IR 05000461/2016009: Difference between revisions

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]]

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

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

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

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]]

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

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

SFP cooling design were unrealistic. The licensee captured these issues into the

SFP cooling; conducted operator training; and provided refined procedural guidance to ensure the control room

AR 02739012,

AR 03943566, and

AR 02741909; reasonably demonstrated that SFP makeup sources would be available

to cope with a prolonged loss of

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

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]]

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

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 ) .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 , 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

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]]

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 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

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

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

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

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-

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

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

TCA and captured them in the

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]]

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 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. The licensee entered these issues into the

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

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

AR 03943566, this violation is being treated as an

AR 02739012,

AR 02741909, and

NCV , consistent with Section 2.3.2 of the

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]]

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

AR 3948042. The proposed corrective action to achieve compliance at the time of this inspection was to request a

AR 2742439 and

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

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

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]]

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

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 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

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

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.]]

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

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 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

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

CDBI - Questions

IR 2736193 Actions Coding of

ACITS 11/30/16

AR 2742439

CDBI Need Action to Track Resolution of

CDBI Question 11/17/16 AR2741299

CDBI [[]]

NRC Observation of

CPS 3317.01 and

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.

HVAC M E02-0VC99 Control Room

HVAC Return Fan A U E02-0

VC ) Control Room

VC 99 Control Room

VC ) Radiation Detectors and Isolation Signal Initiation Logic R E02-0VC99 Control Room

VC System Alarm Circuit Part 1 U E02-0

VC )

VC 99 Control Room

VC System Alarm Circuit Part 2 T E02-0

VC )

VC 99 Control Room