ML17180A534
| 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) | |
See also: IR 05000454/2017009
Text
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 RidsNrrDorlLpl3 RidsNrrPMByron Resource
RidsNrrDirsIrib Resource
Cynthia Pederson
Darrell Roberts Richard Skokowski Allan Barker
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
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 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 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 OTHER DOCUMENTS Number Description or Title Date or RevisionLER 455-2016-001 Manual Reactor Trip due to Circuit Breaker Failure that Caused Actuation of Feedwater
Hammer Prevention System with Automatic
Isolation of Feedwater to Two Steam
Generators and Low Steam Generator Levels 10/12/2016 PR No. 16 009 Revision to Technical Requirements Manual (TRM) Technical Surveillance Requirement
TSR 3.3.k.2 LEFM Channel Calibration 07/12/2016 TB-13-7 Westinghouse Technical Bulletin Solid State Protection New Design Universal Logic
Board and Safeguards Driver Board 48 Vdc Input 12/10/2013 PROCEDURES Number Description or Title Date or Revision1BOSR 4.11.3-1 Unit One Pressurizer PORV Accumulator Pressure Decay Test 7 2BGP 100-1 Plant Heatup 58 2BGP 100-3 Power Ascension 93
2BGP 100-4 Power Descension 50
2BGP 100-4T4 Reactor Trip Post Response Guideline 7 BAR 1-12-D7 PZR PORV SUP PRESS HIGH LOW 7 ER-AA-425 Implementation of the Technical Specification Surveillance Frequency Control Program 1 ER-AA-425-1005 Monitoring the Effects of Changes to the Surveillance Frequency Control Program (SFCP) 1 LS-AA-104 Exelon 50.59 Review Process 10 OP-MW-201-007 Fire Protection System Impairment Control 7 RP-BY-301-1001 Radiological Air Sampling Program 13
6 LIST OF ACRONYMS USED CFR Code of Federal Regulations IEEE Institute of Electrical & Electronics Engineers LOOP Loss of Offsite Power NEI Nuclear Energy Institute
NRC U.S. Nuclear Regulatory Commission
RG Regulatory Guide
SDP Significance Determination Process
SFCP Surveillance Frequency Control Program SSC Structure, System, or Component STI Surveillance Test Interval