IR 05000397/2021012

Design Basis Assurance Inspection (Programs) Inspection Report 05000397/2021012

ML21306A196
Person / Time
Site:	Columbia
Issue date:	11/03/2021
From:	Vincent Gaddy Division of Reactor Safety IV
To:	Scheutz R Energy Northwest
References
IR 2021012
Download: ML21306A196 (13)

Text

November 3, 2021

SUBJECT:

COLUMBIA GENERATING STATION - DESIGN BASIS ASSURANCE INSPECTION (PROGRAMS) INSPECTION REPORT 05000397/2021012

Dear Mr. Scheutz:

On September 30, 2021, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Columbia Generating Station 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 finding of very low safety significance (Green) is documented in this report. This finding involved a violation of NRC requirements. We are treating this violation as a non-cited violation (NCV) consistent with Section 2.3.2 of the Enforcement Policy.

If you contest the violation or the 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 IV; the Director, Office of Enforcement; and the NRC Resident Inspector at Columbia Generating Station.

If you disagree with a cross-cutting aspect assignment 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 U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region IV; and the NRC Resident Inspector at Columbia Generating Station.

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, Signed by Gaddy, Vincent on 11/03/21 Vincent G. Gaddy, Chief Engineering Branch 1 Division of Reactor Safety Docket No. 05000397 License No. NPF-21

Enclosure:

As stated

Inspection Report

Docket Number: 05000397 License Number: NPF-21 Report Number: 05000397/2021012 Enterprise Identifier: I-2021-012-0002 Licensee: Energy Northwest Facility: Columbia Generating Station Location: Richland, WA Inspection Dates: September 13, 2021 to September 30, 2021 Inspectors: W. Cullum, Reactor Inspector D. Reinert, Senior Reactor Inspector F. Thomas, Reactor Inspector Approved By: Vincent G. Gaddy, Chief Engineering Branch 1 Division of Reactor Safety

SUMMARY

The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees performance by conducting a design basis assurance inspection (programs) inspection at Columbia Generating Station, 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 perform a cause evaluation after exceeding an administrative limit during local leak rate test of a containment isolation valve Cornerstone Significance Cross-Cutting Report Aspect Section Barrier Integrity Green [P.2] - 71111.21N.

NCV 05000397/2021012-01 Evaluation 02 Open The inspectors identified a Green finding and associated non-cited violation of 10 CFR 50,

Appendix B, Criterion V, "Procedures," when the licensee failed to follow procedures during local leak rate testing of a containment isolation valve.

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.

Starting on March 20, 2020, in response to the National Emergency declared by the President of the United States on the public health risks of the coronavirus (COVID-19), inspectors were directed to begin telework. In addition, regional baseline inspections were evaluated to determine if all or a portion of the objectives and requirements stated in the IP could be performed remotely. If the inspections could be performed remotely, they were conducted per the applicable IP. In some cases, portions of an IP were completed remotely and on site. The inspections documented below met the objectives and requirements for completion of the IP.

REACTOR SAFETY

71111.21N.02 - Design-Basis Capability of Power-Operated Valves Under 10 CFR 50.55a Requirements POV Review (IP Section 03)

The inspectors:

a. Determined whether the sampled POVs are being tested and maintained in accordance with NRC regulations along with the licensees commitments and/or licensing bases.

Specific Guidance b. Determined whether the sampled POVs are capable of performing their design-basis functions.

c. Determined whether testing of the sampled POVs is adequate to demonstrate the capability of the POVs to perform their safety functions under design-basis conditions.

d. Evaluated maintenance activities including a walkdown of the sampled POVs (if accessible).

(1) High Pressure Core Spray Discharge Isolation Valve, HPCS-V-4

(2) Reactor Core Isolation Cooling Turbine Steam Supply Isolation Valve, RCIC-V-45

(3) Reactor Core Isolation Cooling Discharge to Reactor Pressure Vessel Isolation Valve, RCIC-V-13

(4) Residual Heat Removal Pump 2B Minimum Flow Control Valve, RHR-V-64B

(5) Drywell Floor Drain Discharge Containment Isolation Valve, FDR-V-3

(6) Residual Heat Removal Suppression Chamber Spray Header Isolation Valve, RHR-V-27B

(7) Reactor Closed Cooling Supply to Drywell Cooling Loads Isolation Valve, RCC-V-5

(8) Reactor Feedwater Loop A Supply to Reactor Pressure Vessel Isolation Valve, RFW-V-65A

(9) Main Steam Isolation Valve 28C, MS-V-28C

(10) Reactor Core Isolation Cooling Turbine Steam Supply Inboard Containment Isolation Valve, RCIC-V-63

(11) Residual Heat Removal Shutdown Cooling Suction Isolation Valve, RHR-V-9

(12) Main Steam Atmospheric Depressurization System Safety Relief Valve 5C, MS-RV-5C

INSPECTION RESULTS

Failure to perform a cause evaluation after exceeding an administrative limit during local leak rate test of a containment isolation valve Cornerstone Significance Cross-Cutting Report Aspect Section Barrier Integrity Green [P.2] - 71111.21N.0 NCV 05000397/2021012-01 Evaluation 2 Open The inspectors identified a Green finding and associated non-cited violation of 10 CFR 50, Appendix B, Criterion V, "Procedures," when the licensee failed to follow procedures during local leak rate testing of a containment isolation valve.

Description:

The inspectors reviewed corrective action documents associated with the two most recent local leak rate tests (LLRTs) of valve RHR-V-27B, the Residual Heat Removal to suppression chamber spray header outboard containment isolation valve. Both tests initially demonstrated leakage rates that exceeded the valve's administrative leakage limit. Inspectors also reviewed three licensee procedures relevant to local leak rate testing of the valve: LLRT-01, "Primary Containment Leakage Rate Testing Program," TSP-CONT-R801, "Containment Isolation Valve and Penetration Leak Test Program," and TSP-RHR/X25B-C801, "LLRT of RHR-V-27B." Inspectors noted that LLRT-01 Section 3.4 and TSP-CONT-R801 Step 8.2.10.d both require a cause determination to be performed following a local leak rate test which results in leakage above the administrative limit. A cause determination was not performed for either of the two most recent LLRTs of RHR-V-27B. In both instances the licensee flushed the line and reperformed the LLRT which yielded leakage rates below the administrative limit. Additionally, procedure TSP-RHR/X25B-C801 Step 7.1.22 requires that if a LLRT leak rate is greater than the administrative limit, an evaluation of overall containment leakage is to be performed. This evaluation was not performed for either adverse test of RHR-V-27B.

Corrective Actions: The licensee documented the failure to perform a cause determination for leakage found above the administrative limit during local leak rate testing on RHR-V-27B in the corrective action program. The licensee intends to perform a cause evaluation to determine the reason for exceeding the administrative limit during the LLRT. Additionally, the licensee plans to evaluate the as found results for the LLRT and compare against the Technical Specification limit for overall containment leakage to determine past operability and reportability implications.

Corrective Action References: Action Requests 425456, 425605

Performance Assessment:

Performance Deficiency: The failure to follow site procedures when performing local leak rate testing for containment isolation valves is a performance deficiency. Specifically, the licensee failed to follow procedure LLRT-01 Section 3.4 and procedure TSP-CONT-R801 Step 8.2.10.d which state that a cause determination shall be performed if a valve exceeds the administrative limit during a local leak rate test. The administrative limit was exceeded during a LLRT performed on May 22, 2019 and again on June 4, 2021 for the RHR-V-27B Residual Heat Removal to suppression chamber spray header outboard containment isolation valve. A cause determination was not performed on either occasion.

Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Human Performance attribute of the Barrier Integrity cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, multiple procedure adherence deficiencies led to a substantive and meaningful reduction in overall containment leakage margin. The overall containment leakage Technical Specification limit is 121,536 sccm. The LLRT performed on May 22, 2019 documented a leakage rate of 40,000 sccm, and the June 4, 2021 the LLRT performed on June 4, 2021 found a maximum leakage of 70,000 sccm. This most recent LLRT represented a 57% reduction in margin to the total containment leakage Technical Specification limit.

Significance: The inspectors assessed the significance of the finding using Appendix A, The Significance Determination Process for Findings At-Power. Using Exhibit 3 - Barrier Integrity Screening Questions, the inspectors determined the finding was of very low safety significance (Green) since leakage past RHR-V-27B did not represent an actual open pathway in the physical integrity of reactor containment. The suppression pool spray line represents a closed system which takes suction from the suppression pool and discharges back to the suppression pool.

Cross-Cutting Aspect: P.2 - Evaluation: The organization thoroughly evaluates issues to ensure that resolutions address causes and extent of conditions commensurate with their safety significance. The most recent example of not performing a cause determination following an LLRT test with leakage above the administrative limit occurred on June 4, 2021. This falls within the nominal three-year period for present performance. The P.2 -

Evaluation cross-cutting aspect was chosen because the licensee failed to evaluate the condition of the failed LLRT. The licensee went straight from identification of excessive valve leakage to implementing a resolution by flushing the valve and reperforming the LLRT. The failure to evaluate the condition and jumping straight to a resolution is the proximate cause of not following the LLRT procedures.

Enforcement:

Violation: Title 10 of the Code of Federal Regulations, Part 50, Appendix B, Criterion V, "Procedures," requires, in part, that "Activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings."

Columbia Generating Station Procedure LLRT-01, Primary Containment Leakage Rate Testing Program, Revision 8, Section 3.4, Corrective Action, states, In the event of failure to meet specified acceptance criteria or to perform tests at intervals required under this plan, a CR shall be initiated. A cause determination should be performed, and corrective actions identified that focus on activities that eliminate the cause of the failure and prevent recurrence.

Columbia Generating Station Procedure TSP-CONT-R801, Containment Isolation Valve and Penetration Leak Test Program, Revision 17, Step 8.2.10.d states, If leakage is confirmed above the Administrative Leakage Limit, the Test Coordinator shall initiate a CR to evaluate valve leakage and to perform/document a cause determination.

Contrary to the above, on May 22, 2019 and on June 4, 2021 the licensee did not accomplish activities affecting quality in accordance with procedures of a type appropriate to the circumstances. Specifically, the licensee did not perform a cause determination after RHR-V-27B exceeded the administrative limit during local leak rate testing as required by procedure LLRT-01 Section 3.4, Revision 8 and procedure TSP-CONT-R801 Step 8.2.10.d, Revision

17.

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

EXIT MEETINGS AND DEBRIEFS

The inspectors verified no proprietary information was retained or documented in this report.

On September 30, 2021, the inspectors presented the design basis assurance inspection (programs) inspection results to Robert E. Scheutz and other members of the licensee staff.

DOCUMENTS REVIEWED

Inspection Type Designation Description or Title Revision or

Procedure Date

71111.21N.02 Calculations 216-92-003 Calculation for Weak Link Analysis for HPCS Valve 4 3

216-92-011 Weak Link Analysis for Valve No. LPCS-FCV-11, RHR- 1

FCV-64A,B,C (Fisher 3" Globe Valve)

C106-92-03.02 HPCS System MOV Design Basis Review 3

C106-92-03.03 Calculation for RHR Motor Operated Valve Design Basis 5

Review

C106-92-03.04 Service Water System MOV Design Basis Review 2

C106-92-03.05 WNP-2 RCIC System MOV Design Basis Review 3

C106-92-03.06 WNP-2 RCC System MOV Design Basis Review 1

CE-02-92-51 Analysis and Testing of Limitorque Torque Switches 0

Model SMB-0 thru 5

EI-02-91-04 Motor Thermal Overload (TOL) and Branch Circuit Fuse 0

Sizing Calculation

EI-02-93-04 Overcurrent Protection of Primary Containment Electrical 6

Penetrations

MA 21233 Equipment Qualification Report Duragear Model AVI-1 or 03/23/2005

Bettis NCB415-SR80 Operator

MA 21329 Operator Sizing Calculation for Enertech 3 Inch ANSI 04/15/2005

Class 300 Permaseat Valve

ME-02-00-13 EDR and DFR System Air Operated Valve Functional and 1

MEDP Calculation

ME-02-02-25 AC Gate Valves - MOV Thrust and Setpoint Calculation 6

ME-02-02-26 DC Gate and Globe Valves - MOV Thrust and Set-point 5

Calculation

ME-02-02-27 AC Rising Stem Valves - MOV Thrust and Set-point 5

Calculation

ME-02-95-34 Design Basis Thrust Calculation and Evaluation for 0

RHR-V-8 and RHR-V-9

ME-02-96-20 Temperature Effects of Valves Due to Nearby Heat 0

Sources

ME-02-96-21 MOV Pressure Locking Calculation 0

Corrective Action Action Request 279896, 309945, 366179, 366705, 410264, 419863,

Inspection Type Designation Description or Title Revision or

Procedure Date

Documents (AR) 393944, 421357, 424059, 289329, 366537, 367036,

367502, 367504, 367530, 367865, 378102, 380303,

393351, 393368, 394383, 395337, 395811, 419594,

21674, 422430, 422457, 422728, 424019, 420455,

419150, 419153, 419222, 419703, 419760, 419849,

419743, 415632, 415637, 416151, 289329, 366537,

367036, 367502, 367504, 367530, 367865, 378102,

380303, 393351, 393368, 394383, 395337, 395811,

415594, 421674, 422430, 422457, 422728, 424019,

2728, 408029, 415202, 419849, 419350, 420455,

23243, 355027, 384974, 367718, 366755, 418147,

311280, 385706, 404477, 414806, 415059, 422219,

360516, 360585, 363807, 372892, 391846, 394794

Corrective Action Action Request 425215, 425216, 425219, 425229, 425230, 425246,

Documents (AR) 425248, 425298, 425302, 425303, 425304, 425307,

Resulting from 425456, 425605

Inspection

Drawings 6E051 Reactor Core Isolation Cooling System Annunciators 16

Sheet 1

807E173TC Elementary Diagram RCIC [Reactor Core Isolation 41

Cooling] System

E528-36 MCC [Motor Control Center] Equip. [Equipment] Overload 32

Summary

EWD-6E-055 Electrical Wiring Diagram Reactor Core Isolation Cooling 19

System MOV RCIC-V-63 (E51-F063)

EWD-6E-055A Electrical Wiring Diagram Reactor Core Isolation Cooling 4

System MOV RCIC-V-63 (E51-F063)

EWD-9E-011 Electrical Wiring Diagram Residual Heat Removal System 18

MOV RHR-V-9 (E12-F009)

Engineering EC 12162 Design Evaluation - Table 9 Overload Selection

Changes EC 12163 Design Evaluation - Table 10 Overload Heater Selection

EC 15991 RHR-V-64B Failure to Open on Flow Reduction (MSPI 0

Functions Impact)

EC 17506 Main Steam Relief Valve Hydranuts 01

Inspection Type Designation Description or Title Revision or

Procedure Date

EC 18425 MSRV Nozzle Ring Set Screw OEM Re-Design 00

EC 18452 SW-M-V/12A Motor/Actuator Replacement 0

EC 18783 RHR-MO-9 Gear Change for RHR-V-9 000

EC 8804 CMR To Revise Motor Operator Weight On RHR-V-9 For 000

Calculation 8.14.107, Rev. 10

Miscellaneous AR 374139374139 Pre-NRC Power Operated Valve (POV) Inspection 03/08/2021

Focused Self-Assessment Report

EES-5 General Fuse Selection Criteria and the Electrical 012

Protection of 460 VAC and 125-250 VDC Motors

IST Program Inservice Testing Program Plan Fourth Ten-Year 5.001

Plan Inspection Interval

PSA-AOV-IR- Risk Ranking of Air Operated Valves 1

0001

PSA-MOV-IR- Motor Operated Valve Importance Ranking in Support of 4

0001 the MOV Periodic Verification Program

QID 361020 Seismic and Hydrodynamic Qualification of Anchor Darling 12/18/1985

Globe Valve

TM-2019 Summary of Equipment Qualification Environmental 14

Profiles

TM-2096 Design Valve Factor Criteria used in GL-89-10 Motor 1

Operated Valve Calculations

Procedures 10.24.235 Air Operated Valve Testing and Calibration 11

10.25.132 Thrust Adjustment and Diagnostic Analysis of Motor 34

Operated Valves

10.25.4 Lubrication and Inspection of Limitorque MOVs 29

10.25.74 Testing Motor Operated Valve Motors and Controls 33

8.4.73 MOV Design Basis Testing Evaluation 11/03/2004

LLRT-01 Primary Containment Leakage Rate Testing Program 8

MES-10 Motor Operated Valve Sizing and Switch Setting 10

MOVPP-01 MOV Periodic Verification Program Plan 9

OSP-RCIC/IST- RCIC Operability Test 66

Q701

TSP-CONT-R801 Containment Isolation Valve and Penetration Leak Test 17

Program

Inspection Type Designation Description or Title Revision or

Procedure Date

TSP-RHR/X25B- LLRT of RHR-V-27B 3

C801

Work Orders Work Order (WO) 01112272, 01195276, 02008781, 02042108, 02082275,

2107548, 02160241, 02144982, 02151067, 02160242,

2040351, 02150081, 0216869301, 0216869401,

216869501, 01021307, 01194404, 02048827,

216491001, 01143406, 01177117, 0204280601,

215106801, 0216316801, 0216316901, 01081211,

01188057, 02115790, 0211257501, 021125701,

214452001, 0214809601, 01142628, 02045314,

2082280, 0211082701, 01171199, 02079580, 02145029,

211201501, 01171199, 02048825, 0215464501,

2134383, 02153283, 2145451, 02142995, 02148076

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