ML20196A405

Rept of Facility Changes,Tests & Experiments,Per 10CFR50.59 for 970601-980531. with

ML20196A405
Person / Time
Site:	Waterford
Issue date:	05/31/1998
From:	Ewing E ENTERGY OPERATIONS, INC.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
W3F1-98-0182, W3F1-98-182, NUDOCS 9811270139
Download: ML20196A405 (118)

Text

l y Entergy Operttions,Inc.

Killona. LA 70066 Tel 504 739 6242 Er C. Ewnng,111 yr ga;sajstysnewaar uan W3F1-98-0192 A4.05 PR November 24 1998 U.S. Nuclear Regulatory Commission Attn: Document Control Desk W: thington, D.C. 20555 Eu tect: Waterford 3 SES Docket No. 50-382 License No. NPF-38 Report of Facility Changes, Tests, and Experiments

• Gentlemen:

Enclosed is the Report of Facility Changes, Tests, and Experiments for Waterford 3, which is submitted pursuant ,o 10CFR50.59, which covers the period from June 1, 1997, through May 31,1998. The report also includes summaries of commitment  !

changes made pursuant to the NEl " Guideline for Managing NRC Commitments." I If you have any questions regarding this report, please contact me at (504) 739-6242.

Very truly yours, j

E.C. E g

. Director

/

Nuclear Safety & Regulatory Affairs i

ECE/ELUrtk

Enclosures:

50.59 Summary Report i

50.59 Report -JE 4[ "

cc: E.W. Merschoff (NRC Region IV), C.P. Patel (NRC-NRR),

J. Smith, N.S. Reynolds, NRC Resident inspectors Office c . s 3 p, 9811270139 990531 PDR ADOCK 05000302 R PDR

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I l

l Report of Facility Changes, Tests, and Experiments ,

W3F1-98-0192 Page 2 ' '

November 24,1998 i

(w/o Enclosure) <

ccMail: J.R. McGaha (M-ECH-65)

C.M. Dugger (W-GSB-300)  :

T.R. Leonard . (W-MSB4-300)

J.G. Hoffpauir (W-MSB4-336)

R.F. Burski (W-GSB-305)

A.J. Wrape (W-GSB-315)

E.P. Perkins (W-GSB-318)

R.E. Allen (W-GSB-102)

J.J. Zabritski (W-ADM-567)

M.K. Brandon (W-GSB-318)

M. Kansler (M-ECH-66)

(w/ Enclosure) bec: Waterford 3 Records Center (W-GSB-100)

Licensing Green Folder File i

l i --

9 ENTERGY OPERATIONS, INC.

WATERFORD 3 SES DOCKET NO. 50-382 LICENSE NO. NPF-38 REPORT OF FACILITY CHANGES. TESTS. AND EXPERIMENTS PER 10CFR50.59 JUNE 1.1997 THROUGH MAY 31,1998

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SUMMARY

This report provides the Waterford 3 Facility Changes made pursuant to 10CFR50.59(a)(1). The report covers the period from June 1,1997, through May 31, ,

1998. None of the items in the report were found to involve an unreviewed safety question.

Section I identifies acronyms used in the report.

Section 11 of the report identifies 88 Facility Changes which consist of: 6 Design Changes (DCs),14 Condition Identification / Work Authorizations (Cl/WAs), 3 Temporary

' Alteration Requests (TARS),9 License Document Change Requests (LDCRs),22 Miscellaneous Evaluations,30 Commitment Changes, and 4 Engineering Requests (ERs).

Section 111 of the report identifies 12 Procedure Changes which consist of 7 Plant Procedures and 5 Special Test Procedures (STPs).

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WATERFORD 3  !

10CFR50.59 REPORT ENTERGY OPERATIONS, INC. i JUNE 1,~1997 THROUGH MAY 31,1998 TABLE OF CONTENTS -

item- Page -

No. No.

1. LIST OF ACRONYMS 1

- II. FACILITY CHANGES A; - DESIGN CHANGES r 1. DC-3470, Auxiliary Component Cooling Water System Waterhammer,

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6 Revision 1

2. DC-3472, Hot and Cold Leg RTD Noise Abatement 7

3. DC-3502, Reduce Bypass L'eakage from Penetrations 53 and 65, 8 Revision 1

4. DC-3518, Main Con' denser Air Evacuation ' System improvements, .9

. Revision 3 -

5. DC-3536,.!nstallation of Control Room Emergency Filter Outside 10 Air intakes

6. DC-3539, Vortex Breakers for Refueling Water Storage Pool and -11 Condensate Storage Pool, Revision 0 and Revision 1

. B. CONDITION IDENTIFICATION / WORK AUTHORIZATION (Cl/WA)

1. . Cl-303041/WA-01147691, Essential Chillers DrainsNents Additions 12

' 2.i . Cl-310358/WA-01160028, Rerating of Component Cooling Water 13 Header to the Containment Fan Coolers Between inlet and Outlet Isolation Valves

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3. Cl-310627/WA-01160233, Modify FPC Purification Line 7FS3-8. 14 Supports and Equipment

4. Cl-310814/WA-01160476, Use-as-is - Core Protection Calculator 15

. Cabinet, CP-22, Filter

5. Cl-311171/WA-01161072, Resetting of Relief Valves ACC-121 A(B) 16 on the Shell Side of the CCW Heat Exchangers

6. Cl-310186/WA-01162268, Uprate Component Cooling Water in the 17 Area of the Dry Cooling Towers

. 7. Cl-312040/WA-01162451, Reroute Chiller 'B' Control Power Cable 18 1

8. Cl-311992/WA-01162852, Add Fire Detection in Walkway Between - 19  ;

Fuel Handling Building and Reactor Auxiliary Building at Elevation +21' l

9. Cl-312316/WA-01162893, Appendix R Lighting Upgrade 20  :

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10. Cl-312692/WA-01163755 and Cl-3126914WA-01163756,1/2" 21 l Socket Weld Modification and Shortening 1/2" Vent Line Tubing, Revision 0 and Revision 1

11. Cl-313735/WA-01165734, HRA Containment Isolation Valve 22 Position Indication

12. Cl-313736/WA-01165735, HRA Containment isolation Valve 23 Position indication

13. Cl-313818/WA-01165801, Addition of a CIAS Contact in Series 24 with the Opening Circuit of CAR-201B 14 Cl-314004/WA-01167913, Addition of a CIAS Contact in Series 25 with the Opening Circuit of CAR-201 A C. TEMPORARY ALTERATION REQUEST (TAR)

1. TAR-97-016, Startup Transformer 26 ii

item Page No. No.

2. TAR-97-018, Disable Diversion of Vacuum Pump Exhaust Header 28 to RAB Normal Ventilation

3. TAR-97-004, Addition of Portable Ion Exchange Vessels to the 29 ,

ACCW Filtration Skids, Revision 1 D. LICENSING DOCUMENT CHANGE REQUESTS (LDCR) I l

1. LDCR-97-0114, Revises FSAR Section 3.9.1.2 30 1

2. LDCR-97-0194, Revises FSAR Sections 3.1.37,6.0,6.2.5, 31 6.5.3.1, 7.3.1.1.9.3, and 15.6.3.3.5.3 l

3. LDCR-97-0205, Revises FSAR Sections 9.2.2.2.1 and 9.3.8.2 32

4. LDCR-97-0230, Revises FSAR Table 9.2-8 33

5. LDCR-97-0234, Revises FSAR Section 6.2.1.5 and Figure 6.2-30A 34

6. LDCR-97-0243, Revises FSAR Section 6.2.1.5 35

7. LDCR-97-0249, Revises FSAR Section 9.2.5 and Table 9.2-9 36

8. LDCR-97-0250, Revises FSAR Section 9.3.1.2 37

9. LDCR-97-0260, Revises FSAR Section 3.8.3.8 38 E. MISCELLANEOUS EVALUATIONS

1. Technical Requirements Manual 3/4.6.1.5, Minimum Containment 39 Air Temperature

2. Calculation EC-M89-004, Water Levels Inside Containment (Post- 40 LOCA), Revision 3

3. Calculation EC-E90-006, EDG Loading and Fuel Oil Consumption, 41 Revision 2, Changes 12 and 13

4. Calculation EC-S97-016, Emergency Feedwater Minimum 42 Flow Requirement iii

ltem Page No. No.

5. Calculation EC-192-019, Plant Protection System Setpoint Uncertainty 43 Calculation, Revision 1, Change 1

6. Technical Requirements Manual Change 97-008, Specification for 44 the Auxiliary Boiler Fuel Oil Storage Tank

7. SPEER-9501473, Check Valve for HRA-128B 45

8. SPEER-9701670, Alternate Replacement Evaluation for instrument 46 Air Compressor 'A' Motor

9. Technical Requirements Manual Change 97-012, Reclassify the HRA 47 .

Containment isolation Valves i

10. Change to Technical Specification Bases 3.6.3, Containment '48 Isolation Valves  !

11. Technical Requirements Manual Change 97-011, Allowable Outage 49 i Time for Certain 'B' Train Equipment i

12. Calculation EC-S96-011, "LOCA Offsite and Control Room 50  !

Radiological Dose Consequences"

13. Change to Technical Specification Bases 3.3.1, Reactor Trip _ 51  !

Breaker Channel Operability )

14. Core Operating Limits Report for Cycle 9, Revision 2 52

15. Technical Requirements Manual Change 98-001, Containment 53 Penetration Conductor Over-Current Protective Devices

16. Change to Technical. Specification Bases 3.7.12, Essential Services 54 Chilled Water System

' 17. Technical Requirements Manual Change 98-004, Fire Protection 55 Surveillance Frequency

, 18. Spent Fuel Pool Criticality Analysis 56 1

19. Technical Requirements Manual Change 98-003, Thermal 57 Overload Devices and Bypass Devices

l Item Page No. No.

20. SPC-97-003-0, HPSI and LPSI Bearing Cooling Water Low Flow Alarm 58

21. SPC-97-008-0, BAMTs LoLo Level Alarm 59

22. SPC-97-013-0, Control Room Air Intake Chlorine Detection 60 F. COMMITMENT CHANGES

1. Component Cooling Water Makeup Commitments 61

2. Performance of Barrier Analysis for All Condition Reports 62

3. Approval of Root Cause Analyses and Corrective Action Plans 63

4. Tracking of Corrective Action Implementation 64

5. Replace Root Cause investigation and Delete Procedure Review 65 as Part of Corrective Action

6. Designation of a Corrective Action Review Board 66

7. Closing Condition Reports to Station Modification Requests 67

8. Preparation and Processing of Purchasing Documents 68

9. Inspection of Air Operated Valves 69

10. Surveillance Testing of the Steam Driven Emergency 70  ;

Feedwater Pump

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11. Position of Chilled Water Valve CHW-823 71

12. Relocation of EDG Starting and Control Circuit Post-Maintenance 72 Testing Requirements

13. INPO Human Performance Program Training 73

14. Bagging Contaminated Valve Lineup Sheets 74

15. Implement the Site-Wide Engineering Request Process 75 v

i Itam Page  !

. ' Nc. No.  !

L l .16. Quarterly Human Performance Awareness Days 76 l

l 17. Maintenance Department Review of CE-001-025 77 L 18. Performance of Root Cause Determinations -78

19. Corrective Action Review Board Review of Root Cause Determinations '9 i

20. Quality Assurance and In-House Events Analysis Verification 80 l

of Corrective Actions l

21. Individual Ownership of Performance improvement Plan Items 81 l

22. Increase in Time for initial Operability Determinations 82 {

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23. Responses to inspection Reports . 83 1

24. -Conduct Training for Trenders and Management 84 )

25. Fire Detection for the 'Q' Deck Area 85

26. - Plant Operations issue: Failure to identify Misaligned Valves 86 and Switches  ;

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27. Performance improvement Plan 87  ;

28. Tracking of Weekly Technical Specification Surveillances 88 j l

29. Reduced Inventory Training 89 l l

30. - Station Modification Request CC-024 90 G.- ENGINEERING REQUESTS

1. ER-W3-97-0043-00, Engineering Review of New CPC Software for 91 Elimination of Unnecessary Trips Due to Dropping CEA #2 or #3

2. ER-W3-97-0286-00-00, Reassignment of CEA #3 From a Four-finger 92

- Subgroup to a Five-finger Subgroup to Prevent "CEA Deviation" Alarms l

with All Rods inserted l

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Item Page No. No.

3. ER _W3-98-0190-00-00, Switchgear Ventilation System Battery Room 93 i Exhaust Fan Differential Pressure

- 4. ER-W3-98-0558-00-00, Replace Valve RWM-1255 with a Ball Valve 94 '

Ill, PROCEDURES ,

A. ' PLANT PROCEDURES

1. PE-004-024, ACCW and CCW System Flow Balance, Revision 0 95

2. CE-002-006, Maintaining Reactor Coolant Chemistry, Revision 9 96

3. UNT-006-021, Pump and Valve Inservice Testing, Revision 3 97

4. RW-TEM-001, Containment Fan Cooler Discharge Tank 98 Processing, Revision 0 ; ,

5. UNT-005-013, Fire Protection Program, Revision 6 99

. 6. RW-TEM-002, Spent Resin Pump Room Resin Recovery, Revision 0 100

7. CE-002-001, Maintaining Steam Generator Chemistry, Revision 13 - 101 B. SPECIAL TEST PROCEDURES (STP)

1. STP-99003536, Acceptance Test for DC-3536 102

2. STP-01160772, HVF H&V Room Exhaust Fan Test with Single 103

- Gravity Damper Failure

3. STP-01160647, SVS Battery Fan Room Test with Single Failure 104

'4, STP-01162204, Functional Test of the DCT Fans ESFAS Start 105 from the Auxiliary Control Room '

5. _ STP-99003492, Acceptance Test for DC-3492, Revision 1 106 l

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SUMMARY

This report provides the Waterford 3 Facility Changes made pursuant to 10CFR50.59(a)(1). The report covers the period from June 1,1997, through May 31, 1998. None of the items in the report were found to involve an unreviewed safety question.

Section I identifies acronyms used in the report.

Section ll of the report identifies 88 Facility Changes which consist of: 6 Design Changes (DCs),14 Condition identification / Work Authorizations (Cl/WAs),3 Temporary Alteration Requests (TARS),9 License Document Change Requests (LDCRs),22 Miscellaneous Evaluations,30 Commitment Changes, and 4 Engineering Requests (ERs).

Section ill of the report identifies 12 Procedure Changes which consist of: 7 Plant Procedures and 5 Special Test Procedures (STPs).

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L 1. LIST OF ACRONYMS-i l

l l ACRONYM DEFINITION .

l -ACCW Auxiliary Component Cooling Water i

AE ' Air Evacuation -

AOT Allowed Outage Time AOV- Air Operated Valve .

ASME American Society of Mechanical Engineers ASTM American Society of Testing and Materials

! ATWS Anticipated Transient Without Scram L BAM' Boric Acid Makeup l BAMT Boric Acid Makeup Tanks i BRGM Broad Range Gas Monitors r

CAR' Containment Atmosphere Release

! CARB- Corrective Action Review Board

CCEF Commitment Change Evaluation Form L CCW ' Component Cooling Water CEA Control Element Assembly CEAC: Control Element Assembly Calculator CEDM- ' Control Element Drive Mechanism l- ._CEDMCS . Control Element Drive Mechanism Control System CFC . Containment Fan Coolers

" cfm Cubic Feet per Minute  ;

L CGCS Combustible Gas Control System i

-CHW Chilled Water

, Cl Condition Identification CIAS Containment isolation Actuation Signal CIV Close Intercept Valve L, COLR~ Core Operating Limits Report COLSS' Core Operating Limits Supervisory System .  !

. -CPC Core Prclection Calculator )

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i CR ' ' Condition' Report CRB Condition Review Board CREFU Control Room Emerg<3ncy Filtration Unit . ~l

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CRG Condition Review Group .

CRS- Control _ Room Superintendent

-CS Containment Spray CSAS ' Containment Spray Actuation Signal CSP Condensate Storage Pool CVAS. Controlled Ventilation Area System  !

CVC Chemical and Volume Control - I DBA Design Basis Accident  !

DBD Design Basis Document DCT Dry Cooling Tower I DEH Digital Electro-Hydraulic Control System '!

DEI Dose Equivalent lodina 1 DP ' Differential Pressure s c ECCS . Emergency Core Cooling System EDG Emergency Diesel Generator-EFW Emergency Feedwater EMI : > Electro-magnetic Interference-EQ Environmental Qualification -

ER- Engineering Request ERF- Energy Redistribution Factor ESFl Engineered Safety Features ESFAS- Engineered Safeguards Features Actuation System F__ Fahrenheit  !

P' FHB Fuel Handling Building FOST- Fuel Oil Storage Tanks  ;

.FPC- Fuel Pool Cooling h FSAR Final Safety Analysis Report n: -- gpm - ' Gallons per Minute

[ H&V 1 Heating and Ventilation -

p l- '

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,-,--.4- . > - 3 .-w-

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HP Health Physics a

HPSI High Pressure Safety injection HRA Hydrogen Recombiner Analyzer HVC Control Room HVAC HVF Fuel Handling Building HVAC HX Heat Exchanger lA . Instrument Air lHEA In-House Events Analysis  !

IST Inservice Testing LBLOCA Large Break Loss of Coolant Accident LCO Limiting Condition for Operation i I

LDCR Licensing Document Change Request LE Licensing Engineer LIR Licensing Information Request LLRT Local Leak Rate Test LOCA Loss of Coolant Accident LOOP ' Loss of Offsite Power LPSI Low Pressure Safety injection LWM Liquid Waste Management MCC Motor Control Center MFLB Main Feedwater Line Break MMIS Material Management Information System MOV Motor Operated Valve MR Millirem MS Main Steam MSLB Main Steam Line Break

- N/E Normal / Emergency NPO Nuclear Plant Operator NPSH Net Positive Suction Head ODCM Offsite Dose Calculation Manual i PASS Post-accident Sampling System

' PCT Peak Clad Temperature i

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PElR Problem Evaluation /Information Request i PIP Performance improvement Program PLHGR Peak Linear Heat Generation Rate PNPO~ Primary Nuclear Plant Operator PO. . Purchase Order- -

ppm Parts per Million PPS Plant Protection System

- psig -- Pounds per Square Inch Gauge PWR Pressurized Water Reactor

-QA Quality Assurance RAB Reactor Auxiliary Building

. RAS- Recirculation Actuation Signal;  ;

RCA- Radiologically Controlled Area - -t RCA Root Cause Analysis RCD ~ Root Cause Determination j

'RCP . Reactor Coolant Pump'  !

RCS- Reactor Coolant System i

'RDB Reload Data Block  !

RPS; - Reactor Protection System RT Repetitive Task  !

RTB- Reactor Trip Breaker RTD Resistance Thermal Detector RWM - Radwaste Management System RWSP. _ Refueling Water Storage Pool .

SAF' L Single Active Failure SDCHX Shut Down Cooling Heat Exchanger SFP.. Spent Fuel Pool

SFSR Spent Fuel Storage Rack .

SG Steam Generator

, SGTR- Steam Generator Tube Rupture SI Safety injection 1

SIAS , Safety injection' Actuation Signal

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r SIT Safety injection Tank SMR Station Modification Request SNPO Secondary Nuclear Plant Operator ,

SPEER Spare Part Equivalency Evaluation Report SPC Setpoint Change SS' Shift Superintendent SSC Structure, System, or Component SSD Safe Shutdown SSF1 Safety System Functional inspection STP Special Test Procedure STS Static Transfer Switch SUPS ~ Static Uninterruptible Power Supply SUT Startup Transformer SVS Cable Vault and Switchgear HVAC TAR Temporary Alteration Request f l- TCCW : Turbine Component Cooling Water TCM Tool Contamination Monitor t TRM Technical Requirements Manual TS Technical Specifications

. UAT. Unit Auxiliary Transformer

.- UHS ' Ultimate Heat Sink USQ Unreviewed Safety Question WA' Work Authorization WCT Wet Cooling Tower wg Water Gauge RHSV Reheat Stop Valve IV Intercept Valve I.. TV Throttle Valve GV Governor Valve OPC Overspeed Protection Control LDA Load Drop Anticipation

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' 11. FACILITY CHANGES A; DESIGN CHANGES

1. . DC-3470. Auxiliary Component Coolina Water System Waterhammer.

Revision 1 DESCRIPTION OF CHANGE

- ACCW Pump 'A' and 'B' control switches will be changed out and replaced with a maintained stop and spring return to Normal start switch. This change will prevent the 4 auto start of an ACCW pump when the control switch is placed in the Stop position. In addition, control room annunciation is added to the ACCW pump circuit to make the operator aware that the pump is unavailable for use.

REASON FOR CHANGE The present circuit configuration for the ACCW pumps allows the possible auto start of an ACCW pump after it has been secured by turning the switch to the Stop position and spring return to Normal position. Preventing an auto start of the ACCW pump when the control switch is in the Stop position is desired during surveillance testing and maintenance. This change will prevent the condition described above from occurring.

SAFETY EVALUATION The ACCW ; system is not considered an initiator of any accident described in the FSAR. It functions to mitigate the consequences of an accident by dissipating heat removed from the reactor and its auxiliaries after an accident. This function will not be changed and no consequences of any accident will be affected. The change does not affect the operation of the pump or any other important-to-safety equipment. No new system connections are required and no new accident modes are created. No protective boundary is affected and no margin of safety is reduced.

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- 2. DC-3472L Hot and Cold Leo RTD Noise Abatement QESCRIPTION OF CHANGE j l'  !

The proposed change modifies the shield grounding scheme for the Hot and Cold Leg l . RTD cables. It also improves cable and shielding configurations subject to EMI and separates power conductors from RTD circuit conductors.

l j L REASON FOR CHANGE 1

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This change is being made to reduce the various anomalies associated with the Hot l

. and Cold Leg RTD circuits. I SAFETY EVALUATION I The safety evaluation has concluded that all of the proposed changes have no adverse impact on safe operation of the plant.- The proposed changes do not modify electrical circuit interconnections that would alter the operation of the plant. The proposed EMI 1 improvements are all passive in nature and do not add any active components that would potentially alter the intended function of the systems. The proposed changes will not cause nor do they affect any accidents described in the FSAR. No margin of safety )

is affected by the changes and there are no USQs associated with these changes.

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3. DC-3502. Reduce Bvoass Leakaae from Penetrations 53 and 65. Revision 1 DESCRIPTION OF CHANGE l Revision 1 of DC-3502 revises the acceptance test to remove valve LRT-2100 and the new welded cap outside containment on penetration 65. It performs the acceptance test for penetration 53 under special test procedure STP-99003502 rather than procedure STA-001-004. It removes the requirement for valve LRT-400 to be locked closed. It also adds a statement about License Amendment 128.

l REASON FOR CHANGE The non-essential instrument lines from Penetrations 53 and 65 communicate directly with the containment atmosphere. Each line routes from containment to a solenoid globe valve that closes automatically on a CIAS. An excess flow check valve is located downstream of the automatic valves. The tubing for these non-essential instrument lines up to and including the excess flow check valves is ASME Section Ill, Class 2, seismic Category 1. The remaining portions of the lines are non-safety and although seismically supported, these lines downstream of the isolation valves are not classified as seismic Category 1. The lines terminate at cabinet C-4, which is located outside the area exhausted by the CVAS. Postulating a single active failure of valve CVR-401A or B and a tube rupture on the non-safety part of the non-essential instrument lines, bypass leakage would be limited by an excess flow check valve, CVR-402A or B.

SAFETY EVALUATION The proposed change enhances plant safety by reducing the potential bypass leakage through penetrations 53 and 65. This proposed modification of Penetrations 53 and 65 non-essential lines with both containment isolation valves CVR-400 (formerly CVR-4018) and CVR-401 (formerly CVR-401 A) outside containment has been evaluated per Licensing Amendment Request NPF-38-181. License Amendment 128 has been granted by the NRC for the proposed modification. According to the safety evaluation, the change in power supplies for valves CVR-400 and CVR-401, the implementation activities associated with DC-3502, and the effect of DC-3502 as it relates to GL 96-06 l do not create a USQ.

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4. DC 3518. Main Condenser Air Evacuation System improvements. Revision 3 i DESCRIPTION OF CHANGE i

The proposed change will increase the seal water level in 'B' and 'C' Air Evacuation moisture separators by raising the overflow discharge float valve approximately 5"; i l install a level switch connected to a solenoid operated condensate make-up valve to j replace the existing float operated make-up valve and maintain the increased level in the separator; install an auto-prime line between the first and second stages of the air evacuation pumps; and replace the 20 amp breakers with 15 amp.

REASON FOR CHANGE l

During acceptance testing, the new vacuum pumps experienced problems when

. started. The separator seal water level dropped to a point where the seal water pump discharge pressure surges and the pump cavitates until level is restored in the separator. The second stage loses its prime. The vacuum pump motor climbs up to 160 amps when it should be at approximately 120 amps. The second stage runs hot and the necessary vacuum cannot be achieved.

SAFETY EVALUATION Loss of Condenser Vacuum is an accident that could be affected by the proposed change; hcwever, the proposed change reduces that chance by increasing the efficiency and capacity of the AE system. SG Tube Rupture could cause the secondary systems to become radioactive; however, the leakage paths would be redirected -)

automatically upon detection of high radiation so the consequences would not be  ;

increased. In addition, the three vacuum pumps are sized to account for the increased . i volume of air that must be removed in the winter versus a lower amount in the summer.

The volume increase is caused by a decrease in condenser pressure in the winter.  ;

Therefore, this change will not change the maximum mass flow of air from the condenser. Because the proposed change more efficiently enables the AE system to i maintain condenser air and non-condensables at a low level, the TCCW system, Main Condenser, and AE system are not adversely affected by the change. No new system interfaces are created by the change. No protective boundaries are affected and no margin of safety is reduced.

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5. DC-3536. Installation of Control Room Emeraency Filter Outside Air intakes DESCRIPTION OF CHANGE This modification will install four backdraft dampers for the S-8 Control Room Emergency Filtration Units, one for each of the four Emergency Outside Air intake Paths.

REASON FOR CHANGE Corrective action document 97-1255 identi'isd that a single failure that results in either valve HVC-213A or HVC-213B failing open may cause both trains of CREFU to become inoperable. The inoperability is due to the inability to positively pressurize the Control Room envelope to 1/8" wg with airflow of less than or equal to 200 cfm.

SAFETY EVALUATION According to the safety evaluation, installation of backdraft dampers will ensure that a single failure will not cause both trains of CREFU to become inoperable and re-establishes original design. The proposed changes will not reduce the margin of safety as defined in the basis of any TS or safety analysis and no USQs are created.

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6. DC-3539. Vortex Breakers for Refuelina Water Storaae Pool and Condensate Storaae Pool. Revision 0 and Revision 1 DESCRIPTION OF CHANGE The proposed change will add vortex breakers at each of the three 6" suction lines from the CSP and two 24" ECCS suction lines from the RWSP. Revision 0 is limited in scope to the installation of vortex breakers on the CSP suction lines. Revision 1 of DC-3539 authorizes installation of vortex breakers in the RWSP.

REASON FOR CHANGE As part of a TS change request, the CSP lower voitex limit was established based upon analysis in support of maintaining an overall pool level that includes instrument uncertainty. Following questions from NRR concerning the vortexing analysis, a scale model test was performed. Without a vortex breaker, the model showed that the calculated level at which vortexing could occur was incorrect. It was determined that the RWSP could also potentially vortex at its low level.

SAFETY EVALUATION According to the safety evaluation, failure of neither the RWSP nor the CSP is postulated to initiate any FSAR Chapter 15 accident. CSP inventory is used as the primary source for EFW and as a makeup source for CCW makeup. RWSP is the primary source of water for the Safeguards pumps, HPSI, LPSI, and CS. Installation of the vortex breakers will increase the reliability of both the RWSP and the CSP to meet these functions. During installation of the vortex breakers, water will be maintained in the RWSP and the Boric Acid Makeup Tanks will be utilized in lieu of the RWSP during shutdown. The CSP is not credited in any accident scenario in Mode 5. Therefore, the consequences of any accident will not be increased. The important-to-safety equipment that could be affected by installation of the vortex breakers is the HPSI pumps, CS pumps, LPSI pumps, Charging pumps, EFW pumps, and CCW makeup pumps. There is no increase in probability of malfunction of these pumps since the vortex breakers are designed safety-related, seismic 1 and the breakers will prevent air entrainment and possible vortexing. The vortex breakers will also not adversely affect system head or NPSH available for these pumps. The vortex breakers will be enclosed in the existing suction screens to prevent ingestion of any debris in the pools.  ;

Therefore, there will be no increase in consequences of malfunction of systems taking suction from the RWSP or CSP as previously analyzed in the FSAR. No new system l interconnections are created and no new accidents are created. Addition of the vortex breakers will ensure the minimum required volume is available in the RWSP. The volume displaced by the breakers (1.2 gallons) will have no adverse affect on available ,

CSP or RWSP inventory. Until a TS Change Request to CSP minimum level is l approved, the CSP level will be maintained at greater than the current TS level by j administrative control. Thus, no margin of safety for the RWSP or the CSP will be 1 adversely affected.

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1 B. CONDITION IDENTIFICATION / WORK AUTHORIZATION (Cl/WA) i

1. Cl-303041/WA-01147691L Essential Chillers DrainsNents Additions l

DESCRIPTION  :

The proposed change replaces the vents and drain plugs on the Essential Chiller Condenser and Evaporator with drain and vent valves.

REASON FOR CHANGE To provide venting and draining control during maintenance activities. 6 SAFETY EVALUATION The CHW and CCW Systems are not considered to be the initiator of any accidents evaluated in the FSAR. There are no accidents whose consequences will be affected -

. by the addition of the drains and vent valves. To preclude any adverse affect on important-to-safety equipment, the new components will be seismically qualified and -

designed to ASME Class 3 requirements. No new system interactions or connections are required and no new failure methods will be created.- No margin of safety for either the CHW or CCW systems is affected by the proposed change. Therefore, there is no USQ associated with this change.

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2. Cl-310358/WA-01160028. Reratino of Component Coolina Water Header to the Containment Fan Coolers Between inlet and Outlet isolation Valves DESCRIPTION The proposed change rerates the CCW system components near the CFC area between the inlet and outlet isolation valves to a design pressure of 200 psig and a design temperature of 263* F.

REASON FOR CHANGE During LOCA conditions, Containment pressure can reach a maximum of 44 psig. The relief valves upstream of the CFCs are set at 121 psig. Considering static head due to elevation difference and margin for relief valve set pressure, some segment of the CCW piping near the CFC can be conservatively subjected to a pressure of 200 psig and a temperature of 263* F when one of the CFCs is isolated concurrently with a LOCA condition.

SAFETY EVALUATION The safety evaluation concludes that no USQ exists as a result of rerating of CCW components near the CFC area. The probability or consequences of an accident or malfunction of equipment important-to-safety are not increased by this change and no margin of safety is reduced. These conclusions are made based on: 1) the uprated pressure and temperature do not exceed the design limits of the components in I accordance with the requirements of ASME Section Ill; 2) the required pipe minimum wall thickness at the proposed rerate condition, as delineated in ASME Section 111, does j not exceed the current pipe wall thickness for the listed material; 3) all components are ,

seismically qualified at the proposed design pressure and temperature; and 4) the coils I for the CFCs are designed for a pressure of 200 psig and a temperature of 300* F.

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3. Cl-310627/WA-01160233. Modify FPC Purification Line 7FS3-8 Supports and Eauipment DESCRIPTION This repair package adds support FSRR-4012, modifies support FSRR-188, removes supports FSRH-186 and FSSH-190, and replaces the flange bolts in the affected line with higher strength bolting.

REASON FOR CHANGE Theso changes are required in order to have the SFP purification pump suction line seismically designed from the SFP to the suction flange of the SFP Purification Pump.

The modifications to the fuel pool purification suction line ensure the fuel pool cooling suction line is not uncovered by crediting manual action to close valve FS-309 during and/or after a seismic event.

SAFETY EVALUATION The results of the safety evaluation conclude that no margin of safety is reduced and no USO exists. There are no accidents previously evaluated in the FSAR that are adversely affected by this change. Calculation EC-P97-018 documents the seismic analysis of the SFP purification pump suction piping line and supports qualification of the SFP suction line to ensure continuous availability of the Fuel Pool Cooling system.

The SFP purification system is not required for nor does it support safe shutdown of the plant. No new system interactions are created by this change. The proposed change rather eliminates a failure mode during and/or after a seismic event. TS 3/4.9.11 states 1 that at least 23 feet of water shall be maintained over the top of irradiated fuel assemblies seated in the storage racks whenever irradiated fuel assemblies are in the SFP. The proposed change preserves this margin by ensuring the availability of valve FS-309 to isolate the purification line prior to draining down to the elevation of the  ;

siphon breakers in order to maintain more water in the SFP during and/or after a seismic event.

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4. Cl-310814/WA-01160476. Use-as-is - Core Protection Calculator Cabinet. CP- '

22. Filter DESCRIPTION The proposed change will allow filter use in t'ne Cabinet CP-22 to continue as-is.

REASON FOR CHANGE An extra polyurethane foam filter began being used in Cabinet CP-22 in 1990. The filter is used to prevent additional dust from entering the cabinet. The concern was that the filter restricted flow through the cabinet and decreased the cooling capacity of tho

- blower. Tests show the additional filter does not decrease the flow through the cabinet.

This change allows use of the filter to continue and adds the filter to necessary design documentation.

SAFETY EVALUATION '

The critical characteristic of the filter is the differential pressure that forms across it and its potential to restrict flow. Tests have been run on the cabinet, both with and without the filter, and the results indicated the filter does not affect flow through the cabinet.

The function of the filter is to keep dust out of the electronics of the CPCs and CEACs.

The existing filter is capable of dust collection and does not adversely affect the CPCs or CEACs. No USQ is created by this change.

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5. Cl-311171/WA-01161072. Resettina of Relief Valves ACC-121 A(B) on the Shell .

Side of the CCW Heat Exchanaers DESCRIPTION The proposed change will reset relief valves ACC-121 A(B) to 95 psig so that ACCW components at lower elevations are protected from thermal overpressurit.ation.  ;

REASON FOR CHANGE

! When the ACCW pump is secured and valve ACC-126A or B is closed, the CCW water in the tube side of the CCW heat exchanger will heat up the ACCW water in the shell side. This causes ACCW system pressure to rise, Relief valves ACC-121 A(B), with a set pressure of 125 psig, provide the thermal relief protection for the CCW heat exchanger. Because of the static head difference, when system pressure at the heat ,

exchanger rises to approximately 100 psig. th> system pressure at the lower elevations will exceed the design pressure of the ACCW system. Therefore, ACC-121 A(B) with ,

! their present setting do not provide thermal overpressure relief protection at lower ,

elevations.

l SAFETY EVALUATION No USQ is created by this setpoint change. No margin of safety as defined in the Basis for any TS is reduced. These conclusions are based on: 1) the setpoint pressure does not exceed the ASME Section !!! design limits of the components; 2) Calculation EC-M94-015 demonstrates that the required relief capacity of the valve at the new set '

pressure is less than the valve capacity; 3) the maximum pressure at the valve is less

- than the maximum blowdown pressure; and 4) the replacement spring is made and recommended by the valve manufacturer and is the same material as the one being replaced. No accidents or important-to-safety equipment are adversely affected by this change.

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6. Cl-310186/WA-01162268. Ucrate Component Coolina Water in the Area of the Drv Coolina Towers L

DESCRIPTION l The prooosed change rerates the design pressure of the CCW system in the area of .

l the DCTs to 150 psig, including piping, valves, instruments, and DCTs.

l- REASON FOR CHANGE  !

L During normal operation at low flows (i.e., below 4000 gpm), the CCW design pressure L of 125 psig may be exceeded. -

SAFETY EVALUATION l

l L The results of the safety evaluation conclude that no USQ is created. The proposed L uprate of CCW piping and components does not increase the probability or  ;

consequences of any accident or equipment malfunction. The uprated pressure does  !

l not_ exceed the design limits of the components in accordance with ASME Section Ill.

1 The pipe minimum wall thickness at the proposed design pressure and current design temperature do not exceed ASME Section lil, NC-3641.1 requirements. All j components are seismically qualified at the proposed design pressure and the DCT '

l cooling coil tubes cre qualified for a design pressure of 155 psig. No protective  ;

L boundaries are affected by this change and no margin of safety is reduced.  ;

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7, . Cl-312040/WA-01162451. Reroute Chiller 'B' Control Power Cable DESCRIPTION Reroute Chiller 'B' control power cable from the SUPS power panel to a safety-related non-SUPS power panel.

REASON FOR CHANGE Due to the current design of the safety-related SUPS, a cable vault / Control Room fire could cause successive shorts of the SUPS, thereby losing the control power feed to Chiller 'B'. Chiller 'B' is required for safe shutdown following a cable vault / Control Room fire. -

SAFETY EVALUATION The proposed change has no impact on any accident initiators. Rather, it ensures that because of the inherent design of SUPS which would result in a shutdown of Chiller 'B',

the Chiller will be available in the event of an Appendix R fire. The only adverse impact of this change is that Chiller 'B' control power will not be available for approximately 10 seconds following an Appendix R fire. This momentary loss of control power presents no adverse impact on any accident or equipment because control power will be restored immediately upon restart of EDG 'B' and the Chiller compressor is not sequenced on to the EDG until approximately 168 seconds. No protective boundaries are impacted by this change and no margin of safety is reduced.

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8. Cl-311992/WA-01162852. Add Fire Detection in Walkway Between Fuel Handlina Buildina and Reactor Auxiliary Buildina at Elevation +21' r DESCRIPTION The proposed change adds thermal fire detectors to the walkway and equipment removal area between the FHB and RAB at elevation +21'. Two conduit seals will be .

. impaired to allow rework of a spare conduit. The seals will be reinstalled after the conduit is reworked and cables are pulled.

REASON FOR CHANGE The' existing cable configuration is such that a fire involving opposite train non-safety shutdown cables could potentially cr.use a loss of function of SUPS units.

SAFETY EVALUATION The proposed change does not impact any accidents or important-to-safety equipment.

The installation has been evaluated for seismic concerns. There are no seismic II/I problems so no new accident or equipment malfunction is created. No protective boundaries are affected and no margin of safety is reduced.

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9. Cl-312316/WA-01162893. Accendix R Liahtina Uoorade l

pESCRIPTION Inadequate SSD lighting for normal access to the chillers was identified. A walkdown of all requi, red SSD equipment identified several other areas where additional Appendix R lighting should be provided.

REASON FOR CHANGE '

Bring the plant into compliance with the FSAR and Appendix R requirements for SSD emergency lighting.

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There are no accidents whose probability or consequences are affected by this addition j of Appendix R lighting. The changs is confined to existing normal / emergency (N/E) l~

' lighting circuits that are supplied from the EDGs via the Class 1E motor control centers.

l isolation has been provided by the double breaker scheme in accordance with RG 1.75 l

requirements. There are negligible changes to EDG fuel oil consumption (which are accounted for in previous conservatisms for N/E lighting) and heatloed to both the 'A' and 'B' electrical switchgear rooms (accounted for by calculation). No new system interconnections are required, no new failure methods are created, and no important-to-safety equipment is affected. No protective boundaries are affected and no margin of safety reduced by this change.

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10. Cl-312692/WA-01163755 and Cl-312691/WA-01163756.1/2" Socket Weld Modification and Shortenina 1/2" Vent Line Tubino, Revision 0 and Revision 1 1

DESCRIPTION The proposed change will: a) shorten the 1/2" tubing connection between the 1 1/2" orifice flange and the 1/2" Dragon valve; b) shorten the 1/2" tubing connection between the Dragon valve and the capped end; c) use 1/2" tubing with higher wall thickness; d) add a support to the 1-1/2" tubing; and, e) strengthen the welding of the 1/2" tubing to the orifice flange, and the 0.120" tubing to the valve to prevent high frequency fatigue failure due to the Charging pump vibrations. The new weld configuration consists of the increased fillet weld leg size in the axial direction to 1/4" as suggested by the EPRI report TR-107455.

REASON FOR CHANGE A laboratory test result concluded that the identified crack was the result of high frequency fatigue failure. A fatigue crack may have initiated at the weld root defect (lack of fusica) propagated by low stress high cycle fatigue. The intent of this change is to strengthen the welded connection between the orifice flange and the 1/2" vent tubing per EPRI report TR-107455 to reduce high frequency fatigue failure, caused by the Charging pump vibrations. Design engineering decided to shorten the 1/2" tubing connection between the orifice flange connection and the Dragon valve and between the Dragon valve and the capped end to achieve minimum cantilever length. Design engineering further recommended to use higher wall thickness tubing (0.120" Vs.0.083") from the valve to the flange and using the same increased weld (axial) size to the valve end also. The addition of the tubing clamp support and increased wall  ;

thickness will also stabilize the vent line vibrations.

SAFETY EVALUATION 4 The proposed socket weld modification, the shortening of the existing 1/2" tubing length  ;

and increased weld (axial) size, and increased wall thickness do not have any impact on the function of the vent line. This is an enhancement to the present socket welded configuration and the tubing cantilever length of the vent line. The addition of the tubing clamp support will also stabilize the vent line vibrations. These vent lines provide added assurance that during venting activities (after maintenance) all the air is  ;

effectively removed from the Charging Pump and the associated piping. This safety '

evaluation concludes that the changes proposed by this WA repair package will not reduce the margin of safety as defined in the basis of any TS or safety analysis and no USQs are created.

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11.' Cl-313735/WA-01165734. HRA Containment Isolation Valve Position Indication DESCRIPTION The proposed change will re-wire the limit switches of containment isolation valves HRA-109A, HRA-110A, and HRA-126A to provide positive indication of valve status to the control room.

REASON FOR CHANGE The current scheme where the three containment isolation valves are wired in series to the same set of position indication lights does not adequately inform the control room operator of valve position (s).

SAFETY EVALUATION The HRA system is not the initiator of any accident described in the FSAR and thus t cannot increase the probability of accident occurrence or consequences. Since this is a change to position indication only, it will not impact the ability of the containment  ;

isolation valves to perform their safety function. No new system interconnections are l required for this change and no new accidents or equipment malfunctions are created.  !

No protective boundary is affected by this change and no USQ is created. ,

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12. Cl-313736/WA-01165735. HRA Containment Isolation Valve Position Indication l

DESCRIPTION The proposed change will re-wire the limit switches of containment isolation valves HRA-1098, HRA-1108, and HRA-126B to provide positive indication of valve status to

i. the control room.

. REASON FOR CHANGE l

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L . The current scheme where the three containment isolation valves are wired in series to the same set of position indication lights does not adequately inform the control room l operator of valve position (s).

l SAFETY EVALUATION - 1 I

The HRA system is not the initiator of any accident described in the FSAR and thus j cannot increase the probability of accident occurrence er consequences. Since this is  ;

l a change to position _ indication only, it will not impact the ability of the containment l

L isolation valves to perform their safety function. No new system interconnections are l required for this change and no new accidents or equipment malfunctions are created. )

No protective boundary is affected by this change and no USQ is created.  ;

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l De na u of CAR 0B DESCRIPTION l This repair will add a normally open contact from the CIAS relay in series with the l opening circuit of valve CAR-2018 to ensure the valve closes with a CIAS signal. This l repair also removes the CIAS override signal to valve CAR-2018.

l REASON FOR CHANGE'

[- With the current configuration, a single failure of the opening relay may prevent valve ,

l CAR-201B from closing on a CIAS signal.

i' SAFETY EVALUATION l

The evaluation has determined there is no USQ associated with this change. There p are no accidents that are initiated by this valve and adding the contact will ensure the

! valve closes in the event of a' failed opening relay. The ability of this isolation valve to ,

maintain containment integrity during and post-accident will not be affected and no >

accident consequences will be increased. Neither will any equipment malfunction be l affected by this change. No new failure modes or system interactions will be created l by this change and no margin of safety will be reduced.

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14. Cl-314004/WA-01167913. Addition of a CIAS Contact in Series with the Openino Circuit of CAR-201 A DESCRIPTION This repair will add a normally open contact from the CIAS relay in series with the opening circuit of valve CAR-201A to ensure the valve closes with a CIAS signal. This repair also removes the CIAS override signal to valve CAR-201 A.

REASON FOR CHANGE With the current configuration, a single failure of the opening relay may prevent valve CAR-201 A from closing on a CIAS signal.

SAFETY EVALUATION The evaluation has determined there is no USQ associated with this change. There are no accidents that are initiated by this valve and adding the contact will ensure the valve closes in the event of a failed opening relay. The ability of this isolation valve to maintain containment integrity during and post-accident will not be affected and no accident consequences will be increared. Neither will any equipment malfunction be affected by this change. No new failure modes or system interactions will be created by this change and no margin of safety will be reduced.

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C. TEMPORARY ALTERATION REQUEST (TAR)

1. TAR-97-016. Startup Transformer DESCRIPTION The proposed change replaces the existing SUT 'B'. The replacement meets plant characteristics except the 4.16 kV winding is in a " wye" configuration. This will necessitate not utilizing the transfer scheme between the UAT and the SUT. The 'B' train will be fed from the SUT.

REASON FOR CHANGE The existing SUT 'B' failed internally.

SAFETY EVALUATION The replacement will affect all accidents that involve a LOOP. Analysis has shown that both the existing and replacement SUT provide the required bus voltage and limit the fault current to within equipment ratings for the 3B1 bus and the 382 bus, which powers the safety buses. In addition, the failure modes for both transformers are identical.

Therefore, there will be no increase in accident probability or cmaequences. The primary differences between the replacement and oxisting transformers are: 1) 4160v secondary winding is " wye" for the replacement and " delta" for the existing; 2) impedance values are higher for replacement windings; 3) voltage rating of replacement windings is lower; and 4) replacement transformer uses an "under oil" sudden pressure relay rather than a " gas type" for the existing. These changes were evaluated and it was determined the replacement transformer will not increase the probability of occurrence of an important-to-safety equipment malfunction. Installation of the replacement transformer maintains the TS required physically independent circuits; therefore, there is no increase in the consequences of an equipment malfunction. The following action.s were taken to ensure the differences between the transformers previously identified will not create the possibility of a different accident type than previously evaluated in the FSAR: 1) evaluation to verify Waterford 3 will operate as designed and licensed while being fed from the new transformer; 2) arrangement of the 382 switchgear current transformers to correctly interface with the SUT 'B' current; 3) relay settings changed to maintain system protection; 4) bus ducts temporarily redesigned to be adaptable to replacement transformer bushings; 5) detectors and deluge nozzles repositioned to provide fire protection; 6) operations and i maintenance procedures revised to reflect differences between replacement and

- existing transformers; 7) drawings created or revised to account for differences. The l only new credible failure mode for the 'B' train AC safety busses is damage which could

result from being cross-connected to the 'A' train components which will be electrically 1 out of phase. . This will be prevented by

a) cross connect breakers are electrically

- interlocked; b) procedure OP-006-001 requires dead bus transfer; c) caution tags will be placed on all affected breakers; d) automatic fast dead transfer and the manual hot-i bus transfer between the UAT 'B' and the SUT 'B' will be defeated by operating without  ;

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,. . the UAT 'B' breakers until a comparable SUT 'B' is installed. No protective boundaries l f

i are affected and no margin of safety is reduced by this change. l i

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2. TAR-97-018. Disable Diversion of Vacuum Pumo Exhaust Header to RAB Normal Ventilation DESCRIPTION The proposed change disables the diversion of the vacuum pump exhaust header to i

.the RAB normal ventilation ductwork. This eliminates a problem in which, during a SIAS, the RAB normal ventilation fans would be secured and the radioactive noncondensable gases and vapor mixture from the Condenser Vacuum Pumps would be pumped throughout the RAB ductwork.

REASON FOR CHANGE -

The design of the RAB normal ventilation is such that the E-22 ventilation fans will be secured upon a SIAS. The condenser vacuum pumps will then pump potentially radioactive gases and vapor throughout the RAB non-safety ductwork when a SlAS I occurs. This temporary alteration will prevent potentially radioactive main condenser -

evacuation system gases from being pumped throughout the RAB, SAFETY EVALUATION '

Rerouting condenser off-gas so that it cannot go to the plant stack for Cycle 9 neither ,

causes events previously analyzed to exceed off-site dose limits, nor causes the  !

amount of activity released to exceed the amount given by the NRC as acceptable.

There is no USQ associated with this change. 4 1

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3. ' TAR-97-004. Addition of Portable .1 Exchanae Vessels to the ACCW Filtration Skids. Revision 1 DESCRIPTION The proposed change involves placement of portable ion exchange units on the ACCW "A" and ACCW "B" filtration / chemical addition skids.

REASON FOR CHANGE Zinc must be removed to less than 1.0 ppm prior to discharge of basins to the Circulating Water system.

SAFETY EVALUATION The safety evaluation concludes that the addition of temporary ion exchange vessels to each ACCW basin will not reduce the level of performance of the ACCW system. The ACCW system will be able to perform its design function as described in the FSAR, during normal operation, and during design basis accidents. Mitigation of a MSLB or LOCA will not be adversely affected. No accident will be caused by addition of this temporary alteration. The ion exchangers are secured and will not pose a civil threat.

No USQ is created.

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l D. LICENSING DOCUMENT CHANGE REQUESTS (LDCR)

1. LDCR-97-0114. Revises FSAR Section 3.9.

1.2 DESCRIPTION

Revises FSAR Section 3.9.1.2 to create new Sections 3.9.1.2.1.16 and 3.9.1.2.1.17 to add to the list of engineering computer software programs that are used to perform safety-related calculations. Corrects Section 3.9.1.2.1.15 that was incorrectly labeled 3.9.1.2.15.

REASON FOR CHANGE ,

NUREG-0800, Section 3.9.1, requires a description and verification of all computer programs used in Seismic I, Code and non-Code analyses.

SAFETY EVALUATION According to the safety evaluation, the proposed change does not reduce the margin of safety as defined in the Basis for any TS or safety analysis and does not create a USQ.

All the computer software programs added to the FSAR were maintained under an approved QA program. These QA programs were based on guidelines outlined by the nuclear industry to meet the necessary requirements for use in safety-related calculations. Validation / verification was performed initially by the developer and again upon installation.

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2.~ LDCR-97-0194. Revises FSAR Sections 3.1.37. 6.0. 6.2.5. 6.5.3.1. 7.3.1.1.9.3.

and 15.6.3.3.

5.3 DESCRIPTION

r The proposed change revises FSAR Sections 3.1.37,6.0,6.2.5,6.5.3.1,7.3.1.1.9.3, and 15.6.3.3.5.3 to clarify the design and licensing basis of the CAR system with respect to post-LOCA hydrogen control.

REASON FOR CHANGE l

This resolves a design basis discrepancy between W3 DBD-005, ' Containment Gas Control and Measurement', and FSAR Section 6.2, which incorrectly states CARS is a backup to the Hydrogen Recombiner system. This statement implies that CARS is capable of providing a method for hydrogen control through purging containment at i post-LOCA high-pressure conditions. The DBD states that CARS can only be operated when containment pressure is +2" wg (0.072 psig) or less.

SAFETY EVALUATION The CGCS detects and limits hydrogen accumulation to safe levels following a LOCA but does not initiate any accidents nor increase the probability of them occurring. Each j hydrogen recombiner is 100% redundant and there is no requirement for CARS to act as a backup. Therefore, removing the description of CARS as a backup will not increase the consequences of any accident described in the FSAR. No equipment modification is required, no new method of operation of the CGCS is required, and no new interactions are required for this change. Therefore, no important-to-safety l equipment is affected. The change does not affect any protective boundary nor does it l reduce any margin of safety. l l

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3. LDCR-97-0205. Revises FSAR Sections 9.2.2.2.1 and 92.8.2 DESCRIPTION

, PASS is designed with two sample coolers (cooled by CCW) to reduce RCS sample l temperature prior to collection and analysis. Due to an operability concern with CCW, F

operation of PASS was changed to allow operation with only one sample cocler. The proposed change revises FSAR Sections 9.2.2.2.1 and 9.3.8.2 to reflect this change in i PASS operation.

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The original design basis of PASS _was based on operating two sample coolers. A calculation provided the failure modes and effects analysis for this mode of operation i because the non-nuclear safety portion of CCW was aligned to the safety class portion. l However, the calculation was based on an incorrect input that there is a 15% margin in l the CSP in the event of a line break. Since operating with two coolers could no longer - .

I be justified, system operation was changed to allow operation with one sample cooler.

l SAFETY EVALUATION l

PASS is a non-safety' system designed to provide a means for sampling and analyzing reactor coolant following an accident to determine the extent of core damage. The proposed change allows PASS to be operated with only one sample cooler instead of l two. Since PASS is not required to meet single failure criteria, this is not a single l failure issue. Since sample temperature may be increased, a calculation was l performed which determined the_ increase does not impact PASS and its sampling .

. capability, in addition, the sample temperature does not initiate or contribute to any i accident. Demand on CCW Train 'A' is not changed, therefore there is no impact on Train 'A' temperature and flow. Demand on Train 'B' is diminished, as this train is isolated from PASS by a safety-related isolation valve and a check valve. The overall impact of this reduction on CCW is insignificant and in the conservative direction.

Therefore, no analysis is required to evaluate the change and there is no impact on the operation of important-to-safety equipment. There are no margins of safety associated with operation of PASS.-

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4. LDCR-97-0230. Revises FSAR Table 9.2-8 l DESCRIPTION l

The proposed change revises the DCT tube design pressure from 125 psig to 155 psig.

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REASON FOR CHANGE

, Corrective action document 96-1553 identified a condition where the DCT design l pressure of 125 psig may be exceeded when continuously operating the CCW system i at low flows (i.e., below 4000 gpm). It is possible that during normal operation it may become necessary to reduce CCW flow below 4000 gpm which may result ir operating the DCT above design pressure for extended periods of time. The original design pressure of 125 psig was not a limiting condition, but selected based on assumed operating conditions. ,

SAFETY EVALUATION The safety evaluation concludes that revising FSAR Table 9.2-8 to reflect the new DCT I design pressure of155 psig does not result in any USQs. The revised ASME Code Data Report demonstrates that all CCW pressure retaining portions of the DCT meet their original design requirements in accordance with ASME Code Section Ill,1974  ;

Edition through Summer 74 addenda. LDCR-97-0230 will not reduce the margin of 1

safety nor prevent the DCT from performing its safety function.

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5. LDCR-97-0234. Revises FSAR Section 6.2.1.5 and Fioure 6.2-30A DESCRIPTION Revision of FSAR Section 6.2.1.5, Fan Cooler Heat Removal Capacity, to reflect change in CFC capacity assumed in the analysis. Revise Figure 6.2-30A to reflect the increased fan cooler capacity used in the revised analysis.

REASON FOR CHANGE The core reflood rate during a LBLOCA is a function of containment pressure.

Lowering containment pressure decreases the core reflood rate. Following performance of the Waterford 3 Cycle 9 LBLOCA ECCS analysis, the CFC capacity and LPSI pump flow assumed in the analysis were found to be non-conservative. The impact of increased CFC capacity and LPSI pump flow on the LBLOCA ECCS performance analysis is to lower containment pressure and decrease the core reflood rate and thus increase the predicted severity of the accident. Calculation EC-S97-013 documents the evaluation of the impact of the increase in CFC and LPSI pump capacity on the LBLOCA ECCS performance analysis for Cycle 9.

SAFETY EVALUATION There are no USQs associated with the increases in CFC and LPSI pump capacity in the ECCS performance analysis for Waterford 3 Cycle 9 or with the proposed changes to the FSAR. The CFC and LPSI pumps act to mitigate the consequences of a LBLOCA but do not initiate any accidents. The change in CFC and LPSI pump capacity decreases containment pressure which could reduce the core reflood rate and affect the severity of the LBLOCA accident. However, a reduction in an over'y conservative assumed containment external heat transfer coefficient offsets the increase in capacity so the severity is not increased. The proposed change does not alter the operation of the CFC or LPSI pumps or the operation of any other equipment important-to-safety and no new system interconnections are required. No margin of safety is affected by this change.

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6. [DCR-97-0243. Revise FSAR Section 6.2.1.5 L DESCRIPTION -

! Calculation EC-S97-013, Revision 1, evaluates the impact of an increase in CFC capacity, including airflow of 45,000 cfm, on the ECCS performance minimum l containment pressure calculation in FSAR Section 6.2.1.5. FSAR Figure 6.2-30A1,

" Updated Fan Cooler Heat Removal Capacity," is added to reflect this change in CFC capacity for Cycle 9 and all subsequent cycle analyses. i 1

REASON FOR CHANGE i

The core reflood rate during a LBLOCA is a function of containment pressure. l Lowering containment pressure decreases the core reflood rate. Following performance of the Waterford 3 Cycle 9 LBLOCA ECCS analysis, the CFC capacity -

assumed in the analysis was found to be non-conservative. The impact of increased CFC capacity on the LBLOCA ECCS performance analysis is to lower containment pressure and decrease the core reflood rate and thus increase the predicted severity of the accident. Calculation EC-S97-013, Revision 1, documents ABB-CEs evaluation of i the impact of the increase in CFC, inc;uding a higher airflow than design, on the '

LBLOCA ECCS performance analysis for Cycle 9. Revision 0 of the calculation allows. I for a reduction in the assumed containment total external heat transfer coefficient to offset an increase in CFC heat removal due to a lower CCW temperature. Revision 1 uses this new heat transfer coefficient and determines the PCT. The new PCT is 2165* ,

F, as opposed to 2170 F, and maximum local cladding oxidation is reduced frorn I 8.39% to 8.24%. Therefore, the ABB-CE calculation analysis of record remains bounding.

! SAFETY EVALUATION

There are no USQs associated with the increase in CFC capacity in the ECCS l performance analysis for Waterford 3 Cycle 9 or with the proposed changes to the l UFSAR. ' The CFCs act to mitigate the consequences of a LBLOCA but do not initiate l any accidents. The change in CFC capacity in the LBLOCA ECCS performance analysis decreases containment pressure which reduces the core reflood rate and by itself would increase the severity of the LBLOCA accident. However, a reduction in assumed containment external heat transfer coefficient offsets the increase in CFC capacity so the severity is not increased. The proposed change does not alter the

,. - physical structure or operation of the CFC or the operation of any other equipment

! important-to-safety and no new system interconnections are required. Therefore, no

. margin of safety is affected.

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I l 7.' LDCR-97-0249. Revises FSAR Section 9.2.5 and Table 9.2-9 >

l L DESCRIPTION This change revises FSAR Section 9.2.5 to remove the discussion of once-through cooling during a tornado event and revises Table 9.2-9 to update the tornado heat dissipation reqyirements. -

REASON FOR CHANGE The water consumption documented in calculation MN(Q)-9-17 did not include the l automatic makeup to the CCW Surge Tank that may be required during a tornado 1 l event. In addition, the calculation did not include the non-essential heat load from the

. SFP that will be restored at approximately 11 hours1.273148e-4 days <br />0.00306 hours <br />1.818783e-5 weeks <br />4.1855e-6 months <br /> into the event.'  ;

! SAFETY EVALUATION l ' The UHS, which consists of the DCTs, the WCTs, and the water stored in the WCT L basins,'is not an initiator of any accident previously evaluated in the FSAR. In addition, l- analysis has shown that these changes _will not affect the ability of the UHS to fulfill its

[. heat dissipation capability. Thus no accident probability or consequences are affected l l - by the change. The calculation has also shown that the amount of water required ,

l following a tornado event is available with margin. In addition, the inventory from CCW using gravity drain paths can be used with installed piping and valves. No other replenishment equipment is necessary Any postulated malfunction of the UHS or the i components cooled by UHS will result in the same consequences as currently i evaluated Thus, no important-to-safety equipment or the consequences of its failure l are affected by this change. No new system interconnections are required for this ,

change and the failure modes associated with the affected equipment remain i unchanged. No margin of safety is reduced by the proposed change and no USQs are created.

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8. LDCR-97-0250. Revises FSAR Section 9.3.

1.2 DESCRIPTION

Revises FSAR Section 9.3.1.2 to accurctely reflect that not all safety-related valves have air filters installed as part of their topworks.

REASON FOR CHANGE The FSAR states that safety-related valves that are air operated are equipped with air filters. There are currently 41 safety-related valves which do not have air filters installed for their actuators. Valves EFW-223A(B), EFW-224A(B), and MS-116A(B) have air filters installed but their associated volume boosters do not have an air filter in-line. The affected safety-related valves without air filters are in non-compliance with the FSAR.

SAFETY EVALUATION The safety-related valves without air filters are supplied with high quality instrument air which meets or exceeds the valve actuator suppliers minimum air quality standards.

Plant operation with safety-related valves that do not have point-of-use filters installed at their actuators does not reduce the margin of safety assumed in the plant safety analysis. There are no accidents whose probability or consequences are affected by valves that lack air filters. Since the valves without filters are supplied with air that exceeds vendor quality standards, no important-to-safety equipment is affected by this change. No new interconnections are required by this change, thus no new accidents or malfunctions are created.

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9. LDCR-97-0260. Revises FSAR Section 3.8.

3.8 DESCRIPTION

The proposed change revises FSAR Section 3.8.3.8 for the type of grout from ASTM C476 to ASTM C270.

REASON FOR CHANGE ASTM C476, which is the standard specification for grout for masonry covers two types of grout, fine and coarse, for use in construction of masonry structures. The correct specification is ASTM C270, mortar for unit masonry, which contains a table for compressive strength of cubes for mortar types.

SAFETY EVALUATION This is a change to the specification for mortar identified in the FSAR. It does not affect any accidents or equipment important-to-safety described in the FSAR. No new system interactions or connections are required and no new accident modes are created. The change does not affect any margin of safety.

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E. MISCELLANEOUS EVALUATIONS

1. Technical Reauirements Manual 3/4.6.1.5 Minimum Containment Air Temperature DESCRIPTION The proposed change reduces the minimum containment air temperature from 100* F to 90* F.

REASON FOR CHANGE To validate the Analysis of Record for periods when containment temperature drops below 100* F but remains above 90* F.

SAFETY EVALUATION Containment air temperature does not cause any accident and the reduction in containment temperature does not impact any accident precursors. Evaluation of the LBLOCA determined that no increase in consequences of an accident is expected due to a penalty factor applied to the COLSS calculated PLHGR. No important-to-safety equipment is affected and no new system interactions or connections are created.

' Therefore, the possibility or consequences of an equipment malfunction are not increased. PCT, maximum cladding oxidation, and core-wide oxidation remain bounded by the analysis of record and within acceptance limits; therefore, no margin of safety is reduced.

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2. Calculation EC-M89-004. Water Levels inside Containment (Post-LOCAL Revision 3 I

DESCRIPTION This calculation determines the maximum RWSP level so that the maximum post-LOCA flood level would be below the elevation of required equipment and instruments. Post-l LOCA water sources include the RCS, SITS, all of the RWSP water, and all of the contents of the BAMTs.

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' REASON FOR CHANGE l l

If the RWSP is filled to the 100% level, the post-LOCA flood level would be such that l the Steam Generator level transmitter GG-ILT-1115A would be submerged and part of i

the CFCs AH-13B SB and AH-13D-SB would be submerged. Therefore, a limit on  !

maximum RWSP level is required. 1 1

SAFETY EVALUATION i The RSWP does not act as the initiator for any accident described in the FSAR. It is required to deliver water to the RCS following a SIAS and to the Containment following a CSAS. By calculation, all Modes 1,2, and 3 required instrumentation remains unsubmerged.- For Mode 4 only Steam Generator level transmitter SG-ILT-1115A is slightly submerged. . For all modes, the lowest CFC coils are slightly submerged.

l However, there is no affect on the consequences of any accident or on any important-to-safety equipment as a result of these occurrences. No new equipment connections are created and no protective boundaries are affected by this change.

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3. Calculation EC-E90-006. EDG Loadina and Fuel Oil Consumption. Revision 2.

Chanaes 12 and 13 DESCRIPTION Calculation EC-E90-006, " Emergency Diesel Generator Loading and Fuel Oil Consumption", establishes the design basis EDG loading and fuel oil consumption following design basis accidents with a LOOP. It establishes the basis for determining that the TS minimum volume of fuel oil is adequate for 7 days of operation with a 10%

margin as provided in the requirements and regulatory positions of RG 1.137 and ANSI N195. The following changes to this calculation are addressed in this safety evaluation: 1) conversion of program for calculation from FoxPro 2.0 to MS Access 2.0 by BCP Technical Services; 2) fuel oil consumption during normal testing of the EDGs was revised by Technical Manual 447001225 R57, Volume 1, Tab 3.15 and OP-009-002, Revision 16, and OP-903-068, Revision 12; 3) addition of motor control center MCC 3AB313 feeder cable losses; 4) loading of lighting panels LP-333 and LP-334.

REASON FOR CHANGE Previous revisions of calculation EC-E90-006 and the FSAR indicated the Fuel Oil Storage System did not meet all the provisions and requirements of RG 1.137 and ANSI N195. In addition, the NRC identified that Waterford 3 had not conservatively demonstrated fulfillment of the commitment concerning EDG fuel oil storage. In addition, the safety evaluation for an FSAR change did not adequately address the 10% reduction in the required fuel oil storage capacity from that discussed in the TS Bases, which may have constituted a USQ.

SAFETY EVALUATION The safety evaluation determined that no USQ exists due to the changes made to this calculation. The safety evaluation demonstrates the EDG fuel oil storage system capacity is adequate for 7 days of operation following a limiting design basis accident, including ESF loads, with a 10% margin as required by RG 1.137 and ANSI N195. The EDGs and EDG support systems are not the initiators of any accidents, nor will the i changes initiate any accidents. The consequences of an accident may have been increased if one EDG is not available for the full 7 days. However, the changes made ensure the availability of the standby power for 7 days with the required 10% margin.

The probability or consequences of equipment malfunction are not affected because operation of the ESF equipment and support systems is unchanged. No new actions are required by this change. Cross-tying of the EDG fuel oil storage tanks, transferring fuel oil between the EDGs, and replenishing EDG fuel oil following an accident are the interactions or connections of interest. These actions do not introduce any new manual or immediate actions and are performed using approved procedures or direction from the Technical Support Center. Therefore, the possibility of a malfunction of a different type is not created. Calculation EC-E90-006 has demonstrated that for all accidents concurrent with a LOOP, the margin of safety is maintained.

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4. Calculation EC-S97-016. Emeraency Feedwater Minimum Flow Reauirement DESCRIPTION This calculation establishes the minimum EFW flow requirement as 575 gpm at 1102 psig versus the current flow of 700 gpm at 1163 psig. The affected FSAR sections and TS 3/4.7.1.2 Bases will be revised to incorporate the EFW operating point defined therein. This calculation is not being performed to replace the Chapter 15 analysis, but rather to provide a more realistic yet conservative plant behavior under accident conditions.

REASON FOR CHANGE The turbine-driven EFW pump or the two motor-driven EFW pumps must be capable of providing 100% of EFW system flow. The equipment in the plant may not support the current minimum EFW flow requirement of 700 gpm. New analyses have been performed to determine a revised minimum required EFW flow rate.

SAFETY EVALUATION This safety evaluation concludes that the proposed reduction in required EFW flow rate will not create a USQ. Analyses show that the EFW system will continue to meet its design function with reduced flow. The EFW turbine-driven pump or the two motor-driven pumps will continue to provide well above 100% of required EFW flow for the limiting EFW demand event (Feedwater line break).

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5. Calculation EC-192-019. Plant Protection System Setooint Uncertainty Calculation. Revision 1. Chanae 1 DESCRIPTION This change justifies using an assumed RWSP Low Level RAS analytical limit of 7.0%,

which is conservative with respect to the RAS level required to prevent vortexing (5%).

REASON FOR CHANGE The current RWSP Low Level RAS analytical limit of 7.43% (SER dated 9/5/95) was not based on any analysis. It was based on subtracting the total loop uncertainty from the trip value of 10%. This limit was established to be consistent with other PPS setpoints that have an analytical limit. However, other PPS setpoint analytical limits are based on parameter values used in specific analyses. In July 1997, the level above which vortexing will not occur was established at 5% RWSP level. This calculation is revised to change the RWSP Low Level RAS analytical limit to a limit based on the parameter's (RSWP level) use in a specific analysis.

SAFETY EVALUATION The proposed change does not create a USQ. The RWSP Low Level RAS trip setpoint and allowable value are not affected by this change, and the change is consistent with the methodology stated in the Bases section for TS 3/4.3.2. These values ensure that the RWSP Low Level RAS signal will be initiated prior to the RWSP going below the analytical limit. The revised analytical limit for RWSP Low Level RAS (7%) is conservative with respect to the RWSP level (5%) required to ensure the acceptance criteria of the ECCS Performance Analysis for Peak Clad Temperature are met.

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I l 6. Technical Reauirements Manual Chanae 97-008. Specification for the Auxiliary ,

i Boiler Fuel Oil Storace Tank '

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! DESCRIPTION  :

The proposed change adds TRM Specification 3/4.8.1 to provide additional fuel oil in  !'

order to satisfy the requirements of ANSI N195 to have 10% margin when the time dependent load method is used to determine the minimum amount of fuel oil stored  :

onsite. I ne proposed specification requires the Auxiliary Boiler FOST to be Operable.

L Operability of the tank is ensured by: 1) verification of the level once per 31 days; 2) i l- verification of transfer equipment availability once per 92 days; and 3) verification of i l fuel oil chemistry consistent with TS surveillance requirements for the EDG FOSTs. If l

.these requirements cannot la met, Actions require: 1) restore limit; 2) justification for continued operation with an out-of-limit parameter; 3) verification that sufficient fuel oil  ;

is onsite and available ta . 'ee' ANSI N195; or 4) action initiated to obtain sufficient fuel oil to meet ANSI N195 e prements.

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REASON FOR CHANGE Waterford 3 cannot currently meet the ANSI N195 requirement that fuel oil stored on L site have a 10% margin above the time dependent load calculation.

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l SAFETY EVALUATION l

The addition of this TRM specification does not create a USQ. This change adds a l

new requirement to the TRM to establish contrc!s on the Auxiliary Boiler FOST to ensure the design requirement for fuel oil stored on site meets ANSI N195 (10% ,

margin). The existing TS for the EDG FOST volume is sufficient to operate the EDGs for seven days assuming the worse case accident (LOCA with LOOP) but with only a 1% margin. This change increases that margin to 10%.

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7. SPEER-9501473. Check Valve for HRA-128B DESCRIPTION The proposed change provides a replacement check valve for component HRA-128B.

This requires a change in the manufacturer's name in FSAR Table 3.9-9 and the material in Table 6.2-33.

REASON FOR CHANGE This change is required because this is a containment isolation valve, and the existing valve did not pass the required LLRT.

SAFETY EVALUATION The change does not affect any accident in the FSAR; therefore, no accident probabilities or consequences are affected. The valve has been evaluated to be equivalent to the original. The change does not affect the operation of the hydrogen analyzer and the valve is required to meet Type C testing requirements for containment isolation valves. No new system interactions or connections are required by this i change. Evaluation of equivalence and Type C testing ensure the valve is capable of l

- maintaining containment isolation. No margin of safety is impacted by this change.

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l 8. SPEER-9701670. Alternate Replacement Evaluation for Instrument Air Comoressor 'A' Motor l DESCRIPTION Alternate replacement for the Instrument Air Compressor 'A' motor.

REASON FOR CHANGE The original motor is obsolete.

l SAFETY EVALUATION This item is considered non-safety, non-quality, non-seismic that will be installed in a non-EQ environment. There are no mechanical components affected by this change.

The electrical components affected by the change are the Engineered Safety Features Bus and the Plant Auxiliary Distribution system. All existing circuit logic / interlocks / trips

• will remain as they are. A sligt.t increase in load on the 3A31-S bus has been l evaluated and determined to be acceptable. The affect on the EDG has been i evaluated and determined to be acceptable. Complete loss of lA during full power

operation or under accident conditions does not reduce the ability of the RPS or ESF and their supporting systems to safety shutdown the reactor or to mitigate the consequences of an accident. This change does not create a USQ. 1 l

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9. Technical Reauirements Manual Chanae 97-012. Reclassify the HRA Containment Isolation Valves DESCRIPTION The proposed change will reclassify the HRA containment isolation valves HRA-109A(B), HRA-110A(B), and HRA-126A(B) as Manual / Remote Manual containment isolation valves in TRM Table 3.6-2. In addition, the valves will be administratively maintained as locked closed containment isolation valves. This change will ensure Penetrations 66 and 67 meet the criteria of GDC 54 while maintaining compliance with GDC 56.

REASON FOR CHANGE J

The HRA containment penetrations must meet GDC 54 and GDC 56 criteria for containment isolation piping systems connecting directly to the containment atmosphere. Electrical relay contact K203A provides a path for power to the penetration isolation valves HRA-109A, HRA-110A, and HRA-126A. Likewise, relay 1 contact K203B provides a path for HRA-109B, HRA-1108, and HRA-126-B. I SAFETY EVALUATION It is concluded that the proposed change does not create a USQ. No physical change to the plant will be made and the change will not affect the operation or reliability of any  ;

system. The proposed change will classify containment isolation valves HRA-10SA(B),  ;

HRA-110A(B), and HRA-126A(B) as locked closed containment isolation valves. The change will not impact 10CFR50, Appendix J, leak rate testing classification or j acceptance criteria for the isolation valves, nor will the change create any new pathway l for release of radioactive material. The change does not alter the HRA system or I prevent the HRA system from performing its safety function. The change will ensure the containment isolation function of the HRA penetrations meet single active failure criteria thereby complying with the containment isolation provisions given in GDC 54 and 56. The change ensures that the HRA system meets the design, quality assurance, redundancy, energy source, and instrumentation requirements for an l engineered safety feature. Also, the HRA system itself will not introduce safety problems that may affect containment integrity, as required by RG 1.7 and GDC 41.

The change will, therefore, not increase the probability or consequences of an accident ,

or malfunction of equipment. Also, the margin of safety as defined in the TS Bases and  ;

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10. Chanae to Technical Specification Bases 3.6.3. Containment Isolation Valves DESCRIPTION The proposed change revises the Bases for TS 3.6.3 to clarify the TS actions with l respect to penetrations with multiple flow paths.

i REASON FOR CHANGE The current TS do not specifically address penetrations with multiple flow paths. The i

proposed change does not affect the initiators or the mitigators of any analyzed event and does not create any new system interactions or the way the plant is being operated. The affected penetrations are still required to meet GDC-54 and GDC-55 through GDC-57, as applicable, for all the unisolated flow paths, and be isolated with valves that meet the leakage rate acceptance criteria. This ensures the dose rates at the site boundary are within the limits assumed in the accident analyses. No USQ is created.

l SAFETY EVALUATION

' The proposed change does not affect the initiators or the mitigators of any analyzed event and does not create any new system interactions or the way the plant is being operated. The affected penetrations are still required to meet GDC-54 and GDC-55 through GDC-57, as applicable, for all the unisolated flow paths, and be isolated with valves that meet the leakage rate acceptance criteria. This ensures the dose rates at

the site boundary are within the limits assumed in the accident analyses. No USO is

! created.

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11. Technical Reauirements Manual Chance 97-011. Allowed Outaae Time for Certain 'B' Train Eauipment DESCRIPTION i The proposed change will create a TRM limit to dictate the allowable period of time that certain 'B' train equipment may be out of service. I REASON FOR CHANGE ,

Certain 'B' train equipment is necessary to complete remote shutdown during a fire in the control room / cable vault areas. Some of this equipment is duplicated on the 'AB' i- train. However, the equipment on the 'AB' train is not dedicated for use for remote shutdown due to a fire in the areas of concern and, therefore, is not isolated from the j control room by isolation switches during such a fire. Currently, during an outage on 'B' train equipment, no AOT has been established. This change will limit the period of vulnerability by establishing an acceptable period of time that 'B' train equipment applicable to this condition may be out of service.

SAFETY EVALUATION i

The TSs allow an outage of 7 consecutive days for the instrumentation located on remote safe shutdown panel LCP-43. Seven consecutive days is a period of time the station is licensed to operate while vulnerable to a control room / cable vault fire

requiring remote shutdown. Failure of theinAaentation located on LCP-43 would

substantially jeopardize the ability of the plant to complete a safe shutdown from the remote location. Thus the AOT for LCP-43 has similar basis (protection for a control room / cable vault fire) and bounds application to the 'B' train equipment also dedicated e . for use to complete remote shutdown. The introduction of this limit to the TRM will only i govern requirements related to the AOT of applicable 'B' train equipment from the control room / cable vault fire scenario ventagepoint. Operability of this equipment from the accident analysis vantage will continue to be governed by the TS on which this change will have no impact. Overall, this change limits the allowable period of time that applicable 'B' train equipment may be out of service. No USQ is created.

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12. Calculation EC-S96-011. "LOCA Offsite and Control Room Radioloaical Dose Conseauences" l

DESCRIPTION Calculation EC-S96-011 calculates the post-LOCA design basis offsite and cortrol room radiological doses.

l REASON FOR CHANGE The current design basis LOCA dose calculation accounts for dose contributions due to containment gas leakage and 8 gpm backleakage to the RWSP. Severalissues

regarding equipment performance that have an impact on the post-LOCA offsite and control room doses have come about since the last revision to the LOCA dose calculation. To account for these possible deficiencies, calculation EC-S96-011 incorporates a 10 cfm unfiltered air inleakage to the control room, assumes a total sump liquid leakage of 1 gpm in the RAB-CVAS area, and assumes that the maintenance hatch seal system fails 7 days after the accident. Since the test results i

have shown that the backleakage to the RWSP is very small compared to the 8 gpm assumed in the previous calculation, the backleakage assumption was reduced to 5 gpm.

SAFETY EVALUATION l

The LOCA dose calculation based on the assumptions in the calculation have resulted l in offsite and control room doses that are within acceptance limits of 10CFR100 and GDC 19. There is no USQ associated with the proposed change.

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i-l 13. Chance to Technical Specification Bases 3.3.1. Reactor Trio Breaker Channel

! Operability -

DESCRIPTION The proposed change revises TS Bases 3.3.1 to clarify the Operability requirements for an RTB Channel.

REASON FOR CHANGE The Operability requirements for an RTB channel are ,mt clear in TS 3.3.1 or the Bases of TS 3.3.1 with respect to the reactor trip breakers.

SAFETY EVALUATION l The proposed change does not create a USQ. Neither the design nor configuration of the plant is being changed nor are physical changes being made to plant SSCs. No ,

margin of safety is being reduced and all assumptions of the accident analyses are being preserved. The ATWS system is not affected by this change because it does not  !

rely on the RTBs to trip the unit. Also, a seismic event will not have an adverse consequence if the RTBs are racked out or open, because it cannot render the remaining RTBs incapable c' opening.

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14. Core Ooeratino Limits Report for Cycle 9. Revision 2 C

DESCRIPTION 3

l- The proposed change revises COLR Figures 6 and 7 to reduce the allowable PLHGR ,

i by 0.2 kW/ft. This change will compensate for an error in the ERF used in the LOCA l analysis.

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j. REASON FOR CHANGE  !

$ An ABB-CE Infobulletin,97-04, Revision 1, stated that the ERF used by ABB-CE in the i

! LOCA analysis did not directly reflect the effects of moderator voiding which may occur ,

. at several times during the LOCA. The ERF is used in the LOCA analysis to calculate 3

the PCT during the LOCA. A larger value of the ERF will result in a higher value for PCT during the LOCA for the same initial value of PLHGR. Therefore, a reduced  ;

i- ' PLHGR can accommodate for a higher ERF value, maintaining the PCT at its original  ;

value. To correct for the 0.9% error in ERF, COLR Figures 6 and 7 will incorporate the recommended 0.2 kW/ft penalty in PLHGR. This action will maintain a PCT of 2177* F. #

' SAFETY EVALUATION

[ According to the safety evaluation, PLHGR is not the cause of any accidents previously

! evaluated in the FSAR and does not increase the probability of any accident.  ;

j Reducing the PLHGR to compensate for the error in ERF ensures that the PCT in the l l event of a LBLOCA remains at or below the bounding cycle analysis value. No 4

important to safety equipment will be affected by this change and no new system-interactions or connections are required. Therefore, no new accident or equipment i/ .

malfunction will be created. The error in ERF affects only the LBLOCA analysis and limiting PLHGR ensures that PCT will remain below the bounding analysis value;

therefore, no margin of safety will be reduced.

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15. Technical Recuirements Manual Chanae 98-001. Containment Penetration Conductor Over-Current Protective Devices DESCRIPTION -

!. .The proposed change revises the responsibility matrix to show Design Engineering L Electrical responsible for electrical power systems, adds missing character in .

'~ component number, revises component number, changes reference drawing, and  !

revises a note to better describe a component.

. REASON FOR CHANGE l

identify group responsible for electrical power systems, make editorial changes, and  ;

j- implement action from an engineering request.  ;

1 l SAFETY EVALUATION '

l The proposed changes are primarily editorial in nature. While they do reassign some j

TRM responsibilities and make minor changes, they do not affect any accidents or l equipment important to safety described in the FSAR. These changes do not affect the i physical design or configuration of the plant or any margin of safety or protective J boundary. i l

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16. Chanae to Technical Specification Bases 3.7.12. Essential Services Chilled l Water System

! DESCRIPTION I l The proposed change revises TS 3.7.12 Bases to clarify the operability requirements for the Essential Services CHW. The Bases will be revised to describe normal CHW l system operation and address system conditions required to be met during the 31-day  ;

. surveillance requirement. The Bases will also discuss why the TS surveillance l l requirement temperature and flowrate may not be aoplicable during normal system i operation as a result of insufficient heat load on the CHW system.

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L REASON FOR CHANGE j l

The current TS Bases do not clearly differentiate the operability requirements that '

l ensure the system functions as required during a DBA, since the heat load is greater l during a DBA than during normal operations. Due to instrument uncertainty, the water 1 temperature at the chiller outlet is sometimes higher than the surveillance requirement.

Also, the flowrate is occasionally less than the surveillance requirement. This change will alleviate any confusion over the chiller operating requirements.

i SAFETY EVALUATION l

l This change is an administrative change to clarify the TS Bases regarding operation of l the CHW system during Normal and DBA conditions. The change does not increase l the probability or consequences of an accident previously evaluated in the FSAR '

l because no initiator or mitigator of any analyzed event will be affected. Neither the design nor configuration of the plant is being changed and no physical changes are being made to plant SSCs. All assumptions of the accident analyses are being 1 l preserved so that no margin of safety will be reduced.

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17. Technical Reauirements Manual Chanae 98-004. Fire Protection Surveillance Freauency DESCRIPTION The proposed change increases the frequency of testing of the fire detection system from once per 6 months to once per 12 months.

REASON FOR CHANGE A new microprocessor driven fire detection system was installed for which the manufacturer recommends functional testing on an annual frequency. The current edition of NFPA-72, " National Fire Alarm Code (1996)" specifies a frequency of every 12 months.

SAFETY EVALUATION The proposed change does not create a USQ. Increasing the testing frequency will not increase the probability of any FSAR analyzed accident. There is also no change in accident consequences as a result of this change. The new fire detection system is microprocessor driven and provides for continuous self-checking features that enhance the reliability and operation of the system and reduce the probability of maloperation.

Thus no equipment important-to-safety probability or consequences are affected by this change. No new system interactions or connections are required and no new failure modes are created by this change in testing frequency. No protective boundary or margin of safety is affected by this change.

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L 18. Soent Fuel Pool Criticality Analysis L

DESCRIPTION This is an evaluation of the Waterford 3 response to the NRC request of licensees to

! assess their Boraflex rack degradation and assess the capability of the Boraflex to meet criticality acceptance criteria. The previous criticality analyses considered the effects of various accident configurations and concluded that, when credit is taken for a minimum' soluble boron concentration in the SFP water, the NRC criticality criteria were met. The rzew criticality analysis demonstrates that even with greater Boraflex

!- degradation than previously assumed, the SFSRs continue to m9et the 0.95 keff limit.

REASON FOR CHANGE The SFSR criticality analysis has been updated to account for Boraflex panel degradation.

' SAFETY EVALUATION l The evaluation has demonstrated that no USQ exists for the SFSR criticality analysis.

All design basis events were found to either be bounded by the Reference Analyses or to be within the appropriate NRC acceptance criteria. The accident scenarios considered were the design basis fuel handling accident and the spent fuel cask drop into the SFP.

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19. . Technical Reauirements Manual Chanae 98-003. Thermal Overload Devices and Bvoass Devices DESCRIPTION The proposed change revises TRM Table 3.8-2 to incorporate safety-related valves omitted from the table. 4 REASON FOR CHANGE Per TS 3.8.4.2, all safety-related MOV thermal overload devices and bypass devices are to be operable. TRM Table 3.8-2 identifies all safety-related MOVs and identifies whether the thermal overloads are bypassed. Valves MS-401 A(B), SI-125A(B), SI-  !

135A(B), and SI-412A(B) should have been incorporated into this table.

SAFETY EVALUATION This is an administrative change to add valves to Table 3.8-2 of the TRM. No  !

accidents or important-to-safety equipment will be affected by this change since there is no direct impact on the physical plant. This change identifies whether or not the valves added to the table have their thermal overloads bypassed and whether they are included in the surveillance test program. The test program will then ensure each valve can perform its particular safety function. '

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i . 20. SPC-97-003-0. HPSI and LPSI Bearina Coolina Water Low Flow Alarm l l

DESCRIPTION i

- This setpoint change incorporates calculation EC-M97-001, calculated minimum CCW l - flow to the HPSI and LPSI bearing and seal coolers. The minimum flow is 10 gpm at )

[ . design basis accident conditions, which is bounded by the current numbers in the t

analysis and FSAR Table 9.2-3 which dictates a maximum flow of 20 gpm to the HPSI L and LPSI bearing and seal coolers. The new setpoint of 10.25 gpm +/-0.25 is within the maximum value give in FSAR Table 9.2-3.

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! REASON FOR CHANGE L An.SSFI identified there was no basis found to indicate 8 gpm is sufficient CCW flow to the LPSI and HPSI pumps under accident conditions. Calculation EC-M97-001 L ' identified required minimum CCW flow to be 10 gpm under accident conditions. 1 L SAFETY EVALUATION l There are no accidents caused or affected by this change; therefore, there is no ,

l increase in accident probability or consequences. This change does not increase the  !

L likelihood or consequences of an equipment failure. Rather, it ensures that the i - minimum bearing and seal cooling water flow (CCW) is adequate for the HPSI and y LPSI pumps. No new system interactions are created by this change so no new accident or equipment malfunctions are created. No protective boundaries are 1 l affected, no margin of safety is reduced, and no USQ is created.

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21. SPC-97-008-0. BAMTs LoLo Level Alarm i DESCRIPTION The proposed change adjusts the Lolo level alarm setpoint for the BAMTs from 21 inches to 7 inches.

I REASON FOR CHANGE l

Ensures that, including uncertainties, the required amount of boric acid solution from the BAMTs assumed in the safety analysis is injected into the RCS prior to operations securing the BAM pumps. It also ensures the LoLo level alarm will occur when l appropriate (considering instrument uncertainty) and that sufficient margin is available to prevent the possibility of vortexing the BAM pumps.

SAFETY EVALUATION The proposed change to adjust the LoLo level setpoint alarm on the BAMTs will ensure that sufficient borated water is injected into the RCS following the limiting event of a LOOP with a loss of letdown (BTP 5-1). The proposed setpoint will also provide the operators with sufficient time to secure the pumps prior to air intake (and possible failure). The proposed change does not create any new system interactions or connections and does not increase the probability or consequences of any previously analyzed or unanalyzed accident scenarios. The proposed change does not result in a reduction of any margin of safety.

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22. ~ SPC-97-013-0. Control Room Air intake Chlorine Detection '

DESCRIPTION The proposed change decreases the chlorine monitor setpoint analytical value from 3 ppm to 2 ppm, the calibrated control room isolation setpoint (accounting for uncertainties) from 1.5 ppm to 0.6 ppm, and the calibrated alarm setpoint from 1.0 ppm to 0.4 ppm.

REASON FOR CHANGE Reperformance of the chlorine habitability analysis has determined the 3 ppm analytical limit for a chlorine accident does not ensure a control room isolation response time of less than 4 seconds as required by RG 1.95. The new analysis accounts for variations in control room airflow rates and temperature effects on chlorine monitor response time. The revised analyticallimit of 2.0 seconds ensures the control room dolation will occur within 4 seconds.

SAFETY EVALUATION The chlorine monitors do not interface with piping systems or controls to systems assumed to be initiators in the accident analyses. There are no radiological consequences expected for a chemical release. The change reduces the alarm setpoint and increases the lovel of protection in the control room. The setpoint change does not affect the consequences of a monitor malfunction since the results would be the same regardless of the setpoint - operator incapacitation and inability to reach safe shutdown conditions. No new system interconnection is required thus no new accidents or malfunctions are created. No protective boundaries are affected and no USQ is created.

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l l1 F. COMMITMENT CHANGES

1. Component Coolina Water Makeuo Commitments

SUMMARY

' Rather than modify the ACCW system to meet the existing required design basis flows, a new (lower) design basis ACCW flow to the CCWHX will be determined. The Essential Chiller design flow will remain the same.

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2. -Performance of Barrier Analysis for All Condition Reports

SUMMARY

This change deletes the requirement to perform a barrier analysis for every condition report. The CRG determines the classification, priority, and assignment of CRs. The barrier analysis may be used but it is not the only method to identify apparent causes.

In addition, it may not be practical to apply the barrier analysis method to each CR since other methods may be better suited. This change has no impact on the ability of any SSC to perform its safety function and does not negatively impact the ability of plant personnel to ensure the SSC is capable of performing its intended safety function.

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3. Approval of Root Cause Analyses and Corrective Action Plans

SUMMARY

This change revises the requirement for all root causes analyses and their associated corrective action plan to be reviewed by the CRB for approval. All RCAs and corrective action plans will be presented to a CARB for approval. This change will better utilize management resources and will continue to require that appropriate management review and approve RCAs and corrective action plans. The CRB Chairperson will appoint the CARB members and chairperson. This administrative change will have no impact on the ability of any SSC to perform its safety function or the ability of plant personnel to ensure the SSC is capable of performing its intended safety function.

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4. Trackina of Corrective Action Implementation

SUMMARY

This change eliminates a duplicate commitment tracking system for verification of i implementation of corrective action in W2.501. This commitment was established for l the tracking of active commitments. The corrective action program is required by l l 10CFR50, Appendix B, Criterion XVI, and therefore this commitment application is

! redundant. This administrative change will have no impact on the ability of any SSC to l perform its safety function or the ability of plant personnel to ensure the SSC is capable of performing its intended safety function.

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5. Replace Root Cause Investiaation and Delete Procedure Review as Part of I Corrective Action

SUMMARY

This change deletes the commitment for Root Cause Investigation criteria and the requirement for procedures to be reviewed as part of corrective action for every occurrence. RCAs are performed for all significant adverse conditions unless specifically waived by the CRB. In addition, analytical tools cause procedures to be reviewed and revised as necessary. However, these tools do not specifically require procedure review since circumstances may vary. There are sufficient barriers in place (CRB, CARB, lHEA, and department management) to ensure procedure reviews when appropriate. This administrative change will have no impact on the ability of any SSC to perform its safety function or the ability of plant personnel to ensure the SSC is capable of performing its intended safety function.

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6. Desianation of a Corrective Action Review Board

SUMMARY

This change allows the CRB (hereafter to be called the CRG) to designate a CARB chairperson and for the CARB to review and approve root cause analyses or corrective action plans for Category 1 condition reports. This administrative change will have no

' impact on the ability of any SSC to perform its safety function or the ability of plant personnel to ensure the SSC is capable of performing its intended safety function.

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7. Closina Condition Reports to Station Modification Reauests

SUMMARY

This change deletes the commitment that allows the practice of closing CRs to SMRs.

Site procedure W2.501 provides procedural guidance for adequate closure of CRs. It is not the practice of Waterford 3 to close CRs to SMRs. This violation described closing PElRs improperly. Credit was taken for the CR closure process; however, guidance to this level of detail is impractical. Adequate closure of CRs is bounded by 10CFR50, Appendix B, Criterion XVl; therefore, this commitment is considered redundant. This administrative change will have no impact on the ability of any SSC to perform its safety function or the ability of plant personnel to ensure the SSC is capable of performing its intended safety function.

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8. Preparation and Processino of Purchasino Documents

SUMMARY

This change deletes the requirement for sending an acknowledgement copy of the PO to the vendor and maintaining a log of those copies At one time, vendors were required by the standard PO form to provide an acknowledgement copy. The acknowledgement copy was used to ensure consistency for the format of vendor responses. A log was also maintained of acknowledgment copies sent to vendors.

Incorporation of MMIS allows computerized tracking of PO revisions and verification of vendor acknowledgement during receipt inspection. Purchasing now updates MMIS to reflect the latest PO revision that prints on the receipt inspector's Inspection Report.

The receipt inspector is required to verify the PO number and revision on the vendor's documentation to confirm the vendor's acknowledgement of the latest revision.

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l 9. Insoection of Air Ooerated Valves l

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SUMMARY

1 This change reduces the number of AOVs to be inspected. Forty valves will not be '

r inspected because they are located in radiation or contamination areas and significant

!. dose would be received in making the inspections. The results of the other 168 .

l' inspections only generated 12 Cis, so there is no generic problem with fasteners. In l' addition, no operability concerns were identified during any of the inspections.

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10. - Surveillance Testina of the Steam Driven Emeraency Feedwater Pump

SUMMARY

Surveillance testing of the steam driven EFW pump required, as a prerequisite, that the steam supply piping heat tracing be energized and the piping temperature be greater than 280 F. As a result of a corrective action document, an engineering calculation concluded that degraded heat trace of the EFW steam supply piping will not have an adverse affect on the steam driven EFW pump as long as all piping is maintained greater than 230 F and the average weighted temperature is greater than 350 F.

These values preclude the potential for standing water in the steam supply piping and limit the start-up condensate load of the steam driven EFW pump to no more than 5%

moisture content by volume. By meeting these criteria, it is ensured that a heat trace

- failure will not adversely affect the steam driven EFW pump.  ;

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11. Position of Chilled Water Valve CHW-823

SUMMARY

System pre-operational test procedures for all safety-related systems were compared to Operations department system operating procedures to ensure valves which were throttled during testing were also throttled, if required, in the system standby valve lineup attachments. This resulted in a commitment to lock CHW-823 one turn open.

This change deletes that commitment. Design Change DC-3468 failed open most of the temperature control valves in the CHW System, requiring the flow balance of the system to be changed. This resulted in a new required position for CHW-823 as identified by special test procedure and a resulting change in the standby valve lineup attachments. While CHW-823 will still be throttled to maintain capacity flow to air handling units, it will no longer be maintained one turn open. The position of CHW-823 will now be maintained in OP-002-004 and OP-100-009 in accordance with the most recent results of the CHW flow balance procedure. The new position per the most recent flow balance will maintain the same capacity flows to all air handling units in the CVC/EFW subloop of CHW Train 'B' as were present before DC-3468 was implemented.

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12. Relocation of EDG Startina and Control Circuit Post-Maintenance Testina ,

Reavirements l

SUMMARY

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, This change relocates the requirement for providing additional guidance on post-  !

. maintenance tests that should be performed following maintenance on the EDG starting

l. and control circuits from UNT-005-020, " Post Maintenance Testing", to the Planning ,

l= Information Guide Notebook. No requirements are being deleted, only moved to a  !

L different procedure.

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13. INPO Human Performance Proaram Trainina

SUMMARY

The Waterford 3 PIP originally committed to train plant workers using the INPO Human

- Performance Program. Waterford 3 has entered into a contract with Performance Improvement International to provide assistance in human error reduction, including

' providing training. .This training is equivalent to the INPO training and will be used

. instead of duplicating efforts.

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1 14.- Baooino Contaminated Valve Lineuo Sheets

SUMMARY

This change deletes the requirement from UNT-004-009 for contaminated sheets to be photocopied upon exit from the RCA or for those hard copies to be verified line-by-line.

Stronger RCA controls now require all workers to monitor items for contamination using the TCM. If the TCM were to alarm, the worker would contact HP for further guidar:ce.

This change has no affect on any SSC or on the ability of plant personnel to ensure an SSC is capable of performing its intended function.

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. 15. Imolement the Site-Wide Enaineerina Reauest Process .

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SUMMARY

This change to PIP ltem 4.A.1 extends the due date from 12/31/97 to 3/31/98 to implement a site-wide ER process. Due to the extended outage, procedures W4.104, W4.105, and W4.201 did not receive adequate inter-departmental review. Comments continue to be received from various departments. In addition, it was decided to i assemble a team of engineers to develop examples of ER responses. This will result in a process tabletop walk through, improved procedures and guide, an improved engineering training plan, and enhanced global ownership of the ER process. This

delay of three months will not have any impact on any SSC or on the ability of plant

. personne! to ensure an SSC is capable of performing its intended function.

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1 Quarterly Human Performance Awareness Da.yj! j

SUMMARY

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This change deletes the requirement for holding quarterly Human Performance i Awareness days. When the PlP was written, the process by which Waterford 3 intended to improve human peiformance was not fully developed. As planning proceeded, it was determined that a more structured approach was needed. Waterford j 3 has therefore contracted with Performance improvement International to provide a i more extensive and structured process.

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17. Maintenance Department Review of CE-001-025

SUMMARY

~ This is a change to the commitment due date from 9/30/97 until 11/25/97 for review of procedure CE-001-025.with Maintenance Department personnel. Several maintenance ,

personnel are at Riverbend Station for shared resources and are not available for

.g training.

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l: 18. Performance of Root Cause Determinations 1

SUMMARY

j. This change adds performing an RCD for some significant condition reports based on l whether the CR is designated as category "A", "B", or "C". It also changes the name of

! the Condition Review Board to Condition Review Group. Both RCAs and RCDs will be performed by qualified root cause evaluators, the final product still requires management approval, and any corrective actions must be accepted by the responsible

- department. This administrative change has no impact on any SSC.

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19.. Corrective Action Review Board Review of Root Cause Determinations L -

SUMMARY

1 This change adds the RCD to the information presented to the CARB 'or approval and

, adds a reference to the CRG (formerly the CRB). Both RCAs and RCL's will bo l

L performed by_ qualified root cause evaluators, the final product still requires l management approval, and any corrective actions must be accepted by the responsible j

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20. Quality Assurance and In-House Events Analysis Verification of Corrective  :

Actions

SUMMARY

. f This change makes QA responsible for verifying the effectiveness of corrective actions -

through the audit process and gives IHEA personnel the responsibility for verifying the  ;

effectiveness and completeness of corrective actions for significant CRs. An organizational restructure moved the corrective action program under the Director, Nuclear Safety & Regulatory Affairs, with lHEA being the group responsible for the

. - corrective action program.

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21. Individual Ownership of Performance Imorovement Plan items

SUMMARY

The PIP was developed for the purpose of improving Waterford's regulatory performance and restoring Waterford to a position of leadership in the industry. One of l the important aspects of the PIP is individual ownership of each of the strategies by which the plan is implemented. This change allows for reassignment of PIP items due to organizational changes without having to complete a CCEF for each change.

Persons occupying positions may change due to management development, rotation, promotion, or attrition; however, individual ownership of each PIP item will be maintained to ensure PIP items are implemented as committed to the NRC.

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22. Increase in Time for Initial Operability Determinations

SUMMARY

This change increases the time for an initial operability determination from 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to a time consistent with the appropriate TS LCO action requirements or 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, whichever is greater. This change complies with Generic Letter 91-18 and IM 9900 guidance. It allows for the establishment of a reasonable time for the initial engineering evaluation to be developed and approved. The time between the validity determination of the nonconforming / qualification situation and completion of the initial engineering evaluation will be commensurate with the safety significance of the nonconforming / degraded condition. The time will be consistent with the appropriate TS LCO action requirement time limits or 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, whichever is greater. This change will not negatively impact the ability of any SSC to perform its safety function.

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23. Responses to Inspection Reports

SUMMARY

This change deletes certain tasks from the commitment for Licensing review of NRC inspection reports and corrective action verification. Originally, a licensing action plan was developed using an LIR that sometimes required input from other departments. ,

The LIR was responded to and certified by the responsible department by means of the Task Review and Certification form. The response was reviewed by an LE and approved by appropriate management prior to submittal to the NRC. Corrective actions were accomplished by review by the LE and further validated by Operations QA. Due to process changes and reorganization, these tasks are obsolete. The Director, Nuclear Safety & Regulatoy Affairs, is now responsible for coordinating review and preparation of responses to NRC Inspection Reports with the task being performed by the Licensing Department. Corrective action verification is done by means of a Commitment Closure Verification Form from the responsible department that then receives a confirmatory review by the LE.

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24. Conduct Trainino for Trenders and Manaoement P

SUMMARY

This_ changes the procedure for training responsible trenders and their management i= from UNT-007-025 to W2.111 due to procedure W2.111 now being the controlling plant

! '- trending procedure. This is an administrative process that is not used to ensure SSCs j are capable of performing their intended function.  :

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25. Fire Detection for the 'O' Deck Area

SUMMARY

This is a change in the completion date for installation of fire detection on the 'O' deck area from 12/31/97 to 2/28/98 since the final work scope was not completed until 12/2/97. An estimate for complete installation of 17 fire detectors and associated conduit and cables was 300 hours0.00347 days <br />0.0833 hours <br />4.960317e-4 weeks <br />1.1415e-4 months <br /> with 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> required for detector setup and testing. This change does not affect the action to be taken. The committed to action is a compensatory measure that is not applicable until Refuel Outage 9 or the next major outage when a significant amount of materials are staged or transported to the 'O' Deck area.

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26. Plant Operations issue: Failure to identify Misalianed Valves and Switches

SUMMARY

Initially, OP-100-007 was enhanced to add management's expectations of a requirement for two board walkdowns per shift per control room watchstander and to add a sign off for one board walkdown per control room watchstander to the appropriate turnover sheets. Subsequent review of the process by Operations Management has resulted in modification and clarification of management's expectations for control board walkdowns. The original commitment is not changed; however, the comments section associated with control board walkdowns is revised as follows: 1) The PNPO and SNPO will walk down their boards each shift; 2) Following manipulations of safety related components and systems, a check to verify components will be performed by another NPO or the CRS; 3) Once per shift, the SS should pedorm a board walkdown concentrating on safety related controls manipulated during the shift;

4) There will not be a requirement for watchstanders to sign for board walkdowns.

Board walkdowns do not directly impact the ability of an SSC to perform its function or the ability of plant personnel to ensure the SSC is capable of performing its intended function.

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27. Performance improvement Plan

SUMMARY

This is a change in the due dates for the subject Performance improvement Plan commitments as follows: A-24086, from 6/30/98 until 3/31/00; A-24106 from 1/31/98 until 9/30/98; A-24107, from 1/31/98 until 1/31/99; A-24108, from 1/31/98 until RF10; A-24161, from 12/31/98 until RF10; A-24162, from 12/31/98 until RF09; A-24163, from 12/31/98 until RF09; A-24164, deleted; A-24165, deleted; A-24167, from 12/31/98 until RF09; A-24169, from 3/31/98 to 12/31/99; A-24170, from 9/30/98 until RF09; A-24401, from 3/31/99 to RF09; A-24221, from 12/31/98 to 12/31/99.

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28. Trackina of Weekiv Technical Specification Surveillances

SUMMARY

The change revises this commitment to allow completion of weekly TS Maintenance Department Surveillances to be tracked by the rotating 12-week maintenance schedule and the TS Surveillance late date report. This commitment had required the Control Room staff to track TS surveillances performed by Maintenance and Operations using procedure OP-903-001. This was achieved using a surveillance to track the Battery Bank weekly surveillances and the BRGM weekly surveillances. Waterford 3 has established a rotating 12-week maintenance schedule that controls RTs The work packages for RTs are generated by the responsible discipline planner for performance on the due date. The TS Coordinator tracks completion of all TS surveillances and generates a late date report to ensure completion of required surveillances. Tracking of Maintenance Department Surveillances by the Control Room staff would thus be redundant.

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l l 29. Reduced Inventory Trainina l

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SUMMARY

The commitment originally required the Training Department to develop a refresher course for reduced inventory training to be conducted for personnel who have not received pre-refueling training within six months of operating the plant in a reduced inventory condition. This change removes that restriction and allows the training to be conducted within the normal two year training cycle. The six month restriction places an unnecessary burden on the Training and Operations departments if the plant is forced to shut down and go into a reduced inventory condition on short notice and it has been greater than six months since the training has been given.

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30. . Station Modification Reauest CC-024

SUMMARY

1 This modification of the Temporary Chill Water Supply lines was originally identified as '

long term corrective action in LER 96-015-01 However, during subsequent site review,

- it was determined the action is an enhancement and not required corrective action.

. This change was discussed with Mr; G. Pick'of Region IV in May 1998 1

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G. ENGINEERING REQUESTS

1. ER-W3-97-0043-00. Enaineerina Review of New CPC Software for Elimination of Unnecessary Trios Due to Drocoina CEA #2 or #3 DESCRIPTION A change to the existing RDB and Database for the CPCs will change how the positions of CEA #2 or #3 are interpreted which will eliminate unnecessary trips. CEA Subgroup #23 Position will not be a valid CEA subgroup position because CEA #2 and CEA #3 are no longer considered subgroup target rods. There has not been any  :

change to the actual subgroup assignments within the CEDMCS or any change to

. physical plant equipment.

- REASON FOR CHANGE -

As presently configured, the CPCs installed at Waterford 3 will generate an _

unnecessary reactor trip upon the drop of either CEA #2 or #3. The proposed change will prevent this unnecessary reactor trip.

SAFETY EVALUATION The proposed change to the CPC constants only affects the interpretation of which  !

individual CEA positions are associated with which CEA subgroups within the CEAC l and CPC algorithms. There has not been any change to the actual subgroup l assignments within the CEDMCS or any change to physical plant equipment. The l

proposed change does not affect any accident but does preserve all necessary responses to CEA deviations. No important to-safety equipment is affected and no consequences increased. There has been no physical change that would introduce the ,

possibility of a different type accident than previously evaluated. The response of the l CPCs to plant evolutions has been preserved so that all events that rely on the CPCs  !

' to initiate a reactor trip to preserve the margin of safety are unaffected.

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2. ER-W3-97-0286-00-00. Reassianment of CEA #3 From a Four-finaer Subaroup  !

to a Five-finaer Subaroup to Prevent "CEA Deviation" Alarms with All Rods l Inserted l

- DESCRIPTION i l

This software change to the existing RDB will move CEA #3 from a four-finger CEA subgroup #22 and reassign it to a five-finger subgroup #12. This change will also correct the CEA rod position display system to correctly identify CEA #3 with subgroup L #12 and CEA #2 with subgroup #13. There has not been any change to the actual subgroup assignments within the CEDMCS or any change to physical plant equipment.

REASON FOR CHANGE l

A prior change to the existing RDB for the CPCs changed how the positions of CEA #2 l and #3 were interpreted. This was done to eliminate unnecessary reactor trips. CEA J

~ #2 and #3 were reassigned to subgroups #13 and #22. CEA subgroup #23 is no longer

- a CPC target rod subgroup. During this software revision, CEA #3, which is a five-finger CEA was inadvertently assigned to a four-finger CEA subgroup #22. Subgroup ,

1. 22 now had five CEAs assigned: four four-finger and one five-finger. With all rods l inserted, subgroup #22 can insert the four four-finger CEAs to approximately 9 inches. ;

By design, the four-finger CEAs that straddle two fuel bundles cannot be inserted less )

than 9 inches. CEA #3 is a five-finger CEA that can be fully inserted. This mismatch within subgroup #22 with all rods inserted resulted in "CEA Deviations" alarms on both CEA calculators.

SAFETY EVALUATION There has not been any change to the actual CEA subgroup assignments within the CEDMCS or any change to physical plant equipment. Therefore, no accidents are affected by this change. The reload analysis for Cycle 9 has demonstrated that the

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inward deviation of subgroup 12 has a less adverse impact on the core power distribution than the most adverse single CEA drop. Therefore, there is no increase in accident consequences as a result of this change. Since there has been no actual change to the physical plant and no new interactions are required, there is no increase in either the probability or consequences of an equipment malfunction and no new accidents or equipment malfunctions are created. The response of the CPCs to plant evolutions has been preserved by the changes so that all events that rely on the CPCs to initiate a reactor trip are unaffected.

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3. ER-W3-98-0190-00-00. Switchaear Ventilation System Batterv Room Exhaust Fan Differential Pressure l

DESCRIPTION This change revises the setpoint for Battery Room 'B' Exhaust Fan 'A' differential pressure switch to eliminate the Control Room nuisance alarm. 1 BEASON FOR CHANGE Battery Room 'B' Exhaust Fan 'A' flow instrumentation measures the DP between the inlet of the fan and downstream of the discharge damper. All remaining flow nstrumentation for the battery rooms measures the DP between the inlet and the discharge of the fans. All of the flow instrumentation installed on the Battery Room Exhaust Fans has a setpoint of 0.6 inwc decreasing. The configuration of Battery Room 'B' Exhaust Fan 'A', with the current setpoint, causes nuisance alarms in the Control Room.

SAFETY EVALUATION The accidents potentially affected are the LOOP and LOOP with SAF. However, the I affected component and system do not initiate either of these accidents nor increase their consequences. The affected instrument does not initiate any automatic functions or affect the performance of the fan. Therefore, this change will not increase the probability or consequences of an equipment malfunction. No new system interconnections are required and no new failure modes are created. No protective boundaries are affected and no margin of safety reduced.

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'4. ER-W3-98-0558-00-00. Reolace Valve RWM-1255 with a Ball Valve DESCRIPTION l -This change replaces the manual globe valve, RWM-1255, with a manual ball valve.

[ REASON FOR CHANGE Build up of spent resin prevented this valve from fully closing which led to a spent resin spill.

l SAFETY EVALUATION The evaluation concludes that this change will not reduce the margin of safety and no USQ is created. FSAR Section 11.4.5.1.d states " process valves are plug type valves with ultra high molecular weight polyethylene seats." The application for this L' configuration change is not a process valv.e; it is a drain valve whose function is to

. facilitate draining of the Spent Resin Transfer Pump and associated piping to a floor drain. The replacement ball valve without polyethylene seats is evaluated to be equivalent to the original valve and meets licensing basis requirements.

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. lli. PROCEDURES

- A. PLANT PROCEDURES

1. PE-004-024. ACCW and CCW System Flow Balance. Revision 0 DESCRIPTION ACCW and CCW system flow balance test.

REASON FOR CHANGE

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- This procedure will periodically verify that each safety-related component cooled by the

. ACCW and CCW systems receives the proper flow during accident conditions. This will be accomplished by putting each system in its accident lineup and documenting indicated flow through each component. This test may be performed in Modes 1 L through 6.

l SAFETY EVALUATION

~ The function of the ACCW and CCW systems will not be changed for this test. The test l will verify that the components cooled by the ACCW and CCW systems receive the

, proper design flow. The CCW system removes heat from mechanical components and heat exchangers during normal plant operation. During accident conditions (LOCA, MSLB, or MFLB) the CCW system supplies cooling water to the Essential Services L

Loop (CFCs, EDGs, HPSI, LPSI, CS pump bearing coolers, SDCHX, and Essential l . Chillers) and along with ACCW rejects heat to the atmosphere via the Cooling Towers.

During an accident, cooling water to the Essential Chillers is supplied by ACCW when

. CCW temperature is above 102* F. On an SlAS, two 100% redundant CCW Trains are l_ formed with the nonessential loop (RCPs and CEDM coolers) supplied by CCW Train

A'. On a CSAS, cooling water _to the nonessential seismic loop is isolated. This test may be performed in any plant mode. During power operation, the nonessential seismic and nonessential nonseismic loops will be isolated from the train being tested.

Based on the safety evaluation, no USQ exists.

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2. CE-002-006. Maintainina Reactor Coolant Chemistry. Revision 9 DESCRIPTION i

This change is a result of higher RCS lithium concentrations that will be necessary i during startup and power ascension beginning with Cycle 9. RCS lithium concentration limits are currently 0.2 to 2.2 ppm. This revision proposes a new lithium control band of l

- 0.2 to 2.96 ppm. This increase in lithium concentration will result in increased tritium production in the RCS.

REASON FOR CHANGE -

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' To. incorporate recommendations of Revision 3 of the EPRI PWR Primary Water- j Chemistry Guidelines and to incorporate power uprate changes. For power uprate and l Cycle 9 operation, a boron concentration increase is necessary to control the increased l

reactivity. ' Since boron lowers the pH, lithium will have to be increased to maintain pH l above 6.9, which is the limit specified by EPRI and ABB-CE. Changes made to tritium l

production rates are based upon power uprate expected values and bound current i operation; therefore, utilizing power uprate results is conservative. l SAFETY EVALUATION Plant safety is not impacted and there are no USQs created as a result of this change. )

Operation with increased. lithium concentration will still be within the EPRI guidelines for control of primary water chemistry. The tritium production increase evaluation shows an approximately 23% increase in tritium production. The increase in primary l

coolant tritium activity due to a higher. lithium concentration will not provide any I significant input into the baseline for any accident scenarios. The doses in accidents 'l

' are primarily due to release of fission products contained within the core, with .i radiciodine being the major contributor to dose. The chiid thyroid inhalation dose factors for tritium and lodine 131 were compared. lodine 131 has a dose factor of 1.62 E+7 mrem / year / microcurie / cubic meter. The dose factor for tritium is 1.12 E+3 mrem / year / microcurie / cubic meter which is a factor of 14464 times lower. Accident analyses currently utilize the TS 3.4,7 specific RCS activity of 1 microcurie / gram del as the maximum allowed RCS activity during operation. Therefore, any change in the actual RCS activity will be limited to and bounded by this TS limit. l l

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3. UNT-006-021. Pumo and Valve Inservice Testino. Revision 3 .

DESCRIPTION UNT-006-021 is being revised to implement the Second Ten-Year Pump and Valve IST Program which has been updated to comply with the requirements of ASME/ ANSI OM Code,1987 Edition (OM-1987), including 1988 Addenda (Oma-1988). The change

- also reflects administrative changes made based on organization or other Waterford 3 changes.

I REASON FOR CHANGE

-'10CFR50 55a(f)(4)(ii) requires that Waterford 3 must update its IST Program to comply with the requirements of ASME/ ANSI OM Code,1987 Edition (OM-1987), including 1988 Addenda (Oma-1988).

SAFETY EVALUATION 7 The proposed change does not result in a USQ. Waterford 3 is required to update the IST program to the 1989 Code for the second 120-month interval. This change implements and clarifies _the requirements of the 1989 Code. This change incorporates changes in the responsibilities that have occurred since the last revision to the procedure. This is only a change to the administrative procedure for the IST Program.

The individual implementing procedures for IST each receive their own 50.59 review.

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4. RW-TEM-001. Containment Fan Cooler Discharae Tank Processina. Revicion 0 DESCRIPTION This temporary procedure provides instructions for removing radionuclides from the CFC cleaning solution in the tank in the yard outside the RAB.

REASON FOR CHANGE Removing the radionuclides will greatly reduce the cost of disposing of the cleaning solution.

SAFETY EVALUATION i

This activity does not require modification of any plant system nor does it affect any accident previously evaluated in the FSAR. It does not affect any important-to-safety equipment and no new system interconnections are required. No margin of safety is affected and no protective boundary is changed.

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j 5. UNT-005-013. Fire Protection Proaram. Revision 6 DESCRIPTION l

The proposed revision updates the references and organization titles and changes the format in accordance with sitewide guidelines.

REASON FOR CHANGE i

Implement corrective action and procedure improvements.

SAFETY EVALUATION The proposed revision implements administrative changes only and does not affect the probability or consequences of a fire. There are no direct interface or control actions for plant equipment specified by this procedure. No margin of safety or protective ,

boundary is affected by this change and no USQ is created.  !

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6. RW-TEM-002. Spent Resh Pumo Room Resin Recoverv. Revision 0 DESCRIPTION 1

l . This procedure provides guidance for the recovery of spent resin in the Spent Resin l Pump Room area, transferring the resin, ard packaging the resin in an acceptable shipping container.

REASON FOR CHANGE-

. Procedural guidance was required due to the complexity of recovering the spent resin from the Spent Resin Pump Room.

l j SAFETY EVALUATION i

The only accident that could potentially be affected by this procedure is a Liquid Waste system leak or failure. The accident is a complete failure of all non-safety and non-seismic Category 1 equipment in the LWM system and is assumed to occur as a result of a safe shutdown earthquake. This failure is assumed to result in the simultaneous release of all liquids in the system tanks to the RAB. The amount of water to be added to LWM will be controlled during the dewatering process so the volume will not exceed the storage capacity of the LWM system. The activity levels of the water will be no different than the wastewater introduced to LWM during normal dewatering of the

. Spent Resin Tank. The equipment to be used is not connected to the RCS ard is not L required for accident mitigation or for safe shutdowrt Prior to staiting resin removal, a L ' floor drain plug will be installed to prevent inadvertent draining of resin through the

[ floor drain system from the RWM pump room. The affected pipe services only the RWM pump room and does not service any other areas with important-to-safety-equipment. If a resin line were to fail, the water would go to floor drains that empty Sump #9 but would not affect any equipment or access to it. A calculation was

. performed which demonstrated that if all the activity were to exit through the plant stack, the total dose would be 0.222 MR. This is bounded by the ODCM limit of 15 MR.

There is no USQ associated with this activity.

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b 7. .Q. E-002-001. Maintainino Steam Generator Chemistry. Revision 13 '

i j DESCRIPTION '

- The procedure is being changed to incorporate recommendations of EPRI's PWR Secondary Chemistry Guidelines, Revision 4, and clarify how a blowdown discharge to

, circulating water is to be carried out. This will require a revision of the Steam l Generator Blowdown Chem; 'ry Limits found in FSAR Table 10.3-2.

.! REASON FOR CHANGE

, Upgrade the procedure to the new recommendations in EPRI's PWR Secondary )

i Chemistry Guidelines, Revision 4. Other changes are editorial, administrative, human i i . factors, and corrective actions. ,

1 SAFETY EVALUATION ,

The evaluation has determined that there is no USQ associated with this procedure I revision. The changes will provide tighter controls on steam generator chemistry and I g maximize the integrity of the steam generator tubes and supports. These changes will l not increase either the probability or consequences of an SGTR. No important-to- l

] safety equipment will be adversely affected by these procedure changes. Chloride and )

[ sodium play a key role in the initiation of intergranular attack and stress corrosion i cracking. Having lower limits for these parameters will minimize the likelihood of tube degradation. No new system interactions or connections are required and no new j . equipment' malfunction will be created. No protective boundary or margin of safety will be affected by these changes.

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B. SPECIAL TEST PROCEDURES (STPF l

1. STP-99003536. Acceptance Test for DC-3536  !

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DESCRIPTION  !

.This special test procedure will ensure the installation of DC-3536, installation of four  :

backdraft dampers to the HVC Systems Emergency Outside Air Intakes REASON FOR CHANGE DC-3536 is being installed due to field testing which showed a single failure which

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' results in either_ HVC-213A or HVC-213B failing open may cause both trains of CREFU l to become inoperable. 1 SAFETY EVALUATION l

This special test will be performed in Mode 5, during the time when both trains can be out of service. To declare the units operable again, this special test will have to be successfully performed. No accidents or their consequences are affected by this test.

No equipmentf other than the CREFU, is affected by this test, which will verify the units

- can fulfill their design basis function. No new system interactions or connections are

-evired. However, since the South Side does not contain a BRGM and a Toxic Event musi be considered, the following precautions will be taken: 1) the Meteorological

Tower Sigma Theta reading must be greater than 5 degrees; 2) at least one RAB +7

elevation air handling unit must be operating to ensure ambient air is flowing into the BRGMs; 3) the Chemical Hot Line will be checked just prior to opening the South Side Outside Air Intake. These actions should ensure adequate isolation should a Toxic Gas Event take place while pressurizing the Control Room Envelope for this special test when using the South Side Outside Air Intake No r.mgin of ufety is reduced as a result of this special test and no USQ is involved.

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2. STP-01160772, HVF H&V Room Exhaust Fan Test with Sinole Gravity Damper

! Failure pf1QRIPTION '

This special test will determine the effect of a single active failure of a gravity damper  !

on the safety function of FHB +1 H&V Room Ventilation. ,

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REASON FOR CHANGE ,

This test is being performed to verify that a single active failure of a FHB M&V Room i exhaust fan gravity damper will not prevent the system from performing its safety  !

function.

SAFETY EVALUATION l

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' The FHB Ventilation System is required to respond to a Fuel Handling Accident. This change will not affect the ability of the system to perform its intended function since both units of FHB Emergency Ventilation will be declared inoperable. To preclude the l possibility of an event which might increase the consequences , - an accident, {

movement of fuel within the SFP and crane operation with loads over the SFP will be L suspended during this test. The test will require each gravity damper at the discharge of the secured exhaust fan to be held open for a limited time to quantify exhaust air flow i with a single gravity damper failed open. Precautions will_be taken in the test l

. procedure to prevent the motor for the secured H&V room exhaust fan from tripping.  !

Thus the probability of an equipment malfunction will not be increased. No protective ,

boundaries'or margin of safety will be affected by this test.

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L . 3. STP-01160647. SVS Batterv Fan Roc m Test with Sinale Failure i u ,

{ DESCRIPTION l

L This test will determine the effect of a single active failure (failure to close) of a gravity l damper on the safety function of the Battery Room Exhaust system. In addition, this  !

test will also balance Battery Room exhaust flow rates to within design values.

l: REASON FOR CHANGE To verify that a single active failure c,f a Battery Room exhaust fan grsvity damper or Battery Fan Room exhaust fan gravity damper will not prevent the system from j performing its safety related function. l l

.EAFETY EVALUATION ,

l This test will result in a momentary decrease in Battery Room and Battery Fan Room exhaust flow. However, an engineering calculation determined that with no ventilaGon, i it will take 58 hours6.712963e-4 days <br />0.0161 hours <br />9.589947e-5 weeks <br />2.2069e-5 months <br /> for the 'A' and 'B' Battery Rooms to reach a combustible level of i

hydrogen. It will take even longer for the 'AB' Battery Room (86 hours9.953704e-4 days <br />0.0239 hours <br />1.421958e-4 weeks <br />3.2723e-5 months <br />) and the  !

Computer Battery Room (269 hours0.00311 days <br />0.0747 hours <br />4.447751e-4 weeks <br />1.023545e-4 months <br />), To preclude an increase in probability or

, consequences of an accident, this test limits the cumulative time for reduced exhaust

[ _ flow to 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. The secured fan will be declared inoperable during the test. Since a l= single fan is capable of exhausting the room by itself, there will be no increase in the l probability or consequences of an equipment failure. Running the test in the approved j -. configuration does not create a new accident or equipment malfunction. No protective j- boundary will be affected by this test and no matgin of safety will be reduced.

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4. STP-01162204. Functional Test of the DCT Fans ESFAS Start from the Auxiliarv ,

Control Room i DESCRIPTION i'

This special test will verify operation of the S6X1 contact to auto start the DCT fans durirg simulated LOOP concurrent with transfer of operational control to the Remote ,

Sh'.tdown Panel (LCP-43).

RE3 SON FOR CHANGE l

. Testing of this contact is required by Generic Letter 96-01.  !

SAFETY EVALUATION p

No USQ is created by performing this test. Failure of either the CCW or ACCW systems will not initiate any accident. However, these systems are required to mitigate the consequences of a LOCA or a MSLB. To ensure no accident or equipment is affected and no margin of safety reduced during this test, the affected trains of CCW and ACCW will be considered inoperable. The 100% capacity, redundant trains of CCW and ACCW will remain operable and available to mitigate the consequences of  !

any accident or event. No new connections or system intera,ctions will be required and i no protective boundary will be affected.

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5. STP-99003492. Acceptance Test for DC-3492. Revision 1 DESCRIPTION The CCW Filtration Skid installed by DC-3492 will be placed into service to ensure that the components r~mally cooled by CCW are not adversely affected. CC-413A will be failed open to simulate operation of an EDG. OP-903-118 for CC-200A/CC-727 and CC-200B/CC-563 will be performed to verify that the amount of water diverted to the CCW filters does not adversely affect system operation when the non-safety header isolation valves are closed during quarterly IST valve stroke testing.

. REASON FOR CHANGE The filtration skid installed by DC-3492 requires testing to ensure that the flow diverted to the filters does not adversely affect the components normally cooled by CCW.

SAFETY EVALUATION Placing the CCW filtration skid into service to perform this test will have no affect on the likelihood of a LOCA or MSLB occurring or on the consequences of either accident.

The CCW non-safety header is automatically isolated from the CCW 'A' and 'B' headers; therefore, this test will have no adverse impact on the accident response of the CCW system. The CCW filtration skid was installed on the CCW non-safety header in place of the abandoned waste concentrator. Since the CCW non-safety header automatically isolates if the header leak exceeds makeup capability, no new accidents are created. No protective boundaries are changed and no margin of safety reduced by this special test.

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