IR 05000263/2019011

Triennial Inspection of Evaluation of Changes, Tests and Experiments Baseline Inspection Report 05000263/2019011

ML19289B473
Person / Time
Site:	Monticello
Issue date:	10/16/2019
From:	Robert Daley Engineering Branch 3
To:	Church C Northern States Power Company, Minnesota
References
IR 2019011
Download: ML19289B473 (11)

Text

ber 16, 2019

SUBJECT:

MONTICELLO NUCLEAR GENERATING PLANTTRIENNIAL INSPECTION OF EVALUATION OF CHANGES, TESTS AND EXPERIMENTS BASELINE INSPECTION REPORT 05000263/2019011

Dear Mr. Church:

On September 4, 2019, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Monticello Nuclear Generating Plant and discussed the results of this inspection with you and other members of your staff. The results of this inspection are documented in the enclosed report.

One Severity Level IV violation without an associated finding is documented in this report.

We are treating this violation as a Non-Cited Violation (NCV) consistent with Section 2.3.2.a of the Enforcement Policy.

If you contest the violation or significance or severity of the violation documented in this inspection report, 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; and the NRC Resident Inspector at Monticello Nuclear Generating Plant. This letter, its enclosure, and your response (if any) will be made available for public inspection and copying at http://www.nrc.gov/reading-rm/adams.html and at the NRC Public Document Room in accordance with Title 10 of the Code of Federal Regulations 2.390, Public Inspections, Exemptions, Requests for Withholding.

Sincerely,

/RA/

Robert C. Daley, Chief Engineering Branch 3 Division of Reactor Safety Docket No. 05000263 License No. DPR-22

Enclosure:

As stated

Inspection Report

Docket Number: 05000263 License Number: DPR-22 Report Number: 05000263/2019011 Enterprise Identifier: I-2019-011-0040 Licensee: Northern States Power Company - Minnesota Facility: Monticello Nuclear Generating Plant Location: Monticello, MN Inspection Dates: June 3, 2019, to June 7, 2019 Inspectors: M. Holmberg, Senior Reactor Inspector J. Robbins, Reactor Inspector D. Szwarc, Senior Reactor Inspector Approved By: Robert C. Daley, Chief Engineering Branch 3 Division of Reactor Safety Enclosure

SUMMARY

The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees performance by conducting a triennial inspection of evaluation of changes, tests and experiments baseline inspection at Monticello Nuclear Generating Plant in accordance with the Reactor Oversight Process. The Reactor Oversight Process is the NRCs program for overseeing the safe operation of commercial nuclear power reactors. Refer to https://www.nrc.gov/reactors/operating/oversight.html for more information.

List of Findings and Violations Failure to Report Changes in Peak Clad Temperature for Postulated Loss-of-Coolant Accidents Cornerstone Significance Cross-Cutting Report Aspect Section Not Applicable NCV 05000263/2019011-01 Not Applicable 71111.17T Open/Closed The inspectors identified a Severity Level IV Violation of Title 10 of the Code of Federal Regulations (CFR) 50.46, Acceptance Criteria for Emergency Core Cooling Systems for Lightwater Nuclear Power Reactors for the licensee's failure to report to the NRC changes in the limiting peak clad temperature (PCT) for postulated loss-of-coolant- accident (LOCA)scenarios identified in revision 2 of calculation 13-055 Core Spray and LPCI Flow Delivered to Reactor Vessel for Safety Analysis.

Additional Tracking Items None.

INSPECTION SCOPES

Inspections were conducted using the appropriate portions of the inspection procedures (IPs)in effect at the beginning of the inspection unless otherwise noted. Currently approved IPs with their attached revision histories are located on the public website at http://www.nrc.gov/reading-rm/doc-collections/insp-manual/inspection-procedure/index.html.

Samples were declared complete when the IP requirements most appropriate to the inspection activity were met consistent with Inspection Manual Chapter (IMC) 2515, Light-Water Reactor Inspection Program - Operations Phase. The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel to assess licensee performance and compliance with Commission rules and regulations, license conditions, site procedures, and standards.

REACTOR SAFETY

71111.17T - Evaluations of Changes, Tests, and Experiments Sample Selection (IP Section 02.01)

The inspectors reviewed the following evaluations, screenings, and/or applicability determinations for Title 10 of the Code of Federal Regulations (CFR) 50.59:

(1) SCR-10-0143, Implement the Crossflow Ultrasonic Flow Measurement System, Revision 2

(2) SCR-14-0147, Issue 120V Distribution Panel Voltage Drop Analysis, Revision 0

(3) SCR-15-0131, Operation with HPCI Steam Line Drain Trap Bypass (CV-2043) in the Open Position and Annunciator 3-B-10 Non-Functional, Revision 2

(4) SCR-15-0378 (Evaluation), Replace Standby Gas Treatment Controllers to Improve Reliability, Revision 1

(5) SCR-15-0438, Freeze Seal for SW-37 Replacement, Revision 0

(6) SCR-16-0043 (Evaluation), ECCS-LOCA Analysis Input Changes, Revision 0

(7) SCR-16-0076, Incorporate Updated MOV DC Curves into MOV Calculations, Revision 1

(8) SCR-16-0096, Add Steps to Reduce Impact on LS-23-90 when Unisolating HPCI Steam Lines, Revision 0

(9) SCR-16-0119, Owners Acceptance of Zachery Nuclear Core Spray and Residual Heat Removal (RHR) Piping, Revision 0

(10) SCR-16-0162, Revise USAR 12.02 for Clarification and to Better Align with NEI 08-05 FSAR Update Guidance, Revision 1

(11) SCR-16-0169, Evaluate Refuel Floor Loading when DSC Fuel Cask Staged on Floor, Revision 0

(12) SCR-16-0199, Operate System with Increased Levels in Tank, Revision 2

(13) SCR-16-0213, Turbine Bypass Valve Technical Specification Bases Clarification, Revision 1

(14) SCR-16-0307, Internal Flooding Analysis, Revision 0

(15) SCR-16-0354, 13 RHR Pump Motor Cooler Flow Rate During Surveillance Testing, Revision 0

(16) SCR-16-0377, Replace HPCI Level Switch LS-23-98 Topworks and Disable Auto Pumpdown Feature, Revision 1

(17) SCR-17-0054, Evaluation of Localized Thinning for 2017 Essential Service Water (ESW) Piping Inspections, Revision 0

(18) SCR-17-0147, Revision to USAR Section 8.8.7 Cables, Revision 0

(19) SCR-17-0154, PS17-3" HELB Crack/Break Elimination, Revision 0

(20) SCR-17-034 (Evaluation), Increase GE14 Nuclear Fuel from 35 to 37 GWD/MTU for Core Average End of Cycle Exposure for Radiological Consequences Evaluations, Revision 0

(21) SCR-18-0090, HPCI Room Transient Temperature, Revision 0

(22) SCR-18-0132, Outboard C MSIV Fast Closure, Revision 0

(23) SCR-18-0146, Monticello Calculation 04-133, Outboard Main Steam Isolation Valve Calculation Update, Revision 0

(24) SCR-19-0021, Revise Division 2 RHRSW Pump and Valve Tests, Revision

INSPECTION RESULTS

Failure to Report Changes in Peak Clad Temperature for Postulated Loss-of-Coolant Accidents Cornerstone Severity Cross-Cutting Report Aspect Section Not Severity Level IV Not 71111.17T Applicable NCV 05000263/2019011-01 Applicable Open/Closed The inspectors identified a Severity Level IV Violation of Title 10 of the Code of Federal Regulations (CFR) 50.46 Acceptance Criteria for Emergency Core Cooling Systems for Lightwater Nuclear Power Reactors for the licensee's failure to report to the NRC changes in the limiting peak clad temperature (PCT) for postulated loss-of-coolant- accident (LOCA)scenarios identified in revision 2 of calculation 13-055 Core Spray and LPCI Flow Delivered to Reactor Vessel for Safety

Analysis.

Description: On February 12, 2016, the licensee issued revision 0A of calculation 1180, ECCS [Emergency Core Cooling System]-LOCA [Loss-of-Coolant-Accident]

SAFER/GESTR, to accept vendor General Electric Hitachi (GEH) records as site quality assurance records including GEH analysis 0000-0163-2998-R0 Monticello ECCS LOCA Evaluation for Modified Low Pressure ECCS Injection Performance Curves (LPCS [Low Pressure Core Spray] and LPCI [Low Pressure Coolant Injection]).

On April 13, 2016, the licensee issued Revision 2 to calculation 13-055, Core Spray and LPCI Flow Delivered to Reactor Vessel for Safety Analysis, which provided LPCS and LPCI reactor vessel injection flow versus reactor vessel-to-torus differential pressure curves suitable for use as replacements for those used in current safety analyses and as new inputs in future safety analyses (applicable to GEH 14C fuel in service at the Monticello Nuclear Generating Plant). In calculation 13-055, the licensee concluded that the GEH supplemental analysis 0000-0163-2998-R0 demonstrated improved run out flow for LPCS and resulted in acceptable margins for PCT in design basis LOCA and small break LOCA scenarios. Additionally, in calculation 13-055 and GEH analysis 0000-0163-2998-R0, the licensee identified various scenarios that changed PCT by more than 50 °F for the limiting LOCA transients from that recorded for the last acceptable model, but none of the PCT changes resulted in exceeding the PCT limit of 2200°F identified in 10 CFR 50.46. Subsequently, the licensee confirmed that in 2016, the revised LPCS pump flow parameters were selected and incorporated into Updated Final Safety Analysis Report Table 14.7-8, Core Spray System Parameters that were consistent with Case 2 from the GEH supplemental analysis 0000-0163-2998-R0 and which resulted in a -14 °F change in PCT.

NRC regulation 10 CFR 50.46 identifies the acceptance criteria for emergency core cooling systems in lightwater nuclear power reactors and requires that cooling system performance be calculated in accordance with an acceptable evaluation model for a number of postulated LOCAs of different sizes, locations, and other properties to ensure the most severe postulated LOCA is evaluated. This regulation also requires that for each change or error discovered in the evaluation model or in the application of such a model, the effect on the limiting ECCS analysis, be reported to the NRC at least annually. In this case, the inspectors identified that the licensee had made changes associated with application of the ECCS model and had not reported this result in an annual report to the NRC. Specifically, in Revision 2 of calculation 13-055, the licensee adopted revised LPCS pump performance curves for use in ECCS scenarios that impacted the limiting PCT analysis and did not report this information to the NRC.

Corrective Actions: The licensee entered the issue in their corrective action system on June 4, 2019.

Corrective Action References: 501000028099, 50.59 missed 10 CFR 50.46 report

Performance Assessment:

The inspectors determined this violation was associated with a minor performance deficiency. Consistent with the guidance in IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, the inspectors answered no to the more-than-minor screening questions. The inspectors also reviewed the examples of minor issues in IMC 0612, Appendix E, Examples of Minor Issues and found no examples related to this issue.

Enforcement:

The Reactor Oversight Process (ROP)s significance determination process does not specifically consider the regulatory process impact in its assessment of licensee performance. Therefore, it is necessary to address this violation which impedes the NRCs ability to regulate using traditional enforcement to adequately deter non-compliance.

Severity: In accordance with examples included in Section 6.9.d of the NRC Enforcement Policy, this issue was screened as a Severity Level IV Violation.

Violation: Title 10 CFR 50.46 Acceptance Criteria for Emergency Core Cooling Systems for Lightwater Nuclear Power Reactors Title 10 CFR 50.46a(3)

(ii) requires in part, for each change to or error discovered in an acceptable evaluation model or in the application of such a model that affects the temperature calculation, the license, shall report the nature of the change or error and its estimated effect on the limiting ECCS analysis to the Commission at least annually as specified in 10 CFR 50.4 or 10 CFR 52.3 of this chapter, as applicable.

Contrary to the above, between April 13, 2016, and September 4, 2019, the licensee failed to report the nature of a change in the application of a model that affects the temperature calculation, and its estimated effect on the limiting ECCS analysis to the Commission at least annually. Specifically, on April 13, 2016, in Revision 2 of calculation 13-055, the licensee evaluated and accepted modified ECCS injection pump performance curves that resulted in a -14F change to the PCT and failed to report the nature of the change or error and its estimated effect on the limiting ECCS analysis to the Commission.

Enforcement Action: This violation is being treated as an non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.

EXIT MEETINGS AND DEBRIEFS

The inspectors confirmed that proprietary information was controlled to protect from public disclosure.

• On September 4, 2019, the inspectors presented the triennial inspection of evaluation of changes, tests and experiments baseline inspection results to Mr. Christopher Church, and other members of the licensee staff.

• On June 7, 2019, the inspectors presented the Interim Exit inspection results to Mr. Don Barker and other members of the licensee staff.

DOCUMENTS REVIEWED

Inspection Type Designation Description or Title Revision or

Procedure Date

71111.17T Calculations01-174 Minimum Required RHRSW Pressure at RHR Heat 3

Exchanger

07-045 RHR Pump Motor Modei5K511DT5410 Cooling Coil 2A

Minimum Flow Evaluation

13-055 Core Spray and LPCI Flow Delivered to Reactor Vessel For 2

Safety Analyses14-014 Distribution Panel Y20 Voltage Drop Analysis 0

16-072 480V Coordination Study 0

Calculation 11- ECCS-LOCA SAFER/GESTR 0A

180

Corrective Action 500000087543 50.59 - Formal Evaluation for HPCI Drain Line 03/16/2016

Documents 500001516105 50.59 - HPCI SR Test Inconsistent with TS Bases 03/17/2016

500001560841 Cables not protected from fault on the circuit 06/12/2017

501000004050 EC28685 missing required 50.59 screening 10/19/2017

501000010972 50.59 Screening Graded a 1 at QRT 04/19/2018

501000011895 50.59 AD Need Revision 05/11/2018

501000014807 Outboard MSIV Calculation 04-133 Issues 07/25/2018

601000000116 PS17-3" HELB Cracks/Breaks Elimination 08/22/2017

Corrective Action 501000028099 19-50.59-Missed 10 CFR 50.46 Report 06/04/2019

Documents 501000028218 19-50.59-Piping Corrosion Allowance Doc 06/06/2019

Resulting from 501000028233 19-50.59-Typo in Superseded Calculation 14-014 06/06/2019

Inspection 501000028239 19-50.59 Replaced ESW pipe below Calculated Tmin 06/06/2019

501000028260 19-50.59 Pipe Replaced after Calculation Remaining Service 06/06/2019

Life

501000028261 19-50.59 SW-MIC Service Life Calculation 06/06/2019

Drawings NH-36249 P&ID High Pressure Coolant Injection System (Steam Side) 84

NX-13142-43 RCIC Primary Steam, 6

Engineering 25889 OPERATING WITH HPCI CV-2043 (STEAM TRAP 000

Changes BYPASS) OPEN

26866 EVALUATION OF IMPACT OF HPCI DRAIN LINE BYASS 000

FLOW TO CONDENSER

Inspection Type Designation Description or Title Revision or

Procedure Date

27905 REPLACE HPCI LEVEL SWITCH LS-23-98 TOPWORKS 0

AND DISABLE AUTO PUMP DOWN FEATURE

Engineering 0000-0163-2998- Monticello ECCS LOCA Evaluation for Modified Low 0

Evaluations R0 Pressure ECCS Injection Performance Curves (LPCS and

LPCI)

608000000236 MSIV Steam Leakage Evaluation 0

SCR-15-0408 Replace RHR Pressure Switches PS-10-101A/B/C/D 1

SCR-17-189 Replace DPI-2994C and DPI-2994D 0

Miscellaneous 3830-17-041 Fire Protection Change Review for PCR 01550124 4

Regulatory Issue NRC Staff Position on use of the Westinghouse Crossflow 09/27/2007

Summary 2007- Ultrasonic Flow Meter for Power Uprate OR Power Recovery

Procedures 0255-05-IA-1-2 B RHRSW QUARTERLY PUMP AND VALVE 87

TESTS

255-05-III-2A COMPREHENSIVE 12 RHRSW PUMP AND 33

VALVE TESTS

C.6-003-B-02 HPCI TURBINE EXH HI DRAIN POT LEVEL 5

C.6-003-B-10 HPCI TURBINE INLET HI DRAIN POT LEVEL 9

8