IR 05000456/2008008

IR 05000456-08-008(DRS); 05000457-08-008(DRS); on 08/25/2008 09/12/2008; Braidwood Station, Units 1 & 2; Evaluations of Changes, Tests, or Experiments and Permanent Plant Modifications

ML082880242
Person / Time
Site:	Braidwood
Issue date:	10/13/2008
From:	Hills D E NRC/RGN-III/DRS/EB1
To:	Pardee C G AmerGen Energy Co
References
IR-08-008
Download: ML082880242 (20)

Text

October 13, 2008

Mr. Charles President and Chief Nuclear Officer (CNO), Exelon Nuclear Chief Nuclear Officer (CNO), AmerGen Energy Company, LLC 4300 Winfield Road Warrenville, IL 60555

SUBJECT: BRAIDWOOD STATION, UNITS 1 AND 2 EVALUATIONS OF CHANGES, TESTS, OR EXPERIMENTS AND PERMANENT PLANT MODIFICATIONS BASELINE INSPECTION REPORT 05000456/2008008(DRS); 05000457/2008008(DRS)

Dear Mr. Pardee:

On September 12, 2008, the U.S. Nuclear Regulatory Commission (NRC) completed the evaluations of changes, tests, or experiments and permanent plant modifications inspection at your Braidwood Station, Units 1 and 2. The enclosed report documents the inspection results, which were discussed on September 12, 2008, with Mr. Larry Coyle 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. Based on the results of this inspection, no findings of significance were identified. 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 (PARS) component of NRC's document system (ADAMS). ADAMS is accessible from the NRC Website at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Sincerely,/RA/ David E. Hills, Chief Engineering Branch 1 Division of Reactor Safety Docket Nos. 50-456; 50-457 License Nos. NPF-72; NPF-77

Enclosure:

Inspection Report 05000456/2008008(DRS); 5000457/2008008(DRS)

w/Attachment:

Supplemental Information cc w/encl: Site Vice President - Braidwood Station Plant Manager - Braidwood Station Regulatory Assurance Manager - Braidwood Station Chief Operating Officer and Senior Vice President Senior Vice President - Midwest Operations Senior Vice President - Operations Support Vice President - Licensing and Regulatory Affairs Director - Licensing and Regulatory Affairs Manager Licensing - Braidwood, Byron and LaSalle Associate General Counsel Document Control Desk - Licensing Assistant Attorney General J. Klinger, State Liaison Officer, Illinois Emergency Management Agency Chairman, Illinois Commerce Commission

SUMMARY OF FINDINGS

IR 05000456/2008008(DRS); 05000457/2008008(DRS); 08/25/2008 - 09/12/2008; Braidwood Station, Units 1 & 2; Evaluations of Changes, Tests, or Experiments and Permanent Plant Modifications. The inspection covered a two-week announced baseline inspection on evaluations of changes, tests, or experiments and permanent plant modifications. The inspection was conducted by three regional based engineering inspectors. Based on the results of this inspection, no findings of significance were identified. The significance of most findings is indicated by their color (Green, White, Yellow, Red) using Inspection Manual Chapter (IMC) 0609, "Significance Determination Process" (SDP). Findings for which the SDP does not apply may be Green or 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. NRC-Identified

and Self-Revealed Findings

Cornerstone: Initiating Events

No findings of significance were identified.

Cornerstone: Mitigating Systems No findings of significance were identified. Cornerstone:

Barrier Integrity No findings of significance were identified.

B. Licensee-Identified Violations

No findings of significance were identified.

2

REPORT DETAILS

REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity

1R17 Evaluations of Changes, Tests, or Experiments and Permanent Plant Modifications

.1 Evaluations of Changes, Tests, or Experiments

a. Inspection Scope

From August 25, 2008, through September 12, 2008, the inspectors reviewed 10 evaluations performed pursuant to 10 CFR 50.59 to determine if the evaluations were adequate and that prior NRC approval was obtained as appropriate. The inspectors also reviewed 22 screenings where licensee personnel had determined that a 10 CFR 50.59 evaluation was not necessary. In regard to the changes reviewed where no 10 CFR 50.59 evaluation was performed, the inspectors verified that the changes did not meet the threshold to require a 10 CFR 50.59 evaluation. The evaluations and screenings were chosen based on risk significance, safety significance, and complexity. Documents reviewed are listed in the attachment to this report. 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 constitutes 10 samples of evaluations and 22 samples of changes as defined in Inspection Procedure 71111.17-05.

b. Findings

During this inspection, the NRC senior resident inspector requested the team's assistance with their review of the 10 CFR 50.59 evaluation EC361637 (FDRP 23-003)

"Abandon the Upper Cable Spreading Room Carbon Dioxide (CO 2) System," Revision 0. The results of that review will be documented in the Braidwood Integrated Inspection Report 2008004.

.2 Permanent Plant Modifications

a. Inspection Scope

From August 25, 2008, through September 12, 2008, the inspectors reviewed 13 permanent plant modifications that had been installed in the plant during the last three years. The modifications were chosen based upon risk significance, safety significance, and complexity. As per Inspection Procedure 71111.17, one modification 3was chosen that affected the design bases and functioning of interfacing systems as well as introducing the potential for common cause failures. The inspectors reviewed the modifications to verify that the completed design changes were in accordance with the specified design requirements, and the licensing bases, and to confirm that the changes did not adversely affect any systems' safety function. Design and post-modification testing aspects were verified to ensure the functionality of the modification, its associated system, and any support systems. The inspectors also verified that the modifications performed did not place the plant in an increased risk configuration. 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 constitutes 13 samples as defined in Inspection Procedure 71111.17-05.

b. Findings

(1) Temporary/Permanent Conversion of Lead Shielding on Piping Systems

Introduction:

The inspectors identified an unresolved item (URI) concerning seismic Category I pipe supports associated with the safety injection (SI) and residual heat removal (RH) piping systems. Design documents for the 2SI06 and 2RH01 piping subsystems' pipe supports were not sufficiently detailed to demonstrate compliance with the American Institute of Steel Construction (AISC) Manual of Steel Construction Code and the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code.

Description:

The SI and RH piping systems are part of the emergency core cooling system (ECCS). The Braidwood Updated Final Safety Analysis Report (UFSAR), Section 6.3.1, stated the primary function of the ECCS is to remove the stored and fission product decay heat from the reactor during accident conditions. The ECCS also provides shutdown capability for design basis accidents by means of boron injection. The ECCS was classified as a safety class II system designed to meet seismic category I requirements. The inspectors reviewed Calculation BRW-97-0827-M, "Piping Evaluation for Lead Shielding Installation on Subsystem 2SI06 Piping per Temporary Lead Shielding Request (TSR) No.95-153, 96-018,96-045, 96-053, and 97-120," Revision 0 and Minor Revision 0A. The purpose of Revision 0 was to evaluate the affect of the temporary lead shielding installed on the 2SI06 piping subsystem (i.e., lead shielding installed on sections of pipelines 2RH01BA-12", 2RH01BB-12", 2RH01CA-16", 2RH01CB-16", 2SI06BA-24", and 2SI06BB-24") by the TSRs. The purpose of Minor Revision 0A was to evaluate the affect of converting the temporary lead shielding to permanent lead shielding. In addition, Minor Revision 0A identified that based on recent industry concerns the lead shielding weighed up to 10 percent more than was previously analyzed. Pipe stresses were determined from loads and load combinations due to internal pressure, pipe system dead weight, pipe thermal expansion and seismic excitation. The 2SI06 and 2RH01 piping subsystems were designed to the ASME Boiler and 4Pressure Vessel Code,Section III, Subsection NC and ND, 1977 Edition up to and including the 1979 Addenda. The associated pipe supports were designed to the AISC Manual of Steel Construction Code and the ASME Boiler and Pressure Vessel Code,Section III, Subsection NF, 1977 Edition through Summer 1979 Addenda. The seismic response spectra analysis of the piping subsystem was analyzed using the ASME Code Case N-411, "Alternative Damping Values for Seismic Analysis of Classes 1, 2, and 3 Piping Sections,Section III, Division 1." As specified in the licensee's UFSAR, Section 3.7, Table 3.7-1, "Damping Values," the ASME Code Case N-411 may be used for alternative damping values when NRC conditions as defined in Regulatory Guide 1.84, "Design and Fabrication Code Case Acceptability ASME Section III Division I," Revision 24 are met. In Calculation BRW-97-0827-M, Revision 0, the licensee's qualification of the pipe supports were based on their review of Calculation 13.2.29, "Structural Calculation for Mechanical Component Support [Pipe Support Number]," Revision 2 and the use of engineering judgment. The licensee concluded that sufficient margin existed in the pipe support design such that the supports would be able to withstand the increased loads from the installed lead shielding. The inspectors reviewed the Structural Calculation for Mechanical Component Support

[Pipe Support Number] contained in Calculation 13.2.29, Revision 2, for the following: Pipe Support Number Pipe Support Number Pipe Support Number 2SI06309X 2SI06328X 2SI06342X 2SI06310X 2SI06335X 2SI06345X 2SI06311G 2SI06336X 2SI06351X 2SI06316X 2SI06337X 2SI06358X 2SI06318X 2SI06340S 2SI06360X The inspectors determined that the engineering judgment used in Calculation BRW-97-0827-M, Revision 0 was not valid and the aforementioned pipe supports could exceed their design basis and operability acceptance limits. Also, a condition specified in ASME Code Case N-411 stated "This Code Case is not appropriate for analyzing the dynamic response of piping systems using supports designed to dissipate energy by yielding." The licensee initiated issue report (IR) 00816677, "NRC MOD/50.59 Inspection - 2SI06 Piping Subsystem Support," dated September 11, 2008, to address this issue. In response to IR 00816677, the licensee's prompt operability determination concluded through analysis that the fillet weld connection between the attachment plate and the embedment plate for pipe support 2SI06316X exceeded design basis and operability limits. Subsequently, the licensee performed a walkdown to field verify the actual fillet weld size of this connection. The actual fillet weld size was determined to be 7/16" thick, which was greater than the 1/4" thick fillet weld size used in the analysis. The licensee determined that the pipe support 2SI06316X fillet weld connection exceeded the design basis limits but was within operability acceptance limits. Further analysis by the licensee showed pipe supports 2SI06328X, 2SI06340S, 2SI06345X and 2SI06358X exceeded their design basis limits. The concrete 5expansion anchor bolt evaluation for each pipe support resulted in a factor of safety of less than four but greater than two. A factor of safety of greater than two satisfies the operability requirements specified in procedure OP-AA-108-115, "Operability Determinations (CM-1)," Revision 6. Pipe support 2SI06318X was determined to have no margin for design basis acceptance limits and pipe support 2SI06337X had minimal design basis margin remaining. The licensee determined that the pipe supports 2SI066309X, 2SI06310X, 2SI06311G, 2SI06335X, 2SI06336X, 2SI06342X, 2SI06351X and 2SI06360X met design basis and operability acceptance limits. This issue is considered unresolved pending the licensee's response to address the inspectors' request for additional information (RAI) regarding the following inspectors' concerns:

• The licensee's prompt operability determination was made without incorporating the additional 10 percent weight and installed locations of the temporary lead shielding identified by Calculation BRW-97-0827-M, Minor Revision 0A into the licensee's pipe stress computer analysis program. These changes have the potential to affect all pipe supports and anchors on the 2SI06 and 2RH01 piping subsystems.

• There are approximately 58 pipe supports and five pipe anchors associated with the 2SI06 piping subsystem, which were not verified to determine their design basis acceptability.

• Pipe supports and anchors for the 2RH01 piping subsystem were never verified to determine their design basis acceptability.

• A condition specified in the ASME Code Case N-411 states "When used for reconciliation work or for support optimization of existing designs, the effects of increased motion on existing clearances and on line mounted equipment should be checked." This condition along with the other four ASME Code N-411 specified conditions were not addressed in the pipe stress analysis for either the 2SI06 or 2RH01 piping subsystems. At the end of this inspection, the licensee stated that the SI and RH piping systems' pipe stress computer analysis program would be revised to reflect the RAI concerns and the results provided to the NRC for review within three months of September 12, 2008. (URI 05000457/2008008-01(DRS))

4OA2 Identification and Resolution of Problems

.1 Routine Review of Condition Reports

a. Inspection Scope

From August 25, 2008, through September 12, 2008, the inspectors reviewed corrective action process documents that identified or were related to 10 CFR 50.59 evaluations 6and permanent plant modifications. The inspectors reviewed these documents to evaluate the effectiveness of corrective actions related to permanent plant modifications and evaluations for changes, tests, or experiments issues. 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 team are listed in the attachment to this report.

b. Findings

No findings of significance were identified.

4OA6 Management Meetings

.1 Exit Meeting Summary

On September 12, 2008, the inspectors presented the inspection results to Mr. Larry Coyle and other members of the licensee staff. The licensee acknowledged the issues presented. The inspectors confirmed that none of the potential report input discussed was considered proprietary.

.2 Interim Exit Meetings

No interim exits meetings were conducted. ATTACHMENT:

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee

L. Coyle, Plant Manager

C. Furlow, Design Engineering

G. Golwitzer, Regulatory Assurance

J. Gosnell, Design Engineering

D. Gustofson, Design Engineering

D. Ibrahim, Design Engineering

J. Knight, Nuclear Oversight

T. McCool, Operations

J. Morales, Design Engineering

R. Gadbois, Maintenance Director

J. Odeen, Project Management Director

J. Petty, Regulatory Assurance

D. Riedinger, Design Engineering Manager

B. Schipiour, Work Management Director

M. Smith, Engineering Director Nuclear Regulatory Commission

B. Dickson, Senior Resident Inspector

A. Garmoe, Resident Inspector

J. Heath, Resident Inspector

M. Perry, Illinois Resident Inspector

Attachment

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened

05000457/2008008-01(DRS) URI RAI To Determine Adequacy of Pipe Supports Designed for Design Basis Loading Conditions with Lead Shielding Installed on SI and RH Subsystems (Section 1R17.2b.(1))

Closed and

Discussed

None

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 Date or Revision

051336 Addendum Piping Stress Rpt ESW Sys 2SX13 003D

051790 Piping Stress Rpt for Rx Coolant PZR Sys 2RY09 00E

2890 Doc of Qual Calcs for Class 1 Supports in Piping Sys 2RY09-PZR Safety & Relief Vlvs

000B

067084 Doc of Qual Calcs for Class 1 Supports in Piping Subsystem 2SI25 Boron Injection Sys

01A 6.5.7-BYR02-064 Check Structural Adequacy of New Removable Platform (walkway) in U2 Refueling Pool

13.1.31-BRW-07-0099-S Structural Eval of Existing Pipe Support M-1SI21047G for Revised Loads

13.1.34-BRW-06-0182-S Eval of Support 1SX66007X for Revised Loads 0 13.1.34-BRW-06-0183-S Eval of Support 1SX66006R for Revised Loads 0

13.1.34-BRW-06-0184-S Eval of Support 1SX66009R for Revised Loads 0 Structural Calculations for the Following Mechanical Component Supports:

2SI06309X 2SI06328X 2SI06342X 2SI06310X

2SI06335X 2SI06345X 2SI06311G 2SI06336X 2SI06351X 2SI06316X 2SI06337X 2SI06358X 13.2.29 2SI06318X 2SI06340S 2SI06360X

13.2.31-BRW-07-0119-S Qual of Existing Pipe Support M-2SI25011R 0 13.2.31-BRW-07-0120-S Qual of Existing Pipe Support M-2SI25012X 0 13.2.31-BRW-07-0121-S Qual of Existing Pipe Support M-2SI25013X 0

13.4.11.2-BRW-07-0127-S Structural Eval of Existing Support M-1PSEH024S001T

14.1.14-BRW-2007-0031-S Qual of Pipe Supports M-1ES35043R, M-1ES35044R, M-1ES35045R & M-1ES35046R

14.1.14-BRW-2007-0032-S Qual of Pipe Supports M-1ES37047D & M-1ES38048D

14.2.1-BRW-07-0066-S Qual of Pipe Support M-2ABF22021G 0 14.2.35-BRW-07-0011-S Qual of Pipe Support 2WXF153020T 1 19-T-6 DG Loading During LOOP/LOCA 6A 24.2.1 Mech Component Supports-Design Control Summary for Plant Mod & Station Support Work

ATD-0026 Combustible Fire Loads 9-08

BRW-07-0003-M Seismic Qual of 1.5" & 2" SI Throttle Vlvs (U1 & 2 0

Attachment CALCULATIONS

Number Description or Title Date or Revision Tag Nos. SI8810A-D, SI8816A-D & SI8822A-D)

BRW-07-0102-M Eval/Characterization of Through Wall Leakage from Line 2SX27DA-10 per Code Case N-513-1

BRW-97-0827-M Piping Eval for Lead Shielding Installation on Subsystem 2SI06 Piping per Temp Lead Shielding Request TSR95-153, 96-018/045/053

& 97-120 0

BRW-97-0827-M Piping Eval Lead Shielding on Subsystem 2SI06 0A

NED-I-EIC-0004 PZR Press Protection Channel Error Analysis 4

SITH-1 RWST Level Setpoints 6 & 7

SG-BRW-96-414-M Piping Stress Rpt for Subsystem 1FW02 02A

SG-BRW-96-415-M Piping Stress Rpt for Subsystem 1FW03 03A

SG-BRW-96-416-M Piping Stress Rpt for Subsystem 1FW04 02A

SG-BRW-96-429-M Piping Stress Rpt for Subsystem 1SD26 02A

SG-BRW-96-430-M Piping Stress Rpt for Subsystem 1FW05 02A

SG-BRW-96-431-M Piping Stress Rpt for Subsystem 1SD27 01A

SG-BRW-96-432-M Piping Stress Rpt for Subsystem 1SD28 03A

SG-BRW-96-433-M Piping Stress Rpt for Subsystem 1SD29 02A

CORRECTIVE ACTION PROGRAM DOCUMENTS ISSUED DURING INSPECTION

Number Description or Title Date or Revision

00811051 NRC Inspection Calc Request - Revisions Unavailable August 26, 2008

00815101 NRC Mod/50.59 Insp - Rev 2 to BWOP

CV-16 Did Not Specify Limitation Procedure Use to One Operating Cycle September 8, 2008

00815166 Minor Discrepancies in Fire Combustible Loading Calc September 8, 2008

00815707 NRC Mod/50.59 Insp - Issue with

BRW-E-2007-122 September 9, 2008

00816038 NRC Mod/50.59 Insp - Calc Assumption September 10, 2008

00816539 NRC Mod/50.59 Insp - Calc Discrepancy September 11, 2008

00816619 NRC Mod/50.59 Insp - Concern Actions for EDG Freq Issue September 11, 200800816677 NRC Mod/50.59 Insp - 2SI06 Piping Subsystem Support September 11, 200800817175 NRC Observations Following Mod/50.59 Inspection September 12, 2008

CORRECTIVE ACTION PROGRAM DOCUMENTS ISSUED PRIOR TO INSPECTION

Number Description or Title Date or Revision

00629351 CDBI FASA - DG Freq Variations Not Addressed in Calcs May 14, 2007

00641501 Missing M&TE Selection Sheets for Calibration June 18, 2007

Attachment DRAWINGS Number Description or Title Date or Revision

20E-1-4012A Key Diagram 120Vac Instrument Bus 111 (1IP01J) ESF Division II - Chanel I

P 20E-1-4030IP01 Schematic Diagram 7.5kVA Fixed Freq Inverter for Instrument Bus 111 (1IP05E)

N

50.59 EVALUATIONS

Number Description or Title Date or Revision

BRW-E-2007-122 Implement Results of RWST Vortex Testing July 25, 2007

BRW-E-2007-147 Modifications to SI Throttle Vlvs 1SI8810A-D, 1SI8816A-D, 1SI8822A-S

BRW-E-2007-161 Change In-Core Decay Time for A1R13 from 100Hrs to 78Hrs 0

BRW-E-2007-168 Alternate Load Paths for Heavy Loads for A1R13 0

BRW-E-2007-221 Revise BwOPCV-16 to Temporarily Allow PZR Aux Spray Operation

BRW-E-2008-2 Modifications to the SI Throttle Vlvs 2SI8810A-D, 2SI8816A-D, 2SI8822A-D

BRW-E-2008-77 Alternate Load Paths for Heavy Loads for A2R13 0

BRW-E-2008-105 Operability Eval 07-008, Related to Reduction in Operator Action Time Required to Prevent Overfilling of a Ruptured

SG During an SGTR Event

BRW-E-2008-135 Power Ascension Testing of New Digital Control Sys

0

EC361637 (FDRP 23-003) Abandon the UCSR Carbon Dioxide (CO

2) Sys 0

MODIFICATIONS

Number Description or Title Date or Revision

EC0000356371 Reactor Head Hoist Upgrade 0 EC0000358829 ECCS Sump Screen Project to Resolve

GSI 191 Replace Fiber Insulation w/MRI - SG & Piping Below

EL 429ft

EC0000358934 U2 MCR Safety Related Recorder Replacements 0 EC0000359960 Rewire Seq of Events Recorder Cabinet to Correct Power Supply Failure Alarm

EC0000360141 Replace SI Throttle Vlv Trim, Bonnet, Stem, & Manual Operators & Elimination of Downstream Orifice Plates to Support

GSI 191 (U1)

EC0000360143 Replace SI Throttle Vlv Trim, Bonnet, Stem, & Manual Operators & Elimination of Downstream Orifice Plates to Support

GSI 191 (U2)

EC0000362458 Reactor Head Hoist Upgrade 0 EC0000363113 Replace 480V Feed Breakers for MCC Dist Panels for 0

Attachment MODIFICATIONS

Number Description or Title Date or Revision Coordination EC0000363956 Setpoint Change for 1CV8119 Relief Vlv 1 EC0000365465 Replace U1 Barton Transmitter EQ Seals in Containment 2 EC0000366583 Implement Results of RWST Vortex Testing August 31, 2007

EC0000366605 Implement Results of RWST Vortex Testing U1 & U2 RWST Empty Level Change from 7 - 9 percent July 27, 2007 EC0000366628 Revise U1 PZR Ch Dev Alarm from 2 - 3 percent of Full Power 0

OPERABILITY EVALUATIONS

Number Description or Title Date or Revision

07-005 Through-Wall Leakage of Line 2SX27DA-10 0

PROCEDURES

Number Description or Title Date or Revision

1BCA-1.1 Loss of Emergency Coolant Recirculation U1 105 1BwCA-1.1 Loss of Emergency Coolant Recirculation U1 200

1BCA-1.3 Sump Blockage Control Room Guideline U1 2 1BEP

ES-1.3 Transfer to Cold Leg Recirculation U1 105 1BwCA-1.3 Sump Blockage Control Room Guideline U1 200

1BwEP

ES-1.3 Transfer to Cold Leg Recirculation U1 200

BwOP

CV-16

PZR Aux Spray Operation 2

CC-AA-102 Design Input & Config Change Impact Screening 15

CC-AA-103 Config Change Control for Perm Phy Plant Changes 17

CC-AA-309 Control of Design Analysis 7

CC-AA-309-1001 Guidelines for Prep & Processing Design Analyses 4

CC-AA-309-1011 General Station Piping Analysis 2

OP-AA-108-115 Operability Determinations (CM-1) 6

REFERENCES

Number Description or Title Date or Revision53904 Equivalency Eval: Replacement of Obsolete Westinghouse SSPS Input Relays with Potter & Brumfield Relays

56327 Equivalency Eval: Replacement of Certain Potter & Brumfield Relays with Magnecraft Relays

61168 Equivalency Eval: Replacement of CR Vent Fan Mtrs with Mtrs that Have Vacuum Press Impregnated Insulation Sys

61753 Equivalency Eval May 2, 2008 ASME Code Case Eval Criteria for Temp Acceptance of Flaws in Moderate March 28, 2001

Attachment REFERENCES

Number Description or Title Date or RevisionN-513-1 Energy Class 2 or 3 Piping Sect XI, Div 1 BwSD-FP-23 U1 Sys Demonstration Halon 1301 Fire Protection 0 BwSD-FP-63 U2 Sys Demonstration Halon 1301 Fire Protection 0

EC369677 Tech Eval Documenting Effect of EDG Freq Variations on ECCS Pump Flows.

EC366629 Tech Eval to Evaluate CV, RH, & SI Pmp Performance Due to EDG Freq Shift Post-LOCA (59.2-60.8 Hz)

RG 1.84 Design & Fab Code Case Acceptability ASME Sect III Div I 24

50.59 SCREENINGS

Number Description or Title Date or Revision

BRW-S-2007-005 Design Change EC363874, TS Basis Change 07-001, TRM Change 07-001, FDRP 23-007 & UFSAR Change

DRP 12-008

BRW-S-2007-034 EC363956, Setpoint Change for 1CV8119 Relief Vlv 0

BRW-S-2006-056 Permanently Remove Existing Ladder & Associated Diagonal Braces at U1 Containment Equip Hatch

BRW-S-2007-061 TRM Change Request 07-004 0

BRW-S-2007-066 Revise NED -

MSD-032; Byron & Braidwood CR Heat Load Calc

BRW-S-2007-084 Movement of Heavy Loads in the Fuel Bldg & Spent Fuel Pool in Support of the Rack G Repair Project

BRW-S-2007-091 Turb Governor & Throttle Vlv Testing Freq Change 0

BRW-S-2007-093 Temp Remove Strut 2SX13002X from 2SX27DA-10 Install Leak Mitigation Clamp w/ 1/2" Vlv Tap Over Pinhole Leak

BRW-S-2007-099 Rx Trip or SI (EP-0) Response to Nuclear Power Generation/ATWS(FR-S.1)

BRW-S-2007-120 Runway Beams Extension for 2HC22G in U2 Containment 0

BRW-S-2007-124 Increase PR Ch Dev Setpoint to Prevent Nuisance Alarms 0

BRW-S-2007-137 New Lifting Lug at the U1 & U2 Containment Equip Hatch Penetration Sleeves EC366588 & EC366589

BRW-S-2007-149 U2 Containment Refueling Pool Platform 0

BRW-S-2007-153 Revise Power Dissension Procedures Due to Replacement of DEH Control Sys

BRW-S-2007-160 Revise Aux Bldg HVAC Sys Shutdown Procedure 0

BRW-S-2007-176 Fuel Bldg Crane Heavy Load Lifts >2000lbs & <30,000lbs Using the Aux Hook/BwFP

FH-20T4

BRW-S-2007-177 Fuel Bldg Crane Heavy Load Lifts Using the 125 Ton Main Hook/BwFP

FH-20T5

BRW-S-2007-207 Revise DG Alignment to Standby Condition Procedure 0

BRW-S-2008-075 Operation with Higher RH Letdown Flow Rates 0

BRW-S-2008-090 Eliminate Snubber 2RY09060S on PZR PORV Line to 0

Attachment 50.59 SCREENINGS

Number Description or Title Date or Revision Eliminate Interferences During WOL Installation

BRW-S-2008-096 UFSAR Change Describe Additional Analysis for CS Pmps 0

BRW-S-2008-119 Adverse Cooling Lake Conditions 0

Attachment

LIST OF ACRONYMS

USED [[]]

ADAMS Agency Wide Document Access and Management System

AISC American Institute of Steel Construction

ASME American Society of Mechanical Engineers

CFR Code of Federal Regulations

CNO Chief Nuclear Officer

DRP Division of Reactor Projects

DRS Division of Reactor Safety

ECCS Emergency Core Cooling System

IMC Inspection Manual Chapter

IR Inspection Report and Issue Report

MOD Modification

NEI Nuclear Energy Institute

NRC [[]]

U.S. Nuclear Regulatory Commission

NRR Office of Nuclear Reactor Regulation

OA Other Activities

PARS Publicly Available Records System

RAI Request for Additional Information

RH Residual Heat Removal

ROP Reactor Oversight Process

SDP Significance Determination Process

SI Safety Injection

TS Technical Specifications

TSR Temporary Lead Shielding Request

UFSAR Updated Final Safety Analysis Report URI Unresolved Item