IR 05000254/2018003

NRC Integrated Inspection Report 05000254/2018003, 05000265/2018003, 07200053/2018001 and 07200053/2018002

ML18297A071
Person / Time
Site:	Quad Cities
Issue date:	10/24/2018
From:	Karla Stoedter NRC/RGN-III
To:	Bryan Hanson Exelon Generation Co, Exelon Nuclear
References
IR 2018001, IR 2018002, IR 2018003
Download: ML18297A071 (22)

Text

October 24, 2018

SUBJECT:

QUAD CITIES NUCLEAR POWER STATION, UNITS 1 AND 2NRC INTEGRATED INSPECTION REPORT 05000254/2018003; 05000265/2018003; 07200053/2018001 AND 07200053/2018002

Dear Mr. Hanson:

On September 30, 2018, the U.S. Nuclear Regulatory Commission (NRC) completed an integrated inspection at your Quad Cities Nuclear Power Station, Units 1 and 2. On October 2, 2018, the NRC inspectors discussed the results of this inspection with Mr. K. Ohr, Site Vice President, and other members of your staff. The results of this inspection are documented in the enclosed report.

Based on the results of this inspection, the NRC has identified one issue that was evaluated under the risk significance determination process as having very low safety significance (Green). Additionally, the NRC has identified one issue that was evaluated using the NRC Enforcement Policy and Enforcement Manual (traditional enforcement) and was determined to be of very low safety significance (Severity Level IV). The NRC has also determined that two violations are associated with these issues. Because the licensee initiated condition reports to address these issues, these violations are being treated as a Non-Cited Violations (NCVs),

consistent with Section 2.3.2 of the Enforcement Policy. The NCVs are described in the subject inspection report.

If you contest the violations or significance of these NCVs, 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 the Quad Cities Nuclear Power 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 10 CFR 2.390, Public Inspections, Exemptions, Requests for Withholding.

Sincerely,

/RA/

Karla Stoedter, Chief Branch 1 Division of Reactor Projects

Docket Nos. 50-254; 50-265;72-053 License Nos. DPR-29; DPR-30

Enclosure:

IR 05000254/2018003; 05000265/2018003; 07200053/2018001; 07200053/2018002 cc: Distribution via LISTSERV

SUMMARY

The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees performance by conducting an integrated quarterly inspection at Quad Cities Nuclear Generating Station, Units 1 and 2, 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. Findings and violations being considered in the NRCs assessment are summarized in the table below.

List of Findings and Violations

Failure to Maintain the Design Basis for Residual Heat Removal Torus Suction Valve Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000254/2018003-01 Closed N/A 71111.15 -

Operability Determinations and Functionality Assessments The inspectors identified a Green finding and associated Non-Cited Violation of Title 10 of the Code of Federal Regulations (10 CFR) 50, Appendix B, Criterion III, Design Control, when the licensee performed an in-field adjustment to the torque switch settings on the Unit 1 residual heat removal (RHR) torus suction valve 1-1001-7C and failed to ensure measures were established to assure the valve could continue to meet its design basis requirements.

Failure to Follow Procedures for Forced Helium Dehydration of a Multipurpose Canister Cornerstone Significance Cross-Cutting Aspect Report Section N/A Severity Level IV (Traditional Enforcement)

NCV 05000254/2018003-02; 05000265/2018003-02; 07200053/2018002-01 Closed N/A 60855.1 - Operation of an Independent Spent Fuel Storage Installation at Operating Plants The inspectors identified a Severity Level IV Non-Cited Violation of 10 CFR 72.150 when the licensee failed to follow procedures for the setup of the multipurpose canister (MPC) forced helium dehydration (FHD) system. Specifically, during the setup for processing MPCs during the 2018 Independent Spent Fuel Storage Installation (ISFSI) loading campaign, the licensee failed to follow procedure OU-MW-671-200, MPC Processing FHD for BWRs, Revision 1,

Attachment 9, Step 1.2.1, which connected inlet and outlet hosing between the FHD skid and FHD manifold.

Additional Tracking Items

None

TABLE OF CONTENTS

PLANT STATUS

INSPECTION SCOPES

................................................................................................................

REACTOR SAFETY

.................................................................................................................. 4 External Flooding

OTHER ACTIVITIES - BASELINE

............................................................................................

OTHER ACTIVITIES

TEMPORARY INSTRUCTIONS, INFREQUENT AND ABNORMAL

INSPECTION RESULTS

..............................................................................................................

EXIT MEETINGS AND DEBRIEFS

............................................................................................ 15

DOCUMENTS REVIEWED

......................................................................................................... 15

PLANT STATUS

Unit 1

The unit was operated at or near full rated thermal power from July 1 to September 26, 2018,

with the exception of planned power reductions for turbine testing, control rod pattern

adjustments, and other short-term power changes as requested by the transmission system

operator. On September 26, 2018, the unit experienced a turbine load rejection and an

automatic SCRAM from 100 percent power during switchyard restoration activities associated

with off-site, 345 kilovolt (kV), Line 0401. The unit was placed into a hot shutdown condition

and entered Forced Outage Q1F66. On September 28, 2018, the unit entered Mode 2

Startup, and reached Mode 1, Power Operation, the same day. On September 29, 2018,

operators synchronized the units main turbine generator to the grid, and operators continued

power ascension to full rated thermal power. On September 30, 2018, the unit reached full

power where it operated at or near for the remainder of the inspection period.

Unit 2

The unit was operated at or near full rated thermal power for the entire inspection period, with

the exception of planned power reductions for turbine testing, control rod pattern adjustments,

and other short-term power changes as requested by the transmission system operator.

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

ProgramOperations Phase. The inspectors performed plant status activities described in

IMC 2515 Appendix D, Plant Status, and conducted routine reviews using IP 71152, Problem

Identification and Resolution. 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

71111.01Adverse Weather Protection

External Flooding (1 Sample)

The inspectors evaluated readiness to cope with external flooding on August 8, 2018.

71111.04Equipment Alignment

Partial Walkdown (3 Samples)

The inspectors evaluated system configurations during partial walkdowns of the following

systems/trains:

(1) Unit 1 High Pressure Coolant Injection (HPCI) system during Unit 1 Reactor Core

Isolation Cooling (RCIC) system planned maintenance and testing, on July 11, 2018;

(2) Unit 1B subsystem during Unit 1A Residual Heat Removal System (RHR)/ RHRSW

(residual heat removal service water) subsystems planned maintenance, on

July 23, 2018; and

(3) Unit 1A RHRSW system during Unit 1C RHRSW system planned maintenance, on

August 14, 2018.

71111.05AQFire Protection Annual/Quarterly

Quarterly Inspection (4 Samples)

The inspectors evaluated fire protection program implementation in the following selected

areas:

(1) Fire Zone SBO-4, Unit 2 Station Blackout (SBO) Diesel Generator Room,

Elevation 595-0, SBO Building, on July 25, 2018;

(2) Fire Zone 8.2.7.E, Unit 2 Turbine Building, Elevation 615-6, North Mezzanine Floor, on

August 10, 2018;

(3) Fire Zone 9.3, Unit 1/2 Reactor Building, Elevation 595-0, 1/2 Diesel Generator, on

September 28, 2018; and

(4) Fire Zone 1.1.1.2, Unit 1 Reactor Building, Elevation 595-0, Ground Floor,

September 28, 2018.

71111.06Flood Protection Measures

Internal Flooding (1 Sample)

The inspectors evaluated internal flooding mitigation protections in the U2 RHRSW vaults on

August 2, 2018.

71111.07Heat Sink Performance

Heat Sink (1 Sample)

The inspectors evaluated the Unit 2 emergency diesel generator (EDG) jacket cooling water

heat exchanger performance during the weeks of July 18, 2018, and August 31, 2018.

71111.11Licensed Operator Requalification Program and Licensed Operator Performance

Operator Requalification (1 Sample)

The inspectors observed and evaluated operators in the control room simulator during

requalification training on September 17, 2018.

Operator Performance (1 Sample)

The inspectors observed and evaluated operators on Unit 1 respond to a temperature control

valve failure in the reactor building closed cooling water (RBCCW) system and increasing drywell

temperatures, on August 24, 2018. The inspectors observed and evaluated operators on Unit 2

perform testing and troubleshooting of the C RHR pump following planned maintenance on

September 20, 2018.

71111.12Maintenance Effectiveness

Routine Maintenance Effectiveness (1 Sample)

The inspectors evaluated the effectiveness of routine maintenance activities associated

with the following equipment and/or safety significant function:

(1) Turbine control valves.

71111.13Maintenance Risk Assessments and Emergent Work Control (3 Samples)

The inspectors evaluated the risk assessments for the following planned and emergent

work activities:

(1) Emergent work associated with the degraded Delta T and high temperatures on the

Unit 2 EDG that resulted in online risk change to yellow during the week of

July 16, 2018;

(2) Site safety stand down and work stoppage due to increased number of

OSHA-recordable injuries, emergent work associated with Unit 1 reactor water cleanup

(RWCU) isolation, and subsequent reactor water chemistry excursion during the week of

July 31, 2018; and

(3) Severe thunderstorms impacting work activities and both units online risk change to

yellow during the weeks of August 19, 2018, and September 4, 2018.

71111.15Operability Determinations and Functionality Assessments (6 Samples)

The inspectors evaluated the following operability determinations and functionality

assessments:

(1) Issue Report (IR) 4148330: U2 EDG Trip During QCOS 6600-42, 06/19/2018;

(2) IR 4162608: 2B CS Room Floor Drain Valve Failed QCOS 0020.04, 08/08/2018;

(3) IR 4166453: Drywell Equipment Drain Sump Pump Flow is Degraded, 08/23/2018;

(4) IR 1416606: Galling Potential Identified on 1-1001-7C/D (Unit 1 RHR Suction Valves),

09/21/2012;

(5) IR 4170073: Unit 1 HPCI Flow Controller Erratic, 09/05/2018; and

(6) Engineering Change (EC) 624284: Control Room Envelope Heat Load Analysis for Loss

of One TXV [thermal expansion valve], 07/12/2018.

71111.19Post Maintenance Testing (4 Samples)

The inspectors evaluated the following post maintenance tests:

(1) Unit 2 Emergency diesel generator load test following emergent maintenance associated

with degraded heat exchanger performance on July 19, 2018;

(2) Unit 1/2 B Diesel Driven Fire Pump Annual Capacity Test following annual planned

maintenance and water pump replacement on August 1, 2018;

(3) Unit 1/2 A Diesel Driven Fire Pump Annual Capacity Test following annual planned

maintenance and replacement of fire pump discharge relief valve on August 16, 2018;

and

(4) Unit 2C RHR pump operability test following planned system maintenance on

September 20, 2018.

71111.22Surveillance Testing

The inspectors evaluated the following surveillance tests:

Routine (2 Samples)

(1) QCOS 6600-43: Unit 1/2 Emergency Diesel Generator Load Test, on July 23, 2018; and

(2) QCOS 6620-01: [Unit 2] SBO Diesel Generator (DG) 1(2) Quarterly Load Test, on

August 2, 2018.

In-service (2 Samples)

(1) QCOS 1100-07: Unit 1A Standby Liquid Control (SBLC) Pump Flow Rate Test, on

July 23, 2018; and

(2) QCOS 1300-05: [Unit 2] RCIC Pump Operability Test, on August 21, 2018.

OTHER ACTIVITIES - BASELINE

71151Performance Indicator Verification (4 Samples)

The inspectors verified licensee performance indicators submittals listed below:

(1) MS05: Safety System Functional Failures (SSFFs) Sample-2 Samples (July 1, 2017-

June 30, 2018); and

(2) MS06: Emergency AC Power Systems-2 Samples (July 1, 2017-June 30, 2018).

71152Problem Identification and Resolution

Annual Follow-Up of Selected Issues (1 Sample)

The inspectors reviewed the licensees implementation of its corrective action program

related to the following issue:

(1) IR 4148330: Unit 2 EDG Tripped on High Temperature and IR 4156181: Unit 2 EDG

Elevated Temps During Monthly Load Test.

71153Follow-Up of Events and Notices of Enforcement Discretion

Events (2 Samples)

(1) On July 31, 2018, the inspectors evaluated the licensees response to an isolation of the

Unit 1 RWCU system (Group 3 isolation) coincident with the receipt of a half Group I

primary containment system isolation signal. The licensee identified that a main steam

tunnel temperature switch failure caused the signals to occur. All equipment responded

as expected following the temperature switch failure. Additionally, the inspectors

evaluated the licensees response to a subsequent Unit 1 reactor water chemistry

excursion that occurred on that same day, while the licensee was returning the RWCU

system to service.

(2) On September 26, 2018, Unit 1 experienced a turbine load rejection and automatic reactor SCRAM from full power. The inspectors evaluated the licensees response to

the SCRAM and follow-up actions, including Forced Outage Q1F66 activities that

occurred from September 26 to September 28, 2018. This sample included control room

observations of shutdown and startup activities.

OTHER ACTIVITIESTEMPORARY INSTRUCTIONS, INFREQUENT AND ABNORMAL

60853On-Site Fabrication of Components and Construction of an Independent Spent Fuel

Storage Installation

From November 20, 2017, to August 8, 2018, the inspector evaluated the licensees

fabrication of components and its construction of a second independent spent fuel storage

installation pad. This new ISFSI pad is designed to hold 114 HI-STORM 100 casks, which

will be built to Certificate of Compliance 72-1014, Amendment 8, Revision 1. Results of

earlier inspections related to this ISFSI pad were already documented in NRC Integrated

Inspection Report 05000254/2017004, 05000265/2017004, and 07200053/2017001.

Specifically, the inspector evaluated:

(1) the modulus of elasticity of the subgrade material of the ISFSI pad by reviewing plate

load test reports;

(2) the procurement of ISFSI pad components in accordance with specifications by

reviewing rebar certifications and the ISFSI pad concrete mix design;

(3) the installation of rebar and the preparation for pouring concrete by performing a

walkdown of the installed rebar for Sections B and H of the ISFSI pad on

June 12, 2018;

(4) the delivery, placement, testing, finishing, and curing of concrete by observing the

licensees concrete pour activities for Sections B and H on June 13, 2018;

(5) the compressive strength of the ISFSI pad by reviewing compressive strength test

reports for concrete cylinders taken from concrete placed into sections A, C, G, and J of

the ISFSI pad; and

(6) the 72.48 screening performed for the installation of the second ISFSI pad.

60855.1Operation of an Independent Spent Fuel Storage Installation at Operating Plants

The inspector evaluated the licensees operation of the independent spent fuel storage

installation from August 13 to September 25, 2018. Specifically, the inspector evaluated:

(1) the material and radiological condition of the ISFSI and loaded storage casks through

independently observing the structural condition, and both performing independent

radiological surveys and reviewing the licensees radiological surveys;

(2) a heavy lift of the loaded transfer cask from the spent fuel pool to the cask processing

area;

(3) welding of the root and intermediate passes of an MPC;

(4) non-destructive evaluations of the lid to shell weld of an MPC;

(5) processing of an MPC for storage, including drying using FHD;

(6) the fuel selection and loading for the MPC loaded on August 27, 2018; and

(7) changes to the 72.212 report and changes screened and evaluated using

CFR 72.48 and 50.59.

INSPECTION RESULTS

71111.15Operability Determinations and Functionality Assessments

Failure to Maintain the Design Basis for Residual Heat Removal Torus Suction Valve

Cornerstone

Significance

Cross-Cutting

Aspect

Report Section

Mitigating Systems

Green

NCV 05000254/2018003-01

Closed

Not Applicable

IP 71111.15 -

Operability

Determinations and

Functionality

Assessments

The inspectors identified a Green finding and associated Non-Cited Violation (NCV) of Title

of the Code of Federal Regulations (10 CFR) 50, Appendix B, Criterion III, Design

Control, when the licensee performed an in-field adjustment to the torque switch settings on

RHR torus suction valve 1-1001-7C and failed to ensure measures were established to

assure the valve could continue to meet its design basis requirements.

Description:

In 2012, the licensee identified a potential galling concern for Unit 1 RHR torus suction

valves 1-1001-7C and 1-1001-7

D. Specifically, the licensee documented the Joint Owners

Group Motor Operated Valve Program indicated that valves with stainless steel discs in

sliding contact with stainless steel seats are susceptible to galling when exposed to

differential pressure forces and temperatures greater than 120 degrees Fahrenheit.

Valves 7C and 7D have similar stainless steel discs and seats (these are the only valves with

this susceptibility). Testing was performed by the Electric Power Research Institute (EPRI)

and published in EPRI Report 1000885, Friction and Galling Performance of SA-351-CF8

Stainless Steel. The report discussed a galling concern for the loaded faces of the valve that

could potentially prevent valve movement, depending on temperature and differential

pressure conditions.

The licensee documented the concern in IR 1416606, Galling Potential Identified

on 1-1001-7C/D, on September 21, 2012. The operability evaluation documented in

IR 1416606 determined the valves would maintain their ability to function up to 60 psig reactor

steam dome pressure, when in Mode 3. Under normal (Mode 1) operating conditions, the 7C

and 7D valves are open and not subject to differential pressure forces of concern (torus

pressure on one side and pump suction pressures on other). When the licensee transitions

into Mode 3 and shutdown cooling is required to be placed into service, the 7C and 7D valves

would need to be closed. Once they are closed and shutdown cooling (SDC) is placed into

operation, the valves are subject to reactor pressures on one side and torus pressures on the

other side of the valves. Depending on the operating conditions during Mode 3 (i.e. steam

dome pressure), the valve could be subject to galling that may impact the valves ability to

open.

Technical Specification Limiting Condition for Operation (LCO) 3.4.7, Residual Heat Removal

(RHR) Shutdown Cooling SystemHot Shutdown, requires SDC to be operable in Mode 3

with steam dome pressure below the RHR cut-in permissive pressure limit (93 psig

decreasing, in accordance with QCOP 1000-05, Shutdown Cooling Operation).

Technical Specification LCO 3.5.1, Emergency Core Cooling SystemsOperating, requires

that the low pressure coolant injection (LPCI) mode of RHR is Operable in Modes 1, 2, and 3.

The issue of concern identified in IR 1416606 is that should the RHR 7C and 7D valves be

operated (normally open and would need to be closed for SDC operation) above 60 psig

steam dome pressure, there was a concern the differential pressure across the valves would

be high enough that galling of the valve seat and disc could cause the valves to become

inoperable and render the LPCI mode of RHR inoperable.

As a corrective action from IR 1416606, the licensee made a procedure change to

QCOP 1000-05, which limited operators from placing the B loop of SDC into operation until

reactor pressure was less than or equal to 60 psig (7C and 7D valves are in the B loop). The

licensee had also planned corrective actions to replace the valves during the next Refueling

Outage (RFO) Q1R23, in March 2015. For circumstances outside of the licensees control,

the 7C and 7D valves were unable to be replaced in RFO Q1R23, and the project to replace

the valves was rescheduled for RFO Q1R25, in March 2019.

The inspectors identified the galling issue for the 7C and 7D valves during a Plant Health

Committee presentation in August 2018. After reviewing IR 1416606 and supporting

calculations and documentation, the inspectors asked clarifying questions to the licensee. In

researching the inspectors questions, the licensee identified that in RFO Q1R23 in

March 2015, the licensee made in-field torque switch adjustments to the 7C valve, which

changed the friction factor of the valve in a non-conservative direction. The licensee

calculated the new maximum steam dome pressure the valve would be able to reliably

perform its function was 45 psig reactor steam dome pressure (versus the 60 psig previously

calculated in 2012). The licensee captured the issue in IR 4171765, MO 1-1001-7C May

Not Open During Mode 3 LOCA [loss-of-coolant accident], dated September 11, 2018.

The sizing calculation (Midacalc) of record for these valves, QUA-1-1001-7C and

QUA-1-1001-7D, Revision 6, used a maximum required opening differential pressure of

psid, as calculated in QDC-1000-M-1318, Residual Heat Removal System Combined

DBD and DP Calculation, Revision 000. Calculation QDC-1000-M-1318 uses an incorrect

assumption for the worse case opening condition for the valve because it failed to account for

the Mode 3 loss of coolant accident with SDC in service. Following the identification of the

issue in IR 1416606, the licensee failed to update the differential pressure calculation to

reflect the most restrictive design basis conditions for the valve. Additionally during the 2015

RFO Q1R23, the licensee failed to recognize that the torque switch adjustments made to the

7C valve would have an impact on the valves operation.

Corrective Action: Licensee made a change to procedure QCOP 1000-05 to reflect the more

conservative reactor steam dome pressure limit of 45 psig prior to placing the B loop of SDC

into service. The licensee has a planned work order to replace the Unit 7C and 7D valves in

RFO Q1R25 in 2019.

Corrective Action Reference: IR 4171765: MO 1-1001-7C May Not Open During Mode 3

LOCA

Performance Assessment:

Performance Deficiency: The licensee performed an in-field adjustment to the torque switch

settings on RHR torus suction valve 1-1001-7C and failed to ensure measures were

established to assure the valve could continue to meet its design basis requirements.

Screening: The inspectors determined the performance deficiency was more than minor

because it adversely affected the Design Control attribute of the Mitigating Systems

Cornerstone objective to ensure the availability, reliability, and capability of systems that

respond to initiating events to prevent undesirable consequences (i.e., core damage).

Specifically, the failure to assure the design basis was maintained had the potential to lead to,

or could have resulted in, the Unit 1B loop LPCI mode of RHR not being available to perform

its function under certain design basis operating conditions.

Significance: The inspectors assessed the significance of the finding using SDP Appendix G,

Shutdown Operations. The inspectors determined the finding was of very low safety

significance (Green) because the LPCI Mode of RHR never lost its ability to function and the

inspectors answered yes to question A.1 in Exhibit 3, Mitigating System Screening

Questions.

Cross-cutting Aspect: The inspectors did not identify a cross-cutting aspect related to this

issue because it was greater than 3 years old and did not reflect current licensee

performance.

Enforcement:

Violation: Title 10 CFR Part 50, Appendix B, Criterion III, Design Control, requires, in part,

that measures be established to assure that applicable regulatory requirements and the

design basis are correctly translated into specifications, drawings, procedures, and

instructions. It further requires that design changes, including field changes, be subject to

design control measures commensurate with those applied to the original design.

Contrary to the above, during RFO Q1R23 in March 2015, the licensee performed a field

change and failed to subject it to design control measures commensurate with the original

design. Specifically, the licensee performed an in-field adjustment to the torque switch

settings on RHR torus suction valve 1-1001-7C and failed to ensure measures were

established to assure the valve could continue to meet its design basis requirements, which

are, to be operated at any time during Mode 3 and continue to be able to perform its LPCI

function to open.

Disposition: This violation is being treated as a NCV, consistent with Section 2.3.2 of the

Enforcement Policy.

60855.1Operation of an Independent Spent Fuel Storage Installation at Operating Plants

Failure to Follow Procedures for Forced Helium Dehydration of a Multipurpose Canister

Cornerstone

Severity

Cross-Cutting

Aspect

Report Section

Not Applicable

Severity Level IV (Traditional

Enforcement)

NCV 05000254/2018003-02;

05000265/2018003-02;

200053/2018002-01

Closed

Not Applicable

60855.1 -

Operation of an

Independent Spent

Fuel Storage

Installation at

Operating Plants

The inspectors identified a Severity Level IV NCV of 10 CFR 72.150 when the licensee failed

to follow procedures for the setup of the MPC FHD system. Specifically, during the setup for

processing MPCs during the 2018 ISFSI loading campaign, the licensee failed to follow

procedure OU-MW-671-200, MPC Processing FHD for BWRs, Revision 1, Attachment 9,

Step 1.2.1, which connected inlet and outlet hosing between the FHD skid and FHD manifold.

Description:

On August 30, 2018, the inspector conducted an equipment alignment (configuration control)

walkdown of the FHD system operating on the sixth MPC of the licensees 2018 ISFSI loading

campaign.

The FHD system is used to remove the remaining moisture in an MPC after the bulk water

has been drained. The system functions by circulating heated dry helium through the MPC.

The helium evaporates the water remaining in the MPC, and the moisture-laden helium is

then routed back to the FHD system skid where a demoisturizer removes the entrained water.

Over the previous 2 days, the FHD system had been operating at a reduced flow rate through

the MPC because the system motor kept tripping on thermal overload when operators

attempted to run the system at the normal flow rate. The reduced flow rate precluded the

licensee from demonstrating the sixth MPC had been dried in accordance with the applicable

Technical Specification (TS) dryness limits of Surveillance Requirement (SR) 3.1.1.1.

During the configuration control review, the inspector identified the inlet and outlet hosing

from the FHD skid to the FHD manifold had been switched, resulting in reversed flow through

the MPC. The inspector subsequently identified this line-up was not in accordance with

licensee procedure OU-MW-671-200, MPC Processing FHD for BWRs, Revision 1,

9, Step 1.2.1. Since the hoses did not need to be disconnected during the drying

of the first 5 MPCs, the swapped hoses had been in place since the beginning of the

campaign on June 4, 2018.

As part of its investigation following the inspectors identification of the swapped hoses, the

licensee determined that during the initial processing of the first MPC the temperature

indications on the inlet and outlet temperature gauges had given unexpected readings and

so operators switched the labels on the gauges. A corrective action document, however,

had not been generated for the label issue, an action not in accordance with licensee

procedure PI-AA-120, Issue Identification and Screening Process, Revision 8. In addition,

after operators switched the labels, an independent verification that the labels had been

placed on the correct component using verification practices in accordance with licensee

procedure OP-AA-116-101, Equipment Labeling, Revision 2, was not performed.

Corrective Actions: The licensee entered LCO 3.1.1, Condition A, as it determined there was

not reasonable assurance that SR 3.1.1.1 had been met without further evaluation for the five

previously loaded MPCs. The licensee subsequently performed EC 625388, DCS FHD

Configuration Discrepancy, which concluded that SRs 3.1.1.1, 3.1.1.2, and 3.1.1.3 were

indeed met even with reversed flow. The licensee identified the configuration of the internal

drain tube connected to the drain port on the first five MPCs was different than the

configuration of the tube in the sixth MP

C. The licensee postulated that this difference, in

conjunction with thermal expansion of the tube due to the reversed flow of heated helium,

resulted in the need for the reduced flow in the sixth MPC.

The licensee shut down the FHD system, and placed the MPC in a known safe configuration.

The licensee then met with staff and reinforced the need for procedure use and adherence.

The procedure was then revised to allow switching back the hoses, and the revised procedure

was successfully tested on a mock-up MPC that did not contain nuclear fuel. The licensee

subsequently switched the hoses on the loaded sixth MPC and completed the drying without

incident.

Corrective Action References: IRs 04168081 and 04168791

Performance Assessment:

In accordance with Section 2.2 of the Enforcement Policy and Inspection Manual

Chapter 0612, Appendix B, Issue Screening, ISFSIs are not subject to the Significance

Determination Process and are not subject to the Reactor Oversight Process, so violations

identified at ISFSIs are assessed using traditional enforcement. Traditional enforcement

violations are not assessed for cross-cutting aspects.

Enforcement:

Violation: Title 10 CFR 72.150 requires the licensee to prescribe activities affecting quality by

documented instructions, procedures, or drawings of a type appropriate to the circumstances

and requires that these instructions, procedures, and drawings be followed.

Contrary to the above, from June 4 to August 30, 2018, the licensee failed to follow

instructions, procedures, or drawings of a type appropriate to the circumstances. Specifically,

the licensee failed to follow procedure OU-MW-671-200, MPC Processing Forced Helium

Dehydration for BWRs, Revision 1, Attachment 9, Step 1.2.1. Specifically, the licensee did

not connect the inlet and outlet hoses for the forced helium dehydration system as specified in

the procedure.

Severity: The inspectors assessed the significance of the violation using the NRC

Enforcement Policy and Enforcement Manual. Consistent with the guidance in

Section 1.2.6.D of the Enforcement Manual, if a violation does not fit an example in the

Enforcement Policy Violation Examples, it should be assigned a severity level: (1)

commensurate with its safety significance; and (2) informed by similar violations addressed in

the Violation Examples. The inspectors found no similar violations in the violation examples.

Subsequent determinations identified that no design basis limits were exceeded due to this

issue, therefore the violation was determined to be of very low safety significance (Severity

Level IV).

Enforcement Action: This violation is being treated as a NCV, consistent with Section 2.3.2 of

the Enforcement Policy.

71152Problem Identification and Resolution

Observation

71152Annual Sample Review

IR 4148330: Unit 2 EDG Tripped on High Temperature and IR 4156181: Unit 2 EDG

Elevated Temps During Monthly Load Test

On June 19, 2018, the Unit 2 EDG tripped on high jacket water cooling temperature during its

monthly surveillance test. This high temperature trip is bypassed when there is an emergency

start signal present. The licensees troubleshooting efforts did not identify any conclusive

evidence for the cause of the jacket water cooling issue. Based on a support and refute

matrix, the licensee decided to replace the engine-driven jacket water pump and performed

inspections and cleaning of the jacket water heat exchangers.

On July 16, 2018, during the next scheduled surveillance test, the licensee identified concerns

with jacket water temperatures using thermography the licensee had been performing based

on the issues identified in June. The licensee noted the engine was able to pass all

surveillance requirements during the July 16, 2018, test; however, the licensee decided to

take the diesel out of service and perform additional inspections. During a borescope

examination of the heat exchanger, the licensee identified evidence of micro-fouling that could

have been impacting the ability of the heat exchanger to transfer heat. The licensee

performed a more extensive cleaning of the heat exchanger and subsequent testing showed a

significant improvement in heat exchanger performance. The Unit 2 EDG was declared

operable and returned to service on July 19, 2018.

Based on the two events described, the licensee performed an equipment corrective action

program evaluation (ECAPE) under IR 4148330. In this evaluation, the licensee determined

the most likely cause of the heat exchanger fouling was due to dredging that had occurred in

April 2018, around the same time the Unit 2 EDG 24-hour endurance run surveillance test

took place. The higher amounts of suspended sediment present at the intake, due to the

dredging, made its way into the heat exchanger. The inspectors noted that heat exchanger

inspection methods could vary (i.e. visual vs. borescope) and the acceptance criteria for the

various types of fouling could be subjective due to differences between heat exchanger

inspectors. Additionally, the licensee could perform a variety of cleaning methods depending

on the inspection results. With the type of fouling identified in July, the licensee had to use

more abrasive techniques (than were employed in June) in order to clean the heat exchanger

effectively. The inspectors determined that this event was an example of a weakness in the

licensees ability to identify a condition adverse to quality (heat exchanger degradation).

The inspectors noted several additional planned corrective actions from the ECAPE,

including: revising the guidance for heat exchanger inspections to better identify micro-

fouling; update and provide guidance for proper cleaning of EDG heat exchangers in order to

maximize efficiency and maintain maximum design margin; and planning work orders to

inspect the other EDGs as part of their extent of condition evaluation. The licensee did not

expect the other EDGs (Units 1 and 1/2) to be impacted based on temperature trends from

completed surveillances.

EXIT MEETINGS AND DEBRIEFS

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

disclosure. No proprietary information was documented in this report.

On August 8, 2018, the inspector presented the ISFSI pad inspection results to Mr.

K. Ohr,

Site Vice President, and other members of the licensee staff via telephone conference.

On October 2, 2018, the inspectors presented the quarterly integrated inspection and ISFSI

inspection results to Mr.

K. Ohr, Site Vice President, and other members of the licensee

staff.

DOCUMENTS REVIEWED

71111.04Equipment Alignment

- QOM 1-1000-04; Unit 1 RHR Valve Checklist; Revision 11

- Drawing M-37; Diagram of RHR Service Water Piping; Revision BI

- IR 4166051; NRC ID: Debris in 1A RHRSW HP INBD Bearing Housing; 08/21/2018

- QOM 1-1000-05; U1 RHR Service Water Valve Checklist; Revision 23

71111.05AQFire Protection Annual/Quarterly

- Quad Cities Generating Station Pre-fire Plan; April 2014

- Quad Cities Generating Station Pre-fire Plan; February 2017

71111.07Heat Sink Performance

- EC 625010; U2 EDG Projected Engine Outlet Water Temperature at Maximum Load;

Revision 0

- Drawing M-69, Sheet 3; Diagram of Service Water Piping Diesel Generator Cooling Water;

Revision R

- Drawing M-22, Sheet 3; Diagram of Service Water Piping Diesel Generator Cooling Water;

Revision AA

- Drawing M-22, Sheet 5, Diagram of Service Water Piping; Revision Z

- IR 4156181; U2 EDG Elevated Temps During Monthly Load Test; 07/16/2018

- ECAPE 4148330; U2 EDG Trip During QCOS 6600-42; 08/22/2018

- ER-AA-340; GL 89-13 Program Implementing Procedure; Revision 9

- ER-AA-340-1002; Service Water Heat Exchanger Inspection Guide; Revision 7

- ER-AA-340-1001; GL 89-13 Program Implementation Instructional Guide; Revision 11

- ER-AA-340-2000; Balance-of-Plant Heat Exchanger Inspection, Testing and Maintenance

Guide; Revision 9

- Calc No. QDC-6700-E-1503; Calculation Page; Revision 011

- QCMMS 6600-07; Emergency Diesel Generator Heat Exchanger Service Water Side Clean

and Inspection; Revision 6

- IR 4148330; U2 EDG Trip During QCOS 6600-42; 06/19/2018

- WO 1916398; Open/Inspect Diesel Generator HX; 03/22/2018

- WO 1916397; Open/Inspect Diesel Generator HX; 03/22/2018

- QCOS 6600-57; Unit 2 Diesel Generator Endurance and Margin/Full Load Reject/Hot Restart

Test; Revision 8

- QCOS 5750-09; ECCS Room and DGCWP Cubicle Cooler Monthly Surveillance; Revision 37

71111.12Maintenance Effectiveness

- ER-AA-310-1003; Maintenance Rule - Performance Criteria Selection; Revision 5

71111.13Maintenance Risk Assessments and Emergent Work Control

- Work Week Safety Profiles

71111.15Operability Determinations and Functionality Assessments

- EC 624284; Control Room Envelope Heat Load Analysis for Loss of One TXV; 07/12/2018

- IR 4062552; B Train CREV Superheat Value High; 10/13/2017

- WO 1918839-01; MM Adjust Temp Exp Valve 0-5741-341A on B CR HVAC Syst;

10/13/2017

- Drawing 931126AY; 14-150# Gate Valve, BW, Motor Operated; Revision 1

- EC 04-01284M; Replacement of MO1-1001-7C Valve; 08/11/1995

- IR 1416606; Galling Potential Identified on 1-1001-7C/D; 09/21/2012

- IR 2496550; MO 1-1001-7C and 1-1001-7D Not Replaced During Q1R23; 05/06/2015

- Letter 0124827; Procurement of Replacement Valves for Residual Heat Removal (RHR)

System MO-1-1001-7A, B, C, D Valves; 02/21/1994

- QUA-1-1001-7C; MIDACALC Results, QUA-1-1001-7C (QUA-1), AC Motor Operated

GL96-05 Gate Valve; Revision 6

- QUA-1-1001-7D; MIDACALC Results, QUA-1-1001-7D (QUA-1), AC Motor Operated

GL96-05 Gate Valve; Revision 6

- WR Q60582; 1D RHR Pump Suction Valve Replacement; 09/06/1994

- IR 4170073; U1 HPCI FIC 1-2340-1 Erratic When in Standby; 09/05/2018

- CC-AA-201; Plant Barrier Control Program; Revision 12

- Drawing M-85; Diagram of Reactor Building Equipment Drains; Revision CN

- QCOP 0020-04; Use of Plugs in Reactor Building Floor Drain Lines; Revision 1

- QCOS 0020-04; Reactor Building Floor Drain Sump Ball Valve Leakage Testing; Revision 3

- QCTP 0130-11; Internal Flood Protection Program; Revision 5

- EC 625010; U2 EDG Projected Engine Outlet Water Temperature at Maximum Load;

Revision 0

- ECAP 4148330; U2 EDG Trip During QCOS 6600-42; 08/09/2018

- IR 4148330; U2 EDG Trip During QCOS 6600-42; 06/19/2018

- Letter from Engine Systems, Inc. to Exelon EDG/SBODG Senior Systems Manager;

09/05/2018

71111.19Post Maintenance Testing

- WO 4676246; (LR) Diesel Fire Pump B Capacity Test; 08/03/2018

- WO 4598327-01; Perform Diesel Fire Pump Capacity Test; 08/14/2018

- QCOS 5750-09; ECCS Room and DGCWP Cubicle Cooler Monthly Surveillance; Revision 37

- QCOS 6600-57; Unit 2 Diesel Generator Endurance And Margin/Full Load Reject /Hot Restart

Test; Revision 8

- Drawing M-27, Sheet 1; Diagram of Fire Protection System; Revision QN

- QCMMS 4100-33; 1/2-4101B Diesel Driven Fire Pump Annual Capacity Test; Revision 33

- QCOS 6600-57; Unit 2 Diesel Generator Endurance Margin/Full Load Reject/Hot Restart;

Revision 8

71111.20Refueling and Other Outage Activities

- QCGP 1-1; Normal Unit 1 Startup; Revision 112

- EC 622330; Q1C25 Startup Sequence and Notch Worth Evaluation 7500 through 14500

MWD/MTU; Revision 0

71111.22Surveillance Testing

- QCOS 1100-07; SBLC Pump Flow Rate Test; Revision 38

- QCOS 6600-43; Unit 1/2 Emergency Diesel Generator Load Test; Revision 48

- IR 4166083; RCIC Pressure Indicator Fluctuation; 08/21/2018

- QCOS 1300-05; RCIC Pump Operability Test; Revision 56

- WO 4789995-01; RCIC Pump Operability Test (IST); 08/21/2018

71151Performance Indicator Verification

- EC 0000397569-000; Review of Loss of Secondary Containment DP Noted in IR 1629224;

04/10/2014

- ECR 0000434663; Review EC Eval 397569 Against IR 4099281; 04/23/2018

- IR 4099281-04; Received Unexpected U2 Fuel Pool Radiation Alarm; 03/23/2018

71152Problem Identification and Resolution

- IR 4160504; NSR Fuse Replaced; 07/31/2018

- IR 4166305; OM Code Non-Conformance-Relief Valve Exceeded 10 YR Limit; 08/22/2018

- IST-QDC-Plan; 10CFR50.55a Relief Request; RV-01

- ECAPE 4160251-06; Unexpected Alarm 901-5 D9, Chan A Main Steam Tunnel Hi Temp;

09/20/2018

- IR 4166684; OOT, 2-1705-16A, Trend Code B4; 08/23/2018

- IR 4166687; OOT, 2-1705-16B, Trend Code B4; 08/23/2018

- IR 4170073; U1 HPCI FIC 1-2340-01 Erratic When in Standby; 09/05/2018

- Drawing 4E-6613A; Schematic Diagram Motor Operated Valves 0-2901-6 and 0-2901-7

Safe Shutdown System; Revision H

- QCOP 2900-01; Safe Shutdown Makeup Pump System Preparation for Standby Operation;

Revision 39

71153Follow-Up of Events and Notices of Enforcement Discretion

- CY-AB-120-100; Reactor Water Chemistry; Revision 19

- CY-QC-160-200; Chemistry Excursions and Adverse Trends/Conditions; Revision 2

- IR 4160251; Unexpected Alarm 901-5 D9, Chan A Main Steam Tunnel Hi Temp; 07/31/2018

- IR 4160631; Unit 1 Resin Intrusion Event; 08/01/2018

- IR 4161425; 4.0 Critique for Unit 1 Chemistry Excursion; 08/03/2018

- IR 4177502; Unit 1 Reactor Scram from 100% Reactor Power; 09/27/2018

- Unit 1 Operator Logs from 07/31/2018 to 08/01/2018

- QCOP 6400-35; Performing Transmission Switching Orders; Revision 15

- QCIS 0200-20; Unit 1 Reactor High Pressure Automatic Blowdown Calibration and Functional

Test; Revision 16

- Drawing 4E-1338; Schematic Diagram Generator and Transformer Tripping Relays Primary

System; Revision AX

60853On-site Fabrication of Components and Construction of an ISFSI

- IR 04147163; NRC ID. ISFSI Concrete Density Test Not Performed; 06/14/2018

- Attachment to Letter No. 2017-QDC-205; S&L Review of Terracons Plate Load Test Report

dated 10/11/2017; 10/17/2017

- CC-AA-103-100; Configuration Change Control for Permanent Physical Plant Changes;

Revision 0

- Certificate of Accreditation for Terracon Consultants, Inc. in Bettendorf, Iowa, USA;

06/22/2017

- Certifications for Terracon Concrete Field Testing Personnel

- EC 405175; Dry Cask Storage Project Installation of ISFSI Pad 2; Revision 000

- EC 405175; WPI Attachment 2; Concrete Preparation & Execution Checklist; Revision 000

- PI-AA-120; Issue Identification and Screening Process; Revision 8

- PI-AA-125; Corrective Action Program (CAP) Procedure; Revision 6

- Quad Cities Generating Station ISFSI Expansion Project Specific ISFSI Pours A-C & G-J

Plan; 05/24/2018

- QC-72.48-S-17-0001; Dry Cask Storage Project Installation of ISFSI Pad 2; 06/12/2017

- QCNPS 10 CFR 72.212 Evaluation Report; Revision 12

- RRTI-2144-011; Response to Request for Technical Information (RRTI) Holtec International;

Revision 0

- Specification Q-2052; ISFSI Expansion ISFSI Pad, Final Grading and Misc. Concrete

Structures; 05/01/2017

- 07161047.0146; Concrete Compressive Strength Test Report; 07/02/2018

- 07161047.0149; Concrete Compressive Strength Test Report; 07/02/2018

- Terracon Project No. 07161047; Plate Load Test Submittal; 09/19/2017

- CM-001-08-10; Terracon Quality Management System Manual; Revision 4.3

- Submittal 2 Rev 1; Rebar Shops Revised ISFSI PAD; 07/14/2017

- Submittal 3; Rebar Shops REVISED Apron; 07/14/2017

- Submittal 3 Cert; Rebar Certifications + Tests; 10/04/2017

- Terracon Project No. 07161047; Plate Load Test Report; 10/11/2017

- Submittal 3 Cert R1; Rebar Test Certifications; 10/30/2017

- Submittal 16 R7; Re: Mix #41 Concrete Submittal; 04/18/2018

- EC 405175; Revision Summary including AWA 1 through AWA 14; 05/30/2018

- ACI 311.6-09; Specification for Ready Mixed Concrete Testing Services; November 2009

- 07161047.0160; Concrete Cylinder Density Report; 06/28/2018

- HI-2022444; HI-STORM 100 System FSAR; Revision 11.1

- IR 04075816; ISFSI Expansion Mud Mat Compressive Strength; 11/17/2017

60855.1Operation of an Independent Spent Fuel Storage Installation at Operating Plants

- QC-00-18-00205; 2018 Spent Fuel Dry Cask Storage Project ALARA Plan; 05/07/2018

- QDC-0836-N-2323; Quad Cities Fuel Selection Package for 2018 CampaignISFSI;

04/24/2018

- SA Report 4099184; Quad Cities Spent Fuel Loading Campaign Readiness Assessment;

04/20/2018

- QCNPS 10 CRR 72.212 Evaluation Report; Revision 13

- WO 04588877; ISFSI Lift Yoke Inspection; 01/19/2018

- WO 04592088; ISFSI MPC Lift Cleat Inspection; 01/04/2018

- WO 04643972; EM Reactor Building Ovehead Crane; 05/04/2018

- ISFSI Corrective Action Program Documents; 2017-2018

- Reactor Building Overhead Crane Corrective Action Program Documents; 2017-2018

- RP-AA-305; Holtec HI-TRAC Radiation Surveys; Revision 000

- RP-AA-306; Holtec HI-STORM Radiation Surveys; Revision 000

- RP-AA-307; Holtec ISFSI Radiation Survey; Revision 000

- QCFHP 0800-82; MPC Unloading Operations; Revision 006

- QCFHP 0800-72; HI-STORM Processing; Revision 011

- QCFHP 0800-70; HI-TRAC Loading Operations; Revision 20

- OU-MW-671-200; MPC Processing Forced Helium Dehydration for BWRs; Revision 001

- OU-MW-671-200; MPC Processing Forced Helium Dehydration for BWRs; Revision 002

- OU-MW-671-200; MPC Processing Forced Helium Dehydration for BWRs; Revision 002A

- PI-CNSTR-OP-H-EXE-01; Closure Welding of Holtec Multi-Purpose Canisters at Exelon

Facilities; Revision 18

- MSLT-MPC-EXELON-ON-MW; Helium Mass Spectrometer Leak Test Procedure

Multi-Purpose Canister; Revision 2

- H2-MON-002; Hydrogen Monitoring for Holtec Canisters; Revision 10

- HI-2022966; Forced Helium Dehydrator Sourcebook; Revision 4

- AR 04163723; U-2 Refuel Bridge Main Hoist Load Cell Abnormal Indication; 08/13/2018

- AR 04163993; U-2 Refuel Bridge Main Trolley Position Indication Erratic; 08/13/2018

- AR 04168081; FHD Unit Trips Due to High Temperature Alarm; 08/29/2018

- AR 04168791; DCS-FHD System Processing Configuration Discrepancy; 08/30/2018

- PI-AA-120; Issue Identification and Screening Process; Revision 8

- OP-AA-116-101; Equipment Labeling; Revision X