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| number = ML17013A253
| number = ML17013A253
| issue date = 01/12/2017
| issue date = 01/12/2017
| title = Clinton Power Station - NRC Component Design Bases Inspection, Inspection Report 05000461/2016009
| title = NRC Component Design Bases Inspection, Inspection Report 05000461/2016009
| author name = Jeffers M
| author name = Jeffers M
| author affiliation = NRC/RGN-III/DRS/EB2
| author affiliation = NRC/RGN-III/DRS/EB2
| addressee name = Hanson B C
| addressee name = Hanson B
| addressee affiliation = Exelon Generation Co, LLC, Exelon Nuclear
| addressee affiliation = Exelon Generation Co, LLC, Exelon Nuclear
| docket = 05000461
| docket = 05000461
Line 18: Line 18:


=Text=
=Text=
{{#Wiki_filter:
{{#Wiki_filter:UNITED STATES ary 12, 2017
[[Issue date::January 12, 2017]]


Mr. Bryan Senior VP, Exelon Generation Company, LLC President and CNO, Exelon Nuclear 4300 Winfield Road Warrenville, IL 60555
==SUBJECT:==
CLINTON POWER STATION - NRC COMPONENT DESIGN BASES INSPECTION, INSPECTION REPORT 05000461/2016009


SUBJECT: CLINTON POWER STATION - NRC COMPONENT DESIGN BASES INSPECTION, INSPECTION REPORT 05000461/2016009
==Dear Mr. Hanson:==
On December 1, 2016, the U.S. Nuclear Regulatory Commission (NRC) completed a Component Design Bases Inspection at your Clinton Power Station. Because Exelon Generation Company, LLC, certified to the NRC that it decided to permanently cease power operations at Clinton Power Station by June 1, 2017, on letter titled Certification of Permanent Cessation of Power Operations, dated June 20, 2016, this Component Design Bases Inspection was adjusted to perform a more detailed assessment of performance in areas potentially impacted by the proposed shutdown. This inspection adjustment was consistent with the guidance contained in Inspection Manual Chapter 2515, Light-Water Reactor Inspection Program-Operations Phase, dated February 1, 2016. The enclosed report documents the results of this inspection, which were discussed on December 1, 2016, with Mr. B. Kapellas, Plant Manager, and other members of your staff.
 
Based on the results of this inspection, six NRC-identified findings of very-low safety significance were identified. These findings involved violations of NRC requirements. However, because of their very-low safety significance, and because the issues were entered into your Corrective Action Program, the NRC is treating the issues as Non-Cited Violations in accordance with Section 2.3.2 of the NRC Enforcement Policy.
 
If you contest the subject or severity of these Non-Cited-Violations, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001, with copies to the Regional Administrator, Region III; the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC Resident Inspector at the Clinton Power Station. In addition, if you disagree with the cross-cutting aspect assigned to any finding 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 III, and the NRC Resident Inspector at the Clinton Power Station. In accordance with Title 10 of the Code of Federal Regulations (10 CFR) 2.390, Public Inspections, Exemptions, Requests for Withholding, of the NRC's "Rules of Practice," a copy of this letter, its enclosure, and your response (if any) will be available electronically for public inspection in the NRCs Public Document Room or from the Publicly Available Records (PARS)
component of the NRC's Agencywide Documents Access and Management System (ADAMS).


==Dear Mr. Hanson:==
ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
On December 1, 2016, the U.S. Nuclear Regulatory Commission (NRC) completed a Component Design Bases Inspection at your Clinton Power Station. Because Exelon Generation Company, LLC, certified to the NRC that it decided to permanently cease power operations at Clinton Power Station by June 1, 2017, Certification of Permanent Cessation of Power OperationsJune 20, 2016, this Component Design Bases Inspection was adjusted to perform a more detailed assessment of performance in areas potentially impacted by the proposed shutdown. This inspection adjustment was consistent with the guidance contained in Inspection Manual Chapter -Water Reactor Inspection Program-The enclosed report documents the results of this inspection, which were discussed on December 1, 2016, with Mr. B. Kapellas, Plant Manager, and other members of your staff. Based on the results of this inspection, six NRC-identified findings of very-low safety significance were identified. These findings involved violations of NRC requirements. However, because of their very-low safety significance, and because the issues were entered into your Corrective Action Program, the NRC is treating the issues as Non-Cited Violations in accordance with Section 2.3.2 of the NRC Enforcement Policy. If you contest the subject or severity of these Non-Cited-Violations, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001, with copies to the Regional Administrator, Region III; the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC Resident Inspector at the Clinton Power Station. In addition, if you disagree with the cross-cutting aspect assigned to any finding 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 III, and the NRC Resident Inspector at the Clinton Power Station. In accordance with Title 10 of the Code of Federal Regulations of this letter, its enclosure, and your response (if any) will be available electronically for public ent Room or from the Publicly Available Records (PARS) component of the NRC's Agencywide Documents Access and Management System (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).


Sincerely,/RA/ Mark Jeffers, Chief Engineering Branch 2 Division of Reactor Safety Docket No. 50461 License No. NPF62
Sincerely,
/RA/
Mark Jeffers, Chief Engineering Branch 2 Division of Reactor Safety Docket No. 50-461 License No. NPF-62


===Enclosure:===
===Enclosure:===
IR 05000461/2016009 cc: Distribution via LISTSERV Enclosure U.S. NUCLEAR REGULATORY COMMISSION REGION III Docket No: 50461 License No: NPF62 Report No: 05000461/2016009 Licensee: Exelon Generation Company, LLC Facility: Clinton Power Station Location: Clinton, IL Dates: October 31 - December 1, 2016 Inspectors: Robbins Approved by: M. Jeffers, Chief Engineering Branch 2 Division of Reactor Safety Table of Contents  
IR 05000461/2016009
 
REGION III==
Docket No: 50-461 License No: NPF-62 Report No: 05000461/2016009 Licensee: Exelon Generation Company, LLC Facility: Clinton Power Station Location: Clinton, IL Dates: October 31 - December 1, 2016 Inspectors: N. Féliz Adorno, Senior Reactor Inspector, Lead J. Benjamin, Senior Reactor Inspector, Operations L. Rodriguez, Reactor Inspector, Mechanical J. Robbins, Reactor Inspector, Electrical Approved by: M. Jeffers, Chief Engineering Branch 2 Division of Reactor Safety Enclosure
 
Table of Contents


=SUMMARY=
=SUMMARY=
Line 39: Line 51:


==REACTOR SAFETY==
==REACTOR SAFETY==
6
 
{{a|1R21}}
{{a|1R21}}
==1R21 Component Design Bases Inspection==
==1R21 Component Design Bases Inspection==
{{IP sample|IP=IP 71111.21}}
{{IP sample|IP=IP 71111.21}}
==OTHER ACTIVITIES==
==OTHER ACTIVITIES==
20
 
{{a|4OA2}}
      {{a|4OA2}}
==4OA2 Identification and Resolution of Problems 20 4OA6==
==4OA2 Identification and Resolution of Problems==
Management Meeting 26  
 
      {{a|4OA6}}
==4OA6 Management Meeting 26==


=SUPPLEMENTAL INFORMATION=
=SUPPLEMENTAL INFORMATION=


==KEY POINTS OF CONTACT==
==KEY POINTS OF CONTACT==
Attachment  
Attachment
 
==LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED==
==LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED==
Attachment  
Attachment
 
==LIST OF DOCUMENTS REVIEWED==
==LIST OF DOCUMENTS REVIEWED==
2
2
==LIST OF ACRONYMS==
: [[USED]] [[11]]
: [[SUMMAR]] [[Y Inspection Report 05000461/2016009, 10/31/2016  12/01/2016; Clinton Power Station; Component Design Bases Inspection. The inspection was a 3-week onsite baseline inspection that focused on the design of components. Because Exelon Generation Company, LLC, certified to the]]
: [[U.S.]] [[Nuclear Regulatory Commission (]]
: [[NRC]] [[) that it decided to permanently cease power operations at Clinton Power Station by June 1, 2017, this Component Design Bases Inspection was adjusted to perform a more detailed assessment of performance in areas potentially impacted by the proposed shutdown, as allowed by Inspection Manual Chapter (IMC) -Water Reactor Inspection Program-The inspection was conducted by a team of four regional engineering inspectors. Six Green findings were identified by the team. These findings were considered Non-Cited Violations (NCVs) of]]
: [[NRC]] [[regulations. The significance of inspection findings is indicated by their color (i.e., greater than Green, or Green, White, Yellow, Red) and determined using]]
: [[IMC]] [[Process, April 29, 2015. Cross-cutting aspects are determined using]]
: [[IMC]] [[0310, -Cutting Areas,d December 4, 2014. All violations of]]
: [[NRC]] [[November 1, 2016. The NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-,dated July 2016. NRC-Identified and Self-Revealed Findings Cornerstone:  Mitigating Systems Green:  The team identified a finding of very-low safety significance (Green) and an associated]]
: [[NCV]] [[of Title 10 of the Code of Federal Regulations (]]
: [[CFR]] [[), Part 50, Appendix ify that the incapability of the residual heat removal (RHR) design to support Technical Specifications (TS) operability requirements was a condition adverse to quality. Specifically, when reactor water temperature was greater than 150 degrees Fahrenheit,]]
: [[RHR]] [[could not be realigned from shutdown cooling mode of operations to provide the]]
: [[TS]] [[required functions of the emergency core cooling system, suppression pool cooling, containment spray, and feedwater leakage control system. The licensee captured this issue in their Corrective Action Program (CAP) as Action Request (AR) 02742439 and]]
: [[AR]] [[03948042, and planned to submit a License Amendment Request to align]]
: [[TS]] [[requirements with the design capabilities. The performance deficiency was determined to be more-than-minor because it was associated with the Mitigating Systems cornerstone attribute of design control and adversely affected the associated cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the performance deficiency resulted in voluntarily declaring]]
: [[TS]] [[functions inoperable when performing shutdown cooling operations, which did not ensure the associated mitigating systems availability or capability to respond to an initiating event. The team determined that this finding was of very low safety significance (Green). Specifically, there were no known instances where the finding:  (1) represented a loss of system safety function; (2) represented an actual loss of safety function of at least a single train or two separate safety systems out-of-service for greater than their]]
TS allowed outage time; (3) involved non-TS trains of
equipment; (4) involved a degradation of a functional
: [[RHR]] [[auto-isolation on low reactor vessel level; (5) impacted external event protection; or (6) involved fire brigade issues. The team did not identify a cross-cutting aspect associated with this finding because it did not reflect current licensee performance since the performance deficiency occurred more than 3 years ago.  (Section 4]]
: [[OA]] [[2.b(1)) Cornerstone:  Barrier Integrity Green. The team identified a finding of very-low safety significance (Green) and an associated]]
: [[NCV]] [[of 10]]
: [[CFR]] [[Part 50, Appendix B, licensee failure to use a technically appropriate analytical methodology in the control room radiological habitability calculation. Specifically, the licensee used a methodology that inappropriately characterized the control room heating, ventilation and air-conditioning (HVAC) system outside air intake design resulting in a calculated control room dose following a loss of coolant accident that exceeded the applicable limit. The licensee captured this issue in their]]
: [[CAP]] [[as]]
: [[AR]] [[02742442, completed an operability evaluation, and issued an]]
: [[NRC]] [[event notification. The performance deficiency was determined to be more-than-minor because it was associated with the Barrier Integrity cornerstone attribute of design control and affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the performance deficiency resulted in the control room expected dose following a loss of coolant accident to exceed the applicable limits prompting an operability evaluation. The finding screened as of very-low safety significance (Green) because it only represented a degradation of the radiological barrier function provided for the control room. Specifically, the finding did not affect the control room barrier function against smoke or a toxic atmosphere. The team did not identify a cross-cutting aspect associated with this finding because it was not confirmed to reflect current performance due to the age of the performance deficiency. Specifically, the affected calculations were performed more than 3 years ago.  (Section 1R21.3.b(1)) Green. The team identified a finding of very-low safety significance (Green) and an associated]]
: [[NCV]] [[of Paragraph (b)(2)(i) of]]
: [[10 CFR]] [[50.65scope non-safety related mitigating structure, systems, and components (]]
: [[SSC]] [[) used within an emergency operating procedure (EOP) into Maintenance Rule Program. Specifically, an]]
: [[EOP]] [[used spent fuel pool (]]
: [[SFP]] [[) low-level and high-temperature parameters as distinct entry criteria but the associated components were not included in the scope of the Maintenance Rule Program. The licensee captured the team concerns in their]]
: [[CAP]] [[as]]
: [[AR]] [[02736193, performed an extent of condition to identify any other]]
: [[SSC]] [[addition to the]]
: [[EOP]] [[requiring them to be added to the Maintenance Rule Program scope, and initiated plans to incorporate the affected]]
: [[SSC]] [[into the Maintenance Rule Program scope. The performance deficiency was determined to be more-than-minor because it was associated with the Barrier Integrity cornerstone attribute of]]
SSC performance and affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, a key aspect of the Maintenance Rule is to ensure that maintenance activities are performed in a manner that provide reasonable assurance that SSCs within its scope perform reliably and are capable of providing their intended
Maintenance Rule function(s). In the case of the
: [[SFP]] [[temperature instruments, the licensee was not performing preventive maintenance to ensure that degradation, such as instrument drift, did not adversely affect their ability to detect and alarm]]
: [[EOP]] [[entry conditions such that mitigating actions could be implemented to preserve secondary containment. The finding screened as of very-low safety significance (Green) because it only represented a degradation of the radiological barrier function provided for the control room. Specifically, the finding did not cause]]
: [[SFP]] [[temperature to exceed the maximum analyzed limit, a detectible release of radionuclides, water inventory to decrease below the analyzed limit, or an adverse effect to the]]
: [[SFP]] [[neutron absorber or fuel loading pattern. The team determined that the finding had a cross-cutting aspect in the area of human performance because the licensee did not use a systematic process for evaluating and implementing changes when updating the affected]]
: [[EOP]] [[in 2015.  (Section 1R21.3.b(2)) [H.3] Severity Level]]
: [[IV.]] [[The team identified a Severity Level-IV]]
: [[NCV]] [[of 10]]
: [[CFR]] [[50.68, for the licensee failure to amend the Updated Final Safety Analysis Report (UFSAR) to indicate they chose to comply with]]
: [[10 CFR]] [[50.68(b). Specifically, in 2005, the licensee chose to comply with 10]]
: [[CFR]] [[50.68(b) but did not amend the]]
: [[UFSAR]] [[following the issuance of the associated license amendment. The licensee captured this issue in their]]
: [[CAP]] [[as]]
: [[AR]] [[02741851, reasonably confirmed compliance with 10]]
: [[CFR]] [[50.68(b) requirements (1) through (7) was maintained, and initiated plans to update the]]
: [[UFSAR]] [[to specifically indicate that Clinton Power Station chose to comply with 10]]
: [[CFR]] [[50.68(b). The Significance Determination Process does not specifically consider the impact to the regulatory process in its assessment of licensee performance. Therefore, it was necessary to address this violation, which potentially impacts regulate, using traditional enforcement to adequately deter non-compliance. Specifically, failure to update the]]
: [[UFSAR]] [[challenges the regulatory process because it serves as a reference document used, in part, for recurring safety analyses, evaluating License Amendment Request, and in preparation for and conduct of inspection activities. The team determined the traditional enforcement violation was a Severity Level-]]
: [[IV]] [[violation in accordance with Section 6.1.d.3 of the Enforcement Policy because the un-updated]]
: [[UFSAR]] [[had not been used to evaluate a facility or procedure change that resulted in a condition evaluated as having low-to-moderate or greater safety significance by the Significance Determination Process. However, it had a material impact on safety or licensed activities. Specifically, the un-updated]]
: [[UFSAR]] [[could be used to perform evaluations of facility or procedure changes, which would have the potential to result in unacceptable conditions and/or regulatory decisions. Traditional enforcement violations are not assessed for cross-cutting aspects.  (Section 1R21.3.b(3)) Green. The team identified a finding of very-low safety significance (Green) and an associated]]
: [[NCV]] [[of 10]]
: [[CFR]] [[Part licensee failure to verify the adequacy of design assumptions related to time critical operator actions made in calculations associated with the control room]]
: [[HVAC]] [[and]]
: [[RHR]] [[emergency]]
: [[SFP]] [[cooling functions. Subsequently, it was determined that operators did not fully understand the control room]]
: [[HVAC]] [[system operational demands and that the operational assumptions of the]]
: [[RHR]] [[emergency]]
: [[SFP]] [[cooling design were unrealistic. The licensee captured these issues into the]]
: [[CAP]] [[as]]
: [[AR]] [[02739012,]]
: [[AR]] [[03943566, and]]
AR 02741909; reasonably demonstrated that SFP makeup sources would be available
to cope with a prolonged loss of
: [[SFP]] [[cooling; conducted operator training; and provided refined procedural guidance to ensure the control room]]
: [[HVAC]] [[system would be operated consistent with the design assumptions. The performance deficiency was determined to be more-than-minor because it was associated with the Barrier Integrity cornerstone attribute of human performance and adversely affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the pilot validations of the control room]]
: [[HVAC]] [[system operational assumptions demonstrated a significant reduction in margin due to, in part, a lack of operator understanding of the operational assumptions. Additionally, a preliminary review of procedures associated with]]
: [[SFP]] [[cooling and]]
: [[RHR]] [[determined the operational assumptions of the calculation related to]]
: [[RHR]] [[emergency]]
: [[SFP]] [[cooling were not bounding. The team determined that this finding was of very low safety significance (Green). Specifically, the control room]]
: [[HVAC]] [[system finding example only represented a degradation of the radiological barrier function provided for the control room in that it did not affect the control room barrier function against smoke or a toxic atmosphere. In addition, the finding example related to emergency]]
: [[SFP]] [[cooling did not cause]]
: [[SFP]] [[temperature to exceed the maximum analyzed limit, a detectible release of radionuclides, water inventory to decrease below the analyzed limit, or an adverse effect to the]]
: [[SFP]] [[neutron absorber or fuel loading pattern. The team determined that the finding had a cross-cutting aspect in the area of Human Performance because the operation and engineering organizations did not effectively communicate and coordinate their respective roles in developing the control room]]
: [[HVAC]] [[system validation in a manner that supported nuclear safety.  (Section 1R21.6.b(1)) [H.4] Green:  The team identified a finding of very-low safety significance (Green), and an associated]]
: [[NCV]] [[of 10]]
: [[CFR]] [[Part 50, Appthe identification of a significant design error associated with the control room]]
: [[HVAC]] [[system. Specifically, the licensee did not identify the affected safety function, and promptly restore or confirm system operability. The licensee captured these issues into the]]
: [[CAP]] [[as]]
: [[AR]] [[03948266 and performed a preliminary engineering evaluation using another alternative analytical methodology that reasonably determined the control room]]
: [[HVAC]] [[system remained operable. The performance deficiency was determined to be more-than-minor because it was associated with the Barrier Integrity cornerstone attribute of human performance and adversely affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the performance deficiency resulted in a condition where reasonable doubt on the operability of the control room]]
: [[HVAC]] [[system remained following the identification of a significant design error. The finding screened as of very-low safety significance (Green) because it only represented a degradation of the radiological barrier function provided for the control room. Specifically, the finding did not affect the control room barrier function against smoke or a toxic atmosphere. The team identified that the finding had a cross-cutting aspect in the area of Human Performance because the licensee did not provide training to maintain a knowledgeable workforce that would facilitate an adequate implementation of the operability evaluation process following the identification of a non-conforming design-related issue.  (Section 4]]
: [[OA]] [[2.b(2)) [H.9]]]
: [[REPORT]] [[]]
: [[DETAIL]] [[S 1. REACTOR]]
: [[SAFETY]] [[Cornerstones:  Initiating Events, Mitigating Systems, and Barrier Integrity 1R21 Component Design Bases Inspection (71111.21) .1 Introduction  The objective of the Component Design Bases Inspection is to verify that design bases have been correctly implemented for the selected risk-significant components and that operating procedures and operator actions are consistent with the design and licensing bases. As plants age, their design bases may be difficult to determine and an important design feature may be altered or disabled during a modification. The Probabilistic Risk Assessment (]]
: [[PRA]] [[) model assumes the capability of safety systems and components to perform their intended safety function successfully. This inspectable area verifies aspects of the Initiating Events, Mitigating Systems, and Barrier Integrity cornerstones for which there are no indicators to measure performance. Exelon Generation Company, LLC, certified to the]]
: [[U.S.]] [[Nuclear Regulatory Commission (]]
NRC) that it decided to permanently cease power operations at Clinton Power Station by June 1, 2017, on letter titled, dated June 20, 2016. As a result, this Component Design Bases Inspection was adjusted to perform a more detailed assessment of performance in areas potentially impacted by the planned shutdown, as allowed by Inspection Manual Chapter (IMC) -Water Reactor Inspection Program-Operations  Specific documents reviewed during the inspection are listed in the Attachment to this report. .2 Inspection Sample Selection Process The team used the guidance contained in IMC 2515 to adjust the inspection sample selection process and criteria described in Inspection Procedure 71111.21 to perform a more detailed assessment of performance in areas potentially impacted by the planned shutdown. Based on this approach, a number of samples were selected for the inspection. The team also considered equipment reliability issues in the selection of samples for detailed review. These included items such as performance test results, significant corrective actions, repeated maintenance activities, Maintenance Rule (a)(1) status, system health reports, and resident inspector input of problem areas/equipment. Consideration was also given to the uniqueness and complexity of the design, operating experience, and the available defense in depth margins. A summary of the reviews performed and the specific inspection findings identified are included in the following sections of the report. The team also identified procedures and modifications for review that were associated with the selected components. In addition, the team selected operating experience issues associated with the selected component samples.
This inspection constituted 20 samples (i.e., 11 components and 9 operating experience) as defined in Inspection Procedure 71111.21-05 and as adjusted using the guidance contained in
: [[IMC]] [[2515 for power reactors preparing for transition to decommissioning phase. .3 Component Design a. Inspection Scope The team reviewed the Updated Final Safety Analysis Report (]]
UFSAR), Technical Specifications (TS), design basis documents, drawings, calculations, and other available design basis information, to determine the performance requirements of the selected components. The team used applicable industry standards, such as the American Society of Mechanical Engineers Code and Institute of Electrical and Electronics Engineers The team reviewed the selected components design to assess their capability to perform their required functions and support proper operation of the associated systems. Examples of attributes that were needed for a component to perform its required function included process medium, energy sources, control systems, operator actions, and heat removal. The attributes that verified component condition and tested component capability were appropriate and consistent with the design bases may have included installed configuration, system operation, detailed design, system testing, equipment and environmental qualification, equipment protection, component inputs and outputs, operating experience, and component degradation. For each of the components selected, the team reviewed the maintenance history, preventive maintenance (PM) activities, system health reports, operating experience-related information, vendor manuals, electrical and mechanical drawings, and licensee corrective action documents. Field walkdowns were conducted for all accessible components selected to assess material condition, including age-related degradation, configuration, potential vulnerability to hazards, and consistency between the as-built condition and the design. In addition, the team interviewed licensee personnel from multiple disciplines such as operations, engineering, and maintenance. Other attributes reviewed are included as part of the scope for each individual component. The following 11 components (samples) were reviewed:  Control Room Ventilation Non-Modulating Dampers (0VC01YA/B, 0VC02YA/B, 0VC03YA/B, 0VC04YA/B, 0VC06YA/B, 0VC08YA/B, 0VC09YA/B, 0VC10YA/B, 0VC11YA/B, 0VC48YA/B, 0VC49YA/B, 0VC69Y, 0VC70Y, 0VC81YA/B, 0VC114YA/B, and 0VC115YA/B):  The team reviewed calculations related to control room pressurization, leakage, and radiological habitability to assess the dampers capability to perform their function. In addition, the team reviewed the scope of the control room envelope to assess the design capability to permit access and occupancy of the control room under accident conditions without exceeding the applicable radiological limits. The team also reviewed test procedures and completed tests to assess the associated methodology, acceptance criteria, and test results. In addition, the team reviewed calculations for degraded voltage, control logic, and cable ampacity.
Control Room Chiller (0VC13CA/B):  The team reviewed the control room chiller thermal analysis and control room heat up calculations to assess its capability to maintain temperature within design limits. In addition, the team reviewed the implementation of the Generic Letter (GL) 89-13 Program and its commitments associated with this heat exchanger. Specifically, the team reviewed inspect-and-clean and eddy current test procedures and completed surveillances to assess the associated methodologies, acceptance criteria, and test results. Control Room Makeup and Supply Air Filter Units (0VC09SA/B and
: [[0VC]] [[07SA/B):  The team reviewed calculations related to control room pressurization and radiological habitability to assess the consistency of applicable assumptions with the filter design parameters. In addition, the team reviewed test procedures and completed tests to assess the associated methodology, acceptance criteria, and test results. These tests included those related to the charcoal absorber penetration and system bypass, and methyl iodide penetration. In addition, it included those tests related to the high efficiency particulate air filter penetration and system bypass, and pressure drop. Control Room Chilled Water Pump (OVC08A/B):  The team reviewed hydraulic calculations related to net positive suction head and pump minimum required flow to assess the pump capability to perform its required function. In addition, the team reviewed calculations related to pump motor power requirements. Spent Fuel Pool:  The team reviewed calculations and control measures associated with spent fuel pool (SFP) inventory, including siphoning prevention design features, control of temporary hoses, and minimum required water level. In addition, the team reviewed calculations and control measures, including monitoring activities, associated with]]
: [[SFP]] [[temperature and water chemistry controls. The team also reviewed maintenance activities intended to manage the health of the]]
: [[SFP]] [[, including the liner. Lastly, the team reviewed load drop analyses and the]]
: [[SFP]] [[heavy load operational restrictions. Spent Fuel Pool Cooling Pump (1]]
: [[FC]] [[02PA/B):  The team reviewed the following hydraulic calculations to assess the pump capability to respond to design basis events: pump minimum required flow, minimum required net positive suction head, vortexing, and pump motor cooling minimum required flow. In addition, the team reviewed analyses associated with gas intrusion, such as makeup tank minimum water level setpoint and instrument design configuration. Test procedures and completed surveillances were also reviewed, including quarterly and comprehensive in-service testing, to assess the associated acceptance criteria and test results. The team also reviewed analyses associated with internal flooding due to postulated pipe failures to assess challenges to the pump, motor, or power required for proper operation. In addition, the team reviewed calculations for degraded voltage, control logic, and cable ampacity. Spent Fuel Pool Cooling Surge Tank (1FC01TA/B):  The team reviewed]]
: [[SFP]] [[surge tank inventory and temperature calculations to assess the tank capability to supply the]]
SFP cooling pumps with an adequate water supply. In addition, the team reviewed inventory control design and operational features, including normal and emergency makeup capabilities, and tank level instrument setpoints.
Spent Fuel Pool Cooling Heat Exchanger (1FC01AA/B):  The team reviewed heat transfer calculations and analyses to assess the heat exchanger capability to respond to design basis events. The review included an assessment of cooling water and
: [[SFP]] [[water flow rates and temperatures, tube plugging limits, and heat transfer capacity. In addition, the team reviewed chemistry controls in place for the heat exchangers. Test procedures, completed thermal performance tests, and water chemistry reports were reviewed to assess the associated acceptance criteria, methodology, and test results. Spent Fuel Pool Racks (1F16E002):  The team reviewed criticality analyses and control measures associated with the]]
SFP racks, including the associated safety analyses, rack geometric arrangement, fuel placement controls, and neutron absorbing materials. In addition, the team reviewed seismic analyses to assess the rack design capability to prevent an adverse geometric reconfiguration during an earthquake. Shutdown Service Water Pump (1SX01PA/B):  The team reviewed service water (SX) pump calculations and analysis, including minimum required system flow, pump cooling, and flood protection. In addition, the team reviewed test procedures and completed tests, including quarterly and comprehensive inservice testing, to assess the associated methodology, acceptance criteria, and test results. Additionally, the team reviewed information related to underground cable monitoring and testing associated with this component. Lastly, the team reviewed pump motor calculations and analysis associated with voltage drop, degraded voltage, minimum required voltage, cable ampacity, and protective devices. 125 Volts Direct Current Batteries (1DC01/2E):  The team reviewed calculations and analyses related to battery loads, division separation, battery sizing and capacity, and electrical isolation between class 1E and non-1E. This review was performed to assess the battery capability to support the design basis required voltage requirements of the 125 Volts Direct Current safety-related loads under both normal and design basis accident conditions. In addition, the team reviewed discharge calculations to assess the minimum predicted voltage capability to support inverter operation. The team also reviewed a sample of completed surveillance tests, service duty discharge tests, and age management activities to assess the associated acceptance criteria, methodology, and test results. b. Findings (1) Non-Conservative Control Room Radiological Habitability Assessment Introduction:  The team identified a finding of very-low safety significance (Green) and an associated Non-Cited Violation (NCV) of Title 10 of the Code of Federal Regulations (CFR), Part 50, Appendix B, Criterion III, use a technically appropriate analytical methodology in the control room radiological habitability calculation. Specifically, the licensee used a methodology that inappropriately characterized the control room heating, ventilation, and air-conditioning (HVAC) system outside air intake design resulting in a calculated control room dose following a loss of coolant accident (LOCA) that exceeded the applicable limit.
Description:  stated, under accident conditions and whole body doses are less than those set by Criterion 19 of
: [[10 CFR]] [[Part , provided to permit occupancy and access to the control room without receiving more than  Calculation C-Accident]]
: [[UR]] [[evision 6, evaluated, in part, the radiological consequence of a]]
: [[LOCA]] [[to the control room habitability. On November 10, 2016, the team noted that C-020 used the methodology described in Regulatory Guide 1.194, Section 3.3.2, which stated, room] ventilation system configurations that have two outside air intakes, each of which meets applicable design criteria of an engineered safeguards feature (]]
: [[ESF]] [[), including single-  This methodology credited the ability to select the intake exposed to the lowest dose allowing for certain calculated dose concentrations to be reduced by a factor of 4. However, the control room]]
: [[HVAC]] [[system dual outside air intakes did not meet the single-failure criterion. Elimination of the reduction factor resulted in a higher calculated control room dose following a]]
: [[LOCA]] [[which exceeded the 5 rem limit. The licensee captured the team concerns in their Corrective Action Program (CAP) as Action Request (AR) 02742442. As an immediate corrective action, the licensee completed an operability evaluation that determined the control room remained operable by, in part, crediting actual as found values associated with key controlling parameters instead of licensing basis limits. In addition, the licensee issued]]
: [[NRC]] [[Event Notification 52377, pursuant to 10 Notification Requirements for Operating Nuclear Power Reactors, because the incorrect method used for the control room habitability calculation resulted in an unanalyzed condition. The proposed corrective action to restore compliance at the time of this inspection included revising the affected calculation and performing an apparent cause evaluation. Analysis:  The team determined that the failure to use an analytical methodology that was technically appropriate to characterize the control room]]
: [[HVAC]] [[system outside air intake design in the control room radiological habitability assessments was contrary to]]
: [[10 CFR]] [[Part 50, Appendix B, Criterion]]
: [[IIID]] [[esign Controldeficiency. The performance deficiency was determined to be more-than-minor because it was associated with the Barrier Integrity cornerstone attribute of design control and affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the performance deficiency resulted in the control room expected dose following a]]
: [[LOCA]] [[to exceed the applicable limits prompting an operability evaluation. The team determined the finding could be evaluated using the Significance Determination Process (]]
SDP) Attachment 060Because the finding impacted the Barrier Integrity cornerstone, the team screened the Findings At-Pow-low safety significance (Green) because it
only represented a degradation of the radiological barrier function provided for the control room. Specifically, the finding did not affect the control room barrier function against smoke or a toxic atmosphere. The team did not identify a cross-cutting aspect associated with this finding because it was not confirmed to reflect current performance due to the age of the performance deficiency. Specifically, the affected calculations were performed more than 3 years ago. Enforcement:  in part, that the licensee provide for verifying or checking the adequacy of design, such as by the performance of design reviews, by the use of alternate or simplified calculational methods, or by the performance of a suitable testing program. Contrary to the above, as of December 1, 2016, the licensee failed to verify the adequacy of the control room
: [[HVAC]] [[system design, a safety-related system, structure, and component (]]
: [[SSC]] [[). Specifically, the licensee used a methodology that was not applicable to the control room]]
: [[HVAC]] [[system dual outside air intake design in the control room radiological habitability calculation resulting in a calculated dose that exceeded the applicable limit. The licensee was still evaluating its planned corrective actions. However, the team determined that the continued non-compliance does not present an immediate safety concern because the licensee reasonably determined that the control room envelope remained operable as discussed in Section 4]]
: [[OA]] [[2.1.b(2) of this report. Because this violation was of very-low safety significance and was entered into the  02742442, this violation is being treated as an NCV, consistent with Section 2.3.2 of the]]
: [[NRC]] [[Enforcement Policy.  (]]
: [[NCV]] [[05000461/2016009-01; Non-Conservative Control Room Radiological Habitability Assessment) (2) Failure to Scope Spent Fuel Pool Temperature and Level Instruments into the Maintenance Rule Program Introduction:  The team identified a finding of very-low safety significance (Green) and an associated]]
: [[NCV]] [[of 10]]
: [[CFR]] [[50.65 for Monitoring the Effectiveness of e licensee failure to scope nonsafety-related mitigating]]
: [[SSC]] [[used within an emergency operating procedure (]]
: [[EOP]] [[) into the Maintenance Rule Program. Specifically, an]]
: [[EOP]] [[used]]
: [[SFP]] [[low-level and high-temperature parameters as distinct entry criteria but the associated components were not included in the scope of the Maintenance Rule Program. Description:  In May of 2015, the licensee added]]
: [[SFP]] [[temperature above 150 degrees Fahrenheit and]]
: [[SFP]] [[level below elevation 753 feet and 4 inches as two new distinct entry criteria in Revision 30 of EOP-8, Secondary Containment Control.-8 was revised to include additional actions if]]
: [[SFP]] [[level could not be maintained above elevation 753 feet and 4 inches. The new]]
: [[EOP]] [[entry criteria and decision making steps were added based on the Boiling Water Reactor Owner Group guidance documents. These documents were revised based upon Institute of Nuclear Power Operations Industry Event Report 11-and Makeup,an associated elevated importance of maintaining adequate decay heat removal to the]]
: [[SFP.]] [[Additionally, the]]
UFSAR described that the 150 degrees Fahrenheit temperature limit was set to assure that the auxiliary building environment does not exceed equipment environmental limits.
On around November 2, 2016, the team noted the
: [[SSC]] [[relied upon to detect when these new]]
: [[EOP]] [[entry criteria were met were not included in the scope of the Maintenance Rule Program. Specifically, the licensee relied on control room annunciator alarm 5040-pent Fuel Pool Storage, to detect when the]]
: [[SFP]] [[temperature]]
: [[EOP]] [[entry criterion was met. This alarm was set at 150 degrees Fahrenheit and used]]
: [[SFP]] [[temperature recorder 1]]
: [[TR]] [[-FC079 to display the output temperature sensed by]]
: [[SFP]] [[thermocouple temperature instrument 1]]
: [[TEFC]] [[079. During this inspection, the licensee could not locate any documented]]
: [[PM]] [[or testing performed on the]]
: [[SFP]] [[temperature instrument and recorder other than a]]
: [[PM]] [[to perform replacement of the paperless recorder disk (Ref:]]
: [[AR]] [[02741764). The licensee informed the team that the instrument was calibrated upon installation in January 1999 and the]]
: [[PM]] [[was retired on March 16, 2001, based, in part, upon the components not being within the scope of the Maintenance Rule Program, drift not reducing system reliability, and being coded in Procedure]]
: [[ER]] [[-AA-210, ,non-critical run-to-maintenance component (Ref:]]
: [[PMER]] [[# 01-00793). Similarly, the licensee relied on control room annunciator alarm 5040-pent Fuel Pool,to detect when the]]
: [[SFP]] [[level]]
: [[EOP]] [[entry criterion was met. This alarm used]]
: [[SFP]] [[level instruments 1LS-FC078 and associated level switches. The]]
: [[SFP]] [[level switches had a 2-year instrument check]]
: [[PM]] [[activity. The team consulted with the Office of Nuclear Reactor Regulations (NRR), and determined these instruments met the Maintenance Rule Program scoping criteria and were not required as a result of]]
: [[NRC]] [[Order]]
: [[EA]] [[-12- The licensee captured the team concerns in their]]
: [[CAP]] [[as A R02736193. As an immediate corrective action, the licensee performed an extent of condition to identify any other]]
: [[SSC]] [[addition to the]]
: [[EOP]] [[requiring them to be added to the Maintenance Rule Program scope. The proposed corrective action to restore compliance at the time of this inspection was to incorporate the]]
: [[SFP]] [[temperature and level instruments into the Maintenance Rule Program scope. Analysis:  The team determined that the lice-related]]
: [[SFP]] [[temperature and level components into the Maintenance Rule Program was contrary to 10]]
: [[CFR]] [[50.65(b)(2)(i) and was a performance deficiency. The performance deficiency was determined to be more than minor because it was associated with the Barrier Integrity cornerstone attribute of]]
: [[SSC]] [[performance and affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, a key aspect of the Maintenance Rule is to ensure that maintenance activities are performed in a manner that provide reasonable assurance that]]
: [[SSC]] [[within its scope perform reliably and are capable of providing their intended Maintenance Rule function(s). In the case of the]]
: [[SFP]] [[temperature instruments, the licensee was not performing]]
: [[PM]] [[to ensure that degradation, such as instrument drift, did not adversely affect their ability to detect and alarm EOP-8 entry conditions such that mitigating actions could be implemented to preserve secondary containment. The team determined the finding could be evaluated using the]]
: [[SDP]] [[in accordance with the Barrier Integrity cornerstone, the team screened the finding through]]
IMC 0609, -
screened as of very-low safety significance (Green) because it did not cause
: [[SFP]] [[temperature to exceed the maximum analyzed limit, a detectible release of radionuclides, water inventory to decrease below the analyzed limit, or an adverse effect to the]]
: [[SFP]] [[neutron absorber or fuel loading pattern. The team determined that the finding had a cross-cutting aspect in the area of Human Performance because the licensee did not use a systematic process for evaluating and implementing changes when updating EOP-8 in 2015.  [H.3] Enforcement:  Title]]
: [[10 CFR]] [[50.65Requirements for Monitoring the Effectiveness of requires, in part, that the scope of the Monitoring Program specified in 10]]
: [[CFR]] [[50.65(a)(1) includes nonsafety-related]]
: [[SSC]] [[that are relied upon to mitigate accidents or transients or are used in plant]]
: [[EOP]] [[. Contrary to the above, since May 15, 2015, the licensee failed to scope nonsafety-related]]
: [[SSC]] [[that are relied upon to mitigate accidents or transients or are used in plant]]
: [[EOP]] [[into the Monitoring Program. Specifically,]]
: [[SFP]] [[temperature and level instrumentation that are used in]]
: [[EOP]] [[-8 were not included in the scope of the monitoring program established by the licensee pursuant]]
: [[10 CFR]] [[50.65(a)(1). The licensee was still evaluating its planned corrective actions. However, the team determined that the continued non-compliance does not present an immediate safety concern because the finding did not involve an actual degraded condition. Because this violation was of very-low safety significance and was entered into the]]
: [[AR]] [[02736193, this violation is being treated as an NCV, consistent with Section 2.3.2 of the]]
: [[NRC]] [[Enforcement Policy.  (]]
: [[NCV]] [[05000461/2016009-02; Failure to Scope Spent Fuel Pool Temperature and Level Instruments into the Maintenance Rule Program) (3) Failure to Amend the Updated Final Safety Analysis Report Indicating Choice to Comply with]]
: [[10 CFR]] [[50.68(b) Introduction:  The team identified a Severity Level (]]
: [[SL]] [[)-IV]]
: [[NCV]] [[of 10]]
: [[CFR]] [[50.68, Paragraph (b)(8), for the licensee failure to amend the]]
: [[UFSAR]] [[to indicate they had chosen to comply with 10]]
: [[CFR]] [[50.68(b). Specifically, in 2005, the licensee chose to comply with]]
: [[10 CFR]] [[50.68(b) but did not amend the]]
: [[UFSAR]] [[following the issuance of the associated license amendment. Description:  In 1998, the]]
: [[NRC]] [[amended its regulations to give power reactor licensees the option of either meeting the criticality accident requirements of 10]]
: [[CFR]] [[70.24, ity or]]
: [[10 CFR]] [[Part 50.68, which shares the same title. On August 18, 2004, the licensee submitted a License Amendment Request (]]
: [[LAR]] [[) RS-04-Fuel Storage Expansion to the]]
: [[NRC]] [[to revise]]
: [[TS]] [[. This]]
: [[LAR]] [[was associated with the replacement of some of the]]
: [[SFP]] [[storage racks with a new design. Attachment 5, Section 2.3.g, of the]]
: [[LAR]] [[indicated that the new racks were intended to meet, in part, 10]]
: [[CFR]] [[50.68 requirements. On October 31, 2005, the]]
: [[NRC]] [[approved the]]
: [[LAR]] [[in License Amendment No. 170. Section associated Safety Evaluation Report indicated the]]
: [[NRC]] [[reviewed the]]
LAR against 10 CFR 50.68(b) requirements. However, during this inspection period, the team
noted the
: [[UFSAR]] [[choice to comply with 10]]
: [[CFR]] [[50.68(b) as required by]]
: [[10 CFR]] [[50.68(b)(8). In addition, the]]
: [[UFSAR]] [[did not include implicit or explicit descriptions associated with compliance to]]
: [[10 CFR]] [[50.68(b)(1), 10]]
: [[CFR]] [[50.68(b)(5), and]]
: [[10 CFR]] [[50.68(b)(7). The licensee captured this issue in their]]
: [[CAP]] [[as]]
: [[AR]] [[02741851. As an immediate corrective action, the licensee reasonably confirmed compliance to 10]]
: [[CFR]] [[50.68(b) requirements (1) through (7) was maintained. The proposed corrective action to achieve compliance with]]
: [[10 CFR]] [[50.68(b)(8) at the time of this inspection was to update the]]
: [[UFSAR]] [[to specifically indicate that Clinton Power Station chose to comply with]]
: [[10 CFR]] [[50.68(b). Analysis:  The team determined this violation was associated with a minor performance deficiency. Specifically, the licensee failure to amend the]]
: [[UFSAR]] [[to indicate they chose to comply with]]
: [[10 CFR]] [[50.68(b) was contrary to 10]]
: [[CFR]] [[50.68(b)(8) and was a performance deficiency. The performance deficiency was of minor safety significance because the team  the minor screening questions in block 3 of]]
: [[IMC]] [[0612, Appendix B. The]]
: [[SDP]] [[does not specifically consider the impact to the regulatory process in its assessment of licensee performance. Therefore, it was necessary to address this violation which potentially impacts enforcement to adequately deter non-compliance. Specifically, a failure to update the]]
: [[UFSAR]] [[challenges the regulatory process because it serves as a reference document used, in part, for recurring safety analyses, evaluating]]
: [[LAR]] [[, and in preparation for and conduct of inspection activities. As a result, the violation was evaluated using]]
: [[NRC]] [[Enforcement Policy, dated November 1, 2016. The team determined the violation was a]]
: [[SL]] [[-IV violation in accordance with Section 6.1.d.3 of the Enforcement Policy because the un-updated]]
: [[UFSAR]] [[had not been used to evaluate a facility or procedure change that resulted in a condition evaluated as having low-to-moderate or greater safety significance by the]]
: [[SDP.]] [[However, it had a material impact on safety or licensed activities. Specifically, the un-updated]]
: [[UFSAR]] [[could be used to perform evaluations of facility or procedure changes, which would have the potential to result in unacceptable conditions and/or regulatory decisions. Traditional enforcement violations are not assessed for cross-cutting aspects. Enforcement:  Title 10]]
: [[CFR]] [[50.68(b)(8), requires licensees to amend the]]
: [[UFSAR]] [[no later than the next update required by 10]]
: [[CF]] [[indicating that the licensee had chosen to comply with]]
: [[10 CFR]] [[50.68(b). Title 10]]
: [[CFR]] [[50.71(e) requires, in part, that licensees periodically update the]]
: [[UFSAR]] [[as provided in Paragraph (e)(4). Title 10]]
: [[CFR]] [[50.71(e)(4) requires, in part, that]]
: [[UFSAR]] [[revisions be filed annually or 6 months after each refueling outage provided the interval between successive updates does not exceed 24 months. Contrary to the above, since 2007, the licensee failed to amend the]]
: [[UFSAR]] [[no later than the next update required by]]
: [[10 CFR]] [[50.71(e) to indicate they had chosen to comply with 10]]
: [[CFR]] [[50.68(b). Specifically, the licensee did not amend their]]
: [[UFSAR]] [[within the timeframe required by 10]]
CFR 50.71(e) following the issuance of License Amendment No. 170 on October 31, 2005, which approved a change where the licensee chose to comply with 10 CFR 50.68(b).
The licensee was still evaluating its planned corrective actions. However, the team determined that the continued non-compliance does not present an immediate safety concern because the licensee reasonably confirmed compliance with
: [[10 CFR]] [[50.68(b) requirements (1) through (7) was maintained. Because this was a]]
: [[SL]] [[-IV violation, and was entered into the licAR 02741851, this violation is being treated as a NCV, consistent with Section 2.3.2 of the]]
: [[NRC]] [[Enforcement Policy.  (]]
: [[NCV]] [[05000461/2016009-03; Failure to Amend the Updated Final Safety Analysis Report Indicating Choice to Comply with]]
: [[10 CFR]] [[50.68(b)) .4 Operating Experience a. Inspection Scope The team reviewed  nine operating experience issues (samples) to assess the licensee evaluation and resolution of]]
: [[NRC]] [[generic concerns. The operating experience issues listed below were reviewed in depth as part of this inspection:]]
: [[GL]] [[2007-]]
: [[GL]] [[95--Related Power-O  Information Notice (IN) 09--Absorbing Materials in]]
: [[IN]] [[11--Conservative Criticality Safety Analyses for Fuel Storage;  Bulletin 94-adequate]]
: [[GL]] [[85--  Bulletin 96-Reactor Core, or Over Safety-Re]]
: [[IN]] [[88-]]
: [[IN]] [[14- The team also assessed the accuracy of the licensee response to]]
: [[GL]] [[16-01, Monitoring of Neutron-AbsorbinBecause this response was still under the review of]]
NRR, the team did not assess the associated evaluation and resolution of this generic issue. Thus, this review did not constitute an inspection sample. b. Findings No findings were identified. 
.5 Modifications a. Inspection Scope The team reviewed three permanent plant modifications related to selected risk-significant components to verify that the design bases, licensing bases, and performance capability of the components had not been degraded through modifications. The modifications listed below were reviewed as part of this inspection effort:  Engineering Change (EC) Documentation, Electrical Drawings, and Seismic QualificatiRevision 1;  - Revision 1; and  - b. Findings No findings were identified. .6 Operating Procedure Accident Scenarios a. Inspection Scope The team performed a detailed reviewed of selected procedures associated with the inspection samples. For these procedures, in plant action were walked down with a licensed operator and any interfaces with other departments were evaluated. The procedures were compared to UFSAR, design assumptions, and training materials to assess their consistency. The following operating procedures were reviewed in detail:  e-Fire  Revision 4a;  3208 Revision 15c;  Revision 31e;  Revision 10e;  ,  Reactor Vessel Pool/Spent Fuel Pool, Revision 0d;  Revision 2e;  Revision 8e;  Revision 26e;  Revision 30; and  EOP- Revision 30.
For the procedures listed, time critical operator actions were reviewed for reasonableness. This review included observation of licensed operator crews actions during the performance of a failure of control room
: [[HVAC]] [[system to shift to the high radiation mode scenario on the station simulator to assess operator knowledge level, procedure quality, availability of special equipment where required, and capability to perform time critical operator actions within the required time. In addition, the team evaluated operations interfaces with other departments. The following operator actions were reviewed:  Two]]
: [[SFP]] [[pump and two]]
: [[SFP]] [[heat exchanger operation;  Makeup to the]]
: [[SFP]] [[surge tank with cycled condensate water;  Makeup to the]]
: [[SFP]] [[with fire protection water;  Makeup to the]]
: [[SFP]] [[with SX;  Upper containment overflow makeup to the SFP;  Component cooling water]]
: [[SFP]] [[heat exchanger outlet throttling operation;  Swapping]]
: [[SFP]] [[heat exchanger cooling from component cooling water to SX;  Aligning Residual Heat Removal]]
: [[RHR]] [[for emergency]]
: [[SFP]] [[cooling mode; and  Control room]]
: [[HVAC]] [[system failure to shift to the high radiation mode during a]]
: [[LOCA.]] [[b. Findings (1) Failure to Verify the Adequacy of Design Assumptions Related to Time Critical Operator Actions Introduction:  The team identified a finding of very-low safety significance (Green) and an r the licensee failure to verify the adequacy of design assumptions related to time critical operator actions. Specifically, the licensee failed to verify operator action time response assumptions made in calculations to support control room]]
: [[HVAC]] [[and]]
: [[RHR]] [[capability to cool the]]
: [[SFP.]] [[Description:  The licensee established Revision 3 of Procedure]]
OP-AA-102-106, (TCAs) that were assumed to be performed within a specified time frame in design calculations and associated licensing basis. The program established the standards under which the various TCAs are scoped into the program, controlled, validated, and documented. The validation process was developed by both the operations and engineering departments, and was structured in a manner to ensure that credited operator actions can be performed under the limiting conditions existing within the current licensing basis (e.g., bounding event, limiting single failure, minimum staffing, equipment accessibility).
On December 10, 2015, the licensee performed a validation (i.e.,
: [[TCA]] [[-7) intended to confirm operators would identify the failure of the control room]]
: [[HVAC]] [[train in service to realign from the normal to the high radiation operating mode, and take manual action to start and align the standby control room]]
: [[HVAC]] [[train in high radiation mode within 20 minutes consistent with the assumptions of Calculation C-CoRevision 6. This calculation evaluated, in part, the radiological consequences of a]]
: [[LOCA]] [[to the control room operators. However, during this inspection, the team noted the validation was not performed with corequired a minimum of 2 senior reactor operators and 2 reactor operators. However,]]
: [[TCA]] [[-7 was performed with the normal control room staff compliment of 3 senior reactor operators and 3 reactor operators. The team also noted the validation was not performed with a]]
: [[LOCA.]] [[In addition, the licensee did not address these differences as required by Procedure]]
: [[OP]] [[-AA-102-106, Steps 4.1.7, 4.3.4, 4.3.10, and 4.3.13, and Attachment 1. The team was concerned because minimum staffing during a]]
: [[LOCA]] [[would add significant complexities and competing operational priorities which would likely increase the as-found response time of 12 minutes. The licensee captured the]]
: [[TCA]] [[-7 issues in the]]
: [[CAP]] [[as]]
: [[AR]] [[02739012 and performed a pilot TCA-7 validation during this inspection with minimum staffing and a LOCA, which the inspection team observed. The licensee determined the]]
: [[TCA]] [[was completed in 14 minutes. However, the team noted the licensee declared the]]
: [[TCA]] [[as completed prematurely. Specifically, the licensee determined the]]
: [[TCA]] [[was completed when the standby control room]]
: [[HVAC]] [[train was started in high radiation mode. However, the Calculation C-20 assumptions included proper alignment of the intake damper, which occurred at approximately 19 minutes. The licensee captured this issue in the]]
: [[CAP]] [[as]]
: [[AR]] [[03943566. Corrective actions included creating an assignment to develop a new TCA-7 validation scenario, developing a new smart card procedure defining the necessary steps credited in the TCA, and holding crew tail gate discussions to raise awareness to ensure that these actions would be performed within the assumed time for an actual event. In addition, the licensee identified similar issues associated with other]]
: [[TCA]] [[and captured them in the]]
: [[CAP]] [[as]]
: [[AR]] [[02740663,]]
: [[AR]] [[02741339,]]
: [[AR]] [[02740908, and]]
: [[AR]] [[02740900. The team also noted a separate issue involving the licensee failure to recognize a]]
: [[TCA]] [[related to cooling the]]
: [[SFP]] [[using]]
: [[RHR.]] [[Specifically, Revision 12 of]]
: [[UFSAR]] [[9.1.3, pent Fuel Pool , in part, that a train of]]
: [[RHR]] [[can be used for emergency]]
: [[SFP]] [[cooling but prohibits usage when the]]
: [[RHR]] [[system is required for decay heat removal. Calculation 1]]
: [[FC]] [[32Confirm that the Volume of Water in the Spent Fuel Pool Is Such that there Is Enough Heat Absorption Capability to Allow Sufficient Time for Switching Over to the Residual Heat Removal System for Emergency CoolingRevision 0A, evaluated the]]
: [[SFP]] [[heat up during a loss of the normal]]
: [[SFP]] [[cooling system at power and assumed that it would take approximately 3.5 hours to shutdown and cooldown the plant to permit]]
: [[RHR]] [[usage for]]
: [[SFP]] [[cooling and an additional 1.0 hour to align the]]
: [[RHR]] [[train for this purpose. However, the team noted that a]]
TCA had not been identified, developed, and validated as required by OP-AA-102-106, Steps 2.5, 3.1.2, and 3.2.1. Additionally, based upon a review of the associated procedures and plant walkdowns, the team questioned if the assumed response times were realistic. The licensee reviewed the associated procedures and estimated that the
actions could be completed in 6.5 to 9.5 hours. The licensee entered these issues into the
: [[CAP]] [[as]]
: [[AR]] [[02741909. The proposed corrective action to achieve compliance at the time of this inspection was to develop an associated]]
: [[TCA]] [[and revise Calculation 1]]
: [[FC]] [[32. Analysis:  The team determined that the failure to verify the adequacy of design assumptions related to time critical operator actions was contrary to Procedure OP-AA-102-106 and was a performance deficiency. The performance deficiency was determined to be more-than-minor because it was associated with the Barrier Integrity cornerstone attribute of human performance and adversely affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the pilot validations of the control room]]
: [[HVAC]] [[system operational assumptions demonstrated a significant reduction in margin due to, in part, a lack of operator understanding of the operational assumptions. Additionally, a preliminary review of procedures associated with]]
: [[SFP]] [[cooling and]]
: [[RHR]] [[determined the operational assumptions of the calculation related to]]
: [[RHR]] [[emergency]]
: [[SFP]] [[cooling were not bounding. The team determined the finding could be evaluated using the]]
: [[SDP]] [[in accordance with  Because the finding impacted the Barrier Integrity cornerstone, the team screened the finding through]]
: [[IMC]] [[0609, r Findings At-on The team determined that this finding was of very low safety significance (Green). Specifically, the]]
: [[TCA]] [[-7 finding example only represented a degradation of the radiological barrier function provided for the control room in that it did not affect the control room barrier function against smoke or a toxic atmosphere. In addition, the finding example related to emergency]]
: [[SFP]] [[cooling did not cause]]
: [[SFP]] [[temperature to exceed the maximum analyzed limit, a detectible release of radionuclides, water inventory to decrease below the analyzed limit, or an adverse effect to the]]
: [[SFP]] [[neutron absorber or fuel loading pattern. The team determined that the finding had a cross-cutting aspect in the area of Human Performance because the operation and engineering organizations did not effectively communicate and coordinate their respective roles in developing]]
: [[TCA]] [[-7 in a manner that supported nuclear safety.  [H.4] Enforcement:  Title]]
: [[10 CFR]] [[Part in part, that the licensee provide for verifying or checking the adequacy of design, such as by the performance of design reviews, by the use of alternate or simplified calculational methods, or by the performance of a suitable testing program. Contrary to the above, as of December 1, 2016, the licensee failed to verify the adequacy of the design as evidenced by the following examples:  The licensee did not verity the control room]]
HVAC design, a safety-related SSC, would be capable of maintaining adequate control room habitability following a LOCA and a single failure. Specifically, the design relied on time critical operator action assumptions that were not verified to be correct using the most limiting conditions.
The licensee did not verify the
: [[RHR]] [[design, a safety-related]]
: [[SSC]] [[, would be capable of providing emergency]]
: [[SFP]] [[cooling. Specifically, the design relied on time critical operator action assumptions that were not verified to be correct and were subsequently determined to be unrealistic during this inspection. The licensee was still evaluating its planned corrective actions. However, the team determined that the continued non-compliance does not present an immediate safety concern because the finding did not result in actual]]
: [[SFP]] [[degraded conditions and the licensee reasonably demonstrated that]]
: [[SFP]] [[makeup sources would be available to cope with a prolonged loss of]]
: [[SFP]] [[cooling. In addition, the licensee immediately conducted operator training and provided control room operators refined procedural guidance to ensure the control room]]
: [[HVAC]] [[system would be operated consistent with the design assumptions. Because this violation was of very-low safety significance and was entered into the]]
: [[CAP]] [[as]]
: [[AR]] [[02739012,]]
: [[AR]] [[02741909, and]]
: [[AR]] [[03943566, this violation is being treated as an]]
: [[NCV]] [[, consistent with Section 2.3.2 of the]]
: [[NRC]] [[Enforcement Policy.  (]]
: [[NCV]] [[05000461/2016009-04, Failure to Verify the Adequacy of Design Assumptions Related to Time Critical Operator Actions) 4.]]
: [[OTHER]] [[]]
: [[ACTIVI]] [[TIES 4OA2 Identification and Resolution of Problems .1 Review of Items Entered Into the Corrective Action Program a. Inspection Scope The team reviewed a sample of problems identified by the licensee associated with the selected inspection samples and that were entered into the]]
: [[CAP.]] [[The team reviewed  engineering issues and the effectiveness of corrective actions related to engineering issues. In addition, corrective action documents written on issues identified during the inspection were reviewed to assess the incorporation of the problem into the]]
: [[CAP.]] [[The specific corrective action documents reviewed by the team are listed in the attachment to this report. b. Findings (1) Failure to Promptly Identify that the Incapability of the Residual Heat Removal Design to Support Technical Specifications Operability Requirements Was a Condition Adverse to Quality Introduction:  The team identified a finding of very-low safety significance (Green) and an associated]]
: [[NCV]] [[of 10]]
: [[CFR]] [[Part 50, Appendix B, Crithe licensee failure to promptly identify that the incapability of the]]
: [[RHR]] [[design to support]]
: [[TS]] [[operability requirements was a condition adverse to quality (CAQ). Specifically, when reactor water temperature was greater than 150 degree Fahrenheit,]]
: [[RHR]] [[could not be realigned from shutdown cooling (]]
SDC) mode of operations to provide the TS required functions of the emergency core cooling system, suppression pool cooling, containment spray, and feedwater leakage control system.
Description:  During this inspection, the team noted the licensee was aware that the design of the
: [[RHR]] [[system could not support certain]]
: [[TS]] [[limiting condition of operations (LCOs) while it was aligned for]]
: [[SDC]] [[but failed to recognize this as a]]
: [[CAQ.]] [[This was esidual Heat Removal Shutdown Cooling Operations & Fuel Pool Cooling and AssistRevision 10e, directed operators to declare multiple]]
: [[TS]] [[]]
: [[LCO]] [[inoperable when operating]]
: [[RHR]] [[in]]
: [[SDC]] [[and with reactor water temperature greater than 150 degrees Fahrenheit. These]]
: [[LCO]] [[Emergency Core Cooling Systems-Operating;Emergency Core Cooling Systems-Shutdown;]]
: [[LCO]] [[3.6.1.9,  Residual Heat Removal documented this practice. The team was concerned because the licensee failed to recognize that the inability to maintain the operability of multiple SSCs, as required by TS, was a]]
: [[CAQ]] [[that needed to be corrected. The licensee confirmed the]]
: [[LCO]] [[were declared inoperable under the described conditions because valves 1E12-F004A(B) would need to open to realign]]
: [[RHR]] [[from]]
: [[SDC]] [[mode to support the other]]
: [[LCO]] [[functions. However, these valves would not be capable of opening at water temperatures greater than 150 degrees Fahrenheit due to pressure locking/thermal binding concerns. Specifically, the licensee documented the -105681, dated November 29, 1995, as part of their efforts to respond to]]
: [[NRC]] [[]]
: [[GL]] [[95-Locking and Thermal Binding of Safety-Related Power-However, binding as a]]
: [[CAQ]] [[because they incorrectly concluded:  (1) the valves did not have a safety function to open when in]]
: [[SDC]] [[operations; and (2) they could revise their]]
: [[SDC]] [[procedure to voluntarily declare the valve associated with the]]
: [[SDC]] [[train in operation as inoperable under conditions where the valve would be susceptible to pressure locking/thermal binding. In addition, the licensee confirmed that opening the valves at water temperatures greater than 212 degrees Fahrenheit would have the potential to lead to steam void formation. For example, the water inside of the pipe would flash to steam following]]
: [[RHR]] [[realignment from]]
: [[SDC]] [[to Emergency Core Cooling Systems mode of operation at water temperatures higher than saturation conditions to respond to a shutdown]]
: [[LOCA]] [[, leading to, in part, water hammer concerns. The team discussed this issue with]]
: [[NRR]] [[and reviewed historical licensing basis documents. The team also reviewed the]]
: [[TS]] [[Basis and noted the basis for]]
: [[LCO]] [[3.0.2 stated: The Completion Times of the Required Actions are also applicable when a system or component is removed from service intentionally. The reasons for intentionally relying on the]]
ACTIONS include, but are not limited to, performance of Surveillances, preventive maintenance, corrective maintenance, or investigation of operational problems. Entering ACTIONS for these reasons must be done in a manner that does not compromise safety. Intentional entry into ACTIONS should not be made for operational convenience. Alternatives that would not result in redundant equipment being inoperable should be used instead.
As a result, it was determined the intent of the affected
: [[TS]] [[]]
: [[LCO]] [[was, in part, to ensure the specified]]
: [[SSC]] [[were operable and that it was not acceptable to rely on]]
: [[TS]] [[required actions and associated completion times as corrective actions for]]
: [[CAQ]] [[that are known and expected. The licensee captured this issue in their]]
: [[CAP]] [[as]]
: [[AR]] [[2742439 and]]
: [[AR]] [[3948042. The proposed corrective action to achieve compliance at the time of this inspection was to request a]]
: [[LAR]] [[to align]]
: [[TS]] [[with the plant design capabilities. Analysis:  The team determined the failure to promptly identify that the incapability of the]]
: [[RHR]] [[design to support]]
: [[TS]] [[operability requirements was a]]
: [[CAQ]] [[was contrary to 10]]
: [[CFR]] [[deficiency. The performance deficiency was determined to be more-than-minor because it was associated with the Mitigating Systems cornerstone attribute of design control and adversely affected the associated cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the performance deficiency resulted in voluntarily declaring]]
: [[TS]] [[functions inoperable when performing]]
: [[SDC]] [[operations, which did not ensure the associated mitigating systems availability or capability to respond to an initiating event. The team determined the finding could be evaluated using the]]
: [[SDP]] [[in accordance with]]
: [[IMC]] [[Because the finding impacted the Mitigating Systems cornerstone during shutdown conditions the team Significance Determination Process Phase 1 Initial Screening and Characterization of The team determined that this finding was of very-low safety significance (Green). Specifically, there were no known instances where the finding:  (1) represented a loss of system safety function; (2) represented an actual loss of safety function of at least a single train or two separate safety systems out-of-service for greater than its]]
: [[TS]] [[allowed outage time; (3) involved non-]]
: [[TS]] [[trains of equipment; (4) involved a degradation of a functional]]
: [[RHR]] [[auto-isolation on low reactor vessel level; (5) impacted external event protection; or (6) involved fire brigade issues. The team did not identify a cross-cutting aspect associated with this finding because it did not reflect current licensee performance since the performance deficiency occurred more than 3 years ago. Enforcementrequires, in part, that]]
: [[CAQ]] [[, such as deficiencies and non-conformances are promptly identified and corrected. Contrary to the above, since November 29, 1995, the licensee failed to promptly identify and correct a]]
: [[CAQ.]] [[Specifically, the licensee recognized that valves 1E12-F004A(B) could not be opened to support]]
: [[TS]] [[]]
: [[LCO]] [[3.5.1, 3.5.2, 3.6.1.7, 3.6.1.9,  and 3.6.2.3 under certain conditions while]]
: [[RHR]] [[was aligned for]]
: [[SDC.]] [[However, the licensee did not identify this condition as a]]
CAQ and, as a result, did not capture it in the CAP and correct it.
At the time of this inspection, the licensee was still evaluating its planned corrective actions. However, the team determined that the continued non-compliance did not present an immediate safety concern because the associated
: [[TS]] [[allowable outage times provided reasonable safety assurance if the proposed]]
: [[LAR]] [[is not submitted and approved by the next plant shutdown. Because this violation was of very-low safety significance, and was entered into the  2742439 and]]
: [[AR]] [[3948042, this violation is being treated as a]]
: [[NCV]] [[, consistent with Section 2.3.2 of the]]
: [[NRC]] [[Enforcement Policy.  (]]
: [[NCV]] [[05000461/2016009-05; Failure to Promptly Identify that the Incapability of the Residual Heat Removal Design to Support Technical Specification Operability Requirements Was a Condition Adverse to Quality) (2) Failure to Follow the Operability Determination Process Following the Identification of a Control Room Heating, Ventilation and Air-Conditioning System Design Issue Introduction:  The team identified a finding of very-low safety significance (Green), and evaluation procedure after the identification of a significant design error associated with the control room]]
: [[HVAC]] [[system. Specifically, the licensee did not identify the affected safety function, and promptly restore or confirm system operability. Description:  During this inspection, the licensee created]]
: [[AR]] [[02742442 to capture the team concerns documented in Section 1R21.3.b(1) of this report associated with a control room]]
: [[HVAC]] [[system design error. As a result, the licensee performed an operability evaluation that concluded the control room]]
: [[HVAC]] [[system remained operable. However, the team noted the operability evaluation had the following significant weaknesses that reasonably challenged its conclusion:  The licensee failed to recognize that operability is impacted whenever a]]
: [[TS]] [[]]
: [[SSC]] [[cannot perform its specified function. Specifically, the evaluation recognized that the issue represented a non-conforming design issue that could affect the system capability to maintain the maximum control room occupancy dose within the required limit during a]]
: [[LOCA.]] [[However, it determined that the issue did not impact operability because the issue did not result in physical equipment degradation. The licensee credited compensatory actions that did not restore operability. Specifically, operations created a standing order for operators to don respirators during a]]
LOCA. The entry conditions of the first two revisions of the standing order were only applicable when the control room envelope boundary was inoperable. However, the design error did not affect control room envelope boundary. Thus, the standing order would not be executed during the conditions applicable for the non-conforming condition. In addition, the team noted that the use of respirators could not restore or establish the control room HVAC system operability because they do not support the system capability to perform its specified safety function. The licensee credited an alternative analytical methodology for operability that was not technically appropriate. Specifically, the operability evaluation was based, in part, on an alternative analysis that determined the system would
perform its function by assuming that containment spray, which was not previously credited, would be operated in a specific manner. However, the licensee did not create an instruction (e.g., standing order) to ensure operators would operate containment spray in a manner that was consistent with the analysis assumptions. In addition, the licensee did not evaluate the effects on containment of operating containment spray in the assumed manner. Specifically, the team was concerned about the possibility of overcooling containment resulting in unanalyzed negative pressure conditions. The licensee captured this issue in their
: [[CAP]] [[as]]
: [[AR]] [[03948266. Corrective actions included performing a preliminary engineering evaluation using another alternative analytical methodology that reasonably determined the control room]]
: [[HVAC]] [[system remained operable. At the time of this inspection, the licensee was still processing the results of this preliminary evaluation through their formal operability determination process. Analysis:  The team determined that the failure to follow the operability evaluation procedure after the identification of a significant design error associated with the control room]]
: [[HVAC]] [[system was contrary to]]
: [[10 CFR]] [[Part 50, Appendix B, Criterion V, a performance deficiency. The performance deficiency was determined to be more-than-minor because it was associated with the Barrier Integrity cornerstone attribute of human performance and adversely affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the performance deficiency resulted in a condition where reasonable doubt on the operability of the control room]]
: [[HVAC]] [[system remained following the identification of a significant design error. The team determined the finding could be evaluated using the]]
: [[SDP]] [[in accordance with]]
: [[IMC]] [[06the Barrier Integrity cornerstone, the team screened the finding through]]
: [[IMC]] [[0609 e Significance Determination Process for Findings At-on screened as of very-low safety significance (Green) because it only represented a degradation of the radiological barrier function provided for the control room. Specifically, the finding did not affect the control room barrier function against smoke or a toxic atmosphere. The team identified that the finding had a cross-cutting aspect in the area of Human Performance because the licensee did not provide training to maintain a knowledgeable workforce that would facilitate an adequate implementation of the operability evaluation process following the identification of a non-conforming design-related issue.  [H.9]  Enforcement:  Title ions, Procedures, and Drawings,requires, in part, that activities affecting quality be prescribed by documented procedures of a type appropriate to the circumstances and be accomplished in accordance with these procedures. The licensee established Procedure]]
: [[OP]] [[-AA-108-115, Revision 17, as the implementing procedure for determinations of operability of safety-related]]
: [[SSC]] [[included in]]
TSs, an activity affecting quality.
Contrary to the above, from November 18 to December 1, 2016, the licensee failed to follow procedure
: [[OP]] [[-AA-108-115 as evidenced by the following examples:  Step measures may be perable]]
: [[SSC]] [[to an operable but degraded or non-Step action, either physical or administrative, that is taken to maintain or enhance an operable but degraded or nonconforming]]
: [[SSC]] [[to ensure its specified safety function  Contrary to these steps, the licensee used a compensatory action that did not ensure the specified safety function of the nonconforming]]
: [[SSC]] [[could be performed. Specifically, operations created a standing order for operators to don respirators that would not be executed during the conditions applicable for the non-conforming condition. In addition, the use of respirators could not restore or establish the control room]]
: [[HVAC]] [[system operability because they do not support the system capability to perform its specified safety function. Step ty from a detailed examination Step component, or device shall be operable or have operability when its capable of considered operable, an]]
: [[SSC]] [[must be capable of performing the safety functions specified by its design, within the required range of design physical conditions,  Contrary to these steps, the licensee did not determine operability of the control room]]
HVAC when there was reasonable doubt it would be capable of performing its safety function following the discovery of a nonconforming design. Specifically, the licensee assumed that operability could only be challenged by physical equipment degradation. technically appropriate to characterize the SSC involved, the nature of the degraded or nonconforming condition, and specific facility  Contrary to this step, the licensee used an analytical method that was not technically appropriate to characterize the specific facility design. Specifically, the analytical method assumed that containment spray would be operated in a specific manner that was not assured by procedures and the licensee did not create a temporary instruction to ensure operators would operate containment spray in a manner that was consistent with the analysis assumptions. In addition, the assumed containment spray operation had the potential to overcool containment and this effect was not evaluated. At the time of this inspection, the licensee was still evaluating its planned corrective actions. However, the team determined that the continued non-compliance did not present an immediate safety concern because the licensee reasonably concluded the system was operable based upon a preliminary engineering analysis.
Because this violation was of very-low safety significance, and was entered into the
: [[AR]] [[02742442, this violation is being treated as a]]
: [[NCV]] [[, consistent with Section 2.3.2 of the]]
: [[NRC]] [[Enforcement Policy.  (]]
: [[NCV]] [[05000461/2016009-06; Failure to Follow the Operability Determination Process Following the Identification of a Control Room Heating, Ventilation and Air-Conditioning System Design Issue) 4OA6  Management Meeting .1 Exit Meeting Summary On December 1, 2016, the team presented the inspection results to Mr.]]
: [[B.]] [[Kapellas, and other members of the licensee staff. The licensee acknowledged the issues presented. The team confirmed that several documents reviewed were considered proprietary and were handled in accordance with the]]
NRC policy related to proprietary information. ATTACHMENT:  SUPPLEMENTAL INFORMATION
: [[SUPPLE]] [[MENTAL INFORMATION]]
: [[KEY]] [[]]
: [[POINTS]] [[]]
: [[OF]] [[]]
: [[CONTAC]] [[T Licensee]]
: [[B.]] [[Kapellas, Plant Manager S. Gackstetter, Site Engineering Director C. Dunn, Operations Director D. Shelton, Regulatory Assurance Manager M. Heger, Senior Manager Design Engineering D. Avery, Regulatory Assurance Representative U.S. Nuclear Regulatory Commission M. Jeffers, Chief, Engineering Branch 2 N. Féliz Adorno, Senior Reactor Inspector]]
: [[LIST]] [[]]
: [[OF]] [[]]
: [[ITEMS]] [[OPENED, CLOSED,]]
: [[AND]] [[]]
: [[DISCUS]] [[SED Opened and Closed 05000461/2016009-01]]
: [[NCV]] [[Non Conservative Control Room Radiological Habitability Assessment (Section 1R21.3.b(1)) 05000461/2016009-02]]
: [[NCV]] [[Failure to Scope]]
: [[SFP]] [[Temperature and Level Instruments into the Maintenance Rule Program (Section 1R21.3.b(2)) 05000461/2016009-03]]
: [[NCV]] [[Failure to Amend the]]
: [[UFSAR]] [[Indicating Choice to Comply with 10]]
: [[CFR]] [[50.68(b)  (Section 1R21.3.b(3)) 05000461/2016009-04]]
: [[NCV]] [[Failure to Verify the Adequacy of Design Assumptions Related to Time Critical Operator Actions (Section 1R21.6.b(1)) 05000461/2016009-05]]
: [[NCV]] [[Failure to Promptly Identify that the Incapability of the]]
: [[RHR]] [[Design to Support]]
: [[TS]] [[Operability Requirements Was a]]
: [[CAQ]] [[(Section 4]]
: [[OA]] [[2.b(1)) 05000461/2016009-06]]
: [[NCV]] [[Failure to Follow the Operability Determination Process Following the Identification of a Control Room]]
: [[HVAC]] [[System Design Issue (Section 4OA2.b(2))]]
: [[LIST]] [[]]
: [[OF]] [[DOCUMENTS REVIEWED The following is a list of documents reviewed during the inspection. Inclusion on this list does not imply that the]]
: [[NRC]] [[inspectors reviewed the documents in their entirety, but rather, that selected sections of portions of the documents were evaluated as part of the overall inspection effort. Inclusion of a document on this list does not imply]]
: [[NRC]] [[acceptance of the document or any part of it, unless this is stated in the body of the inspection report. CALCULATIONS Number Description or Title Date or Revision VC-91 Minimum Available and Maximum Required]]
: [[NPSH]] [[for the]]
: [[VC]] [[System Chilled Water Pumps 0 VC-82 Control Room Pressurization with Two Open Floor Drains 0 VC-68 Min./Max. Flow & Density Determination for]]
: [[VC]] [[0]]
: [[VC]] [[-40 Allowable Tolerances for]]
: [[VC]] [[Airflows 0 C-020 Reanalysis of Loss of Coolant Accident Using Alternate Source Terms 6 C-022 Site Boundary and Control Room Dose Following a]]
: [[FHA]] [[in Containment Using]]
: [[AST]] [[0]]
: [[VC]] [[-86 Evaluation of Control Room Chillers for]]
: [[SX]] [[Acceptance Criteria 1]]
: [[EQ]] [[-CL006 Environmental Qualification of The Okonite Company Low Voltage Power and Control Cables 04/01/03 EQ-CL007 Environmental Qualification of Okonite]]
: [[5KV]] [[Power Cables Splice Tapes and Tapes Configurations 24]]
: [[EC]] [[]]
: [[399907 SX]] [[Cable Submergence Evaluation 0]]
: [[EC]] [[392511 Fuel Pool Cooling Pump Trip Reliability  Low Suction Pressure]]
: [[1 EC]] [[370372 Review and Approve Spare]]
: [[SX]] [[Motor Documentation,  Electrical Drawings, and Seismic Qualification]]
: [[1 EC]] [[401731 Division 1 Battery Inter-Rack Jumper Support 0 19-D-28 Review of Division 1]]
: [[DC]] [[System Review 1A 15 19-AK-13]]
: [[CPS]] [[Load Control Calculation 3 19-]]
: [[AK]] [[-13 LOCA09-1E fed by RAT, 138kV and 345kV System at 105%, 1E buses at 4300V via SVC, Fault Currents for Bus 1AP07E and 1AP09E 3 19-AN-08 4160V]]
: [[ESF]] [[Switchgear Buses 1A1 and 1B1 Motor Relay Settings 4 01]]
: [[FC]] [[43 Thermal-Hydraulic Analysis, Holtec Calc HI-2033116 2A 01FC07 Allowable Value for]]
: [[FC]] [[Pump Trip 3A 01]]
: [[FC]] [[25]]
: [[FC]] [[]]
: [[HX]] [[Performance 5 01FC42 Criticality Safety Analysis for Clinton, HI-2033135 2 01SX52 Process Type Heat Exchanger Acceptance 0 IP-M-0563 Determination of Allowable Leak Rates and Loss of]]
: [[UHS]] [[Volume from Shutdown Service Water (]]
: [[SX]] [[) Boundary Valves 2B HI-2033124 Spent Fuel Storage Expansion at Clinton Power Station 2 IP-F-0096 Nuclear Evaluation of High Density Spent Fuel Storage Racks at Low Temperature 0 EC390921 Fuel Pool Cooling and Cleanup]]
: [[HX]] [[Test Data and Performance Evaluation-1]]
: [[FC]] [[01AA 0 IP-M-0197 Evaluation of Potential Inleakage through Screen House Penetration Seals below Probable Maximum Flood Level 0 IP-M-0825]]
: [[FC]] [[Surge Tank Vortex Evaluation 0]]
EC 265230 Re-Rack Project Safe Load Path Deviation 1
: [[CORREC]] [[TIVE]]
: [[ACTION]] [[]]
: [[DOCUME]] [[NTS (GENERATED]]
: [[DUE]] [[]]
: [[TO]] [[]]
: [[THE]] [[]]
: [[INSPEC]] [[TION) Number Description or Title Date AR02735956]]
: [[CDBI]] [[Calculation]]
: [[VC]] [[-91 Has an Incorrect Term 11/02/16 AR02740900 CDBI:]]
: [[ADMIN]] [[Gaps Identified in Review of]]
: [[OP]] [[-CL-102-106-1001 11/14/16 AR02740908 CDBI: OP-AA-102-106 Attachment 1 Info Needs Updated 11/14/16 AR02741020]]
: [[CDBI]] [[2016:]]
: [[NRC]] [[Observation for]]
: [[ATI]] [[889611-01 11/14/16]]
: [[AR]] [[02741299 CDBI:]]
: [[NRC]] [[Observation of]]
: [[CPS]] [[3317.01 and EOP-8 Directions 11/15/16 AR02741339]]
: [[CDBI]] [[11/15/16]]
: [[AR]] [[02741520]]
: [[CDBI]] [[]]
: [[FC]] [[Surge Tank Vortexing Observation 11/15/16 AR02741555]]
: [[2016 CDBI]] [[]]
: [[FC]] [[Surge Tank Levels M05S are Misleading 11/15/16 AR02735934 CDBI:  Bolting on New Div]]
: [[2 SX]] [[Pump Requires Coating 11/02/16]]
: [[AR]] [[02735962]]
: [[CDBI]] [[]]
: [[ID]] [[: Door 1DR1-16 Will Not Self-Close and Latch 11/02/16 AR02736120]]
: [[CDBI]] [[Unapproved Document Used as Reference in Calc]]
: [[VC]] [[-19 11/02/16 AR02736193 2016 CDBI: Spent Fuel Pool Instruments Not in the]]
: [[MT]] [[Rule 11/02/16]]
: [[AR]] [[02736537]]
: [[CDBI]] [[]]
: [[ID]] [[: Housekeeping Issues in Fuel Building 11/03/16 AR02736756 CDBI:]]
: [[VC]] [[Locker Room Exhaust Dampers Incorrectly Classified 11/03/16]]
: [[AR]] [[02737143]]
: [[CDBI]] [[M05-1102 Sht1 Needs Updated Failure Mechanism of Damper 11/04/16]]
: [[AR]] [[02737150]]
: [[CDBI]] [[]]
: [[DC]] [[-ME-06-CP Has Incorrect Addition Control Room Leakage 11/04/16 AR02737155]]
: [[CDBI]] [[Drawing Rev Has a Note Removed that Is Still Referenced 11/04/16]]
: [[AR]] [[02737285 2016 CDBI: Housekeeping Equip Stored on Elec Junction Box 11/04/16 AR02739012]]
: [[CDBI]] [[]]
: [[ID]] [[: Validation Errors of Operator Response Time Action 11/09/16 AR02739642]]
: [[CDBI]] [[2016-]]
: [[WO]] [[Task Not Cancel Requested When Work Was Completed 11/10/16 AR02740040 CDBI: Statement in]]
: [[USAR]] [[Not Updated for]]
: [[SFP]] [[Re-Rerack 11/11/16 AR02741764]]
: [[CDBI]] [[2016: Spent Fuel Pool Instrumentation Being Accurate 11/16/16]]
: [[AR]] [[02741851]]
: [[NRC]] [[]]
: [[CDBI]] [[Identified]]
: [[USAR]] [[Update Needed 11/16/16]]
: [[AR]] [[02741862]]
: [[CDBI]] [[]]
: [[NRC]] [[Questions Staffing on]]
: [[TCA]] [[Validations 11/16/16]]
: [[AR]] [[02741909]]
: [[CDBI]] [[Potential New]]
: [[TCA]] [[for]]
: [[FC]] [[Cooling from]]
: [[RHR]] [[A System 11/16/16 AR02741937]]
: [[NRC]] [[]]
: [[CDBI]] [[Generic Letter Response GL2016-01 Needs Update 11/16/16 AR02742297]]
: [[CDBI]] [[Calculation C-022 Needs to be Updated with Current Data 11/15/16]]
: [[AR]] [[02742333]]
: [[CDBI]] [[2016:]]
: [[EC]] [[]]
: [[399907 EC]] [[Eval for Submerged Cables 11/17/16]]
: [[AR]] [[02742439]]
: [[CDBI]] [[Need Action to Track Resolution of]]
: [[CDBI]] [[Question 11/17/16 AR02742442 CDBI:  Inappropriate Calculation Method for]]
: [[CR]] [[Habitability 11/17/16]]
: [[AR]] [[02742446 CDBI: Coding of Corrective Actions Not]]
: [[IAW]] [[]]
: [[PI]] [[-AA-125 11/17/16 AR02742785]]
: [[CDBI]] [[Revise 01]]
: [[FC]] [[25 to Include Analysis with Tube Plugging 11/18/16 AR03943566]]
: [[CDBI]] [[Observations of Piloted]]
: [[TCA]] [[-7 Performance 11/21/16 AR03943594]]
: [[CDBI]] [[]]
: [[DWG]] [[V54417-600 SH2 of 7 Contains Typo]]
: [[CCP]] [[11/21/16]]
: [[AR]] [[03943717 CDBI: Timeliness of]]
: [[NRC]] [[Report 11/21/16]]
: [[AR]] [[03944282]]
: [[CDBI]] [[Fire Protection System Lesson Plan Not in]]
: [[EDMS]] [[11/22/16 AR03946256]]
: [[CDBI]] [[5040.02 Window 2D and 2E Need Refinement 11/28/16]]
: [[AR]] [[03946310 CDBI: Revise]]
: [[CPS]] [[2700.12/13 for Valve Throttling Positioning 11/28/16]]
: [[AR]] [[03946387]]
: [[CDBI]] [[]]
: [[VC]] [[Tracer Gas Test Trending 11/28/16 AR03946468]]
: [[CDBI]] [[Past]]
: [[IR]] [[Returned to Ops for OP/Functionality Review 11/28/16 AR03946485]]
: [[CDBI]] [[Incorrect]]
: [[SFP]] [[Elevation in]]
: [[TS]] [[4.3.2 11/28/16]]
: [[AR]] [[03946490]]
: [[CDBI]] [[Bundle and Core Design Verif Guide Missing Info 11/28/16]]
: [[AR]] [[03946974]]
: [[CDBI]] [[]]
: [[FC]] [[Check Valve]]
: [[IST]] [[Flow Rate Issue 11/28/16]]
: [[AR]] [[03946979]]
: [[CDBI]] [[No Spent Fuel Pool Siphon Breaker Sizing Calculation 11/29/16]]
: [[AR]] [[03948042]]
: [[CDBI]] [[]]
: [[RHR]] [[]]
: [[TS]] [[3.5.1 Mode 3 Applicability 12/01/16]]
: [[CORREC]] [[TIVE]]
: [[ACTION]] [[]]
: [[DOCUME]] [[NTS (REVIEWED]]
: [[DURING]] [[]]
: [[THE]] [[INSPECTION) Number Description or Title Date AR02484166]]
: [[VC]] [[Damper 49]]
: [[YB]] [[Has Leakage Through Blades 04/13/15 AR02484163]]
: [[VC]] [[Damper 48]]
: [[YB]] [[Has Leakage Through Blades 04/13/15 AR02631112 Damper Blade Not Making Contact 02/24/16 AR02631463]]
: [[VC]] [[A Outlet Temp 0]]
: [[TIVC]] [[421]]
: [[HOOS]] [[02/25/16]]
: [[AR]] [[02700958 Unexpected]]
: [[VC]] [[Chiller B Trip 08/04/16]]
: [[AR]] [[01153798]]
: [[VC]] [[Filtered Air In-Leakage Trend Increasing 12/17/10]]
: [[AR]] [[02712999]]
: [[2016 CDBI]] [[]]
: [[FASA]] [[,]]
: [[NRC]] [[]]
: [[IN]] [[2009-26 and IN2011-03 Discrepancies 09/07/16 AR02713000]]
: [[2016 CDBI]] [[]]
: [[FASA]] [[- 1FC01AA:]]
: [[FC]] [[]]
: [[HX]] [[Not in]]
: [[GL]] [[89-13 Program 09/07/16]]
: [[AR]] [[01509794 1CC075B;]]
: [[CCW]] [[Tank Lowers When]]
: [[SX]] [[Aligned to]]
: [[FC]] [[]]
: [[HX]] [[05/02/13 ATI01206369-01]]
: [[NRC]] [[Information Notice 2011-03: Nonconservative Criticality Safety Analyses for Fuel Storage 06/01/11]]
: [[AR]] [[01674529]]
: [[OPEX]] [[]]
: [[NRC]] [[Info Notice 2014-09 - Spent Fuel Misloading Events 06/24/14 AR00181452]]
: [[OPEX]] [[]]
: [[OE]] [[15859 Salem Tritium Leak from Fuel Pool /]]
: [[IN]] [[2004-05 10/17/03]]
: [[ATI]] [[01406618-01]]
: [[NRC]] [[]]
: [[IN]] [[2012- 11/08/12 AR01616833 SFP,]]
: [[IFTS]] [[to Cask Loading Pool Potential for Loss of Water 02/04/14]]
: [[AR]] [[00987616]]
: [[IN]] [[2009-26 Degradation of Neutron Absorbing Materials in]]
: [[SFP]] [[11/02/09 AR01377717 Initial Results from Div.]]
: [[2 SX]] [[Flow Balance 06/14/12]]
: [[AR]] [[01037486]]
: [[2010 CDBI]] [[]]
: [[FASA]] [[Identifies Flow Balance Procedure Weakness 03/02/10 AR01074788]]
: [[CDBI]] [[]]
: [[NRC]] [[Inspector Challenges Information in]]
: [[IR]] [[1037486 05/28/10]]
: [[AR]] [[2596101 1FC004A  Contingency]]
: [[WO]] [[needed for Valve Repair 12/04/15]]
: [[AR]] [[2453592 1FC02PB Pump Outboard Bearing Oil Leak 02/11/15 AR1294780 3317.01]]
: [[2 FC]] [[Pump and]]
: [[HX]] [[Operation  11/27/11 AR2740900 CDBI:  Admin Gaps Identified in Review OP-CL-102-1001 11/14/16 AR2741862 CDBI:  Limiting Staff for]]
: [[TCA]] [[11/16/16]]
: [[AR]] [[1558342]]
: [[FP]] [[]]
: [[FASA]] [[:  OP-CL-102-106-1001 Does Not Contain All]]
: [[TCA]] [[09/13/13]]
: [[AR]] [[2722442 Inappropriate Calculation Method for]]
: [[CR]] [[Habitability 11/17/16]]
: [[AR]] [[2741299]]
: [[CDBI]] [[]]
: [[CPS]] [[3317.01 and EOP-8 Conflicting Information 11/15/16 AR2742446]]
: [[CDBI]] [[Coding of Corrective Actions Not]]
: [[IAW]] [[PI-AA-125 11/17/16 AR2741339]]
: [[CDBI]] [[11/15/16]]
: [[AR]] [[1674529]]
: [[OPEX]] [[]]
: [[NRC]] [[]]
: [[IN]] [[2014-09  Spent Fuel Misloading Events 06/24/14]]
: [[AR]] [[2414160]]
: [[OPEX]] [[Eval for]]
: [[NRC]] [[]]
: [[IN]] [[2014-14  Potential Safety Enhancement 11/19/14]]
: [[AR]] [[2725126 Perform]]
: [[EOC]] [[Review for New]]
: [[MR]] [[Function VF-01 10/06/16 AR2740663 Validation Errors of Operator Response Time Actions 11/13/16 AR3948266]]
: [[CDBI]] [[Operability Evaluation for]]
: [[MRC]] [[Habitability Observation 12/01/16 AR2739012]]
: [[CDBI]] [[]]
: [[ID]] [[: Validation Errors of Operator Response Time Action 11/09/16 AR2741339]]
: [[CDBI]] [[11/15/16]]
: [[AR]] [[2740908]]
: [[CDBI]] [[]]
: [[OP]] [[-AA-102-106 Attachment 1 Info Needs Updated 11/14/16 AR3947667]]
: [[CDBI]] [[Questions]]
: [[IR]] [[2741909 Action Coding of]]
: [[ACIT]] [[11/30/16]]
: [[AR]] [[2741909]]
: [[CDBI]] [[Potential New]]
: [[TCA]] [[for]]
: [[FC]] [[Cooling from]]
: [[RHR]] [[A System 11/16/16 AR2736193]]
: [[CDBI]] [[Spent Fuel Pool Instruments not in]]
: [[MR]] [[Scope 11/02/16 AR3948042]]
: [[CDBI]] [[]]
: [[RHR]] [[]]
: [[TS]] [[3.5.1. Mode 3 Applicability 12/01/16]]
: [[AR]] [[2739012]]
: [[CDBI]] [[]]
: [[ID]] [[:  Validation Errors of Operator Response Time Action  11/09/16 AR3947456]]
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: [[AR]] [[2742439]]
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: [[CDBI]] [[]]
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: [[EOP]] [[-8 Directions 11/15/16 AR2736537]]
: [[CDBI]] [[]]
ID Misc Tools Laying Loose Around Protected Pump 11/02/16
: [[DRAWIN]] [[GS Number Description or Title Revision M05-1102 Sh.]]
: [[1 P&ID]] [[Control Room]]
: [[HVAC]] [[U M05-1102 Sh.]]
: [[2 P&ID]] [[Control Room]]
: [[HVAC]] [[J M05-1102 Sh.]]
: [[3 P&ID]] [[Control Room]]
: [[HVAC]] [[N M05-1102 Sh.]]
: [[4 P&ID]] [[Control Room]]
: [[HVAC]] [[M E02-0VC99 Control Room]]
: [[HVAC]] [[System (]]
: [[VC]] [[) Control Room]]
: [[HVAC]] [[Return Fan A U E02-0]]
: [[VC]] [[99 Control Room]]
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: [[VC]] [[) Radiation Detectors and Isolation Signal Initiation Logic R E02-0VC99 Control Room]]
: [[HVAC]] [[System (]]
: [[VC]] [[)]]
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: [[VC]] [[)]]
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: [[VC]] [[) Control Room]]
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: [[MCC]] [[1C (1]]
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: [[ERAT]] [[]]
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: [[II]] [[Fuel Pool Rack Layout 10 M05-1048 Sht. 9 P&]]
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DOCUMENTS (SCREENINGS/SAFETY EVALUATIONS) Number Description or Title Revision CL-2011-S-027 GNF2 Fuel Transition 1 CL-2006-S-051 Spent Fuel Pool Re-Rack Phase 2 0  MISCELLANEOUS  Number Description or Title Date  Response to Generic Letter 2007-01 12/07/07 Letter Y-105681 Susceptibility Evaluation Criteria Operational Screening 11/29/95
: [[MISCEL]] [[LANEOUS  Number Description or Title Date Letter U-602553 -Thermal Binding of Safety-Related Power-Operated Gate Valves  02/09/96 WCC-EXN-LH1-15-001 Clinton Unit 1 Cycle 17 Bundle Design Reports 09/04/15 CPS-16-026]]
: [[NRC]] [[]]
: [[GL]] [[2016-01, Monitoring of Neutron Absorbing Materials in Spent Fuel Pools, Response Information 10/03/16 Letter Y-209452 Closure Package for]]
: [[IEIN]] [[88- 10/06/88 Letter Y-210816 Closure Package for]]
: [[IEIN]] [[88- 06/16/89 Letter RS-16-207 Response to Generic Letter 2016-01 11/03/16 Letter Y-216546 Closure Package for]]
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: [[NRC]] [[]]
: [[IN]] [[87-13 Review 11/02/87 SE-LOR-501 Operator Response Time Validation Scenario 2 for TCA-7 11/15/15]]
: [[TCA]] [[7 Validation]]
: [[VC]] [[Manual Start Stby Train and Alignment 12/10/15 SE-LOR-501 Operator Response Time Validation Scenario 2 for TCA-7 11/15/15]]
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: [[EIN]] [[1]]
: [[TRFC]] [[079 01/28/99 Data Sht. TE009 Thermo Electric qTE-FC079/80]]
: [[TE]] [[Data Sheet 07/07/78]]
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: [[EC]] [[370372 Review and Approve Spare]]
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: [[OPERAB]] [[ILITY EVALUATIONS  Number Description or Title Date or Revision EC407244 Operability of the Main Control Room Ventilation System 0 EC389732 Evaluate Past Operability/Reportability of Div. 2 Components With as Found Flow Below Design Values 09/07/12 EC407232 Evaluate the Effectiveness of]]
: [[SCBA]] [[in the]]
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: [[HEPA]] [[Filter Test Data Sheet 27 9866.02 VG/VC Charcoal Adsorber Leak Test 33 9866.02F001 Charcoal Adsorber Leak Test Data Sheet 32 9070.02 Control Room]]
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: [[PROCED]] [[URES  Number Description or Title Revision 9070.02D001 Control Room]]
: [[HVAC]] [[High Rad, Initiation Functional Data Sheet 29b 3402.01 Control Room]]
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: [[1 SX]] [[System Flow Balance Verification 9 3703.02 Fuel Handling Platform (F11) Operations 19d 4303.02 Abnormal Lake Level 12c]]
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: [[HVAC]] [[System Division 1 33 OP-CL-102-106-1001 Operator Response Time Program At]]
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VC Hi Rad Initiation 00
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: [[SX]] [[Pump Oper. Test (SX Pump A) 08/10/16 WO1915452 9069.01B20]]
: [[OP]] [[]]
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: [[WORK]] [[]]
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: [[WORK]] [[]]
: [[DOCUME]] [[NTS  Number Description or Title Date WO01702351 Division 2 Battery Service Test 04/07/15 WO01914535 Division 2 Battery Quarterly Surveillance 07/07/16 WO01953137 Division 2 Battery Weekly Pilot Cell Check 09/22/16 WO1471402 Perform]]
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: [[LIST]] [[]]
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: [[USED]] [[]]
: [[ADAMS]] [[Agencywide Document Access Management System]]
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: [[CAQ]] [[Condition Adverse to Quality]]
: [[CFR]] [[Code of Federal Regulations]]
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: [[EOP]] [[Emergency Operating Procedure]]
: [[GL]] [[Generic Letter]]
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: [[NRR]] [[Office of Nuclear Reactor Regulations]]
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: [[UFSAR]] [[Updated Final Safety Analysis Report]]
: [[B.]] [[Hanson -3- Letter to Brian Hanson from Mark Jeffers dated January 12, 2017]]
: [[SUBJEC]] [[T:  CLINTON]]
: [[POWER]] [[]]
: [[STATIO]] [[N -]]
: [[NRC]] [[]]
: [[COMPON]] [[ENT]]
: [[DESIGN]] [[]]
: [[BASES]] [[INSPECTION, INSPECTION REPORT 05000461/2016009]]
}}
}}

Latest revision as of 18:27, 19 December 2019

NRC Component Design Bases Inspection, Inspection Report 05000461/2016009
ML17013A253
Person / Time
Site: Clinton Constellation icon.png
Issue date: 01/12/2017
From: Jeffers M
NRC/RGN-III/DRS/EB2
To: Bryan Hanson
Exelon Generation Co, Exelon Nuclear
References
IR 2016009
Download: ML17013A253 (41)


Text

UNITED STATES ary 12, 2017

SUBJECT:

CLINTON POWER STATION - NRC COMPONENT DESIGN BASES INSPECTION, INSPECTION REPORT 05000461/2016009

Dear Mr. Hanson:

On December 1, 2016, the U.S. Nuclear Regulatory Commission (NRC) completed a Component Design Bases Inspection at your Clinton Power Station. Because Exelon Generation Company, LLC, certified to the NRC that it decided to permanently cease power operations at Clinton Power Station by June 1, 2017, on letter titled Certification of Permanent Cessation of Power Operations, dated June 20, 2016, this Component Design Bases Inspection was adjusted to perform a more detailed assessment of performance in areas potentially impacted by the proposed shutdown. This inspection adjustment was consistent with the guidance contained in Inspection Manual Chapter 2515, Light-Water Reactor Inspection Program-Operations Phase, dated February 1, 2016. The enclosed report documents the results of this inspection, which were discussed on December 1, 2016, with Mr. B. Kapellas, Plant Manager, and other members of your staff.

Based on the results of this inspection, six NRC-identified findings of very-low safety significance were identified. These findings involved violations of NRC requirements. However, because of their very-low safety significance, and because the issues were entered into your Corrective Action Program, the NRC is treating the issues as Non-Cited Violations in accordance with Section 2.3.2 of the NRC Enforcement Policy.

If you contest the subject or severity of these Non-Cited-Violations, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001, with copies to the Regional Administrator, Region III; the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC Resident Inspector at the Clinton Power Station. In addition, if you disagree with the cross-cutting aspect assigned to any finding 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 III, and the NRC Resident Inspector at the Clinton Power Station. In accordance with Title 10 of the Code of Federal Regulations (10 CFR) 2.390, Public Inspections, Exemptions, Requests for Withholding, of the NRC's "Rules of Practice," a copy of this letter, its enclosure, and your response (if any) will be available electronically for public inspection in the NRCs Public Document Room or from the Publicly Available Records (PARS)

component of the NRC's Agencywide Documents Access and Management System (ADAMS).

ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Sincerely,

/RA/

Mark Jeffers, Chief Engineering Branch 2 Division of Reactor Safety Docket No. 50-461 License No. NPF-62

Enclosure:

IR 05000461/2016009

REGION III==

Docket No: 50-461 License No: NPF-62 Report No: 05000461/2016009 Licensee: Exelon Generation Company, LLC Facility: Clinton Power Station Location: Clinton, IL Dates: October 31 - December 1, 2016 Inspectors: N. Féliz Adorno, Senior Reactor Inspector, Lead J. Benjamin, Senior Reactor Inspector, Operations L. Rodriguez, Reactor Inspector, Mechanical J. Robbins, Reactor Inspector, Electrical Approved by: M. Jeffers, Chief Engineering Branch 2 Division of Reactor Safety Enclosure

Table of Contents

SUMMARY

................................................................................................................................ 2

REPORT DETAILS

REACTOR SAFETY

1R21 Component Design Bases Inspection

OTHER ACTIVITIES

4OA2 Identification and Resolution of Problems

4OA6 Management Meeting 26

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Attachment

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Attachment

LIST OF DOCUMENTS REVIEWED

2