ML16113A346: Difference between revisions

From kanterella
Jump to navigation Jump to search
(Created page by program invented by StriderTol)
(StriderTol Bot change)
 
Line 17: Line 17:


=Text=
=Text=
{{#Wiki_filter:UNITED STATES
{{#Wiki_filter:UNITED STATES  
                              NUCLEAR REGULATORY COMMISSION
NUCLEAR REGULATORY COMMISSION  
                                                REGION III
REGION III  
                                    2443 WARRENVILLE RD. SUITE 210
2443 WARRENVILLE RD. SUITE 210  
                                          LISLE, IL 60532-4352
LISLE, IL 60532-4352  
                                            April 19, 2016
April 19, 2016  
Mr. Peter A. Gardner
Site Vice President
Mr. Peter A. Gardner  
Monticello Nuclear Generating Plant
Site Vice President  
Northern States Power Company, Minnesota
Monticello Nuclear Generating Plant  
2807 West County Road 75
Northern States Power Company, Minnesota  
Monticello, MN 55362-9637
2807 West County Road 75  
SUBJECT: MONTICELLO NUCLEAR GENERATING PLANT - EVALUATIONS OF CHANGES,
Monticello, MN 55362-9637  
              TESTS, AND EXPERIMENTS AND PERMANENT PLANT MODIFICATIONS
SUBJECT: MONTICELLO NUCLEAR GENERATING PLANT - EVALUATIONS OF CHANGES,  
              BASELINE INSPECTION REPORT 05000263/2016008
TESTS, AND EXPERIMENTS AND PERMANENT PLANT MODIFICATIONS  
Dear Mr. Gardner:
BASELINE INSPECTION REPORT 05000263/2016008  
On March 24, 2016, the U.S. Nuclear Regulatory Commission (NRC) completed an Evaluations
Dear Mr. Gardner:  
of Changes, Tests, and Experiments and Permanent Plant Modifications inspection at your
On March 24, 2016, the U.S. Nuclear Regulatory Commission (NRC) completed an Evaluations  
Monticello Nuclear Generating Plant. The enclosed inspection report documents the inspection
of Changes, Tests, and Experiments and Permanent Plant Modifications inspection at your  
results which were discussed on March 24, 2016, with Mr. M. Lingenfelter and other members
Monticello Nuclear Generating Plant. The enclosed inspection report documents the inspection  
of your staff.
results which were discussed on March 24, 2016, with Mr. M. Lingenfelter and other members  
The inspection examined activities conducted under your license as they relate to safety and
of your staff.  
compliance with the Commissions rules and regulations and with the conditions of your license.
The inspection examined activities conducted under your license as they relate to safety and  
The inspectors reviewed selected procedures and records, observed activities, and interviewed
compliance with the Commissions rules and regulations and with the conditions of your license.
personnel.
The inspectors reviewed selected procedures and records, observed activities, and interviewed  
No findings were identified during this inspection.
personnel.  
In accordance with Title 10 of the Code of Federal Regulations (10 CFR) 2.390, Public
No findings were identified during this inspection.  
Inspections, Exemptions, Requests for Withholding, of the NRC's "Rules of Practice," a copy
In accordance with Title 10 of the Code of Federal Regulations (10 CFR) 2.390, Public  
of this letter, its enclosure, and your response (if any) will be available electronically for public
Inspections, Exemptions, Requests for Withholding, of the NRC's "Rules of Practice," a copy  
inspection in the NRCs Public Document Room or from the Publicly Available Records (PARS)
of this letter, its enclosure, and your response (if any) will be available electronically for public  
component of the NRC's Agencywide Documents Access and Management System (ADAMS).
inspection in the NRCs Public Document Room or from the Publicly Available Records (PARS)  
ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html
component of the NRC's Agencywide Documents Access and Management System (ADAMS).
(the Public Electronic Reading Room).
ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html  
                                                Sincerely,
(the Public Electronic Reading Room).  
                                                /RA/
Sincerely,  
                                                Robert C. Daley, Chief
                                                Engineering Branch 3
/RA/  
                                                Division of Reactor Safety
Docket No. 50-263
Robert C. Daley, Chief  
License No. DPR-22
Engineering Branch 3  
Enclosure:
Division of Reactor Safety  
IR 05000263/2016008
Docket No. 50-263  
cc: Distribution via LISTSERV
License No. DPR-22  
Enclosure:  
IR 05000263/2016008  
cc: Distribution via LISTSERV  


          U. S. NUCLEAR REGULATORY COMMISSION
Enclosure
                          REGION III
U. S. NUCLEAR REGULATORY COMMISSION  
Docket No:           50-263
REGION III  
License No:         DPR-22
Docket No:  
Report No:           05000263/2016008
50-263  
Licensee:           Northern States Power Company, Minnesota
License No:  
Facility:           Monticello Nuclear Generating Plant
DPR-22  
Location:           Monticello, MN
Report No:  
Dates:               February 29 thru March 24, 2016
05000263/2016008  
Inspectors:         Alan Dahbur, Senior Reactor Inspector (Lead)
Licensee:  
                    Jorge J. Corujo-Sandín, Reactor Inspector
Northern States Power Company, Minnesota  
                    Michael A. Jones, Reactor Inspector
Facility:  
Approved by:         Robert C. Daley, Chief
Monticello Nuclear Generating Plant
                    Engineering Branch 3
Location:  
                    Division of Reactor Safety
Monticello, MN  
                                                                  Enclosure
Dates:  
February 29 thru March 24, 2016  
Inspectors:  
Alan Dahbur, Senior Reactor Inspector (Lead)  
Jorge J. Corujo-Sandín, Reactor Inspector  
Michael A. Jones, Reactor Inspector  
Approved by:  
Robert C. Daley, Chief  
Engineering Branch 3  
Division of Reactor Safety  


                                        SUMMARY
2
Inspection Report 05000263/2016008; 02/29/2016 - 03/24/2016; Monticello Nuclear Generating
SUMMARY  
Plant; Evaluations of Changes, Tests, and Experiments and Permanent Plant Modifications.
Inspection Report 05000263/2016008; 02/29/2016 - 03/24/2016; Monticello Nuclear Generating  
This report covers a 2-week announced baseline inspection on evaluations of changes,
Plant; Evaluations of Changes, Tests, and Experiments and Permanent Plant Modifications.  
tests, and experiments and permanent plant modifications. The inspection was conducted
This report covers a 2-week announced baseline inspection on evaluations of changes,  
by Region III based engineering inspectors. The U.S. Nuclear Regulatory Commissions
tests, and experiments and permanent plant modifications. The inspection was conducted  
program for overseeing the safe operation of commercial nuclear power reactors is described
by Region III based engineering inspectors. The U.S. Nuclear Regulatory Commissions  
in NUREG-1649, Reactor Oversight Process, Revision 5, dated February 2014.
program for overseeing the safe operation of commercial nuclear power reactors is described  
        NRC-Identified and Self-Revealed Findings
in NUREG-1649, Reactor Oversight Process, Revision 5, dated February 2014.  
        No findings were identified.
NRC-Identified and Self-Revealed Findings  
        Licensee-Identified Violations
No findings were identified.  
        No violations were identified.
Licensee-Identified Violations  
                                              2
No violations were identified.  


                                      REPORT DETAILS
3
1.   REACTOR SAFETY
REPORT DETAILS  
      Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity
1.  
1R17 Evaluations of Changes, Tests, and Experiments and Permanent Plant Modifications
REACTOR SAFETY  
      (71111.17T)
Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity  
.1   Evaluation of Changes, Tests, and Experiments
1R17 Evaluations of Changes, Tests, and Experiments and Permanent Plant Modifications  
  a. Inspection Scope
(71111.17T)  
      The inspectors reviewed 7 safety evaluations performed pursuant to Title 10, Code of
.1  
      Federal Regulations (CFR), Part 50, Section 59, to determine if the evaluations were
Evaluation of Changes, Tests, and Experiments  
      adequate, and that prior U.S. Nuclear Regulatory Commission (NRC) approval was
a.  
      obtained as appropriate. The inspectors also reviewed 16 screenings where licensee
Inspection Scope  
      personnel had determined that a 10 CFR 50.59 evaluation was not necessary. The
The inspectors reviewed 7 safety evaluations performed pursuant to Title 10, Code of  
      inspectors reviewed these documents to determine if:
Federal Regulations (CFR), Part 50, Section 59, to determine if the evaluations were  
            the changes, tests, and experiments performed were evaluated in accordance
adequate, and that prior U.S. Nuclear Regulatory Commission (NRC) approval was  
              with 10 CFR 50.59 and that sufficient documentation existed to confirm that a
obtained as appropriate. The inspectors also reviewed 16 screenings where licensee  
              license amendment was not required;
personnel had determined that a 10 CFR 50.59 evaluation was not necessary. The  
            the safety issue requiring the change, tests or experiment was resolved;
inspectors reviewed these documents to determine if:  
            the licensee conclusions for evaluations of changes, tests, and experiments were
              correct and consistent with 10 CFR 50.59; and
the changes, tests, and experiments performed were evaluated in accordance  
            the design and licensing basis documentation was updated to reflect the change.
with 10 CFR 50.59 and that sufficient documentation existed to confirm that a  
      The inspectors used, in part, Nuclear Energy Institute Document 96-07, Guidelines for
license amendment was not required;  
      10 CFR 50.59 Implementation, Revision 1, to determine acceptability of the completed
      evaluations, and screenings. The Nuclear Energy Institute document was endorsed by
the safety issue requiring the change, tests or experiment was resolved;  
      the NRC in Regulatory Guide 1.187, Guidance for Implementation of 10 CFR 50.59,
      Changes, Tests, and Experiments, dated November 2000. The inspectors also
the licensee conclusions for evaluations of changes, tests, and experiments were  
      consulted Part 9900 of the NRC Inspection Manual, 10 CFR Guidance for
correct and consistent with 10 CFR 50.59; and  
      10 CFR 50.59, Changes, Tests, and Experiments.
      This inspection constituted 4 samples of evaluations and 13 samples of screenings
the design and licensing basis documentation was updated to reflect the change.  
      and/or applicability determinations as defined in Inspection Procedure 71111.17-04.
The inspectors used, in part, Nuclear Energy Institute Document 96-07, Guidelines for  
  b. Findings
10 CFR 50.59 Implementation, Revision 1, to determine acceptability of the completed  
      (Open) Unresolved Item 05000263/2016008-01, Failure to provide acceptable Alternate
evaluations, and screenings. The Nuclear Energy Institute document was endorsed by  
      Methods of Decay Heat Removal
the NRC in Regulatory Guide 1.187, Guidance for Implementation of 10 CFR 50.59,  
      Introduction: The inspectors identified an Unresolved Item associated with Technical
Changes, Tests, and Experiments, dated November 2000. The inspectors also  
      Specification (TS) 3.4.8, Residual Heat Removal (RHR) Shutdown Cooling System -
consulted Part 9900 of the NRC Inspection Manual, 10 CFR Guidance for  
      Cold Shutdown. Specifically, the licensee failed to verify that the capability of the
10 CFR 50.59, Changes, Tests, and Experiments.  
      alternate methods of decay heat removal described in Operations Manual C.4-B.03.04.A,
This inspection constituted 4 samples of evaluations and 13 samples of screenings  
      Loss of Normal Shutdown Cooling, were adequate to combat a loss of shutdown cooling
and/or applicability determinations as defined in Inspection Procedure 71111.17-04.  
      resulting from the loss of one or two RHR subsystems while in MODE 4 with high decay
b.  
      heat load.
Findings  
                                                3
(Open) Unresolved Item 05000263/2016008-01, Failure to provide acceptable Alternate  
Methods of Decay Heat Removal  
Introduction: The inspectors identified an Unresolved Item associated with Technical  
Specification (TS) 3.4.8, Residual Heat Removal (RHR) Shutdown Cooling System -  
Cold Shutdown. Specifically, the licensee failed to verify that the capability of the  
alternate methods of decay heat removal described in Operations Manual C.4-B.03.04.A,  
Loss of Normal Shutdown Cooling, were adequate to combat a loss of shutdown cooling  
resulting from the loss of one or two RHR subsystems while in MODE 4 with high decay  
heat load.  


Description: The Limiting Condition for Operation (LCO) 3.4.8 of TS Residual Heat
4
Removal Shutdown Cooling System - Cold Shutdown, required in Mode 4, two RHR
Description: The Limiting Condition for Operation (LCO) 3.4.8 of TS Residual Heat  
shutdown cooling subsystems shall be operable, and, with no recirculation pump in
Removal Shutdown Cooling System - Cold Shutdown, required in Mode 4, two RHR  
operation, at least one RHR shutdown cooling subsystem shall be in operation. The
shutdown cooling subsystems shall be operable, and, with no recirculation pump in  
TS Bases Section 3.4.8, indicated that an operable RHR shutdown cooling subsystem
operation, at least one RHR shutdown cooling subsystem shall be in operation. The  
consisted of one operable RHR pump, one heat exchanger, the associated piping and
TS Bases Section 3.4.8, indicated that an operable RHR shutdown cooling subsystem  
valves, and the necessary portions of the RHR Service Water System System capable
consisted of one operable RHR pump, one heat exchanger, the associated piping and  
of providing cooling water to the heat exchanger. The TS Bases Section 3.4.8 further
valves, and the necessary portions of the RHR Service Water System System capable  
indicated that the two subsystems have a common suction source and were allowed to
of providing cooling water to the heat exchanger. The TS Bases Section 3.4.8 further  
have a common heat exchanger and common discharge piping. Thus, to meet the LCO,
indicated that the two subsystems have a common suction source and were allowed to  
both pumps in one loop or one pump in each of the two loops must be operable. Since
have a common heat exchanger and common discharge piping. Thus, to meet the LCO,  
the piping and heat exchangers were passive components that were assumed not to fail,
both pumps in one loop or one pump in each of the two loops must be operable. Since  
they were allowed to be common to both subsystems.
the piping and heat exchangers were passive components that were assumed not to fail,  
When TS 3.4.8, LCO could not be met, Condition A, for one or two RHR shutdown
they were allowed to be common to both subsystems.
cooling subsystems inoperable, the Required Action was to, verify an alternate
When TS 3.4.8, LCO could not be met, Condition A, for one or two RHR shutdown  
method of decay heat removal was available for each inoperable RHR shutdown
cooling subsystems inoperable, the Required Action was to, verify an alternate  
cooling subsystem. The completion time for the required action was 1 hour, and
method of decay heat removal was available for each inoperable RHR shutdown  
once per 24 hours thereafter. The TS Bases 3.4.8 for Condition A indicated that with
cooling subsystem. The completion time for the required action was 1 hour, and  
one of the two required RHR shutdown cooling subsystems inoperable, the remaining
once per 24 hours thereafter. The TS Bases 3.4.8 for Condition A indicated that with  
subsystem was capable of providing the required decay heat removal. However, the
one of the two required RHR shutdown cooling subsystems inoperable, the remaining  
overall reliability was reduced, therefore, an alternate method of decay heat removal
subsystem was capable of providing the required decay heat removal. However, the  
must be provided. With both RHR shutdown cooling subsystems inoperable, an
overall reliability was reduced, therefore, an alternate method of decay heat removal  
alternate method of decay heat removal must be provided in addition to that provided
must be provided. With both RHR shutdown cooling subsystems inoperable, an  
for the initial RHR shutdown cooling subsystem inoperability. This was to ensure the
alternate method of decay heat removal must be provided in addition to that provided  
re-establishment of backup decay heat removal capabilities, similar to the requirements
for the initial RHR shutdown cooling subsystem inoperability. This was to ensure the  
of the LCO. The bases further stated that the required cooling capacity of the alternate
re-establishment of backup decay heat removal capabilities, similar to the requirements  
method should be ensured by verifying (by calculation or demonstration) its capability to
of the LCO. The bases further stated that the required cooling capacity of the alternate  
maintain or reduce temperature. Alternate methods that can be used included (but not
method should be ensured by verifying (by calculation or demonstration) its capability to  
limited to) the Reactor Water Cleanup System by itself or using feed and bleed in
maintain or reduce temperature. Alternate methods that can be used included (but not  
combination with Control Rod Drive System or Condensate/Feed Systems.
limited to) the Reactor Water Cleanup System by itself or using feed and bleed in  
Abnormal Procedure, Operations Manual C.4-B.03.04.A, Loss of Normal Shutdown
combination with Control Rod Drive System or Condensate/Feed Systems.  
Cooling, provided instructions for establishing alternate methods for decay heat
Abnormal Procedure, Operations Manual C.4-B.03.04.A, Loss of Normal Shutdown  
removal. The inspectors noticed that except for the alternate method as described
Cooling, provided instructions for establishing alternate methods for decay heat  
below in the G-EK-1-45, the licensee was not able to show by calculation or
removal. The inspectors noticed that except for the alternate method as described  
demonstration that the systems and methods credited in this procedure would be
below in the G-EK-1-45, the licensee was not able to show by calculation or  
capable of providing sufficient heat removal capability or appropriate levels of
demonstration that the systems and methods credited in this procedure would be  
redundancy as required by TS 3.4.8.
capable of providing sufficient heat removal capability or appropriate levels of  
The G-EK-1-45 was a General Electric Letter to Northern States Power, Subject: Cold
redundancy as required by TS 3.4.8.  
Shutdown Capability Report, dated April 22, 1981. This letter provided a report which
The G-EK-1-45 was a General Electric Letter to Northern States Power, Subject: Cold  
described the capability of the Monticello Nuclear Generating Plant to achieve cold
Shutdown Capability Report, dated April 22, 1981. This letter provided a report which  
shutdown using only safety class systems and assuming the worst single failure. The
described the capability of the Monticello Nuclear Generating Plant to achieve cold  
alternate shutdown decay heat removal method used in the report credited combinations
shutdown using only safety class systems and assuming the worst single failure. The  
of the RHR pumps and heat exchangers in the suppression pool cooling mode of RHR
alternate shutdown decay heat removal method used in the report credited combinations  
to ensure suppression pool water temperatures were below the design limit. This
of the RHR pumps and heat exchangers in the suppression pool cooling mode of RHR  
method utilized the core spray system and safety relief valves to circulate reactor
to ensure suppression pool water temperatures were below the design limit. This  
inventory to remove decay heat from the reactor.
method utilized the core spray system and safety relief valves to circulate reactor  
                                          4
inventory to remove decay heat from the reactor.  


    The inspectors noted that calculations supporting the above alternate strategy utilized an
5
    RHR subsystem that could be inoperable and/or unavailable and therefore may not be
The inspectors noted that calculations supporting the above alternate strategy utilized an  
    credited to comply with TS 3.4.8. Specifically, the inspectors were concerned that while
RHR subsystem that could be inoperable and/or unavailable and therefore may not be  
    the plant was in mode 4, with a credited one subsystem inoperable, the licensees
credited to comply with TS 3.4.8. Specifically, the inspectors were concerned that while  
    credited alternate decay heat removal method that relied on an RHR subsystem, to
the plant was in mode 4, with a credited one subsystem inoperable, the licensees  
    perform the required suppression pool cooling function. The inspectors were concerned
credited alternate decay heat removal method that relied on an RHR subsystem, to  
    that relying on the only operable RHR subsystem for the alternate method did not meet
perform the required suppression pool cooling function. The inspectors were concerned  
    the intent of the TS requirement as described in the TS Bases. Furthermore, the
that relying on the only operable RHR subsystem for the alternate method did not meet  
    inspectors noticed for Mode 4 with two RHR subsystems inoperable, the licensee failed
the intent of the TS requirement as described in the TS Bases. Furthermore, the  
    to verify by calculation or demonstrations that two additional redundant alternate decay
inspectors noticed for Mode 4 with two RHR subsystems inoperable, the licensee failed  
    heat removal methods existed with sufficient capacity to maintain the average reactor
to verify by calculation or demonstrations that two additional redundant alternate decay  
    coolant temperature below 212 degrees Fahrenheit.
heat removal methods existed with sufficient capacity to maintain the average reactor  
    During the inspection, the licensee indicated that the Boiling Reactor Owners Group was
coolant temperature below 212 degrees Fahrenheit.  
    in the process of developing a draft TS Task Force Traveler to address the requirement
During the inspection, the licensee indicated that the Boiling Reactor Owners Group was  
    of TS 3.4.8 and its Bases.
in the process of developing a draft TS Task Force Traveler to address the requirement  
    Based on the information above, the inspectors were concerned that the plant
of TS 3.4.8 and its Bases.
    Operations Manual was inadequate and failed to include alternate decay heat removal
Based on the information above, the inspectors were concerned that the plant  
    methods that would enable the licensee to comply with the requirement of TS 3.4.8. The
Operations Manual was inadequate and failed to include alternate decay heat removal  
    Operations Manual was required per TS 5.4.1 Procedures, which required that written
methods that would enable the licensee to comply with the requirement of TS 3.4.8. The  
    procedures shall be established, implemented, and maintained covering the emergency
Operations Manual was required per TS 5.4.1 Procedures, which required that written  
    operating procedures. The inspectors determined that this issue was unresolved
procedures shall be established, implemented, and maintained covering the emergency  
    pending the actions by the licensee and the Boiling Reactor Owners Group and the NRC
operating procedures. The inspectors determined that this issue was unresolved  
    review of these actions. The licensee entered the inspectors concerns into their
pending the actions by the licensee and the Boiling Reactor Owners Group and the NRC  
    Corrective Action Program as AR 01516098. (URI 05000263/2016008-01, Failure to
review of these actions. The licensee entered the inspectors concerns into their  
    provide acceptable Alternate Methods of Decay Heat Removal)
Corrective Action Program as AR 01516098. (URI 05000263/2016008-01, Failure to  
.2   Permanent Plant Modifications
provide acceptable Alternate Methods of Decay Heat Removal)  
  a. Inspection Scope
.2  
    The inspectors reviewed seven permanent plant modifications that had been installed
Permanent Plant Modifications  
    in the plant during the last 3 years. This review included in-plant walkdowns portions
a.  
    of the high-pressure coolant injection steam drain line system, the Emergency Diesel
Inspection Scope  
    Generator Fuel Oil Transfer System, including the Diesel Fuel Oil pump house, the new
The inspectors reviewed seven permanent plant modifications that had been installed  
    diesel fuel oil pumps installed in the day tank room, and portions of the fuel oil storage
in the plant during the last 3 years. This review included in-plant walkdowns portions  
    tank tornado missile protection modifications. The modifications were selected based
of the high-pressure coolant injection steam drain line system, the Emergency Diesel  
    upon risk significance, safety significance, and complexity. The inspectors reviewed the
Generator Fuel Oil Transfer System, including the Diesel Fuel Oil pump house, the new  
    modifications selected to determine if:
diesel fuel oil pumps installed in the day tank room, and portions of the fuel oil storage  
            the supporting design and licensing basis documentation was updated;
tank tornado missile protection modifications. The modifications were selected based  
            the changes were in accordance with the specified design requirements;
upon risk significance, safety significance, and complexity. The inspectors reviewed the  
            the procedures and training plans affected by the modification have been
modifications selected to determine if:  
              adequately updated;
            the test documentation as required by the applicable test programs has been
the supporting design and licensing basis documentation was updated;  
              updated; and
            post-modification testing adequately verified system operability and/or
the changes were in accordance with the specified design requirements;  
              functionality.
                                              5
the procedures and training plans affected by the modification have been  
adequately updated;  
the test documentation as required by the applicable test programs has been  
updated; and  
post-modification testing adequately verified system operability and/or  
functionality.  


      The inspectors also used applicable industry standards to evaluate acceptability of the
6
      modifications. The list of modifications and other documents reviewed by the inspectors
The inspectors also used applicable industry standards to evaluate acceptability of the  
      is included as an Attachment to this report.
modifications. The list of modifications and other documents reviewed by the inspectors  
      This inspection constituted eight permanent plant modification samples as defined in
is included as an Attachment to this report.  
      Inspection Procedure 71111.17-04.
This inspection constituted eight permanent plant modification samples as defined in  
4.   OTHER ACTIVITIES
Inspection Procedure 71111.17-04.  
4OA2 Problem Identification and Resolution
4.  
.1   Routine Review of Condition Reports
OTHER ACTIVITIES  
  a. Inspection Scope
4OA2 Problem Identification and Resolution  
      The inspectors reviewed several corrective action process documents that identified or
.1  
      were related to Title 10 of the Code of Federal Regulations, Part 50.59 evaluations and
Routine Review of Condition Reports  
      permanent plant modifications. The inspectors reviewed these documents to evaluate
a.  
      the effectiveness of corrective actions related to permanent plant modifications and
Inspection Scope  
      evaluations of changes, tests, and experiments. In addition, corrective action
The inspectors reviewed several corrective action process documents that identified or  
      documents written on issues identified during the inspection were reviewed to verify
were related to Title 10 of the Code of Federal Regulations, Part 50.59 evaluations and  
      adequate problem identification and incorporation of the problems into the corrective
permanent plant modifications. The inspectors reviewed these documents to evaluate  
      action system. The specific corrective action documents that were sampled and
the effectiveness of corrective actions related to permanent plant modifications and  
      reviewed by the inspectors are listed in the attachment to this report.
evaluations of changes, tests, and experiments. In addition, corrective action  
  b. Findings
documents written on issues identified during the inspection were reviewed to verify  
      No findings were identified.
adequate problem identification and incorporation of the problems into the corrective  
4OA6 Management Meetings
action system. The specific corrective action documents that were sampled and  
.1   Exit Meeting Summary
reviewed by the inspectors are listed in the attachment to this report.  
      On March 24, 2016, the inspectors presented the inspection results to Mr. M. Lingenfelter,
b.  
      and other members of the licensee staff. The licensee personnel acknowledged the
Findings  
      inspection results presented and did not identify any proprietary content. The inspectors
No findings were identified.  
      confirmed that all proprietary material provided to the inspection team was identified and
4OA6 Management Meetings  
      will be dispositioned in accordance with applicable processes.
.1  
ATTACHMENT: SUPPLEMENTAL INFORMATION
Exit Meeting Summary
                                                6
On March 24, 2016, the inspectors presented the inspection results to Mr. M. Lingenfelter,  
and other members of the licensee staff. The licensee personnel acknowledged the  
inspection results presented and did not identify any proprietary content. The inspectors  
confirmed that all proprietary material provided to the inspection team was identified and  
will be dispositioned in accordance with applicable processes.
ATTACHMENT: SUPPLEMENTAL INFORMATION


                              SUPPLEMENTAL INFORMATION
Attachment
                                  KEY POINTS OF CONTACT
SUPPLEMENTAL INFORMATION  
Licensee
KEY POINTS OF CONTACT  
M. Lingenfelter, Director of Engineering
Licensee  
A. Gonnering, Design Engineering
M. Lingenfelter, Director of Engineering  
M. Kelly, Performance Assurance Manager
A. Gonnering, Design Engineering  
J. Gausman, Engineering
M. Kelly, Performance Assurance Manager  
A. Ward, Regulatory Affairs Manager
J. Gausman, Engineering  
T. Hurrle, Design Engineering Manager
A. Ward, Regulatory Affairs Manager  
B. Halvorson, Engineering
T. Hurrle, Design Engineering Manager  
A. Kouba, Regulatory Affairs
B. Halvorson, Engineering
D. Alstad, Design Engineer
A. Kouba, Regulatory Affairs  
E. Watzel, Electrical Design Engineering Supervisor
D. Alstad, Design Engineer  
U.S. Nuclear Regulatory Commission
E. Watzel, Electrical Design Engineering Supervisor  
P. Zurawski, Senior Resident Inspector
U.S. Nuclear Regulatory Commission  
P. LaFlamme, Acting Senior Resident Inspector
P. Zurawski, Senior Resident Inspector  
D. Krause, Resident Inspector
P. LaFlamme, Acting Senior Resident Inspector  
                    LIST OF ITEMS OPENED, CLOSED AND DISCUSSED
D. Krause, Resident Inspector  
Opened
LIST OF ITEMS OPENED, CLOSED AND DISCUSSED  
                                Failure to provide acceptable Alternate Methods of Decay Heat
Opened
05000263/2016008-01 URI
05000263/2016008-01 URI
                                Removal (Section 1R17.1b)
Failure to provide acceptable Alternate Methods of Decay Heat  
Closed and Discussed
Removal (Section 1R17.1b)  
None
Closed and Discussed  
                                  LIST OF ACRONYMS USED
None  
ADAMS         Agencywide Documents Access and Management System
LIST OF ACRONYMS USED  
CFR           Code of Federal Regulations
ADAMS  
LCO           Limiting Condition for Operation
Agencywide Documents Access and Management System  
NRC           U.S. Nuclear Regulatory Commission
CFR  
PARS           Publicly Available Records System
Code of Federal Regulations  
RHR           Residual Heat Removal
LCO  
TS             Technical Specifications
Limiting Condition for Operation  
                                                                                    Attachment
NRC  
U.S. Nuclear Regulatory Commission  
PARS  
Publicly Available Records System  
RHR  
Residual Heat Removal  
TS  
Technical Specifications  


                                  LIST OF DOCUMENTS REVIEWED
2
The following is a list of documents reviewed during the inspection. Inclusion on this list does
LIST OF DOCUMENTS REVIEWED  
not imply that the NRC inspectors reviewed the documents in their entirety, but rather, that
The following is a list of documents reviewed during the inspection. Inclusion on this list does  
selected sections of portions of the documents were evaluated as part of the overall inspection
not imply that the NRC inspectors reviewed the documents in their entirety, but rather, that  
effort. Inclusion of a document on this list does not imply NRC acceptance of the document or
selected sections of portions of the documents were evaluated as part of the overall inspection  
any part of it, unless this is stated in the body of the inspection report.
effort. Inclusion of a document on this list does not imply NRC acceptance of the document or  
10 CFR 50.59 EVALUATIONS
any part of it, unless this is stated in the body of the inspection report.
Number             Description or Title                                                 Revision
10 CFR 50.59 EVALUATIONS  
SCR-12-0559       HPCI Logic Change to Provide Margin to MO-2035 and #16 Battery           1
Number  
SCR-13-0554       External Flooding Protection Strategy Change                             0
Description or Title  
SCR-15-0202       Evaluation of EPG/SAG, Revision 3                                         0
Revision  
SCR-16-0024       Disconnect Faulty 46-19 PIP Over-Travel Input                             0
SCR-12-0559  
10 CFR 50.59 SCREENINGS
HPCI Logic Change to Provide Margin to MO-2035 and #16 Battery
Number             Description or Title                                                 Revision
1  
SCR-13-0696       Revise EDG Base Tank Fuel Oil Level Calculation 90-023                   0
SCR-13-0554  
SCR-14-0074       Time Delay Relay 97-29 and 97-31 Setpoint Change                         0
External Flooding Protection Strategy Change
SCR-14-0413       Temp Rev to C.6-006-A-01and C.6-006-A-02
0  
SCR-14-0415       USAR-06.06 Revision                                                       0
SCR-15-0202  
SCR-14-0421       EC 23981 EDG Fuel Oil Tank Vent Lines Missile Protection                 0
Evaluation of EPG/SAG, Revision 3  
SCR-14-0512       Safety and Seismic Classification of the DG/RF and DG/RV Relays           0
0  
SCR-14-0542       RHRSW and Emergency Service Water TS Bases Changes
SCR-16-0024  
SCR-14-0591       Fuel Oil Separation                                                       4
Disconnect Faulty 46-19 PIP Over-Travel Input  
SCR-14-0593       Revise Calculation 94-086 on SRV Accumulation Allowable                   0
0  
                  Leakage Rates
SCR-15-0093       EDG ESW Basket Strainer Modification
10 CFR 50.59 SCREENINGS  
SCR-15-0115       C.4-B.09.02.A Revision to Resolve CAP AR 01465720                         1
Number  
SCR-15-0193       Room Heat Up Calculation Revisions for SBO
Description or Title  
SCR-15-0291       Revise Maximum Volume of EDG Base Tank in 90-023                         0
Revision  
SCR-15-0292       Diesel Pump House Heat Up Calculation                                     0
SCR-13-0696  
CALCULATIONS
Revise EDG Base Tank Fuel Oil Level Calculation 90-023  
Number             Description or Title                                               Revision
0  
03-089             Inservice Testing Acceptance Criteria                                     3
SCR-14-0074  
09-106             CSP Motor-Oil and Bearing Operating Temperatures without                 1
Time Delay Relay 97-29 and 97-31 Setpoint Change  
                  Cooling Water
0  
09-176             Evaluation for Debris Disposition in Supply Pipe and Motor Cooler         0
SCR-14-0413  
                  Tube
Temp Rev to C.6-006-A-01and C.6-006-A-02  
09-178             Time to reach the RHRSW Pump Motor Cooling Line Strainer                 0
                  Limiting Pressure Differential
SCR-14-0415  
14-025             Instrument Setpoint Calculation - Time Delay for Transfer to EDG         0
USAR-06.06 Revision  
                  on Loss of Voltage
0  
90-023             EC 23085 - EDG Fuel Oil Train Separation                                 3
SCR-14-0421  
92-224             Emergency Diesel Generator Loading                                       6A
EC 23981 EDG Fuel Oil Tank Vent Lines Missile Protection  
94-086             Max Allowed Leakage Rates and Test Acceptance Criteria for SRV           5
0  
                                                    2
SCR-14-0512  
Safety and Seismic Classification of the DG/RF and DG/RV Relays  
0  
SCR-14-0542  
RHRSW and Emergency Service Water TS Bases Changes  
SCR-14-0591  
Fuel Oil Separation
4  
SCR-14-0593  
Revise Calculation 94-086 on SRV Accumulation Allowable  
Leakage Rates  
0
SCR-15-0093  
EDG ESW Basket Strainer Modification  
SCR-15-0115  
C.4-B.09.02.A Revision to Resolve CAP AR 01465720  
1  
SCR-15-0193  
Room Heat Up Calculation Revisions for SBO  
SCR-15-0291  
Revise Maximum Volume of EDG Base Tank in 90-023  
0  
SCR-15-0292  
Diesel Pump House Heat Up Calculation  
0  
CALCULATIONS  
Number  
Description or Title  
Revision  
03-089  
Inservice Testing Acceptance Criteria  
3  
09-106  
CSP Motor-Oil and Bearing Operating Temperatures without  
Cooling Water  
1
09-176  
Evaluation for Debris Disposition in Supply Pipe and Motor Cooler  
Tube  
0
09-178  
Time to reach the RHRSW Pump Motor Cooling Line Strainer  
Limiting Pressure Differential  
0
14-025  
Instrument Setpoint Calculation - Time Delay for Transfer to EDG  
on Loss of Voltage  
0
90-023  
EC 23085 - EDG Fuel Oil Train Separation  
3  
92-224  
Emergency Diesel Generator Loading  
6A  
94-086  
Max Allowed Leakage Rates and Test Acceptance Criteria for SRV  
5  


CORRECTIVE ACTION PROGRAM DOCUMENTS INITIATED DURING INSPECTION
3
Number     Description or Title                                               Date
CORRECTIVE ACTION PROGRAM DOCUMENTS INITIATED DURING INSPECTION  
1510936     Incomplete EC Record Copy                                       02/03/2016
Number  
1514133     Clarification for USAR Section 8.4.1.3                         03/01/2016
Description or Title  
1514202     Page missing from WO 00491265 Record                           03/02/2016
Date  
1514369     Screening SCR 14-0421 Answered Question Incorrectly             03/03/2016
1510936  
1514464     EDG Building Roof FOI 91-0265                                   03/03/2016
Incomplete EC Record Copy  
1515054     Bases for Procedure A.6 Contain Incorrect Statements           03/09/2016
02/03/2016  
1515688     Signs of Leakage around FO-11-3                                 03/15/2016
1514133  
1515716     NRC not Provided with Latest Copy of EC23085                   03/15/2016
Clarification for USAR Section 8.4.1.3  
1515907     Formal Evaluation for HPCI Drain Line Bypass Flow               03/16/2016
03/01/2016  
1515939     Question Raised on CRD 46-19                                   03/16/2016
1514202  
1516098     Actions for when LCO 3.4.8, RA A.1 not met Unclear             03/17/2016
Page missing from WO 00491265 Record  
1516101     Core Spray Motor Cooling Design Basis Question                 03/17/2016
03/02/2016  
1516105     HPCI SR Test Inconsistent with TS Bases                         03/17/2016
1514369  
1516106     RCIC Surveillance Required Test Inconsistent with TS Bases     03/17/2016
Screening SCR 14-0421 Answered Question Incorrectly  
CORRECTIVE ACTION PROGRAM DOCUMENTS REVIEWED
03/03/2016  
Number     Description or Title                                               Date
1514464  
952310     M91064A Quarterly Backflushing of Residual Heat Removal         07/27/1991
EDG Building Roof FOI 91-0265  
            system and Core Spray Pump Motors
03/03/2016  
01196451   CDBI EDG Base Tank Volume Calculation CA 90-023                 09/03/2003
1515054  
01355853   Update UFSAR for External Flooding Description Discrepancy     10/22/2012
Bases for Procedure A.6 Contain Incorrect Statements  
01414416   Diesel Fuel Oil Temperature in Fuel Oil Transfer House is not   12/17/2013
03/09/2016  
            Known
1515688  
01420875-03 Condition Evaluation on EDG Base Tank Level Issues             04/04/2014
Signs of Leakage around FO-11-3  
01424477   Appendix R Fire Strategy for Fire Area XII incorrect           03/27/2014
03/15/2016  
1478798     EG Transfer Relay not Classified as Safety Related             05/13/2015
1515716  
1484554     RHRSW-29-2 Handwheel/Stem Sheared off                           06/29/2015
NRC not Provided with Latest Copy of EC23085  
1502700     Catastrophic fail of MO-1900                                   11/19/2015
03/15/2016  
DRAWINGS
1515907  
Number     Description or Title                                             Revision
Formal Evaluation for HPCI Drain Line Bypass Flow  
NF-36175   Single Line Diagram - Station Connections                           85
03/16/2016  
104B2506   Connection Diagram - Control Rod Drive Position Indicator Probe
1515939  
NH-46250   P&ID - High Pressure Coolant Injection System                       83
Question Raised on CRD 46-19  
NE-36399-9 Essential Bus Transfer Circuit - Division I                         77
03/16/2016  
NE-36399-9B Essential Bus Transfer Circuits - Division II                       78
1516098  
NF-36061   Equipment Location - Turbine Building EL 951-0                   76
Actions for when LCO 3.4.8, RA A.1 not met Unclear  
NF-36750   Standby Diesel Generator Building                                   8
03/17/2016  
NH-36241-1 Reactor Pressure Relief P&ID                                       78
1516101  
NH-36051   P&ID Diesel Oil System                                             85
Core Spray Motor Cooling Design Basis Question  
NH-178639-1 Levee Alignment and Bin Wall Plan                                   4
03/17/2016  
NF-119034-1 #11/#12 DG Fuel Oil System Isometric                               78
1516105  
NH-36253   P&ID Standby Liquid Control System                                 80
HPCI SR Test Inconsistent with TS Bases  
NH-36249   P&ID (Steam Side) High Pressure Coolant Injection System           82
03/17/2016  
NX-13142-42 Primary Steam & HPCI System                                         78
1516106  
                                          3
RCIC Surveillance Required Test Inconsistent with TS Bases  
03/17/2016  
CORRECTIVE ACTION PROGRAM DOCUMENTS REVIEWED  
Number  
Description or Title  
Date  
952310  
M91064A Quarterly Backflushing of Residual Heat Removal  
system and Core Spray Pump Motors  
07/27/1991
01196451  
CDBI EDG Base Tank Volume Calculation CA 90-023  
09/03/2003  
01355853  
Update UFSAR for External Flooding Description Discrepancy  
10/22/2012  
01414416  
Diesel Fuel Oil Temperature in Fuel Oil Transfer House is not  
Known
12/17/2013  
01420875-03  
Condition Evaluation on EDG Base Tank Level Issues  
04/04/2014  
01424477  
Appendix R Fire Strategy for Fire Area XII incorrect  
03/27/2014  
1478798  
EG Transfer Relay not Classified as Safety Related  
05/13/2015  
1484554  
RHRSW-29-2 Handwheel/Stem Sheared off  
06/29/2015  
1502700  
Catastrophic fail of MO-1900  
11/19/2015  
DRAWINGS  
Number  
Description or Title  
Revision  
NF-36175  
Single Line Diagram - Station Connections  
85  
104B2506  
Connection Diagram - Control Rod Drive Position Indicator Probe
NH-46250  
P&ID - High Pressure Coolant Injection System  
83  
NE-36399-9  
Essential Bus Transfer Circuit - Division I  
77  
NE-36399-9B  
Essential Bus Transfer Circuits - Division II  
78  
NF-36061  
Equipment Location - Turbine Building EL 951-0  
76  
NF-36750  
Standby Diesel Generator Building  
8  
NH-36241-1  
Reactor Pressure Relief P&ID  
78  
NH-36051  
P&ID Diesel Oil System  
85  
NH-178639-1  
Levee Alignment and Bin Wall Plan  
4  
NF-119034-1  
#11/#12 DG Fuel Oil System Isometric  
78  
NH-36253  
P&ID Standby Liquid Control System  
80  
NH-36249  
P&ID (Steam Side) High Pressure Coolant Injection System  
82  
NX-13142-42  
Primary Steam & HPCI System  
78  


MODIFICATIONS
4
Number         Description or Title                                           Revision
MODIFICATIONS  
EC-14065       RHRSW Motor Cooler Strainers                                     1
Number  
EC-20887       LT-5200 River Level Setpoint Change for Upper Value
Description or Title  
EC-21934       Evaluation of Corrosion Found in the 11 EDG Coolant               0
Revision  
              Expansion Tank
EC-14065  
EC-21999       Equivalency Evaluation: RHRSW-17 is the emergency injection       0
RHRSW Motor Cooler Strainers  
              check valve for the RHR to RSW crosstie
1  
EC-22008       Monticello 125V #12 Battery Modified Performance Test Profile     0
EC-20887  
EC-22414       SQUG Evaluation of Diesel Oil Service Pump P-77                   0
LT-5200 River Level Setpoint Change for Upper Value  
EC-23085       EDG Fuel Oil Train Separation                                     0
EC-23272       Revise EDG Base Tank Fuel Oil Level Calc 90-023
EC-21934  
EC-23616       Revise Setpoints for Relays 97-29 and 97-31                       0
Evaluation of Corrosion Found in the 11 EDG Coolant  
EC-23857       Recirc Pump Seal Water Piping                                     0
Expansion Tank  
EC-25889       Operating with HPCI CV-2043 (Steam Trap Bypass) Open             0
0
EC-25266       EDG Fuel Oil Separation                                           0
EC-21999  
OTHER DOCUMENTS
Equivalency Evaluation: RHRSW-17 is the emergency injection  
                                                                              Date or
check valve for the RHR to RSW crosstie  
Number        Description or Title                                           Revision
0
10040-A-020   Technical Specification for Steel Roof Deck                       2
EC-22008  
FOI 91-0265   Qualification of the EDG Building Roof for Accumulated Snow   04/18/1994
Monticello 125V #12 Battery Modified Performance Test Profile  
              Load
0  
FG-E-SE-03     50.59 Resource Manual                                             5
EC-22414  
WO 00505386-30 EC23085 Pre-Op Testing Division I                             05/06/2015
SQUG Evaluation of Diesel Oil Service Pump P-77  
WO 00505386-29 EC23085 Pre-Op Testing Division II                           04/26/2015
0  
257HA354       Technical Specification for High Pressure Coolant Injection       2
EC-23085  
              System
EDG Fuel Oil Train Separation
G-EK-1-45     Cold Shutdown Capability Report                               04/22/1981
0  
SRI 95-002     Core Spray Pump Motor Without Water Cooling                   09/28/1995
EC-23272  
EE 25506       RFO27 Decay Heat Evaluation
Revise EDG Base Tank Fuel Oil Level Calc 90-023  
PROCEDURES
Number         Description or Title                                           Revision
EC-23616  
0075           Control Rod Drive Coupling Test                                   19
Revise Setpoints for Relays 97-29 and 97-31  
C.06-006-C-01 Diesel Oil Storage Tank T-44 Hi Low Level                         6
0  
C.06-006-C-02 Diesel Oil Storage Tank T-44 Low-Low Level                       6
EC-23857  
C.06-006-C-03 Division 1 EDG P-160A & P-160C Not Running                       6
Recirc Pump Seal Water Piping  
C.06-006-C-06 Diesel Gen Tank T-160A Level/Flow Low                             4
0  
2014-02       Turbine Building Outside                                         27
EC-25889  
A.6           Acts of Nature                                                   53
Operating with HPCI CV-2043 (Steam Trap Bypass) Open  
0255-17-ID-1   Master Alternate Nitrogen System Tests                           25
0  
0255-17-ID-15 SRV RV-71D and RV-2-71G Pneumatic Supply Leakage Test             13
EC-25266  
Ops Man       Loss of Normal Shutdown Cooling                                  15
EDG Fuel Oil Separation  
C.4-B.03.04.A
0  
1339           ECCS Pump Motor Cooler Flush                                     35
9111-01       Shutdown Cooling Division I Protected System Ticket Checklist     6
OTHER DOCUMENTS  
2270           Critical Safety System Checklist                                 11
Number
OWI-02.03     Operator Rounds, Turbine Building West                           64
Description or Title  
Ops Man B.     Core Spray Cooling System                                        42
Date or
03.01-05
Revision  
0255-05-1A-1-2 B RHR SW Quarterly Pump and Valve Test                           82
10040-A-020  
                                            4
Technical Specification for Steel Roof Deck  
2  
FOI 91-0265  
Qualification of the EDG Building Roof for Accumulated Snow  
Load
04/18/1994  
FG-E-SE-03  
50.59 Resource Manual  
5  
WO 00505386-30 EC23085 Pre-Op Testing Division I  
05/06/2015  
WO 00505386-29 EC23085 Pre-Op Testing Division II  
04/26/2015  
257HA354  
Technical Specification for High Pressure Coolant Injection  
System  
2
G-EK-1-45  
Cold Shutdown Capability Report  
04/22/1981  
SRI 95-002  
Core Spray Pump Motor Without Water Cooling  
09/28/1995  
EE 25506  
RFO27 Decay Heat Evaluation
PROCEDURES  
Number  
Description or Title  
Revision  
0075  
Control Rod Drive Coupling Test  
19  
C.06-006-C-01  
Diesel Oil Storage Tank T-44 Hi Low Level  
6  
C.06-006-C-02  
Diesel Oil Storage Tank T-44 Low-Low Level  
6  
C.06-006-C-03  
Division 1 EDG P-160A & P-160C Not Running  
6  
C.06-006-C-06  
Diesel Gen Tank T-160A Level/Flow Low
4  
2014-02  
Turbine Building Outside  
27  
A.6  
Acts of Nature  
53  
0255-17-ID-1  
Master Alternate Nitrogen System Tests  
25  
0255-17-ID-15  
SRV RV-71D and RV-2-71G Pneumatic Supply Leakage Test  
13  
Ops Man  
C.4-B.03.04.A  
Loss of Normal Shutdown Cooling
15
1339
ECCS Pump Motor Cooler Flush  
35  
9111-01  
Shutdown Cooling Division I Protected System Ticket Checklist  
6  
2270  
Critical Safety System Checklist  
11  
OWI-02.03
Operator Rounds, Turbine Building West  
64  
Ops Man B.  
03.01-05  
Core Spray Cooling System
42
0255-05-1A-1-2  
B RHR SW Quarterly Pump and Valve Test  
82  


                                                                      April 19, 2016
Mr. Peter A. Gardner
April 19, 2016  
Site Vice President
Mr. Peter A. Gardner  
Monticello Nuclear Generating Plant
Site Vice President  
Northern States Power Company, Minnesota
Monticello Nuclear Generating Plant  
2807 West County Road 75
Northern States Power Company, Minnesota  
Monticello, MN 55362-9637
2807 West County Road 75  
SUBJECT: MONTICELLO NUCLEAR GENERATING PLANT - EVALUATIONS OF CHANGES, TESTS,
Monticello, MN 55362-9637  
                    AND EXPERIMENTS AND PERMANENT PLANT MODIFICATIONS BASELINE
SUBJECT: MONTICELLO NUCLEAR GENERATING PLANT - EVALUATIONS OF CHANGES, TESTS,  
                    INSPECTION REPORT 05000263/2016008
AND EXPERIMENTS AND PERMANENT PLANT MODIFICATIONS BASELINE  
Dear Mr. Gardner:
INSPECTION REPORT 05000263/2016008  
On March 24, 2016, the U.S. Nuclear Regulatory Commission (NRC) completed an Evaluations of
Dear Mr. Gardner:  
Changes, Tests, and Experiments and Permanent Plant Modifications inspection at your Monticello
On March 24, 2016, the U.S. Nuclear Regulatory Commission (NRC) completed an Evaluations of  
Nuclear Generating Plant. The enclosed inspection report documents the inspection results which were
Changes, Tests, and Experiments and Permanent Plant Modifications inspection at your Monticello  
discussed on March 24, 2016, with Mr. M. Lingenfelter and other members of your staff.
Nuclear Generating Plant. The enclosed inspection report documents the inspection results which were  
The inspection examined activities conducted under your license as they relate to safety and compliance
discussed on March 24, 2016, with Mr. M. Lingenfelter and other members of your staff.  
with the Commissions rules and regulations and with the conditions of your license. The inspectors
The inspection examined activities conducted under your license as they relate to safety and compliance  
reviewed selected procedures and records, observed activities, and interviewed personnel.
with the Commissions rules and regulations and with the conditions of your license. The inspectors  
No findings were identified during this inspection.
reviewed selected procedures and records, observed activities, and interviewed personnel.  
In accordance with Title 10 of the Code of Federal Regulations (10 CFR) 2.390, Public Inspections,
No findings were identified during this inspection.  
Exemptions, Requests for Withholding, of the NRC's "Rules of Practice," a copy of this letter, its
In accordance with Title 10 of the Code of Federal Regulations (10 CFR) 2.390, Public Inspections,  
enclosure, and your response (if any) will be available electronically for public inspection in the NRCs
Exemptions, Requests for Withholding, of the NRC's "Rules of Practice," a copy of this letter, its  
Public Document Room or from the Publicly Available Records (PARS) component of the NRC's
enclosure, and your response (if any) will be available electronically for public inspection in the NRCs  
Agencywide Documents Access and Management System (ADAMS). ADAMS is accessible from the
Public Document Room or from the Publicly Available Records (PARS) component of the NRC's  
NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
Agencywide Documents Access and Management System (ADAMS). ADAMS is accessible from the  
                                                                          Sincerely,
NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).  
                                                                          /RA/
Sincerely,  
                                                                          Robert C. Daley, Chief
                                                                          Engineering Branch 3
/RA/  
                                                                          Division of Reactor Safety
Robert C. Daley, Chief  
Docket No. 50-263
Engineering Branch 3  
License No. DPR-22
Division of Reactor Safety  
Enclosure:
Docket No. 50-263  
IR 05000263/2016008
License No. DPR-22  
cc: Distribution via LISTSERV
Enclosure:  
DISTRIBUTION:
  IR 05000263/2016008  
Jeremy Bowen                                             Darrell Roberts                                  DRPIII
cc: Distribution via LISTSERV  
RidsNrrPMPalisades Resource                               Richard Skokowski                                DRSIII
DISTRIBUTION:  
RidsNrrDorlLpl3-1 Resource                               Allan Barker                                     Jim Clay
Jeremy Bowen  
RidsNrrDirsIrib Resource                                  Carole Ariano                                    Carmen Olteanu
RidsNrrPMPalisades Resource  
Cynthia Pederson                                          Linda Linn                                        ROPreports.Resource@nrc.gov
RidsNrrDorlLpl3-1 Resource
ADAMS Accession Number ML16113A346
RidsNrrDirsIrib Resource
    Publicly Available                       Non-Publicly Available                       Sensitive                 Non-Sensitive
Cynthia Pederson
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
Darrell Roberts
  OFFICE             RIII                                 RIII                           RIII                               RIII
Richard Skokowski
  NAME               ADahbur:cl                           RDaley
Allan Barker  
  DATE               04/19/16                             04/19/16
Carole Ariano
                                                            OFFICIAL RECORD COPY
Linda Linn
DRPIII
DRSIII
Jim Clay  
Carmen Olteanu  
ROPreports.Resource@nrc.gov  
ADAMS Accession Number ML16113A346  
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  
RIII  
NAME  
ADahbur:cl  
RDaley  
DATE  
04/19/16  
04/19/16  
OFFICIAL RECORD COPY
}}
}}

Latest revision as of 00:55, 10 January 2025

Evaluations or Changes, Tests, and Experiments and Permanent Plant Modifications Baseline Inspection Report 05000263/2016008
ML16113A346
Person / Time
Site: Monticello Xcel Energy icon.png
Issue date: 04/19/2016
From: Robert Daley
Engineering Branch 3
To: Gardner P
Northern States Power Company, Minnesota
References
IR 2016008
Download: ML16113A346 (12)
v • d • e


See also: IR 05000263/2016008

Text

UNITED STATES

NUCLEAR REGULATORY COMMISSION

REGION III

2443 WARRENVILLE RD. SUITE 210

LISLE, IL 60532-4352

April 19, 2016

Mr. Peter A. Gardner

Site Vice President

Monticello Nuclear Generating Plant

Northern States Power Company, Minnesota

2807 West County Road 75

Monticello, MN 55362-9637

SUBJECT: MONTICELLO NUCLEAR GENERATING PLANT - EVALUATIONS OF CHANGES,

TESTS, AND EXPERIMENTS AND PERMANENT PLANT MODIFICATIONS

BASELINE INSPECTION REPORT 05000263/2016008

Dear Mr. Gardner:

On March 24, 2016, the U.S. Nuclear Regulatory Commission (NRC) completed an Evaluations

of Changes, Tests, and Experiments and Permanent Plant Modifications inspection at your

Monticello Nuclear Generating Plant. The enclosed inspection report documents the inspection

results which were discussed on March 24, 2016, with Mr. M. Lingenfelter and other members

of your staff.

The inspection examined activities conducted under your license as they relate to safety and

compliance with the Commissions rules and regulations and with the conditions of your license.

The inspectors reviewed selected procedures and records, observed activities, and interviewed

personnel.

No findings were identified during this inspection.

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, 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/

Robert C. Daley, Chief

Engineering Branch 3

Division of Reactor Safety

Docket No. 50-263

License No. DPR-22

Enclosure:

IR 05000263/2016008

cc: Distribution via LISTSERV

Enclosure

U. S. NUCLEAR REGULATORY COMMISSION

REGION III

Docket No:

50-263

License No:

DPR-22

Report No:

05000263/2016008

Licensee:

Northern States Power Company, Minnesota

Facility:

Monticello Nuclear Generating Plant

Location:

Monticello, MN

Dates:

February 29 thru March 24, 2016

Inspectors:

Alan Dahbur, Senior Reactor Inspector (Lead)

Jorge J. Corujo-Sandín, Reactor Inspector

Michael A. Jones, Reactor Inspector

Approved by:

Robert C. Daley, Chief

Engineering Branch 3

Division of Reactor Safety

2

SUMMARY

Inspection Report 05000263/2016008; 02/29/2016 - 03/24/2016; Monticello Nuclear Generating

Plant; Evaluations of Changes, Tests, and Experiments and Permanent Plant Modifications.

This report covers a 2-week announced baseline inspection on evaluations of changes,

tests, and experiments and permanent plant modifications. The inspection was conducted

by Region III based engineering inspectors. The U.S. Nuclear Regulatory Commissions

program for overseeing the safe operation of commercial nuclear power reactors is described

in NUREG-1649, Reactor Oversight Process, Revision 5, dated February 2014.

NRC-Identified and Self-Revealed Findings

No findings were identified.

Licensee-Identified Violations

No violations were identified.

3

REPORT DETAILS

1.

REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity

1R17 Evaluations of Changes, Tests, and Experiments and Permanent Plant Modifications

(71111.17T)

.1

Evaluation of Changes, Tests, and Experiments

a.

Inspection Scope

The inspectors reviewed 7 safety evaluations performed pursuant to Title 10, Code of

Federal Regulations (CFR), Part 50, Section 59, to determine if the evaluations were

adequate, and that prior U.S. Nuclear Regulatory Commission (NRC) approval was

obtained as appropriate. The inspectors also reviewed 16 screenings where licensee

personnel had determined that a 10 CFR 50.59 evaluation was not necessary. The

inspectors reviewed these documents to determine if:

the changes, tests, and experiments performed were evaluated in accordance

with 10 CFR 50.59 and that sufficient documentation existed to confirm that a

license amendment was not required;

the safety issue requiring the change, tests or experiment was resolved;

the licensee conclusions for evaluations of changes, tests, and experiments were

correct and consistent with 10 CFR 50.59; and

the design and licensing basis documentation was updated to reflect the change.

The inspectors used, in part, Nuclear Energy Institute Document 96-07, Guidelines for

10 CFR 50.59 Implementation, Revision 1, to determine acceptability of the completed

evaluations, and screenings. The Nuclear Energy Institute document was endorsed by

the NRC in Regulatory Guide 1.187, Guidance for Implementation of 10 CFR 50.59,

Changes, Tests, and Experiments, dated November 2000. The inspectors also

consulted Part 9900 of the NRC Inspection Manual, 10 CFR Guidance for

10 CFR 50.59, Changes, Tests, and Experiments.

This inspection constituted 4 samples of evaluations and 13 samples of screenings

and/or applicability determinations as defined in Inspection Procedure 71111.17-04.

b.

Findings

(Open) Unresolved Item 05000263/2016008-01, Failure to provide acceptable Alternate

Methods of Decay Heat Removal

Introduction: The inspectors identified an Unresolved Item associated with Technical

Specification (TS) 3.4.8, Residual Heat Removal (RHR) Shutdown Cooling System -

Cold Shutdown. Specifically, the licensee failed to verify that the capability of the

alternate methods of decay heat removal described in Operations Manual C.4-B.03.04.A,

Loss of Normal Shutdown Cooling, were adequate to combat a loss of shutdown cooling

resulting from the loss of one or two RHR subsystems while in MODE 4 with high decay

heat load.

4

Description: The Limiting Condition for Operation (LCO) 3.4.8 of TS Residual Heat

Removal Shutdown Cooling System - Cold Shutdown, required in Mode 4, two RHR

shutdown cooling subsystems shall be operable, and, with no recirculation pump in

operation, at least one RHR shutdown cooling subsystem shall be in operation. The

TS Bases Section 3.4.8, indicated that an operable RHR shutdown cooling subsystem

consisted of one operable RHR pump, one heat exchanger, the associated piping and

valves, and the necessary portions of the RHR Service Water System System capable

of providing cooling water to the heat exchanger. The TS Bases Section 3.4.8 further

indicated that the two subsystems have a common suction source and were allowed to

have a common heat exchanger and common discharge piping. Thus, to meet the LCO,

both pumps in one loop or one pump in each of the two loops must be operable. Since

the piping and heat exchangers were passive components that were assumed not to fail,

they were allowed to be common to both subsystems.

When TS 3.4.8, LCO could not be met, Condition A, for one or two RHR shutdown

cooling subsystems inoperable, the Required Action was to, verify an alternate

method of decay heat removal was available for each inoperable RHR shutdown

cooling subsystem. The completion time for the required action was 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, and

once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter. The TS Bases 3.4.8 for Condition A indicated that with

one of the two required RHR shutdown cooling subsystems inoperable, the remaining

subsystem was capable of providing the required decay heat removal. However, the

overall reliability was reduced, therefore, an alternate method of decay heat removal

must be provided. With both RHR shutdown cooling subsystems inoperable, an

alternate method of decay heat removal must be provided in addition to that provided

for the initial RHR shutdown cooling subsystem inoperability. This was to ensure the

re-establishment of backup decay heat removal capabilities, similar to the requirements

of the LCO. The bases further stated that the required cooling capacity of the alternate

method should be ensured by verifying (by calculation or demonstration) its capability to

maintain or reduce temperature. Alternate methods that can be used included (but not

limited to) the Reactor Water Cleanup System by itself or using feed and bleed in

combination with Control Rod Drive System or Condensate/Feed Systems.

Abnormal Procedure, Operations Manual C.4-B.03.04.A, Loss of Normal Shutdown

Cooling, provided instructions for establishing alternate methods for decay heat

removal. The inspectors noticed that except for the alternate method as described

below in the G-EK-1-45, the licensee was not able to show by calculation or

demonstration that the systems and methods credited in this procedure would be

capable of providing sufficient heat removal capability or appropriate levels of

redundancy as required by TS 3.4.8.

The G-EK-1-45 was a General Electric Letter to Northern States Power, Subject: Cold

Shutdown Capability Report, dated April 22, 1981. This letter provided a report which

described the capability of the Monticello Nuclear Generating Plant to achieve cold

shutdown using only safety class systems and assuming the worst single failure. The

alternate shutdown decay heat removal method used in the report credited combinations

of the RHR pumps and heat exchangers in the suppression pool cooling mode of RHR

to ensure suppression pool water temperatures were below the design limit. This

method utilized the core spray system and safety relief valves to circulate reactor

inventory to remove decay heat from the reactor.

5

The inspectors noted that calculations supporting the above alternate strategy utilized an

RHR subsystem that could be inoperable and/or unavailable and therefore may not be

credited to comply with TS 3.4.8. Specifically, the inspectors were concerned that while

the plant was in mode 4, with a credited one subsystem inoperable, the licensees

credited alternate decay heat removal method that relied on an RHR subsystem, to

perform the required suppression pool cooling function. The inspectors were concerned

that relying on the only operable RHR subsystem for the alternate method did not meet

the intent of the TS requirement as described in the TS Bases. Furthermore, the

inspectors noticed for Mode 4 with two RHR subsystems inoperable, the licensee failed

to verify by calculation or demonstrations that two additional redundant alternate decay

heat removal methods existed with sufficient capacity to maintain the average reactor

coolant temperature below 212 degrees Fahrenheit.

During the inspection, the licensee indicated that the Boiling Reactor Owners Group was

in the process of developing a draft TS Task Force Traveler to address the requirement

of TS 3.4.8 and its Bases.

Based on the information above, the inspectors were concerned that the plant

Operations Manual was inadequate and failed to include alternate decay heat removal

methods that would enable the licensee to comply with the requirement of TS 3.4.8. The

Operations Manual was required per TS 5.4.1 Procedures, which required that written

procedures shall be established, implemented, and maintained covering the emergency

operating procedures. The inspectors determined that this issue was unresolved

pending the actions by the licensee and the Boiling Reactor Owners Group and the NRC

review of these actions. The licensee entered the inspectors concerns into their

Corrective Action Program as AR 01516098. (URI 05000263/2016008-01, Failure to

provide acceptable Alternate Methods of Decay Heat Removal)

.2

Permanent Plant Modifications

a.

Inspection Scope

The inspectors reviewed seven permanent plant modifications that had been installed

in the plant during the last 3 years. This review included in-plant walkdowns portions

of the high-pressure coolant injection steam drain line system, the Emergency Diesel

Generator Fuel Oil Transfer System, including the Diesel Fuel Oil pump house, the new

diesel fuel oil pumps installed in the day tank room, and portions of the fuel oil storage

tank tornado missile protection modifications. The modifications were selected based

upon risk significance, safety significance, and complexity. The inspectors reviewed the

modifications selected to determine if:

the supporting design and licensing basis documentation was updated;

the changes were in accordance with the specified design requirements;

the procedures and training plans affected by the modification have been

adequately updated;

the test documentation as required by the applicable test programs has been

updated; and

post-modification testing adequately verified system operability and/or

functionality.

6

The inspectors also used applicable industry standards to evaluate acceptability of the

modifications. The list of modifications and other documents reviewed by the inspectors

is included as an Attachment to this report.

This inspection constituted eight permanent plant modification samples as defined in

Inspection Procedure 71111.17-04.

4.

OTHER ACTIVITIES

4OA2 Problem Identification and Resolution

.1

Routine Review of Condition Reports

a.

Inspection Scope

The inspectors reviewed several corrective action process documents that identified or

were related to Title 10 of the Code of Federal Regulations, Part 50.59 evaluations and

permanent plant modifications. The inspectors reviewed these documents to evaluate

the effectiveness of corrective actions related to permanent plant modifications and

evaluations of changes, tests, and experiments. In addition, corrective action

documents written on issues identified during the inspection were reviewed to verify

adequate problem identification and incorporation of the problems into the corrective

action system. The specific corrective action documents that were sampled and

reviewed by the inspectors are listed in the attachment to this report.

b.

Findings

No findings were identified.

4OA6 Management Meetings

.1

Exit Meeting Summary

On March 24, 2016, the inspectors presented the inspection results to Mr. M. Lingenfelter,

and other members of the licensee staff. The licensee personnel acknowledged the

inspection results presented and did not identify any proprietary content. The inspectors

confirmed that all proprietary material provided to the inspection team was identified and

will be dispositioned in accordance with applicable processes.

ATTACHMENT: SUPPLEMENTAL INFORMATION

Attachment

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee

M. Lingenfelter, Director of Engineering

A. Gonnering, Design Engineering

M. Kelly, Performance Assurance Manager

J. Gausman, Engineering

A. Ward, Regulatory Affairs Manager

T. Hurrle, Design Engineering Manager

B. Halvorson, Engineering

A. Kouba, Regulatory Affairs

D. Alstad, Design Engineer

E. Watzel, Electrical Design Engineering Supervisor

U.S. Nuclear Regulatory Commission

P. Zurawski, Senior Resident Inspector

P. LaFlamme, Acting Senior Resident Inspector

D. Krause, Resident Inspector

LIST OF ITEMS OPENED, CLOSED AND DISCUSSED

Opened 05000263/2016008-01 URI

Failure to provide acceptable Alternate Methods of Decay Heat

Removal (Section 1R17.1b)

Closed and Discussed

None

LIST OF ACRONYMS USED

ADAMS

Agencywide Documents Access and Management System

CFR

Code of Federal Regulations

LCO

Limiting Condition for Operation

NRC

U.S. Nuclear Regulatory Commission

PARS

Publicly Available Records System

RHR

Residual Heat Removal

TS

Technical Specifications

2

LIST OF DOCUMENTS REVIEWED

The following is a list of documents reviewed during the inspection. Inclusion on this list does

not imply that the NRC inspectors reviewed the documents in their entirety, but rather, that

selected sections of portions of the documents were evaluated as part of the overall inspection

effort. Inclusion of a document on this list does not imply NRC acceptance of the document or

any part of it, unless this is stated in the body of the inspection report.

10 CFR 50.59 EVALUATIONS

Number

Description or Title

Revision

SCR-12-0559

HPCI Logic Change to Provide Margin to MO-2035 and #16 Battery

1

SCR-13-0554

External Flooding Protection Strategy Change

0

SCR-15-0202

Evaluation of EPG/SAG, Revision 3

0

SCR-16-0024

Disconnect Faulty 46-19 PIP Over-Travel Input

0

10 CFR 50.59 SCREENINGS

Number

Description or Title

Revision

SCR-13-0696

Revise EDG Base Tank Fuel Oil Level Calculation 90-023

0

SCR-14-0074

Time Delay Relay 97-29 and 97-31 Setpoint Change

0

SCR-14-0413

Temp Rev to C.6-006-A-01and C.6-006-A-02

SCR-14-0415

USAR-06.06 Revision

0

SCR-14-0421

EC 23981 EDG Fuel Oil Tank Vent Lines Missile Protection

0

SCR-14-0512

Safety and Seismic Classification of the DG/RF and DG/RV Relays

0

SCR-14-0542

RHRSW and Emergency Service Water TS Bases Changes

SCR-14-0591

Fuel Oil Separation

4

SCR-14-0593

Revise Calculation 94-086 on SRV Accumulation Allowable

Leakage Rates

0

SCR-15-0093

EDG ESW Basket Strainer Modification

SCR-15-0115

C.4-B.09.02.A Revision to Resolve CAP AR 01465720

1

SCR-15-0193

Room Heat Up Calculation Revisions for SBO

SCR-15-0291

Revise Maximum Volume of EDG Base Tank in 90-023

0

SCR-15-0292

Diesel Pump House Heat Up Calculation

0

CALCULATIONS

Number

Description or Title

Revision 03-089

Inservice Testing Acceptance Criteria

3 09-106

CSP Motor-Oil and Bearing Operating Temperatures without

Cooling Water

1 09-176

Evaluation for Debris Disposition in Supply Pipe and Motor Cooler

Tube

0 09-178

Time to reach the RHRSW Pump Motor Cooling Line Strainer

Limiting Pressure Differential

0 14-025

Instrument Setpoint Calculation - Time Delay for Transfer to EDG

on Loss of Voltage

0 90-023

EC 23085 - EDG Fuel Oil Train Separation

3 92-224

Emergency Diesel Generator Loading

6A 94-086

Max Allowed Leakage Rates and Test Acceptance Criteria for SRV

5

3

CORRECTIVE ACTION PROGRAM DOCUMENTS INITIATED DURING INSPECTION

Number

Description or Title

Date

1510936

Incomplete EC Record Copy

02/03/2016

1514133

Clarification for USAR Section 8.4.1.3

03/01/2016

1514202

Page missing from WO 00491265 Record

03/02/2016

1514369

Screening SCR 14-0421 Answered Question Incorrectly

03/03/2016

1514464

EDG Building Roof FOI 91-0265

03/03/2016

1515054

Bases for Procedure A.6 Contain Incorrect Statements

03/09/2016

1515688

Signs of Leakage around FO-11-3

03/15/2016

1515716

NRC not Provided with Latest Copy of EC23085

03/15/2016

1515907

Formal Evaluation for HPCI Drain Line Bypass Flow

03/16/2016

1515939

Question Raised on CRD 46-19

03/16/2016

1516098

Actions for when LCO 3.4.8, RA A.1 not met Unclear

03/17/2016

1516101

Core Spray Motor Cooling Design Basis Question

03/17/2016

1516105

HPCI SR Test Inconsistent with TS Bases

03/17/2016

1516106

RCIC Surveillance Required Test Inconsistent with TS Bases

03/17/2016

CORRECTIVE ACTION PROGRAM DOCUMENTS REVIEWED

Number

Description or Title

Date

952310

M91064A Quarterly Backflushing of Residual Heat Removal

system and Core Spray Pump Motors

07/27/1991

01196451

CDBI EDG Base Tank Volume Calculation CA 90-023

09/03/2003

01355853

Update UFSAR for External Flooding Description Discrepancy

10/22/2012

01414416

Diesel Fuel Oil Temperature in Fuel Oil Transfer House is not

Known

12/17/2013

01420875-03

Condition Evaluation on EDG Base Tank Level Issues

04/04/2014

01424477

Appendix R Fire Strategy for Fire Area XII incorrect

03/27/2014

1478798

EG Transfer Relay not Classified as Safety Related

05/13/2015

1484554

RHRSW-29-2 Handwheel/Stem Sheared off

06/29/2015

1502700

Catastrophic fail of MO-1900

11/19/2015

DRAWINGS

Number

Description or Title

Revision

NF-36175

Single Line Diagram - Station Connections

85

104B2506

Connection Diagram - Control Rod Drive Position Indicator Probe

NH-46250

P&ID - High Pressure Coolant Injection System

83

NE-36399-9

Essential Bus Transfer Circuit - Division I

77

NE-36399-9B

Essential Bus Transfer Circuits - Division II

78

NF-36061

Equipment Location - Turbine Building EL 951-0

76

NF-36750

Standby Diesel Generator Building

8

NH-36241-1

Reactor Pressure Relief P&ID

78

NH-36051

P&ID Diesel Oil System

85

NH-178639-1

Levee Alignment and Bin Wall Plan

4

NF-119034-1

  1. 11/#12 DG Fuel Oil System Isometric

78

NH-36253

P&ID Standby Liquid Control System

80

NH-36249

P&ID (Steam Side) High Pressure Coolant Injection System

82

NX-13142-42

Primary Steam & HPCI System

78

4

MODIFICATIONS

Number

Description or Title

Revision

EC-14065

RHRSW Motor Cooler Strainers

1

EC-20887

LT-5200 River Level Setpoint Change for Upper Value

EC-21934

Evaluation of Corrosion Found in the 11 EDG Coolant

Expansion Tank

0

EC-21999

Equivalency Evaluation: RHRSW-17 is the emergency injection

check valve for the RHR to RSW crosstie

0

EC-22008

Monticello 125V #12 Battery Modified Performance Test Profile

0

EC-22414

SQUG Evaluation of Diesel Oil Service Pump P-77

0

EC-23085

EDG Fuel Oil Train Separation

0

EC-23272

Revise EDG Base Tank Fuel Oil Level Calc 90-023

EC-23616

Revise Setpoints for Relays 97-29 and 97-31

0

EC-23857

Recirc Pump Seal Water Piping

0

EC-25889

Operating with HPCI CV-2043 (Steam Trap Bypass) Open

0

EC-25266

EDG Fuel Oil Separation

0

OTHER DOCUMENTS

Number

Description or Title

Date or

Revision

10040-A-020

Technical Specification for Steel Roof Deck

2

FOI 91-0265

Qualification of the EDG Building Roof for Accumulated Snow

Load

04/18/1994

FG-E-SE-03

50.59 Resource Manual

5

WO 00505386-30 EC23085 Pre-Op Testing Division I

05/06/2015

WO 00505386-29 EC23085 Pre-Op Testing Division II

04/26/2015

257HA354

Technical Specification for High Pressure Coolant Injection

System

2

G-EK-1-45

Cold Shutdown Capability Report

04/22/1981

SRI 95-002

Core Spray Pump Motor Without Water Cooling

09/28/1995

EE 25506

RFO27 Decay Heat Evaluation

PROCEDURES

Number

Description or Title

Revision

0075

Control Rod Drive Coupling Test

19

C.06-006-C-01

Diesel Oil Storage Tank T-44 Hi Low Level

6

C.06-006-C-02

Diesel Oil Storage Tank T-44 Low-Low Level

6

C.06-006-C-03

Division 1 EDG P-160A & P-160C Not Running

6

C.06-006-C-06

Diesel Gen Tank T-160A Level/Flow Low

4

2014-02

Turbine Building Outside

27

A.6

Acts of Nature

53

0255-17-ID-1

Master Alternate Nitrogen System Tests

25

0255-17-ID-15

SRV RV-71D and RV-2-71G Pneumatic Supply Leakage Test

13

Ops Man

C.4-B.03.04.A

Loss of Normal Shutdown Cooling

15

1339

ECCS Pump Motor Cooler Flush

35

9111-01

Shutdown Cooling Division I Protected System Ticket Checklist

6

2270

Critical Safety System Checklist

11

OWI-02.03

Operator Rounds, Turbine Building West

64

Ops Man B.

03.01-05

Core Spray Cooling System

42

0255-05-1A-1-2

B RHR SW Quarterly Pump and Valve Test

82

April 19, 2016

Mr. Peter A. Gardner

Site Vice President

Monticello Nuclear Generating Plant

Northern States Power Company, Minnesota

2807 West County Road 75

Monticello, MN 55362-9637

SUBJECT: MONTICELLO NUCLEAR GENERATING PLANT - EVALUATIONS OF CHANGES, TESTS,

AND EXPERIMENTS AND PERMANENT PLANT MODIFICATIONS BASELINE

INSPECTION REPORT 05000263/2016008

Dear Mr. Gardner:

On March 24, 2016, the U.S. Nuclear Regulatory Commission (NRC) completed an Evaluations of

Changes, Tests, and Experiments and Permanent Plant Modifications inspection at your Monticello

Nuclear Generating Plant. The enclosed inspection report documents the inspection results which were

discussed on March 24, 2016, with Mr. M. Lingenfelter and other members of your staff.

The inspection examined activities conducted under your license as they relate to safety and compliance

with the Commissions rules and regulations and with the conditions of your license. The inspectors

reviewed selected procedures and records, observed activities, and interviewed personnel.

No findings were identified during this inspection.

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, 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/

Robert C. Daley, Chief

Engineering Branch 3

Division of Reactor Safety

Docket No. 50-263

License No. DPR-22

Enclosure:

IR 05000263/2016008

cc: Distribution via LISTSERV

DISTRIBUTION:

Jeremy Bowen

RidsNrrPMPalisades Resource

RidsNrrDorlLpl3-1 Resource

RidsNrrDirsIrib Resource

Cynthia Pederson

Darrell Roberts

Richard Skokowski

Allan Barker

Carole Ariano

Linda Linn

DRPIII

DRSIII

Jim Clay

Carmen Olteanu

ROPreports.Resource@nrc.gov

ADAMS Accession Number ML16113A346

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

RIII

NAME

ADahbur:cl

RDaley

DATE

04/19/16

04/19/16

OFFICIAL RECORD COPY

Retrieved from "https://kanterella.com/w/index.php?title=ML16113A346&oldid=5173541"