ML16292A811
| ML16292A811 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 10/17/2016 |
| From: | Darrell Roberts Region 3 Administrator |
| To: | Bryan Hanson Exelon Generation Co, Exelon Nuclear |
| References | |
| EA-16-155 IR 2016007 | |
| Download: ML16292A811 (7) | |
See also: IR 05000373/2016007
Text
UNITED STATES
NUCLEAR REGULATORY COMMISSION
REGION III
2443 WARRENVILLE RD. SUITE 210
LISLE, IL 60532-4352
October 17, 2016
Mr. Bryan C. Hanson
Senior VP, Exelon Generation Company, LLC
President and CNO, Exelon Nuclear
4300 Winfield Road
Warrenville, IL 60555
SUBJECT: RESPONSE TO DISPUTED NON-CITED VIOLATION; LASALLE COUNTY
STATION, UNITS 1 AND 2 - INSPECTION REPORT 05000373/2016007;
Dear Mr. Hanson:
On July 22, 2016, Mr. William J. Trafton, Site Vice President, LaSalle County Station, provided
a response (Agencywide Documents Access and Management System [ADAMS] Accession
Number ML16204A307) to an U.S. Nuclear Regulatory Commission (NRC) inspection report
issued on June 22, 2016, concerning activities conducted at the LaSalle County Station.
Specifically, the July 22, 2016, letter contested one of the Non-Cited Violations (NCVs)
contained in the Inspection Report, namely NCV 05000373/2016007-04; 05000374/2016007-04,
Alternate Shutdown Procedures Failed to Ensure RCIC [Reactor Core Isolation Cooling] MOVs
[Motor Operated Valves] Supply Breakers Were Closed. By our letter (ML16231A395) dated
August 18, 2016, the NRC acknowledged Mr. Traftons letter and advised him that NRC staff was
evaluating his reply and would inform him of the results of our evaluation.
In the July 22, 2016, letter your staff stated that the NRC had inappropriately evaluated the
current Exelon Generating Company analysis and approach as a performance deficiency and
stated that the NCV should be rescinded. Your staff provided the following specific bases for
contesting the NCV:
1. The NRCs postulated fire-induced circuit failures involving shorts are outside the
published and accepted standards.
2. The valves identified by the NRC in the NCV [1(2)E51-F019 and 1E51-F059] do not
fall under the trip reset requirements since their function is not necessary to support
the systems fire safe shutdown capability.
The NRC conducted a detailed review of your response and the applicable regulatory
requirements, in accordance with Part I, Section 2.2.8, of the NRC Enforcement Manual. The
evaluation was conducted by a knowledgeable individual independent from the NRC staff in the
Division of Reactor Safety who originally identified the violation and issued the inspection report.
After careful consideration of the basis for your denial of the NCV, the NRC has determined that
B. Hanson -2-
the NCV is valid as stated in the original inspection report. The NRC staff, in its independent
assessment, has determined that the issue qualifies as a performance deficiency due to the
published and accepted standards and the LaSalle Safety Evaluation Report, NUREG-0519,
which required the consideration of fire-induced circuit failures involving shorts. Details of the
evaluation are provided in the enclosure to this letter.
In accordance with Title 10 of the Code of Federal Regulations (10 CFR) 2.390, Public
Inspections, Exemptions, Requests for Withholding, of the NRC's Rules of Practice, a copy of
this letter, its enclosure, your staffs July 22, 2016 letter, and your response (if any) will be
available electronically for public inspection in the NRCs Public Document Room or from the
Publicly Available Records (PARS) component of the NRC's Agencywide Documents Access
and Management System (ADAMS). ADAMS is accessible from the NRC Web site at
http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
Sincerely,
/RA/
Darrell J. Roberts
Deputy Regional Administrator
Docket Nos. 50-373; 50-374
Enclosure:
Independent Assessment of LaSalle Contested Violation
cc: Distribution via LISTSERV
Independent Assessment of LaSalle Contested Violation
The independent assessment of the LaSalle County Station contested Non-Cited Violation
(NCV)05000373/2016007-04; 050i00374/2016007-04, Alternate Shutdown Procedures Failed
to Ensure RCIC [Reactor Core Isolation Cooling] MOVs [Motor Operated Valves] Supply
Breakers Were Closed, concluded that the NCV in the original inspection report was valid.
A review of the NCV is provided below.
The independent assessment included a review of the following documents:
Calculation L-004017, 250 Vdc Breaker Fuse Coordination for RCIC, Revision 000
Generic Letter 86-10, Implementation of Fire Protection Requirements, April 24, 1986
LaSalle County Station, Units 1 and 2 - Response to NRC Component Design Basis
Inspection, Inspection Report 05000373/2016007; 05000374/2016007, July 22, 2016
(ADAMS Accession No. ML16204A307)
Letter form S. Collins, NRC, to R. Beedle, Nuclear Energy Institute, Assessment of
NEI Concerns Regarding NRC Information Notice 92-18 Potential for Loss of
Remote Shutdown Capability During a Control Room Fire, March 11, 1997
(ADAMS Accession No. ML003716454)
Letter from M. Jeffers, NRC, to W. Trafton, LaSalle County Station, LaSalle County
Station Acknowledgment of Disputed Violation of NRC Inspection Report 05000373/2016007; 05000374/2016007, August 18, 2016 (ADAMS Accession No.
NRC Inspection Report 05000373/2016007; 05000374/2016007, LaSalle County
Station, Units 1 and 2 - NRC Component Design Bases Inspection, June 22, 2016
(ADAMS Accession No. ML16174A094)
NUREG-0519, Safety Evaluation Report Related to the Operation of LaSalle County
Station Units 1 and 2, March 1981
NUREG-0800, Standard Review Plan, July 1981
Operating Abnormal Procedure LOA-FX-101, Unit 1 Safe Shutdown with a Fire in
the Control Room or AEER, Revision 27
Operating Abnormal Procedure LOA-FX-201, Unit 2 Safe Shutdown with a Fire in
the Control Room or AEER, Revision 29
Title 10, Code of Federal Regulations, Part 50, Appendix R, Fire Protection Program
for Nuclear Power Facilities Operating Prior to January 1, 1979
In the July 22, 2016 letter, Exelon claimed that the valves identified by the NRC in the NCV
were not required to achieve and maintain safe shutdown; therefore, would not be required to
be proceduralized. Exelon further stated that the U.S. Nuclear Regulatory Commission (NRC)
had not identified or produced any documented requirement, position, or guidance applicable to
LaSalle County Station that specifically describes the additional short circuits they [the NRC] are
contending must be analyzed. Exelon also stated that, Absent a documented standard for
which it was reasonable for LaSalle to foresee, this issue does not qualify as a performance
deficiency.
Enclosure
Exelon disagreed with two of the NRCs key positions that formed the basis for the NCV.
Exelon stated that:
1. The NRC postulated fire-induced circuit failures involving shorts are outside the
published and accepted standards.
2. The valves identified by the NRC in the NCV [1(2)E51-F019 and 1E51-F059] do not
fall under the trip reset requirements since their function is not necessary to support
the systems fire safe shutdown capability.
Exelon Position 1
Exelons position is that the finding/non-cited violation is based on the NRC inspectors
postulating fire-induced circuit failures that are outside the scope of the current requirements or
any NRC endorsed industry guidance applicable to LaSalle County Station. Specifically, Exelon
stated that the inspectors identified short circuit cable lengths shorter than those analyzed in
calculation L-004017 (250 VDC Breaker Fuse Coordination For Reactor Core Isolation Cooling
(RCIC)) by postulating shorts between cables associated with the valves in question
(i.e., 1(2)E51-F019 or F059) and another valve from the same power source, or shorts between
cables associated with these valves and the ground, and cables associated with other valves
and the ground that would end up with a short circuit via the ground.
Exelon further stated that the NRC position on the methodology used in calculation L-004017
was based on Nuclear Energy Institute (NEI) Document 00-01, Guidance for Post-Fire Safe
Shutdown Circuit Analysis, which LaSalle is not currently committed to. Exelon stated that,
Neither during the inspection, nor in the inspection report, has the NRC provided reference to
any formal NRC guidance or endorsed industry document that supports their position that these
additional shorts need to be postulated. Exelon further stated that the NRC had not clearly
described their basis for the performance deficiency.
Response to Exelons Position 1
Section III.G.2 of Title 10 of the Code of Federal Regulations (CFR), Part 50, Appendix R
requires, in part, that, where cables or equipment, including associated non-safety circuits that
could prevent operation or cause maloperation due to hot shorts, open circuits, or shorts to
ground, of redundant trains of systems necessary to achieve and maintain hot shutdown
conditions are located within the same fire area outside of primary containment, one of the
following means of ensuring that one of the redundant trains is free of fire damage shall be
provided Further, for those fire areas not meeting the criteria of III.G.2 Sections III.G.3 and
III.L apply.Section III.L.7 states, in part, that, The safe shutdown equipment and systems for
each fire area shall be known to be isolated from associated non-safety circuits in the fire area
so that hot shorts, open circuits, or shorts to ground in the associated circuits will not prevent
operation of the safe shutdown equipment.
Plants licensed on or after January 1, 1979 (which includes LaSalle) are not required to meet
the requirements of Appendix R. The NRC staff reviewed these plants Fire Protection
Programs against the Standard Review Plan, NUREG-0800. The Standard Review Plan,
Section C.5c(7), contains nearly identical language to Section III.L.7 of Appendix R and states,
in part, that, The safe shutdown equipment and systems for each fire area should be known to
be isolated from associated circuits in the fire area so that hot shorts, open circuits, or shorts to
ground in the associated circuits will not prevent operation of the safe shutdown equipment.
2
The NRC developed Generic Letter 86-10, Implementation of Fire Protection Requirements, to
provide guidance and interpretations of Appendix R requirements. In section 5.3.1, Circuit
Failure Modes, the industry asked, What circuit failure modes must be considered in
identifying circuits associated by spurious actuation? The NRC responded, in part, that:
Sections III.G.2 and III.L.7 of Appendix R define the circuit failure modes as hot shorts,
open circuits, and shorts to ground. For consideration of spurious actuations, all
possible functional failure states must be evaluated, that is, the component could be
energized or de-energized by one or more of the above failure modes. Therefore,
valves could fail open or closed
The discussion of Generic Letter 86-10 is included here because the licensee specifically
referred to the generic letter in the July 22, 2016, response letter and the generic letter contains
the standard fire protection license condition that most, if not all, licensees previously adopted.
As such, the generic letter serves as a guidance document that is applicable to LaSalle County
Station.
The documents discussed above are included in order to show that the NRC staff position
regarding the need for licensees to consider hot shorts has been consistent for the past 35 years.
Regarding the LaSalle County Station licensing basis, the NRC discussed the need to consider
hot shorts in the LaSalle Safety Evaluation Report, NUREG-0519, dated March 1981. In
Section 9.5.3, Alternate Shutdown, of the Safety Evaluation Report, the NRC stated the
following:
(2) For the design basis fire affecting the control room, cable spreading room, or remote
shutdown locations, we require electrical circuits between these control locations to be
sufficiently isolated so that both safe (hot and cold) shutdown capability will not be lost at
both locations, To assure this shutdown capability, we required the applicant to provide
the following information.
(d) The results of an analysis that demonstrates that failure (open, ground, or hot
short) of each circuit identified will not affect the capability to achieve safe
shutdown.
In the licensees July 22, 2016, letter Exelon stated that the NRC based their conclusions
regarding the licensees methodology in calculation L-004017 on guidance provided in NEI 00-01.
However, in the Inspection Report 05000373/2016007; 05000374/2016007 write-up of the issue,
the NRC inspectors did not refer to NEI 00-01. The inspectors based their conclusion on
regulatory requirements and long-standing guidance.
The NRC staff has concluded, in its independent review of Exelon Position 1, that the fire-induced
circuit failures involving shorts that the NRC inspectors postulated in the inspection report are
within the published and accepted standards and that the basis for the performance deficiency is
valid. The preceding discussion provides a 35-year history of regulations and published staff
positions and guidance supporting the need for licensees to consider fire-induced hot shorts.
3
Exelon Position 2
Exelons position is that the valves identified by the NRC in the NCV [1(2)E51-F019 and
1E51-F059] do not fall under the trip reset requirements since their function is not necessary to
support the systems fire safe shutdown capability. The LaSalle Fire Protection Report credits
RCIC injection to the reactor pressure vessel to support fire safe shutdown. Therefore, as
described in the requirements above, it is not necessary to provide alternate instructions for
these valves in the procedures.
Response to Exelons Position 2
In the letter Exelon stated that, valves 1(2)E51-F019 and F059 do not impact the RCIC
injection into the reactor and are not essential for the RCIC system to perform this credited fire
safe shutdown function. Exelon stated that a spurious opening of valves 1(2)E51-F059 would
not divert water from the RCIC injection path because that valve is in series with the normally
closed 1(2)E51-F022 valve. They also stated that the RCIC minimum flow valve 1(2)E51-F019
is normally closed and should it spuriously open the RCIC system would be able to maintain the
desired flow rate.
The concern that the NRC inspectors documented in the inspection report was that the AOPs
did not include alternative instructions to verify that breakers for valves 1(2)E51-F019 and F059
would close in the event the circuit breakers open, preventing the MOVs from being operated.
If these valves could not be opened the centrifugal RCIC pump could be damaged as a result of
the pump deadheading and overheating in a very short period of time. This event could occur
upon startup of the RCIC turbine and pump and fire damage to circuits associated with valves
1(2)E51-F019 and F059 due to the hot shorts. The previous response to Exelon Position 1
validated that LaSalle was required to consider the impacts of hot shorts.
By not providing alternate instructions for the operators in the AOPs in the event that valves
1(2)E51-F019 and F059 could not be opened LaSalle did not ensure that fire damage to circuits
associated with those valves would not affect the capability to achieve safe shutdown. The
failure to open those valves could result in damage to the RCIC system, which is the credited
system for alternate shutdown from the remote shutdown panel.
The NRC staff has concluded, in its independent review of Exelon Position 2, that the valves
1(2)E51-F019 and 1E51-F059 do need to be considered under the trip reset requirements
because their failure to open could result in damage to RCIC system components and an
inability of the RCIC system to perform its safe shutdown actions.
Overall Conclusion
In conclusion, the NRC staff has determined that the issue qualifies as a performance deficiency
due to the published and accepted standards (as documented in Title 10 of the Code of Federal
Regulations (CFR), Part 50, Appendix R; NUREG-0800; and Generic Letter 86-10) and the
LaSalle Safety Evaluation Report, NUREG-0519, which required the consideration of fire-induced
circuit failures involving shorts. The regulatory requirements, staff positions, and guidance on
the need to consider fire-induced hot shorts have been well documented over the past 35 years.
4
B. Hanson -2-
the NCV is valid as stated in the original inspection report. The NRC staff, in its independent
assessment, has determined that the issue qualifies as a performance deficiency due to the
published and accepted standards and the LaSalle Safety Evaluation Report, NUREG-0519,
which required the consideration of fire-induced circuit failures involving shorts. Details of the
evaluation are provided in the enclosure to this letter.
In accordance with Title 10 of the Code of Federal Regulations (10 CFR) 2.390, Public
Inspections, Exemptions, Requests for Withholding, of the NRC's Rules of Practice, a copy of
this letter, its enclosure, your staffs July 22, 2016 letter, and your response (if any) will be
available electronically for public inspection in the NRCs Public Document Room or from the
Publicly Available Records (PARS) component of the NRC's Agencywide Documents Access
and Management System (ADAMS). ADAMS is accessible from the NRC Web site at
http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
Sincerely,
/RA/
Darrell J. Roberts
Deputy Regional Administrator
Docket Nos. 50-373; 50-374
Enclosure:
Independent Assessment of LaSalle Contested Violation
cc: Distribution via LISTSERV
DISTRIBUTION:
RidsNrrDorlLpl3-2 Resource
RidsNrrPMLaSalle
RidsNrrDirsIrib Resource
Cynthia Pederson
DRPIII
DRSIII
ADAMS Accession Number ML16292A811
Publicly Available Non-Publicly Available Sensitive Non-Sensitive
To receive a copy of this document, indicate in the concurrence box "C" = Copy without attach/encl "E" = Copy with attach/encl "N" = No copy
OFFICE RIII RIII RIII OE RIII RIII
NAME DSzwarc:cl MJeffers KLambert for RArrighi via email KOBrien DRoberts
RSkokowski
DATE 09/14/16 10/03/16 10/04/16 09/20/16 10/12/16 10/17/16
OFFICIAL RECORD COPY