IR 05000237/2015007

IR 05000237/2015007; 05000249/2015007; on 05/11/2015 - 05/29/2015; Dresden Nuclear Power Station; Evaluations of Changes, Tests, and Experiments and Permanent Plant Modifications (Gmh)

ML15183A063
Person / Time
Site:	Dresden
Issue date:	07/01/2015
From:	Dariusz Szwarc Engineering Branch 3
To:	Bryan Hanson Exelon Generation Co, Exelon Nuclear
References
IR 2015007
Download: ML15183A063 (20)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION REGION III 2443 WARRENVILLE RD. SUIT E 210 LISLE, IL 60532

-4352 July 1, 2015 Mr. Bryan Senior VP, Exelon Generation Company, LLC President and CNO, Exelon Nuclear 4300 Winfield Road Warrenville, IL 60555

SUBJECT: DRESDEN NUCLEAR POWER STATION - EVALUATION S OF CHANGES, TESTS, AND EXPERIMENTS AND PERMANENT PLANT MODIFICATIONS BASELINE INSPECTION REPORT 05000237/2015007; 05000249/2015007

Dear Mr. Hanson:

On May 29, 2015, the U.S. Nuclear Regulatory Commission (NRC) completed an Evaluations of Changes, Tests, and Experiments and Permanent Plant Modifications inspection at your Dresden Nuclear Power Station

. The enclosed inspection report documents the inspection results, which were discussed on May 29, 2015, with Mr. Shane Marik, and other members of your staff.

The inspection examined activities conducted under your license as they relate to safety and compliance with the Commission's rules and regulations

, and with the conditions of your license. The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel.

NRC inspectors documented t wo NRC-identified finding s of very-low safety significance (Green) in this report

. These findings were determined to involve violation s of NRC requirements.

However, because of the ir very-low safety significance

, and because the se issues were entered into your Corrective Action Program, the NRC is treating the se issues as Non-Cited Violations (NCVs) in accordance with Section 2.3.2, of the NRC Enforcement Policy.

If you contest the subject or severity of the 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, U.S. Nuclear Regulatory Commission, Washington, DC 20555

-0001; and the NRC Resident Inspector at Dresden Nuclear Power Station

. In addition, if you disagree with the cross

-cutting aspect assigned to any finding 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 Regional Administrator, Region III, and the NRC Resident Inspector at Dresden Nuclear Power Station

. 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 NRC's 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/

Dariusz Szwarc, Acting Chief Engineering Branch 3 Division of Reactor Safety Docket Nos. 50

-237, 50-249;72-037 License Nos. DPR

-19; DPR-25

Enclosure:

Inspection Report 0500 0237/2015007; 05000249/2015007 cc w/encl:

Distribution via LISTSERV Enclosure U. S. NUCLEAR REGULATORY COMMISSION REGION III Docket No s: 50-237; 50-249 License No s: DPR-19; DPR-25 Report No:

05000237/2015007; 05000249/2015007 Licensee:

Exelon Generation Company, LLC Facility:

Dresden Nuclear Power Station Location:

Morris, IL Dates: May 11 - 29, 2015 Inspectors:

George M. Hausman, Senior Engineering Inspector (Lead)

Jorge J. Corujo-Sandin, Engineering Inspector Mark T. Jeffers, Engineering Inspector Observer:

Christopher A. Hunt, Reactor Engineer Approved by:

Dariusz Szwarc, Acting Chief Engineering Branch 3 Division of Reactor Safety 2

SUMMARY

Inspection

Report 05000237/2015007; 05000249/2015007; 05/11/2015 - 05/29/2015; Dresden Nuclear Power Station

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.

Two findings of very-low safety significance were identified by the inspectors. The findings were considered Non-Cited Violations of U.S. Nuclear Regulatory Commission (NRC)regulations. The significance of most findings is indicated by their color (i.e., greater than Green, or Green, White, Yellow, Red) using Inspection Manual Chapter (IMC)0609, "Significance Determination Process (SDP).

" Cross-cutting aspects were determined using IMC 0310, "Aspect s within the Cross-Cutting Areas." Findings for which the SDP does not apply may be Green

, or be assigned a severity level after NRC management review. All violations of NRC requirements are dispositioned in accordance with the NRC's Enforcement Policy dated J uly 9, 2013. The NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG

-1649, "Reactor Oversight Process," Revision 5, dated February 2014

. Cornerstone

Mitigating Systems

Green.

The inspectors identified a finding of very-low safety significance

, and an associated Non-Cited Violation (NCV) of Title 10, Code of Federal Regulations (CFR), Part 50, Appendix B, Criterion III, "Design Control

," for the licensee's failure to ensure that applicable regulatory requirements and the isolation condenser

's (IC's) design bases were correctly translated into procedures. Specifically, the licensee added steps to the IC control procedures which directed operators to secure the IC in order to prevent the water level in the shell from going below 3.5 feet. The added steps would result in the IC being shutdown when required to operate per the IC's design bases. The licensee entered th e issue into their Corrective Action Program (CAP) as Action Request 02506445, "NRC MOD/5059 Inspection:

ISCO [Isolation Condenser]

Operating Procedures," dated May 28, 2015.

The performance deficiency was determined to be more than minor because the finding was associated with the Mitigating Systems cornerstone attribute of Procedure Quality

, and affected the cornerstone

's objective of ensuring the capability of systems that respond to initiating events to prevent undesirable consequences (i.e.,

core damage). Specifically, the inadequate procedures would drive the operators to stop the IC during a design bases event and prevent the IC from performing its design function of removing decay heat from the reactor.

The finding ha s a cross-cutting aspect in the area of Human Performance; Teamwork, because the licensee did not communicate and coordinate activities within and across organizational boundaries to ensure nuclear safety is maintained. Specifically, the Operations Department failed to communicate and coordinate with the Engineering Department when developing the procedural changes.

[H.4] (Section 1R17.1b)

Green.

The inspectors identified a finding of very-low safety significance

, and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," for the failure to account for increased fuel oil consumption during the development of the Emergency Diesel Generator (EDG)

Calculation 10553-CALC-07, "Dresden Station 3 Emergency Diesel Generators Endurance Calculations," Revision 2, which resulted in non-conservative Technical Specifications (TS). Specifically, the licensee failed to account for the increased fuel oil consumption at an EDG frequency of 61.2 Hertz (Hz), and ensure that the minimum fuel oil level in the EDG day tanks, as required per TS 3.8.1.4, was adequate to support the EDGs' mission time at 110 percent for one hour. The licensee entered the issue into their CAP as Action Request 02506869, "NRC MOD/5059 Inspection:

Emergency Diesel Generator Fuel Consumption," dated May 28, 2015.

The performance deficiency was determined to be more than minor because the finding was associated with the Mitigating Systems cornerstone attribute of design control and affected the cornerstone's objective of ensuring the capability of systems that respond to initiating events to prevent undesirable consequences (i.e.,

core damage). Specifically, the licensee failed to account for the increased fuel oil consumption resulting from operation at a higher EDG frequency

. Therefore, the licensee did not ensure that th e minimum fuel oil level in the day tanks, as required per TS 3.8.1.4, was adequate to support the EDGs' mission time at 110 percent for one hour. This finding ha s a cross-cutting aspect in the area of Problem Identification and Resolution

Identification

, because the licensee did not did not thoroughly evaluate the EDG fuel oil consumption when considering EDG frequency variation.

Specifically, the licensee failed to translate applicable design bases into specifications which resulted in non-conservative TS. [P.1] (Section 1R17.2b)4

REPORT DETAILS

REACTOR SAFETY

Cornerstone s: Initiating Events, Mitigating Systems, and Barrier Integrity 1R17 Evaluation s of Changes, Tests, and Experiments and Permanent Plant Modifications (71111.17 T)

.1 Evaluation

of Changes, Tests, and Experiments

a. Inspection Scope

The inspectors reviewed 7 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 15 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 (NEI) 96-07, "Guidelines for 10 CFR 50.59 Implementation," Revision 1, to determine acceptability of the completed evaluations, and screenings. The NEI document was endorsed by the NRC in Regulatory Guide (RG) 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 7 samples of evaluations

, and 15 samples of screenings and/or applicability determinations as defined in Inspection Procedure (

IP) 71111.17-04.

b. Findings

Procedure Revisions Result ed in Isolation Condenser Unable to Meet Design Basi s Introduction

The inspectors identified a finding of very

-low safety significance (Green)

, and an associated Non

-Cited Violation (NCV) of 10 CFR, Part 50, Appendix B, Criterion III, "Design Control

," for the licensee's failure to ensure that applicable regulatory requirements and the isolation condenser's (IC's) design bas es were correctly translated into procedures. Specifically, the licensee added steps to the IC control procedures which directed operators to secure the IC in order to prevent the water level in the shell from going below 3.5 feet. The added steps would result in the IC being shutdown when required to operate per the IC's design bases

.

5 Description

The safety

-related IC system functions as a heat sink for decay heat removal from the reactor vessel following a reactor scram

, and isolation from the main condenser. Each IC (one per unit) consists of two tube bundles immersed in a large water storage tank.

The IC system operates by natural circulation. During operation

, the IC tube side will contain reactor coolant water

, and the shell side will contain clean demineralized water. The inspector s observed that the licensee made changes to three IC control procedures. The procedure change s directed operators to secure the IC on a low

-shell side level condition. Specifically, the IC would be secured if the shell side level c ould not be maintained above 3.5 feet. At this height, the water level in the shell is just above the top of the tube bundles. The purpose of the procedure change was to prevent uncovering of the tube bundles in order to protect the IC for future availability (i.e., when the shell side level could be restored to ensure the IC would be available to respond to a Beyond Design Basis External Event

). The procedures affected were as follows

DOP 1300-02, "Automatic Operation of Isolation Condenser," Revision 25, Step G.4.a; DOP 1300-03, "Manual Operation of the Isolation Condenser," Revision 34, Step G.9.a; and DAN 902(3)-3D-4, "Isolation Condenser Level Hi/Low Annunciator Respond Procedure

," Revision 12, Step B.2.a. In the Updated Final Safety Analysis Report (UFSAR), Section 5.4.6 and the Technical Specification s (TS) Bases 3.5.3, the IC design basi s was describe d as: (1) remove 252.5 Million British thermal units (

MBtu)/hour, which is equivalent to the decay heat rate 8.8 minutes after the scram; and (2) provide sufficient decay heat removal capability for 20 minutes of operation without makeup water to the shell. The TS require s a number of surveillance requirements (SRs) be performed to ensure these bases are met, including the following:

SR 3.5.3.1:

Verify the IC 6 feet, and shellside water 210°F; and SR 3.5.3.4:

Verify the IC system heat removal capability to remove design heat load [i.e., 252.5MBtu/hour]. The licensee maintains the shell water level abov e 7 feet via administrative controls and monitors the temperature in an effort to remain well below the temperature limits. However, the licensing bases of the isolation condense r states that water level of the shell is expected to go below 3.5 feet in order to mitigate a credited event under design bases conditions.

As a result, procedural guidance

, which would prevent the isolation condenser shell water level from going below 3.5 feet, would preclude the component from performing its design function under design bases conditions.

As part of immediate corrective actions the licensee marked the procedures for review in order to determine required changes. In addition, the licensee demonstrated the shell water temperature was monitored in order to maintain it well below TS requirements and the water level was administratively maintained at or above 7 feet. This provides additional margin to the IC to perform its design function prior to reaching the 3.5 foot limit.

6 The licensee documented the inspectors' concern under AR 02506445, "NRC Mod/5059 Inspection:

ISCO [Isolation Condenser]

Operating Procedures," dated May 28, 2015.

The licensee plans to evaluate the IC procedural changes and determine what modifications are needed. In addition, the licensee is considering performing an Apparent Cause Evaluation to evaluate the concern

. The inspectors concluded that procedural changes were performed by the Operations Department without engaging the Engineering Department to ensure there were no adverse impact s to the IC or associated design and licensing bases.

Analysis:

The inspectors determined that the licensee added steps to the IC control procedures which directed operators to secure the IC in order to prevent the water level in the shell from going below 3

.5 feet, which was contrary to

10 CFR, Part 50, Appendix B, Criterion III, "Design Control

," and was a performance deficiency. Specifically, the added steps would result in the IC being shutdown when required to operate per the IC's design bases

. The performance deficiency was determined to be more than minor because the finding was associated with the Mitigating Systems cornerstone attribute of Procedure Quality

, and affected the cornerstone's objective of ensuring the availability, reliability and capability of the systems that respond to initiating events to prevent undesirable consequences (i.e.,

core damage). Specifically, the inadequate procedures would drive the operators to stop the IC during a design bases event and prevent the IC from performing its design basis function of removing decay heat from the reactor.

In accordance with Inspection Manual Chapter (IMC)0609, "Significance Determination Process," Attachment 0609.04, "Initial Characterization of Findings

," Table 2, the inspectors determined the finding affected the Mitigating Systems cornerstone.

As a result, the inspectors determined the finding could be evaluated using Appendix A, "The Significance Determination Process for Findings At

-Power," Exhibit 2, for the Mitigating Systems cornerstone. The performance deficiency affected the design or qualification of mitigating structure s, systems, and component s (SSC); however, the SSC maintained its operability or functionality as applicable.

Specifically, the licensee administratively maintains the shell water level at or above 7 feet and the temperature is maintained below 210 degrees Fahrenheit. This provides additional margin to the IC to perform its design function prior to reaching the 3.5 foot limit. Therefore, the inspectors answered "yes" to the Mitigating Systems Screening Question A.1 in Exhibit 2, and screened the finding as having very

-low safety significance (Green).

The finding has a cross-cutting aspect in the area of Human Performance

Teamwork because the licensee did not communicate and coordinate activities within and across organizational boundaries to ensure nuclear safety is maintained. Specifically, the Operations Department failed to communicate and coordinate with the Engineering Department when developing the procedural changes. [H.4]

Enforcement

Title 10 CFR, Part 50, Appendix B, Criterion III, "Design Control," requires, in part, that measures shall be established to assure that applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures, and instructions

. Contrary to the above, from April 22, 2013, to May 29, 2015, the licensee failed to ensure that applicable regulatory requirements and the IC's design basis were correctly translated into procedures. Specifically, the licensee added steps to the 7 IC control procedures which directed operators to secure the IC in order to prevent the water level in the shell from going below 3.5 feet. The added steps would result in the IC being shutdown when required to operate per the IC's design bases.

This violation is being treated as an NCV, consistent with Section 2.3.2 of the Enforcement Policy because it was of very

-low safety significance and was entered into the licensee's CAP as Action Request 02506445, "NRC MOD/5059 Inspection:

ISCO Operating Procedures," dated May 28, 2015.

The licensee will continue to administratively control level of the shell at or above seven feet to provide additional margin to the IC to perform its function. (NCV 05000237/2015007

-01; NCV 05000249/2015007

-01, Procedure Revisions Resulted in Isolation Condenser Unable to Meet Design Basis)

.2 Permanent

Plant Modifications

a. Inspection Scope

The inspectors reviewed nine permanent plant modifications that had been installed in the plant during the last 3 years. This review included in

-plant walkdowns for portions of the modified High Pressure Coolant Injection (HPCI)system to assess recent replacement of the auxiliary oil pump.

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.

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 constitut ed nine permanent plant modification samples as defined in IP 71111.17-04.

b. Findings

Emergency Diesel Generator Usable Fuel Calculations Failed to Consider Appropriate Emergency Diesel Generator Frequency Variations

Introduction:

The inspectors identified a finding of very low safety significance (Green) and an associated NCV of 10 CFR, Part 50, Appendix B, Criterion III, "Design Control," for the failure to account for increased fuel oil consumption during the development of the Emergency Diesel Generator (EDG)endurance calculations which resulted in non-conservative TS. Specifically, the licensee failed to account for the increased fuel oil consumption at an EDG frequency of 61.2 Hertz (Hz), and ensure that the minimum fuel oil level in the EDG day tanks, as required per TS 3.8.1.4, was adequate to support the EDGs' mission time at 110 percent for one hour.

8 Description

The TSs SR 3.8.1.2 allows an EDG frequency tolerance of +/-

2 percent.

This tolerance was based on RG 1.9, "Application and Testing of Safety

-Related Diesel Generators in Nuclear Power Plants

," Revision 4, requirements that the EDG frequency recover to within +/-

2 percent of 60 Hz (i.e., 58.8 - 61.2 Hz) within a specified period during the sequencing of loads on the bus.

Therefore, the EDGs could operate at a frequency of 61.2 Hz, which would be the worst

-case scenario for loading of the EDGs.

Additionally, TS SR 3.8.1.4 verifies adequate level of fuel oil in the day tank and the bulk storage tanks. The level selected is to ensure adequate fuel oil for a minimum of one hour of EDG operation at 110 percent of full load for the day tank and approximately two days at 100 percent full load for the bulk storage tanks.

The levels identified by TS SR 3.8.1.4 ar e 205 gallons and 10,000 gallons for the day tank and bulk storage tank, respectively.

During review of Calculation 10553-CALC-07, "Dresden Station Emergency Diesel Generators Endurance Calculations," Revision 2, the inspectors questioned why the licensee based their fuel consumption on the EDGs operating at 110 percent at 60 Hz rather than 61.2 Hz as allowed by TS SR 3.8.1.2. The higher frequency would result in higher fuel consumption; therefore, would be more conservative.

Specifically, the estimated fuel consumption at 110 percent loading at 61.2 Hz would be approximately 211.3 gallons/hour. The calculation did identify that the 2 percent frequency tolerance would result in higher fuel consumption; however, the more conservative frequency was only applied to calculating EDG loading at 100 percent. The calculation of the day tank level at 110 percent only considered the less conservative 60 Hz frequency.

The inspectors determined that the EDGs could operate at a steady state frequency up to 61.2 Hz according to TS SR 3.8.1.2. This would result in a higher fuel consumption that would exceed the TS 3.8.1.4 minimum volumetric fuel requirements.

The TS 3.8.1.4 minimum fuel requirements were based on operating the EDGs at a frequency up to 60 Hz, rather than 61.2 Hz, which resulted in non-conservative TS.

The inspectors discussed the issue with the licensee and identified that the licensee ha d administrative procedures that would limit the frequency of the EDG to 60.5 Hz and would ensure the day tank level remained greater than 350 gallons. The Dresden Procedure DGA

-12, "Partial or Complete Loss of AC Power," Revision 73, ensures operators maintain frequency of the EDGs between 59

.5 to 60.5 Hz.

Additionally, Dresden Surveillance Procedure DOS 6600-14, "Diesel Oil Transfer Pump Operation and Fuel Consumption Test," Revision 20, requires operators to maintain at least 350 gallons in the day tank. Therefore, the EDG would remain capable of performing its specified safety function. However, even with the administrative limits, the minimum fuel requirements identified in TS SR 3.8.1.4 would remain non

-conservative since the fuel consumption would still be higher at the 60.5 Hz which is not represented in TS SR 3.8.1.4. The licensee captured this issue and entered it into their CAP as Action Request 02506869. The licensee intends to evaluate the effect of the increased frequency on their EDG Calculations.

Analysis:

The inspectors determined that the licensee

's failure to account for increased fuel oil consumption during the development of the EDG Calculation 10553-CALC-07, "Dresden Station Emergency Diesel Generators Endurance Calculations," Revision 2, resulted in non

-conservative TS and was contrary to 10 CFR, Part 50, Appendix B, Criterion III, "Design Control

," and was a performance deficiency. Specifically, the 9 licensee failed to account for the increased fuel oil consumption at an EDG frequency of 61.2 Hz, and ensure that the minimum fuel oil level in the EDG day tanks, as required per TS 3.8.1.4, was adequate to support the EDGs' mission time at 110 percent for one hour. The performance deficiency was determined to be more than minor because the finding was associated with the Mitigating Systems cornerstone attribute of design control and affected the cornerstone

's objective of ensuring the capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage)

. Specifically, the licensee failed to account for the increased fuel oil consumption resulting from operation at a higher EDG frequency

. Therefore, the licensee did not ensure that the minimum fuel oil level in the day tanks, as required per TS 3.8.1.4, was adequate to support the EDGs' mission time at 110 percent for one hour

. In accordance with IMC 0609, "Significance Determination Process," Attachment 609.04, "Initial Characterization of Findings," Table 2, the inspectors determined the finding affected the Mitigating Systems cornerstone. As a result, the inspectors determined the finding could be evaluated using Appendix A, "The Significance Determination Process for Findings At

-Power," Exhibit 2, for the Mitigating Systems cornerstone.

The performance deficiency affected the design or qualification of a mitigating SSC; however, the SSC maintained its operability or functionality as applicable. Specifically, the licensee was able to demonstrate that adequate fuel in the storage tanks would be available to support the EDGs mission time when operating at the administratively controlled higher frequency limit specified in procedures. Therefore, the inspectors answered "no" to all the Mitigating Systems Screening Questions in Exhibit 2, and screened the finding as having very

-low safety significance (Green). This finding has a cross

-cutting aspect in the area of Problem Identification and Resolution

Identification because the licensee did not thoroughly evaluate the EDG fuel oil consumption when considering EDG frequency variation. Specifically, the licensee failed to translate applicable design bases into specifications which resulted in non-conservative TS. [P.1]

Enforcement

Title 10 CFR, Part 50, Appendix B, Criterion III, "Design Control," requires, in part, that measures shall be established to ensure that applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures, and instructions

. Contrary to the above, from May 25, 2011, until May 29, 2015, the licensee failed to ensure that applicable regulatory requirements and the design basis were correctly translated into specifications during development of the EDG Calculation 10553

-CALC-07, "Dresden Station Emergency Diesel Generators Endurance Calculations," Revision 2, which resulted in non-conservative TS

. Specifically, the licensee failed to account for the increased fuel oil consumption at an EDG frequency of 61.2 Hz, and ensure that the minimum fuel oil level in the EDG day tanks, as required per TS 3.8.1.4, was adequate to support the EDGs' mission time at 110 percent for one hour.

This violation is being treated as an NCV, consistent with Section 2.3.2 of the Enforcement Policy because it was of very

-low safety significance, and was entered into the licensee's CAP as Action Reque st 02506869, "NRC MOD/5059 Inspection:

EDG Fuel Consumption," dated May 28, 2015. The licensee is evaluating the effect of the increased frequency on their EDG calculations, and will continue to administratively 10 control level in the fuel oil day tank high enough to ensure the SSC remains operable. (NCV 05000249/2015007-02; NCV 05000237/

2015007-02, EDG Usable Fuel Calculations Failed to Consider Appropriate EDG Frequency Variations)

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 10 CFR 50.59 evaluations and permanent plant modifications. The inspectors reviewed these documents to evaluate the effectiveness of corrective actions related to permanent p lant 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 May 29, 2015, the inspector s presented the inspection results to Mr. Shane Marik

, 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 reviewed during the inspection was returned to the licensee staff.

ATTACHMENT:

SUPPLEMENTAL INFORMATION

SUPPLEMENTAL

INFORMATI

ON KEY POINTS OF CONTAC

T Licensee

G. Baxa, Regulatory Assurance

M. Budelier, Senior Manager Design Engineering

D. Doggett, Regulatory Assurance

D. Eaman, Design Instrumentation and Control

(I&C) Manager

B. Franzen, Regulatory Assurance

Manager

R. Gaston, Licensing Manager

- Corporate

T. Griffith, Senior Licensing Engineer

B. Kapellas, Maintenance Director

B. Madderom, Design Electrical Manager

S. Marik, Site Vice President

G. Morrow, Operations Directo

r

P. O'Brien, Site Corrective Action Program

Manager

R. Osgood, Senior Nuclear Site Communications Specialist

E. Rogers, Nuclear Oversight

R. Schmidt, Chemistry Manager

B. Surges, Work Control

D. Walker, NRC Coordinator

J. Washko, Plant Manager

D. Wolverton, Design Mechanical Manager

U.S. Nuclear Regulatory Commission K. O'Brien, Director, Division of

Reactor Safety

G. Roach, Senior Resident Inspector

LIST OF ITEMS OPENED, CLOSED AND DISCUSS

ED Opened and Closed

05000237/2015007-01; 05000249/2015007

-01 NCV Procedure Revisions Resulted in Isolation Condenser Unable to Meet Design Basis

(Section 1R17.1b.)05000237/2015007-02; 05000249/2015007

-02 NCV EDG Usable Fuel Calculations Failed to Consider Appropriate EDG Frequency Variations

(Section 1R17.2b.) Discussed None

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.

ANALYSIS (ENGINEERING)

Number Description or Title

Date or Revision

DR-27D-M-002 Dresden SBO Building Ventilation Air Requirement

DRE13-0004 Use of U2/3 or U3 DG Cooling H2O Pump As An Alternate M/U Source During Design Basis Flood

EC 391519 Cumulative Effect of FM on Dresden

U3 Reactor Vessel and

Connected Systems

- D3R22 00 EC 391643 Alternate

IC, RPV and SFP M

/U H2O Source 09 EC 396217 D2R23 - Cumulative Effect of

FM on the Reactor Vessel and Connected Systems

GEH Proprietary

0000-0130-8389-R1 ICF Task T0305:

RPV Flow Induced Vibration

GEH Proprietary

0000-0140-6826-R1 ICF Task T0318:

Piping Flow Induced Vibration

ASSESSMENTS

Number Description or Title

Date or Revision

AR 02419476 Unauthorized TCC Installed for Discharge Canal Temperature Recorder

January 20, 2014 AR 02469215-04 Inadequate 50.59 Review

U2/U3 ASD Modification

April 15, 2015 10 CFR 50.59 EVALUATIONS

Number Description or Title

Date or Revision

2012-02-001 Temporary N

Inerting Gas Supply

2012-06-001 TAT TR32 Open Phase Detection Protective Relay Circuit Installation

, Revision 0 October 17, 2012 2012-11-001 Cumulative Effects of

FM on Dresden

U3 Reactor Vessel and Connected Systems

- D3R22 00 2013-09-001 Dresden Increased Core Flow Implementation

, Revision 0 October 8, 2013 2013-10-001 Temporarily Bypass

U2 RWCU Trips

2013-11-001 Disable Fuel Pool Cooling Pump Trips, Revision 0 November 8, 2013 2013-11-003 D2R23 - Cumulative Effects of

FM on Dresden

U2 Reactor Vessel and Connected Systems

- D3R22 00 10 CFR 50.59 SCREENINGS

Number Description or Title

Date or Revision

2012-0007 DG RM Air Temp High

10 CFR 50.59 SCREENINGS

Number Description or Title

Date or Revision

2012-0039 Repair Air U3 SBO Fuel Day Tank Room Exhaust

Fan 3-5790-6007B, Revision

February 27, 2012 2012-0967 Floods 00 2012-1002 Engine Start Air Pressure Low or Locked Out or Air Valve Closed or Not Full

2013-0025 Alternate IC, RPV

& Spent Fuel M

/U H2O Source 00 2013-0028 Automatic Operation of the IC

2013-0029 Secure IC on

Hi/Low Level

2013-0030 Manual Operation of the IC

2013-0032 IC M/U Pumps Local Operation

2013-0144 U2 RAT TR 22 Sudden Pressure Relay

Out of 3 Trip Logic, Revision

July 11, 2013 2014-0095 HPCI System Standby Operation

2014-0168 SFP Instrumentation

- Fukushima, Revision

April 8, 2015 2014-0181 Replace Relay CR

-102 Panel

253-11, Revision

October 31, 2014 2014-0205 Inspect Discharge Check Valve 2

-2099-970B 00 2015-0042 Issue Calculation Analyzing TOL Relays in a Degraded Voltage

Condition, Revision

February 24, 2015 CALCULATIONS

Number Description or Title

Date or Revision

0101-0072-01 Dresden IC Heat Transfer Calculation

9389-46-19-2 Calculation for DG 2 Loading Under Design Bases Accident Condition, Revision

April 4, 2007 10553-CALC-07 Dresden EDGs Endurance Calculations

, Revision 2 May 25, 2011 DRE07-0005 Determination of Connected Loading on 120/240Vac ES Bus Dist. Panel 2-49 Powered from UPS at Panel 2-63, Revision

February 23, 2015 DRE13-001 Validation of TOL Relay Sizes Subject to a Degraded Voltage Condition, Revision

February 24, 2015 EC 388351 Temporary N

Inerting Gas Supply

CORRECTIVE ACTION PROGRAM DOCUMENTS

(ARs) ISSUED DURING INSPECTION

Number Description or Title

Date or Revision

2493498 PORC Approved 10 CFR 50.59 Evaluation Cannot Be Located

April 30, 2015 02499058 NRC INSP:

Parts Evaluation Errors

May 11, 2015 02499647 NRC INSP:

Typo in 50.59 Screening 2012

-0007 & DAN 923-74-045 May 12, 2015 02506445 NRC MOD/5059 Inspection:

IC Operating Procedures

May 28, 2015 02506845 NRC MOD/5059 Inspection:

Observation Improvement to DOS

20-01 May 28, 2015

CORRECTIVE ACTION PROGRAM DOCUMENTS

(ARs) ISSUED DURING INSPECTION

Number Description or Title

Date or Revision

2506869 NRC MOD/5059 Inspection:

EDG Fuel Consumption

May 28, 2015 CORRECTIVE ACTION PROGRAM DOCUMENTS

(ARs) REVIEWED Number Description or Title

Date or Revision

00191696 Uncovering of the IC

Tubes December 18, 2003 00210558 Additional Proof Needed to Conclude Cold M/U Splash March 24, 2004 00919110 NRC Concern

- EDG Fuel Oil Storage Tank Alarm Setpoint Margin

May 13, 2009 01405505 1B CCCT Supply Pump Found Tripped

August 27, 2012 01456015 Procedure Enhancement Regarding IC Level

December 27, 2012 01533278 C/O Requires a 50.59 Screening

July 13, 2013 02386093 Mod/50.59 2014 FASA

- EDG Fuel Consumption

September

25, 2014 02388710 Dresden Susceptible to Similar NRC Violation Issued to Fermi

September

30, 2014 02420155 C/O 118604 Has Been Placed for >90

Days December 3, 2014 02469215 Inadequate 50.59 Eval for ASD Modifications

March 16, 2015 DRAWINGS Number Description or Title

Date or Revision

2E-2302A Key Diagram 4160

/480V SWGRs & 480V MCCs Y 12E-2304 Key Diagram 4160V SWGRs 23-1 and 24-1 W 12E-2306 Key Diagram

-Reactor Building 480V SWGR 28 & 29 AE 12E-2328 Single Line Diagram Emergency Power System

O 12E-2509, Sheet 1 Schematic Diagram Primary Containment Isolation System Clean

-Up System Isolation Logic

AY & AZ M-22 Diagram of SW Piping EO M-27 Diagram of Core Spray Piping

AAN M-30, Sheet

Diagram of RWCU System AAP M-355 Diagram of SW Piping SI M-375 Diagram of Fire Protection Piping

M M-517 DG Engine Cooling

H2O System I MODIFICATIONS

Number Description or Title

Date or Revision

EC 347256 Replacement of Solenoid Valves 2(3)-1601-58, -59, -61, -62 with EQ Qualified Solenoid Valves

April 3, 2012 EC 380369 Revise Setpoints for EDG Fuel Oil Storage Tanks Level Switches

EC 383034 IC Valves Cable Reroutes

MODIFICATIONS

Number Description or Title

Date or Revision

EC 388981 Modify Supports on CREV RCU 2/3

-9400-102 Skid/Frame to Facilitate Replacement of Valves

EC 390811 Rewire MOV 3

-3901 Circuitry to Support MSO Project, Revision

November 14, 2012 EC 391291 Rewire MOV 2

-1501-22B Circuitry to Support MSO Project, Revision

November 30, 2012 EC 398999 SFP Instrumentation

- Fukushima

IEE 81788 IEE for US Electric Motor, CAT.

ID 1414246-4 September

2, 2013 IEE 81789 IEE for US Electric Motor, CAT. ID

1452042-4 March 26, 2012 OTHER DOCUMENTS

Number Description or Title

Date or Revision

EC 0000346716

Review and Approve EQ ASCO Solenoid Valve Model NP8316A54E to Replace Commercial Non-EQ Solenoid Valve HB8316D14 for 2(3)-1601-58, 59, 61, & 62.

February 5, 2004 EC 0000347323

Review and Approve Seismic Qualification of Solenoid Valve Model NP8316A54E to Replace Solenoid Valve HB8316D14 for 2(3)

-1601-58, 59, 61, & 62. EC 346716 Addressed EQ Portion Only

February 13, 2004 GEH-OLNC-0000-0104-3152-02-R0 On-Line NobleChemTM (OLNC) Application Technical Safety Evaluation for Dresden

U2 00 NEDC-33635P On-Line NobleChemTM (OLNC) Application Technical Safety Evaluation for Dresden

U3 01 WO 99016611 D3 6Y PM Replace SR/EQ Solenoid on N

M/U VLV 1601-59 November 21, 2006 PROCEDURES

Number Description or Title

Date or Revision

CC-AA-112 Temporary Configuration

Changes 22 CC-AA-304 Component Classification

CC-AA-309-1002 Key Calculation Identification and Improvement

DAN 902(3)-3D-4 IC Level Hi/Low Annunciator Respond Procedure

DAN DG2(3)(2/3)AC

-2 Engine Start Air Press Low or Locked Out or Air Valve Closed or Not Full Open Annunciator Response Procedure

DES 0040-08 ASCO Solenoid Valve Surveillance/Replacement

DGA-12 Partial or Complete Loss of AC Power

DIS 6600-01 DG Starting Air Press Instrumentation Calibration

DOA 0010-04 Floods 34 DOP 1300-02 Automatic Operation of IC

& 26 DOP 1300-03 Manual Operation of IC

& 35 DOP 1300-09 IC M/U Pump Local Operation

DOP 2300-01 HPCI System Standby Operation

PROCEDURES

Number Description or Title

Date or Revision

DOS 0500-05 Calculation of Core Thermal Power

DOS 6600-01 DG Surveillance Tests

28 DOS 6600-14 Diesel Oil Transfer Pump Operation and Fuel Consumption Test

EP-AA-1004 Addendum 3 Dresden EAL Tables

OP-AA-109-101-1002 Clearance

and Tagging Quarterly Audit

SM-AA-300 Procurement Engineering Support Activities

SM-AA-300-1001 Procurement Engineering Process and Responsibilities

REFERENCES

Number Description or Title

Date or Revision

NRR Letter to Commonwealth Edison Company

Safety Evaluation By NRR Related to Amendment

140 to FOL

No. DPR-19 September

21, 1995

LIST OF ACRONYMS USE

D ADAMS Agencywide Documents Access and Management System

CALC Calculation

CAP Corrective Action Program

CFR Code of Federal Regulations

CNO Chief Nuclear Officer

DRS Division of Reactor Safety

EDG Emergency Diesel

Generator

HPCI High Pressure Coolant Injection

Hz Hertz IC Isolation Condenser

IMC Inspection Manual Chapter

IP Inspection Procedure

ISCO Isolation Condenser

LLC Limited Liability Company

MBtu Million British Thermal Units

NCV Non-Cited Violation

NEI Nuclear Energy Institute

NRC U.S. Nuclear Regulatory Commission

NRR Office of Nuclear Reactor Regulation

NUREG NRC Technical Report Designation

PARS Public Available Records

RG Regulatory Guide

SR Surveillance Requirement

SSC Structures, Systems and Components TS Technical Specifications

UFSAR Updated Final Safety Analysis Report

B. Hanson -2- 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 NRC's 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/

Dariusz Szwarc, Acting Chief Engineering Branch

Division of Reactor Safety

Docket Nos. 50

-237, 50-249;72-037 License Nos. DPR

-19; DPR-25 Enclosure:

Inspection Report

05000237/2015007;

05000249/2015007 cc w/encl:

Distribution via LISTSERV DISTRIBUTION

w/encl: Kimyata MorganButler

RidsNrrDorlLpl3

-2 Resource

RidsNrrPMDresden Resource

RidsNrrDirsIrib Resource

Cynthia Pederson

Darrell Roberts

Richard Skokowski

Allan Barker Carole Ariano

Linda Linn

DRPIII DRSIII Jim Clay Carmen Olteanu

ROPreports.Resource@nrc.gov

ADAMS Accession Number ML15183A063

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To receive a copy of this document, indicate in the concurrence box "C" = Copy without attach/encl "E" = Copy with attach/encl "N" = No copy

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