IR 05000280/2013005: Difference between revisions

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=Text=
=Text=
{{#Wiki_filter:UNITED STATES NUCLEAR REGULATORY COMMISSION REGION II 245 PEACHTREE CENTER AVENUE NE, SUITE 1200 ATLANTA, GEORGIA 30303-1257 February 10, 2014  
{{#Wiki_filter:UNITED STATES uary 10, 2014


Mr. David President and Chief Nuclear Officer Virginia Electric and Power Company Innsbrook Technical Center 5000 Dominion Boulevard Glen Allen, VA 23060-6711
==SUBJECT:==
 
SURRY POWER STATION - NRC INTEGRATED INSPECTION REPORT 05000280/2013005, 05000281/2013005
SUBJECT: SURRY POWER STATION - NRC INTEGRATED INSPECTION REPORT 05000280/2013005, 05000281/2013005


==Dear Mr. Heacock:==
==Dear Mr. Heacock:==
On December 31, 2013, the United States Nuclear Regulatory Commission (NRC) completed an inspection at your Surry Power Station Units 1 and 2. The enclosed inspection report documents the inspection findings which were discussed on January 29, 2014, with Mr. R. Simmons and other members of your staff. Inspectors documented the results of this inspection in the enclosed inspection report.
On December 31, 2013, the United States Nuclear Regulatory Commission (NRC) completed an inspection at your Surry Power Station Units 1 and 2. The enclosed inspection report documents the inspection findings which were discussed on January 29, 2014, with Mr. R.


This report documents two findings of very low sa fety significance (Green), both of which were determined to be violations of NRC requirements.
Simmons and other members of your staff. Inspectors documented the results of this inspection in the enclosed inspection report.


Additionally, one licensee-identified violation which was determined to be of very low safety significance is listed in this report. However, because of the very low safety significance of these issues and because they were entered into  
This report documents two findings of very low safety significance (Green), both of which were determined to be violations of NRC requirements. Additionally, one licensee-identified violation which was determined to be of very low safety significance is listed in this report. However, because of the very low safety significance of these issues and because they were entered into your corrective action program, the NRC is treating these as non-cited violations (NCV)
 
consistent with Section 2.3.2.a of the NRC Enforcement Policy.
your corrective action program, the NRC is tr eating these as non-cited violations (NCV) consistent with Section 2.3.2.a of the NRC Enforcement Policy.


If you contest the violations or significance of these NCVs, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the Nuclear Regulatory Commission, ATTN.: Document Control Desk, Washington DC 20555-0001; with copies to the Regional Administrator Region II; the Director, Office of Enforcement, United States Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC Resident Inspector at the Surry Power Station.
If you contest the violations or significance of these NCVs, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the Nuclear Regulatory Commission, ATTN.: Document Control Desk, Washington DC 20555-0001; with copies to the Regional Administrator Region II; the Director, Office of Enforcement, United States Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC Resident Inspector at the Surry Power Station.
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Additionally, if you disagree with a cross-cutting aspect assignment in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your disagreement, to the Regional Administrator, Region II, and the NRC Resident Inspector at the Surry Power Station. As a result of the Safey Culture Common Language Initiative, the terminology and coding of cross-cutting aspects were revised beginning in calendar year (CY) 2014. New cross-cutting aspects identified in CY 2014 will be coded under the latest revision to IMC 0310. Cross-cutting aspects identified in the last six months of 2013 using the previous terminology will be converted to the latest revision in accordance with the cross-cutting themes and potential substantive cross-cutting issues in accordance with IMC 0305 starting with the CY 2014 mid-cycle assessment review.
Additionally, if you disagree with a cross-cutting aspect assignment in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your disagreement, to the Regional Administrator, Region II, and the NRC Resident Inspector at the Surry Power Station. As a result of the Safey Culture Common Language Initiative, the terminology and coding of cross-cutting aspects were revised beginning in calendar year (CY) 2014. New cross-cutting aspects identified in CY 2014 will be coded under the latest revision to IMC 0310. Cross-cutting aspects identified in the last six months of 2013 using the previous terminology will be converted to the latest revision in accordance with the cross-cutting themes and potential substantive cross-cutting issues in accordance with IMC 0305 starting with the CY 2014 mid-cycle assessment review.


In accordance with Title 10 of the Code of Federal Regulations 2.390, "Public Inspections, Exemptions, Requests for Withholding," of the NRC's "Rules of Practice," a copy of this letter, its enclosure, and your response (if any) will be available electronically for public inspection in the NRC's Public Document Room or from the Publicly Available Records (PARS) component of the NRC's Agencywide Documents Access and Managem ent System (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
In accordance with Title 10 of the Code of Federal Regulations 2.390, Public Inspections, Exemptions, Requests for Withholding, of the NRCs Rules of Practice, a copy of this letter, its enclosure, and your response (if any) will be available electronically for public inspection in the NRCs Public Document Room or from the Publicly Available Records (PARS) component of the NRCs Agencywide Documents Access and Management System (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).


Sincerely,
Sincerely,
/RA/
/RA/
Michael F. King, Chief Reactor Projects Branch 5 Division of Reactor Projects  
Michael F. King, Chief Reactor Projects Branch 5 Division of Reactor Projects Docket Nos.: 50-280, 50-281 License Nos.: DPR-32, DPR-37
 
Docket Nos.: 50-280, 50-281 License Nos.: DPR-32, DPR-37  


===Enclosure:===
===Enclosure:===
Inspection Report 05000280/2013005, 05000281/2013005 w/Attachment: Supplemental Information  
Inspection Report 05000280/2013005, 05000281/2013005 w/Attachment: Supplemental Information


REGION II==
REGION II==
 
Docket Nos.: 50-280, 50-281 License Nos.: DPR-32, DPR-37 Report No: 05000280/2013005, 05000281/2013005 Licensee: Virginia Electric and Power Company (VEPCO)
Docket Nos.: 50-280, 50-281 License Nos.: DPR-32, DPR-37  
Facility: Surry Power Station, Units 1 and 2 Location: 5850 Hog Island Road Surry, VA 23883 Dates: October 1, 2013 through December 31, 2013 Inspectors: P. McKenna, Senior Resident Inspector J. Nadel, Resident Inspector D. Bacon, Senior Operations Engineer (Section 1R11)
 
Report No: 05000280/2013005, 05000281/2013005 Licensee: Virginia Electric and Power Company (VEPCO)  
 
Facility: Surry Power Station, Units 1 and 2  
 
Location: 5850 Hog Island Road Surry, VA 23883  
 
Dates: October 1, 2013 through December 31, 2013  
 
Inspectors: P. McKenna, Senior Resident Inspector J. Nadel, Resident Inspector D. Bacon, Senior Operations Engineer (Section 1R11)
A. Butcavage, Reactor Inspector (Section 1R08)
A. Butcavage, Reactor Inspector (Section 1R08)
R. Carrion, Senior Reactor Inspector (Section 1R08)
R. Carrion, Senior Reactor Inspector (Section 1R08)
R. Hamilton, Senior Health Physicist (Section 2RS8) R. Kellner, Health Physicist (Sections 2RS1, 40A1) J. Laughlin, Emergency Preparedness Inspector (Section 1EP4) J. Rivera-Ortiz, Senior Reactor Inspector (Section 1R08)
R. Hamilton, Senior Health Physicist (Section 2RS8)
K. Roche, Reactor Operations Engineer  
R. Kellner, Health Physicist (Sections 2RS1, 40A1)
 
J. Laughlin, Emergency Preparedness Inspector (Section 1EP4)
Approved by: Michael F. King, Chief Reactor Projects Branch 5 Division of Reactor Projects  
J. Rivera-Ortiz, Senior Reactor Inspector (Section 1R08)
 
K. Roche, Reactor Operations Engineer Approved by: Michael F. King, Chief Reactor Projects Branch 5 Division of Reactor Projects Enclosure
Enclosure  


=SUMMARY OF FINDINGS=
=SUMMARY OF FINDINGS=
IR 05000280/2013005, 05000281/2013005; 10/01/2013-12/31/2013; Surry Power Station, Units 1 and 2: Plant Modifications; Radioactive Solid Waste Processing and Radioactive Material  
IR 05000280/2013005, 05000281/2013005; 10/01/2013-12/31/2013; Surry Power Station, Units and 2: Plant Modifications; Radioactive Solid Waste Processing and Radioactive Material Handling, Storage, and Transportation The report covered a three month period of inspection by resident inspectors and region based inspectors. One NRC identified and one self-revealing findings were identified and both were determined to be a non-cited violations (NCVs). The significance of most findings is indicated by their color (Green, White, Yellow, Red) using Inspection Manual Chapter (IMC) 0609,
 
Significance Determination Process (SDP). The cross-cutting aspect was determined using IMC 0310, Components Within The Cross-Cutting Areas. Findings for which the SDP does not apply may be Green or be assigned a severity level after NRC management review. The NRCs program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process Revision 4, dated December 2006.
Handling, Storage, and Transportation  
 
The report covered a three month period of inspection by resident inspectors and region based inspectors. One NRC identified and one self-revealing findings were identified and both were determined to be a non-cited violations (NCVs). The significance of most findings is indicated by their color (Green, White, Yellow, Red) using Inspection Manual Chapter (IMC) 0609,  
"Significance Determination Process" (SDP). The cross-cutting aspect was determined using IMC 0310, "Components Within The Cross-Cutting Areas."  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.


===Cornerstone: Barrier Integrity===
===Cornerstone: Barrier Integrity===
: '''Green.'''
: '''Green.'''
An NRC-identified Green NCV of 10 CFR 50, Appendix B, Criterion III, "Design Control," was identified for the licensee's failure to adequately protect safety-related low head safety injection system (LHSI) piping from a tornado missile. Specifically, the licensee installed a mechanical piping connection to the Unit 1 LHSI system as part of a design change (DC) and did not provide tornado missile protection for the portions of piping that were installed above the 28.6 feet elevation in the safeguards valve pit and connected directly to the LHSI piping. The issue was documented in the licensee's corrective action program (CAP) as condition report (CR) 533401.
An NRC-identified Green NCV of 10 CFR 50, Appendix B, Criterion III, Design Control, was identified for the licensees failure to adequately protect safety-related low head safety injection system (LHSI) piping from a tornado missile. Specifically, the licensee installed a mechanical piping connection to the Unit 1 LHSI system as part of a design change (DC) and did not provide tornado missile protection for the portions of piping that were installed above the 28.6 feet elevation in the safeguards valve pit and connected directly to the LHSI piping. The issue was documented in the licensees corrective action program (CAP) as condition report (CR) 533401.


The licensee's failure to protect the Unit 1 LHSI system piping against external missile hazards when the piping was modified by the diverse and flexible coping strategies (FLEX) mechanical piping connection as part of DC SU-12-00022 was a performance deficiency (PD) that was within the licensee's ability to foresee and correct.
The licensees failure to protect the Unit 1 LHSI system piping against external missile hazards when the piping was modified by the diverse and flexible coping strategies (FLEX) mechanical piping connection as part of DC SU-12-00022 was a performance deficiency (PD) that was within the licensees ability to foresee and correct.


The inspectors determined that the performance deficiency (PD) was more than minor because it was associated with the design control attribute of the Barrier Integrity Cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. Specifically, the welded piping connection installed by DC SU-12-00022 located between containment and containment isolation valve 1-SI-MOV-1890A, "A" LHSI hot leg discharge isolation valve, was susceptible to failure from the impact of a tornado generated missile. Using Manual Chapter 0609.04, Initial Characterization of Findings, Table 2, dated June 19, 2012, the finding was determined to affect the reactor containment barrier and the Barrier Integrity Cornerstone. The inspectors screened the finding using Manual Chapter 0609 Appendix A, "Significance Determination Process (SDP) for Findings at-Power", and determined the finding was of very low safety significance (Green) because the finding did not represent an actual open pathway in the physical integrity of reactor containment. This finding has a cross-cutting aspect in the work control component of the human performance area, H.3(b); because the licensee failed to address both the impact of changes in the work scope on the plant and to use adequate interdepartmental coordination during the design process. (Section 1R18)    
The inspectors determined that the performance deficiency (PD) was more than minor because it was associated with the design control attribute of the Barrier Integrity Cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. Specifically, the welded piping connection installed by DC SU-12-00022 located between containment and containment isolation valve 1-SI-MOV-1890A, A LHSI hot leg discharge isolation valve, was susceptible to failure from the impact of a tornado generated missile. Using Manual Chapter 0609.04, Initial Characterization of Findings,
Table 2, dated June 19, 2012, the finding was determined to affect the reactor containment barrier and the Barrier Integrity Cornerstone. The inspectors screened the finding using Manual Chapter 0609 Appendix A, Significance Determination Process (SDP) for Findings at-Power, and determined the finding was of very low safety significance (Green) because the finding did not represent an actual open pathway in the physical integrity of reactor containment. This finding has a cross-cutting aspect in the work control component of the human performance area, H.3(b); because the licensee failed to address both the impact of changes in the work scope on the plant and to use adequate interdepartmental coordination during the design process. (Section 1R18)


===Cornerstone: Public Radiation Safety===
===Cornerstone: Public Radiation Safety===
: '''Green.'''
: '''Green.'''
A self-revealing non-cited violation of 10 CFR 20.1802, "Control of Material not in Storage", was identified for the licensee's failure to maintain control and constant surveillance of licensed radioactive material in a controlled or unrestricted area (Health Physics (HP) technical services area of the administration building) that was not in storage. The material that was initially unaccounted for was an Americium-241 check source with an activity of 0.02 micro-Curies, used to perform routine function checks on iSolo alpha/beta counter. The issued was documented in the licensee's corrective action program (CAP) as condition report (CR) 523692.
A self-revealing non-cited violation of 10 CFR 20.1802, Control of Material not in Storage, was identified for the licensees failure to maintain control and constant surveillance of licensed radioactive material in a controlled or unrestricted area (Health Physics (HP) technical services area of the administration building) that was not in storage. The material that was initially unaccounted for was an Americium-241 check source with an activity of 0.02 micro-Curies, used to perform routine function checks on iSolo alpha/beta counter. The issued was documented in the licensees corrective action program (CAP) as condition report (CR) 523692.


The licensee's failure to control and maintain constant surveillance of licensed material that is in a controlled or unrestricted area and that is not in storage was a performance deficiency (PD). The PD was more than minor because it was associated with the Program and Process attribute of the Public Radiation Safety Cornerstone and affected the cornerstone objective of ensuring adequate protection of public health and safety from exposure to radioactive materials released into the public domain. Using Manual Chapter 0609, Appendix D, "Public Radiation Safety SDP," this finding determined to be was of very low safety significance (Green) in that the public radiation exposure was not greater than 0.005 rem (5 millirem). The inspectors determined that cross-cutting issue H.4(b), "The licensee defines and effectively communicates expectations regarding procedural compliance and personnel follow procedures," was applicable for this violation because the radiation protection (RP) technician had failed to follow procedure HP-1033.148, "Canberra iSolo: Performance Checks", step 6.4.3 which states: "Ensure check source is removed from Canberra iSolo and returned to designated storage location". (Section 2RS8)
The licensees failure to control and maintain constant surveillance of licensed material that is in a controlled or unrestricted area and that is not in storage was a performance deficiency (PD). The PD was more than minor because it was associated with the Program and Process attribute of the Public Radiation Safety Cornerstone and affected the cornerstone objective of ensuring adequate protection of public health and safety from exposure to radioactive materials released into the public domain. Using Manual Chapter 0609, Appendix D, Public Radiation Safety SDP, this finding determined to be was of very low safety significance (Green) in that the public radiation exposure was not greater than 0.005 rem (5 millirem). The inspectors determined that cross-cutting issue H.4(b), The licensee defines and effectively communicates expectations regarding procedural compliance and personnel follow procedures, was applicable for this violation because the radiation protection (RP) technician had failed to follow procedure HP-1033.148, Canberra iSolo: Performance Checks, step 6.4.3 which states: Ensure check source is removed from Canberra iSolo and returned to designated storage location. (Section 2RS8)
One violation of very low safety significance, which was identified by the licensee, was reviewed by the inspectors. Corrective actions taken or planned by the licensee have been entered into the licensee's corrective action program. This violation and the respective corrective actions are listed in Section 4OA7 of this report.
One violation of very low safety significance, which was identified by the licensee, was reviewed by the inspectors. Corrective actions taken or planned by the licensee have been entered into the licensees corrective action program. This violation and the respective corrective actions are listed in Section 4OA7 of this report.


=REPORT DETAILS=
=REPORT DETAILS=
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==REACTOR SAFETY==
==REACTOR SAFETY==
Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity  
Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity


{{a|1R01}}
{{a|1R01}}
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====a. Inspection Scope====
====a. Inspection Scope====
The inspectors reviewed the licensee's preparations for seasonal cold weather.
The inspectors reviewed the licensees preparations for seasonal cold weather.
 
Inspection focused on verification of design features and implementation of the licensees procedure for cold weather conditions, 0-OSP-ZZ-001, Cold Weather Preparation, Revision 14. The inspectors walked down key structures including the turbine and auxiliary buildings, safeguards buildings, the emergency switchgear rooms, and emergency battery rooms and verified HVAC systems were operating properly and that area temperatures remained within design requirements specified in the UFSAR.


Inspection focused on verification of design features and implementation of the licensee's procedure for cold weather conditions, 0-OSP-ZZ-001, "Cold Weather Preparation," Revision 14. The inspectors walked down key structures including the turbine and auxiliary buildings, safeguards buildings, the emergency switchgear rooms, and emergency battery rooms and verified HVAC systems were operating properly and that area temperatures remained within design requirements specified in the UFSAR. The mitigating systems reviewed during this inspection include: the auxiliary feedwater systems, the refueling water storage tanks, emergency diesel generators, alternate alternating current (AAC) diesel generator, and emergency switchgear.
The mitigating systems reviewed during this inspection include: the auxiliary feedwater systems, the refueling water storage tanks, emergency diesel generators, alternate alternating current (AAC) diesel generator, and emergency switchgear.


====b. Findings====
====b. Findings====
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====a. Inspection Scope====
====a. Inspection Scope====
The inspectors performed a detailed walkdown and inspection of the Unit 1 containment spray system to verify the system was properly aligned and capable of performing its safety function, and to assess its materi al condition. During the walkdown, the inspectors verified breaker positions were in the proper alignment, component labeling was accurate, hangers and supports were functional, and local indications were accurate. Recent testing history was also reviewed to verify that standby components were performing within their design. The plant health report, system drawings, condition reports, the UFSAR, and TS were reviewed and outstanding deficiencies were verified to be properly classified and not affect system operability and capability to perform its  
The inspectors performed a detailed walkdown and inspection of the Unit 1 containment spray system to verify the system was properly aligned and capable of performing its safety function, and to assess its material condition. During the walkdown, the inspectors verified breaker positions were in the proper alignment, component labeling was accurate, hangers and supports were functional, and local indications were accurate. Recent testing history was also reviewed to verify that standby components were performing within their design. The plant health report, system drawings, condition reports, the UFSAR, and TS were reviewed and outstanding deficiencies were verified to be properly classified and not affect system operability and capability to perform its safety function. The inspectors reviewed the corrective action program to verify equipment alignment issues were being identified and resolved.
 
safety function. The inspectors reviewed the corrective action program to verify equipment alignment issues were being identified and resolved.


====b. Findings====
====b. Findings====
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====a. Inspection Scope====
====a. Inspection Scope====
The inspectors conducted tours of the areas listed below that is important to reactor safety to verify the licensee's implementation of fire protection requirements as described in fleet procedures CM-AA-FPA-100, "Fire Protection/Appendix R (Fire Safe Shutdown) Program," Revision 8, CM-AA-FPA-101, "Control of Combustible and Flammable Materials," Revision 6, and CM-AA-FPA-102, "Fire Protection and Fire Safe Shutdown Review and Preparation Process and Design Change Process," Revision 6. The reviews were performed to evaluate the fire protection program operational status and material condition and the adequacy of: 1) control of transient combustibles and ignition sources; 2) fire detection and suppression capability; 3) passive fire protection features; 4) compensatory measures established for out-of-service, degraded or inoperable fire protection equipment, systems, or features; and 5) procedures, equipment, fire barriers, and systems so that post-fire capability to safely shutdown the plant is ensured. The inspectors reviewed the corrective action program to verify fire protection deficiencies were being identified and properly resolved.
The inspectors conducted tours of the areas listed below that is important to reactor safety to verify the licensees implementation of fire protection requirements as described in fleet procedures CM-AA-FPA-100, Fire Protection/Appendix R (Fire Safe Shutdown) Program, Revision 8, CM-AA-FPA-101, Control of Combustible and Flammable Materials, Revision 6, and CM-AA-FPA-102, Fire Protection and Fire Safe Shutdown Review and Preparation Process and Design Change Process, Revision 6.
 
The reviews were performed to evaluate the fire protection program operational status and material condition and the adequacy of: 1) control of transient combustibles and ignition sources; 2) fire detection and suppression capability; 3) passive fire protection features; 4) compensatory measures established for out-of-service, degraded or inoperable fire protection equipment, systems, or features; and 5) procedures, equipment, fire barriers, and systems so that post-fire capability to safely shutdown the plant is ensured. The inspectors reviewed the corrective action program to verify fire protection deficiencies were being identified and properly resolved.
* Unit 1 Containment
* Unit 1 Containment
* 27.6 foot level of the Auxiliary Building
* 27.6 foot level of the Auxiliary Building
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====a. Inspection Scope====
====a. Inspection Scope====
The inspectors reviewed the internal flood protection measures and procedural controls established to address potential flooding in the Unit 1 and 2 turbine buildings, the emergency switchgear rooms, and mechanical equipment rooms 3 and 4 during the operation of temporary modification SU-11-00017, "CCHX SW Inlet Pipe Repair & 1-SW-MOV-102A/B Replacements" which installed temporary service water (SW) piping in the Unit 1 turbine building. The inspectors conducted a walkdown of the affected areas to observe and assess the condition of the installed flood dikes, floor drain backflow preventers, the sealing of holes and penetrations between flood areas, the adequacy of water tight doors, the operability of flooding alarms, and the installed sump pumps.
The inspectors reviewed the internal flood protection measures and procedural controls established to address potential flooding in the Unit 1 and 2 turbine buildings, the emergency switchgear rooms, and mechanical equipment rooms 3 and 4 during the operation of temporary modification SU-11-00017, CCHX SW Inlet Pipe Repair & 1-SW-MOV-102A/B Replacements which installed temporary service water (SW) piping in the Unit 1 turbine building. The inspectors conducted a walkdown of the affected areas to observe and assess the condition of the installed flood dikes, floor drain backflow preventers, the sealing of holes and penetrations between flood areas, the adequacy of water tight doors, the operability of flooding alarms, and the installed sump pumps.


Additionally, the inspectors verified that the required actions of the safety evaluation related to license amendment number 279 were being accomplished by the licensee.
Additionally, the inspectors verified that the required actions of the safety evaluation related to license amendment number 279 were being accomplished by the licensee.
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====a. Inspection Scope====
====a. Inspection Scope====
The inspectors reviewed the licensee's heat exchanger program document, 0-MCM-0812-01, "Component Cooling Heat Exchanger (CCHX) Inspection and Cleaning,"
The inspectors reviewed the licensees heat exchanger program document, 0-MCM-0812-01, Component Cooling Heat Exchanger (CCHX) Inspection and Cleaning, Revision 18, trending data maintained by the system engineer, maintenance rule information, specific commitments, and design basis information. Specific focus was placed on the operation and surveillance testing of the component cooling heat exchangers during the installation and use of the temporary service water jumper piping during Unit 1 Fall 2013 refueling outage. The inspectors reviewed licensee performance of surveillance procedure 1-NSP-CC-005, Component Cooling Heat Exchanger Tests Using the Temporary Monitoring System, Revision 1, which is a test designed to measure the capability of the component cooling heat exchangers while they are being fed from the temporary service water jumper piping and the normal supply piping was out of service. The inspectors reviewed testing procedures and test results to confirm that the component cooling heat exchangers were still able to perform their functions and that planned corrective actions were appropriate. The inspectors verified that significant heat exchanger performance issues were being entered into the licensees CAP and appropriately addressed.
Revision 18, trending data maintained by the system engineer, maintenance rule information, specific commitments, and design basis information. Specific focus was placed on the operation and surveillance testing of the component cooling heat exchangers during the installation and use of the temporary service water jumper piping during Unit 1 Fall 2013 refueling outage. The inspectors reviewed licensee performance of surveillance procedure 1-NSP-CC-005, "Component Cooling Heat Exchanger Tests Using the Temporary Monitoring System," Revision 1, which is a test designed to measure the capability of the component cooling heat exchangers while they are being fed from the temporary service water jumper piping and the normal supply piping was out of service. The inspectors reviewed testing procedures and test results to confirm that the component cooling heat exchangers were still able to perform their functions and that planned corrective actions were appropriate. The inspectors verified that significant heat exchanger performance issues were being entered into the licensee's CAP and appropriately addressed.


====b. Findings====
====b. Findings====
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====a. Inspection Scope====
====a. Inspection Scope====
From October 28, 2013, through November 1, 2013, the inspectors conducted an on-site review of the implementation of the licensee's Inservice Inspection (ISI) program for monitoring degradation of the reactor coolant system, steam generator (SG) tubes, emergency feedwater systems, risk-significant piping and components and containment systems in Unit 1.
From October 28, 2013, through November 1, 2013, the inspectors conducted an on-site review of the implementation of the licensees Inservice Inspection (ISI) program for monitoring degradation of the reactor coolant system, steam generator (SG) tubes, emergency feedwater systems, risk-significant piping and components and containment systems in Unit 1.


Non-Destructive Examination (NDE) Activities and Welding Activities: The inspectors reviewed associated records and directly observed the following non-destructive examinations required by the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPVC) to verify compliance with Section XI and Section V of the ASME BPVC of record for Surry Power Station Unit 1 (1998 Edition with 2000 Addenda). The inspectors also verified that any relevant indications and defects were  
Non-Destructive Examination (NDE) Activities and Welding Activities: The inspectors reviewed associated records and directly observed the following non-destructive examinations required by the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPVC) to verify compliance with Section XI and Section V of the ASME BPVC of record for Surry Power Station Unit 1 (1998 Edition with 2000 Addenda). The inspectors also verified that any relevant indications and defects were dispositioned in accordance with the requirements of the ASME BPVC or an NRC-approved alternative requirement.
 
dispositioned in accordance with the requirements of the ASME BPVC or an NRC-approved alternative requirement.
* Liquid penetrant (PT) examination of reinforcing plate weld number 11448-WMKS-RH-E-1A/1-RH-E-1A/1-A06, 1A residual heat removal (RHR) inlet side heat exchanger
* Liquid penetrant (PT) examination of reinforcing plate weld number 11448-WMKS-RH-E-1A/1-RH-E-1A/1-A06, 1A residual heat removal (RHR) inlet side heat exchanger
* Ultrasonic testing (UT) examination of elbow-to-pipe weld number 11448-WMKS-1105B9/3-CH-71/1-BS, 3-inch diameter charging line of charging pump 1-C
* Ultrasonic testing (UT) examination of elbow-to-pipe weld number 11448-WMKS-1105B9/3-CH-71/1-BS, 3-inch diameter charging line of charging pump 1-C
* Visual testing (VT-3) examination of pipe support 11448-MKS-1105B9 The inspectors also reviewed a sample of other augmented or industry initiative examinations performed during the Fall 2013 refueling outage. The inspectors observed a portion of an augmented, enhanced visual testing (EVT-1), associated with the industry's Materials Reliability Program (MRP)-227-A, "Pressurized Water Reactor Internals Inspection and Evaluation Guidelines."  Specifically, the inspectors in conjunction with an NDE Level III qualified inspector, reviewed video recordings of the core barrel upper girth weld visual inspection of the inside diameter of the weld between the 180° and 270° azimuth locations to determine if the examination and disposition of indications were consistent with the MRP-227-A guidelines. The inspectors interviewed plant personnel to verify that the current scope of the Reactor Internals MRP-227 inspection activities included all the primary component inspections listed in the MRP-227 tables for a Westinghouse designed unit.
* Visual testing (VT-3) examination of pipe support 11448-MKS-1105B9 The inspectors also reviewed a sample of other augmented or industry initiative examinations performed during the Fall 2013 refueling outage. The inspectors observed a portion of an augmented, enhanced visual testing (EVT-1), associated with the industrys Materials Reliability Program (MRP)-227-A, Pressurized Water Reactor Internals Inspection and Evaluation Guidelines. Specifically, the inspectors in conjunction with an NDE Level III qualified inspector, reviewed video recordings of the core barrel upper girth weld visual inspection of the inside diameter of the weld between the 180° and 270° azimuth locations to determine if the examination and disposition of indications were consistent with the MRP-227-A guidelines. The inspectors interviewed plant personnel to verify that the current scope of the Reactor Internals MRP-227 inspection activities included all the primary component inspections listed in the MRP-227 tables for a Westinghouse designed unit.


During non-destructive surface and volumetric examinations performed since the previous RFO, the licensee did not identify any recordable indications that were accepted for continued service through analytical methods. Therefore, no NRC review was completed for this attribute of the inspection procedure.
During non-destructive surface and volumetric examinations performed since the previous RFO, the licensee did not identify any recordable indications that were accepted for continued service through analytical methods. Therefore, no NRC review was completed for this attribute of the inspection procedure.


The inspectors reviewed the following pressure boundary welds completed for risk-significant systems since the last Unit 1 RFO to evaluate if the licensee applied the pre-service non-destructive examinations and acceptance criteria required by the construction code and ASME BPVC Section XI. In addition, the inspectors reviewed the work orders (WO), welding procedure specifications, welder qualifications, welding material certification and supporting weld procedure qualification records, to evaluate if the weld procedures were qualified in accordance with the requirements of the construction code and ASME BPVC, Section IX.
The inspectors reviewed the following pressure boundary welds completed for risk-significant systems since the last Unit 1 RFO to evaluate if the licensee applied the pre-service non-destructive examinations and acceptance criteria required by the construction code and ASME BPVC Section XI. In addition, the inspectors reviewed the work orders (WO), welding procedure specifications, welder qualifications, welding material certification and supporting weld procedure qualification records, to evaluate if the weld procedures were qualified in accordance with the requirements of the construction code and ASME BPVC, Section IX.
* WO 38103090652, weld number 42-WS-13-10, 42-inch diameter service water system line into the component c ooling water heat exchanger (CCHX)
* WO 38103090652, weld number 42-WS-13-10, 42-inch diameter service water system line into the component cooling water heat exchanger (CCHX)
* WO 38103279050, 1-1/2-inch diameter CC-346-151, weld number 137 (Sockolet to pipe) Pressure Vessel Upper Head Penetration Inspection Activities: For the Unit 1 reactor pressure vessel upper head, no examination was required pursuant to 10 CFR 50.55a(g)(6)(ii)(D) for the Fall 2013 RFO. Therefore, no NRC review was completed for this inspection procedure attribute.
* WO 38103279050, 1-1/2-inch diameter CC-346-151, weld number 137 (Sockolet to pipe)
Pressure Vessel Upper Head Penetration Inspection Activities: For the Unit 1 reactor pressure vessel upper head, no examination was required pursuant to 10 CFR 50.55a(g)(6)(ii)(D) for the Fall 2013 RFO. Therefore, no NRC review was completed for this inspection procedure attribute.


Boric Acid Corrosion Control (BACC): The inspectors reviewed the licensee's BACC program activities to ensure implementation with commitments made in response to NRC Generic Letter 88-05, "Boric Acid Corrosion of Carbon Steel Reactor Pressure Boundary," and applicable industry guidance documents. Specifically, the inspectors performed an on-site record review of procedures and the results of the licensee's containment walk-down inspections performed during the Fall 2013 RFO (S1R25). The inspectors also interviewed the BACC program owner, conducted an independent walk-
Boric Acid Corrosion Control (BACC): The inspectors reviewed the licensees BACC program activities to ensure implementation with commitments made in response to NRC Generic Letter 88-05, Boric Acid Corrosion of Carbon Steel Reactor Pressure Boundary, and applicable industry guidance documents. Specifically, the inspectors performed an on-site record review of procedures and the results of the licensees containment walk-down inspections performed during the Fall 2013 RFO (S1R25). The inspectors also interviewed the BACC program owner, conducted an independent walk-down of portions of the containment to evaluate compliance with the licensees BACC program requirements, and verified that degraded or non-conforming conditions, such as boric acid leaks, were properly identified, evaluated, and corrected in accordance with the licensees BACC and corrective action programs, and were consistent with the requirements of ASME BPVC Section XI and 10 CFR Part 50, Appendix B, Criterion XVI.
 
down of portions of the containment to ev aluate compliance with the licensee's BACC program requirements, and verified that degraded or non-conforming conditions, such as boric acid leaks, were properly identified, evaluated, and corrected in accordance with the licensee's BACC and corrective action programs, and were consistent with the requirements of ASME BPVC Section XI and 10 CFR Part 50, Appendix B, Criterion XVI.


Specifically the inspectors reviewed the results and evaluations associated with the following boric acid indications:
Specifically the inspectors reviewed the results and evaluations associated with the following boric acid indications:
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* CR-529802, Boric Acid Identified on 1-RC-ICV-3026
* CR-529802, Boric Acid Identified on 1-RC-ICV-3026
* CR-529813, Boric Acid Identified on 1-RC-ICV-3085
* CR-529813, Boric Acid Identified on 1-RC-ICV-3085
* CR-529937, Boric Acid Identified on 1-RC-ICV-3142 Steam Generator (SG) Tube Inspection Activities. The inspectors reviewed the eddy current examination activities performed in Unit 1 SG "B" during RFO S1R25 to verify compliance with the licensee's TS, ASME BPVC Section XI, and Nuclear Energy Institute (NEI) 97-06, "Steam Generator Program Guidelines."  The inspectors interviewed licensee personnel and vendor staff responsible for the SG inspection project and reviewed documentation associated with the SG inspections and integrity assessments as described below.
* CR-529937, Boric Acid Identified on 1-RC-ICV-3142 Steam Generator (SG) Tube Inspection Activities. The inspectors reviewed the eddy current examination activities performed in Unit 1 SG B during RFO S1R25 to verify compliance with the licensees TS, ASME BPVC Section XI, and Nuclear Energy Institute (NEI) 97-06, Steam Generator Program Guidelines. The inspectors interviewed licensee personnel and vendor staff responsible for the SG inspection project and reviewed documentation associated with the SG inspections and integrity assessments as described below.


The inspectors reviewed the scope of the eddy current examinations to verify that known and potential areas of tube degradation were inspected. The inspectors also verified that inspection scope expansion criteria were implemented based on inspection results as directed by the "Electric Power Research Institute (EPRI) Pressurized Water Reactor Steam Generator Examination Guidelines", Revision 7. The inspectors selected a sample of degradation mechanisms from the "Unit 1 Steam Generator Degradation Assessment" and verified that the in-situ pressure testing criteria were determined in accordance with the EPRI Guidelines. Additionally, the inspectors reviewed eddy current indication reports to determine whether tubes with relevant indications were appropriately screened for in-situ pressure testing. The inspectors' review included the implementation of tube repair criteria and planned repair methods to verify they were consistent with plant TS and industry guidelines.
The inspectors reviewed the scope of the eddy current examinations to verify that known and potential areas of tube degradation were inspected. The inspectors also verified that inspection scope expansion criteria were implemented based on inspection results as directed by the Electric Power Research Institute (EPRI) Pressurized Water Reactor Steam Generator Examination Guidelines, Revision 7. The inspectors selected a sample of degradation mechanisms from the Unit 1 Steam Generator Degradation Assessment and verified that the in-situ pressure testing criteria were determined in accordance with the EPRI Guidelines. Additionally, the inspectors reviewed eddy current indication reports to determine whether tubes with relevant indications were appropriately screened for in-situ pressure testing. The inspectors review included the implementation of tube repair criteria and planned repair methods to verify they were consistent with plant TS and industry guidelines.


The inspectors compared the recent eddy current examination results with the last "Condition Monitoring and Operational Assessment Report" for SG "B" to assess the licensee's prediction capability for maximum tube degradation and number of tubes with indications. The inspectors verified that the licensee's evaluation was conservative and that current examination results were bound by the operational assessment projections. Since the licensee performed eddy current tube examinations only in SG "B" during the Fall 2013 RFO, the inspectors reviewed the last operational assessment for SGs "A" and "C" to verify the licensee met the inspection frequency established in the plant TS and had evaluated the acceptability of these SGs to meet the tube integrity performance criteria until the next scheduled inspection.
The inspectors compared the recent eddy current examination results with the last Condition Monitoring and Operational Assessment Report for SG B to assess the licensees prediction capability for maximum tube degradation and number of tubes with indications. The inspectors verified that the licensees evaluation was conservative and that current examination results were bound by the operational assessment projections.
 
Since the licensee performed eddy current tube examinations only in SG B during the Fall 2013 RFO, the inspectors reviewed the last operational assessment for SGs A and C to verify the licensee met the inspection frequency established in the plant TS and had evaluated the acceptability of these SGs to meet the tube integrity performance criteria until the next scheduled inspection.


The inspectors also compared past examination results discussed in the latest degradation assessment with the recent eddy current examination results to verify that new degradation mechanisms were identified and evaluated before plant startup. The review of eddy current examination results included the disposition of potential loose part indications on the SG secondary side to verify that corrective actions for evaluating and retrieving loose parts were consistent with the EPRI Guidelines. The inspectors also reviewed a sample of primary-to-secondary leakage data for the last Unit 1 operating cycle to obtain reasonable assurance that operational leakage in all three SGs remained below the detection or action level thresholds during the previous operating cycle.
The inspectors also compared past examination results discussed in the latest degradation assessment with the recent eddy current examination results to verify that new degradation mechanisms were identified and evaluated before plant startup. The review of eddy current examination results included the disposition of potential loose part indications on the SG secondary side to verify that corrective actions for evaluating and retrieving loose parts were consistent with the EPRI Guidelines. The inspectors also reviewed a sample of primary-to-secondary leakage data for the last Unit 1 operating cycle to obtain reasonable assurance that operational leakage in all three SGs remained below the detection or action level thresholds during the previous operating cycle.
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In addition, the inspectors reviewed documentation for a sample of eddy current data analysts, eddy current probes, and eddy current testers to verify these were qualified to detect the existing and potential degradation mechanisms applicable to Surry SG tubes.
In addition, the inspectors reviewed documentation for a sample of eddy current data analysts, eddy current probes, and eddy current testers to verify these were qualified to detect the existing and potential degradation mechanisms applicable to Surry SG tubes.


This review included a sample of site-specific examination technique specification sheets (ETSSs) to ensure that their qualification and site-specific implementation were consistent with Appendix H or I of the "EPRI Pressurized Water Reactor Steam Generator Examination Guidelines," Revision 7. The selected ETSSs for review were based on degradation mechanisms of interest to the inspectors based on plant-specific and industry operating experience. The inspectors selected bobbin probe ETSSs qualified for detection and sizing of loose part wear in tube freespan locations, thinning in tube support plates and top-of-tubesheet, and wear at tube support plates and anti-vibration bars. The inspectors also selected rotating and array probe ETSSs for detection of circumferential and axial stress corrosion cracking on the internal and external surfaces of the tubes at the expansion transition area. The inspectors also reviewed a sample of eddy current data with a qualified data analyst to confirm that data analysis was performed in accordance with the ETSSs and site-specific analysis guidelines. The sample of eddy current data selected for review was for the following tubes in SG "B": R34C58 (Bobbin Probe), R40C25 (Bobbin Probe), R40C50 (Array-Probe), and R45C48 (Array Probe).
This review included a sample of site-specific examination technique specification sheets (ETSSs) to ensure that their qualification and site-specific implementation were consistent with Appendix H or I of the EPRI Pressurized Water Reactor Steam Generator Examination Guidelines, Revision 7. The selected ETSSs for review were based on degradation mechanisms of interest to the inspectors based on plant-specific and industry operating experience. The inspectors selected bobbin probe ETSSs qualified for detection and sizing of loose part wear in tube freespan locations, thinning in tube support plates and top-of-tubesheet, and wear at tube support plates and anti-vibration bars. The inspectors also selected rotating and array probe ETSSs for detection of circumferential and axial stress corrosion cracking on the internal and external surfaces of the tubes at the expansion transition area. The inspectors also reviewed a sample of eddy current data with a qualified data analyst to confirm that data analysis was performed in accordance with the ETSSs and site-specific analysis guidelines. The sample of eddy current data selected for review was for the following tubes in SG B: R34C58 (Bobbin Probe), R40C25 (Bobbin Probe), R40C50 (Array-Probe), and R45C48 (Array Probe).


Based on the review of eddy current examination results for SG "B" and interviews with the licensee, the inspectors confirmed that no new degradation mechanisms were identified, no eddy current scope expansion was required, none of the SG tubes examined met the criteria for in-situ pressure testing, and none of the indications left in-service required repair. Furthermore, the inspectors interviewed licensee staff and reviewed the inspection plan/procedure and a sample of inspection results for the inspection conducted in the Unit 1 SG "B" secondary side internals, to verify that potential areas of degradation based on site-specific operating experience were inspected, and appropriate corrective actions were taken to address degradation indications. The sample of inspection results consisted of pictures showing the condition of accessible surfaces of the steam drum, uppermost tube support plate, anti-vibration bars, J-nozzle welds, feedwater rings, feedwater ring supports, primary separator swirl vanes, and lower bank secondary separators.
Based on the review of eddy current examination results for SG B and interviews with the licensee, the inspectors confirmed that no new degradation mechanisms were identified, no eddy current scope expansion was required, none of the SG tubes examined met the criteria for in-situ pressure testing, and none of the indications left in-service required repair. Furthermore, the inspectors interviewed licensee staff and reviewed the inspection plan/procedure and a sample of inspection results for the inspection conducted in the Unit 1 SG B secondary side internals, to verify that potential areas of degradation based on site-specific operating experience were inspected, and appropriate corrective actions were taken to address degradation indications. The sample of inspection results consisted of pictures showing the condition of accessible surfaces of the steam drum, uppermost tube support plate, anti-vibration bars, J-nozzle welds, feedwater rings, feedwater ring supports, primary separator swirl vanes, and lower bank secondary separators.


Identification and Resolution of Problems
Identification and Resolution of Problems: The inspectors reviewed a sample of corrective action program documents associated with ISI issues to verify that the licensee was identifying problems at an appropriate threshold and entering them in the corrective action program for resolution. The inspectors performed this review to ensure compliance with 10CFR Part 50, Appendix B, Criterion XVI, Corrective Action, requirements. The sample of corrective action documents selected for review also included the licensees evaluation of recent operating experience information applicable to the plant. The corrective action documents reviewed by the inspectors are listed in the report attachment.
: The inspectors reviewed a sample of corrective action program documents associated with ISI issues to verify that the licensee was identifying problems at an appropriate threshold and entering them in the corrective action program for resolution. The inspectors performed this review to ensure compliance with 10CFR Part 50, Appendix B, Criterion XVI, "Corrective Action," requirements. The sample of corrective action documents selected for review also included the licensee's evaluation of recent operating experience information applicable to the plant. The corrective action documents reviewed by the inspectors are listed in  
 
the report attachment.


====b. Findings====
====b. Findings====


=====Introduction:=====
=====Introduction:=====
The inspectors identified an unresolved item related to the inspection of the reactor pressure vessel (RPV) component supports as required by ASME BPVC Section XI, for which additional information is needed to determine if the issue of concern represents a performance deficiency or a violation of the regulatory  
The inspectors identified an unresolved item related to the inspection of the reactor pressure vessel (RPV) component supports as required by ASME BPVC Section XI, for which additional information is needed to determine if the issue of concern represents a performance deficiency or a violation of the regulatory requirements.
 
requirements.


=====Description:=====
=====Description:=====
The code of record for the current ISI program at Surry Power Station Unit 1 is the 1998 Edition of the ASME BPVC Section XI with the 2000 addenda. This Code edition includes inspection requirements for both nuclear class 1 piping and vessel supports (Subsection IWF) and their attachment welds (Subsection IWB). Subsection IWB, Table IWB-2500-1, item number B10.10, describes the examination requirements for welded attachments for vessels, piping, pumps, and valves. Note 1 of Table IWB-2500-1 states that attachment welds (weld buildup) on nozzles that are in compression under normal load conditions and provide only component support are excluded from the surface examination requirements. The note also provides additional conditions to identify what type welded attachment configurations require inspection. Table IWB-2500-1 also references Figures IWB-2500-13, -14 and -15 to further describe the examination requirements.
The code of record for the current ISI program at Surry Power Station Unit 1 is the 1998 Edition of the ASME BPVC Section XI with the 2000 addenda. This Code edition includes inspection requirements for both nuclear class 1 piping and vessel supports (Subsection IWF) and their attachment welds (Subsection IWB). Subsection IWB, Table IWB-2500-1, item number B10.10, describes the examination requirements for welded attachments for vessels, piping, pumps, and valves. Note 1 of Table IWB-2500-1 states that attachment welds (weld buildup) on nozzles that are in compression under normal load conditions and provide only component support are excluded from the surface examination requirements. The note also provides additional conditions to identify what type welded attachment configurations require inspection. Table IWB-2500-1 also references Figures IWB-2500-13, -14 and -15 to further describe the examination requirements.


The inspectors noted that the scope of the Surry Unit 1 ISI program for the inspection of the nuclear class 1 RPV supports did include the requirements for the IWF portion of the ASME Section XI code required inspections. However, the inspectors identified that the licensee excluded the surface examination requirements for the RPV support attachment welds required by Table IWB-2500-1, item number B10.10 based on the exemptions provided by Note 1 of the table. The licensee's position was that the surface examinations are not required based on the exclusion criteria provided in Note 1 for attachment welds under compressive loads during normal conditions and the configurations described in Figures IWB-2500-13, -14 and -15.
The inspectors noted that the scope of the Surry Unit 1 ISI program for the inspection of the nuclear class 1 RPV supports did include the requirements for the IWF portion of the ASME Section XI code required inspections. However, the inspectors identified that the licensee excluded the surface examination requirements for the RPV support attachment welds required by Table IWB-2500-1, item number B10.10 based on the exemptions provided by Note 1 of the table. The licensees position was that the surface examinations are not required based on the exclusion criteria provided in Note 1 for attachment welds under compressive loads during normal conditions and the configurations described in Figures IWB-2500-13, -14 and -15.


The inspectors reviewed design basis documents for the Unit 1 RPV supports and identified that the normal loading conditions of the supports included both compressive and shear loads. The inspectors determined that additional information and discussion with the NRC Office of Nuclear Reactor Regulation (NRR) staff was required, in order to determine if the licensee's interpretation and implementation of the exemptions in Table IWB-2500-1 were in compliance with the ASME BPVC Section XI. Therefore, the NRR and Region II staff agreed to submit a Task Interface Agreement (TIA), which could involve the submittal of a formal inquiry to the applicable ASME BPVC committee to request an interpretation of the examination requirements and exemptions in Table IWB-2500-1 for welded attachments for vessels and piping. The NRC initiated TIA-2014-02 to determine the staff's position on whether the configuration of the RPV supports at Surry meets the exclusion criteria in ASME BPVC Section XI.
The inspectors reviewed design basis documents for the Unit 1 RPV supports and identified that the normal loading conditions of the supports included both compressive and shear loads. The inspectors determined that additional information and discussion with the NRC Office of Nuclear Reactor Regulation (NRR) staff was required, in order to determine if the licensees interpretation and implementation of the exemptions in Table IWB-2500-1 were in compliance with the ASME BPVC Section XI. Therefore, the NRR and Region II staff agreed to submit a Task Interface Agreement (TIA), which could involve the submittal of a formal inquiry to the applicable ASME BPVC committee to request an interpretation of the examination requirements and exemptions in Table IWB-2500-1 for welded attachments for vessels and piping. The NRC initiated TIA-2014-02 to determine the staffs position on whether the configuration of the RPV supports at Surry meets the exclusion criteria in ASME BPVC Section XI.


This issue remains unresolved until the resolution of TIA-2014-02 to determine if the issue of concern represents a performance deficiency or a violation of regulatory requirements. This issue is identified as URI 05000280/2013005-01, Application of ASME Section XI, Table IWB 2500-1, Item B10.10, Inspection Requirements and Note 1 Exemptions.
This issue remains unresolved until the resolution of TIA-2014-02 to determine if the issue of concern represents a performance deficiency or a violation of regulatory requirements. This issue is identified as URI 05000280/2013005-01, Application of ASME Section XI, Table IWB 2500-1, Item B10.10, Inspection Requirements and Note 1 Exemptions.
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====a. Inspection Scope====
====a. Inspection Scope====
The inspectors observed and evaluated a licensed operator simulator exercise given on November 26, 2013. The scenario was intended to exercise the entire operations crew and assess the ability of the operators to react correctly to multiple failures. The inspectors observed the crew's performance to determine whether the crew met the scenario objectives; accomplished the critical tasks; demonstrated the ability to take timely action in a safe direction and to prioritize, interpret, and verify alarms; demonstrated proper use of alarm response, abnormal, and emergency operating procedures; demonstrated proper command and control; communicated effectively; and appropriately classified events per the emergency plan. The inspectors observed the post training critique to determine that weaknesses or improvement areas revealed by the training were captured by the instructor and reviewed with the operators.
The inspectors observed and evaluated a licensed operator simulator exercise given on November 26, 2013. The scenario was intended to exercise the entire operations crew and assess the ability of the operators to react correctly to multiple failures. The inspectors observed the crews performance to determine whether the crew met the scenario objectives; accomplished the critical tasks; demonstrated the ability to take timely action in a safe direction and to prioritize, interpret, and verify alarms; demonstrated proper use of alarm response, abnormal, and emergency operating procedures; demonstrated proper command and control; communicated effectively; and appropriately classified events per the emergency plan. The inspectors observed the post training critique to determine that weaknesses or improvement areas revealed by the training were captured by the instructor and reviewed with the operators.


====b. Findings====
====b. Findings====
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====a. Inspection Scope====
====a. Inspection Scope====
During the inspection period, the inspectors conducted observations of licensed reactor operator activities to ensure consistency with licensee procedures and regulatory requirements. For the following activities, the inspectors observed the following elements of operator performance: 1) operator compliance and use of plant procedures including technical specifications; 2) control board component manipulations; 3) use and interpretation of plant instrumentation and alarms; 4) documentation of activities; 5) management and supervision of activities; and 6) control room communications.
During the inspection period, the inspectors conducted observations of licensed reactor operator activities to ensure consistency with licensee procedures and regulatory requirements. For the following activities, the inspectors observed the following elements of operator performance: 1) operator compliance and use of plant procedures including technical specifications; 2) control board component manipulations; 3) use and interpretation of plant instrumentation and alarms; 4) documentation of activities; 5) management and supervision of activities; and 6) control room communications.
* Unit 1 shutdown for RFO 1R25.
* Unit 1 shutdown for RFO 1R25.
* Unit 1 draining of reactor coolant to flange level during RFO 1R25.
* Unit 1 draining of reactor coolant to flange level during RFO 1R25.
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====a. Inspection Scope====
====a. Inspection Scope====
On February 8, 2013, the licensee completed the annual requalification operating examinations required to be administered to all licensed operators in accordance with Title 10 of the Code of Federal Regulations 55.59(a)(2), "Requalification requirements,"
On February 8, 2013, the licensee completed the annual requalification operating examinations required to be administered to all licensed operators in accordance with Title 10 of the Code of Federal Regulations 55.59(a)(2), Requalification requirements, of the NRCs Operators Licenses. The inspectors performed an in-office review of the overall pass/fail results of the individual operating examinations and the crew simulator operating examinations in accordance with Inspection Procedure (IP) 71111.11, Licensed Operator Requalification Program and Licensed Operator Performance. The results were compared to the thresholds established in Section 3.02, Requalification Examination Results, of IP 71111.11.
of the NRC's "Operators' Licenses."  The inspectors performed an in-office review of the overall pass/fail results of the individual operating examinations and the crew simulator operating examinations in accordance with Inspection Procedure (IP) 71111.11, "Licensed Operator Requalification Program and Licensed Operator Performance."  The results were compared to the thresholds established in Section 3.02, "Requalification Examination Results," of IP 71111.11.


====b. Findings====
====b. Findings====
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====a. Inspection Scope====
====a. Inspection Scope====
For the three equipment issues described in the condition reports listed below, the inspectors evaluated the effectiveness of the corresponding licensee's preventive and corrective maintenance. The inspectors performed a detailed review of the problem history and associated circumstances, evaluated the extent of condition reviews, as required, and reviewed the generic implications of the equipment and/or work practice problem(s). Inspectors performed walkdowns of the accessible portions of the system, performed in-office reviews of procedures and evaluations, and held discussions with system engineers. The inspectors compared the licensee's actions with the requirements of the Maintenance Rule (10 CFR 50.65), station procedures ER-AA-MRL-10, "Maintenance Rule Program," Revision 5, and ER-AA-MRL-100, "Implementing the Maintenance Rule," Revision 6.
For the three equipment issues described in the condition reports listed below, the inspectors evaluated the effectiveness of the corresponding licensee's preventive and corrective maintenance. The inspectors performed a detailed review of the problem history and associated circumstances, evaluated the extent of condition reviews, as required, and reviewed the generic implications of the equipment and/or work practice problem(s). Inspectors performed walkdowns of the accessible portions of the system, performed in-office reviews of procedures and evaluations, and held discussions with system engineers. The inspectors compared the licensees actions with the requirements of the Maintenance Rule (10 CFR 50.65), station procedures ER-AA-MRL-10, Maintenance Rule Program, Revision 5, and ER-AA-MRL-100, Implementing the Maintenance Rule, Revision 6.
* CR 496625, Loss of Unit 2 'B' DC Bus and Vital Buses 2-IV and 2-II
* CR 496625, Loss of Unit 2 'B' DC Bus and Vital Buses 2-IV and 2-II
* CR 523699, 1-FP-P-2, Diesel Driven Fire Pump tripping on overspeed
* CR 523699, 1-FP-P-2, Diesel Driven Fire Pump tripping on overspeed
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====a. Inspection Scope====
====a. Inspection Scope====
The inspectors evaluated, as appropriate, the three activities listed below for the following: 1) the effectiveness of the risk assessments performed before maintenance activities were conducted; 2) the management of risk; 3) that, upon identification of an unforeseen situation, necessary steps were taken to plan and control the resulting emergent work activities; and, 4) that maintenance risk assessments and emergent work problems were adequately identified and resolved. The inspectors verified that the licensee was complying with the requirements of 10 CFR 50.65(a)(4) and the data output from the licensee's safety monitor associated with the risk profile of Units 1 and 2. The inspectors reviewed the corrective action program to verify deficiencies in risk assessments were being identified and properly resolved.
The inspectors evaluated, as appropriate, the three activities listed below for the following: 1) the effectiveness of the risk assessments performed before maintenance activities were conducted; 2) the management of risk; 3) that, upon identification of an unforeseen situation, necessary steps were taken to plan and control the resulting emergent work activities; and, 4) that maintenance risk assessments and emergent work problems were adequately identified and resolved. The inspectors verified that the licensee was complying with the requirements of 10 CFR 50.65(a)(4) and the data output from the licensees safety monitor associated with the risk profile of Units 1 and 2.
* Unit 1 and Unit 2 risk during "1J" bus degraded voltage and logic testing
 
The inspectors reviewed the corrective action program to verify deficiencies in risk assessments were being identified and properly resolved.
* Unit 1 and Unit 2 risk during 1J bus degraded voltage and logic testing
* Unit 1 and Unit 2 risk during component cooling heat exchanger (CCHX) SW jumper in service, 1B battery out of service and Unit 1 core offload in progress
* Unit 1 and Unit 2 risk during component cooling heat exchanger (CCHX) SW jumper in service, 1B battery out of service and Unit 1 core offload in progress
* Unit 1 and Unit 2 risk during the removal and return to service of the "C" reserve station service transformers (RSST) with Unit 1 at lowered RCS inventory
* Unit 1 and Unit 2 risk during the removal and return to service of the "C" reserve station service transformers (RSST) with Unit 1 at lowered RCS inventory
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====a. Inspection Scope====
====a. Inspection Scope====
The inspectors reviewed the six operability evaluations listed below, affecting risk-significant mitigating systems, to assess as appropriate: 1) the technical adequacy of the evaluations; 2) whether continued system operability was warranted; 3) whether other existing degraded conditions were considered; 4) if compensatory measures were involved, whether the compensatory measures were in place, would work as intended, and were appropriately controlled; and 5) where continued operability was considered unjustified, the impact on TS Limiting Conditions for Operation and the risk significance.
The inspectors reviewed the six operability evaluations listed below, affecting risk-significant mitigating systems, to assess as appropriate: 1) the technical adequacy of the evaluations; 2) whether continued system operability was warranted; 3) whether other existing degraded conditions were considered; 4) if compensatory measures were involved, whether the compensatory measures were in place, would work as intended, and were appropriately controlled; and 5) where continued operability was considered unjustified, the impact on TS Limiting Conditions for Operation and the risk significance.


The inspectors' review included verification that operability determinations were made as specified in OP-AA-102, "Operability Determination," Revision 10. The inspectors reviewed the licensee's corrective action program to verify deficiencies in operability determinations were being identified and corrected.
The inspectors review included verification that operability determinations were made as specified in OP-AA-102, Operability Determination, Revision 10. The inspectors reviewed the licensees corrective action program to verify deficiencies in operability determinations were being identified and corrected.
* CR 528665, Units 1 and 2 containment spray & recirculation spray check valve counterweights.
* CR 528665, Units 1 and 2 containment spray & recirculation spray check valve counterweights.
* CR 529007, 1-CH-P-1B, "B" charging pump gearbox observed oil foaming.
* CR 529007, 1-CH-P-1B, B charging pump gearbox observed oil foaming.
* CR 532802, Unit 1 "C" loop room drain clogged on containment closeout walkdown.
* CR 532802, Unit 1 "C" loop room drain clogged on containment closeout walkdown.
* CRs 532841 and 532718, 1-VS-MOV-100C/D, containment purge exhaust valves, have leakage greater than 1-OPT-CT-201 acceptance criteria.
* CRs 532841 and 532718, 1-VS-MOV-100C/D, containment purge exhaust valves, have leakage greater than 1-OPT-CT-201 acceptance criteria.
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====a. Inspection Scope====
====a. Inspection Scope====
The inspectors reviewed temporary modification, SU-11-00017, "CCHX SW Inlet Pipe Repair & 1-SW-MOV-102A/B Replacements" to verify that the modification did not affect system operability or availability as described by the TS and UFSAR. In addition, the inspectors verified that the temporary modification was in accordance with CM-AA-TDC-204, "Temporary Modifications," Revision 3, and for the related work package, that adequate controls were in place, procedures and drawings were updated, and post-installation tests verified the operability of the affected systems.
The inspectors reviewed temporary modification, SU-11-00017, CCHX SW Inlet Pipe Repair & 1-SW-MOV-102A/B Replacements to verify that the modification did not affect system operability or availability as described by the TS and UFSAR. In addition, the inspectors verified that the temporary modification was in accordance with CM-AA-TDC-204, Temporary Modifications, Revision 3, and for the related work package, that adequate controls were in place, procedures and drawings were updated, and post-installation tests verified the operability of the affected systems.


====b. Findings====
====b. Findings====
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====a. Inspection Scope====
====a. Inspection Scope====
The inspectors reviewed a permanent plant modification design change (DC) SU-12-00022, "FLEX BDB Mechanical Connections" which is associated with Fukushima Mitigating Strategies (FMS) related modifications. The inspection scope for this modification was restricted to those elements necessary to satisfy the stated objectives of IP 71111.18, specifically:
The inspectors reviewed a permanent plant modification design change (DC) SU-12-00022, FLEX BDB Mechanical Connections which is associated with Fukushima Mitigating Strategies (FMS) related modifications. The inspection scope for this modification was restricted to those elements necessary to satisfy the stated objectives of IP 71111.18, specifically:
* To verify that modifications have not affected the safety functions of important safety systems
* To verify that modifications have not affected the safety functions of important safety systems
* To verify that the current design bases, licensing bases, and performance capability of risk significant SSCs have not been degraded through modifications; and
* To verify that the current design bases, licensing bases, and performance capability of risk significant SSCs have not been degraded through modifications; and
* To verify that modifications performed during increased risk-significant configurations did not place the plant in an unsafe condition.
* To verify that modifications performed during increased risk-significant configurations did not place the plant in an unsafe condition.


The inspectors also verified that the permanent modification was in accordance with CM-AA-DDC-201, "Design Changes," Revision 12. The inspection did not address whether the associated FMS modification satisfactorily addressed the objectives of NRC Order EA-12-049, "Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis (BDB) External Events.
The inspectors also verified that the permanent modification was in accordance with CM-AA-DDC-201, Design Changes, Revision 12. The inspection did not address whether the associated FMS modification satisfactorily addressed the objectives of NRC Order EA-12-049, Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis (BDB) External Events.


====b. Findings====
====b. Findings====


=====Introduction:=====
=====Introduction:=====
An NRC-identified Green NCV of 10 CFR 50, Appendix B, Criterion III, "Design Control," was identified for the licensee's failure to adequately protect safety-related low head safety injection system (LHSI) piping from a tornado missile.
An NRC-identified Green NCV of 10 CFR 50, Appendix B, Criterion III, Design Control, was identified for the licensees failure to adequately protect safety-related low head safety injection system (LHSI) piping from a tornado missile.


Specifically, the licensee installed a mechanical piping connection to the Unit 1 LHSI system as part of a DC and did not provide tornado missile protection for the portions of piping that were installed above the 28.6 foot' elevation in the safeguards valve pit and connected directly to the LHSI piping.
Specifically, the licensee installed a mechanical piping connection to the Unit 1 LHSI system as part of a DC and did not provide tornado missile protection for the portions of piping that were installed above the 28.6 foot elevation in the safeguards valve pit and connected directly to the LHSI piping.


=====Description:=====
=====Description:=====
The licensee completed installation of the Unit 1 physical plant connections of DC SU-12-00022, "SPS 1 and 2 FLEX Mechani cal Connections" during the fall 2013 refueling outage. The design change is part of the licensee's compliance with NRC order EA-12-049, "Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for BDB External Events" in accordance with NEI 12-06, "Diverse and Flexible Coping Strategies (FLEX) Implementation Guide."  One of the mechanical connections of this modification is designed to allow for high pressure make-up/borated water injection into the Reactor Coolant System during a BDB event. This was accomplished through a welded pipe connection on the existing 6 inch LHSI piping located between containment and the normally closed containment isolation valve 1-SI-MOV-1890A, "A" LHSI hot leg discharge valve. The 3 inch FLEX piping traversed from this connection, on the 13.9 foot elevation of the safeguards valve pit, through the floor of the two upper levels to the grade level 28.6 foot elevation at the top of the safeguards valve pit, where it is terminated with a hose connection.
The licensee completed installation of the Unit 1 physical plant connections of DC SU-12-00022, SPS 1 and 2 FLEX Mechanical Connections during the fall 2013 refueling outage. The design change is part of the licensees compliance with NRC order EA-12-049, Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for BDB External Events in accordance with NEI 12-06, Diverse and Flexible Coping Strategies (FLEX) Implementation Guide. One of the mechanical connections of this modification is designed to allow for high pressure make-up/borated water injection into the Reactor Coolant System during a BDB event. This was accomplished through a welded pipe connection on the existing 6 inch LHSI piping located between containment and the normally closed containment isolation valve 1-SI-MOV-1890A, A LHSI hot leg discharge valve. The 3 inch FLEX piping traversed from this connection, on the 13.9 foot elevation of the safeguards valve pit, through the floor of the two upper levels to the grade level 28.6 foot elevation at the top of the safeguards valve pit, where it is terminated with a hose connection.


The residents questioned the adequacy of this design and noted that the new pipe was only supported by a single seismic restraint on the 28.6 foot elevation, which is per design, but also results in a higher load on the welded LHSI piping connection from externally initiated events. As a result, it was identified that the portion of piping installed above the 28.6 foot elevation is susceptible to a tornado generated missile because the roof of this level of the safeguards valve pit is corrugated metal that was never designed to protect against tornado missiles. The licensee could not immediately determine potential loading from such an impact on the safety-related LHSI piping two floors below.
The residents questioned the adequacy of this design and noted that the new pipe was only supported by a single seismic restraint on the 28.6 foot elevation, which is per design, but also results in a higher load on the welded LHSI piping connection from externally initiated events. As a result, it was identified that the portion of piping installed above the 28.6 foot elevation is susceptible to a tornado generated missile because the roof of this level of the safeguards valve pit is corrugated metal that was never designed to protect against tornado missiles. The licensee could not immediately determine potential loading from such an impact on the safety-related LHSI piping two floors below.
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The licensee evaluated the concern in CR 533401 and performed an immediate operability determination (IOD). The IOD concluded that the piping was operable based on engineering judgment and also non-conforming to the design bases. Initially, a prompt operability determination (POD) was assigned to perform an analysis of the postulated tornado missile impact on the LHSI System piping. However, the POD was cancelled before it was due, because the licensee decided to cut and remove a portion of the FLEX piping below the 28.6 foot elevation to eliminate the possibility of a tornado missile impact affecting the LHSI piping. The design will be reworked prior to the planned Unit 2 modification in spring of 2014.
The licensee evaluated the concern in CR 533401 and performed an immediate operability determination (IOD). The IOD concluded that the piping was operable based on engineering judgment and also non-conforming to the design bases. Initially, a prompt operability determination (POD) was assigned to perform an analysis of the postulated tornado missile impact on the LHSI System piping. However, the POD was cancelled before it was due, because the licensee decided to cut and remove a portion of the FLEX piping below the 28.6 foot elevation to eliminate the possibility of a tornado missile impact affecting the LHSI piping. The design will be reworked prior to the planned Unit 2 modification in spring of 2014.


Dominion procedure CM-AA-DDC-201, "Design Changes", Rev. 12 step 3.3.5.h required the following during installation of the DC: "If practical and station conditions allow, then walkdown the DC to ensure it is being installed in accordance with the DC and no adverse consequences are created by the modification."  The design change document, SU-12-00022, section 3.0 stated, in part, "All new piping, valves, tee connections etc., are designed to withstand design bases external flooding, storms with high winds (hurricanes, tornadoes, etc.,) and associated missiles."  Because the design modification exposed the LHSI system piping to a tornado missile hazard, the issue was placed in the corrective action program as CR 533401.
Dominion procedure CM-AA-DDC-201, Design Changes, Rev. 12 step 3.3.5.h required the following during installation of the DC: If practical and station conditions allow, then walkdown the DC to ensure it is being installed in accordance with the DC and no adverse consequences are created by the modification. The design change document, SU-12-00022, section 3.0 stated, in part, All new piping, valves, tee connections etc.,
are designed to withstand design bases external flooding, storms with high winds (hurricanes, tornadoes, etc.,) and associated missiles. Because the design modification exposed the LHSI system piping to a tornado missile hazard, the issue was placed in the corrective action program as CR 533401.


Through discussions with Dominion corporate engineering staff, the residents learned that the lack of missile protection at the 28.6 foot elevation was discovered during the 30% design review. As a result, the planned location of the safety related boundary valve (1-SI-500) was moved two floors down from its original designed position above the floor level. However, despite this change, the forces that could be transmitted to safety related equipment below floor level were overlooked. Any break in 3 inch piping between 1-SI-500 and the LHSI piping because of an external missile would affect containment integrity. The licensee attributed this oversight to the lack of participation and coordination with the civil engineering department during this portion of the design review and associated plant walkdowns.
Through discussions with Dominion corporate engineering staff, the residents learned that the lack of missile protection at the 28.6 foot elevation was discovered during the 30% design review. As a result, the planned location of the safety related boundary valve (1-SI-500) was moved two floors down from its original designed position above the floor level. However, despite this change, the forces that could be transmitted to safety related equipment below floor level were overlooked. Any break in 3 inch piping between 1-SI-500 and the LHSI piping because of an external missile would affect containment integrity. The licensee attributed this oversight to the lack of participation and coordination with the civil engineering department during this portion of the design review and associated plant walkdowns.


=====Analysis:=====
=====Analysis:=====
The inspectors concluded that the licensee's failure to protect the Unit 1 LHSI system piping against external missile hazards when the piping was modified by the FLEX mechanical piping connection as part of DC SU-12-00022 was a PD that was within the licensee's ability to foresee and correct. Specifically, Dominion procedure CM-AA-DDC-201, Rev. 12 step 3.3.5.h requires confirmation that a modification is installed in accordance with the DC and no adverse unintended consequences are created; while DC SU-12-00022 further requires the mechanical piping modifications to be able to withstand tornado missiles. The inspectors determined that the PD was more than minor because it was associated with the design control attribute of the Barrier Integrity Cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. Specifically, the welded piping connection installed by DC SU-12-00022 located between containment and containm ent isolation valve 1-SI-MOV-1890A, "A" LHSI hot leg discharge isolation valve, was susceptible to failure from the impact of a tornado generated missile.
The inspectors concluded that the licensees failure to protect the Unit 1 LHSI system piping against external missile hazards when the piping was modified by the FLEX mechanical piping connection as part of DC SU-12-00022 was a PD that was within the licensees ability to foresee and correct. Specifically, Dominion procedure CM-AA-DDC-201, Rev. 12 step 3.3.5.h requires confirmation that a modification is installed in accordance with the DC and no adverse unintended consequences are created; while DC SU-12-00022 further requires the mechanical piping modifications to be able to withstand tornado missiles. The inspectors determined that the PD was more than minor because it was associated with the design control attribute of the Barrier Integrity Cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. Specifically, the welded piping connection installed by DC SU-12-00022 located between containment and containment isolation valve 1-SI-MOV-1890A, A LHSI hot leg discharge isolation valve, was susceptible to failure from the impact of a tornado generated missile.


Using Manual Chapter 0609.04, Initial Characterization of Findings, Table 2, dated June 19, 2012, the finding was determined to affect the reactor containment barrier and the Barrier Integrity Cornerstone. The inspectors screened the finding using Manual Chapter 0609 Appendix A, "Significance Determination Process (SDP) for Findings at-Power", and determined the finding was of very low safety significance (Green) because the finding did not represent an actual open pathway in the physical integrity of reactor containment. This finding has a cross-cutting aspect in the work control component of the human performance area, H.3(b); because the licensee failed to address both the impact of changes in the work scope on the plant and to use adequate interdepartmental coordination during the design process.
Using Manual Chapter 0609.04, Initial Characterization of Findings, Table 2, dated June 19, 2012, the finding was determined to affect the reactor containment barrier and the Barrier Integrity Cornerstone. The inspectors screened the finding using Manual Chapter 0609 Appendix A, Significance Determination Process (SDP) for Findings at-Power, and determined the finding was of very low safety significance (Green) because the finding did not represent an actual open pathway in the physical integrity of reactor containment. This finding has a cross-cutting aspect in the work control component of the human performance area, H.3(b); because the licensee failed to address both the impact of changes in the work scope on the plant and to use adequate interdepartmental coordination during the design process.


Enforcement
=====Enforcement:=====
:  10 CFR 50, Appendix B, Criterion III, Design Control, requires, in part, that design changes, including field changes, shall be subject to design control measures commensurate with those applied to the original design. Contrary to the above, on November 21, 2013, the licensee did not assure that changes to the LHSI piping made by DC SU-2012-00022 were subject to the same design control measures as the original piping design. Specifically, the licensee installed a mechanical piping connection to the LHSI System as part of this design change and did not ensure that the new piping was protected from tornado missiles such that existing LHSI piping would remain unaffected. Because the licensee entered the issue into their corrective action program as CR 533401 and the finding is Green, this violation is being treated as an NCV, consistent with Section 2.3.2 of the NRC Enforcement Policy: NCV 05000280/2013005-02, Failure to Missile Protect Beyond Design Bases FLEX Modification to Low Head Safety Injection Piping.
10 CFR 50, Appendix B, Criterion III, Design Control, requires, in part, that design changes, including field changes, shall be subject to design control measures commensurate with those applied to the original design. Contrary to the above, on November 21, 2013, the licensee did not assure that changes to the LHSI piping made by DC SU-2012-00022 were subject to the same design control measures as the original piping design. Specifically, the licensee installed a mechanical piping connection to the LHSI System as part of this design change and did not ensure that the new piping was protected from tornado missiles such that existing LHSI piping would remain unaffected. Because the licensee entered the issue into their corrective action program as CR 533401 and the finding is Green, this violation is being treated as an NCV, consistent with Section 2.3.2 of the NRC Enforcement Policy: NCV 05000280/2013005-02, Failure to Missile Protect Beyond Design Bases FLEX Modification to Low Head Safety Injection Piping.


===.2 Permanent Modification SU-13-01019===
===.2 Permanent Modification SU-13-01019===


====a. Inspection Scope====
====a. Inspection Scope====
The inspectors reviewed a permanent plant modification DC SU-13-01019, "BDB-Flex Power for Essential Instrumentation and Equipment" which is associated with FMS related modifications. The inspection scope for this modification was restricted to those elements necessary to satisfy the stated objectives of IP 71111.18, specifically:
The inspectors reviewed a permanent plant modification DC SU-13-01019, BDB-Flex Power for Essential Instrumentation and Equipment which is associated with FMS related modifications. The inspection scope for this modification was restricted to those elements necessary to satisfy the stated objectives of IP 71111.18, specifically:
* To verify that modifications have not affected the safety functions of important safety systems
* To verify that modifications have not affected the safety functions of important safety systems
* To verify that the current design bases, licensing bases, and performance capability of risk significant SSCs have not been degraded through modifications; and
* To verify that the current design bases, licensing bases, and performance capability of risk significant SSCs have not been degraded through modifications; and
* To verify that modifications performed during increased risk-significant configurations did not place the plant in an unsafe condition.
* To verify that modifications performed during increased risk-significant configurations did not place the plant in an unsafe condition.


The inspectors also verified that the permanent modification was in accordance with CM-AA-DDC-201, "Design Changes," Revision 12. The inspection did not address whether the associated FMS modification satisfactorily addressed the objectives of NRC Order EA-12-049, "Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for BDB External Events.
The inspectors also verified that the permanent modification was in accordance with CM-AA-DDC-201, Design Changes, Revision 12. The inspection did not address whether the associated FMS modification satisfactorily addressed the objectives of NRC Order EA-12-049, Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for BDB External Events.


====b. Findings====
====b. Findings====
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====a. Inspection Scope====
====a. Inspection Scope====
The inspectors reviewed six post maintenance test procedures and/or test activities for selected risk-significant mitigating systems listed below, to assess whether: 1) the effect of testing on the plant had been adequately addressed by control room and/or engineering personnel; 2) testing was adequate for the maintenance performed;           3) acceptance criteria were clear and adequately demonstrated operational readiness consistent with design and licensing basis documents; 4) test instrumentation had current calibrations, range, and accuracy consistent with the application; 5) tests were performed as written with applicable prerequisites satisfied; 6) jumpers installed or leads lifted were properly controlled; 7) test equipment was removed following testing; and     8) equipment was returned to the status required to perform in accordance with VPAP-2003, "Post Maintenance Testing Program," Revision 14.
The inspectors reviewed six post maintenance test procedures and/or test activities for selected risk-significant mitigating systems listed below, to assess whether: 1) the effect of testing on the plant had been adequately addressed by control room and/or engineering personnel; 2) testing was adequate for the maintenance performed; 3) acceptance criteria were clear and adequately demonstrated operational readiness consistent with design and licensing basis documents; 4) test instrumentation had current calibrations, range, and accuracy consistent with the application; 5) tests were performed as written with applicable prerequisites satisfied; 6) jumpers installed or leads lifted were properly controlled; 7) test equipment was removed following testing; and 8) equipment was returned to the status required to perform in accordance with VPAP-2003, Post Maintenance Testing Program, Revision 14.
* 1-OPT-EG-001, Rev. 58, Emergency Diesel Generator (EDG) #1 Monthly Start Exercise Test, following replacement of the fast start reset relay
* 1-OPT-EG-001, Rev. 58, Emergency Diesel Generator (EDG) #1 Monthly Start Exercise Test, following replacement of the fast start reset relay
* 1-EPT-0102-02, Rev. 4 and 1-EPT-0103-02, Rev. 14, Monthly and Quarterly Voltage Check of the 1B Station Battery, following replacement of battery cell #17 and discharge testing of the 1B station battery
* 1-EPT-0102-02, Rev. 4 and 1-EPT-0103-02, Rev. 14, Monthly and Quarterly Voltage Check of the 1B Station Battery, following replacement of battery cell #17 and discharge testing of the 1B station battery
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====a. Inspection Scope====
====a. Inspection Scope====
The inspectors reviewed the station's work schedule and outage risk plan for the Unit 1 refueling outage, which was conducted October 20 through November 21, 2013, to confirm that the licensee had appropriately considered risk, industry experience, and previous site-specific problems in developing and implementing a plan that assured maintenance of defense-in-depth. The inspectors used IP 71111.20, "Refueling and Outage Activities," to observe portions of the maintenance and startup activities to verify that the licensee maintained defense-in-depth commensurate with the outage risk plan and applicable TS. The inspectors monitored licensee controls over the outage activities listed below.
The inspectors reviewed the stations work schedule and outage risk plan for the Unit 1 refueling outage, which was conducted October 20 through November 21, 2013, to confirm that the licensee had appropriately considered risk, industry experience, and previous site-specific problems in developing and implementing a plan that assured maintenance of defense-in-depth. The inspectors used IP 71111.20, Refueling and Outage Activities, to observe portions of the maintenance and startup activities to verify that the licensee maintained defense-in-depth commensurate with the outage risk plan and applicable TS. The inspectors monitored licensee controls over the outage activities listed below.
* Licensee configuration management, including daily outage reports, to evaluate maintenance of defense-in-depth commensurate with the outage risk plan for key safety functions and compliance with the applicable TS when taking equipment out of service.
* Licensee configuration management, including daily outage reports, to evaluate maintenance of defense-in-depth commensurate with the outage risk plan for key safety functions and compliance with the applicable TS when taking equipment out of service.
* Implementation of clearance activities and confirmation that tags were properly hung and equipment appropriately configured to safely support the work or testing.
* Implementation of clearance activities and confirmation that tags were properly hung and equipment appropriately configured to safely support the work or testing.
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For the eight surveillance tests listed below, the inspectors examined the test procedures, witnessed testing, or reviewed test records and data packages, to determine whether the scope of testing adequately demonstrated that the affected equipment was functional and operable, and that the surveillance requirements of TS were met. The inspectors also determined whether the testing effectively demonstrated that the systems or components were operationally ready and capable of performing their intended safety functions.
For the eight surveillance tests listed below, the inspectors examined the test procedures, witnessed testing, or reviewed test records and data packages, to determine whether the scope of testing adequately demonstrated that the affected equipment was functional and operable, and that the surveillance requirements of TS were met. The inspectors also determined whether the testing effectively demonstrated that the systems or components were operationally ready and capable of performing their intended safety functions.


In-Service Testing
In-Service Testing:
:
* 1-OPT-FW-003, Rev. 44, Turbine Driven Auxiliary Feedwater Pump Surveillance Test Reactor Coolant System Leak Rate Determination:
* 1-OPT-FW-003, Rev. 44, Turbine Driven Auxiliary Feedwater Pump Surveillance Test Reactor Coolant System Leak Rate Determination
* 2-OPT-RC-10.0, Rev. 40, Reactor Coolant Leakage - Computer Calculated Appendix J Leak Rate Determination:
:
* 2-OPT-RC-10.0, Rev. 40, Reactor Coolant Leakage - Computer Calculated Appendix J Leak Rate Determination
:
* 1-OPT-CT-201, Rev. 21, Containment Isolation Valve Local Leak Rate Testing (Type C Containment Testing) for 1-CS-13, Penetration 64
* 1-OPT-CT-201, Rev. 21, Containment Isolation Valve Local Leak Rate Testing (Type C Containment Testing) for 1-CS-13, Penetration 64
* 1-OPT-CT-201, Rev. 21, Containment Isolation Valve Local Leak Rate Testing (Type C Containment Testing) for 1-CC-TV-105A, Penetration 25
* 1-OPT-CT-201, Rev. 21, Containment Isolation Valve Local Leak Rate Testing (Type C Containment Testing) for 1-CC-TV-105A, Penetration 25
* 1-OPT-CT-201, Rev. 21, Containment Isolation Valve Local Leak Rate Testing (Type C Containment Testing) for 1-SS-TV-102B, Penetration 56B Surveillance Testing
* 1-OPT-CT-201, Rev. 21, Containment Isolation Valve Local Leak Rate Testing (Type C Containment Testing) for 1-SS-TV-102B, Penetration 56B Surveillance Testing:
:
* 2-NSP-SI-001, Rev. 1, Ultrasonic Examination of Safety Injection Piping
* 2-NSP-SI-001, Rev. 1, Ultrasonic Examination of Safety Injection Piping
* 1-OPT-RC-10.1, Rev. 11, Reactor Coolant Leakage Walkdown at Cold Shutdown
* 1-OPT-RC-10.1, Rev. 11, Reactor Coolant Leakage Walkdown at Cold Shutdown
* 1-OPT-ZZ-002, Rev. 37, ESF Actuation with Undervoltage and Degraded Voltage - 1J Bus
* 1-OPT-ZZ-002, Rev. 37, ESF Actuation with Undervoltage and Degraded Voltage -
1J Bus


====b. Findings====
====b. Findings====
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==RADIATION SAFETY==
==RADIATION SAFETY==
Cornerstones: Occupational Radiation Safety (OS) and Public Radiation Safety (PS)  
Cornerstones: Occupational Radiation Safety (OS) and Public Radiation Safety (PS)
 
{{a|2RS1}}
{{a|2RS1}}
==2RS1 Radiological Hazard Assessment and Exposure Controls==
==2RS1 Radiological Hazard Assessment and Exposure Controls==


====a. Inspection Scope====
====a. Inspection Scope====
Hazard Assessment and Instructions to Workers: During facility tours, the inspectors observed labeling of radioactive material and postings for radiation areas, high radiation areas (HRA), locked HRAs (LHRA), very HRAs (VHRA), radioactive material storage areas, and contaminated areas established within the radiologically controlled area (RCA) of the Unit 1 (U1) and Unit 2 (U2) Auxiliary Buildings, U1 containment, the Independent Spent Fuel Storage Installation (ISFSI), and radioactive waste processing and storage locations. The inspectors independently measured radiation dose rates or directly observed conduct of licensee radi ation surveys for RCA areas in the U1 containment, U1 and U2 Auxiliary buildings, and ISFSI. The inspectors reviewed survey records for several plant areas including surveys for alpha emitters, airborne radioactivity, and pre-job surveys for selected U1 Refueling 25 (U1R25) tasks. The inspectors also discussed changes to plant operations that could contribute to changing radiological conditions since the last inspection. For selected U1R25 outage jobs, the inspectors attended pre-job mockups and briefings and reviewed radiation work permit (RWP) details to assess communication of radiological control requirements and current radiological conditions to workers. Selected U1R25 work activities included U1 sliding foot inspection (reactor vessel support sliding foot and floor plate), U1 fuel transfer cart repairs, U1 Non-Destructive Examination (NDE) and ISI inspections, lower internals lift and set, 10 year ISI girth weld inspection, transfer canal work, and routine RP job  
Hazard Assessment and Instructions to Workers: During facility tours, the inspectors observed labeling of radioactive material and postings for radiation areas, high radiation areas (HRA), locked HRAs (LHRA), very HRAs (VHRA), radioactive material storage areas, and contaminated areas established within the radiologically controlled area (RCA) of the Unit 1 (U1) and Unit 2 (U2) Auxiliary Buildings, U1 containment, the Independent Spent Fuel Storage Installation (ISFSI), and radioactive waste processing and storage locations. The inspectors independently measured radiation dose rates or directly observed conduct of licensee radiation surveys for RCA areas in the U1 containment, U1 and U2 Auxiliary buildings, and ISFSI. The inspectors reviewed survey records for several plant areas including surveys for alpha emitters, airborne radioactivity, and pre-job surveys for selected U1 Refueling 25 (U1R25) tasks. The inspectors also discussed changes to plant operations that could contribute to changing radiological conditions since the last inspection. For selected U1R25 outage jobs, the inspectors attended pre-job mockups and briefings and reviewed radiation work permit (RWP) details to assess communication of radiological control requirements and current radiological conditions to workers. Selected U1R25 work activities included U1 sliding foot inspection (reactor vessel support sliding foot and floor plate), U1 fuel transfer cart repairs, U1 Non-Destructive Examination (NDE) and ISI inspections, lower internals lift and set, 10 year ISI girth weld inspection, transfer canal work, and routine RP job coverage.


coverage.
Hazard Control and Work Practices: The inspectors observed and evaluated access barrier effectiveness for selected LHRA and VHRA locations to include the U1 and U2 Auxiliary Buildings and U1 containment. Changes to procedural guidance for LHRA and VHRA controls were discussed with RP supervisors. Controls and their implementation for storage of irradiated material within the spent fuel pool were reviewed and discussed in detail. Established radiological controls (including airborne controls) were evaluated for selected tasks, including U1 sliding foot inspection, and U1 fuel transfer cart repairs.
 
Hazard Control and Work Practices: The inspectors observed and evaluated access barrier effectiveness for selected LHRA and VHRA locations to include the U1 and U2 Auxiliary Buildings and U1 containment. Changes to procedural guidance for LHRA and VHRA controls were discussed with RP supervisors. Controls and their implementation for storage of irradiated material within the spent fuel pool were reviewed and discussed in detail. Established radiological controls (including airborne controls) were evaluated for selected tasks, including U1 sliding foot inspection, and U1 fuel transfer cart repairs.


In addition, licensee controls for areas where dose rates could change significantly as a result of refueling operations were reviewed, observed, and discussed including controlling access to areas adjacent to the fuel transfer canal in the U1 Auxiliary Building and containment.
In addition, licensee controls for areas where dose rates could change significantly as a result of refueling operations were reviewed, observed, and discussed including controlling access to areas adjacent to the fuel transfer canal in the U1 Auxiliary Building and containment.
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Occupational workers adherence to selected RWPs and RP technician proficiency in providing job coverage were evaluated through direct observations and interviews with licensee staff. Electronic dosimeter (ED) alarm set points and worker stay times were evaluated against area radiation survey results. Worker response to select ED dose rate alarms was evaluated. For selected U1R25 HRA and LHRA tasks involving significant dose rate gradients, the use and placement of whole body and extremity dosimetry to monitor worker exposure was discussed with licensee staff.
Occupational workers adherence to selected RWPs and RP technician proficiency in providing job coverage were evaluated through direct observations and interviews with licensee staff. Electronic dosimeter (ED) alarm set points and worker stay times were evaluated against area radiation survey results. Worker response to select ED dose rate alarms was evaluated. For selected U1R25 HRA and LHRA tasks involving significant dose rate gradients, the use and placement of whole body and extremity dosimetry to monitor worker exposure was discussed with licensee staff.


Control of Radioactive Material: The inspectors observed surveys of material and personnel being released from the RCA and U1 containment using portable radiation  
Control of Radioactive Material: The inspectors observed surveys of material and personnel being released from the RCA and U1 containment using portable radiation survey instruments, hand and foot monitors, small article monitors, personnel contamination monitors, and portal monitor instruments. The inspectors observed current calibration labels and source check information for selected radiation and air sampling and monitoring instruments located at the RCA release point, in the Auxiliary Building, and U1 containment. The inspectors reviewed documentation of equipment sensitivity, alarm setpoints, discussed release program guidance with RP staff and observed source response testing of selected monitoring instruments located at the RCA release point. The inspectors also reviewed records of leak tests on selected sealed sources and discussed nationally tracked source transactions with RP staff.


survey instruments, hand and foot monitors, small article monitors, personnel contamination monitors, and portal monitor instruments. The inspectors observed current calibration labels and source check information for selected radiation and air sampling and monitoring instruments located at the RCA release point, in the Auxiliary Building, and U1 containment.
Problem Identification and Resolution: CAP documents associated with radiological hazard assessment and control were reviewed and assessed. The inspectors evaluated the licensees ability to identify and resolve the issues in accordance with procedure PI-AA-200, Corrective Action Program, Rev. 21. The inspectors also evaluated the scope of the licensees internal audit program and reviewed recent assessment results.


The inspectors reviewed documentation of equipment sensitivity, alarm setpoints, discussed release program guidance with RP staff and observed source response testing of selected monitoring instruments located at the RCA release point. The inspectors also reviewed records of leak tests on selected sealed sources and discussed nationally tracked source transactions with RP staff.
RP activities were evaluated against the requirements of USFAR Section 11; TS Section 6.4; 10 CFR Parts 19 and 20; and approved licensee procedures. Licensee programs for monitoring materials and personnel released from the RCA were evaluated against 10 CFR Part 20 and IE Circular 81-07, Control of Radioactively Contaminated Material.


Problem Identification and Resolution:  CAP documents associated with radiological hazard assessment and control were reviewed and assessed. The inspectors evaluated the licensee's ability to identify and resolve the issues in accordance with procedure PI-AA-200, Corrective Action Program, Rev. 21. The inspectors also evaluated the scope of the licensee's internal audit program and reviewed recent assessment results.
Documents reviewed are listed in the Attachment.
 
RP activities were evaluated against the requirements of USFAR Section 11; TS Section 6.4; 10 CFR Parts 19 and 20; and approved licensee procedures. Licensee programs for monitoring materials and personnel released from the RCA were evaluated against 10 CFR Part 20 and IE Circular 81-07, "Control of Radioactively Contaminated Material."  Documents reviewed are listed in the Attachment.


====b. Findings====
====b. Findings====
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====a. Inspection Scope====
====a. Inspection Scope====
Waste Processing and Characterization: During inspector walk-downs, accessible sections of the liquid and solid radioactive waste (radwaste) processing systems were assessed for material condition and confor mance with system design diagrams.
Waste Processing and Characterization: During inspector walk-downs, accessible sections of the liquid and solid radioactive waste (radwaste) processing systems were assessed for material condition and conformance with system design diagrams.


Inspected equipment included radwaste storage tanks; resin transfer piping, resin and filter packaging components; and abandoned evaporator equipment. The inspectors discussed component function, processing system changes, and radwaste program implementation with licensee staff.
Inspected equipment included radwaste storage tanks; resin transfer piping, resin and filter packaging components; and abandoned evaporator equipment. The inspectors discussed component function, processing system changes, and radwaste program implementation with licensee staff.


The radionuclide characterizations for 2012 and 2013 for selected waste streams were reviewed and discussed with radioactive material control (RMC) staff. For primary resin, reactor coolant system filters and dry active waste the inspectors evaluated analyses for hard-to-detect nuclides, reviewed the use of scaling factors, and examined quality assurance comparison results between licensee waste stream characterizations and outside laboratory data. Waste stream mixing and concentration averaging methodology for resins and filters was evaluated and discussed with RMC staff. The inspectors also reviewed the licensee's procedural guidance for monitoring changes in waste stream isotopic mixtures.
The radionuclide characterizations for 2012 and 2013 for selected waste streams were reviewed and discussed with radioactive material control (RMC) staff. For primary resin, reactor coolant system filters and dry active waste the inspectors evaluated analyses for hard-to-detect nuclides, reviewed the use of scaling factors, and examined quality assurance comparison results between licensee waste stream characterizations and outside laboratory data. Waste stream mixing and concentration averaging methodology for resins and filters was evaluated and discussed with RMC staff. The inspectors also reviewed the licensees procedural guidance for monitoring changes in waste stream isotopic mixtures.


Radioactive Material Storage: During walk-downs of indoor and outdoor radioactive material storage areas, the inspectors observed the physical condition and labeling of storage containers and the posting of radioactive material areas. The inspectors also reviewed licensee procedural guidance for storage and monitoring of radioactive  
Radioactive Material Storage: During walk-downs of indoor and outdoor radioactive material storage areas, the inspectors observed the physical condition and labeling of storage containers and the posting of radioactive material areas. The inspectors also reviewed licensee procedural guidance for storage and monitoring of radioactive material.


material.
Transportation: There were no significant shipments during the week of inspection; however, the inspectors did review shipping procedure requirements and discussed preparation of shipping documents, package marking and labeling, and interviewed shipping technicians regarding Department of Transportation (DOT) regulations.
 
Transportation: There were no significant shipments during the week of inspection; however, the inspectors did review shipping procedure requirements and discussed preparation of shipping documents, package marking and labeling, and interviewed shipping technicians regarding Department of Transportation (DOT) regulations.


Selected shipping records were reviewed for consistency with licensee procedures and compliance with NRC and DOT regulations. The inspectors reviewed emergency response information, DOT shipping package classification, waste classification, radiation survey results, and evaluated whether receiving licensees were authorized to accept the packages. Licensee procedures for handling shipping containers were compared to Certificate of Compliance requirements and manufacturer recommendations. In addition, training records for selected individuals currently qualified to ship radioactive material were reviewed.
Selected shipping records were reviewed for consistency with licensee procedures and compliance with NRC and DOT regulations. The inspectors reviewed emergency response information, DOT shipping package classification, waste classification, radiation survey results, and evaluated whether receiving licensees were authorized to accept the packages. Licensee procedures for handling shipping containers were compared to Certificate of Compliance requirements and manufacturer recommendations. In addition, training records for selected individuals currently qualified to ship radioactive material were reviewed.


Radwaste processing activities and equipment configuration were reviewed for compliance with the licensee's Process Control Program and UFSAR, Chapter 11.
Radwaste processing activities and equipment configuration were reviewed for compliance with the licensees Process Control Program and UFSAR, Chapter 11.


Waste stream characterization analyses were reviewed against regulations detailed in 10 CFR Part 20, 10 CFR Part 61, and guidance provided in the Branch Technical Position on Waste Classification (1983). Radioactive material and waste storage activities were reviewed against the requirements of 10 CFR Part 20. Transportation program implementation was reviewed against regulations detailed in 10 CFR Part 20, 10 CFR Part 71, 49 CFR Parts 172-178, as well as the guidance provided in NUREG-1608. Training activities were assessed against 49 CFR Part 172 Subpart H.
Waste stream characterization analyses were reviewed against regulations detailed in 10 CFR Part 20, 10 CFR Part 61, and guidance provided in the Branch Technical Position on Waste Classification (1983). Radioactive material and waste storage activities were reviewed against the requirements of 10 CFR Part 20. Transportation program implementation was reviewed against regulations detailed in 10 CFR Part 20, 10 CFR Part 71, 49 CFR Parts 172-178, as well as the guidance provided in NUREG-1608. Training activities were assessed against 49 CFR Part 172 Subpart H.
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Documents reviewed are listed in the Attachment.
Documents reviewed are listed in the Attachment.


Problem Identification and Resolution: The inspectors reviewed CRs in the area of radwaste processing and transportation. The inspectors evaluated the licensee's ability to identify and resolve the issues in accordance with procedure PI-AA-200. The inspectors also evaluated the scope of the licensee's internal audit program and reviewed recent assessment results. Documents reviewed are listed in the Attachment.
Problem Identification and Resolution: The inspectors reviewed CRs in the area of radwaste processing and transportation. The inspectors evaluated the licensees ability to identify and resolve the issues in accordance with procedure PI-AA-200. The inspectors also evaluated the scope of the licensees internal audit program and reviewed recent assessment results. Documents reviewed are listed in the Attachment.


====b. Findings====
====b. Findings====


=====Introduction:=====
=====Introduction:=====
A self-revealing Green NCV of 10 CFR 20.1802, "Control of Material not in Storage" was identified for the licensee's failure to control and maintain constant surveillance of licensed material that is in a controlled or unrestricted area and that is not in storage.
A self-revealing Green NCV of 10 CFR 20.1802, Control of Material not in Storage was identified for the licensees failure to control and maintain constant surveillance of licensed material that is in a controlled or unrestricted area and that is not in storage. Specifically, an Americium-241 check source was not returned to the designated storage location.
 
Specifically, an Americium-241 check source was not returned to the designated storage location.


=====Description:=====
=====Description:=====
On August 19, 2013, an RP technician trainee tasked with performing instrument function checks discovered that the Americium-241 (Am-241) check source was missing from the source kit. The source with 0.02 micro-Curies of Am- 241 was used to perform function checks for the iSolo alpha/beta sample counter in the health physics technical services area in the administration building. The source is greater than the exempt quantity for Am-241 and therefore must be controlled. The RP technician trainee assigned to perform function checks discovered that the 0.02 micro-Curie iSolo check source had been left inside the instrument, in the health physics technical support area. The source had not been returned to the proper storage location after being used on August 18, 2013.
On August 19, 2013, an RP technician trainee tasked with performing instrument function checks discovered that the Americium-241 (Am-241) check source was missing from the source kit. The source with 0.02 micro-Curies of Am- 241 was used to perform function checks for the iSolo alpha/beta sample counter in the health physics technical services area in the administration building. The source is greater than the exempt quantity for Am-241 and therefore must be controlled. The RP technician trainee assigned to perform function checks discovered that the 0.02 micro-Curie iSolo check source had been left inside the instrument, in the health physics technical support area. The source had not been returned to the proper storage location after being used on August 18, 2013.


On August 20, 2013, the RP technician trainee notified the RP technician that had last used the source that it had not been returned to the designated storage location. The RP Technician subsequently notified his supervision. Supervision established that the technician had not complied with procedure HP-1033.148, "Canberra iSolo: Performance Checks," step 6.4.3 which states: "Ensure check source is removed from Canberra iSolo and returned to designated storage location."  Supervision also identified that the current revision of procedure C-HP-1071.010, "Control of Radioactive Sources," implied but did not explicitly direct the source to be returned to the designated storage location after use. The source was outside the designated storage for approximately 20 hours. The licensee placed this issue in the CAP as CR 523692.
On August 20, 2013, the RP technician trainee notified the RP technician that had last used the source that it had not been returned to the designated storage location. The RP Technician subsequently notified his supervision. Supervision established that the technician had not complied with procedure HP-1033.148, Canberra iSolo: Performance Checks, step 6.4.3 which states: Ensure check source is removed from Canberra iSolo and returned to designated storage location. Supervision also identified that the current revision of procedure C-HP-1071.010, Control of Radioactive Sources, implied but did not explicitly direct the source to be returned to the designated storage location after use. The source was outside the designated storage for approximately 20 hours. The licensee placed this issue in the CAP as CR 523692.


=====Analysis:=====
=====Analysis:=====
The inspectors determined that the licensee's failure to control and maintain constant surveillance of licensed material that is in a controlled or unrestricted area and that is not in storage was a performance deficiency (PD). The PD was more than minor because it was associated with the Program and Process attribute of the Public Radiation Safety Cornerstone and affected the cornerstone objective of ensuring adequate protection of public health and safety from exposure to radioactive materials released into the public domain. Using Manual Chapter 0609, Appendix D, "Public Radiation Safety SDP," this finding determined to be was of very low safety significance (Green) in that the public radiation exposure was not greater than 0.005 rem (5 millirem).
The inspectors determined that the licensees failure to control and maintain constant surveillance of licensed material that is in a controlled or unrestricted area and that is not in storage was a performance deficiency (PD). The PD was more than minor because it was associated with the Program and Process attribute of the Public Radiation Safety Cornerstone and affected the cornerstone objective of ensuring adequate protection of public health and safety from exposure to radioactive materials released into the public domain. Using Manual Chapter 0609, Appendix D, Public Radiation Safety SDP, this finding determined to be was of very low safety significance (Green) in that the public radiation exposure was not greater than 0.005 rem (5 millirem).


The inspectors determined that cross-cutting issue H.4(b), "The licensee defines and effectively communicates expectations regarding procedural compliance and personnel follow procedures." was applicable for this violation because the RP Technician had failed to follow procedure HP-1033.148, "Canberra iSolo: Performance Checks", step 6.4.3 which states: "Ensure check source is removed from Canberra iSolo and returned to designated storage location".
The inspectors determined that cross-cutting issue H.4(b), The licensee defines and effectively communicates expectations regarding procedural compliance and personnel follow procedures. was applicable for this violation because the RP Technician had failed to follow procedure HP-1033.148, Canberra iSolo: Performance Checks, step 6.4.3 which states: Ensure check source is removed from Canberra iSolo and returned to designated storage location.


=====Enforcement:=====
=====Enforcement:=====
Section 10 CFR 20.1802,"Control of Material not in Storage", requires that the licensee control and maintain constant surveillance of licensed material that is in a controlled or unrestricted area and that is not in storage. Contrary to the above, on August 18, 2013, the licensee failed to maintain control of licensed radioactive material in an unrestricted area and not in storage, in that an Americium 241 check source was found in the iSolo alpha/beta counter in the Health Physics Technical Services area in the Administration Building. However, because the licensee documented this issue in its corrective action program (CR 523692) and because the violation is of very low safety  
Section 10 CFR 20.1802,Control of Material not in Storage, requires that the licensee control and maintain constant surveillance of licensed material that is in a controlled or unrestricted area and that is not in storage. Contrary to the above, on August 18, 2013, the licensee failed to maintain control of licensed radioactive material in an unrestricted area and not in storage, in that an Americium 241 check source was found in the iSolo alpha/beta counter in the Health Physics Technical Services area in the Administration Building. However, because the licensee documented this issue in its corrective action program (CR 523692) and because the violation is of very low safety significance, it is being treated as an NCV, consistent with Section 2.3.2.a of the NRC Enforcement Policy. (NCV 05000280, 05000281/2013-003, Failure to Maintain Control of Licensed Radioactive Material that was not in Storage).


significance, it is being treated as an NCV, consistent with Section 2.3.2.a of the NRC Enforcement Policy.  (NCV 05000280, 05000281/2013-003, Failure to Maintain Control of Licensed Radioactive Material that was not in Storage).
===Cornerstone: Emergency Preparedness===
 
===Cornerstone: Emergency Preparedness===


1EP4 Emergency Action Level and Emergency Plan Changes
1EP4 Emergency Action Level and Emergency Plan Changes


====a. Inspection Scope====
====a. Inspection Scope====
The NSIR headquarters staff performed an in-office review of the latest revisions of various Emergency Plan Implementing Pr ocedures (EPIPs) and the Emergency Plan located under ADAMS accession numbers ML123420028, ML13029A221, and ML130370771, as listed in the Attachment.
The NSIR headquarters staff performed an in-office review of the latest revisions of various Emergency Plan Implementing Procedures (EPIPs) and the Emergency Plan located under ADAMS accession numbers ML123420028, ML13029A221, and ML130370771, as listed in the Attachment.


The licensee determined that in accordance with 10 CFR 50.54(q), the changes made in the revisions resulted in no reduction in the effectiveness of the Plan, and that the revised Plan continued to meet the requirements of 10 CFR 50.47(b) and Appendix E to 10 CFR Part 50. The NRC review was not documented in a safety evaluation report and did not constitute approval of licensee-generated changes; therefore, these revisions aresubject to future inspection. Documents reviewed are listed in the Attachment. This inspection activity satisfied one inspection sample for the emergency action level and emergency plan changes on an annual basis.
The licensee determined that in accordance with 10 CFR 50.54(q), the changes made in the revisions resulted in no reduction in the effectiveness of the Plan, and that the revised Plan continued to meet the requirements of 10 CFR 50.47(b) and Appendix E to 10 CFR Part 50. The NRC review was not documented in a safety evaluation report and did not constitute approval of licensee-generated changes; therefore, these revisions aresubject to future inspection. Documents reviewed are listed in the Attachment. This inspection activity satisfied one inspection sample for the emergency action level and emergency plan changes on an annual basis.
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====a. Inspection Scope====
====a. Inspection Scope====
The inspectors performed a periodic review of the six following Unit 1 and 2 PIs to assess the accuracy and completeness of the submitted data and whether the performance indicators were calculated in accordance with the guidance contained in NEI 99-02, "Regulatory Assessment Performance Indicator Guideline," Revision 7. The inspection was conducted in accordance with NRC IP 71151, "Performance Indicator Verification."  Specifically, the inspectors reviewed the Unit 1 and Unit 2 data reported to the NRC for the period October 1, 2012, through September 30, 2013. Documents  
The inspectors performed a periodic review of the six following Unit 1 and 2 PIs to assess the accuracy and completeness of the submitted data and whether the performance indicators were calculated in accordance with the guidance contained in NEI 99-02, Regulatory Assessment Performance Indicator Guideline, Revision 7. The inspection was conducted in accordance with NRC IP 71151, Performance Indicator Verification. Specifically, the inspectors reviewed the Unit 1 and Unit 2 data reported to the NRC for the period October 1, 2012, through September 30, 2013. Documents reviewed included applicable NRC inspection reports, licensee event reports, operator logs, station performance indicators, and related CRs.
 
reviewed included applicable NRC inspection reports, licensee event reports, operator logs, station performance indicators, and related CRs.
* Unit 1 & 2 High Pressure Injection MSPI
* Unit 1 & 2 High Pressure Injection MSPI
* Unit 1 & 2 Residual Heat Removal MSPI
* Unit 1 & 2 Residual Heat Removal MSPI
Line 532: Line 495:


====a. Inspection Scope====
====a. Inspection Scope====
The inspectors reviewed the Occupational Exposure Control Effectiveness PI results for the Occupational Radiation Safety Cornerstone from May, 2012 through October, 2013. The inspectors also reviewed ED alarm logs and CRs related to controls for exposure significant areas.
The inspectors reviewed the Occupational Exposure Control Effectiveness PI results for the Occupational Radiation Safety Cornerstone from May, 2012 through October, 2013.
 
The inspectors also reviewed ED alarm logs and CRs related to controls for exposure significant areas.


====b. Findings====
====b. Findings====
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==4OA2 Identification and Resolution of Problems==
==4OA2 Identification and Resolution of Problems==


===.1 Daily Reviews of items Entered into the Corrective Action Program===
===.1 Daily Reviews of items Entered into the Corrective Action Program:===
:


====a. Inspection Scope====
====a. Inspection Scope====
As required by NRC 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 the licensee's CAP. This review was accomplished by reviewing daily CR report summaries and periodically attending daily CR Review Team meetings.
As required by NRC 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 the licensees CAP. This review was accomplished by reviewing daily CR report summaries and periodically attending daily CR Review Team meetings.


====b. Findings====
====b. Findings====
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====a. Inspection Scope====
====a. Inspection Scope====
The inspectors performed an in-depth review of the licensee's apparent cause analysis and corrective actions associated with CR 503764, 15H3 main generator output breaker for EDG #1 failure to operate (trip) from the Main Control Room during monthly surveillance testing. The inspectors assessed the licensee's problem identification threshold, cause analyses, extent of condition reviews, compensatory actions, and the prioritization and timeliness of the licensee's corrective actions to determine whether the licensee was appropriately identifying, characterizing, and correcting problems associated with this issue and whether the planned or completed corrective actions were appropriate. The inspectors compared the actions taken to the requirements of the licensee's CAP as specified in procedure, PI-AA-200, "Corrective Action Program,"
The inspectors performed an in-depth review of the licensees apparent cause analysis and corrective actions associated with CR 503764, 15H3 main generator output breaker for EDG #1 failure to operate (trip) from the Main Control Room during monthly surveillance testing. The inspectors assessed the licensees problem identification threshold, cause analyses, extent of condition reviews, compensatory actions, and the prioritization and timeliness of the licensees corrective actions to determine whether the licensee was appropriately identifying, characterizing, and correcting problems associated with this issue and whether the planned or completed corrective actions were appropriate. The inspectors compared the actions taken to the requirements of the licensees CAP as specified in procedure, PI-AA-200, Corrective Action Program, Revision 21 and 10 CFR 50, Appendix B. In addition, the inspectors reviewed the CAP for similar issues, and interviewed engineering personnel to assess the effectiveness of the implemented corrective actions.
 
Revision 21 and 10 CFR 50, Appendix B. In addition, the inspectors reviewed the CAP for similar issues, and interviewed engineering personnel to assess the effectiveness of the implemented corrective actions.


====b. Findings and observations====
====b. Findings and observations====
No findings were identified.
No findings were identified.


The licensee determined that the apparent cause of the failure was due to a failed conductor in the circuit breaker control cable. The conductor failure was attributed to mechanical failure of the cable jacket and conductor insulation due to post-installation stressors such as sharp bends in conduit and proximity to cable tray edges. The licensee's immediate corrective actions resulted in the identification and use of a spare conductor in the same cable to restore the breaker and the EDG to operability. Other corrective actions included further testing of the failed conductor to pinpoint the failure location, the replacement of the cable with a new cable that was routed separately, and the creation of a work order to remove the failed conductor cable and perform a failure analysis. The inspectors reviewed the corrective actions completed to date and noted that although further cable testing narrowed down the possible cable sections where the fault originates, the exact location could not be verified due to physical access restrictions. The inspectors noted that the corrective action to create a work order to remove the failed conductor cable and perform a failure analysis had been completed at the time of this inspection, but the work order was still in the planning stages. Inspectors also noted that this failure represented the second conductor in the same cable to fail within a seven month time period; a fact which supports the conclusion that a mechanical failure of the cable jacket is occurring somewhere along the routing path.
The licensee determined that the apparent cause of the failure was due to a failed conductor in the circuit breaker control cable. The conductor failure was attributed to mechanical failure of the cable jacket and conductor insulation due to post-installation stressors such as sharp bends in conduit and proximity to cable tray edges. The licensees immediate corrective actions resulted in the identification and use of a spare conductor in the same cable to restore the breaker and the EDG to operability. Other corrective actions included further testing of the failed conductor to pinpoint the failure location, the replacement of the cable with a new cable that was routed separately, and the creation of a work order to remove the failed conductor cable and perform a failure analysis. The inspectors reviewed the corrective actions completed to date and noted that although further cable testing narrowed down the possible cable sections where the fault originates, the exact location could not be verified due to physical access restrictions. The inspectors noted that the corrective action to create a work order to remove the failed conductor cable and perform a failure analysis had been completed at the time of this inspection, but the work order was still in the planning stages. Inspectors also noted that this failure represented the second conductor in the same cable to fail within a seven month time period; a fact which supports the conclusion that a mechanical failure of the cable jacket is occurring somewhere along the routing path.
 
The licensee did route a new EDG control cable during the Fall 2013 Unit 1 refueling outage. The inspectors reviewed the work packages and associated EDG testing to verify that the installation of the new cable and other corrective actions taken were effective. The inspectors did not identify any additional issues from this review. The inspectors determined the licensees evaluation of the issue appropriately identified the apparent and contributing causes. Additionally, the inspectors determined that the corrective actions developed as a result of the apparent cause analysis were reasonable and commensurate with the safety significance of the EDG system.


The licensee did route a new EDG control cable during the Fall 2013 Unit 1 refueling outage. The inspectors reviewed the work packages and associated EDG testing to verify that the installation of the new cable and other corrective actions taken were effective. The inspectors did not identify any additional issues from this review. The inspectors determined the licensee's evaluation of the issue appropriately identified the apparent and contributing causes. Additionally, the inspectors determined that the corrective actions developed as a result of the apparent cause analysis were reasonable and commensurate with the safety significance of the EDG system.
===.3 Annual Sample: Review of CR 508616, AAC (Alternate A/C) Mechanical Overspeed Trip===


===.3 Annual Sample: Review of CR 508616, AAC (Alternate A/C) Mechanical Overspeed Trip System Failed===
System Failed


====a. Inspection Scope====
====a. Inspection Scope====
The inspectors performed an in-depth review of the licensee's apparent cause analysis and corrective actions associated with CR 508616, AAC Mechanical Overspeed Trip System Failed. The inspectors assessed the licensee's problem identification threshold, cause analyses, extent of condition reviews, compensatory actions, and the prioritization and timeliness of the licensee's corrective actions to determine whether the licensee was appropriately identifying, characterizing, and correcting problems associated with this issue and whether the planned or completed corrective actions were appropriate. The inspectors compared the actions taken to the requirements of the licensee's CAP as specified in procedure, PI-AA-200, "Corrective Action Program," Revision 21 and 10 CFR 50, Appendix B. In addition, the inspectors reviewed the corrective action program for similar issues, and interviewed engineering personnel to assess the effectiveness of the implemented corrective actions.
The inspectors performed an in-depth review of the licensees apparent cause analysis and corrective actions associated with CR 508616, AAC Mechanical Overspeed Trip System Failed. The inspectors assessed the licensees problem identification threshold, cause analyses, extent of condition reviews, compensatory actions, and the prioritization and timeliness of the licensees corrective actions to determine whether the licensee was appropriately identifying, characterizing, and correcting problems associated with this issue and whether the planned or completed corrective actions were appropriate. The inspectors compared the actions taken to the requirements of the licensees CAP as specified in procedure, PI-AA-200, Corrective Action Program, Revision 21 and 10 CFR 50, Appendix B. In addition, the inspectors reviewed the corrective action program for similar issues, and interviewed engineering personnel to assess the effectiveness of the implemented corrective actions.


====b. Findings and observations====
====b. Findings and observations====
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The licensee concluded in ACE 019423 that the apparent cause of the AAC mechanical overspeed trip mechanism failure was likely an internal part failure in the Woodward Mechanical Overspeed Controller. The cause determination was limited due to the fact that the overspeed controller was not removed or disassembled during troubleshooting.
The licensee concluded in ACE 019423 that the apparent cause of the AAC mechanical overspeed trip mechanism failure was likely an internal part failure in the Woodward Mechanical Overspeed Controller. The cause determination was limited due to the fact that the overspeed controller was not removed or disassembled during troubleshooting.


Through discussions with plant engineering personnel at the time of the failure, the inspectors learned that there was no replac ement governor available for the AAC EDG and a Reasonable Assurance of Safety (RAS) was created to allow for long term operation without a functioning mechanical overspeed trip system. However, a part has since become available and current corrective actions are planned to replace the governor and restore the mechanical overspeed trip function during a maintenance package in mid-2014. The inspectors reviewed the CRs, RAS, and vendor documents associated with this failure and did not identify any regulatory concerns. However, the inspectors did note that the reliance on the electrical overspeed trip mechanism until mid-2014 represents a reduction in the original design safety margin of the AAC EDG.
Through discussions with plant engineering personnel at the time of the failure, the inspectors learned that there was no replacement governor available for the AAC EDG and a Reasonable Assurance of Safety (RAS) was created to allow for long term operation without a functioning mechanical overspeed trip system. However, a part has since become available and current corrective actions are planned to replace the governor and restore the mechanical overspeed trip function during a maintenance package in mid-2014. The inspectors reviewed the CRs, RAS, and vendor documents associated with this failure and did not identify any regulatory concerns. However, the inspectors did note that the reliance on the electrical overspeed trip mechanism until mid-2014 represents a reduction in the original design safety margin of the AAC EDG.


The inspectors also noted that the licensee self-identified the fact that the electrical overspeed function that is being relied upon had not previously been tested or verified to function at the correct setpoint. As a result, four year PM testing and eight year PM replacement activities were created. However, the inspectors noted that neither PM will be performed in the interim until the planned overspeed trip replacement activities in  
The inspectors also noted that the licensee self-identified the fact that the electrical overspeed function that is being relied upon had not previously been tested or verified to function at the correct setpoint. As a result, four year PM testing and eight year PM replacement activities were created. However, the inspectors noted that neither PM will be performed in the interim until the planned overspeed trip replacement activities in mid-2014.
 
mid-2014.


The inspectors determined that the corrective actions developed as a result of the apparent cause analysis were reasonable and commensurate with the safety significance of the AAC EDG system. The inspectors did not identify any additional issues from this review.
The inspectors determined that the corrective actions developed as a result of the apparent cause analysis were reasonable and commensurate with the safety significance of the AAC EDG system. The inspectors did not identify any additional issues from this review.


===.4 Annual Sample:===
===.4 Annual Sample: Review of Operator Work Arounds===
Review of Operator Work Arounds


====a. Inspection Scope====
====a. Inspection Scope====
The inspectors performed a review regarding the licensee's assessments and corrective actions for operator workarounds (OWAs). The inspectors reviewed the cumulative effects of the licensee's OWAs and licensee procedure OP-AA-1700, "Operations Aggregate Impact", Revision 6. The inspectors reviewed the data package associated with this procedure which included an evaluation of the cumulative effects of the OWAs on the operator's ability to safely operate the plant and effectively respond to abnormal and emergency plant conditions. The inspectors reviewed and monitored licensee planned and completed corrective actions to address underlying equipment issues causing the OWAs. The inspectors also evaluated OWAs against the requirements of the licensee's CAP as specified in PI-AA-200, "Corrective Action," Revisions 21, 10 CFR 50, Appendix B, and OP-AA-100, "Conduct of Operations," Revision 25.
The inspectors performed a review regarding the licensees assessments and corrective actions for operator workarounds (OWAs). The inspectors reviewed the cumulative effects of the licensees OWAs and licensee procedure OP-AA-1700, Operations Aggregate Impact, Revision 6. The inspectors reviewed the data package associated with this procedure which included an evaluation of the cumulative effects of the OWAs on the operators ability to safely operate the plant and effectively respond to abnormal and emergency plant conditions. The inspectors reviewed and monitored licensee planned and completed corrective actions to address underlying equipment issues causing the OWAs. The inspectors also evaluated OWAs against the requirements of the licensees CAP as specified in PI-AA-200, Corrective Action, Revisions 21, 10 CFR 50, Appendix B, and OP-AA-100, "Conduct of Operations," Revision 25.


====b. Findings and Observations====
====b. Findings and Observations====
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These observations took place during both normal and off-normal plant working hours.
These observations took place during both normal and off-normal plant working hours.


These quarterly resident inspector observations of security force personnel and activities did not constitute any additional inspection samples. Rather, they were considered an integral part of the inspectors' normal plant status review and inspection activities.
These quarterly resident inspector observations of security force personnel and activities did not constitute any additional inspection samples. Rather, they were considered an integral part of the inspectors normal plant status review and inspection activities.


====b. Findings====
====b. Findings====
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=====Exit Meeting Summary=====
=====Exit Meeting Summary=====
On January 29, 2014, the inspection results were presented to Mr. R. Simmons and other members of his staff, who acknowledged the findings. The inspectors asked the licensee whether any of the material examined during the inspection should be  
On January 29, 2014, the inspection results were presented to Mr. R. Simmons and other members of his staff, who acknowledged the findings. The inspectors asked the licensee whether any of the material examined during the inspection should be considered proprietary.
 
considered proprietary.


{{a|4OA7}}
{{a|4OA7}}
Line 621: Line 580:


The following finding of very low significance (Green) was identified by the licensee and is a violation of NRC requirements which meets the criteria of the NRC Enforcement Policy, for being dispositioned as an NCV.
The following finding of very low significance (Green) was identified by the licensee and is a violation of NRC requirements which meets the criteria of the NRC Enforcement Policy, for being dispositioned as an NCV.
* 10 CFR 50, Appendix B, Criterion III, "Design Control," requires, in part, that measures shall be established to assure that applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures, and instructions and that deviations from such standards are controlled. Contrary to the above, on October 8, 2013, the licensee discovered a failure to control the angle of the counterweight arms on Unit 1 and Unit 2 containment spray (CS) and recirculation spray (RS) check valves, CS-13, CS-24, RS-11 and RS-17 such that the arms were no longer vertical when the valves were full open. In addition, no documented evaluation of the valves to open fully under design flow conditions could be found. Further evaluation required the licensee to change the Unit 2 acceptance criteria for the CS pump integrated service test (IST) to increase the minimum required pump differential pressure that is required to meet the design basis flow requirements.
* 10 CFR 50, Appendix B, Criterion III, "Design Control, requires, in part, that measures shall be established to assure that applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures, and instructions and that deviations from such standards are controlled. Contrary to the above, on October 8, 2013, the licensee discovered a failure to control the angle of the counterweight arms on Unit 1 and Unit 2 containment spray (CS) and recirculation spray (RS) check valves, CS-13, CS-24, RS-11 and RS-17 such that the arms were no longer vertical when the valves were full open. In addition, no documented evaluation of the valves to open fully under design flow conditions could be found. Further evaluation required the licensee to change the Unit 2 acceptance criteria for the CS pump integrated service test (IST) to increase the minimum required pump differential pressure that is required to meet the design basis flow requirements.


This finding is of very low safety significance (Green) because the position of the counterweight arms did not affect the design or qualification of the CS system and it did not represent a loss of system or train safety function.
This finding is of very low safety significance (Green) because the position of the counterweight arms did not affect the design or qualification of the CS system and it did not represent a loss of system or train safety function. The position of the Unit 1 counterweight arms was corrected during the fall 2013 RFO. This issue has been entered into the licensees CAP as CR 528665.


The position of the Unit 1 counterweight arms was corrected during the fall 2013 RFO. This issue has been entered into the licensee's CAP as CR 528665.
ATTACHMENT:  
 
ATTACHMENT:


=SUPPLEMENTAL INFORMATION=
=SUPPLEMENTAL INFORMATION=
Line 633: Line 590:
SUPPLEMENTAL INFORMATION
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
KEY POINTS OF CONTACT
===Licensee Personnel===
===Licensee Personnel===
: [[contact::L. Baker]], Manager Training  
: [[contact::L. Baker]], Manager Training
: [[contact::J. Eggart]], Manager, Radiation Protection & Chemistry  
: [[contact::J. Eggart]], Manager, Radiation Protection & Chemistry
: [[contact::B. Garber]], Supervisor, Station Licensing  
: [[contact::B. Garber]], Supervisor, Station Licensing
: [[contact::P. Harris]], Supervisor Radiological Analysis and Instrumentation  
: [[contact::P. Harris]], Supervisor Radiological Analysis and Instrumentation
: [[contact::A. Harrow]], Manager, Organizational Effectiveness  
: [[contact::A. Harrow]], Manager, Organizational Effectiveness
: [[contact::J. Henderson]], Engineering Manager  
: [[contact::J. Henderson]], Engineering Manager
: [[contact::J. Hopkins]], Supervisor Rad Material Control  
: [[contact::J. Hopkins]], Supervisor Rad Material Control
: [[contact::L. Hilbert ]], Manager, Outage and Planning  
: [[contact::L. Hilbert ]], Manager, Outage and Planning
: [[contact::R. Johnson]], Manager, Operations  
: [[contact::R. Johnson]], Manager, Operations
: [[contact::L. Lane]], Site Vice President  
: [[contact::L. Lane]], Site Vice President
: [[contact::D. Lawrence]], Director, Station Safety and Licensing  
: [[contact::D. Lawrence]], Director, Station Safety and Licensing
: [[contact::T. Mayer]], Steam Generator Program Owner  
: [[contact::T. Mayer]], Steam Generator Program Owner
: [[contact::C. Olsen]], Director, Station Engineering  
: [[contact::C. Olsen]], Director, Station Engineering
: [[contact::J. Pollard]], Licensing Engineer  
: [[contact::J. Pollard]], Licensing Engineer
: [[contact::L. Ragland]], Supervisor Radiation Protection Operations  
: [[contact::L. Ragland]], Supervisor Radiation Protection Operations
: [[contact::R. Scanlan]], Manager, Maintenance  
: [[contact::R. Scanlan]], Manager, Maintenance
: [[contact::K. Sloane]], Plant Manager  
: [[contact::K. Sloane]], Plant Manager
: [[contact::M. Smith]], Manager, Nuclear Oversight  
: [[contact::M. Smith]], Manager, Nuclear Oversight
: [[contact::M. True]], ISI/BACCP Site Owner  
: [[contact::M. True]], ISI/BACCP Site Owner
: [[contact::E. Turko]], ISI/NDE Supervisor  
: [[contact::E. Turko]], ISI/NDE Supervisor
: [[contact::N. Turner]], Supervisor, Emergency Preparedness  
: [[contact::N. Turner]], Supervisor, Emergency Preparedness
: [[contact::C. Wray]], NDE Coordinator
: [[contact::C. Wray]], NDE Coordinator
LIST OF ITEMS OPENED, CLOSED AND DISCUSSED
LIST OF ITEMS OPENED, CLOSED AND DISCUSSED
Opened
Opened
05000280/2013005-01               URI Application of ASME Section XI, Table IWB 2500-1, Item B10.10, Inspection Requirements and Note 1 Exemptions (Section 1R08)
05000280/2013005-01             URI         Application of ASME Section XI, Table IWB 2500-1,
Item B10.10, Inspection Requirements and Note 1
Exemptions (Section 1R08)
Opened and Closed
Opened and Closed
05000280/2012005-02               NCV Failure to Missile Protect Beyond Design Bases FLEX Modification to Low Head Safety Injection
05000280/2012005-02             NCV         Failure to Missile Protect Beyond Design Bases
FLEX Modification to Low Head Safety Injection
Piping (Section 1R18)
Piping (Section 1R18)
05000280, 281/2012005-03       NCV Failure to Maintain Control of Licensed Radioactive Material that was not in Storage (Section 2RS8)  
05000280, 281/2012005-03         NCV         Failure to Maintain Control of Licensed Radioactive
 
Material that was not in Storage (Section 2RS8)
List of Documents Reviewed
List of Documents Reviewed
Section 1R01: Adverse Weather Protection
Section 1R01: Adverse Weather Protection
Procedures
Procedures
0-EPM-1303-01, Freeze Protection Inspection, Rev. 20
0-EPM-1303-01, Freeze Protection Inspection, Rev. 20
0-OP-ZZ-021, Severe Weather Preparation, Rev. 2
0-OP-ZZ-021, Severe Weather Preparation, Rev. 2
0-OSP-ZZ-001, Cold Weather Preparation, Rev. 14
0-OSP-ZZ-001, Cold Weather Preparation, Rev. 14
Condition Reports (*NRC Identified) 500113     532539     532542     532543     532544    532545    533512     533537     534776
Condition Reports (*NRC Identified)
534779*   535462
500113 532539 532542 532543               532544    532545    533512 533537 534776
Section 1R04: Equipment Alignment
534779* 535462
Section 1R04: Equipment Alignment
Procedures
Procedures
0-MOP-AAC-002, Return to Service of the AAC Diesel Generator, Rev. 19
0-MOP-AAC-002, Return to Service of the AAC Diesel Generator, Rev. 19
0-MOP-AAC-001, Removal from Service of the AAC Diesel Generator, Rev. 17
0-MOP-AAC-001, Removal from Service of the AAC Diesel Generator, Rev. 17
0-OP-AAC-001A, AAC Diesel Generator Systems Alignment, Rev. 7 0-OP-FC-001A, Spent Fuel Pit Cooling System Alignment, Rev. 7 1-OP-CS-001A, Containment Spray System Alignment, Rev. 7 1-OPT-FW-003, Turbine Driven Auxiliary Feedwater Pump 1-FW-P-2, Rev. 44  
0-OP-AAC-001A, AAC Diesel Generator Systems Alignment, Rev. 7
 
0-OP-FC-001A, Spent Fuel Pit Cooling System Alignment, Rev. 7
Drawings 11448-FM-068A SH 3, Flow/Valve Operating Number Diagram Feedwater System Surry Unit 1, Rev. 60 11448-FM-084A SH 2, Flow/Valve Operating Number Diagram Containment Spray System   Unit 1, Rev. 49
1-OP-CS-001A, Containment Spray System Alignment, Rev. 7
 
1-OPT-FW-003, Turbine Driven Auxiliary Feedwater Pump 1-FW-P-2, Rev. 44
Condition Reports (*NRC Identified) 532794    533114*
Drawings
Section 1R05: Fire Protection
11448-FM-068A SH 3, Flow/Valve Operating Number Diagram Feedwater System Surry Unit 1,
Rev. 60
11448-FM-084A SH 2, Flow/Valve Operating Number Diagram Containment Spray System
Unit 1, Rev. 49
Condition Reports (*NRC Identified)
2794 533114*
Section 1R05: Fire Protection
Procedures
Procedures
0-FS-FP-161, Auxiliary Building - Elevation 27 Feet - 6 Inches, Rev. 2
0-FS-FP-161, Auxiliary Building - Elevation 27 Feet - 6 Inches, Rev. 2
0-FS-FP-116, Control Room - Elevation 27 Feet - 6 Inches, Rev. 4
0-FS-FP-116, Control Room - Elevation 27 Feet - 6 Inches, Rev. 4
1-FS-FP-109, Battery Room 1A - Unit 1 Elevation 27 Feet - 6 Inches, Rev. 2 2-FS-FP-168, Turbine Basement - Unit 2 Elevation 9 Feet - 6 Inches, Rev. 1
1-FS-FP-109, Battery Room 1A - Unit 1 Elevation 27 Feet - 6 Inches, Rev. 2
Condition Reports (*NRC Identified) 535598*     535607*     535609*     535338     535490*  
2-FS-FP-168, Turbine Basement - Unit 2 Elevation 9 Feet - 6 Inches, Rev. 1
 
Condition Reports (*NRC Identified)
Section 1R06: Flood Protection
535598* 535607* 535609* 535338               535490*
Section 1R06: Flood Protection
Procedures
Procedures
0-AP-13.00, Turbine Building or MER 3 Flooding, Rev. 28
0-AP-13.00, Turbine Building or MER 3 Flooding, Rev. 28
Line 695: Line 662:
1-MOP-SW-003, Operation of the Service Water Jumper, Rev. 2
1-MOP-SW-003, Operation of the Service Water Jumper, Rev. 2
Condition Reports (*NRC Identified)
Condition Reports (*NRC Identified)
530896*     531322
530896* 531322
Attachment
 
Other Documents
Other Documents
2-615, Virginia Electric and Power Company Surry Power Station Units 1 and 2 Proposed License Amendment Request Regarding Temporary
2-615, Virginia Electric and Power Company Surry Power Station Units 1 and 2 Proposed
Service Water Jumper to the Component Cooling Heat Exchangers, 09/26/12 Safety Evaluation Related to Amendment No. 279 to Renewed Facility Operating License Nos. DPR-32 and DPR-37 Virginia Electric and Power Company Surry Power Station, Unit Nos. 1
License Amendment Request Regarding Temporary Service Water Jumper to the Component
Cooling Heat Exchangers, 09/26/12
Safety Evaluation Related to Amendment No. 279 to Renewed Facility Operating License Nos.
DPR-32 and DPR-37 Virginia Electric and Power Company Surry Power Station, Unit Nos. 1
and 2 Docket Nos. 50-280 and 50-281, 09/23/13
and 2 Docket Nos. 50-280 and 50-281, 09/23/13
Section 1R07: Heat Sink Performance
Section 1R07: Heat Sink Performance
Procedures
Procedures
0-NSP-CC-005, CCHX Tests Using the Temporary Monitoring System, Rev. 1
0-NSP-CC-005, CCHX Tests Using the Temporary Monitoring System, Rev. 1
Condition Reports
Condition Reports
530290     528965     529922     530896    532319     533016    533442  
530290 528965         529922   530896    532319 533016    533442
 
Other Documents
Other Documents
VEPCO Heat Exchanger Specification Sheet for Component Cooling Water Heat Exchanger Proto-HX Heat Exchanger Calculation Reports for Surry Unit 1 RFO 25 ETE-SU-2012-1016, Component Cooling Water Heat Exchanger Performance Testing
VEPCO Heat Exchanger Specification Sheet for Component Cooling Water Heat Exchanger
Section 1R08: Inservice Inspection Activities
Proto-HX Heat Exchanger Calculation Reports for Surry Unit 1 RFO 25
ETE-SU-2012-1016, Component Cooling Water Heat Exchanger Performance Testing
Section 1R08: Inservice Inspection Activities
Procedures
Procedures
0-NSP-RC-003, Visual examination of Reactor Pressure Vessel Bottom Mounted Instrumentation (BMI), Surry Unit 1 & 2, Rev. 2 54-ISI-370-003, Nondestructive Examination Procedure, Areva NP, Remote Underwater Visual examination of Westinghouse Reactor Pressure Vessel Internals for Pressurized Water
0-NSP-RC-003, Visual examination of Reactor Pressure Vessel Bottom Mounted
Reactors in accordance with MRP-228 (Inspection Standard for PWR Internals) Areva 03-9177821, Secondary Side Visual Inspection Plan and Procedure for Dominion, Surry 1R25, Rev. 3 Areva ETSS_BOB001_MIZ80_R0, Surry Unit 1, Outage 1R25, 10/25/13
Instrumentation (BMI), Surry Unit 1 & 2, Rev. 2
54-ISI-370-003, Nondestructive Examination Procedure, Areva NP, Remote Underwater Visual
examination of Westinghouse Reactor Pressure Vessel Internals for Pressurized Water
Reactors in accordance with MRP-228 (Inspection Standard for PWR Internals)
Areva 03-9177821, Secondary Side Visual Inspection Plan and Procedure for Dominion, Surry
1R25, Rev. 3
Areva ETSS_BOB001_MIZ80_R0, Surry Unit 1, Outage 1R25, 10/25/13
Areva ETSS_RPC001_MIZ80_R0, Surry Unit 1, Outage 1R25, 10/25/13
Areva ETSS_RPC001_MIZ80_R0, Surry Unit 1, Outage 1R25, 10/25/13
Areva ETSS_XP001_2X19_MIZ80_R0, Surry Unit 1, Outage 1R25, 10/25/13
Areva ETSS_XP001_2X19_MIZ80_R0, Surry Unit 1, Outage 1R25, 10/25/13
SRY-SGPMS-002, Surry Site Specific Eddy Current Analysis Guidelines, Surry 1R25,   October 2013
SRY-SGPMS-002, Surry Site Specific Eddy Current Analysis Guidelines, Surry 1R25,
October 2013
Engineering/Technical Evaluations
Engineering/Technical Evaluations
Areva Document 51-9210968-000, Surry Unit 1 1R25 - EPRI Appendix H/I Eddy Current Technique Review, Rev. 0 ETE-SU-2010-0029, Condition Monitoring and Operational Assessment Unit 1 CMOA Fall 2010, Rev. 0 ETE-SU-2012-0011, Steam Generator Condition
Areva Document 51-9210968-000, Surry Unit 1 1R25 - EPRI Appendix H/I Eddy Current
Monitoring and Operational Assessment Surry Unit 1, Rev. 0 ETE-SU-2013-0052, Steam Generator Degradation Assessment, Rev. 0  
Technique Review, Rev. 0
ETE-SU-2010-0029, Condition Monitoring and Operational Assessment Unit 1 CMOA Fall 2010,
Rev. 0
ETE-SU-2012-0011, Steam Generator Condition Monitoring and Operational Assessment Surry
Unit 1, Rev. 0
ETE-SU-2013-0052, Steam Generator Degradation Assessment, Rev. 0
Drawings
B&W Drawing No. 134816E, Primary Inlet Nozzle, Section A-A, Rev. 3
DWG No. 1B79662, Calibration Standard ADVB-031-96, Rev. 1
DWG No. 1B79667, Calibration Standard ADVB-036-96, Rev. 1


Drawings B&W Drawing No. 134816E, Primary Inlet Nozzle, Section A-A, Rev. 3 DWG No. 1B79662, Calibration Standard ADVB-031-96, Rev. 1
DWG No. 1B80274, Calibration Standard EP5-011-098, 02/28/98
DWG No. 1B79667, Calibration Standard ADVB-036-96, Rev. 1
DWG No. 1B80279, Calibration Standard EP5-016-098, 02/28/98
Attachment DWG No. 1B80274, Calibration Standard EP5-011-098, 02/28/98 DWG No. 1B80279, Calibration Standard EP5-016-098, 02/28/98 DWG No. 1B81049, Eddy Current Sizing Standards Assembly & Detail, Rev. 0
DWG No. 1B81049, Eddy Current Sizing Standards Assembly & Detail, Rev. 0
DWG No. 9103696B, Surry .875 X 0.50 ASME-AVB-Expansion Cal Std, Rev. 0
DWG No. 9103696B, Surry .875 X 0.50 ASME-AVB-Expansion Cal Std, Rev. 0
DWG No. 9103697B, Surry .875 X 0.50 ASME-AVB-Expansion Cal Std, Rev. 0
DWG No. 9103697B, Surry .875 X 0.50 ASME-AVB-Expansion Cal Std, Rev. 0
DWG No. 9103698B, Surry .875 X 0.50 EDM Cal Std As Built Drawing, Rev. 0
DWG No. 9103698B, Surry .875 X 0.50 EDM Cal Std As Built Drawing, Rev. 0
DWG No. 9103699B, Surry .875 X 0.50 EDM Cal Std As Built Drawing, Rev. 0 DWG No. CB02221.DWG, PDI Alternative ASME Calibration Block, Rev. 0 Stone and Webster Drawing No. 11448-FV-7A, Reactor Neutron Shield Tank Assembly, Rev. 6
DWG No. 9103699B, Surry .875 X 0.50 EDM Cal Std As Built Drawing, Rev. 0
Stone and Webster Drawing No. 11448-FV-7D, Reactor Neutron Shield Tank Sheet 3, Reactor Sliding Foot Assembly Section, Rev. 6
DWG No. CB02221.DWG, PDI Alternative ASME Calibration Block, Rev. 0
Stone and Webster Drawing No. 11448-FV-7A, Reactor Neutron Shield Tank Assembly, Rev. 6
Stone and Webster Drawing No. 11448-FV-7D, Reactor Neutron Shield Tank Sheet 3, Reactor
Sliding Foot Assembly Section, Rev. 6
Corrective Action Documents
Corrective Action Documents
CR 473993, Boric Acid on 1-RC-PCV-1455A (Body to Bonnet)
CR 473993, Boric Acid on 1-RC-PCV-1455A (Body to Bonnet)
CR 474000, Boric Acid on 1-SI-85 (Body to Bonnet)
CR 474000, Boric Acid on 1-SI-85 (Body to Bonnet)
CR 474220, ISI Rejection of Integral Attachment Weld Due to Incomplete Weld
CR 474220, ISI Rejection of Integral Attachment Weld Due to Incomplete Weld
CR 474792, Improper Thread Engagement Was Noted on Anchor Bolt for Constant Support CR 474284, ISI Rejection of Pipe Support for Structural Corrosion and Thread Engagement CR 474808, ISI Rejection of Pipe Support for Incorrect Clearances, Corrosion, and Weld Loss
CR 474792, Improper Thread Engagement Was Noted on Anchor Bolt for Constant Support
CR 474284, ISI Rejection of Pipe Support for Structural Corrosion and Thread Engagement
CR 474808, ISI Rejection of Pipe Support for Incorrect Clearances, Corrosion, and Weld Loss
CR 474875, Improper Clearances and Missing Field Welds on Pipe Hanger
CR 474875, Improper Clearances and Missing Field Welds on Pipe Hanger
CR 476209, Steam Generator Riser Barrel Discoloration
CR 476209, Steam Generator Riser Barrel Discoloration
CR 476520, Foreign Object Identified During FOSAR Inspection
CR 476520, Foreign Object Identified During FOSAR Inspection
CR478573, Boric Acid on 1-CH-P-1C Inboard and Outboard Seals CR 489853, SGMP has issued a Needed Guidance per NEI 03-08 CR 496300, ACE 019326, Foreign Object Identified in 2-RC-E-1A
CR478573, Boric Acid on 1-CH-P-1C Inboard and Outboard Seals
CR 489853, SGMP has issued a Needed Guidance per NEI 03-08
CR 496300, ACE 019326, Foreign Object Identified in 2-RC-E-1A
CR 501733, Submit WO to Perform 1-RC-E-1B Steam Drum Inspection for FME
CR 501733, Submit WO to Perform 1-RC-E-1B Steam Drum Inspection for FME
CR 502848, Submit WO to Perform 1-RC-E-1C Steam Drum Inspection for FME
CR 502848, Submit WO to Perform 1-RC-E-1C Steam Drum Inspection for FME
CR 529738, Boric Acid Leaking from 1-SI-495 (Drain Valve) (Active) CR 529764, Boric Acid Discovered on 1-SI-102 (Packing) (Inactive) CR 529775, Boric Acid on 1-RC-P-1A (Seal and Flange)
CR 529738, Boric Acid Leaking from 1-SI-495 (Drain Valve) (Active)
CR 529764, Boric Acid Discovered on 1-SI-102 (Packing) (Inactive)
CR 529775, Boric Acid on 1-RC-P-1A (Seal and Flange)
CR 529783, Boric Acid Leakage on 01-RC-HCV-1557C
CR 529783, Boric Acid Leakage on 01-RC-HCV-1557C
CR 529802, Boric Acid Identified on 1-RC-ICV-3026. (Packing/Bonnet) (Inactive)
CR 529802, Boric Acid Identified on 1-RC-ICV-3026. (Packing/Bonnet) (Inactive)
CR 529813, Boric Acid Identified on 1-RC-ICV-3085 (Fitting) (Inactive)
CR 529813, Boric Acid Identified on 1-RC-ICV-3085 (Fitting) (Inactive)
CR 529823, Boric Acid Identified in Drain Catch Downstream of 1-RC-ICV-3066/3067 CR 529937, Boric Acid Identified on 1-RC-ICV-3142 (Packing) (Active) CR 530267, Loose bolt on Rx Support
CR 529823, Boric Acid Identified in Drain Catch Downstream of 1-RC-ICV-3066/3067
CR 529937, Boric Acid Identified on 1-RC-ICV-3142 (Packing) (Active)
CR 530267, Loose bolt on Rx Support
CR 530542, ISI Rejectable Conditions Noted on Spring Support 1-WFPD-H006B
CR 530542, ISI Rejectable Conditions Noted on Spring Support 1-WFPD-H006B
CR 530547, ISI Noted Condition on Spring Support 1-WFPD-H006A
CR 530547, ISI Noted Condition on Spring Support 1-WFPD-H006A
CR 532007, Corrosion Observed on 1-RC-FC-1482C CR 532305, Boric Acid on 1-RC-FC-1480B (Plug) CR 533049, Open Item from NRC ISI Inspection  
CR 532007, Corrosion Observed on 1-RC-FC-1482C
 
CR 532305, Boric Acid on 1-RC-FC-1480B (Plug)
CR 533049, Open Item from NRC ISI Inspection
Other Documents
Other Documents
Areva Certificate of Calibration 39936, Eddy Current Tester MIZ-80 S/N 024, 08/15/13
Areva Certificate of Calibration 39936, Eddy Current Tester MIZ-80 S/N 024, 08/15/13
Areva Certificate of Calibration 39938, Eddy Current Tester MIZ-80 S/N 051, 08/15/13 Areva Certificate of Calibration 39939, Eddy Current Tester MIZ-80 S/N 059, 08/15/13 Areva Certificate of Calibration 40136, Eddy Current Tester MIZ-80 S/N 011, 08/28/13
Areva Certificate of Calibration 39938, Eddy Current Tester MIZ-80 S/N 051, 08/15/13
Areva Certificate of Calibration 39939, Eddy Current Tester MIZ-80 S/N 059, 08/15/13
Areva Certificate of Calibration 40136, Eddy Current Tester MIZ-80 S/N 011, 08/28/13
Areva Document 51-9208139-000, Surry Unit 1 - 1R25 ECT Inspection Plan, Rev. 0
Areva Document 51-9208139-000, Surry Unit 1 - 1R25 ECT Inspection Plan, Rev. 0
Attachment Areva S1R25-Memo-001, Surry 1 1R25 Qualified Data Analysts, 10/26/13 ASME Boiler and Pressure Vessel Code Case, N-722-1, ASME Approved date, 01/26/09 CEM-0142, Sensitivity Evaluation of the Structural Adequacy of the Reactor Vessel Sliding Foot Assemblies for the loss of Multiple Cap Screws, SPS Unit-1, 11/15/13 Certificate of Visual Examination, ID Number C6756, Expires August 5, 2014
 
Areva S1R25-Memo-001, Surry 1 1R25 Qualified Data Analysts, 10/26/13
ASME Boiler and Pressure Vessel Code Case, N-722-1, ASME Approved date, 01/26/09
CEM-0142, Sensitivity Evaluation of the Structural Adequacy of the Reactor Vessel Sliding Foot
Assemblies for the loss of Multiple Cap Screws, SPS Unit-1, 11/15/13
Certificate of Visual Examination, ID Number C6756, Expires August 5, 2014
Certificate of Visual Examination, ID Number Y4315, Expires August 26, 2014
Certificate of Visual Examination, ID Number Y4315, Expires August 26, 2014
Certification ID Number CY4315, Areva Incorporated, Certificate of Personnel Qualification, Rev. 30 Certification ID Number C6756, Areva Incorporated, Certificate of Personnel Qualification,
Certification ID Number CY4315, Areva Incorporated, Certificate of Personnel Qualification,
Rev. 30
Certification ID Number C6756, Areva Incorporated, Certificate of Personnel Qualification,
Rev. 23
Rev. 23
Certified Materials Test Report for Spool arc 65, Class ER70S-2 weld wire for GTAW
Certified Materials Test Report for Spool arc 65, Class ER70S-2 weld wire for GTAW
Dominion Corporate Welding Manual, Performance Qualification Test Methods, Performance Qualification Method No. 501A, GTAW ER-AP-BAC-10, Boric Acid Corrosion Control Program, Rev. 10
Dominion Corporate Welding Manual, Performance Qualification Test Methods, Performance
ER-AP-BAC-101, Boric Acid Corrosion Control Program (BACCP) Inspections, Rev. 9 ER-AP-BAC-102, Boric Acid Corrosion Control Program (BACCP) Evaluations, Rev. 10 IOD-000191, Immediate Operability Determination for CR 530267, 11/15/13 Inservice Inspection Owner's Activity Report for Surry Unit 1 Refueling Outage S1R24 NDE Personnel Certification Records:
Qualification Method No. 501A, GTAW
ER-AP-BAC-10, Boric Acid Corrosion Control Program, Rev. 10
ER-AP-BAC-101, Boric Acid Corrosion Control Program (BACCP) Inspections, Rev. 9
ER-AP-BAC-102, Boric Acid Corrosion Control Program (BACCP) Evaluations, Rev. 10
IOD-000191, Immediate Operability Determination for CR 530267, 11/15/13
Inservice Inspection Owners Activity Report for Surry Unit 1 Refueling Outage S1R24
NDE Personnel Certification Records:  
: [[contact::B. Knox]],  
: [[contact::B. Knox]],  
: [[contact::K. Newell]],  
: [[contact::K. Newell]],  
Line 764: Line 778:
: [[contact::M. Story]],  
: [[contact::M. Story]],  
: [[contact::M. Brown]],  
: [[contact::M. Brown]],  
: [[contact::P. Anderson]],  
: [[contact::P.
Anderson]],  
: [[contact::J. Jacobs]],  
: [[contact::J. Jacobs]],  
: [[contact::A. Jensen]],  
: [[contact::A. Jensen]],  
Line 770: Line 785:
: [[contact::B. Zollner]],  
: [[contact::B. Zollner]],  
: [[contact::T. Thomas]],  
: [[contact::T. Thomas]],  
: [[contact::D. Strickland]],  
: [[contact::D. Strickland]], M. True
: [[contact::M. True NDE Personnel Qualification and Certification Record for VT Examinations for leakage on lower head in accordance with N-722]], 2013 Lower Head Examiner, 3/12/2013 Program Health Report, Steam Generator, Period Q3-2013, 10/23/2013 SAR000480, Boric Acid Corrosion Control Program (BACCP) Formal Self-Assessment, 10/25/10 OPEX002605, IN 2010-05 Management of Steam Generator Loose Parts Automated Eddy Current Data Analysis, 02/11/10 OPEX002748, IN 10-21 Crack-Like Indication in the U-bend Region of a Thermally Treated Alloy 600 Steam Generator Tube, 10/15/10 OPEX003137, IN 12-07 Tube-to-Tube Contact Resulting in Wear in Once-Through Steam
NDE Personnel Qualification and Certification Record for VT Examinations for leakage on lower
Generators, 08/01/12 OPEX003176, IN 13-11 Crack-Like Indications at Dents/Dings and in the Free span Region of Thermally Treated Alloy 600 Steam Generator Tubes, 7/11/2013 SAR-000841, Welder Qualification, 09/29/09 Surry Power Station Unit-1, October 2013 Maintenance Outage Reactor Internals Inspection Schedule, 10/09/13 WCAP-15988-NP, Generic Guidance for an Effective Boric Acid Inspection Program for Pressurized Water Reactors, Rev. 2 Welding Technique Sheet for Welding Technique Number 103, GTAW Manual Welding Zetec Certificate of Authenticity, Areva/Surry 1, 10/02/13 Zetec Certificate of Conformance, Shipment ID 14143, Contract 1013053297, 10/10/13
head in accordance with N-722, 2013 Lower Head Examiner, 3/12/2013
Program Health Report, Steam Generator, Period Q3-2013, 10/23/2013
SAR000480, Boric Acid Corrosion Control Program (BACCP) Formal Self-Assessment,
10/25/10
OPEX002605, IN 2010-05 Management of Steam Generator Loose Parts Automated Eddy
Current Data Analysis, 02/11/10
OPEX002748, IN 10-21 Crack-Like Indication in the U-bend Region of a Thermally Treated
Alloy 600 Steam Generator Tube, 10/15/10
OPEX003137, IN 12-07 Tube-to-Tube Contact Resulting in Wear in Once-Through Steam
Generators, 08/01/12
OPEX003176, IN 13-11 Crack-Like Indications at Dents/Dings and in the Free span Region of
Thermally Treated Alloy 600 Steam Generator Tubes, 7/11/2013
SAR-000841, Welder Qualification, 09/29/09
Surry Power Station Unit-1, October 2013 Maintenance Outage Reactor Internals Inspection
Schedule, 10/09/13
WCAP-15988-NP, Generic Guidance for an Effective Boric Acid Inspection Program for
Pressurized Water Reactors, Rev. 2
Welding Technique Sheet for Welding Technique Number 103, GTAW Manual Welding
Zetec Certificate of Authenticity, Areva/Surry 1, 10/02/13
Zetec Certificate of Conformance, Shipment ID 14143, Contract 1013053297, 10/10/13
Zetec Certificate of Conformance, Shipment ID 14366, Contract 1013061214, 09/26/13
Zetec Certificate of Conformance, Shipment ID 14366, Contract 1013061214, 09/26/13
Section 1R11: Licensed Operator Requalification Program
Section 1R11: Licensed Operator Requalification Program
Procedures
Procedures
1-GOP-2.7, Unit Shutdown, Power Decrease from Allowable Power to Unit Offline for Refueling Outage, Rev. 21 1-OP-RC-004, Draining the RCS to Reactor Flange Level, Rev. 34
1-GOP-2.7, Unit Shutdown, Power Decrease from Allowable Power to Unit Offline for Refueling
Attachment 1-OP-RX-009, Dilution to Critical Conditions Following Refueling, Rev 21 OP-AA-100, Conduct of Operations, Rev. 25
Outage, Rev. 21
Section 1R12: Maintenance Effectiveness
1-OP-RC-004, Draining the RCS to Reactor Flange Level, Rev. 34
 
1-OP-RX-009, Dilution to Critical Conditions Following Refueling, Rev 21
OP-AA-100, Conduct of Operations, Rev. 25
Section 1R12: Maintenance Effectiveness
Procedures
Procedures
0-MPT-0704-01, Diesel Fire Pump Engine Inspection, Rev. 17
0-MPT-0704-01, Diesel Fire Pump Engine Inspection, Rev. 17
0-OPT-FP-009, Diesel Driven Fire Protection Water Pump 1-FP-P-2, Rev. 22 CM-AA-FPA-100, Fire Protection Appendix R Program, Rev. 8 ER-AA-MRL-10, Maintenance Rule Program, Rev. 5
0-OPT-FP-009, Diesel Driven Fire Protection Water Pump 1-FP-P-2, Rev. 22
ER-AA-MRL-100, Implementing Maintenance Rule, Rev. 6  
CM-AA-FPA-100, Fire Protection Appendix R Program, Rev. 8
 
ER-AA-MRL-10, Maintenance Rule Program, Rev. 5
ER-AA-MRL-100, Implementing Maintenance Rule, Rev. 6
Condition Reports
Condition Reports
463678     481455     504410     504427      520001     523968    528885     529424     534028     534105     535283     535566  
463678 481455       504410     504427      520001   523968    528885 529424 534028
 
534105 535283       535566
Other Documents
Other Documents
ACE 19512, AAC Jacket Water Heater Failed to Heat, 08/30/13 DC-SU-12-01121, Replacement of Fuses for AAC Diesel Jacket Keep Warm Heater 0-BCW-HTR-1, Rev. 0 ME-0977, Fire Protection: Fire Main Loop Pressure Calculation, Rev. 0
ACE 19512, AAC Jacket Water Heater Failed to Heat, 08/30/13
SDBD-SPS-FP, System Design Basis Document for Fire Protection System SPS, Rev. 12 System Health Report, Black Out Diesel System, Q3-13
DC-SU-12-01121, Replacement of Fuses for AAC Diesel Jacket Keep Warm Heater 0-BCW-
HTR-1, Rev. 0
ME-0977, Fire Protection: Fire Main Loop Pressure Calculation, Rev. 0
SDBD-SPS-FP, System Design Basis Document for Fire Protection System SPS, Rev. 12
System Health Report, Black Out Diesel System, Q3-13
RCE 000027, Diesel Fire Pump Overspeed Trip, 08/06/07
RCE 000027, Diesel Fire Pump Overspeed Trip, 08/06/07
Work Orders
Work Orders
38046801201     38046801301
38046801201     38046801301
Section 1R13: Work Control
Section 1R13: Work Control
Procedures
Procedures
ADM-OU-SU-201, Shutdown Safety Assessment Checklist, Rev. 8
ADM-OU-SU-201, Shutdown Safety Assessment Checklist, Rev. 8
Line 799: Line 842:
1-OPT-ZZ-002, ESF Actuation with Undervoltage and Degraded Voltage 1J Bus, Rev. 37
1-OPT-ZZ-002, ESF Actuation with Undervoltage and Degraded Voltage 1J Bus, Rev. 37
Other Documents
Other Documents
Shutdown Risk Review Emergent Work Evaluation & High Risk Evolution Contingency Plan -   "C" RSST Switching at Lowered Inventory, 11/08/13
Shutdown Risk Review Emergent Work Evaluation & High Risk Evolution Contingency Plan -
Section 1R15: Operability Determinations and Functionality Assessments
C RSST Switching at Lowered Inventory, 11/08/13
 
Section 1R15: Operability Determinations and Functionality Assessments
Procedures
Procedures
1-OPT-CS-004, Containment Spray Check Valves, Rev. 8
1-OPT-CS-004, Containment Spray Check Valves, Rev. 8
2-OPT-CS-002, Containment Spray System Test, Rev. 16
2-OPT-CS-002, Containment Spray System Test, Rev. 16
1-OPT-CT-201, Containment Isolation Valve Local Leak Rate Testing (Type C Containment Testing), Rev. 21
1-OPT-CT-201, Containment Isolation Valve Local Leak Rate Testing (Type C Containment
CM-AA-2, Design Change Expectations, Rev. 0 DNES-AA-GN-1001, Engineering Review, Rev. 3 CM-AA-DDC-201, Design Changes, Rev. 12  
Testing), Rev. 21
CM-AA-2, Design Change Expectations, Rev. 0
DNES-AA-GN-1001, Engineering Review, Rev. 3
CM-AA-DDC-201, Design Changes, Rev. 12


Attachment
Condition Reports
Condition Reports
28665     530886     532703     530267    529007   517409     518809    532802    454236 454039    
28665 530886       532703   530267    529007 517409     518809    532802    454236
 
454039
Other Documents
Other Documents
CE-1642, The Effect of Reactor Pressure Vessel (RPV) Head Replacement on the RPV Support System (Neutron Shield Tank), Surry Power Station Units 1 and 2, Rev. 0 CE-1653, Review of Structural Adequacy of the Reactor Vessel Support Sliding Foot Assemblies - Surry Units 1 and 2, Rev. 0 ETE-CME-2013-0010, Evaluation of SPS Check Valves 1-RS-11, 1-RS-17, 1-CS-13, 1-CS-24, 2-RS-11, 2-RS-17, 2-CS-13, and 2-CS-24 in their Current Configuration 2-CS-P-1A and 2-CS-P-1B Performance Test History, 09/11-11/13 ETE-SU-2012-1015, Oil Foaming in the Charging Pump Sump, Rev. 0 SM-1476, Surry GOTHIC Analysis of NPSH Available for the LHSI and RS Pumps, Addendum   H, Rev. 1
CE-1642, The Effect of Reactor Pressure Vessel (RPV) Head Replacement on the RPV Support
Work Order 38103146704  
System (Neutron Shield Tank), Surry Power Station Units 1 and 2, Rev. 0
 
CE-1653, Review of Structural Adequacy of the Reactor Vessel Support Sliding Foot
Section 1R18: Plant Modifications
Assemblies - Surry Units 1 and 2, Rev. 0
ETE-CME-2013-0010, Evaluation of SPS Check Valves 1-RS-11, 1-RS-17, 1-CS-13, 1-CS-24,
2-RS-11, 2-RS-17, 2-CS-13, and 2-CS-24 in their Current Configuration
2-CS-P-1A and 2-CS-P-1B Performance Test History, 09/11-11/13
ETE-SU-2012-1015, Oil Foaming in the Charging Pump Sump, Rev. 0
SM-1476, Surry GOTHIC Analysis of NPSH Available for the LHSI and RS Pumps, Addendum
H, Rev. 1
Work Order 38103146704
Section 1R18: Plant Modifications
Procedures
Procedures
1-OPT-CT-201, Containment Isolation Valve Local Leak Rate Testing (Type C Containment Testing), Rev. 21
1-OPT-CT-201, Containment Isolation Valve Local Leak Rate Testing (Type C Containment
Testing), Rev. 21
CM-AA-2, Design Change Expectations, Rev. 0
CM-AA-2, Design Change Expectations, Rev. 0
DNES-AA-GN-1001, Engineering Review, Rev. 3 CM-AA-DDC-201, Design Changes, Rev. 12
DNES-AA-GN-1001, Engineering Review, Rev. 3
CM-AA-DDC-201, Design Changes, Rev. 12
Temporary Modifications
Temporary Modifications
SU-11-00017, CCHX SW Inlet Pipe Repair & 1-SW-MOV-102A/B Replacements, Rev. 0
SU-11-00017, CCHX SW Inlet Pipe Repair & 1-SW-MOV-102A/B Replacements, Rev. 0
Line 826: Line 881:
SU-12-00005, SPS Flex Equipment, Rev. 0
SU-12-00005, SPS Flex Equipment, Rev. 0
SU-12-00022, Flex BDB Mechanical Connections, Rev. 0
SU-12-00022, Flex BDB Mechanical Connections, Rev. 0
SU-13-01019, BDB-Flex Power for Essential Instrumentation and Equipment, Rev. 7  
SU-13-01019, BDB-Flex Power for Essential Instrumentation and Equipment, Rev. 7
 
Other Documents
Other Documents
ETE-CEP-2012-0005, Design and Licensing Basis Review of the Surry Seismic and Flooding Requirements Related to the March 12, 2012, NRC 50.54(f) Request for Information, Rev. 2 ETE-CEP-2012-0011, Beyond Design Basis - FLEX Strategy Overall Integrated Plan Basis Document, Rev. 0 ETE-CEP-2012-0020, Surry Units 1 & 2 - Reliable Spent Fuel Pool Instrumentation Project
ETE-CEP-2012-0005, Design and Licensing Basis Review of the Surry Seismic and Flooding
Requirements Related to the March 12, 2012, NRC 50.54(f) Request for Information, Rev. 2
ETE-CEP-2012-0011, Beyond Design Basis - FLEX Strategy Overall Integrated Plan Basis
Document, Rev. 0
ETE-CEP-2012-0020, Surry Units 1 & 2 - Reliable Spent Fuel Pool Instrumentation Project
Documentation, Rev. 0
Documentation, Rev. 0
ETE-CEP-2012-0030, Input Required For Mechanical Work Scope to Facilitate Implementation of Phase 1 Coping Strategy of NEI-12-06 "Diverse and Flexible Coping Strategies (FLEX) Implementation Guide For Surry Power Station", Rev. 1 50.59 Screening Form for DC-SU-00022, Flex BDB Mechanical Connections
ETE-CEP-2012-0030, Input Required For Mechanical Work Scope to Facilitate Implementation
Condition Reports (*NRC Identified) 530290     530512     530896*     530935     531248     532319    532478    532712     533401*
of Phase 1 Coping Strategy of NEI-12-06 Diverse and Flexible Coping Strategies
Attachment
  (FLEX) Implementation Guide For Surry Power Station, Rev. 1
50.59 Screening Form for DC-SU-00022, Flex BDB Mechanical Connections
Condition Reports (*NRC Identified)
530290 530512 530896* 530935               531248   532319    532478    532712   533401*
 
Work Orders
Work Orders
38103284558     38103312079
38103284558       38103312079
Drawings 11448-FM-071A SH 2, Flow/Valve Diagram Circulating & Service Water Systems Surry Unit 1, Rev. 103
Drawings
11448-FM-071A SH 4, Flow/Valve Diagram Circulating & Service Water Systems Surry Unit 1, Rev. 36 11448-FM-089A SH 2, Flow/Valve Operating Numbers Diagram Safety Injection System Surry Unit 1, Rev. 55
11448-FM-071A SH 2, Flow/Valve Diagram Circulating & Service Water Systems Surry
11448-FM-089B SH 4, Flow/Valve Operating Numbers Diagram Safety Injection System Surry Unit 1, Rev. 25 1200022-1-M-803, Piping Isometric BDB Modification RCS Make-Up Injection Connection Surry Unit 1, Rev. 4
Unit 1, Rev. 103
Section 1R19: Post Maintenance Testing
11448-FM-071A SH 4, Flow/Valve Diagram Circulating & Service Water Systems Surry
Unit 1, Rev. 36
11448-FM-089A SH 2, Flow/Valve Operating Numbers Diagram Safety Injection System Surry
Unit 1, Rev. 55
11448-FM-089B SH 4, Flow/Valve Operating Numbers Diagram Safety Injection System Surry
Unit 1, Rev. 25
200022-1-M-803, Piping Isometric BDB Modification RCS Make-Up Injection Connection Surry
Unit 1, Rev. 4
Section 1R19: Post Maintenance Testing
Procedures
Procedures
0-OP-SW-007, Emergency Service Water Pump 1-SW-P-1A Comprehensive Test, Rev. 14 1-OPT-FW-003, Turbine Driven Auxiliary Feedwater Pump 1-FW-P-2, Rev. 44
0-OP-SW-007, Emergency Service Water Pump 1-SW-P-1A Comprehensive Test, Rev. 14
1-OPT-FW-003, Turbine Driven Auxiliary Feedwater Pump 1-FW-P-2, Rev. 44
0-OPT-VS-006, Flow Switches FS-VS-117A and FS-VS-117B Test, Rev. 10
0-OPT-VS-006, Flow Switches FS-VS-117A and FS-VS-117B Test, Rev. 10
0-OP-VS-002, Auxiliary Building Ventilation System, Rev. 21
0-OP-VS-002, Auxiliary Building Ventilation System, Rev. 21
1-OPT-CT-201, Containment Isolation Valve Local Leak Rate Testing (Type C Containment Testing), Rev. 21 1-PT-18.8, Charging Pump Service Water Performance, Rev. 35
1-OPT-CT-201, Containment Isolation Valve Local Leak Rate Testing (Type C Containment
1-EPT-106-02, Main Station Battery 1B Service Test, Rev. 23  
Testing), Rev. 21
 
1-PT-18.8, Charging Pump Service Water Performance, Rev. 35
1-EPT-106-02, Main Station Battery 1B Service Test, Rev. 23
Condition Reports
Condition Reports
535056     522094     530792    531460    531495
535056 522094       530792    531460    531495
Maintenance Orders/Work Orders
Maintenance Orders/Work Orders
38103178809     38103248693
38103178809 38103248693
Section 1R20: Refueling and Outage Activities
Section 1R20: Refueling and Outage Activities
Procedures
Procedures
1-GOP-2.7, Unit Shutdown, Power Decrease from Allowable Power to Unit Offline for Refueling Outage, Rev. 24 1-OP-RC-004, Draining the RCS to Reactor Flange Level, Rev. 34
1-GOP-2.7, Unit Shutdown, Power Decrease from Allowable Power to Unit Offline for Refueling
1-OP-RX-009, Dilution to Critical Conditions Following Refueling, Rev 21.  
Outage, Rev. 24
 
1-OP-RC-004, Draining the RCS to Reactor Flange Level, Rev. 34
Condition Reports (*NRC Identified) 529782    529912     530494     530507     531543     531608     531860     532200     532217     532232    532236     532733     532774*   532782*   532802    532824*   533322     533347  
1-OP-RX-009, Dilution to Critical Conditions Following Refueling, Rev 21.
 
Condition Reports (*NRC Identified)
29782 529912 530494 530507 531543                   531608   531860   532200 532217
2232 532236 532733 532774* 532782*                 532802    532824*   533322 533347
Other Documents
Other Documents
Surry Unit 1 2013 RFO Shutdown Risk Review Report, Rev. 0  
Surry Unit 1 2013 RFO Shutdown Risk Review Report, Rev. 0


Attachment
Section 1R22: Surveillance Testing
Section 1R22: Surveillance Testing
Procedures
Procedures
1-OPT-CT-201, Containment Isolation Valve Local Leak Rate Testing (Type C Containment Testing), Rev. 21
1-OPT-CT-201, Containment Isolation Valve Local Leak Rate Testing (Type C Containment
Testing), Rev. 21
1-OPT-ZZ-002, ESF Actuation with Undervoltage and Degraded Voltage - 1J Bus, Rev. 37
1-OPT-ZZ-002, ESF Actuation with Undervoltage and Degraded Voltage - 1J Bus, Rev. 37
2-OPT-RC-10.01, Reactor Coolant Leakage - Manually Calculated, Rev. 16
2-OPT-RC-10.01, Reactor Coolant Leakage - Manually Calculated, Rev. 16
2-OPT-RC-10.00, Reactor Coolant Leakage - Computer Calculated, Rev. 40
2-OPT-RC-10.00, Reactor Coolant Leakage - Computer Calculated, Rev. 40
Condition Reports
Condition Reports
25641     529887     529437     529900     530027    530047     530048     531506     532624
25641 529887       529437   529900   530027    530047   530048   531506     532624
2668     533118     533152     534915
2668 533118       533152   534915
Work Orders
Work Orders
38103237838  
38103237838
 
Other Documents
Other Documents
EWR 91-011, RC Leakage Calculation Revisions/Surry/1&2, 03/28/91 Technical Report NE-1381, Evaluation of Surry Power Station RCS Leak Rate Calculation,   Rev. 0
EWR 91-011, RC Leakage Calculation Revisions/Surry/1&2, 03/28/91
Section 2RS1: Radiological Hazard Assessment and Exposure Controls
Technical Report NE-1381, Evaluation of Surry Power Station RCS Leak Rate Calculation,
Rev. 0
Section 2RS1: Radiological Hazard Assessment and Exposure Controls
Procedures, Guidance Documents, and Manuals
Procedures, Guidance Documents, and Manuals
C-HP-1032.051, Airborne Radioactivity Counting and Analysis, Rev. 2 C-HP-1032.080, Controlled Area and Unrestricted Area Radiological Surveys, Rev. 9 HP-1032.110, Standard Radiation Monitoring & Dose Rate Trending, Rev. 0
C-HP-1032.051, Airborne Radioactivity Counting and Analysis, Rev. 2
C-HP-1032.080, Controlled Area and Unrestricted Area Radiological Surveys, Rev. 9
HP-1032.110, Standard Radiation Monitoring & Dose Rate Trending, Rev. 0
PI-AA-200, Corrective Action Program, Rev. 21
PI-AA-200, Corrective Action Program, Rev. 21
RP-AA-106, Radiological Work Control Program, Rev. 2
RP-AA-106, Radiological Work Control Program, Rev. 2
RP-AA-109, Radiological Survey Program, Rev. 0 RP-AA-220, Radiological Survey Scheduling, Rev. 1 RP-AA-222, Radiation Surveys, Rev. 1
RP-AA-109, Radiological Survey Program, Rev. 0
RP-AA-220, Radiological Survey Scheduling, Rev. 1
RP-AA-222, Radiation Surveys, Rev. 1
RP-AA-223, Contamination Surveys, Rev. 3
RP-AA-223, Contamination Surveys, Rev. 3
RP-AA-226, Alpha Monitoring, Rev. 3
RP-AA-226, Alpha Monitoring, Rev. 3
RP-AA-233, Control of General License Devices, Rev. 0
RP-AA-233, Control of General License Devices, Rev. 0
RP-AA-240, Discrete Radioactive Particle Control, Rev. 1 RP-AA-275, Radiological Risk Assessment Process, Rev. 0 VPAP-2101, Radiation Protection Program, Rev. 34
RP-AA-240, Discrete Radioactive Particle Control, Rev. 1
0-HSP-INST-002, DAC Value Calculations and Instrument Sensitivity Determination, Rev. 0 0-HPT-ISFSI-001, Independent Spent Fuel Storage Installation (ISFSI) Quarterly Surveillance, Rev. 13 0-HPT-ISFSI-002, NUHOMS Dr
RP-AA-275, Radiological Risk Assessment Process, Rev. 0
y Spent Fuel Storage System Surveillance, Rev. 4 0-HPT-ISFSI-003, CASTOR V/21, CASTOR X/33, NAC-128, MC-10 & TN-32 Dry Storage Cask Surveillance Requirements, Rev. 1 0-HPT-LKTEST-001, Health Physics Source Leak Test, Rev. 6
VPAP-2101, Radiation Protection Program, Rev. 34
0-HSP-INST-002, DAC Value Calculations and Instrument Sensitivity Determination, Rev. 0
0-HPT-ISFSI-001, Independent Spent Fuel Storage Installation (ISFSI) Quarterly Surveillance,
Rev. 13
0-HPT-ISFSI-002, NUHOMS Dry Spent Fuel Storage System Surveillance, Rev. 4
0-HPT-ISFSI-003, CASTOR V/21, CASTOR X/33, NAC-128, MC-10 & TN-32 Dry Storage Cask
Surveillance Requirements, Rev. 1
0-HPT-LKTEST-001, Health Physics Source Leak Test, Rev. 6
Records and Data
Records and Data
Air Sample Count Room Log, 5/19/2013 to 10/24/2013
Air Sample Count Room Log, 5/19/2013 to 10/24/2013
Air Sample Results, U-1 Transfer Canal, Sample ID 13-2518-1021-1651, 10/21/2013
Air Sample Results, U-1 Transfer Canal, Sample ID 13-2518-1021-1651, 10/21/2013
Attachment ALARA Review Evaluations: 13-008, 2013 U-1 RFO RP Routine & Radioactive Material Control Support, 9/23/2013 13-013, 2013 U-1 NDE Inspections, 9/25/2013
 
ALARA Review Evaluations:
13-008, 2013 U-1 RFO RP Routine & Radioactive Material Control Support, 9/23/2013
13-013, 2013 U-1 NDE Inspections, 9/25/2013
13-021, 2013 U-1 RFO Lower Internals Lift & Set, Draft
13-021, 2013 U-1 RFO Lower Internals Lift & Set, Draft
13-022, 2013 U-1 RFO Sliding Foot inspection, 9/29/2013
13-022, 2013 U-1 RFO Sliding Foot inspection, 9/29/2013
13-023, 2013 U-1 RFO Fuel Transfer Cart repairs, 9/25/2013
13-023, 2013 U-1 RFO Fuel Transfer Cart repairs, 9/25/2013
13-025, 2013 U-1 RFO 10 Year lSI MRP-227 Girth Weld Inspections, 9/25/2013 Argos Radon rejection setting recommendations, email from vendor, 4/24/2012 Gamma Spectrum Analysis, Sample ID 23-OCT-2013-0085, 10/23/2013
13-025, 2013 U-1 RFO 10 Year lSI MRP-227 Girth Weld Inspections, 9/25/2013
Argos Radon rejection setting recommendations, email from vendor, 4/24/2012
Gamma Spectrum Analysis, Sample ID 23-OCT-2013-0085, 10/23/2013
List of Non-Fuel items stored in the Spent Fuel Pool, 10/1/2013
List of Non-Fuel items stored in the Spent Fuel Pool, 10/1/2013
List of the 10 most exposure significant work areas in the plant, 10/2/2013
List of the 10 most exposure significant work areas in the plant, 10/2/2013
List of 2013 Unit 1 Outage Radiation Work Permits, 10/3/2013 List of Alarm Setpoints for RCA Exit and Release Monitors (Personnel Contamination, Portal, and Small Article Monitors), 10/24/2013 NTS Annual Inventory Report for Surry Power Station, 1/8/2013
List of 2013 Unit 1 Outage Radiation Work Permits, 10/3/2013
List of Alarm Setpoints for RCA Exit and Release Monitors (Personnel Contamination, Portal,
and Small Article Monitors), 10/24/2013
NTS Annual Inventory Report for Surry Power Station, 1/8/2013
Preliminary Committed Effective Dose Equivalent (CEDE) - Inhalation Intake Data, 10/23/2013
Preliminary Committed Effective Dose Equivalent (CEDE) - Inhalation Intake Data, 10/23/2013
Radiation Protection, Health Physics Operations Turnover Report, 11/23/2013
Radiation Protection, Health Physics Operations Turnover Report, 11/23/2013
Radiation Work Permits (RWPs): 13-0-2102, RP Outage Support, Rev. 0
Radiation Work Permits (RWPs):
13-0-2102, RP Outage Support, Rev. 0
13-0-2109, NDE/ISI Inspections, Rev. 0
13-0-2109, NDE/ISI Inspections, Rev. 0
13-0-2505, Transfer Canal Activities, Rev. 0
13-0-2505, Transfer Canal Activities, Rev. 0
13-0-2516, Lower Internals Lift & Set - Unit 1, Rev. 0
13-0-2516, Lower Internals Lift & Set - Unit 1, Rev. 0
13-0-2517, Sliding Foot Inspection, Rev. 0 13-0-2518, Fuel Transfer Cart Repairs, Rev. 0 13-0-2520, U-1 10 Year lSI MRP-227 Girth Weld Inspections, Rev. 0
13-0-2517, Sliding Foot Inspection, Rev. 0
Work Order (WO) 38103164478, Inventory of Nationally Tracked Sources by Health Physics, 1/8/2013 WO 3810333089, 0-HPT-ISFSI-001, Independent Spent Fuel Storage Installation (ISFSI) Quarterly Surveillance, 7/15/2013 WO 38103366702, 0-HPT-ISFSI-001, Independent Spent Fuel Storage Installation (ISFSI) Quarterly Surveillance, 10/15/2013 Radiological Surveys:
13-0-2518, Fuel Transfer Cart Repairs, Rev. 0
13-0-2520, U-1 10 Year lSI MRP-227 Girth Weld Inspections, Rev. 0
Work Order (WO) 38103164478, Inventory of Nationally Tracked Sources by Health Physics,
1/8/2013
WO 3810333089, 0-HPT-ISFSI-001, Independent Spent Fuel Storage Installation (ISFSI)
Quarterly Surveillance, 7/15/2013
WO 38103366702, 0-HPT-ISFSI-001, Independent Spent Fuel Storage Installation (ISFSI)
Quarterly Surveillance, 10/15/2013
Radiological Surveys:
Aux Bldg -2' GATE 1 & 2 IX alley, Survey M-20130119-2 and M-20130820-2, 1/18/2013 and
Aux Bldg -2' GATE 1 & 2 IX alley, Survey M-20130119-2 and M-20130820-2, 1/18/2013 and
8/20/2013 Auxiliary Building Tool Room Survey, Survey M-20130923-7, 9/23/2013 and 9/16/2013 Auxiliary Building 27' Process Vent line on overhead, Survey 20120813-2, 8/13/2012
8/20/2013
Fuel Building 27', 15', 6', Survey M-20130801-2, 8/1/2013 and 9/11/2013
Auxiliary Building Tool Room Survey, Survey M-20130923-7, 9/23/2013 and 9/16/2013
Fuel Building 45', Survey M-20130801-1, 8/1/2013 and 9/11/2013
Auxiliary Building 27 Process Vent line on overhead, Survey 20120813-2, 8/13/2012
ISFSI Pad #1, 8/22/13 ISFSI Pad #3/NUHOMS Bunker, 8/23/2013, 9/4/2013 and 9/9/2013 Service Building Tool Room Survey, 10/15/2013 and 10/22/2013
Fuel Building 27, 15, 6, Survey M-20130801-2, 8/1/2013 and 9/11/2013
Fuel Building 45, Survey M-20130801-1, 8/1/2013 and 9/11/2013
ISFSI Pad #1, 8/22/13
ISFSI Pad #3/NUHOMS Bunker, 8/23/2013, 9/4/2013 and 9/9/2013
Service Building Tool Room Survey, 10/15/2013 and 10/22/2013
U-1 Containment Transfer Canal, 10/21/2013
U-1 Containment Transfer Canal, 10/21/2013
U-1 Containment Reactor Cavity, 10/21/2013
U-1 Containment Reactor Cavity, 10/21/2013
U-1 Containment - Incore Sump Room, 10/22/2013
U-1 Containment - Incore Sump Room, 10/22/2013
U-1 Containment "B" Loop Room, Contamination Survey of ISI Equipment, 10/23/2013 U-1 Charging Pumps -2', Survey M-20130825-1, 8/24/2013, 9/29/2013 and 10/22/13 U-1 2" elevation, "A", "B", and "C" Charging pump Cubes, 1/22/2013U-1 13' Pipe Chase Knockout Drum Pre and Post Shielding, Survey M-20120803-4, 8/3/2012
U-1 Containment B Loop Room, Contamination Survey of ISI Equipment, 10/23/2013
Attachment U-1 -2" Elevation Knockout Drum drain line, 8/2/2012 and 8/7/2012 U-1 13' Cable Vault, Surveys M-20120803-5, M-20120807-4, and M-20120807-8, 8/2/2012, 8/3/2012 and 8/7/2012 U1R25 Daily Exposure Status, 10/21/2013
U-1 Charging Pumps -2, Survey M-20130825-1, 8/24/2013, 9/29/2013 and 10/22/13
U-1 2 elevation, A, B, and C Charging pump Cubes, 1/22/2013U-1 13 Pipe Chase
Knockout Drum Pre and Post Shielding, Survey M-20120803-4, 8/3/2012


U-1 -2 Elevation Knockout Drum drain line, 8/2/2012 and 8/7/2012
U-1 13 Cable Vault, Surveys M-20120803-5, M-20120807-4, and M-20120807-8, 8/2/2012,
8/3/2012 and 8/7/2012
U1R25 Daily Exposure Status, 10/21/2013
Corrective Action Program Documents
Corrective Action Program Documents
Apparent Cause Evaluation, ACE 019215, Aux. Building Dose Rates increase during Resin Transfer, 8/29/2012
Apparent Cause Evaluation, ACE 019215, Aux. Building Dose Rates increase during Resin
Apparent Cause Evaluation, ACE019240, Groundwater Protection Program Criteria Exceeded, 10/4/2012 Apparent Cause Evaluation, ACE019241, Special Nuclear Material not stored in the Protected
Transfer, 8/29/2012
Area, 10/12/2012 Nuclear Oversight Audit 12-06, "Radiological Protection/Process Control Program/Chemistry", 9/20/2012 Self-Assessment, SAA001754 "Radioactive Source Control", 8/16/2012
Apparent Cause Evaluation, ACE019240, Groundwater Protection Program Criteria Exceeded,
Self-Assessment, SAR001710, "Radiological Protection Technician Fundamentals", 12/10/2012
10/4/2012
Self-Assessment, SAR002281, "Radioactive C
Apparent Cause Evaluation, ACE019241, Special Nuclear Material not stored in the Protected
ontamination Control Program", 9/18/2013
Area, 10/12/2012
Nuclear Oversight Audit 12-06, Radiological Protection/Process Control Program/Chemistry,
9/20/2012
Self-Assessment, SAA001754 Radioactive Source Control, 8/16/2012
Self-Assessment, SAR001710, Radiological Protection Technician Fundamentals, 12/10/2012
Self-Assessment, SAR002281, Radioactive Contamination Control Program, 9/18/2013
Condition Reports
Condition Reports
2059     484956     485152    485847     510859    514234     523692
2059 484956       485152    485847   510859    514234   523692
2RS8: Radioactive Material Processing and Transportation
2RS8: Radioactive Material Processing and Transportation
Procedures, Manuals, and Guides
Procedures, Manuals, and Guides
0-HSP-RESIN-001, Primary Resin Transfer Activities, Rev. 1 C-HP-1071.010, Control of Radioactive Sources, Rev. 6
0-HSP-RESIN-001, Primary Resin Transfer Activities, Rev. 1
C-HP-1071.010, Control of Radioactive Sources, Rev. 6
C-HP-1071.030, Receiving Radioactive Material, Rev.3
C-HP-1071.030, Receiving Radioactive Material, Rev.3
C-HP-1071.040, Packaging and Shipment of Radioactive Material, Rev. 10
C-HP-1071.040, Packaging and Shipment of Radioactive Material, Rev. 10
C-HP-1072.010, Packaging Radioactive Waste, Rev. 2 C-HP-1072.030, Computer Programs for Radwaste and Radioactive Material, Rev. 1 C-HP-1072.040, Radioactive Waste Disposal Usi
C-HP-1072.010, Packaging Radioactive Waste, Rev. 2
ng the Barnwell Disposal Facility, Rev. 9 C-HP-1072.050, Radioactive Waste Transfer To Licensed Waste Processors, Rev. 9
C-HP-1072.030, Computer Programs for Radwaste and Radioactive Material, Rev. 1
C-HP-1072.070, Radioactive Waste Disposal Usi
C-HP-1072.040, Radioactive Waste Disposal Using the Barnwell Disposal Facility, Rev. 9
ng the Energy Solutions Containerized Waste
C-HP-1072.050, Radioactive Waste Transfer To Licensed Waste Processors, Rev. 9
Facility, Rev. 5 C-HP-1072.071, Radioactive Waste Disposal Using the Energy Solutions Bulk Waste Facility, Rev. 5 HP-1071.021, Storing Radioactive Material Outside the Protected Area, Rev. 10
C-HP-1072.070, Radioactive Waste Disposal Using the Energy Solutions Containerized Waste
HP-1071.022, Placing High Integrity Containers (HICs) Into Storage at the SRF HIC Storage Area, Rev.0 HP-1072.020, Sampling, Analyzing, and Classifying Radioactive Waste, Rev. 6 HP-1072.082, Radioactive Waste Transfer To Studsvik Processing Facility, Erwin, TN, Rev. 2 RP-AA-108, Radioactive Material Control Program, Rev. 2   3.4.5.i
Facility, Rev. 5
C-HP-1072.071, Radioactive Waste Disposal Using the Energy Solutions Bulk Waste Facility,
Rev. 5
HP-1071.021, Storing Radioactive Material Outside the Protected Area, Rev. 10
HP-1071.022, Placing High Integrity Containers (HICs) Into Storage at the SRF HIC Storage
Area, Rev.0
HP-1072.020, Sampling, Analyzing, and Classifying Radioactive Waste, Rev. 6
HP-1072.082, Radioactive Waste Transfer To Studsvik Processing Facility, Erwin, TN, Rev. 2
RP-AA-108, Radioactive Material Control Program, Rev. 2 3.4.5.i
RP-AA-233, Control of General License Devices, Rev. 0
RP-AA-233, Control of General License Devices, Rev. 0
RP-AA-231, Radiological Control Areas, Rev.3
RP-AA-231, Radiological Control Areas, Rev.3
RP-AA-232, Radioactive Material Control, Rev. 4
RP-AA-232, Radioactive Material Control, Rev. 4
PI-AA-200, Corrective Action, Rev. 21  
PI-AA-200, Corrective Action, Rev. 21


Attachment Shipping Records and Radwaste Data
Shipping Records and Radwaste Data
Shipments: SH-2013-020     (Type A)
Shipments:
SH-2013-026     (Type A)
SH-2013-020 (Type A)
SV-2012-2         (Type B)
SH-2013-026 (Type A)
D-2013-1           (LSA)
SV-2012-2       (Type B)
D-2013-6           (SCO)
D-2013-1       (LSA)
Waste Streams: 1-FC-FL-1B 2013 IL    Fuel Pit Skimmer Filter
D-2013-6       (SCO)
DAW 2012 IL               Dry Active Waste Smears
Waste Streams:
Blend Tank 2012 IL     Blended Class B Primary Resin Catch Tank 2012 IL     Blended Class A Primary Resin RO 2012                     Surry Radwaste Facility (SRF) Liquid Waste
1-FC-FL-1B 2013 IL    Fuel Pit Skimmer Filter
EBT 2012                   SRF Evaporator Bottoms Tank Liquid
DAW 2012 IL           Dry Active Waste Smears
1-CH-FL-2 2012 IL     Unit 1 Reactor Coolant Filter
Blend Tank 2012 IL   Blended Class B Primary Resin
2-CH-FL-2 2012 IL     Unit 2 Reactor Coolant Filter
Catch Tank 2012 IL   Blended Class A Primary Resin
RO 2012               Surry Radwaste Facility (SRF) Liquid Waste
EBT 2012             SRF Evaporator Bottoms Tank Liquid
1-CH-FL-2 2012 IL     Unit 1 Reactor Coolant Filter
2-CH-FL-2 2012 IL     Unit 2 Reactor Coolant Filter
Condition Reports
Condition Reports
456050     485152   485014     485847     521582    523692
456050 485152       485014     485847   521582    523692
Section 1EP4: Emergency Action Level and Emergency Plan Changes
Section 1EP4: Emergency Action Level and Emergency Plan Changes
Change Packages
Change Packages
Emergency Plan, Revs. 57 and 58 Emergency Action Level Technical Bases Document, Rev 3
Emergency Plan, Revs. 57 and 58
Emergency Action Level Technical Bases Document, Rev 3
Evacuation Time Estimate Study Update
Evacuation Time Estimate Study Update
Section 4OA1: Performance Indicator Verification
Section 4OA1: Performance Indicator Verification
Procedures, Guidance Documents, and Manuals
Procedures, Guidance Documents, and Manuals
RP-AA-111, "Monitoring and Improving R
RP-AA-111, Monitoring and Improving Radiological Performance, Rev. 2
adiological Performance", Rev. 2 RP-AA-112, "Radiation Safety Performance Indicator Reporting", Rev. 4
RP-AA-112, Radiation Safety Performance Indicator Reporting, Rev. 4
PI-AA-100-1000, "Performance Indicators", Rev. 4  
PI-AA-100-1000, Performance Indicators, Rev. 4
 
Records and Data
Records and Data
Regulatory Assessment Performance Indicators, Radiological Protection Surry Power Station, for May 2012 through April 2013
Regulatory Assessment Performance Indicators, Radiological Protection Surry Power Station,
for May 2012 through April 2013
Condition Reports
Condition Reports
475959     476101     494281     498211
475959 476101 494281 498211
Section 4OA1: Performance Indicator Verification
Section 4OA1: Performance Indicator Verification
Procedures
Procedures
ER-AA-SPI-1001, Implementation of the Consolidat
ER-AA-SPI-1001, Implementation of the Consolidated Data Entry (CDE) Reporting for Mitigating
ed Data Entry (CDE) Reporting for Mitigating   System Performance Index (MSPI), Rev. 2
System Performance Index (MSPI), Rev. 2
ER-AA-SPI-1002, Maintaining the MSPI Basis Document, Rev. 1
ER-AA-SPI-1002, Maintaining the MSPI Basis Document, Rev. 1
Other Documents
Other Documents
Monthly PI Reports with Associated Data, September 2012 to September 2013
Monthly PI Reports with Associated Data, September 2012 to September 2013
Technical Report SE-0006, NRC MSPI Basis Document Surry Power Station, Rev. 1
Technical Report SE-0006, NRC MSPI Basis Document Surry Power Station, Rev. 1
Attachment
 
Section 4OA2: Identification and Resolution of Problems
Section 4OA2: Identification and Resolution of Problems
Procedures
Procedures
OP-AA-100, Conduct of Operations, Rev. 25
OP-AA-100, Conduct of Operations, Rev. 25
SUADM-O-26, Administrative Control of Operational Components, Rev. 6 1-OPT-EG-001, EDG #1 Monthly Start Exercise Test, Rev. 59  
SUADM-O-26, Administrative Control of Operational Components, Rev. 6
 
1-OPT-EG-001, EDG #1 Monthly Start Exercise Test, Rev. 59
Condition Reports
Condition Reports
508823     508316     535283    503764     532802
508823 508316     535283    503764   532802
Other Documents
Other Documents
ACE 019423, Failed ACC Mechanical Overspeed Trip System, 04/30/13
ACE 019423, Failed ACC Mechanical Overspeed Trip System, 04/30/13
ACE 019381, 15H3 Failure During 1-OPT-EG-001, 03/26/13 RTE P-SURR-335912, AAC Electronic Speed Switch, 09/26/13 RAS 000226, Failed AAC Mechanical Overspeed Trip System, 03/21/13
ACE 019381, 15H3 Failure During 1-OPT-EG-001, 03/26/13
U1/U2 Operations Aggregate Impact report  
RTE P-SURR-335912, AAC Electronic Speed Switch, 09/26/13
 
RAS 000226, Failed AAC Mechanical Overspeed Trip System, 03/21/13
U1/U2 Operations Aggregate Impact report
LIST OF ACRONYMS
LIST OF ACRONYMS
 
AAC   Alternating Current
AAC Alternating Current
ACE   Apparent Cause Evaluation
ACE Apparent Cause Evaluation
ADAMS Agencywide Document Access and Management System
ADAMS Agencywide Document Access and Management System AFW Auxiliary Feedwater
AFW   Auxiliary Feedwater
ALARA As Low As Reasonably Achievable BACC  Boric Acid Corrosion Control BDB Beyond Design Basis
ALARA As Low As Reasonably Achievable
BACC  Boric Acid Corrosion Control
BDB   Beyond Design Basis
BPVC  Boiler and Pressure Vessel Code
BPVC  Boiler and Pressure Vessel Code
CAP Corrective Action Program
CAP   Corrective Action Program
CC Component Cooling CDE Consolidated Data Entry CEDE   Committed Effective Dose Equivalent
CC   Component Cooling
CFR Code of Federal Regulations
CDE   Consolidated Data Entry
CR Condition Report
CEDE Committed Effective Dose Equivalent
CS Containment Spray DC Design Change DOT Department of Transportation
CFR   Code of Federal Regulations
ED Electronic Dosimeter
CR   Condition Report
EDG Emergency Diesel Generator
CS   Containment Spray
EVT Enhanced Visual Testing
DC   Design Change
EPIP  Emergency Plan Implementing Procedures EPRI  Electric Power Research Institute ETSS  Examination Technique Specification Sheet
DOT   Department of Transportation
ED   Electronic Dosimeter
EDG   Emergency Diesel Generator
EVT   Enhanced Visual Testing
EPIP  Emergency Plan Implementing Procedures
EPRI  Electric Power Research Institute
ETSS  Examination Technique Specification Sheet
FLEX  Diverse and Flexible Coping Strategies
FLEX  Diverse and Flexible Coping Strategies
FMS Fukushima Mitigating Strategies
FMS   Fukushima Mitigating Strategies
HIC High Integrity Container HP Health Physics HRA High Radiation Area
HIC   High Integrity Container
HX Heat Exchanger
HP   Health Physics
IMC Inspection Manual Chapter
HRA   High Radiation Area
IOD Immediate Operability Determination
HX   Heat Exchanger
IP Inspection Procedure ISFSI Independent Spent Fuel Storage Installation ISI In-Service Inspection
IMC   Inspection Manual Chapter
IST Integrated Service Test
IOD   Immediate Operability Determination
IP   Inspection Procedure
ISFSI Independent Spent Fuel Storage Installation
ISI   In-Service Inspection
IST   Integrated Service Test
LHRA  Locked High Radiation Area
LHRA  Locked High Radiation Area
LHSI  Low Head Safety Injection LP Liquid Penetrant MSPI  Mitigating Systems Performance Indicator
LHSI  Low Head Safety Injection
NCV Non-cited Violation
LP   Liquid Penetrant
NDE Non-Destructive Examination
MSPI  Mitigating Systems Performance Indicator
NEI Nuclear Energy Institute
NCV   Non-cited Violation
NPSH  Net Positive Suction Head NRC Nuclear Regulatory Commission NRR Office of Nuclear Reactor Regulation
NDE   Non-Destructive Examination
OD Operability Determination
NEI   Nuclear Energy Institute
Attachment OS Occupational Radiation Safety OWA Operator Work Arounds PARS Publicly Available Records
NPSH  Net Positive Suction Head
PD Performance Defiency
NRC   Nuclear Regulatory Commission
PI Performance Indicator
NRR   Office of Nuclear Reactor Regulation
PM Periodic Maintenance
OD   Operability Determination
POD Prompt Operability Determination PS Public Radiation Safety RAS Reasonable Assurance of Safety
 
RCA Radiologically Controlled Area
OS     Occupational Radiation Safety
RCE Root Cause Evaluation
OWA   Operator Work Arounds
RCS Reactor Coolant System RFO Refueling Outage RHR Residual Heat Removal
PARS   Publicly Available Records
RMC Radioactive Material Control
PD     Performance Defiency
RP Radiation Protection
PI     Performance Indicator
RPV Reactor Pressure Vessel RS Recirculation Spray RSST Reserve Service Station Transformer
PM     Periodic Maintenance
RTP Rated Thermal Power
POD   Prompt Operability Determination
RWP Radiation Work Permit
PS     Public Radiation Safety
SDP Significance Determination Process
RAS   Reasonable Assurance of Safety
SG Steam Generator SRF Surry Radwaste Facility SSC System, Structure or Component
RCA   Radiologically Controlled Area
SW Service Water
RCE   Root Cause Evaluation
RCS   Reactor Coolant System
RFO   Refueling Outage
RHR   Residual Heat Removal
RMC   Radioactive Material Control
RP     Radiation Protection
RPV   Reactor Pressure Vessel
RS     Recirculation Spray
RSST   Reserve Service Station Transformer
RTP   Rated Thermal Power
RWP   Radiation Work Permit
SDP   Significance Determination Process
SG     Steam Generator
SRF   Surry Radwaste Facility
SSC   System, Structure or Component
SW     Service Water
TDAFWP Turbine Driven Auxiliary Feedwater Pump
TDAFWP Turbine Driven Auxiliary Feedwater Pump
TIA Task Interface Agreement TS Technical Specifications UFSAR  Updated Final Safety Analysis Report
TIA   Task Interface Agreement
URI Unresolved Item
TS     Technical Specifications
UT Ultrasonic Testing
UFSAR  Updated Final Safety Analysis Report
VEPCO Virginia Electric and Power Company
URI   Unresolved Item
VHRA Very High Radiation Area VPAP Virginia Power Administrative Procedure VT Visual Testing
UT     Ultrasonic Testing
WO Work Order
VEPCO Virginia Electric and Power Company
VHRA   Very High Radiation Area
VPAP   Virginia Power Administrative Procedure
VT     Visual Testing
WO     Work Order
Attachment
}}
}}

Revision as of 08:26, 4 November 2019

IR 05000280-13-005, 05000281-13-005; 10/01/2013 - 12/31/2013; Surry Power Station, Units 1 and 2: Plant Modifications; Radioactive Solid Waste Processing and Radioactive Material Handling, Storage, and Transportation
ML14041A449
Person / Time
Site: Surry  Dominion icon.png
Issue date: 02/10/2014
From: Mark King
NRC/RGN-II/DRP/RPB5
To: Heacock D
Virginia Electric & Power Co (VEPCO)
References
IR-13-005
Download: ML14041A449 (50)
v • d • e


Text

UNITED STATES uary 10, 2014

SUBJECT:

SURRY POWER STATION - NRC INTEGRATED INSPECTION REPORT 05000280/2013005, 05000281/2013005

Dear Mr. Heacock:

On December 31, 2013, the United States Nuclear Regulatory Commission (NRC) completed an inspection at your Surry Power Station Units 1 and 2. The enclosed inspection report documents the inspection findings which were discussed on January 29, 2014, with Mr. R.

Simmons and other members of your staff. Inspectors documented the results of this inspection in the enclosed inspection report.

This report documents two findings of very low safety significance (Green), both of which were determined to be violations of NRC requirements. Additionally, one licensee-identified violation which was determined to be of very low safety significance is listed in this report. However, because of the very low safety significance of these issues and because they were entered into your corrective action program, the NRC is treating these as non-cited violations (NCV)

consistent with Section 2.3.2.a of the NRC Enforcement Policy.

If you contest the violations or significance of these NCVs, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the Nuclear Regulatory Commission, ATTN.: Document Control Desk, Washington DC 20555-0001; with copies to the Regional Administrator Region II; the Director, Office of Enforcement, United States Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC Resident Inspector at the Surry Power Station.

Additionally, if you disagree with a cross-cutting aspect assignment in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your disagreement, to the Regional Administrator, Region II, and the NRC Resident Inspector at the Surry Power Station. As a result of the Safey Culture Common Language Initiative, the terminology and coding of cross-cutting aspects were revised beginning in calendar year (CY) 2014. New cross-cutting aspects identified in CY 2014 will be coded under the latest revision to IMC 0310. Cross-cutting aspects identified in the last six months of 2013 using the previous terminology will be converted to the latest revision in accordance with the cross-cutting themes and potential substantive cross-cutting issues in accordance with IMC 0305 starting with the CY 2014 mid-cycle assessment review.

In accordance with Title 10 of the Code of Federal Regulations 2.390, Public Inspections, Exemptions, Requests for Withholding, of the NRCs Rules of Practice, a copy of this letter, its enclosure, and your response (if any) will be available electronically for public inspection in the NRCs Public Document Room or from the Publicly Available Records (PARS) component of the NRCs Agencywide Documents Access and Management System (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Sincerely,

/RA/

Michael F. King, Chief Reactor Projects Branch 5 Division of Reactor Projects Docket Nos.: 50-280, 50-281 License Nos.: DPR-32, DPR-37

Enclosure:

Inspection Report 05000280/2013005, 05000281/2013005 w/Attachment: Supplemental Information

REGION II==

Docket Nos.: 50-280, 50-281 License Nos.: DPR-32, DPR-37 Report No: 05000280/2013005, 05000281/2013005 Licensee: Virginia Electric and Power Company (VEPCO)

Facility: Surry Power Station, Units 1 and 2 Location: 5850 Hog Island Road Surry, VA 23883 Dates: October 1, 2013 through December 31, 2013 Inspectors: P. McKenna, Senior Resident Inspector J. Nadel, Resident Inspector D. Bacon, Senior Operations Engineer (Section 1R11)

A. Butcavage, Reactor Inspector (Section 1R08)

R. Carrion, Senior Reactor Inspector (Section 1R08)

R. Hamilton, Senior Health Physicist (Section 2RS8)

R. Kellner, Health Physicist (Sections 2RS1, 40A1)

J. Laughlin, Emergency Preparedness Inspector (Section 1EP4)

J. Rivera-Ortiz, Senior Reactor Inspector (Section 1R08)

K. Roche, Reactor Operations Engineer Approved by: Michael F. King, Chief Reactor Projects Branch 5 Division of Reactor Projects Enclosure

SUMMARY OF FINDINGS

IR 05000280/2013005, 05000281/2013005; 10/01/2013-12/31/2013; Surry Power Station, Units and 2: Plant Modifications; Radioactive Solid Waste Processing and Radioactive Material Handling, Storage, and Transportation The report covered a three month period of inspection by resident inspectors and region based inspectors. One NRC identified and one self-revealing findings were identified and both were determined to be a non-cited violations (NCVs). The significance of most findings is indicated by their color (Green, White, Yellow, Red) using Inspection Manual Chapter (IMC) 0609,

Significance Determination Process (SDP). The cross-cutting aspect was determined using IMC 0310, Components Within The Cross-Cutting Areas. Findings for which the SDP does not apply may be Green or be assigned a severity level after NRC management review. The NRCs program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process Revision 4, dated December 2006.

Cornerstone: Barrier Integrity

Green.

An NRC-identified Green NCV of 10 CFR 50, Appendix B, Criterion III, Design Control, was identified for the licensees failure to adequately protect safety-related low head safety injection system (LHSI) piping from a tornado missile. Specifically, the licensee installed a mechanical piping connection to the Unit 1 LHSI system as part of a design change (DC) and did not provide tornado missile protection for the portions of piping that were installed above the 28.6 feet elevation in the safeguards valve pit and connected directly to the LHSI piping. The issue was documented in the licensees corrective action program (CAP) as condition report (CR) 533401.

The licensees failure to protect the Unit 1 LHSI system piping against external missile hazards when the piping was modified by the diverse and flexible coping strategies (FLEX) mechanical piping connection as part of DC SU-12-00022 was a performance deficiency (PD) that was within the licensees ability to foresee and correct.

The inspectors determined that the performance deficiency (PD) was more than minor because it was associated with the design control attribute of the Barrier Integrity Cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. Specifically, the welded piping connection installed by DC SU-12-00022 located between containment and containment isolation valve 1-SI-MOV-1890A, A LHSI hot leg discharge isolation valve, was susceptible to failure from the impact of a tornado generated missile. Using Manual Chapter 0609.04, Initial Characterization of Findings,

Table 2, dated June 19, 2012, the finding was determined to affect the reactor containment barrier and the Barrier Integrity Cornerstone. The inspectors screened the finding using Manual Chapter 0609 Appendix A, Significance Determination Process (SDP) for Findings at-Power, and determined the finding was of very low safety significance (Green) because the finding did not represent an actual open pathway in the physical integrity of reactor containment. This finding has a cross-cutting aspect in the work control component of the human performance area, H.3(b); because the licensee failed to address both the impact of changes in the work scope on the plant and to use adequate interdepartmental coordination during the design process. (Section 1R18)

Cornerstone: Public Radiation Safety

Green.

A self-revealing non-cited violation of 10 CFR 20.1802, Control of Material not in Storage, was identified for the licensees failure to maintain control and constant surveillance of licensed radioactive material in a controlled or unrestricted area (Health Physics (HP) technical services area of the administration building) that was not in storage. The material that was initially unaccounted for was an Americium-241 check source with an activity of 0.02 micro-Curies, used to perform routine function checks on iSolo alpha/beta counter. The issued was documented in the licensees corrective action program (CAP) as condition report (CR) 523692.

The licensees failure to control and maintain constant surveillance of licensed material that is in a controlled or unrestricted area and that is not in storage was a performance deficiency (PD). The PD was more than minor because it was associated with the Program and Process attribute of the Public Radiation Safety Cornerstone and affected the cornerstone objective of ensuring adequate protection of public health and safety from exposure to radioactive materials released into the public domain. Using Manual Chapter 0609, Appendix D, Public Radiation Safety SDP, this finding determined to be was of very low safety significance (Green) in that the public radiation exposure was not greater than 0.005 rem (5 millirem). The inspectors determined that cross-cutting issue H.4(b), The licensee defines and effectively communicates expectations regarding procedural compliance and personnel follow procedures, was applicable for this violation because the radiation protection (RP) technician had failed to follow procedure HP-1033.148, Canberra iSolo: Performance Checks, step 6.4.3 which states: Ensure check source is removed from Canberra iSolo and returned to designated storage location. (Section 2RS8)

One violation of very low safety significance, which was identified by the licensee, was reviewed by the inspectors. Corrective actions taken or planned by the licensee have been entered into the licensees corrective action program. This violation and the respective corrective actions are listed in Section 4OA7 of this report.

REPORT DETAILS

Summary of Plant Status

Unit 1 operated at or near rated thermal power (RTP) from the beginning of the inspection period until October 20, 2013, when it was shutdown to begin a planned refueling outage (RFO).

It remained offline until November 21, when the main turbine generator was synchronized to the grid. On November 24, the unit reached full RTP and operated there for the remainder of the inspection period.

Unit 2 operated at or near full RTP throughout the inspection period.

REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity

1R01 Adverse Weather Protection

.1 Seasonal Readiness Reviews for Cold Weather

a. Inspection Scope

The inspectors reviewed the licensees preparations for seasonal cold weather.

Inspection focused on verification of design features and implementation of the licensees procedure for cold weather conditions, 0-OSP-ZZ-001, Cold Weather Preparation, Revision 14. The inspectors walked down key structures including the turbine and auxiliary buildings, safeguards buildings, the emergency switchgear rooms, and emergency battery rooms and verified HVAC systems were operating properly and that area temperatures remained within design requirements specified in the UFSAR.

The mitigating systems reviewed during this inspection include: the auxiliary feedwater systems, the refueling water storage tanks, emergency diesel generators, alternate alternating current (AAC) diesel generator, and emergency switchgear.

b. Findings

No findings were identified.

1R04 Equipment Alignment

.1 Partial Walkdown

a. Inspection Scope

The inspectors conducted three equipment alignment partial walkdowns to evaluate the operability of selected redundant trains or backup systems, listed below, with the other train or system inoperable or out of service. The inspectors reviewed the functional systems descriptions, Updated Final Safety Analysis Report (UFSAR), system operating procedures, and Technical Specifications (TS) to determine correct system lineups for the current plant conditions. The inspectors performed walkdowns of the systems to verify that critical components were properly aligned and to identify any discrepancies which could affect operability of the redundant train or backup system.

  • Spent fuel pool cooling following Unit 1 core offload.
  • AAC (Station Blackout) Diesel following an unplanned equipment outage.

b. Findings

No findings were identified.

.2 Complete Walkdown

a. Inspection Scope

The inspectors performed a detailed walkdown and inspection of the Unit 1 containment spray system to verify the system was properly aligned and capable of performing its safety function, and to assess its material condition. During the walkdown, the inspectors verified breaker positions were in the proper alignment, component labeling was accurate, hangers and supports were functional, and local indications were accurate. Recent testing history was also reviewed to verify that standby components were performing within their design. The plant health report, system drawings, condition reports, the UFSAR, and TS were reviewed and outstanding deficiencies were verified to be properly classified and not affect system operability and capability to perform its safety function. The inspectors reviewed the corrective action program to verify equipment alignment issues were being identified and resolved.

b. Findings

No findings were identified.

1R05 Fire Protection

.1 Quarterly Fire Protection Reviews

a. Inspection Scope

The inspectors conducted tours of the areas listed below that is important to reactor safety to verify the licensees implementation of fire protection requirements as described in fleet procedures CM-AA-FPA-100, Fire Protection/Appendix R (Fire Safe Shutdown) Program, Revision 8, CM-AA-FPA-101, Control of Combustible and Flammable Materials, Revision 6, and CM-AA-FPA-102, Fire Protection and Fire Safe Shutdown Review and Preparation Process and Design Change Process, Revision 6.

The reviews were performed to evaluate the fire protection program operational status and material condition and the adequacy of: 1) control of transient combustibles and ignition sources; 2) fire detection and suppression capability; 3) passive fire protection features; 4) compensatory measures established for out-of-service, degraded or inoperable fire protection equipment, systems, or features; and 5) procedures, equipment, fire barriers, and systems so that post-fire capability to safely shutdown the plant is ensured. The inspectors reviewed the corrective action program to verify fire protection deficiencies were being identified and properly resolved.

  • Unit 1 Containment
  • 27.6 foot level of the Auxiliary Building
  • Main Control Room
  • 1A Battery Room
  • Unit 2 Turbine Building Basement

b. Findings

No findings were identified.

1R06 Flood Protection Measures

a. Inspection Scope

The inspectors reviewed the internal flood protection measures and procedural controls established to address potential flooding in the Unit 1 and 2 turbine buildings, the emergency switchgear rooms, and mechanical equipment rooms 3 and 4 during the operation of temporary modification SU-11-00017, CCHX SW Inlet Pipe Repair & 1-SW-MOV-102A/B Replacements which installed temporary service water (SW) piping in the Unit 1 turbine building. The inspectors conducted a walkdown of the affected areas to observe and assess the condition of the installed flood dikes, floor drain backflow preventers, the sealing of holes and penetrations between flood areas, the adequacy of water tight doors, the operability of flooding alarms, and the installed sump pumps.

Additionally, the inspectors verified that the required actions of the safety evaluation related to license amendment number 279 were being accomplished by the licensee.

b. Findings

No findings were identified.

1R07 Heat Sink Performance

a. Inspection Scope

The inspectors reviewed the licensees heat exchanger program document, 0-MCM-0812-01, Component Cooling Heat Exchanger (CCHX) Inspection and Cleaning, Revision 18, trending data maintained by the system engineer, maintenance rule information, specific commitments, and design basis information. Specific focus was placed on the operation and surveillance testing of the component cooling heat exchangers during the installation and use of the temporary service water jumper piping during Unit 1 Fall 2013 refueling outage. The inspectors reviewed licensee performance of surveillance procedure 1-NSP-CC-005, Component Cooling Heat Exchanger Tests Using the Temporary Monitoring System, Revision 1, which is a test designed to measure the capability of the component cooling heat exchangers while they are being fed from the temporary service water jumper piping and the normal supply piping was out of service. The inspectors reviewed testing procedures and test results to confirm that the component cooling heat exchangers were still able to perform their functions and that planned corrective actions were appropriate. The inspectors verified that significant heat exchanger performance issues were being entered into the licensees CAP and appropriately addressed.

b. Findings

No findings were identified.

1R08 Inservice Inspection Activities

a. Inspection Scope

From October 28, 2013, through November 1, 2013, the inspectors conducted an on-site review of the implementation of the licensees Inservice Inspection (ISI) program for monitoring degradation of the reactor coolant system, steam generator (SG) tubes, emergency feedwater systems, risk-significant piping and components and containment systems in Unit 1.

Non-Destructive Examination (NDE) Activities and Welding Activities: The inspectors reviewed associated records and directly observed the following non-destructive examinations required by the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPVC) to verify compliance with Section XI and Section V of the ASME BPVC of record for Surry Power Station Unit 1 (1998 Edition with 2000 Addenda). The inspectors also verified that any relevant indications and defects were dispositioned in accordance with the requirements of the ASME BPVC or an NRC-approved alternative requirement.

  • Ultrasonic testing (UT) examination of elbow-to-pipe weld number 11448-WMKS-1105B9/3-CH-71/1-BS, 3-inch diameter charging line of charging pump 1-C
  • Visual testing (VT-3) examination of pipe support 11448-MKS-1105B9 The inspectors also reviewed a sample of other augmented or industry initiative examinations performed during the Fall 2013 refueling outage. The inspectors observed a portion of an augmented, enhanced visual testing (EVT-1), associated with the industrys Materials Reliability Program (MRP)-227-A, Pressurized Water Reactor Internals Inspection and Evaluation Guidelines. Specifically, the inspectors in conjunction with an NDE Level III qualified inspector, reviewed video recordings of the core barrel upper girth weld visual inspection of the inside diameter of the weld between the 180° and 270° azimuth locations to determine if the examination and disposition of indications were consistent with the MRP-227-A guidelines. The inspectors interviewed plant personnel to verify that the current scope of the Reactor Internals MRP-227 inspection activities included all the primary component inspections listed in the MRP-227 tables for a Westinghouse designed unit.

During non-destructive surface and volumetric examinations performed since the previous RFO, the licensee did not identify any recordable indications that were accepted for continued service through analytical methods. Therefore, no NRC review was completed for this attribute of the inspection procedure.

The inspectors reviewed the following pressure boundary welds completed for risk-significant systems since the last Unit 1 RFO to evaluate if the licensee applied the pre-service non-destructive examinations and acceptance criteria required by the construction code and ASME BPVC Section XI. In addition, the inspectors reviewed the work orders (WO), welding procedure specifications, welder qualifications, welding material certification and supporting weld procedure qualification records, to evaluate if the weld procedures were qualified in accordance with the requirements of the construction code and ASME BPVC,Section IX.

Pressure Vessel Upper Head Penetration Inspection Activities: For the Unit 1 reactor pressure vessel upper head, no examination was required pursuant to 10 CFR 50.55a(g)(6)(ii)(D) for the Fall 2013 RFO. Therefore, no NRC review was completed for this inspection procedure attribute.

Boric Acid Corrosion Control (BACC): The inspectors reviewed the licensees BACC program activities to ensure implementation with commitments made in response to NRC Generic Letter 88-05, Boric Acid Corrosion of Carbon Steel Reactor Pressure Boundary, and applicable industry guidance documents. Specifically, the inspectors performed an on-site record review of procedures and the results of the licensees containment walk-down inspections performed during the Fall 2013 RFO (S1R25). The inspectors also interviewed the BACC program owner, conducted an independent walk-down of portions of the containment to evaluate compliance with the licensees BACC program requirements, and verified that degraded or non-conforming conditions, such as boric acid leaks, were properly identified, evaluated, and corrected in accordance with the licensees BACC and corrective action programs, and were consistent with the requirements of ASME BPVC Section XI and 10 CFR Part 50, Appendix B, Criterion XVI.

Specifically the inspectors reviewed the results and evaluations associated with the following boric acid indications:

  • CR-532007, Corrosion observed on 1-RC-FC1482C
  • CR-529937, Boric Acid Identified on 1-RC-ICV-3142 Steam Generator (SG) Tube Inspection Activities. The inspectors reviewed the eddy current examination activities performed in Unit 1 SG B during RFO S1R25 to verify compliance with the licensees TS, ASME BPVC Section XI, and Nuclear Energy Institute (NEI) 97-06, Steam Generator Program Guidelines. The inspectors interviewed licensee personnel and vendor staff responsible for the SG inspection project and reviewed documentation associated with the SG inspections and integrity assessments as described below.

The inspectors reviewed the scope of the eddy current examinations to verify that known and potential areas of tube degradation were inspected. The inspectors also verified that inspection scope expansion criteria were implemented based on inspection results as directed by the Electric Power Research Institute (EPRI) Pressurized Water Reactor Steam Generator Examination Guidelines, Revision 7. The inspectors selected a sample of degradation mechanisms from the Unit 1 Steam Generator Degradation Assessment and verified that the in-situ pressure testing criteria were determined in accordance with the EPRI Guidelines. Additionally, the inspectors reviewed eddy current indication reports to determine whether tubes with relevant indications were appropriately screened for in-situ pressure testing. The inspectors review included the implementation of tube repair criteria and planned repair methods to verify they were consistent with plant TS and industry guidelines.

The inspectors compared the recent eddy current examination results with the last Condition Monitoring and Operational Assessment Report for SG B to assess the licensees prediction capability for maximum tube degradation and number of tubes with indications. The inspectors verified that the licensees evaluation was conservative and that current examination results were bound by the operational assessment projections.

Since the licensee performed eddy current tube examinations only in SG B during the Fall 2013 RFO, the inspectors reviewed the last operational assessment for SGs A and C to verify the licensee met the inspection frequency established in the plant TS and had evaluated the acceptability of these SGs to meet the tube integrity performance criteria until the next scheduled inspection.

The inspectors also compared past examination results discussed in the latest degradation assessment with the recent eddy current examination results to verify that new degradation mechanisms were identified and evaluated before plant startup. The review of eddy current examination results included the disposition of potential loose part indications on the SG secondary side to verify that corrective actions for evaluating and retrieving loose parts were consistent with the EPRI Guidelines. The inspectors also reviewed a sample of primary-to-secondary leakage data for the last Unit 1 operating cycle to obtain reasonable assurance that operational leakage in all three SGs remained below the detection or action level thresholds during the previous operating cycle.

In addition, the inspectors reviewed documentation for a sample of eddy current data analysts, eddy current probes, and eddy current testers to verify these were qualified to detect the existing and potential degradation mechanisms applicable to Surry SG tubes.

This review included a sample of site-specific examination technique specification sheets (ETSSs) to ensure that their qualification and site-specific implementation were consistent with Appendix H or I of the EPRI Pressurized Water Reactor Steam Generator Examination Guidelines, Revision 7. The selected ETSSs for review were based on degradation mechanisms of interest to the inspectors based on plant-specific and industry operating experience. The inspectors selected bobbin probe ETSSs qualified for detection and sizing of loose part wear in tube freespan locations, thinning in tube support plates and top-of-tubesheet, and wear at tube support plates and anti-vibration bars. The inspectors also selected rotating and array probe ETSSs for detection of circumferential and axial stress corrosion cracking on the internal and external surfaces of the tubes at the expansion transition area. The inspectors also reviewed a sample of eddy current data with a qualified data analyst to confirm that data analysis was performed in accordance with the ETSSs and site-specific analysis guidelines. The sample of eddy current data selected for review was for the following tubes in SG B: R34C58 (Bobbin Probe), R40C25 (Bobbin Probe), R40C50 (Array-Probe), and R45C48 (Array Probe).

Based on the review of eddy current examination results for SG B and interviews with the licensee, the inspectors confirmed that no new degradation mechanisms were identified, no eddy current scope expansion was required, none of the SG tubes examined met the criteria for in-situ pressure testing, and none of the indications left in-service required repair. Furthermore, the inspectors interviewed licensee staff and reviewed the inspection plan/procedure and a sample of inspection results for the inspection conducted in the Unit 1 SG B secondary side internals, to verify that potential areas of degradation based on site-specific operating experience were inspected, and appropriate corrective actions were taken to address degradation indications. The sample of inspection results consisted of pictures showing the condition of accessible surfaces of the steam drum, uppermost tube support plate, anti-vibration bars, J-nozzle welds, feedwater rings, feedwater ring supports, primary separator swirl vanes, and lower bank secondary separators.

Identification and Resolution of Problems: The inspectors reviewed a sample of corrective action program documents associated with ISI issues to verify that the licensee was identifying problems at an appropriate threshold and entering them in the corrective action program for resolution. The inspectors performed this review to ensure compliance with 10CFR Part 50, Appendix B, Criterion XVI, Corrective Action, requirements. The sample of corrective action documents selected for review also included the licensees evaluation of recent operating experience information applicable to the plant. The corrective action documents reviewed by the inspectors are listed in the report attachment.

b. Findings

Introduction:

The inspectors identified an unresolved item related to the inspection of the reactor pressure vessel (RPV) component supports as required by ASME BPVC Section XI, for which additional information is needed to determine if the issue of concern represents a performance deficiency or a violation of the regulatory requirements.

Description:

The code of record for the current ISI program at Surry Power Station Unit 1 is the 1998 Edition of the ASME BPVC Section XI with the 2000 addenda. This Code edition includes inspection requirements for both nuclear class 1 piping and vessel supports (Subsection IWF) and their attachment welds (Subsection IWB). Subsection IWB, Table IWB-2500-1, item number B10.10, describes the examination requirements for welded attachments for vessels, piping, pumps, and valves. Note 1 of Table IWB-2500-1 states that attachment welds (weld buildup) on nozzles that are in compression under normal load conditions and provide only component support are excluded from the surface examination requirements. The note also provides additional conditions to identify what type welded attachment configurations require inspection. Table IWB-2500-1 also references Figures IWB-2500-13, -14 and -15 to further describe the examination requirements.

The inspectors noted that the scope of the Surry Unit 1 ISI program for the inspection of the nuclear class 1 RPV supports did include the requirements for the IWF portion of the ASME Section XI code required inspections. However, the inspectors identified that the licensee excluded the surface examination requirements for the RPV support attachment welds required by Table IWB-2500-1, item number B10.10 based on the exemptions provided by Note 1 of the table. The licensees position was that the surface examinations are not required based on the exclusion criteria provided in Note 1 for attachment welds under compressive loads during normal conditions and the configurations described in Figures IWB-2500-13, -14 and -15.

The inspectors reviewed design basis documents for the Unit 1 RPV supports and identified that the normal loading conditions of the supports included both compressive and shear loads. The inspectors determined that additional information and discussion with the NRC Office of Nuclear Reactor Regulation (NRR) staff was required, in order to determine if the licensees interpretation and implementation of the exemptions in Table IWB-2500-1 were in compliance with the ASME BPVC Section XI. Therefore, the NRR and Region II staff agreed to submit a Task Interface Agreement (TIA), which could involve the submittal of a formal inquiry to the applicable ASME BPVC committee to request an interpretation of the examination requirements and exemptions in Table IWB-2500-1 for welded attachments for vessels and piping. The NRC initiated TIA-2014-02 to determine the staffs position on whether the configuration of the RPV supports at Surry meets the exclusion criteria in ASME BPVC Section XI.

This issue remains unresolved until the resolution of TIA-2014-02 to determine if the issue of concern represents a performance deficiency or a violation of regulatory requirements. This issue is identified as URI 05000280/2013005-01, Application of ASME Section XI, Table IWB 2500-1, Item B10.10, Inspection Requirements and Note 1 Exemptions.

1R11 Licensed Operator Requalification Program

.1 Resident Inspector Quarterly Review

a. Inspection Scope

The inspectors observed and evaluated a licensed operator simulator exercise given on November 26, 2013. The scenario was intended to exercise the entire operations crew and assess the ability of the operators to react correctly to multiple failures. The inspectors observed the crews performance to determine whether the crew met the scenario objectives; accomplished the critical tasks; demonstrated the ability to take timely action in a safe direction and to prioritize, interpret, and verify alarms; demonstrated proper use of alarm response, abnormal, and emergency operating procedures; demonstrated proper command and control; communicated effectively; and appropriately classified events per the emergency plan. The inspectors observed the post training critique to determine that weaknesses or improvement areas revealed by the training were captured by the instructor and reviewed with the operators.

b. Findings

No findings were identified.

.2 Resident Inspector Observation of Control Room Operations

a. Inspection Scope

During the inspection period, the inspectors conducted observations of licensed reactor operator activities to ensure consistency with licensee procedures and regulatory requirements. For the following activities, the inspectors observed the following elements of operator performance: 1) operator compliance and use of plant procedures including technical specifications; 2) control board component manipulations; 3) use and interpretation of plant instrumentation and alarms; 4) documentation of activities; 5) management and supervision of activities; and 6) control room communications.

  • Unit 1 shutdown for RFO 1R25.
  • Unit 1 startup at the end of RFO 1R25.

b. Findings

No findings were identified.

.3 Annual Review of Licensee Requalification Examination Results

a. Inspection Scope

On February 8, 2013, the licensee completed the annual requalification operating examinations required to be administered to all licensed operators in accordance with Title 10 of the Code of Federal Regulations 55.59(a)(2), Requalification requirements, of the NRCs Operators Licenses. The inspectors performed an in-office review of the overall pass/fail results of the individual operating examinations and the crew simulator operating examinations in accordance with Inspection Procedure (IP) 71111.11, Licensed Operator Requalification Program and Licensed Operator Performance. The results were compared to the thresholds established in Section 3.02, Requalification Examination Results, of IP 71111.11.

b. Findings

No findings were identified.

1R12 Maintenance Effectiveness

a. Inspection Scope

For the three equipment issues described in the condition reports listed below, the inspectors evaluated the effectiveness of the corresponding licensee's preventive and corrective maintenance. The inspectors performed a detailed review of the problem history and associated circumstances, evaluated the extent of condition reviews, as required, and reviewed the generic implications of the equipment and/or work practice problem(s). Inspectors performed walkdowns of the accessible portions of the system, performed in-office reviews of procedures and evaluations, and held discussions with system engineers. The inspectors compared the licensees actions with the requirements of the Maintenance Rule (10 CFR 50.65), station procedures ER-AA-MRL-10, Maintenance Rule Program, Revision 5, and ER-AA-MRL-100, Implementing the Maintenance Rule, Revision 6.

  • CR 496625, Loss of Unit 2 'B' DC Bus and Vital Buses 2-IV and 2-II
  • CR 523699, 1-FP-P-2, Diesel Driven Fire Pump tripping on overspeed
  • CR 534028, Breaker for AAC diesel prelube and jacket heater pump tripped

b. Findings

No findings were identified.

1R13 Maintenance Risk Assessments and Emergent Work Control

a. Inspection Scope

The inspectors evaluated, as appropriate, the three activities listed below for the following: 1) the effectiveness of the risk assessments performed before maintenance activities were conducted; 2) the management of risk; 3) that, upon identification of an unforeseen situation, necessary steps were taken to plan and control the resulting emergent work activities; and, 4) that maintenance risk assessments and emergent work problems were adequately identified and resolved. The inspectors verified that the licensee was complying with the requirements of 10 CFR 50.65(a)(4) and the data output from the licensees safety monitor associated with the risk profile of Units 1 and 2.

The inspectors reviewed the corrective action program to verify deficiencies in risk assessments were being identified and properly resolved.

  • Unit 1 and Unit 2 risk during 1J bus degraded voltage and logic testing
  • Unit 1 and Unit 2 risk during component cooling heat exchanger (CCHX) SW jumper in service, 1B battery out of service and Unit 1 core offload in progress

b. Findings

No findings were identified.

1R15 Operability Determinations and Functionality Assessments

a. Inspection Scope

The inspectors reviewed the six operability evaluations listed below, affecting risk-significant mitigating systems, to assess as appropriate: 1) the technical adequacy of the evaluations; 2) whether continued system operability was warranted; 3) whether other existing degraded conditions were considered; 4) if compensatory measures were involved, whether the compensatory measures were in place, would work as intended, and were appropriately controlled; and 5) where continued operability was considered unjustified, the impact on TS Limiting Conditions for Operation and the risk significance.

The inspectors review included verification that operability determinations were made as specified in OP-AA-102, Operability Determination, Revision 10. The inspectors reviewed the licensees corrective action program to verify deficiencies in operability determinations were being identified and corrected.

  • CR 529007, 1-CH-P-1B, B charging pump gearbox observed oil foaming.
  • CR 532802, Unit 1 "C" loop room drain clogged on containment closeout walkdown.
  • CRs 532841 and 532718, 1-VS-MOV-100C/D, containment purge exhaust valves, have leakage greater than 1-OPT-CT-201 acceptance criteria.
  • CR 532703, Unit 1 sliding foot support pad on "B" loop hot leg has a loose cap screw

b. Findings

No findings were identified.

1R18 Plant Modifications

.1 Temporary Modification SU-11-00017

a. Inspection Scope

The inspectors reviewed temporary modification, SU-11-00017, CCHX SW Inlet Pipe Repair & 1-SW-MOV-102A/B Replacements to verify that the modification did not affect system operability or availability as described by the TS and UFSAR. In addition, the inspectors verified that the temporary modification was in accordance with CM-AA-TDC-204, Temporary Modifications, Revision 3, and for the related work package, that adequate controls were in place, procedures and drawings were updated, and post-installation tests verified the operability of the affected systems.

b. Findings

No findings were identified.

.2 Permanent Modification SU-12-00022

a. Inspection Scope

The inspectors reviewed a permanent plant modification design change (DC) SU-12-00022, FLEX BDB Mechanical Connections which is associated with Fukushima Mitigating Strategies (FMS) related modifications. The inspection scope for this modification was restricted to those elements necessary to satisfy the stated objectives of IP 71111.18, specifically:

  • To verify that modifications have not affected the safety functions of important safety systems
  • To verify that the current design bases, licensing bases, and performance capability of risk significant SSCs have not been degraded through modifications; and
  • To verify that modifications performed during increased risk-significant configurations did not place the plant in an unsafe condition.

The inspectors also verified that the permanent modification was in accordance with CM-AA-DDC-201, Design Changes, Revision 12. The inspection did not address whether the associated FMS modification satisfactorily addressed the objectives of NRC Order EA-12-049, Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis (BDB) External Events.

b. Findings

Introduction:

An NRC-identified Green NCV of 10 CFR 50, Appendix B, Criterion III, Design Control, was identified for the licensees failure to adequately protect safety-related low head safety injection system (LHSI) piping from a tornado missile.

Specifically, the licensee installed a mechanical piping connection to the Unit 1 LHSI system as part of a DC and did not provide tornado missile protection for the portions of piping that were installed above the 28.6 foot elevation in the safeguards valve pit and connected directly to the LHSI piping.

Description:

The licensee completed installation of the Unit 1 physical plant connections of DC SU-12-00022, SPS 1 and 2 FLEX Mechanical Connections during the fall 2013 refueling outage. The design change is part of the licensees compliance with NRC order EA-12-049, Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for BDB External Events in accordance with NEI 12-06, Diverse and Flexible Coping Strategies (FLEX) Implementation Guide. One of the mechanical connections of this modification is designed to allow for high pressure make-up/borated water injection into the Reactor Coolant System during a BDB event. This was accomplished through a welded pipe connection on the existing 6 inch LHSI piping located between containment and the normally closed containment isolation valve 1-SI-MOV-1890A, A LHSI hot leg discharge valve. The 3 inch FLEX piping traversed from this connection, on the 13.9 foot elevation of the safeguards valve pit, through the floor of the two upper levels to the grade level 28.6 foot elevation at the top of the safeguards valve pit, where it is terminated with a hose connection.

The residents questioned the adequacy of this design and noted that the new pipe was only supported by a single seismic restraint on the 28.6 foot elevation, which is per design, but also results in a higher load on the welded LHSI piping connection from externally initiated events. As a result, it was identified that the portion of piping installed above the 28.6 foot elevation is susceptible to a tornado generated missile because the roof of this level of the safeguards valve pit is corrugated metal that was never designed to protect against tornado missiles. The licensee could not immediately determine potential loading from such an impact on the safety-related LHSI piping two floors below.

The licensee evaluated the concern in CR 533401 and performed an immediate operability determination (IOD). The IOD concluded that the piping was operable based on engineering judgment and also non-conforming to the design bases. Initially, a prompt operability determination (POD) was assigned to perform an analysis of the postulated tornado missile impact on the LHSI System piping. However, the POD was cancelled before it was due, because the licensee decided to cut and remove a portion of the FLEX piping below the 28.6 foot elevation to eliminate the possibility of a tornado missile impact affecting the LHSI piping. The design will be reworked prior to the planned Unit 2 modification in spring of 2014.

Dominion procedure CM-AA-DDC-201, Design Changes, Rev. 12 step 3.3.5.h required the following during installation of the DC: If practical and station conditions allow, then walkdown the DC to ensure it is being installed in accordance with the DC and no adverse consequences are created by the modification. The design change document, SU-12-00022, section 3.0 stated, in part, All new piping, valves, tee connections etc.,

are designed to withstand design bases external flooding, storms with high winds (hurricanes, tornadoes, etc.,) and associated missiles. Because the design modification exposed the LHSI system piping to a tornado missile hazard, the issue was placed in the corrective action program as CR 533401.

Through discussions with Dominion corporate engineering staff, the residents learned that the lack of missile protection at the 28.6 foot elevation was discovered during the 30% design review. As a result, the planned location of the safety related boundary valve (1-SI-500) was moved two floors down from its original designed position above the floor level. However, despite this change, the forces that could be transmitted to safety related equipment below floor level were overlooked. Any break in 3 inch piping between 1-SI-500 and the LHSI piping because of an external missile would affect containment integrity. The licensee attributed this oversight to the lack of participation and coordination with the civil engineering department during this portion of the design review and associated plant walkdowns.

Analysis:

The inspectors concluded that the licensees failure to protect the Unit 1 LHSI system piping against external missile hazards when the piping was modified by the FLEX mechanical piping connection as part of DC SU-12-00022 was a PD that was within the licensees ability to foresee and correct. Specifically, Dominion procedure CM-AA-DDC-201, Rev. 12 step 3.3.5.h requires confirmation that a modification is installed in accordance with the DC and no adverse unintended consequences are created; while DC SU-12-00022 further requires the mechanical piping modifications to be able to withstand tornado missiles. The inspectors determined that the PD was more than minor because it was associated with the design control attribute of the Barrier Integrity Cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. Specifically, the welded piping connection installed by DC SU-12-00022 located between containment and containment isolation valve 1-SI-MOV-1890A, A LHSI hot leg discharge isolation valve, was susceptible to failure from the impact of a tornado generated missile.

Using Manual Chapter 0609.04, Initial Characterization of Findings, Table 2, dated June 19, 2012, the finding was determined to affect the reactor containment barrier and the Barrier Integrity Cornerstone. The inspectors screened the finding using Manual Chapter 0609 Appendix A, Significance Determination Process (SDP) for Findings at-Power, and determined the finding was of very low safety significance (Green) because the finding did not represent an actual open pathway in the physical integrity of reactor containment. This finding has a cross-cutting aspect in the work control component of the human performance area, H.3(b); because the licensee failed to address both the impact of changes in the work scope on the plant and to use adequate interdepartmental coordination during the design process.

Enforcement:

10 CFR 50, Appendix B, Criterion III, Design Control, requires, in part, that design changes, including field changes, shall be subject to design control measures commensurate with those applied to the original design. Contrary to the above, on November 21, 2013, the licensee did not assure that changes to the LHSI piping made by DC SU-2012-00022 were subject to the same design control measures as the original piping design. Specifically, the licensee installed a mechanical piping connection to the LHSI System as part of this design change and did not ensure that the new piping was protected from tornado missiles such that existing LHSI piping would remain unaffected. Because the licensee entered the issue into their corrective action program as CR 533401 and the finding is Green, this violation is being treated as an NCV, consistent with Section 2.3.2 of the NRC Enforcement Policy: NCV 05000280/2013005-02, Failure to Missile Protect Beyond Design Bases FLEX Modification to Low Head Safety Injection Piping.

.2 Permanent Modification SU-13-01019

a. Inspection Scope

The inspectors reviewed a permanent plant modification DC SU-13-01019, BDB-Flex Power for Essential Instrumentation and Equipment which is associated with FMS related modifications. The inspection scope for this modification was restricted to those elements necessary to satisfy the stated objectives of IP 71111.18, specifically:

  • To verify that modifications have not affected the safety functions of important safety systems
  • To verify that the current design bases, licensing bases, and performance capability of risk significant SSCs have not been degraded through modifications; and
  • To verify that modifications performed during increased risk-significant configurations did not place the plant in an unsafe condition.

The inspectors also verified that the permanent modification was in accordance with CM-AA-DDC-201, Design Changes, Revision 12. The inspection did not address whether the associated FMS modification satisfactorily addressed the objectives of NRC Order EA-12-049, Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for BDB External Events.

b. Findings

No findings were identified.

1R19 Post Maintenance Testing

a. Inspection Scope

The inspectors reviewed six post maintenance test procedures and/or test activities for selected risk-significant mitigating systems listed below, to assess whether: 1) the effect of testing on the plant had been adequately addressed by control room and/or engineering personnel; 2) testing was adequate for the maintenance performed; 3) acceptance criteria were clear and adequately demonstrated operational readiness consistent with design and licensing basis documents; 4) test instrumentation had current calibrations, range, and accuracy consistent with the application; 5) tests were performed as written with applicable prerequisites satisfied; 6) jumpers installed or leads lifted were properly controlled; 7) test equipment was removed following testing; and 8) equipment was returned to the status required to perform in accordance with VPAP-2003, Post Maintenance Testing Program, Revision 14.

  • 1-EPT-0102-02, Rev. 4 and 1-EPT-0103-02, Rev. 14, Monthly and Quarterly Voltage Check of the 1B Station Battery, following replacement of battery cell #17 and discharge testing of the 1B station battery
  • 1-OPT-FW-003, Rev. 44, Turbine Driven Auxiliary Feedwater (AFW) Pump 1-FW-P-2 Periodic Test, following periodic maintenance (PM)
  • 1-PT-18.8, Rev. 35, Charging Service Water Pump Performance Test on 1-SW-P-10A, following pump seal replacement.
  • 0-OPT-VS-006, Rev. 10, Flow Switches FS-VS-117A and FS-VS-117B Test, following flow switch replacement on 1-VS-F-58A, the auxiliary building filter exhaust fan
  • 0-OPT-SW-007, Rev.16, Emergency Service Water Pump 1-SW-P-1A Comprehensive Test, following pump replacement

b. Findings

No findings were identified.

1R20 Refueling and Other Outage Activities

a. Inspection Scope

The inspectors reviewed the stations work schedule and outage risk plan for the Unit 1 refueling outage, which was conducted October 20 through November 21, 2013, to confirm that the licensee had appropriately considered risk, industry experience, and previous site-specific problems in developing and implementing a plan that assured maintenance of defense-in-depth. The inspectors used IP 71111.20, Refueling and Outage Activities, to observe portions of the maintenance and startup activities to verify that the licensee maintained defense-in-depth commensurate with the outage risk plan and applicable TS. The inspectors monitored licensee controls over the outage activities listed below.

  • Licensee configuration management, including daily outage reports, to evaluate maintenance of defense-in-depth commensurate with the outage risk plan for key safety functions and compliance with the applicable TS when taking equipment out of service.
  • Implementation of clearance activities and confirmation that tags were properly hung and equipment appropriately configured to safely support the work or testing.
  • Controls over the status and configuration of electrical systems to ensure that TS and outage safety plan requirements were met, and controls over switchyard activities.
  • Controls over activities that could affect reactivity.
  • Spent fuel cooling operations to verify that outage work was not impacting the ability of the operations staff to operate the spent fuel cooling system during and after core offload.
  • The control of containment penetrations and containment entries to verify that the licensee controlled those penetrations and activities in accordance with the appropriate TS and could achieve/maintain containment closure for required conditions.
  • Licensee identification and resolution of problems related to forced outage activities.

b. Findings

No findings were identified.

1R22 Surveillance Testing

a. Inspection Scope

For the eight surveillance tests listed below, the inspectors examined the test procedures, witnessed testing, or reviewed test records and data packages, to determine whether the scope of testing adequately demonstrated that the affected equipment was functional and operable, and that the surveillance requirements of TS were met. The inspectors also determined whether the testing effectively demonstrated that the systems or components were operationally ready and capable of performing their intended safety functions.

In-Service Testing:

  • 2-OPT-RC-10.0, Rev. 40, Reactor Coolant Leakage - Computer Calculated Appendix J Leak Rate Determination:
  • 2-NSP-SI-001, Rev. 1, Ultrasonic Examination of Safety Injection Piping
  • 1-OPT-ZZ-002, Rev. 37, ESF Actuation with Undervoltage and Degraded Voltage -

1J Bus

b. Findings

No findings were identified.

RADIATION SAFETY

Cornerstones: Occupational Radiation Safety (OS) and Public Radiation Safety (PS)

2RS1 Radiological Hazard Assessment and Exposure Controls

a. Inspection Scope

Hazard Assessment and Instructions to Workers: During facility tours, the inspectors observed labeling of radioactive material and postings for radiation areas, high radiation areas (HRA), locked HRAs (LHRA), very HRAs (VHRA), radioactive material storage areas, and contaminated areas established within the radiologically controlled area (RCA) of the Unit 1 (U1) and Unit 2 (U2) Auxiliary Buildings, U1 containment, the Independent Spent Fuel Storage Installation (ISFSI), and radioactive waste processing and storage locations. The inspectors independently measured radiation dose rates or directly observed conduct of licensee radiation surveys for RCA areas in the U1 containment, U1 and U2 Auxiliary buildings, and ISFSI. The inspectors reviewed survey records for several plant areas including surveys for alpha emitters, airborne radioactivity, and pre-job surveys for selected U1 Refueling 25 (U1R25) tasks. The inspectors also discussed changes to plant operations that could contribute to changing radiological conditions since the last inspection. For selected U1R25 outage jobs, the inspectors attended pre-job mockups and briefings and reviewed radiation work permit (RWP) details to assess communication of radiological control requirements and current radiological conditions to workers. Selected U1R25 work activities included U1 sliding foot inspection (reactor vessel support sliding foot and floor plate), U1 fuel transfer cart repairs, U1 Non-Destructive Examination (NDE) and ISI inspections, lower internals lift and set, 10 year ISI girth weld inspection, transfer canal work, and routine RP job coverage.

Hazard Control and Work Practices: The inspectors observed and evaluated access barrier effectiveness for selected LHRA and VHRA locations to include the U1 and U2 Auxiliary Buildings and U1 containment. Changes to procedural guidance for LHRA and VHRA controls were discussed with RP supervisors. Controls and their implementation for storage of irradiated material within the spent fuel pool were reviewed and discussed in detail. Established radiological controls (including airborne controls) were evaluated for selected tasks, including U1 sliding foot inspection, and U1 fuel transfer cart repairs.

In addition, licensee controls for areas where dose rates could change significantly as a result of refueling operations were reviewed, observed, and discussed including controlling access to areas adjacent to the fuel transfer canal in the U1 Auxiliary Building and containment.

Occupational workers adherence to selected RWPs and RP technician proficiency in providing job coverage were evaluated through direct observations and interviews with licensee staff. Electronic dosimeter (ED) alarm set points and worker stay times were evaluated against area radiation survey results. Worker response to select ED dose rate alarms was evaluated. For selected U1R25 HRA and LHRA tasks involving significant dose rate gradients, the use and placement of whole body and extremity dosimetry to monitor worker exposure was discussed with licensee staff.

Control of Radioactive Material: The inspectors observed surveys of material and personnel being released from the RCA and U1 containment using portable radiation survey instruments, hand and foot monitors, small article monitors, personnel contamination monitors, and portal monitor instruments. The inspectors observed current calibration labels and source check information for selected radiation and air sampling and monitoring instruments located at the RCA release point, in the Auxiliary Building, and U1 containment. The inspectors reviewed documentation of equipment sensitivity, alarm setpoints, discussed release program guidance with RP staff and observed source response testing of selected monitoring instruments located at the RCA release point. The inspectors also reviewed records of leak tests on selected sealed sources and discussed nationally tracked source transactions with RP staff.

Problem Identification and Resolution: CAP documents associated with radiological hazard assessment and control were reviewed and assessed. The inspectors evaluated the licensees ability to identify and resolve the issues in accordance with procedure PI-AA-200, Corrective Action Program, Rev. 21. The inspectors also evaluated the scope of the licensees internal audit program and reviewed recent assessment results.

RP activities were evaluated against the requirements of USFAR Section 11; TS Section 6.4; 10 CFR Parts 19 and 20; and approved licensee procedures. Licensee programs for monitoring materials and personnel released from the RCA were evaluated against 10 CFR Part 20 and IE Circular 81-07, Control of Radioactively Contaminated Material.

Documents reviewed are listed in the Attachment.

b. Findings

No findings were identified.

2RS8 Radioactive Solid Waste Processing and Radioactive Material Handling, Storage, and

Transportation

a. Inspection Scope

Waste Processing and Characterization: During inspector walk-downs, accessible sections of the liquid and solid radioactive waste (radwaste) processing systems were assessed for material condition and conformance with system design diagrams.

Inspected equipment included radwaste storage tanks; resin transfer piping, resin and filter packaging components; and abandoned evaporator equipment. The inspectors discussed component function, processing system changes, and radwaste program implementation with licensee staff.

The radionuclide characterizations for 2012 and 2013 for selected waste streams were reviewed and discussed with radioactive material control (RMC) staff. For primary resin, reactor coolant system filters and dry active waste the inspectors evaluated analyses for hard-to-detect nuclides, reviewed the use of scaling factors, and examined quality assurance comparison results between licensee waste stream characterizations and outside laboratory data. Waste stream mixing and concentration averaging methodology for resins and filters was evaluated and discussed with RMC staff. The inspectors also reviewed the licensees procedural guidance for monitoring changes in waste stream isotopic mixtures.

Radioactive Material Storage: During walk-downs of indoor and outdoor radioactive material storage areas, the inspectors observed the physical condition and labeling of storage containers and the posting of radioactive material areas. The inspectors also reviewed licensee procedural guidance for storage and monitoring of radioactive material.

Transportation: There were no significant shipments during the week of inspection; however, the inspectors did review shipping procedure requirements and discussed preparation of shipping documents, package marking and labeling, and interviewed shipping technicians regarding Department of Transportation (DOT) regulations.

Selected shipping records were reviewed for consistency with licensee procedures and compliance with NRC and DOT regulations. The inspectors reviewed emergency response information, DOT shipping package classification, waste classification, radiation survey results, and evaluated whether receiving licensees were authorized to accept the packages. Licensee procedures for handling shipping containers were compared to Certificate of Compliance requirements and manufacturer recommendations. In addition, training records for selected individuals currently qualified to ship radioactive material were reviewed.

Radwaste processing activities and equipment configuration were reviewed for compliance with the licensees Process Control Program and UFSAR, Chapter 11.

Waste stream characterization analyses were reviewed against regulations detailed in 10 CFR Part 20, 10 CFR Part 61, and guidance provided in the Branch Technical Position on Waste Classification (1983). Radioactive material and waste storage activities were reviewed against the requirements of 10 CFR Part 20. Transportation program implementation was reviewed against regulations detailed in 10 CFR Part 20, 10 CFR Part 71, 49 CFR Parts 172-178, as well as the guidance provided in NUREG-1608. Training activities were assessed against 49 CFR Part 172 Subpart H.

Documents reviewed are listed in the Attachment.

Problem Identification and Resolution: The inspectors reviewed CRs in the area of radwaste processing and transportation. The inspectors evaluated the licensees ability to identify and resolve the issues in accordance with procedure PI-AA-200. The inspectors also evaluated the scope of the licensees internal audit program and reviewed recent assessment results. Documents reviewed are listed in the Attachment.

b. Findings

Introduction:

A self-revealing Green NCV of 10 CFR 20.1802, Control of Material not in Storage was identified for the licensees failure to control and maintain constant surveillance of licensed material that is in a controlled or unrestricted area and that is not in storage. Specifically, an Americium-241 check source was not returned to the designated storage location.

Description:

On August 19, 2013, an RP technician trainee tasked with performing instrument function checks discovered that the Americium-241 (Am-241) check source was missing from the source kit. The source with 0.02 micro-Curies of Am- 241 was used to perform function checks for the iSolo alpha/beta sample counter in the health physics technical services area in the administration building. The source is greater than the exempt quantity for Am-241 and therefore must be controlled. The RP technician trainee assigned to perform function checks discovered that the 0.02 micro-Curie iSolo check source had been left inside the instrument, in the health physics technical support area. The source had not been returned to the proper storage location after being used on August 18, 2013.

On August 20, 2013, the RP technician trainee notified the RP technician that had last used the source that it had not been returned to the designated storage location. The RP Technician subsequently notified his supervision. Supervision established that the technician had not complied with procedure HP-1033.148, Canberra iSolo: Performance Checks, step 6.4.3 which states: Ensure check source is removed from Canberra iSolo and returned to designated storage location. Supervision also identified that the current revision of procedure C-HP-1071.010, Control of Radioactive Sources, implied but did not explicitly direct the source to be returned to the designated storage location after use. The source was outside the designated storage for approximately 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />. The licensee placed this issue in the CAP as CR 523692.

Analysis:

The inspectors determined that the licensees failure to control and maintain constant surveillance of licensed material that is in a controlled or unrestricted area and that is not in storage was a performance deficiency (PD). The PD was more than minor because it was associated with the Program and Process attribute of the Public Radiation Safety Cornerstone and affected the cornerstone objective of ensuring adequate protection of public health and safety from exposure to radioactive materials released into the public domain. Using Manual Chapter 0609, Appendix D, Public Radiation Safety SDP, this finding determined to be was of very low safety significance (Green) in that the public radiation exposure was not greater than 0.005 rem (5 millirem).

The inspectors determined that cross-cutting issue H.4(b), The licensee defines and effectively communicates expectations regarding procedural compliance and personnel follow procedures. was applicable for this violation because the RP Technician had failed to follow procedure HP-1033.148, Canberra iSolo: Performance Checks, step 6.4.3 which states: Ensure check source is removed from Canberra iSolo and returned to designated storage location.

Enforcement:

Section 10 CFR 20.1802,Control of Material not in Storage, requires that the licensee control and maintain constant surveillance of licensed material that is in a controlled or unrestricted area and that is not in storage. Contrary to the above, on August 18, 2013, the licensee failed to maintain control of licensed radioactive material in an unrestricted area and not in storage, in that an Americium 241 check source was found in the iSolo alpha/beta counter in the Health Physics Technical Services area in the Administration Building. However, because the licensee documented this issue in its corrective action program (CR 523692) and because the violation is of very low safety significance, it is being treated as an NCV, consistent with Section 2.3.2.a of the NRC Enforcement Policy. (NCV 05000280, 05000281/2013-003, Failure to Maintain Control of Licensed Radioactive Material that was not in Storage).

Cornerstone: Emergency Preparedness

1EP4 Emergency Action Level and Emergency Plan Changes

a. Inspection Scope

The NSIR headquarters staff performed an in-office review of the latest revisions of various Emergency Plan Implementing Procedures (EPIPs) and the Emergency Plan located under ADAMS accession numbers ML123420028, ML13029A221, and ML130370771, as listed in the Attachment.

The licensee determined that in accordance with 10 CFR 50.54(q), the changes made in the revisions resulted in no reduction in the effectiveness of the Plan, and that the revised Plan continued to meet the requirements of 10 CFR 50.47(b) and Appendix E to 10 CFR Part 50. The NRC review was not documented in a safety evaluation report and did not constitute approval of licensee-generated changes; therefore, these revisions aresubject to future inspection. Documents reviewed are listed in the Attachment. This inspection activity satisfied one inspection sample for the emergency action level and emergency plan changes on an annual basis.

b. Findings

No findings were identified.

OTHER ACTIVITIES

4OA1 Performance Indicator (PI) Verification

1. Mitigating Systems (MS) Safety Cornerstone

a. Inspection Scope

The inspectors performed a periodic review of the six following Unit 1 and 2 PIs to assess the accuracy and completeness of the submitted data and whether the performance indicators were calculated in accordance with the guidance contained in NEI 99-02, Regulatory Assessment Performance Indicator Guideline, Revision 7. The inspection was conducted in accordance with NRC IP 71151, Performance Indicator Verification. Specifically, the inspectors reviewed the Unit 1 and Unit 2 data reported to the NRC for the period October 1, 2012, through September 30, 2013. Documents reviewed included applicable NRC inspection reports, licensee event reports, operator logs, station performance indicators, and related CRs.

  • Unit 1 & 2 High Pressure Injection MSPI
  • Unit 1 & 2 Cooling Water MSPI

b. Findings

No findings were identified.

2. Occupational Radiation Safety Cornerstone

a. Inspection Scope

The inspectors reviewed the Occupational Exposure Control Effectiveness PI results for the Occupational Radiation Safety Cornerstone from May, 2012 through October, 2013.

The inspectors also reviewed ED alarm logs and CRs related to controls for exposure significant areas.

b. Findings

No findings were identified.

4OA2 Identification and Resolution of Problems

.1 Daily Reviews of items Entered into the Corrective Action Program:

a. Inspection Scope

As required by NRC 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 the licensees CAP. This review was accomplished by reviewing daily CR report summaries and periodically attending daily CR Review Team meetings.

b. Findings

No findings were identified.

.2 Annual Sample: Review of CR 503764, Closed Light for 15H3 in the MCR Went Out

During Performance of 1-OPT-EG-001

a. Inspection Scope

The inspectors performed an in-depth review of the licensees apparent cause analysis and corrective actions associated with CR 503764, 15H3 main generator output breaker for EDG #1 failure to operate (trip) from the Main Control Room during monthly surveillance testing. The inspectors assessed the licensees problem identification threshold, cause analyses, extent of condition reviews, compensatory actions, and the prioritization and timeliness of the licensees corrective actions to determine whether the licensee was appropriately identifying, characterizing, and correcting problems associated with this issue and whether the planned or completed corrective actions were appropriate. The inspectors compared the actions taken to the requirements of the licensees CAP as specified in procedure, PI-AA-200, Corrective Action Program, Revision 21 and 10 CFR 50, Appendix B. In addition, the inspectors reviewed the CAP for similar issues, and interviewed engineering personnel to assess the effectiveness of the implemented corrective actions.

b. Findings and observations

No findings were identified.

The licensee determined that the apparent cause of the failure was due to a failed conductor in the circuit breaker control cable. The conductor failure was attributed to mechanical failure of the cable jacket and conductor insulation due to post-installation stressors such as sharp bends in conduit and proximity to cable tray edges. The licensees immediate corrective actions resulted in the identification and use of a spare conductor in the same cable to restore the breaker and the EDG to operability. Other corrective actions included further testing of the failed conductor to pinpoint the failure location, the replacement of the cable with a new cable that was routed separately, and the creation of a work order to remove the failed conductor cable and perform a failure analysis. The inspectors reviewed the corrective actions completed to date and noted that although further cable testing narrowed down the possible cable sections where the fault originates, the exact location could not be verified due to physical access restrictions. The inspectors noted that the corrective action to create a work order to remove the failed conductor cable and perform a failure analysis had been completed at the time of this inspection, but the work order was still in the planning stages. Inspectors also noted that this failure represented the second conductor in the same cable to fail within a seven month time period; a fact which supports the conclusion that a mechanical failure of the cable jacket is occurring somewhere along the routing path.

The licensee did route a new EDG control cable during the Fall 2013 Unit 1 refueling outage. The inspectors reviewed the work packages and associated EDG testing to verify that the installation of the new cable and other corrective actions taken were effective. The inspectors did not identify any additional issues from this review. The inspectors determined the licensees evaluation of the issue appropriately identified the apparent and contributing causes. Additionally, the inspectors determined that the corrective actions developed as a result of the apparent cause analysis were reasonable and commensurate with the safety significance of the EDG system.

.3 Annual Sample: Review of CR 508616, AAC (Alternate A/C) Mechanical Overspeed Trip

System Failed

a. Inspection Scope

The inspectors performed an in-depth review of the licensees apparent cause analysis and corrective actions associated with CR 508616, AAC Mechanical Overspeed Trip System Failed. The inspectors assessed the licensees problem identification threshold, cause analyses, extent of condition reviews, compensatory actions, and the prioritization and timeliness of the licensees corrective actions to determine whether the licensee was appropriately identifying, characterizing, and correcting problems associated with this issue and whether the planned or completed corrective actions were appropriate. The inspectors compared the actions taken to the requirements of the licensees CAP as specified in procedure, PI-AA-200, Corrective Action Program, Revision 21 and 10 CFR 50, Appendix B. In addition, the inspectors reviewed the corrective action program for similar issues, and interviewed engineering personnel to assess the effectiveness of the implemented corrective actions.

b. Findings and observations

No findings were identified.

The licensee concluded in ACE 019423 that the apparent cause of the AAC mechanical overspeed trip mechanism failure was likely an internal part failure in the Woodward Mechanical Overspeed Controller. The cause determination was limited due to the fact that the overspeed controller was not removed or disassembled during troubleshooting.

Through discussions with plant engineering personnel at the time of the failure, the inspectors learned that there was no replacement governor available for the AAC EDG and a Reasonable Assurance of Safety (RAS) was created to allow for long term operation without a functioning mechanical overspeed trip system. However, a part has since become available and current corrective actions are planned to replace the governor and restore the mechanical overspeed trip function during a maintenance package in mid-2014. The inspectors reviewed the CRs, RAS, and vendor documents associated with this failure and did not identify any regulatory concerns. However, the inspectors did note that the reliance on the electrical overspeed trip mechanism until mid-2014 represents a reduction in the original design safety margin of the AAC EDG.

The inspectors also noted that the licensee self-identified the fact that the electrical overspeed function that is being relied upon had not previously been tested or verified to function at the correct setpoint. As a result, four year PM testing and eight year PM replacement activities were created. However, the inspectors noted that neither PM will be performed in the interim until the planned overspeed trip replacement activities in mid-2014.

The inspectors determined that the corrective actions developed as a result of the apparent cause analysis were reasonable and commensurate with the safety significance of the AAC EDG system. The inspectors did not identify any additional issues from this review.

.4 Annual Sample: Review of Operator Work Arounds

a. Inspection Scope

The inspectors performed a review regarding the licensees assessments and corrective actions for operator workarounds (OWAs). The inspectors reviewed the cumulative effects of the licensees OWAs and licensee procedure OP-AA-1700, Operations Aggregate Impact, Revision 6. The inspectors reviewed the data package associated with this procedure which included an evaluation of the cumulative effects of the OWAs on the operators ability to safely operate the plant and effectively respond to abnormal and emergency plant conditions. The inspectors reviewed and monitored licensee planned and completed corrective actions to address underlying equipment issues causing the OWAs. The inspectors also evaluated OWAs against the requirements of the licensees CAP as specified in PI-AA-200, Corrective Action, Revisions 21, 10 CFR 50, Appendix B, and OP-AA-100, "Conduct of Operations," Revision 25.

b. Findings and Observations

No findings were identified. In general, the inspectors verified that the licensee has identified operator workaround problems at an appropriate threshold, entered them in the corrective action program, and has proposed or implemented appropriate corrective actions.

4OA5 Other Activities

.1 Quarterly Resident Inspector Observations of Security Personnel and Activities

a. Inspection Scope

During the inspection period, the inspectors conducted observations of security force personnel and activities to ensure that the activities were consistent with the licensee security procedures and regulatory requirements relating to nuclear plant security.

These observations took place during both normal and off-normal plant working hours.

These quarterly resident inspector observations of security force personnel and activities did not constitute any additional inspection samples. Rather, they were considered an integral part of the inspectors normal plant status review and inspection activities.

b. Findings

No findings were identified.

4OA6 Meetings, Including Exit

Resident Inspector

Exit Meeting Summary

On January 29, 2014, the inspection results were presented to Mr. R. Simmons and other members of his staff, who acknowledged the findings. The inspectors asked the licensee whether any of the material examined during the inspection should be considered proprietary.

4OA7 Licensee-Identified Violation

The following finding of very low significance (Green) was identified by the licensee and is a violation of NRC requirements which meets the criteria of the NRC Enforcement Policy, for being dispositioned as an NCV.

  • 10 CFR 50, Appendix B, Criterion III, "Design Control, requires, in part, that measures shall be established to assure that applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures, and instructions and that deviations from such standards are controlled. Contrary to the above, on October 8, 2013, the licensee discovered a failure to control the angle of the counterweight arms on Unit 1 and Unit 2 containment spray (CS) and recirculation spray (RS) check valves, CS-13, CS-24, RS-11 and RS-17 such that the arms were no longer vertical when the valves were full open. In addition, no documented evaluation of the valves to open fully under design flow conditions could be found. Further evaluation required the licensee to change the Unit 2 acceptance criteria for the CS pump integrated service test (IST) to increase the minimum required pump differential pressure that is required to meet the design basis flow requirements.

This finding is of very low safety significance (Green) because the position of the counterweight arms did not affect the design or qualification of the CS system and it did not represent a loss of system or train safety function. The position of the Unit 1 counterweight arms was corrected during the fall 2013 RFO. This issue has been entered into the licensees CAP as CR 528665.

ATTACHMENT:

SUPPLEMENTAL INFORMATION

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee Personnel

L. Baker, Manager Training
J. Eggart, Manager, Radiation Protection & Chemistry
B. Garber, Supervisor, Station Licensing
P. Harris, Supervisor Radiological Analysis and Instrumentation
A. Harrow, Manager, Organizational Effectiveness
J. Henderson, Engineering Manager
J. Hopkins, Supervisor Rad Material Control
L. Hilbert, Manager, Outage and Planning
R. Johnson, Manager, Operations
L. Lane, Site Vice President
D. Lawrence, Director, Station Safety and Licensing
T. Mayer, Steam Generator Program Owner
C. Olsen, Director, Station Engineering
J. Pollard, Licensing Engineer
L. Ragland, Supervisor Radiation Protection Operations
R. Scanlan, Manager, Maintenance
K. Sloane, Plant Manager
M. Smith, Manager, Nuclear Oversight
M. True, ISI/BACCP Site Owner
E. Turko, ISI/NDE Supervisor
N. Turner, Supervisor, Emergency Preparedness
C. Wray, NDE Coordinator

LIST OF ITEMS OPENED, CLOSED AND DISCUSSED

Opened

05000280/2013005-01 URI Application of ASME Section XI, Table IWB 2500-1,

Item B10.10, Inspection Requirements and Note 1

Exemptions (Section 1R08)

Opened and Closed

05000280/2012005-02 NCV Failure to Missile Protect Beyond Design Bases

FLEX Modification to Low Head Safety Injection

Piping (Section 1R18)

05000280, 281/2012005-03 NCV Failure to Maintain Control of Licensed Radioactive

Material that was not in Storage (Section 2RS8)

List of Documents Reviewed

Section 1R01: Adverse Weather Protection

Procedures

0-EPM-1303-01, Freeze Protection Inspection, Rev. 20

0-OP-ZZ-021, Severe Weather Preparation, Rev. 2

0-OSP-ZZ-001, Cold Weather Preparation, Rev. 14

Condition Reports (*NRC Identified)

500113 532539 532542 532543 532544 532545 533512 533537 534776

534779* 535462

Section 1R04: Equipment Alignment

Procedures

0-MOP-AAC-002, Return to Service of the AAC Diesel Generator, Rev. 19

0-MOP-AAC-001, Removal from Service of the AAC Diesel Generator, Rev. 17

0-OP-AAC-001A, AAC Diesel Generator Systems Alignment, Rev. 7

0-OP-FC-001A, Spent Fuel Pit Cooling System Alignment, Rev. 7

1-OP-CS-001A, Containment Spray System Alignment, Rev. 7

1-OPT-FW-003, Turbine Driven Auxiliary Feedwater Pump 1-FW-P-2, Rev. 44

Drawings

11448-FM-068A SH 3, Flow/Valve Operating Number Diagram Feedwater System Surry Unit 1,

Rev. 60

11448-FM-084A SH 2, Flow/Valve Operating Number Diagram Containment Spray System

Unit 1, Rev. 49

Condition Reports (*NRC Identified)

2794 533114*

Section 1R05: Fire Protection

Procedures

0-FS-FP-161, Auxiliary Building - Elevation 27 Feet - 6 Inches, Rev. 2

0-FS-FP-116, Control Room - Elevation 27 Feet - 6 Inches, Rev. 4

1-FS-FP-109, Battery Room 1A - Unit 1 Elevation 27 Feet - 6 Inches, Rev. 2

2-FS-FP-168, Turbine Basement - Unit 2 Elevation 9 Feet - 6 Inches, Rev. 1

Condition Reports (*NRC Identified)

535598* 535607* 535609* 535338 535490*

Section 1R06: Flood Protection

Procedures

0-AP-13.00, Turbine Building or MER 3 Flooding, Rev. 28

0-NSP-CC-005, CCHX Tests Using the Temporary Monitoring System, Rev. 1

1-MOP-SW-003, Operation of the Service Water Jumper, Rev. 2

Condition Reports (*NRC Identified)

530896* 531322

Other Documents

2-615, Virginia Electric and Power Company Surry Power Station Units 1 and 2 Proposed

License Amendment Request Regarding Temporary Service Water Jumper to the Component

Cooling Heat Exchangers, 09/26/12

Safety Evaluation Related to Amendment No. 279 to Renewed Facility Operating License Nos.

DPR-32 and DPR-37 Virginia Electric and Power Company Surry Power Station, Unit Nos. 1

and 2 Docket Nos. 50-280 and 50-281, 09/23/13

Section 1R07: Heat Sink Performance

Procedures

0-NSP-CC-005, CCHX Tests Using the Temporary Monitoring System, Rev. 1

Condition Reports

530290 528965 529922 530896 532319 533016 533442

Other Documents

VEPCO Heat Exchanger Specification Sheet for Component Cooling Water Heat Exchanger

Proto-HX Heat Exchanger Calculation Reports for Surry Unit 1 RFO 25

ETE-SU-2012-1016, Component Cooling Water Heat Exchanger Performance Testing

Section 1R08: Inservice Inspection Activities

Procedures

0-NSP-RC-003, Visual examination of Reactor Pressure Vessel Bottom Mounted

Instrumentation (BMI), Surry Unit 1 & 2, Rev. 2

54-ISI-370-003, Nondestructive Examination Procedure, Areva NP, Remote Underwater Visual

examination of Westinghouse Reactor Pressure Vessel Internals for Pressurized Water

Reactors in accordance with MRP-228 (Inspection Standard for PWR Internals)

Areva 03-9177821, Secondary Side Visual Inspection Plan and Procedure for Dominion, Surry

1R25, Rev. 3

Areva ETSS_BOB001_MIZ80_R0, Surry Unit 1, Outage 1R25, 10/25/13

Areva ETSS_RPC001_MIZ80_R0, Surry Unit 1, Outage 1R25, 10/25/13

Areva ETSS_XP001_2X19_MIZ80_R0, Surry Unit 1, Outage 1R25, 10/25/13

SRY-SGPMS-002, Surry Site Specific Eddy Current Analysis Guidelines, Surry 1R25,

October 2013

Engineering/Technical Evaluations

Areva Document 51-9210968-000, Surry Unit 1 1R25 - EPRI Appendix H/I Eddy Current

Technique Review, Rev. 0

ETE-SU-2010-0029, Condition Monitoring and Operational Assessment Unit 1 CMOA Fall 2010,

Rev. 0

ETE-SU-2012-0011, Steam Generator Condition Monitoring and Operational Assessment Surry

Unit 1, Rev. 0

ETE-SU-2013-0052, Steam Generator Degradation Assessment, Rev. 0

Drawings

B&W Drawing No. 134816E, Primary Inlet Nozzle, Section A-A, Rev. 3

DWG No. 1B79662, Calibration Standard ADVB-031-96, Rev. 1

DWG No. 1B79667, Calibration Standard ADVB-036-96, Rev. 1

DWG No. 1B80274, Calibration Standard EP5-011-098, 02/28/98

DWG No. 1B80279, Calibration Standard EP5-016-098, 02/28/98

DWG No. 1B81049, Eddy Current Sizing Standards Assembly & Detail, Rev. 0

DWG No. 9103696B, Surry .875 X 0.50 ASME-AVB-Expansion Cal Std, Rev. 0

DWG No. 9103697B, Surry .875 X 0.50 ASME-AVB-Expansion Cal Std, Rev. 0

DWG No. 9103698B, Surry .875 X 0.50 EDM Cal Std As Built Drawing, Rev. 0

DWG No. 9103699B, Surry .875 X 0.50 EDM Cal Std As Built Drawing, Rev. 0

DWG No. CB02221.DWG, PDI Alternative ASME Calibration Block, Rev. 0

Stone and Webster Drawing No. 11448-FV-7A, Reactor Neutron Shield Tank Assembly, Rev. 6

Stone and Webster Drawing No. 11448-FV-7D, Reactor Neutron Shield Tank Sheet 3, Reactor

Sliding Foot Assembly Section, Rev. 6

Corrective Action Documents

CR 473993, Boric Acid on 1-RC-PCV-1455A (Body to Bonnet)

CR 474000, Boric Acid on 1-SI-85 (Body to Bonnet)

CR 474220, ISI Rejection of Integral Attachment Weld Due to Incomplete Weld

CR 474792, Improper Thread Engagement Was Noted on Anchor Bolt for Constant Support

CR 474284, ISI Rejection of Pipe Support for Structural Corrosion and Thread Engagement

CR 474808, ISI Rejection of Pipe Support for Incorrect Clearances, Corrosion, and Weld Loss

CR 474875, Improper Clearances and Missing Field Welds on Pipe Hanger

CR 476209, Steam Generator Riser Barrel Discoloration

CR 476520, Foreign Object Identified During FOSAR Inspection

CR478573, Boric Acid on 1-CH-P-1C Inboard and Outboard Seals

CR 489853, SGMP has issued a Needed Guidance per NEI 03-08

CR 496300, ACE 019326, Foreign Object Identified in 2-RC-E-1A

CR 501733, Submit WO to Perform 1-RC-E-1B Steam Drum Inspection for FME

CR 502848, Submit WO to Perform 1-RC-E-1C Steam Drum Inspection for FME

CR 529738, Boric Acid Leaking from 1-SI-495 (Drain Valve) (Active)

CR 529764, Boric Acid Discovered on 1-SI-102 (Packing) (Inactive)

CR 529775, Boric Acid on 1-RC-P-1A (Seal and Flange)

CR 529783, Boric Acid Leakage on 01-RC-HCV-1557C

CR 529802, Boric Acid Identified on 1-RC-ICV-3026. (Packing/Bonnet) (Inactive)

CR 529813, Boric Acid Identified on 1-RC-ICV-3085 (Fitting) (Inactive)

CR 529823, Boric Acid Identified in Drain Catch Downstream of 1-RC-ICV-3066/3067

CR 529937, Boric Acid Identified on 1-RC-ICV-3142 (Packing) (Active)

CR 530267, Loose bolt on Rx Support

CR 530542, ISI Rejectable Conditions Noted on Spring Support 1-WFPD-H006B

CR 530547, ISI Noted Condition on Spring Support 1-WFPD-H006A

CR 532007, Corrosion Observed on 1-RC-FC-1482C

CR 532305, Boric Acid on 1-RC-FC-1480B (Plug)

CR 533049, Open Item from NRC ISI Inspection

Other Documents

Areva Certificate of Calibration 39936, Eddy Current Tester MIZ-80 S/N 024, 08/15/13

Areva Certificate of Calibration 39938, Eddy Current Tester MIZ-80 S/N 051, 08/15/13

Areva Certificate of Calibration 39939, Eddy Current Tester MIZ-80 S/N 059, 08/15/13

Areva Certificate of Calibration 40136, Eddy Current Tester MIZ-80 S/N 011, 08/28/13

Areva Document 51-9208139-000, Surry Unit 1 - 1R25 ECT Inspection Plan, Rev. 0

Areva S1R25-Memo-001, Surry 1 1R25 Qualified Data Analysts, 10/26/13

ASME Boiler and Pressure Vessel Code Case, N-722-1, ASME Approved date, 01/26/09

CEM-0142, Sensitivity Evaluation of the Structural Adequacy of the Reactor Vessel Sliding Foot

Assemblies for the loss of Multiple Cap Screws, SPS Unit-1, 11/15/13

Certificate of Visual Examination, ID Number C6756, Expires August 5, 2014

Certificate of Visual Examination, ID Number Y4315, Expires August 26, 2014

Certification ID Number CY4315, Areva Incorporated, Certificate of Personnel Qualification,

Rev. 30

Certification ID Number C6756, Areva Incorporated, Certificate of Personnel Qualification,

Rev. 23

Certified Materials Test Report for Spool arc 65, Class ER70S-2 weld wire for GTAW

Dominion Corporate Welding Manual, Performance Qualification Test Methods, Performance

Qualification Method No. 501A, GTAW

ER-AP-BAC-10, Boric Acid Corrosion Control Program, Rev. 10

ER-AP-BAC-101, Boric Acid Corrosion Control Program (BACCP) Inspections, Rev. 9

ER-AP-BAC-102, Boric Acid Corrosion Control Program (BACCP) Evaluations, Rev. 10

IOD-000191, Immediate Operability Determination for CR 530267, 11/15/13

Inservice Inspection Owners Activity Report for Surry Unit 1 Refueling Outage S1R24

NDE Personnel Certification Records:

B. Knox,
K. Newell,
R. Vojvodic,
M. Story,
M. Brown,
P.

Anderson,

J. Jacobs,
A. Jensen,
T. Coburn,
B. Zollner,
T. Thomas,
D. Strickland, M. True

NDE Personnel Qualification and Certification Record for VT Examinations for leakage on lower

head in accordance with N-722, 2013 Lower Head Examiner, 3/12/2013

Program Health Report, Steam Generator, Period Q3-2013, 10/23/2013

SAR000480, Boric Acid Corrosion Control Program (BACCP) Formal Self-Assessment,

10/25/10

OPEX002605, IN 2010-05 Management of Steam Generator Loose Parts Automated Eddy

Current Data Analysis, 02/11/10

OPEX002748, IN 10-21 Crack-Like Indication in the U-bend Region of a Thermally Treated

Alloy 600 Steam Generator Tube, 10/15/10

OPEX003137, IN 12-07 Tube-to-Tube Contact Resulting in Wear in Once-Through Steam

Generators, 08/01/12

OPEX003176, IN 13-11 Crack-Like Indications at Dents/Dings and in the Free span Region of

Thermally Treated Alloy 600 Steam Generator Tubes, 7/11/2013

SAR-000841, Welder Qualification, 09/29/09

Surry Power Station Unit-1, October 2013 Maintenance Outage Reactor Internals Inspection

Schedule, 10/09/13

WCAP-15988-NP, Generic Guidance for an Effective Boric Acid Inspection Program for

Pressurized Water Reactors, Rev. 2

Welding Technique Sheet for Welding Technique Number 103, GTAW Manual Welding

Zetec Certificate of Authenticity, Areva/Surry 1, 10/02/13

Zetec Certificate of Conformance, Shipment ID 14143, Contract 1013053297, 10/10/13

Zetec Certificate of Conformance, Shipment ID 14366, Contract 1013061214, 09/26/13

Section 1R11: Licensed Operator Requalification Program

Procedures

1-GOP-2.7, Unit Shutdown, Power Decrease from Allowable Power to Unit Offline for Refueling

Outage, Rev. 21

1-OP-RC-004, Draining the RCS to Reactor Flange Level, Rev. 34

1-OP-RX-009, Dilution to Critical Conditions Following Refueling, Rev 21

OP-AA-100, Conduct of Operations, Rev. 25

Section 1R12: Maintenance Effectiveness

Procedures

0-MPT-0704-01, Diesel Fire Pump Engine Inspection, Rev. 17

0-OPT-FP-009, Diesel Driven Fire Protection Water Pump 1-FP-P-2, Rev. 22

CM-AA-FPA-100, Fire Protection Appendix R Program, Rev. 8

ER-AA-MRL-10, Maintenance Rule Program, Rev. 5

ER-AA-MRL-100, Implementing Maintenance Rule, Rev. 6

Condition Reports

463678 481455 504410 504427 520001 523968 528885 529424 534028

534105 535283 535566

Other Documents

ACE 19512, AAC Jacket Water Heater Failed to Heat, 08/30/13

DC-SU-12-01121, Replacement of Fuses for AAC Diesel Jacket Keep Warm Heater 0-BCW-

HTR-1, Rev. 0

ME-0977, Fire Protection: Fire Main Loop Pressure Calculation, Rev. 0

SDBD-SPS-FP, System Design Basis Document for Fire Protection System SPS, Rev. 12

System Health Report, Black Out Diesel System, Q3-13

RCE 000027, Diesel Fire Pump Overspeed Trip, 08/06/07

Work Orders

38046801201 38046801301

Section 1R13: Work Control

Procedures

ADM-OU-SU-201, Shutdown Safety Assessment Checklist, Rev. 8

WM-AA-100, Work Management, Rev. 22

1-OPT-ZZ-002, ESF Actuation with Undervoltage and Degraded Voltage 1J Bus, Rev. 37

Other Documents

Shutdown Risk Review Emergent Work Evaluation & High Risk Evolution Contingency Plan -

C RSST Switching at Lowered Inventory, 11/08/13

Section 1R15: Operability Determinations and Functionality Assessments

Procedures

1-OPT-CS-004, Containment Spray Check Valves, Rev. 8

2-OPT-CS-002, Containment Spray System Test, Rev. 16

1-OPT-CT-201, Containment Isolation Valve Local Leak Rate Testing (Type C Containment

Testing), Rev. 21

CM-AA-2, Design Change Expectations, Rev. 0

DNES-AA-GN-1001, Engineering Review, Rev. 3

CM-AA-DDC-201, Design Changes, Rev. 12

Condition Reports

28665 530886 532703 530267 529007 517409 518809 532802 454236

454039

Other Documents

CE-1642, The Effect of Reactor Pressure Vessel (RPV) Head Replacement on the RPV Support

System (Neutron Shield Tank), Surry Power Station Units 1 and 2, Rev. 0

CE-1653, Review of Structural Adequacy of the Reactor Vessel Support Sliding Foot

Assemblies - Surry Units 1 and 2, Rev. 0

ETE-CME-2013-0010, Evaluation of SPS Check Valves 1-RS-11, 1-RS-17, 1-CS-13, 1-CS-24,

2-RS-11, 2-RS-17, 2-CS-13, and 2-CS-24 in their Current Configuration

2-CS-P-1A and 2-CS-P-1B Performance Test History, 09/11-11/13

ETE-SU-2012-1015, Oil Foaming in the Charging Pump Sump, Rev. 0

SM-1476, Surry GOTHIC Analysis of NPSH Available for the LHSI and RS Pumps, Addendum

H, Rev. 1

Work Order 38103146704

Section 1R18: Plant Modifications

Procedures

1-OPT-CT-201, Containment Isolation Valve Local Leak Rate Testing (Type C Containment

Testing), Rev. 21

CM-AA-2, Design Change Expectations, Rev. 0

DNES-AA-GN-1001, Engineering Review, Rev. 3

CM-AA-DDC-201, Design Changes, Rev. 12

Temporary Modifications

SU-11-00017, CCHX SW Inlet Pipe Repair & 1-SW-MOV-102A/B Replacements, Rev. 0

Permanent Modifications

SU-12-00005, SPS Flex Equipment, Rev. 0

SU-12-00022, Flex BDB Mechanical Connections, Rev. 0

SU-13-01019, BDB-Flex Power for Essential Instrumentation and Equipment, Rev. 7

Other Documents

ETE-CEP-2012-0005, Design and Licensing Basis Review of the Surry Seismic and Flooding

Requirements Related to the March 12, 2012, NRC 50.54(f) Request for Information, Rev. 2

ETE-CEP-2012-0011, Beyond Design Basis - FLEX Strategy Overall Integrated Plan Basis

Document, Rev. 0

ETE-CEP-2012-0020, Surry Units 1 & 2 - Reliable Spent Fuel Pool Instrumentation Project

Documentation, Rev. 0

ETE-CEP-2012-0030, Input Required For Mechanical Work Scope to Facilitate Implementation

of Phase 1 Coping Strategy of NEI-12-06 Diverse and Flexible Coping Strategies

(FLEX) Implementation Guide For Surry Power Station, Rev. 1

50.59 Screening Form for DC-SU-00022, Flex BDB Mechanical Connections

Condition Reports (*NRC Identified)

530290 530512 530896* 530935 531248 532319 532478 532712 533401*

Work Orders

38103284558 38103312079

Drawings

11448-FM-071A SH 2, Flow/Valve Diagram Circulating & Service Water Systems Surry

Unit 1, Rev. 103

11448-FM-071A SH 4, Flow/Valve Diagram Circulating & Service Water Systems Surry

Unit 1, Rev. 36

11448-FM-089A SH 2, Flow/Valve Operating Numbers Diagram Safety Injection System Surry

Unit 1, Rev. 55

11448-FM-089B SH 4, Flow/Valve Operating Numbers Diagram Safety Injection System Surry

Unit 1, Rev. 25

200022-1-M-803, Piping Isometric BDB Modification RCS Make-Up Injection Connection Surry

Unit 1, Rev. 4

Section 1R19: Post Maintenance Testing

Procedures

0-OP-SW-007, Emergency Service Water Pump 1-SW-P-1A Comprehensive Test, Rev. 14

1-OPT-FW-003, Turbine Driven Auxiliary Feedwater Pump 1-FW-P-2, Rev. 44

0-OPT-VS-006, Flow Switches FS-VS-117A and FS-VS-117B Test, Rev. 10

0-OP-VS-002, Auxiliary Building Ventilation System, Rev. 21

1-OPT-CT-201, Containment Isolation Valve Local Leak Rate Testing (Type C Containment

Testing), Rev. 21

1-PT-18.8, Charging Pump Service Water Performance, Rev. 35

1-EPT-106-02, Main Station Battery 1B Service Test, Rev. 23

Condition Reports

535056 522094 530792 531460 531495

Maintenance Orders/Work Orders

38103178809 38103248693

Section 1R20: Refueling and Outage Activities

Procedures

1-GOP-2.7, Unit Shutdown, Power Decrease from Allowable Power to Unit Offline for Refueling

Outage, Rev. 24

1-OP-RC-004, Draining the RCS to Reactor Flange Level, Rev. 34

1-OP-RX-009, Dilution to Critical Conditions Following Refueling, Rev 21.

Condition Reports (*NRC Identified)

29782 529912 530494 530507 531543 531608 531860 532200 532217

2232 532236 532733 532774* 532782* 532802 532824* 533322 533347

Other Documents

Surry Unit 1 2013 RFO Shutdown Risk Review Report, Rev. 0

Section 1R22: Surveillance Testing

Procedures

1-OPT-CT-201, Containment Isolation Valve Local Leak Rate Testing (Type C Containment

Testing), Rev. 21

1-OPT-ZZ-002, ESF Actuation with Undervoltage and Degraded Voltage - 1J Bus, Rev. 37

2-OPT-RC-10.01, Reactor Coolant Leakage - Manually Calculated, Rev. 16

2-OPT-RC-10.00, Reactor Coolant Leakage - Computer Calculated, Rev. 40

Condition Reports

25641 529887 529437 529900 530027 530047 530048 531506 532624

2668 533118 533152 534915

Work Orders

38103237838

Other Documents

EWR 91-011, RC Leakage Calculation Revisions/Surry/1&2, 03/28/91

Technical Report NE-1381, Evaluation of Surry Power Station RCS Leak Rate Calculation,

Rev. 0

Section 2RS1: Radiological Hazard Assessment and Exposure Controls

Procedures, Guidance Documents, and Manuals

C-HP-1032.051, Airborne Radioactivity Counting and Analysis, Rev. 2

C-HP-1032.080, Controlled Area and Unrestricted Area Radiological Surveys, Rev. 9

HP-1032.110, Standard Radiation Monitoring & Dose Rate Trending, Rev. 0

PI-AA-200, Corrective Action Program, Rev. 21

RP-AA-106, Radiological Work Control Program, Rev. 2

RP-AA-109, Radiological Survey Program, Rev. 0

RP-AA-220, Radiological Survey Scheduling, Rev. 1

RP-AA-222, Radiation Surveys, Rev. 1

RP-AA-223, Contamination Surveys, Rev. 3

RP-AA-226, Alpha Monitoring, Rev. 3

RP-AA-233, Control of General License Devices, Rev. 0

RP-AA-240, Discrete Radioactive Particle Control, Rev. 1

RP-AA-275, Radiological Risk Assessment Process, Rev. 0

VPAP-2101, Radiation Protection Program, Rev. 34

0-HSP-INST-002, DAC Value Calculations and Instrument Sensitivity Determination, Rev. 0

0-HPT-ISFSI-001, Independent Spent Fuel Storage Installation (ISFSI) Quarterly Surveillance,

Rev. 13

0-HPT-ISFSI-002, NUHOMS Dry Spent Fuel Storage System Surveillance, Rev. 4

0-HPT-ISFSI-003, CASTOR V/21, CASTOR X/33, NAC-128, MC-10 & TN-32 Dry Storage Cask

Surveillance Requirements, Rev. 1

0-HPT-LKTEST-001, Health Physics Source Leak Test, Rev. 6

Records and Data

Air Sample Count Room Log, 5/19/2013 to 10/24/2013

Air Sample Results, U-1 Transfer Canal, Sample ID 13-2518-1021-1651, 10/21/2013

ALARA Review Evaluations:

13-008, 2013 U-1 RFO RP Routine & Radioactive Material Control Support, 9/23/2013

13-013, 2013 U-1 NDE Inspections, 9/25/2013

13-021, 2013 U-1 RFO Lower Internals Lift & Set, Draft

13-022, 2013 U-1 RFO Sliding Foot inspection, 9/29/2013

13-023, 2013 U-1 RFO Fuel Transfer Cart repairs, 9/25/2013

13-025, 2013 U-1 RFO 10 Year lSI MRP-227 Girth Weld Inspections, 9/25/2013

Argos Radon rejection setting recommendations, email from vendor, 4/24/2012

Gamma Spectrum Analysis, Sample ID 23-OCT-2013-0085, 10/23/2013

List of Non-Fuel items stored in the Spent Fuel Pool, 10/1/2013

List of the 10 most exposure significant work areas in the plant, 10/2/2013

List of 2013 Unit 1 Outage Radiation Work Permits, 10/3/2013

List of Alarm Setpoints for RCA Exit and Release Monitors (Personnel Contamination, Portal,

and Small Article Monitors), 10/24/2013

NTS Annual Inventory Report for Surry Power Station, 1/8/2013

Preliminary Committed Effective Dose Equivalent (CEDE) - Inhalation Intake Data, 10/23/2013

Radiation Protection, Health Physics Operations Turnover Report, 11/23/2013

Radiation Work Permits (RWPs):

13-0-2102, RP Outage Support, Rev. 0

13-0-2109, NDE/ISI Inspections, Rev. 0

13-0-2505, Transfer Canal Activities, Rev. 0

13-0-2516, Lower Internals Lift & Set - Unit 1, Rev. 0

13-0-2517, Sliding Foot Inspection, Rev. 0

13-0-2518, Fuel Transfer Cart Repairs, Rev. 0

13-0-2520, U-1 10 Year lSI MRP-227 Girth Weld Inspections, Rev. 0

Work Order (WO) 38103164478, Inventory of Nationally Tracked Sources by Health Physics,

1/8/2013

WO 3810333089, 0-HPT-ISFSI-001, Independent Spent Fuel Storage Installation (ISFSI)

Quarterly Surveillance, 7/15/2013

WO 38103366702, 0-HPT-ISFSI-001, Independent Spent Fuel Storage Installation (ISFSI)

Quarterly Surveillance, 10/15/2013

Radiological Surveys:

Aux Bldg -2' GATE 1 & 2 IX alley, Survey M-20130119-2 and M-20130820-2, 1/18/2013 and

8/20/2013

Auxiliary Building Tool Room Survey, Survey M-20130923-7, 9/23/2013 and 9/16/2013

Auxiliary Building 27 Process Vent line on overhead, Survey 20120813-2, 8/13/2012

Fuel Building 27, 15, 6, Survey M-20130801-2, 8/1/2013 and 9/11/2013

Fuel Building 45, Survey M-20130801-1, 8/1/2013 and 9/11/2013

ISFSI Pad #1, 8/22/13

ISFSI Pad #3/NUHOMS Bunker, 8/23/2013, 9/4/2013 and 9/9/2013

Service Building Tool Room Survey, 10/15/2013 and 10/22/2013

U-1 Containment Transfer Canal, 10/21/2013

U-1 Containment Reactor Cavity, 10/21/2013

U-1 Containment - Incore Sump Room, 10/22/2013

U-1 Containment B Loop Room, Contamination Survey of ISI Equipment, 10/23/2013

U-1 Charging Pumps -2, Survey M-20130825-1, 8/24/2013, 9/29/2013 and 10/22/13

U-1 2 elevation, A, B, and C Charging pump Cubes, 1/22/2013U-1 13 Pipe Chase

Knockout Drum Pre and Post Shielding, Survey M-20120803-4, 8/3/2012

U-1 -2 Elevation Knockout Drum drain line, 8/2/2012 and 8/7/2012

U-1 13 Cable Vault, Surveys M-20120803-5, M-20120807-4, and M-20120807-8, 8/2/2012,

8/3/2012 and 8/7/2012

U1R25 Daily Exposure Status, 10/21/2013

Corrective Action Program Documents

Apparent Cause Evaluation, ACE 019215, Aux. Building Dose Rates increase during Resin

Transfer, 8/29/2012

Apparent Cause Evaluation, ACE019240, Groundwater Protection Program Criteria Exceeded,

10/4/2012

Apparent Cause Evaluation, ACE019241, Special Nuclear Material not stored in the Protected

Area, 10/12/2012

Nuclear Oversight Audit 12-06, Radiological Protection/Process Control Program/Chemistry,

9/20/2012

Self-Assessment, SAA001754 Radioactive Source Control, 8/16/2012

Self-Assessment, SAR001710, Radiological Protection Technician Fundamentals, 12/10/2012

Self-Assessment, SAR002281, Radioactive Contamination Control Program, 9/18/2013

Condition Reports

2059 484956 485152 485847 510859 514234 523692

2RS8: Radioactive Material Processing and Transportation

Procedures, Manuals, and Guides

0-HSP-RESIN-001, Primary Resin Transfer Activities, Rev. 1

C-HP-1071.010, Control of Radioactive Sources, Rev. 6

C-HP-1071.030, Receiving Radioactive Material, Rev.3

C-HP-1071.040, Packaging and Shipment of Radioactive Material, Rev. 10

C-HP-1072.010, Packaging Radioactive Waste, Rev. 2

C-HP-1072.030, Computer Programs for Radwaste and Radioactive Material, Rev. 1

C-HP-1072.040, Radioactive Waste Disposal Using the Barnwell Disposal Facility, Rev. 9

C-HP-1072.050, Radioactive Waste Transfer To Licensed Waste Processors, Rev. 9

C-HP-1072.070, Radioactive Waste Disposal Using the Energy Solutions Containerized Waste

Facility, Rev. 5

C-HP-1072.071, Radioactive Waste Disposal Using the Energy Solutions Bulk Waste Facility,

Rev. 5

HP-1071.021, Storing Radioactive Material Outside the Protected Area, Rev. 10

HP-1071.022, Placing High Integrity Containers (HICs) Into Storage at the SRF HIC Storage

Area, Rev.0

HP-1072.020, Sampling, Analyzing, and Classifying Radioactive Waste, Rev. 6

HP-1072.082, Radioactive Waste Transfer To Studsvik Processing Facility, Erwin, TN, Rev. 2

RP-AA-108, Radioactive Material Control Program, Rev. 2 3.4.5.i

RP-AA-233, Control of General License Devices, Rev. 0

RP-AA-231, Radiological Control Areas, Rev.3

RP-AA-232, Radioactive Material Control, Rev. 4

PI-AA-200, Corrective Action, Rev. 21

Shipping Records and Radwaste Data

Shipments:

SH-2013-020 (Type A)

SH-2013-026 (Type A)

SV-2012-2 (Type B)

D-2013-1 (LSA)

D-2013-6 (SCO)

Waste Streams:

1-FC-FL-1B 2013 IL Fuel Pit Skimmer Filter

DAW 2012 IL Dry Active Waste Smears

Blend Tank 2012 IL Blended Class B Primary Resin

Catch Tank 2012 IL Blended Class A Primary Resin

RO 2012 Surry Radwaste Facility (SRF) Liquid Waste

EBT 2012 SRF Evaporator Bottoms Tank Liquid

1-CH-FL-2 2012 IL Unit 1 Reactor Coolant Filter

2-CH-FL-2 2012 IL Unit 2 Reactor Coolant Filter

Condition Reports

456050 485152 485014 485847 521582 523692

Section 1EP4: Emergency Action Level and Emergency Plan Changes

Change Packages

Emergency Plan, Revs. 57 and 58

Emergency Action Level Technical Bases Document, Rev 3

Evacuation Time Estimate Study Update

Section 4OA1: Performance Indicator Verification

Procedures, Guidance Documents, and Manuals

RP-AA-111, Monitoring and Improving Radiological Performance, Rev. 2

RP-AA-112, Radiation Safety Performance Indicator Reporting, Rev. 4

PI-AA-100-1000, Performance Indicators, Rev. 4

Records and Data

Regulatory Assessment Performance Indicators, Radiological Protection Surry Power Station,

for May 2012 through April 2013

Condition Reports

475959 476101 494281 498211

Section 4OA1: Performance Indicator Verification

Procedures

ER-AA-SPI-1001, Implementation of the Consolidated Data Entry (CDE) Reporting for Mitigating

System Performance Index (MSPI), Rev. 2

ER-AA-SPI-1002, Maintaining the MSPI Basis Document, Rev. 1

Other Documents

Monthly PI Reports with Associated Data, September 2012 to September 2013

Technical Report SE-0006, NRC MSPI Basis Document Surry Power Station, Rev. 1

Section 4OA2: Identification and Resolution of Problems

Procedures

OP-AA-100, Conduct of Operations, Rev. 25

SUADM-O-26, Administrative Control of Operational Components, Rev. 6

1-OPT-EG-001, EDG #1 Monthly Start Exercise Test, Rev. 59

Condition Reports

508823 508316 535283 503764 532802

Other Documents

ACE 019423, Failed ACC Mechanical Overspeed Trip System, 04/30/13

ACE 019381, 15H3 Failure During 1-OPT-EG-001, 03/26/13

RTE P-SURR-335912, AAC Electronic Speed Switch, 09/26/13

RAS 000226, Failed AAC Mechanical Overspeed Trip System, 03/21/13

U1/U2 Operations Aggregate Impact report

LIST OF ACRONYMS

AAC Alternating Current

ACE Apparent Cause Evaluation

ADAMS Agencywide Document Access and Management System

AFW Auxiliary Feedwater

ALARA As Low As Reasonably Achievable

BACC Boric Acid Corrosion Control

BDB Beyond Design Basis

BPVC Boiler and Pressure Vessel Code

CAP Corrective Action Program

CC Component Cooling

CDE Consolidated Data Entry

CEDE Committed Effective Dose Equivalent

CFR Code of Federal Regulations

CR Condition Report

CS Containment Spray

DC Design Change

DOT Department of Transportation

ED Electronic Dosimeter

EDG Emergency Diesel Generator

EVT Enhanced Visual Testing

EPIP Emergency Plan Implementing Procedures

EPRI Electric Power Research Institute

ETSS Examination Technique Specification Sheet

FLEX Diverse and Flexible Coping Strategies

FMS Fukushima Mitigating Strategies

HIC High Integrity Container

HP Health Physics

HRA High Radiation Area

HX Heat Exchanger

IMC Inspection Manual Chapter

IOD Immediate Operability Determination

IP Inspection Procedure

ISFSI Independent Spent Fuel Storage Installation

ISI In-Service Inspection

IST Integrated Service Test

LHRA Locked High Radiation Area

LHSI Low Head Safety Injection

LP Liquid Penetrant

MSPI Mitigating Systems Performance Indicator

NCV Non-cited Violation

NDE Non-Destructive Examination

NEI Nuclear Energy Institute

NPSH Net Positive Suction Head

NRC Nuclear Regulatory Commission

NRR Office of Nuclear Reactor Regulation

OD Operability Determination

OS Occupational Radiation Safety

OWA Operator Work Arounds

PARS Publicly Available Records

PD Performance Defiency

PI Performance Indicator

PM Periodic Maintenance

POD Prompt Operability Determination

PS Public Radiation Safety

RAS Reasonable Assurance of Safety

RCA Radiologically Controlled Area

RCE Root Cause Evaluation

RCS Reactor Coolant System

RFO Refueling Outage

RHR Residual Heat Removal

RMC Radioactive Material Control

RP Radiation Protection

RPV Reactor Pressure Vessel

RS Recirculation Spray

RSST Reserve Service Station Transformer

RTP Rated Thermal Power

RWP Radiation Work Permit

SDP Significance Determination Process

SG Steam Generator

SRF Surry Radwaste Facility

SSC System, Structure or Component

SW Service Water

TDAFWP Turbine Driven Auxiliary Feedwater Pump

TIA Task Interface Agreement

TS Technical Specifications

UFSAR Updated Final Safety Analysis Report

URI Unresolved Item

UT Ultrasonic Testing

VEPCO Virginia Electric and Power Company

VHRA Very High Radiation Area

VPAP Virginia Power Administrative Procedure

VT Visual Testing

WO Work Order

Attachment

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