IR 05000413/2009006: Difference between revisions

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{{Adams
{{Adams
| number = ML093270459
| number = ML091250239
| issue date = 11/20/2009
| issue date = 05/04/2009
| title = IR 05000413-09-006, 05000414-09-006, on 7/27/2009 - 08/27/2009; Catawba Nuclear Station, Units 1 and 2; Component Design Basis Inspection
| title = Catawba Nuclear Station - Component Design Bases Inspection - NRC Inspection Report 0500413-09-006 and 0500414-09-006
| author name = Jones D A
| author name = Desai B B
| author affiliation = NRC/RGN-II/DRS/EB1
| author affiliation = NRC/RGN-II/DRS/EB1
| addressee name = Morris J R
| addressee name = Morris J R
| addressee affiliation = Duke Energy Carolinas, LLC, Duke Power Co
| addressee affiliation = Duke Energy Carolinas, LLC, Duke Power Co
| docket = 05000413, 05000414
| docket = 05000413, 05000414, 07200045
| license number = NPF-035, NPF-052
| license number = NPF-035, NPF-052
| contact person =  
| contact person =  
| document report number = IR-09-006
| document report number = IR-09-006
| document type = Inspection Report
| document type = Inspection Report, Letter
| page count = 37
| page count = 6
}}
}}


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=Text=
=Text=
{{#Wiki_filter:
{{#Wiki_filter:
[[Issue date::November 20, 2009]]
[[Issue date::May 4, 2009]]


Mr. J. Site Vice President Duke Power Company, LLC d/b/a Duke Energy Carolinas, LLC Catawba Nuclear Station 4800 Concord Road York, SC 29745-9635
Mr. J. Site Vice President Duke Power Company, LLC d/b/a Duke Energy Carolinas, LLC Catawba Nuclear Station 4800 Concord Road York, SC 29745-9635


SUBJECT: CATAWBA NUCLEAR STATION - NRC COMPONENT DESIGN BASES INSPECTION - INSPECTION REPORT 05000413/2009006 AND 05000414/2009006
SUBJECT: CATAWBA NUCLEAR STATION - COMPONENT DESIGN BASES INSPECTION - NRC INSPECTION REPORT 05000413/2009006 AND 05000414/2009006


==Dear Mr. Morris:==
==Dear Mr. Morris:==
On August 27, 2009, U. S. Nuclear Regulatory Commission (NRC) completed an inspection at your Catawba Nuclear Station Units 1 and 2. The enclosed inspection report documents the inspection results, which were discussed on August 27, 2009, with Mr. Tom Ray and other members of your staff. A re-exit was conducted by telephone on October 15, 2009 with Mr. Randy Hart and other members of your staff.
The purpose of this letter is to notify you that the U.S. Nuclear Regulatory Commission (NRC)
Region II staff will conduct a component design bases inspection at your Catawba Nuclear Station during the period of July 6 - August 28, 2009. The inspection team will be led by W. R. Lewis, a Senior Reactor Inspector from the NRC's Region II Office. This inspection will be conducted in accordance with the baseline inspection procedure, Procedure 71111.21, Component Design Bases Inspection, issued August 19, 2008.


The inspection examined activities conducted under your license as they relate to safety and compliance with the Commission's rules and regulations and with the conditions of your license. The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel.
The inspection will evaluate the capability of risk significant / low margin components to function as designed and to support proper system operation. The inspection will also include a review of selected operator actions, operating experience, and modifications.


This report documents one NRC-identified finding of very low safety significance (Green) which was determined to be a violation of NRC requirements. However, because of the very low safety significance and because it is entered into your corrective action program, the NRC is treating the finding as a non-cited violation (NCV) consistent with Section VI.A.1 of the NRC Enforcement Policy. If you contest this NCV, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the Nuclear Regulatory Commission, ATTN.: Document Control Desk, Washington DC 20555-0001; with copies to the Regional Administrator, Region II; the Director, Office of Enforcement, United States Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC resident inspector at the Catawba Nuclear Plant. In addition, if you disagree with the characterization of 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 II, and the NRC resident inspector at the Catawba Nuclear Plant. The information you provide will be considered in accordance with Inspection Manual Chapter 0305.
During a telephone conversation on April 27, 2009, Mr. Lewis confirmed, with Mr. Mark Sawicki of your staff, arrangements for an information gathering site visit and the three-week onsite inspection. The schedule is as follows:


In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter, its enclosure, and your response (if any) will be available electronically for public inspection in the DEC 2 NRC Public Document Room or from the Publicly Available Records (PARS) component of the NRC's document system (ADAMS). ADAMS is accessible from the NRC web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
* Information gathering visit: Week of July 6 - 10, 2009.


Sincerely,/RA/ David A. Jones, Acting Chief Engineering Branch 1 Division of Reactor Safety Docket Nos.: 50-413, 50-414 License Nos.: NPF-35, NPF-52
* Onsite weeks: Weeks of July 27 - 31, August 10 - 14, and August 24 - 28, 2009.


===Enclosure:===
The purpose of the information gathering visit is to meet with members of your staff to identify risk-significant components and operator actions. Information and documentation needed to support the inspection will also be identified and gathered or requested. Mr. G. MacDonald, a Region II Senior Reactor Analyst, may accompany Mr. Lewis during the information gathering visit to review probabilistic ri sk assessment data and identify risk significant components which will be examined duri ng the inspection.
Inspection Report 05000413/2009006, 05000414/2009006


===w/Attachment:===
DEC 2 The enclosure lists documents that will be needed prior to the information gathering visi t. Please contact Mr. Lewis with any questions prior to preparing materials listed in the enclosure and provide the referenced information to the Region II office by June 19, 2009. The inspectors will try to minimize your administrative burden by specifically identifying only those documents required for the inspection preparation.
Supplemental Information cc w/encl: (See page 3)


_________________________ X G SUNSI REVIEW COMPLETE OFFICE RII:DRS RII:DRS RII:DRS RII:DRS RII:DRS CONTRACTOR CONTRACTOR SIGNATURE R A R A R A R A R A R A R ANAME R. LEWIS R. BERRYMAN D. MAS R. PATTERSONA. ALLEN C. EDWARDS O. MAZZONI DATE 11/19/2009 11/19/2009 11/19/2009 11/19/2009 11/19/2009 11/19/2009 11/19/2009 E-MAIL COPY? YES NO YES NO YES NO YES NO YES NO YES NO YES NO OFFICE RII:DRP RII:DRS SIGNATURE R A R A NAME J. BARTLEY D. JONES DATE 11/20/2009 11/120/2009 E-MAIL COPY? YES NO YES NO YES NO YES NO YES NO YES NO YES NO DEC 3 cc w/encl: Randy D. Hart Regulatory Compliance Manager Duke Energy Carolinas, LLC Electronic Mail Distribution R. L. Gill, Jr. Manager Nuclear Regulatory Issues & Industry Affairs Duke Energy Carolinas, LLC Electronic Mail Distribution Dhiaa M. Jamil Group Executive and Chief Nuclear Officer Duke Energy Carolinas, LLC Electronic Mail Distribution
During the information gathering visit, the team leader will also discuss the following inspection support administrative details: office space; site, plant and information system access; required resources; and information exchange protocol.


Kathryn B. Nolan Senior Counsel Duke Energy Carolinas, LLC 526 South Church Street-EC07H Charlotte, NC 28202 Lisa F. Vaughn Associate General Counsel Duke Energy Carolinas, LLC 526 South Church Street-EC07H Charlotte, NC 28202 Senior Resident Inspector U.S. Nuclear Regulatory Commission Catawba Nuclear Station U.S. NRC 4830 Concord Road York, SC 29745
In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter and its enclosure will be available electronically for public ins pection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of NRC's document system (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).


David A. Repka Winston Strawn LLP Electronic Mail Distribution North Carolina MPA-1 Suite 600 P.O. Box 29513 Raleigh, NC 27525-0513 Susan E. Jenkins Section Manager, Division of Waste Management Bureau of Land and Waste Management S.C. Department of Health and Environmental Control Electronic Mail Distribution R. Mike Gandy Division of Radioactive Waste Mgmt. S.C. Department of Health and Environmental Control Electronic Mail Distribution W. Lee Cox III (Acting) Section Chief Radiation Protection Section N.C. Department of Environmental Commerce & Natural Resources Electronic Mail Distribution
Thank you for your cooperation in this matter. If you have any questions regarding the information requested or the inspection, please contact Mr. Lewis at (404) 562-4541 or me at (404) 562-4519.


Vanessa Quinn Federal Emergency Management Agency 500 C Street, SW Room 840 Washington, DC 20472
Sincerely,/RA/


Steve Weatherman, Operations Analyst North Carolina Electric Membership Corporation Electronic Mail Distribution
Binoy B. Desai, Chief Engineering Branch 1 Division of Reactor Safety Docket Nos.: 50-413, 50-414, 72-045 License Nos.: NPF-35, NPF-52


County Manager of York County York County Courthouse York, SC 29745 Piedmont Municipal Power Agency Electronic Mail Distribution Peggy Force Assistant Attorney General State of North Carolina P.O. Box 629 Raleigh, NC 27602 DEC 4 Letter to J. from David Jones dated November 20, 2009.
===Enclosure:===
Information Request For Catawba Nuclear Station Component Design Bases Inspection


SUBJECT: CATAWBA NUCLEAR STATION - NRC COMPONENT DESIGN BASES INSPECTION - INSPECTION REPORT 05000413/2009006 AND 05000414/2009006 Distribution w/encl:
cc w/encl: (See page 3)
RIDSNRRDIRS PUBLIC A. Adams, NRR RidsNrrPMCatawba Resource


Enclosure U. S. NUCLEAR REGULATORY COMMISSION REGION II
_________________________ X SUNSI REVIEW COMPLETE OFFICE RII:DRS RII:DRS SIGNATURE NAME W.R. Lewis B. Desai DATE 04/ /2009 04/ /2009 E-MAIL COPY? YES NO YES NO YES NO YES NO YES NO YES NO YES NO DEC 3 cc w/encls:
Randy D. Hart Regulatory Compliance Manager Duke Power Company, LLC d/b/a Duke Energy Carolinas, LLC Electronic Mail Distribution


Docket Nos.: 50-413, 50-414 License Nos.: NPF-35, NPF-52
R. L. Gill, Jr.


Report Nos.: 05000413/2009006, 05000414/2009006 Licensee: Duke Energy Carolinas, LLC  
Manager Nuclear Regulatory Issues & Industry Affairs Duke Power Company, LLC d/b/a Duke Energy Carolinas, LLC Electronic Mail Distribution


Facility: Catawba Nuclear Station, Units 1 and 2 Location: York, SC 29745
Dhiaa M. Jamil Group Executive and Chief Nuclear Officer Duke Energy Carolinas, LLC Electronic Mail Distribution


Dates: July 27 - August 27, 2009
Kathryn B. Nolan Senior Counsel Duke Energy Corporation 526 South Church Street-EC07H Charlotte, NC 28202


Inspectors: R. Lewis, Senior Reactor Inspector (Lead) R. Berryman, Senior Reactor Inspector D. Mas-Penaranda, Reactor Inspector R. Patterson, Reactor Inspector A. Allen, Reactor Inspector (Trainee)
Lisa F. Vaughn Associate General Counsel Duke Energy Corporation 526 South Church Street-EC07H Charlotte, NC 28202
O. Mazzoni, Contractor C. Edwards, Contractor Approved by: David Jones, Acting Chief Engineering Branch 1 Division of Reactor Safety Enclosure


=SUMMARY OF FINDINGS=
Senior Resident Inspector Duke Energy Corporation Catawba Nuclear Station U.S. NRC 4830 Concord Road York, SC 29745
IR 05000413/2009006, 05000414/2009006; 7/27/2009 - 08/27/2009; Catawba Nuclear Station, Units 1 and 2; Component Design Basis Inspection.


This inspection was conducted by a team of four NRC inspectors from the Region II office, one trainee, and two NRC contract inspectors. One finding of very low safety significance (i.e. Green) was identified during this inspection. The significance of most findings is indicated by their color (Green, White, Yellow, Red) using IMC 0609, "Significance Determination Process" (SDP). Findings for which the SDP does not apply may be Green or be assigned a severity level after NRC management review. The NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, "Reactor Oversight Process," (ROP) Revision 4, dated December 2006.
David A. Repka Winston Strawn LLP Electronic Mail Distribution


===Cornerstone: Mitigating Systems===
North Carolina MPA-1 Suite 600 P.O. Box 29513 Raleigh, NC 27525-0513


Green:  The team identified a non-cited violation of 10 CFR 50.65(a)(1) for the licensee's failure to monitor the turbine-driven auxiliary feedwater pump (CAPT) sump valves for Units 1 and 2. PIPs C-09-05020 and C-09-04390 initiated immediate corrective actions, including testing of the subject valves during the inspection, wherein valve 1WL848 failed to stroke. Additionally, the licensee increased the maintenance category of the affected components and made procedural modifications to provide positive valve position controls.
Susan E. Jenkins Director, Division of Waste Management Bureau of Land and Waste Management S.C. Department of Health and Environmental Control Electronic Mail Distribution


The team determined that the licensee's failure to monitor the performance and condition of Valve 1WL848 was a performance deficiency. This finding is more than minor because it is associated with equipment performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to perform periodic testing or preventative maintenance resulted in a lack of reasonable assurance that the valves would perform their function of protecting CAPT. The team determined that the finding is of very low safety significance (Green) using the SDP because the finding did not represent an actual loss of safety function. This finding was reviewed for cross-cutting aspects and none were identified since the performance deficiency was not indicative of current licensee performance. (Section 1R21.2.5)
R. Mike Gandy Division of Radioactive Waste Mgmt.


=REPORT DETAILS=
S.C. Department of Health and Environmental Control Electronic Mail Distribution


==REACTOR SAFETY==
Beverly O. Hall Chief, Radiation Protection Section Department of Environmental Health N.C. Department of Environmental Commerce & Natural Resources Electronic Mail Distribution
Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity
{{a|1R21}}
==1R21 Component Design Bases Inspection==
{{IP sample|IP=IP 71111.21}}
===.1 Inspection Sample Selection Process===


The team selected risk significant components and operator actions for review using information contained in the licensee's Probabilistic Risk Assessment (PRA). In general, this included components and operator actions that had a risk achievement worth factor greater than 1.3 or Birnbaum value greater than 1 X10
Elizabeth McMahon Assistant Attorney General S.C. Attorney General's Office P.O. Box 11549 Columbia, SC 29211
-6. The components selected were generally located within the following systems:  Component Cooling, Auxiliary Feedwater, Service Water, Residual Heat Removal, Charging, Safety Injection, 4kV Distribution, Hydrogen Mitigation, and Standby Shutdown Facility. The sample included 19 components, 3 operating experience items, and 5 operator actions. The team performed a margin assessment and detailed review of the selected risk-significant components to verify that the design bases had been correctly implemented and maintained. This design margin assessment considered original design issues, margin reductions due to modifications, or margin reductions identified as a result of material condition issues. Equipment reliability issues were also considered in the selection of components for detailed review. These reliability issues included items related to failed performance test results, significant corrective action, repeated maintenance, maintenance rule (a)1 status, RIS 05-020 (formerly GL 91-18) conditions, NRC resident inspector input of problem equipment, System Health Reports, industry operating experience and licensee problem equipment lists. Consideration was also given to the uniqueness and complexity of the design, operating experience, and the available defense in depth margins. An overall summary of the reviews performed and the specific inspection findings identified is included in the following sections of the report.


===.2 Results of Detailed Reviews===
Vanessa Quinn Federal Emergency Management Agency 500 C Street, SW Room 840 Washington, DC 20472


===.2.1 2KCPUA1, 2A1 Component Cooling Pump===
Steve Weatherman, Operations Analyst North Carolina Electric Membership Corporation Electronic Mail Distribution
 
====a. Inspection Scope====
The inspectors reviewed the plant technical specifications (TS), updated final safety analysis (UFSAR), design basis documents (DBDs) and associated system lesson plans to establish an overall understanding of the design bases of the component. Design calculations (i.e. net positive suction head (NPSH), total dynamic head (TDH), pump curves, vortex, seismic qualification, component cooling system flow distribution) and site procedures were reviewed against these bases to verify design assumptions have been appropriately translated into these documents. The team requested and reviewed system modifications over the life of the component to verify that the subject modifications did not serve to degrade the component's performance capability and were appropriately incorporated into relevant drawings and procedures. Equipment ratings and specifications were reviewed against design basis requirements to ensure that the equipment qualification is suitable for all expected conditions. A component walkdown was conducted to verify that the installed configuration will support design basis function under accident/event conditions and have been maintained to be consistent with design assumptions. Test procedures and more recent test results were reviewed against design basis documents to verify that acceptance criteria for tested parameters were supported by calculations or other engineering documents and that individual tests and/or analyses validated component operation under accident/event conditions.
 
Vendor documentation, system health reports, preventive and corrective maintenance history, and corrective action system documents were reviewed in order to verify that potential degradation was monitored or prevented and the component replacement was consistent with in service/equipment qualification life. The team reviewed the licensee's response to BL 88-04, Potential Safety-Related Pump Loss, in order to verify that applicable insights from operating experience have been applied to the selected components.
 
====b. Findings====
No findings of significance were identified.
 
===.2.2 1RNPUB, 1B Nuclear Service Water Pump===
 
====a. Inspection Scope====
The inspectors reviewed the plant TS, UFSAR, DBDs and associated system lesson plans to establish an overall understanding of the design bases of the component. Design calculations (i.e., CNC 1223.24-00-0027 and CNC 1223.24-00-0058) and site procedures were reviewed against these bases to verify design assumptions have been appropriately translated into these documents. The team requested and reviewed system modifications over the life of the component to verify that the subject modifications did not serve to degrade the component's performance capability and were appropriately incorporated into relevant drawings and procedures. Equipment ratings and specifications were reviewed against design basis requirements to ensure that the equipment qualification was suitable for all expected conditions. A component walkdown was conducted to verify that the installed configuration would support design basis function under accident/event conditions and had been maintained to be consistent with design assumptions. Test procedures and more recent test results were reviewed against design basis documents to verify that acceptance criteria for tested parameters were supported by calculations or other engineering documents and that individual tests and/or analyses validated component operation under accident/event conditions.
 
Vendor documentation, system health reports, preventive and corrective maintenance history, and corrective action system documents were reviewed in order to verify that potential degradation was monitored or prevented and the component replacement was consistent with in-service/equipment qualification life.
 
====b. Findings====
No findings of significance were identified.
 
===.2.3 2CAPUA, 2A Motor-Driven Auxiliary Feedwater Pump===
 
====a. Inspection Scope====
The inspectors reviewed the plant TS, UFSAR, DBDs and associated system lesson plans to establish an overall understanding of the design bases of the component. Design calculations (i.e. minimum flow and test acceptance) and site procedures were reviewed against these bases to verify design assumptions have been appropriately translated into these documents. The team requested and reviewed system modifications over the life of the component to verify that the subject modifications did not serve to degrade the component's performance capability and were appropriately incorporated into relevant drawings and procedures. A component walkdown was conducted to verify that the installed configuration would support design basis function under accident/event conditions and have been maintained to be consistent with design assumptions. Control panel indicators were observed and operating procedures reviewed to verify that component operation and alignments were consistent with design and licensing basis assumptions. Additionally, the inspectors reviewed time critical operator actions where those actions supported design basis assumptions or conclusions. Alternate flow paths and water sources, as well as possible diversion paths, were reviewed to verify that the process medium would be available and unimpeded during an accident. Test procedures and more recent test results were reviewed against design basis documents to verify that acceptance criteria for tested parameters were supported by calculations or other engineering documents and that individual tests and/or analyses validated component operation under accident/event conditions. Vendor documentation, system health reports, preventive and corrective maintenance history, and corrective action system documents were reviewed in order to verify that potential degradation was monitored or prevented and the component replacement was consistent with in service/equipment qualification life.
 
====b. Findings====
No findings of significance were identified.
 
===.2.4 2KDHXA, Emergency Diesel Generator Jacket Cooling Water Heat Exchanger===
 
====a. Inspection Scope====
The inspectors reviewed the plant TS, UFSAR, DBDs and associated system lesson plans to establish an overall understanding of the design bases of the component. Design calculations (i.e., CNC 1223.59-01-0005) and site procedures were reviewed against these bases to verify design assumptions have been appropriately translated into these documents. The team requested and reviewed system modifications over the life of the component to verify that the subject modifications did not serve to degrade the component's performance capability and were appropriately incorporated into relevant drawings and procedures. A component walkdown was conducted to verify that the installed configuration would support design basis function under accident/event conditions and had been maintained to be consistent with design assumptions. External event analyses were reviewed against design specifications and requirements in order to verify that the equipment was adequately protected. Test procedures and more recent test results were reviewed against design basis documents to verify that acceptance criteria for tested parameters were supported by calculations or other engineering documents and that individual tests and/or analyses validated component operation under accident/event conditions. Vendor documentation, system health reports, preventive and corrective maintenance history, and corrective action system documents were reviewed in order to verify that potential degradation was monitored or prevented and the component replacement was consistent with in service/equipment qualification life.
 
====b. Findings====
No findings of significance were identified.
 
===.2.5 1WL848, Floor Drain Sump D Discharge to Turbine Bldg Sump Air-Operated Valve===
 
====a. Inspection Scope====
The inspectors reviewed the plant TS, UFSAR, DBDs and associated system lesson plans to establish an overall understanding of the design bases of the component. Design calculations (i.e., site flooding analysis) and site procedures were reviewed against these bases to verify design assumptions have been appropriately translated into these documents. The team requested and reviewed system modifications over the life of the component to verify that the subject modifications did not serve to degrade the component's performance capability and were appropriately incorporated into relevant drawings and procedures. A component walkdown was conducted to verify that the installed configuration would support design basis function under accident/event conditions and have been maintained to be consistent with design assumptions. Control panel indicators were observed and operating procedures reviewed to verify that component operation and alignments were consistent with design and licensing basis assumptions. Where manual actions were credited, the inspector's review included the availability of any necessary tools or accessibility aids. Additionally, the inspectors reviewed time critical operator actions where those actions supported design basis assumptions or conclusions. Test procedures and more recent test results were reviewed against design basis documents to verify that acceptance criteria for tested parameters were supported by calculations or other engineering documents and that individual tests and/or analyses validated component operation under accident/event conditions. Vendor documentation, system health reports, preventive and corrective maintenance history, and corrective action system documents were reviewed in order to verify that potential degradation was monitored or prevented and the component replacement was consistent with in service/equipment qualification life.
 
====b. Findings====
 
=====Introduction:=====
The team identified a Green non-cited violation of 10 CFR 50.65(a)(1) for the licensee's failure to monitor, in accordance with established criteria, the performance and condition of the turbine-driven auxiliary feedwater pump (CAPT) sump valves for Units 1 and 2.
 
=====Description:=====
Valve 1/2WL848 is a normally open, fail as-is, locally controlled, plug valve with a pneumatic operator and a packless stem. The valve is located in the common discharge header from the auxiliary feedwater (CA) pump pit sumps.
 
Valve 1/2WL848, when open, permits the discharge of the CAPT sumps to the turbine building sump when less than pre-determined levels of radiation are sensed by radiation monitor 1/2EMF52. Valve 1/2WL848 closes upon receipt of a high radiation signal so that effluent from floor drain sump D would not be discharged to the turbine building sump. To ensure that any release to the environment would be kept as low as reasonably achievable (ALARA), the closure of valve 1/2WL848 results in the diversion of this potentially contaminated stream to the residual heat removal (ND) and containment spray (NS) pump rooms sump through valve 1/2WL847, a normally closed valve of the same type which opens upon the high radiation signal. Valve 1/2WL848, if failed closed, could result in the overflow of the CAPT sump and eventually affect the continued operation of the CAPT. The CAPT, a safety-related component, requires a small amount of cooling flow (typically less than 15gpm) which creates a constant input to the CAPT pit sump. The conservative estimate of the time for increased water level to impact CAPT operation was 86.8 minutes following failure of the sump system. The team noted that since plant start-up the licensee failed to inspect, test or perform preventative maintenance on 1/2WL848 and 1/2WL847. During the inspection, the licensee performed performance test (PT/1/A/4700/020) and Valve 1WL848 failed to stroke closed. When the closed pushbutton was pressed, the valve remained in the full open position and air was leaking from both above and below the valve's air actuator. In several attempts, the valve would not move from the full open position. The licensee also tested valve 1WL847; it stroked successfully open and closed. 1/2WL847 is normally closed to isolate the flow path to the ND/NS auxiliary building sump. The team determined that the licensee failed to provide reasonable assurance that valve 1WL848 was capable of fulfilling its intended function. During the inspection the licensee repaired the 1WL848 actuator and retested it satisfactorily. Also the licensee re-categorized the valves in the discharge flow path from risk category C (run-to-failure) to risk category A (highly risk-significant) and enhanced procedural guidance to ensure a flow path will always be available from the CAPT pit sump.
 
=====Analysis:=====
The team determined that the licensee's failure to monitor the performance and condition of valve 1/2WL848 was a performance deficiency. This finding was more than minor because it was associated with equipment performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to perform periodic testing or preventative maintenance resulted in a lack of reasonable assurance that the valves would perform their safety important function in protecting the CAPT. Using Inspection Manual Chapter 0609, "Significance Determination Process," Phase 1 Worksheet, the team determined that the finding is of very low safety significance (Green) because the finding did not represent an actual loss of safety function. This finding was reviewed for cross-cutting aspects and none were identified since the performance deficiency was not indicative of current licensee performance.
 
=====Enforcement:=====
10 CFR 50.65(a)(1) states, in part, that licensees shall monitor the performance or condition of structures, systems and components (SSCs) within the scope of the rule as defined by 10 CFR 50.65(b), against license established goals, in a manner sufficient to provide reasonable assurance that such SSCs are capable of fulfilling their intended function. 10 CFR 50.65(a)(2) states that monitoring as specified in paragraph (a)(1)- is not required where it has been demonstrated that the performance or condition of a SSC is being effectively controlled through the performance of appropriate preventative maintenance, such that the SSC remains capable of performing its intended function. Contrary to the above, the licensee failed to demonstrate that the performance or condition of the CAPT pit sump discharge valve had been effectively controlled through the performance of appropriate preventive maintenance, testing or inspection and did not monitor against licensee established goals. Specifically, since plant start-up the licensee failed to monitor the performance and condition of Valve 1/2WL848, an expectation which was enacted by regulation effective July 10, 1996. Because this finding is of very low safety significance and because it was entered into the licensee's corrective action program as PIP C-09-05020 and PIP C-09-04390, this violation is being treated as an NCV, consistent with the NRC Enforcement Policy:  NCV 05000413, 414/2009006-001, Failure to monitor the turbine-driven auxiliary feedwater pump sump valves for units 1 and 2.
 
===.2.6 2ND28A, ND Supply to NV and 2A NI Pumps Motor-Operated Valve===
 
====a. Inspection Scope====
The inspectors reviewed the plant TS, UFSAR, DBDs and associated system lesson plans to establish an overall understanding of the design bases of the component. Design calculations (i.e., CNC 1205.19-00-0060) and site procedures were reviewed against these bases to verify design assumptions have been appropriately translated into these documents. The inspector reviewed the calculations for the degraded voltage at the motor-operated valve (MOV) terminals, to ensure the proper voltage was utilized in the team's review of MOV torque calculations. The team requested and reviewed system modifications over the life of the component to verify that the subject modifications did not serve to degrade the component's performance capability and were appropriately incorporated into relevant drawings and procedures. A component walkdown was conducted to verify that the installed configuration would support design basis function under accident/event conditions and had been maintained to be consistent with design assumptions. Control room indicators were observed and operating procedures reviewed to verify that component operation and alignments were consistent with design and licensing basis assumptions. Additionally, the inspectors reviewed time critical operator actions where those actions supported design basis assumptions or conclusions. Test procedures and more recent test results were reviewed against design basis documents to verify that acceptance criteria for tested parameters were supported by calculations or other engineering documents and that individual tests and/or analyses validated component operation under accident/event conditions. Vendor documents, system health reports, preventive and corrective maintenance history, and corrective action system documents were reviewed in order to verify that potential degradation was monitored or prevented and the component replacement was consistent with in service/equipment qualification life.
 
====b. Findings====
No findings of significance were identified.
 
===.2.7 2NI136B, ND Supply to NI Pump 2B Motor-Operated Valve===
 
====a. Inspection Scope====
The inspectors reviewed the plant TS, UFSAR, DBDs and associated system lesson plans to establish an overall understanding of the design bases of the component. Design calculations (i.e., CNC 1205.19-00-0021) and site procedures were reviewed against these bases to verify design assumptions have been appropriately translated into these documents. The inspector reviewed the calculations for the degraded voltage at the MOV terminals, to ensure the proper voltage was utilized in the team's review of MOV torque calculations. The team requested and reviewed system modifications over the life of the component to verify that the subject modifications did not serve to degrade the component's performance capability and were appropriately incorporated into relevant drawings and procedures. A component walkdown was conducted to verify that the installed configuration would support design basis function under accident/event conditions and had been maintained to be consistent with design assumptions. Control room indicators were observed and operating procedures reviewed to verify that component operation and alignments were consistent with design and licensing basis assumptions. Additionally, the inspectors reviewed time critical operator actions where those actions supported design basis assumptions or conclusions. Test procedures and more recent test results were reviewed against design basis documents to verify that acceptance criteria for tested parameters were supported by calculations or other engineering documents and that individual tests and/or analyses validated component operation under accident/event conditions. Vendor documents, system health reports, preventive and corrective maintenance history, and corrective action system documents were reviewed in order to verify that potential degradation was monitored or prevented and the component replacement was consistent with in service/equipment qualification life.
 
====b. Findings====
No findings of significance were identified.
 
===.2.8 1NV865, Standby M/U Pump Suction from Transfer Tube Motor-Operated Valve===
 
====a. Inspection Scope====
The inspectors reviewed the plant TS, UFSAR, DBDs and associated system lesson plans to establish an overall understanding of the design bases of the component. Design calculations (i.e., operating parameters and MOV calculations) and site procedures were reviewed against these bases to verify design assumptions have been appropriately translated into these documents. The inspector reviewed the calculations for the degraded voltage at the MOV terminals, to ensure the proper voltage was utilized in the team's review of MOV torque calculations. The team requested and reviewed system modifications over the life of the component to verify that the subject modifications did not serve to degrade the component's performance capability and were appropriately incorporated into relevant drawings and procedures. A component walkdown was conducted to verify that the installed configuration would support design basis function under accident/event conditions and have been maintained to be consistent with design assumptions. Test procedures and more recent test results were reviewed against design basis documents to verify that acceptance criteria for tested parameters were supported by calculations or other engineering documents and that individual tests and/or analyses validated component operation under accident/event conditions. Vendor documents, system health reports, preventive and corrective maintenance history, and corrective action system documents were reviewed in order to verify that potential degradation was monitored or prevented and the component replacement was consistent with in service/equipment qualification life.
 
====b. Findings====
No findings of significance were identified.
 
===.2.9 1NV872A, Standby M/U Pump Filter Outlet Motor-Operated Valve===
 
====a. Inspection Scope====
The inspectors reviewed the plant TS, UFSAR, DBDs and associated system lesson plans to establish an overall understanding of the design bases of the component. Design calculations (i.e., operating parameters and MOV calculations) and site testing procedures were reviewed against these bases to verify design assumptions have been appropriately translated into these documents. The inspector reviewed the calculations for the degraded voltage at the MOV terminals, to ensure the proper voltage was utilized in the team's review of MOV torque calculations. The team requested and reviewed system modifications over the life of the component to verify that the subject modifications did not serve to degrade the component's performance capability and were appropriately incorporated into relevant drawings and procedures. A component walkdown was conducted to verify that the installed configuration would support design basis function under accident/event conditions and have been maintained to be consistent with design assumptions. Test procedures and more recent test results were reviewed against design basis documents to verify that acceptance criteria for tested parameters were supported by calculations or other engineering documents and that individual tests and/or analyses validated component operation under accident/event conditions. Vendor documentation, system health reports, preventive and corrective maintenance history, and corrective action system documents were reviewed in order to verify that potential degradation was monitored or prevented and the component replacement was consistent with in service/equipment qualification life.
 
====b. Findings====
No findings of significance were identified.
 
===.2.10 1NV813, ND Pump Discharge to NI Pump Suction Check Valve===
 
====a. Inspection Scope====
The inspectors reviewed the plant TS, UFSAR, DBDs and associated system lesson plans to establish an overall understanding of the design bases of the component. Site procedures were reviewed against these bases to verify design assumptions have been appropriately translated into these documents. A component walkdown was conducted to verify that the installed configuration would support design basis function under accident/event conditions and had been maintained to be consistent with design assumptions. External event analyses were reviewed against design specifications and requirements in order to verify that the equipment was adequately protected. Test procedures and more recent test results were reviewed against design basis documents to verify that acceptance criteria for tested parameters were supported by calculations or other engineering documents and that individual tests and/or analyses validated component operation under accident/event conditions. Vendor documentation, system health reports, preventive and corrective maintenance history, and corrective action system documents were reviewed in order to verify that potential degradation was monitored or prevented and the component replacement was consistent with in service/equipment qualification life.
 
====b. Findings====
No findings of significance were identified.
 
===.2.11 1WL850, CA Pump Room Sump Discharge Header Check Valve===
 
====a. Inspection Scope====
The inspectors reviewed the plant TS, UFSAR, DBDs and associated system lesson plans to establish an overall understanding of the design bases of the component. Design calculations, site testing procedures and vendor documentation were reviewed against these bases to verify design assumptions have been appropriately translated into these documents. The team requested and reviewed system modifications over the life of the component to verify that the subject modifications did not serve to degrade the component's performance capability and were appropriately incorporated into relevant drawings and procedures. A component walkdown was conducted to verify that the installed configuration would support design basis function under accident/event conditions and have been maintained to be consistent with design assumptions. Test procedures and more recent test results were reviewed against design basis documents to verify that acceptance criteria for tested parameters were supported by calculations or other engineering documents and that individual tests and/or analyses validated component operation under accident/event conditions. System health reports, preventive and corrective maintenance history, and corrective action system documents were reviewed in order to verify that potential degradation was monitored or prevented and the component replacement was consistent with in service/equipment qualification life.
 
====b. Findings====
No findings of significance were identified.
 
===.2.12 2ETA, 4160VAC Essential Power Switchgear===
 
====a. Inspection Scope====
The inspectors reviewed the plant TS, UFSAR, DBDs and associated system lesson plans to establish an overall understanding of the design bases of the component. Voltage and short circuit calculations, as well as switchgear test, maintenance and operational procedures were reviewed to verify that design bases and design assumptions have been appropriately translated into design calculations and procedures. Testing procedures and recent results were reviewed to verify that acceptance criteria for tested parameters were supported by calculations or other engineering documents to ensure that design and licensing bases were met and to verify that individual tests and/or analyses validated component operation under accident/event conditions. The inspectors conducted a walkdown of the switchgear, reviewed vendor manuals and construction drawings, and performed alignment verifications to verify that the component's installed configuration would support its design basis function under accident/event conditions and that the equipment was properly protected. Control wiring diagrams and DC loading calculations were reviewed to verify that component inputs and outputs were suitable for application and would be acceptable under accident/event conditions. System health reports, component maintenance history and licensee corrective action program reports were reviewed to verify that potential degradation was monitored or prevented and the component replacement was consistent with in service/equipment qualification life. Environmental qualification documents and procurement specifications were reviewed to verify that equipment qualification was suitable for the environment expected under all conditions.
 
====b. Findings====
No findings of significance were identified.
 
===.2.13 1ETA06, 1A1 KC Pump Motor Power Supply Breaker===
 
====a. Inspection Scope====
The inspectors reviewed the plant TS, UFSAR, DBDs and associated system lesson plans to establish an overall understanding of the design bases of the component.
 
Coordination and motor starting curves were reviewed, along with short circuit calculations, and maintenance and testing procedures, to verify that design bases and design assumptions have been appropriately translated into design calculations and procedures. Control wiring diagrams and DC loading calculations were reviewed to verify that component inputs and outputs were suitable for application and would be acceptable under accident/event conditions. System health reports, component maintenance history and licensee corrective action program reports were reviewed to verify that potential degradation was monitored or prevented and the component replacement was consistent with in service/equipment qualification life. Environmental qualification documents and procurement specifications were reviewed to verify that equipment qualification was suitable for the environment expected under all conditions.
 
====b. Findings====
No findings of significance were identified.
 
===.2.14 1EBA, 1A Essential Battery Bank===
 
====a. Inspection Scope====
The inspectors reviewed the plant TS, UFSAR, DBDs and associated system lesson plans to establish an overall understanding of the design bases of the component.
 
Battery sizing, loading, and voltage calculations were reviewed, as were maintenance and operational procedures, in order to verify that design bases and design assumptions have been appropriately translated into design calculations and procedures. Test procedures and recent testing results were reviewed in order to verify that acceptance criteria for tested parameters were supported by calculations or other engineering documents to ensure that design and licensing bases were met and that individual tests and or analyses validated component operation under accident/event conditions. Battery room temperature evaluations were reviewed to verify that the equipment qualification was suitable for the environment expected under all conditions. The inspectors conducted a walkdown of the batteries and their associated chargers, reviewed vendor manuals and construction drawings, and performed focused field inspections to verify that the component's installed configuration would support its design basis function under accident/event conditions and that the equipment was properly protected.
 
Interviews with system engineers and maintenance personnel were conducted, system health reports, component maintenance history and licensee corrective action program reports were reviewed to verify that potential degradation was monitored or prevented and the component replacement was consistent with in service/equipment qualification life.
 
====b. Findings====
No findings of significance were identified.
 
===.2.15 Standby Shutdown Facility (SSF) Aux Diesel===
 
====a. Inspection Scope====
The inspectors reviewed the plant TS, UFSAR, DBDs and associated system lesson plans to establish an overall understanding of the design bases of the component. Loading, voltage drop, and short circuit calculations were reviewed, as were maintenance and operational procedures, in order to verify that design bases and design assumptions have been appropriately translated into design calculations and procedures. Coordination capability was not specifically addressed. Test procedures and recent testing results were reviewed in order to verify that acceptance criteria for tested parameters were supported by calculations or other engineering documents to ensure that design and licensing bases were met and that individual tests and or analyses validated component operation under accident/event conditions. The inspectors conducted a walkdown of the engine and associated support systems and distribution, reviewed vendor manuals and construction drawings, and performed focused field inspections to verify that the component's installed configuration would support its design basis function under accident/event conditions and that the equipment was properly protected.
 
Interviews with system engineers and maintenance personnel were conducted, system health reports, component maintenance history and licensee corrective action program reports were reviewed to verify that potential degradation was monitored or prevented and the component replacement was consistent with in service/equipment qualification life.
 
====b. Findings====
No findings of significance were identified.
 
===.2.16 EHM, Hydrogen Mitigation System===
 
====a. Inspection Scope====
The inspectors reviewed the plant TS, UFSAR, DBDs and associated system lesson plans to establish an overall understanding of the design bases of the component. System one line diagrams and control circuit drawings were reviewed to develop an understanding of system architecture. Test procedures and recent testing results were reviewed in order to verify that acceptance criteria for tested parameters were supported by calculations or other engineering documents to ensure that design and licensing bases were met and that individual tests and or analyses validated component operation under accident/event conditions. The inspectors conducted a walkdown of the system, reviewed vendor manuals and construction drawings, and performed focused field inspections to verify that the component's installed configuration would support its design basis function under accident/event conditions and that the equipment was properly protected.
 
Interviews with system engineers and maintenance personnel were conducted, system health reports, component maintenance history and licensee corrective action program reports were reviewed to verify that potential degradation was monitored or prevented and the component replacement was consistent with in service/equipment qualification life.
 
====b. Findings====
No findings of significance were identified.
 
===.2.17 1NC31B, Unit 1 Pressurizer PORV Isolation===
 
====a. Inspection Scope====
The inspectors reviewed the plant TS, UFSAR, DBDs and associated system lesson plans to establish an overall understanding of the design bases of the component. System one line diagrams and control circuit drawings were reviewed to develop an understanding of system architecture. Test procedures and recent testing results were reviewed in order to verify that acceptance criteria for tested parameters were supported by calculations or other engineering documents to ensure that design and licensing bases were met and that individual tests and or analyses validated component operation under accident/event conditions. The inspectors conducted a walkdown of the system, reviewed vendor manuals and construction drawings, and performed focused field inspections to verify that the component's installed configuration would support its design basis function under accident/event conditions and that the equipment was properly protected.
 
Interviews with system engineers and maintenance personnel were conducted, system health reports, component maintenance history and licensee corrective action program reports were reviewed to verify that potential degradation was monitored or prevented and the component replacement was consistent with in service/equipment qualification life.
 
====b. Findings====
No findings of significance were identified.
 
===.2.18 1NDPG5040/5050, ND Pump Minimum Flow===
 
====a. Inspection Scope====
The inspectors reviewed the plant TS, UFSAR, DBDs and associated system lesson plans to establish an overall understanding of the design bases of the component. Instrument control drawings, setpoint and loop uncertainty calculations were reviewed to verify that design bases and design assumptions have been appropriately translated. Instrument modifications were reviewed to verify that the performance capability of the component had not been degraded. Calibration procedures and testing results were reviewed to verify that acceptance criteria for tested parameters were supported by calculations or other engineering documents and that individual tests and or analyses validated component operation under accident/event conditions. The inspectors conducted a walkdown of the instrument, reviewed vendor manuals and construction drawings, and performed alignment verifications to verify that the component's installed configuration would support its design basis function under accident/event conditions and that the equipment was properly protected. Interviews with system engineers and maintenance personnel were conducted, system health reports, component maintenance history and licensee corrective action program reports were reviewed to verify that potential degradation was monitored or prevented and the component replacement was consistent with in service/equipment qualification life.
 
====b. Findings====
No findings of significance were identified.
 
===.2.19 1NVLT5760/5761, Volume Control Tank Level===
 
====a. Inspection Scope====
The inspectors reviewed the plant TS, UFSAR, DBDs and associated system lesson plans to establish an overall understanding of the design bases of the component. Instrument control drawings, setpoint and loop uncertainty calculations were reviewed to verify that design bases and design assumptions have been appropriately translated. Instrument modifications were reviewed to verify that the performance capability of the component had not been degraded. Calibration procedures and testing results were reviewed to verify that acceptance criteria for tested parameters were supported by calculations or other engineering documents and that individual tests and or analyses validated component operation under accident/event conditions. The inspectors conducted a walkdown of the instrument, reviewed vendor manuals and construction drawings, and performed alignment verifications to verify that the component's installed configuration would support its design basis function under accident/event conditions and that the equipment was properly protected. Interviews with system engineers and maintenance personnel were conducted, system health reports, component maintenance history and licensee corrective action program reports were reviewed to verify that potential degradation was monitored or prevented and the component replacement was consistent with in service/equipment qualification life.
 
====b. Findings====
No findings of significance were identified.
 
===.3 Review of Low Margin Operator Actions===
 
====a. Inspection Scope====
The team performed a margin assessment and detailed review of five risk significant and time critical operator actions. Where possible, margins were determined by the review of the assumed design basis and UFSAR response times and performance times documented by job performance measures (JPMs). For the selected components and operator actions, the team performed an assessment of the Emergency Procedures (EPs), Abnormal Procedures (APs), control room alarm response procedures, and other operations procedures to determine the adequacy of the procedures and availability of equipment required to complete the actions. Operator actions were observed on the plant simulator and during plant walk downs. The following operator actions were observed on the licensee's operator training simulator:
* EP actions to establish high pressure recirculation flow per EP/1/A/5000/ES-1.3, Transfer to Cold Leg Recirculation
* EP actions to energize hydrogen igniters per EP/1/A/5000/E-0, Reactor Trip or Safety Injection
* EP actions to recognize and close the block valve in the event of a stuck open pressurizer PORV per EP/1/A/5000/E-0, Reactor Trip or Safety Injection
* Recover from a loss of nuclear service water (RN) due to the failure of an RN pump discharge valve to open per OP/1/A/6100/010M, Panel 1AD-12, A/1, RN Pump A Flow Hi/Lo and OP/0/A/6400/006C, Nuclear Service Water System. Additionally, the team walked down, "table-topped" and investigated the following operational scenario:
* AP actions to align back-up cooling to the 1A NV pump per AP/1/A/5500/021, Loss of Component Cooling, Enclosure 4
 
====b. Findings====
No findings of significance were identified.
 
===.4 Review of Industry Operating Experience===
 
====a. Inspection Scope====
The team reviewed selected operating experience issues that had occurred at domestic and foreign nuclear facilities for applicability at the Catawba Nuclear Plant. The team performed an independent applicability review for issues that were identified as applicable to the Catawba Nuclear Plant and were selected for a detailed review. The issues that received a detailed review by the team included:
* IN 86-37, Degradation of Station Batteries
* IN 06-15, Vibration-Induced Degradation and Failure of Safety-Related Valves
* IN 07-05, Vertical Deep Draft Pump Shaft and Coupling Failures
 
====b. Findings====
No findings of significance were identified.
 
==OTHER ACTIVITIES==
{{a|4OA6}}
==4OA6 Meetings, Including Exit On August 27, 2009, the team presented the inspection results to Mr. Tom Ray and other members of the licensee staff.==
A re-exit was conducted by telephone on October 15, 2009, with Mr. Randy Hart and other members of the licensee staff, following the successful isochronous testing of your standby shutdown facility (SSF) emergency diesel generator and the team's review of circumstances surrounding your SSF coordination calculations. No proprietary information was reviewed as part of the inspection.


ATTACHMENT:  SUPPPLEMENTAL INFORMATION 
County Manager of York County York County Courthouse York, SC 29745


=SUPPLEMENTAL INFORMATION=
Piedmont Municipal Power Agency Electronic Mail Distribution


==KEY POINTS OF CONTACT==
Peggy Force Assistant Attorney General State of North Carolina P.O. Box 629 Raleigh, NC 27602 DEC 4 Letter to J. from Binoy B. Desai, dated May 4, 2009.


===Licensee personnel===
SUBJECT: CATAWBA NUCLEAR STATION - COMPONENT DESIGN BASES INSPECTION - NRC INSPECTION REPORT 05000413/2009006 AND 05000414/2009006 Distribution w/encl
:
: Jon Thompson, NRR PUBLIC RidsNrrPMCatawba Resource Institute of Nuclear Power Operations (INPO)
: [[contact::T. Brooks]], Operations Requalification Supervisor
Enclosure INFORMATION REQUEST FOR CATAWBA NUCLEAR STATION COMPONENT DESIGN BASES INSPECTION
: [[contact::W. Smith]], Operations Instructor
: [[contact::R. Hart]], Licensing Manager
: [[contact::T. Ray]], Engineering Manager
: [[contact::M. Perry]], Power Supervisor, RES
===NRC personnel===
: [[contact::A. Sabich]], Former Senior Resident Inspector
: [[contact::A. Hutto]], Senior Resident Inspector


==LIST OF ITEMS==
Please provide the information electronically in ".pdf" files, Excel, or other searchable format on CDROM. The CDROM should be indexed and possibly hyperlinked to facilitate ease of use. Information in "lists" should contain enough information to be easily understood to someone who has knowledge of pressurized water reactor technology.


===OPENED, CLOSED AND DISCUSSED===
1. From your most recent probabilistic safety analysis (PSA) excluding external events and fires, please provide: a. Two risk rankings of components from your site-specific probabilistic safety analysis (PSA) - one sorted by Risk Achievement Worth (RAW), and the other sorted by Birnbaum Importance. b. A list of the top 500 cutsets.


===Opened and Closed===
2. From your most recent probabilistic safety analysis (PSA) including external events and fires, please provide: a. Two risk rankings of components from your site-specific probabilistic safety analysis (PSA) - one sorted by Risk Achievement Worth (RAW), and the other sorted by Birnbaum Importance. b. A list of the top 500 cutsets.
: 05000413, 414/2009006-01 NCV Failure to monitor the turbine-driven auxiliary feedwater pump sump valves for units 1 and 2  (Section 1R21.2.5)


==LIST OF DOCUMENTS REVIEWED==
3. Risk ranking of operator actions from your site specific PSA sorted by RAW. Provide copies of your human reliability worksheets for these items. 4. Any pre-existing evaluation or list of components and calculations with low design margins (i.e., pumps closest to the design limit for flow or pressure, diesel generator close to design required output, heat exchangers close to rated design heat removal, MOV risk-
Licensing Documents
margin rankings, etc.). 5. A list of station applicability evaluations/reviews performed and document ed in the station corrective action program in the past two years for industry events, critical equipment failures, and safety related equipment vul nerabilities (as communicated by NRC generic communications, industry communications, 10 CFR Part 21 notifications, etc.). 6. A list of operability evaluations completed within the last two years, sorted by associated component or system. 7. A list of common-cause failures of components that have occurred at Catawba Nuclear Station and have been identified within the last five years. 8. A list of equipment currently planned for upgrade/improvement by the site (e.g. "ONE" List), including a description of the reason(s) why each component (i.e. not a programmatic or system level concern) is on that list and summaries (if available) of your plans to address those reasons. 9. A list of equipment currently in RIS 05-020 (formerly GL 91-18) status, or in MR (a)(1) status. 10. Contact information for a person to discuss PRA information prior to the information gathering trip: name, title, phone number, and e-mail address.
: TS, Current UFSAR, Current
: SER and Supplements Design Basis Documents (Functional System Descriptions)
: CNS-0165.EHM-00-0001, Error! Reference source not found.Error! Reference source not found., Rev. 8
: CNS-115.01-EPC-0001, 4.16KV Essential Auxiliary Power System (EPC) and Class 1E DG Protective Relaying and Metering System (ERN), Rev. 9
: CNS-1554.NV-00-0001 Chemical and Volume Control (NV) System, Rev. 36
: CNS-1560.SS-00-0001, SSF, Standby Shutdown Facility Aux Diesel System, Rev. 25
: CNS-1561.ND-00-0001, Residual Heat Removal (ND) System, Rev. 29
: CNS-1562.NI-00-0001, Design basis Specification for Safety Injection System, Rev. 39
: CNS-1565.WL-00-0001 Liquid Waste System, Rev. 34, 04/25/2009
: CNS-1573.KC-00-0001, Design Basis Specification for the Component Cooling System (KC), Rev. 34 
: Attachment
: CNS-1574.RN-00-0001, Nuclear Service Water System (RN), Rev. 52
: CNS-1592.CA-00-0001, Auxiliary Feedwater (CA) System, Rev. 37
: CNS-1605.VI-00-0001, Instrument Air (VI) System, Rev. 21 
===Drawings===
: CN-1499-CA.07-00, Instrument Detail Motor Driven Auxiliary Feedwater Pump to Steam Generator Flow Control, Rev. 11CN-1554-1.6, Flow Diagram of Chemical and Volume Control System (NV), Rev. 20
: CN-1499-ND.01.04-01, Catawba Nuclear Station Unit 1 Instrument Location auxiliary Building EL. 543'-0", Rev. 86
: CN-1499-ND.01-00, Instrument detail RHR Pump Minimum Flow Control, Rev. 8
: CN-1550-1.0, Symbols for Diagrams, Rev. 23
: CN-1550-1.1, Symbols for Diagrams, Rev. 14
: CN-1550-2.0, Symbols for Diagrams, Rev. 14
: CN-1553-1.1, Flow Diagram of Reactor Coolant System (NC), Rev. 21
: CN-1554-1.7, Flow Diagram of Chemical and Volume Control System (NV), Rev. 28
: CN-1554-1.8, Flow Diagram of Chemical & Volume Control System(NV), Rev.8, 06/19/07
: CN-1561-1.0, Flow Diagram of Residual Heat Removal System (ND), Rev. 30
: CN-1561-1.1, Flow Diagram of Residual Heat Removal System (ND), Rev. 24
: CN-1562-1.2, Flow Diagram of Safety Injection System (NI), Rev. 30
: CN-1562-1.3, Flow Diagram of Safety Injection System (NI), Rev. 16
: CN-1565-2.2, Liquid Radwaste System Flow Diagram, Rev.36, 01/30/08
: CN-1571-1.0, Flow Diagram of Refueling Water System (FW), Rev. 29
: CN-1573-2.2, Flow Diagram of Component Cooling System (KC), Rev. 7
: CN-1574-1.0, Flow Diagram of Nuclear Service Water System (RN), Rev. 52
: CN-1574-1.2, Flow Diagram of Nuclear Service Water System (RN), Rev. 49
: CN-1601-2.1, Flow Diagram of Drinking Water System (YD), Rev. 27
: CN-1601-2.2, Flow Diagram of Drinking Water System (YD), Rev. 6
: CN-1601-2.3, Flow Diagram of Drinking Water System (YD), Rev. 22
: CN-1601-2.4, Flow Diagram of Drinking Water System (YD), Rev. 9
: CN-1703-03.24, One Line Diagram, 600V Non Essential Load Center #1SLXG Standby Shutdown facility, Rev. 3
: CN-1703-08.01, Three Line Diagram, 250/125VDC Aux. Power System (ETM) Standby Shutdown, Rev. 13
: CN-1703-08.02, One Line Diagram, 250/125VDC and 120VAC Aux. Power System/ETM and ETL, Rev. 15
: CN-1938-01.01, Computer Cable Routing Outdoor Area General Plan, Rev. 64
: CN-1938-06.01, Computer Cable Routing Outdoor Area Sections & Detail, Rev. 9
: CN-2499-CA.07-00, Instrument Detail Motor Driven Auxiliary Feedwater Pump to Steam Generator Flow Control, Rev. 9
: CN-2554-1.6, Flow Diagram of Chemical and Volume Control System (NV), Rev. 15
: CN-2554-1.7, Flow Diagram of Chemical and Volume Control System (NV), Rev. 19
: CN-2561-1.0, Flow Diagram of Residual Heat Removal System (ND), Rev. 33
: CN-2561-1.1, Flow Diagram of Residual Heat Removal System (ND), Rev. 24
: CN-2562-1.2, Flow Diagram of Safety Injection System (NI), Rev. 30
: CN-2562-1.3, Flow Diagram of Safety Injection System (NI), Rev. 24 
: Attachment
: CN-2571-1.0, Flow Diagram of Refueling Water System (FW), Rev. 23
: CN-2573-1.0, Flow Diagram of Component Cooling (KC) System, Rev. 24
: CN-2573-1.1, Flow Diagram of Component Cooling (KC) System, Rev. 15
: CN-2573-1.2, Flow Diagram of Component Cooling (KC) System, Rev. 13
: CN-2573-1.3, Flow Diagram of Component Cooling (KC) System, Rev. 11
: CN-2573-1.4, Flow Diagram of Component Cooling (KC) System, Rev. 12
: CN-2573-1.5, Flow Diagram of Component Cooling (KC) System, Rev. 11
: CN-2573-1.7, Flow Diagram of Component Cooling (KC) System, Rev. 12
: CN-2573-2.0, Flow Diagram of Component Cooling (KC) System, Rev. 7
: CN-2573-2.1, Flow Diagram of Component Cooling (KC) System, Rev. 7
: CN-2573-2.2, Flow Diagram of Component Cooling (KC) System, Rev. 4
: CN-2573-2.3, Flow Diagram of Component Cooling (KC) System, Rev. 6
: CN-2592-1.1, Flow Diagram of Auxiliary Feedwater System (CA), Rev. 30
: CN-2605-1.12, Flow Diagram of Instrument Air System (VI), Rev. 14
: CN-2609-1.0, Flow Diagram of Diesel Generator Engine Cooling Water System (KD), Rev. 16
: CNEE-0111-02.39, Elementary Diagram 600V Non Essential Load Center No. 1SLXG, compt 4A, 4B incoming bkr., P.T. & Synch. Ckts., Rev. 6
: CNEE-0111-02.43, Elementary Diagram 600V Non-Essential Load Center No. 1SLXG, Cpt.5B Diesel Generator "S", Fdr. Bkr., Rev. 7
: CNEE-0111-02.44, Elementary Diagram 600V Non-Essential Load Center No. 1SLXG, Cpt.5B Diesel Generator "S", Fdr. Bkr., Rev. 6
: CNEE-01114-00.10, Elementary Diagram Diesel Generator No. 1A Load sequencer (Part 8) committed & accelerated sequence circuits, Rev.10
: CNEE-0115-01.06, Elementary diagram 4160V switchgear 2ETA Unit 6 Component Cooling Water PMP MTR. 1A1, Rev. 8
: CNEE-0115-01.13, Elementary Diagram Nuclear Service Water System (RN) RN Pump B Discharge Isolation Valve 1RN038B, Rev. 11
: CNEE-0115-01.20, Elementary Diagram 4160V Switchgear 1ETA Breaker Failure, Mode Selector and Degraded Bus Voltage Circuit, Rev. 17
: CNEE-0115-01.34, Elementary Diagram 4160V Switchgear Unit #14 Nuclear Service Water Pump Motor 1B (1PMTR0156), Rev. 8
: CNEE-0115-01.40, Elementary Diagram 4160V Switchgear 1ETB Breaker Failure, Mode Selector and Degraded Bus Voltage Circuit, Rev. 16
: CNEE-0115-01.41, Elementary Diagram 4KV Switchgear 1ETA and 1ETB Breaker Internal Controls, Rev, 0
: CNEE-0141-01.04, Elementary Diagram Residual Heat Removal System (ND) ND pump 1A Miniflow Valve 1ND025A, Rev. 9
: CNEE-0141-01.10, Elementary Diagram Residual Heat Removal System (ND) ND pump 1B Miniflow Valve 1ND059B, Rev. 8
: CNEE-0144-01.01, Elementary Diagram SSF Diesel Controls and Instrumentation Sys (EQD) Misc. Diesel Controls and Alarms, Rev. 10
: CNEE-0150-01.10, Elementary Diagram Reactor Coolant System (NC) PZR Power Operated Relief Isolation Valve 1NC031B, Rev. 15
: CNEE-0151-01.31, Elementary Diagram Safety Injection System (NI) Containment Sump Line 1A Isolation Valve 1NI185A, Rev. 16
: CNEE-0157-03.13, Elementary Diagram Chemical & Volume Control Sys (NV) Solenoid Valves - Aux. Bldg. 1NV153A, 1NV172A, Rev. 13 
: Attachment
: CNEE-0160-02.30-02, WL Solenoid Valves 1WL847&1WL848, Rev.7, 01/16/85
: CNEE-0165-02.01, Elementary Diagram Hydrogen Mitigation System  (EHM) Ignitor Box Group "A" Alternate Source, Rev. 6
: CNEE-0165-02.02, Elementary Diagram Hydrogen Mitigation System  (EHM) Ignitor Box Group "B" Normal Source, Rev. 5
: CNEE-0215-01.03-01, Elementary diagram 4160V switchgear 2ETA Unit 3 NOR. INC. BKR. Transf. 2ATC, Rev. 5
: CNEE-0215-01.04-01, Elementary diagram 4160V switchgear 2ETA Unit 4
: STD-BY INC. BKR. Transf. SATA, Rev. 5
: CNEE-0215-01.06, Elementary diagram 4160V switchgear 2ETA Unit 6 Component Cooling Water PMP MTR. 2A1, Rev. 5
: CNEE-0215-01.13, Elementary diagram 4160V switchgear 2ETA Unit 13 Auxiliary Feedwater PMP MTR. 2A, Rev. 5
: CNEE-0215-01.18-01, Elementary Diagram 4160V Switchgear 2ETA Unit 18 Diesel Generator A2, Rev. 7
: CNEE-0215-01.20, Elementary Diagram 4160V Switchgear 2ETA Breaker Failure, Mode Selector and Degraded Bus Voltage Circuit, Rev. 10
: CNEE-0215-01.40, Elementary Diagram 4160V Switchgear 2ETB Breaker Failure, Mode Selector and Degraded Bus Voltage Circuit, Rev. 11
: CNEE-0215-01.41, Elementary Diagram 4160V Switchgear 2ETA and 2ETB Breaker Internal Controls, Rev. 0
: CNEE-0241-01.01, Elementary Diagram (ND) Residual Heat Removal System NC Loop 2 Supply to ND Train 2B Isolation Valve 2 ND001B, Rev. 13
: CNEE-0241-01.02, Elementary Diagram (ND) Residual Heat Removal System NC Loop 2 Supply to ND Train 2A Isolation Valve 2 ND002A, Rev. 14
: CNEE-0241-01.02-01, Elementary Diagram (ND) Residual Heat Removal System NC Loop 2 To ND Train 2A Isolation Valve 2
: ND002A, Rev. 8
: CNEE-0241-01.05, Elementary Diagram Residual Heat Removal System (ND) Supply to NV & 2A NI Pumps Valve 2
: ND028A, Rev. 7
: CNEE-0251-01.07, Elementary Diagram Residual Heat Removal System (ND) Supply to Safety Injection Sys (NI) Pumps 2B Valve 2
: NI136B, Rev. 10
: CNEE-0251-01.10, Elementary Diagram Safety Injection System (NI) Safety Injection Pump
: Miniflow Header to F.W. Valve 2NI147B, Rev. 7
: CNEE-0251-01.28, Elementary Diagram Safety Injection System (NI) Safety Injection Pump 2B Miniflow Line Isolation Valve 2NI144A, Rev. 4
: CNEE-0255-02.08, Elementary Diagram Refueling Water System (FW) ND Pump 2A Suction From FWST Valve 2
: FW027A, Rev. 10
: CNEE-0259-01.05, Elementary Diagram Containment Spray System (NS) Safety Injection Spray Isolation Valve 2NS043A, Rev. 6
: CNEE-0265-02.01, Elementary Diagram Hydrogen Mitigation System  (EHM) Ignitor Box Group "A" Normal Source, Rev. 7
: CNEE-0265-02.01-01, Elementary Diagram Hydrogen Mitigation System  (EHM) Ignitor Box Group "A" Alternate Source, Rev. 0
: CN-FSAR-FIG.08-001, Onsite Power System, Rev. 2
: CN-FSAR-FIG.08-021, Essential and Blackout Auxiliary Power Systems, Rev. 3
: CN-FSAR-FIG.08-19, Non-Class 1E AC Power Systems, Rev. 1
: CN-FSAR-FIG.6-136.01, ECCS Process Flow Diagram (NV) System, Rev. 0 
: Attachment
: CN-FSAR-FIG.6-136.02, ECCS Process Flow Diagram (NI) System, Rev. 0
: CN-FSAR-FIG.6-136.03, ECCS Process Flow Diagram (ND) System, Rev. 0
: CNM-1201.05-0282, Service water Pump Vendor Manual, 4/15/07
: CNM-1201.05-0555 001, Bingham Type "VTM" Pump Outline, Rev. DF
: CNM-1201.05-299, Nuclear Service Water Pump Certified Performance Test Curve, 5/8/79
: CNM-1205.00-0152 001, Valve Assembly, 8 Inch, 300 lb, Swing Check Valve, Rev. A
: CNM-1205.00-0207 001, Valve Assembly, 8 Inch, 600 lb, Gate Valve, 11/6/07
: CNM-1205-00-1652, Borg-Warner Seismic Report
: NSR 76130, 1/27/83
: CNM-1301.00-0237 006, Diesel Generator Instruction Manual, 12/8/98
: CNM-1301.00-0285, Seismic Qualification of Transamerica Delaval Diesel Generator Set, 10/25/82
: CNM-1301.02-0016, sh 1, DC schematics 700kW, 346/600V, 60 Hz Generator Control Panel, Rev. D4
: CNM-1301.02-0016, sh 2, A.C. Elementary Diagram, Rev. D5
: CNM-2201.05-0045, MDAFW Pump Curves Document, Rev.0, 05/14/80
: CNM-2201.05-0046, MDAFW Pump Curve Data Sheet, Rev.0, 05/14/80
: CNSF-1574-RN.01, Summary Flow Diagram, Nuclear Service Water System (RN), Rev. 6
: CNSF-2592-CA.01 Auxiliary Feedwater System Flow Diagram, Rev.5, 10/20/01
: CNTC-2573-KC.P001-01, Component Cooling Pumps Test Acceptance Criteria, Rev. 3
: CNTC-2573-KC.P002-01, KC Minimum Flow Lines, Rev. 2
: CNTC-2573-KC.P003-01, Component Cooling Pump Matching, Rev 0
: CNTC-2573-KC.S001-01, KC System Nonessential Header Flow Balance, Rev 0
: CNTC-2573-KC.S001-02, KC System Nonessential Header Flow Balance, Rev 3
: CNTC-2592-CA-V001-02 CA Test Acceptance Criteria, Rev.2, 03/26/93 Fisher Controls Drawing 54A7671, 657-ET Diaphragm Actuated Control Valve, Rev. DJ
: Corrective Action Documents (PIPs)
: C-03-01608, KD HX 2B ECT identified crack-like indications on 8 tubes, 3/13/03 C-03-04625, RN pump 1A UBOC flow long term declining trend, 8/20/03 C-04-01569, KD HX trends indicate potential degraded thermal performance, 3/31/04
: C-07-01547 CA Discharge MOV Failure to Re-Open after Testing, 05/01/2007 C-07-02443, 2A1 KC pump Outboard seal though has a 65 dpm leak, 05/16/07 C-07-02475, Problems were found with breaker 1ETB14 in the performance of its PM on 17MAY07 by crew 462, 05/17/2007 C-07-04311, 2A1 KC Pump Outboard Bearing Seal Leakage is at 60 DPM, 08/19/07
: C-07-06881, Observation of data for
: SWGDT-B pressure, 11/10/2007 C-07-06890, Discovered loose test tee on 2NVLT5761 (VCT level), 11/11/2007 C-07-07610, KC pump 2A1 outboard seal leak rate observed as approx 50 dpm, 12/22/07 C-08-00284, KC pump 2A1 outboard seal leakage is transitioning from drops per minute and approaching a small steady stream, 02/18/2008 C-08-00356, Leakage on 2A1 KC pump OB bearing has doubled since last night shift from about 88 dpm to about 190 dpm, 01/20/08 C-08-00487, 2B NV pump balance line pressure above rounds limit of 80 psig, 01/26/2008 C-08-00562, Acceptance Criteria not met, 2B ND pump Flow reading was high, 01/31/2008
: C-08-00880, 1ETA 18 Breaker Racked in Indication failed to operate when OPS racked breaker into cubicle, 02/13/2008 
: Attachment
: C-08-01465, Found during performance of Unit 1, Channel 1 18 month Channel Calibration, 03/11/2008 C-08-03245, Black substance/coating found in water box & on tube sheet of 1A KD HX, 5/26/08 C-08-03729, Perform cal under W/O
: 1723634, 06/12/2008
: C-08-04289, Unplanned entry into Tech Specs 3.7.8 for loss of 1B RN pump, 7/12/08 C-08-04393, RN flow to B YC chiller exceeded Hi alarm setpoint of 2000 gpm for approximately 2 minutes while throttling flow for 1B RN Head Curve Testing. Maximum flow per OAC was 2049.1 gpm, 07/18/2008 C-08-04399, Need to establish new baseline values for 1B RN pump, 7/19/08
: C-08-04691, Unit 1 VCT level indication is failing low, 08/04/2008 C-08-04906, Invlt5761 remote bellows isolation valves and bellows not labeled, 08/14/2008 C-08-05074, During performance of 2A ND IWP the 2A ND pump failed to start, 08/22/2008 C-08-05137, Breaker 1ETA 17 failed as found minimum DC voltage testing for a close operation, 08/26/2008 C-08-05147, 1A RN pump motor vibration is noticeable different than the other pumps when in service. NLO's need some INFO as to how much vibration is acceptable when performing rounds, 08/27/2008 C-08-05937, The Catawba Vital I&C Battery system is nearing the end of its available load growth capacity for future load additions, 10/06/2008 C-08-06130, Results of RN Lake Discharge Pipe UT Examination, 109/16/08 C-08-06329, 1B2 KC Pump Outboard seal leakage >30gpm upon pump start, 10/29/2008
: C-08-06396, Unit 2 dilutions and blended makeups show unexpected VCT level changes, 11/02/2008 C-08-06840, the capacity margin for loads growth seems to run out on the 125 Volts DC Vital I&C Battery System, 12/02/2008 C-09-01380, 2
: ETA-2 Breaker has cracked arc chutes, discovered during PM inspection, 03/08/2009
: C-09-01902, PT/2/A/4400/003 A KC Train 2A Performance Test was aborted due to no local flow indication on 2A1 KC Pump start with 2KC-6 "2A1 KC Pump Discharge" throttle 90% CLOSED per procedure, 3/24/2009 C-09-01903, Performed IWP on 2A1 KC pump after pump maintenance. The data taken is considered new baseline values, 03/22/2009 C-09-02770, Excessive noise on suction of 2A1 KC Pump. 04/22/2009 C-09-03784 , The inboard oil bubbler on the 2A1 KC Pump is discolored and appears milky, 06/21/2009 C-09-03910, KC pump 2B1 4160V feeder has damaged armor cable with exposed insulation, 06/27/2009 C-09-04390 WL Sump Pump Check Valve Inservice Test & PM Changes, 07/23/2009 C-09-04844, Determine if any operator actions are credited to control AFW flow during LOOP C-09-04847- SSF Load and Voltage Calculation, minor clarifications and elimination of inconsistencies -CDBI items 266, and 369, 8/14/09
: C-09-05020 1WL 848 AOV Failed to stroke closed during PT, 08/21/2009 C-09-05061- (SSF Load Center Breaker Coordination and Diesel Generator Relay Setting Calculation) - CDBI item 367, 8/25/09 C-09-05073- SSF D/G Testing Issue (Load Bank Testing and Power Factor value) -CDBI items 293, 294, 370, and 371, 8/25/09
: C-93-00703, KC Pump Calculation uses incorrect pump performance data, 08/25/1993 
: Attachment
: C-99-01308, Review of
: CNC-1223.24-00-0017,
: CNC-1223.24-00-0018 &
: CNC-1223.24-00-0019, 4/13/99 G-06-00154, Bussmann - Vendor Notification # 42021. Potential Part 21, 04/04/2006 G-07-00269, Three Site Response to
: IN 2007-05, Vertical Deep Draft Pump Shaft and Coupling Failures, 3/22/07 G-08-00341, IN08-2 Findings Identified during the Component Design Bases Inspections, 04/07/2008 M-06-3978 Valve Working Group to access
: IN 06-15, 12/07/06
: Completed Surveillances
: CNM 1301.02-0061, Full Load Test (Start up test data, SSF D/G, isochronous mode) , dated 3/24/80 IP/1/A/3170/003 A, 1EHM  -PFM Insp of Trn 1A H2 Ignitors, completed 11/10/08, 2/3/09, 3/4/09 and 4/28/09
: OST-1080, Auxiliary Feedwater Pump 1SX-SAB Full Flow completed September 5, 2007
: OST-1214, ESW System Operability Train A Quarterly Interval Modes 1-2-3-4-5-6-Defueled, completed 2/4/08, 2/5/08 2/25/08, 5/3/06, 5/19/07, 6/17/06, 7/27/06, 8/18/07, 9/9/06, 11/10/07, and 12/3/06
: OST-1215, ESW System Operability Train B Quarterly Interval Modes 1-2-3-4-5-6-Defueled, completed 1/17/07, 2/15/08, 3/11/08, 3/17/08, 4/17/07, 5/16/06, 7/6/07, 7/29/06, ; September 24, 2007; October 21, 2006; December 29, 2007 PT/0/A/4400/022 B, RN Pump IST Comprehensive, 7/18/08 PT/0/A/4400/022 B, RN Pump IST Quarterly, 1/24/09, 3/29/09 & 6/17/09 PT/1/A/4200/013 H, NI/NV Check Valve Test, completed 5/26/05 and 5/28/08
: PT/2/A/4200/013 F ND Valve Inservice Test, completed 10/6/07, 4/1/09 and 4/7/09
: PT/2/A/4200/013 G NI Valve Inservice Test, completed 4/1/06, 10/13/07 & 3/28/09 PT/2/A/4400/003 A, Component Cooling (KC) Train 2A Performance Test completed 2/15/08, 5/09/08, 8/01/08, 10/24/08, 1/16/09, 3/22/09, 5/09/09, 7/03/09 PT/2/A/4400/06 E, 2A KD Heat Exchanger Heat Capacity Trending 6/96 thru 3/09 
===Calculations===
: CNC 1223.23-00-0016, Component Cooling Surge Tank Verification, Rev. 6
: CNC-1205.19-00-0026
: GL 89-10 MOV Calculation, NV System: 1(2)NV865A, Rev.5
: CNC-1205.19-00-0153
: GL 89-10 MOV Calculation, NV System: 1(2)NV872A, Rev.1
: CNC-1205.41-00-0003, Auxiliary Feedwater Control Valve (1&2CA36, 40, 44, 48, 52, 56, 60, and 64) Accumulator Sizing, Rev. 0
: CNC-1205.41-00-0022, 1(2)
: CA 040, 044, 056, 060 and
: 1(2) 036, 048, 052, 064 Actuator Capability Evaluation Supporting Air Operated Valve (AOV) Program, Rev. 0
: CNC-1206.03-00-0001 Flood Levels for Structures Outside of the Reactor Building, Rev.4, 07/15/83
: CNC-1206.03-00-0109, Moderate Energy Spray Interaction Study, NSW Pump House, Rev. 4
: CNC-1210.04-00-0016, RHR Pump Miniflow Control Instrument Loop Accuracy Calculation NDPG5040/5050, Rev. 7
: CNC-1210.04-00-0096, ASME Section XI Inservice Pump Testing Instrumentation Accuracies, Rev. 5 
: Attachment
: CNC-1210.04-00-0131, VCT PCS Level Instrument Uncertainty, Rev. 0
: CNC-1223.04-00-0041 Parameters for Standby Makeup Pump Filter Outlet Valve, Rev.3, 12/28/93
: CNC-1223.04-00-0057 Operating Parameters for Valves 1(2)NV-865A
: CNC-1223.04-00-0107, VCT NPSH and Low-Low Level Auto-swap Setpoint, Rev. 0
: CNC-1223.11-00-0011, Design Parameters for RHR Valves 1 & 2
: ND 28A, Rev. 6
: CNC-1223.12-00-0036, Operating Parameters for Valves 1(2)
: NI-136B, Rev. 4
: CNC-1223.12-00-0061, NI and NV Check Valve Test Acceptance Criteria Verification, Rev. 3
: CNC-1223.13-00-0038, Unit 2 CA Condensate Source Vortexing Determination, Rev. 0
: CNC-1223.23-00-0011, Component Cooling System Test Acceptance Criteria, Rev. 17
: CNC-1223.23-00-0022, Component Cooling Heat Load and Flow Requirements, Rev. 11
: CNC-1223.23-00-0033, Supporting Calc for Response to NRC IE Bulletin 88-04, "Potential Safety Related Pump Loss" (Component Cooling Pumps), Rev. 13
: CNC-1223.23-00-0041. Evaluation of KC Pump Operation Near Maximum Tested Flow Rates, Rev. 1
: CNC-1223.24-00-0027, Flow Distribution Model of the RN System, Rev. 6
: CNC-1223.24-00-0052, Mechanical Design Inputs Calculation for NSM
: CN-50492/00, Nuclear Service Water (RN) System Chemical Treatment, Rev. 1
: CNC-1223.24-00-0058, Effect of RN System Flushing on Operability, Rev. 3
: CNC-1223.24-00-0060, RN Single Supply Header Operation Evaluation
: CNC-1223.24-00-0064, RN Single Supply Header Operation Design Basis Rev. [??]
: CNC-1223.42-00-0013 CA System Test Acceptance Criteria, Rev.0
: CNC-1223.42-00-0014 CA Pumps Mini-Flow Design Verification, Rev.1
: CNC-1223.42-00-0043 Unit 2 System Response to Chp.15 Accidents, Rev.5
: CNC-1223.59-01-0005, RN Flow and Fouling Acceptance Criteria on KD Heat Exchangers, Rev. 15
: CNC-1381.05-00-0011, 125VDC Vital Instrumentation and Control Power System Battery and Battery Charger Sizing Calculation, Rev. 8
: CNC-1381.05-00-0012, 4160 Volt Essential Auxiliary Power System Switchgear Relay Setting, Rev. 14
: CNC-1381.05-00-0014, 4160/600V Essential Load Center Transformer Feeder Relay Settings and Essential Load Center Breaker Coordination, Rev. 7
: CNC-1381.05-00-0017, Class 1E Diesel Protective Relaying and Sequence Undervoltage Relay Settings, Rev. 15
: CNC-1381.05-00-0108, Verification of Equipment Qualification Under Degraded Conditions, Rev. 1
: CNC-1381.05-00-0147, U1/2, 4.16KV Essential Auxiliary Power System (EPC) Diesel-Generator (D/G) LOCA/LOOP (Blackout) Loading Analysis, Rev. 10
: CNC-1381.05-00-0149, Vital DC System Voltage Analysis, Rev. 15
: CNC-1381.05-00-0198, Catawba Unit 1, 6.9kV, 4.16kV, 600V Auxiliary Power Systems Safety-Related Voltage Analysis, Rev. 9
: CNC-1381.05-00-0199, U2, 6.9kV, 4.16kV & 600V Auxiliary Power System Safety- Related Voltage Analysis, Rev. 14
: CNC-1381.05-00-0209, Catawba Unit 1 ETAP Power Station Auxiliary Power System Fault Study, Rev. 3
: CNC-1381.05-00-0210, Catawba Unit 2 ETAP Power Station Auxiliary Power System Fault Study, Rev. 5 
: Attachment
: CNC-1381.06-00-0035, 250/125 VDC SSF Aux Power System Battery and Battery Charger Sizing Calc, Rev. V5
: CNC-1381.06-00-0036, Standby Shutdown Facility Load Center Breaker Coordination and Diesel Generator Relay Setting Calculation, Rev. 07
: CNC-1381.06-00-0060, ETAP Power Station Loading/Voltage Calculation For Catawba SSF D/G And SSF Auxiliary Power System, Rev. 3
: CNM 1201.05-0272, Component Cooling Pump Seismic Rpt and Nzl Loads, Rev. 6
: CNM 1205.01-0064.001 Kerotest Instruction Manual, 02/03/87
: CNM 1205.19-0083.001 Rotork Vendor Manual, Rev. D6
: CNM-1205.00-1103 Operation and Maintenance Manual OMM2057, dtd. 03/26/1981
: CNM-1205.00-1498 Seismic Analysis of Motor Operator Globe Valves, Rev.B
: CNTC-2592-CA.V001-03, MPD 2A (Flow to SGs A or B) Test Acceptance Criteria Sheet, Rev. 4 
===Procedures===
: AP/0/A/5500/022, Loss of Instrument Air, Rev. 28
: AP/1/A/5500/002, Turbine Generator Trip, Rev. 29 AP/1/A/5500/003, Load Rejection, Rev. 35 AP/1/A/5500/009, Rapid Down Power, Rev. 25 AP/1/A/5500/011, Pressurizer Pressure Anomalies, Rev. 020 AP/1/A/5500/020, Loss of Nuclear Service Water, Rev. 36
: AP/1/A/5500/021, Loss of Component Cooling, Rev. 35 AP/2/A/5500/007, Loss of Normal Power, Rev. 54
: EDM 101, Engineering Calculations/Analyses, Rev. 14
: EDM 601, Engineering Change, Rev. 9 EP/1/A/5000/E-0, Reactor Trip or Safety Injection, Rev. 34
: EP/1/A/5000/E-0, Reactor Trip or Safety Injection, Rev. 35 EP/1/A/5000/ES-1.3, Transfer to Cold Leg Recirculation, Rev. 21 EP/1/A/5000/F-0, Critical Safety Function Status Trees - Containment, Rev. 7 IP/0/A/3816/010, Barton Model 580 and 581 DP Switch Calibration, Rev. 29 IP/0/A/3820/001B, Limitorque Operator Preventive Maintenance, Rev.006
: IP/0/A/4974/021, Procedure for Motor On-Line Testing, Rev. 7 IP/0/A/4974/032, Induction Motor Inspection and Testing, Rev. 11 IP/0/B/3817/013 E, Rosemount Model 3051S (Scalable) Series Transmitters Configuration/Calibration Check and Maintenance Procedure, Rev. 04 IP/1/A/3144/001 B, Calibration Procedure for ND Miniflow Control Pressure Switches, Rev. 20
: IP/1/B/3222/023 A, Volume Control Tank Level Cabinet 5, Loop 1NVLT5761 (LT-112) Calibration, Rev. 25 IP/1/B/3222/023 B, Volume Control Tank Level Cabinet 8, Loop 1NVLT5760 (LT-185) Calibration, Rev. 22 MP/0/A/7150/004 A, Component Cooling Pumps Rotating Element Assembly, Rev. 12 MP/0/A/7600/038, Borg Warner Bonnet Hung Swing Check Valve Low Pressure, Rev. 17 MP/0/A/7600/108, Anderson Greenwood Wafer Check Valve Maintenance, Rev. 017
: MP/0/A/7650/002, Lubrication Of Station Equipment, Rev. 25 MP/0/A/7650/056 D, Diesel Generator Jacket Water Heat Exchanger Corrective Maintenance, Rev. 15 MP/0/A/7650/166, HX/CH Tube Pull & Replacement, Rev. 4 
: Attachment
: OP/0/A/6400/006 C, Nuclear Service Water System, Rev. 266 OP/0/A/6400/006C, Nuclear Service Water System, Rev. 266 OP/1/A/6100/010G, Panel 1AD-6, A/8, PZR HI PRESS ALERT, Rev. 63 OP/1/A/6100/010J, Panel 1AD-9, D/8, FWST 2/4 LO LEVEL, Rev. 65
: OP/1/A/6100/010M, Panel 1AD-12, A/1(4), RN Pump A(B) Flow Hi/Lo, Rev. 40 OP/1/A/6350/005, Change Hold, Alternate AC Power Sources, Rev. 67 OP/1/A/6550/002, Diesel Generator Operation, Rev. 144 OP/1/B/6100/009S, Annunciator Response for Standby Shutdown Facility, Rev.022
: OP/1/B/6100/010F, Annunciator Response for Panel 1AD-5, Rev. 053 OP/1/B/6100/010Y, Annunciator Response for Radiation Monitoring Panel, Rev. 041 OP/2/A/6100/010M, Panel 2AD-12, A/1(4), RN Pump A(B) Flow Hi/Lo, Rev. 30 PT/0/A/4200/017 A, Standby Shutdown Facility Diesel Test, Rev. 20 PT/0/A/4200/019, Standby Shutdown Facility Diesel Generator Operating Parameters, Rev. 4 PT/0/A/4400/008 B, RN Flow Balance Train B, Rev. 48 PT/0/B/4450/015 A, Air Quality Periodic Test for the VI System, Rev. 011 PT/0/B/4450/015 B, Air Quality Periodic Test for the VI System, Rev. 0219 PT/1/A/4200/001 I, As Found Containment Isolation Valve Leak Rate Test, Rev. 011
: PT/1/A/4200/0013 B, NV Valve Inservice Test, Rev. 034 PT/1/A/4200/013 H, NI/NV Check Valve Test, Rev. 33 PT/1/A/4700/020, WL Sump Pump Check Valve Inservice Test, Rev. 010 PT/1/A/4700/020, WL Sump Pump Check Valve Inservice Test, Rev. 011 PT/2/A/4200/013 F ND Valve Inservice Test, Rev. 25
: PT/2/A/4200/013 G NI Valve Inservice Test, Rev. 34 PT/2/A/4200/079, CA Flow Control Valve Air Accumulator Leakage Test, Rev. 8 PT/2/A/4250/003 A, MDAFW Pump 2A Performance Test, Rev. 045 PT/2/A/4250/003E, CA System Discharge Control Valve Throttling Procedure, Rev. 35 PT/2/A/4400/003 A, Component Cooling (KC) Train 2A Performance Test, Rev. 42
: PT/2/A/4400/006 E, KD Heat Exchanger 2A Heat Capacity Test, Rev. 23 RP/0/A/5000/001, Classification of Emergency, Rev. 020 
===Work Orders===
: 01792301, 2KC PU A1 : Flush O/B Mechanical Seal With Warm Demin Water, 1/18/2008
: 01797121, 2KC PU A1 : Flush O/B Mechanical Seal With Warm Demin Water, 3/17/2008
: 01874676, 2KC PU A1- Sample And Change Oil For I/B And O/B Pump Bearings, 6/30/2009
: 01736300, MP/0/A/7650/002 Verify proper oil level after pump, dtd 7/3/2007
: 0184219301, MP/0/A/7650/002 Verify proper oil level after pump, dtd 5/4/2009
: 01703659, 2VC
: PS 5530 PFM 2CRA-AHU-1 Inst. Calibration, 03/28/2007
: 01723634, CE101196, 1NV
: LT 5761: Replace Channel 1 VCT Level Transmitter, 05/22/2008
: 01735109, 1VC
: PS 5530 PFM 1CRA-AHU-1 Inst. Calibration, 07/31/2007
: 01755847, 1RN
: TE 7690: PERF CHAR Test on Spare Thermocouple Conductor, 08/26/2009
: 01804482, 2VC
: PS 5530 PFM 2CRA-AHU-1 Inst. Calibration, 06/15/2009
: 01816772, 2ERN PFM Routine Test 27 relays on 2B D/G, 0714/2008
: 98031306, 2EPC Refurbish Breaker 2ETA-3 (4160 VAC), 09/14/19989
: 96098715, 2EPC Insp/Maintenance on 2ETA Bus, 04/02/1997
: 98120782, 1EBA Performance Test, 05/03/1999
: 98122654, 2EPC SWAP BKR 2ETA-18 W/Refurbed BK, 08/29/2000 
: Attachment
: 98465171, 2EPC PM Breaker 2ETA-13 (2A CA PMP MTR), 02/12/2002
: 98534825, 2EPC BK
: ETA-4 PM Breaker, 10/21/2002
: 98518305, 2EPC PM 2ETA-3, 5, 10, 16 Protective Relays, 03/06/2003
: 98650069, 1EBA Performance Test, 05/09/2004
: 98642861, 2EPC BK
: ETA-3 PM Breaker, 08/30/2004
: 98656044, 1EPC BK
: ETA-6 PM Breaker, 05/12/2005
: 98766017, Perform CAL 1CRA-AHU-1 Instrumentation, 02/13/2006
: 98717611, 2EPC BK
: ETA-18 PM Breaker, 04/02/2006
: 01122453, 1ND
: PG 5050: CAL ND PMP 1B Miniflow Cont, 08/17/2006
: 01129815, 1ND
: PG 5040: CAL ND PMP 1A Miniflow Cont, 08/29/2006
: 01703997, 2KC PU A1- Perform PM Inspection, 10/16/2006
: 01700517, 1NV PFM CH CAL VCT LVL (1NVP5760), 10/18/2006
: 01702387 01, MOV test for 1NI136B, 11/20/06
: 01720633 01, 1KD HX end bell leak repair, 11/25/06
: 01707027, 1NV PFM CH CAL VCT LVL (1NVP5761), 04/10/2007
: 01716114, 1EPL BA EBA PFM Vital Battery Service Test, 05/08/2007
: 01707459, 1EPC - Routine Test 1ETB SWGR Relays, 05/16/2007
: 01744979, 1EPC BK
: ETB-14: PM Breaker, 05/17/2007
: 01748590, 2EPC BK
: ETA-13 PM Breaker, 05/31/2007
: 01735398, 1EPL-VITAL Batt Charger ECA CAP Test, 08/28/2007
: 01746283, 2KC PU A1- Perform PM Inspection, 10/14/2007
: 01082722, CE73053; Revise Span for 1NDPG5040, 01/16/2008
: 01758391, 1ND
: PG 5050: CAL ND PMP 1B Miniflow Cont, 01/29/2008
: 01796627, 1EPC BK
: ETA-18; I/R Breaker T.O.C>, 02/13/2008
: 01813512, 2ERN PFM Routine Test 27 Relays on 2B D/G, 06/17/2008
: 01814963, 2ERN PFM Routine Test 27 Relays on 2A D/G, 06/26/2008
: 01778230, 1NV PFM CH CAL VCT LVL (1NVP5760), 07/14/2008
: 01818313, 2ERN PFM Routine test 27 relays on 2A D/G, 07/21/2008
: 01793755, 2KC PU A1- Perform PM Inspection, 7/30/2008
: 01816612, 2EPC BK
: ETA-4 PM Breaker, 08/18/2008
: 01818466, 1EPC BK
: ETA-17: PM Breaker, 08/26/2008
: 0173861901 1NV:Replace Optical Isolators(1NV872A), dtd. 9/17/08
: 01806734, 1NV PFM CH CAL VCT LVL (1NVP5761), 10/01/2008
: 01837102, Refurbish Spare 5-HK Breaker, 12/10/2008
: 0183713101, Performed PM per procedure/TAM, dtd. 12/14/2008
: 01810401, 1EPL BA EBA PFM Vital Battery Service Test, 12/16/2008
: 01850544, 1EPC BK
: ETA-6 Swap Breaker W/refurbed BKR, 02/12/2009
: 01822040, 2EPC BK
: ETA-2: PM Breaker, 03/08/2009
: 01820231, 2EPC PM 2ETA-3,5,10,16 Protective Relays, 03/14/2009
: 01787948, 2KC PU A1: Replace Outboard Mechanical Seal, 3/18/2009
: 01820443, 2KC PU A1- Perform Outage PM Inspection, 3/21/2009
: 01820496, 2KC PU A1- Refurbish Pump, 3/21/2009
: 01857255, Performed AFW PT Acceptance criteria met, dtd. 5/1/2009
: 01801315, 1ND
: PG 5050: CAL ND PMP 1B Miniflow Cont, 05/21/2009
: 01801316, 1ND
: PG 5040: CAL ND PMP 1A Miniflow Cont, 06/03/2009
: 01824860, 1NV P5761 Unit 1VCT Level CH 1 Slowly Failing Low, 08/12/2008
: Attachment 
===Miscellaneous===
: 1B RN Oil Analysis Data, 12/08, 3/09 & 6/09 1EBA Vendor Test, 06/27/1995
: ASME Section XI Code, 1983 Ed. Attachment 1 to 50.59 Evaluation for NSM
: CN-50492/0, Nuclear Service Water (RN) System Chemical Treatment, 10/19/00 Catawba Check Valve Condition Monitoring Program, 4/29/02
: CNM 1201.05-0273, Component Cooling Water Pumps, Rev D24
: CNM 1205.04-0086.001 I/B Nuclear Safety Related Plug Valves, Rev. 4
: CNM 1225,00-0068.001 Check Valve Vendor Manual
: CNM 2201.05-0031, Component Cooling Pump Certified H/Q Curves, Rev D2
: CNM-1210.00-0039.001, NSSS Precautions, Limitations, and Setpoints, Rev. D49
: CNM-1210.04-0276.001, Technical Manual for Model 580A-0 Differential Pressure Indicating Switch, 05/12/1983
: CNM-1210.04-0421.001, Product Manual for Rosemount Model 3051 Smart Pressure Transmitter Family, Rev. 0
: CNM-1210.04-0503.001, Rosemount Transmitters models 3051C, 3051T, 1151, and 2088 Manual, 01/25/1999
: CNM-1312-02-00-0054, ABB Installation/ Maintenance Instructions, Metal-Clad Medium-Voltage Power Switchgear, Type 5HK, 7.5HK, Rev. 6
: CNM-1312-02-00-0054, Key Interlock General Information Dimensions, Rev. 0
: CNM-1318.00-00004-001, Environmental Qualification Test Report/ Analysis Summary, 02/12/1986
: CNM-1356.01-0026.001, Install and Maintenance Manual for Vital Battery & Racks Unit 1, rev.
: D4
: CNS 1201.05-00-0006, Component Cooling Pumps, Rev 6
: CNS-1201.05-00-0001, Nuclear Service Water Pumps, 12/1/08
: CNS-1205.00-00-0005, Nuclear Safety Related Stainless Steel Valves, 12/29/82
: CNS-1465.00-00-0007, Plant Seismic Design, Rev. 2 Coupling Correspondence with Johnston Pump, B Felker w/ E. Huffaker, 10/22/03
: DC-3.12, Cable Ampacity Design Criteria, Rev. 2 Duke Power Generic Letter 96-05 Program Plan, Rev. 2
: EQMM-1393.01-A01-01 Environmental Qualification Maintenance Manual, 4/16/03
: EQMM-1393.01-G04-00, Environmental Qualification Maintenance Manual for Westinghouse Motors, Rev. 7
: EQMM-1393.01-N01-01, Environmental Qualification Maintenance Manual, Rev. 5 IEEE Std 450-2002, Recommended Practice for Maintenance, testing, and Replacement of Vented Lead -Acid Batteries for Stationary applications, 04/03/2003 IEEE Std 485-1997, Recommended Practice for Sizing Lead-Acid Batteries for Stationary applications, 03/20/1997 Letter (draft), T. Hamilton (CN) to Sulzer Pump, Subject: Potential Part 21 Reportability, File:
: CN-211.55, 9/25/08 Letter, B. Woodford (Cooper Energy Services) to G. Yezefski (CN), Subject: Diesel Generator Engine Heat Rejection, 11/1/2000 Letter, R. Kayler (CN) to B. Woodford (Cooper Energy Services), Subject: Diesel Generator Jacket Water Cooler Fouling Factor Design Data, 5/13/04 
: Attachment Maintenance Rule Scoping Documentation NRC Bulletin No. 88-04: Potential Safety-Related Pump Loss NSD229 Evaluation and Reporting of Potential Defects Per 10CFR21, Rev.4, 10/20/08 Nuclear Policy Manual, Directive 229, Evaluation and Reporting of Potential Defects and Noncompliance per 10 CFR Part 21, Rev. 4 OMb Code, 1971 Ed.
: OP-CN-PSS-KC, Component Cooling Lesson Plan, Rev. 57
: OP-CN-PSS-KC-004, JPM for Alternate Cooling to 1A NV Pump, Rev. 8 Operator Awareness Review, Recent Changes to E-0 (Reactor Trip or Safety Injection) Unit 1 Rev. 35, Unit 2 Rev. 33 PO #
: CN-35364, Rebuild of Nuclear Service Water Pumps, 10/28/98 RE - 3.02 Electrical Engineering Criteria, Relaying - 600 and 480 VAC Auxiliary Systems - Equipment Protection Settings, Rev. 4 RN Pump 1B dp trend data 8/15/07 thru 6/17/09
: SDQA-00055-CNS, KD Heat Capacity, Rev. 5 Service Water System Program Manual, Rev. 9
: Vendor Manual: SSF Diesel Generator Instruction Manual Volume II, Duke File Number
: CNM 1301.02-0058.002, Rev. D10 Vendor Manual¸
: CNM 1308.01-0021.001,Inductrol Regulator Type Airs Single Phase Motor-Operated Manual or Automatic Control" Rev. D1, Vendor Manual¸
: CNM 1341.01-0001.001, "Tayco Engineering Hydrogen Assembly and Qualification Report" , Rev. D3
: Modifications Engineering Change
: 0000095020,
: CD 101434 - Power Train A Hydrogen Igniters From the SSF, Rev. 000 Minor Modification CD201435, Power from Unit 2A Hydrogen Igniters from SSF Diesel, dated 7/25/08
: CN-11372, Rev 00, Modification to revise the run-out set-points for the Component Cooling (KC) single pump and two pump operation
: CNCE-72672, Perform various MOV operator replacements and MOV testing to support
: GL 89-10 & 96-05, 9/9/04
: CD-100548, Replace MOV operators due to Rotork Part 21 issue (Ref. PIP C-04-2668), 11/7/06 EC101196, Replace 1NVLT5761 Channel 1 VCT Level Transmitter, 06/14/2008 EC200902, Replace 2NVLT5761 Level Transmitter, 12/14/2006
: System Health Reports
: CA - Auxiliary Feedwater Health Report 2008Q4 EPC - 4KV Unit 1/2 Health Report 2008Q3 EPL - 125VDC Vital I&C Power System Health Report 2009Q1 KC - Component Cooling Health Report 2008Q4
: ND - Residual Heat Removal Health Report 2008Q2 NI  - Safety Injection Health Report 2008Q3 NV - Chemical And Volume Control Health Report 2009Q1 RN - Nuclear Service Water Health Report 2009Q1 
: Attachment Valves and Heat Exchangers Executive Report 4
th Quarter/2008 Valves and Heat Exchangers Summary Report 2009 2
nd Quarter
: PIPs Initiated as a Result of Inspection
: C-09-04390, Valves 1/2WL-847/848 do not have objective evidence of periodic testing or maintenance for the life of the plant,
: C-09-04685, Field Walkdown Identified Damaged Capillary Armor for 1/2NVLT5760, 8/5/09 C-09-04813, Unit 1 Aux Feed Control Valve Accumulator Drawings not in NAS, 8/12/09
: C-09-04822, Incorrect stroke times for 1/2ND28A shown on flow diagrams
: CN-1561-1.0 &
: CN-2561-1.0, 8/12/09 C-09-04847, SSF Load and Voltage Calculation, minor clarifications and elimination of inconsistencies, 8/13/09 C-09-04969, A reference cited in
: CNC-1223.23-00-0041, Evaluation of KC Pump Operation Near Maximum Tested Flow Rates, Rev 1, does not point to the correct source document, 8/19/09 C-09-05007, Documentation discrepancy in 4160VAC switchgear EQ documentation, 8/20/09 C-09-05020, Valve 1WL848 failed to stroke closed, 8/21/09 C-09-05061, SSF Load Center Breaker Coordination and Diesel Generator Relay Setting Calculation concerns, 8/24/09 C-09-05073, SSF D/G Testing Issue Load Bank Testing and Power Factor value concerns, 8/25/09 C-09-05125, Revision of the horsepower (hp) used in
: CNM 1201.05-0272, Rev 6, to determine the shaft torque, 8/26/09
: C-09-05132,
: CNC-1223.24-00-0058 &
: CNC-1223.24-00-0060 need to demonstrate that pump curves used are bounded by IST limits, 8/27/09 C-09-05137, Convective heat losses from KD HX not included in determination of maximum heat load for fouling and tube plugging limits, 8/27/09 C-09-05142, In reference to pump failure documented in C-08-4289, Part 21 notification to pump vendor been may not have been provided, 8/27/09 C-09-05144, Clarification of the Basis for Aux Bldg Room Temperatures, 8/27/09
: C-09-05146, Apparent non-conservative (low) weight used in seismic analysis for 1(2) ND028A & 1(2) NI0136B, 8/27/09 C-09-05147,
: CNC-1223.24-00-0027 does not confirm that throttle valve positions calculated match up with actual field settings, 8/27/09
}}
}}

Revision as of 02:26, 20 September 2018

Catawba Nuclear Station - Component Design Bases Inspection - NRC Inspection Report 0500413-09-006 and 0500414-09-006
ML091250239
Person / Time
Site: Catawba  Duke Energy icon.png
Issue date: 05/04/2009
From: Desai B B
NRC/RGN-II/DRS/EB1
To: Morris J R
Duke Energy Carolinas, Duke Power Co
References
IR-09-006
Download: ML091250239 (6)
v • d • e


Text

May 4, 2009

Mr. J. Site Vice President Duke Power Company, LLC d/b/a Duke Energy Carolinas, LLC Catawba Nuclear Station 4800 Concord Road York, SC 29745-9635

SUBJECT: CATAWBA NUCLEAR STATION - COMPONENT DESIGN BASES INSPECTION - NRC INSPECTION REPORT 05000413/2009006 AND 05000414/2009006

Dear Mr. Morris:

The purpose of this letter is to notify you that the U.S. Nuclear Regulatory Commission (NRC)

Region II staff will conduct a component design bases inspection at your Catawba Nuclear Station during the period of July 6 - August 28, 2009. The inspection team will be led by W. R. Lewis, a Senior Reactor Inspector from the NRC's Region II Office. This inspection will be conducted in accordance with the baseline inspection procedure, Procedure 71111.21, Component Design Bases Inspection, issued August 19, 2008.

The inspection will evaluate the capability of risk significant / low margin components to function as designed and to support proper system operation. The inspection will also include a review of selected operator actions, operating experience, and modifications.

During a telephone conversation on April 27, 2009, Mr. Lewis confirmed, with Mr. Mark Sawicki of your staff, arrangements for an information gathering site visit and the three-week onsite inspection. The schedule is as follows:

  • Information gathering visit: Week of July 6 - 10, 2009.
  • Onsite weeks: Weeks of July 27 - 31, August 10 - 14, and August 24 - 28, 2009.

The purpose of the information gathering visit is to meet with members of your staff to identify risk-significant components and operator actions. Information and documentation needed to support the inspection will also be identified and gathered or requested. Mr. G. MacDonald, a Region II Senior Reactor Analyst, may accompany Mr. Lewis during the information gathering visit to review probabilistic ri sk assessment data and identify risk significant components which will be examined duri ng the inspection.

DEC 2 The enclosure lists documents that will be needed prior to the information gathering visi t. Please contact Mr. Lewis with any questions prior to preparing materials listed in the enclosure and provide the referenced information to the Region II office by June 19, 2009. The inspectors will try to minimize your administrative burden by specifically identifying only those documents required for the inspection preparation.

During the information gathering visit, the team leader will also discuss the following inspection support administrative details: office space; site, plant and information system access; required resources; and information exchange protocol.

In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter and its enclosure will be available electronically for public ins pection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of NRC's document system (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Thank you for your cooperation in this matter. If you have any questions regarding the information requested or the inspection, please contact Mr. Lewis at (404) 562-4541 or me at (404) 562-4519.

Sincerely,/RA/

Binoy B. Desai, Chief Engineering Branch 1 Division of Reactor Safety Docket Nos.: 50-413, 50-414,72-045 License Nos.: NPF-35, NPF-52

Enclosure:

Information Request For Catawba Nuclear Station Component Design Bases Inspection

cc w/encl: (See page 3)

_________________________ X SUNSI REVIEW COMPLETE OFFICE RII:DRS RII:DRS SIGNATURE NAME W.R. Lewis B. Desai DATE 04/ /2009 04/ /2009 E-MAIL COPY? YES NO YES NO YES NO YES NO YES NO YES NO YES NO DEC 3 cc w/encls:

Randy D. Hart Regulatory Compliance Manager Duke Power Company, LLC d/b/a Duke Energy Carolinas, LLC Electronic Mail Distribution

R. L. Gill, Jr.

Manager Nuclear Regulatory Issues & Industry Affairs Duke Power Company, LLC d/b/a Duke Energy Carolinas, LLC Electronic Mail Distribution

Dhiaa M. Jamil Group Executive and Chief Nuclear Officer Duke Energy Carolinas, LLC Electronic Mail Distribution

Kathryn B. Nolan Senior Counsel Duke Energy Corporation 526 South Church Street-EC07H Charlotte, NC 28202

Lisa F. Vaughn Associate General Counsel Duke Energy Corporation 526 South Church Street-EC07H Charlotte, NC 28202

Senior Resident Inspector Duke Energy Corporation Catawba Nuclear Station U.S. NRC 4830 Concord Road York, SC 29745

David A. Repka Winston Strawn LLP Electronic Mail Distribution

North Carolina MPA-1 Suite 600 P.O. Box 29513 Raleigh, NC 27525-0513

Susan E. Jenkins Director, Division of Waste Management Bureau of Land and Waste Management S.C. Department of Health and Environmental Control Electronic Mail Distribution

R. Mike Gandy Division of Radioactive Waste Mgmt.

S.C. Department of Health and Environmental Control Electronic Mail Distribution

Beverly O. Hall Chief, Radiation Protection Section Department of Environmental Health N.C. Department of Environmental Commerce & Natural Resources Electronic Mail Distribution

Elizabeth McMahon Assistant Attorney General S.C. Attorney General's Office P.O. Box 11549 Columbia, SC 29211

Vanessa Quinn Federal Emergency Management Agency 500 C Street, SW Room 840 Washington, DC 20472

Steve Weatherman, Operations Analyst North Carolina Electric Membership Corporation Electronic Mail Distribution

County Manager of York County York County Courthouse York, SC 29745

Piedmont Municipal Power Agency Electronic Mail Distribution

Peggy Force Assistant Attorney General State of North Carolina P.O. Box 629 Raleigh, NC 27602 DEC 4 Letter to J. from Binoy B. Desai, dated May 4, 2009.

SUBJECT: CATAWBA NUCLEAR STATION - COMPONENT DESIGN BASES INSPECTION - NRC INSPECTION REPORT 05000413/2009006 AND 05000414/2009006 Distribution w/encl

Jon Thompson, NRR PUBLIC RidsNrrPMCatawba Resource Institute of Nuclear Power Operations (INPO)

Enclosure INFORMATION REQUEST FOR CATAWBA NUCLEAR STATION COMPONENT DESIGN BASES INSPECTION

Please provide the information electronically in ".pdf" files, Excel, or other searchable format on CDROM. The CDROM should be indexed and possibly hyperlinked to facilitate ease of use. Information in "lists" should contain enough information to be easily understood to someone who has knowledge of pressurized water reactor technology.

1. From your most recent probabilistic safety analysis (PSA) excluding external events and fires, please provide: a. Two risk rankings of components from your site-specific probabilistic safety analysis (PSA) - one sorted by Risk Achievement Worth (RAW), and the other sorted by Birnbaum Importance. b. A list of the top 500 cutsets.

2. From your most recent probabilistic safety analysis (PSA) including external events and fires, please provide: a. Two risk rankings of components from your site-specific probabilistic safety analysis (PSA) - one sorted by Risk Achievement Worth (RAW), and the other sorted by Birnbaum Importance. b. A list of the top 500 cutsets.

3. Risk ranking of operator actions from your site specific PSA sorted by RAW. Provide copies of your human reliability worksheets for these items. 4. Any pre-existing evaluation or list of components and calculations with low design margins (i.e., pumps closest to the design limit for flow or pressure, diesel generator close to design required output, heat exchangers close to rated design heat removal, MOV risk-

margin rankings, etc.). 5. A list of station applicability evaluations/reviews performed and document ed in the station corrective action program in the past two years for industry events, critical equipment failures, and safety related equipment vul nerabilities (as communicated by NRC generic communications, industry communications, 10 CFR Part 21 notifications, etc.). 6. A list of operability evaluations completed within the last two years, sorted by associated component or system. 7. A list of common-cause failures of components that have occurred at Catawba Nuclear Station and have been identified within the last five years. 8. A list of equipment currently planned for upgrade/improvement by the site (e.g. "ONE" List), including a description of the reason(s) why each component (i.e. not a programmatic or system level concern) is on that list and summaries (if available) of your plans to address those reasons. 9. A list of equipment currently in RIS 05-020 (formerly GL 91-18) status, or in MR (a)(1) status. 10. Contact information for a person to discuss PRA information prior to the information gathering trip: name, title, phone number, and e-mail address.

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