Text

Byron Station, Units 1 and 2 - Evaluations of Changes, Tests, and Experiments Baseline Inspection Report 05000454/2017009; 05000455/2017009
	ML17180A534
Person / Time
Site:	Byron
Issue date:	06/29/2017
From:	Daley R C; 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 2-EVALUATIONS 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 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 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 2-EVALUATIONS OF CHANGES, TESTS, AND EXPERIMENTS BASELINE INSPECTION REPORT 05000454/2017009;

05000455/2017009

DISTRIBUTION: Jeremy Bowen

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RidsNrrPMByron Resource

RidsNrrDirsIrib Resource

Cynthia Pederson

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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

Enclosure 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

2 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 NRC's Enforcement Policy dated

November 1, 2016. The NRC's 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 licensee's 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)

3 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 licensee's 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.

4 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."

5 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 intervals-However, 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 operable-An 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 licensee's 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.

6 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.

7 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 pr

ior 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 licensee's 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

Attachment 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; 05000455/2017009-01 NCV Failure to Perform 10 CFR 50.59 Evaluation for UFSAR Change (Section 1R17.1b)

2 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 Revision6G-14-003 Implementation of WCAP-15063-P-A, 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"

0 6G-14-004 Establishing a Nitrogen Blanket on the Volume Control Tank (VCT)

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

Use in Reactor Coolant Pump Seal

Configurations

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

Blank Off TB Air Intake

0 6G-16-006 Reroute AF Diesel Pump Combustion Air Intake to 364' General Area Unit 1 & 2

0 6G-16-007 Temporarily Defeat FW Water Hammer Prevention System (WHPS) FW Isolation

Signals During Normal Power Operation For

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

(EC 406958)

0 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 Revision6D-15-002 UFSAR Update to Reflect Current LOCA Design Inputs for Unit 2

1 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 Reactive Stuck Rod Worths to Use in the

Event of Untrippable RCCAs

0 6E-14-017 Revise Diesel Generator and Integrated Safeguards LOOP/ESF Surveillance Test

0

3 10 CFR 50.59 SCREENINGS

Number Description or Title Date or RevisionSurveillance Frequency from 18 Months to 18

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

Cubicle Cooler

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

Propagation

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

Condensate Booster Pumps Lube Oil

Pressure Switches (EC 398037)

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

Condensate Booster Pumps Lube Oil

Pressure Switches (EC 398036)

0 6E-14-052 SSPS Wiring Changes Needed to Address Westinghouse Technical Bulletin TB-13-7 Solid State Protection System New Design Universal Logic Board and Safeguards Driver

Board 48 Vdc Input (EC 397531)

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

026087) 0 6E-15-035 Increase Pressurizer PORV Accumulator Tank Operating Pressure to Increase Margin

for PORV Operation (Unit 1)

0 6E-16-001 UFSAR Update of the Diesel-Generator Fuel Oil Storage and Transfer System Description

0 6E-16-061 Technical Requirements Manual (TRM) Technical Surveillance (PR No.16-009)

0 6E-16-097 Appendix J Scope Reduction 0

CALCULATIONS

Number Description or Title Date or RevisionBY-13-003 Diesel Generator and Integrated Safeguards LOOP/ESF Surveillance Test Interval

Evaluation

0 BYR 10-053 / BRW-10-0033-1 Calculation Feedwater Pressure Uncertainty for Input to LEFM CheckPlus System

0 BYR10-054 ER-800 Bounding Uncertainty Analysis for Thermal Power Determination at Byron Unit 1 Using the LEFM CheckPlus System

3 BYR10-055 ER-801 Bounding Uncertainty Analysis for Thermal Power Determination at Byron Unit 2

Using the LEFM CheckPlus System

2

4 CORRECTIVE ACTION PROGRAM DOCUMENTS INITIATED DURING INSPECTION

Number Description or Title Date or Revision4002071 50.59 Screening Form for BOP DO-16 Missing 04/24/2017 4017182 NRC Question on 50.59 Screening for DRP 15-073 06/01/2017

CORRECTIVE ACTION PROGRAM DOCUMENTS REVIEWED

Number Description or Title Date or Revision1500984 2B DG Shut Down Earlier than Desired During Sequence Test

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

not Met 04/12/2013 1673329 Concern About Testing Performed For HELB Dampers 06/19/2014 2467656 Issues Identified in Engineering Evaluation 03/12/2015 2496142 CDBI - 50.59 And DRP Did not Explicitly Evaluate GDC 5

05/05/2015 2727378 MCC 234V4 not Energized Subsequent U-2 Rx Trip 10/12/2016 3978965 2017 50.59 FASA Identifies 50.59 Requires

Revision

02/27/2017

DRAWINGS

Number Description or Title Date or RevisionM-129, Sheet 1C Diagram of Containment Spray, Unit 2 09/01/2000 M-138, Sheet 2 Diagram of Chemical & Volume Control &

Boron Thermal Regen, Unit 2

09/22/2000 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 &

Boron Thermal Regen, Unit 1

03/05/1998 M-64, Sheet 5 Diagram of Chemical & Volume Control &

Boron Thermal Regen, Unit 1

02/24/1999

ENGINEERING CHANGES

Number Description or Title Date or Revision397531 SSPS Wiring Changes Needed to Address Westinghouse Technical Bulletin TB-13-7

Solid State Protection System Logic Board

and Safeguards Driver Board 48 Vdc Input

0 398036 Modify Logic for Unit 1 Condensate and

Condensate Booster Pumps Lube Oil

Pressure Switches

1

5 ENGINEERING CHANGES

Number Description or Title Date or Revision398037 Modify Logic for Unit 2 Condensate and

Condensate Booster Pump Lube Oil

Pressure Switches

1 406958 Temporarily Defeat FW Water Hammer Prevention System (WHPS) FW Isolation Signals During Normal Power Operation for

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

0 617642 Temporarily Defeat FW Water Hammer Prevention System (WHPS) FW Isolation Signals During Normal Power Operation for Steam Generators 2A/2B/2C/2D (Braidwood)

01/09/2017

ML17180A534

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