IR 05000498/2007007: Difference between revisions

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{{Adams
{{Adams
| number = ML072350335
| number = ML080450543
| issue date = 08/22/2007
| issue date = 02/13/2008
| title = South Texas Project Electric Generating Station, Units 1 and 2 - Information Request for an NRC Biennial Baseline Component Design Bases Inspection 05000498-07-007; and 05000499-07-007
| title = IR 05000498-07-007 and 05000499-07-007; on 09/24/2007 - 01/22/2008; South Texas Project, Units 1 and 2; NRC Inspection Procedure 71111.21, Component Design Bases Inspection.
| author name = Jones W B
| author name = Bywater R L
| author affiliation = NRC/RGN-IV/DRS
| author affiliation = NRC/RGN-IV/DRS/EB1
| addressee name = Sheppard J J
| addressee name = Sheppard J J
| addressee affiliation = South Texas Project Nuclear Operating Co
| addressee affiliation = South Texas Project Nuclear Operating Co
Line 11: Line 11:
| contact person =  
| contact person =  
| document report number = IR-07-007
| document report number = IR-07-007
| document type = Letter
| document type = Inspection Report, Letter
| page count = 10
| page count = 35
}}
}}


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=Text=
=Text=
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[[Issue date::August 22, 2007]]
[[Issue date::February 13, 2008]]


James J. Sheppard, President and Chief Executive Officer STP Nuclear Operating Company P.O. Box 289 Wadsworth, TX 77483
James J. Sheppard, President and Chief Executive Officer STP Nuclear Operating Company P.O. Box 289 Wadsworth, TX 77483


SUBJECT: SOUTH TEXAS PROJECT ELECTRIC GENERATING STATION, UNITS 1 and 2 - INFORMATION REQUEST FOR AN NRC BIENNIAL BASELINE COMPONENT DESIGN BASES INSPECTION 05000498/2007007; AND 05000499/2007007
SUBJECT: SOUTH TEXAS PROJECT ELECTRIC GENERATING STATION, UNITS 1 AND 2 - NRC COMPONENT DESIGN BASES INSPECTION REPORT 05000498/2007007 AND 05000499/2007007


==Dear Mr. Sheppard:==
==Dear Mr. Sheppard:==
On September 24, 2007, the NRC will begin a biennial baseline Component Design Bases Inspection at the South Texas Project Electric Generating Station, Units 1 and 2. A team of six inspectors plus a team leader, will perform this 3-Week inspection. This inspection will be performed in accordance with revised NRC Baseline Inspection Procedure 71111.21 and replaces the biennial Safety System Design and Performance Capability inspection.
On November 26, 2007, the U.S. Nuclear Regulatory Commission (NRC)
completed onsite portions of a component design bases inspection at your South Texas Project Electric Generating Station, Units 1 and 2. The preliminary results were discussed with you and members of your staff on November 26, 2007. After additional in-office inspection, a telephonic exit was conducted on January 22, 2008. The enclosed report documents our inspection findings.


The Component Design Bases Inspection focuses on components that have high risk and low design margins. The components to be reviewed during this baseline inspection will mainly be identified during an information gathering visit and during the subsequent in-office preparation week. In addition, a number of risk significant operator actions and operating experience issues associated with the component samples will also be selected for review.
This 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 team reviewed selected procedures and records, observed activities, and interviewed cognizant plant personnel.


The inspection will include 4-weeks onsite, including the information gathering site visit and 3-weeks of onsite inspection. The inspection will consist of six NRC inspectors, of which five will focus on engineering and one on operations. The current inspection schedule is as follows:
The report documents six NRC identified findings, each involving a violation of NRC requirements. All of the findings were evaluated under the risk significance determination process as having very low safety significance (Green). Because of their very low safety significance and because they are entered into your corrective action program, these violations are being treated as noncited violations, consistent with Section VI.A of the Enforcement Policy. If you contest the subject or significance of any of these noncited violations, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001, with copies to the Regional Administrator, U.S. Nuclear Regulatory Commission, Region IV, 611 Ryan Plaza Drive, Suite 400, Arlington, Texas 76011; the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC Resident Inspector at the South Texas Project Electric Generating Station, Units 1 and 2.
Onsite information gathering visit: Week of September 4, 2007 Onsite weeks: September 24, 2007, October 15 and October 22, 2007 The purpose of the information gathering visit is to meet with members of your staff to identify potential risk-significant components and operator actions. The lead inspector will also request a tour of the plant with a member of your probabilistic risk analyst or operations staff. Additional information and documentation needed to support the inspection will also be identified. A Region IV senior reactor analyst will accompany the lead inspector during the information gathering visit, to review probabilistic risk assessment data and assist in identifying risk significant components, which will be reviewed during the inspection.


STP Nuclear Operating Company 2Experience with previous baseline design inspections of similar depth and length has shown that these types of inspections are extremely resource intensive, both for the NRC inspectors and the licensee staff. In order to minimize the inspection impact on the site and to ensure a productive inspection, we have enclosed a request for information needed for the inspection. The request has been divided into three groups. The first group lists information necessary for the information gathering visit and for general preparation. This information should be available to the Regional office by no later than August 29, 2007. Insofar as possible, this information should be provided electronically to the lead inspector. Since the inspection will be concentrated on high risk/low margin components, calculations associated with your list of high risk components should be available for the inspectors to review during the information gathering visit to assist in our selection of components based on available design margin.
STP Nuclear Operating Company - 2 -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 inspection 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).


The second group of documents being requested is those items that the team will need access to while onsite and after components are selected. The third group lists information necessary to aid the inspection team in tracking issues identified as a result of the inspection. It is requested that this information be provided to the lead inspector as the information is generated during the inspection. It is important that all of these documents are up to date and completed in order to minimize the number of additional documents requested during the preparation and/or the onsite portions of the inspection. In order to facilitate the inspection, we request that a contact individual be assigned to each inspector to ensure information requests, questions, and concerns are addressed in a timely manner.
Sincerely,/RA/
 
Russell L. Bywater, Chief Engineering Branch 1 Division of Reactor Safety
 
Dockets: 50-498; 50-499 Licenses: NPF-76; NPF-80
 
===Enclosures:===
NRC Inspection Report 05000498/2007007
 
and 05000499/2007007
 
===w/Attachment:===
Supplemental Information
 
cc w/enclosures:
 
E. D. Halpin Site Vice President STP Nuclear Operating Company South Texas Project Electric Generating Station P.O. Box 289 Wadsworth, TX 77483
 
Ken Coates Plant General Manager STP Nuclear Operating Company South Texas Project Electric Generating Station P.O. Box 289 Wadsworth, TX 77483
 
S. M. Head, Manager, Licensing STP Nuclear Operating Company P.O. Box 289, Mail Code: N5014 Wadsworth, TX 77483
 
C. T. Bowman STP Nuclear Operating Company - 3 -General Manager, Oversight STP Nuclear Operating Company P.O. Box 289 Wadsworth, TX 77483
 
Marilyn Kistler Sr. Staff Specialist, Licensing STP Nuclear Operating Company P.O. Box 289, Mail Code 5014 Wadsworth, TX 77483
 
C. M. Canady City of Austin Electric Utility Department 721 Barton Springs Road
 
Austin, TX 78704
 
J. J. Nesrsta/R. K. Temple/
E. Alercon/Kevin Pollo City Public Service Board P.O. Box 1771
 
San Antonio, TX 78296
 
Jon C. Wood Cox Smith Matthews 112 E. Pecan, Suite 1800
 
San Antonio, TX 78205
 
A. H. Gutterman, Esq.
 
Morgan, Lewis & Bockius 1111 Pennsylvania Avenue NW Washington, DC 20004
 
Director, Division of Compliance & Inspection Bureau of Radiation Control Texas Department of State Health Services 1100 West 49th Street
 
Austin, TX 78756
 
Brian Almon Public Utility Commission William B. Travis Building P.O. Box 13326 1701 North Congress Avenue
 
Austin, TX 78701-3326
 
STP Nuclear Operating Company - 4 -Environmental and Natural Resources Policy Director P.O. Box 12428
 
Austin, TX 78711-3189
 
Judge, Matagorda County Matagorda County Courthouse
 
1700 Seventh Street Bay City, TX 77414
 
Anthony Jones, Chief Inspector Texas Department of Licensing and Regulation Boiler Program P.O. Box 12157
 
Austin, TX 78711
 
Susan M. Jablonski Office of Permitting, Remediation and Registration Texas Commission on Environmental Quality MC-122, P.O. Box 13087
 
Austin, TX 78711-3087
 
Ted Enos 4200 South Hulen
 
Suite 422 Fort Worth, TX 76109
 
Steve Winn/Christine Jacobs/
Eddy Daniels/Marty Ryan NRC Energy, Inc.
 
211 Carnegie Center Princeton, NJ 08540
 
INPO Records Center 700 Galleria Parkway Atlanta, GA 30339-3064
 
Lisa R. Hammond, Chief Technological Hazards Branch National Preparedness Division FEMA Region VI 800 N. Loop 288
 
Denton, TX 76209
 
STP Nuclear Operating Company - 5 -
 
STP Nuclear Operating Company - 6 -Electronic distribution by RIV: Regional Administrator (EEC)
DRP Director (DDC)
DRS Director (RJC1)
DRS Deputy Director (ACC)
Senior Resident Inspector (JLD5)
Branch Chief, DRP/A (CEJ1)
Senior Project Engineer, DRP/A (TRF)
Team Leader, DRP/TSS (CJP)
RITS Coordinator (MSH3)
DRS STA (DAP)
D. Pelton, OEDO RIV Coordinator (DLP1)
 
ROPreports STP Site Secretary (HLW1)
 
SUNSI Review Completed: ___Y__ ADAMS: Yes No Initials: ___WSifre___ Publicly Available Non-Publicly Available Sensitive Non-Sensitive
 
SRI:EB1 RI:PBC RI:EB1 OE:OB C:EB1 C:PBA C:EB1 WSifre/lm b MChambers SMakor GApger RLBywate r CEJohnson RLBywater /RA/ /RA/ /RA/ /RA/ /RA/ /RA/ /RA/ 1/2/08 12/18/08 1/2/08 1/2/08 2/13/8 2/12/08 2/13/08 STP Nuclear Operating Company - 7 -OFFICIAL RECORD COPY T=Telephone E=E-mail F=Fax
 
Enclosure - 1 -
 
=SUMMARY OF FINDINGS=
IR 05000498/2007007 and 05000499/2007007; September 24, 2007 through January 22, 2008;
 
South Texas Project Electric Generating Station, Units 1 and 2; NRC Inspection Procedure 71111.21, "Component Design Bases Inspection."
 
The report covered a 4-week period of onsite inspection and additional in-office inspection performed by six region-based inspectors and two contractors. The inspection identified six Green noncited violations. The significance of most findings is indicated by its color (Green, White, Yellow, Red) using Inspection Manual Chapter 0609, "Significance Determination Process."  Findings for which the significance determination process does not apply may be Green or be assigned a severity level after NRC management review. The NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, "Reactor Oversight Process," Revision 3, dated July 2000.
 
===A. NRC - Identified Findings===
 
===Cornerstone: Mitigating Systems===
: '''Green.'''
The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," having very low safety significance for the failure to specify in a design calculation allowable relay setpoint tolerances. Specifically, the licensee failed to specify and verify in the relay setpoint calculations the relay setpoint tolerances used in the calibration test procedures. The issue was documented in the corrective action program as Condition Record 07-15443.
 
The finding was determined to be more than minor because it is associated with the Mitigating Systems cornerstone attribute of "Design Control."  It impacts the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events and prevent undesirable consequences. The failure to verify the effects of relay setpoint tolerances on relay coordination time intervals could have resulted in a loss-of-relay coordination and could lead to either a loss of power to safety-related components or lead to a potential for compromising other equipment on a single fault that the relay was designed to isolate. Using Manual Chapter 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations," Phase 1 screening, the finding screened as having very low safety significance (Green) because the condition did not represent a loss of safety function of a system or a train.  (Section 1R21.b.1)
: '''Green.'''
The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," having very low safety significance for the failure to include all potential loads in the standby diesel generator fuel oil sizing calculation. Specifically, the licensee did not account for increased standby diesel 
 
- 2 -generator fuel oil usage resulting from the addition of manual electrical loads during the 7-day mission run time. The licensee entered this finding into their corrective action program as Condition Record 07-15592. The licensee subsequently demonstrated that the spent fuel pool cooling pumps would be the only additional manual loads actually used during the 7 days of operation in the bounding design basis scenario and that there were additional conservative assumptions in the sizing calculation to demonstrate sufficient margin.
 
The finding was determined to be more than minor because it is associated with the Mitigating Systems cornerstone attribute of "Design Control."  It impacts the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events and prevent undesirable consequences. Using Inspection Manual Chapter 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Pow er Situations," Phase 1 screening, the finding screened as having very low safety significance (Green)because it was a design or qualification deficiency confirmed not to result in loss of operability or functionality.  (Section 1R21.b.2)
: '''Green.'''
The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criteria III, "Design Control," of very low safety significance for the failure to translate design basis information into specifications and procedures. Specifically, a non-conservative system pressure was used as an input to an engineering design calculation for the auxiliary feedwater outside containment isolation valves. This finding has been entered into the licensee's corrective action program as Condition Record 07-15455.
 
The finding was determined to be more than minor because it is associated with the Mitigating Systems cornerstone attribute of "Design Control."  It impacts the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events and prevent undesirable consequences. Using Inspection Manual Chapter 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Pow er Situations," Phase 1 screening, the finding screened as having very low safety significance (Green)because it did not represent a loss safety function of a system or a train.  (Section 1R21.b.3)
: '''Green.'''
The team identified a noncited violation of Technical Specification Surveillance Requirement 4.8.1.1.2.E.11, having very low safety significance for the licensee's failure to adequately perform the technical specification surveillance requirement. Specifically, the licensee failed to verify the loading times of the essential chillers in order to verify the automatic load sequence timer was operable. This issue was entered into the licensee's corrective action program as Condition Records 07-14903 and 07-14959.
 
The finding was determined to be more than minor because it is associated with the Mitigating Systems cornerstone attribute of "Design Control."  It impacts the 
 
- 3 -cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events and prevent undesirable consequences. Using Inspection Manual Chapter 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Pow er Situations," Phase 1 screening, the finding screened as having very low safety significance (Green)because it did not represent a loss of safety function of a system or a train.
 
(Section 1R21.b.4)
: '''Green.'''
The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XI, "Test Control," having very low safety significance for the licensee's failure to implement a test program to assure that all installed safety-related molded case circuit breakers will perform satisfactorily in service.
 
Specifically, the licensee had not adequately exercised or subjected to periodic testing all of the 125V dc molded case circuit breakers since initial plant operation. The licensee entered the finding into their corrective action program as Condition Record 07-15817.
 
The finding was determined to be more than minor because it is associated with the Mitigating Systems cornerstone attribute of "Equipment Performance."  It impacts the cornerstone objective of ensuring the availability, reliability, capability of systems that respond to initiating events and prevent undesirable consequences. Using Inspection Manual Chapter 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations," Phase 1 screening, the finding screened as having very low safety significance (Green) because it did not result in a loss of safety function of a system or a train.  (Section 1R21.b.5)
: '''Green.'''
The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criteria III, "Design Control," of very low safety significance for the failure to adequately translate design basis information into specifications and procedures. Specifically, measurement instrument uncertainties were not included in the determination of minimum allowed high head safety injection pump and low head safety injection pump developed head values used during periodic technical specification surveillance testing. The licensee entered the finding into their corrective action program as Condition Record 07-15752.
 
The finding was determined to be more than minor because it is associated with the Mitigating Systems cornerstone attribute of "Design Control."  It impacts the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events and prevent undesirable consequences. Using Inspection Manual Chapter 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Pow er Situations," Phase 1 screening, the finding screened as having very low safety significance (Green)because it did not result in a loss of safety function of a system or a train.
 
(Section 1R21.b.6)
 
- 4 -B. Licensee-Identified Findings
 
None.
 
- 5 -U.S. NUCLEAR REGULATORY COMMISSION REGION IV
 
Dockets:  05000498, 05000499
 
Licenses:
NPF-76, NPF-80
 
Report:  05000498/2007007; 05000499/2007007
 
Licensee:
STP Nuclear Operating Company
 
Facility:
South Texas Project Electric Generating Station, Units 1 and 2
 
Location  FM 521 - 8 miles west of Wadsworth Wadsworth, Texas 77483
 
Dates:  September 24, 2007 through January 22, 2008k
 
Inspectors:
W. Sifre, Senior Reactor Inspector, Engineering Branch 1 M. Chambers, Resident Inspector, Branch C B. Henderson, Reactor Inspector, Engineering Branch 1 S. Makor, Reactor Inspector, Engineering Branch 1 S. Rutenkroger, Reactor Inspector, Engineering Branch 1 G. Apger, Operations Engineer, Operations Branch
 
Contractors:
H. Anderson, Mechanical Contractor J. Chiloyan, Electrical Contractor
 
Approved By:
Russell L. Bywater, Chief Engineering Branch 1 Division of Reactor Safety
 
=REPORT DETAILS=
 
==REACTOR SAFETY==
Inspection of component design bases verifies the initial design and subsequent modifications and provides monitoring of the capability of the selected components and operator actions to perform their design bases functions. As plants age, their design bases may be difficult to determine and an important design feature may be altered or disabled during a modification. The plant risk assessment model assumes the capability of safety systems and components to perform their intended safety function successfully. This inspectable area verifies aspects of the Initiating Events, Mitigating Systems and Barrier Integrity cornerstones for which there are no indicators to measure performance.
 
In addition to performing the baseline inspection, the team reviewed actions taken by the licensee in response to previously identified significant issues associated with engineering performance.
{{a|1R21}}
==1R21 Component Design Bases Inspection==
{{IP sample|IP=IP 71111.21}}
The team selected risk-significant components and operator actions for review using information contained in the licensee
=s probabilistic risk assessment. In general, this included components and operator actions that had a risk achievement worth factor greater than two or Birnbaum importance value greater than 1E-6.
 
====a. Inspection Scope====
To verify that the selected components would function as required, the team reviewed design basis assumptions, calculations, and procedures. In some instances, the team performed independent calculations to verify the appropriateness of the licensee engineers' conclusions. The team also verified that the condition of the components was consistent with the design bases and that the tested capabilities met the required criteria.
 
The team reviewed maintenance work records, corrective action documents, and industry operating experience information to verify that licensee personnel considered degraded conditions and their impact on the components. For the review of operator actions, the team observed operators during simulator scenarios associated with the selected components, as well as observing simulated actions in the plant.
 
The team performed a margin assessment and detailed review of the selected risk-significant components to verify that the design bases have been correctly implemented and maintained. This design margin assessment considered original design issues, margin reductions due to modification, 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 included items such as failed 
- 7 -performance test results; significant corrective actions; repeated maintenance; 10 CFR 50.65(a)1 status; operable, but degraded, conditions; NRC re sident 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.
 
The components selected for review were:
* 345/138,13.8kV Standby Transformer ST002A
* 13.8kV/4/16 Auxiliary Engineered Safety Feature Transformer E2B
* 4.16kV Engineered Safety Feature BUS E2B
* Standby Diesel Generator 22
* 4.16kV/480 V Load Center Transformer E2B
* 480V Load Center E2B2
* 125V DC Battery and Charger Train B
* Electrical Auxiliary Building HVAC
* 10KVA Inverter E1V 2201
* Auxiliary Feedwater Motor Driven Pump 23
* Auxiliary Feedwater Turbine Driven Pump Td 14
* Auxiliary Feedwater Valve 0019
* Steam Generator Power Operated Relief Valve 2A
* High Pressure Safety Injection Pump 2A
* Low Pressure Safety Injection Pump 2A
* Refueling Water Storage Tank
* Essential Cooling Water Pump 2A
* Essential Chilled Water Pump 2A The selected operator actions were:
* Opening electrical auxiliary building doors and start of smoke purge on loss of ventilation to switchgear rooms.
* Isolation of a faulted steam generator.
* Initiation of reactor coolant system depressurization.
* Tripping of the reactor coolant pumps.
* Diagnosis of a steam generator tube rupture to start appropriate procedures.
* Starting auxiliary feedwater if engineered safety features actuation system fails during a control room fire.
 
The operating experience issues were: 
- 8 -
* NRC Information Notice (IN) 2006-06, "Loss-of-Offsite Power and Station Blackout Are More Probable During Summer Period."
* NRC IN 2007-09, "Equipment Operability Under Degraded Voltage Conditions."
* NRC Generic Letter 89-10, "Consideration of the Results of NRC-Sponsored Tests of Motor-Operated Valves."
* NRC IN 2006-18, "Significant Loss of Safety-Related Electrical Power at Forsmark, Unit 1, in Sweden."
* NRC IN 2007-27, "Recurring Events Involving Emergency Diesel Generator Operability."
 
====b. Findings====
b.1. Failure to Specify Setpoint Calibration Limits in Relay Setpoint Calculations
 
=====Introduction.=====
The team identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," for the failure to specify in a design calculation the allowable relay setpoint tolerances stated in the licensee's relay setpoint calibration test procedures. Under postulated electrical fault or overload conditions, the lack of adequate relay coordinating time intervals between relay operating characteristics would lead to spurious tripping and to either a loss of power to safety-related components or lead to a potential for compromising other equipment on a single fault that the relay was designed to isolate.
 
=====Description.=====
During the review of licensee's completed protective relay trip setpoint calibration test procedures, relay setting records and relay setting calculations to verify whether the applied relay settings were consistent with the designed basis calculations, the team noted that the acceptance criteria for the allowable values of relay setpoints stated in calibration test Procedures PM EM-2-03000814, WAN 274021 and relay setting sheets were neither specified nor verified in the design basis relay setting Calculation EC-5029, "4.16kV Switchgear Relay Setting."  Following discovery, the licensee performed a preliminary evaluation for affected components using the worst-case scenario of relay setpoint tolerances stated on the relay setting records and concluded that the affected components would still perform their required safety functions in the event of an electrical fault. The issue was documented in licensee's corrective action program as Condition Record 07-15443.
 
=====Analysis.=====
The licensee's failure to specify relay setpoint tolerances and verify the effects on coordination margin in relay setpoint calculations for relays used on 4.16kV emergency safety feature switchgears was a performance deficiency. The finding was determined to be more than minor because it is associated with the Mitigating Systems cornerstone attribute of "Design Control."  It impacts the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating 
- 9 -events and prevent undesirable consequences. The failure to verify the effects of relay setpoint tolerances on relay coordination time intervals could have resulted in a loss-of-relay coordination and could lead to either a loss of power to safety-related components or lead to a potential for compromising other equipment on a single fault that the relay was designed to isolate. Using Manual Chapter 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations", Phase 1 screening, the finding screened as having very low safety significance (Green) because the condition had not resulted in a loss of safety function of a system or a train. This finding was reviewed for crosscutting aspects and none were identified.
 
=====Enforcement.=====
Part 50 of Title 10 of the Code of Federal Regulations, Appendix B, Criterion III, "Design Control," requires, in part, that design control measures provide for verifying or checking the adequacy of design, such as by the performance of design reviews, by the use of alternate or simplified calculational methods, or by performance of a suitable testing program.
 
Contrary to the above, the licensee's design control measures failed to either specify the relay setpoint tolerances or verify the adequacy of the design for safety-related 4160V electrical distribution system to ensure that the trip settings of the protective relays were adequate to ensure selective tripping in the event of a fault. Specifically, the team identified that the licensee failed to specify and verify in the relay setpoint calculations the relay setpoint tolerances used in the calibration test procedures. Because this violation was of very low safety significance and has been entered into the licensee's corrective action program as Condition Record 07-15443, it is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy: NCV 05000498; 499/2007007-01, Failure to Specify Setpoint Calibration Limits in Relay Setpoint Calculations.
 
b.2. Failure to Consider Manual Loads for Fuel Oil Storage Tank Sizing Calculation
 
=====Introduction.=====
The team identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," for the licensee's failure to account for manual electrical loads in determining fuel oil usage during the standby diesel generators' seven day mission time for the fuel oil storage tank sizing calculation.
 
=====Description.=====
The Final Safety Analysis Report, Revi sion 0, stated that the fuel oil storage tanks were sized to have sufficient capacity to provide for continuous operation of the diesel generators for 7 days at their continuous rating, (i.e., 5935 kW). The licensee revised the Updated Final Safety Analysis Report (UFSAR) on December 9, 1992, to replace the loading at the standby diesel generator continuous rating with the "engineered safety features load requirements."  However, the documented review contained in Unreviewed Safety Question Evaluation 91-0031 and Calculation MC-6256, "Sizing of SDG FOST," Revision 0, both discussed including all the non-engineered safety features loads listed in UFSAR, Table 8.3-3, as part of the fuel and storage tank sizing requirement.
 
- 10 -In particular, the Unreviewed Safety Question Evaluation 91-0031 stated, "This [including all the listed non-engineered safety features loads] is in accordance with the ANSI N195 Standard which states, 'If the design includes provision for an operator to supply power to equipment other than the minimum required for the plant condition, such additional load(s) shall be included in the calculation of required fuel oil storage capacity."  Regulatory Guide 1.137, "Fuel Oil Systems for Standby Diesel Generators,"
Revision 1, dated October 1979, refers to the requirements described in ANSI N195-1976, "Fuel Oil Systems for Standby Diesel-Generators," to be a method acceptable to the NRC staff for complying with the Commission's regulations regarding diesel fuel oil systems for standby diesel generators and assurance of adequate diesel fuel oil quality. The safety evaluation report originally prepared for South Texas Project Electric Generating Station used ANSI N195 as the standard to evaluate the acceptability of the fuel oil storage tank design and sizing.
 
Since the UFSAR, as revised, did not discuss the additional manual loads, which must be considered in order to evaluate the fuel oil storage tank sizing, Calculation MC-6256, Revision 0, was ultimately re vised in Revision 3, dated October 3, 1996, to remove consideration of all manual loads. Therefore, beginning with that revision the design basis non-conservatively removed consideration of expected actual plant operations with respect to manual loads during the bounding design basis accident analysis.
 
The team interviewed engineering and operations personnel in order to determine what equipment from UFSAR, Table 8.3-3, would be supplied power other than the minimum required for the plant condition. These interviews revealed a range of possible equipment, which could be utilized since the operations philosophy would be to exceed the minimum required for the plant condition in order to place the plant in as safe a condition as possible. The upper range of potential manually loaded equipment would have resulted in exceeding the minimum technical specification fuel oil volume requirement of 60,500 gallons during the 7-day mission time of the standby diesel generators during the worst-case design basis accident considered. However, in further discussions, licensee personnel balanced the operations philosophy with the 7-day fuel oil requirements considered as part of the design basis event and concluded the spent fuel pool cooling pumps would be the only additional manual loads utilized during the bounding scenario. The fuel oil storage tank sizing calculation included additional conservative assumptions regarding expected pump operation during design basis accident scenarios. For example, the calculation assumed auxiliary feedwater, high head safety injection, and containment spray pumps would be run continuously for 48 hours following a large break loss of coolant accident. Therefore, the licensee recalculated the fuel oil storage tank sizing using more realistic assumptions with respect to load profile and determined sufficient fuel oil margin does exist with all design basis conditions considered. This issue was entered into the licensee's corrective action program as
 
Condition Record 07-15592.
 
=====Analysis.=====
The team determined that the failure to account for manual electrical loads in determining fuel oil usage during the standby diesel generators' 7-day mission time for the fuel oil storage tank sizing calculation was a performance deficiency. The finding 
- 11 -was determined to be more than minor because it is associated with the Mitigating Systems cornerstone attribute of "Design Control."  It impacts the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events and prevent undesirable consequences. Not accounting for the additional manual loads increases the likelihood that the required inventory of fuel oil for a 7-day mission time would not be available.
 
Using Inspection Manual Chapter 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations," Phase 1 screening, the finding screened as having very low safety significance (Green) because it was a design or qualification deficiency confirmed not to result in loss of operability or functionality.
 
This finding was reviewed for crosscutting aspects and none were identified.
 
=====Enforcement.=====
Part 50 of Title 10 of the Code of Federal Regulations, Appendix B, Criterion III, "Design Control," requires, in part, that measures shall be established to assure that applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures, and instructions.
 
Contrary to the above, the licensee had not correctly translated design basis information into the standby diesel generator fuel oil tank sizing analysis. Specifically, the licensee failed to translate the loading and usage associated with additional manual loads, reasonably expected to be utilized during the bounding design basis accident, into Calculation MC-6256, Revision 4. Because this violation was of very low safety significance and has been entered into the licensee's corrective action program as Condition Record 07-15592, it is being treated as a noncited violati on consistent with Section VI.A of the NRC Enforcement Policy: NCV 05000498; 499/2007007-02, Manual Loads not Considered for Fuel Oil Storage Tank Sizing Calculation.
 
b.3. Failure to Use Correct Design Inputs in Determination of the Weak Link for the Auxiliary  Feedwater System Outside Containment Isolation Motor Operated Valves.
 
=====Introduction.=====
The team identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criteria III, "Design Control," for the failure to translate design basis information into specifications and procedures. Specifically, the team identified that a non-conservative system pressure was used as an input to the engineering design calculation for the auxiliary feedwater outside containment isolation valves (MC-6204 Document Change Notice MC-145 issued 7/31/1992, Revise Motor-Operator Valve Thrust and Torque Calculation for AF-19, AF-48, AF-65, and AF-85).
 
=====Description.=====
The team identified that the pressure loading calculation in the motor-operated valve weak link analysis for the auxiliary feedwater outside containment isolation valves used a system pressure of 1250 psig. This value was based on the steam generator power-operated relief valves in the main steam systems being normally set at 1225 psig for normal operation and an additional 25 psig was added to the nominal steam generator power-operated relief valve set point to allow for any set point uncertainty.
 
- 12 -This did not take into account accident conditions that result in the backpressure from the main steam system being greater than 1250 psig.
 
In response to the team's questions that the pressure could be greater than 1250 psig, the licensee issued Condition Record 07-15455-4, "Discussion Paper; Re-perform Weak Link Calculation at 1324 psid and 200°F," received October 24, 2007; and Condition Record 07-15455, "Discussion Paper; Weak Link Discussion of Motor-Operated Valves During Normal and Accident Operation," received October 15, 2007. The licensee determined that an increase in steam generator pressure greater than normal operating pressure would occur during certain design bases accident conditions. The appropriate input to the calculation was determined to be a steam generator pressure of 1324 psig, which allows for a 1 percent margin for setting tolerance and 2 percent for pressure drop in the piping connecting the safety valves to the steam generator from the lowest safety valve set point of 1285 psig. With the revised 1324 psig value and the original assumed valve temperature of 200°F the new weak link calculation resulted in two of the eight auxiliary feedwater outside containment isolation valves (one in each unit) having a torque switch setting that exceeded the weak link calculated set point in the close direction. The weak link for these valves is the valve seat. Valve thrust plus system pressure exceeding the valve seat strength could result in thrusting the valve disc into the seat and failure of the valve.
 
The licensee subsequently provided the following justification for the operability of the valves using the 1324 psid accident pressure.  "From a review of all accidents that result in an increase in Steam Generator pressures also result in the starting of the auxiliary feedwater pumps. The auxiliary feedwater system water supply has a design temperature range of 32°F to 120°F. Single failure criteria states that one of the auxiliary feedwater pumps may not start, however it is NOT creditable for a pump to not start and to have sufficient back leakage to raise the temperature of the outside containment isolation valve to 200°F at the same time. Therefore, the maximum abnormal temperature is 170°F."  The licensee determined that the weak link calculation at 1324 psid and 170°F results in adequate margin between current thrust settings of all eight auxiliary feedwater outside containment isolation valves and the calculated weak link stresses of the valve seats to assure operability under accident conditions.
 
=====Analysis.=====
The failure to use a conservative design input in the engineering analysis was a performance deficiency. The finding was determined to be more than minor because it is associated with the Mitigating Systems cornerstone attribute of "Design Control."  It impacts the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events and prevent undesirable consequences. Using Inspection Manual Chapter 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations," Phase 1 screening, the finding screened as having very low safety significance (Green) because it did not represent a loss of safety function of a system or a train. This finding was reviewed for crosscutting aspects and none were identified.
 
- 13 -Enforcement. Part 50 of Title 10 of the Code of Federal Regulations, Appendix B, Criterion III, "Design Control," states, in part, that measures shall be established to assure that design basis are correctly translated into specifications and procedures. Contrary to the above, in Calculation MC-145, the licensee did not use a conservative pressure input necessary to prevent damage to auxiliary feedwater outside containment isolation valves during a design basis event. Because this violation was of very low safety significance and has been entered into the licensee's corrective action program as Condition Record 07-15455), it is being treated as a noncited viola tion consistent with Section VI.A of the NRC Enforcement Policy: NCV 05000498; 499/2007007-03, Failure to Use Correct Design Inputs in Determination of the Weak Link for the Auxiliary Feedwater System Outside Containment Isolation Motor Operated Valves.


The lead inspector for this inspection is Mr. Wayne Sifre. We understand that our licensing engineer contact for this inspection is Mr. Jim Morris of your organization. If there are any questions about the inspection or the requested materials, please contact the lead inspector at (817) 860-8193 or via e-mail at wcs@nrc.gov.
b.4. Surveillance Procedure Lacked Check for Timing of Chiller Loading on the Bus


In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter will be made available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of NRC's Agencywide Documents Access and Management System (ADAMS). ADAMS is accessible from the Electronic Reading Room page of the NRC's public Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
=====Introduction.=====
The team identified a Green noncite d violation of Technical Specification Surveillance Requirement 4.8.1.1.2.E.11, for the licensee's failure to adequately perform the technical specification surveillance requirement. Specifically, the licensee failed to verify the loading times of the essential chillers in order to verify the automatic load sequence timer was operable.


Sincerely,/RA/  
=====Description.=====
Technical Specification Surveillance Requirement 4.8.1.1.2.E.11 requires "Verifying that the automatic load sequence timer is OPERABLE with the first sequenced load verified to be loaded between 1.0 second and 1.6 seconds, and all other load blocks within +/- 10% of its design interval."  The team requested to review the strip chart data recorded from the surveillance tests that demonstrated this surveillance requirement had been performed satisfactorily. Licensee personnel recognized, however, that the actual loading times referenced in the surveillance requirement had not been included in the measurements. Procedure 0PSP02-SF-001A, "ESF Diesel Sequencer Timing Test Train A," Revision 11 (Trains B and C similar), only tests the time that the sequence timer demands breaker closure and does not measure and/or record the actual load times.
 
The licensee entered Technical Specification 4.0.3 for all three trains of standby diesel generators for both units, allowing 24 hours to fully perform the surveillances successfully. By reviewing the strip chart recorder data for the last loss-of-offsite power and loss-of-offsite power with emergency safety features actuation testing of the standby diesel generators, the licensee verified Technical Specification Surveillance Requirement 4.8.1.1.2.E.11 was successfully met for Standby Diesel Generators 11, 21, and 22. In the case of Standby Diesel Generators 13 and 23, the recorded information had a time resolution loss due to switching of recording speeds during the test. The licensee performed a risk evaluation to delay the complete performance of the surveillance test until the next scheduled time (the next outage scheduled for Spring 2008). The team reviewed this assessment and agreed with its conclusions since the data that was available fully supported the equipment being able to perform its safety function.
 
- 14 -However, for Standby Diesel Generator 12, a review of the strip chart data revealed that Essential Chiller 12B had loaded on the bus at 168 seconds versus the design interval of 270 seconds. This condition had not been discovered in prior surveillance testing because Procedure 0PSP02-SF-001A did not contain instructions to verify the timing of relays outside of the sequence timer itself.
 
The licensee declared Standby Diesel Generator 12 inoperable at 09:45 on October 5, 2007, entering Technical Specification 3.8.1.1, Actions B and D.
 
The cause of the timing discrepancy was isolated to a 35 second blocking circuit external to the chiller that would not prevent the chiller from performing its design safety function. As such, the safety functions of the sequence timer, the standby diesel generator, and Essential Chiller 12B were not adversely affected by the condition, nor would those safety functions be impacted by starting/loading times of the essential chillers between 65 and 270 seconds. The licensee revised the design documents referencing the loading time of the essential chillers to be between 65 and 270 seconds. Once completed, the surveillance testing was declared successful, and the licensee declared Standby Diesel Generator 12 operable at 18:25 on October 11, 2007. This issue was entered into the licensee's corrective action program as Condition Records 07-14903 and 07-14959.
 
=====Analysis.=====
The team determined that the failure to adequately perform Technical Specification Surveillance Requirement 4.8.1.1.2.E.11 was a performance deficiency. The finding was determined to be more than minor because it is associated with the Mitigating Systems cornerstone attribute of "Design Control."  It impacts the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events and prevent undesirable consequences. Not fully performing the required surveillances increases the likelihood that the standby diesel generators and supported equipment would not perform their design safety functions when needed. Using Inspection Manual Chapter 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations," Phase 1 screening, the finding screened as having very low safety significance (Green) because the finding did not represent a loss of safety function of the sequence timer, standby diesel generator, or the essential chiller. This finding was reviewed for crosscutting aspects and none were identified.
 
=====Enforcement.=====
Technical Specification Surveillance Requirement 4.8.1.1.2.E.11 requires "Verifying that the automatic load sequence timer is OPERABLE with the first sequenced load verified to be loaded between 1.0 second and 1.6 seconds, and all other load blocks within +/- 10% of its design interval."  Contrary to the above, the licensee failed to verify the actual loading times of the sequenced loads. Specifically, the licensee only verified the time that the sequence timer demands breaker closure and did not perform the "verified to be loaded" requirement. Because the violation was of very low safety significance and has been entered into the licensee's corrective action program as
 
Condition Records 07-14903 and 07-14959, it is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy: NCV 05000498; 
- 15 -499/2007007-04, Surveillance Procedure Lacked Check for Timing of Chiller Loading on the Bus.
 
b.5. Inadequate Test Program for 125V DC Molded Case Circuit Breakers
 
=====Introduction.=====
The team identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion XI, "Test Control," for the failure to implement a test program to assure that all installed safety-related molded case circuit breakers will perform satisfactorily in service. Specifically, the licensee had not adequately exercised or subjected to periodic testing all of the 125V dc molded case circuit breakers since initial plant operation.
 
=====Description.=====
During the review of surveillance tests for the Auxiliary Feedwater Motor-Operated Valve 0019, the team discovered that the molded case circuit breaker had not been exercised or subjected to testing since the initial plant operation. In addition, further inspection discovered that the majority of 125V dc-fed molded case circuit breakers were also not exercised or subjected to periodic testing since installation in 1986. The types of molded case circuit breakers undergoing any type of preventative testing/maintenance included battery chargers, distribution panels, and inverters since they were infrequently cycled by other maintenance activities. Conversely, the breakers that fed loads to standby diesel generator field flash, reactor trip switchgear, 4.16kV switchgear control power and emergency safety feature load sequencers appeared to have not been tested since it was assumed that they were cycled in other maintenance activities.
 
The team noted that the licensee performed tests on molded case circuit breakers to satisfy Information Notice IEN 93-64 and ensure that molded case circuit breakers installed remained functional during plant operations. Following the test was an engineering evaluation acknowledging that molded case circuit breakers were subject to potential age-related degradation, which could result in a failure to trip in accordance with the published time-current characteristic curves because of various factors, such as grease hardening. In 2001, the licensee decided that the sample size for the dc-fed loads
 
indicated that limited failures in the test population did not warrant a pre-established test program. Essentially, credit was taken for circuit breakers being cycled as a part of other maintenance programs, but it was realized that these tests performed on breakers, failed to actually cycle the breaker. In fact the handswitch was used to open and close the valve.
 
- 16 -Updated Final Safety Analysis Report, Section 8.3.2.1.4, provides for "Periodic testing Class 1E dc power system equipment is performed in accordance with Regulatory Guide 1.32 to verify its ability to perform its safety function."  Information Notice 93-64, "Periodic Testing and Preventative Maintenance of Molded Case Circuit Breakers,"
stated, "Detecting or assessing degradation could only be accomplished through appropriate periodic testing and monitoring."  The team found that the licensee's evaluation and approach to the industry experience, design life, potential common mode failures, and component age concerns were not addressed in the test program. The licensee entered this finding into their corrective action program as Condition Record 07-15817.
 
=====Analysis.=====
The team determined that the lack of periodic testing on all of the dc molded case circuit breakers was a performance deficiency. The finding was determined to be more than minor because it is associated with the Mitigating Systems cornerstone attribute of "Equipment Performance."  It impacts the cornerstone objective of ensuring the availability, reliability, capability of systems that respond to initiating events and prevent undesirable consequences. Using Inspection Manual Chapter 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations,"
Phase 1 screening, the finding screened as having very low safety significance (Green)because it did not result in a loss of safety function of a system or train. This finding was reviewed for crosscutting aspects and none were identified.
 
=====Enforcement.=====
Part 50 of Title 10 of the Code of Federal Regulations
, Appendix B, Criterion XI, "Test Control," stated, in part, that test programs shall be established to assure that all testing required to demonstrate that structures, systems and components will perform satisfactorily in service. Contrary to the above, the licensee failed to implement a test program to assure all installed safety-related molded case circuit breakers will perform satisfactorily in service. Because this violation was of very low safety significance and has been entered into the licensee's corrective action program as Condition Record 07-15817, it is being treated as a noncited violati on consistent with Section VI.A of the NRC Enforcement Policy: NCV 05000498; 499/2007007-05, Inadequate Test Program for 125V DC Molded Case Circuit Breakers.


William B. Jones, Chief Engineering Branch 1 Division of Reactor Safety Dockets: 50-498; 50-499 Licenses: NPF-76; NPF-80
b.6. Failure to Incorporate Instrument Uncertainties into Surveillance Requirements for Technical Specification Limiting Condition for Operation 3.5.2 (Surveillance Requirement 4.5.2.f)


STP Nuclear Operating Company 3 cc w/enclosure: E. D. Halpin Site Vice President STP Nuclear Operating Company P.O. Box 289 Wadsworth, TX 77483 Ken Coates Plant General Manager STP Nuclear Operating Company P.O. Box 289 Wadsworth, TX 77483 S. M. Head, Manager, Licensing STP Nuclear Operating Company P.O. Box 289, Mail Code: N5014 Wadsworth, TX 77483
=====Introduction.=====
The team identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criteria III, "Design Control," for the failure to adequately translate design basis information into specifications and procedures. Specifically, measurement instrument uncertainties were not included in the determination of minimum allowed high head safety injection pump and low head safety injection pump developed head values during periodic technical specification surveillance testing.


C. T. Bowman General Manager, Oversight STP Nuclear Operating Company P.O. Box 289 Wadsworth, TX 77483 Marilyn Kistler Sr. Staff Specialist, Licensing STP Nuclear Operating Company P.O. Box 289, Mail Code 5014 Wadsworth, TX 77483 C. M. Canady City of Austin Electric Utility Department 721 Barton Springs Road Austin, TX 78704 J. J. Nesrsta/R. K. Temple/
- 17 -Description. Technical Specification Limiting Condition for Operation 3.5.2, Surveillance Requirement 4.5.2.f.1 for the high head safety injection pump, and Surveillance Requirement 4.5.2.f.2 for the low head safety injection pump require:
E. Alercon/Kevin Pollo City Public Service Board P.O. Box 1771 San Antonio, TX 78296 Jon C. Wood Cox Smith Matthews 112 E. Pecan, Suite 1800 San Antonio, TX 78205
For the High Head Safety Injection pumps, verification that the pump develops a differential pressure on recirculation flow when tested pursuant to Technical Specification 4.0.5 greater than or equal to 1480 psid.


STP Nuclear Operating Company 4A. H. Gutterman, Esq. Morgan, Lewis & Bockius 1111 Pennsylvania Avenue NW Washington, DC 20004 Institute of Nuclear Power Operations (INPO)Records Center 700 Galleria Parkway SE, Suite 100 Atlanta, GA 30339 Director, Division of Compliance & Inspection Bureau of Radiation Control Texas Department of State Health Services 1100 West 49th Street Austin, TX 78756 Brian Almon Public Utility Commission William B. Travis Building P.O. Box 13326 1701 North Congress Avenue Austin, TX 78701-3326 Environmental and Natural Resources Policy Director P.O. Box 12428 Austin, TX 78711-3189 Judge, Matagorda County Matagorda County Courthouse 1700 Seventh Street Bay City, TX 77414 Terry Parks, Chief Inspector Texas Department of Licensing and Regulation Boiler Program P.O. Box 12157 Austin, TX 78711
For the Low Head Safety Injection pumps, verification that the pump develops a differential pressure on recirculation flow when tested pursuant to Technical Specification 4.0.5 greater than or equal to 286 psid.


Susan M. Jablonski Office of Permitting, Remediation and Registration Texas Commission on Environmental Quality MC-122, P.O. Box 13087 Austin, TX 78711-3087 Ted Enos 4200 South Hulen Suite 422 Fort Worth, TX 76109
Upon review of Surveillance Procedures 0PSP03-SI-0001 and 0PSP03-SI-0004, the team identified that the pump developed head acceptance criteria in the procedures did not include consideration of measurement instrument uncertainties and were numerically equal to the technical specification values. As a result, there was no documented assurance that the recorded current and historical surveillance test results would demonstrate pump developed heads above the required minimum technical specification requirements when measurement instrument uncertainties were taken into consideration. Therefore, the technical specification surveillance test acceptance criteria were non-conservative.


STP Nuclear Operating Company 5Steve Winn/Christine Jacobs/ Eddy Daniels/Marty Ryan NRC Energy, Inc. 211 Carnegie Center Princeton, NJ 08540
The team reviewed Design Basis Document 5Z529ZB01025, "Technical Specification/ Limiting Conditions for Operation Design Basis Document," Revision 2, and determined that it had erroneously stated for both high-head safety injection pumps and low-head safety injection pumps that "This value is a conservative, nominal value and needs no additional instrument uncertainty margin. This value is acceptable for use. This value is only used in this application (Technical Specifications 4.5.2.f.1 and Technical Specification 4.5.2.f.2)."


Lisa R. Hammond, Chief Technological Hazards Branch National Preparedness Division FEMA Region VI 800 N. Loop 288 Denton, TX 76209
The licensee issued Condition Record 07-15752. The condition record stated that "The pump test procedures currently use the technical specification values as the low limit for operability and should be revised. The most recent performance of all safety injection pumps meets the "upward adjusted" low limits."


STP Nuclear Operating Company 6Electronic distribution by RIV: Regional Administrator (BSM1) DRP Director (ATH) DRS Director (DDC) DRS Deputy Director (RJC1) Senior Resident Inspector (JLD5) Branch Chief, DRP/A (CEJ1) Senior Project Engineer, DRP/A (TRF) Team Leader, DRP/TSS (CJP) RITS Coordinator (MSH3) SUNSI Review Completed: __Y____ADAMS:
=====Analysis.=====
# Yes G No Initials: __WCS____
The failure to include consideration of measurement instrument uncertainties, in relation to the instrumentation utilized in periodic surveillance tests to measure the pump developed head, into the technical specification surveillance test acceptance criteria was a performance deficiency. The finding was determined to be more than minor because it is associated with the Mitigating Systems cornerstone attribute of "Design Control."  It impacts the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events and prevent undesirable consequences. Using Inspection Manual Chapter 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Pow er Situations," Phase 1 screening, the finding screened as having very low safety significance (Green) because it 
# Publicly Available G Non-Publicly Available G Sensitive
- 18 -did not represent a loss of safety system function. This finding was reviewed for crosscutting aspects and none were identified.
# Non-Sensitive SRI:EB1 C:EB1 WCSifre/lmb WBJones
/RA/ /Ra/ 08/22/2007 08/22/2007 OFFICIAL RECORD COPY T=Telephone E=E-mail F=Fax Enclosure 1 INFORMATION REQUEST FOR COOPER NUCLEAR STATION COMPONENT DESIGN BASES INSPECTION (CDBI)
Inspection Report: 05000498/2007007; 05000499/2007007


Information Gathering Dates: September 4-5, 2007 On-site Inspection Dates: September 24-28, 2007, October 15-26, 2007.
=====Enforcement.=====
Part 50 of Title 10 of the Code of Federal Regulations, Appendix B, Criterion III, states, in part, that measures shall be established to assure that design bases are correctly translated into specifications and procedures. Contrary to the above, the licensee did not conservatively account for the effect of instrument uncertainty in development of acceptance criteria for the technical specification surveillance values for Technical Specification Limiting Condition for Operation 3.5.2. Thus, the minimum allowed high head safety injection and low head safety injection pump developed head had not been definitively demonstrated during surveillance testing to exceed the minimum Technical Specification limiting condition for operation values. Because this violation was of very low safety significance and has been entered into the licensee's corrective action program as Condition Record 07-15752, it is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy: NCV 05000498; 499/2007007-06, Failure to Incorporate Instrument Uncertainties into Surveillance Requirements for Technical Specification Limiting Condition for Operation 3.5.2 (Specifically Surveillance Requirement 4.5.2.f).


Inspection Procedure: IP 71111.21, "Component Design Bases Inspection" Lead Inspector/Team Leader: Wayne C. Sifre 817-860-8193 wcs@nrc.gov
==OTHER ACTIVITIES==
{{a|4OA5}}
==4OA5 Other Activities==


I. Information Requested Prior to the information Gathering Visit The following information is requested by September 29, 2007, or sooner, to facilitate inspection preparation. If you have any questions regarding this information, please call the lead inspector as soon as possible. (Please provide the information electronically in ".pdf" files, Excel, or other searchable formats. The information should contain descriptive names, and be indexed and hyperlinked to facilitate ease of use. Information in "lists" should contain enough information to be easily understood by someone who has knowledge of pressurized water reactor technology).
a.1 Unresolved Item Associated with the Effect of Standby Diesel Generator Technical Specification Voltage Variation on Supplied Equipment


1. An excel spreadsheet of equipment basic events (with definitions) including importance measures (RAW, FV) from your internal events probabilistic risk assessment (PRA), including risk ranking of top 50 components from your site specific PRA sorted by Risk Achievement Worth (RAW). Include values for Birnbaum Importance, Risk Reduction Worth (RRW), and Fussell-Veseley (FV) (as applicable).
=====Introduction.=====
The team identified an unresolved item associated with the steady state output voltage supplied by the standby diesel generators is allowed to vary by Technical Specification 3/4.8.1 from 3744 V to 4576 V (+/- 10%) during a loss of offsite power event. Specifically, the licensee has not analyzed for the effect of this full variation.


2. Provide a list of the top 500 cut-sets from your PRA
=====Description.=====
The design analysis assumed maximum supplied voltage variations based upon offsite power supplies which were analyzed to vary less than the technical specification allowed steady state variation for the standby diesel generators. Components throughout the plant would be adversely affected by either an undervoltage or overvoltage condition.


3. Copies of PRA system notebooks 4. An excel spreadsheet of PRA human action basic events or risk ranking of operator actions from your site specific PSA sorted by RAW and FV. Provide copies of your human reliability worksheets for these items.
Since this is a very broad issue that encompasses components powered from the standby diesel generator during a design basis event, the licensee will require significant time to evaluate its effects. Although available safety margins will be less, the degree of this effect is not yet known since the effect of the variation varies upon the analyzed parameter and currently analyzed margins vary significantly. The actual safety function of equipment is not expected to be compromised since the standby diesel generators are presently controlled to a tighter band of voltage operation than allowed by technical specifications and review of the surveillance testing of the standby diesel generators confirms this tighter band is currently being maintained.


5. If you have an External Events or Fire PSA Model, provide the information requested in items 1-3 for external events and fire.
- 19 - Once the licensee has evaluated the effect of the allowed steady state voltage variation and determined the degree of safety margin impact throughout the plant, the NRC can complete the inspection of that analysis in order to close this issue. The licensee has


6. Any Pre-existing evaluation or list of components and associated 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 etc.).
documented this issue in Condition Record 07-15554 and the item is unresolved pending the licensee's completion of its analysis and NRC review: URI05000498; 499/2007007-07, Effect of Standby Diesel Generator Technical Specification Voltage Variation on Supplied Equipment.
Enclosure 2 7. List of high risk maintenance rule systems/components and functions; based on engineering or expert panel judgment.


8. A list of operating experience evaluations for the last 2 years.
a.2 Unresolved Item Involving Combined Adverse Conditions not considered in Fuel Oil Storage Tank Sizing


9. 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 generators close to design required output, heat exchangers close to rated design heat removal etc.)
=====Introduction.=====
The team identified an unresolved item involving accounting for the combined effect of vortexing and standby diesel generator frequency variations in fuel oil storage tank sizing.


10. A list of permanent and temporary modifications sorted by component identified in Item 1.
=====Description.=====
Calculation MC-6256, "Sizing of Standby Diesel Generator Fuel Oil Storage Tank," Revision 4, determined a total 7-day fuel oil requirement of 51,500 gallons, comparing this value with a technical specification requirement of 60,500 gallons. However, this calculation did not consider the effects of vortexing or generator frequency variations. Condition Record 97-14434-10 included an evaluation of fuel oil vortexing completed as part of a "Review of Safety Related Tanks (other than Refueling Water Storage Tank & Auxiliary Feedwater Storage Tank) for Vortexing Concerns."  Separately, Calculation EC-5100, Standby Generator Transient Response Model," Revision 2, contained an evaluation performed under Condition Record 97-13089-1 in order to "Perform Evaluation of Electrical Frequency Variations on Mechanical Fluid Systems."


11. Information of any common cause failure of components experienced in the last 5 years at your facility.
The vortexing evaluation determined that 13.5 inches of fuel oil volume would be susceptible to excessive air entrainment, representing 4120 gallons of unusable fuel oil with a 7-day fuel oil requirement of 55,360 gallons (referencing Calculation MC-6038, "Standby Diesel Generator Fuel Oil Storage Tank Level Setting Calculation."  The total required volume would therefore be 59,480 gallons.


12. List of current "operator workarounds."
The frequency effects evaluation determined that "Standby Diesel Generator load would increase by roughly 6% because the majority of load consists of pumps and fans with primarily friction system loads."  The evaluation then compared this 6 percent increase in load with the standby diesel generator fuel oil storage tank calculated margin of more than 10 percent.


13. A list of the design calculations which provide the design margin information for components included in item 1. (Calculations should be available during the information gathering visit).
However, the vortexing evaluation had already effectively reduced the majority of the analyzed margin with a remaining 1020 gallons of fuel oil between 59,480 gallons and the technical specification requirement of 60,500 gallons. Therefore, applying a 6 percent increase in standby diesel generator load in addition to considering vortexing effects would have exceeded the technical specification requirement under those analyzed conditions.


14. List of root cause evaluations associate with component failures or design issues initiated/completed in the last 5 years.
- 20 - In addition, the most recent fuel oil storage tank sizing calculation determined a 7 day fuel oil requirement of 51,500 gallons. As discussed in the finding "Manual Loads not Considered for Fuel Oil Storage Tank Sizing Calculation," this requirement neglected manual loads during the 7 days for which provision would be made to use during a design basis event. A bounding analysis considering the actual anticipated manual loads, in addition to the vortexing reduction and increased load frequency effect, exceeds the minimum technical specification requirement.


15. Current management and engineering organizational charts.
The fuel oil storage tank sizing calculation included additional conservative assumptions regarding expected pump operation during design basis accident scenarios. For example, the calculation assumed auxiliary feedwater, high-head safety injection, and containment spray pumps would be run continuously for 48 hours following a large break loss-of-coolant accident. The licensee recalculated the fuel oil storage tank sizing using more realistic assumptions with respect to load profile and determined sufficient fuel oil margin does exist with all design basis conditions considered. This issue was entered into the licensee's corrective action program as Condition Request 07-14398 and 07-15592.


16. South Texas Project IPEEE, if available electronically.
After further discussions with staff from the NRC Office of Nuclear Reactor Regulation, the team concluded that this issue of failure to account for the combined effect of vortexing and standby diesel generator frequency variation in the fuel oil storage tank sizing would remain open as an unresolved item. Additional NRC staff review was necessary to determine whether the issue was acceptable, whether it was a finding, or whether it constituted a deviation or violation. Pending completion of this review, this item is unresolved: URI 05000498; 499/2007007-08, Combined Adverse Conditions not Considered in Fuel Oil Storage Tank Sizing.


17. Mechanical piping drawings for:
{{a|4OA6}}
Engineered Safety Features Emergency core cooling Systems Emergency Diesel Generators 18. Electrical one-line drawings for:
==4OA6 Meetings, Including Exit==


Off-site power/switchyard supplies Normal ac power systems Emergency ac/dc power systems including, 120Vac power, and 125Vdc/24Vdc safety class systems 19. List of any common-cause failures of components in the last 3 years.
On October 26, 2007, the team leader presented the preliminary inspection results to Mr. E. Halpin, Site Vice President, and other members of the South Texas Project staff. After additional offsite and onsite inspection a preliminary exit meeting was conducted on November 26, 2007, with Mr. J.


Enclosure 3 II. Information Requested to be Available on the First Day of Inspection (September 24, 2007) 1. List of condition reports (corrective action documents) associated with each of the selected components for the last 5 years.
Sheppard, President and Chief Executive officer and other members of the licensee's staff. After additional in-office inspection, a telephonic exit was conducted on January 22, 2008. The licensee acknowledged the findings during each meeting. While some proprietary information was reviewed during this inspection, no proprietary information was included in this report.


2. The corrective maintenance history associated with each of the selected components for the last 2 years.
A-1


3. Copies of calculations associated with each of the selected component (if not previously provided), excluding data files. Please review the calculations and also provide copies of reference material (such as drawings, engineering requests, and vendor letters).
=SUPPLEMENTAL INFORMATION=


4. Copies of operability evaluations associated with each of the selected components and plans for restoring operability, if applicable.
==KEY POINTS OF CONTACT==


5. Copies of selected operator workaround evaluations associated with each of he selected components and plans for resolution, if applicable.
===Licensee personnel===
: [[contact::C. Bowman]], General Manager Oversight
: [[contact::K. Coats]], Plant General Manager
: [[contact::R. Engen]], Manager, Maintenance Engineering
: [[contact::E. Halpin]], Site Vice President
: [[contact::S. Head]], Manager, Licensing
: [[contact::K. House]], Manager, Design Engineering
: [[contact::B. Jenewein]], Manager, Testing/Programs Engineering
: [[contact::R. Lovell]], Manager, Industrial Alliances
: [[contact::M. Meier]], General manager Station Support
: [[contact::J. Mertink]], Manager, Operations
: [[contact::M. Murray]], Manager, Systems Engineering
: [[contact::G. Powell]], Manager, Site Engineering
: [[contact::D. Rencurrel]], Vise President, Engineering
: [[contact::M. Ruvalcaba]], Supervisor, Engineering
: [[contact::J. Sheppard]], President and Chief Executive Officer
: [[contact::D. Towler]], Manager, Quality


6. Copies of any open temporary modifications associated with each of the selected components, if applicable.
===NRC personnel===
: [[contact::W. Jones]], Chief, Engineering Branch 1
: [[contact::J. Dixon]], Senior Resident Inspector


7. Trend data on the selected electrical/mechanical components' performance for last 3 years (for example, pumps' performance including in-service testing, other vibration monitoring, oil sample results, etc., as applicable).
==LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED==


8. A copy of any internal/external self-assessments and associated corrective action documents generated in preparation for the inspection.
===Opened===


9. A copy of engineering/operations related audits completed in the last 2 years.
URI05000498; 499/2007007-07 URI Effect of Standby Diesel Generator Technical Specification Voltage Variation on Supplied


10. List of motor-operated valves (MOVs) in the valve program , design margin and risk ranking.
Equipment URI05000498; 499/2007007-08 URI Combined Adverse Conditions not Considered in Fuel Oil Storage Tank Sizing


11. List of air-operated valves (AOVs) in the valve program, design and risk ranking.
===Opened and Closed===


12. SSCs in the maintenance rule (a)(1) category.
NCV05000498; 499/2007007-01 NC
V Failure to Specify Setpoint Calibration Limits in Relay Setpoint Calculations


13. Site top ten issues list.
NCV05000498; 499/2007007-02 NC
V Manual Loads not Considered for Fuel Oil Storage Tank Sizing Calculation 


14. Provide list of PRA assumptions regarding operator actions and the associated procedures.
NCV05000498; 499/2007007-03 NC
V Failure to Use Correct Design Inputs in Determination of the Weak Link for the Auxiliary
Feedwater System Outside Containment Isolation
Motor Operated Valves


15. List of licensee contacts for the inspection team with pager or phone numbers.
NCV05000498; 499/2007007-04 NC
V Surveillance Procedure Lacked Check for Timing of Chiller Loading on the Bus


Enclosure 4III. Information Requested to be provided throughout the inspection.
NCV05000498; 499/2007007-05 NC
V Inadequate Test Program for 125V DC Molded Case Circuit Breakers


1. Copies of any corrective action documents generated as a result of the team's questions or queries during this inspection.
NCV05000498; 499/2007007-06 NC
V Failure to Incorporate Instrument Uncertainties into Surveillance Requirements for Technical
Specification Limiting Condition for Operation 3.5.2
(Specifically Surveillance Requirement 4.5.2.f)


2, Copies of the list of questions submitted by the team members and the status /resolution of the information requested (provide daily during the inspection to each team member).
==LIST OF DOCUMENTS REVIEWED==
===Calculations===
: Number Title Revision/Date
: MC-5694 Auxiliary Feedwater System Failure Modes and Effects Analysis 3
: EC-5008 Class 1E Battery, Battery Charger and Inverter Sizing 13
: EC-5100 Standby Diesel Generator Transient Response Model 2
: MC-6213
: GNL 89-10 Calc for MOV 1AF0019 6
: ZC-7038 Loop Uncertainty Calcul ation for QDPS Cabinet Temperature Instrumentation
: EC-5001 Fault Analysis 6
: MC-6462 DVAC Calculation for DC Motor MOVs 0
: EC-5031 480 Volt Load Centers 7
: EC-5018 Short Circuit Current Analysis - Class 1E 125 VDC and Non-Class 1E 250, 125, and 48 VDC Systems
: EC-5003-02 Cable Ampacity in Underground Ducts 8
: EC-5037 Maximum Allowable Length of AC Power Cables 4
: EC-5033 Protection Non 1E 48 VDC, 125 VDC & 250 VDC and Class
: IE 125 VDC Systems
: EC-5004 Cable Ampacity
: 7 3Q159MC6038 SDBY DG FOST Level Setting Calculation 2 5Q159MC5912 NPSH on the Fuel Oil Transfer Pump
===Calculations===
: Number Title Revision/Date
: EC-5008 Class 1E Battery, Battery Charger and Inverter Sizing 13
: EC-5033 Protection Non 1E 48 VDC, 125 VDC & 250 VDC and Class 1E 125 VDC Systems
: EC-5018 Short Circuit Current Analysis - Class 1E 125 VDC and Non-Class 1E 250, 125 and 48 VDC Systems
: EC-5100 Standby Diesel Generator Transient Response Model 2
: MC-5037 RWST Volumes & Limits 9
: MC-6256 Sizing of SDG FOST 4
: MDCN 89219-75 Protection - DC System (CB & RLY Settings) 01/15/97
: ZC-7029 Loop Uncertainty Calculation for Standby Diesel Generator Fuel Oil Storage Tank Level Monitoring Instrumentation
: EC-5000 Voltage Regulation Study 12
: EC-5003-6 Cable Ampacity in Underground Ducts-Data Sheets 11
: EC-5004 Cable Ampacities 7
: EC-5014 Maximum Length of Control Cables 4
: EC-5022 Transformer Neutral Grounding Resistor sizing 2
: EC-5020 Main Transformer Sizing Calculation 3
: EC-5024 Diesel Generator Neutral Grounding 2
: EC-5028 Protection 13.8 KV Switchgear 9
: EC-5029 4.16 KV Switchgear Relay Setting 5
: EC-5030 Class 1E Diesel Generation Protection 1
: EC-5034 Standby Transformer Protection 3
: EC-5036 DC Cable Sizing 7
: EC-5039 Control Cable Size Verification 0
: EC-5052 Degraded and Undervoltage Protection 6
: CC-06425 1997 Emergency Cooling Pond Sediment Calculation 0
: CC-09959 2002 Emergency Cooling Pond Sediment Calculation 0 FRSS/CWBS-C-121 TGX Minimum and Maximum Safeguards 07/13/87
: MC-5430 Emergency Cooling Water Intake Structure Cooling and Heating Loads
: MC-05860 Containment Emergency Sump Performance 1
: MC-6220 SI & CS Pump NPSH 4
: MC-6251 Essential Cooling Water Transient Analysis 0
: MC-6412 Essential Chilled Water Load 1
: MC-06482 Essential Chilled Water / EAB HVAC Design Basis Loads with Capacity of 300 Tons per Train
: PFD-FTE-285 Standard Single and Twin Units - 4XL Model Fluid Systems Process Flow Diagrams and Piping Design
: Requirements
: V-EC-1330 Motor Operated Valve (MOV) Evaluation (A2SIMOV0031A)
: ZC-7024 Loop Uncertainty Calculation for RWST Level 2
===Calculations===
: Number Title Revision/Date Monitoring Instrumentation 2N129MC5519 Pressure Drop Evaluation for the Safety Injection System 0 2N129MC5815 RWST Vacuum Potential 0 2N129MC6091 Minimum Flow Orifices SI System - Low Head - Line 3"SI1302PB2
: 3N129HMC6100 Evaluate Safety Related Pump Miniflows per NRC Bulletin 88-04
: 3R289MC5429 Head Losses Calculation - Essential Cooling Water System 1 3R289MC5633 Essential Cooling Water Pump Submergence 2 3V110MC5234 Expansion Tank Sizing for Essential Chilled Water System 2 5N129MC5519 Pressure Drop Evaluation for the Safety Injection System 0 5R289MC5812 Essential Cooling Water (ECW) Hydraulic Network Analysis (HNA)
: 88-EW-002 ECW Pump Discharge and Suction Pressure 
===Calculations===
: 3Q159MC6038 Standby Diesel Generator Fuel Oil Storage Tank Level Setting Calculation
: 5Q159MC5912 NPSH on the Fuel Oil Transfer Pump 0 3N129HMC6100 Evaluate Safety Related Pump Miniflows 0 3S149MC5051 Auxiliary Feedwater Pump Discharge Pressure 4 3S149MC5861 Auxiliary Feedwater Pump Design TDH (Total Discharge Head), Flow Rate and Pump Runout
: 3S149MC5057 Maximum and Minimum Flow Requirements of the AFW System
: 3L482MC6204 GNL (Generic Letter) 89-10 Calculation for MOVD2AFMOV0019 (Weak Link Analysis)
: DCN
: MC-145 Revision of Mc 6204 to Incorporate Yield Stress Values for Weak Link Calculation at Design Temperatures.
: 7/31/1992
: MC-6163 Penetration Seals for HELBA (High Energy Line Break Analysis) and Flooding
: MC-5557 IVC (Isolation Valve Cubicle) Flood Analysis 8
: ZC-7029 FOST (Fuel Oil Storage Tank) Wide Range Level Indicating Loops, page 7.
: MC-6256 Sizing of Standby Diesel Generator FOST 4
: Condition Records
: 07-14379 07-14383 07-14422 07-14463 07-14520 07-15473
: 07-15752 07-14425 07-14423 07-15418 07-15443 07-3647 
: 07-3745 07-4926 07-5278 07-5465 07-5652 07-7917
: 07-14387 07-14398 07-14663 07-14903 07-14959 07-15449
: 07-15554 07-15592 07-15568 07-15586 07-15801 07-15828
: 07-16309 07-15817 07-15847 07-15702 07-11491 07-11517
: 07-15499 06-16616 06-10937 06-4423 06-16998 05-2442
: 05-15477 05-14112 05-15251 04-5149 04-15477 04-15476
: 04-4428 03-928 03-8824 03-14479 03-9356 01-15847
: 97-14434
===Drawings===
: Number Title Revision/Date 
: 00009E0VAAB#2 Single Line Diagram Vital 120V AC Distribution Panels
: DP 1202,
: DP 1203
: 00009E0PMAK#2 Single Line Diagram Class-IE Motor Control Center E2A4
: EAB 18 00009E0DJAB#2 Single Line Diagram 125 V DC 1E Distribution Switchboard E2D11
: 00009E0AF14#1 E/D AFW Turbine Pump 14
: MOV 0019 11 00009E0PLAB#2 Single Line Diagram, 480 V Class 1E Load Center E2B 12
: 00009E0PKAA#2 Single line Diagram, 4.16KV Class 1E Switchgear E2A 10
: 00009E0EW01#2 Elementary Diagram, Essential Cooling Water Pumps 2A, 2B, & 2C
: 00009E0DJAB#1 Single Line Diagram, 125VDC 1E Distribution Switchboard E1D11
: 00009E0PMAD#2 Single Line Diagram, 480 V Class 1E Motor Control Center 21 00009E0VAAB#1 Single Line Diagram Vital 120V AC Distribution Panels
: DP 1202,
: DP 1203
: 00009E0PMAD#2 Single Line Diagram 480 V Class 1E Motor Control Center E2B1 (EAB)
: 00009E0HE13#2 Elementary Diagram E.A.B. HVAC Return Fans, FN001, FN002 & FN003
: 00009E0HE09#2 Elementary Diagram E.A.B. HVAC Main Supply Air Vent Fans, FN014, FN015, & FN016
: 00009E0AAAB#1 Single Line Diagram Class 1E 125V DC & 120V Vital AC Non-Class 1E 48V, 125V, 250V, DC, & 120V Vital AC Non-Class 1E Inverter Power for Computer
: 208V/120V AC Regulated Power
: 00009E0DJAA#1 Single Line Diagram 125VDC Class 1E Distribution SWBD. E1A11 (Channel I) (E.A.B.)
: 00009E0DJAB#1 Single Line Diagram 125V DC Class 1E Distr. Switchboard E1D11 (Channel II) (EAB)
: 00009E0PMAD#2 Single Line Diagram, 480 V Class 1E Motor Control Center 21 00000E0AAAA Main One Line Diagram for Units No. 1 & 2 (site 19 
: A-6Drawings Number Title Revision/Date
: Electrical Di stribution) 00009E0DJAA#1 Single Line Diagram 125VDC Class 1E Distribution SWBD. E1A11 (Channel I) (E.A.B.)
: 00009E0DJAC#1 Single Line Diagram 125V DC Class 1E Distribution SWBD E1B11 (Channel III) (EAB)
: 00009E0DJAD#1 Single Line Diagram 125V DC Class 1E Distribution SWBD E1C11 (Channel IV) (EAB)
: 00009E0DJAE#1 Single Line Diagram 125V DC Class 1E Distribution Panels PL039A, PL039B, PL039C, PL040A (EAB)
: 00009E-PKAB-01
#2 Single Line Dwg 4.16KV Class 1E Switchgear 9 00009EOPCAB #2 Single Line Dwg 13.8KV Switchgear 2GA 14 00009EOPLAB #2 Single Line Dwg 480V Class 1E Load Center E2B 16
: 00009EOPC21 #2 Elementary Diagram 13.8KV Standby Bus 2G Supply BKR
: ST-260 from #2 Standby XFMR, Sheet1
: 00009EOPC25 #2 Elementary Diagram 13.8KV Standby Bus 2G Supply BKR
: ST-280 from #1 Standby XFMR, Sheet 1
: 00009EOPK03 #2 Elementary Diagram 4.16KV Feeder to 480V Loadcenter Transformer E2A2, E2B2, E2C2, Sheet 1
: 00009EOPK02 #2 Elementary Diagram 4.16KV Feeder to 480V Loadcenter Transformer E2A1, E2B1, E2C1, Sheet 1
: 00009EOPK01 #2 Elementary Diagram 4.16KV ESF Bus E2A, E2B, E2C Supply Breaker Control, Sheet 1
: 00009EODG01 #2 Elementary Diagram Standby Diesel Generator DG22 4.16KV Feeder Breaker, Sheet 3
: 00009EOPC19 #2 Elementary Diagram 13.8KV 2G Aux and Standby Bus Tie BKR T-240, Sheet 1
: 3V119V10002#1 P&ID - HVAC / Essential Chilled Water System 13 3V119V10003#1 P&ID - HVAC / Essential Chilled Water System 18 3V119V10004#1 P&ID - HVAC / Essential Chilled Water System 9 4352-00006JF 125V DC Distribution Switchboard E1D11 G
: 4352-00004JF Class 1E 125 VDC Distribution Switchboard E1D11 BOM F 5Q159F00045#1, Sheet 1 Piping & Instrumentation Diagram Standby Diesel Generator Fuel Oil Storage & Transfer System
: 5Q159F00045#1, Sheet 2 Piping & Instrumentation Diagram Standby Diesel Fuel Oil 10 5Q159F22540#1 Piping & Instrumentation Diagram Standby Diesel Jacket Water 20 5Q159F22541#1 Piping & Instrumentation Diagram Standby Diesel Cooling Water
: 5Q159F22542#1 Piping & Instrumentation Diagram Standby Diesel Lube Oil 19 
: A-7Drawings Number Title Revision/Date 
: 5Q159F22543#1 Piping & Instrumentation Diagram Standby Diesel Air Intake & Exhaust
: 5Q159F22544#1 Piping & Instrumentation Diagram Standby Diesel Starting Systems & Alarms
: 5Q159F22545#1 Piping & Instrumentation Diagram Standby Diesel Shutdown System
: 5Q159F22546#1 Piping & Instrumentation Diagram Standby Diesel Starting Air
: 9-E-DJAF-01#1 Single Line Diagram 125V DC Class 1E Distribution Panels PL139A, PL139B, PL139C (DGB)  
: 5R289Z42077#2 Essential Cooling Water Pumps Logic Diagram 13 5R289Z42081#2 ECW Pump Discharge Valves Logic Diagram System: EW 8 5V119Z41572 E.A.B. HVAC Return Fans Logic Diagram 12 5S149Z40136 AFW Turbine Pump Isolation Valve Logic Diagram 9
: 5S142F00024 Piping & Instrumentation Diagram Auxiliary Feedwater 10 5S141F00024 Piping & Instrumentation Diagram Auxiliary Feedwater 11 5S142F00024 Sheet Piping and Instrument Drawing for Auxiliary Feedwater System 10 5S142F00024 Sheet Piping and Instrument Drawing for Auxiliary Feedwater System 10 5G-15-9-P-0053 Composite Piping - Isolation Valves Cubicle Plan at El.
: 34'-0" 3 5N129F05013#1 P&ID - Safety Injection System 27 5N129F05016#1 P&ID - Safety Injection System 14
: 5N169F20000#1 P&ID - Residual Heat Removal System 24
: 5R289F05038#1 P&ID - Essential Cooling Water System - Train A
: 13
: 5R289F05039#1 Sheet 1, P&ID - Essential Cooling Water System 16
: 5V119V10001#1 P&ID - HVAC / Essential Chilled Water System 31 5V119V10001#2 P&ID - HVAC / Essential Chilled Water System 31 5V119V25000#2 P&ID - HVAC / Electrical Auxiliary Building Main Area System 16 5V159V00027#2 P&ID - HVAC Miscellaneous Buildings Essential Cooling Water Intake Structure
: 6P-20-0-M-0031 General Arrangement Drawing - Essential Cooling Water Intake & Discharge Structures
: 8114-01036-WU Vendor Dwg L.V.M.E. "DS" SWGR C 8121-01023-GU Vendor Dwg Indoor Metal-Clad SWGR 5HK B
: 9G069F10006 #2 Piping and Instrumentation Diagram Isolation Valves Cubicles Building Sump Pump & Drains System for oily


3. Reference materials (available electronically and as needed during all on-site weeks): * General set of plant drawings * IPE/PRA report * Procurement documents for components selected * Plant procedures (normal, abnormal, emergency, surveillance, etc.) * Technical specifications * Updated Final Safety Analysis Report
waste. 6 9-M-06-9-B-0177 Plumbing Isolation Valve Cubicle Building Floor Plan 2 
* Vendor manuals
: A-8Drawings Number Title Revision/Date
: EL. 34'-0" Area 11 9-M-06-9-B-0175 Plumbing Isolation Valve Cubicle Building Embedment Plan EL. Area 11
: 9EAF14-01#1 Elementary Diagram Aux Feedwater Turbine Pump 14 Isolation
: MOV-0019 
===Procedures===
: Number Title Revision/Date
: 0POP02-AM-0001 ERFDADS Computer 120 VAC
: UPS 13 0POP02-AE-0004 120 VAC ESF Vital Distribution Power Supplies 25 0POP02-AF-0001 Auxiliary Feedwater 24
: 0POP02-HE-0001 Electrical Auxiliary Building HVAC System 26 0POP02-EW-0001 Essential Cooling Water Operations 41 0POP04-HE-0001 Loss of EAB or Control Room HVAC 7 1POP09-AN-03M2 Annunciator Lampbox 1-03M-2 Response Instructions 26 2POP09-AN-03M2 Annunciator Lampbox 2-03M-2 Response Instructions 23 0POP09-AN-22M3 Annunciator Lampbox 22M03 Response Instructions 20 0POP09-AN-02M3 Annunciator Lampbox 2M03 Response Instructions 19 0PGP03-ZE-0073 Molded Case Circuit Breaker Testing Program 2 0PMP05-NA-0004 Molded Case Breaker Test 25
: 0PSP03-HE-0001 Control Room Emergency Ventilation System
: 0PSPS03-AF-0010 Auxiliary Feedwater System Valve Operability Test 21 0PSP03-EW-0017 Essential Cooling Water System Train A Testing 24 0PSP03-AF-0011 Auxiliary Feed Flow Verification 7
: 0PSP03-SP-0019D Turbine Driven Auxiliary Feedwater Actuation and Response Time Test
: 0POP01-ZO-0009 Ground Isolation 0 0POP02-DG-0001 Emergency Diesel Generator 11(21) 42
: 0POP02-DG-0002 Emergency Diesel Generator 12(22) 48
: 0POP02-DG-0003 Emergency Diesel Generator 13(23) 45
: 0POP02-EE-0001 ESF (Class 1E) DC Disctribution System 16 0POP04-AE-0001 First Response to Loss of Any or All 13.8 KV or 4.16 KV Bus 34 0POP04-DJ-0001 Loss of Class 1E 125 VDC Power 20 0POP05-E0-EC00 Loss of All AC Power 18
: 0POP09-AN-0102 Annunciator Lampbox 1(2)-102 Response Instructions 12 0POP09-AN-0104 Annunciator Lampbox 1(2)-104 Response Instructions 12 0POP09-AN-0106 Annunciator Lampbox 1(2)-106 Response Instructions 12 0POP09-AN-02M3 Annunciator Lampbox 2M03 Response Instructions 19 0POP09-AN-03M3 Annunciator Lampbox 3M03 Response Instructions 22 0POP09-AN-22M3 Annunciator Lampbox 22M03 Response Instructions 20 0PSP02-SF-0001A ESF Diesel Sequencer Timing Test Train A 11
===Procedures===
: Number Title Revision/Date
: 0PSP03-DG-0001 Standby Diesel 11(21) Operability Test 33
: 0PSP03-DG-0002 Standby Diesel 12(22) Operability Test 31
: 0PSP03-DG-0003 Standby Diesel 13(23) Operability Test 34
: 0PSP03-DG-0007 Standby Diesel 11(21) LOOP Test 20
: 0PSP03-DG-0008 Standby Diesel 12(22) LOOP Test 18
: 0PSP03-DG-0009 Standby Diesel 13(23) LOOP Test 20
: 0PSP03-DG-0013 Standby Diesel 11(21) LOOP - ESF Actuation Test 20 0PSP03-DG-0014 Standby Diesel 12(22) LOOP - ESF Actuation Test 19 0PSP03-DG-0015 Standby Diesel 13(23) LOOP - ESF Actuation Test 21 0PSP03-ZQ-0028 Operator Logs 98
: 0PSP06-DJ-0001 125 Volt Class 1E Battery 7 Day Surveillance Test 28 0PSP06-DJ-0002 125 Volt Class 1E Battery Quarterly Surveillance Test 19 0PSP06-DJ-0003 125 Volt Class 1E Battery Surveillance Test 13 0PSP06-DJ-0006 Battery Charger 8 Hour Load Verification 19
: 0PSP06-DJ-0007 125 Volt Class 1E Battery Combined Service and Performance Surveillance Test
: 1POP09-AN-03M2 Annunciator Lampbox 1-03M-2 Response Instructions 26 2POP09-AN-03M2 Annunciator Lampbox 2-03M-2 Response Instructions 23
: IP-3.20Q Interdepartmental Procedures 10CFR50.59 Evaluations 4
: 0POP01-ZQ-0022 Plant Operations Shift Routines 52
: 0PSP03-DG-0016 Standby Diesel 11(21) Twenty-Four Hour Load Test 23
: 0PSP06-PK-0001 4.16KV Class 1E Undervoltage Relay Channel Calibration/TADOT-Channel 1
: 0PSP06-PK-0002 4.16KV Class 1E Undervoltage Relay Channel Calibration/TADOT-Channel 2
: 0PSP06-PK-0003 4.16KV Class 1E Undervoltage Relay Channel Calibration/TADOT-Channel 3
: 0PSP06-PK-0004 4.16KV Class 1E Undervoltage Relay Channel Calibration/TADOT-Channel 4
: 0PSP06-PK-005 4.16KV Class 1E Degraded Voltage Relay Calibration/TADOT-Channel 1
: 0PSP06-PK-006 4.16KV Class 1E Degraded Voltage Relay Calibration/TADOT-Channel 2
: 0PSP06-PK-007 4.16KV Class 1E Degraded Voltage Relay Calibration/TADOT-Channel 3
: 0PSP06-PK-008 4.16KV Class 1E Degraded Voltage Relay Calibration/TADOT-Channel 4
: 0P0P02-CH-0005 Essential Chiller Operation 44
: 0P0P03-CH-0001 Essential Chilled Water Pump 11A(21A) Inservice Test 15 0PSP03-EW-0008 Essential Cooling Water Pump 1A(2A) Reference Values Measurement
: 0PSP03-EW-0017 Essential Cooling Water System Train A Testing 24
: 0PSP03-CH-0004 Essential Chilled Water Pump 11A(12A) Reference 7
===Procedures===
: Number Title Revision/Date Values Measurement 0PSP03-SI-0001 Low Head Safety Injection Pump 1A(2A) Inservice Test 13 0PSP03-SI-0004 High Head Safety Injection Pump 1A (2A) Inservice Test 12 0PSP03-SI-0007 Low Head Safety Injection Pump 1A(2A) Reference Values Measurement
: 0PSP03-SI-0010 High Head Safety Injection Pump 1A(2A) Reference Values Measurement
: 0PSP03-SI-0026 High Head Safety Injection Pump Flow Rate Measurement
: 0PSP03-SI-0027 Low Head Safety Injection Pump Flow Rate Measurement
: WAN 127044 Preventive Maintenance 4160V Switchgear E1A CUB 1 0
: WAN 147480 Preventive Maintenance AUX ESF XFMR E1A To 4.16KV ESF BUS E1A
: WAN 177227 PM Inspect Breaker, Unit 2 STBY XFMR TO STBY Bus
: 2G 7
: WAN 220394 PM Inspect Breaker, Unit 2 STBY XFMR To STBY BUS
: 2G 2
: WAN 256076 13.8KV To 4160 VAC ESF Transformer E2B 1
: WAN 257849 480V Load Center E2B 7
: WAN 265895 PM Unit 2 Standby Transformer 1
: WAN 268512 PM Inspect Breaker, To 4.16KV AUX ESF XFMR E2B 8
: WAN 274021 PM Calibrate Relays, To 4.16 KV AUX ESF XFMR E2B 0 
===Miscellaneous Documents===
: Number Title
: Revision/Date
: 4E519EB1108
===Design Basis Document===
: 4.16KV AC Power (PK) System 3 4E53EB1109 Design Basis Document Class 1E AC Power (PL/PM) System 2 4E549EB01110 Design Basis Document, Class 1E Vital 120V AC System 2 4E510EQ1005 Design Criteria Class 1E AC Power Distribution 8 4E520EQ1006 Class 1E 125 Vdc Design Criteria 6 5R289MB1006 Design Basis Document Essential Cooling Water System 5 5S149MB01016 Design Basis Document Auxiliary Feedwater System 5
: 5V119VB01022 Design Basis Document HE/HE (CRE) System 4 5E540EL5031
: Electrical Setpoint Index 3
: STP Interconnection Agreement 8/15/2002
: ERCOT Operating Procedure Manual Transmission & Security Desk
: 8454-00017-KV 4160/480V Transformer E2B1 Nameplate A 8454-00014-KV 4160/480V Transformer E2B2 Nameplate A 
: A-11Miscellaneous Documents Number Title
: Revision/Date
: 8074-01024-WM 13.8KV/4160V Transformer E2B Nameplate A 8074-01004-WM 13.8KV/4160V Transformer E2B Test Data A
: 8394-00037-ZF 25/13.8KV Unit 2 Auxiliary Transformer Nameplate G
: 8045-01007-WB 362.25/13.8KV Standby Transformer 2 Nameplate A
: 8045-01005-WB 362.25/13.8KV Standby Transformer 2 Test Data A
: VTD - W120-250 VTD - Maintenance Program Manual for Safety Related Type DS Low Voltage Metal Enclosed Switchgear
: VTD-B455-0047
: VTD-Installation/Maintenance Instructions for Metal Clad Medium Voltage Power Switchgear Type 5HK
: VTD-B455-0042
: VTD-Installation/Maintenance Instructions Medium Voltage Power Circuit Breakers Type 5HK
: NRC Letter to Mr. Bradford M. Radimer, Chairman, IEEE Battery Working Group
: 01/11/90
: Reply to Notice of Violation 9235-03 Regarding a Failure to Fully Test Essential Chiller ESF Loading Timing
: Sequence 04/02/93
: South Texas Project, Units 1 and 2 - Issuance of Amendments RE: Technical Specification 3/4.8.2 for Batteries and DC Systems (TAC Nos. MD0333 and
: MD0334) 07/20/07 5Q159MB1023
: 4A.1.6 Fluid System Margins, Page 4A-25 3 ANSI N195-1976 American National Standard Fuel Oil Systems for Standby Diesel-Generators
: 04/12/76
: CN-2824 Revise UFSAR Table 9.4-1 "Normal Parameters Temp." to Reflect a "Normal Parameters Temp" Range of 70°F to
: 77°F Instead of the 73°F to 77°F Range Currently Listed
: 2/27/06 DCP# 04-5388 Install Diodes Across Battery Chargers E2C11-1 and E2C11-2 Alarm Relays
: 04/20/04 DCP# 04-6544 Replace Float/Equalize Sw itches of Class 1E Battery Chargers 09/08/05 NSAC 125 Guideline for 10
: CFR 50.59 Safety Evaluations 06/1989
: NUREG-0800 USNRC Standard Review Plan 2
: NUREG/CR-2792 An Assessment of Residual Heat Removal and Containment Spray Pump Performance Under Air and Debris Ingesting Conditions
: 09/1982
: PR-880403 Weekly Surveillance Performed Incorrectly 10/06/88
: PR-910030 Surveillance Performed and Reviewed Incorrectly 02/24/92
: PRA-07-010 Unit 1 ESF
: DG 13 and Unit 2 ESF
: DG 23 TS Surveillance 4.8.1.1.2.e.11 Not Fully Performed
: 10/05/07 Regulatory Guide
: 1.137 Fuel Oil Systems for Standby Diesel Generators 0 Regulatory Guide Fuel Oil Systems for Standby Diesel Generators 1 
: 2Miscellaneous Documents Number Title
: Revision/Date
: 1.137
: SPR 910145 Operability Questionable on Class 1E Batteries 10/12/92
: SPR 921482 Sequence Start Times for Essential Chiller 21B did not Meet Design Value Anticipated Start Times
: 2/11/92 USQE 91-0031 FSAR
: CN-1725 06/07/91 VTD-W120-0678
: AB De-Ion Circuit Breakers Time Current Characteristics Curves for Standard and Mark 75
: Thermal Magnetic Circuit Breakers
: VTD W120-0152 Type MME Magnetic Contactor 0
: VTD-W120-0216 AB De-Ion Circuit Breakers Standard Seltronic Mark 75 and Tri-Pac Designs
: VTD-W120-0300 Qualified Display Processing System (QDPS) 0 R289XG170BHY Essential Cooling Water Induction Motor Data Sheet
: 05/21/82 B03050-0008H4 Vendor Drawing AMTEK 10 KVA Inverter
: A
: B03050-0007H4 Vendor Drawing AMTEK 10 KVA Inverter C
: B03050-0005H4 Vendor Drawing AMTEK 10 KVA Rectifier B
: VTD-A363-0021 VTD - 10 KVA Inverter 3
: 2-E-EM-0822 Configuration Control Package 00
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Revision as of 16:13, 10 November 2018

IR 05000498-07-007 and 05000499-07-007; on 09/24/2007 - 01/22/2008; South Texas Project, Units 1 and 2; NRC Inspection Procedure 71111.21, Component Design Bases Inspection.
ML080450543
Person / Time
Site: South Texas  STP Nuclear Operating Company icon.png
Issue date: 02/13/2008
From: Bywater R L
Region 4 Engineering Branch 1
To: Sheppard J J
South Texas
References
IR-07-007
Download: ML080450543 (35)
v • d • e


Text

February 13, 2008

James J. Sheppard, President and Chief Executive Officer STP Nuclear Operating Company P.O. Box 289 Wadsworth, TX 77483

SUBJECT: SOUTH TEXAS PROJECT ELECTRIC GENERATING STATION, UNITS 1 AND 2 - NRC COMPONENT DESIGN BASES INSPECTION REPORT 05000498/2007007 AND 05000499/2007007

Dear Mr. Sheppard:

On November 26, 2007, the U.S. Nuclear Regulatory Commission (NRC)

completed onsite portions of a component design bases inspection at your South Texas Project Electric Generating Station, Units 1 and 2. The preliminary results were discussed with you and members of your staff on November 26, 2007. After additional in-office inspection, a telephonic exit was conducted on January 22, 2008. The enclosed report documents our inspection findings.

This 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 team reviewed selected procedures and records, observed activities, and interviewed cognizant plant personnel.

The report documents six NRC identified findings, each involving a violation of NRC requirements. All of the findings were evaluated under the risk significance determination process as having very low safety significance (Green). Because of their very low safety significance and because they are entered into your corrective action program, these violations are being treated as noncited violations, consistent with Section VI.A of the Enforcement Policy. If you contest the subject or significance of any of these noncited violations, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001, with copies to the Regional Administrator, U.S. Nuclear Regulatory Commission, Region IV, 611 Ryan Plaza Drive, Suite 400, Arlington, Texas 76011; the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC Resident Inspector at the South Texas Project Electric Generating Station, Units 1 and 2.

STP Nuclear Operating Company - 2 -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 inspection 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).

Sincerely,/RA/

Russell L. Bywater, Chief Engineering Branch 1 Division of Reactor Safety

Dockets: 50-498; 50-499 Licenses: NPF-76; NPF-80

Enclosures:

NRC Inspection Report 05000498/2007007

and 05000499/2007007

w/Attachment:

Supplemental Information

cc w/enclosures:

E. D. Halpin Site Vice President STP Nuclear Operating Company South Texas Project Electric Generating Station P.O. Box 289 Wadsworth, TX 77483

Ken Coates Plant General Manager STP Nuclear Operating Company South Texas Project Electric Generating Station P.O. Box 289 Wadsworth, TX 77483

S. M. Head, Manager, Licensing STP Nuclear Operating Company P.O. Box 289, Mail Code: N5014 Wadsworth, TX 77483

C. T. Bowman STP Nuclear Operating Company - 3 -General Manager, Oversight STP Nuclear Operating Company P.O. Box 289 Wadsworth, TX 77483

Marilyn Kistler Sr. Staff Specialist, Licensing STP Nuclear Operating Company P.O. Box 289, Mail Code 5014 Wadsworth, TX 77483

C. M. Canady City of Austin Electric Utility Department 721 Barton Springs Road

Austin, TX 78704

J. J. Nesrsta/R. K. Temple/

E. Alercon/Kevin Pollo City Public Service Board P.O. Box 1771

San Antonio, TX 78296

Jon C. Wood Cox Smith Matthews 112 E. Pecan, Suite 1800

San Antonio, TX 78205

A. H. Gutterman, Esq.

Morgan, Lewis & Bockius 1111 Pennsylvania Avenue NW Washington, DC 20004

Director, Division of Compliance & Inspection Bureau of Radiation Control Texas Department of State Health Services 1100 West 49th Street

Austin, TX 78756

Brian Almon Public Utility Commission William B. Travis Building P.O. Box 13326 1701 North Congress Avenue

Austin, TX 78701-3326

STP Nuclear Operating Company - 4 -Environmental and Natural Resources Policy Director P.O. Box 12428

Austin, TX 78711-3189

Judge, Matagorda County Matagorda County Courthouse

1700 Seventh Street Bay City, TX 77414

Anthony Jones, Chief Inspector Texas Department of Licensing and Regulation Boiler Program P.O. Box 12157

Austin, TX 78711

Susan M. Jablonski Office of Permitting, Remediation and Registration Texas Commission on Environmental Quality MC-122, P.O. Box 13087

Austin, TX 78711-3087

Ted Enos 4200 South Hulen

Suite 422 Fort Worth, TX 76109

Steve Winn/Christine Jacobs/

Eddy Daniels/Marty Ryan NRC Energy, Inc.

211 Carnegie Center Princeton, NJ 08540

INPO Records Center 700 Galleria Parkway Atlanta, GA 30339-3064

Lisa R. Hammond, Chief Technological Hazards Branch National Preparedness Division FEMA Region VI 800 N. Loop 288

Denton, TX 76209

STP Nuclear Operating Company - 5 -

STP Nuclear Operating Company - 6 -Electronic distribution by RIV: Regional Administrator (EEC)

DRP Director (DDC)

DRS Director (RJC1)

DRS Deputy Director (ACC)

Senior Resident Inspector (JLD5)

Branch Chief, DRP/A (CEJ1)

Senior Project Engineer, DRP/A (TRF)

Team Leader, DRP/TSS (CJP)

RITS Coordinator (MSH3)

DRS STA (DAP)

D. Pelton, OEDO RIV Coordinator (DLP1)

ROPreports STP Site Secretary (HLW1)

SUNSI Review Completed: ___Y__ ADAMS: Yes No Initials: ___WSifre___ Publicly Available Non-Publicly Available Sensitive Non-Sensitive

SRI:EB1 RI:PBC RI:EB1 OE:OB C:EB1 C:PBA C:EB1 WSifre/lm b MChambers SMakor GApger RLBywate r CEJohnson RLBywater /RA/ /RA/ /RA/ /RA/ /RA/ /RA/ /RA/ 1/2/08 12/18/08 1/2/08 1/2/08 2/13/8 2/12/08 2/13/08 STP Nuclear Operating Company - 7 -OFFICIAL RECORD COPY T=Telephone E=E-mail F=Fax

Enclosure - 1 -

SUMMARY OF FINDINGS

IR 05000498/2007007 and 05000499/2007007; September 24, 2007 through January 22, 2008;

South Texas Project Electric Generating Station, Units 1 and 2; NRC Inspection Procedure 71111.21, "Component Design Bases Inspection."

The report covered a 4-week period of onsite inspection and additional in-office inspection performed by six region-based inspectors and two contractors. The inspection identified six Green noncited violations. The significance of most findings is indicated by its color (Green, White, Yellow, Red) using Inspection Manual Chapter 0609, "Significance Determination Process." Findings for which the significance determination process does not apply may be Green or be assigned a severity level after NRC management review. The NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, "Reactor Oversight Process," Revision 3, dated July 2000.

A. NRC - Identified Findings

Cornerstone: Mitigating Systems

Green.

The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," having very low safety significance for the failure to specify in a design calculation allowable relay setpoint tolerances. Specifically, the licensee failed to specify and verify in the relay setpoint calculations the relay setpoint tolerances used in the calibration test procedures. The issue was documented in the corrective action program as Condition Record 07-15443.

The finding was determined to be more than minor because it is associated with the Mitigating Systems cornerstone attribute of "Design Control." It impacts the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events and prevent undesirable consequences. The failure to verify the effects of relay setpoint tolerances on relay coordination time intervals could have resulted in a loss-of-relay coordination and could lead to either a loss of power to safety-related components or lead to a potential for compromising other equipment on a single fault that the relay was designed to isolate. Using Manual Chapter 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations," Phase 1 screening, the finding screened as having very low safety significance (Green) because the condition did not represent a loss of safety function of a system or a train. (Section 1R21.b.1)

Green.

The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," having very low safety significance for the failure to include all potential loads in the standby diesel generator fuel oil sizing calculation. Specifically, the licensee did not account for increased standby diesel

- 2 -generator fuel oil usage resulting from the addition of manual electrical loads during the 7-day mission run time. The licensee entered this finding into their corrective action program as Condition Record 07-15592. The licensee subsequently demonstrated that the spent fuel pool cooling pumps would be the only additional manual loads actually used during the 7 days of operation in the bounding design basis scenario and that there were additional conservative assumptions in the sizing calculation to demonstrate sufficient margin.

The finding was determined to be more than minor because it is associated with the Mitigating Systems cornerstone attribute of "Design Control." It impacts the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events and prevent undesirable consequences. Using Inspection Manual Chapter 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Pow er Situations," Phase 1 screening, the finding screened as having very low safety significance (Green)because it was a design or qualification deficiency confirmed not to result in loss of operability or functionality. (Section 1R21.b.2)

Green.

The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criteria III, "Design Control," of very low safety significance for the failure to translate design basis information into specifications and procedures. Specifically, a non-conservative system pressure was used as an input to an engineering design calculation for the auxiliary feedwater outside containment isolation valves. This finding has been entered into the licensee's corrective action program as Condition Record 07-15455.

The finding was determined to be more than minor because it is associated with the Mitigating Systems cornerstone attribute of "Design Control." It impacts the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events and prevent undesirable consequences. Using Inspection Manual Chapter 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Pow er Situations," Phase 1 screening, the finding screened as having very low safety significance (Green)because it did not represent a loss safety function of a system or a train. (Section 1R21.b.3)

Green.

The team identified a noncited violation of Technical Specification Surveillance Requirement 4.8.1.1.2.E.11, having very low safety significance for the licensee's failure to adequately perform the technical specification surveillance requirement. Specifically, the licensee failed to verify the loading times of the essential chillers in order to verify the automatic load sequence timer was operable. This issue was entered into the licensee's corrective action program as Condition Records 07-14903 and 07-14959.

The finding was determined to be more than minor because it is associated with the Mitigating Systems cornerstone attribute of "Design Control." It impacts the

- 3 -cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events and prevent undesirable consequences. Using Inspection Manual Chapter 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Pow er Situations," Phase 1 screening, the finding screened as having very low safety significance (Green)because it did not represent a loss of safety function of a system or a train.

(Section 1R21.b.4)

Green.

The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XI, "Test Control," having very low safety significance for the licensee's failure to implement a test program to assure that all installed safety-related molded case circuit breakers will perform satisfactorily in service.

Specifically, the licensee had not adequately exercised or subjected to periodic testing all of the 125V dc molded case circuit breakers since initial plant operation. The licensee entered the finding into their corrective action program as Condition Record 07-15817.

The finding was determined to be more than minor because it is associated with the Mitigating Systems cornerstone attribute of "Equipment Performance." It impacts the cornerstone objective of ensuring the availability, reliability, capability of systems that respond to initiating events and prevent undesirable consequences. Using Inspection Manual Chapter 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations," Phase 1 screening, the finding screened as having very low safety significance (Green) because it did not result in a loss of safety function of a system or a train. (Section 1R21.b.5)

Green.

The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criteria III, "Design Control," of very low safety significance for the failure to adequately translate design basis information into specifications and procedures. Specifically, measurement instrument uncertainties were not included in the determination of minimum allowed high head safety injection pump and low head safety injection pump developed head values used during periodic technical specification surveillance testing. The licensee entered the finding into their corrective action program as Condition Record 07-15752.

The finding was determined to be more than minor because it is associated with the Mitigating Systems cornerstone attribute of "Design Control." It impacts the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events and prevent undesirable consequences. Using Inspection Manual Chapter 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Pow er Situations," Phase 1 screening, the finding screened as having very low safety significance (Green)because it did not result in a loss of safety function of a system or a train.

(Section 1R21.b.6)

- 4 -B. Licensee-Identified Findings

None.

- 5 -U.S. NUCLEAR REGULATORY COMMISSION REGION IV

Dockets: 05000498, 05000499

Licenses:

NPF-76, NPF-80

Report: 05000498/2007007; 05000499/2007007

Licensee:

STP Nuclear Operating Company

Facility:

South Texas Project Electric Generating Station, Units 1 and 2

Location FM 521 - 8 miles west of Wadsworth Wadsworth, Texas 77483

Dates: September 24, 2007 through January 22, 2008k

Inspectors:

W. Sifre, Senior Reactor Inspector, Engineering Branch 1 M. Chambers, Resident Inspector, Branch C B. Henderson, Reactor Inspector, Engineering Branch 1 S. Makor, Reactor Inspector, Engineering Branch 1 S. Rutenkroger, Reactor Inspector, Engineering Branch 1 G. Apger, Operations Engineer, Operations Branch

Contractors:

H. Anderson, Mechanical Contractor J. Chiloyan, Electrical Contractor

Approved By:

Russell L. Bywater, Chief Engineering Branch 1 Division of Reactor Safety

REPORT DETAILS

REACTOR SAFETY

Inspection of component design bases verifies the initial design and subsequent modifications and provides monitoring of the capability of the selected components and operator actions to perform their design bases functions. As plants age, their design bases may be difficult to determine and an important design feature may be altered or disabled during a modification. The plant risk assessment model assumes the capability of safety systems and components to perform their intended safety function successfully. This inspectable area verifies aspects of the Initiating Events, Mitigating Systems and Barrier Integrity cornerstones for which there are no indicators to measure performance.

In addition to performing the baseline inspection, the team reviewed actions taken by the licensee in response to previously identified significant issues associated with engineering performance.

1R21 Component Design Bases Inspection

Main article: IP 71111.21

The team selected risk-significant components and operator actions for review using information contained in the licensee

=s probabilistic risk assessment. In general, this included components and operator actions that had a risk achievement worth factor greater than two or Birnbaum importance value greater than 1E-6.

a. Inspection Scope

To verify that the selected components would function as required, the team reviewed design basis assumptions, calculations, and procedures. In some instances, the team performed independent calculations to verify the appropriateness of the licensee engineers' conclusions. The team also verified that the condition of the components was consistent with the design bases and that the tested capabilities met the required criteria.

The team reviewed maintenance work records, corrective action documents, and industry operating experience information to verify that licensee personnel considered degraded conditions and their impact on the components. For the review of operator actions, the team observed operators during simulator scenarios associated with the selected components, as well as observing simulated actions in the plant.

The team performed a margin assessment and detailed review of the selected risk-significant components to verify that the design bases have been correctly implemented and maintained. This design margin assessment considered original design issues, margin reductions due to modification, 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 included items such as failed

- 7 -performance test results; significant corrective actions; repeated maintenance; 10 CFR 50.65(a)1 status; operable, but degraded, conditions; NRC re sident 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.

The components selected for review were:

  • 345/138,13.8kV Standby Transformer ST002A
  • 13.8kV/4/16 Auxiliary Engineered Safety Feature Transformer E2B
  • 4.16kV Engineered Safety Feature BUS E2B
  • Standby Diesel Generator 22
  • 4.16kV/480 V Load Center Transformer E2B
  • 480V Load Center E2B2
  • 125V DC Battery and Charger Train B
  • Electrical Auxiliary Building HVAC
  • 10KVA Inverter E1V 2201
  • High Pressure Safety Injection Pump 2A
  • Low Pressure Safety Injection Pump 2A
  • Refueling Water Storage Tank
  • Essential Cooling Water Pump 2A
  • Essential Chilled Water Pump 2A The selected operator actions were:
  • Opening electrical auxiliary building doors and start of smoke purge on loss of ventilation to switchgear rooms.
  • Diagnosis of a steam generator tube rupture to start appropriate procedures.
  • Starting auxiliary feedwater if engineered safety features actuation system fails during a control room fire.

The operating experience issues were:

- 8 -

  • NRC IN 2007-09, "Equipment Operability Under Degraded Voltage Conditions."
  • NRC IN 2006-18, "Significant Loss of Safety-Related Electrical Power at Forsmark, Unit 1, in Sweden."

b. Findings

b.1. Failure to Specify Setpoint Calibration Limits in Relay Setpoint Calculations

Introduction.

The team identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," for the failure to specify in a design calculation the allowable relay setpoint tolerances stated in the licensee's relay setpoint calibration test procedures. Under postulated electrical fault or overload conditions, the lack of adequate relay coordinating time intervals between relay operating characteristics would lead to spurious tripping and to either a loss of power to safety-related components or lead to a potential for compromising other equipment on a single fault that the relay was designed to isolate.

Description.

During the review of licensee's completed protective relay trip setpoint calibration test procedures, relay setting records and relay setting calculations to verify whether the applied relay settings were consistent with the designed basis calculations, the team noted that the acceptance criteria for the allowable values of relay setpoints stated in calibration test Procedures PM EM-2-03000814, WAN 274021 and relay setting sheets were neither specified nor verified in the design basis relay setting Calculation EC-5029, "4.16kV Switchgear Relay Setting." Following discovery, the licensee performed a preliminary evaluation for affected components using the worst-case scenario of relay setpoint tolerances stated on the relay setting records and concluded that the affected components would still perform their required safety functions in the event of an electrical fault. The issue was documented in licensee's corrective action program as Condition Record 07-15443.

Analysis.

The licensee's failure to specify relay setpoint tolerances and verify the effects on coordination margin in relay setpoint calculations for relays used on 4.16kV emergency safety feature switchgears was a performance deficiency. The finding was determined to be more than minor because it is associated with the Mitigating Systems cornerstone attribute of "Design Control." It impacts the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating

- 9 -events and prevent undesirable consequences. The failure to verify the effects of relay setpoint tolerances on relay coordination time intervals could have resulted in a loss-of-relay coordination and could lead to either a loss of power to safety-related components or lead to a potential for compromising other equipment on a single fault that the relay was designed to isolate. Using Manual Chapter 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations", Phase 1 screening, the finding screened as having very low safety significance (Green) because the condition had not resulted in a loss of safety function of a system or a train. This finding was reviewed for crosscutting aspects and none were identified.

Enforcement.

Part 50 of Title 10 of the Code of Federal Regulations, Appendix B, Criterion III, "Design Control," requires, in part, that design control measures provide for verifying or checking the adequacy of design, such as by the performance of design reviews, by the use of alternate or simplified calculational methods, or by performance of a suitable testing program.

Contrary to the above, the licensee's design control measures failed to either specify the relay setpoint tolerances or verify the adequacy of the design for safety-related 4160V electrical distribution system to ensure that the trip settings of the protective relays were adequate to ensure selective tripping in the event of a fault. Specifically, the team identified that the licensee failed to specify and verify in the relay setpoint calculations the relay setpoint tolerances used in the calibration test procedures. Because this violation was of very low safety significance and has been entered into the licensee's corrective action program as Condition Record 07-15443, it is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy: NCV 05000498; 499/2007007-01, Failure to Specify Setpoint Calibration Limits in Relay Setpoint Calculations.

b.2. Failure to Consider Manual Loads for Fuel Oil Storage Tank Sizing Calculation

Introduction.

The team identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," for the licensee's failure to account for manual electrical loads in determining fuel oil usage during the standby diesel generators' seven day mission time for the fuel oil storage tank sizing calculation.

Description.

The Final Safety Analysis Report, Revi sion 0, stated that the fuel oil storage tanks were sized to have sufficient capacity to provide for continuous operation of the diesel generators for 7 days at their continuous rating, (i.e., 5935 kW). The licensee revised the Updated Final Safety Analysis Report (UFSAR) on December 9, 1992, to replace the loading at the standby diesel generator continuous rating with the "engineered safety features load requirements." However, the documented review contained in Unreviewed Safety Question Evaluation 91-0031 and Calculation MC-6256, "Sizing of SDG FOST," Revision 0, both discussed including all the non-engineered safety features loads listed in UFSAR, Table 8.3-3, as part of the fuel and storage tank sizing requirement.

- 10 -In particular, the Unreviewed Safety Question Evaluation 91-0031 stated, "This [including all the listed non-engineered safety features loads] is in accordance with the ANSI N195 Standard which states, 'If the design includes provision for an operator to supply power to equipment other than the minimum required for the plant condition, such additional load(s) shall be included in the calculation of required fuel oil storage capacity." Regulatory Guide 1.137, "Fuel Oil Systems for Standby Diesel Generators,"

Revision 1, dated October 1979, refers to the requirements described in ANSI N195-1976, "Fuel Oil Systems for Standby Diesel-Generators," to be a method acceptable to the NRC staff for complying with the Commission's regulations regarding diesel fuel oil systems for standby diesel generators and assurance of adequate diesel fuel oil quality. The safety evaluation report originally prepared for South Texas Project Electric Generating Station used ANSI N195 as the standard to evaluate the acceptability of the fuel oil storage tank design and sizing.

Since the UFSAR, as revised, did not discuss the additional manual loads, which must be considered in order to evaluate the fuel oil storage tank sizing, Calculation MC-6256, Revision 0, was ultimately re vised in Revision 3, dated October 3, 1996, to remove consideration of all manual loads. Therefore, beginning with that revision the design basis non-conservatively removed consideration of expected actual plant operations with respect to manual loads during the bounding design basis accident analysis.

The team interviewed engineering and operations personnel in order to determine what equipment from UFSAR, Table 8.3-3, would be supplied power other than the minimum required for the plant condition. These interviews revealed a range of possible equipment, which could be utilized since the operations philosophy would be to exceed the minimum required for the plant condition in order to place the plant in as safe a condition as possible. The upper range of potential manually loaded equipment would have resulted in exceeding the minimum technical specification fuel oil volume requirement of 60,500 gallons during the 7-day mission time of the standby diesel generators during the worst-case design basis accident considered. However, in further discussions, licensee personnel balanced the operations philosophy with the 7-day fuel oil requirements considered as part of the design basis event and concluded the spent fuel pool cooling pumps would be the only additional manual loads utilized during the bounding scenario. The fuel oil storage tank sizing calculation included additional conservative assumptions regarding expected pump operation during design basis accident scenarios. For example, the calculation assumed auxiliary feedwater, high head safety injection, and containment spray pumps would be run continuously for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> following a large break loss of coolant accident. Therefore, the licensee recalculated the fuel oil storage tank sizing using more realistic assumptions with respect to load profile and determined sufficient fuel oil margin does exist with all design basis conditions considered. This issue was entered into the licensee's corrective action program as

Condition Record 07-15592.

Analysis.

The team determined that the failure to account for manual electrical loads in determining fuel oil usage during the standby diesel generators' 7-day mission time for the fuel oil storage tank sizing calculation was a performance deficiency. The finding

- 11 -was determined to be more than minor because it is associated with the Mitigating Systems cornerstone attribute of "Design Control." It impacts the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events and prevent undesirable consequences. Not accounting for the additional manual loads increases the likelihood that the required inventory of fuel oil for a 7-day mission time would not be available.

Using Inspection Manual Chapter 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations," Phase 1 screening, the finding screened as having very low safety significance (Green) because it was a design or qualification deficiency confirmed not to result in loss of operability or functionality.

This finding was reviewed for crosscutting aspects and none were identified.

Enforcement.

Part 50 of Title 10 of the Code of Federal Regulations, Appendix B, Criterion III, "Design Control," requires, in part, that measures shall be established to assure that applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures, and instructions.

Contrary to the above, the licensee had not correctly translated design basis information into the standby diesel generator fuel oil tank sizing analysis. Specifically, the licensee failed to translate the loading and usage associated with additional manual loads, reasonably expected to be utilized during the bounding design basis accident, into Calculation MC-6256, Revision 4. Because this violation was of very low safety significance and has been entered into the licensee's corrective action program as Condition Record 07-15592, it is being treated as a noncited violati on consistent with Section VI.A of the NRC Enforcement Policy: NCV 05000498; 499/2007007-02, Manual Loads not Considered for Fuel Oil Storage Tank Sizing Calculation.

b.3. Failure to Use Correct Design Inputs in Determination of the Weak Link for the Auxiliary Feedwater System Outside Containment Isolation Motor Operated Valves.

Introduction.

The team identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criteria III, "Design Control," for the failure to translate design basis information into specifications and procedures. Specifically, the team identified that a non-conservative system pressure was used as an input to the engineering design calculation for the auxiliary feedwater outside containment isolation valves (MC-6204 Document Change Notice MC-145 issued 7/31/1992, Revise Motor-Operator Valve Thrust and Torque Calculation for AF-19, AF-48, AF-65, and AF-85).

Description.

The team identified that the pressure loading calculation in the motor-operated valve weak link analysis for the auxiliary feedwater outside containment isolation valves used a system pressure of 1250 psig. This value was based on the steam generator power-operated relief valves in the main steam systems being normally set at 1225 psig for normal operation and an additional 25 psig was added to the nominal steam generator power-operated relief valve set point to allow for any set point uncertainty.

- 12 -This did not take into account accident conditions that result in the backpressure from the main steam system being greater than 1250 psig.

In response to the team's questions that the pressure could be greater than 1250 psig, the licensee issued Condition Record 07-15455-4, "Discussion Paper; Re-perform Weak Link Calculation at 1324 psid and 200°F," received October 24, 2007; and Condition Record 07-15455, "Discussion Paper; Weak Link Discussion of Motor-Operated Valves During Normal and Accident Operation," received October 15, 2007. The licensee determined that an increase in steam generator pressure greater than normal operating pressure would occur during certain design bases accident conditions. The appropriate input to the calculation was determined to be a steam generator pressure of 1324 psig, which allows for a 1 percent margin for setting tolerance and 2 percent for pressure drop in the piping connecting the safety valves to the steam generator from the lowest safety valve set point of 1285 psig. With the revised 1324 psig value and the original assumed valve temperature of 200°F the new weak link calculation resulted in two of the eight auxiliary feedwater outside containment isolation valves (one in each unit) having a torque switch setting that exceeded the weak link calculated set point in the close direction. The weak link for these valves is the valve seat. Valve thrust plus system pressure exceeding the valve seat strength could result in thrusting the valve disc into the seat and failure of the valve.

The licensee subsequently provided the following justification for the operability of the valves using the 1324 psid accident pressure. "From a review of all accidents that result in an increase in Steam Generator pressures also result in the starting of the auxiliary feedwater pumps. The auxiliary feedwater system water supply has a design temperature range of 32°F to 120°F. Single failure criteria states that one of the auxiliary feedwater pumps may not start, however it is NOT creditable for a pump to not start and to have sufficient back leakage to raise the temperature of the outside containment isolation valve to 200°F at the same time. Therefore, the maximum abnormal temperature is 170°F." The licensee determined that the weak link calculation at 1324 psid and 170°F results in adequate margin between current thrust settings of all eight auxiliary feedwater outside containment isolation valves and the calculated weak link stresses of the valve seats to assure operability under accident conditions.

Analysis.

The failure to use a conservative design input in the engineering analysis was a performance deficiency. The finding was determined to be more than minor because it is associated with the Mitigating Systems cornerstone attribute of "Design Control." It impacts the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events and prevent undesirable consequences. Using Inspection Manual Chapter 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations," Phase 1 screening, the finding screened as having very low safety significance (Green) because it did not represent a loss of safety function of a system or a train. This finding was reviewed for crosscutting aspects and none were identified.

- 13 -Enforcement. Part 50 of Title 10 of the Code of Federal Regulations, Appendix B, Criterion III, "Design Control," states, in part, that measures shall be established to assure that design basis are correctly translated into specifications and procedures. Contrary to the above, in Calculation MC-145, the licensee did not use a conservative pressure input necessary to prevent damage to auxiliary feedwater outside containment isolation valves during a design basis event. Because this violation was of very low safety significance and has been entered into the licensee's corrective action program as Condition Record 07-15455), it is being treated as a noncited viola tion consistent with Section VI.A of the NRC Enforcement Policy: NCV 05000498; 499/2007007-03, Failure to Use Correct Design Inputs in Determination of the Weak Link for the Auxiliary Feedwater System Outside Containment Isolation Motor Operated Valves.

b.4. Surveillance Procedure Lacked Check for Timing of Chiller Loading on the Bus

Introduction.

The team identified a Green noncite d violation of Technical Specification Surveillance Requirement 4.8.1.1.2.E.11, for the licensee's failure to adequately perform the technical specification surveillance requirement. Specifically, the licensee failed to verify the loading times of the essential chillers in order to verify the automatic load sequence timer was operable.

Description.

Technical Specification Surveillance Requirement 4.8.1.1.2.E.11 requires "Verifying that the automatic load sequence timer is OPERABLE with the first sequenced load verified to be loaded between 1.0 second and 1.6 seconds, and all other load blocks within +/- 10% of its design interval." The team requested to review the strip chart data recorded from the surveillance tests that demonstrated this surveillance requirement had been performed satisfactorily. Licensee personnel recognized, however, that the actual loading times referenced in the surveillance requirement had not been included in the measurements. Procedure 0PSP02-SF-001A, "ESF Diesel Sequencer Timing Test Train A," Revision 11 (Trains B and C similar), only tests the time that the sequence timer demands breaker closure and does not measure and/or record the actual load times.

The licensee entered Technical Specification 4.0.3 for all three trains of standby diesel generators for both units, allowing 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to fully perform the surveillances successfully. By reviewing the strip chart recorder data for the last loss-of-offsite power and loss-of-offsite power with emergency safety features actuation testing of the standby diesel generators, the licensee verified Technical Specification Surveillance Requirement 4.8.1.1.2.E.11 was successfully met for Standby Diesel Generators 11, 21, and 22. In the case of Standby Diesel Generators 13 and 23, the recorded information had a time resolution loss due to switching of recording speeds during the test. The licensee performed a risk evaluation to delay the complete performance of the surveillance test until the next scheduled time (the next outage scheduled for Spring 2008). The team reviewed this assessment and agreed with its conclusions since the data that was available fully supported the equipment being able to perform its safety function.

- 14 -However, for Standby Diesel Generator 12, a review of the strip chart data revealed that Essential Chiller 12B had loaded on the bus at 168 seconds versus the design interval of 270 seconds. This condition had not been discovered in prior surveillance testing because Procedure 0PSP02-SF-001A did not contain instructions to verify the timing of relays outside of the sequence timer itself.

The licensee declared Standby Diesel Generator 12 inoperable at 09:45 on October 5, 2007, entering Technical Specification 3.8.1.1, Actions B and D.

The cause of the timing discrepancy was isolated to a 35 second blocking circuit external to the chiller that would not prevent the chiller from performing its design safety function. As such, the safety functions of the sequence timer, the standby diesel generator, and Essential Chiller 12B were not adversely affected by the condition, nor would those safety functions be impacted by starting/loading times of the essential chillers between 65 and 270 seconds. The licensee revised the design documents referencing the loading time of the essential chillers to be between 65 and 270 seconds. Once completed, the surveillance testing was declared successful, and the licensee declared Standby Diesel Generator 12 operable at 18:25 on October 11, 2007. This issue was entered into the licensee's corrective action program as Condition Records 07-14903 and 07-14959.

Analysis.

The team determined that the failure to adequately perform Technical Specification Surveillance Requirement 4.8.1.1.2.E.11 was a performance deficiency. The finding was determined to be more than minor because it is associated with the Mitigating Systems cornerstone attribute of "Design Control." It impacts the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events and prevent undesirable consequences. Not fully performing the required surveillances increases the likelihood that the standby diesel generators and supported equipment would not perform their design safety functions when needed. Using Inspection Manual Chapter 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations," Phase 1 screening, the finding screened as having very low safety significance (Green) because the finding did not represent a loss of safety function of the sequence timer, standby diesel generator, or the essential chiller. This finding was reviewed for crosscutting aspects and none were identified.

Enforcement.

Technical Specification Surveillance Requirement 4.8.1.1.2.E.11 requires "Verifying that the automatic load sequence timer is OPERABLE with the first sequenced load verified to be loaded between 1.0 second and 1.6 seconds, and all other load blocks within +/- 10% of its design interval." Contrary to the above, the licensee failed to verify the actual loading times of the sequenced loads. Specifically, the licensee only verified the time that the sequence timer demands breaker closure and did not perform the "verified to be loaded" requirement. Because the violation was of very low safety significance and has been entered into the licensee's corrective action program as

Condition Records 07-14903 and 07-14959, it is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy: NCV 05000498;

- 15 -499/2007007-04, Surveillance Procedure Lacked Check for Timing of Chiller Loading on the Bus.

b.5. Inadequate Test Program for 125V DC Molded Case Circuit Breakers

Introduction.

The team identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion XI, "Test Control," for the failure to implement a test program to assure that all installed safety-related molded case circuit breakers will perform satisfactorily in service. Specifically, the licensee had not adequately exercised or subjected to periodic testing all of the 125V dc molded case circuit breakers since initial plant operation.

Description.

During the review of surveillance tests for the Auxiliary Feedwater Motor-Operated Valve 0019, the team discovered that the molded case circuit breaker had not been exercised or subjected to testing since the initial plant operation. In addition, further inspection discovered that the majority of 125V dc-fed molded case circuit breakers were also not exercised or subjected to periodic testing since installation in 1986. The types of molded case circuit breakers undergoing any type of preventative testing/maintenance included battery chargers, distribution panels, and inverters since they were infrequently cycled by other maintenance activities. Conversely, the breakers that fed loads to standby diesel generator field flash, reactor trip switchgear, 4.16kV switchgear control power and emergency safety feature load sequencers appeared to have not been tested since it was assumed that they were cycled in other maintenance activities.

The team noted that the licensee performed tests on molded case circuit breakers to satisfy Information Notice IEN 93-64 and ensure that molded case circuit breakers installed remained functional during plant operations. Following the test was an engineering evaluation acknowledging that molded case circuit breakers were subject to potential age-related degradation, which could result in a failure to trip in accordance with the published time-current characteristic curves because of various factors, such as grease hardening. In 2001, the licensee decided that the sample size for the dc-fed loads

indicated that limited failures in the test population did not warrant a pre-established test program. Essentially, credit was taken for circuit breakers being cycled as a part of other maintenance programs, but it was realized that these tests performed on breakers, failed to actually cycle the breaker. In fact the handswitch was used to open and close the valve.

- 16 -Updated Final Safety Analysis Report, Section 8.3.2.1.4, provides for "Periodic testing Class 1E dc power system equipment is performed in accordance with Regulatory Guide 1.32 to verify its ability to perform its safety function." Information Notice 93-64, "Periodic Testing and Preventative Maintenance of Molded Case Circuit Breakers,"

stated, "Detecting or assessing degradation could only be accomplished through appropriate periodic testing and monitoring." The team found that the licensee's evaluation and approach to the industry experience, design life, potential common mode failures, and component age concerns were not addressed in the test program. The licensee entered this finding into their corrective action program as Condition Record 07-15817.

Analysis.

The team determined that the lack of periodic testing on all of the dc molded case circuit breakers was a performance deficiency. The finding was determined to be more than minor because it is associated with the Mitigating Systems cornerstone attribute of "Equipment Performance." It impacts the cornerstone objective of ensuring the availability, reliability, capability of systems that respond to initiating events and prevent undesirable consequences. Using Inspection Manual Chapter 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Power Situations,"

Phase 1 screening, the finding screened as having very low safety significance (Green)because it did not result in a loss of safety function of a system or train. This finding was reviewed for crosscutting aspects and none were identified.

Enforcement.

Part 50 of Title 10 of the Code of Federal Regulations

, Appendix B, Criterion XI, "Test Control," stated, in part, that test programs shall be established to assure that all testing required to demonstrate that structures, systems and components will perform satisfactorily in service. Contrary to the above, the licensee failed to implement a test program to assure all installed safety-related molded case circuit breakers will perform satisfactorily in service. Because this violation was of very low safety significance and has been entered into the licensee's corrective action program as Condition Record 07-15817, it is being treated as a noncited violati on consistent with Section VI.A of the NRC Enforcement Policy: NCV 05000498; 499/2007007-05, Inadequate Test Program for 125V DC Molded Case Circuit Breakers.

b.6. Failure to Incorporate Instrument Uncertainties into Surveillance Requirements for Technical Specification Limiting Condition for Operation 3.5.2 (Surveillance Requirement 4.5.2.f)

Introduction.

The team identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criteria III, "Design Control," for the failure to adequately translate design basis information into specifications and procedures. Specifically, measurement instrument uncertainties were not included in the determination of minimum allowed high head safety injection pump and low head safety injection pump developed head values during periodic technical specification surveillance testing.

- 17 -Description. Technical Specification Limiting Condition for Operation 3.5.2, Surveillance Requirement 4.5.2.f.1 for the high head safety injection pump, and Surveillance Requirement 4.5.2.f.2 for the low head safety injection pump require:

For the High Head Safety Injection pumps, verification that the pump develops a differential pressure on recirculation flow when tested pursuant to Technical Specification 4.0.5 greater than or equal to 1480 psid.

For the Low Head Safety Injection pumps, verification that the pump develops a differential pressure on recirculation flow when tested pursuant to Technical Specification 4.0.5 greater than or equal to 286 psid.

Upon review of Surveillance Procedures 0PSP03-SI-0001 and 0PSP03-SI-0004, the team identified that the pump developed head acceptance criteria in the procedures did not include consideration of measurement instrument uncertainties and were numerically equal to the technical specification values. As a result, there was no documented assurance that the recorded current and historical surveillance test results would demonstrate pump developed heads above the required minimum technical specification requirements when measurement instrument uncertainties were taken into consideration. Therefore, the technical specification surveillance test acceptance criteria were non-conservative.

The team reviewed Design Basis Document 5Z529ZB01025, "Technical Specification/ Limiting Conditions for Operation Design Basis Document," Revision 2, and determined that it had erroneously stated for both high-head safety injection pumps and low-head safety injection pumps that "This value is a conservative, nominal value and needs no additional instrument uncertainty margin. This value is acceptable for use. This value is only used in this application (Technical Specifications 4.5.2.f.1 and Technical Specification 4.5.2.f.2)."

The licensee issued Condition Record 07-15752. The condition record stated that "The pump test procedures currently use the technical specification values as the low limit for operability and should be revised. The most recent performance of all safety injection pumps meets the "upward adjusted" low limits."

Analysis.

The failure to include consideration of measurement instrument uncertainties, in relation to the instrumentation utilized in periodic surveillance tests to measure the pump developed head, into the technical specification surveillance test acceptance criteria was a performance deficiency. The finding was determined to be more than minor because it is associated with the Mitigating Systems cornerstone attribute of "Design Control." It impacts the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events and prevent undesirable consequences. Using Inspection Manual Chapter 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for At-Pow er Situations," Phase 1 screening, the finding screened as having very low safety significance (Green) because it

- 18 -did not represent a loss of safety system function. This finding was reviewed for crosscutting aspects and none were identified.

Enforcement.

Part 50 of Title 10 of the Code of Federal Regulations, Appendix B, Criterion III, states, in part, that measures shall be established to assure that design bases are correctly translated into specifications and procedures. Contrary to the above, the licensee did not conservatively account for the effect of instrument uncertainty in development of acceptance criteria for the technical specification surveillance values for Technical Specification Limiting Condition for Operation 3.5.2. Thus, the minimum allowed high head safety injection and low head safety injection pump developed head had not been definitively demonstrated during surveillance testing to exceed the minimum Technical Specification limiting condition for operation values. Because this violation was of very low safety significance and has been entered into the licensee's corrective action program as Condition Record 07-15752, it is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy: NCV 05000498; 499/2007007-06, Failure to Incorporate Instrument Uncertainties into Surveillance Requirements for Technical Specification Limiting Condition for Operation 3.5.2 (Specifically Surveillance Requirement 4.5.2.f).

OTHER ACTIVITIES

4OA5 Other Activities

a.1 Unresolved Item Associated with the Effect of Standby Diesel Generator Technical Specification Voltage Variation on Supplied Equipment

Introduction.

The team identified an unresolved item associated with the steady state output voltage supplied by the standby diesel generators is allowed to vary by Technical Specification 3/4.8.1 from 3744 V to 4576 V (+/- 10%) during a loss of offsite power event. Specifically, the licensee has not analyzed for the effect of this full variation.

Description.

The design analysis assumed maximum supplied voltage variations based upon offsite power supplies which were analyzed to vary less than the technical specification allowed steady state variation for the standby diesel generators. Components throughout the plant would be adversely affected by either an undervoltage or overvoltage condition.

Since this is a very broad issue that encompasses components powered from the standby diesel generator during a design basis event, the licensee will require significant time to evaluate its effects. Although available safety margins will be less, the degree of this effect is not yet known since the effect of the variation varies upon the analyzed parameter and currently analyzed margins vary significantly. The actual safety function of equipment is not expected to be compromised since the standby diesel generators are presently controlled to a tighter band of voltage operation than allowed by technical specifications and review of the surveillance testing of the standby diesel generators confirms this tighter band is currently being maintained.

- 19 - Once the licensee has evaluated the effect of the allowed steady state voltage variation and determined the degree of safety margin impact throughout the plant, the NRC can complete the inspection of that analysis in order to close this issue. The licensee has

documented this issue in Condition Record 07-15554 and the item is unresolved pending the licensee's completion of its analysis and NRC review: URI05000498; 499/2007007-07, Effect of Standby Diesel Generator Technical Specification Voltage Variation on Supplied Equipment.

a.2 Unresolved Item Involving Combined Adverse Conditions not considered in Fuel Oil Storage Tank Sizing

Introduction.

The team identified an unresolved item involving accounting for the combined effect of vortexing and standby diesel generator frequency variations in fuel oil storage tank sizing.

Description.

Calculation MC-6256, "Sizing of Standby Diesel Generator Fuel Oil Storage Tank," Revision 4, determined a total 7-day fuel oil requirement of 51,500 gallons, comparing this value with a technical specification requirement of 60,500 gallons. However, this calculation did not consider the effects of vortexing or generator frequency variations. Condition Record 97-14434-10 included an evaluation of fuel oil vortexing completed as part of a "Review of Safety Related Tanks (other than Refueling Water Storage Tank & Auxiliary Feedwater Storage Tank) for Vortexing Concerns." Separately, Calculation EC-5100, Standby Generator Transient Response Model," Revision 2, contained an evaluation performed under Condition Record 97-13089-1 in order to "Perform Evaluation of Electrical Frequency Variations on Mechanical Fluid Systems."

The vortexing evaluation determined that 13.5 inches of fuel oil volume would be susceptible to excessive air entrainment, representing 4120 gallons of unusable fuel oil with a 7-day fuel oil requirement of 55,360 gallons (referencing Calculation MC-6038, "Standby Diesel Generator Fuel Oil Storage Tank Level Setting Calculation." The total required volume would therefore be 59,480 gallons.

The frequency effects evaluation determined that "Standby Diesel Generator load would increase by roughly 6% because the majority of load consists of pumps and fans with primarily friction system loads." The evaluation then compared this 6 percent increase in load with the standby diesel generator fuel oil storage tank calculated margin of more than 10 percent.

However, the vortexing evaluation had already effectively reduced the majority of the analyzed margin with a remaining 1020 gallons of fuel oil between 59,480 gallons and the technical specification requirement of 60,500 gallons. Therefore, applying a 6 percent increase in standby diesel generator load in addition to considering vortexing effects would have exceeded the technical specification requirement under those analyzed conditions.

- 20 - In addition, the most recent fuel oil storage tank sizing calculation determined a 7 day fuel oil requirement of 51,500 gallons. As discussed in the finding "Manual Loads not Considered for Fuel Oil Storage Tank Sizing Calculation," this requirement neglected manual loads during the 7 days for which provision would be made to use during a design basis event. A bounding analysis considering the actual anticipated manual loads, in addition to the vortexing reduction and increased load frequency effect, exceeds the minimum technical specification requirement.

The fuel oil storage tank sizing calculation included additional conservative assumptions regarding expected pump operation during design basis accident scenarios. For example, the calculation assumed auxiliary feedwater, high-head safety injection, and containment spray pumps would be run continuously for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> following a large break loss-of-coolant accident. The licensee recalculated the fuel oil storage tank sizing using more realistic assumptions with respect to load profile and determined sufficient fuel oil margin does exist with all design basis conditions considered. This issue was entered into the licensee's corrective action program as Condition Request 07-14398 and 07-15592.

After further discussions with staff from the NRC Office of Nuclear Reactor Regulation, the team concluded that this issue of failure to account for the combined effect of vortexing and standby diesel generator frequency variation in the fuel oil storage tank sizing would remain open as an unresolved item. Additional NRC staff review was necessary to determine whether the issue was acceptable, whether it was a finding, or whether it constituted a deviation or violation. Pending completion of this review, this item is unresolved: URI 05000498; 499/2007007-08, Combined Adverse Conditions not Considered in Fuel Oil Storage Tank Sizing.

4OA6 Meetings, Including Exit

On October 26, 2007, the team leader presented the preliminary inspection results to Mr. E. Halpin, Site Vice President, and other members of the South Texas Project staff. After additional offsite and onsite inspection a preliminary exit meeting was conducted on November 26, 2007, with Mr. J.

Sheppard, President and Chief Executive officer and other members of the licensee's staff. After additional in-office inspection, a telephonic exit was conducted on January 22, 2008. The licensee acknowledged the findings during each meeting. While some proprietary information was reviewed during this inspection, no proprietary information was included in this report.

A-1

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee personnel

C. Bowman, General Manager Oversight
K. Coats, Plant General Manager
R. Engen, Manager, Maintenance Engineering
E. Halpin, Site Vice President
S. Head, Manager, Licensing
K. House, Manager, Design Engineering
B. Jenewein, Manager, Testing/Programs Engineering
R. Lovell, Manager, Industrial Alliances
M. Meier, General manager Station Support
J. Mertink, Manager, Operations
M. Murray, Manager, Systems Engineering
G. Powell, Manager, Site Engineering
D. Rencurrel, Vise President, Engineering
M. Ruvalcaba, Supervisor, Engineering
J. Sheppard, President and Chief Executive Officer
D. Towler, Manager, Quality

NRC personnel

W. Jones, Chief, Engineering Branch 1
J. Dixon, Senior Resident Inspector

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened

URI05000498; 499/2007007-07 URI Effect of Standby Diesel Generator Technical Specification Voltage Variation on Supplied

Equipment URI05000498; 499/2007007-08 URI Combined Adverse Conditions not Considered in Fuel Oil Storage Tank Sizing

Opened and Closed

NCV05000498; 499/2007007-01 NC

V Failure to Specify Setpoint Calibration Limits in Relay Setpoint Calculations

NCV05000498; 499/2007007-02 NC

V Manual Loads not Considered for Fuel Oil Storage Tank Sizing Calculation

NCV05000498; 499/2007007-03 NC

V Failure to Use Correct Design Inputs in Determination of the Weak Link for the Auxiliary

Feedwater System Outside Containment Isolation

Motor Operated Valves

NCV05000498; 499/2007007-04 NC

V Surveillance Procedure Lacked Check for Timing of Chiller Loading on the Bus

NCV05000498; 499/2007007-05 NC

V Inadequate Test Program for 125V DC Molded Case Circuit Breakers

NCV05000498; 499/2007007-06 NC

V Failure to Incorporate Instrument Uncertainties into Surveillance Requirements for Technical

Specification Limiting Condition for Operation 3.5.2

(Specifically Surveillance Requirement 4.5.2.f)

LIST OF DOCUMENTS REVIEWED

Calculations

Number Title Revision/Date
MC-5694 Auxiliary Feedwater System Failure Modes and Effects Analysis 3
EC-5008 Class 1E Battery, Battery Charger and Inverter Sizing 13
EC-5100 Standby Diesel Generator Transient Response Model 2
MC-6213
GNL 89-10 Calc for MOV 1AF0019 6
ZC-7038 Loop Uncertainty Calcul ation for QDPS Cabinet Temperature Instrumentation
EC-5001 Fault Analysis 6
MC-6462 DVAC Calculation for DC Motor MOVs 0
EC-5031 480 Volt Load Centers 7
EC-5018 Short Circuit Current Analysis - Class 1E 125 VDC and Non-Class 1E 250, 125, and 48 VDC Systems
EC-5003-02 Cable Ampacity in Underground Ducts 8
EC-5037 Maximum Allowable Length of AC Power Cables 4
EC-5033 Protection Non 1E 48 VDC, 125 VDC & 250 VDC and Class
IE 125 VDC Systems
EC-5004 Cable Ampacity
7 3Q159MC6038 SDBY DG FOST Level Setting Calculation 2 5Q159MC5912 NPSH on the Fuel Oil Transfer Pump

Calculations

Number Title Revision/Date
EC-5008 Class 1E Battery, Battery Charger and Inverter Sizing 13
EC-5033 Protection Non 1E 48 VDC, 125 VDC & 250 VDC and Class 1E 125 VDC Systems
EC-5018 Short Circuit Current Analysis - Class 1E 125 VDC and Non-Class 1E 250, 125 and 48 VDC Systems
EC-5100 Standby Diesel Generator Transient Response Model 2
MC-5037 RWST Volumes & Limits 9
MC-6256 Sizing of SDG FOST 4
MDCN 89219-75 Protection - DC System (CB & RLY Settings) 01/15/97
ZC-7029 Loop Uncertainty Calculation for Standby Diesel Generator Fuel Oil Storage Tank Level Monitoring Instrumentation
EC-5000 Voltage Regulation Study 12
EC-5003-6 Cable Ampacity in Underground Ducts-Data Sheets 11
EC-5004 Cable Ampacities 7
EC-5014 Maximum Length of Control Cables 4
EC-5022 Transformer Neutral Grounding Resistor sizing 2
EC-5020 Main Transformer Sizing Calculation 3
EC-5024 Diesel Generator Neutral Grounding 2
EC-5028 Protection 13.8 KV Switchgear 9
EC-5029 4.16 KV Switchgear Relay Setting 5
EC-5030 Class 1E Diesel Generation Protection 1
EC-5034 Standby Transformer Protection 3
EC-5036 DC Cable Sizing 7
EC-5039 Control Cable Size Verification 0
EC-5052 Degraded and Undervoltage Protection 6
CC-06425 1997 Emergency Cooling Pond Sediment Calculation 0
CC-09959 2002 Emergency Cooling Pond Sediment Calculation 0 FRSS/CWBS-C-121 TGX Minimum and Maximum Safeguards 07/13/87
MC-5430 Emergency Cooling Water Intake Structure Cooling and Heating Loads
MC-05860 Containment Emergency Sump Performance 1
MC-6220 SI & CS Pump NPSH 4
MC-6251 Essential Cooling Water Transient Analysis 0
MC-6412 Essential Chilled Water Load 1
MC-06482 Essential Chilled Water / EAB HVAC Design Basis Loads with Capacity of 300 Tons per Train
PFD-FTE-285 Standard Single and Twin Units - 4XL Model Fluid Systems Process Flow Diagrams and Piping Design
Requirements
V-EC-1330 Motor Operated Valve (MOV) Evaluation (A2SIMOV0031A)
ZC-7024 Loop Uncertainty Calculation for RWST Level 2

Calculations

Number Title Revision/Date Monitoring Instrumentation 2N129MC5519 Pressure Drop Evaluation for the Safety Injection System 0 2N129MC5815 RWST Vacuum Potential 0 2N129MC6091 Minimum Flow Orifices SI System - Low Head - Line 3"SI1302PB2
3N129HMC6100 Evaluate Safety Related Pump Miniflows per NRC Bulletin 88-04
3R289MC5429 Head Losses Calculation - Essential Cooling Water System 1 3R289MC5633 Essential Cooling Water Pump Submergence 2 3V110MC5234 Expansion Tank Sizing for Essential Chilled Water System 2 5N129MC5519 Pressure Drop Evaluation for the Safety Injection System 0 5R289MC5812 Essential Cooling Water (ECW) Hydraulic Network Analysis (HNA)
88-EW-002 ECW Pump Discharge and Suction Pressure

Calculations

3Q159MC6038 Standby Diesel Generator Fuel Oil Storage Tank Level Setting Calculation
5Q159MC5912 NPSH on the Fuel Oil Transfer Pump 0 3N129HMC6100 Evaluate Safety Related Pump Miniflows 0 3S149MC5051 Auxiliary Feedwater Pump Discharge Pressure 4 3S149MC5861 Auxiliary Feedwater Pump Design TDH (Total Discharge Head), Flow Rate and Pump Runout
3S149MC5057 Maximum and Minimum Flow Requirements of the AFW System
3L482MC6204 GNL (Generic Letter) 89-10 Calculation for MOVD2AFMOV0019 (Weak Link Analysis)
DCN
MC-145 Revision of Mc 6204 to Incorporate Yield Stress Values for Weak Link Calculation at Design Temperatures.
7/31/1992
MC-6163 Penetration Seals for HELBA (High Energy Line Break Analysis) and Flooding
MC-5557 IVC (Isolation Valve Cubicle) Flood Analysis 8
ZC-7029 FOST (Fuel Oil Storage Tank) Wide Range Level Indicating Loops, page 7.
MC-6256 Sizing of Standby Diesel Generator FOST 4
Condition Records
07-14379 07-14383 07-14422 07-14463 07-14520 07-15473
07-15752 07-14425 07-14423 07-15418 07-15443 07-3647
07-3745 07-4926 07-5278 07-5465 07-5652 07-7917
07-14387 07-14398 07-14663 07-14903 07-14959 07-15449
07-15554 07-15592 07-15568 07-15586 07-15801 07-15828
07-16309 07-15817 07-15847 07-15702 07-11491 07-11517
07-15499 06-16616 06-10937 06-4423 06-16998 05-2442
05-15477 05-14112 05-15251 04-5149 04-15477 04-15476
04-4428 03-928 03-8824 03-14479 03-9356 01-15847
97-14434

Drawings

Number Title Revision/Date
00009E0VAAB#2 Single Line Diagram Vital 120V AC Distribution Panels
DP 1202,
DP 1203
00009E0PMAK#2 Single Line Diagram Class-IE Motor Control Center E2A4
EAB 18 00009E0DJAB#2 Single Line Diagram 125 V DC 1E Distribution Switchboard E2D11
00009E0AF14#1 E/D AFW Turbine Pump 14
MOV 0019 11 00009E0PLAB#2 Single Line Diagram, 480 V Class 1E Load Center E2B 12
00009E0PKAA#2 Single line Diagram, 4.16KV Class 1E Switchgear E2A 10
00009E0EW01#2 Elementary Diagram, Essential Cooling Water Pumps 2A, 2B, & 2C
00009E0DJAB#1 Single Line Diagram, 125VDC 1E Distribution Switchboard E1D11
00009E0PMAD#2 Single Line Diagram, 480 V Class 1E Motor Control Center 21 00009E0VAAB#1 Single Line Diagram Vital 120V AC Distribution Panels
DP 1202,
DP 1203
00009E0PMAD#2 Single Line Diagram 480 V Class 1E Motor Control Center E2B1 (EAB)
00009E0HE13#2 Elementary Diagram E.A.B. HVAC Return Fans, FN001, FN002 & FN003
00009E0HE09#2 Elementary Diagram E.A.B. HVAC Main Supply Air Vent Fans, FN014, FN015, & FN016
00009E0AAAB#1 Single Line Diagram Class 1E 125V DC & 120V Vital AC Non-Class 1E 48V, 125V, 250V, DC, & 120V Vital AC Non-Class 1E Inverter Power for Computer
208V/120V AC Regulated Power
00009E0DJAA#1 Single Line Diagram 125VDC Class 1E Distribution SWBD. E1A11 (Channel I) (E.A.B.)
00009E0DJAB#1 Single Line Diagram 125V DC Class 1E Distr. Switchboard E1D11 (Channel II) (EAB)
00009E0PMAD#2 Single Line Diagram, 480 V Class 1E Motor Control Center 21 00000E0AAAA Main One Line Diagram for Units No. 1 & 2 (site 19
A-6Drawings Number Title Revision/Date
Electrical Di stribution) 00009E0DJAA#1 Single Line Diagram 125VDC Class 1E Distribution SWBD. E1A11 (Channel I) (E.A.B.)
00009E0DJAC#1 Single Line Diagram 125V DC Class 1E Distribution SWBD E1B11 (Channel III) (EAB)
00009E0DJAD#1 Single Line Diagram 125V DC Class 1E Distribution SWBD E1C11 (Channel IV) (EAB)
00009E0DJAE#1 Single Line Diagram 125V DC Class 1E Distribution Panels PL039A, PL039B, PL039C, PL040A (EAB)
00009E-PKAB-01
  1. 2 Single Line Dwg 4.16KV Class 1E Switchgear 9 00009EOPCAB #2 Single Line Dwg 13.8KV Switchgear 2GA 14 00009EOPLAB #2 Single Line Dwg 480V Class 1E Load Center E2B 16
00009EOPC21 #2 Elementary Diagram 13.8KV Standby Bus 2G Supply BKR
ST-260 from #2 Standby XFMR, Sheet1
00009EOPC25 #2 Elementary Diagram 13.8KV Standby Bus 2G Supply BKR
ST-280 from #1 Standby XFMR, Sheet 1
00009EOPK03 #2 Elementary Diagram 4.16KV Feeder to 480V Loadcenter Transformer E2A2, E2B2, E2C2, Sheet 1
00009EOPK02 #2 Elementary Diagram 4.16KV Feeder to 480V Loadcenter Transformer E2A1, E2B1, E2C1, Sheet 1
00009EOPK01 #2 Elementary Diagram 4.16KV ESF Bus E2A, E2B, E2C Supply Breaker Control, Sheet 1
00009EODG01 #2 Elementary Diagram Standby Diesel Generator DG22 4.16KV Feeder Breaker, Sheet 3
00009EOPC19 #2 Elementary Diagram 13.8KV 2G Aux and Standby Bus Tie BKR T-240, Sheet 1
3V119V10002#1 P&ID - HVAC / Essential Chilled Water System 13 3V119V10003#1 P&ID - HVAC / Essential Chilled Water System 18 3V119V10004#1 P&ID - HVAC / Essential Chilled Water System 9 4352-00006JF 125V DC Distribution Switchboard E1D11 G
4352-00004JF Class 1E 125 VDC Distribution Switchboard E1D11 BOM F 5Q159F00045#1, Sheet 1 Piping & Instrumentation Diagram Standby Diesel Generator Fuel Oil Storage & Transfer System
5Q159F00045#1, Sheet 2 Piping & Instrumentation Diagram Standby Diesel Fuel Oil 10 5Q159F22540#1 Piping & Instrumentation Diagram Standby Diesel Jacket Water 20 5Q159F22541#1 Piping & Instrumentation Diagram Standby Diesel Cooling Water
5Q159F22542#1 Piping & Instrumentation Diagram Standby Diesel Lube Oil 19
A-7Drawings Number Title Revision/Date
5Q159F22543#1 Piping & Instrumentation Diagram Standby Diesel Air Intake & Exhaust
5Q159F22544#1 Piping & Instrumentation Diagram Standby Diesel Starting Systems & Alarms
5Q159F22545#1 Piping & Instrumentation Diagram Standby Diesel Shutdown System
5Q159F22546#1 Piping & Instrumentation Diagram Standby Diesel Starting Air
9-E-DJAF-01#1 Single Line Diagram 125V DC Class 1E Distribution Panels PL139A, PL139B, PL139C (DGB)
5R289Z42077#2 Essential Cooling Water Pumps Logic Diagram 13 5R289Z42081#2 ECW Pump Discharge Valves Logic Diagram System: EW 8 5V119Z41572 E.A.B. HVAC Return Fans Logic Diagram 12 5S149Z40136 AFW Turbine Pump Isolation Valve Logic Diagram 9
5S142F00024 Piping & Instrumentation Diagram Auxiliary Feedwater 10 5S141F00024 Piping & Instrumentation Diagram Auxiliary Feedwater 11 5S142F00024 Sheet Piping and Instrument Drawing for Auxiliary Feedwater System 10 5S142F00024 Sheet Piping and Instrument Drawing for Auxiliary Feedwater System 10 5G-15-9-P-0053 Composite Piping - Isolation Valves Cubicle Plan at El.
34'-0" 3 5N129F05013#1 P&ID - Safety Injection System 27 5N129F05016#1 P&ID - Safety Injection System 14
5N169F20000#1 P&ID - Residual Heat Removal System 24
5R289F05038#1 P&ID - Essential Cooling Water System - Train A
13
5R289F05039#1 Sheet 1, P&ID - Essential Cooling Water System 16
5V119V10001#1 P&ID - HVAC / Essential Chilled Water System 31 5V119V10001#2 P&ID - HVAC / Essential Chilled Water System 31 5V119V25000#2 P&ID - HVAC / Electrical Auxiliary Building Main Area System 16 5V159V00027#2 P&ID - HVAC Miscellaneous Buildings Essential Cooling Water Intake Structure
6P-20-0-M-0031 General Arrangement Drawing - Essential Cooling Water Intake & Discharge Structures
8114-01036-WU Vendor Dwg L.V.M.E. "DS" SWGR C 8121-01023-GU Vendor Dwg Indoor Metal-Clad SWGR 5HK B
9G069F10006 #2 Piping and Instrumentation Diagram Isolation Valves Cubicles Building Sump Pump & Drains System for oily

waste. 6 9-M-06-9-B-0177 Plumbing Isolation Valve Cubicle Building Floor Plan 2

A-8Drawings Number Title Revision/Date
EL. 34'-0" Area 11 9-M-06-9-B-0175 Plumbing Isolation Valve Cubicle Building Embedment Plan EL. Area 11
9EAF14-01#1 Elementary Diagram Aux Feedwater Turbine Pump 14 Isolation
MOV-0019

Procedures

Number Title Revision/Date
0POP02-AM-0001 ERFDADS Computer 120 VAC
UPS 13 0POP02-AE-0004 120 VAC ESF Vital Distribution Power Supplies 25 0POP02-AF-0001 Auxiliary Feedwater 24
0POP02-HE-0001 Electrical Auxiliary Building HVAC System 26 0POP02-EW-0001 Essential Cooling Water Operations 41 0POP04-HE-0001 Loss of EAB or Control Room HVAC 7 1POP09-AN-03M2 Annunciator Lampbox 1-03M-2 Response Instructions 26 2POP09-AN-03M2 Annunciator Lampbox 2-03M-2 Response Instructions 23 0POP09-AN-22M3 Annunciator Lampbox 22M03 Response Instructions 20 0POP09-AN-02M3 Annunciator Lampbox 2M03 Response Instructions 19 0PGP03-ZE-0073 Molded Case Circuit Breaker Testing Program 2 0PMP05-NA-0004 Molded Case Breaker Test 25
0PSP03-HE-0001 Control Room Emergency Ventilation System
0PSPS03-AF-0010 Auxiliary Feedwater System Valve Operability Test 21 0PSP03-EW-0017 Essential Cooling Water System Train A Testing 24 0PSP03-AF-0011 Auxiliary Feed Flow Verification 7
0PSP03-SP-0019D Turbine Driven Auxiliary Feedwater Actuation and Response Time Test
0POP01-ZO-0009 Ground Isolation 0 0POP02-DG-0001 Emergency Diesel Generator 11(21) 42
0POP02-DG-0002 Emergency Diesel Generator 12(22) 48
0POP02-DG-0003 Emergency Diesel Generator 13(23) 45
0POP02-EE-0001 ESF (Class 1E) DC Disctribution System 16 0POP04-AE-0001 First Response to Loss of Any or All 13.8 KV or 4.16 KV Bus 34 0POP04-DJ-0001 Loss of Class 1E 125 VDC Power 20 0POP05-E0-EC00 Loss of All AC Power 18
0POP09-AN-0102 Annunciator Lampbox 1(2)-102 Response Instructions 12 0POP09-AN-0104 Annunciator Lampbox 1(2)-104 Response Instructions 12 0POP09-AN-0106 Annunciator Lampbox 1(2)-106 Response Instructions 12 0POP09-AN-02M3 Annunciator Lampbox 2M03 Response Instructions 19 0POP09-AN-03M3 Annunciator Lampbox 3M03 Response Instructions 22 0POP09-AN-22M3 Annunciator Lampbox 22M03 Response Instructions 20 0PSP02-SF-0001A ESF Diesel Sequencer Timing Test Train A 11

Procedures

Number Title Revision/Date
0PSP03-DG-0001 Standby Diesel 11(21) Operability Test 33
0PSP03-DG-0002 Standby Diesel 12(22) Operability Test 31
0PSP03-DG-0003 Standby Diesel 13(23) Operability Test 34
0PSP03-DG-0007 Standby Diesel 11(21) LOOP Test 20
0PSP03-DG-0008 Standby Diesel 12(22) LOOP Test 18
0PSP03-DG-0009 Standby Diesel 13(23) LOOP Test 20
0PSP03-DG-0013 Standby Diesel 11(21) LOOP - ESF Actuation Test 20 0PSP03-DG-0014 Standby Diesel 12(22) LOOP - ESF Actuation Test 19 0PSP03-DG-0015 Standby Diesel 13(23) LOOP - ESF Actuation Test 21 0PSP03-ZQ-0028 Operator Logs 98
0PSP06-DJ-0001 125 Volt Class 1E Battery 7 Day Surveillance Test 28 0PSP06-DJ-0002 125 Volt Class 1E Battery Quarterly Surveillance Test 19 0PSP06-DJ-0003 125 Volt Class 1E Battery Surveillance Test 13 0PSP06-DJ-0006 Battery Charger 8 Hour Load Verification 19
0PSP06-DJ-0007 125 Volt Class 1E Battery Combined Service and Performance Surveillance Test
1POP09-AN-03M2 Annunciator Lampbox 1-03M-2 Response Instructions 26 2POP09-AN-03M2 Annunciator Lampbox 2-03M-2 Response Instructions 23
IP-3.20Q Interdepartmental Procedures 10CFR50.59 Evaluations 4
0POP01-ZQ-0022 Plant Operations Shift Routines 52
0PSP03-DG-0016 Standby Diesel 11(21) Twenty-Four Hour Load Test 23
0PSP06-PK-0001 4.16KV Class 1E Undervoltage Relay Channel Calibration/TADOT-Channel 1
0PSP06-PK-0002 4.16KV Class 1E Undervoltage Relay Channel Calibration/TADOT-Channel 2
0PSP06-PK-0003 4.16KV Class 1E Undervoltage Relay Channel Calibration/TADOT-Channel 3
0PSP06-PK-0004 4.16KV Class 1E Undervoltage Relay Channel Calibration/TADOT-Channel 4
0PSP06-PK-005 4.16KV Class 1E Degraded Voltage Relay Calibration/TADOT-Channel 1
0PSP06-PK-006 4.16KV Class 1E Degraded Voltage Relay Calibration/TADOT-Channel 2
0PSP06-PK-007 4.16KV Class 1E Degraded Voltage Relay Calibration/TADOT-Channel 3
0PSP06-PK-008 4.16KV Class 1E Degraded Voltage Relay Calibration/TADOT-Channel 4
0P0P02-CH-0005 Essential Chiller Operation 44
0P0P03-CH-0001 Essential Chilled Water Pump 11A(21A) Inservice Test 15 0PSP03-EW-0008 Essential Cooling Water Pump 1A(2A) Reference Values Measurement
0PSP03-EW-0017 Essential Cooling Water System Train A Testing 24
0PSP03-CH-0004 Essential Chilled Water Pump 11A(12A) Reference 7

Procedures

Number Title Revision/Date Values Measurement 0PSP03-SI-0001 Low Head Safety Injection Pump 1A(2A) Inservice Test 13 0PSP03-SI-0004 High Head Safety Injection Pump 1A (2A) Inservice Test 12 0PSP03-SI-0007 Low Head Safety Injection Pump 1A(2A) Reference Values Measurement
0PSP03-SI-0010 High Head Safety Injection Pump 1A(2A) Reference Values Measurement
0PSP03-SI-0026 High Head Safety Injection Pump Flow Rate Measurement
0PSP03-SI-0027 Low Head Safety Injection Pump Flow Rate Measurement
WAN 127044 Preventive Maintenance 4160V Switchgear E1A CUB 1 0
WAN 147480 Preventive Maintenance AUX ESF XFMR E1A To 4.16KV ESF BUS E1A
WAN 177227 PM Inspect Breaker, Unit 2 STBY XFMR TO STBY Bus
2G 7
WAN 220394 PM Inspect Breaker, Unit 2 STBY XFMR To STBY BUS
2G 2
WAN 256076 13.8KV To 4160 VAC ESF Transformer E2B 1
WAN 257849 480V Load Center E2B 7
WAN 265895 PM Unit 2 Standby Transformer 1
WAN 268512 PM Inspect Breaker, To 4.16KV AUX ESF XFMR E2B 8
WAN 274021 PM Calibrate Relays, To 4.16 KV AUX ESF XFMR E2B 0

Miscellaneous Documents

Number Title
Revision/Date
4E519EB1108

Design Basis Document

4.16KV AC Power (PK) System 3 4E53EB1109 Design Basis Document Class 1E AC Power (PL/PM) System 2 4E549EB01110 Design Basis Document, Class 1E Vital 120V AC System 2 4E510EQ1005 Design Criteria Class 1E AC Power Distribution 8 4E520EQ1006 Class 1E 125 Vdc Design Criteria 6 5R289MB1006 Design Basis Document Essential Cooling Water System 5 5S149MB01016 Design Basis Document Auxiliary Feedwater System 5
5V119VB01022 Design Basis Document HE/HE (CRE) System 4 5E540EL5031
Electrical Setpoint Index 3
STP Interconnection Agreement 8/15/2002
ERCOT Operating Procedure Manual Transmission & Security Desk
8454-00017-KV 4160/480V Transformer E2B1 Nameplate A 8454-00014-KV 4160/480V Transformer E2B2 Nameplate A
A-11Miscellaneous Documents Number Title
Revision/Date
8074-01024-WM 13.8KV/4160V Transformer E2B Nameplate A 8074-01004-WM 13.8KV/4160V Transformer E2B Test Data A
8394-00037-ZF 25/13.8KV Unit 2 Auxiliary Transformer Nameplate G
8045-01007-WB 362.25/13.8KV Standby Transformer 2 Nameplate A
8045-01005-WB 362.25/13.8KV Standby Transformer 2 Test Data A
VTD - W120-250 VTD - Maintenance Program Manual for Safety Related Type DS Low Voltage Metal Enclosed Switchgear
VTD-B455-0047
VTD-Installation/Maintenance Instructions for Metal Clad Medium Voltage Power Switchgear Type 5HK
VTD-B455-0042
VTD-Installation/Maintenance Instructions Medium Voltage Power Circuit Breakers Type 5HK
NRC Letter to Mr. Bradford M. Radimer, Chairman, IEEE Battery Working Group
01/11/90
Reply to Notice of Violation 9235-03 Regarding a Failure to Fully Test Essential Chiller ESF Loading Timing
Sequence 04/02/93
South Texas Project, Units 1 and 2 - Issuance of Amendments RE: Technical Specification 3/4.8.2 for Batteries and DC Systems (TAC Nos. MD0333 and
MD0334) 07/20/07 5Q159MB1023
4A.1.6 Fluid System Margins, Page 4A-25 3 ANSI N195-1976 American National Standard Fuel Oil Systems for Standby Diesel-Generators
04/12/76
CN-2824 Revise UFSAR Table 9.4-1 "Normal Parameters Temp." to Reflect a "Normal Parameters Temp" Range of 70°F to
77°F Instead of the 73°F to 77°F Range Currently Listed
2/27/06 DCP# 04-5388 Install Diodes Across Battery Chargers E2C11-1 and E2C11-2 Alarm Relays
04/20/04 DCP# 04-6544 Replace Float/Equalize Sw itches of Class 1E Battery Chargers 09/08/05 NSAC 125 Guideline for 10
CFR 50.59 Safety Evaluations 06/1989
NUREG-0800 USNRC Standard Review Plan 2
NUREG/CR-2792 An Assessment of Residual Heat Removal and Containment Spray Pump Performance Under Air and Debris Ingesting Conditions
09/1982
PR-880403 Weekly Surveillance Performed Incorrectly 10/06/88
PR-910030 Surveillance Performed and Reviewed Incorrectly 02/24/92
PRA-07-010 Unit 1 ESF
DG 13 and Unit 2 ESF
DG 23 TS Surveillance 4.8.1.1.2.e.11 Not Fully Performed
10/05/07 Regulatory Guide
1.137 Fuel Oil Systems for Standby Diesel Generators 0 Regulatory Guide Fuel Oil Systems for Standby Diesel Generators 1
2Miscellaneous Documents Number Title
Revision/Date
1.137
SPR 910145 Operability Questionable on Class 1E Batteries 10/12/92
SPR 921482 Sequence Start Times for Essential Chiller 21B did not Meet Design Value Anticipated Start Times
2/11/92 USQE 91-0031 FSAR
CN-1725 06/07/91 VTD-W120-0678
AB De-Ion Circuit Breakers Time Current Characteristics Curves for Standard and Mark 75
Thermal Magnetic Circuit Breakers
VTD W120-0152 Type MME Magnetic Contactor 0
VTD-W120-0216 AB De-Ion Circuit Breakers Standard Seltronic Mark 75 and Tri-Pac Designs
VTD-W120-0300 Qualified Display Processing System (QDPS) 0 R289XG170BHY Essential Cooling Water Induction Motor Data Sheet
05/21/82 B03050-0008H4 Vendor Drawing AMTEK 10 KVA Inverter
A
B03050-0007H4 Vendor Drawing AMTEK 10 KVA Inverter C
B03050-0005H4 Vendor Drawing AMTEK 10 KVA Rectifier B
VTD-A363-0021 VTD - 10 KVA Inverter 3
2-E-EM-0822 Configuration Control Package 00
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