RC-18-0080, Virgil C. Summer Nuclear Station, Unit 1, Updated Final Safety Analysis Report, Chapter 14, Initial Tests and Operation

Virgil C. Summer Nuclear Station, Unit 1, Updated Final Safety Analysis Report, Chapter 14, Initial Tests and Operation

ML18221A211
Person / Time
Site:	Summer
Issue date:	05/31/2018
From:	South Carolina Electric & Gas Co
To:	Office of Nuclear Reactor Regulation
Shared Package
ML18221A142	List: ML18221A134 ML18221A145 ML18221A147 ML18221A151 ML18221A152 ML18221A153 ML18221A154 ML18221A155 ML18221A156 ML18221A157 ML18221A158 ML18221A159 ML18221A160 ML18221A161 ML18221A162 ML18221A165 ML18221A166 ML18221A167 ML18221A168 ML18221A169 ML18221A171 ML18221A172 ML18221A175 ML18221A178 ML18221A180 ML18221A181 ML18221A182 ML18221A206 ML18221A210 ML18221A211 ML18221A212 ML18221A213 ML18221A214 ML18221A215 RC-18-0080, Updated Final Safety Analysis Report, Chapter 9, Figures 9.5-1 Through 9.5-11 - Redacted RC-18-0080, Virgil C. Summer Nuclear Station Unit 1 Submittal of Updated Final Safety Analysis Report, Updated Through May 31, 2018, Cover Sheet RC-18-0080, Virgil C. Summer Nuclear Station, Unit 1, Updated Final Safety Analysis Report, Approved FSAR Revision Notices RC-18-0080, Virgil C. Summer Nuclear Station, Unit 1, Updated Final Safety Analysis Report, Chapter 1, Introduction and General Description of Plant RC-18-0080, Virgil C. Summer Nuclear Station, Unit 1, Updated Final Safety Analysis Report, Chapter 10.0, Steam and Power Conversion System RC-18-0080, Virgil C. Summer Nuclear Station, Unit 1, Updated Final Safety Analysis Report, Chapter 11.0, Radioactive Waste Management RC-18-0080, Virgil C. Summer Nuclear Station, Unit 1, Updated Final Safety Analysis Report, Chapter 12.0, Radiation Protection RC-18-0080, Virgil C. Summer Nuclear Station, Unit 1, Updated Final Safety Analysis Report, Chapter 13, Conduct of Operations RC-18-0080, Virgil C. Summer Nuclear Station, Unit 1, Updated Final Safety Analysis Report, Chapter 14, Initial Tests and Operation RC-18-0080, Virgil C. Summer Nuclear Station, Unit 1, Updated Final Safety Analysis Report, Chapter 15, Accident Analyses RC-18-0080, Virgil C. Summer Nuclear Station, Unit 1, Updated Final Safety Analysis Report, Chapter 16, Technical Specifications RC-18-0080, Virgil C. Summer Nuclear Station, Unit 1, Updated Final Safety Analysis Report, Chapter 17, Quality Assurance RC-18-0080, Virgil C. Summer Nuclear Station, Unit 1, Updated Final Safety Analysis Report, Chapter 18, Aging Management Programs and Activities RC-18-0080, Virgil C. Summer Nuclear Station, Unit 1, Updated Final Safety Analysis Report, Chapter 3.0, Design of Structures, Components, Equipment and Systems RC-18-0080, Virgil C. Summer Nuclear Station, Unit 1, Updated Final Safety Analysis Report, Chapter 4.0, Reactor RC-18-0080, Virgil C. Summer Nuclear Station, Unit 1, Updated Final Safety Analysis Report, Chapter 5.0, Reactor Coolant System ... further results
References
RC-18-0080
Download: ML18221A211 (155)

Text

TABLE OF CONTENTS Section Title Page 14.0 INITIAL TESTS AND OPERATION 14.1-1 14.1 TEST PROGRAM 14.1-1 14.1.1 ADMINISTRATIVE PROCEDURES (TESTING) 14.1-2 14.1.1.1 Development of Procedures 14.1-2 14.1.1.2 Execution of Test Procedures 14.1-5 14.1.1.3 Review, Evaluation, and Approval of Test Results 14.1-6 14.1.1.4 Personnel Responsibilities and Qualifications 14.1-6 14.1.1.5 Test Records 14.1-10 14.1.2 ADMINISTRATIVE PROCEDUR ES (MODIFICATIONS) 14.1-10 14.1.2.1 Test Initiated Changes for Systems and Components 14.1-11 14.1.2.2 Test Procedure Change and Modifications 14.1-11 14.1.3 Test Objectives and Procedures 14.1-12 14.1.3.1 Phase I (Construction, Component and Subsystem Functional Tests) 14.1-12 14.1.3.2 Phase II (Pre-Operational Tests) 14.1-12 14.1.3.3 Phase III (Fuel Loading and Pre-Critical Test) 14.1-15 14.1.3.4 Phase III (Power Ascension Test) 14.1-16 14.1.4 Core Loading and Initial Operation 14.1-17 14.1.4.1 Core Loading 14.1-17 14.1.4.2 Postloading Tests 14.1-19 14.1.4.3 Initial Criticality 14.1-20 14.1.4.4 Low Power Testing 14.1-20 14.1.4.5 Power Ascension 14.1-21 14.1.5 ADMINISTRATIVE PROCEDURES (SYSTEM OPERATION) 14.1-22 14.2 AUGMENTATION OF STAFF FOR INITIAL TEST AND OPERATION 14.2-1 14.2.1 ORGANIZATIONAL FUNCTIONS, RESPONSIBILITIES AND AUTHORITIES 14.2-1 14.2.1.1 Plant Operating Staff 14.2-1 14.2.1.2 Westinghouse Electric Corporation 14.2-1 14.2.1.3 Gilbert Associates, Inc.

14.2-1 14.2.2 INTERRELATIONSHIPS AND INTERFACES 14.2-2 14.2.3 PERSONNEL FUNCTIONS, RESPONSIBILITIES AND AUTHORITIES 14.2-2 14.2.3.1 Plant Staff 14-2-2 14.2.3.2 Westinghouse Electric Corporation 14.2-2 14.2.3.3 Gilbert Associates, Inc. 14-2-2 14.2.4 PERSONNEL QUALIFICATIONS 14.2-2 14-i Reformatted Per Amendment 02-01 LIST OF TABLES Table Title Page 14.1-1 Fire Protection System 14.1-23 14.1-2 Containment Isolation Valves Leakage Rate Test 14.1-24 14.1-3 Containment Penetrations Leakage Rate Test 14.1-25 14.1-4 Containment Air Locks Leakage Rate Test 14.1-26 14.1-5 Reactor Building Structural Acceptance Test 14.1-27 14.1-6 Containment Integrated Leakage Rate Test 14.1-28 14.1-7 Unit Auxiliary, Emergency Auxiliary and Engineered Safety Features Transformers 14.1-29 14.1-8 7200 Volt Electrical System 14.1-30 14.1-9 480 Volt Buses 14.1-32 14.1-10 480 Volt Motor Control Centers 14.1-33 14.1-11 120 Volt A-C 14.1-34 14.1-12 D-C System 14.1-35 14.1-13 Plant Paging and Communication System 14.1-37 14.1-14 Diesel Fuel Oil Transfer and Storage System 14.1-38 14.1-15 Emergency Diesel Generators 14.1-39 14.1-16 Response to Loss of Instrument Air 14.1-41 14.1-17 Auxiliary Building Ventilation Sy stem (Radioactive Portion) 14.1-42 14.1-18 Fuel Handling Building Ventilation System (Radioactive Portion) 14.1-43 14.1-19 Process and Area Radiation Monitoring System 14.1-44 14.1-20 Component Cooling Water System 14.1-45 14.1-21 Service Water System 14.1-46 14.1-22 Boric Acid Batching and Transfer System 14.1-47 14.1-23 Heat Tracing for Boron Injection Tank and Associated Piping 14.1-48 14.1-24 Reactor Building Cooling System 14.1-49 14.1-25 Chemical and Volume Control System 14.1-50 14.1-26 Pressurizer Relief Tank 14.1-51 14.1-27 Reactor Coolant System Heatup for Hot Functional Testing 14.1-52 14.1-28 Systems Thermal Expansion 14.1-54 14.1-29 Hot Functional Testing 14.1-55 14.1-30 Emergency Feedwater System 14.1-56 14.1-31 Engineered Safety Features Circuitry 14.1-57 14.1-32 Nuclear Sampling System 14.1-58 14-ii Reformatted Per Amendment 02-01 LIST OF TABLES (Continued)

Table Title Page 14.1-33 Reactor Coolant System Cooldown from Hot Functional Testing 14.1-59 14.1-34 Residual Heat Removal System Flow Tests 14.1-60 14.1-35 Nuclear Instrumentation System 14.1-62 14.1-36 Control Building Ventilation Systems 14.1-63 14.1-37 Core Loading Instrumentation 14.1-65 14.1-38 Reactor Components and Fuel Handling Tools and Fixtures 14.1-66 14.1-39 Fuel Transfer System 14.1-67 14.1-40 Safety Injection High Head Flow Balancing Test 14.1-68 14.1-41 Safety Injection Accumulator Blowdown Test 14.1-70 14.1-42 Spent Fuel Cooling System 14.1-71 14.1-42a Spent Fuel Cooling System 14.1-72 14.1-43 Containment Isolation System 14.1-73 14.1-44 Reactor Protection Operational Check 14.1-74 14.1-45 Engineered Safety Features System Operational Check 14.1-75 14.1-46 Integrated Engineered Safety Features Test 14.1-76 14.1-47 Reactor Building Spray System 14.1-77 14.1-48 Leak Detection Monitoring System 14.1-79 14.1-49 Post Accident Hydrogen Removal System 14.1-80 14.1-50 Radioactive Waste Disposal System 14.1-81 14.1-51 Boron Thermal Regeneration System 14.1-83 14.1-52 Reactor Protection System Time Response Measurement 14.1-84 14.1-53 Initial Fuel Loading 14.1-86 14.1-54 Incore Movable Detectors 14.1-87 14.1-55 Rod Drop Time Measurement 14.1-88 14.1-56 Rod Drive Mechanism Timing 14.1-89 14.1-57 Rod Position Indication 14.1-90 14.1-58 Reactor Coolant System Flow Measurement 14.1-91 14.1-59 Reactor Coolant System Flow Coastdown 14.1-92 14.1-60 Resistance Temperature Detector Bypass Loop Flow Verification 14.1-93 14.1-61 Reactor Vessel O-Ring Leak Test 14.1-94 14.1-62 Pressurizer Spray and Heater Capability and Setting Continuous Spray Flow 14.1-95 14-iii Reformatted Per Amendment 02-01 LIST OF TABLES (Continued)

Table Title Page 14.1-63 Water Quality Test 14.1-96 14.1-64 Initial Criticality 14.1-97 14.1-65 Low Power Test 14.1-98 14.1-65a Augmented Low-Power Test 14.1-101 14.1-66 Incore Movable Detector and Thermocouple Mapping at Power 14.1-103 14.1-67 Power Coefficient and Power Defect Measurement 14.1-104 14.1-68 Effluent Radiation Monitor Test 14.1-105 14.1-69 Radiation Shielding Survey 14.1-106 14.1-70 Process Computer 14.1-107 14.1-71 Thermal Power Measurements and Instrument Calibration 14.1-108 14.1-72 Automatic Control Systems Checkout 14.1-109 14.1-73 Plant Response to Step Load Changes 14.1-110 14.1-74 Pseudo Rod Ejection Test 14.1-111 14.1-75 Rod Drop Test 14.1-112 14.1-76 Below-Bank Rod Test 14.1-113 14.1-77 Plant Trip From 100% Power 14.1-114 14.1-78 Loss of Offsite Power 14.1-115 14.1-79 Shutdown from Outside the Control Room 14.1-116 14.1-79a Emergency Lighting 14.1-117 14.1-79b Heat Tracing for Safety-Related Outdoor Piping 14.1-118 14.1-79c Pressure Boundary Integrity Test 14.1-119 14.1-79d Seismic Instrumentation 14.1-120 14.1-80 Power Ascension Test Program 14.1-121 14.1-81 Initial Test Program Schedule 14.1-122 14.1-82 Generic Flush Procedure 14.1-123 14.1-83 Generic Hydrostatic/Pneumatic Test 14.1-124 14.1-84 Instrument Control Procedure 14.1-125 14.1-85 Functional Test 14.1-126 14.1-86 Steam Generator Power Operated Relief Valve 14.1-127 14.1-87 Condensate System 14.1-128 14.1-88 Feedwater System 14.1-129 14.1-89 Main Condenser Dump Valves 14.1-130 14-iv Reformatted Per Amendment 02-01 LIST OF TABLES (Continued)

Table Title Page 14.1-90 Circulating Water System 14.1-131 14.1-91 Chemical Feed System 14.1-132 14.1-92 Nuclear Blowdown Processing System 14.1-133 14.1-93 Control Rod Drive 14.1-134 14.1-94 Miscellaneous Plant Drains 14.1-135 14.1-95 Fuel Handling Building Pool Liner Leak Test 14.1-136 14.1-96 Reactor Building Ventilation "Post Accident Operation" 14.1-137 14.1-97 ESF Equipment Rooms Cooling Systems 14.1-138 14.1-98 S. I. Accumulator Discharge Valve Functional Test 14.1-139 14.1-99 ECCS Check Valve Leak Testing System Operational Test 14.1-141 14.1-100 S. I. Accumulator Check Valve Hot Operational Test 14.1-142 14.1-101 Instrument Air System 14.1-143 14.1-102 Pressurizer Pressure and Level Control 14.1-144 14-v Reformatted Per Amendment 02-01 LIST OF EFFECTIVE PAGES (LEP)

The following list delineates pages to Chapter 14 of the Virgil C. Summer Nuclear Station Final Safety Analysis Report which are currently through April 2004. The latest changes to pages and figures are indicated below by Revision Number (RN) in the Amendment column along with the Revision Number and date for each page and figure included in the Final Safety Analysis Report.

Page / Fig.No. Amend. No.

Date Page / Fig.No. Amend. No.

Date 14-vi Reset April 2004 Page 14-i 02-01 May 2002 14-ii 02-01 May 2002 14-iii 02-01 May 2002 14-iv 02-01 May 2002 14-v 02-01 May 2002 14-vi Reset April 2004 14-vii Reset April 2004 14-viii Reset April 2004 Page 14-1-1 RN04-007 March 2004 14.1-2 02-01 May 2002 14.1-3 02-01 May 2002 14.1-4 02-01 May 2002 14.1-5 02-01 May 2002 14.1-6 02-01 May 2002 14.1-7 02-01 May 2002 14.1-8 02-01 May 2002 14.1-9 02-01 May 2002 14.1-10 02-01 May 2002 14.1-11 02-01 May 2002 14.1-12 02-01 May 2002 14.1-13 02-01 May 2002 14.1-14 02-01 May 2002 14.1-15 02-01 May 2002 14.1-16 02-01 May 2002 14.1-17 02-01 May 2002 14.1-18 02-01 May 2002 14.1-19 02-01 May 2002 14.1-20 02-01 May 2002 14.1-21 02-01 May 2002 14.1-22 02-01 May 2002 14.1-23 97-01 August 1997 14.1-24 97-01 August 1997 14.1-25 97-01 August 1997 Page 14.1-26 97-01 August 1997 14.1-27 97-01 August 1997 14.1-28 97-01 August 1997 14.1-29 97-01 August 1997 14.1-30 97-01 August 1997 14.1-31 97-01 August 1997 14.1-32 97-01 August 1997 14.1-33 97-01 August 1997 14.1-34 97-01 August 1997 14.1-35 97-01 August 1997 14.1-36 97-01 August 1997 14.1-37 97-01 August 1997 14.1-38 97-01 August 1997 14.1-39 97-01 August 1997 14.1-40 97-01 August 1997 14.1-41 97-01 August 1997 14.1-42 97-01 August 1997 14.1-43 97-01 August 1997 14.1-44 97-01 August 1997 14.1-45 97-01 August 1997 14.1-46 97-01 August 1997 14.1-47 97-01 August 1997 14.1-48 97-01 August 1997 14.1-49 97-01 August 1997 14.1-50 97-01 August 1997 14.1-51 97-01 August 1997 14.1-52 97-01 August 1997 14.1-53 97-01 August 1997 14.1-54 97-01 August 1997 14.1-55 97-01 August 1997 14.1-56 97-01 August 1997 14.1-57 97-01 August 1997 14.1-58 97-01 August 1997 LIST OF EFFECTIVE PAGES (Continued)

Page / Fig.No. Amend. No.

Date Page / Fig.No. Amend. No.

Date Page 14.1-59 02-01 May 2002 14.1-60 02-01 May 2002 14.1-61 02-01 May 2002 14.1-62 97-01 August 1997 14.1-63 97-01 August 1997 14.1-64 97-01 August 1997 14.1-65 97-01 August 1997 14.1-66 97-01 August 1997 14.1-67 02-01 May 2002 14.1-68 02-01 May 2002 14.1-69 02-01 May 2002 14.1-70 02-01 May 2002 14.1-71 97-01 August 1997 14.1-72 97-01 August 1997 14.1-73 97-01 August 1997 14.1-74 97-01 August 1997 14.1-75 97-01 August 1997 14.1-76 97-01 August 1997 14.1-77 97-01 August 1997 14.1-78 97-01 August 1997 14.1-79 97-01 August 1997 14.1-80 97-01 August 1997 14.1-81 97-01 August 1997 14.1-82 97-01 August 1997 14.1-83 97-01 August 1997 14.1-84 97-01 August 1997 14.1-85 97-01 August 1997 14.1-86 97-01 August 1997 14.1-87 97-01 August 1997 14.1-88 97-01 August 1997 14.1-89 97-01 August 1997 14.1-90 97-01 August 1997 14.1-91 97-01 August 1997 14.1-92 97-01 August 1997 14.1-93 97-01 August 1997 14.1-94 97-01 August 1997 14.1-95 97-01 August 1997 Page 14.1-96 97-01 August 1997 14.1-97 97-01 August 1997 14.1-98 97-01 August 1997 14.1-99 97-01 August 1997 14.1-100 97-01 August 1997 14.1-101 97-01 August 1997 14.1-102 97-01 August 1997 14.1-103 97-01 August 1997 14.1-104 97-01 August 1997 14.1-105 97-01 August 1997 14.1-106 97-01 August 1997 14.1-107 97-01 August 1997 14.1-108 97-01 August 1997 14.1-109 97-01 August 1997 14.1-110 97-01 August 1997 14.1-111 97-01 August 1997 14.1-112 97-01 August 1997 14.1-113 97-01 August 1997 14.1-114 97-01 August 1997 14.1-115 97-01 August 1997 14.1-116 97-01 August 1997 14.1-117 97-01 August 1997 14.1-118 97-01 August 1997 14.1-119 97-01 August 1997 14.1-120 97-01 August 1997 14.1-121 97-01 August 1997 14.1-122 97-01 August 1997 14.1-123 97-01 August 1997 14.1-124 97-01 August 1997 14.1-125 97-01 August 1997 14.1-126 97-01 August 1997 14.1-127 97-01 August 1997 14.1-128 97-01 August 1997 14.1-129 97-01 August 1997 14.1-130 97-01 August 1997 14.1-131 97-01 August 1997 14.1-132 97-01 August 1997 14-vii Reset April 2004 LIST OF EFFECTIVE PAGES (Continued)

Page / Fig.No. Amend. No.

Date Page / Fig.No. Amend. No.

Date Page 14.1-133 97-01 August 1997 14.1-134 97-01 August 1997 14.1-135 97-01 August 1997 14.1-136 97-01 August 1997 14.1-137 97-01 August 1997 14.1-138 97-01 August 1997 14.1-139 97-01 August 1997 14.1-140 97-01 August 1997 14.1-141 02-01 May 2002 14.1-142 02-01 May 2002 14.1-143 97-01 August 1997 14.1-144 97-01 August 1997 14.1-145 97-01 August 1997 14.2-1 97-01 August 1997 14.2-2 97-01 August 1997 14-viii Reset April 2004 RN 04-007 NOTE 14.0 Chapter 14, in its entirety, is being retained for historical purposes only.

14.0 INITIAL TESTS AND OPERATION An extensive test program is conducted by SCE&G to verify that the systems, components, and structures which make up Virgil C. Summer Nuclear Station will perform as they were designed.

The overall test objective is to assure that the operation of Virgil C. Summer Nuclear Station will not endanger the health and safety of the public. The ultimate responsibility fo r the initial tests and operation belongs to SCE&G. 14.1 TEST PROGRAM The initial test program of Vi rgil C. Summer Nuclear Station is in three phases; Phase I, component or construction tests; Phase II, acceptance and preoperational tests; and, Phase III, core loading and initial startup tests.

Component or construction tests are performed to ensure that components and subsystems meet their functional requirements. The Phase I test program consists of hydrostatic test, flushing, electrical che cks, initial operation of equipment, and other specific tests on individual items of equipment.

Acceptance and preoperational tests demonstrate that systems and structures can perform their intended functions.

These tests start with system or subsystem turnovers from construction and continue through core loading.

Core loading and initial star tup tests begin only after the plant operating license has been issued. It begins with core loading and continues through initial criticality, ascension to power, and is complete when the plant is fully licensed for commercial operation. The results of t hese tests show that the plant follows its predicted nuclear parameters and can be operated at rated capacity without endangering the health or safety of the public.

Some of the tests described in Regulator y Guide 1.68 are considered preoperational tests, but because of their requirements, are performed during star tup testing. An example is scram performance of control rods. This can only be done after the core is loaded and the vessel head is on; although th is test appears in the preoperational section of the guide.

To assure quality control, procedures used in testing are written and approved in accordance with Section 14.1.

1. Changes made to appr oved procedures are in accordance with Section 14.1.

2. The use of plant pr ocedures is discussed in Section 14.1.5. Compliance with regulatory guides is discussed in Appendix 3A.

14.1-1 Reformatted Per Amendment 02-01 The Manager Virgil C. Summer Nuclear Stat ion is responsible for the development, administration, and conduct of the test progr am. The test program is conducted by a startup group which is organized specifically for performing t he initial tests and startup of the plant. After commercial operation this group will be disbanded. The startup group is described in more det ail in Section 14.1.1.4.

The initial test program schedule is presented as Table 14.1-81.

14.1.1 ADMINISTRATIVE PROCEDURES (TESTING) 14.1.1.1 Development of Procedures The preparation of Phase I, II, and III test procedures is t he responsibility of the Plant Manager or his designee (qualification described in Sect ion 13.1.3). The Startup Supervisor may assume these duties for Phase I, II, and III while the startup group is active. Procedures which are contracted to other organizations are written in accordance with the terms of the contract but are reviewed as if written by SCE&G Nuclear Operations. Phase II and III test procedures as described in this Chapter are approved by the Manager Virgil C. Summer Nu clear Station. Acceptance criteria and performance requirements are determined by SCE&G Nuclear Operations by evaluating data furnished by Westinghouse, Gilbert, other vendors, and contractors. Information received from other departments of SCE&G is given close evaluation in the determination of acceptance criteria and performance requirements.

The preparation of Phase I Te st Instructions is the responsibility of the Startup Supervisor, who will designate this responsibilit y to the Test Supervisors. The Test Supervisor will prepare the original test in struction following the guidelines and test methods set forth in generic instructions. If more guidelines are considered necessary, the Test Supervisor will follow specified formats depending on the type of instruction being written (Hydro, Electrical, I&C, Flush). During preparation of the Phase I Test Instruction, the Test Supervisor will use the following references as applicable:

Virgil C. Summer Nuclear Station, FSAR , Westinghouse System Descriptions, Gilbert Design Descriptions, Westinghouse NSSS Startup Manuals, applicable vendor Instruction Manuals, and other sources as necessary. Afte r preparation of the test instruction is completed, the Test Supervi sor will submit the instruction to the Lead Systems Supervisor, who will review and approve the test instruction.

The preparation of Phase II Test Procedures as described in Section 14.1.3.2 is the responsibility of the Startup supervisor who wil l designate this responsibility to the Test Supervisors. The Test Supervisor(s) prepares the assigned procedure in the prescribed format and submits the procedure for distribution. Copies will be sent to the Westinghouse Startup Represent ative, the Gilbert Associates Startup Representative, the Startup Supervisor, and t he Manager, Virgil C. Summer Nuclear Station, or their designated alternates, each having at least the qualifications of a Lead Systems Supervisor as described in Section 14.1.1.4, for review. The Startup Supervisor may also send the procedure to designated groups or persons for review. These designated groups or persons may include Nuclear Engineering personnel, Quality Assurance 14.1-2 Reformatted Per Amendment 02-01 personnel, the Assistant Managers Operations, Maintenance Services, and/or Technical Support. Minimum qualifications for Nuclear Engineering and plant personnel assigned to review Phase II Test Procedures are provided in Section 13.1.

The reviewers examine the procedure for co rrectness and comment as necessary. The procedure with comments is then returned to the originator for resolution/incorporation of comments. When comments are resolved/incorporated, the procedure is forwarded to the Startup Supervisor. The Startup Supervisor or his designee signs the procedure recommending approval. The procedure is tr ansmitted to the Plan t Manager or his designee for his approval.

The preparation of Phase III Test Procedures will be the responsibility of the Assistant Manager Technical Support or his designated alternate, who will assign the

responsibility for the preparation of individu al test procedures to plant or support personnel. The originator(s) prepares the assigned procedur e in the prescribed format and submits the procedure for distribution. The procedure(s) will be reviewed and approved as per Plant Administrative Procedures (FSAR Chapter 13 and Technical Specifications Section 6).

Test procedures are available to NRC inspection personnel 30 days prior to the scheduled performance of the activity, but not less than 90 days prior to the scheduled core loading date. NRC possession of procedures shall not impede revision, review, or refinement of the procedures.

The format for Phase II and III te st procedures is as follows:

TITLE PAGE The title page contains the station name, procedure title and number, revision number, and signature and date for approval.

1.0 PURPOSE

This section contains the general purpose and objectives of the test.

2.0 REFERENCES

This section contains the references us ed to develop the test procedure such as FSAR, system descriptions, operating procedures, applicable standards and codes, drawings, manufacturer's literatur e, etc. Where possible, revision numbers and dates are shown.

14.1-3 Reformatted Per Amendment 02-01

3.0 PREREQUISITES

This section contains title and number of pertinent tests or port ions of tests which must be completed prior to conducting th is test and/or tests which must be carried out concurrently with this test.

This section also contains the plant st atus required to conduc t the tests such as pressure, temperature, levels, etc. and out line of system status, such as special valve lineups, test equipment installation, temporary equipment installation, temporary equipment, and identification of any equipment that should or should not be operating for equipment or personnel safety.

4.0 SPECIAL TEST EQUIPMENT This section lists special equipment, other than normal system equipment and instrumentation, requir ed to conduct the test.

5.0 LIMITATIONS AND PRECAUTIONS This section contains design and safety limits for equipment and any special limitations and precautions needed for safety of personnel or equipment needed to assure that the required re sults are obtained during the test.

6.0 TEST METHOD The test method consists of one or more sections containing step-by-step instructions to accomplish the purpose of the test.

Appropriate inspection points are established to record the accomplishment of the test objectives and test requirements.

Test method ensures that the system or plant is placed in a safe condition after a test has been completed and that nons tandard arrangements are restored to their standard condition and that proper veri fication and records are maintained.

7.0 DATA REQUIREMENTS This section consists of the necessary instructions to assure that the required data is obtained by outlining requirements that can clearly be associated with steps of the test method.

Preplanned data sheets are used as much as possible with provisions for name of the person responsible for recording data, type of observation performed, and acknowledgment of the a cceptability of the data.

The data are compared with the acceptance criteria for the test to determine acceptability.

14.1-4 Reformatted Per Amendment 02-01 8.0 ACCEPTANCE CRITERIA This section lists those qualitative or quantitative requirements or limits contained in the design documents against which test results are evaluated.

14.1.1.2 Execution of Test Procedures Prior to the start of the initial testing program, the Virgil C. Summer Nuclear Station Operations Department arranges itself into a startup group which is more suited for the initial testing program. Th is organization is headed by the Manager Virgil C. Summer Nuclear Station. A Startup Supervisor w ho reports to the Manager Virgil C. Summer Nuclear Station organizes and directs the startup program. T he Startup Supervisor appoints a test supervisor for each test.

More than one test supervisor may be appointed. The test supervisor ensures that t he test is conducted according to the test procedure. Qualified test personnel are assigned by the Startup Supervisor to perform the tests. The personnel perform the test work under the supervision of the test supervisor.

Prior to fuel loading all Phase I (com ponent and construction) tests and Phase II (Pre-Operational and Acceptanc e) tests are scheduled to be conducted and the results reviewed and approved in accordance with approved administrative procedures.

A limited number of test def iciencies may exist at the scheduled core loading date. These test deficiencies will be evaluated and categorized.

Appropriate hold points in the Phase III (Initi al Startup) Test Progr am will be identified. The applicable deficiencies will be resolved prior to proceeding with the Phase III Test Program. During the power ascension phase of testing, certain data are analyzed by the test supervisor and the results are reviewed by the Plant Manager or his designee before power is increased to the next higher plateau. Analysis of data includes checking radial flux for symmetry, verifying that axial flux is within predicted values, checking effluent radioactivity, monitoring system operation, and checking power level and nuclear instrumentation by using a heat balance. Befo re power is increase d to a higher plateau, the high flux trip is reset to a value no greater than 20 percent beyond the next power level. Approval of the Plant Manager or his designat ed alternate must be obtained before increasing power to the next plateau. Approval of the completed procedure documentation is not required for continuat ion of the startup program or power increases.

When conducting the test, the test supervisor ensures that the full intent of the procedure is met. If the te st supervisor determines the need for either a procedure or plant change, it is accomplishe d as outlined in Section 14.1.2.

The test supervisor shall notify the Startup Supervisor when the test has been completed.

14.1-5 Reformatted Per Amendment 02-01 14.1.1.3 Review, Evaluation, and Approval of Test Results After a Phase I test is completed and the data is compiled, the Test Supervisor will review the test results and comp are them to the applic able Acceptance Criteria. If he is satisfied that the test result s are satisfactory, he will submit the test to the Lead Systems Supervisor. The Lead Systems Supervisor will be responsible for reviewing the test results and comparing them to the applicable Acceptance Criteria. If he is satisfied that the test results are acceptable, he will approve the test data. If t he test results are not satisfactory, the Lead Systems Supervisor will recommend to the Startup Supervisor what actions are necessary.

After a Phase II test is completed and the data is compiled, the Startup Supervisor will assign the Westinghouse Star tup Representative, the G ilbert Associates Startup Representative, or their designated alternates, each having qualifications of a Lead Systems Supervisor as described in Section 14.

1.1.4, to review the test results. The assigned personnel will review the test results and compare them to the applicable acceptance criteria specified in the procedure, and forward the test results to the Startup Supervisor. The Startup Supervisor or designated al ternate will then review the test results and, if acceptabl e, sign the test results recommending approval to the Manager Virgil C. Summer Nuclea r Station. The test resu lts are then sent to the Manager Virgil C. Summer Nuclear Station for his review and approval. If test results are not acceptable, the pers onnel assigned to review the data recommends to the Startup Supervisor what actions are necessa ry. These actions are outlined in more detail in Section 14.1.2.1.

After a Phase III Test is completed and the data is compiled, the Lead Phase III Test Supervisor will place the review of the completed test procedure on the agenda of the Plant Safety Review Committee Meeting for consideration for approval of the test results. If the review of the test results during the Plant Safety Review Committee Meeting show that the test results are acceptable, the Plant Sa fety Review Committee Chairman signs a procedure recommending approv al, the procedure is transmitted to the Plant Manager or his designat ed alternate for his approval.

14.1.1.4 Personnel Responsibilities and Qualifications Personnel who will manage, supervis e, or execute any of the Ph ase II or III tests of the initial test program are inco rporated into the Virgil C.

Summer Nuclear Station startup group. This group consists of the M anager, Virgil C. Summer Nuclear Station, a Startup Supervisor, Lead Systems Supervisor, test supervisors and test personnel. To the extent practical, these individuals come from the plant operating, technical, and maintenance personnel. This allows the utilization of the knowledge these persons have acquired from their training and practical experiences. Utilization of plant personnel during testing further enhances their knowledge of the plant and aids in their training.

14.1-6 Reformatted Per Amendment 02-01 SCE&G also utilizes qualified startup personnel from other organizations as members of the startup group. These people constitute a temporary additi on to the plant staff which provides the extra people needed during startup to handle the increased work load.

These additional people may be obtained from contractors for the specific purpose of starting up the Virgil C.

Summer Nuclear Station.

In addition to the startup group, SCE&G may utilize vendor service personnel to provide expert advice and assistance in check out, startup, and testing of their equipment.

Technical services are available from Westinghouse, Gilbert, and Daniel.

The Manager Virgil C. Summer Nuclear Station, in addition to his other duties outlined in Section 13.1.2.2, has overa ll responsibility for Phase II and Phase III of the initial test program. He is responsible for the development of the test procedures, the administration and conduct of the test, and the review and actions taken on the test results. The qualifications of the Plant Manager are outlined in Section 13.1.3.

The Startup Supervisor is under the supervision of t he Manager Virgil C. Summer Nuclear Station and ensures effective adminis trative control and implementation of the test program. The Startup Supervisor meets one of the follo wing qualifications.

1. A graduate of a four-year engineering or science college or university plus five years of experience in operation, testing, or inspection of power plant, nuclear plant, heavy industrial, or other similar equipment or facilities. At least two years of this experience should be associated with nuclear facilities; or if not, the individual shall have training sufficient to acquaint hi m thoroughly with the safety aspects of a nuclear facility.

2. A high school graduate or equivalent, plus 10 years of experience in operation, testing, or inspection of power plant, nucl ear plant, heavy industrial, or other similar equipment or facilities. At least two years of this ex perience should be associated with nuclear facilities; or if not, the individual shall have training sufficient to acquaint him thoroughly with the safety aspects of a nuclear facility.

See Appendix 3A for details of co mpliance with Regulatory Guide 1.58.

The Startup Supervisor's duties and responsibilities include, but are not limited to the following:

1. Supervision and coordi nation of the activities of the startup group.

2. Responsibility for the master startup schedule.

3. Assignment for review responsibilities for individual test result s of members of the startup group.

14.1-7 Reformatted Per Amendment 02-01

4. Review of test data and test procedure modification in accordance with established administrative procedures.

5. Recommendations to Nuclear Engineering on any request for construction or engineering changes or modifications dete rmined to be necessary by results of a test. 6. Liaison with contractors and vendors and coordination of any activities relative to the test program.

7. The preparation and maintenan ce of the startup manual.

8. The review of Phase II procedures and test results.

A Lead Systems Supervisor has the responsib ility of coordinat ing test personnel assigned to work for him. He supervi ses any training needed for the personnel assigned. A Lead Systems Supervisor may be assigned to be the Test Supervisor on an individual test or any num ber of tests. Lead Systems Supervisors shall meet the qualifications of the Start up Supervisor. Lead Systems Supervisors shall receive indoctrination on startup administrative procedures prior to beginning work.

The Lead Systems Supervisor responsibilities include, but are not limited to the following:

1. Establish system work loads and res ponsibility for Test Supervisors with the Startup Organization.

2. Attend meetings with other groups (i.e. construction) as needed.

3. Coordinate turned over system work schedules.

4. Review system turnover packages for completeness of documentation.

5. Review turnover packages for closure and completeness to ens ure that the system is ready for testing.

6. Coordinate activities of supplie rs/contractors (when applicable) onsite.

7. Coordinate all flushing and cleani ng efforts on his assigned systems.

8. Assure that all system deficiencies (design or test) are properly documented, reported, and corrected as expediti ously as possible through the Startup Supervisor.

9. Provide technical guidance for des igned systems of testing and operation as required.

14.1-8 Reformatted Per Amendment 02-01

10. Assure that the necessary test equipment supplies are available.

11. He is responsible for coordination of the activities of the Startup Test Supervisors assigned to his group. He will assist t he Startup Supervisor in scheduling tests and work completion.

12. The review of Phase II procedures and test results.

13. Assign Test Supervisors to t he preparation of test procedures.

A test supervisor has the responsibility of coordinating the test personnel assigned to work for him. He supervises any training ne eded for the specific test assigned. A test supervisor may be assigned to be the test supervisor on an individual test or any number of tests. He is assigned a copy of the test procedur e on which all data and acknowledgments are recorded.

Test supervisors shall meet one of the following qualifications:

1. A graduate of four-year engi neering or science college or university, plus two years experience in operations, testing, or inspection of power plant, nuclear plant, heavy industrial, or other similar equipment or fac ilities. (For Phase III, at least one of the two years experience must be applicab le nuclear power plant experience).

2. A high school graduate or equivalent, plus four years experience in operations, testing, or inspection of power plant, nucl ear plant, heavy industrial, or other similar equipment or facilities. (For Phase III, five years experience, at least two of which must be applicable nuclear power plant experience). Test supervisors shall receive indoctrination on startup administr ative procedures prior to beginning work.

See Appendix 3A for details of co mpliance with Regulatory Guide 1.58.

A Test Supervisor's duties and responsibilit ies include, but ar e not limited to, the following:

1. Coordinate with construction to have the necessary temporary piping installed to facilitate flushing activities and other tem porary arrangements to facilitate startup.

2. Carry out system inspection and participate in walkdowns.

3. Assure that the necessary Electrical, Mechanical, and Instrument checkouts have been completed and documented.

4. In the initial operation (if applicable) obt ain and record the necessary data. Report, record, and correct deficiencies as required.

14.1-9 Reformatted Per Amendment 02-01

5. Present test reports and system data files to the Lead System Supervisor for inclusion in the startup files.

6. Document all deficiencies noted, corrections made, and verifications performed during testing activities.

7. Document all print changes, etc. to ensure final prints are correct and data file is complete.

8. Ensure that approved flushing proc edures are implemen ted and allowed.

9. Calculate the approximate quant ities of water required to minimize the use of water and coordinate the use of pum ps and the availability of Electrical Power.

10. Walkdown the system with the cognizant personnel using the flushing P & ID's to check valve line up prior to flushing. Supervise the flushing operation.

11. Check for and document, by signing off the steps of the pr ocedure, such as metering devices, orifice plates, valve internals, temporary strainers, blind flanges, piping, and the isolation of sensitive instruments that ar e removed from the system to facilitate flushing.

12. Verify that the mechanical items are being flushed in accordance with the limits and precautions specified by the procedur e so that contaminants and/or flow velocities will not adversely affect subsequent operation or damage the equipment.

13. In conjunction with SCE&G Chemistry Depar tment verify that the proper chemicals at the designated concentration and temperature are being used in the system

lay-up condition.

14.1.1.5 Test Records After a test is completed and the results are approved, the master test copy and other permanent information relative to the test will be filed as part of the plant's permanent file in the permanent records room as appropriate. This gives plant personnel access to information on systems, components, etc., from plant startup to serve as a baseline for evaluating performance at any futu re time in the plant's life.

14.1.2 ADMINISTRATIVE PROC EDURES (MODIFICATIONS)

When the need arises for changes in plant systems or components or in test procedures, administrative procedures are followed.

14.1-10 Reformatted Per Amendment 02-01 14.1.2.1 Test Initiated Chan ges for Systems and Components A minor change in a system, component, or c ontrol is one which does not involve design intent of the system.

If, during a test, there arises a need for a minor change in the field, the change, authorized by the test supervisor, is made by qualified personnel.

Minor changes are indicated with the test data. These minor changes must be approved by the Lead Systems Supervisor befor e the test results may be approved. The test supervisor initiates changes in documents which the minor change affects and notifies appropriate design organization for review.

A major change in a system, component, or control is one which could alter design intent of the system. When personnel designated to review test results, the test supervisor or the Startup Supervisor, indicate the necessity of major changes, the Startup Supervisor assigns qualified person nel to review the data at hand and make recommendations to SCE&G Production Engine ering for resolution and approval. The Production Engineering personnel studying the necessity of a major change may seek the assistance of vendor, contra ctor, or others. Before resulting modifications are started they must have the written approval of the Manager Virg il C. Summer Nuclear Station or his designee. Additional tests or retests shall be rescheduled by the Startup Supervisor. These retests shall be conducted using approved procedures. Modifications or maintenance initiated by test results and the reason they were initiated are noted by the Startup Supervisor.

The Manager Virgil C. Summer Nuclear Stati on is notified when test results are not acceptable. Vendors, contra ctors, and others as necessary are notified of any developments and followup activities concerning unacceptable test results.

14.1.2.2 Test Procedure Change and Modifications Modifications to approved Phase II test procedur es are designated either minor or major

modifications. A minor modification does not involve changes in scope, intent, or acceptance limits. Minor modification may be effected with approval of the test supervisor. The modification mu st be noted on the test procedure.

Other Phase II procedure modifications are des ignated as major modi fications. These modifications are incorporated into the test procedure by an individual designated by the

Startup Supervisor. (If t he startup group is not in e ffect, the Technical Support Engineering Supervisor assumes his duties).

The revised procedure goes through the review and approval steps of an original procedure.

Phase III procedures will be changed and modified in accordance with Technical

Specifications.

14.1-11 Reformatted Per Amendment 02-01 14.1.3 TEST OBJECTIVES AND PROCEDURES 14.1.3.1 Phase I (Construction, Component, and Subsystem Functional Tests)

NOTE: Phase I tests are generic in nat ure and are written to demonstrate individual components or subsystems meet their functional requirements.

The test abstracts furnished below de scribe the general objectives, test method, and acceptance criteria for each type of component or subsystem being tested utilizing a Phase I Test Procedure to demonstrate that the component or subsystem meets its functional requirements.

1. Generic Flush Proced ure (see Table 14.1-82).

2. Generic Hydrostatic/Pneumatic Test Procedure (see Table 14.1-83).

3. Instrument Control Procedure (see Table 14.1-84).

4. Generic Functional Test Procedure (see Table 14.1-85).

14.1.3.2 Phase II (Pre-Operational Tests)

1. Fire Protection System (see Table 14.1-1).

2. Containment Isolation Valves L eakage Rate Test (see Table 14.1-2).

3. Containment Penetrations Leaka ge Rate Test (see Table 14.1-3).

4. Containment Air Locks Leakage Ra te Test (see Table 14.1-4).

5. Reactor Building Structural A cceptance Test (see Table 14.1-5).

6. Containment Integrated Leakage Rate Test (see Table 14.1-6).

7. Unit Auxiliary, Emergency Auxiliary, and Engineered Safety Features Transformers (see Table 14.1-7).

8. 7200 Volt Electrical System (see Table 14.1-8).

9. 480 Volt Buses (see Table 14.1-9).

10. 480 Volt Motor Control C enters (see Table 14.1-10).

11. 120 Volt A-C (see Table 14.1-11).

12. D-C System (see Table 14.1-12).

14.1-12 Reformatted Per Amendment 02-01

13. Plant Paging and Communication System (see Table 14.1-13).

14. Diesel Fuel Oil Transfer and St orage System (see Table 14.1-14).

15. Emergency Diesel Generat ors (see Table 14.1-15).

16. Response to Loss of Instru ment Air (see Table 14.1-16).

17. Auxiliary Building Ventilation System (Radioactive Portion) (see Table 14.1-17).

18. Fuel Handling Building Ventilation System (Radioactive Portion) (see Table 14.1-18).

19. Process and Area Radiation Monitoring System (see Table 14.1-19).

20. Component Cooling Water S ystem (see Table 14.1-20).

21. Service Water System (see Table 14.1-21).

22. Boric Acid Batching and Trans fer System (see Table 14.1-22).

23. Heat Tracing for Boron Injection Tank and Associated Piping (see Table 14.1-23).

24. Reactor Building Ventilation Systems (see Table 14.1-24).

25. Chemical and Volume Control System (see Table 14.1-25).

26. Pressurizer Relief Tank (see Table 14.1-26).

27. Reactor Coolant System Heatup for Hot Functional Testing (see Table 14.1-27).

28. Systems Thermal Expansion (see Table 14.1-28).

29. Hot Functional Testing (see Table 14.1-29).

30. Emergency Feedwater Syst em (see Table 14.1-30).

31. Engineered Safety Features Circuitry (see Table 14.1-31).

32. Nuclear Sampling System (see Table 14.1-32).

33. Reactor Coolant System Cool down From Hot Functional Testing (see Table 14.1-33).

34. Residual Heat Removal System (see Table 14.1-34).

14.1-13 Reformatted Per Amendment 02-01

35. Nuclear Instrumentation System (see Table 14.1-35).

36. Control Building Ventilation System (see Table 14.1-36).

37. Deleted.

38. Reactor Components and Fuel Handling T ools and Fixtures (see Table 14.1-38).

39. Fuel Transfer System (see Table 14.1-39).

40. Safety Injection Pumps Operat ional Test (see Table 14.1-40).

41. Safety Injection Accumulator Blowdown (see Table 14.1-41).

42. Spent Fuel Cooling System (see Table 14.1-42).

43. Containment Isolation System (see Table 14.1-43).

44. Reactor Protection Operational Check (see Table 14.1-44).

45. Engineered Safety Features System Operational Check (see Table 14.1-45).

46. Integrated Engineered Safety F eatures Test (see Table 14.1-46).

47. Reactor Building Spray System (see Table 14.1-47).

48. Leak Detection Monitoring System (see Table 14.1-48).

49. Post Accident Hydrogen Removal System (see Table 14.1-49).

50. Radioactive Waste Disposal System (see Table 14.1-50).

51. Boron Thermal Regeneration System (see Table 14.1-51).

52. Reactor Protection System Time Response Measurement (s ee Table 14.1-52).

53. Emergency Lighting (see Table 14.1-79a).

54. Heat Tracing for Safety Related Outdoor Piping (see Table 14.1-79b).

55. Pressure Boundary Integrit y Test (see Table 14.1-79c).

56. Seismic Instrumentation (see Table 14.1-79d).

57. Steam Generator Powe r Operated Relief Valv e (see Table 14.1-86).

14.1-14 Reformatted Per Amendment 02-01

58. Condensate System (see Table 14.1-87).

59. Main Condenser Dump Valves (see Table 14.1-89).

60. Circulating Water Syst em (see Table 14.1-90).

61. Chemical Feed System (see Table 14.1-91).

62. Nuclear Blowdown Processi ng System (see Table 14.1-92).

63. Control Rod Drive (see Table 14.1-93).

64. Miscellaneous Plant Drai ns (see Table 14.1-95).

65. Fuel Handling Building Pool Li ner Leak Test (see Table 14.1-95).

66. Reactor Building Ventilation Post A ccident Operation (see Table 14.1-96).

67. ESF Equipment Rooms Coolin g Systems (see Table 14.1-97).

68. S.I. Accumulator Discharge Valve F unctional Test (see Table 14.1-98).

69. ECCS Check Valve Leak Testing Syst em Operational Test (see Table 14.1-99).

70. S.I. Accumulator Check Valve Hot Operational Test (see Table 14.1-100).

71. Instrument Air Syst em (see Table 14.1-101).

72. Pressurizer Pressure and Level Control (see Table 14.1-102).

73. Process Computer (see Table 14.1-70).

14.1.3.3 Phase III (Fuel Loadi ng and Pre-Critical Test)

These tests are listed in the order in whic h they are most likely to be performed:

1. Core Loading Instrumentation (see Table 14.1-37).

2. Initial Fuel Loading (see Table 14.1-53).

3. Incore Movable Detectors (see Table 14.1-54).

4. Rod Drop Time Measurement (see Table 14.1-55).

5. Rod Drive Mechanism Timing (see Table 14.1-56).

6. Rod Position Indication (see Table 14.1-57).

14.1-15 Reformatted Per Amendment 02-01

7. Reactor Coolant System Flow Measurement (see Table 14.1-58).

8. Reactor Coolant System Flow Coastdown (see Table 14.1-59).

9. Resistance Temperature Detector Bypass Loop Flow Verification (see Table 14.1-60).

10. Reactor Vessel O-Ring Leak Test (see Table 14.1-61).

11. Pressurizer Spray and Heater Capabili ty and Setting Continuous Spray Flow (see Table 14.1-62).

12. Water Quality Test (see Table 14.1-63).

14.1.3.4 Phase III (Power Ascension Test)

These tests are listed in the order in wh ich they are most likely to be performed.

1. Initial Criticalit y (see Table 14.1-64).

2. Low Power Test (see Table 14.1-65).

a. Augmented Low Power Test (see Table 14.1-65a).

3. Incore Movable Detector and Thermocouple Mapping at Power (see Table 14.1-66).

4. Power Coefficient and Power Defect Measurement (see Table 14.1-67).

5. Effluent Radiation Monito r Test (see Table 14.1-68).

6. Radiation Shielding Survey (see Table 14.1-69).

7. Thermal Power Measurements and Instrument Calibration (see Table 14.1-71).

8. Automatic Control Systems Checkout (see Table 14.1-72).

9. Feedwater System (see Table 14.1-88).

10. Plant Response to Step Load Changes (see Table 14.1-73).

11. Pseudo Rod Ejection Test (see Table 14.1-74).

12. Rod Drop Test (see Table 14.1-75).

13. Below-Bank Rod Test (see Table 14.1-76).

14.1-16 Reformatted Per Amendment 02-01

14. Plant Loss of Electrical Load (see Table 14.1-77).

15. Loss of Offsite Power (see Table 14.1-78).

16. Shutdown from Outside the Control Room (see Table 14.1-79).

Table 14.1-80 indicates power levels at which the above tests are performed.

14.1.4 CORE LOADING AN D INITIAL OPERATION Core loading begins when all prerequisite system tests and operations are satisfactorily completed and the NRC operating license received. Upon completion of core loading, the reactor upper internals and pressure vessel head are installed and additional mechanical and electrical tests are performed.

The reactor is then ready for its initial criticality. After the initial criticality, low power tests, and power ascension test will commence. The purpose of these tests is to establish the operational characteristics of the unit and core, to acquire data for the pro per calibration of setpoints, and to ensure that operation conforms to the license requirements.

14.1.4.1 Core Loading Before starting core loading, the precore load ing tests must be complete, the plant shall have a NRC operating license, and there must be an appr opriately licensed operating staff. The Plant M anager is responsible for the conduc t of core loading and all core loading personnel. Technical assistance will be provided by Westinghouse during the initial core loading operation. During this time, containment integrity and security must be maintained through the use of established pr ocedures. The overall process of initial core loading is, in general, directed from t he operating floor of the reactor building.

The as-loaded core configuration is specif ied as part of the core design studies conducted well in advance of plant startup and as such is not subject to change at startup. In the event that mechanical damage is sustained during core loading operations by a fuel assembly of a type for which no spare is available onsite, an alternate core scheme whose characteristics closely approximate t hose of the initially prescribed pattern will be determined.

The core is assembled in the reactor vesse l, submerged in wate r containing enough dissolved boric acid to maintain a calculated core effective multiplication constant of 0.95 or lower. The refueling cavity is dry during initial core loadi ng. Core moderator chemistry conditions (particularly boron concentration) are prescribed in the core loading procedure document and are verified periodically by chemical analysis of moderator samples taken prior to and during core loading operation.

14.1-17 Reformatted Per Amendment 02-01 Core loading instrumentation consists of two permanently in stalled source range (pulse type) nuclear channels and two temporary inco re source range channels plus a third temporary channel which can be used as a spare. The permanent channels, when responding, are monitored in the control room by licensed pl ant operators; the temporary channels are install ed in the containment structur e and are monitored by fuel loading personnel. At least one permanent channel is equipped with an audible count rate indicator heard in the reactor building and the control room. Both plant channels have the capability of displaying the neutron flux level on strip chart recorders. The temporary channels indicate on rate meters with a minimum of one channel recorded on a strip chart recorder. Mini mum count rates of 1/2 counts per second, attributable to core neutrons, are required on at least two (i.e., temporary and/or permanent source range detectors) available nuclear source ch annels at all times following installation of the initial nucleus of eight fuel assemblies.

A response check of nuclear instruments to a neutron source shall be performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> prior to loading of the core, or resumption of loading if delay is for more than 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

At least two neutron sources are introduced into the core loading program to ensure a neutron population of a minimu m of 1/2 counts/sec for adequat e monitoring of the core.

Fuel assemblies together with inserted components (control rod assemblies, burnable poison inserts, source spider, or thimble plugging devices) are placed in the reactor vessel one at a time according to a previously established and approved sequence which was developed to provide reliable core monitoring with minimum possibility of core mechanical damage. The core loading procedure documents include a detailed tabular check sheet which prescribes and verifies the successive movements of each fuel assembly and its specified inserts from its initial position in the fuel racks to its final position in the core. Multiple checks are made of component serial numbers and types at successive transfer points to guard against possible inadvertent exchanges or substitution of components and fuel assembly status boards are maintained throughout the core loading operations.

An initial nucleus of eight fuel assemblies, the first of which contains a neutron source, is the minimum source-fuel nucleus which permits subsequent meaningful inverse count rate monitoring. This initial nucleus is determined by calculation and previous experience to be mark edly subcritical (K eff 0.95) under the required conditions of loading. 02-01 Subsequent fuel additions are accompanied by detailed neutron count rate monitoring to determine that the count rate does not increase excessively and that the extrapolated inverse count rate ratio is not decreasing fo r unexplained reasons. The results of each loading step are evaluated by SCE&G and tec hnical advisors before the next prescribed step is started.

14.1-18 Reformatted Per Amendment 02-01 Criteria for safe loading require that loading operations st op immediately if:

1. An unanticipated increase in the neutron count rates by a factor of 2 occurs on all responding nuclear channels during any single loading step after the initial nucleus of 8 fuel assemblies is loaded (excluding ant icipated change due to detector and/or source movement).

2. The neutron count rate on any individual nuclear channel increases by a factor of five during any single loading step after the initial nucleus of 8 fuel assemblies is loaded (excluding anticipated changes due to detector and/or source movements).

3. A decrease in boron concentration gr eater than 20 ppm is determined from two successive samples of reactor coolant system water until the decrease is explained.

An alarm in the Reactor Building and Cont rol Room is coupled to the source range channels with a setpoint at five times the cu rrent count rate. This alarm automatically alerts the fuel loading personnel to an indication of high count rate and requires an immediate stop of all operations until the situation is eval uated by SCE&G and technical advisors.

Core loading procedures specify alignment of fluid systems to prevent inadvertent dilution of the reactor coolant, restrict the movement of fuel to preclude the possibility of mechanical damage, prescribe the conditions under which loading can proceed, identify chains of responsibility and authority, and provide for cont inuous and complete fuel and core component accountability.

14.1.4.2 Postloading Tests Upon completion of core loading, the reactor upper internals and pressure vessel head are installed and additional mec hanical and electrical tests ar e performed prior to initial criticality. The final pressure tests are conducted after filling and venting is completed.

Mechanical and electrical tests are performed on the control rod drive mechanisms.

These tests include a complete operational checkout of the mechanisms. Checks are made to ensure that the control rod asse mbly position indicator coil stacks are connected to rod drive mechanism coils.

Tests are performed on the reactor trip circui ts to test manual trip operation and actual control rod assembly drop time s are measured for each cont rol rod assembly. By use of dummy signals, the Reactor Control and Protection System is made to produce trip signals for the various plant abnormalities that require tripping.

14.1-19 Reformatted Per Amendment 02-01 At all times that the control rod drive mechanisms ar e being tested, the boron concentration in the coolant-moderator is la rge enough such that the shutdown margin requirements specified in the Technical Specifications are met. During individual RCCA or RCC bank motion, source range instrumentation is monitored for unexpected changes in core reactivity.

A complete functional electrical and mechanical check is made of the incore nuclear flux mapping system at operating temperature and pressure.

14.1.4.3 Initial Criticality Initial criticality is established by s equentially withdrawing the shutdown and control groups of control rod assemblies from the co re, leaving the last withdrawn control group inserted far enough in the core to provide effe ctive control when criticality is achieved, and then continuously diluting the heavily borated reactor coolant until the chain reaction is self-sustaining. Successive stages of control rod assembly group withdrawal and of boron concentration reduction are monitored by observing changes in neutron count rate as indicated by the regular source range nucl ear instrumentation as functions of group position during rod motion and, subsequently, of reactor coolant boron concentration and primary water addition to the Reactor Coolan t System during dilution. Throughout this period samples of the primar y coolant are obtained and analyz ed for boron concentration.

Primary safety reliance is based on inverse c ount rate ratio monitoring as an indication of the nearness and rate of approach to crit icality of the core during control rod assembly group withdrawal and during reacto r coolant boron dilution. The rate of approach is reduced as the reactor approaches extrapolated critica lity to ensure that effective control is maintained at all times.

Written procedures specify alignment of fl uid systems to allow controlled starting, stopping, and adjustment of the rate of the approach to crit icality. These procedures also identify chains of responsibility and authority during initial criticality.

14.1.4.4 Low Power Testing A prescribed program of reactor physics measurements is undertaken to verify that the basic static and kinetic characte ristics of the core are as ex pected and that the values of the kinetic coefficients assumed in the safeguards analysis are indeed conservative.

The measurements are made at low power and primarily at or near operating temperature and pressure. Meas urements, to include verification of calculated values of control rod assembly group reactivity wo rths, of moderator te mperature coefficient under various core conditions, of differential boron concentration reactivity worth, and of critical boron concentrations as functions of control rod assembly group configuration are made. In addition, measur ements of the relative powe r distributions are made.

Concurrent tests are conducted on the in strumentation includ ing the source and intermediate range nuclear channels.

14.1-20 Reformatted Per Amendment 02-01 In accordance with NUREG 0737, item I.G.1, a discussion of the special Low Power Testing to be performed at the Virgil C.

Summer Station has been submitted to the NRC under separate cover letters dated 10/31/80, 12/2/80, and 12/22/80.

Detailed procedures are prepared to specif y the sequence of tests and measurements to be conducted and the conditions under which each is to be performed to ensure both safety of operation and the relevancy and consistency of the results obtained. If significant deviations from design predictions exist, unacceptable behavior is revealed, or apparent anomalies develop, a review by qualified personnel will be performed. The Plant Manager may determine that a review by the Plant Safety Review Committee is appropriate prior to increasing power to the next testing plateau.

14.1.4.5 Power Ascension When the operating characterist ics of the reactor and plant are verified by low power testing, a program of power level escalation in successive stages brings the plant to its full rated power level. Prior to starting to the next higher power plateau, as defined by the startup test progr am, management approval s hall be obtained, also, the high flux trip shall be reset to a value no greater than 20 percent beyond the next power level. During the power escalation, a predete rmined test program is followed to verify that the reactor and plant are performing as expected. T he minimum test requirements for each successive stage of power escalation are specified.

Measurements are made to determine the relati ve power distribution in the core as functions of power level and control assembly group position.

Secondary system heat balances ensure that the several indications of power level are consistent and provide bases for calibration of the power range nuclear channels. The ability of the Reactor Control System to respond effectivel y to signals from primary and secondary instrumentation under a variet y of conditions encountered in normal operations is verified.

At prescribed power levels, the dynamic response characteristics of the Reactor Coolant and Steam Systems ar e evaluated. These responses are evaluated for step

load changes of 10 percent and a plant loss of electrical load from 100 percent power.

Adequacy of radiation shielding is verified by gamma and neutron radiation surveys inside the containment and throughout the station site. Periodic sampling is performed to verify coolant c hemistry and activity.

The sequence of tests, meas urements, and intervening operations is described in the power escalation procedure with specific details relating to the conduct of the several tests and measurements.

14.1-21 Reformatted Per Amendment 02-01 14.1.5 ADMINISTRATIVE PROCEDURES (SYSTEM OPERATION)

Whenever the test allows, the plant operati ng procedures for Virgil C. Summer Nuclear Station will be incorporated into the test procedures. This will help in evaluating plant operating procedures. If during a test it is found that the operat ing procedure is not adequate, the test procedure will be c hanged in the manner described in Section 14.1.2.2; and the operating procedure shall be changed in accordance with Chapter 13. Any operating procedure, or part of one, which is referenced in a test procedure must be approved.

14.1-22 Reformatted Per Amendment 02-01 14.1-23AMENDMENT 97-01AUGUST 1997TABLE 14.1-1FIRE PROTECTION SYSTEM1.0ObjectiveDemonstrate that the Fire Protection System is capable of providing adequate fireprotection under all conditions, including loss of power to the motor driven pump.2.0PrerequisitesFire Protection System installation and component checks completed.3.0Test Methods3.1Align the subsystems for operation and establish normal flow paths forwater systems.3.2Check operation of water systems and CO 2 system.3.3Verify pump heads and flow rates under both normal and emergency powerconditions of the water systems.3.4Verify operation and response of detector systems.4.0Acceptance Criteria4.1Alarms, interlocks, and detection devices function as per designrequirements.4.2System is capable of providing protection in accordance with applicable fireprotection codes.4.3Where appropriate, flows, pressures, and spray patterns are as per designrequirements.

14.1-24AMENDMENT 97-01AUGUST 1997TABLE 14.1-2CONTAINMENT ISOLATION VALVES LEAKAGE RATE TEST1.0ObjectiveDemonstrate that the leakage from containment isolation valves and gasketedblind flanges required to be tested by 10 CFR 50, Appendix J is within allowable limits.2.0Prerequisites2.1Portions of the systems containing isolation valves and/or gasketed blindflanges have been successfully pressure tested.2.2Motor operated and air operated containment isolation valves have beenfunctionally tested.2.3Portions of the systems containing isolation valves and/or gasketed blindflanges have been properly drained and vented.2.4Containment isolation valves have been closed using their normal mode ofoperation.3.0Test Methods3.1A pressure of not less than P a is applied against the isolation valve orgasketed blind flange.3.2Measurement of valve leakage is made using a rotameter array or otherequivalent instrumentation.4.0Acceptance Criteria4.1Containment isolation valve and gasketed blind flange leakage rates arewithin allowable design limits.

14.1-25AMENDMENT 97-01AUGUST 1997TABLE 14.1-3CONTAINMENT PENETRATIONS LEAKAGE RATE TEST1.0ObjectiveDemonstrate that the leakage from containment penetrations required to be testedby 10 CFR 50, Appendix J is within allowable limits.2.0Prerequisites2.1Penetrations have been properly installed.3.0Test Methods3.1A pressure of not less than P a is applied to the penetration.3.2Measurement of penetration leakage is made using a rotameter array orother equivalent instrumentation.4.0Acceptance Criteria4.1Containment penetration leakage rates are within allowable design limits.

14.1-26AMENDMENT 97-01AUGUST 1997TABLE 14.1-4CONTAINMENT AIR LOCKS LEAKAGE RATE TEST1.0ObjectiveDemonstrate that leakage from the containment per sonnel air locks andequipment hatch is within allowable limits.2.0Prerequisites2.1Containment personnel air locks and equipment hatch have been properlyinstalled.3.0Test Methods3.1A pressure of not less than P a is applied to the seals of the containmentpersonnel air lock and equipment hatch.3.2Measurement of seal leakage is made using a rotameter array or otherequivalent instrumentation.4.0Acceptance Criteria4.1Containment personnel air lock and equipment hatch leakage is withinallowable design limits.

14.1-27AMENDMENT 97-01AUGUST 1997TABLE 14.1-5REACTOR BUILDING STRUCTURAL ACCEPTANCE TEST1.0ObjectiveVerify the structural integrity of the Reactor Building.2.0Prerequisites2.1Reactor Building penetrations installed and penetration leak tests (Type Btest) completed.2.2Isolation valve leak test (Type C test) completed2.3Reactor Building Ventilation Systems operable to extent required to controlReactor Building internal temperature.3.0Test Methods3.1Prior to initial fuel loading, the Reactor Building will be subjected to apressure equivalent to 115 percent of the Reactor Building designed pressure. This test demonstrates that the Reactor Building is capable of resisting the postulated accident pressure. In addition, by measuring the structural response and comparing the results with analytical predictions, the test verifies that the structure does behave as anticipated.3.2Instrumentation, measuring systems, pressurization procedure,deformation, strain and temperature measurements, crack pattern mapping, and data acquisition schedules for the preoperational structural acceptance test will be in accordance with the approved test procedure4.0Acceptance CriteriaThe Reactor Building meets structural integrity design requirements.

14.1-28AMENDMENT 97-01AUGUST 1997TABLE 14.1-6CONTAINMENT INTEGRATED LEAKAGE RATE TEST1.0ObjectiveDemonstrate that the containment leakage rate following a postulated loss ofcoolant accident (LOCA) is within allowable limits.2.0Prerequisites2.1Reactor Building structural acceptance test completed 2.2Affected systems lined up in their post accident mode.3.0Test Methods3.1The containment is pressurized at a rate of 5 psi per hour or less to P a asdefined in the approved test procedure.3.2Containment temperature is held essentially constant for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> while theinternal temperature, pressure, and dewpoint are closely monitored.3.3Using the perfect gas law, the change in the containment air mass is computed.3.4The leakage rate in percent/day is computed from the changes incontainment air mass.3.5A known leak is superimposed on the containment to verify instrumentationaccuracy.4.0Acceptance Criteria4.1Integrated leakage rate is within allowable design limits.4.2Instrumentation accuracy is verified to be within allowable design limits.

14.1-29AMENDMENT 97-01AUGUST 1997TABLE 14.1-7UNIT AUXILIARY, EMERGENCY AUXILIARY ANDENGINEERED SAFETY FEATURES TRANSFORMERS1.0Objective1.1Demonstrate the capability of the unit auxiliary, emergency auxiliary andengineered safety features transformers to supply electrical power to the 7200 volt buses.1.2Verify operation of protection devices and functional operation of controls and interlocks.2.0Prerequisites2.1230 kV substation operational and energized.

2.2Meters, relays, and protective devices calibrated and tested2.3125 volt d-c available.

2.4Erection work on transformers and switchgear completed 2.5Transformer oil and fan systems tested and in service.

2.6Isolated phase bus tested and ready for service.

2.7Breaker controls and transfer scheme verified.

2.8PT and CT circuits checked for polarity and continuity.

2.97200 volt breakers racked out.

2.10115 kV supply line energized.3.0Test Methods3.1Simulate signals to temperature controls and verify operation of transformeroil pumps and fans.3.2Simulate signals to verify annunciators for transformer protective devices.

3.3Energize transformers and verify phase rotation.4.0Acceptance CriteriaTransformers provide reliable source of electrical power to 7200 volt buses inaccordance with design requirements and Section 8.3.

14.1-30AMENDMENT 97-01AUGUST 1997TABLE 14.1-87200 VOLT ELECTRICAL SYSTEM1.0Objective1.1Verify that 7200 volt buses can be energized from their respective normaland alternate sources.1.2Verify that electrical and mechanical interlocks function properly2.0Prerequisites2.1Breakers in the 7200 volt buses are racked out and tagged.

2.2Meters, relays, and protective devices calibrated and tested2.3125 volt d-c available.

2.4Phase rotation checked on 7200 volt buses.

2.5Unit auxiliary, emergency auxiliary, and ESF transformers energized.3.0Test Methods3.1Rack in and close 7200 volt breakers to energize associated 7200 volt buses3.2Record voltage and verify phase relationship.

3.3Shift buses to alternate power sources as applicable and verify phaserelationship.3.4Verify 7200 Volt BOP Buses will transfer from normal source to alternatesource if a fault occurs on normal supply.3.5Verify 7200 Volt ESF Buses will transfer from normal source to emergencydiesel generator if voltage is lost on normal supply4.0Acceptance Criteria4.1The 7200 volt buses are capable of being energized from their normal andalternate sources and that proper phase relationship is exhibited.4.2System interlocks and alarms function properly.4.3BOP buses transfer from normal source to alternate source if a fault occurson normal supply.

14.1-31AMENDMENT 97-01AUGUST 1997TABLE 14.1-8 (Continued)7200 VOLT ELECTRICAL SYSTEM4.4ESF buses will transfer from normal source to emergency diesel generatorif voltage is lost on normal supply 14.1-32AMENDMENT 97-01AUGUST 1997TABLE 14.1-9480 VOLT BUSES1.0Objective1.1Verify that 480 volt buses can be energized from their normal and alternate sources.1.2Verify that electrical and mechanical interlocks function properly2.0Prerequisites2.1Breakers in the 480 volt buses are racked out and tagged.

2.2Meters, relays, and protective devices calibrated and tested2.3125 volt d-c and 7.2 kV buses energized.

2.4Phase rotation checked on 480 volt buses.3.0Test Methods3.1Close 7.2 kV breakers to energize load center transformers and buses.3.2Measure voltage and verify phase relationship.

3.3Shift buses to alternate power sources as applicable and verify phaserelationship.3.4Simulate loss of power.4.0Acceptance Criteria4.1480 volt buses are capable of being energized from their normal andalternate sources, and proper phase relationship is exhibited.4.2The 480 volt buses respond correctly to a loss of power.

4.3Interlocks and alarms function properly.

14.1-33AMENDMENT 97-01AUGUST 1997TABLE 14.1-10480 VOLT MOTOR CONTROL CENTERS1.0Objective1.1Verify that 480 volt motor control centers can be energized from theirnormal sources.1.2Verify that electrical and mechanical interlocks function properly2.0Prerequisites2.1All breakers in 480 volt motor control centers are racked out and tagged.

2.2Meters, relays, and protective devices calibrated and tested2.3125 volt d-c available.

2.4480 volt buses energized.

2.5Phase rotation checked on motor control centers.3.0Test Methods3.1Rack in and close motor control center supply breakers.3.2Measure voltage and verify phase relationship.

3.3Simulate loss of power.4.0Acceptance Criteria4.1Motor control centers are capable of being energized from their normalsources, and proper phase relationship is exhibited.4.2The 480 volt motor control centers respond correctly to a loss of power.

4.3System interlocks and alarms function properly.

14.1-34AMENDMENT 97-01AUGUST 1997TABLE 14.1-11120 VOLT A-C1.0ObjectiveDemonstrate the capabilities of the 120 volt vital instrument power system and the120 volt a-c regulated instrument power system to supply power to instrumentation and control loads under normal and emergency conditions at full load.2.0Prerequisites2.1120 volt instrument power systems installation and component checkscompleted.2.2480 volt MCC available.2.3125 volt d-c system operable3.0Test Methods3.1Energize 120 volt instrument power buses from their normal power sourcesand load to full load using a load bank.3.2Demonstrate ability to transfer each vital instrument bus manually to abackup instrument bus and back to its static inverter.3.3Trip the normal power supplies to the static inverters. Verify automatictransfer to alternate dc source. Verify transfer back to normal supply when re-energized.3.4Demonstrate ability to transfer each instrument panel manually to itsalternate source.4.0Acceptance Criteria4.1Vital buses or panels can be manually transferred to alternate sources.4.2Vital inverters will transfer to the alternate 125V DC input source upon lossof normal 480V AC input source. Normal 120V AC is maintained during the transfer.4.3System interlocks and alarms function properly.4.4Vital AC inverters will supply their design load at 60 amps.

14.1-AMENDMENT 97-01AUGUST 1997 35TABLE 14.1-12D-C SYSTEM1.0ObjectiveDemonstrate the capability of the d-c system to provide a source of reliable,uninterruptable d-c power for normal and emergency instrumentation, control, and power loads.2.0Prerequisites2.1480 volt a-c power available.

2.2Battery Room Ventilation System operable.

2.3Batteries, battery chargers, and d-c distribution system, including protectivedevices, installation, and component checks completed.2.4D-C breakers are open.3.0Test Methods3.1Energize the battery chargers and verify this normal feed to bus.3.2Verify alarms and interlocks.

3.3Discharge the batteries at a controlled rate and determine ampere-hourcapacity.3.4Adjust chargers to supply d-c load and charge batteries simultaneously.3.5De-energize battery chargers while the applicable buses are carrying theirnormal station load to verify that battery will maintain load.3.6Verify that 125 volt bus can be fed from the backup charger upon loss ofnormal charger (manual transfer).3.7Verify ground detection by connecting variable load resistor to ground.4.0Acceptance Criteria4.1System interlocks and alarms function properly.4.2Batteries are capable of supplying plant dc power upon de-energization oftheir chargers.

14.1-AMENDMENT 97-01AUGUST 1997 36TABLE 14.1-12 (Continued)D-C SYSTEM4.3Battery chargers are capable of maintaining normal bus loads concurrentwith charging the batteries.4.4Bus can be supplied by backup charger.

14.1-37AMENDMENT 97-01AUGUST 1997TABLE 14.1-13PLANT PAGING AND COMMUNICATION SYSTEM1.0Objective1.1Demonstrate the adequacy of the plant paging systems intracommunicationbetween all local stations in each separate system, and interconnection of the plant PABX system to the commercial telephone service and SCE&G microwave system.1.2Demonstrate that the radiation emergency (site evacuation) alarm can beheard from any location in the plant under all required conditions.1.3Demonstrate operability of Reactor Building evacuation alarm and firealarm.2.0Prerequisites2.1Communications systems installation and component checks completed.

2.2Sound levels established for locations where noise levels might interferewith communication.3.0Test Methods3.1Test the portable stations, page/party line phone station, and redundantcommunication phone stations for proper operation.3.2Test interconnection of the plant PABX system to the commercialtelephone service, and SCE&G microwave system.3.3Verify alarms.

3.4Shift PABX equipment to alternate power sources and verify operation.4.0Acceptance Criteria4.1Communication systems provide for paging, normal plant communications,shutdown communications, interconnection to SCE&G microwave system and commercial telephone service and alarm signaling in accordance with design requirements and Section 9.5.2.4.2Evacuation alarm can be heard from any location in the plant.

4.3Paging systems and alarms will not produce sound pressure levels above130 db level where personnel will conduct normal plant communications.

14.1-38AMENDMENT 97-01AUGUST 1997TABLE 14.1-14DIESEL FUEL OIL TRANSFER AND STORAGE SYSTEM1.0ObjectiveDemonstrate that the Diesel Fuel Oil Transfer and Storage System supplies fueloil to the diesel fuel oil day tanks and verify modes of operation.2.0Prerequisites2.1Fire protection is available.

2.2Diesel Fuel Oil System installation and component checks completed.3.0Test Methods3.1Align system for operation and establish normal transfer flow paths.3.2Verify capability to transfer fuel oil to each day tank.

3.3Verify alarms and automatic pump operations.4.0Acceptance CriteriaFuel transfer capability of the system meets the design requirements.

14.1-39AMENDMENT 97-01AUGUST 1997TABLE 14.1-15EMERGENCY DIESEL GENERATORS1.0Objective1.1Demonstrate manual start and synchronization of the diesel generators.1.2Demonstrate load-carrying capacity of diesel generators.

1.3Demonstrate fuel consumption of diesels.

1.4Demonstrate proper operation of diesel auxiliaries during the load carryingcapacity test.2.0Prerequisites2.1Station batteries charged and d-c control power available.

2.2Relays calibrated and all normal bus protective devices checked and inservice.2.3Diesel engine auxiliary systems installation and component checkscompleted as specified.2.4Diesel room ventilation and fire protection available.

2.5Diesel Fuel Oil System operable.3.0Test Methods3.1Demonstrate manual start and synchronization of each diesel generator,including synchronizing the diesel generator unit to offsite power while the unit is connected to the emergency load, isolating the diesel generator unit, and restoring it to standby status.3.2Verify timing of diesel generators starting sequence and available for load.

3.3Verify capability to control diesel generators in local and remote operation.3.4The diesel generators will be operated at continuous rated load for 22hours. The load will then be increased to the two hour rating and held atthis condition for a period of two hours. During this 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period, the indicated readings from supervisory instrumentation will be recorded toestablish base line operating conditions.

14.1-40AMENDMENT 97-01AUGUST 1997TABLE 14.1-15 (Continued)EMERGENCY DIESEL GENERATORS3.5Conduct load-carrying duration test.3.6Verify load rejection and overspeed trip.

3.7Demonstrate diesel generator reliability by performing 69/N starts (where Nis the number of diesel generator units).4.0Acceptance Criteria4.1Diesel generators function in maintaining the 7200 volt ESF buses asdesigned.4.2Diesel generators do not overspeed when load is removed.

4.3Each redundant onsite power source and its load group can functionwithout any dependence upon any other redundant load group or portion thereof.4.4Diesel auxiliaries perform their design requirements.

4.5Diesel generators demonstrate required reliability.

14.1-41AMENDMENT 97-01AUGUST 1997TABLE 14.1-16RESPONSE TO LOSS OF INSTRUMENT AIR1.0ObjectiveDemonstrate that pneumatically operated "Active" valves fail to their safe positionon a loss of instrument air.2.0Prerequisites2.1Instrument Air System installation and component checks completed.

2.2Associated systems completed to the extent necessary to allow conduct of this test.2.3System has been blown down and cleanliness requirements met.3.0Test MethodsNote: Specific valves and systems may be tested individually.3.1Align system for normal operation.

3.2Reduced instrument air pressure to zero psig.

3.3Observe the response of pneumatically operated "Active" valves during lossof air pressure and record the position to which each valve fails.3.4Air operated valves equipped with safety related volume tanks will beisolated from the instrument air header and stroked to verify their operability.4.0Acceptance Criteria4.1Pneumatically operated "Active" valves fail to their "safe" position on loss ofinstrument air.4.2Components with "safety related" air volume tanks operate in accordancewith design.

14.1-42AMENDMENT 97-01AUGUST 1997TABLE 14.1-17AUXILIARY BUILDING VENTILATION SYSTEM(RADIOACTIVE PORTION)1.0Objective1.1Demonstrate the capabilities of the Auxiliary Building Ventilation System toprovide for control and disposal of airborne radioactivity.1.2Confirm the proper operation of system interlocks and controls.2.0Prerequisites2.1Auxiliary Building Ventilation System installation and component checkscompleted.2.2Associated systems completed to the extent necessary to allow conduct of this test.3.0Test Methods3.1Align system for normal operation.3.2Verify building internal negative pressure differential with respect toatmosphere.3.3Test positioning on pneumatically operated dampers.

3.4Simulate isolation actuation signal and observe system response.

3.5Verify fan and damper, interlocks and permissives.4.0Acceptance Criteria4.1System provides for control and disposal of airborne radioactivity inaccordance with design requirements and Section 9.4.4.2Fans and dampers, function as per design.

4.3Auxiliary Building internal pressure is less than atmospheric.

14.1-43AMENDMENT 97-01AUGUST 1997TABLE 14.1-18FUEL HANDLING BUILDING VENTILATION SYSTEM(RADIOACTIVE PORTION) 1.0 2.0Prerequisites2.1Fuel Handling Building Ventilation System installation and componentchecks completed.2.2Associated systems completed to the extent necessary to allow conduct of this test.3.0Test Methods3.1Align system for normal operation.3.2Verify fan and damper, interlocks and permissives.3.3Test positioning on pneumatically operated dampers.

3.4Simulate isolation actuation signal and observe system response.

3.5Verify building internal negative pressure differential with respect toatmosphere.4.0Acceptance Criteria4.1System provides for control and disposal of airborne radioactivity inaccordance with design requirements and Section 9.4.4.2Fans and dampers, function as per design.

4.3Building internal pressure with respect to atmospheric pressure is perdesign.

14.1-44AMENDMENT 97-01AUGUST 1997TABLE 14.1-19PROCESS AND AREA RADIATION MONITORING SYSTEM1.0ObjectiveDemonstrate the capability of the Proce ss and Area Radiation Monitoring Systemsto monitor effectively the levels of radiation in the plant or its effluents and to initiate isolation and alarms as required.2.0Prerequisites2.1Process and Area Radiation Monitoring Systems installation andcomponent checks completed.2.2Associated systems completed to the extent necessary to allow theconduct of this test.3.0Test Methods3.1Align system for normal operation. Position valves in associated systemsas necessary to allow response to isolation signals.3.2Verify proper functioning of system detectors by utilizing test sources andother procedures as appropriate.3.3Verify proper system response to simulated alarm conditions by monitoringcontroller outputs, alarm indications, and the operation of isolation valves where possible.4.0Acceptance CriteriaSystem effectively monitors and responds to levels of radiation in the plant areasand effluents in accordance with design requirements and Sections 11.4, 12.1.4, and 12.2.4.

14.1-45AMENDMENT 97-01AUGUST 1997TABLE 14.1-20COMPONENT COOLING WATER SYSTEM1.0ObjectiveDemonstrate the capability of the Component Cooling System to supply designflow to system components.2.0Prerequisites2.1Component Cooling (CC) System installation and component checkscompleted.2.2Adequate supply of demineralized water available.

2.3Associated systems completed to the extent necessary to allow conduct of this test.3.0Test Methods3.1Verify the ability of the CC pumps to deliver design low and high speedflows to system components.3.2Verify the ability of the CC system to perform its design functions for SI andrecirculation modes of operation.3.3Verify CC system controls operate as designed.4.0Acceptance Criteria4.1System flows, pressures, and automatic functions are in accordance withdesign requirements identified on the applicable flow diagrams.4.2System interlocks, instrumentation, and alarms function properly.

14.1-46AMENDMENT 97-01AUGUST 1997TABLE 14.1-21SERVICE WATER SYSTEM1.0ObjectiveDemonstrate the capability of the Service Water System to provide adequatecooling water.2.0Prerequisites2.1Service Water System installation and component checks completed.2.2Associated systems completed to the extent necessary to allow conduct of this test.3.0Test Methods3.1Align system for normal operation.

3.2Verify system flow rates and system control.4.0Acceptance Criteria4.1System flows, pressures, and automatic functions are in accordance withdesign requirements of Section 9.2.1.

14.1-47AMENDMENT 97-01AUGUST 1997TABLE 14.1-22BORIC ACID BATCHING AND TRANSFER SYSTEM1.0ObjectiveVerify proper functioning of equipment and instrumentation utilized in batching,storage, transfer, and recirculation of boric acid solutions.2.0Prerequisites2.1Boric Acid System installation and component checks completed.2.2Adequate supply of grade A water available.

2.3Steam supply available to batching tank jacket heater.

2.4Associated systems completed to the extent necessary to allow conduct of this test.3.0Test Methods3.1Align system for normal operation.

3.2Verify boric acid tank and batching tank level setpoints, controller functions,and steam delivery to batching tank heaters.3.3Verify capability of boric acid transfer pumps to deliver water from thebatching tank to the boric acid tanks and to recirculate each boric acid tank.3.4Verify capability of supplying the charging pump suction header.4.0Acceptance Criteria4.1System provides for batching, storage, transfer, and recirculation flow pathsin accordance with Section 9.3.4.4.2Interlocks, automatic functions, alarms, flows, and pressures are withindesign limits.

14.1-48AMENDMENT 97-01AUGUST 1997TABLE 14.1-23HEAT TRACING FOR BORON INJECTION TANK AND ASSOCIATED PIPING1.0ObjectiveDemonstrate the ability of the Heat Tracing System to maintain propertemperature control in the various piping systems involved with the boron injection

tank.2.0Prerequisites2.1Heat Tracing System installation and component checks completed.

2.2Associated systems completed to the extent necessary to allow theconduct of this test.3.0Test Methods3.1Energize Heat Tracing System.3.2Monitor temperatures maintained by each heat tracing circuit with thesystem in a static condition.3.3Place boron injection recirculation pump in operation and establish transferflow path.3.4Monitor temperatures maintained by each heat tracing circuit.4.0Acceptance CriteriaEach heat tracing circuit maintains temperature within allowable design limits.

14.1-49AMENDMENT 97-01AUGUST 1997TABLE 14.1-24REACTOR BUILDING COOLING SYSTEM1.0Objective1.1Demonstrate the operation of the Reactor Building cooling equipment fornormal plant operating conditions.1.2Demonstrate the capacity of the Reactor Building cooling units to maintainarea temperatures within design limits during normal operating conditions.2.0Prerequisites2.1Reactor Building Cooling System installation and component checkscompleted.2.2Service Water and Industrial Cooling Water Systems operable2.3Associated systems completed to the extent necessary to allow conduct of this test.3.0Test Methods3.1Align system for normal operation.3.2Verify fan, damper, and cooling unit controls, interlocks, and permissives.3.3With the plant at, or near, normal operating conditions, survey variousareas to verify that temperatures do not exceed design limits.4.0Acceptance Criteria4.1System interlocks, and controls function properly.4.2Temperature survey does not show any hot spots or exceed design limits.

14.1-50AMENDMENT 97-01AUGUST 1997TABLE 14.1-25CHEMICAL AND VOLUME CONTROL SYSTEM1.0ObjectiveDemonstrate that the Chemical and Volume Control System (CVCS) performs asrequired during plant operation.2.0Prerequisites2.1CVCS installation and component checks completed.

2.2Reactor Coolant System at the condition specified in the approved testprocedure.2.3Adequate supply of Grade A water available in refueling water storage tankor reactor makeup tank.2.4Associated systems completed to the extent necessary to allow conduct of this test.3.0Test Methods3.1Align CVCS for normal operation and establish normal flow paths.3.2Verify capacities of letdown orifices and pressure drop of reactor coolantfilter.3.3Check operation of the letdown line temperature and pressure controllerswith the demineralizers bypassed.3.4Verify operation of excess letdown and seal water subsystems.

3.5Verify flow rates and pressure drops of demineralizers.

3.6Verify charging pumps' flow rates and the seal water flow rate for eachreactor coolant pump.3.7Verify volume control tank level controller operation. Check reactormakeup control system response to inventory changes of volume control tank.4.0Acceptance CriteriaSystem performance, interlocks, and automatic functions are in accordance withSection 9.3.4.

14.1-51AMENDMENT 97-01AUGUST 1997TABLE 14.1-26PRESSURIZER RELIEF TANK1.0ObjectiveVerify that the pressurizer relief tank provides for adequate control of thedischarge from the primary reliefs and safety valves.2.0Prerequisites2.1Hydrostatic test of pressurizer relief tank completed 2.2Pressurizer relief tank installation checks completed.2.3Radioactive waste disposal system completed to the extent necessary toallow conduct of this test.2.4Adequate supply of grade A water available.

2.5Adequate supply of nitrogen available.3.0Test Methods3.1Verify alarms, interlock operations, and spray flow control.3.2Demonstrate ability to maintain nitrogen blanket in pressurizer relief tank.

3.3Verify transfer flow paths from pressurizer relief tank4.0Acceptance CriteriaPressurizer relief tank provides disposal of primary plant coolant discharge inaccordance with design requirements and Section 5.5.11.

14.1-52AMENDMENT 97-01AUGUST 1997TABLE 14.1-27REACTOR COOLANT SYSTEM HEATUP FOR HOT FUNCTIONAL TESTING1.0ObjectivePerform functional checks on the Reactor Coolant System and associatedsystems components and instrumentation required to bring the plant from a cold shutdown condition to normal operating temperature and pressure.2.0Prerequisites2.1Reactor Coolant System and supporting systems valve lineups for normaloperation completed and normal flow paths established.2.2Reactor Coolant System cold hydrostatic test completed.2.3Preoperational and acceptance tests completed as necessary2.4Instrumentation and control checkouts and calibrations completed asnecessary.2.5Secondary system ready to receive steam and return feedwater to thesteam generators.2.6Diesel generators fully operable. Batteries and battery chargers are inservice.2.7Systems completed to the extent necessary to allow conduct of this test.3.0Test Methods3.1Establish specified charging and letdown flow rate and seal water flow tothe reactor coolant pumps.3.2Operate reactor coolant pumps.

3.3Energize pressurizer heaters and conduct solid system pressure controldemonstration.3.4Perform chemistry adjustment demonstrations.

3.5Form pressurizer steam bubble.

3.6At approximately 100F intervals, stabilize system parameters and recordrequired data, measurements, and observations for incore thermocouple and RTD cross calibration, reactor coolant pump vibration measurements, and Reactor Coolant System thermal expansion measurements.

14.1-53AMENDMENT 97-01AUGUST 1997TABLE 14.1-27 (Continued)REACTOR COOLANT SYSTEM HEATUP FOR HOT FUNCTIONAL TESTING3.7Verify ability to maintain steam generator levels by operation of theatmospheric steam dump and the Emergency Feedwater System.3.8Check operability of pressurizer power operated relief valves, spray valves,and steam generator atmospheric steam dump valves by either placing in service or performing functional check.3.9Main steam power operated relief valves are functionally checked.4.0Acceptance CriteriaSystems, components, instrumentation, and controls function within allowabledesign limits.

14.1-54AMENDMENT 97-01AUGUST 1997TABLE 14.1-28SYSTEMS THERMAL EXPANSION1.0Objective1.1Verify that the applicable systems piping can expand without obstructionduring initial heatup to normal operating conditions.1.2Confirm that systems piping and components return to their approximatebaseline cold position after cooldown to ambient conditions.2.0Prerequisites2.1To commence with the Reactor Coolant System heatup for hot functionaltesting.2.2Hanger lock pins removed and expansion clearances set to the proper coldvalues.2.3Reference points for measurements established.2.4Piping supports are verified for proper installation prior to heatup forthermal expansion measurements.3.0Test Methods3.1Record cold baseline data.3.2Obtain a set of measurements during Reactor Coolant System heatup atapproximately 250, 350, 450, and 550F.3.3Upon completion of Reactor Coolant System cooldown, obtain a set ofmeasurements.4.0Acceptance Criteria4.1Piping movements do not cause piping rubs or interference with otherequipment.4.2Piping movements do not cause undue stresses as determined byinspection.4.3Piping and components return to approximate baseline position oncooldown.

14.1-55AMENDMENT 97-01AUGUST 1997TABLE 14.1-29HOT FUNCTIONAL TESTING1.0ObjectivePerform functional checks on the Reactor Coolant System and associatedsystems components and instrumentation required during normal hot plant operation.2.0Prerequisites2.1Reactor Coolant System (RCS) heatup completed, RCS conditions of 525-557F and 2235 psig being maintained.2.2Systems completed to the extent necessary to allow conduct of this test.3.0Test Methods3.1Check the response, stability, and general control characteristics of thepressure control system.3.2Perform other tests which required the Reactor Coolant System to be atnormal operating no-load temperature and pressure as outlined in the hot functional testing procedures.4.0Acceptance CriteriaSystems, components, instrumentation, and controls function within allowabledesign limits.

14.1-56AMENDMENT 97-01AUGUST 1997TABLE 14.1-30EMERGENCY FEEDWATER SYSTEM1.0ObjectiveDemonstrate that the Emergency Feedwater System is capable of providingadequate quantities of feedwater for the removal of decay heat.2.0Prerequisites2.1Emergency Feedwater System installation and component checkscompleted.2.2Condensate storage tank adequately filled to supply the required water forthe duration of the test.3.0Test Methods3.1Verify flow rates of the pumps and system control.3.2Verify that the turbine driven emergency feedwater pump cold startssuccessfully five successive times.4.0Acceptance Criteria4.1Emergency flow capability of the system meets the design requirements.4.2System interlocks and alarms function within allowable design limits.4.3The turbine driven emergency feedwater pump cold starts successfully fivesuccessive times.

14.1-57AMENDMENT 97-01AUGUST 1997TABLE 14.1-31ENGINEERED SAFETY FEATURES CIRCUITRY1.0Objective1.1Verify proper operation of motor and air operated valves to the differentESF signals and various interlocks.1.2Verify proper operation of breakers to ESF signals.2.0Prerequisites2.1Associated systems have been checked 2.2Necessary breakers are racked out and other conditions are as outlined in test.3.0Test MethodsSimulate various ESF signals and note component responses to signals.4.0Acceptance CriteriaValves and breaker respond properly to actuation signals.

14.1-58AMENDMENT 97-01AUGUST 1997TABLE 14.1-32NUCLEAR SAMPLING SYSTEM1.0ObjectiveVerify that a quantity of representative fluid can be obtained from each samplingpoint.2.0Prerequisites2.1RCS at operating pressure and temperature.2.2Installation checks complete.3.0Test Methods3.1From each nuclear sampling point take a sample.3.2Check operation of remotely and manually operated valves.4.0Acceptance CriteriaFlows are adequate for gathering samples. Remote valve actuation signals getproper response.

14.1-59Reformatted PerAmendment 02-01TABLE 14.1-33REACTOR COOLANT SYSTEM COOLDOWN FROM HOT FUNCTIONAL TESTING1.0ObjectivePerform functional checks on the Reactor Coolant System and associated systemscomponents and instrumentation required to bring the plant to the cooled-down,depressurized condition.2.0Prerequisites2.1Reactor Coolant System (RCS) hot functional testing completed, RCSconditions of 525-557F and 2235 psig being maintained.2.2Reactor makeup water storage tank contains sufficient quantity of grade Awater to accommodate the contraction of the primary coolant during cooldown.2.3Specified systems completed to the extent necessary to allow conduct of this test.3.0Test Methods3.1Demonstrate the Reactor Coolant System degassing procedure3.2Secure two reactor coolant pumps and commence plant cooldown bydecreasing the set pressure of the steam dump valves.3.3Record data as required for incore thermocouple and RTD cross calibration.3.4When reactor coolant temperature and pressure are below 350F and 425psig, place the Residual Heat Removal System in operation.3.5Collapse the steam bubble.3.6Continue pressurizer and Reactor Coolant System cooldown to 140F andreduce pressure to 50 psig.3.7Establish conditions for Reactor Coolant System draining.4.0Acceptance CriteriaSystems, components, instrumentation, and controls function within allowable design limits.

14.1-60Reformatted PerAmendment 02-01TABLE 14.1-34RESIDUAL HEAT REMOVAL SYSTEM FLOW TESTS1.0Objective1.1To record actual RHR pump data (flow versus head) and compare this to themanufacturer's pump curve.1.2To demonstrate proper flow rates in the cold leg injection, cold legrecirculation, and hot leg recirculation modes.1.3To demonstrate that motor operated valves in the LHSI system strokeproperly against maximum differential pressure.1.4Verify proper operation of flow control valves (603A & B, 605A & B).1.5To verify the friction loss in the sump suction lines to the RHR pumps andverify vortex control.1.6To demonstrate the automatic opening of the RHR sump isolation valves(8811A & B, 8812A & B) on receipt of a RWST lo-lo level signal coincidentwith a safety injection signal.1.7To demonstrate proper RHR system flow rate in the cooldown lineup (RHRpumps taking suction from the RCS loops and discharging back into the RCS cold legs with the reactor vessel head removed).1.8To demonstrate that interlocks associated with the following valves operateproperly; 8701A, 8701B, 8702A, 8702B, 8706A, and 8706B.2.0Prerequisites2.1Reactor vessel head is not installed and the internals are not in the vessel.2.2Sufficient water available in the RWST to conduct the test.3.0Test Methods3.1Each RHR pump will be run at various flow rates to record pressure and flowdata (motor currents and vibration data will be recorded at full flow).3.2Each RHR pump will be run in the cold leg injection, cold leg recirculation(with suction from the RWST instead of the containment sump), and hot leg recirculation (with suction from the RWST instead of the containment sump) modes to record flow rates.

14.1-61Reformatted PerAmendment 02-01TABLE 14.1-34 (Continued)RESIDUAL HEAT REMOVAL SYSTEM FLOW TESTS3.3Each motor operated valve in the low head safety injection system will bestroked against maximum differential pressure.3.4The RHR pumps will be run at various flow rates using flow control valves603A & B, and 605A & B.3.5Each RHR recirculation sump will be flooded with water and each RHR pumpwill be operated taking suction from its respective sump to verify frictionlosses in each sump suction line and to demonstrate vortex control.3.6Each of the sump isolation valves (8811A & B and 8812A & B) will receive alo-lo RWST level coincident with a safety injection signal to verify that each of the four valves open.3.7Each of the RHR Pumps will take a suction from the RCS (with the level inthe vessel at least one foot above the nozzles) and discharge back into the RCS cold legs.3.8Demonstrate that the interlocks associated with valves 8701A, 8701B, 8702A,8702B, 8706A, and 8706B operate properly.4.0Acceptance Criteria4.1Pump performance characteristics and required flow rates in the injection andrecirculation modes will be in accordance with design requirements.4.2Each motor operated valve will be observed locally for proper operation andtimed to verify the valve strokes in less than the maximum allowable time.4.3Flow control valves will be monitored to verify no abnormal system flowoscillations.4.4Actual friction loss in the sump suction lines will be compared to thecalculated friction losses for acceptability. Each sump will be monitored during pump operation to verify that no vortices are formed.4.5RHR system flow rate in the cooldown lineup will be in accordance withdesign requirements.4.6Valve interlocks operate as required by design.

02-01 14.1-62Reformatted PerAmendment 02-01TABLE 14.1-35NUCLEAR INSTRUMENTATION SYSTEM1.0ObjectiveVerify that the Nuclear Instrumentation System performs the required indicationand control functions through the source, intermediate, and power ranges of operation.2.0Prerequisites2.1Nuclear Instrumentation System installed with calibration and initialalignments completed.2.2System energized for stabilization prior to commencing this test.2.3Systems complete as required for conducting this test.3.0Test MethodsUsing the installed test facilities, verify proper performance of instrumentation,including output signals to the Reactor Protection System, and remote indications.4.0Acceptance CriteriaSystem performance is in accordance with Section 7.2.

14.1-63AMENDMENT 97-01AUGUST 1997TABLE 14.1-36CONTROL BUILDING VENTILATION SYSTEMS1.0Objective1.1Demonstrate the operation of the following Control Building HVAC systemsduring normal and abnormal plant operating conditions.1.1.1Control Room Ventilation 1.1.2Relay Room Ventilation 1.1.3Computer Room Ventilation 1.1.4Controlled Access Ventilation1.2Confirm the proper operation of system equipment interlocks and controls.1.3Measure the total Control Room boundary leak rate.2.0Prerequisites2.1Individual Control Building Ventilation systems installation and componentchecks completed.2.2Associated systems completed to the extent necessary to allow conduct of this test.2.3Control Room pressure boundaries are sealed properly3.0Test Methods3.1Verify fan, damper, heater, humidifier, and cooling unit controls, interlocks,and permissives, for all systems.3.2Verify isolation damper operation on applicable systems.

3.3Verify actuations from radiation monitors and safety injection signals onapplicable systems.3.4Verify that Control Room internal pressure is greater than atmosphericpressure with a maximum of 400 cfm admitted from outside air.4.0Acceptance Criteria4.1The dampers and fans respond to recirculation signals in accordance withSections 6.5.1 and 9.4 for applicable systems.

14.1-64AMENDMENT 97-01AUGUST 1997TABLE 14.1-36 (Continued)CONTROL BUILDING VENTILATION SYSTEMS4.2Fans, dampers, heaters, humidifiers, and cooling units function correctly foreach system.4.3Control Room is maintained at positive (1/8") differential pressure withrespect to atmosphere in the normal and emergency modes of operation.4.4Control Room leak rate does not exceed its maximum design value.

14.1-65AMENDMENT 97-01AUGUST 1997TABLE 14.1-37CORE LOADING INSTRUMENTATION1.0ObjectiveVerify proper operation of the source range instrumentation channels prior to fuelloading operations.2.0Prerequisites2.1Temporary source range instrumentation installation checks completed.2.2Permanent source range channels operable.3.0Test Methods3.1Perform calibration of each source range channel.3.2Verify response of each channel to a neutron source.

3.3Verify audible signal from at least one permanent channel available incontrol room and Reactor Building.4.0Acceptance CriteriaInstrumentation provides monitoring of source range neutron level for loading fuelas required by the Technical Specifications.

14.1-66AMENDMENT 97-01AUGUST 1997TABLE 14.1-38REACTOR COMPONENTS AND FUEL HANDLING TOOLS AND FIXTURES1.0ObjectiveVerify the adequacy of the special equipment required for refueling operations.2.0PrerequisitesEquipment to be checked out is onsite and inspected in accordance with theroutine receiving inspection.3.0Test Methods3.1Check each tool for smooth performance and complete actuation.3.2Check adequacy of locating devices, guides, and chambers.

3.3Verify operation of interlocks and/or safety devices.3.4Verify load test lifting devices.4.0Acceptance CriteriaEquipment provides for safe handling of fuel assemblies and reactor components.NOTES:1.Test may be conducted with a dummy fuel element when a fuelelement is necessary for test.2.Testing was conducted on the polar crane in accordance withstandard crane testing procedures during the construction of the station. This test included:a.Hoists upper and lower limit switches verified.

b.Load test to 125%.

c.Operational performance under load.

The polar crane is under the administrative controls of the nuclearoperations department.

14.1-67Reformatted PerAmendment 02-01TABLE 14.1-39FUEL TRANSFER SYSTEM1.0ObjectiveProvide functional demonstration of the Fuel Transfer System and fuel handlingtools prior to initial core load.2.0Prerequisites2.1Reactor components and fuel handling tools and fixtures test completed.2.2Fuel Transfer System installation and component checks completed.2.3Reactor vessel head and upper internals stored in the refueling positions.2.4Dummy fuel assembly stored in a new fuel storage rack.3.0Test MethodsWith canal drained, conduct the various fuel handling evolutions with the dummy fuel assembly.4.0Acceptance CriteriaSystem provides for storage, transfer, and handling of fuel assemblies as designed.NOTE:Test may be conducted with a dummy fuel element when an element isneeded.

14.1-68Reformatted PerAmendment 02-01TABLE 14.1-40SAFETY INJECTION HIGH HEAD FLOW BALANCING TEST1.0Objective1.1To establish proper positioning of the high head cold leg injection andrecirculation throttle valve (8996A, B, C, 8994A, B, C) to balance injectionflows to each loop and to limit charging pump runout flow.1.2To establish proper positioning of the high head hot leg injection andrecirculation throttle valves (8989A, B, C, 8991A, B, C) to balance injection flows to each loop and to limit charging pump runout flow.1.3To record actual charging pump data (flow versus head) and compare this tomanufacturer's pump curve.1.4To demonstrate that motor operated valves in the high head S. I. Systemstroke properly against maximum expected differential pressure.1.5To demonstrate that the charging pumps are capable of taking a suction fromthe RHR pumps.2.0Prerequisites2.1Reactor vessel head is not installed and the internals are not in the vessel.2.2Sufficient water available in RWST to conduct test.3.0Test Methods3.1Each charging pump will be run through the cold leg injection path. Throttlevalves will be adjusted to provide the minimum required branch line flows from the pump whose flow path has the highest resistance (lowest flow) to the throttle valves. The throttle valves will then be locked in position and flow from each pump will be run through the branch lines to verify that minimum flow rates are still met and charging pump runout flow is limited. This method of testing will also be used for positioning throttle valves in the cold leg recirculation and hot leg recirculation flow paths.3.2Each charging pump will be run at various flow rates through the high headinjection lines to record pressure and flow data (motor currents and vibrations data will be recorded at full flow).

02-01 02-01 14.1-69Reformatted PerAmendment 02-01TABLE 14.1-40 (Continued)SAFETY INJECTION HIGH HEAD FLOW BALANCING TEST3.3Each charging pump will be operated while taking suction from an RHRpump. The charging pumps will be lined up to provide flow through one of thefour high head safety injection flow paths.4.0Acceptance CriteriaPump performance characteristics and minimum required flow rates in the injectionand recirculation flow paths will be in accordance with design requirements. Eachmotor operated valve will be observed locally for proper operation and timed to verify the valve strokes in less than the maximum allowable time. Each charging pump will be observed for any abnormal vibration under full flow when takingsuction from the RHR pumps.

14.1-70Reformatted PerAmendment 02-01TABLE 14.1-41S. I. ACCUMULATOR BLOWDOWN TEST1.0Objective1.1To demonstrate accumulator blowdown and verify that the accumulatorblowdown data is within the acceptable L/D range for each accumulator.1.2Time the opening of each accumulator discharge valve (8808A, B, & C) underflow conditions.2.0Prerequisites2.1The reactor vessel head is not installed and the internals are not in thevessel.3.0Test Methods3.1Fill each accumulator to a level consistent with vendor supplied accumulatorblowdown prerequisites.3.2Connect a brush recorder to accumulator pressure and level instrumentationand to the accumulator discharge valve indication circuitry.3.3Pressurize each accumulator to 100 psig.3.4Start the brush recorder and then open the accumulator discharge valve.3.5Using data obtained during accumulator blowdowns, calculate the L/D foreach accumulator discharge line.4.0Acceptance CriteriaThe L/D values for each accumulator discharge line are within design limits andthat each accumulator discharge valve opens in less than the maximum allowable stroke time.4.1Pressure and level alarm setpoints are proper.4.2Valve opening times and blowdown rates satisfy design requirements.

14.1-71AMENDMENT 97-01AUGUST 1997TABLE 14.1-42SPENT FUEL COOLING SYSTEM1.0ObjectiveDemonstrate flow capabilities of the Spent Fuel Cooling System2.0Prerequisites2.1Spent Fuel Cooling System installation and component checks completed.

2.2Adequate supply of grade B water available.

2.3No fuel is stored in the spent fuel pool.3.0Test Methods3.1Demonstrate filling and draining of the spent fuel pool, fuel transfer canal,cask loading area, and the refueling water storage tank.3.2Demonstrate circulation through demineralizer loop, heat exchangers, andskimmer loop.4.0Acceptance CriteriaSystem provides for filling, draining, and/or purification of the water in the spentfuel pool, fuel transfer canal, the cask loading area, and refueling water storage tank in accordance with Section 9.1.3.

14.1-72AMENDMENT 97-01AUGUST 1997TABLE 14.1-42aSPENT FUEL COOLING SYSTEM1.0ObjectiveDemonstrate flow capabilities of the Spent Fuel Cooling System2.0Prerequisites2.1Spent Fuel Cooling System installation and component checks.

2.2Reactor coolant drain tank and pumps installation and component checkscompleted.2.3Adequate supply of grade B water available.3.0Test Methods3.1Demonstrate filling and draining of the refueling canal and refueling cavity.3.2Demonstrate circulation through demineralizer loop and skimmer loop.4.0Acceptance CriteriaSystem provides for filling, draining, and/or purification of the water in the refuelingcanal and refueling cavity, in accordance with Section 9.1.3.

14.1-73AMENDMENT 97-01AUGUST 1997TABLE 14.1-43CONTAINMENT ISOLATION SYSTEM1.0ObjectiveDemonstrate the capability of the containment isolation system to respondproperly to an isolation signal.2.0Prerequisites2.1Containment isolation system installation and component checkscompleted.2.2Associated systems completed to the extent necessary to allow the conductof this test.2.3Containment isolation system and the applicable isolation valves inassociated systems aligned for normal operation.3.0Test Methods3.1Verify component response to Phase A containment isolation by manuallyinitiating a Phase A containment isolation signal from the MCB.3.2Verify component response to Phase B containment isolation by manuallyinitiating a spray actuation signal from the MCB.4.0Acceptance CriteriaSystem response to containment isolation actuation signals is in accordance withSection 6.2.4.

14.1-74AMENDMENT 97-01AUGUST 1997TABLE 14.1-44REACTOR PROTECTION OPERATIONAL CHECK1.0ObjectiveVerify the correct installation and proper operation of the reactor trip portion of theReactor Protection System.2.0Prerequisites2.1Reactor plant in cold shutdown condition.

2.2Instrumentation and Reactor Protective Systems installation checks andcalibrations completed.3.0Test Methods3.1Utilizing the appropriate train test panels, conduct individual tests of eachtrain's tripping logic.3.2Conduct overall logic test for both trains.4.0Acceptance CriteriaSystem performance is in accordance with Section 7.2 and the approved testprocedure.

14.1-75AMENDMENT 97-01AUGUST 1997TABLE 14.1-45ENGINEERED SAFETY FEATURES SYSTEMOPERATIONAL CHECK1.0ObjectiveVerify the operation of the ESF logic systems for all conditions of trip logic.2.0Prerequisites2.1Instrumentation and ESF systems installation checks and calibrationcompleted2.2Reactor plant in cold shutdown condition prior to core loading..3.0Test Methods3.1Conduct individual train logic tests.3.2Conduct overall logic test for both trains.

3.3Verify redundant tripping of each ESF channel through to the relay orcontroller that actuates the ESF device.4.0Acceptance CriteriaSystem performance is in accordance with Section 7.3.

14.1-76AMENDMENT 97-01AUGUST 1997TABLE 14.1-46INTEGRATED ENGINEERED SAFETY FEATURES TEST1.0ObjectiveDemonstrate the capability of the ESF equipment during a simulated accidentcondition to function in the proper sequence of manner within acceptable parameters.2.0Prerequisites2.1ESF systems checks and component checks completed.2.2Emergency diesel generators are fully operations.

2.3Systems to be tested are filled and aligned in recirculation with theappropriate storage tanks.3.0Test Methods3.1Verify the train A ESF load group assignment and component response toan ESF signal by defeating the automatic start of the train B diesel generator and de-energizing the train B ESF 7200 volt bus 1DB, then manually initiating a safety injection and spray actuation from the MCB.3.2Verify the train B ESF load group assignment and component response toan ESF signal by defeating the automatic start of the train A diesel generator and de-energizing the train A ESF 7200 volt bus 1DA, then manually initiating a safety injection and spray actuation signal from the MCB.3.3Verify the train A & B ESF component response to ESF signals with offsitepower by manually initiating a safety injection and spray actuation signal from the MCB.3.4Verify the train A & B ESF component response to ESF signals with onlyonsite power by simultaneously de-energizing offsite power and manually initiating a safety injection and spray actuation signal from the MCB. Diesel generator starting time and loading sequence will be recorded.4.0Acceptance Criteria4.1Emergency diesel generators respond within acceptable design limits..4.2ESF systems and components respond as required.

14.1-77AMENDMENT 97-01AUGUST 1997TABLE 14.1-47REACTOR BUILDING SPRAY SYSTEM1.0Objective1.1To record actual Reactor Building Spray pump data (flow vs. head) andcompare this to the manufacturer's pump curve.1.2To demonstrate that motor operated valves in the Reactor Building SpraySystem stroke properly against maximum differential pressure.1.3To determine the friction loss in the sump suction lines to each ReactorBuilding Spray pump and verify vortex control.1.4To demonstrate that the spray ring nozzles are free of obstructions.

1.5To record the NaOH drawdown rates using various combinations of ECCS pumps.2.0Prerequisites2.1Reactor Building Spray System installation and component checkscompleted.2.2Sufficient grade A water available in the refueling water storage tank andsodium hydroxide storage tank.3.0Test Methods3.1Each Reactor Building Spray pump will be run at various flow rates torecord pressure and flow data (motor currents and vibration data will berecorded at full flow).3.2Each motor operated valve in the Reactor Building Spray System will bestroked against maximum differential pressure.3.3Each Reactor Building Spray pump recirculation sump will be flooded withwater and each Reactor Building Spray pump will be run taking a suctionfrom its respective sump to determine friction losses in each pump suction line and to demonstrate vortex control.3.4Air will be forced through each spray nozzle to verify that nozzles are freeof obstructions.3.5Various combinations of ECCS pumps will be run and the NaOH drawdownrates will be recorded for each combination.

14.1-78AMENDMENT 97-01AUGUST 1997TABLE 14.1-47 (Continued)REACTOR BUILDING SPRAY SYSTEM4.0Acceptance Criteria4.1Actual pump performance characteristics will be in accordance with designrequirements.4.2Each motor operated valve will be observed locally for proper oper ation andtimed to verify the valve strokes in less than the maximum allowable time.4.3Actual friction loss in the sump suction lines will be compared to thecalculated friction losses for acceptability. Each sump will be monitored during pump operation to verify no vortexes are formed.4.4Each spray nozzle will be checked to ensure it passes air.

4.5Actual NaOH drawdown rates will be analyzed for acceptability.

14.1-79AMENDMENT 97-01AUGUST 1997TABLE 14.1-48LEAK DETECTION MONITORING SYSTEM1.0ObjectiveDemonstrate system capability of detecting the presence of significant leakagefrom the reactor coolant loops to the Reactor Building atmosphere during normal operations.2.0Prerequisites2.1Leak Detection Monitoring System installation and component checkscompleted.2.2Associated system completed to the extent necessary to allow the conductof this test.3.0Test Methods3.1Verify proper functioning of containment air particulate monitor andradioactive gas monitor detectors by exposure to standard test sources.3.2Verify monitor's flow rates and associated controls, indications, and alarms.

3.3Verify proper functioning of leak detection instrumentation associated withthe system.4.0Acceptance CriteriaSystem provides for monitoring of RCS leakage within design acceptance limits.

14.1-80AMENDMENT 97-01AUGUST 1997TABLE 14.1-49POST ACCIDENT HYDROGEN REMOVAL SYSTEM1.0ObjectiveDemonstrate the capability of the Post Accident Hydrogen Removal System toprovide adequate flow and heat for removal of combustible gases.2.0PrerequisitesPost Accident Hydrogen Removal System installation and component checkscompleted.3.0Test Methods3.1Verify remote actuation.3.2Demonstrate ability to obtain atmospheric samples from each sample point.3.3Verify proper operation of each hydrogen recombiner.4.0Acceptance CriteriaHydrogen recombiners function as designed.

14.1-81AMENDMENT 97-01AUGUST 1997TABLE 14.1-50RADIOACTIVE WASTE DISPOSAL SYSTEM1.0ObjectiveDemonstrate the ability of the Radioactive Waste Disposal System to providecontrolled handling and disposal of solid, liquid, and gaseous radioactive wastes..2.0Prerequisites2.1Solid Waste Processing, Liquid Waste Processing, and Gaseous WasteProcessing Systems installation and component checks completed.2.2Associated systems completed to the extent necessary to allow the conductof this test.3.0Test Methods3.1The system will be tested in a series of functional Phase I tests asdescribed in Table 14.1-85 for the following subsystems:a.Reactor coolant drain tank and pumps b.Waste evaporator feed pump and holdup tank c.Floor drain tank and pump d.Waste evaporator condensate tank and pump e.Waste evaporator f.Chemical drain tank and pump g.Spent resin storage and transfer h.Laundry and hot shower tank and pump I.Waste monitor tank and pump j.Waste evaporator concentrates tank and pump K.Radwaste solidification module lExcess waste holdup tank and pump m.Decontamination pit collection tank and pump n.Nuclear drain sump pumps 14.1-82AMENDMENT 97-01AUGUST 1997TABLE 14.1-50 (Continued)RADIOACTIVE WASTE DISPOSAL SYSTEMo.Waste gas compressors and control valvesp.Waste gas hydrogen recombiners q.Waste gas decay tank drain pump4.0Acceptance CriteriaSystem provides controlled handling and disposal of radioactive wastes inaccordance with Sections 11.2, 11.3, and 11.5.

14.1-83AMENDMENT 97-01AUGUST 1997TABLE 14.1-51BORON THERMAL REGENERATION SYSTEM1.0ObjectiveOperationally checkout the Boron Thermal Regeneration System (BTRS) andoperate the system with letdown flow.2.0Prerequisites2.1BTRS installation and component checks completed.

2.2The Reactor Coolant System at normal operating temperature and pressure.

2.3Associated systems completed to the extent necessary to allow the conductof this test.3.0Test Methods3.1Align the system for normal operation.3.2Operate the system in the dilution and the boration modes.4.0Acceptance CriteriaSystem flow, pressures, and temperatures are within design limits.

14.1-84AMENDMENT 97-01AUGUST 1997TABLE 14.1-52REACTOR PROTECTION SYSTEM TIME RESPONSE MEASUREMENT1.0ObjectiveVerify the Reactor Protection System response times and function of each trippath including sensor response time. Response time measurements will beverified for all sensors for which response time measurements are required by Technical Specifications, Tables 3.3-2 and 3.3-5.2.0Prerequisites2.1To be performed prior to initial fuel loading.

2.2Instrumentation and Reactor Protective System installation checks andcalibrations completed.3.0Test Methods3.1Utilizing test panels and temporary instrumentation as required, measurethe time response and verify the functioning of each trip path in the reactor protective circuitry by simulating the sensor input to the process protection cabinets.For ESF trip paths, the time response measured will include the sensorinput to the process protection cabinets through the energization of the slave relay in the solid state protection cabinets.Component response times (i.e., the valves travel to their required position,pump discharge pressures reach their required values, etc.) will be obtained from the pre-operational test performed on the systems for which the component is a part.Diesel generator starting and sequence loading delays will be obtained fromthe Integrated Engineered Safety Features Test (Table 14.1-46).Sensors response times will be obtained by vendor testing, by onsite benchtesting, or by in-place testing.The response times measured by this test and those sensor timeresponses obtained as outlined above and the component time responseobtained from preoperational tests will be algebraically summed to obtainan overall response time for each trip path.Delays in sensing lines will be determined analytically.

14.1-85AMENDMENT 97-01AUGUST 1997TABLE 14.1-52 (Continued)REACTOR PROTECTION SYSTEM TIME RESPONSE MEASUREMENT4.0Acceptance CriteriaResponse times of the individual "trip paths" including process to sensor couplingdelay are less than the maximum allowable times specified in the Technical Specifications, Tables 3.3-2 and 3.3-5.

14.1-86AMENDMENT 97-01AUGUST 1997TABLE 14.1-53INITIAL FUEL LOADING1.0ObjectiveTo accomplish initial fuel loading in a safe, orderly manner.2.0Prerequisites2.1All the tests required to be performed before fuel loading are completed.

2.2The Residual Heat Removal System is maintaining the coolant in constantrecirculation.

2.3Boron concentration is sufficient to maintain Keff .95.2.4Five source range neutron detectors are installed; two permanentlyinstalled, two temporary incore, and a third temporary to act as a spare.2.5Containment integrity has been established.3.0Test Methods3.1An initial nucleus of eight fuel assemblies with a source is loaded.3.2Using an inverse neutron count rate plot as a guide to insure safe loading,fuel assemblies are placed in the core until core loading is complete.4.0Acceptance CriteriaThe core is loaded in the specified configuration barring mechanical damage to anassembly.

14.1-87AMENDMENT 97-01AUGUST 1997TABLE 14.1-54INCORE MOVABLE DETECTORS1.0ObjectiveVerify proper response of the individual channels of instrumentation and the abilityto accurately position the detectors of the Incore Movable Detector System.2.0Prerequisites2.1Incore Movable Detector System installation and component checkscompleted.2.2"Manual local" operation has been checked using a dummy cable.2.3Core installed.

2.4Gas Purge System and Leak Detection System installation and componentchecks completed.3.0Test Methods3.1Align system for normal operation.3.2Verify proper operation of all transfer devices, isolation valves, safety andlimit switches, and readout and control equipment.3.3Compare position readouts with observed position of detectors.4.0Acceptance Criteria4.1System provides mapping capability as described in Section 7.7.

14.1-88AMENDMENT 97-01AUGUST 1997TABLE 14.1-55ROD DROP TIME MEASUREMENT1.0ObjectiveDetermine the drop time for each full length control rod at no flow cold conditionsand at full flow hot conditions. Also, the slowest rod and the fastest rod are tripped 10 times at no flow/cold conditions and at full flow/hot conditions.2.0Prerequisites2.1Core installed and reactor vessel head in place.

2.2Boron concentration equal to or greater than that required for refuelingshutdown.2.3Rod Position Indication System operable.2.4Both source range protection channels available in the control room.3.0Test Methods3.1Withdraw selected bank to the fully withdrawn position.3.2Conduct individual rod drop tests, recording rod drop time, rod travel time,and other specified data.3.3Repeat for all banks of full length rods in required conditions of flow andtemperature.3.4Drop the slowest rod and the fastest rod 10 times at the required flow andtemperature conditions.4.0Acceptance Criteria4.1Drop time for all rods is less than the maximum value specified in theTechnical Specifications.4.2Operation of the dashpot for all rods, as indicated by recorder traces, willbe verified.

14.1-89AMENDMENT 97-01AUGUST 1997TABLE 14.1-56ROD DRIVE MECHANISM TIMING1.0ObjectiveVerify proper timing of each Rod Control System slave cycler and conduct anoperational check of each full length control rod drive mechanism.2.0Prerequisites2.1All full length control rod drive mechanism equipment installed with rodcontrol cluster assemblies attached.2.2Reactor Coolant System filled and vented.2.3Boron concentration equal to or greater than that required for refuelingshutdown.2.4Baseline count rates established for each source range channel.2.5Test is to be performed at cold and hot standby conditions.3.0Test Methods3.1Verify the timing of each power cabinet's slave cycler.3.2Conduct individual mechanism operational checks by withdrawing andinserting each mechanism a specified number of steps while obtaining an oscillograph trace.4.0Acceptance CriteriaMechanism timing and operational checks verified to be within acceptable designlimits.

14.1-90AMENDMENT 97-01AUGUST 1997TABLE 14.1-57ROD POSITION INDICATION1.0Objective1.1Demonstrate that the Rod Position Indication System performs the requiredindication and alarm functions for each full length rod control cluster assembly.1.2Demonstrate performance of the full length rod control cluster assembliesover their full range of travel.2.0Prerequisites2.1Reactor Coolant System at normal operating no load temperature and pressure.2.2Boron concentration equal to or greater than that required for refuelingshutdown.2.3Cold shutdown alignment and adjustments of Rod Position IndicationSystem completed.3.0Test Methods3.1The rods are withdrawn in groups and their indicated positions arecompared with the group step counters.4.0Acceptance Criteria4.1Indicators and alarms function in accordance with Section 7.7.

14.1-91AMENDMENT 97-01AUGUST 1997TABLE 14.1-58REACTOR COOLANT SYSTEM FLOW MEASUREMENT1.0ObjectiveObtain the data to compute actual Reactor Coolant System flow rates as theyrelate to the design flow rates.2.0Prerequisites2.1Core installed.

2.2Reactor plant is in hot standby condition with all control rods fully inserted.

2.3Reactor coolant pumps operable.3.0Test Methods 3.1Measure loop temperatures, loop elbow tap p's, and reactor coolant pumpinput power and speed for various configurations of reactor coolant pumps.3.2Compute actual Reactor Coolant System flow rate. Density variation ofcold leg fluid will be accounted for in the data reduction method to providedirect comparison to full power conditions.4.0Acceptance Criteria4.1Reactor Coolant System flow rates are determined to be greater than orequal to the thermal design minimum less than or equal to the mechanical design maximum as per FSAR Table 5.1-1.4.2If the criteria as described in 4.1 is not met, the power level will berestricted to the rated thermal power which the measured Reactor Coolant System flow rate will support.

14.1-92AMENDMENT 97-01AUGUST 1997TABLE 14.1-59REACTOR COOLANT SYSTEM FLOW COASTDOWN1.0Objective1.1Measure the rate at which Reactor Coolant System flow changessubsequent to reactor coolant pump stops and starts.1.2Measure time delays associated with the loss of flow accident.2.0Prerequisites2.1Core installed.

2.2Reactor plant is in hot standby condition with all control rods fully inserted.3.0Test Methods3.1Selectively trip reactor coolant pumps from various configurations of pumpoperation.3.2Measure required flow data and response times for each configuration ofpump operation.4.0Acceptance Criteria4.1Time delays associated with the loss of flow accident are within the designvalues.4.2Rate of change of reactor coolant flow is within the design limits for thevarious pump configurations.

14.1-93AMENDMENT 97-01AUGUST 1997TABLE 14.1-60RESISTANCE TEMPERATURE DETECTOR BYPASSLOOP FLOW VERIFICATION1.0ObjectiveTo ensure the flow necessary to achieve the design objective for reactor coolanttransport time in each resistance temperature detector (RTD) bypass loop and to verify flow setpoint.2.0Prerequisites2.1The Reactor Coolant System is at normal operating no load temperatureand pressure.2.2Installed pipe measurements have been made and necessary constructionand tests are finished.3.0Test Methods3.1Measure flow in RTD bypass loops3.2Reduce flow to verify alarm setpoints.

3.3Increase flow and verify alarm setpoints.4.0Acceptance Criteria4.1Flow gives a transport time of 1 second or less in the RTD bypass loopssuch that the total time response given in Technical Specification Table 3.3-2, item 7, is not exceeded.4.2Alarms actuate and clear at proper setpoints.

14.1-94AMENDMENT 97-01AUGUST 1997TABLE 14.1-61REACTOR VESSEL O-RING LEAK TEST1.0ObjectiveVerify that there is no leakage past the reactor vessel head and vessel sealfollowing installation of the reactor vessel head after core loading.2.0Prerequisites2.1Core installed, reactor vessel head installed, and reactor vessel head studstorqued.2.2Reactor Coolant System pressure integrity verified in accordance withASME Code prior to core loading.3.0Test Methods3.1Establish normal operating no load temperature and pressure conditions forReactor Coolant System.3.2Increase system pressure to 100 psi above operating pressure and checkfor leakage past the head and vessel seal.4.0Acceptance CriteriaAcceptable leakage past reactor vessel head and vessel seal is determined.

14.1-95AMENDMENT 97-01AUGUST 1997TABLE 14.1-62PRESSURIZER SPRAY AND HEATER CAPABILITY ANDSETTING CONTINUOUS SPRAY FLOW1.0Objective1.1Establish proper continuous spray flow rates.

1.2Verify pressurizer normal control spray effectiveness.

1.3Verify pressurizer heater effectiveness.2.0Prerequisites2.1Core installed.

2.2Plant is in shutdown condition at approximately the normal operating noload temperature and pressure.3.0Test Methods 3.1Adjust continuous spray flow rates such that T between the pressurizerand spray lines is less than or equal to 200F and the spray line low-temperature alarms are clear.3.2Check normal control spray effectiveness by spraying down toapproximately 2000 psig.3.3Check heater effectiveness by energizing heaters with power operatedrelief valves in "close" and spray and level controls in manual. Allow pressure to increase to approximately 2300 psig.4.0Acceptance Criteria4.1Continuous spray flow adjusted in accordance with design requirements.4.2Heater and normal control spray effectiveness are in accordance withdesign requirements.

14.1-96AMENDMENT 97-01AUGUST 1997TABLE 14.1-63WATER QUALITY TEST1.0ObjectiveVerify acceptable water quality of Reactor Coolant System fill and makeup waterprior to initial criticality.2.0Prerequisites2.1Reactor Coolant System filled and vented in preparation for initial criticality.

2.2Reactor makeup water storage tank at operating level.3.0Test Methods3.1Sample Reactor Coolant System and analyze for chlorides, conductivity,total dissolved solids, pH, clarity, and fluorides.3.2Sample Reactor Makeup Water System and analyze for the above.4.0Acceptance CriteriaAnalyses are within the limits specified in the Technical Specifications.

14.1-97AMENDMENT 97-01AUGUST 1997TABLE 14.1-64INITIAL CRITICALITY1.0ObjectiveTo bring the reactor critical.2.0Prerequisites2.1All tests to be performed before the initial criticality have been performed.

2.2Reactor Coolant System is at normal operating temperature and pressure.3.0Test Methods3.1Withdraw shutdown bank.3.2Withdraw control banks leaving the last bank far enough in to provideeffective control.3.3Dilute the reactor coolant until critical.

3.4Achieve stead-state hot zero power by using control rod movement.4.0Acceptance CriteriaThe reactor achieves criticality in an orderly, safe manner.

14.1-98AMENDMENT 97-01AUGUST 1997TABLE 14.1-65LOW POWER TEST1.0Objective1.1Verify nuclear instrumentation overlap.

1.2Verify point of adding heat.

1.3Measure rods out boron concentration.

1.4Determine moderator temperature coefficient.

1.5Determine integral and differential worths for sequenced control banks.

1.6Determine differential boron worth.

1.7Measure ejected control cluster assembly worth at hot zero power.2.0Prerequisites2.1The Reactor Coolant System is in the hot zero power condition with thereactor critical.2.2Reactor Coolant System temperature is being maintained.2.3Required signals for data collection and recording are available.3.0Test Methods3.1The neutron flux level will be increased by outward control rod motion andthe nuclear instrumentation overlap recorded. Adjustments will be made asnecessary to insure minimum overlap as described in the Technical Specifications.3.2The neutron flux level will be increased by outward control rod motion untiltemperature feedback effects are noted. The upper limit for zero power physics testing is defined as approximately one decade below this level.3.3The all rods out, critical boron concentration is determined by measuringthe just critical boron concentration with bank D near the fully withdrawn position. The amount of reactivity held down by bank D is then dynamically determined by withdrawal of bank D, noting the amount of reactivity inserted and converting this value to an equivalent amount of boron.

14.1-99AMENDMENT 97-01AUGUST 1997TABLE 14.1-65 (Continued)LOW POWER TEST3.4The moderator temperature coefficient for various boron concentrations isobtained by dynamically measuring the reactivity change due to a temperature change in the primary system.3.5The sequenced bank differential rod worth is determined by either boratingthe Reactor Coolant System while withdrawing the control banks or by diluting the Reactor Coolant System while inserting the control banks to maintain nominal system criticality. Integral worth is then determined from the differential reactivity data.3.6Differential boron worth at hot zero power is determined by obtaining andanalyzing reactor coolant samples for boron content in conjunction with control bank movement to maintain nominal criticality during boration.

Boron concentration as a function of time in combination with integrated reactivity as a function of time is used to plot reactivity versus boron concentration, the slope of which yields differential boron worth.3.7Ejected rod cluster control assembly worth at hot zero power is determinedby obtaining a critical configuration with the sequenced rod banks at their insertion limit as defined in the Technical Specifications. The most reactive inserted rod is withdrawn to maintain nominal criticality during boration.

The reactivity addition is determined by summing the differential reactivity insertions as the rod is withdrawn to its withdrawal limit.4.0Acceptance Criteria4.1Nuclear instrumentation overlap meets the minimum requirements.

4.2The all rods out, critical boron concentration is within 50 ppm of designprediction.4.3The moderator temperature coefficient is negative under the allowedconditions of normal critical operation.4.4Measured integral worths for control banks are within 10% of predictedworths and total worth of all rods less the most reactive rod exceeds the shutdown reactivity requirements throughout the fuel cycle.

4.5Differential boron worth, over the range measured, is within 10 percent ofthe predicted value.

14.1-100AMENDMENT 97-01AUGUST 1997TABLE 14.1-65 (Continued)LOW POWER TEST4.6The measured integral reactivity worth of the RCCA withdrawn duringthe rod ejection test, and the resultant hot channel factor, F Q, are lessthan or equal to the values used in the safety analysis with adequateallowance for measurement uncertainty.

14.1-101AMENDMENT 97-01AUGUST 1997TABLE 14.1-65aAUGMENTED LOW-POWER TEST1.0Objective1.1Demonstrate decay heat removal capability of natural circulation.1.2Determine the depressurization rate following a loss of pressurizer heatersfor evaluation of the loss of all AC emergency procedure.1.3Demonstrate to the plant operators the effect of increased charging flowand reduced steam generator pressure on the saturation margin.1.4Demonstrate under simulated loss of offsite power conditions that decayheat can be removed via the steam generators by maintaining steam generator level with the Emergency Feedwater system.1.5Demonstrate under simulated loss of onsite and offsite power conditionsthat decay heat can be removed via the steam generators by maintaining steam generator level with the Emergency Feedwater system.1.6Demonstrate boron mixing and Reactor Coolant System cooldown usingnatural circulation (credit for this objective may be taken for testing completed on a similar plant).1.7Train Operators on a simulator in:a.Natural circulation operationsb.The use of auxiliary pressurizer spray c.Loss of offsite, and loss of offsite and onsite AC power2.0Prerequisites2.1The Reactor Coolant System and all emergency and auxiliary systems areat the conditions specified in the appropriate approved test procedure.2.2Required signals for data collection and recording are available.2.3Reactor Power is maintained less than or equal to five percent.

14.1-102AMENDMENT 97-01AUGUST 1997TABLE 14.1-65a (Continued)AUGMENTED LOW-POWER TEST3.0Test Methods3.1In Hot Standby with the Reactor Coolant Pumps supplying heat input to thesecondary side, simulate removal of onsite AC power sources and operate the plant utilizing manual control and the steam driven EmergencyFeedwater Pump. HVAC to the pump room is isolated to verify pump operability in this environment.3.2After Fuel Loading, but prior to Initial Criticality, establish stable condition at T no Load and 2235 psig with RCP B in operation. Reduce pressure byturning off pressurizer heaters noting depressurization rate. Re-establishheaters and reduce pressure by use of auxiliary spray noting depressurization rate and effect on margin to saturation temperature. At reduced pressure observe the effects of changes in charging flow and steam flow on margin to saturation temperature.3.3With the reactor critical at approximately 3% reactor power, place the plant in natural circulation mode observing the length of time for plant to stabilize,flow distribution, power distribution, and ability to maintain cooling mode.3.4Perform Loss-of-Offsite Power/Station Blackout Test with plant trip form 10-20% Rated Thermal Power. Operate Plant establishing stable conditions in natural circulation using batteries and emergency diesels.3.5Referencing boration and cooldown tests performed at Sequoyah I, NorthAnna II, Farley II, and Diablo Canyon I, verify similar plant response by parameter and plant comparison. Operator training for cooldown on natural circulation provided on a simulator.4.0Acceptance Criteria4.1Natural circulation was established and maintained under steady stateconditions in all cases.4.2Operator training during natural circulation was accomplished.

4.3Adequate boron mixing during RCS cooldown on natural circulation hasbeen proven at Virgil C. Summer Nuclear Station or credit has been taken for this test performed on a similar plant.

14.1-103AMENDMENT 97-01AUGUST 1997TABLE 14.1-66INCORE MOVABLE DETECTOR AND THERMOCOUPLEMAPPING AT POWER1.0ObjectiveTo obtain and analyze core power distributions for various control rodconfigurations at each major power plateau.2.0Prerequisites2.1The reactor is critical at a steady-state power level.

2.2Incore Instrumentation System functional test is complete and the system isoperable.2.3Computer systems are operable as necessary for incore map processing.3.0Test MethodsReactor power level is stabilized and complete incore flux maps are obtained andprocessed.4.0Acceptance CriteriaCore peaking factors are within acceptable limits as defined in the TechnicalSpecifications.

14.1-104AMENDMENT 97-01AUGUST 1997TABLE 14.1-67POWER COEFFICIENT AND POWER DEFECT MEASUREMENT1.0ObjectiveTo determine the differential power coefficient of reactivity and the integral powerdefect.2.0Prerequisites2.1The reactor is in the hot zero power condition with rods in the specifiedmaneuvering band.2.2The instrumentation necessary for collection of data is installed, calibrated,and operable.3.0Test MethodsReactor power is maintained congruent with turbine load demand by control bankadjustment throughout the range of each load change from the hot zero power condition to the hot full power condition. Reactivity increments due to periodic control bank steps are determined and recorded throughout each load change. At selected power levels, conditions are stabilized and a heat balance obtained toaccurately determine core power. Power coefficient and power defect are calculated with data obtained over the range from hot zero power to hot full power.4.0Acceptance Criteria4.1The best estimate of the measured power coefficient as a function ofpower, derived from experimental data, is equal to or more conservative than that used in the accident analysis.4.2The measured power defect is compatible with design predictions.

14.1-105AMENDMENT 97-01AUGUST 1997TABLE 14.1-68EFFLUENT RADIATION MONITOR TEST1.0ObjectiveTo verify the performance of the effluent monitors under actual dischargeconditions. This test is to be performed at each major power plateau.2.0Prerequisites2.1The reactor has been operating for a time sufficient to generaterepresentative effluents.2.2The effluent monitors have been checked against known sources.3.0Test MethodsFollowing standard procedures, the suitability of effluents for discharge is verifiedby radiochemical analysis. Discharge is commenced and the response of effluent monitors is observed and recorded. Effluent is sampled in accordance with established procedures and effluent monitor performance is verified through radiochemical analysis.4.0Acceptance CriteriaThe installed effluent monitors perform in accordance with design standards andproperly indicate the radioactive content of the effluent.

14.1-106AMENDMENT 97-01AUGUST 1997TABLE 14.1-69RADIATION SHIELDING SURVEY1.0ObjectiveTo measure radiation dose levels at preselected points throughout the plant toverify shielding effectiveness.2.0Prerequisites2.1Radiation survey instruments to be used are calibrated against known sources.2.2The reactor is critical at a steady-state power.3.0Test MethodsIn accordance with procedures for radiation surveys, dose levels are measured atpoints throughout the station.4.0Acceptance CriteriaMeasured radiation levels are within the limits for the zone designation of eacharea surveyed.

14.1-107AMENDMENT 97-01AUGUST 1997TABLE 14.1-70PROCESS COMPUTER1.0ObjectiveVerify the output value printout of the Nonsafety-Related process computer forcomputer inputs.2.0PrerequisitesComputer has been functionally checked out.3.0Test MethodsData comparisons are made between simulated process signals and the printoutof the process computer.4.0Acceptance Criteria4.1Errors between simulated process signals and computer output are withinacceptable design limits.4.2Alarm functions are within acceptable design limits.

14.1-108AMENDMENT 97-01AUGUST 1997TABLE 14.1-71THERMAL POWER MEASUREMENTS ANDINSTRUMENT CALIBRATION1.0ObjectiveTo determine the power output of the core at major power plateaus for nuclearinstrumentation calibration.2.0Prerequisites2.1The plant is in a steady-state condition.

2.2Necessary process instruments have been checked and calibrated.

2.3Steam generator blowdown has been stopped.3.0Test Methods3.1For each loop, check to ensure steam flow and feed flow are equal, thenrecord the mass flow rate.3.2Record steam pressure and find steam enthalpy of saturated liquidconditions.3.3Record feedwater temperature and find feedwater enthalpy for saturatedliquid conditions.3.4Using the proper equation determine thermal output.

3.5Calibrate nuclear instrumentation to agree with core thermal output.4.0Acceptance CriteriaInstrumentation is calibrated to indicate core output.

14.1-109AMENDMENT 97-01AUGUST 1997TABLE 14.1-72AUTOMATIC CONTROL SYSTEMS CHECKOUT1.0ObjectiveTo demonstrate that the automatic control systems for pressurizer pressure andlevel control, rod control, steam generator level control, and turbine generator control respond properly to changes in controlling parameters.2.0Prerequisites2.1The reactor and turbine generator are at approximately 25 percent powerwith normal operating parameters.2.2Related instrumentation has been checked and calibrated.3.0Test Methods3.1Using manual control of pressure control devices (sprays, reliefs, andheaters) increases and decreases are made in system pressure. The controls are placed in automatic and the responses noted.3.2Using manual control of pressurizer level control, changes are made fromthe normal level. The controller is placed in automatic and the responses to high and low levels are noted.3.3Using manual rod control, errors in Reactor Coolant System averagetemperature are initiated. The rod control is placed in automatic and the responses to high and low average temperatures are noted.3.4Using manual control of the various controllers (three main feedwater valvecontrollers and the master feedpump controller), steam generator level and feedwater pressure changes are made. The controllers are placed in automatic and the responses noted.3.5With turbine control in automatic, changes in steam pressure are made bymanual manipulations of control rods. Closing of a stop and control valve is used to change steam conditions. The turbine generator responses tothese manipulations are noted.4.0Acceptance Criteria4.1Automatic controllers return plant parameters to normal values withoutexcessive overshoot.4.2The recorded process variables are not more limiting than stated in thesetpoint study document.

14.1-110AMENDMENT 97-01AUGUST 1997TABLE 14.1-73PLANT RESPONSE TO STEP LOAD CHANGES1.0ObjectiveTo demonstrate satisfactory plant response to a 10 percent load change fromvarious power levels.2.0Prerequisites2.1The various control systems have been tested and are in automatic.

2.2All pressurizer and main steam relief and safety valves are operable.

2.3The control rods are in the maneuvering band for the power level existing atthe commencement of the test.2.4Plant conditions are stabilized and pertinent parameters to be measuredare connected to high speed recorders.3.0Test Methods3.1Output is manually reduced at a rate sufficient to simulate a step loadchange equivalent to approximately a 10 percent load decrease.3.2After stabilization of systems, output is manually increased at a ratesufficient to simulate a step load change equivalent to approximately a 10 percent load increase.3.3Pertinent parameters affected by a load change are measured andrecorded.3.4At various power levels, as required by the test procedure, the test isrepeated.4.0Acceptance Criteria4.1Neither the turbine nor the reactor trips, and no initiation of safety injectionis experienced.4.2No pressurizer or main steam relief or safety valves lift.

4.3No operator action is required to restore conditions to steady state.

4.4Control systems maintain their parameters within their design transientlimits per NSSS Supplier Setpoint Study Document.

14.1-111AMENDMENT 97-01AUGUST 1997TABLE 14.1-74PSEUDO ROD EJECTION TEST1.0ObjectiveTo verify the rod worth and hot channel factors assumed in the safety analysis.2.0Prerequisites2.1The reactor is critical and at a steady-state power level of greater than orequal to 10 percent but no more than 50 percent.2.2All of the excore instrumentation channels are operable.2.3The incore detectors and thermocouples are operable3.0Test Methods3.1The affected rod bank is placed at the full power insertion limit.3.2Single rod motion is accomplished by disconnecting the lift coils on all therods in the affected bank except the coil on the selected rod.3.3The selected rod is withdrawn from the core while power and reactorcoolant temperature are held constant by boron concentration changes.3.4Data is gathered by use of the incore detectors and thermocouples andboron concentration.4.0Acceptance CriteriaThe measured worth of the pseudo ejected rod and the associated hot channelfactors are more conservative than those assumed in the safety analyses.

14.1-112AMENDMENT 97-01AUGUST 1997TABLE 14.1-75ROD DROP TEST1.0ObjectiveTo demonstrate the operations of the negative rate trip circuitry in detecting thesimultaneous insertion of two cluster control assemblies.2.0Prerequisites2.1All power range nuclear instrumentation channels are operable.

2.2The reactor is at the steady-state power level specified in the procedurewith the controlling bank near the full power insertion limit.2.3Pertinent parameters to be measured are connected to recording devices.3.0Test Methods3.1All four power range nuclear instrumentation channel positive and negativerate trips are defeated with instrumentation set up to monitor the negative rate trip bistables.3.2Two rods from a common group most difficult to detect by excore detectorsdue to low worth and core location, are simultaneously dropped by removing voltage to both the movable and stationary gripper coils of the designated rod.3.3Following the transient, recorded data is evaluated for system andinstrumentation response.4.0Acceptance Criteria4.1The negative rate trip circuitry is initiated on a minimum of three powerrange nuclear instrumentation channels as a result of simultaneously dropping two control rods.

14.1-113AMENDMENT 97-01AUGUST 1997TABLE 14.1-76BELOW-BANK ROD TEST1.0ObjectiveTo demonstrate the response of the nuclear and incore instrumentation to a rodcluster control assembly below the nominal bank position and to determine hot channel factors associated with this misalignment.2.0Prerequisites2.1All power range nuclear instrumentation channels are operable.

2.2The moveable incore detectors are operable.

2.3Power escalation testing is completed to approximately the 50 percentreactor power level.3.0Test Methods3.1Single rod movement is accomplished by disconnecting the lift coils of allrods in the affected bank except the selected rod.3.2During rod cluster control assembly insertion, power range detectorcurrents, thermocouple maps, and moveable incore detector traces are periodically recorded to demonstrate sensitivity to RCCA misalignment.The power range detector data provides information to relate core quadrant tilt to rod cluster control assembly position. With the RCCA fully misaligned, a moveable detector flux map is obtained to verify resultant core hot channel factors.4.0Acceptance Criteria4.1The measured radial hot channel factor resultant from a single RCCA fullymisaligned from its bank is less than or equal to the value assumed in the safety analysis (Table 15.2-2) and does not exceed the Technical Specification limit for the power level at which the test is performed.4.2Incore and/or nuclear instrumentation is demonstrated to detect anysignificant power maldistribution caused by the misaligned rod cluster control assembly.

14.1-114AMENDMENT 97-01AUGUST 1997TABLE 14.1-77PLANT TRIP FROM 100% POWER1.0Objective1.1To demonstrate the ability of the primary and secondary pl ant and the plantautomatic control systems to sustain a trip from 100% power and to bring the plant to a stable condition following the transient.1.2To determine the overall response time of the reactor coolant hot legresistance temperature detectors.1.3To obtain data which is to be evaluated to determine if changes in thecontrol system setpoints are warranted to improve transient response based on actual plant operation.2.0Prerequisites2.1The various control systems are in the automatic mode and functioningproperly.2.2The reactor is steady state of 100 percent power, with the rods in themaneuvering band.2.3Pressurizer and main steam safety and relief valves are in service andoperable.2.4Pertinent parameters to be measured are connected to recording devices3.0Test Methods3.1Initiate a plant trip by manually initiating a turbine generator trip.3.2Pertinent parameters are recorded on recording devices.

3.3Following the transient, recorded data is evaluated for system andcontroller response and possible abnormalities.4.0Acceptance Criteria4.1The parameters recorded are not more limiting than those in the setpointstudy document.

14.1-115AMENDMENT 97-01AUGUST 1997TABLE 14.1-78LOSS OF OFFSITE POWER1.0ObjectiveTo demonstrate that the necessary equipment, controls, and indication areavailable following the isolation of the Offsite Power Distribution System to remove decay heat from the core using only emergency power supplies.2.0PrerequisitesThe plant is at a steady-state condition with greater than 10 percent (10%)generator output.3.0Test Methods3.1Simulate loss of power by manually de-energizing the offsite power suppliesto all 7.2 Kv buses, and initiating a turbine generator trip.3.2Using approved operating procedures, bring the plant to a hot standbycondition and maintain the plant in a hot standby condition for at least 30 minutes using the only emergency onsite power sources.4.0Acceptance CriteriaThe hot standby condition is achieved and maintained for at least 30 minutesusing only emergency onsite power sources.

14.1-116AMENDMENT 97-01AUGUST 1997TABLE 14.1-79SHUTDOWN FROM OUTSIDE THE CONTROL ROOM1.0ObjectiveThe purpose of this test is to demonstrate the following:1.That the plant can be taken off the line from power operations andmaintained in hot standby from outside the control room.2.That cooldown from hot standby can be initiated and maintained fromoutside the control room by reducing the RCS temperature approximately 50F without exceeding the cooldown rate.3.That the Residual Heat Removal System can be initiated and controlledfrom outside the control room by placing the RHRS in service and reducingthe RCS temperature approximately 50F without exceeding the cooldownrate.2.0Prerequisites2.1The plant is at some power level greater than or equal to 10 percent outputon the generator.2.2Two shifts of operators are onsite, one observes from control room, theother performs the test.3.0Test Methods3.1The plant is tripped off the line from outside the control room.

3.2Using switchgear, manual operators on valves, local indicators, and localpanels of controls and indicators, hot standby (approximately no load temperature) is maintained through the assistance of the plant communication system for a period of at least thirty minutes.3.3The plant is cooled from approximately no-load temperature approximately 50F without exceeding the cooldown rate.3.4RHRS is initiated from outside the control room and the plant is cooledapproximately 50F without exceeding the cooldown rate.4.0Acceptance CriteriaHot standby is maintained by the members of one shift from outside the controlroom for at least 30 minutes.The plant is cooled approximately 50F below the approximately no-load RCStemperature without exceeding the cooldown rate.The RHR system is initiated from outside the control room and is used to reducethe RCS temperature approximately 50F.Note: Steps 3.3 and 3.4 may be performed during hot functional testing.

14.1-117AMENDMENT 97-01AUGUST 1997TABLE 14.1-79aEMERGENCY LIGHTING1.0ObjectiveVerify that the Emergency DC Lighting System meets design requirements.2.0PrerequisitesAll construction checks have been completed.3.0Test Methods3.1Interrupt the standby lighting sources to provide illumination solely from theemergency system.3.2Record footcandle level data at selected locations.4.0Acceptance CriteriaEmergency Lighting System meets requirements of Section 9.5.3.3.

14.1-118AMENDMENT 97-01AUGUST 1997TABLE 14.1-79bHEAT TRACING FOR SAFETY RELATED OUTDOOR PIPING1.0ObjectiveDemonstrate the ability of the Heat Tracing System to maintain propertemperature control in safety related outdoor piping.2.0Prerequisites2.1Heat Tracing System installation and component checks completed.

2.2Associated systems completed to the extent necessary to allow conduct of this test.3.0Test Methods3.1Energize Heat Tracing System.3.2Monitor temperatures maintained by each heat tracing circuit with thesystem in a static condition.4.0Acceptance CriteriaEach heat tracing circuit maintains temperature within design limits.

14.1-119AMENDMENT 97-01AUGUST 1997TABLE 14.1-79cPRESSURE BOUNDARY INTEGRITY TEST1.0ObjectiveDemonstrate pressure boundary integrity conforms to applicable codes.2.0Prerequisites2.1Reactor vessel internals, head, and studs installed.

2.2Required weld non-destructive testing is complete and accepted.3.0Test Methods3.1The Reactor Coolant System is filled with Grade "A" water.

3.2The Reactor Coolant System is brought to hydrostatic test temperatureutilizing the system pumps.3.3The Reactor Coolant System is brought to the required test pressureutilizing the Chemical Volume Control System.4.0Acceptance Criteria4.1The pressure boundary satisfactorily withstands the test pressure asrequired by applicable ASME codes.

14.1-120AMENDMENT 97-01AUGUST 1997TABLE 14.1-79dSEISMIC INSTRUMENTATION1.0ObjectiveVerify operability of the seismic instrumentation.2.0PrerequisitesInstallation and calibration of the seismic instrumentation is complete.3.0Test Methods3.1Verify system response to a simulated signal.3.2Verify operability of system alarms.4.0Acceptance Criteria4.1System response to a simulated signal meets design intent.4.2System alarms function as designed.

14.1-121AMENDMENT 97-01AUGUST 1997TABLE 14.1-80POWER ASCENSION TEST PROGRAMTestLow Power 30%50%75%90%100%Initial CriticalityN/ALow Power TestX Incore Moveable Detector and Thermocouple Mapping at PowerXXXXXX Power Coefficient and Power Defect MeasurementXXXXX Effluent Radiation Monitor TestXXXXXX Radiation Shielding SurveyXXX Thermal Power Measurements and Instrument CalibrationXXXXX Automatic Control Systems CheckoutX Plant Response to Step Load ChangesXXX Pseudo Rod Ejection TestTo be specified in the test Rod Drop TestTo be specified in the test Below-Bank Rod TestX Plant Loss of Electrical Load XLoss of Offsite PowerGreater than or equal to 10% gen. load Shutdown from Outside the Control RoomGreater than or equal to 10% gen. load X- To be performed Note: Tests at each given power test plateau will be reviewed before increasing power to the next test plateau.

14.1-122AMENDMENT 97-01AUGUST 1997TABLE 14.1-81INITIAL TEST PROGRAM SCHEDULEStartup ActivityMonths BeforeCommercial OperationApproximateDuration WeeksReactor Coolant System Cold Hydro151Hot Functional Testing103 Core Loading62 Low Power Test54 Power Ascension38Startup group staffing will begin approximately 30 months prior to commercialoperation. The startup group will be completely staffed and trained approximately 20months prior to commercial operation.

14.1-123AMENDMENT 97-01AUGUST 1997TABLE 14.1-82GENERIC FLUSH PROCEDURE1.0ObjectiveFlush plant systems to the appropriate level class of cleanliness.2.0Prerequisites2.1Construction is completed and the system is turned over to the point to allowflushing.2.2Adequate source of clean water is available.2.3Permanent plant equipment, i.e., pump or temporary flush equipment, i.e.,Hydrolaser is available for the flush.3.0Test Methods3.1Align the system valves to flush the system.

3.2Using permanent plant equipment or Hydrolaser flush the system.4.0Acceptance Criteria4.1By flush cloth or visual observation verify that the system meets the desired cleanliness class.

14.1-124AMENDMENT 97-01AUGUST 1997TABLE 14.1-83GENERIC HYDROSTATIC/PNEUMATIC TEST1.0ObjectiveProve structural integrity of the systems being tested to the design requirementsas specified in the particular design specification.2.0Prerequisites2.1The system to be tested must be completed sufficiently enough byconstruction to alleviate the need for system boundary re-entry and test invalidity.2.2The system is to be flushed, if practical, before the test is to be conducted.2.3The document research and quality control requirements must be completedbefore testing can begin.3.0Test Methods3.1Calibrated test gages are installed at selected locations for test pressureverification.3.2The system is filled with the appropriate hydrostatic or pneumatic testmedium at a temperature higher than the minimum specified in the design specification.3.3By use of test pumps, compressors, or other appropriate means the systemis pressurized to the test pressure and held for the specified time limit, then lowered to the inspection pressure and structural integrity is verified.3.4After the test, the system is drained or vented then it is put in the proper layup condition.4.0Acceptance Criteria4.1No external leakage is allowed except at 1.) temporary connections,installed for the purpose of making the test, 2.) at gaskets, seals, leak offs, etc., on previously tested components, 3.) items noted on the test report as being permitted by specification.4.2The leakage shall not exceed the capacity of the pressure source tomaintain the required test pressure.

14.1-125AMENDMENT 97-01AUGUST 1997TABLE 14.1-84INSTRUMENT CONTROL PROCEDURE1.0Objective1.1Verify and, if required, re-establish the accuracies and control functions ofthe channel sensors, associated signal processing equipment, indicating equipment, and alarms associated with the analog process signals.2.0Prerequisites2.1Process control cabinets are installed and energized.

2.2Process control sensors are installed and connected to the process controlcabinets.2.3Interconnection between the process control cabinets and the main controlboard is complete.3.0Test Methods3.1Using test equipment to simulate the process parameter the correctionconversion, by the sensor, of the process parameter to an analog signal will be verified.3.2Using a signal simulator to simulate an analog signal from the sensor thecorrect processing of the analog signal by the process equipment, setpoints of comparators, and operation of indicating equipment will be verified.4.0Acceptance Criteria4.1The process instrumentation provides intolerance conversion of the processparameter and performs its as-designed indication and control functions.

14.1-126AMENDMENT 97-01AUGUST 1997TABLE 14.1-85FUNCTIONAL TEST1.0ObjectiveTo test the operability of subsystems and non-safety related systems todemonstrate their ability to meet their functional requirements.2.0Prerequisites2.1Specify any tests or portions of tests which must be completed prior toconduct of this test.2.2Specify plant status necessary to start the test.3.0Test Methods3.1Step by step method to accomplish the test.

3.2Specify data to be taken during the test such as; vibration, bearingtemperatures, pressure and flow characteristics, automatic control functions,valve open and closing under actual pressure conditions.3.3Restore the system to a safe post test condition.4.0Acceptance Criteria4.1Specify those qualitative or quantitative requirements that demonstrate thesystems ability to meet its functional requirements.

14.1-127AMENDMENT 97-01AUGUST 1997TABLE 14.1-86STEAM GENERATOR POWER OPERATED RELIEF VALVE1.0Objective1.1Demonstrate that the steam generator power operated relief valves respondas designed to control Reactor Coolant System temperature utilizing steampressure as the input signal.1.2Demonstrate that the steam generator power operated relief valves respondas designed to control Reactor Coolant System temperature utilizing Tavgas the input signal.2.0Prerequisites2.1Phase I testing complete on steam generator power operated relief valvesand associated process instrumentation control loops.3.0Test Methods3.1During hot functional testing verify proper operation of the power operatedrelief valves in the pressure control mode at Reactor Coolant Systemtemperatures of 350F, 450F, and approximately 557F.3.2Verify the proper operation of the steam generator power operated reliefvalves in the temperature control mode by inducing a simulated Tavg signal in the process control racks.4.0Acceptance Criteria4.1Visual observation of the valves stroking to full flow at a reactor coolanttemperature of approximately 557F detects no abnormal movement ofvalves or associated piping.4.2Steam generator power operated relief valves respond as required tosimulated inputs of Tavg.

14.1-128AMENDMENT 97-01AUGUST 1997TABLE 14.1-87CONDENSATE SYSTEM1.0Objective1.1Demonstrate that the Condensate System delivers condensed steam fromthe main condenser hotwell through the low pressure feedwater heaters to the deaerator.1.2Demonstrate that the Hotwell Level Control System automatically makes upand rejects water to the condensate storage tank.1.3Demonstrate the system interlocks and automatic devices perform theirintended functions.2.0Prerequisites2.1Condensate System installation is complete and component checks arecompleted.2.2Deaerator and hotwell are capable of receiving water.3.0Test Methods3.1System is aligned for operation.

3.2Check operation of condensate pump speed controls.

3.3Verify pump head and flow rates.

3.4Verify that the Hotwell Level Control System functions to maintain hotwelllevel.3.5Check operation of pump trips and interlocks.4.0Acceptance Criteria4.1Pump flow capabilities meet or exceed design requirements.4.2Interlocks, pump trips, and protective devices function per designrequirements.4.3Verify control system function per design requirements.

14.1-129AMENDMENT 97-01AUGUST 1997TABLE 14.1-88FEEDWATER SYSTEM1.0Objective1.1Demonstrate that the Feedwater System provides heated and deaeratedwater to the steam generators during normal and transient plant conditions.1.2Demonstrate that plant and component protection devices perform theirintended functions.2.0Prerequisites2.1Feedwater System installation is complete and component checks arecompleted.2.2Steam generators are capable of receiving water.3.0Test Methods3.1System is aligned for operation.

3.2Check operation of feedwater pump turbine speed controls to provideprogrammed differential pressure between the steam generator and feedwater pump.3.3Verify pump head and flow rates.3.4Check operation of pump trips and interlocks.4.0Acceptance Criteria4.1Pump flow capabilities meet or exceed design requirements.4.2Interlocks, pump trips, and protective devices function per designrequirements.4.3Verify control system function per design requirements.

14.1-130AMENDMENT 97-01AUGUST 1997TABLE 14.1-89MAIN CONDENSER DUMP VALVES1.0Objective1.1Demonstrate that the main condenser dump valves respond properly tosimulated temperature and pressure input signals.1.2During hot functional testing, cycle each steam dump valve full open.2.0Prerequisites2.1Phase I testing complete on main condenser dump valves and associatedprocess instrumentation control loops.3.0Test Methods3.1Simulate temperature and pressure inputs to the control circuitry for themain steam dump valves and verify proper system response.3.2During hot functional testing, cycle each steam dump valve fully open andfully closed and verify no abnormal valve or pipe movement.4.0Acceptance Criteria4.1Main condenser dump valves respond properly to simulated temperatureand pressure inputs.4.2Each main steam dump valve cycles fully open and closed with no abnormalvalve or pipe movement.

14.1-131AMENDMENT 97-01AUGUST 1997TABLE 14.1-90CIRCULATING WATER SYSTEM1.0Objective1.1Demonstrate the capability of the Circulating Water System to provideadequate cooling water.2.0Prerequisites2.1Circulating Water System installation, piping flushes, and component checksare completed.2.2Associated systems are completed to the extent necessary to allow conductof this test.3.0Test Methods3.1Align the system for operation.

3.2Verify system protective interlocks 3.3Verify flows and pressures for normal operations.4.0Acceptance Criteria4.1System flows, pressures, and automatic functions are in accordance withdesign requirements.4.2System interlocks function in accordance with design.

14.1-132AMENDMENT 97-01AUGUST 1997TABLE 14.1-91CHEMICAL FEED SYSTEM1.0Objective1.1Demonstrate that the Condensate Chemical Injection System can maintainestablished limits of pH and dissolved oxygen in the Condensate and Feedwater Systems.1.2Demonstrate that the Steam Generator Standby Chemical Injection Systemcan maintain established limits of pH and dissolved oxygen in the secondary side of the steam generators during wet layup.2.0Prerequisites2.1Chemical Injection System installation and component checks are complete.3.0Test Methods3.1Align systems for normal operation and establish flow paths.

3.2Check operation of metering controls for injection systems.4.0Acceptance Criteria4.1Condensate System delivers calibrated volume to the Condensate System.4.2Steam Generator Standby Chemical Injection System delivers a calibratedvolume to the steam generator.

14.1-133AMENDMENT 97-01AUGUST 1997TABLE 14.1-92NUCLEAR BLOWDOWN PROCESSING SYSTEM1.0Objective1.1Demonstrate flow capabilities of the Nuclear Blowdown Processing System.2.0Prerequisites2.1Nuclear Blowdown Processing System installation and component checksare completed.3.0Test Methods3.1Demonstrate circulation through different loops of the Nuclear BlowdownProcessing System and verify ability to demineralize water to CondensateSystem grade water.4.0Acceptance Criteria4.1System provides the capability to demineralize steam generator nuclearblowdown to Condensate System cleanliness criteria.

14.1-134AMENDMENT 97-01AUGUST 1997TABLE 14.1-93CONTROL ROD DRIVE1.0Objective1.1Demonstrate that the circuitry and components comprising the Control RodDrive System will perform their design requirements without control rod drives installed.2.0Prerequisites2.1Wiring and component installation complete.3.0Test Methods3.1Initially energize the rod position indication system and verify proper signalresponse from the detection coils to the indication display.3.2Verify proper control rod drive mechanism polarity and coil currents.3.3Verify proper control rod drive M-G set operation.

3.4Verify proper control rod drive mechanism magnetic coil sequencing.4.0Acceptance Criteria4.1Control rod drive components and circuitry function as required by design.

14.1-135AMENDMENT 97-01AUGUST 1997TABLE 14.1-94MISCELLANEOUS PLANT DRAINS1.0Objective1.1Demonstrate the ability of Miscellaneous Plant Drains System to removewater from the plant.2.0Prerequisites2.1Miscellaneous Plant Drains System installation and component checks arecompleted as necessary.3.0Test Methods3.1Verify capability of the sump pumps to remove water form the sumps.

3.2Verify sump level interlocks to applicable plant equipment.4.0Acceptance Criteria4.1Sump level interlocks to plant equipment function as designed.4.2System provides for water removal in accordance with design requirements.

14.1-136AMENDMENT 97-01AUGUST 1997TABLE 14.1-95FUEL HANDLING BUILDING POOL LINER LEAK TEST1.0Objective1.1Demonstrate water tight integrity of spent fuel pit, fuel transfer canal, andcask loading pit.1.2Demonstrate water tight integrity of the fuel transfer tube isolation valve andblind flange.1.3Demonstrate water tight integrity of spent fuel pit gate and cask loading pitgate.1.4Demonstrate air tight integrity of the inflatable seals on the spent fuel pit andcask loading pit gates.2.0Prerequisites2.1Pool liners have been installed and cleaned2.2Spent fuel racks are not installed.

2.3Adequate source of water is available.3.0Test Methods3.1Fill pools with water and monitor liner leakage.

3.2Install transfer canal isolation gates, drain canal and monitor leakagethrough the gates.4.0Acceptance Criteria4.1Total leakage for liner is less than design maximums.

4.2Fuel transfer tube isolation valve4.2.1Blind flange and valve is less than design maximums.

4.2.2Valve stem is less than design maximums.4.3Isolation gates4.3.1Inflatable seal on gates - no visible bubbles.

4.3.2Spent fuel pit gate is less than design maximums.

4.3.3Cask loading pit gate is less than design maximums.

14.1-137AMENDMENT 97-01AUGUST 1997TABLE 14.1-96REACTOR BUILDING VENTILATION "POST ACCIDENT OPERATION"1.0Objective1.1Demonstrate that the Reactor Building Cooling Units have the capacity tocool the Reactor Building within the design load during the worst postulated accident conditions.1.2Demonstrate during the integrated leak rate test that the Reactor Buildingcooling fan motor over current protection will allow operation of the motorsunder the maximum load.1.3Demonstrate during the integrated leak rate test that the Reactor BuildingCooling Unit bypass damper closes.2.0Prerequisites2.1Installation of components and support systems are complete andoperational.2.2The integrated leak rate test is in progress for objective 1.2.2.3Adjusts fan pitch settings to accommodate increased containment airdensity (for ILRT).3.0Test Methods3.1Align the R. B. Coolers for system operation in the "Post-Accident" mode.3.2Measure the flowrate and temperature drop across the cooling coil air andwater sides at ambient conditions.3.3Calculate the heat removal capacity and extrapolate for worst conditionsusing the coil data sheets and measured data.3.4Align system for operation during the containment integrated leak rate testand monitor the fan motor operation.3.5During the integrated leak rate test, reposition the Reactor Building CoolingUnit dampers from the bypass to the closed position.4.0Acceptance Criteria4.1The cooling units are shown to have the capacity to remove the designedheat load at the worst postulated conditions.4.2The fan motor operates in accordance with design requirements during theintegrated leak rate test.4.3Reactor Building Cooling Unit dampers operate from the bypass to theclosed position.

14.1-138AMENDMENT 97-01AUGUST 1997TABLE 14.1-97ESF EQUIPMENT ROOMS COOLING SYSTEMS1.0Objective1.1Demonstrate that the cooling systems for the following ESF EquipmentRooms have the capacity to cool the areas within design loads during the worst postulated conditions.1.1.1Reactor Bldg. Cooling Units (Normal Mode)1.1.2Charging Pump Room Cooling Unit 1.1.3RHR-Spray Pump Room Cooling Units 1.1.4Safety Related MCC Switchgear Cooling Units 1.1.5Relay Room Cooling System (Recirc. Mode) 1.1.6Control Room Ventilation System (Recirc. Mode) 1.1.7ESF Switchgear Room Cooling 1.1.8Speed Switch Room Cooling 1.1.9Service Water Booster Pump Cooling Units 1.1.10Emergency Feedwater Pump Cooling Units2.0Prerequisites2.1Installation of components and support systems are complete andoperational.3.0Test Methods3.1Align the cooler for system operation and measure the flowrate andtemperature drop across the cooling coil air and water sides at ambient conditions.3.2Calculate the heat removal capacity at ambient conditions. Verify that thecoils are operating on their characteristic curve by comparing test data results to manufacturer's design program data.4.0Acceptance Criteria4.1The cooling units are shown to operate on their characteristic curve, therebyverifying that they have the capacity to remove the design heat load at the worst postulated conditions for each of their respective areas.

14.1-139AMENDMENT 97-01AUGUST 1997TABLE 14.1-98S.I. ACCUMULATOR DISCHARGE VALVE FUNCTIONAL TEST1.0Objective1.1To demonstrate that each accumulator isolation valve (8808 A, B, & C) willopen under the maximum differential pressure conditions of zero RCS pressure and maximum expected accumulator precharge pressure upon receipt of a safety injection signal by each valve.1.2To verify the actual accumulator level matches the indicated accumulatorlevel at the minimum technical specification level.1.3To demonstrate that each accumulator isolation valve (8808 A, B, & C) willopen on an increasing RCS pressure signal.2.0Prerequisites2.1The Reactor Coolant System is depressurized and adequately vented.2.2A source of nitrogen is available to pressurize the accumulators.

2.3A tygon hose is connected to the accumulator for level indication.3.0Test Methods3.1Close valves 8808 A, B, & C and then fill each accumulator with grade Awater. As each accumulator fills, compare the accumulator actual level with the indicated level when actual level reaches the minimum technical specification level. Isolate the tygon hose before pressurizing the accumulators.3.2Pressurize each accumulator to the maximum expected accumulatorprecharge pressure.3.3Initiate a safety injection signal to the control circuitry of each valve.3.4Increase the RCS pressure signal to each valve.4.0Acceptance Criteria4.1That each accumulator isolation valve will open in less than the maximumallowable stroke time under maximum differential pressure conditions.4.2That indicated accumulator level is within tolerance at the actual minimumtechnical specification level in the accumulator.

14.1-140AMENDMENT 97-01AUGUST 1997TABLE 14.1-98 (Continued)S.I. ACCUMULATOR DISCHARGE VALVE FUNCTIONAL TEST4.3That each accumulator isolation valve will open when required on anincreasing RCS pressure signal.

14.1-141Reformatted PerAmendment 02-01TABLE 14.1-99ECCS CHECK VALVE LEAK TESTING SYSTEM OPERATIONAL TEST1.0Objective1.1Demonstrate proper functioning of the ECCS Check Valve Leak TestingSystem and perform a backleakage test on each check valve monitored bythe ECCS Check Valve Leak Testing System.2.0Prerequisites2.1Instrumentation in the ECCS Check Valve Leak Testing System has beencalibrated.2.2Hot functional testing is in progress.3.0Test Methods3.1Perform a leak test on each check valve in the ECCS normally checked usingthe ECCS check valve leak testing system.4.0Acceptance Criteria4.1System operates satisfactorily during leak check tests.

14.1-142Reformatted PerAmendment 02-01TABLE 14.1-100S.I. ACCUMULATOR CHECK VALVE HOT OPERATIONAL TEST1.0Objective1.1To demonstrate that the accumulator discharge check valves (8948A, B & Cand 8956A, B & C) will function under normal operating temperature andpressure conditions.2.0Prerequisites2.1Hot functional testing is in progress.3.0Test Methods3.1Inject or pump water through each check valve during hot functional testing.4.0Acceptance Criteria4.1That each check valve passes water into the RCS during hot functionaltesting. 02-01 14.1-143AMENDMENT 97-01AUGUST 1997TABLE 14.1-101INSTRUMENT AIR SYSTEM1.0Objective1.1Demonstrate the ability of the Instrument Air System to supply instrumentand service air.2.0Prerequisites2.1Instrument Air System installation and component checks are completed asnecessary.2.2System has been blown down and cleanliness requirements met.3.0Test Methods3.1Verify operability of system alarms.3.2Verify operability of system interlocks.

3.3Verify ability of air compressors to maintain system pressure.

3.4Verify start of standby compressor on reduction of air pressure.

3.5Verify isolation of service air on reduction of air pressure.

3.6Verify backup supply to Reactor Building header on reduction of pressure.

3.7Verify operability of dryers.4.0Acceptance Criteria4.1System interlocks and alarms function as designed.4.2Instrument Air System supplies air to Reactor Building Instrument Airservices on loss of Reactor Building Instrument Air.4.3Service air is isolated on loss of Instrument Air.

4.4Air compressors and dryers function as designed.

14.1-144AMENDMENT 97-01AUGUST 1997TABLE 14.1-102PRESSURIZER PRESSURE AND LEVEL CONTROL1.0Objective1.1To demonstrate the response, stability, and control characteristics of thePressurizer Pressure and Level Control System.1.2To demonstrate the operation of the pressurizer power relief valves.2.0Prerequisites2.1The Reactor Coolant System is at hot no-load conditions as required by hotfunctional testing.2.2The pressurizer relief tank is operable.2.3Pressurizer pressure and level process instrumentation has been calibrated.3.0Test Methods3.1Place the pressurizer pressure control in manual and increase (decrease)the setpoint of the automatic controller 50 psig. Switch the pressurizer pressure control back to automatic control and observe the system response as pressure increases (decreases) to the new setpoint.3.2Using manual control, increase the system pressure to the pressurizer highpressure reactor trip value, then lower the pressure to the pressurizer low pressure reactor trip value, and verify the operation of the alarms and interlocks associated with the Pressurizer Pressure Control System.3.3Place the pressurizer level control in manual and decrease (increase) thepressurizer level 5%. Switch the pressurizer level control to automatic and observe the system response as pressurizer level increases (decreases) tothe normal setpoint.3.4Using manual control, increase pressurizer level to the high level reactor trip,then lower the pressurizer level to the low level heater cutout and letdown isolation value, and verify the operation of alarms and interlocks associatedwith the Pressurizer Level Control System.3.5Verify the operation and response time of the pressurizer power relief valvesby opening and closing the valves from the MCB and allowing a blow down for ten (10) seconds or until 1900 psig is reached.

14.1-145AMENDMENT 97-01AUGUST 1997TABLE 14.1-102 (Continued)PRESSURIZER PRESSURE AND LEVEL CONTROL3.6Verify the operation of the pressurizer backup heaters from the control roomevacuation panel.4.0Acceptance Criteria4.1System operation is in accordance with design requirements.

14.2 AUGMENTATION OF STAFF FO R INITIAL TEST AND OPERATION During the initial test program at Virgil C. Summer Nuclear Station, SCE&G is using a special startup organization. In addition to the plant operati ng staff, the startup organization is augmented by representativ es from Westinghouse, Gilbert, and other contractors and vendors as required. Co mpetent technical personnel from other SCE&G departments may be called upon to assist and advise the startup organization as necessary.

14.2.1 ORGANIZATIONAL FUNCT IONS, RESPONSIBILITIES AND AUTHORITIES The functions, responsibilitie s, and authorities of organizations participating in the startup program are discussed in Se ctions 14.2.1.1 through 14.2.1.3.

14.2.1.1 Plant Operating Staff Under the direction of the Manager, Virgil C. Summer Nuclear Station qualified plant personnel to assist in preparation of test procedures and performance of tests. Test procedure preparation includes writing and review of the procedure and approval of the procedure by the Manager, Virgil C. Summe r Nuclear Station. Test performance includes checking of prerequi sites, operation of equipment, and recording of data. All functions requiring an NRC license are performed by licensed personnel.

Prior to receiving nuclear fuel at the plant site, SCE&G will establish industrial security and health physics programs under the control and administrat ion of the Manager, Virgil C. Summer Nuclear Station. Also, under t he direction of the Manager, Virgil C. Summer Nuclear Station, plant operating and maintenance personnel will assume responsibility for operation and maintenance of tested and accepted systems.

14.2.1.2 Westinghouse El ectric Corporation Westinghouse provides onsite technical assistance to SCE&G during the installation, startup, testing, and initial operation of each nuclear steam supply system (NSSS). In this manner, Westinghouse aids the owner and assures that each NSSS is installed, started, tested, and operated in conformance with design intent. Westinghouse onsite personnel provide direct assistance and act as technical liaison with Westinghouse headquarters to resolve problems within the Westinghouse scope. Personnel assignments are made as required by site activities.

14.2.1.3 Gilbert Associates, Inc. Gilbert provides experienced personnel to assist in the startup activities. Liaison is maintained between the Gilbert Reading, PA, offices and field perso nnel to facilitate resolution of problems related to test spec ifications and system operation for which Gilbert has design responsibility.

14.2-1 AMENDMENT 97-01 AUGUST 1997 14.2.2 INTERRELATIONSHIPS AND INTERFACES Onsite representatives of Westinghouse, Gilbert, and other contractors and vendors are responsible to the Manager, Vi rgil C. Summer Nuclear Stati on. If any of the above personnel participate in the perfo rmance of a test, they will be responsible to the person above them in the startup organization, that is a test supervisor or the startup supervisor. For activities related to construction, they will be responsible to the SCE&G

Manager of Construction.

14.2.3 PERSONNEL FUNCTIONS, RESPONSIBILITIES, AND AUTHORITIES The functions, responsibilitie s, and authorities of major augmenting personnel are as discussed in Sections 14.

2.3.1 through 14.2.3.3.

14.2.3.1 Plant Staff Members of the Virgil C. Summer Nuclear Station operating, maintenance, and technical groups who meet the requirements fo r test supervisor or test personnel are utilized in these capacities. Their functions, responsibilities, and authorities are described in Section 14.1.1.4.

14.2.3.2 Westinghouse El ectric Corporation Westinghouse startup specialists assigned to the Virgil C. Summer Nuclear Station act as technical advisors to the Plant Manager and assist him in NSSS startup activities.

Their function and responsibilities include, but are not limited to, assisting the startup group in procurement of pr ocedures, system checkout, functional testing, and coordination of the test program. Resolution of te chnical problems within the Westinghouse scope of supply shall be resolved by the startup specialists by direct liaison with Westinghouse headquarters.

14.2.3.3 Gilbert Associates, Inc. Gilbert personnel assigned to the Virgil C. Summe r Nuclear Station site assist in startup activities. Liaison is maintained between the site personnel and the Gilbert home office

to facilitate resolution of problems and questions which may arise during startup and testing and are related to Gil bert design activities. Gilbert site personnel may assist SCE&G in actual performance of st artup and testing of the plant.

14.2.4 PERSONNEL QUALIFICATIONS Personnel who manage, supervise, or perfo rm preoperational or startup tests shall satisfy the qualifications st ated in Section 14.1.1.4.

14.2-2 AMENDMENT 97-01 AUGUST 1997