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Evaluations or Changes, Tests, and Experiments and Permanent Plant Modifications Baseline Inspection Report 05000263/2016008
	ML16113A346
Person / Time
Site:	Monticello
Issue date:	04/19/2016
From:	Robert Daley; Engineering Branch 3
To:	Gardner P; Northern States Power Company, Minnesota
References
	IR 2016008
	Download: ML16113A346 (12)
	v • d • e

UNITED STATES

NUCLEAR REGULATORY COMMISSION

REGION III

2443 WARRENVILLE RD. SUITE 210

LISLE, IL 60532-4352

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 55362-9637

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 with the Commissions 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 (10 CFR) 2.390, Public

Inspections, Exemptions, Requests for Withholding, of the NRC's "Rules of Practice," a copy

of this letter, its enclosure, and your response (if any) will be available electronically for public

inspection in the NRCs Public Document Room or from the Publicly Available Records (PARS)

component of the NRC's Agencywide Documents Access and Management System (ADAMS).

ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html

(the Public Electronic Reading Room).

Sincerely,

/RA/

Robert C. Daley, Chief

Engineering Branch 3

Division of Reactor Safety

Docket No. 50-263

License No. DPR-22

Enclosure:

IR 05000263/2016008

cc: Distribution via LISTSERV

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

Enclosure

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. The U.S. Nuclear Regulatory Commissions

program for overseeing the safe operation of commercial nuclear power reactors is described

in NUREG-1649, Reactor Oversight Process, Revision 5, dated February 2014.

NRC-Identified and Self-Revealed Findings

No findings were identified.

Licensee-Identified Violations

No violations were identified.

2

REPORT DETAILS

1. REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity

1R17 Evaluations of Changes, Tests, and Experiments and Permanent Plant Modifications

(71111.17T)

.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 96-07, Guidelines for

10 CFR 50.59 Implementation, Revision 1, to determine acceptability of the completed

evaluations, and screenings. The Nuclear Energy Institute document was endorsed by

the NRC in Regulatory Guide 1.187, Guidance for Implementation of 10 CFR 50.59,

Changes, Tests, and Experiments, dated November 2000. The inspectors also

consulted Part 9900 of the NRC Inspection Manual, 10 CFR Guidance for

10 CFR 50.59, Changes, Tests, and Experiments.

This inspection constituted 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

Specification (TS) 3.4.8, Residual Heat Removal (RHR) Shutdown Cooling System -

Cold Shutdown. Specifically, the licensee failed to verify that the capability of the

alternate methods of decay heat removal described in Operations Manual C.4-B.03.04.A,

Loss of Normal Shutdown Cooling, were adequate to combat a loss of shutdown cooling

resulting from the loss of one or two RHR subsystems while in MODE 4 with high decay

heat load.

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Description: The Limiting Condition for Operation (LCO) 3.4.8 of TS Residual 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 cooling water to the heat exchanger. 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 allowed to be common to both subsystems.

When TS 3.4.8, LCO could not be met, Condition A, for one or two RHR shutdown

cooling subsystems inoperable, the Required Action was to, verify an alternate

method of decay heat removal was available for each inoperable RHR shutdown

cooling subsystem. The completion time for the required action was 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , and

once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 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-B.03.04.A, Loss of Normal Shutdown

Cooling, provided instructions for establishing alternate methods for decay heat

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.

4

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

Operations Manual was required per TS 5.4.1 Procedures, which required that 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 the inspectors concerns into 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.

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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.

4. 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

6

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

Failure to provide acceptable Alternate Methods of Decay Heat

05000263/2016008-01 URI

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

Attachment

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 EC 23981 EDG Fuel Oil Tank Vent Lines Missile Protection 0

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 0

Leakage Rates

SCR-15-0093 EDG ESW Basket Strainer Modification

SCR-15-0115 C.4-B.09.02.A Revision to Resolve CAP AR 01465720 1

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 1

Cooling Water

09-176 Evaluation for Debris Disposition in Supply Pipe and Motor Cooler 0

Tube

09-178 Time to reach the RHRSW Pump Motor Cooling Line Strainer 0

Limiting Pressure Differential

14-025 Instrument Setpoint Calculation - Time Delay for Transfer to EDG 0

on Loss of Voltage

90-023 EC 23085 - 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

2

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

1516098 Actions for when LCO 3.4.8, RA A.1 not met Unclear 03/17/2016

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 07/27/1991

system and Core Spray Pump Motors

01196451 CDBI EDG Base Tank Volume Calculation 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 12/17/2013

Known

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 - Turbine Building EL 951-0 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

3

MODIFICATIONS

Number Description or Title Revision

EC-14065 RHRSW Motor Cooler Strainers 1

EC-20887 LT-5200 River Level Setpoint Change for Upper Value

EC-21934 Evaluation of Corrosion Found in the 11 EDG Coolant 0

Expansion Tank

EC-21999 Equivalency Evaluation: RHRSW-17 is the emergency injection 0

check valve for the RHR to RSW crosstie

EC-22008 Monticello 125V #12 Battery Modified Performance Test Profile 0

EC-22414 SQUG Evaluation of Diesel Oil Service Pump P-77 0

EC-23085 EDG Fuel Oil Train Separation 0

EC-23272 Revise EDG Base Tank Fuel Oil Level Calc 90-023

EC-23616 Revise Setpoints for Relays 97-29 and 97-31 0

EC-23857 Recirc Pump Seal Water Piping 0

EC-25889 Operating with HPCI CV-2043 (Steam Trap Bypass) Open 0

EC-25266 EDG Fuel Oil Separation 0

ML16113A346

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