IR 05000266/2009007
| ML100210779 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 01/21/2010 |
| From: | Daley R C NRC/RGN-II/DRS/EB3 |
| To: | Meyer L Point Beach |
| References | |
| IR-09-007 | |
| Download: ML100210779 (19) | |
Text
January 21, 2010
Mr. Larry Meyer Senior Vice President NextEra Energy Point Beach, LLC 6610 Nuclear Road Two Rivers, WI 54241
SUBJECT: POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 EVALUATION OF CHANGES, TESTS, OR EXPERIMENTS AND PERMANENT PLANT MODIFICATIONS BASELINE INSPECTION REPORT 05000266/2009007(DRS); 05000301/2009007(DRS)
Dear Mr. Meyer:
On December 18, 2009, the U.S. Nuclear Regulatory Commission (NRC) completed an Evaluations of Changes, Tests, or Experiments and Permanent Plant Modifications Inspection at your Point Beach Nuclear Plant, Units 1 and 2. The enclosed report documents the inspection results, which were discussed on December 18, 2009, with you and other members of your staff. This inspection examined activities conducted under your license as they relate to safety and to compliance with the Commission's rules and regulations and with the conditions of your license. The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel. Based on the results of this inspection, one NRC-identified finding of very low safety-significance was identified. The finding involved a violation of NRC requirements. However, because of its very low safety-significance, and because the issue was entered into your corrective action program, the NRC is treating the issue as a Non-Cited Violation (NCV) in accordance with Section VI.A.1 of the NRC Enforcement Policy. If you contest the subject or severity of a Non-Cited Violation, 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 a copy to the Regional Administrator, U.S. Nuclear Regulatory Commission - Region III, 2443 Warrenville Road, Suite 210, Lisle, IL 60532-4352; the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; and the Resident Inspector Office at the Point Beach Nuclear Plant. In addition, if you disagree with the characterization of any finding in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your disagreement, to the Regional Administrator, Region III, and the NRC Resident Inspector at the Point Beach Nuclear Plant. The information that you provide will be considered in accordance with Inspection Manual Chapter 0305. In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter, its enclosure, and your response (if any), will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records System (PARS) component of NRC's Agencywide Documents Access and Management System (ADAMS), accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
Sincerely,/RA/
Robert Daley, Chief Engineering Branch 3 Division of Reactor Safety Docket Nos. 50-266; 50-301 License Nos. DPR-24; DPR-27
Enclosure:
Inspection Report No. 05000266/2009007(DRS); 05000301/2009007 (DRS)
w/Attachment:
Supplemental Information cc w/encl: Distribution via ListServ Enclosure U.S. NUCLEAR REGULATORY COMMISSION REGION III Docket No: 05000266; 05000301 License No: DPR-24; DPR-27 Report No: 05000266/2009007(DRS); 05000301/2009007(DRS) Licensee: NextEra Energy Point Beach, LLC Facility: Point Beach Nuclear Plant, Units 1 and 2 Location: Two Rivers, WI Dates: November 30, 2009 through December 18, 2009 Inspectors: A. Dahbur, Senior Reactor Inspector (Lead) J. Neurauter, Senior Reactor Inspector N. Feliz-Adorno, Reactor Inspector Observers: S. Edmonds, NSPDP M. Benson, General Engineer, NSPDP Approved by: R. Daley, Chief Engineering Branch 3 Division of Reactor Safety (DRS)
Enclosure 1
SUMMARY OF FINDINGS
IR 05000266/2009007 (DRS); 05000301/2009007 (DRS); 11/30/2009 - 12/18/2009; Point Beach Nuclear Plant, Units 1 and 2; Evaluations of Changes, Tests, or Experiments and Permanent Plant Modifications. This report covers a two-week announced baseline inspection on Evaluations of Changes, Tests, or Experiments and Permanent Plant Modifications. The inspection was conducted by
Region III based engineering inspectors. Based on the results of this inspection, one finding of very low safety-significance (Green) was identified. The finding was considered a Non-Cited Violation (NCV) of NRC regulations. The significance of most findings is indicated by their color (Green, White, Yellow, Red) using Inspection Manual Chapter (IMC) 0609, "Significance Determination Process" (SDP). Findings for which the SDP does not apply may be Green or be assigned a severity level after NRC management review. The NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, "Reactor Oversight Process," Revision 4, dated December 2006.
A. NRC-Identified
and Self-Revealed Findings
Cornerstone: Mitigating Systems
- Green.
A finding of very low safety-significance (Green) and associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control" was identified by the inspectors for the licensee's failure to adequately calculate the maximum room temperature for G-01 and G-02. Specifically, the licensee's calculation 2005-0054 failed to incorporate the design basis described in Technical Specification (TS) bases 3.8.1 related to the numbers of fire dampers associated with G-01 and G-02 exhaust fans that must be opened to maintain room temperature. The calculation also failed to demonstrate that the temperature stratification close to the combustion air intake filter was acceptable. Instead, the calculation only considered the bulk air temperature in the room. The licensee subsequently entered these concerns into their corrective action program as AR 01162599 and AR 01162759. The finding was determined to be more than minor because the finding was similar to IMC 0612, Appendix E, Example (3.J). The calculation errors were significant in that there was reasonable doubt that the maximum room temperature would not exceed the value of the Vendor Technical manual. The finding impacted the Mitigating System cornerstone of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee did not ensure that the maximum room temperature of EDG-1 and EDG-2 would not exceed 115 degrees (o) Fahrenheit (F), which is required to be maintained to ensure that the EDGs will perform their safety function during a design basis accident when the outside air temperature was 95 o F. The finding was of very low safety-significance based on a Phase 1 screening in accordance with IMC 0609, Appendix A, ASignificance Determination of Reactor Inspection Findings for At-Power Situations.
@ This finding was not associated with a cross-cutting aspect because the finding was not indicative of the licensee's current performance. (Section 1R17.1.b(1))
B. Licensee-Identified Violations
No findings of significance were identified.
2
REPORT DETAILS
REACTOR SAFETY
Cornerstone:
Initiating Events, Mitigating Systems, and Barrier Integrity
1R17 Evaluation of Changes, Tests, or Experiments and Permanent Plant Modifications
.1 Evaluation of Changes, Tests, or Experiments
a. Inspection Scope
From November 30, 2009 through December 18, 2009, the inspectors reviewed 13 safety evaluations (SEs) performed pursuant to 10 CFR 50.59 to determine if the evaluations were adequate and that prior NRC approval was obtained as appropriate. The inspectors also reviewed 13 screenings and one calculation where licensee personnel had determined that a 10 CFR 50.59 evaluation was not necessary. The inspectors reviewed these documents to determine if:
- the changes, tests, or experiments performed were evaluated in accordance with 10 CFR 50.59 and that sufficient documentation existed to confirm that a license amendment was not required;
- the safety issue requiring the change, tests or experiment was resolved;
- the licensee conclusions for evaluations of changes, tests, or experiments were correct and consistent with 10 CFR 50.59; and
- the design and licensing basis documentation was updated to reflect the change. The inspectors used, in part, Nuclear Energy Institute (NEI) 96-07, "Guidelines for 10 CFR 50.59 Implementation," Revision 1, to determine acceptability of the completed evaluations, and screenings. The NEI document was endorsed by the NRC in Regulatory Guide 1.187, "Guidance for Implementation of 10 CFR 50.59, Changes, Tests, and Experiments," dated November 2000. The inspectors also consulted Part 9900 of the NRC Inspection Manual, "10 CFR Guidance for 10 CFR 50.59, Changes, Tests, and Experiments." This inspection constituted 13 samples of evaluations and 13 samples of changes as defined in IP 71111.17-04.
b. Findings
(1) Errors Found in the Room Ventilation Calculation for G-01 and G-02
Introduction:
A finding of very low safety-significance and associated NCV of 10 CFR Part 50, Appendix B, Criterion III, ADesign Control" was identified by the inspectors for the licensee's failure to adequately calculate the maximum room temperature for the Emergency Diesel Generators (EDGs) G-01 and G-02 during post-accident scenarios and assuming an outside temperature of 95 o F. Specifically, the licensee's calculation for the EDG rooms' 3ventilation did not incorporate the design basis described in TS Bases 3.8.1, which indicated that three out of four dampers were required to be open to maintain room temperature for the emergency diesel generators G-01 and G-02. In addition, the calculation also failed to demonstrate that the temperature stratification close to the combustion air intake filter was adequate, instead, the calculation only considered the bulk air temperature in the room.
Description:
Emergency Diesel Generators G-01 and G-02 draw combustion intake air from their respective rooms; therefore, the room temperature directly affected temperature limitations on the diesel. Design basis calculation for EDG ventilation, 2005-0054 "Control Building GOTHIC Temperature Calculation" was prepared to demonstrate the adequacy of the ventilation systems for the rooms housing the G-01 and G-02 in order to maintain an environment of 115 o F for post-accident scenarios assuming a maximum outside air temperature of 95 o F. The temperature limitation of 115 o F was based on the diesel Vendor Technical Manual (VTM 00367) which stated that when the combustion air temperature exceeded 115 o F, the diesel must be derated from its design load. Per calculation 2005-0054, the calculated maximum room temperature value inside the EDG rooms during an accident was 113.3 o F. Based on this result, the licensee determined that the ventilation system for the EDG rooms were adequate to support the design basis loads of the diesels. The licensee completed a 10 CFR 50.59 Screening 2009-010 "Technical Specification 3.8.1 Bases Changes," to revise the stated allowable room temperature in the G01 and G02 room from 120 o F to 115 o F. The change was made to be consistent with the design basis calculation and the diesel Vendor Technical Manual. The inspectors reviewed the Technical Specification Bases 3.8.1 and noted that a specific statement was not considered in the above calculation. Specifically, TS Basis 3.8.1 stated that for G-01 and G-02, three of four fire dampers associated with the diesel room exhaust fans must be open to maintain room temperature. The inspectors were concerned that calculation 2005-0054 was not conservative, in that the air flow measured values used by the calculation were obtained with all dampers open. These measurements were taken per PBTP 157 "Diesel Room Exhaust Flow Measurement." During a walkdown in the G-01 and G-02 rooms, the inspectors also noted that the combustion air intake was close to the un-insulated portion of the exhaust manifold which could have a temperature of 800 o F during diesel operation. The inspectors were also concerned that calculation 2005-0054 was again not conservative in that the calculation only addressed the bulk temperature in the rooms and did not account for temperature stratification close to the combustion air intake filter. This may result in an increase of the combustion air temperature for the diesel that may require a decrease of the diesel loading. On December 4, 2009, the licensee entered this issue into their corrective action program as AR 01162599 "TS Bases 3.8.1 not supported by calculation," and AR 01162759 "DG Room Temperature Calculation has Low Margin." Following the inspectors' identification of these concerns, the licensee revised calculation 2005-0054 using the actual diesel design basis loading, 2820 KW for G-01 and 2818 KW for G-02, which was less than the 2000 hour0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> rating of the diesels (2850 KW). By calculating the temperature using actual loading, the licensee was able to produce more margin when addressing the inspectors' concerns. The licensee demonstrated that using the actual design basis loading resulted in an increase of the maximum allowable G-01 and G-02 room temperatures, 118.8 o F for G-01 and 119 o F for G-02, and thus an increase in the available margin. However, the results of the revised 4calculation also showed that when assuming both fans running and one of the four fire dampers closed for each diesel as indicated in the TS bases, the maximum predicted temperature of the diesel rooms would increase from the previous calculation. However, even though the results of the new calculation reduced margin due to higher calculated temperatures relative to the previous calculation, margin was still available such that the increased temperature was not enough to affect the operability of the diesels. The licensee also performed a walkdown in the G-01 room to determine a temperature distribution throughout the room when the EDG and fans were operating. The specific purpose was to determine whether there was a difference between the bulk heatup temperature of the room and the combustion air inlet. Based on several readings that were taken, the licensee determined that although there were variations in the air temperature in the room, there was no evidence of a temperature differential between the bulk air temperature of the ventilation flow and the temperature at the combustion air inlet.
Analysis:
The inspectors determined that the licensee's failure to adequately calculate the maximum room temperature for emergency diesel rooms for pos-accident was contrary to the requirements of 10 CFR Part 50, Appendix B, Criterion III and was a performance deficiency. Specifically, the licensee did not ensure that the TS basis and the variation of the air temperature were accounted for when calculating the maximum room temperature for G-01 and G-02. The licensee failed to ensure that the maximum room temperature would not exceed 115 o F, which is required to be maintained to ensure that the EDGs will perform their safety function during a design basis accident when the outside air temperature was 95 o F. The finding was determined to be more than minor because the finding was similar to IMC 0612, Appendix E, Example (3.J). Specifically, the errors in the calculation were significant in that although, at the end of the inspection, the licensee was able to demonstrate operability and adequate margin existed; at the time of discovery there was reasonable doubt that the maximum room temperature for G-01 and G-02 would not exceed the value specified in the Vendor Technical Manual which ensure the availability and reliability of the diesel during post design basis accident. The performance deficiency also impacted the Mitigating System cornerstone of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences.
The inspectors determined the finding could be evaluated using the SDP in accordance with IMC 0609, "Significance Determination Process," Attachment 0609.04, "Phase 1 -
Initial Screening and Characterization of Findings," Tables 3b and 4a for the Mitigating Systems Cornerstone. The inspectors determined that the finding was of very low safety-significance (Green) because the finding did not involve a design or qualification deficiency there was no actual loss of safety function, no single train loss of safety function for greater than the TS allowed outage time, and no risk due to external events. The inspectors did not identify a cross-cutting aspect associated with this finding because the finding was not indicative of the licensee's current performance. The licensee change to the Technical Specification bases did not require revision to the existed calculation 2005-0054.
5Enforcement: Title 10 CFR Part 50, Appendix B, Criterion III, ADesign 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, on December 2, 2009, the inspectors identified that the licensee failed to assure that the design basis for the emergency diesel generators were correctly translated into instructions. Specifically, the licensee's calculation 2005-0054 failed to account for the Technical Specification 3.8.1 requirement related to the numbers of fire dampers associated with G-01 and G-02 exhaust fans that must be opened to maintain room temperature. The date of origination for the TS requirement that led to this violation could not be determined. The calculation also failed to demonstrate that the temperature stratification close to the combustion air intake filter was acceptable. Instead, the licensee only calculated the bulk air temperature in the room. Because this violation was of a very low safety-significance and because it was entered into the licensee's corrective action program as AR 01162599 and AR 01162759, this violation is being treated as a Non-Cited Violation (NCV), consistent with Section VI.A.1 of the NRC enforcement policy. (NCV 05000266/2009007-01; 05000301/2009007-01)
.2 Permanent Plant Modifications
a. Inspection Scope
From November 30, 2009 through December 18, 2009, the inspectors reviewed nine permanent plant modifications that had been installed in the plant during the last three years and two calculations that had been also revised during the last three years. This review included in-plant walkdowns for the Component Cooling Water (CCW) heat exchangers; PAB battery and inverter rooms; façade flood barriers; and EDG-1 and EDG-2 rooms and their associated damper areas. The modifications were selected based upon risk significance, safety-significance, and complexity. The modifications selected also included two modifications that were installed to facilitate the extended power uprate (EPU) project. The inspectors reviewed the modifications selected to determine if:
- the supporting design and licensing basis documentation was updated;
- the changes were in accordance with the specified design requirements;
- the procedures and training plans affected by the modification have been adequately updated;
- the test documentation as required by the applicable test programs has been updated;
and
- post-modification testing adequately verified system operability and/or functionality. The inspectors also used applicable industry standards to evaluate acceptability of the modifications. The list of modifications and other documents reviewed by the inspectors is included as an attachment to this report. This inspection constituted 11 permanent plant modification samples as defined in IP 71111.17-04.
6b. Findings No findings of significance were identified.
OTHER ACTIVITIES (OA)
4OA2 Identification and Resolution of Problems
.1 Routine Review of Condition Reports
a. Inspection Scope
From November 30, 2009 through December 18, 2009, the inspectors reviewed Corrective Action Process documents that identified or were related to 10 CFR 50.59 evaluations and permanent plant modifications. The inspectors reviewed these documents to evaluate the effectiveness of corrective actions related to permanent pant modifications and evaluations for changes, tests, or experiments issues. In addition, corrective action documents written on issues identified during the inspection were reviewed to verify adequate problem identification and incorporation of the problems into the corrective action system. The specific corrective action documents that were sampled and reviewed by the inspectors are listed in the attachment to this report.
b. Findings
No findings of significance were identified.
4OA5 Other Activities
.1 Plant Modifications in Support of Extended Power Uprate (EPU)
a. Inspection Scope
From November 30, 2009 through December 18, 2009, the inspectors reviewed the following completed plant modifications that supported EPU:
- Unit 2 replacement motor driven auxiliary feedwater mechanical tie-ins to the service water and auxiliary feedwater systems. Specifically, the inspectors reviewed a sample of the associated engineering change documentation including the 10 CFR 50.59 screening, design calculations, work orders, engineering change requests, and corrective action requests to assure the installed plant change was consistent with the design and licensing bases. The inspectors also walked down the mechanical tie-ins to the service water and feedwater systems to verify the installed piping configurations were consistent with the design and installation documentation.
- Replacement of motor driven auxiliary feedwater electrical and instrumentations tie-ins modification installed during Unit 2 refueling per EC-13401. The inspectors also walked down changes, installed per EC-13401, to the Unit 2 control room panels with the Seismic Qualification Utility Group (SQUG) engineer.
7b. Findings No findings of significance were identified.
4OA6 Meetings
.1 Exit Meeting Summary
On December 18, 2009, the inspectors presented the inspection results to you, and other members of your staff. The licensee personnel acknowledged the inspection results presented and did not identify any proprietary content. The inspectors confirmed that all proprietary material reviewed during the inspection were returned to the licensee staff. ATTACHMENT:
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee
- L. Meyer, Site Vice President
- J. Costedio, Licensing Manager
- M. Durbin, Training Supervisor
- R. Harrsch, Operations Manager
- V. Kanal, Engineering Supervisor (Electrical)
- K. Locke, Licensing Engineering Analyst
- A. Mitchell, System Engineering Manager
- C. Trsziss, Engineering Director
- S. Ruesch, NOS Manager
- B. Woyak, Design Engineering Supervisor (Mechanical)
- P. Wild, Design Engineering Manager Nuclear Regulatory Commission
- S. Burton, Senior Resident Inspector
- A. Sanchez, Reactor Inspector (Trainee)
LIST OF ITEMS
OPENED, CLOSED AND DISCUSSED
Opened and Closed
- 05000301/2009007-01 NCV Errors Found in the Room Ventilation Calculation for G-01 and G-02 (Section 1R17.1.b.(1))
Discussed
None
2
LIST OF DOCUMENTS REVIEWED
The following is a list of documents reviewed during the inspection.
- Inclusion on this list does not imply that the NRC inspectors reviewed the documents in their entirety, but rather, that selected sections of portions of the documents were evaluated as part of the overall inspection effort.
- Inclusion of a document on this list does not imply NRC acceptance of the document or any part of it, unless this is stated in the body of the inspection report.
- CFR 50.59 EVALUATIONS Number Description or TitleDate or Revision
- 07-005 Façade Flooding and Grout Removal December 06, 200708-003 Installation to Replace Portions of Power Cables from 2X-04 to 2A-03 and 2A-04 April 04, 2008 08-005 Revised Procedure 2-SOP-480V-001 April 24, 2008 08-008 Remove Requirement to Backseat Normally Open Valves September 15, 200808-016 Replacement of Unit 1 FW Heaters to Support EPU October 03, 2008 08-018 Impact of Revised Containment Heat Sink Paint Thicknesses on Containment Integrity Analysis November 01, 200808-019 Revised Containment Heat Sink for SLB Containment Integrity Analysis
- November 01, 200808-020 Feedwater Temperature and Volume Increases with Replacement of FW Heaters on SLB Containment Response November 02, 200808-021 Creation of Procedures for Supplemental Ventilation April 03, 2009 08-022 Evaluation of Increased Delay Time for Bus A-01/A02 Undervoltage Reactor Trip February 26, 2009 09-003 UFSAR
- 01141895 Changes to FSAR A.7, "Plant Internal Flooding" May 14, 2009 09-005 PAB Superstructure - Use of Existing Craneway Bolts June 11, 2009 09-010 T.S. Bases B 3.6.5 Revision to Westinghouse Letter
- WEP-06-64, Containment Integrity Evaluation for Increased Paint Thickness on Containment Structures July 31, 2009
- CFR 50.59 SCREENINGS Number Description or TitleDate or Revision
- 05-172 Setpoint Changes for EDG HX June 06, 2007 06-099 Minimum Voltage Requirements for MCC June 14, 2006
- 06-132 DG SW Flow Loop Uncertainty Calculation August 01, 2006 06-257 Revised Calculation RWST Level Setpoint July 18, 2007 07-120 Auxiliary Feedwater Pump Low Suction Pressure Trip
- September 27, 2007
- 310
- CFR 50.59 SCREENINGS Number Description or TitleDate or Revision
- 07-183 Revised Calculation Service Water Flow vs. Temperature Requirements December 13, 2007 08-020 Revised Technical Specifications Bases 3.8.1 and 3.8.2 to Include Offsite Power Back Fed Through X-01 January 20, 2008 08-132 Temporary Cooling of PAB Battery and Inverter Rooms August 06, 2008 08-139 Revised Parametric Values in Technical Specifications Bases October 01, 2008 08-190 Replacement of Unit 1 FW Heater to Support EPU November 5, 2008 08-198 Calculation - SI Accumulator Level Instrument Uncertainty Setpoint October 01, 2008 09-010 Technical Specification 3.8.1 Bases Change January 28, 2009 09-014
- EC-13400: U2R30 Replacement MDAFW Mechanical Tie-Ins
- August 24, 2009 09-016 TRM 4.12 Diesel Fuel Oil Change February 02, 2009 09-174 Procedure Change - TDAFW Pump Overspeed Trip Test Using Air September 18, 2009
- MODIFICATIONS Number Description or TitleDate or RevisionEC-08321 Design Documentation in Support of
- GSI-191 Resolution Revision 0
- EC-08744 Replacement of Obsolete Westinghouse Breaker Revision 0
- EC-10633 Equivalency Evaluation of a Machined 1/2" Threaded Coupling for Use on 1SI-301R-3 Revision 0
- EC-12000 Evaluation of Reactor Vessel Structural Integrity Analysis (EE 2008-0004) Revision 0
- Revision 1
- EC-13163 Fatigue Pro Analysis for Point Beach Units 1 and 2 Revision 0
- CALCULATIONS
- Number Description or TitleDate or Revision
- 2005-0054 Control Building GOTHIC Temperature Calculation Revision 001 2006-0035 Parametric Values Revision 1
- 4CALCULATIONS
- Number Description or TitleDate or Revision
- 2007-0002 Emergency Diesel Generator Frequency Uncertainty Calculation Revision 0 2008-024
- AFWP Room Flood Basis Calculation Revision 0 2009-0012 PAB Craneway Lateral Connection to Column Lines 10 and 13, Rows G to U, Bolting Analysis Revision 0 PNPB-305336-
- S01 Structural Analysis of Central PAB with Crane Load of 125 Tons Revision 001-A
- PBNP-994-18-M01 Determine Flood Water Volume in the U-1 & U-2 Façade Areas Based on Flood Barrier Height of 3'-0" Revision 2
- PBNP-994-18-M03 Determine Maximum Water Flow Rate from 8 Inch Service Water Break in U-2 Façade Revision 0 PBNP-994-18-
- S01 Calculation for Design of Barriers for Façade Flooding Revision 2
- PBCH-05Q-302 FatiguePro Baseline for PBNP 1/2 through August 2003 Revision 4
- FP-PBCH-306 FatiguePro Analysis of Plant Data for Point Beach Units 1 and 2 - 2003 and 2004 Revision 2
- FP-PBCH-307 FatiguePro Analysis of Plant Data for Point Beach Units 1 and 2 - 2005 through 2008 Revision 1
- TDI-6007-07 Vortex, Air Ingestion and Void Fraction Revision 4
- CORRECTIVE ACTION PROGRAM DOCUMENTS ISSUED DURING INSPECTION Number Description or Title Date or Revision
- AR-1162483 Reconcile Differences between FSAR A.1 and 2005-0053 FSAR Appendix A.1 December 02, 2009
- AR-1162487 Reconcile Differences between FSAR Spray PP Levels FSAR Section 6.4 December 02, 2009
- AR-1162465 G-01 and G-02 Fire Dampers are not Classified during Investigation into NRC
- Questions December 02, 2009
- AR-1162468 Structural Drawing not Updated December 02, 2009
- AR-1162494 Drawing C-360 not updated for damaged beam December 02, 2009
- EC-10633 Minimum Wall Discussions December 03, 2009
- AR-1162577 Evaluation 2008-022 Insufficient Detail December 03, 2009
- AR-1162599 T. S. Bases 3.8.1 not supported by calculation December 04, 2009
- AR-1162611 Black/Yellow Tape on Floor in NV KV Switchgear During an NRC Walkdown December 04, 2009
- AR-1162654 FSAR 14.3.4 Temperature Curves Outdated December 03, 2009
- AR-1162632 Current Interrupting Rating Errors on
- QF-0422 Form December 04, 2009
- AR-1162615 Inappropriately Pre-Screening of TRM 2.2 Revision 2 December 04, 2009
- 5CORRECTIVE ACTION PROGRAM DOCUMENTS ISSUED DURING INSPECTION Number Description or Title Date or Revision
- AR-1162604 STPT 2.3 Revision 4, Containment Pressure and LTOP, is Inconsistent with the Current Pressure Temperature Limits Report as Documented in TRM 2.2 December 04, 2009
- AR-1162759 DG Room Temperature Calculation has Low Margin December 07, 2009
- AR-1163042 Clarify FSAR 6.4 Description of Spray Duration December 11, 2009
- AR-1163234 FSAR Table 6.4-9 is Outdated December 15, 2009
- AR-1163243 50.59 Evaluation 2008-018 Incomplete December 15, 2009
- AR-1163278 Code Case Calculation N-392-3 December 16, 2009
- AR-1163306 Scaffold Review not Thoroughly Documented December 16, 2009
- AR-1163310 ICPs Reference Outdated CTS T.S. Values December 16, 2009
- PBF-9114 Improvements (Scaffolding) December 17, 2009
- CORRECTIVE ACTION PROGRAM DOCUMENTS REVIEWED Number Description or TitleDate or Revision
- AR 01134889 Field Condition Does Not Match Piping Isometric Drawing September 2, 2008AR
- 01143683 Fatigue Analysis of Nuclear Power Plant Components - NRC Regulatory Issue Summary 2008-30 February 6, 2009
- PROCEDURES
- Number Description or TitleDate or Revision
- TS 83 Emergency Diesel Generator G-03 Monthly Revision 24 2-SOP-480-001 480V System Normal Operations Revision 10
- OTHER DOCUMENTS/REFERENCES Number Description or TitleDate or Revision Point Beach Nuclear Power Plant Unit 1 Cycle 32 Reload Safety Evaluation Revision 2
- DOCUMENTS RELATED TO EXTENDED POWER UPRATE MODIFICATIONS Number Description or TitleDate or Revision
- SCR 2009-0014 10
- CFR 50.59 Screening:
- 6DOCUMENTS RELATED TO EXTENDED POWER UPRATE MODIFICATIONS Number Description or TitleDate or Revision
- ECN-14628
- ECN 13400-02 Rotate 2" Service Water Tie-In
- Revision 0
- ECN-14704
- ECN-14716
- EB-10-AFU-01 Hilti Bolt Location Revision 0
- ECN 14856 EC for Evaluating Cut Reinforcing Steel Revision 0
- AR 01160194 Rebar Cut in Floor in PAB at 26' Elevation November 2, 2009
- AR 01161009 Failure Investigation Process Established for Repetitive Weld Failure November 11, 2009AR
- 01161630 Cut Reinforcing Bar in AFW Pump Room Wall November 18, 2009AR
- 01161983 Auxiliary Feedwater Weld Installed in Inaccessible Location November 24, 2009
- S-11165-116-04 Auxiliary Building Floor Slab Evaluation for Rebar Cut at Elevation 26'-0" Revision 0 S-11165-134-01 Evaluation of AFW Pump Room Wall F for Rebar Cut Per
- AR 01161630 Revision 0
- WE-300044 Service Water Return from Unit 2 Containment Penetrations Up to and Including the 20"-JB-2 Discharge Header Revision 01-B
- WE-200051 2EB10A-3" Auxiliary Feedwater System from Structural Anchors
- DB3-H11 and
- DB3-2H2 to Containment Penetration P6 (EB10-
- A12) Revision 00-B
- WE-200052 Auxiliary Feedwater System from Structural Anchors
- DB3-2H7 and
- DB3-2H4 to Containment Penetration P5 (EB10-A13) Revision 00-C
- WE-200051S Emergency FW from Penetrations P-5 and P-6 to Anchors H-11, 2H2, 2H4 and 2H7 Revision 00-C
- WO-370105-01 MDAFW Mechanical Tie-Ins to Service Water Per
- EC 13400, 6" Service Water Supply Line October 19, 2009
- WO-370105-02 MDAFW Mechanical Tie-Ins to Service Water Per
- EC 13400, 2" Service Water Return Line October 17, 2009
- WO-370105-09 MDAFW Mechanical Tie-Ins to Service Water Per
- EC 13400, Pre-Fabricate 6" Service Water Supply Tie-In October 16, 2009
- WO-370131-17 Mechanical AFW Tie-In Per
- EC 13400, Pre-October 20, 2009
- 7DOCUMENTS RELATED TO EXTENDED POWER UPRATE MODIFICATIONS Number Description or TitleDate or Revision Fabricate 2FE094036 / AFW Supply to 2HX1A, 3" Tee
- N/A Weld Failure Causal Evaluation November 14, 2009
- 8
LIST OF ACRONYMS
- USED [[]]
- NRC [[]]
- CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter, its enclosure, and your response (if any), will be available electronically for public inspection in the
- PARS ) component of NRC's Agencywide Documents Access and Management System (ADAMS), accessible from the
- NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room). Sincerely, /
RA/ Robert Daley, Chief
Engineering Branch 3 Division of Reactor Safety Docket Nos. 50-266; 50-301 License Nos. DPR-24; DPR-27 Enclosure: Inspection Report No. 05000266/2009007(DRS); 05000301/2009007 (DRS) w/Attachment: Supplemental Information
cc w/encl: Distribution via Listserv