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| issue date = 05/05/2008
| issue date = 05/05/2008
| title = IR 05000220-08-002, 05000410-08-002; 01/01/08 - 03/31/08; Nine Mile Point Nuclear Station, Units 1 and 2; Surveillance Testing
| title = IR 05000220-08-002, 05000410-08-002; 01/01/08 - 03/31/08; Nine Mile Point Nuclear Station, Units 1 and 2; Surveillance Testing
| author name = Dentel G T
| author name = Dentel G
| author affiliation = NRC/RGN-I/DRP/PB1
| author affiliation = NRC/RGN-I/DRP/PB1
| addressee name = Polson K J
| addressee name = Polson K
| addressee affiliation = Nine Mile Point Nuclear Station, LLC
| addressee affiliation = Nine Mile Point Nuclear Station, LLC
| docket = 05000220, 05000410
| docket = 05000220, 05000410
Line 18: Line 18:


=Text=
=Text=
{{#Wiki_filter:UNITED STATES NUCLEAR REGULATORY COMMISSION REGION I 475 ALLENDALE ROAD KING OF PRUSSIA, PA 19406-1415 May 5, 2008  
{{#Wiki_filter:May 5, 2008


Mr. Keith Vice President Nine Mile Point
==SUBJECT:==
 
NINE MILE POINT NUCLEAR STATION - NRC INTEGRATED INSPECTION REPORT 05000220/2008002 and 05000410/2008002
Nine Mile Point Nuclear Station, LLC
 
P.O. Box 63
 
Lycoming, NY 13093
 
SUBJECT: NINE MILE POINT NUCLEAR STATION - NRC INTEGRATED INSPECTION REPORT 05000220/2008002 and 05000410/2008002


==Dear Mr. Polson:==
==Dear Mr. Polson:==
On March 31, 2008, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection  
On March 31, 2008, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your Nine Mile Point Nuclear Station, Units 1 and 2. The enclosed integrated inspection report documents the inspection results discussed on April 11, 2008, with you and members of your staff.
 
at your Nine Mile Point Nuclear Station, Units 1 and 2. The enclosed integrated inspection  
 
report documents the inspection results discussed on April 11, 2008, with you and members of  
 
your staff.
 
The inspection examined activities conducted under your license as they relate to safety and compliance with the Commission's rules and regulations and with the conditions of your license.
 
The inspectors reviewed selected procedures and records, observed activities, and interviewed
 
personnel.
 
This report documents one self-revealing finding of very low safety significance (Green). The
 
finding was determined to involve a violation of NRC requirements. However, because of its very low safety significance and because it was entered into your corrective action program (CAP), the NRC is treating this finding as a non-cited violation in accordance with Section VI.A.1
 
of the NRC's Enforcement Policy. If you contest the non-cited violation noted in this report, you
 
should provide a response with the basis for your denial, within 30 days of the date of this
 
inspection report, to the Nuclear Regulatory Commission, ATTN.: Document Control Desk, Washington, D.C. 20555-0001; with copies to the Regional Administrator, Region I; the Director, Office of Enforcement; U.S. Nuclear Regulatory Commission, Washington, D.C. 20555-001; and the NRC Senior Resident Inspector at Nine Mile Point Nuclear Station.
 
In accordance with 10 CFR Part 2.390 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 NRC Public Document Room or from the Public ly Available Records (PARS) component of the NRC's document 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/ Glenn T. Dentel, Chief
 
Projects Branch 1
 
Division of Reactor Projects
 
Docket No.: 50-220, 50-410 License No.: DPR-63, NPF-69
 
===Enclosure:===
Inspection Report 05000220/2008002 and 05000410/2008002
 
===w/Attachment:===
Supplemental Information cc w/encl: M. Wallace, President, Constellation Generation
 
B. Barron, Senior Vice President and Chief Nuclear Officer
 
C. Fleming, Esquire, Senior Counsel, Constellation Energy Group, LLC


M. Wetterhahn, Esquire, Winston and Strawn
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.


T. Syrell, Director, Licensing, Nine Mile Point Nuclear Station
The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel.


P. Tonko, President and CEO, New York State Energy Research and Development Authority
This report documents one self-revealing finding of very low safety significance (Green). The finding was determined to involve a violation of NRC requirements. However, because of its very low safety significance and because it was entered into your corrective action program (CAP), the NRC is treating this finding as a non-cited violation in accordance with Section VI.A.1 of the NRCs Enforcement Policy. If you contest the non-cited violation noted in this report, you should provide a response with the basis for your denial, within 30 days of the date of this inspection report, to the Nuclear Regulatory Commission, ATTN.: Document Control Desk, Washington, D.C. 20555-0001; with copies to the Regional Administrator, Region I; the Director, Office of Enforcement; U.S. Nuclear Regulatory Commission, Washington, D.C. 20555-001; and the NRC Senior Resident Inspector at Nine Mile Point Nuclear Station.


J. Spath, Program Director, New York State Energy Research and Development Authority
In accordance with 10 CFR Part 2.390 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 NRC Public Document Room or from the Publicly Available Records (PARS) component of the NRCs document system (ADAMS). ADAMS is accessible from the NRC Web Site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).


P. D. Eddy, Electric Division, NYS Department of Public Service
Sincerely,
/RA/


C. Donaldson, Esquire, Assistant Attorney General, New York Department of Law
Glenn T. Dentel, Chief Projects Branch 1 Division of Reactor Projects


Supervisor, Town of Scriba
Docket No.:
50-220, 50-410 License No.: DPR-63, NPF-69


T. Judson, Central NY Citizens Awareness Network
Enclosure:
Inspection Report 05000220/2008002 and 05000410/2008002 w/Attachment: Supplemental Information


D. Katz, Citizens Awareness Network
cc w/encl:
M. Wallace, President, Constellation Generation B. Barron, Senior Vice President and Chief Nuclear Officer C. Fleming, Esquire, Senior Counsel, Constellation Energy Group, LLC M. Wetterhahn, Esquire, Winston and Strawn T. Syrell, Director, Licensing, Nine Mile Point Nuclear Station P. Tonko, President and CEO, New York State Energy Research and Development Authority J. Spath, Program Director, New York State Energy Research and Development Authority P. D. Eddy, Electric Division, NYS Department of Public Service C. Donaldson, Esquire, Assistant Attorney General, New York Department of Law Supervisor, Town of Scriba T. Judson, Central NY Citizens Awareness Network D. Katz, Citizens Awareness Network


=SUMMARY OF FINDINGS=
=SUMMARY OF FINDINGS=
Line 100: Line 53:
Units 1 and 2; Surveillance Testing.
Units 1 and 2; Surveillance Testing.


The report covered a three-month period of inspection by resident inspectors and regional specialist inspectors. One Green non-cited violation (NCV) was identified. The significance of most findings is indicated by their color (Green, White, Yellow, Red) using Inspection Manual  
The report covered a three-month period of inspection by resident inspectors and regional specialist inspectors. One Green non-cited violation (NCV) was identified. The significance of most findings is indicated by their color (Green, White, Yellow, Red) using Inspection Manual Chapter (IMC) 0609, Significance Determination Process (SDP). Findings for which the SDP does not apply may be Green or be assigned a severity level after NRC management review.


Chapter (IMC) 0609, "Significance Determination Process (SDP). Findings for which the SDP does not apply may be Green or be assigned a severity level after NRC management review.
The NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process, Revision 4, dated December 2006.
 
The NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, "Reactor Oversight Process," Revision 4, dated December 2006.
 
===A. NRC-Identified and Self-Revealing Findings===


===NRC-Identified and Self-Revealing Findings===
===Cornerstone: Mitigating Systems===
===Cornerstone: Mitigating Systems===
*
: '''Green.'''
: '''Green.'''
A self-revealing, non-cited violation (NCV) of Technical Specification (TS) 5.4, "Procedures," was identified on January 14, 2008, when technicians improperly performed a surveillance procedure which resulted in isolation of the Unit 2 RCIC system. Specifically, while performing a test of the area temperature instruments that provide high temperature isolation signals for the main steam system, technicians erroneously disconnected an electrical lead associated with the RCIC leak detection system. This resulted in an automatic isol ation of the RCIC system steam supply and the unavailability of RCIC for approximately four hours. Operators immediately recognized the error and halted the surveillance procedure. Technicians reconnected the lead and operators restored RCIC to a normal standby lineup.
A self-revealing, non-cited violation (NCV) of Technical Specification (TS) 5.4,  
"Procedures," was identified on January 14, 2008, when technicians improperly performed a surveillance procedure which resulted in isolation of the Unit 2 RCIC system. Specifically, while performing a test of the area temperature instruments that provide high temperature isolation signals for the main steam system, technicians erroneously disconnected an electrical lead associated with the RCIC leak detection system. This resulted in an automatic isolation of the RCIC system steam supply and the unavailability of RCIC for approximately four hours. Operators immediately recognized the error and halted the surveillance procedure. Technicians reconnected the lead and operators restored RCIC to a normal standby lineup.


The finding was greater than minor because it was associated with the human performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding was determined to be of very low safety significance in accordance with IMC 0609,
The finding was greater than minor because it was associated with the human performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding was determined to be of very low safety significance in accordance with IMC 0609,
Appendix A, "Determining the Significance of Reactor Inspection Findings for At-Power  
Appendix A, "Determining the Significance of Reactor Inspection Findings for At-Power Situations," based on a Phase 3 analysis. The Region I senior reactor analyst (SRA)used the Nine Mile Point Unit 2 Standardized Plant Analysis Risk (SPAR) model and the actual out-of-service time to determine the risk significance. This finding had a cross-cutting aspect in the area of human performance because of the ineffective use of human error prevention techniques (H.4.a per IMC 0305). (Section 1R22)  
 
Situations," based on a Phase 3 analysis. The Region I senior reactor analyst (SRA)  
 
used the Nine Mile Point Unit 2 Standardized Plant Analysis Risk (SPAR) model and the actual out-of-service time to determine the risk significance. This finding had a cross-cutting aspect in the area of human performance because of the ineffective use of human error prevention techniques (H.4.a per IMC 0305). (Section 1R22)  
 
===B. Licensee-Identified Violations===


===Licensee-Identified Violations===
None.
None.
4


=REPORT DETAILS=
=REPORT DETAILS=


===Summary of Plant Status===
===Summary of Plant Status===
Nine Mile Point Unit 1 was operated at full rated thermal power (RTP) throughout the inspection period, with the exception of planned power reductions and recoveries for planned reactor recirculation pump maintenance, control rod testing, and main turbine valve testing.


Nine Mile Point Unit 1 was operated at full rated thermal power (RTP) throughout the inspection
Nine Mile Point Unit 2 began the inspection period at full RTP. Operators performed several planned power reductions and recoveries for control rod pattern adjustments, main turbine and main steam isolation valve testing, and control rod testing. On March 22, the reactor was shut down to commence refueling outage (RFO)
 
period, with the exception of planned power reductions and recoveries for planned reactor
 
recirculation pump maintenance, control rod testing, and main turbine valve testing.
 
Nine Mile Point Unit 2 began the inspection period at full RTP. Operators performed several  
 
planned power reductions and recoveries for control rod pattern adjustments, main turbine and  
 
main steam isolation valve testing, and control rod testing. On March 22, the reactor was shut  
 
down to commence refueling outage (RFO) 11.


==REACTOR SAFETY==
==REACTOR SAFETY==
Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity
Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity  
{{a|1R04}}


{{a|1R04}}
==1R04 Equipment Alignment
==1R04 Equipment Alignment==


==
===.1 Partial System Walkdown (71111.04 - Four samples)===
===.1 Partial System Walkdown (71111.04 - Four samples)===
====a. Inspection Scope====
====a. Inspection Scope====
The inspectors performed four partial system wa lkdowns to verify risk-significant systems were properly aligned for operation. The inspectors verified the operability and alignment  
The inspectors performed four partial system walkdowns to verify risk-significant systems were properly aligned for operation. The inspectors verified the operability and alignment of these risk-significant systems while their redundant trains or systems were inoperable or out of service for maintenance. The inspectors compared system lineups to system operating procedures, system drawings, and the applicable chapters in the updated final safety analysis report (UFSAR). The inspectors verified the operability of critical system components by observing component material condition during the system walkdown.


of these risk-significant systems while thei r redundant trains or systems were inoperable or out of service for maintenance. The ins pectors compared system lineups to system operating procedures, system drawings, and the applicable chapters in the updated final
Documents reviewed during this inspection are listed in the Attachment. The inspectors performed partial walkdowns of the following systems:
 
* Unit 2 'B' residual heat removal (RHR) system, while the Division 1 low pressure emergency core cooling systems ('A' RHR and low pressure core spray) were inoperable for planned maintenance (January 17, 2008);
safety analysis report (UFSAR). The inspectors verified the operability of critical system
 
components by observing component material condition during the system walkdown.
 
Documents reviewed during this inspection are listed in the Attachment. The inspectors  
 
performed partial walkdowns of the following systems:
* Unit 2 'B' residual heat removal (RHR) system, while the Division 1 low pressure emergency core cooling systems ('A' RHR and low pressure core spray) were  
 
inoperable for planned maintenance (January 17, 2008);
* Unit 1 control room air treatment system while 112 and 121 control room chillers were out of service for corrective maintenance (February 1, 2008);
* Unit 1 control room air treatment system while 112 and 121 control room chillers were out of service for corrective maintenance (February 1, 2008);
* Unit 1 core spray system 12 during inservice testing of core spray system 11 (February 26, 2008); and
* Unit 1 core spray system 12 during inservice testing of core spray system 11 (February 26, 2008); and
Line 174: Line 98:


===.2 Complete System Walkdown (71111.04S - One sample)===
===.2 Complete System Walkdown (71111.04S - One sample)===
====a. Inspection Scope====
====a. Inspection Scope====
The inspectors performed a complete walkdown of the Unit 1 emergency cooling system  
The inspectors performed a complete walkdown of the Unit 1 emergency cooling system to identify discrepancies between the existing equipment configuration and that specified in the design documents. During the walkdown, system drawings and operating procedures were used to determine the proper equipment alignment and operational status. The inspectors reviewed the open maintenance work orders (WO) that could affect the ability of the system to perform its functions. Documentation associated with temporary modifications, operator workarounds, and items tracked by plant engineering were also reviewed to assess their collective impact on system operation. In addition, the inspectors reviewed the condition report (CR) database to verify that equipment alignment problems were being identified and appropriately resolved. Documents reviewed during this inspection are listed in the Attachment.


to identify discrepancies between the existing equipment configuration and that specified
====b. Findings====
No findings of significance were identified. {{a|1R05}}


in the design documents. During the walkdown, system drawings and operating
==1R05 Fire Protection (71111.05Q - Six samples)
 
procedures were used to determine the proper equipment alignment and operational
 
status. The inspectors reviewed the open maintenance work orders (WO) that could affect
 
the ability of the system to perform its functions. Documentation associated with
 
temporary modifications, operator workarounds, and items tracked by plant engineering
 
were also reviewed to assess their collective impact on system operation. In addition, the
 
inspectors reviewed the condition report (CR) database to verify that equipment alignment
 
problems were being identified and appropriately resolved. Documents reviewed during this inspection are listed in the Attachment.
 
====b. Findings====
No findings of significance were identified.
{{a|1R05}}
==1R05 Fire Protection (71111.05Q - Six samples)==


====a. Inspection Scope====
====a. Inspection Scope====
The inspectors toured six areas important to reactor safety at NMPNS to evaluate the  
==
 
The inspectors toured six areas important to reactor safety at NMPNS to evaluate the stations control of transient combustibles and ignition sources, and to examine the material condition, operational status, and operational lineup of fire protection systems including detection, suppression, and fire barriers. Documents reviewed for this inspection are listed in the Attachment. The areas inspected included:
station's control of transient combustibles and ignition sources, and to examine the  
 
material condition, operational status, and operational lineup of fire protection systems  
 
including detection, suppression, and fire barriers. Documents reviewed for this inspection  
 
are listed in the Attachment. The areas inspected included:
* Unit 1 train 11 battery and battery board rooms;
* Unit 1 train 11 battery and battery board rooms;
* Unit 1 train 12 battery and battery board rooms;
* Unit 1 train 12 battery and battery board rooms;
Line 219: Line 117:


====b. Findings====
====b. Findings====
No findings of significance were identified.
No findings of significance were identified. {{a|1R08}}


6 1R08 Inservice Inspection Activities (71111.08 - One sample)
==1R08 Inservice Inspection Activities (71111.08 - One sample)


====a. Inspection Scope====
====a. Inspection Scope====
The purpose of this inspection was to assess the effectiveness of the inservice inspection (ISI) program for monitoring degradation of the reactor coolant system (RCS) boundary, risk significant piping system boundaries, and the containment boundary. The inspectors  
==
 
The purpose of this inspection was to assess the effectiveness of the inservice inspection (ISI) program for monitoring degradation of the reactor coolant system (RCS) boundary, risk significant piping system boundaries, and the containment boundary. The inspectors assessed the ISI activities using the criteria specified in the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section XI and applicable NRC regulatory requirements.
assessed the ISI activities using the criteria specified in the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section XI and  
 
applicable NRC regulatory requirements.
 
The inspectors selected a sample of nondestructive examination (NDE) activities for
 
observation or review, and evaluation for compliance with the requirements of ASME Section XI. Also, the inspectors selected samples of activities associated with the
 
repair/replacement of safety related pressure boundary components. The sample selection was based on the inspection procedure objectives, risk significance, and availability. Specifically, the inspec tors focused on components and systems where degradation would result in a significant challenge to the integrity of pressure boundary
 
components.
 
The inspectors performed an observation of one volumetric examination (ultrasonic) and


portions of a surface examination (liquid penetrant). In addition, the inspectors performed
The inspectors selected a sample of nondestructive examination (NDE) activities for observation or review, and evaluation for compliance with the requirements of ASME Section XI. Also, the inspectors selected samples of activities associated with the repair/replacement of safety related pressure boundary components. The sample selection was based on the inspection procedure objectives, risk significance, and availability. Specifically, the inspectors focused on components and systems where degradation would result in a significant challenge to the integrity of pressure boundary components.


a documentation review of a magnetic particle surface examination. The sample selection  
The inspectors performed an observation of one volumetric examination (ultrasonic) and portions of a surface examination (liquid penetrant). In addition, the inspectors performed a documentation review of a magnetic particle surface examination. The sample selection included the following:
 
included the following:
* Ultrasonic test (UT), volumetric examination, weld # 2CSL-26-05-FW005, butt weld, pipe to penetration, core spray system;
* Ultrasonic test (UT), volumetric examination, weld # 2CSL-26-05-FW005, butt weld, pipe to penetration, core spray system;
* Magnetic particle test, surface examination, welds #300 and 301, integral attachments, lugs to pipe, main steam system; and
* Magnetic particle test, surface examination, welds #300 and 301, integral attachments, lugs to pipe, main steam system; and
* Liquid penetrant test, surface examination, welds SW 95, 96, 97 and 98, integral attachments, lugs to pipe, RCS.
* Liquid penetrant test, surface examination, welds SW 95, 96, 97 and 98, integral attachments, lugs to pipe, RCS.


The inspectors performed an evaluation of work activities during a drywell entry and  
The inspectors performed an evaluation of work activities during a drywell entry and visually examined the condition of accessible portions of the containment liner and coatings for peeling, blistering, corrosion, mechanical damage, and other degradation mechanisms. The inspectors noted that two different coatings were apparent on various locations of the internal exposed metallic surfaces of the containment liner. The inspectors reviewed documentation which supported the coating qualification in accordance with ANSI N101.2 and that all coating had been applied in accordance with Regulatory Guide 1.54.


visually examined the condition of accessible portions of the containment liner and  
The inspectors reviewed portions of the in-process remote visual examination of the steam dryer. The inspectors reviewed three CRs initiated as a result of the dryer examination and noted the rejectable indications reported. The indications noted had not been identified during the previous examination (previous outage in 2006). These issues were placed in the corrective action program for engineering evaluation and disposition.


coatings for peeling, blistering, corrosion, mechanical damage, and other degradation
The inspectors selected for review a sample of repair/rework activities which required the development and implementation of an ASME Section XI repair plan. The inspectors reviewed documentation for the weld repair of two pressure boundary valves in ASME Class 3 systems. The inspectors reviewed the ASME Section XI plans, work scope, activity sequence, weld filler metal selection, weld procedure specifications and procedure qualification records, welder qualifications, specified non-destructive tests, acceptance criteria, and post work testing. The following samples were inspected:
* WO 04-08487-00 was initiated for the mechanical and weld repair of globe valve 2IAS-V181 in the instrument air system. The repair involved the disassembly and rebuilding of the valve. The disassembly of the valve required the removal of the body to bonnet weld to access the internals for mechanical rework of the valve seats.


mechanisms. The inspectors noted that two different coatings were apparent on various
Restoration of the body to bonnet weld was required following the completion of the repair and installation of the valve internals.
* WO 05-21585-00 was initiated to facilitate the removal, testing, rebuilding, inspection and re-installation by welding, into the piping system of relief valve 2WCS-RV21A in the reactor water cleanup system. It was necessary to eliminate existing installation welds in order to remove, rebuild, and test the valve. Acceptance testing of the completed valve repair and welding was specified in the repair plan. A visual examination was specified for the installation welds and a system pressure test specified to verify valve and system integrity.


locations of the internal exposed metallic surfaces of the containment liner. The
No sample of a previously identified recordable indication accepted as-is for continued service from the previous and the current outage was available for review during the inspection.


inspectors reviewed documentation which supported the coating qualification in accordance with ANSI N101.2 and that all coating had been applied in accordance with Regulatory Guide 1.54.
====b. Findings====
No findings of significance were identified. {{a|1R11}}


The inspectors reviewed portions of the in-process remote visual examination of the steam
==1R11 Licensed Operator Requalification Program (71111.11Q - Two samples)
 
dryer. The inspectors reviewed three CRs initiated as a result of the dryer examination
 
and noted the rejectable indications reported. The indications noted had not been
 
identified during the previous examination (previous outage in 2006). These issues were
 
placed in the corrective action program for engineering evaluation and disposition.
 
7 The inspectors selected for review a sample of repair/rework activities which required the development and implementation of an ASME Section XI repair plan. The inspectors
 
reviewed documentation for the weld repair of two pressure boundary valves in ASME Class 3 systems. The inspectors reviewed the ASME Section XI plans, work scope, activity sequence, weld filler metal selection, weld procedure specifications and procedure
 
qualification records, welder qualifications, specified non-destructive tests, acceptance
 
criteria, and post work testing. The following samples were inspected:
* WO 04-08487-00 was initiated for the mechanical and weld repair of globe valve 2IAS-V181 in the instrument air system. The repair involved the disassembly and
 
rebuilding of the valve. The disassembly of the valve required the removal of the body
 
to bonnet weld to access the internals for mechanical rework of the valve seats.
 
Restoration of the body to bonnet weld was required following the completion of the
 
repair and installation of the valve internals.
* WO 05-21585-00 was initiated to facilitate the removal, testing, rebuilding, inspection and re-installation by welding, into the piping system of relief valve 2WCS-RV21A in the reactor water cleanup system. It was necessary to eliminate existing installation
 
welds in order to remove, rebuild, and test the valve. Acceptance testing of the
 
completed valve repair and welding was specified in the repair plan. A visual
 
examination was specified for the installation welds and a system pressure test
 
specified to verify valve and system integrity.
 
No sample of a previously identified recordable indication accepted as-is for continued
 
service from the previous and the current outage was available for review during the
 
inspection.
 
====b. Findings====
No findings of significance were identified.
{{a|1R11}}
==1R11 Licensed Operator Requalification Program (71111.11Q - Two samples)==


====a. Inspection Scope====
====a. Inspection Scope====
The inspectors evaluated two simulator scenarios licensed operator requalification training  
==
 
The inspectors evaluated two simulator scenarios licensed operator requalification training program. The inspectors assessed the clarity and effectiveness of communications, the implementation of appropriate actions in response to alarms, the performance of timely control board operation, and the oversight and direction provided by the shift manager.
program. The inspectors assessed the clarity and effectiveness of communications, the  
 
implementation of appropriate actions in response to alarms, the performance of timely  


control board operation, and the oversight and direction provided by the shift manager.
During the scenario, the inspectors also compared simulator performance with actual plant performance in the control room. Documents reviewed for this inspection are listed in the  
 
During the scenario, the inspectors also compared simulator performance with actual plant  
 
performance in the control room. Documents reviewed for this inspection are listed in the  
. The following scenarios were observed:
. The following scenarios were observed:
* On March 17, 2008, the inspectors observed a Unit 2 operations crew during "Just In Time Training" (JITT) in preparation for RFO 11. The crew performed an approach to criticality, discussed the performance of surveillance procedure N2-OSP-EGS-R004, "Operating Cycle Diesel Generator Simulated Loss of Offsite Power with ECCS 8 Division I and II," and discussed plant modifications that would be performed during  
* On March 17, 2008, the inspectors observed a Unit 2 operations crew during Just In Time Training (JITT) in preparation for RFO 11. The crew performed an approach to criticality, discussed the performance of surveillance procedure N2-OSP-EGS-R004, Operating Cycle Diesel Generator Simulated Loss of Offsite Power with ECCS Division I and II, and discussed plant modifications that would be performed during the outage.
* On March 18, 2008, the inspectors observed a Unit 2 operations crew during JITT training in preparation for RFO 11. The crew performed a plant cooldown, including the transition to RHR shutdown cooling in service.


the outage.
====b. Findings====
* On March 18, 2008, the inspectors observed a Unit 2 operations crew during JITT training in preparation for RFO 11. The crew performed a plant cooldown, including
No findings of significance were identified. {{a|1R12}}


the transition to RHR shutdown cooling in service.
==1R12 Maintenance Effectiveness (71111.12Q - Two samples)
 
====b. Findings====
No findings of significance were identified.
{{a|1R12}}
==1R12 Maintenance Effectiveness (71111.12Q - Two samples)==


====a. Inspection Scope====
====a. Inspection Scope====
The inspectors reviewed performance-based problems and the performance and condition  
==
 
The inspectors reviewed performance-based problems and the performance and condition history of selected systems to assess the effectiveness of the maintenance program. The inspectors reviewed the systems to ensure that the stations review focused on proper maintenance rule scoping in accordance with 10 CFR Part 50.65, characterization of reliability issues, tracking system and component unavailability, and 10 CFR Part 50.65 (a)(1) and (a)(2) classification. In addition, the inspectors reviewed the sites ability to identify and address common cause failures and to trend key parameters. Documents reviewed for the inspection are listed in the Attachment. The following two maintenance rule inspection samples were reviewed:
history of selected systems to assess the effectiveness of the maintenance program. The  
 
inspectors reviewed the systems to ensure that the station's review focused on proper  
 
maintenance rule scoping in accordance with 10 CFR Part 50.65, characterization of  
 
reliability issues, tracking system and component unavailability, and 10 CFR Part 50.65 (a)(1) and (a)(2) classification. In addition, the inspectors reviewed the site's ability to  
 
identify and address common cause failures and to trend key parameters. Documents  
 
reviewed for the inspection are listed in the Attachment. The following two maintenance  
 
rule inspection samples were reviewed:
* Unit 1 fire protection systems due to long-standing equipment problems; and
* Unit 1 fire protection systems due to long-standing equipment problems; and
* Unit 2 service water (SW) system due to extended unavailability of the 'E' SW pump.
* Unit 2 service water (SW) system due to extended unavailability of the E SW pump.


====b. Findings====
====b. Findings====
No findings of significance were identified.
No findings of significance were identified. {{a|1R13}}
{{a|1R13}}
==1R13 Maintenance Risk Assessments and Emergent Work Control (71111.13 - Six samples)==


====a. Inspection Scope====
==1R13 Maintenance Risk Assessments and Emergent Work Control (71111.13 - Six samples)
The inspectors evaluated the effectiveness of the maintenance risk assessments required


by paragraph (a)(4) of 10 CFR Part 50.65. The inspectors reviewed equipment logs, work
a.


schedules, and performed plant tours to gain assurance that actual plant configuration
==
Inspection Scope


matched the assessed configuration. Additionally, the inspectors verified that risk  
The inspectors evaluated the effectiveness of the maintenance risk assessments required by paragraph (a)(4) of 10 CFR Part 50.65. The inspectors reviewed equipment logs, work schedules, and performed plant tours to gain assurance that actual plant configuration matched the assessed configuration. Additionally, the inspectors verified that risk management actions for both planned and emergent work were consistent with those described in station procedures. Documents reviewed for the inspection are listed in the  
 
management actions for both planned and emergent work were consistent with those  
 
described in station procedures. Documents reviewed for the inspection are listed in the  
.
.
The inspectors reviewed risk assessments for the activities listed below.
The inspectors reviewed risk assessments for the activities listed below.


Unit 1
Unit 1
* Week of January 21, 2008, that included 112 containment spray quarterly surveillance, an emergent issue with the 112 containment spray raw water pump packing  
* Week of January 21, 2008, that included 112 containment spray quarterly surveillance, an emergent issue with the 112 containment spray raw water pump packing overheating, emergency diesel generator (EDG) 102 monthly surveillance, high drywell pressure instrument trip channel test, and a power reduction to 88 percent to return 11 reactor recirculation pump to service.
 
* Week of January 28, 2008, that included control rod drive (CRD) pump quarterly surveillance, liquid poison system quarterly surveillance, emergency service water pump quarterly surveillance, main steam isolation valve (MSIV) partial stroke testing, and emergent activities to troubleshoot spiking on average power range monitors (APRMs) 12 and 15, and flow oscillations on 11 reactor recirculation pump.
overheating, emergency diesel generator (EDG) 102 monthly surveillance, high drywell pressure instrument trip channel test, and a power reduction to 88 percent to return 11  
* Week of February 12, 2008, that included a two day maintenance period on 11 high pressure coolant injection (HPCI) system, cleaning of 11 turbine building closed loop cooling (TBCLC) heat exchanger, repair of a packing leak on emergency cooling (EC)valve IV-39-11R which rendered 11 EC inoperable for two days, maintenance on vital uninterruptable power supply (UPS) 162A, EDG raw water system quarterly surveillance, securing 11 reactor recirculation pump for maintenance on its associated motor generator, and 11 reactor recirculation flow loop calibration and flow converter calibrations.
 
reactor recirculation pump to service.
* Week of January 28, 2008, that included control rod drive (CRD) pump quarterly surveillance, liquid poison system quarterl y surveillance, emergency service water pump quarterly surveillance, main steam isolation valve (MSIV) partial stroke testing, and emergent activities to troubleshoot spiking on average power range monitors (APRMs) 12 and 15, and flow oscillations on 11 reactor recirculation pump.
* Week of February 12, 2008, that included a two day maintenance period on 11 high pressure coolant injection (HPCI) system, cleaning of 11 turbine building closed loop  
 
cooling (TBCLC) heat exchanger, repair of a packing leak on emergency cooling (EC)  
 
valve IV-39-11R which rendered 11 EC inoperable for two days, maintenance on vital  
 
uninterruptable power supply (UPS) 162A, EDG raw water system quarterly  
 
surveillance, securing 11 reactor recirculation pump for maintenance on its associated  
 
motor generator, and 11 reactor recirculation flow loop calibration and flow converter  
 
calibrations.


Unit 2
Unit 2
* Week of January 7, 2008, that included a power reduction to 80 percent for control rod pattern adjustment, reactor vessel water level low surveillance, main steam line high flow surveillance, and investigation of increased drywell unidentified leakage  
* Week of January 7, 2008, that included a power reduction to 80 percent for control rod pattern adjustment, reactor vessel water level low surveillance, main steam line high flow surveillance, and investigation of increased drywell unidentified leakage concurrent with a reactor recirculation pump motor winding cooler leakage alarm.
* Week of January 21, 2008, that included a two day maintenance period for the HPCS system, HPCS system quarterly surveillance, on-line motor testing and lubrication of the A control rod drive pump, Division 3 EDG monthly surveillance, and emergent maintenance to stop makeup water leakage into the standby liquid control tank.
* Week of January 28, 2008, that included Division 2 EDG monthly surveillance, a power reduction to 70 percent for control rod sequence exchange, MSIV testing, and turbine valve testing, Division 2 standby gas treatment system inoperable for one day for filter medium sampling, C RHR system inoperable for one day for planned maintenance, C RHR system quarterly surveillance, and quarterly test of emergency core cooling systems (ECCS) actuation on high drywell pressure.


concurrent with a reactor recirculation pump motor winding cooler leakage alarm.
====b. Findings====
* Week of January 21, 2008, that included a two day maintenance period for the HPCS system, HPCS system quarterly surveillance, on-line motor testing and lubrication of the 'A' control rod drive pump, Division 3 EDG monthly surveillance, and emergent maintenance to stop makeup water leakage into the standby liquid control tank.
No findings of significance were identified. {{a|1R15}}
* Week of January 28, 2008, that included Division 2 EDG monthly surveillance, a power reduction to 70 percent for control rod sequence exchange, MSIV testing, and turbine valve testing, Division 2 standby gas treatment system inoperable for one day for filter


medium sampling, 'C' RHR system inoperable for one day for planned maintenance,
==1R15 Operability Evaluations (71111.15 - Seven samples)
'C' RHR system quarterly surveillance, and quarterly test of emergency core cooling
 
systems (ECCS) actuation on high drywell pressure.
 
====b. Findings====
No findings of significance were identified.
{{a|1R15}}
==1R15 Operability Evaluations (71111.15 - Seven samples)==


====a. Inspection Scope====
====a. Inspection Scope====
The inspectors evaluated the acceptability of the operability evaluations, the use and  
==
The inspectors evaluated the acceptability of the operability evaluations, the use and control of compensatory measures, and the compliance with TSs. The evaluations were reviewed using criteria specified in NRC Regulatory Issue Summary 2005-20, Revision to Guidance Formerly Contained in NRC Generic Letter 91-18, Information to Licensees Regarding Two NRC Inspection Manual Sections on Resolution of Degraded and Nonconforming Conditions and on Operability, and Inspection Manual Part 9900, Operability Determinations and Functionality Assessments for Resolution of Degraded or Nonconforming Conditions Adverse to Quality or Safety. The inspectors review included verification that the operability determinations were made as specified by Procedure CNG-OP-1.01-1002, Conduct of Operability Determinations / Functionality Assessments.


control of compensatory measures, and the compliance with TSs. The evaluations were
The technical adequacy of the determinations was reviewed and compared to the TSs, UFSAR, and associated design basis documents (DBDs). Documents reviewed for the inspection are listed in the Attachment. The following evaluations were reviewed:
 
reviewed using criteria specified in NRC Regulatory Issue Summary 2005-20, "Revision to
 
Guidance Formerly Contained in NRC Generic Letter 91-18, 'Information to Licensees
 
Regarding Two NRC Inspection Manual Sections on Resolution of Degraded and
 
Nonconforming Conditions and on Operability'," and Inspection Manual Part 9900, "Operability Determinations and Functionality Assessments for Resolution of Degraded or
 
Nonconforming Conditions Adverse to Quality or Safety."  The inspectors' review included
 
verification that the operability determinations were made as specified by Procedure CNG-OP-1.01-1002, "Conduct of Operability Determinations / Functionality Assessments."
 
The technical adequacy of the determinations was reviewed and compared to the TSs, UFSAR, and associated design basis documents (DBDs). Documents reviewed for the  
 
inspection are listed in the Attachment. The following evaluations were reviewed:
* CR 2008-531 concerning turbine first stage bowl pressure switch calibrations at Unit 1;
* CR 2008-531 concerning turbine first stage bowl pressure switch calibrations at Unit 1;
* CR 2006-3751 concerning environmental qualification of Unit 1 RB emergency ventilation damper position indicating switches;
* CR 2006-3751 concerning environmental qualification of Unit 1 RB emergency ventilation damper position indicating switches;
Line 439: Line 212:


====b. Findings====
====b. Findings====
No findings of significance were identified.
No findings of significance were identified. {{a|1R18}}
{{a|1R18}}
 
==1R18 Plant Modifications (71111.18 - One sample)==
==1R18 Plant Modifications (71111.18 - One sample)


====a. Inspection Scope====
====a. Inspection Scope====
The inspectors reviewed Unit 2 permanent modification N2-05-010, "Eliminate Single Point Vulnerability for Main Steam Tunnel Cooling."  The purpose was to reduce the likelihood of  
==
The inspectors reviewed Unit 2 permanent modification N2-05-010, Eliminate Single Point Vulnerability for Main Steam Tunnel Cooling. The purpose was to reduce the likelihood of a high temperature main steam line isolation due to loss of ventilation in the main steam lead enclosure. The inspectors assessed the adequacy of the modification package, including post-modification testing, and verified that applicable design and licensing basis requirements were met and that design margins were not degraded by the change.


a high temperature main steam line isolation due to loss of ventilation in the main steam
====b. Findings====
No findings of significance were identified. {{a|1R19}}


lead enclosure. The inspectors assessed the adequacy of the modification package, 11 including post-modification testing, and verified that applicable design and licensing basis
==1R19 Post Maintenance Testing (71111.19 - Five samples)
 
requirements were met and that design margins were not degraded by the change.
 
====b. Findings====
No findings of significance were identified.
{{a|1R19}}
==1R19 Post Maintenance Testing (71111.19 - Five samples)==


====a. Inspection Scope====
====a. Inspection Scope====
The inspectors reviewed the post maintenance tests listed below to verify that procedures  
==
 
The inspectors reviewed the post maintenance tests listed below to verify that procedures and test activities ensured system operability and functional capability. The inspectors reviewed the test procedure to verify that the procedure adequately tested the safety functions that may have been affected by the maintenance activity, that the acceptance criteria in the procedure were consistent with information in the applicable licensing basis and/or DBDs, and that the procedure had been properly reviewed and approved. The inspectors also witnessed the test or reviewed test data, to verify that the test results adequately demonstrated restoration of the affected safety functions. Documents reviewed for this inspection are listed in the Attachment.
and test activities ensured system operability and functional capability. The inspectors  
 
reviewed the test procedure to verify that the procedure adequately tested the safety  
 
functions that may have been affected by the maintenance activity, that the acceptance  
 
criteria in the procedure were consistent with information in the applicable licensing basis  
 
and/or DBDs, and that the procedure had been properly reviewed and approved. The  
 
inspectors also witnessed the test or reviewed test data, to verify that the test results  
 
adequately demonstrated restoration of the affected safety functions. Documents  
 
reviewed for this inspection are listed in the Attachment.
* Unit 1, WO 07-10842-00 that performed maintenance on the tie breaker between non-vital 600V power board 17A and vital 600V power board 17B. The retest was performed in accordance with N1-OP-30, "4.16KV, 600V, and 480V House Service."
* Unit 1, WO 07-10842-00 that performed maintenance on the tie breaker between non-vital 600V power board 17A and vital 600V power board 17B. The retest was performed in accordance with N1-OP-30, "4.16KV, 600V, and 480V House Service."
* Unit 1, WO 07-08535-00 that repacked emergency condenser steam line drain valve IV-39-11. The retest was performed in accordance with N1-ST-Q4, "Reactor Coolant  
* Unit 1, WO 07-08535-00 that repacked emergency condenser steam line drain valve IV-39-11. The retest was performed in accordance with N1-ST-Q4, Reactor Coolant System Isolation Valves Operability Test.
 
* Unit 1, WO 08-02028-00 that performed maintenance on the reactor protection system motor generator MG-141 voltage regulator. The retest was performed in accordance with N1-OP-48, Motor Generator Sets.
System Isolation Valves Operability Test."
* Unit 1, WO 07-06872-00 that repaired leaking emergency condenser vacuum breaker valve 60.1-28. The retest was performed by overflowing the emergency condenser shell using N1-ST-M2, Emergency Cooling System Makeup Tank Level Control Valves Exercising Test.
* Unit 1, WO 08-02028-00 that performed maintenance on the reactor protection system motor generator MG-141 voltage regulator. The retest was performed in accordance  
* Unit 2, WO 07-01190-00 that performed inspection of the A CRD pump motor. The retest for the circuit breaker rack-out was performed in accordance with N2-OP-30, Control Rod Drive.
 
with N1-OP-48, "Motor Generator Sets."
* Unit 1, WO 07-06872-00 that repaired leaking emergency condenser vacuum breaker valve 60.1-28. The retest was performed by overflowing the emergency condenser shell using N1-ST-M2, "Emergency Cooling System Makeup Tank Level Control  
 
Valves Exercising Test."
* Unit 2, WO 07-01190-00 that performed inspection of the 'A' CRD pump motor. The retest for the circuit breaker rack-out was performed in accordance with N2-OP-30, "Control Rod Drive."


====b. Findings====
====b. Findings====
No findings of significance were identified.
No findings of significance were identified. {{a|1R20}}


12 1R20 Refueling and Other Outage Activities (71111.20 - In Progress)
==1R20 Refueling and Other Outage Activities (71111.20 - In Progress)


====a. Inspection Scope====
====a. Inspection Scope====
The inspectors observed and/or reviewed the following Unit 2 refueling outage activities to  
==
 
The inspectors observed and/or reviewed the following Unit 2 refueling outage activities to verify that operability requirements were met and that risk, industry experience, and previous site-specific problems were considered. The refueling outage and inspection sample were in progress at the end of the inspection period. Documents reviewed for this inspection are listed in the Attachment.
verify that operability requirements were met and that risk, industry experience, and  
* The inspectors reviewed the outage schedule and procedures, and verified that TS-required safety system availability was maintained and shutdown risk was minimized. The inspectors verified that, when specified by NMPNS procedure NIP-OUT-01, Shutdown Safety, contingency plans existed for restoring key safety functions.
 
previous site-specific problems were considered. The refueling outage and inspection  
 
sample were in progress at the end of the inspection period. Documents reviewed for this  
 
inspection are listed in the Attachment.
* The inspectors reviewed the outage schedule and procedures, and verified that TS-required safety system availability was maintained and shutdown risk was  
 
minimized. The inspectors verified that, when specified by NMPNS procedure  
 
NIP-OUT-01, "Shutdown Safety," contingency plans existed for restoring key safety  
 
functions.
* The inspectors observed portions of the plant shutdown and cooldown on March 22, and verified that the TS cooldown rate limits were satisfied.
* The inspectors observed portions of the plant shutdown and cooldown on March 22, and verified that the TS cooldown rate limits were satisfied.
* Through plant tours, the inspectors verified that NMPNS maintained and adequately protected electrical power supplies to safety-related equipment and that TS requirements were met.
* Through plant tours, the inspectors verified that NMPNS maintained and adequately protected electrical power supplies to safety-related equipment and that TS requirements were met.
* The inspectors verified proper alignment and operation of shutdown cooling and other decay heat removal systems. The verification also included reactor cavity and fuel pool makeup paths and water sources, and administrative control of drain down paths.
* The inspectors verified proper alignment and operation of shutdown cooling and other decay heat removal systems. The verification also included reactor cavity and fuel pool makeup paths and water sources, and administrative control of drain down paths.
* The inspectors verified that requirements for refueling operations were met through refuel bridge observations, control room panel walkdowns, and discussions with  
* The inspectors verified that requirements for refueling operations were met through refuel bridge observations, control room panel walkdowns, and discussions with Operations Department personnel.
* After the drywell was open for general access, the inspectors performed an as-found walkdown to identify evidence of RCS leakage and assess the condition of drywell structures, piping, and supports.


Operations Department personnel.
====b. Findings====
* After the drywell was open for general access, the inspectors performed an "as-found" walkdown to identify evidence of RCS leakage and assess the condition of drywell
No findings of significance were identified. {{a|1R22}}


structures, piping, and supports.
==1R22 Surveillance Testing (71111.22 - Eight samples)
 
====b. Findings====
No findings of significance were identified.
{{a|1R22}}
==1R22 Surveillance Testing (71111.22 - Eight samples)==


====a. Inspection Scope====
====a. Inspection Scope====
The inspectors witnessed performance of and/or reviewed test data for risk-significant surveillance tests (STs) to assess whether the components and systems tested satisfied  
==
 
The inspectors witnessed performance of and/or reviewed test data for risk-significant surveillance tests (STs) to assess whether the components and systems tested satisfied design and licensing basis requirements. The inspectors verified that test acceptance criteria were clear, demonstrated operational readiness and were consistent with the DBDs; that test instrumentation had current calibrations and the range and accuracy for the application; and that tests were performed, as written, with applicable prerequisites satisfied. Upon test completion, the inspectors verified that equipment was returned to the status specified to perform its safety function. Documents reviewed for this inspection are listed in the Attachment.
design and licensing basis requirements. The inspectors verified that test acceptance  
 
criteria were clear, demonstrated operational readiness and were consistent with the  
 
DBDs; that test instrumentation had current calibrations and the range and accuracy for 13 the application; and that tests were performed, as written, with applicable prerequisites  
 
satisfied. Upon test completion, the inspectors verified that equipment was returned to the  
 
status specified to perform its safety function. Documents reviewed for this inspection are  
 
listed in the Attachment.


The following STs were reviewed:
The following STs were reviewed:
* N1-ST-M8, "RB Emergency Ventilation System Operability Test;"
* N1-ST-M8, RB Emergency Ventilation System Operability Test;
* N1-ST-Q6C, "Containment Spray System Loop 112 Quarterly Operability Test;"
* N1-ST-Q6C, Containment Spray System Loop 112 Quarterly Operability Test;
* N1-ST-Q21, "Instrument Air Valves Quarterly Test;"
* N1-ST-Q21, Instrument Air Valves Quarterly Test;
* N1-ISP-201-022, "Drywell Water Leak Detection Instrument Channel Test;"
* N1-ISP-201-022, Drywell Water Leak Detection Instrument Channel Test;
* N2-ISP-LDS-R106, "Main Steam Line Tunnel and MSL Lead Enclosure Temperature Instrument Channel Calibration;"
* N2-ISP-LDS-R106, Main Steam Line Tunnel and MSL Lead Enclosure Temperature Instrument Channel Calibration;
* N2-OSP-RHS-Q@006, "RHR System Loop C Pump and Valve Operability Test and System Integrity Test;"
* N2-OSP-RHS-Q@006, "RHR System Loop C Pump and Valve Operability Test and System Integrity Test;"
* N2-OSP-ICS-Q@002, "RCIC Pump and Valve Operability Test and System Integrity Test and ASME XI Functional Test;" and
* N2-OSP-ICS-Q@002, "RCIC Pump and Valve Operability Test and System Integrity Test and ASME XI Functional Test;" and
* N2-OSP-GTS-R001, "Secondary Containment Integrity Test."
* N2-OSP-GTS-R001, Secondary Containment Integrity Test.


====b. Findings====
====b. Findings====
=====Introduction.=====
=====Introduction.=====
A self-revealing Green NCV of TS 5.4, "Procedures," was identified on January 14, 2008, when technicians improperly performed a surveillance procedure which resulted in isolation of the Unit 2 RCIC system. Specifically, technicians erroneously  
A self-revealing Green NCV of TS 5.4, "Procedures," was identified on January 14, 2008, when technicians improperly performed a surveillance procedure which resulted in isolation of the Unit 2 RCIC system. Specifically, technicians erroneously disconnected an electrical lead associated with the RCIC leak detection system, which resulted in an automatic isolation of the RCIC system steam supply.
 
disconnected an electrical lead associated with the RCIC leak detection system, which  
 
resulted in an automatic isolation of the RCIC system steam supply.


=====Description.=====
=====Description.=====
On January 14, 2008, instrument and controls technicians were performing an ST on the area temperature instruments that provide high temperature isolation signals for the main steam system. The surveill ance, N2-ISP-LDS-R106, "Main Steam Line  
On January 14, 2008, instrument and controls technicians were performing an ST on the area temperature instruments that provide high temperature isolation signals for the main steam system. The surveillance, N2-ISP-LDS-R106, "Main Steam Line Tunnel and MSL Lead Enclosure Temperature Instrument Channel Calibration," requires that the associated thermocouple leads be disconnected prior to performing the channel calibration. When the technicians attempted to perform this action (step 7.2.1 of 1), they incorrectly identified the specified terminals. The procedure directed the technicians to disconnect thermocouple leads from 2CEC*PNL707 Bay F, TH405 TB-1, terminals 10 and 11. The lead that was actually disconnected was from terminal 14. One of the technicians had initially questioned the adequacy of their terminal identification since the terminals were not individually labeled. However, they concluded that they had identified the correct terminal and proceeded. The wires that they proceeded to disconnect were thermocouple leads for a temperature instrument that provides area high temperature isolation for the RCIC system. The open circuit created by lifting the first lead resulted in an automatic isolation of the RCIC system steam supply.
 
Tunnel and MSL Lead Enclosure Temperature Instrument Channel Calibration," requires  
 
that the associated thermocouple leads be disconnected prior to performing the channel  
 
calibration. When the technicians attempted to perform this action (step 7.2.1 of 1), they incorrectly identified t he specified terminals. The procedure directed the technicians to disconnect thermocouple leads from 2CEC*PNL707 Bay F, TH405 TB-1, terminals 10 and 11. The lead that was actually disconnected was from terminal 14. One  
 
of the technicians had initially questioned the adequacy of their terminal identification since  
 
the terminals were not individually labeled. However, they concluded that they had  
 
identified the correct terminal and proceeded. The wires that they proceeded to disconnect  
 
were thermocouple leads for a temperature instrument that provides area high temperature isolation for the RCIC system. The open circuit created by lifting the first lead resulted in an automatic isolation of the RCIC system steam supply.


Operators immediately recognized the error and halted the surveillance procedure.
Operators immediately recognized the error and halted the surveillance procedure.


Technicians reconnected the thermocouple, and operators restored RCIC to a normal  
Technicians reconnected the thermocouple, and operators restored RCIC to a normal standby lineup. During the four hours that the RCIC steam supply was isolated, the RCIC system was inoperable and unavailable. The TS allowed outage time for the RCIC system is 14 days.
 
standby lineup. During the four hours that the RCIC steam supply was isolated, the RCIC  


system was inoperable and unavailable. The TS allowed outage time for the RCIC system is 14 days.
The performance deficiency associated with this event was that technicians did not correctly perform a ST procedure, which caused the Unit 2 RCIC system to automatically isolate, rendering the system unavailable to perform its safety function.
 
The performance deficiency associated with this event was that technicians did not  
 
correctly perform a ST procedure, which caus ed the Unit 2 RCIC system to automatically isolate, rendering the system unavailable to perform its safety function.


=====Analysis.=====
=====Analysis.=====
The finding was greater than minor because it was associated with the human performance attribute of the Mitigating Sy stems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that  
The finding was greater than minor because it was associated with the human performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to Initiating Events to prevent undesirable consequences.
 
respond to Initiating Events to prevent undesirable consequences.
 
The inspectors evaluated the significance of this finding using IMC 0609, Appendix A, Phase 1, and determined that a Phase 2 analysis was required because the finding
 
represented an actual loss of the RCIC system safety function for four hours. The Region I
 
SRA determined that a Phase 3 analysis was necessary because the site-specific Phase 2
 
notebook indicated that the finding could be more than of very low safety significance
 
assuming an exposure time of three days. The SRA used the Nine Mile Point Unit 2 SPAR model and the actual four-hour exposure time to determine that the increase in core damage frequency was in the range of 1 core damage accident in 1.25E8 years of reactor
 
operation, or high E-9 per year. The SPAR model dominant cutsets were a station
 
blackout with failure of high pressure injection sources and the inability to restore AC
 
power within 30 minutes. Based on this review, the SRA concluded that the finding was of
 
very low safety significance (Green).
 
The finding had a cross-cutting aspect in the area of human performance because of the
 
ineffective use of human error prevention techniques, in that, although peer checking had


identified a question, that question was not adequately resolved prior to proceeding (H.4.a  
The inspectors evaluated the significance of this finding using IMC 0609, Appendix A, Phase 1, and determined that a Phase 2 analysis was required because the finding represented an actual loss of the RCIC system safety function for four hours. The Region I SRA determined that a Phase 3 analysis was necessary because the site-specific Phase 2 notebook indicated that the finding could be more than of very low safety significance assuming an exposure time of three days. The SRA used the Nine Mile Point Unit 2 SPAR model and the actual four-hour exposure time to determine that the increase in core damage frequency was in the range of 1 core damage accident in 1.25E8 years of reactor operation, or high E-9 per year. The SPAR model dominant cutsets were a station blackout with failure of high pressure injection sources and the inability to restore AC power within 30 minutes. Based on this review, the SRA concluded that the finding was of very low safety significance (Green).


per IMC 0305)  
The finding had a cross-cutting aspect in the area of human performance because of the ineffective use of human error prevention techniques, in that, although peer checking had identified a question, that question was not adequately resolved prior to proceeding (H.4.a per IMC 0305)  


=====Enforcement.=====
=====Enforcement.=====
TS 5.4, "Procedures," states that written procedures shall be established, implemented, and maintained covering the applicable procedures recommended in  
TS 5.4, Procedures, states that written procedures shall be established, implemented, and maintained covering the applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978. Regulatory Guide 1.33, Revision 2, Appendix A, February 1978, Item 8, Procedures for Control of Measuring and Test Equipment and for STs, Procedures, and Calibrations, lists containment isolation tests as an applicable group of tests. Contrary to the above, Unit 2 Instrument Surveillance Procedure N2-ISP-LDS-R106, "Main Steam Line Tunnel and MSL Lead Temperature Instrument Channel Calibration," was not correctly implemented.


Regulatory Guide 1.33, Revision 2, Appendix A, February 1978. Regulatory Guide 1.33, Revision 2, Appendix A, February 1978, Item 8, "Procedures for Control of Measuring and
On January 14, 2008, while attempting to perform Procedure Attachment 1, step 7.2.1, to disconnect field thermocouple leads from 2CEC*PNL707 Bay F, TH405 TB-1, terminals 10 and 11, technicians incorrectly disconnected the lead from terminal 14. This action resulted in an automatic isolation of the RCIC system steam supply. Because this procedural noncompliance is of very low safety significance and was entered into the CAP as CR 2008-332, this violation is being treated as an NCV, consistent with Section VI.A of the NRC Enforcement Policy: NCV 05000410/2008002-01, Failure to Correctly Perform Procedure Caused Inadvertent Isolation of the RCIC Steam Supply.


Test Equipment and for STs, Procedures, and Calibrations," lists containment isolation
===Cornerstone: Emergency Preparedness===
 
1EP6 Drill Evaluation (71114.06 - One sample)
tests as an applicable group of tests. Contrary to the above, Unit 2 Instrument
 
Surveillance Procedure N2-ISP-LDS-R106, "Main Steam Line Tunnel and MSL Lead Temperature Instrument Channel Calib ration," was not correctly implemented.
 
On January 14, 2008, while attempting to perform Procedure Attachment 1, step 7.2.1, to
 
disconnect field thermocouple leads from 2CEC*PNL707 Bay F, TH405 TB-1, terminals 10
 
and 11, technicians incorrectly disconnected the lead from terminal 14. This action
 
resulted in an automatic isolation of the RCIC system steam supply. Because this procedural noncompliance is of very low safety significance and was entered into the CAP as CR 2008-332, this violation is being treated as an NCV, consistent with Section VI.A of the NRC Enforcement Policy: NCV 05000410/2008002-01, Failure to Correctly Perform
 
Procedure Caused Inadvertent Isolation of the RCIC Steam Supply.
 
15Cornerstone:  Emergency Preparedness 1EP6 Drill Evaluation (71114.06 - One sample)


====a. Inspection Scope====
====a. Inspection Scope====
The inspectors completed one emergency drill evaluation inspection sample. The  
The inspectors completed one emergency drill evaluation inspection sample. The inspectors observed simulator, technical support center (TSC), and operations support center activities associated with the Unit 1 emergency planning drill on March 4, 2008. The scenario consisted of a leak from the spent fuel pool (SFP) due to an earthquake during the previous shift, a loss of off-site power (including power to the TSC) with failure of one main steam line to isolate, and a main steam line break outside secondary containment. The inspectors verified that emergency classification declarations and notifications were completed in accordance with 10 CFR 50.72, 10 CFR 50, Appendix E, and the Nine Mile Point emergency plan implementing procedures. Documents reviewed for this inspection are listed in the Attachment.
 
inspectors observed simulator, technical support center (TSC), and operations support  
 
center activities associated with the Unit 1 emergency planning drill on March 4, 2008. The  
 
scenario consisted of a leak from the spent fuel pool (SFP) due to an earthquake during the  
 
previous shift, a loss of off-site power (including power to the TSC) with failure of one main  
 
steam line to isolate, and a main steam line break outside secondary containment. The  
 
inspectors verified that emergency classification declarations and notifications were  
 
completed in accordance with 10 CFR 50.72, 10 CFR 50, Appendix E, and the Nine Mile  
 
Point emergency plan implementing procedures. Documents reviewed for this inspection  
 
are listed in the Attachment.


====b. Findings====
====b. Findings====
Line 666: Line 303:


==RADIATION SAFETY==
==RADIATION SAFETY==
 
===Cornerstone: Occupational Radiation Safety (OS)===
===Cornerstone:===
2OS1 Access Control to Radiologically Significant Areas (71121.01 - Eight samples)
Occupational Radiation Safety (OS)2OS1 Access Control to Radiologically Significant Areas (71121.01 - Eight samples)


====a. Inspection Scope====
====a. Inspection Scope====
Based on the work activities during the Unit 2 refueling outage, the inspectors selected  
Based on the work activities during the Unit 2 refueling outage, the inspectors selected three jobs (drywell scaffold, drywell inservice inspection, and in-vessel visual inspection)being performed in radiation areas, airborne radioactivity areas, or high radiation areas  
 
(<1 R/hr) for observation. The inspectors observed work that was estimated to result in the highest collective doses, involved diving activities in or around spent fuel or highly activated material, or that involved potentially changing (deteriorating) radiological conditions. The inspectors reviewed all radiological job requirements (radiation work permit requirements and work procedure requirements). The inspectors observed job performance with respect to these requirements. The inspectors determined if radiological conditions in the work area were adequately communicated to workers through briefings and postings.
three jobs (drywell scaffold, drywell inservice inspection, and in-vessel visual inspection)  
 
being performed in radiation areas, airborne radioactivity areas, or high radiation areas  
 
(<1 R/hr) for observation. The inspectors observed work that was estimated to result in the  
 
highest collective doses, involved diving activities in or around spent fuel or highly  
 
activated material, or that involved potentially changing (deteriorating) radiological  
 
conditions. The inspectors reviewed all radiological job requirements (radiation work  
 
permit requirements and work procedure requirements). The inspectors observed job  
 
performance with respect to these requirements. The inspectors determined if radiological conditions in the work area were adequately communicated to workers through briefings and postings.
 
During job performance observations, the inspectors verified the adequacy of radiological
 
controls, such as required surveys (including system breach radiation, contamination, and airborne surveys), radiation protection job coverage (including audio and visual
 
surveillance for remote job coverage), and contamination controls. For high radiation work
 
areas with significant dose rate gradients (factor of 5 or more), the inspectors reviewed the
 
application of dosimetry to effectively monitor exposure to personnel.
 
16 During job performance observations, the inspectors observed radiation worker
 
performance with respect to stated radiation protection work requirements. The inspectors
 
determined if workers were aware of the significant radiological conditions in their
 
workplace and the radiation work permit controls/limits in place, and that their performance
 
took into consideration the level of radiological hazards present.
 
During job performance observations, the inspectors observed radiation protection


technician performance with respect to all radiation protection work requirements. The
During job performance observations, the inspectors verified the adequacy of radiological controls, such as required surveys (including system breach radiation, contamination, and airborne surveys), radiation protection job coverage (including audio and visual surveillance for remote job coverage), and contamination controls. For high radiation work areas with significant dose rate gradients (factor of 5 or more), the inspectors reviewed the application of dosimetry to effectively monitor exposure to personnel.


inspectors determined if they were aware of the radiological conditions in their work area
During job performance observations, the inspectors observed radiation worker performance with respect to stated radiation protection work requirements. The inspectors determined if workers were aware of the significant radiological conditions in their workplace and the radiation work permit controls/limits in place, and that their performance took into consideration the level of radiological hazards present.


and the radiation work permit controls/limits, and if their performance was consistent with their training and qualifications with respect to the radiological hazards and work activities.
During job performance observations, the inspectors observed radiation protection technician performance with respect to all radiation protection work requirements. The inspectors determined if they were aware of the radiological conditions in their work area and the radiation work permit controls/limits, and if their performance was consistent with their training and qualifications with respect to the radiological hazards and work activities.


The inspectors identified exposure significant work areas within radiation areas, high  
The inspectors identified exposure significant work areas within radiation areas, high radiation areas (<1 R/hr), or airborne radioactivity areas in the plant and reviewed the associated controls and surveys of these areas to determine if controls (e.g., surveys, postings, barricades) were acceptable. The areas reviewed by the inspectors included the drywell, inside the bioshield, under vessel and on the refueling floor.


radiation areas (<1 R/hr), or airborne radioactivity areas in the plant and reviewed the
With a survey instrument, the inspectors walked down these areas or their perimeters to determine whether prescribed radiation work permits, procedure, and engineering controls were in place, whether surveys and postings were complete and accurate, and whether air samplers were properly located.


associated controls and surveys of these areas to determine if controls (e.g., surveys, postings, barricades) were acceptable. The areas reviewed by the inspectors included the
The inspectors reviewed radiation work permits used to access these and other high radiation areas and identified what work control instructions or control barriers had been specified. The inspectors used plant-specific TS high radiation area requirements as the standard for the necessary barriers. The inspectors reviewed electronic personal dosimeter alarm set points (both integrated dose and dose rate) for conformity with survey indications and plant policy. The inspectors verified that workers knew what actions are required when their electronic personal dosimeter noticeably malfunctions or alarms.


drywell, inside the bioshield, under vessel and on the refueling floor.
The inspectors evaluated performance against the requirements contained in 10 CFR Part 20, Unit 1 TS 6.7, and Unit 2 TS 6.12.
 
With a survey instrument, the inspectors wa lked down these areas or their perimeters to determine whether prescribed radiation work permits, procedure, and engineering controls
 
were in place, whether surveys and postings were complete and accurate, and whether air
 
samplers were properly located.
 
The inspectors reviewed radiation work permits used to access these and other high
 
radiation areas and identified what work control instructions or control barriers had been
 
specified. The inspectors used plant-specific TS high radiation area requirements as the
 
standard for the necessary barriers. The inspectors reviewed electronic personal
 
dosimeter alarm set points (both integrated dose and dose rate) for conformity with survey
 
indications and plant policy. The inspectors verified that workers knew what actions are
 
required when their electronic personal dosimeter noticeably malfunctions or alarms.
 
The inspectors evaluated performance against the requirements contained in 10 CFR Part  
 
20, Unit 1 TS 6.7, and Unit 2 TS 6.12.


====b. Findings====
====b. Findings====
Line 755: Line 330:


====a. Inspection Scope====
====a. Inspection Scope====
The inspectors obtained a list of work activities ranked by actual/estimated exposure that  
The inspectors obtained a list of work activities ranked by actual/estimated exposure that were in progress during the refueling outage and selected three work activities of highest exposure significance (see section 2OS1 above).


were in progress during the refueling outage and selected three work activities of highest
The inspectors reviewed the ALARA work activity evaluations, exposure estimates, and exposure mitigation requirements. The inspectors determined whether procedures, engineering and work controls had been established based on sound radiation protection principles to achieve occupational exposures that were ALARA. The inspectors determined whether the radiological work had been reasonably grouped into work activities based on historical precedence, industry norms, and/or special circumstances.


exposure significance (see section 2OS1 above).
The inspectors compared the results achieved (dose rate reductions, person-rem used)with the intended dose established in ALARA planning for these work activities. The inspectors reviewed, where applicable, inconsistencies between intended and actual work activity doses.


The inspectors reviewed the ALARA work activity evaluations, exposure estimates, and
Based on scheduled work activities and associated exposure estimates, the inspectors selected three work activities in radiation areas, airborne radioactivity areas, or high radiation areas for observation. The inspectors concentrated on work activities that presented the greatest radiological risk to workers. The inspectors evaluated use of ALARA controls for these work activities by evaluating use of engineering controls to achieve dose reductions.


exposure mitigation requirements. The inspectors determined whether procedures, 17 engineering and work controls had been established based on sound radiation protection
The inspectors evaluated Constellations performance against the requirements contained in 10 CFR Part 20.1101.
 
principles to achieve occupational exposures that were ALARA. The inspectors
 
determined whether the radiological work had been reasonably grouped into work
 
activities based on historical precedence, i ndustry norms, and/or special circumstances.
 
The inspectors compared the results achieved (dose rate reductions, person-rem used)
 
with the intended dose established in ALARA planning for these work activities. The
 
inspectors reviewed, where applicable, inconsistencies between intended and actual work
 
activity doses.
 
Based on scheduled work activities and associated exposure estimates, the inspectors selected three work activities in radiation areas, airborne radioactivity areas, or high
 
radiation areas for observation. The inspectors concentrated on work activities that
 
presented the greatest radiological risk to workers. The inspectors evaluated use of
 
ALARA controls for these work activities by evaluating use of engineering controls to achieve dose reductions.
 
The inspectors evaluated Constellation's performance against the requirements contained  
 
in 10 CFR Part 20.1101.


====b. Findings====
====b. Findings====
Line 797: Line 346:


====a. Inspection Scope====
====a. Inspection Scope====
The inspectors identified the types of portable radiation detection instrumentation used for  
The inspectors identified the types of portable radiation detection instrumentation used for job coverage of high radiation area work, other temporary area radiation monitors currently used in the plant, and continuous air monitors associated with jobs with the potential for workers to receive 50 mrem committed effective dose equivalent.


job coverage of high radiation area work, other temporary area radiation monitors currently
The inspectors evaluated performance against the requirements contained in 10 CFR Part 20.1501, 10 CFR Part 20.1703 and 10 CFR Part 20.1704.
 
used in the plant, and continuous air monitors associated with jobs with the potential for
 
workers to receive 50 mrem committed effective dose equivalent.
 
The inspectors evaluated performance against the requirements contained in  
 
10 CFR Part 20.1501, 10 CFR Part 20.1703 and 10 CFR Part 20.1704.


====b. Findings====
====b. Findings====
Line 814: Line 355:
==OTHER ACTIVITIES==
==OTHER ACTIVITIES==
{{a|4OA1}}
{{a|4OA1}}
==4OA1 Performance Indicator Verification (71151 - Four samples)==
==4OA1 Performance Indicator Verification (71151 - Four samples)==
====a. Inspection Scope====
====a. Inspection Scope====
The inspectors sampled NMPNS submittals fo r the performance indicators (PIs) listed below. To verify the accuracy of the PI data reported during that period, the PI definition 18 guidance contained in Nuclear Energy Institute (NEI) 99-02, "Regulatory Assessment  
The inspectors sampled NMPNS submittals for the performance indicators (PIs) listed below. To verify the accuracy of the PI data reported during that period, the PI definition guidance contained in Nuclear Energy Institute (NEI) 99-02, "Regulatory Assessment Indicator Guideline," Revision 5, was used to verify the basis in reporting for each data element.
 
Indicator Guideline," Revision 5, was used to verify the basis in reporting for each data  


element. Cornerstone: Initiating Events
===Cornerstone: Initiating Events===
 
The inspectors reviewed licensee event reports (LERs) and operator logs to determine whether NMPNS accurately reported the number of unplanned scrams at Unit 1 and Unit 2 from July 2007 to December 2007.
The inspectors reviewed licensee event reports (LERs) and operator logs to determine  
 
whether NMPNS accurately reported the num ber of unplanned scrams at Unit 1 and Unit 2 from July 2007 to December 2007.
* Unit 1 and Unit 2 unplanned scrams per 7000 critical hours; and
* Unit 1 and Unit 2 unplanned scrams per 7000 critical hours; and
* Unit 2 and Unit 2 unplanned scrams with complications.
* Unit 2 and Unit 2 unplanned scrams with complications.


====b. Findings====
====b. Findings====
No findings of significance were identified.
No findings of significance were identified. {{a|4OA2}}


{{a|4OA2}}
==4OA2 Identification and Resolution of Problems==
==4OA2 Identification and Resolution of Problems==
 
{{IP sample|IP=IP 71152}}
(71152)


====a. Inspection Scope====
====a. Inspection Scope====
As specified by Inspection Procedure 71152, "Identification and Resolution of Problems,"
As specified by Inspection Procedure 71152, Identification and Resolution of Problems, and in order to help identify repetitive equipment failures or specific human performance issues for follow-up, the inspectors performed a daily screening of items entered into Nine Mile Points CAP. In accordance with the baseline inspection procedures, the inspectors also identified selected CAP items across the initiating events, mitigating systems, and barrier integrity cornerstones for additional follow-up and review. The inspectors assessed the threshold for problem identification, the adequacy of the cause analyses, extent of condition review, operability determinations, and the timeliness of the specified corrective actions.
 
and in order to help identify repetitive equipment failures or specific human performance  


issues for follow-up, the inspectors performed a daily screening of items entered into Nine
The ISI inspectors reviewed a sample of examination reports and CRs initiated during ISI examinations to evaluate Constellations effectiveness in the identification and resolution of problems. The inspectors reviewed CRs 2008-2332, 2008-2345, and 2008-2363, which identified flaws and other nonconforming conditions discovered during this outage. The inspectors verified that the nonconforming conditions identified were reported, characterized, and entered into the CAP.
 
Mile Point's CAP. In accordance with the baseline inspection procedures, the inspectors
 
also identified selected CAP items across t he initiating events, mitigating systems, and barrier integrity cornerstones for additional follow-up and review. The inspectors assessed
 
the threshold for problem identification, the adequacy of the cause analyses, extent of
 
condition review, operability determinations, and the timeliness of the specified corrective
 
actions.
 
The ISI inspectors reviewed a sample of examination reports and CRs initiated during ISI examinations to evaluate Constellation's effectiveness in the identification and resolution of problems. The inspectors reviewed CRs 2008-2332, 2008-2345, and 2008-2363, which  
 
identified flaws and other nonconforming conditions discovered during this outage. The  
 
inspectors verified that the nonconforming conditions identified were reported, characterized, and entered into the CAP.


====b. Findings====
====b. Findings====
No findings of significance were identified.
No findings of significance were identified.
{{a|4OA6}}


{{a|4OA6}}
==4OA6 Meetings, including Exit==
==4OA6 Meetings, including Exit==
===Exit Meeting Summary===
===Exit Meeting Summary===
The inspectors presented the inspection results to Mr. Keith Polson and other members of NMPNS management on April 11, 2008. NMPNS acknowledged that no proprietary information was involved.


The inspectors presented the inspection results to Mr. Keith Polson and other members of
ATTACHMENT:  
 
NMPNS management on April 11, 2008. NM PNS acknowledged that no proprietary information was involved.
 
ATTACHMENT:


=SUPPLEMENTAL INFORMATION=
=SUPPLEMENTAL INFORMATION=


==KEY POINTS OF CONTACT==
==KEY POINTS OF CONTACT==
===Licensee Personnel===
===Licensee Personnel===
: [[contact::K. Polson]], Vice President  
: [[contact::K. Polson]], Vice President  
Line 891: Line 402:


==LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED==
==LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED==
===Opened and Closed===
: 05000410/2008002-01


===Opened and Closed===
NCV  
: 05000410/2008002-01  NCV Failure to Correctly Perform Procedure
Caused Inadvertent Isolation of RCIC Steam


Supply (Section 1R22)  
Failure to Correctly Perform Procedure Caused Inadvertent Isolation of RCIC Steam Supply (Section 1R22)  


===Closed===
===Closed===
: None.  
None.  


===Discussed===
===Discussed===
None.  
None.  


==LIST OF DOCUMENTS REVIEWED==
==LIST OF DOCUMENTS REVIEWED==
==Section 1R04: Equipment Alignment==
: SDBD-204,"Emergency Cooling System Design Basis Document," Revision 07
: N1-OP-13, "Emergency Cooling System," Revision 33
: C-18017-C, "Emergency Cooling System P&I Diagram," Revision 54
: N1-ST-C22, "Emergency Cooling Vent Path Operability Test," Revision 04
: N1-ST-M2, "Emergency Cooling System Makeup Tank level Control Valves Exercising Test," Revision 15
: S14-60.1-F001, Overflow Loop Seal," Revision 00
: N2-OP-31, "Residual Heat Removal System," Revision 17
: N2-VLU-01, "Walkdown Order Valve Lineup and Valv
e Operations," Attachment 31, "N2-OP-31
: Walkdown Valve Lineup," Revision 00
: N2-OP-33, "High Pressure Core Spray System," Revision 07
: N2-VLU-01, "Walkdown Order Valve Lineup and Valv
e Operations," Attachment 33, "N2-OP-33
: Walkdown Valve Lineup," Revision 00
==Section 1R05: Fire Protection==
: NMPNS Unit 1 UFSAR, Appendix 10A, "Fire Hazards Analysis"
: NMPNS Unit 2 UFSAR, Appendix 9A, "Degree of Compliance with Branch Technical Position
: CMEB 9.5-1"
: NMPNS Unit 2 UFSAR, Appendix 9B, "Safe Shutdown Evaluation"
: GAP-INV-02, "Control of Material Storage Areas," Revision 19
==Section 1R08: Inservice Inspection Activities==
: Examination Procedures
: 54-ISI-835-12 R00 Ultrasonic Examination of Ferritic Piping Welds (Manual)
: PDI-UT-1 Rev D PDI Generic Procedure for the UT examination of ferritic pipe welds
: NDEP-PT-3.00 R16 Liquid Penetrant Examination
: NDEP-MT-4.00 R15 Magnetic Particle Examination
: NDEP-VT-2.01 R18 Visual Examination ASME XI
: Examination Reports
: 2-ANP-3.00-08-001 Liquid Penetrant Examination of 2RCS-64-00-SW95-98 Integral Welds (four lugs) 2-ANP-4.00-08-002 Magnetic Particle Examination of 2MSS-01-04-FW300/301 Integral Welds 2-ANP-835-08-005 Ultrasonic Examination, pipe to penetration butt weld, core spray system 
===Work Orders===
: 04-08487-00
: Repair leaking boundary valve 2IAS-V181, disassemble and repair
: 05-21585-00
: Remove, rebuild, calibrate and replace relief valve 2WCS-RV21A
: Welding Procedures
: WPS-8-8-BA-102 Manual gas tungsten arc (GTAW) and shielded metal arc (SMAW)
welding of P8 to P8
: WPS-1-1-BA-101 Manual GTAW and SMAW of P1 to P1 
===Procedure===
: Qualification Records
: PQR N107
: Manual GTAW and SMAW procedure qualification record (P8 to P8)
: PQR N120
: Manual GTAW and SMAW procedure qualification record (P1 to P1)
: PQR N177
: Manual GTAW and SMAW procedure qualification record (P1 to P1)
: PQR N203
: Shielded Metal Arc Welding procedure qualification record (P8 to P8) 
===Drawings===
: PID-32A R16
: Piping Low Pressure Core Spray
: ISI-26-05
: Core Spray ISI Isometric 
===Miscellaneous===
: Examiner 1664 Magnetic Particle Testing Performance Qualification Record
: Examiner 9893 Ultrasonic Testing PDI Performance Qualification Record Welder A, B and C Performance Qualification Records (ASME Section IX)
==Section 1R11: Licensed Operator Requalification Program==
: N2-OP-101A, "Plant Startup," Revision 18
: N2-OSP-EGS-R004, "Operating Cycle Diesel Generator Simulated Loss of Offsite Power with 
: ECCS Division I and II," Revision 08
: N2-OP-101C, "Plant Shutdown," Revision 18
==Section 1R12: Maintenance Effectiveness==
: Unit 2 Integrated Scoping Matrix
: Unit 2 Integrated Performance Criteria Matrix
: Unit 2 High Safety Significance Functions and Related Key Safety Functions, Revision 15
: S-MRM-REL-0101, "Maintenance Rule," Revision 18
: S-MRM-REL-0104, "Maintenance Rule Scope," Revision 1
: S-MRM-REL-0105, "Maintenance Rule Performance Criteria," Revision 1
: CR 2008-728
: CR 2008-912
: CR 2008-981
: CR 2008-1016
: CR 2008-1024
: CR 2008-1837
==Section 1R13: Maintenance Risk Assessments and Emergent Work Control==
: GAP-OPS-117, "Integrated Risk Management," Revision 14
: GAP-PSH-03, "Control of On-line Work Activities," Revision 15
: NAI-PSH-03, "On-line Work Management Process," Revision 11
==Section 1R15: Operability Evaluations==
: CNG-OP-1.01-1002, "Conduct of Operability Determinations / Functionality Assessments," Revision 00
: CR 2008-531
: CR 2008-618
: CR 2008-1276
: CR 2008-1721
: CR 2008-2159
: CR 2008-2176
: CR 2007-7407
: CR 2007-5154
: CR 2006-3751
: CR 2006-436
==Section 1R18: Plant Modifications==
: N2-05-010, "Eliminate Single Point Vulnerability for Main Steam Tunnel Cooling"
==Section 1R19: Post Maintenance Testing==
: GAP-SAT-02, "Pre/Post Maintenance Test Requirements," Revision 26
==Section 1R20: Refueling and Other Outage Activities==
: Outage Schedule Shutdown Safety Review Report for NMP2 Refueling Outage N2R11
: N2-OP-101C, "Plant Shutdown," Revision 15
: NIP-OUT-01, "Shutdown Safety," Revision 20
: GAP-PSH-01, "Work Control," Revision 42
: GAP-OPS-02, "Control of Hazardous Energy, Clearance, and Tagging," Revision 24
: N2-MPM-GEN-903, "Reactor Vessel Disassembly," Revision 02
: N2-FHP-003, "Refueling Manual," Revision 07
: N2-FHP-13.3, "Core Shuffle," Revision 02
: Shutdown Safety Contingency Plan N2R11-001, "Division 1 LOP/LOCA and ECCS Functional Testing" Shutdown Safety Contingency Plan N2R11-002, "Reactor Cavity Floodup" Shutdown Safety Contingency Plan N2R11-003, "Reactor Cavity Drain Down to Mode 4" Shutdown Safety Contingency Plan N2R11-004, "Div ision 1 Electrical Work with 2SFP*P1A 
: Protected" N2-SOP-38, "Loss of Spent Fuel Pool Cooling," Revision 03
: N2-SOP-31, "Loss of Shutdown Cooling," Revision 04
: N2-SOP-31R, "Refueling Operations Alternate Shutdown Cooling," Revision 04
==Section 1R22: Surveillance Testing==
: CNG-HU-1.01, "Human Performance Program," Revision 01
: CNG-HU-1.01-1000, "Human Performance," Revision 02
: CNG-HU-1.01-1001, "Human Performance Tools and Verification Practices," Revision 02
: CNG-HU-1.01-1002, "Pre-Job Briefings and Post-Job Critiques," Revision 02
: GAP-SAT-01, "ST Program," Revision 16
: GAP-OPS-117, "Integrated Risk Management," Revision 14
: NMPNS-IST-001, "Pump and Valve Inse rvice Testing Program," Revision 00
: MDC-11, "Pump Curves and Acceptance Criteria," Revision 14
==Section 1EP6: Drill Evaluation==
: EPIP-EPP-01, "Classification of Emergency Conditions at Unit 1," Revision 17
: EPIP-EPP-20, "Emergency Notifications," Revision 18
: N1-EOP-5, "Secondary Containment Control," Revision 13
: Emergency Preparedness Scenario for the EP Drill to be Conducted on March 4, 2008
==Section 2OS1: Access Control to Radiologically Significant Areas==
: Radiation Work Permits:
: 2505 (Miscellaneous Drywell Maintenance); 2700 (Refuel Floor
: Activities); 2515 (SRV Activities); 2510 (Drywe ll Scaffold); 2502 (Under Vessel Activities); 2507 (Drywell ISI); 2511 (Drywell Insulation)
==Section 2OS2: ALARA Planning and Controls==
: ALARA Reviews:
: 08-2-20 (Refuel Floor Activiti es); 08-2-14 (SRV Activities); 08-2-10 (Drywell Scaffold); 08-2-02 (Under Vessel Activities); 08-2-06 (Drywell ISI); 08-2-11 (Drywell Insulation);
: RFO11 Radiation Protection Pre-Outage ALARA Review
==Section 4OA2: Identification and Resolution of Problems==
===Procedures===
: NIP-ECA-01, "Corrective Action Program," Revision 46 
===Condition Reports===
: 2008-2432
: 2008-2345
: 2008-2463
: 2008-3253
: 2008-2875
: 2008-2882
: 2008-1993
: 2008-1569
: 2008-1578
: 2008-1721
: 2008-1440
: 2008-1271
: 2008-0794
: 2008-0816
: 2008-0860
: 2008-0531
: 2008-0246
==LIST OF ACRONYMS==
: [[AC]] [[alternating current]]
ADAMS Agency Documents
Access Management System
: [[ALARA]] [[as low as reasonably achievable]]
: [[ANSI]] [[American National Standards Institute]]
: [[APRM]] [[average power range monitor]]
: [[ASME]] [[American Society of Mechanical Engineers]]
: [[CAP]] [[corrective action program]]
: [[CFR]] [[Code of Federal Regulations]]
: [[CR]] [[condition report]]
: [[CRD]] [[control rod drive]]
: [[DBD]] [[design basis document]]
: [[EC]] [[emergency cooling]]
: [[ECCS]] [[emergency core cooling system]]
: [[EDG]] [[emergency diesel generator]]
: [[GTAW]] [[gas tungsten arc welding]]
: [[HPCI]] [[high pressure coolant injection]]
: [[HPCS]] [[high pressure core spray]]
: [[IMC]] [[inspection manual chapter]]
: [[ISI]] [[inservice inspection]]
: [[JITT]] [[just in time training]]
: [[LER]] [[licensee event report]]
: [[MSIV]] [[main steam isolation valve]]
: [[NCV]] [[non-cited violation]]
: [[NDE]] [[nondestructive examination]]
: [[NEI]] [[Nuclear Energy Institute]]
: [[NMPNS]] [[Nine Mile Point Nuclear Station,]]
: [[LLC]] [[]]
NRC  Nuclear Regulatory Commission
mrem  millirem
: [[OS]] [[Occupational Radiation Safety]]
: [[PARS]] [[Publicly Available Records]]
: [[PI]] [[performance indicator]]
: [[RCIC]] [[reactor core isolation cooling]]
: [[RCS]] [[reactor coolant system]]
: [[RFO]] [[refueling outage]]
: [[RHR]] [[residual heat removal]]
: [[RTP]] [[rated thermal power]]
: [[SDP]] [[significance determination process]]
: [[SFP]] [[spent fuel pool]]
: [[SLC]] [[standby liquid control]]
: [[SMAW]] [[shielded metal arc welding]]
: [[SPAR]] [[Standardized Plant Analysis Risk]]
: [[SRA]] [[senior reactor analyst]]
: [[ST]] [[surveillance test]]
SW  service water 


A-7TBCLC turbine building closed loop cooling
: [[TS]] [[technical specification]]
: [[TSC]] [[technical support center]]
: [[UFSAR]] [[updated final safety analysis report]]
: [[UPS]] [[uninterruptable power supply]]
: [[UT]] [[ultrasonic test]]
: [[WO]] [[work order]]
}}
}}

Latest revision as of 16:54, 14 January 2025

IR 05000220-08-002, 05000410-08-002; 01/01/08 - 03/31/08; Nine Mile Point Nuclear Station, Units 1 and 2; Surveillance Testing
ML081270471
Person / Time
Site: Nine Mile Point  Constellation icon.png
Issue date: 05/05/2008
From: Glenn Dentel
Reactor Projects Branch 1
To: Polson K
Nine Mile Point
Dentel, G RGN-I/DRP/BR1/610-337-5233
References
IR-08-002
Download: ML081270471 (31)
v • d • e


Text

May 5, 2008

SUBJECT:

NINE MILE POINT NUCLEAR STATION - NRC INTEGRATED INSPECTION REPORT 05000220/2008002 and 05000410/2008002

Dear Mr. Polson:

On March 31, 2008, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your Nine Mile Point Nuclear Station, Units 1 and 2. The enclosed integrated inspection report documents the inspection results discussed on April 11, 2008, with you and 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.

This report documents one self-revealing finding of very low safety significance (Green). The finding was determined to involve a violation of NRC requirements. However, because of its very low safety significance and because it was entered into your corrective action program (CAP), the NRC is treating this finding as a non-cited violation in accordance with Section VI.A.1 of the NRCs Enforcement Policy. If you contest the non-cited violation noted in this report, you should provide a response with the basis for your denial, within 30 days of the date of this inspection report, to the Nuclear Regulatory Commission, ATTN.: Document Control Desk, Washington, D.C. 20555-0001; with copies to the Regional Administrator, Region I; the Director, Office of Enforcement; U.S. Nuclear Regulatory Commission, Washington, D.C. 20555-001; and the NRC Senior Resident Inspector at Nine Mile Point Nuclear Station.

In accordance with 10 CFR Part 2.390 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 NRC Public Document Room or from the Publicly Available Records (PARS) component of the NRCs document 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/

Glenn T. Dentel, Chief Projects Branch 1 Division of Reactor Projects

Docket No.:

50-220, 50-410 License No.: DPR-63, NPF-69

Enclosure:

Inspection Report 05000220/2008002 and 05000410/2008002 w/Attachment: Supplemental Information

cc w/encl:

M. Wallace, President, Constellation Generation B. Barron, Senior Vice President and Chief Nuclear Officer C. Fleming, Esquire, Senior Counsel, Constellation Energy Group, LLC M. Wetterhahn, Esquire, Winston and Strawn T. Syrell, Director, Licensing, Nine Mile Point Nuclear Station P. Tonko, President and CEO, New York State Energy Research and Development Authority J. Spath, Program Director, New York State Energy Research and Development Authority P. D. Eddy, Electric Division, NYS Department of Public Service C. Donaldson, Esquire, Assistant Attorney General, New York Department of Law Supervisor, Town of Scriba T. Judson, Central NY Citizens Awareness Network D. Katz, Citizens Awareness Network

SUMMARY OF FINDINGS

IR 05000220/2008002, 05000410/2008002; 01/01/08 - 03/31/08; Nine Mile Point Nuclear Station,

Units 1 and 2; Surveillance Testing.

The report covered a three-month period of inspection by resident inspectors and regional specialist inspectors. One Green non-cited violation (NCV) was identified. The significance of most findings is indicated by their color (Green, White, Yellow, Red) using Inspection Manual Chapter (IMC) 0609, Significance Determination Process (SDP). Findings for which the SDP does not apply may be Green or be assigned a severity level after NRC management review.

The NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process, Revision 4, dated December 2006.

NRC-Identified and Self-Revealing Findings

Cornerstone: Mitigating Systems

Green.

A self-revealing, non-cited violation (NCV) of Technical Specification (TS) 5.4,

"Procedures," was identified on January 14, 2008, when technicians improperly performed a surveillance procedure which resulted in isolation of the Unit 2 RCIC system. Specifically, while performing a test of the area temperature instruments that provide high temperature isolation signals for the main steam system, technicians erroneously disconnected an electrical lead associated with the RCIC leak detection system. This resulted in an automatic isolation of the RCIC system steam supply and the unavailability of RCIC for approximately four hours. Operators immediately recognized the error and halted the surveillance procedure. Technicians reconnected the lead and operators restored RCIC to a normal standby lineup.

The finding was greater than minor because it was associated with the human performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding was determined to be of very low safety significance in accordance with IMC 0609,

Appendix AProperty "Inspection Manual Chapter" (as page type) with input value "NRC Inspection Manual 0609,</br></br>Appendix A" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process., "Determining the Significance of Reactor Inspection Findings for At-Power Situations," based on a Phase 3 analysis. The Region I senior reactor analyst (SRA)used the Nine Mile Point Unit 2 Standardized Plant Analysis Risk (SPAR) model and the actual out-of-service time to determine the risk significance. This finding had a cross-cutting aspect in the area of human performance because of the ineffective use of human error prevention techniques (H.4.a per IMC 0305). (Section 1R22)

Licensee-Identified Violations

None.

REPORT DETAILS

Summary of Plant Status

Nine Mile Point Unit 1 was operated at full rated thermal power (RTP) throughout the inspection period, with the exception of planned power reductions and recoveries for planned reactor recirculation pump maintenance, control rod testing, and main turbine valve testing.

Nine Mile Point Unit 2 began the inspection period at full RTP. Operators performed several planned power reductions and recoveries for control rod pattern adjustments, main turbine and main steam isolation valve testing, and control rod testing. On March 22, the reactor was shut down to commence refueling outage (RFO)

REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity

==1R04 Equipment Alignment

==

.1 Partial System Walkdown (71111.04 - Four samples)

a. Inspection Scope

The inspectors performed four partial system walkdowns to verify risk-significant systems were properly aligned for operation. The inspectors verified the operability and alignment of these risk-significant systems while their redundant trains or systems were inoperable or out of service for maintenance. The inspectors compared system lineups to system operating procedures, system drawings, and the applicable chapters in the updated final safety analysis report (UFSAR). The inspectors verified the operability of critical system components by observing component material condition during the system walkdown.

Documents reviewed during this inspection are listed in the Attachment. The inspectors performed partial walkdowns of the following systems:

  • Unit 1 control room air treatment system while 112 and 121 control room chillers were out of service for corrective maintenance (February 1, 2008);
  • Unit 1 core spray system 12 during inservice testing of core spray system 11 (February 26, 2008); and

b. Findings

No findings of significance were identified.

.2 Complete System Walkdown (71111.04S - One sample)

a. Inspection Scope

The inspectors performed a complete walkdown of the Unit 1 emergency cooling system to identify discrepancies between the existing equipment configuration and that specified in the design documents. During the walkdown, system drawings and operating procedures were used to determine the proper equipment alignment and operational status. The inspectors reviewed the open maintenance work orders (WO) that could affect the ability of the system to perform its functions. Documentation associated with temporary modifications, operator workarounds, and items tracked by plant engineering were also reviewed to assess their collective impact on system operation. In addition, the inspectors reviewed the condition report (CR) database to verify that equipment alignment problems were being identified and appropriately resolved. Documents reviewed during this inspection are listed in the Attachment.

b. Findings

No findings of significance were identified.

==1R05 Fire Protection (71111.05Q - Six samples)

a. Inspection Scope

==

The inspectors toured six areas important to reactor safety at NMPNS to evaluate the stations control of transient combustibles and ignition sources, and to examine the material condition, operational status, and operational lineup of fire protection systems including detection, suppression, and fire barriers. Documents reviewed for this inspection are listed in the Attachment. The areas inspected included:

  • Unit 1 train 11 battery and battery board rooms;
  • Unit 1 train 12 battery and battery board rooms;
  • Unit 1 containment spray pump room (112, 122), reactor building (RB) 198 and 237 foot elevations;
  • Unit 2 RB 175 foot elevation;
  • Unit 2 RB 196 foot elevation; and
  • Unit 2 steam tunnel;

b. Findings

No findings of significance were identified.

==1R08 Inservice Inspection Activities (71111.08 - One sample)

a. Inspection Scope

==

The purpose of this inspection was to assess the effectiveness of the inservice inspection (ISI) program for monitoring degradation of the reactor coolant system (RCS) boundary, risk significant piping system boundaries, and the containment boundary. The inspectors assessed the ISI activities using the criteria specified in the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI and applicable NRC regulatory requirements.

The inspectors selected a sample of nondestructive examination (NDE) activities for observation or review, and evaluation for compliance with the requirements of ASME Section XI. Also, the inspectors selected samples of activities associated with the repair/replacement of safety related pressure boundary components. The sample selection was based on the inspection procedure objectives, risk significance, and availability. Specifically, the inspectors focused on components and systems where degradation would result in a significant challenge to the integrity of pressure boundary components.

The inspectors performed an observation of one volumetric examination (ultrasonic) and portions of a surface examination (liquid penetrant). In addition, the inspectors performed a documentation review of a magnetic particle surface examination. The sample selection included the following:

  • Magnetic particle test, surface examination, welds #300 and 301, integral attachments, lugs to pipe, main steam system; and

The inspectors performed an evaluation of work activities during a drywell entry and visually examined the condition of accessible portions of the containment liner and coatings for peeling, blistering, corrosion, mechanical damage, and other degradation mechanisms. The inspectors noted that two different coatings were apparent on various locations of the internal exposed metallic surfaces of the containment liner. The inspectors reviewed documentation which supported the coating qualification in accordance with ANSI N101.2 and that all coating had been applied in accordance with Regulatory Guide 1.54.

The inspectors reviewed portions of the in-process remote visual examination of the steam dryer. The inspectors reviewed three CRs initiated as a result of the dryer examination and noted the rejectable indications reported. The indications noted had not been identified during the previous examination (previous outage in 2006). These issues were placed in the corrective action program for engineering evaluation and disposition.

The inspectors selected for review a sample of repair/rework activities which required the development and implementation of an ASME Section XI repair plan. The inspectors reviewed documentation for the weld repair of two pressure boundary valves in ASME Class 3 systems. The inspectors reviewed the ASME Section XI plans, work scope, activity sequence, weld filler metal selection, weld procedure specifications and procedure qualification records, welder qualifications, specified non-destructive tests, acceptance criteria, and post work testing. The following samples were inspected:

  • WO 04-08487-00 was initiated for the mechanical and weld repair of globe valve 2IAS-V181 in the instrument air system. The repair involved the disassembly and rebuilding of the valve. The disassembly of the valve required the removal of the body to bonnet weld to access the internals for mechanical rework of the valve seats.

Restoration of the body to bonnet weld was required following the completion of the repair and installation of the valve internals.

  • WO 05-21585-00 was initiated to facilitate the removal, testing, rebuilding, inspection and re-installation by welding, into the piping system of relief valve 2WCS-RV21A in the reactor water cleanup system. It was necessary to eliminate existing installation welds in order to remove, rebuild, and test the valve. Acceptance testing of the completed valve repair and welding was specified in the repair plan. A visual examination was specified for the installation welds and a system pressure test specified to verify valve and system integrity.

No sample of a previously identified recordable indication accepted as-is for continued service from the previous and the current outage was available for review during the inspection.

b. Findings

No findings of significance were identified.

==1R11 Licensed Operator Requalification Program (71111.11Q - Two samples)

a. Inspection Scope

==

The inspectors evaluated two simulator scenarios licensed operator requalification training program. The inspectors assessed the clarity and effectiveness of communications, the implementation of appropriate actions in response to alarms, the performance of timely control board operation, and the oversight and direction provided by the shift manager.

During the scenario, the inspectors also compared simulator performance with actual plant performance in the control room. Documents reviewed for this inspection are listed in the

. The following scenarios were observed:

  • On March 17, 2008, the inspectors observed a Unit 2 operations crew during Just In Time Training (JITT) in preparation for RFO 11. The crew performed an approach to criticality, discussed the performance of surveillance procedure N2-OSP-EGS-R004, Operating Cycle Diesel Generator Simulated Loss of Offsite Power with ECCS Division I and II, and discussed plant modifications that would be performed during the outage.
  • On March 18, 2008, the inspectors observed a Unit 2 operations crew during JITT training in preparation for RFO 11. The crew performed a plant cooldown, including the transition to RHR shutdown cooling in service.

b. Findings

No findings of significance were identified.

==1R12 Maintenance Effectiveness (71111.12Q - Two samples)

a. Inspection Scope

==

The inspectors reviewed performance-based problems and the performance and condition history of selected systems to assess the effectiveness of the maintenance program. The inspectors reviewed the systems to ensure that the stations review focused on proper maintenance rule scoping in accordance with 10 CFR Part 50.65, characterization of reliability issues, tracking system and component unavailability, and 10 CFR Part 50.65 (a)(1) and (a)(2) classification. In addition, the inspectors reviewed the sites ability to identify and address common cause failures and to trend key parameters. Documents reviewed for the inspection are listed in the Attachment. The following two maintenance rule inspection samples were reviewed:

  • Unit 1 fire protection systems due to long-standing equipment problems; and

b. Findings

No findings of significance were identified.

==1R13 Maintenance Risk Assessments and Emergent Work Control (71111.13 - Six samples)

a.

==

Inspection Scope

The inspectors evaluated the effectiveness of the maintenance risk assessments required by paragraph (a)(4) of 10 CFR Part 50.65. The inspectors reviewed equipment logs, work schedules, and performed plant tours to gain assurance that actual plant configuration matched the assessed configuration. Additionally, the inspectors verified that risk management actions for both planned and emergent work were consistent with those described in station procedures. Documents reviewed for the inspection are listed in the

.

The inspectors reviewed risk assessments for the activities listed below.

Unit 1

  • Week of February 12, 2008, that included a two day maintenance period on 11 high pressure coolant injection (HPCI) system, cleaning of 11 turbine building closed loop cooling (TBCLC) heat exchanger, repair of a packing leak on emergency cooling (EC)valve IV-39-11R which rendered 11 EC inoperable for two days, maintenance on vital uninterruptable power supply (UPS) 162A, EDG raw water system quarterly surveillance, securing 11 reactor recirculation pump for maintenance on its associated motor generator, and 11 reactor recirculation flow loop calibration and flow converter calibrations.

Unit 2

  • Week of January 21, 2008, that included a two day maintenance period for the HPCS system, HPCS system quarterly surveillance, on-line motor testing and lubrication of the A control rod drive pump, Division 3 EDG monthly surveillance, and emergent maintenance to stop makeup water leakage into the standby liquid control tank.

b. Findings

No findings of significance were identified.

==1R15 Operability Evaluations (71111.15 - Seven samples)

a. Inspection Scope

==

The inspectors evaluated the acceptability of the operability evaluations, the use and control of compensatory measures, and the compliance with TSs. The evaluations were reviewed using criteria specified in NRC Regulatory Issue Summary 2005-20, Revision to Guidance Formerly Contained in NRC Generic Letter 91-18, Information to Licensees Regarding Two NRC Inspection Manual Sections on Resolution of Degraded and Nonconforming Conditions and on Operability, and Inspection Manual Part 9900, Operability Determinations and Functionality Assessments for Resolution of Degraded or Nonconforming Conditions Adverse to Quality or Safety. The inspectors review included verification that the operability determinations were made as specified by Procedure CNG-OP-1.01-1002, Conduct of Operability Determinations / Functionality Assessments.

The technical adequacy of the determinations was reviewed and compared to the TSs, UFSAR, and associated design basis documents (DBDs). Documents reviewed for the inspection are listed in the Attachment. The following evaluations were reviewed:

  • CR 2008-531 concerning turbine first stage bowl pressure switch calibrations at Unit 1;
  • CR 2006-3751 concerning environmental qualification of Unit 1 RB emergency ventilation damper position indicating switches;
  • CR 2008-1721 concerning leaking Unit 1 emergency condenser vacuum breaker valve 60.1-28;
  • CR 2007-7404 concerning Unit 2 Division 2 EDG operability with a failed emergency fuel oil solenoid valve;
  • CR 2008-1276 concerning identification of increased post-accident head losses associated with the Unit 2 ECCS suppression pool suction strainers; and
  • CR 2008-2176 concerning out of specification resistance readings on the Unit 2 Division 1 EDG potential transformer fuse/contact linkage assembly.

b. Findings

No findings of significance were identified.

==1R18 Plant Modifications (71111.18 - One sample)

a. Inspection Scope

==

The inspectors reviewed Unit 2 permanent modification N2-05-010, Eliminate Single Point Vulnerability for Main Steam Tunnel Cooling. The purpose was to reduce the likelihood of a high temperature main steam line isolation due to loss of ventilation in the main steam lead enclosure. The inspectors assessed the adequacy of the modification package, including post-modification testing, and verified that applicable design and licensing basis requirements were met and that design margins were not degraded by the change.

b. Findings

No findings of significance were identified.

==1R19 Post Maintenance Testing (71111.19 - Five samples)

a. Inspection Scope

==

The inspectors reviewed the post maintenance tests listed below to verify that procedures and test activities ensured system operability and functional capability. The inspectors reviewed the test procedure to verify that the procedure adequately tested the safety functions that may have been affected by the maintenance activity, that the acceptance criteria in the procedure were consistent with information in the applicable licensing basis and/or DBDs, and that the procedure had been properly reviewed and approved. The inspectors also witnessed the test or reviewed test data, to verify that the test results adequately demonstrated restoration of the affected safety functions. Documents reviewed for this inspection are listed in the Attachment.

  • Unit 1, WO 07-10842-00 that performed maintenance on the tie breaker between non-vital 600V power board 17A and vital 600V power board 17B. The retest was performed in accordance with N1-OP-30, "4.16KV, 600V, and 480V House Service."
  • Unit 1, WO 07-08535-00 that repacked emergency condenser steam line drain valve IV-39-11. The retest was performed in accordance with N1-ST-Q4, Reactor Coolant System Isolation Valves Operability Test.
  • Unit 1, WO 08-02028-00 that performed maintenance on the reactor protection system motor generator MG-141 voltage regulator. The retest was performed in accordance with N1-OP-48, Motor Generator Sets.
  • Unit 1, WO 07-06872-00 that repaired leaking emergency condenser vacuum breaker valve 60.1-28. The retest was performed by overflowing the emergency condenser shell using N1-ST-M2, Emergency Cooling System Makeup Tank Level Control Valves Exercising Test.
  • Unit 2, WO 07-01190-00 that performed inspection of the A CRD pump motor. The retest for the circuit breaker rack-out was performed in accordance with N2-OP-30, Control Rod Drive.

b. Findings

No findings of significance were identified.

==1R20 Refueling and Other Outage Activities (71111.20 - In Progress)

a. Inspection Scope

==

The inspectors observed and/or reviewed the following Unit 2 refueling outage activities to verify that operability requirements were met and that risk, industry experience, and previous site-specific problems were considered. The refueling outage and inspection sample were in progress at the end of the inspection period. Documents reviewed for this inspection are listed in the Attachment.

  • The inspectors reviewed the outage schedule and procedures, and verified that TS-required safety system availability was maintained and shutdown risk was minimized. The inspectors verified that, when specified by NMPNS procedure NIP-OUT-01, Shutdown Safety, contingency plans existed for restoring key safety functions.
  • The inspectors observed portions of the plant shutdown and cooldown on March 22, and verified that the TS cooldown rate limits were satisfied.
  • Through plant tours, the inspectors verified that NMPNS maintained and adequately protected electrical power supplies to safety-related equipment and that TS requirements were met.
  • The inspectors verified proper alignment and operation of shutdown cooling and other decay heat removal systems. The verification also included reactor cavity and fuel pool makeup paths and water sources, and administrative control of drain down paths.
  • The inspectors verified that requirements for refueling operations were met through refuel bridge observations, control room panel walkdowns, and discussions with Operations Department personnel.
  • After the drywell was open for general access, the inspectors performed an as-found walkdown to identify evidence of RCS leakage and assess the condition of drywell structures, piping, and supports.

b. Findings

No findings of significance were identified.

==1R22 Surveillance Testing (71111.22 - Eight samples)

a. Inspection Scope

==

The inspectors witnessed performance of and/or reviewed test data for risk-significant surveillance tests (STs) to assess whether the components and systems tested satisfied design and licensing basis requirements. The inspectors verified that test acceptance criteria were clear, demonstrated operational readiness and were consistent with the DBDs; that test instrumentation had current calibrations and the range and accuracy for the application; and that tests were performed, as written, with applicable prerequisites satisfied. Upon test completion, the inspectors verified that equipment was returned to the status specified to perform its safety function. Documents reviewed for this inspection are listed in the Attachment.

The following STs were reviewed:

  • N1-ST-M8, RB Emergency Ventilation System Operability Test;
  • N1-ST-Q21, Instrument Air Valves Quarterly Test;
  • N1-ISP-201-022, Drywell Water Leak Detection Instrument Channel Test;
  • N2-OSP-RHS-Q@006, "RHR System Loop C Pump and Valve Operability Test and System Integrity Test;"
  • N2-OSP-ICS-Q@002, "RCIC Pump and Valve Operability Test and System Integrity Test and ASME XI Functional Test;" and

b. Findings

Introduction.

A self-revealing Green NCV of TS 5.4, "Procedures," was identified on January 14, 2008, when technicians improperly performed a surveillance procedure which resulted in isolation of the Unit 2 RCIC system. Specifically, technicians erroneously disconnected an electrical lead associated with the RCIC leak detection system, which resulted in an automatic isolation of the RCIC system steam supply.

Description.

On January 14, 2008, instrument and controls technicians were performing an ST on the area temperature instruments that provide high temperature isolation signals for the main steam system. The surveillance, N2-ISP-LDS-R106, "Main Steam Line Tunnel and MSL Lead Enclosure Temperature Instrument Channel Calibration," requires that the associated thermocouple leads be disconnected prior to performing the channel calibration. When the technicians attempted to perform this action (step 7.2.1 of 1), they incorrectly identified the specified terminals. The procedure directed the technicians to disconnect thermocouple leads from 2CEC*PNL707 Bay F, TH405 TB-1, terminals 10 and 11. The lead that was actually disconnected was from terminal 14. One of the technicians had initially questioned the adequacy of their terminal identification since the terminals were not individually labeled. However, they concluded that they had identified the correct terminal and proceeded. The wires that they proceeded to disconnect were thermocouple leads for a temperature instrument that provides area high temperature isolation for the RCIC system. The open circuit created by lifting the first lead resulted in an automatic isolation of the RCIC system steam supply.

Operators immediately recognized the error and halted the surveillance procedure.

Technicians reconnected the thermocouple, and operators restored RCIC to a normal standby lineup. During the four hours that the RCIC steam supply was isolated, the RCIC system was inoperable and unavailable. The TS allowed outage time for the RCIC system is 14 days.

The performance deficiency associated with this event was that technicians did not correctly perform a ST procedure, which caused the Unit 2 RCIC system to automatically isolate, rendering the system unavailable to perform its safety function.

Analysis.

The finding was greater than minor because it was associated with the human performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to Initiating Events to prevent undesirable consequences.

The inspectors evaluated the significance of this finding using IMC 0609, Appendix A, Phase 1, and determined that a Phase 2 analysis was required because the finding represented an actual loss of the RCIC system safety function for four hours. The Region I SRA determined that a Phase 3 analysis was necessary because the site-specific Phase 2 notebook indicated that the finding could be more than of very low safety significance assuming an exposure time of three days. The SRA used the Nine Mile Point Unit 2 SPAR model and the actual four-hour exposure time to determine that the increase in core damage frequency was in the range of 1 core damage accident in 1.25E8 years of reactor operation, or high E-9 per year. The SPAR model dominant cutsets were a station blackout with failure of high pressure injection sources and the inability to restore AC power within 30 minutes. Based on this review, the SRA concluded that the finding was of very low safety significance (Green).

The finding had a cross-cutting aspect in the area of human performance because of the ineffective use of human error prevention techniques, in that, although peer checking had identified a question, that question was not adequately resolved prior to proceeding (H.4.a per IMC 0305)

Enforcement.

TS 5.4, Procedures, states that written procedures shall be established, implemented, and maintained covering the applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978. Regulatory Guide 1.33, Revision 2, Appendix A, February 1978, Item 8, Procedures for Control of Measuring and Test Equipment and for STs, Procedures, and Calibrations, lists containment isolation tests as an applicable group of tests. Contrary to the above, Unit 2 Instrument Surveillance Procedure N2-ISP-LDS-R106, "Main Steam Line Tunnel and MSL Lead Temperature Instrument Channel Calibration," was not correctly implemented.

On January 14, 2008, while attempting to perform Procedure Attachment 1, step 7.2.1, to disconnect field thermocouple leads from 2CEC*PNL707 Bay F, TH405 TB-1, terminals 10 and 11, technicians incorrectly disconnected the lead from terminal 14. This action resulted in an automatic isolation of the RCIC system steam supply. Because this procedural noncompliance is of very low safety significance and was entered into the CAP as CR 2008-332, this violation is being treated as an NCV, consistent with Section VI.A of the NRC Enforcement Policy: NCV 05000410/2008002-01, Failure to Correctly Perform Procedure Caused Inadvertent Isolation of the RCIC Steam Supply.

Cornerstone: Emergency Preparedness

1EP6 Drill Evaluation (71114.06 - One sample)

a. Inspection Scope

The inspectors completed one emergency drill evaluation inspection sample. The inspectors observed simulator, technical support center (TSC), and operations support center activities associated with the Unit 1 emergency planning drill on March 4, 2008. The scenario consisted of a leak from the spent fuel pool (SFP) due to an earthquake during the previous shift, a loss of off-site power (including power to the TSC) with failure of one main steam line to isolate, and a main steam line break outside secondary containment. The inspectors verified that emergency classification declarations and notifications were completed in accordance with 10 CFR 50.72, 10 CFR 50, Appendix E, and the Nine Mile Point emergency plan implementing procedures. Documents reviewed for this inspection are listed in the Attachment.

b. Findings

No findings of significance were identified.

RADIATION SAFETY

Cornerstone: Occupational Radiation Safety (OS)

2OS1 Access Control to Radiologically Significant Areas (71121.01 - Eight samples)

a. Inspection Scope

Based on the work activities during the Unit 2 refueling outage, the inspectors selected three jobs (drywell scaffold, drywell inservice inspection, and in-vessel visual inspection)being performed in radiation areas, airborne radioactivity areas, or high radiation areas

(<1 R/hr) for observation. The inspectors observed work that was estimated to result in the highest collective doses, involved diving activities in or around spent fuel or highly activated material, or that involved potentially changing (deteriorating) radiological conditions. The inspectors reviewed all radiological job requirements (radiation work permit requirements and work procedure requirements). The inspectors observed job performance with respect to these requirements. The inspectors determined if radiological conditions in the work area were adequately communicated to workers through briefings and postings.

During job performance observations, the inspectors verified the adequacy of radiological controls, such as required surveys (including system breach radiation, contamination, and airborne surveys), radiation protection job coverage (including audio and visual surveillance for remote job coverage), and contamination controls. For high radiation work areas with significant dose rate gradients (factor of 5 or more), the inspectors reviewed the application of dosimetry to effectively monitor exposure to personnel.

During job performance observations, the inspectors observed radiation worker performance with respect to stated radiation protection work requirements. The inspectors determined if workers were aware of the significant radiological conditions in their workplace and the radiation work permit controls/limits in place, and that their performance took into consideration the level of radiological hazards present.

During job performance observations, the inspectors observed radiation protection technician performance with respect to all radiation protection work requirements. The inspectors determined if they were aware of the radiological conditions in their work area and the radiation work permit controls/limits, and if their performance was consistent with their training and qualifications with respect to the radiological hazards and work activities.

The inspectors identified exposure significant work areas within radiation areas, high radiation areas (<1 R/hr), or airborne radioactivity areas in the plant and reviewed the associated controls and surveys of these areas to determine if controls (e.g., surveys, postings, barricades) were acceptable. The areas reviewed by the inspectors included the drywell, inside the bioshield, under vessel and on the refueling floor.

With a survey instrument, the inspectors walked down these areas or their perimeters to determine whether prescribed radiation work permits, procedure, and engineering controls were in place, whether surveys and postings were complete and accurate, and whether air samplers were properly located.

The inspectors reviewed radiation work permits used to access these and other high radiation areas and identified what work control instructions or control barriers had been specified. The inspectors used plant-specific TS high radiation area requirements as the standard for the necessary barriers. The inspectors reviewed electronic personal dosimeter alarm set points (both integrated dose and dose rate) for conformity with survey indications and plant policy. The inspectors verified that workers knew what actions are required when their electronic personal dosimeter noticeably malfunctions or alarms.

The inspectors evaluated performance against the requirements contained in 10 CFR Part 20, Unit 1 TS 6.7, and Unit 2 TS 6.12.

b. Findings

No findings of significance were identified.

2OS2 ALARA Planning and Controls (71121.02 - Four samples)

a. Inspection Scope

The inspectors obtained a list of work activities ranked by actual/estimated exposure that were in progress during the refueling outage and selected three work activities of highest exposure significance (see section 2OS1 above).

The inspectors reviewed the ALARA work activity evaluations, exposure estimates, and exposure mitigation requirements. The inspectors determined whether procedures, engineering and work controls had been established based on sound radiation protection principles to achieve occupational exposures that were ALARA. The inspectors determined whether the radiological work had been reasonably grouped into work activities based on historical precedence, industry norms, and/or special circumstances.

The inspectors compared the results achieved (dose rate reductions, person-rem used)with the intended dose established in ALARA planning for these work activities. The inspectors reviewed, where applicable, inconsistencies between intended and actual work activity doses.

Based on scheduled work activities and associated exposure estimates, the inspectors selected three work activities in radiation areas, airborne radioactivity areas, or high radiation areas for observation. The inspectors concentrated on work activities that presented the greatest radiological risk to workers. The inspectors evaluated use of ALARA controls for these work activities by evaluating use of engineering controls to achieve dose reductions.

The inspectors evaluated Constellations performance against the requirements contained in 10 CFR Part 20.1101.

b. Findings

No findings of significance were identified.

2OS3 Radiation Monitoring Instrumentation and Protective Equipment (71121.03 - One sample)

a. Inspection Scope

The inspectors identified the types of portable radiation detection instrumentation used for job coverage of high radiation area work, other temporary area radiation monitors currently used in the plant, and continuous air monitors associated with jobs with the potential for workers to receive 50 mrem committed effective dose equivalent.

The inspectors evaluated performance against the requirements contained in 10 CFR Part 20.1501, 10 CFR Part 20.1703 and 10 CFR Part 20.1704.

b. Findings

No findings of significance were identified.

OTHER ACTIVITIES

4OA1 Performance Indicator Verification (71151 - Four samples)

a. Inspection Scope

The inspectors sampled NMPNS submittals for the performance indicators (PIs) listed below. To verify the accuracy of the PI data reported during that period, the PI definition guidance contained in Nuclear Energy Institute (NEI) 99-02, "Regulatory Assessment Indicator Guideline," Revision 5, was used to verify the basis in reporting for each data element.

Cornerstone: Initiating Events

The inspectors reviewed licensee event reports (LERs) and operator logs to determine whether NMPNS accurately reported the number of unplanned scrams at Unit 1 and Unit 2 from July 2007 to December 2007.

  • Unit 1 and Unit 2 unplanned scrams per 7000 critical hours; and
  • Unit 2 and Unit 2 unplanned scrams with complications.

b. Findings

No findings of significance were identified.

4OA2 Identification and Resolution of Problems

Main article: IP 71152

a. Inspection Scope

As specified by Inspection Procedure 71152, Identification and Resolution of Problems, and in order to help identify repetitive equipment failures or specific human performance issues for follow-up, the inspectors performed a daily screening of items entered into Nine Mile Points CAP. In accordance with the baseline inspection procedures, the inspectors also identified selected CAP items across the initiating events, mitigating systems, and barrier integrity cornerstones for additional follow-up and review. The inspectors assessed the threshold for problem identification, the adequacy of the cause analyses, extent of condition review, operability determinations, and the timeliness of the specified corrective actions.

The ISI inspectors reviewed a sample of examination reports and CRs initiated during ISI examinations to evaluate Constellations effectiveness in the identification and resolution of problems. The inspectors reviewed CRs 2008-2332, 2008-2345, and 2008-2363, which identified flaws and other nonconforming conditions discovered during this outage. The inspectors verified that the nonconforming conditions identified were reported, characterized, and entered into the CAP.

b. Findings

No findings of significance were identified.

4OA6 Meetings, including Exit

Exit Meeting Summary

The inspectors presented the inspection results to Mr. Keith Polson and other members of NMPNS management on April 11, 2008. NMPNS acknowledged that no proprietary information was involved.

ATTACHMENT:

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee Personnel

K. Polson, Vice President
S. Belcher, Plant Manager
R. Dean, Director, Quality and Performance Assessment
J. Laughlin, Manager, Engineering Services
J. Krakuszeski, Manager, Operations
J. Kaminski, Manager, Emergency Preparedness
T. Shortell, Manager, Training
S. Sova, Manager, Radiation Protection
T. Syrell, Director, Licensing
W. Byrne, Manager, Nuclear Security

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened and Closed

05000410/2008002-01

NCV

Failure to Correctly Perform Procedure Caused Inadvertent Isolation of RCIC Steam Supply (Section 1R22)

Closed

None.

Discussed

None.

LIST OF DOCUMENTS REVIEWED

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