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Evaluations of Changes, Tests, and Experiments Baseline Inspection Report 05000454/2017009; 05000455/2017009
	ML17180A534
Person / Time
Site:	Byron
Issue date:	06/29/2017
From:	Robert Daley; Engineering Branch 3
To:	Bryan Hanson; Exelon Generation Co, Exelon Nuclear
References
	IR 2017009
	Download: ML17180A534 (16)
	v • d • e

UNITED STATES

NUCLEAR REGULATORY COMMISSION

REGION III

2443 WARRENVILLE RD. SUITE 210

LISLE, IL 60532-4352

June 29, 2017

Mr. Bryan C. Hanson

Senior VP, Exelon Generation Company, LLC

President and CNO, Exelon Nuclear

4300 Winfield Road

Warrenville, IL 60555

SUBJECT: BYRON STATION, UNITS 1 AND 2EVALUATIONS OF CHANGES, TESTS, AND

EXPERIMENTS BASELINE INSPECTION REPORT 05000454/2017009;

05000455/2017009

Dear Mr. Hanson:

On May 19, 2017, the U.S. Nuclear Regulatory Commission (NRC) completed an Evaluations

of Changes, Tests, and Experiments inspection at your Byron Station. The enclosed inspection

report documents the inspection results which were discussed on June 1, 2017, with

Mr. T. Chalmers 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.

NRC inspectors documented one traditional enforcement Severity Level IV violation in this

report. This traditional enforcement violation was identified with an associated finding.

However, because the issue was a Severity Level IV violation and was entered into your

corrective action program, the NRC is treating the issue as a Non-Cited Violation in accordance

with Section 2.3.2 of the NRC Enforcement Policy.

If you contest the violation or significance of the Non-Cited Violation, 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 Byron Station.

B. Hanson -2-

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/

Robert C. Daley, Chief

Engineering Branch 3

Division of Reactor Safety

Docket Nos. 50-454; 50-455

License Nos. NPF-37; NPF-66

Enclosure:

IR 05000454/2017009; 05000455/2017009

cc: Distribution via LISTSERV

B. Hanson -3-

Letter to Bryan Hanson from Robert C. Daley dated June 29, 2017

SUBJECT: BYRON STATION, UNITS 1 AND 2EVALUATIONS OF CHANGES, TESTS, AND

EXPERIMENTS BASELINE INSPECTION REPORT 05000454/2017009;

05000455/2017009

DISTRIBUTION:

Jeremy Bowen

RidsNrrDorlLpl3

RidsNrrPMByron Resource

RidsNrrDirsIrib Resource

Cynthia Pederson

DRPIII

DRSIII

ROPreports.Resource@nrc.gov

ADAMS Accession Number: ML17180A534

Publicly Available Non-Publicly Available Sensitive Non-Sensitive

OFFICE RIII RIII

NAME DSzwarc:vv RDaley

DATE 06/29/2017 06/29/2017

OFFICIAL RECORD COPY

U. S. NUCLEAR REGULATORY COMMISSION

REGION III

Docket No: 50-454; 50-455

License No: NPF-37; NPF-66

Report No: 05000454/2017009; 05000455/2017009

Licensee: Exelon Generation Company, LLC

Facility: Byron Station, Units 1 and 2

Location: Byron, IL

Dates: May 15 through June 1, 2017

Inspectors: G. Hausman, Senior Reactor Inspector

A. Shaikh, Senior Reactor Inspector

D. Szwarc, Senior Reactor Inspector (Lead)

Approved by: R. Daley, Chief

Engineering Branch 3

Division of Reactor Safety

Enclosure

SUMMARY

Inspection Report 05000454/2017009, 05000455/2017009; 05/15/2017 - 06/01/2017; Byron

Station, Units 1 and 2; Evaluations of Changes, Tests, and Experiments.

This report covers a one-week announced Evaluations of Changes, Tests, and Experiments

baseline inspection. The inspection was conducted by Region III based engineering inspectors.

One violation was identified by the inspectors. The violation, and its associated finding, was

considered a Non-Cited Violation 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 0609, Significance Determination

Process. Findings for which the Significance Determination Process 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 NRCs Enforcement Policy dated

November 1, 2016. The NRCs program for overseeing the safe operation of commercial

nuclear power reactors is described in NUREG-1649, Reactor Oversight Process, Revision 6,

dated July 2016.

Cornerstone: Mitigating Systems

Severity Level IV. The inspectors identified a Severity Level IV, Non-Cited Violation of

10 CFR 50.59, Changes, Tests, and Experiments, Section(d)(1) and an associated

finding of very low safety significance (Green) for the licensees failure to provide a

written evaluation which provided the basis for the determination that a change did not

require a license amendment. Specifically, the licensee failed to provide a basis for why

a change to the surveillance frequencies of emergency diesel generators described in

the Updated Final Safety Analysis Report did not require prior NRC approval.

The inspectors determined that the performance deficiency was more than minor

because the inspectors could not reasonably determine that the changes would not have

ultimately required NRC prior approval. The associated finding screened to Green (very

low safety significance) because it did not result in the loss of operability or functionality.

The diesel generators passed their most recent surveillances. As a result the violation is

categorized as Severity Level IV in accordance with section 6.1.d of the NRC

Enforcement Policy. The issue did not have a cross-cutting aspect because it was not

reflective of current performance. (Section 1R17.1b)

2

REPORT DETAILS

1. REACTOR SAFETY

Cornerstone: Initiating Events, Mitigating Systems, and Barrier Integrity

1R17 Evaluations of Changes, Tests, and Experiments (71111.17T)

.1 Evaluation of Changes, Tests, and Experiments

a. Inspection Scope

The inspectors reviewed 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 screenings and applicability determinations

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 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 23 samples of evaluations, screenings, and/or applicability

determinations as defined in IP 71111.17-05.

b. Findings

Failure to Perform 10 CFR 50.59 Evaluation for Updated Final Safety Analysis Report

Change

Introduction: The inspectors identified a Severity Level IV, Non-Cited Violation of 10

CFR 50.59, Changes, Tests, and Experiments, Section(d)(1) and an associated finding

of very low safety significance (Green) for the licensees failure to provide a written

evaluation which provided the basis for the determination that a change did not require a

license amendment. Specifically, the licensee failed to provide a basis for why a change

to the surveillance frequencies of emergency diesel generators described in the Updated

Final Safety Analysis Report (UFSAR) did not require prior NRC approval.

3

Description: The licensee relocated numerous surveillance frequencies from Technical

Specifications to the licensee-controlled Surveillance Frequency Control Program

(SFCP) since 2010. The relocated surveillance frequencies included emergency diesel

generator surveillances required to be performed during every refueling outage as

specified in Regulatory Guide 1.9, Application and Testing of Safety-Related Diesel

Generators in Nuclear Power Plants, Revision 3.

Appendix A, Application of NRC Regulatory Guides, of the UFSAR listed the applicable

NRC regulatory guides that the licensee complies with. The section covering Regulatory

Guide 1.9 stated that:

Regulatory Guide (RG) 1.9, Revision 3, endorses IEEE [Institute of Electrical and

Electronics Engineers] Standard 387-1984, IEEE Standard Criteria for Diesel

Generator Units Applied as Standby Power Supplies for Nuclear Power

Generating Stations. In addition to this standard, RG 1.9, Revision 3, provides

supplemental regulatory positions. The Licensee complies with these

supplemental regulatory positions in Revision 3 with the following clarifications

regarding:

The licensee changed the last sentence to state:

The Licensee complies with IEEE Standard 387-1984 and these supplemental

regulatory positions in Revision 3 with the following clarifications regarding:

The licensee further added clarification number 8, Regulatory Position C.2.3.2.3,

Refueling Outage Testing stating:

Exception is taken to the statement that the overall emergency diesel generator

unit design capability should be demonstrated at every refueling outage by

performing the tests identified in Table 1 of Regulatory Guide 1.9. Refueling

Outage Testing as identified in Table 1 of Regulatory Guide 1.9 is performed in

accordance with the Technical Specifications, and the test interval may be

supplanted with performance-based, risk-informed test intervals. This statement

in Regulatory Position C.2.3.2.3 is in accordance with Section 6.5.2 of IEEE

Standard 387-1984. By taking exception to Regulatory Position C.2.3.2.3,

exception is also being taken to the statement in Section 6.5.2 of IEEE Standard

387-1984 that the diesel generator unit shall be given one cycle of each of the

specified tests at least once every 18 months to demonstrate its continued

capability of performing its required function.

The licensee documented the UFSAR change in 50.59 Screening 6E-14-017, Revise

Diesel Generator and Integrated Safeguards LOOP [loss of offsite power] / ESF

[engineered safety feature] Surveillance Test Surveillance Frequency from 18 Months to

18 Months Staggered, Revision 0. In the screening, the licensee stated that the,

Proposed change does not have any effect on any SSC [Structure, System, or

Component] described in the UFSAR. They further stated that, The components that

are not directly tested within an eighteen month frequency are justified to not be affected

by the proposed change as evaluated in the evaluation required by NEI 04-10,

Revision 1. The licensee also stated that, The SFCP evaluation per NEI 04-10

ensures the reliability of the SSC to perform its intended design function is not

decreased.

4

The licensee performed evaluation BY-13-003, Diesel Generator and Integrated

Safeguards LOOP ESF Surveillance Test Surveillance Frequency Surveillance Test

Interval (STI) Evaluation, dated March 3, 2014 under the SFCP. This evaluation

provided the basis for extending the surveillance frequencies under the SFCP. In

section C.7 of the evaluation the licensee concluded that:

The Surveillance Frequency Control Program, as approved by the NRC for Byron

and for other nuclear power plants, supplants prescriptive test intervals, such

as those specified in the above standards and guides, with performance-based,

risk-informed test intervalsHowever, a UFSAR change is required to reflect the

fact that the frequency of the Class 1E Diesel Generator and Integrated

Safeguards Test is per the SFCP, and that the station takes exception to the

frequency specified in RG 1.9. Approval of this UFSAR Change Request is

required prior to implementation of this Surveillance Test Interval extension.

The SFCP evaluation does not contain the same questions as a 50.59 evaluation and

it therefore does not replace a 50.59 evaluation in evaluating the applicable criteria of

10 CFR 50.59(c)(2). In the 50.59 screening the licensee referred to the SFCP

evaluation as providing the basis for why the change did not impact the reliability of the

diesel generators. The SFCP evaluation stated that a UFSAR change would need to be

processed to change the references to Regulatory Guide 1.9 and IEEE-384. This

created a circular logic.

The Nuclear Energy Institute stated in guidance document NEI 04-10, Risk-Informed

Technical Specifications Initiative 5b Risk-Informed Method for Control Surveillance

Frequencies, Revision 1 Step 3 in section 3.0 that, the safety analysis acceptance

criteria in the plant licensing basis (e.g., FSAR, supporting analyses) will continue to be

met with the proposed changes to Surveillance Frequencies. Further, step 7 in section

4.0 states, in part, to, Document that assumptions in the plant licensing basis would not

be invalidated when performing the surveillance at the bounding interval limit for the

proposed STI change.

Even though the SFCP allows licensees to change their Technical Specification

surveillance frequencies, licensees are still required to process UFSAR changes per the

50.59 process. Therefore, completion of the STI evaluation does not preclude the need

to perform a 50.59 review. The increase in the surveillance frequency does have an

impact on the reliability of the diesel generators, and as such it is considered to be

adverse. The licensee stated in section C.8 of the SFCP evaluation that, Surveillances

are primarily performed to demonstrate that equipment is operableAn extended

surveillance interval could lead to less conditioning and component degradation, and

some failure mechanisms could become more prominent and increase equipment failure

probabilities. The licensee should have performed a 10 CFR 50.59 evaluation to

determine if the change would have resulted in a more than minimal increase in the

likelihood of occurrence of a malfunction of an SSC important to safety.

Analysis: The inspectors determined that the licensees failure to provide a written

evaluation which provided the basis for the determination that a change did not require a

license amendment was contrary to 10 CFR 50.59(d)(1) and was a performance

deficiency. Specifically, the licensee failed to provide a basis for why a change to the

surveillance frequencies of emergency diesel generators described in the UFSAR did

not require prior NRC approval.

5

The inspectors determined that the performance deficiency was more than minor

because the finding was associated with the Mitigating Systems cornerstone attribute of

Equipment Performance and affected the cornerstone objective of ensuring the

availability, reliability, and capability of systems that respond to initiating events to

prevent undesirable consequences (i.e., core damage). Specifically, by extending the

surveillance frequency of the emergency diesel generators the licensee potentially

affected the reliability of the diesel generators because certain components of the diesel

generators could be affected due to less conditioning.

In addition, the associated violation was determined to be more than minor because the

inspectors could not reasonably determine that the changes would not have ultimately

required NRC prior approval.

Violations of 10 CFR 50.59 are dispositioned using the traditional enforcement process

instead of the Significance Determination Process (SDP) because they are considered

to be violations that potentially impede or impact the regulatory process. This violation is

associated with a finding that has been evaluated by the SDP and communicated with

an SDP color reflective of the safety impact of the deficient licensee performance. The

SDP, however, does not specifically consider the regulatory process impact. Thus,

although related to a common regulatory concern, it is necessary to address the violation

and finding using different processes to correctly reflect both the regulatory importance

of the violation and the safety significance of the associated finding.

In this case, the inspectors determined the finding could be evaluated using the SDP in

accordance with Inspection Manual Chapter 0609, Significance Determination Process.

Using Attachment 0609.04, Initial Characterization of Findings, Table 2 the inspectors

determined that the finding affected the Mitigating Systems cornerstone. As a result, the

inspectors evaluated the finding using Appendix A, The Significance Determination

Process (SDP) for Findings At-Power, Exhibit 2 for the Mitigating Systems cornerstone.

The finding screened to Green (very low safety significance) because it did not result in

the loss of operability or functionality. The diesel generators passed their most recent

surveillances.

In accordance with section 6.1.d of the NRC Enforcement Policy this violation is

categorized as Severity Level IV because the resulting changes were evaluated by the

SDP as having very low safety significance (i.e., green finding).

The inspectors did not identify a cross-cutting aspect associated with the finding

because the finding was not representative of current performance. The licensee

performed the screening over three years prior to the start of the inspection.

Enforcement: Title 10 CFR Part 50.59, Changes, Tests, and Experiments, section

(d)(1) requires the licensee to maintain records of changes in the facility, of changes in

procedures, and of tests and experiments made pursuant to 10 CFR 50.59(c). Title 10 CFR 50.59(d)(1) requires that these records include a written evaluation which provides

the basis for the determination that a change, test, or experiment did not require a

license amendment. Title 10 CFR 50.59(c)(2) requires a licensee to obtain a license

amendment prior to implementing a proposed change, test, or experiment if the change,

test, or experiment would result in more than a minimal increase in the likelihood of

occurrence of a malfunction of an SSC important to safety.

6

Contrary to the above, between February 14, 2014 and June 1, 2017, the licensee failed

to provide a written evaluation which provided the basis for determining that a change,

test, or experiment made pursuant to 10 CFR 50.59(c) did not require a license

amendment. Specifically, the licensee failed to provide a basis for why a change to the

surveillance frequencies of emergency diesel generators described in the Updated Final

Safety Analysis Report did not require prior NRC approval. The licensee did not provide

a basis for why the change would not result in more than a minimal increase in the

likelihood of occurrence of a malfunction of an SSC important to safety.

This violation is being treated as an Non-Cited Violation, consistent with Section 2.3.2 of

the Enforcement Policy because it was a Severity Level IV violation and was entered

into the licensees corrective action program as Action Request 04017182, NRC

Question on 50.59 Screening for DRP 15-073, dated June 1, 2017. The licensee

planned to work to disposition the issue. (NCV 050004542017009-01;

05000455/2017009-01, Failure to Perform 10 CFR 50.59 Evaluation for UFSAR

Change).

4. OTHER ACTIVITIES (OA)

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. The inspectors reviewed these documents to

evaluate the effectiveness of corrective actions related to 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 of significance were identified.

4OA6 Meetings

.1 Exit Meeting Summary

On June 1, 2017, the inspectors presented the inspection results to Mr. T. Chalmers,

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

7

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee

J. Bauer, Corporate Licensing Engineer

T. Chalmers, Plant Manager

G. Contrady, Regulatory Assurance Engineer

Z. Cox, Regulatory Assurance

D. Gullott, Corporate Licensing Manager

C. Keller, Engineering Director

D. Spitzer, Regulatory Assurance Manager

G. Wilhelmsen, Senior Engineering Manager

K. Zlevor, Senior Engineer

L. Zurawski, Regulatory Assurance

U.S. Nuclear Regulatory Commission

R. Daley, Branch Chief, EB3

C. Hunt, Resident Inspector

J. McGhee, Senior Resident Inspector

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened and Closed

05000454/2017009-01; NCV Failure to Perform 10 CFR 50.59 Evaluation for UFSAR

05000455/2017009-01 Change (Section 1R17.1b)

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 Date or Revision

6G-14-003 Implementation of WCAP-15063-P-A, 0

Revision 1, with Errata, Westinghouse

Improved Performance Analysis and Design

Model (PAD 4.0), and WCAP-12610-P-A

and GENPD-404-P-A Addendum 2-A,

Westinghouse Clad Corrosion Model for

ZIRLO and Optimized ZIRLO

6G-14-004 Establishing a Nitrogen Blanket on the 0

Volume Control Tank (VCT)

6G-15-004 Implement Use of Westinghouse SHIELD for 0

Use in Reactor Coolant Pump Seal

Configurations

6G-16-001 Remove AF Diesel Air Intake Elbow and 0

Blank Off TB Air Intake

6G-16-006 Reroute AF Diesel Pump Combustion Air 0

Intake to 364 General Area Unit 1 & 2

6G-16-007 Temporarily Defeat FW Water Hammer 0

Prevention System (WHPS) FW Isolation

Signals During Normal Power Operation For

Steam Generators 2A/2B/2C/2D

(EC 406958)

6G-16-008 Appendix J Scope Reduction 0

6G-17-001 Lost SFP Crimps and RVLIS Pins 0

10 CFR 50.59 SCREENINGS

Number Description or Title Date or Revision

6D-15-002 UFSAR Update to Reflect Current LOCA 1

Design Inputs for Unit 2

6D-15-010 Alternate Main Control Room Ventilation 2

6D-15-016 LCOAR ESF Battery Room Ventilation 7

6D-15-025 Revision to BCB-2 Table 1-6 to Add Most 0

Reactive Stuck Rod Worths to Use in the

Event of Untrippable RCCAs

6E-14-017 Revise Diesel Generator and Integrated 0

Safeguards LOOP/ESF Surveillance Test

2

10 CFR 50.59 SCREENINGS

Number Description or Title Date or Revision

Surveillance Frequency from 18 Months to 18

Months Staggered

6E-14-020 Operation of 1A SX Pump With One-Half of a 0

Cubicle Cooler

6E-14-040 Drill Holes in Line 2CD31AD to Stop Crack 0

Propagation

6E-14-042 Modify Logic for Unit 2 Condensate and 0

Condensate Booster Pumps Lube Oil

Pressure Switches (EC 398037)

6E-14-046 Modify Logic for Unit 1 Condensate and 0

Condensate Booster Pumps Lube Oil

Pressure Switches (EC 398036)

6E-14-052 SSPS Wiring Changes Needed to Address 0

Westinghouse Technical Bulletin TB-13-7

Solid State Protection System New Design

Universal Logic Board and Safeguards Driver

Board 48 Vdc Input (EC 397531)

6E-14-060 Plant Barrier Impairment (Penetration 0

026087)

6E-15-035 Increase Pressurizer PORV Accumulator 0

Tank Operating Pressure to Increase Margin

for PORV Operation (Unit 1)

6E-16-001 UFSAR Update of the Diesel-Generator Fuel 0

Oil Storage and Transfer System Description

6E-16-061 Technical Requirements Manual (TRM) 0

Technical Surveillance (PR No.16-009)

6E-16-097 Appendix J Scope Reduction 0

CALCULATIONS

Number Description or Title Date or Revision

BY-13-003 Diesel Generator and Integrated Safeguards 0

LOOP/ESF Surveillance Test Interval

Evaluation

BYR 10-053 / Calculation Feedwater Pressure Uncertainty 0

BRW-10-0033-1 for Input to LEFM CheckPlus System

BYR10-054 ER-800 Bounding Uncertainty Analysis for 3

Thermal Power Determination at Byron Unit 1

Using the LEFM CheckPlus System

BYR10-055 ER-801 Bounding Uncertainty Analysis for 2

Thermal Power Determination at Byron Unit 2

Using the LEFM CheckPlus System

3

CORRECTIVE ACTION PROGRAM DOCUMENTS INITIATED DURING INSPECTION

Number Description or Title Date or Revision

4002071 50.59 Screening Form for BOP DO-16 04/24/2017

Missing

4017182 NRC Question on 50.59 Screening for DRP 06/01/2017

15-073

CORRECTIVE ACTION PROGRAM DOCUMENTS REVIEWED

Number Description or Title Date or Revision

1500984 2B DG Shut Down Earlier than Desired 04/12/2013

During Sequence Test

1500993 2B DG Sequence Times Acceptance Criteria 04/12/2013

not Met

1673329 Concern About Testing Performed For HELB 06/19/2014

Dampers

2467656 Issues Identified in Engineering Evaluation 03/12/2015

2496142 CDBI - 50.59 And DRP Did not Explicitly 05/05/2015

Evaluate GDC 5

2727378 MCC 234V4 not Energized Subsequent U-2 10/12/2016

Rx Trip

3978965 2017 50.59 FASA Identifies 50.59 Requires 02/27/2017

Revision

DRAWINGS

Number Description or Title Date or Revision

M-129, Sheet 1C Diagram of Containment Spray, Unit 2 09/01/2000

M-138, Sheet 2 Diagram of Chemical & Volume Control & 09/22/2000

Boron Thermal Regen, Unit 2

M-46, Sheet 1C Diagram of Containment Spray, Unit 1 01/08/1998

M-50, Sheet 1B Diagram of Diesel Fuel Oil AP

M-64, Sheet 2 Diagram of Chemical & Volume Control & 03/05/1998

Boron Thermal Regen, Unit 1

M-64, Sheet 5 Diagram of Chemical & Volume Control & 02/24/1999

Boron Thermal Regen, Unit 1

ENGINEERING CHANGES

Number Description or Title Date or Revision

397531 SSPS Wiring Changes Needed to Address 0

Westinghouse Technical Bulletin TB-13-7

Solid State Protection System Logic Board

and Safeguards Driver Board 48 Vdc Input

398036 Modify Logic for Unit 1 Condensate and 1

Condensate Booster Pumps Lube Oil

Pressure Switches

4

ENGINEERING CHANGES

Number Description or Title Date or Revision

398037 Modify Logic for Unit 2 Condensate and 1

Condensate Booster Pump Lube Oil

Pressure Switches

406958 Temporarily Defeat FW Water Hammer 0

Prevention System (WHPS) FW Isolation

Signals During Normal Power Operation for

Steam Generators 2A/2B/2C/2D

617642 Temporarily Defeat FW Water Hammer 01/09/2017

Prevention System (WHPS) FW Isolation

Signals During Normal Power Operation for

Steam Generators 2A/2B/2C/2D (Braidwood)

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