you are requesting or aggregating comments from other persons
for submission to the NRC, then you should inform those persons not to
include identifying or contact information that they do not want to be
publicly disclosed in their comment submission. Your request should
state that the NRC does not routinely edit comment submissions to
remove such information before making the comment submissions available
to the public or entering the comment submissions into ADAMS.
II. Discussion
This report summarizes severe accident-induced consequential steam
generator tube rupture (C-SGTR) analyses recently performed by the
NRC's Office of Nuclear Regulatory Research. The C-SGTRs are
potentially risk-significant events because thermally-induced steam
generator (SG) tube failures caused by hot gases from a damaged reactor
core can result in a containment bypass event and a large release of
fission products to the environment. The main accident scenarios of
interest are those that lead to core damage with high reactor pressure,
dry SG, and low SG pressure (high-dry-low) conditions. A typical
example of such an accident scenario is a station blackout with loss of
auxiliary feedwater. The analyses described in this report include risk
assessment, thermal-hydraulic analyses, and materials behavior
analyses. This work builds on, and updates, previous NRC work.
The current analyses evaluate replacement SGs with thermally-
treated Alloy 600 and Alloy 690 heat exchange tubes and use the latest
tube flaw data available in the 2010 time frame. A main focus of this
work was to compare C-SGTR results for the different SG geometries
associated with Westinghouse and Combustion Engineering plant designs.
It has been previously understood that the geometry of the SG reactor
coolant inlet plenum region and the hot-leg (HL) influences the
temperature of the gases reaching the steam generator tubes during
closed-loop-seal natural circulation conditions. Hotter gases reaching
the SG tube reduce the time before tube failure, which increases the
likelihood of containment bypass. However, if a thermally-induced
failure sufficient to depressurize the reactor coolant system (RCS)
develops in another location, fission product release through failed SG
tubes may be prevented or minimized. Therefore, the possibility of an
earlier failure of other RCS components (such as the reactor coolant
HL) is also considered. Pressure-induced steam generator tube rupture
(SGTR) scenarios, which also may lead to tube failure and subsequent
containment bypass, were also studied, but are deemed to be of lesser
potential impact on overall plant risk.
The methods developed were intended to address the contribution of
thermally-induced SGTR during severe accidents and pressure-induced
SGTR during a number of design-basis accidents. The methods and the
pilot applications were developed in a manner that can establish the
framework to perform a more comprehensive Probabilistic Risk Assessment
that can address the C-SGTR at a level of detail suitable for other NRC
needs.
Dated at Rockville, Maryland, this 26th day of May 2016.
For the Nuclear Regulatory Commission.
Kevin Coyne,
Branch Chief, Probabilistic Risk Assessment Branch, Division of Risk
Analysis, Office of Nuclear Regulatory Research.
[FR Doc. 2016-13387 Filed 6-6-16; 8:45 am]
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