IR 05000266/2015010
| ML15261A631 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 09/16/2015 |
| From: | Robert Daley Engineering Branch 3 |
| To: | Mccartney E Point Beach |
| References | |
| IR 2015010 | |
| Download: ML15261A631 (16) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION REGION III 2443 WARRENVILLE RD. SUIT E 210 LISLE, IL 60532-4352 September 16, 2015 Mr. Eric McCartney Site Vice President NextEra Energy Point Beach, LLC 6610 Nuclear Road Two Rivers, WI 54241 SUBJECT: POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2
- EVALUATION S OF CHANGES, TESTS, AND EXPERIMENTS AND PERMANENT PLANT MODIFICATIONS BASELINE INSPECTION REPORT 05000 266/20 1 5010; 05000 301/20 1 5010 Dea r Mr. McCartney: On August 28, 2015, the U.S. Nuclear Regulatory Commission (NRC) completed an Evaluations of Changes, Tests, and Experiments and Permanent Plant Modifications inspection at your Point Beach Nuclear Plan The enclosed inspection report documents the inspection results which were discussed on August 28, 2015, with you , and other members of your staf The inspection examined activities conducted under your license as they relate to safety and compliance with the Commission's rules and regulations and with the conditions of your licens The inspectors reviewed selected procedures and records, observed activities, and interviewed personne The NRC inspectors documented one finding of very-low safety significance (Green)
in this repor This finding was determined to involve a violation of NRC requirement However, because of its very-low safety significance , and because the issue w as entered into your Corrective Action Program, the NRC is treating the issue as a Non-Cited Violation (NCV) i n accordance with Section 2. of the NRC Enforcement Polic If you contest the subject or severity of the N on-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 copies to the Regional Administrator, Region III; the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC Resident Inspector at the Point Beach Nuclear Plan E. McCartney-2- In accordance with Title 10 of the Code of Federal Regulations (10 CFR) 2.390, "Public Inspections, Exemptions, Requests for Withholding," of the NRC's "Rules of Practice," a copy of this letter, its enclosure, and your response (if any) will be available electronically for public inspection in the NRC's Public Document Room or from the Publicly Available Records (PARS)
component of the NRC's Agencywide Documents Access and Management System (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
Sincerely,
/RA/
Robert C. Daley, Chi ef Engineering Branch 3 Division of Reactor Safety Docket No ; 50-301 License Nos. DPR-24; DPR-27
Enclosure:
I R 05000266/2015010; 05000301/2015010 w/Attachment: Supplemental Information
REGION III==
Docket No: 50-266; 50-301 License No
05000266/2015010; 05000301/2015010 Licensee: NextEra Energy Point Beach, LLC Facility: Point Beach Nuclear Plant, Units 1 and 2 Location: Two Rivers, WI Dates: August 10 - 28, 2015 Inspectors:
Jorge J. Corujo-Sandín, Reactor Inspector (Lead) Mark T. Jeffers, Reactor Inspector Michael A. Jones, Reactor Inspector Approved by:
Robert C. Daley, Chief Engineering Branch 3 Division of Reactor Safety 2
SUMMARY
Inspection
Report 05000 266/20 1 5010, 05000 301/20 1 5010; 08/10/2015
- 08/28/2015
- Point Beach Nuclear Plant, Units 1 and 2
- Evaluations of Changes, Tests, and Experiments and Permanent Plant Modifications.
This report covers a 2-week announced baseline inspection on evaluations of changes, tests, and experiments and permanent plant modifications. The inspection was conducted by Region III based engineering inspectors. One finding was identified by the inspectors. The findin g was considered a Non-Cited Violation (NCV) of U.S. Nuclear Regulatory Commission (NRC) regulations. The significance of most findings is indicated by their color (i.e., Greater than Green, or Green, White, Yellow, Red) using Inspection Manual Chapter (IMC) 0609, "Significance Determination Process."
Cross-cutting aspects were determined using IMC 0310, "Aspects Within the Cross Cutting Areas." Findings for which the Significance Determination Process does not apply may be Green or be assigned a severity level after NRC management review. All violations of NRC requirements are dispositioned in accordance with the NRC's Enforcement Policy dated J ul y 9, 201 3. The NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG
-1649, "Reactor Oversight Process," Revision 5, dated February 2014
.
Cornerstone: Mitigating Systems
- Green.
The inspectors identified a finding of very-low safety significance
, and an associated N CV of Title 10 , Code of Federal Regulations, Part 50, Appendix B, Criterion III, "Design Control," for the licensee's failure to evaluate for potential gas intrusion from the spray additive tank into the containment spray (CS) system during the injection phase of a design-basis accident. As part of immediate corrective actions, the licensee entered the concern in the Corrective Action Process as AR 2068569 , and performed an evaluation which determined no air entrainment is expected to occur during the injection phase.
The performance deficiency was determined to be more than minor because it was associated with the Mitigating System s cornerstone attribute of equipment performance
, and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences.
Specifically, air intrusion into the CS system could affect the operability of the CS pumps by causing degraded performance and/or air binding of the pumps.
The finding screened as having very-low safety significance. Specifically, the finding was a deficiency affecting the design or qualification of a mitigating structure
, system , or component (SSC), however, based on the evaluation performed by the licensee the SSC maintained its operability.
Based on the timeframe of the violation the inspectors did not identify a cross-cutting aspect associated with this finding. (Section 1R17.1b)3
REPORT DETAILS
REACTOR SAFETY
Cornerstone s: Initiating Events, Mitigating Systems, and Barrier Integ ri ty 1R17 Evaluation s of Changes, Tests, and Experiments and Permanent Plant Modifications (71111.17 T)
.1 Evaluation of Changes, Tests, and
Experiments
a. Inspection Scope
The inspectors reviewed 7 safety evaluations performed pursuant to Title 10 , Code of Federal Regulations (CFR), Part 50 , Section 59 , to determine if the evaluations were adequate , and that prior U.S. Nuclear Regulatory Commission (NRC) approval was obtained as appropriate. The inspectors also reviewed 16 screenings 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, and 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, and 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 7 samples of evaluations and 16 samples o f screenings and/or applicability determinations as defined in Inspection Procedure (IP) 71111.17-04.
b. Findings
Failure to Evaluate Containment Spray System for Potential Gas Intrusion
Introduction:
The inspectors identified a finding of very
-low safety significance (Green), and an associated Non
-Cited Violation (NCV) of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," for the licensee's failure to evaluate for potential gas intrusion from the spray additive tank into the containment spray (CS) system during the injection phase of a design-basis accident (DBA).
Description:
On January 11, 2008, the NRC requested each addressee of Generic Letter (GL) 2008-01 to evaluate its Emergency Core Cooling Systems
, decay heat removal, and CS systems licensing basis, design, testing, and corrective actions to 4 ensure that gas accumulation was maintained less than the amount that would challenge the operability of these systems, and take appropriate actions when conditions adverse to quality were identified. The licensee's original actions to address these requests were, in part, to perform design reviews to identify and address gas intrusion mechanisms. Vortexing was among the gas intrusions mechanisms identified by GL 2008-01. Per its design bases, the CS system would operate during both the injection phase and recirculation phase of certain DBA. However, the inspectors identified the licensee had failed to evaluate the CS system against potential gas intrusion during the injection phase, under design bases conditions. Specifically, during the injection phase one of CS design bases function s is to transfer sodium hydroxide (NaOH) from the spray additive tank (SAT) to the containment sump. This function is required to maintain adequate pH level in the sump in order to prevent chloride induced stress corrosion cracking and maintain iodine in the iodate form that will stay in solution. To accomplish this function the CS system is aligned to two separate suction sources, the SAT and the Refueling Water Storage Tank (RWST). The inspectors noted the licensee had evaluated for potential gas intrusion, including vortex prevention, from the RWST but had failed to evaluate for potential draining or vortex formation on the SAT during the injection phase.
The inspectors discussed their concerns with the licensee and preliminary results showed the SAT could be completely drained and air ingested into the CS system. This concern was entered into the Corrective Action Program (CAP) as AR 2068569.
A follow-up evaluation was performed by the licensee and determined the SAT would not drain down and no air entraining vortex should form. As a result the licensee determined the CS system remained operable.
Analysis:
The inspectors determined the failure to evaluate for potential gas intrusion from th e SAT into the CS system during the injection phase of a DBA was contrary to 10 CFR Part 50, Appendix B, Criterion III, "Design Control," and was a performance deficiency.
The performance deficiency was determined to be more than minor because it was associated with the Mitigating System s cornerstone attribute of equipment performance
, and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences.
Specifically, air intrusion into the CS system could affect the operability of the CS pumps by causing degraded performance and/or air bindi ng of the pumps.
The inspectors determined the finding could be evaluated using the Significance Determination Process in accordance with Inspection Manual Chapter (IMC) 0609, "Significance Determination Process," Attachment 0609.04," "Initial Characterization of Findings." Specifically, the inspectors used IMC 0609
, Appendix A
, "Significance Determination Process for Findings At
-Power", Exhibit 2, "Mitigating Systems Screening Questions", issued June 19, 2012
, and answered "yes" to Question A.1. The finding was a deficiency affecting the design or qualification of a mitigating structure
, system , or component (SSC)
, but the SSC maintained its operability. Specifically the licensee evaluated the CS system and determined the SSC remained operable. As a result, the finding screened as having a very
-low safety significance, i.e.
, Green.
5 The inspectors did not identify a cross
-cutting aspect associated with this finding. Specifically, the decision to not evaluate the injection phase was taken around the 2008 timeframe when the licensee was responding to the GL 2008-01. As a result the inspectors determined this finding was not representative of the licensee's current performance.
Enforcement
- Title 10 CFR Part 50, Appendix B, Criterion III, "Design Control," requires, in part, that measures shall be established to assure that design control measures shall provide for verifying or checking the adequacy of design, such as by the performance of design reviews, by the use of alternate or simplified calculation methods, or by the performance of a suitable testing program.
Contrary to the above, as of August 28, 2015, the licensee had not verified the adequacy of the CS design. Specifically, the licensee failed to evaluate for potential gas intrusion into the CS system due to drain down and/or vortexing in the SAT during the injection phase of a DBA.
This violation is being treated as a n NCV, consistent with Section 2.3.2 of the NRC Enforcement Poli cy. The violation was entered into the licensee's CAP as AR 2068569. As part of immediate corrective actions the licensee performed an evaluation which determined the SAT would not drain down and no air entraining vortex should form. As a result the licensee determined the CS system remained operable. (NCV 05000266/2015010
-01; 05000301/2015010
-01, Failure to Evaluate Containment Spray System for Potential Gas Intrusion
).
.2 Permanent Plant Modifications
a. Inspection Scope
The inspectors reviewed seven permanent plant modifications that had been installed in the plant during the last 3 years. This review included in
-plant walkdowns for
- portions of the Spent Fuel Pool Instrumentation modification
- reroute of transformer 1X04 low voltage side power cables modification; portions of the CS system associated with modified procedures; and the (G-05) gas turbine generator including control cabinets, supporting equipment, including batteries, support diesels, piping, supports, and the ai r inlet weather hood structure. The modifications were selected based upon risk significance, safety significance, and complexity. 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.
6 This inspection constituted seven permanent plant modification samples as defined in IP 71111.17-04.
OTHER ACTIVITIES
4OA2 Problem Identification and Resolution
.1 Routine Review of Condition Reports
a. Inspection Scope
The inspectors reviewed several 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 p lant modifications and evaluations of changes, tests, and experiments. 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 were identified.
4OA6 Management Meetings
.1 Exit Meeting
Summary O n August 28, 2015
, the inspector s presented the inspection results to Mr. McCartney , and other members of the licensee staff. The licensee personnel acknowledged the inspection results presented and did not identify any proprietary content. The inspectors confirmed that all proprietary material provided to the inspection team was identified and will be dispositioned in accordance with applicable processes.
ATTACHMENT:
SUPPLEMENTAL INFORMATION
SUPPLEMENTAL INFORMATION KEY POINTS OF CONTAC
T Licensee
- S. Aerts, Performance Improvement Manager
- D. DeBoer, Plant Manager
- F. Eichhorst, Design Engineering Configuration Management Supervisor
- D. Forter, Manager of Projects
- R. Harrsch, Engineering
Director
- D. Jensen, Design Engineering
- K. Kinjerski, Procedures Supervisor
- T. Lohr, Reactor Operator
- E. McCartney, Site Vice President
A. Nash Design Engineering Supervisor
- R. Parker, Chemistry Manager
- M. Rosseau, Electrical / I&C Design Supervisor
- G. Str har sky, NOS Manager
- R. Webber, Operations Director
- P. Wild, Design Engineering Manager
- J. Wilson, Maintenance Director
- B. Woyak, Licensing Manager
U.S. Nuclear Regulatory Commission
- D. Oliver, Senior Resident Inspector
LIST OF ITEMS OPENED, CLOSED AND DISCUSSED Opened and Closed
-01; 05000301/2015010
-01 NCV Failure to Evaluate Containment Spray System for Potential Gas Intrusion
(Section 1R17.1b) Discussed None
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 Title
Date or Revision 2010-002-01 EC 13251 - Reroute of 1X04 Low Voltage Side Power Cables to Façade, Rev. 1
05/22/12 2011-001 1A-01 and 1A-02 Loss of Voltage Relay Setpoint Changes 04/13/11 2011-003-01 Raise EOP Setpoint
O.2 (RWST Level to Stop Spray Pumps) to 17%
- Radiological Consequence
03/19/12 2011-004 Raise EOP Setpoint O.2 (RWST Level to Stop Spray Pumps) to 17%
- Containment Pressure & Temperature
05/29/11 2012-004 EC 260347
- 120V Vital Cable Short Circuit Protection for Units 1 and 2
11/04/12 2013-001 Unit 1 and Unit 2 Auxiliary Feedwater Pump Turbine Replacement
10/02/13 2014-004 EC 277545 "RCP Seal Replacement" 00 10 CFR 50.59 SCREENINGS
Number Description or Title
Date or Revision 2006-0275 Replacement of Potentially Overdutied Breakers and/or Fuses in MCCs 1B
-31, 2B-11, B-33, B-44, B-45, B-46, and B-47 05/02/07 2010-0186 "At-Risk" Installation to Replace Portions of Power Cables from 1X
-04 to Façade
09/23/10 2011-0093 AST Related RWST EOP Setpoint Changes
05/28/11 2012-0077 LIC 1756890, FSAR 6.1 and FSAR 9.2 Revision to Add GL 2008-01 Discussion
05/17/12 2012-0177-01 Replace Fuses D72
-18-07, D72-22-08, D72-22-10 10/13/13 2012-0119 EC 260347
- 120V Vital Cable Short Circuit Protection for Units 1 and 2
10/29/12 2012-0199 Revision to IT08A Following Establishment of 1MS
-2082 IST Acceptance Criteria
2/06/12 2013-0023 CHANGE FSAR TABLE 9.1
-1, COMPONENT COOLING SYSTEM COMPONENT DATA
2/15/13 2013-0043 Revise OI 168 "Emergency Diesel Generator Operability" 07/10/13 2013-0 046 G-05 Vibration Monitor: Enable Trip Multiplier Function for Resonant Speed Range
03/08/13 2014-0033 EC 281270
- Add 60th Cell to Batteries D
-05 and D-06 03/24/14 2014-0203-01 EOP-1.3: Transfer
to Containment Sump Recirculation
- Low Head Injection
2/18/14
10 CFR 50.59 SCREENINGS
Number Description or Title
Date or Revision 2014-0158 EC 27745 "RCP Seal Replacement" 00 2014-0170 Emergency Diesel Generator Operability
09/20/14 2015-0002 Revise AOP
-11-2 Requirements
01/15/15 2015-0004 Revised AOP
-8F for Area Radiation Monitor (72.48)
01/15/15 2013-0024 Revise TRM 3.7.7, OI 70, TS 33, TS 34, AOP13A, AOP
-8F, FSAR 9.6.1, OI I55, PC 97 Parts 1
-8 and 1(2)
-SOP-VNCC-001-4 to Allow 85F SW Inlet Temperature and to Specify Operability Limits on
Low Pump Bay Level for the G01/G02 EDGs and the Lower Elevations CFCs
03/15/13 2015-0029 Revise Normal And Adverse Containment Pressurizer Level Values In EOP 3.1
-3, 3.2, 3.3
2015-0032 Revise AOP
-22 Step 1.c RNO to 60 Hz
04/06/15 CALCULATIONS
Number Description or Title
Date or Revision CN-SEE-III-08-10 Point Beach Units 1 and 2 RHRS Cooldown Analysis for EPU to 1806 MWT NSSS Power
3 CN-TA-07-133 Evaluation of an Increased Delay for the Reactor Coolant Pump Undervoltage
Reactor Trip Function at Point Beach Nuclear Plant Units 1 and 2
N-92-005 125 VDC Coordination Analysis
N-93-056 D05 DC System Sizing, Voltage Drop and Short Circuit Calculation
N-93-059 D106 DC System Sizing, Voltage Drop and Short Circuit
Calculation 08 13-298 2012 CCW Heat Exchangers HX
-012A/B Thermal Performance Test Results Report
0 13-316 2011 CCW Heat Exchangers HX
-012A/B Thermal Performance Test Results Report
0 13-323 2009 SFP Heat Exchanger HX
-013A/B Thermal Performance Test Results 11/14/13 13-326 2012 SFP Heat Exchanger HX
-013A/B Thermal Performance Test Results
11/20/13 2002-0003 Service Water System Design Basis
4 2003-0046 Battery Chargers Sizing and Current Limit Set Point
2005-0037 Spent Fuel Pool Anti
-Siphon Provisions
2005-0054 Control Building GOTHIC Temperature Calculation
2008-005 4.16 KV 7 480 V Loss of Voltage and Underfrequency Relay Setting
2007-0002 Emergency Diesel Generator Frequency Uncertainty Calculation
2009-0012 Extended Power Uprate (EPU) Related EOP Setpoint Calculations
2010-0027 Miscellaneous Level EOP Setpoint Calculations
CALCULATIONS
Number Description or Title
Date or Revision 2010-01716 Eductor Analysis
- Ingestion of Air into Containment Spray Pump Suction
2013-0020 PAB Scenarios for Fukushima Coping
129187-E-0030 Cable Ampacity De
-Rating Calculation
CORRECTIVE ACTION PROGRAM DOCUMENTS INITIATED DURING INSPECTION
Number Description or Title
Date 02068569 2015 MOD/50.59 NRC Inspection Review of SW Temp Increase 08/20/15 02069124 Eval 2013-001 Contains Typos
08/24/15 02069296 NRC 50.59 / MOD Inspection: Analysis of SAT Emptying
08/25/15 02069400 DBD For CC not Maintained
09/28/15 02069630 NRC 50.59 / MOD Inspection: Potential for Error
--- 02069693 OI-169 EDG Operability Potentially Confusing
08/26/15 02069856 TRM 3.7.7 08/27/15 02069859 Discrepancy in CC HX Fouling
08/27/15 CORRECTIVE ACTION PROGRAM DOCUMENTS REVIEWED
Number Description or Title
Date 98-0036 IST Program Does Not List Requirement to Exercise and Stroke Time Spray Additive Tank Valves
01/06/98 0 1221210 Degraded Seat Material on Unit 1 SI
-854 Check Valves
09/15/95 01316511 Appendix R Associated Circuits 120VAC Coordination Concern 04/21/07 01659682 Fuse Replace for D72
-22-08 and -22-10, D72-18-07 and -18-10 06/10/11 01756590 Possible FSAR Update in Response to GL 2008
-01 04/19/12 01794746 Preconditioning Issue in IT
-08A TDAFP Test
08/17/12 0 1891067 COMP-NA CR SITE DRAWING AND PROCEDURE CHANGES POST MODIFICATIONS
07/22/13 01904504 Clarify FSAR 7.4.3 & 10.2 Re: AFWP Suction Transfer to
SW 09/17/13 0 1943408 FSAR Overstates Docketed Design Bases
2/25/14 01950740 DC Calculation Issues Identified
03/22/14 0 1979452 Historical 50.59 Evaluations Did Not Identify FSAR Impact 07/28/14 0 1985361 Additional Performance Issues with 50.59 Products
08/19/15 01996270 CLB Updates are Not Timely
10/05/14 01998626 Revise FSAR Section 8.1.4 to Update Ref. 2
10/13/14 02032684 AOP-22 Unit 1
03/16/15 0 2037428 Plant Changed Without an EC or 50.59
04/24/15 0 2038568 RCP No. 1 Seal EOP Setpoint Not Updated Since 1995
04/07/15
DRAWINGS Number Description or Title
Revision E-6 Sheet 1 Single Line Diagram
- 125V DC Dist. System
MCCK00000143, Sheet 3 P&ID Auxiliary Coolant System
MCCK00000422, Sheet 2 P&ID Auxiliary Coolant System
MESK00000259, Sheet 1 P&ID Safety Injection System
MSFK00000271, Sheet 1 P&ID Auxiliary Coolant System
MSIK00000165, Sheet 2 P&ID Safety
Injection System
MSIK00001248, Sheet 3 P&ID Safety Injection System
MODIFICATIONS
Number Description or Title
Revision EC 00 8075 Replacement of Potentially Overdutied Breakers and/or Fuses in MCCs 1B
-31, 2B-11, B-33, B-44, B-45, B-46, and B-47 00 EC 0 13251 Reroute of 1X04 Low Voltage Side Power Cables to Façade 00 EC 260287 REPLACEMENT OF ACCIDENT FAN COOLER HOSES WITH ALTERNATE DESIGN FLEX
EC 260313 UNIT 2 CONTAINMENT SUMP A LEVEL TRANSMITTER REPLACEMENT
EC 260347 120V Vital Cable Short Circuit Protection for Units 1 and
01 EC 260579 0P-032E-M, SW Pump Motor Equivalency
EC 260795 125 VDC Power Supply Reconfiguration for Unit 1 Turbine Auto
-Stop Trip (AST) / Emergency Trip (ET) Control Circuits
EC 260820 Install Test Switches for 4.16kV Bus Loss of voltage Relays - Unit 2 00 EC 261461 Install Test Switches for 4.16kV Bus Loss of voltage Relays - Unit 1 00 EC 261964 P-032D-M SW PUMP MOTOR REPLACED WITH SPARE MOTOR (+)
EC 270973 Westinghouse/Cutler Hammer DHP
-VR Breaker Enhancements
EC 271690 Evaluation of an Increased Delay for the Reactor Coolant Pump Undervoltage Reactor Trip Function at Point Beach Nuclear Plant Units 1 and 2
-Low Level Alarm Setpoint in STPT 11.1 and Other Affected Documents Per CALC PBNP
-IC-02-004-A 00 EC 276803 NRC Order Fukushima Strategy
- Spent Fuel Pool Instrumentation Upgrade
MODIFICATIONS
Number Description or Title
Revision EC 277206 Replace Fuses D72
-18-07, D72-18-10, D72-22-08, D72-22-10 01 EC 277545 RCP Seal Replacement
EC 281270 D-05 and D-06 Battery Addition
OTHER DOCUMENTS
Number Description or Title
Date or Revision DBD-11 Safety Injection & Containment Spray Pump Design Bases Document
EC 278678 G-05 Vibration Monitor: Enable Trip Multiplier Function for Resonant Speed Range
Evaluation 00456339
- Fast Acting Fuse
2/12/14 FPTE-2007-001 Technical Evaluation of Associated Circuit by Common Power Supply and by Common Enclosure
04/30/07 WEP-98-017 Containment Pressure and Temperature Increase During Recirculation Due to Loos of RHR Heat Exchanger Cooling
03/05/98 WO 40127282 07
H52-DHPVR2000-3; Breaker Overhaul
03/27/14 WO 40202961 01
D-06 Battery Modified Performance Test
01/21/14 WO 40220768
G-05 Apply Vib
Trip Block to Turbine Resonant Speed Range 03/08/13 WO 40301275 01
D-06/Connect 60th Cell into Battery D
-06 Per EC 281270 04/25/14 WO 40301276 01
D-05/Connect 60th Cell into Battery D
-05 06/18/14 WO 40302871 01
D-05 / D-07, Battery Surveillance Tests
07/15/14 WO 40344232 01
25V, Station Tech Spec Batteries Weekly Inspection
08/03/15 2013-0013 Change FSAR Table 9.1
-1, "Component Cooling System Component Data" 02/15/13 PROCEDURES
Number Description or Title
Revision --- Gas Accumulation Management Program
EOP 0.1 Reactor Trip Response
EOP-1.3 Unit 1
Transfer to Containment Sump Recirculation
- Low Head Injection
OI 168 Emergency Diesel Generator Operability
RMP 9046-4 Seven Day Pilot Cell Battery Surveillance
STPT 5.3 Pressurizer Pressure and Level Control
TS 84 Emergency Diesel Generator G
-04 Monthly
LIST OF ACRONYMS USE
D ADAMS Agencywide Documents Access and Management System
CAP Corrective Action Program
CFR Code of Federal
Regulations
DBA Design-Basis Accident
DRP Division of Reactor Project
s DRS Division of Reactor Safety
GL Generic Letter
IMC Inspection Manual Chapter
IP Inspection Procedure
NCV Non-Cited Violation
NEI Nuclear Energy Institute
NRC U.S. Nuclear Regulatory Commission
PARS Public ly Available Records System
RWST Refueling Water Storage Tank
SAT Spray Additive Tank
SSC Structure, System, and Component
E. McCartney
-2- In accordance with Title 10 of the Code of Federal Regulations
(10 CFR) 2.390, "Public Inspections, Exemptions, Requests for Withholding," of the NRC's "Rules of Practice," a copy
of this letter, its enclosure, and your response (if any) will be available electronically for public inspection in the NRC's Public Document Room or from the Publicly Available Records (PARS)
component of the NRC's Agencywide
Documents Access and Management System (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html
(the Public Electronic Reading Room).
Sincerely, /RA/
Robert
- C. Daley, Chi ef Engineering Branch 3
Division of Reactor Safety Docket Nos.
50-266; 50-301 License Nos. DPR-24; DPR-27 Enclosure:
I R 05000266/2015010; 05000301/2015010
w/Attachment: Supplemental Information
cc w/encl: Distribution via LISTSERV
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RidsNrrDirsIrib Resource
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ADAMS Accession Number ML15261A631 Publicly Available
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