ML16113A346
| ML16113A346 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 04/19/2016 |
| From: | Daley R C Engineering Branch 3 |
| To: | Gardner P A Northern States Power Company, Minnesota |
| References | |
| IR 2016008 | |
| Download: ML16113A346 (12) | |
See also: IR 05000263/2016008
Text
April 19, 2016
Mr. Peter Site Vice President Monticello Nuclear Generating Plant Northern States Power Company, Minnesota 2807 West County Road 75 Monticello, MN 553629637
SUBJECT: MONTICELLO NUCLEAR GENERATING PLANT - EVALUATIONS OF CHANGES, TESTS, AND EXPERIMENTS AND PERMANENT PLANT MODIFICATIONS BASELINE INSPECTION REPORT 05000263/2016008
Dear Mr. Gardner:
On March 24, 2016, the U.S. Nuclear Regulatory Commission (NRC) completed an Evaluations of Changes, Tests, and Experiments and Permanent Plant Modifications inspection at your Monticello Nuclear Generating Plant. The enclosed inspection report documents the inspection results which were discussed on March 24, 2016, with Mr. M. Lingenfelter and other members of your staff. The inspection examined activities conducted under your license as they relate to safety and The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel. No findings were identified during this inspection. In accordance with Title 10 of the Code of Federal Regulations of this letter, its enclosure, and your response (if any) will be available electronically for public 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, Chief Engineering Branch 3 Division of Reactor Safety Docket No. 50263 License No. DPR22
Enclosure:
IR 05000263/2016008 cc: Distribution via LISTSERV Enclosure U. S. NUCLEAR REGULATORY COMMISSION REGION III Docket No: 50-263 License No: DPR-22 Report No: 05000263/2016008 Licensee: Northern States Power Company, Minnesota Facility: Monticello Nuclear Generating Plant Location: Monticello, MN Dates: February 29 thru March 24, 2016 Inspectors: Alan Dahbur, Senior Reactor Inspector (Lead) Jorge J. Corujo-Sandín, Reactor Inspector Michael A. Jones, Reactor Inspector Approved by: Robert C. Daley, Chief Engineering Branch 3 Division of Reactor Safety 2
SUMMARY
Inspection Report 05000263/2016008; 02/29/2016 03/24/2016; Monticello Nuclear Generating Plant; 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. program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-, dated February 2014.
NRC-Identified
and Self-Revealed Findings No findings were identified.
Licensee-Identified Violations
No violations were identified.
3
REPORT DETAILS
REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity
1R17 Evaluations of Changes, Tests, and Experiments and Permanent Plant Modifications
.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 Document 96-10 CFR e acceptability of the completed evaluations, and screenings. The Nuclear Energy Institute document was endorsed by CFR 50.59, e inspectors also consulted Part CFR Guidance for 10 CFR This inspection constituted 4 samples of evaluations and 13 samples of screenings and/or applicability determinations as defined in Inspection Procedure 71111.17-04.
b. Findings
(Open) Unresolved Item 05000263/2016008-01, Failure to provide acceptable Alternate Methods of Decay Heat Removal
Introduction:
The inspectors identified an Unresolved Item associated with Technical Shutdown Cooling System Cold alternate methods of decay heat removal described in Operations Manual C.4-B.03.04.A, Loss of Normal Shutdown Cooling,ing resulting from the loss of one or two RHR subsystems while in MODE 4 with high decay heat load.
4
Description:
The Limiting Condition for Operation (LCO) 3.4.8 of TS ual Heat Removal Shutdown Cooling System Cold Shutdown, required in Mode 4, two RHR shutdown cooling subsystems shall be operable, and, with no recirculation pump in operation, at least one RHR shutdown cooling subsystem shall be in operation. The TS Bases Section 3.4.8, indicated that an operable RHR shutdown cooling subsystem consisted of one operable RHR pump, one heat exchanger, the associated piping and valves, and the necessary portions of the RHR Service Water System System capable of providing c The TS Bases Section 3.4.8 further indicated that the two subsystems have a common suction source and were allowed to have a common heat exchanger and common discharge piping. Thus, to meet the LCO, both pumps in one loop or one pump in each of the two loops must be operable. Since the piping and heat exchangers were passive components that were assumed not to fail, they were When TS 3.4.8, LCO could not be met, Condition A,one or two RHR shutdown cooling subsystems inoperable, method of decay heat removal was available for each inoperable RHR shutdown cooling hour, and once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter. The TS Bases 3.4.8 for Condition A indicated that with one of the two required RHR shutdown cooling subsystems inoperable, the remaining subsystem was capable of providing the required decay heat removal. However, the overall reliability was reduced, therefore, an alternate method of decay heat removal must be provided. With both RHR shutdown cooling subsystems inoperable, an alternate method of decay heat removal must be provided in addition to that provided for the initial RHR shutdown cooling subsystem inoperability. This was to ensure the re-establishment of backup decay heat removal capabilities, similar to the requirements of the LCO. The bases further stated that the required cooling capacity of the alternate method should be ensured by verifying (by calculation or demonstration) its capability to maintain or reduce temperature. Alternate methods that can be used included (but not limited to) the Reactor Water Cleanup System by itself or using feed and bleed in combination with Control Rod Drive System or Condensate/Feed Systems. Abnormal Procedure, Operations Manual C.4-removal. The inspectors noticed that except for the alternate method as described below in the G-EK-1-45, the licensee was not able to show by calculation or demonstration that the systems and methods credited in this procedure would be capable of providing sufficient heat removal capability or appropriate levels of redundancy as required by TS 3.4.8.
The G-EK-1-45 was a General Electric Letter to Northern States Power, Subject: Cold Shutdown Capability Report, dated April 22, 1981. This letter provided a report which described the capability of the Monticello Nuclear Generating Plant to achieve cold shutdown using only safety class systems and assuming the worst single failure. The alternate shutdown decay heat removal method used in the report credited combinations of the RHR pumps and heat exchangers in the suppression pool cooling mode of RHR to ensure suppression pool water temperatures were below the design limit. This method utilized the core spray system and safety relief valves to circulate reactor inventory to remove decay heat from the reactor.
5 The inspectors noted that calculations supporting the above alternate strategy utilized an RHR subsystem that could be inoperable and/or unavailable and therefore may not be credited to comply with TS 3.4.8. Specifically, the inspectors were concerned that while the plant was in mode 4, with a credited one subsystem inoperable, the licensees credited alternate decay heat removal method that relied on an RHR subsystem, to perform the required suppression pool cooling function. The inspectors were concerned that relying on the only operable RHR subsystem for the alternate method did not meet the intent of the TS requirement as described in the TS Bases. Furthermore, the inspectors noticed for Mode 4 with two RHR subsystems inoperable, the licensee failed to verify by calculation or demonstrations that two additional redundant alternate decay heat removal methods existed with sufficient capacity to maintain the average reactor coolant temperature below 212 degrees Fahrenheit. During the inspection, the licensee indicated that the Boiling Reactor Owners Group was in the process of developing a draft TS Task Force Traveler to address the requirement of TS 3.4.8 and its Bases. Based on the information above, the inspectors were concerned that the plant Operations Manual was inadequate and failed to include alternate decay heat removal methods that would enable the licensee to comply with the requirement of TS 3.4.8. The Othat written procedures shall be established, implemented, and maintained covering the emergency operating procedures. The inspectors determined that this issue was unresolved pending the actions by the licensee and the Boiling Reactor Owners Group and the NRC review of these actions. The licensee entered to their Corrective Action Program as AR 01516098. (URI 05000263/2016008-01, Failure to provide acceptable Alternate Methods of Decay Heat Removal)
.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 portions of the high-pressure coolant injection steam drain line system, the Emergency Diesel Generator Fuel Oil Transfer System, including the Diesel Fuel Oil pump house, the new diesel fuel oil pumps installed in the day tank room, and portions of the fuel oil storage tank tornado missile protection modifications. 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.
6 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 eight permanent plant modification samples as defined in Inspection Procedure 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 Title 10 of the Code of Federal Regulations, Part 50.59 evaluations and permanent plant modifications. The inspectors reviewed these documents to evaluate the effectiveness of corrective actions related to permanent plant 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
On March 24, 2016, the inspectors presented the inspection results to Mr. M. Lingenfelter, 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
KEY POINTS OF CONTACT
Licensee
- M. Lingenfelter, Director of Engineering
- A. Gonnering, Design Engineering
- M. Kelly, Performance Assurance Manager
- J. Gausman, Engineering
- A. Ward, Regulatory Affairs Manager
- T. Hurrle, Design Engineering Manager
- B. Halvorson, Engineering
- A. Kouba, Regulatory Affairs
- D. Alstad, Design Engineer
- E. Watzel, Electrical Design Engineering Supervisor
- U.S. Nuclear Regulatory Commission P. Zurawski, Senior Resident Inspector
- P. LaFlamme, Acting Senior Resident Inspector
- D. Krause, Resident Inspector
LIST OF ITEMS
OPENED, CLOSED AND DISCUSSED
Opened
- 05000263/2016008-01 URI Failure to provide acceptable Alternate Methods of Decay Heat Removal (Section 1R17.1b) Closed and
Discussed
None LIST OF ACRONYMS USED ADAMS Agencywide Documents Access and Management System CFR Code of Federal Regulations LCO Limiting Condition for Operation NRC U.S. Nuclear Regulatory Commission PARS Publicly Available Records System RHR Residual Heat Removal TS Technical Specifications
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.
- 10
- CFR 50.59 EVALUATIONS Number Description or Title Revision
- SCR-12-0559 HPCI Logic Change to Provide Margin to
- MO-2035 and #16 Battery
- 1
- SCR-13-0554 External Flooding Protection Strategy Change
- 0
- SCR-15-0202 Evaluation of EPG/SAG, Revision 3 0
- SCR-16-0024 Disconnect Faulty 46-19 PIP Over-Travel Input 0
- 10
- CFR 50.59 SCREENINGS Number Description or Title Revision
- SCR-13-0696 Revise EDG Base Tank Fuel Oil Level Calculation 90-023 0
- SCR-14-0074 Time Delay Relay 97-29 and 97-31 Setpoint Change 0
- SCR-14-0413 Temp Rev to C.6-006-A-01and C.6-006-A-02
- SCR-14-0415
- USAR-06.06 Revision 0
- SCR-14-0421
- SCR-14-0512 Safety and Seismic Classification of the DG/RF and DG/RV Relays 0
- SCR-14-0542 RHRSW and Emergency Service Water TS Bases Changes
- SCR-14-0591 Fuel Oil Separation
- 4
- SCR-14-0593 Revise Calculation 94-086 on SRV Accumulation Allowable Leakage Rates 0
- SCR-15-0115 C.4-B.09.02.A Revision to Resolve CAP
- SCR-15-0193 Room Heat Up Calculation Revisions for SBO
- SCR-15-0291 Revise Maximum Volume of EDG Base Tank in 90-023 0
- SCR-15-0292 Diesel Pump House Heat Up Calculation 0
- CALCULATIONS Number Description or Title Revision 03-089 Inservice Testing Acceptance Criteria 3 09-106 CSP Motor-Oil and Bearing Operating Temperatures without Cooling Water 1 09-176 Evaluation for Debris Disposition in Supply Pipe and Motor Cooler Tube 0 09-178 Time to reach the RHRSW Pump Motor Cooling Line Strainer Limiting Pressure Differential 0 14-025 Instrument Setpoint Calculation
- Time Delay for Transfer to EDG on Loss of Voltage 0 90-023
- EDG Fuel Oil Train Separation 3 92-224 Emergency Diesel Generator Loading 6A 94-086 Max Allowed Leakage Rates and Test Acceptance Criteria for SRV 5
- CORRECTIVE ACTION PROGRAM DOCUMENTS INITIATED DURING INSPECTION Number Description or Title Date
- 1510936 Incomplete EC Record Copy 02/03/2016
- 1514133 Clarification for USAR Section 8.4.1.3 03/01/2016
- 1514202 Page missing from
- WO 00491265 Record 03/02/2016
- 1514369 Screening
- SCR 14-0421 Answered Question Incorrectly 03/03/2016
- 1514464 EDG Building Roof
- FOI 91-0265 03/03/2016
- 1515054 Bases for Procedure A.6 Contain Incorrect Statements 03/09/2016
- 1515688 Signs of Leakage around
- FO-11-3 03/15/2016
- 1515716 NRC not Provided with Latest Copy of EC23085 03/15/2016
- 1515907 Formal Evaluation for HPCI Drain Line Bypass Flow 03/16/2016
- 1515939 Question Raised on
- CRD 46-19 03/16/2016
- 1516101 Core Spray Motor Cooling Design Basis Question 03/17/2016
- 1516105 HPCI SR Test Inconsistent with TS Bases 03/17/2016
- 1516106 RCIC Surveillance Required Test Inconsistent with TS Bases 03/17/2016
- CORRECTIVE ACTION PROGRAM DOCUMENTS REVIEWED Number Description or Title Date
- 952310 M91064A Quarterly Backflushing of Residual Heat Removal system and Core Spray Pump Motors 07/27/1991
- CA 90-023 09/03/2003
- 01355853 Update UFSAR for External Flooding Description Discrepancy 10/22/2012
- 01414416 Diesel Fuel Oil Temperature in Fuel Oil Transfer House is not Known 12/17/2013
- 01420875-03 Condition Evaluation on EDG Base Tank Level Issues 04/04/2014
- 01424477 Appendix R Fire Strategy for Fire Area XII incorrect 03/27/2014
- 1478798 EG Transfer Relay not Classified as Safety Related 05/13/2015
- 1484554
- RHRSW-29-2 Handwheel/Stem Sheared off 06/29/2015
- 1502700 Catastrophic fail of
- MO-1900 11/19/2015
- DRAWINGS Number Description or Title Revision
- NF-36175 Single Line Diagram
- Station Connections 85 104B2506 Connection Diagram
- Control Rod Drive Position Indicator Probe
- NH-46250 P&ID
- High Pressure Coolant Injection System 83
- NE-36399-9 Essential Bus Transfer Circuit
- Division I 77
- NE-36399-9B Essential Bus Transfer Circuits
- Division
- II 78
- NF-36061 Equipment Location - 76
- NF-36750 Standby Diesel Generator Building 8
- NH-36241-1 Reactor Pressure Relief P&ID 78
- NH-36051 P&ID Diesel Oil System 85
- NH-178639-1 Levee Alignment and Bin Wall Plan 4
- NF-119034-1 #11/#12 DG Fuel Oil System Isometric 78
- NH-36253 P&ID Standby Liquid Control System 80
- NH-36249 P&ID (Steam Side) High Pressure Coolant Injection System 82
- NX-13142-42 Primary Steam & HPCI System 78
- MODIFICATIONS Number Description or Title Revision
- LT-5200 River Level Setpoint Change for Upper Value
- EC-21999 Equivalency Evaluation:
- RHRSW-17 is the emergency injection check valve for the RHR to RSW crosstie 0
- EC-22008 Monticello 125V #12 Battery Modified Performance Test Profile 0
- 0
- EC-23616 Revise Setpoints for Relays 97-29 and 97-31 0
- EC-23857 Recirc Pump Seal Water Piping 0
- CV-2043 (Steam Trap Bypass) Open 0
- OTHER DOCUMENTS Number Description or Title Date or Revision 10040-A-020 Technical Specification for Steel Roof Deck 2
- FOI 91-0265 Qualification of the EDG Building Roof for Accumulated Snow Load 04/18/1994
- FG-E-SE-03 50.59 Resource Manual 5
- WO 00505386-30 EC23085 Pre-Op Testing Division I 05/06/2015
- WO 00505386-29 EC23085 Pre-Op Testing Division
- II 04/26/2015 257HA354 Technical Specification for High Pressure Coolant Injection System 2 G-EK-1-45 Cold Shutdown Capability Report 04/22/1981
- SRI 95-002 Core Spray Pump Motor Without Water Cooling 09/28/1995
- EE 25506 RFO27 Decay Heat Evaluation
- PROCEDURES Number Description or Title Revision 0075 Control Rod Drive Coupling Test 19 C.06-006-C-01 Diesel Oil Storage Tank T-44 Hi Low Level 6 C.06-006-C-02 Diesel Oil Storage Tank T-44 Low-Low Level 6 C.06-006-C-03 Division 1 EDG P-160A & P-160C Not Running 6 C.06-006-C-06 Diesel Gen Tank T-160A Level/Flow Low
- 4 2014-02 Turbine Building Outside 27 A.6 Acts of Nature 53 0255-17-ID-1 Master Alternate Nitrogen System Tests 25 0255-17-ID-15 SRV
- RV-71D and
- RV-2-71G Pneumatic Supply Leakage Test 13 Ops Man C.4-B.03.04.A Loss of Normal Shutdown Cooling 15 1339
- ECCS Pump Motor Cooler Flush 35 9111-01 Shutdown Cooling Division I Protected System Ticket Checklist 6 2270 Critical Safety System Checklist 11
- OWI-02.03
- Operator Rounds, Turbine Building West 64 Ops Man B. 03.01-05 Core Spray Cooling System 42 0255-05-1A-1-2 B RHR SW Quarterly Pump and Valve Test 82
- April 19, 2016 Mr. Peter A. Gardner Site Vice President Monticello Nuclear Generating Plant Northern States Power Company, Minnesota 2807 West County Road 75 Monticello, MN
- 553629637 SUBJECT:
- MONTICELLO NUCLEAR GENERATING PLANT - EVALUATIONS OF CHANGES, TESTS, AND EXPERIMENTS AND PERMANENT PLANT MODIFICATIONS BASELINE INSPECTION REPORT 05000263/2016008 Dear Mr. Gardner: On March 24, 2016, the U.S. Nuclear Regulatory Commission (NRC) completed an Evaluations of Changes, Tests, and Experiments and Permanent Plant Modifications inspection at your Monticello Nuclear Generating Plant.
- The enclosed inspection report documents the inspection results which were discussed on March 24, 2016, with Mr. M. Lingenfelter and other members of your staff. The inspection examined activities conducted under your license as they relate to safety and compliance rules and regulations and with the conditions of your license.
- The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel. No findings were identified during this inspection.
- In accordance with Title 10 of the Code of Federal Regulations this letter, its enclosure, and your response (if any) will be available electronically for public inspecPublic 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, Chief Engineering Branch 3 Division of Reactor Safety Docket No. 50263 License No. DPR22 Enclosure:
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