IR 05000373/2012007

IR 05000373-12-007 & 05000374-12-007, 09/04/2012 - 09/21/2012, LaSalle County Station, Units 1 and 2, Evaluations of Changes, Tests, or Experiments and Permanent Plant Modifications

ML12303A047
Person / Time
Site:	LaSalle
Issue date:	10/29/2012
From:	Robert Daley Engineering Branch 3
To:	Pacilio M Exelon Generation Co, Exelon Nuclear
Alan Dahbur
References
IR-12-007
Download: ML12303A047 (17)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION

REGION III

2443 WARRENVILLE ROAD, SUITE 210 LISLE, IL 60532-4352 October 29, 2012 Mr. Michael Senior Vice President, Exelon Generation Company, LLC President and Chief Nuclear Officer (CNO), Exelon Nuclear 4300 Winfield Road Warrenville , IL 60555 SUBJECT: LASALLE COUNTY STATION, UNITS 1 AND 2 EVALUATION S OF CHANGES, TESTS, OR EXPERIMENTS AND PERMANENT PLANT MODIFICATIONS BASELINE INSPECTION REPORT 05000373/20 12007; 05000374/20 1200 7

Dear Mr. Pacilio:

On September 21, 2012, the U. S. Nuclear Regulatory Commission (NRC) completed an Evaluations of Changes, Tests, or Experiments and Permanent Plant Modifications inspection at your LaSalle County Station. The enclosed inspection report documents the inspection results

, which were discussed on September 21, 2012, with Mr. P. Karaba 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.

The NRC-identified one traditional enforcement Severity Level IV violation. This traditional enforcement violation was identified with an associated finding. However, because of its ver y low safety significance and because the issue was entered into your Corrective Action Program , the NRC is treating the issue as a Non-Cited Violation (NCV) in accordance with Section 2.3.2 of the NRC Enforcement Policy

. If you contes t the subject or severity of the NCV 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 a copy to the Regional Administrator, U.S. Nuclear Regulatory Commission

- Region III, 2443 Warrenville Road, Suite 210, Lisle, IL 60532-4352; the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; and the Resident Inspector office at the LaSalle County Station. In addition, if you disagree with the cross-cutting aspect assigned to the 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 the LaSalle County Station. In accordance with 10 CFR 2.390 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 Public Document Room or from the Publicly Available Records System (PARS) component of NRC's Agencywide Documents Access and Management System (ADAMS), accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Sincerely,

/RA/ Robert C. Daley, Chie f Engineering Branch 3 Division of Reactor Safety Docket Nos. 50-373; 50-374 License Nos. NPF-11; NPF-18

Enclosure:

Inspection Report 05000373/2012007; 05000374/2012007 w/Attachment: Supplemental Information

REGION III==

Docket No:

05000373; 05000374 License No:

NPF-11; NPF-18 Report No:

05000373/2012007; 05000374/2012007 Licensee: Exelon Generation Company, LLC Facility: LaSalle County Station, Units 1 and 2 Location: Marseilles, IL Dates: September 4 through September 21, 2012 Inspectors:

Alan Dahbur, Senior Reactor Inspector (Lead)

Mohammad Munir, Reactor Inspector Vijay Meghani, Reactor Inspector Approved by:

Robert C. Daley, Chief Engineering Branch 3 Division of Reactor Safety 1 Enclosure

SUMMARY

IR 05000373/2012007; 05000374/2012007; 09/04/2012

- 09/21/2012

LaSalle County Station, Units 1 and 2; Evaluations of Changes, Tests, or Experiments and Permanent Plant Modifications.

This report covers a two-week announced baseline inspection on evaluations of changes, tests, or experiments and permanent plant modifications. The inspection was conducted by Region III based engineering inspectors.

Based on the results of this inspection, one Severity Level IV violation was identified by the inspectors. The findin g was considered a Non-Cited Violation (NCV) of NRC regulations. The significance of the majority of findings are indicated by their color (Green, White, Yellow, Red) using Inspection Manual Chapter (IMC) 0609, "Significance Determination Process" (SDP).

Cross-cutting aspects were determined using IMC 0310, "Components Within the Cross-Cutting Areas." Findings for which the SDP does not apply may be Green or may be assigned a severity level after NRC management review. The NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, "Reactor Oversight Process," Revision 4, dated December 2006.

A.

Cornerstone: Initiating Events

N RC-Identified and Self-Revealed Findings Severity Level IVThe inspectors determined that the violation was more than minor because the finding, if left uncorrected would become a more significant safety concern. Iin addition, the inspector could not reasonably determine that the changes would not have ultimately required NRC prior approval. The inspectors determined that the finding was of very low safety significance (Green) based on a review of the licensee's operability determination and corrective actions for non-conformance to the ASME code requirements issues identified since the deletion of the TRM section. The inspectors determined that the licensee's actions in the four instances did not have any technical safety concerns. This finding had a cross-cutting aspect in the area of Human Performance within the Decision Making component because the licensee did not use conservative assumptions to ensure the proposed activity was safe. Specifically, the licensee mad e an inadequate assumption when they determined that the removal of TRM, Section 3.4.a did not have an adverse effect. H.1(b). (Section 1R17.1.b)

The inspectors identified a Severity Level IV Non-Cited Violation and an associated finding of very low safety significance (Green) of 10 CFR 50.59, "Changes, Tests, and Experiments," Section (d)1 for the licensee's failure to perform a written safety evaluation to demonstrate that the deletion of the Technical Requirements Manual (TRM), Section 3.4.a did not require a license amendment. The licensee entered this issue into their Corrective Action Program and initiated a Standing Order reinstating the TRM Section 3.4.a. B. No violations of significance were identified.

Licensee-Identified Violations

1. REACTOR SAFETY

REPORT DETAILS

Cornerstone: Initiating Events, Mitigating Systems, and Barrier Integrity 1R17 Evaluation

s of Changes, Tests, or Experiments and Permanent Plant Modifications

.1

a. Evaluation of Changes, Tests, or Experiments From September 4, 20 12 through September 21, 2012, the inspectors reviewed seven safety evaluations performed pursuant to Title 10, Code of Federal Regulations (CFR) 50.59 to determine if the evaluations were adequate and that prior NRC

-approval was obtained as appropriate. The inspectors also reviewed seventeen screenings where licensee personnel had determined that a 10 CFR 50.59 evaluation was not necessary. The inspectors reviewed these documents to determine if:

Inspection Scope the changes, tests, or 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, or 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 seven samples of evaluations and seventeen samples of changes as defined in IP 71111.17-04. b.

.1 Findings Introduction

The inspectors identified a Severity Level IV, NCV of 10 CFR 50.59(d)(1), "Changes, Tests, and Experiments," and an associated finding of very low safety significance (Green

) for the licensee's failure to perform a safety evaluation to demonstrate that the removal of the TRM Section 3.4.a did not require a license amendment. Specifically, the licensee did not adequately address the adverse effects of the changes in their screening L09-240. This inadequate screening resulted in the adverse changes not being evaluated under 10 CFR 50.59 change evaluation criteria.Failure to Perform a Written Safety Evaluation for TRM Changes:

3 Enclosure Description

Technical Requirements Manual Section 3.4.a. addressed structural integrity requirements for the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code) Class 1, 2, and 3 components of the Reactor Coolant System (RCS) including surveillance requirements to verify structural integrity in accordance with the In

-Service Inspection/Testing Program. This section also prescribed specific actions with corresponding completion times if the structural integrity of an ASME Code Class 1, 2 or 3 component was found not to be in conformance with the requirements of the Code. The actions included either restoring the component within its limit or isolating it. These requirements were previously stated in Technical Specification (TS) Section 3/4.4.8, before it was relocated to the TRM as a part of the LaSalle Station's Technical Specifications conversion to the Improved Technical Specifications (ITS). The license amendment specified that the change control process for the new TRM requirements Section 3.4.a. was 10 CFR 50.59.

The licensee deleted Section 3.4.a and the corresponding section in the bases from the TRM in 2009 using the 10 CFR 50.59 process per screening L09

-240. The licensee concluded that the change did not have an adverse effect based on a justification that the TRM requirements were a duplicate to the requirements in 10 CFR 50.55a and were being implemented through the requirements in Technical Specification 5.5.7 for the In-Service Testing Program. Based on the justifications used in screening L09

-240 , the licensee screen ed out the change as not needing a full safety evaluation and consequently not needing NRC prior approval. The licensee

's screening also concluded that actions taken, following deletion of the TRM section, for a component found not meeting the structural integrity acceptance criteria was not change d. Specifically, if ASME Code requirements were not met, the impact to system operability and/or functionality would be assessed as a nonconforming condition, in accordance with the Corrective Action Program. The operability of the Structures, Systems, and Components (SSC) would be determined in accordance with the TS definition of operability for TS SSCs or functionality for non TS SSCs following the guidance in Operation Procedure (OP)-AAA-108-115, "Operability Determinations." The inspectors determined, based on a review of the licensee's procedures for corrective actions and operability determinations, that following these procedures would not result in the same actions as those described in the deleted section of the TRM. The inspectors were concerned that for component s found in non

-conformance, the TRM Section 3.4.a require d either restoring the structural integrity within its limits or isolating the components, whereas , the use of the operability determination process could allow the licensee to continue operations without isolating the non-conforming component. The inspectors were concerned that these actions were not covered under 10 CFR 50.55a

, and therefore , in some cases the deletion of the TRM actions could result in an adverse effect on the component reliability. Consequently, since the deletion of the TRM actions resulted in adverse effects, this change required a 50.59 safety evaluation.

Licensee's Procedure LS

-AA-104-1000, 50.59 Resource Manual required that if a change has both positive and adverse effects, the change should be screened in, thus requiring a 50.59 safety evaluation. These procedural requirements are consistent with the Nuclear Energy Institute (NEI) 96

-07, Revision 1, "Guidelines for 10 CFR 50.59 Implementation."

Based on the above, the inspectors determined that the licensee failed to appropriately evaluate the differences between the actions required per the deleted TRM Section 3.4.a and the new process. Specifically, the licensee failed to 4 Enclosure appropriately evaluate the se differences in a safety evaluation or obtain an NRC prior approval/licensee amendment.

The licensee entered the deficiency in their Corrective Actions Program as Action Request (AR) 1416141 and initiated a Standing Order reinstating TRM Section 3.4.a. until identifying the appropriate action for final disposition of this issue.

Analysis:

The inspectors determined that the failure to provide a written safety evaluation to demonstrate that the deletion of TRM Section 3.4.a did not require a licensee amendment was contrary to the requirements of 10 CFR 50.59(d)(1)and was a performance deficiency because an adequate screening would have identified the requirement to perform a safety evaluation. The inspectors determined that the finding was more than minor because the finding, if left uncorrected, would become a more significant safety concern.

Specifically, the deletion of the TRM Section 3.4.a could result in components not meeting the acceptance criteria for structural integrity being left in service, thereby adversely affecting reliability of the RCS components. The inspectors concluded this finding was associated with the Initiating Events Cornerstone.

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 that the finding could be evaluated using the SDP in accordance with IMC 0609, "Significance Determination Process."

Using Attachment 0609.04, "Initial Characterization of Findings," Table 2, the inspectors determined that the finding affected the Initiating Events cornerstone.

As a result, the inspectors evaluated the finding using Appendix A, "The Significance Determination Process (SDP) for Findings At-Power," Exhibit 1 for the Initiating Events cornerstone.

The inspectors answered "NO" to the first two questions in Exhibit 1.

Specifically, the inspectors determined that the finding could n ot result in exceeding the leak rate for a small LOCA and it could n ot have likely affected other systems used to mitigate a LOCA resulting in a total loss of their function , and therefore

, the inspectors screened the finding as Green.

This was based on the inspectors

' review of the four Action Reports issued since the TRM change implementation documenting the non

-conforming conditions related to structural integrity of the ASME components.

Specifically, in two instances the affected components were isolated, in one instance where the leak was not isolable, the plant was brought to cold shutdown as required by the Technical Specifications for implementation of repair, and the remaining case was found to be in compliance with the code on further evaluation.

5 Enclosure 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 determined that the associated finding had a cross-cutting aspect in the Decision Making component of the Human Performance cross

-cutting area because the licensee did not use conservative assumptions to ensure nuclear safety. Specifically, the licensee's failure to perform a safety evaluation was the result of a non

-conservative assumption that the TRM requirements were duplicate to the 10 CFR 50.55a; and therefore, the changes did not have an adverse effect even though the specific required actions listed in the deleted section of the TRM were not included in the new process for addressing structural integrity related non

-conformances.

H.1(b) Enforcement

Title 1 0 CFR 50.59 Section (d)(1) requires, in part , that the licensee maintain records of changes in the facility, of changes in procedures, and of tests and experiments. These records must include a written evaluation which provides the bases for the determination that the change, test, or experiment does not require a license amendment pursuant to Paragraph (c)(2).

Contrary to the above, on December 23, 2009 , the licensee failed to perform a written safety evaluation to demonstrate that the deletion of TRM Section 3.4.a did not require a license amendment. Specifically, the licensee did not adequately address the adverse effects of the changes in their screening L09-240. This inadequate screening resulted in the adverse changes not being evaluated under 10 CFR 50.59 change evaluation criteria.

The 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 Corrective Action Program (AR 1416141

). (NCV 05000373/2012007-01; 05000374/2012007

-01; Failure to Perform a Written Safety Evaluation for TRM Changes

). The associated finding is evaluated separately from the traditional enforcement violation

and therefore, the finding is being assigned a separate Tracking Number (FIN 05000373/2012007-02; 05000374/2012007

-02; Failure to Perform a Written Safety Evaluation for TRM Changes

).

.2 a. Permanent Plant Modifications

From September 4, 2012 through September 21, 2012, the inspectors reviewed seven permanent plant modifications that had been installed in the plant during the last three years. This review included in

-plant walkdowns for portions of the Engineering Change (EC) 383736 "SBLC Test Seismic Fix;" and EC 353398 "Design and Install AB

-TB HELB Barrier." The modifications were selected based upon risk significance, safety significance, and complexity. The inspectors reviewed the modifications selected to determine if: Inspection Scope 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 constituted seven permanent plant modification samples as defined in IP 71111.17-04. b. No findings of significance were identified.

Findings

OTHER ACTIVITIES

(OA)

4OA2

.1 Problem Identification and Resolution

a. Routine Review of Condition Reports From September 4, through September 21, 2012, 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, or 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.

Inspection Scope b. No findings of significance were identified.

Findings

4OA6

.1 Meetings O n September 21, 2012, the inspector s presented the inspection results to Mr. P. Karaba , 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.

Exit Meeting Summary ATTACHMENT:

SUPPLEMENTAL INFORMATION

Attachment

SUPPLEMENTAL INFORMATION KEY POINTS OF CONTAC

T

P. Karaba, Site Vice President

Licensee

H. Vinyard, Plant Manager

J. Miller, Acting

Engineering Director

W. Hilton, Senior Design Engineering Manager

L. Ekern, Nuclear Oversight

S. Shield, Acting Regulatory Assurance Manager

J. Washko, Operations Director

J. Vegara, Regulatory Assurance

J. Pula, Design Engineering

S. Tanton, Design Engineering

E. Zacharias, Design Engineering

R. Ruiz, Senior Resident Inspector

Nuclear Regulatory Commission

R. Daley, Chief, Engineering Branch

3, DRS LIST OF ITEMS OPENED, CLOSED AND DISCUSS

ED 05000373/2012

7-01; 05000374/2012

7-01 Opened and Closed

NCV Failure to Perform a Written Safety Evaluation for TRM Changes Section (Section 1R17.1b) 05000373/2012

7-02; 05000374/2012

7-02 FIN Failure to Perform a Written Safety Evaluation for TRM Changes (Section 1R17.1b) None.Discussed

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.

CALCULATIONS

Number Description or Title

L-003491 Date or Revision

Allowable Air Pocket in Water Filled RHR (LPCI) Piping

Revision 0

L-003623 SBLC Test Tanks Seismic Fix analysis

Revision 000A

L-003226 HELB Barrier Between Aux & Turbine Building on Col Row R North of R18 at Elev. 768'

-0" Revision 1

CORRECTIVE ACTION PROGRAM DOCUMENTS GENERATED Number Description or Title

AR 1409551

Date or Revision

NRC 50.59/MOD Inspection

-Screening L09

-259 0 9/06/2012 AR 1409696 2012 LaSalle 50.59/MOD inspection

- Procedure Enhancements

09/06/2012

AR 1410181

Typo in Screening L10

-128 09/07/2012

AR 1410193 NRC ID'D - Enhancement to LFP

-100-1 0 9/07/2012 AR 1415258

NRC ID'D Incorrect UFSAR Change Package Transmitted to RA

09/19/2012

AR 1415864

NRC ID: Issues Found in Calculation L

-003226 09/20/2012

AR 1416141

Inadequate 10CFR50.59 Screening

09/20/2012

CORRECTIVE ACTION PROGRAM DOCUMENTS REVIEWED

Number Description or Title

AR 984453 Date or Revision

Standby Liquid Control System Performance

10/26/2009

AR 1129847

Seismic Mounting of the SBLC Test Tank

- CDBI Question

10/22/2010

AR 1131668

Design analysis 030015(EMD) RE: SBLC

Test Tank 10/27/2010

AR 1132019

Update RE: Design analysis 030015 (EMD)

& SBLC Test Tank

10/28/2010

AR 1140568

Need ACE to Support SBLC Test Tank LER

11/15/2010

AR 872660 Small Air Void Identified by UT Exam in 2A LPCI Piping

01/27/2009 AR 950501 Need to Include Potential Air Void in Design Basis 0 8/06/2009 AR1322688 Potential Vulnerability Switchyard Single Open Phase Detection

2/03/2012

Attachment

CALCULATIONS

Number Description or Title

AR693116 Date or Revision

CDBI Unresolved Issue Determined to be 50.59 Violation

11/01/2007

DRAWINGS Number Description or Title

1E-0-4412AE Date or Revision

Schematic Diagram Diesel Generator 0 Generator Engine Control System "DG" Part 5 W 1E-0-4412AG Schematic Diagram Diesel Generator 0 Generator Engine Control System "DG" Part 7 X 1E-1-4220AJ Schematic Diagram Residual Heat Removal

System "RH" (E12) Part 9

AD 1E-1-4220AM Schematic Diagram Residual Heat Removal System "RH" (E12) Part 12

X 1E-1-4220AL Schematic Diagram Residual Heat Removal System "RH" (E12) Part 11

Z 1E-1-4220BS Schematic Diagram Residual Heat Removal System "RH" (E12) Part 41

O M-2100 P&ID C&I Details Control Rod Drive System

RD D 1E-1-4206CF Schematic Diagram RCMS, System "RD" Panel 1H13

-P603 RCMS Power Distribution

A 1E-!-4206AQ Schematic Diagram RCMS, System RD Panel 1H13

-P659: RCMS Power Module

C 1E-1-4207AE Schematic Diagram Control Rod Drive Hydraulic System "RD" (C118) Part

F 1E-1-4206CE Schematic Diagram RCMS, System RD,

Panel 1H13

-P603 RCMS Power Distribution

O 1E-1-4206AP Schematic Diagram RCMS, System RD

Panel 1H13

-P659: Power Distribution

C D-22079 Engine Cooling System Diagram 2600 KW Generator Set

H M-99 P & ID Standby Liquid Control System

AC M-145 P & ID Standby Liquid Control System

AE 1E-1-4226AR Schematic Diagram Reactor Core Isolation Cooling System "RI" (E51) Part 16

U 1E-1-4208AF Schematic Diagram Feedwater Control System "FW" Part 6

O 10 CFR 50.59

EVALUATIONS

Number Description or Title

L09-197 Date or Revision

Increase in Control Room and AEER

Makeup Flow Rates

Attachment

CALCULATIONS

Number Description or Title

L09-236 Date or Revision

RCMS Power Module Trip on Unintended Control Rod Motion

Revision 0

L09-245 High Pressure Turbine Nozzle Plate Replacement

2/5/2010 L10-96 50.59 Evaluation for EC 377369 and UFSAR Chg LUCR

-214 8/19/2010 L11-09 Reduce Operator Burdens By Modifying RHR/LPCS Interlock

Revision 0 L12-39 10CFR50.59 Evaluation Associated With

LFP-100-1, Revision 51 2/24/2012 L12-106 Revise Design Analysis for a UHS Temperature of 107

Revision 0 L09-206 Evaluation of Steam Flow Induced Error Impact on the Level 3 Setpoint Analytical Limit Revision 0 10 CFR 50.59 SCREENINGS

Number Description or Title

L08-157 Date or Revision EC 370438

- Install Additional Supports for Service Water Piping for Seismic Qualification

Revision 0

L09-181 Primary Containment Penetration Conductor Over

-Current Protective Devices Revision 0

L09-200 RPS Bus A Transfer

Revision 0

L09-203 Standby Liquid Control (SBLC) Pump Discharge Relief Valves 1(2)C41

-F029A/B Pressure Set Point Change

Revision 0

L09-213 Service Water Pump Auto Start

L09-241 Standby Liquid Control System Pressure

Revision 0

L09-250 Interim Single Loop Operation Baseline Data Gathering and Operability Verification

Revision 0

L10-128 Revise UFSAR and TS Bases for Refueling Equipment Requirements for Single Failure

Revision 0

L10-148 LOS-AA-S101 Revision 0

L10-160 Installing Jumper Around Cell in Division 1, 2 or 3 125 Volt Battery

Revision 9

L10-20 Install Full Current Bypass Electrical Jumper Around Unit 2 Main Power Transformer

Revision 0

L10-49 Removing 345KV Circuit Breaker from Service Revision 0

L10-56 Reclassify 0TI

-DG064B, C and 1(2)

TI-Revision 0

Attachment

CALCULATIONS

Number Description or Title

DG064A, B, C as Non

-Safety-Related Date or Revision

L11-153 Unit 1 (2) AC Power System Abnormal

LOA-AP-101(201) Revision 0

L11-95 50.59 Screening for EC 383736

Revision 0

L12-036 Modify Bus 141Y(241Y), 142Y(242Y) and 143(243) Undervoltage and Degraded Voltage Control room Alarms

Revision 0

L12-18 Unit 1 (2) AC Power System Abnormal

LOA-AP-101(201) Revision 0

MODIFICATIONS

Number Description or Title

EC 353390 Date or Revision

Service Water Pump Auto Start

EC 353398 Design and Install AB

-TB HELB Barrier(Roll Up) Revision 3

EC 368256 EDG K8 Relay Time Delay Set Point Change for 0DGK008, 1DGK008, 1E22K008 and

2E22K008 Revision 0

EC 370438 Install Additional Supports for Service Water Piping for Seismic Qualification

Revision 1

EC 376196 Reduce OPS Burdens by Modifying RHR Interlocks

Revision 0

EC 377369 Allowable Air Void in ECCS

- LPCI Piping

Revision 0

EC 380535 MSO MODs for Valves E51

-F019, 1E51

-F045, and 1E51-F046 to Address NRC RG 1.189

Revision 0

EC 383736 SBLC Test Tanks Seismic Fix

Revision 5

EC 387852 Applicability of Limerick Evaluation Re: NF 500 Mast and "Short Jib Hoist" to LaSalle

Revision 0

EC 388101 Setpoint Change for Feedwater Pump Low Suction Pressure Trip Time Delay Relay

Revision 0

EC 388666 Revise Design Analysis for

Post Accident UHS Temperature of 107

Revision 1 PROCEDURES

Number Description or Title

LOS-RH-M1 Date or Revision

RHR System and RHR WS System

Operability Test For Mode 1,2,3,4 and 5

Revision 25

LOS-RH-01 Filling and Venting the Residual Heat Removal System Revision 44

LFP-100-1 Master Refuel Procedure

Revision 51

OTHER DOCUMENTS

Number Description or Title

LGA-RH-103 Date or Revision

Unit 1 A/B RHR Operations in The LGAS/LSAMGS General Abnormal Revision 10

Attachment

CALCULATIONS

Number Description or Title

Procedure Date or Revision

LEP-DC-114 Installing Jumper Around Cell in Division 1, 2 or 3 125 Volt Battery

Revision 10

OE 09-003 SBLC Pump Discharge Relief Valve 1(2)C41

-F029A/B Revision 0

LPGP-PSTG-01S14 Plant Specific Technical Guidelines Section

14-LGA-Related Hard Cards Administrative Procedure Revision 4

LOA-AP-101 Unit 1 AC Power System Abnormal

Revision 39

LUCR-214 UFSAR Change Request

7/30/2010 OE 10-004 Standby Liquid Control (SBLC) Test Tank

Revision 2

OE07-004 Operability Evaluation for IR 693116

Revision 3

Attachment

LIST OF ACRONYMS USE

D ADAMS Agencywide Documents Access and Management System

AR Action Report ASME American Society of Mechanical Engineers

CFR Code of Federal Regulations

CNO Chief Nuclear Officer

DRS Division of Reactor Safety

EC Engineering Change

FIN Finding IMC Inspection

Manual Chapter

IST Inservice Testing

ITS Improved Technical Specifications

NCV Non-Cited Violation

NEI Nuclear Engineering Institute

NRC U.S. Nuclear Regulatory Commission

PARS Public Available Records System

RCS Reactor Coolant System

SDP Significance Determination Process

SPAR Standardized Plant Analysis Risk

SSC Structures, Systems and Components

SW Service Water

TRM Technical Requirements manual

TS Technical Specification

UFSAR Updated Final Safety Analysis Report

M. Pacilio

-2- In accordance with 10 CFR 2.390 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 Public Document Room or from the Publicly Available Records System (PARS) component of NRC's Agencywide Documents Access and Management System (ADAMS), accessible from the NRC Web site at http://www.nrc.gov/reading

-rm/adams.html

(the Public Electronic Reading Room).

Sincerely, /RA/ Robert

C. Daley, Chief

Engineering Branch 3

Division of Reactor Safety

Docket Nos. 50

-373; 50-374 License Nos. NPF

-11; NPF-18 Enclosure:

Inspection Report 05000373/2012007; 05000374/2012007

w/Attachment: Supplemental Information

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